ML022590378
| ML022590378 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/12/2002 |
| From: | Wilson M Dominion Nuclear Connecticut |
| To: | Conte R NRC/RGN-I/DRS/OSB |
| Conte R | |
| References | |
| -RFPFR, 50-423/02301, ES401-6, NUREG-1021, Rev 8 50-423/02301 | |
| Download: ML022590378 (527) | |
Text
{{#Wiki_filter:ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 1 Tier# 1 Group# I K/A # 001.AK3.02 Importance Rating 4.3 Proposed Question: With the plant at 50% power, the following sequence of events occurs:
- 1. A single Control Bank "D" rod starts to withdraw with no demand signal present.
- 2. The RO places the rod bank selector switch in MANUAL, and the rod motion stops.
- 3. The US enters AOP 3552 "Malfunction of the Rod Drive System".
- 4. The SM refers to Technical Specification section 3/4.1.3 "Movable Control Assemblies".
What is one of the items ensured by the specifications of section 3/4.1.3? A. Acceptable power distribution limits are maintained. B. DNBR is maintained within its assumed steady state envelope of operation. C. Power defect remains within design limits during operation. D. Moderator Temperature Coefficient remains within its analyzed range. Proposed Answer: A Explanation (Optional): distribution The movable control assemblies section of Technical Specifications ensures that: 1) acceptable power MARGIN is maintained. 3) potential effects of rod limits are maintained ("A" correct). 2) minimum SHUTDOWN "C" and "D" wrong). "B" is the basis of section misalignment on associated accident analysis are limited ("B", 3/4.2.5 "DNB Parameters". "D" is a basis for section 3/4.1.1.4 "Minimum Temperature for Criticality". Technical Reference(s): Tech.Spec. 3/4.1.3 Basis (Attach if not previously provided) Proposed references to be provided to applicants during examination: None the Learning MC-05478 Describe the major administrative or procedural precautions and limitations placed on Objective: operation, of the Rod Control System, and the basis for each. Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.43.2 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
I REACTIVITY CONTROL SYSTEMS August 27, 2001 BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES I The specifications of this section ensure that: (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control MILLSTONE - UNIT 3 B 3/4 1-3 Amendment No. JZ, 00, Ah, 7J7, ,Yf7, XPO, 197
REACTIVITY CONTROL SYSTEMS LBDCR 3-17-01 February 14, 2002 BASES MOVABLE CONTROL ASSEMBLIES (Continued) rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within +12 steps at 24, 48, 120, and fully withdrawn position for the Control Banks and 18, 210, and fully withdrawn position for the Shutdown Banks provides assururances that the Digital Rod Position Indicator is operating correctly over the full range of indication. Since the Digital. Rod Position Indication System does not indicate the actual shutdown rod position between 18 step.s and 210 steps, only points in the indicated ranges are picked for verification of agreement with demanded position. The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of peaking, factors and a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation. In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation. The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with Tavg greater than or equal to 551'F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions. Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours with more fre-quent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCOs are satisfied. The Digital Rod Position Indication (DRPI) System is defined as follows:
- Rod position indication as displayed on DRPI display panel (MB4), or
- Rod position indication as displayed by the Plant Process Computer System With the above definition, LCO, 3.1.3.2, "ACTION a." is not applicable with either DRPI display panel or the plant process computer points OPERABLE.
The plant process computer may be utilized to satisfy DRPI System requirements which meets LCO 3.1.3.2, in requiring diversity for determining digital rod position indication. Technical Specification SR 4.1.3.2 determines each digital rod position indicator to be OPERABLE by verifying the Demand Position Indication System and the DRPI System agree within 12 steps at least once each 12 hours, except during the time when the rod position deviation monitor is inoperable, then compare the Demand Position Indication System and the DRPI System at least once each 4 hours. The Rod Deviation Monitor is generated only from the DRPI panel at MB4. Therefore, when rod position indication as displayed by the plant process computer is the only available indication, then perform SURVEILLANCE REQUIREMENTS every 4 hours. MILLSTONE - UNIT 3 B 3/4 1-4 Amendment No. fg, 0884
REACTIVITY CONTROL SYSTEMS LBDCR 3-17-01 February 14, 2002 BASES MOVABLE CONTROL ASSEMBLIES (Continued) Additional surveillance is required to ensure the plant process computer indications are in agreement with those displayed on the DRPI. This additional SURVEILLANCE REQUIREMENT is as follows: Each rod position, indication as. displayed by the plant process computer shall be determined to be OPERABLE by verifying the rod position indication as displayed on the DRPI display panel agrees with the rod position indication as displayed by the plant process computer at least once per 12 hours. The rod position indication, as displayed by DRPI display panel (MB4), is a non-QA system, calibrated on a refueling interval, and used to implement T/S 3.1.3.2. Because the plant process computer receives field data from the same source as the DRPI System (MB4), and is also calibrated on a refueling interval, it fully meets all requirements specified in T/S 3.1.3.2 for rod position. Additionally, the plant process computer provides the same type and level of accuracy as the DRPI System (MB4). The plant process computer does not provide any alarm or rod position. deviation monitoring as does DRPI display panel (MB4). For Specification 3.1.3.3, the ACTION states to immediately open the reactor trip breakers if the LCO is not satisfied, however, it is appropriate to select Data A and Data B individually, to verify satisfactory rod position indication is not available before opening the reactor trip breakers. For Specification 3.1.3.1 ACTIONS b. and c., it is incumbent upon the plant to verify the trippability of the inoperable control rod(s). Trippability is defined in Attachment C to a letter dated December 21, 1984, from E. P. Rahe (Westinghouse) to C. 0. Thomas (NRC). This may be by verification of a control system failure, usually electrical in nature, or that the failure is associated with the control rod stepping mechanism. In the event the plant is unable to verify the rod(s) trippability, it must be assumed to be untrippable and thus falls under the requirements of ACTION a. Assuming a controlled shutdown from 100% RATED THERMAL POWER, this allows approximately 4 hours for this verification. For LCO 3.1.3.6 the control rods shall be limited in insertion as defined in the Core Operating Limits Report (COLR). The BASES for the Rod Insertion Limit (RIL) is located in the COLR (Reference 3.) and the current cycle reload 50.59 evaluation. The applicable I&C calibration procedure (Reference 1.) being current indicates the associated circuitry is OPERABLE. There are conditions when the Lo-Lo and Lo alarms of the RIL Monitor are limited below the RIL, as indicated in COLR Section 2.3, Control Rod Insertion Limits. The RIL Monitor remains OPERABLE because the lead control rod bank still has the Lo and Lo-Lo alarms greater than or equal to the RIL. When rods are at the top of the core, the Lo-Lo alarm is limited below the RIL to prevent spurious alarms. The RIL is equal to the Lo-Lo alarm until the adjustable upper limit setpoint on the RIL Monitor is reached, then the alarm remains at the adjustable upper limit setpoint. When the RIL is in the region above the adjustable upper limit setpoint, the Lo-Lo alarm is below the RIL. MILLSTONE - UNIT 3 B 3/4 1-5 Amendment No. Po, 0884
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 2 Tier # I Group# I K/A # 003.AKI.19 Importance Rating 2.9 Proposed Question: Initial conditions:
- The Plant has been steady at 80% power for several days.
- The core is at middle of life conditions.
- Control bank D rods are at 170 steps.
The following sequence of events occurs:
- 1. One Shutdown Bank A rod drops into the core.
- 2. The dropped rod adds -100 pcm of reactivity.
- 3. The operators fail to take rods to MANUAL
- 4. Bank D rods withdraw 25 steps and restore Tave to program.
What would have been the results if initial control bank D rod height had been 200 steps? A. Rods would withdraw 28 steps and restore Tave to program. B. Rods would withdraw 28 steps, failing to restore Tave to program. C. Rods would withdraw 23 steps, failing to restore Tave to program. D. Rods would withdraw 18 steps and restore Tave to program. Proposed Answer: C Explanation (Optional): When the rod drops, reactor power will decrease, and Tave will decrease since steam demand has not changed. This results in rods moving out, restoring temperature. Differential rod worth for rods below 200 steps is approximately 4 pcm/step. Above 200 steps, differential rod worth is approximately 3 pcm/step, requiring rods to move out about 33 steps ("A" and "D" wrong). Automatic rod withdrawal is blocked at 223 steps by C- 11 ("C" correct, "B" wrong). Technical Differential Rod Worth Curve RE-D-02 (Attach if not previously provided) Reference(s): Monthly Reactivity Data Sheet Proposed references to be provided to applicants during examination: None Learning MC-0000 15 Given one of the below listed failures (partial or complete) of the Rod Control System, Objective: determine the effects on the system and on interrelated systems... Dropped Rod... Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement I
MONTHLY REACTIVITY DATA SHEET Millstone Unit 3 Cycle 8 Median Values for June 2002 (Sheet 1 of 1) Guideline Values: Moderator Temperature Coefficient ............. -36 porn / OF Power Coefficient ........................................... -28 porn / % Power Boron Worth Differential Boron Worth (DBW) ............................ -7.3 porn / ppm Inverse Boron Worth per % Power ............................ 4 ppm / % Power Rod Worth Approximate Differential Rod Worth (above 200 steps) -3.0 pcm / step Approximate Differential Rod Worth (125 to 200 Steps) -4.0 porn / step Boration Gallons Boric acid per ppm RCS [B] increase ........... 9.0 gal BA / ppm Gallons Boric acid per % Power reduction ............. 34 gal BA / % Power Gallons Boric acid per MWe reduction ............. 2.9 gal BA / MWe Dilution Gallons PGS per ppm RCS [B] decrease ............. 244.0 gal PGS / ppm Gallons PGS per % Power increase ............. 932 gal PGS / % Power Gallons PGS per MWe increase ............. 77.6 gal PGS / MWe Auto Makeup Reactivity Correction Factor ............ 0.893 Note: Guideline values are approximate reactivity coefficients intended for licensed operator use at Hot Full Power in performing rapid and real time assessments of reactor responses. Plant curves should be consulted for increased accuracy as time permits. Preparer/Date Reviewer/Date Approver/Date
RE-D-02 25 212-ER-01-0030, Rev. 17
'age \ of Si MP3-08-02 F Differential Rod Worth versus Steps Withdrawn Control Bank D and C in Overlap EOL HFP Equilibrium Xenon ,
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----------------- 225 25 50 75 100 125 150 175 200 0
lBANK Dl 115 140 165 190 215 lBANK C CONTROL BANK POSITION ( STEPS) Preparer/Dat~ p 3/00 Reviewer/Date Page 2 of 3 Approver/Da
Written Examination Question Worksheet Form ES-401-6 ES-401 Question # 3 Examination Outline Cross-reference: Tier # I Group# 1 K/A # 005.AK2.02 Importance Rating 2.6 Proposed Question: occurs: With the plant initially at 100% power, the following sequence of events
- 1. A Control Bank "C" Group 1 rod drops.
- 2. The crew prepares to recover the rod.
- 3. The US directs the RO to operate the ROD DISCONNECT switches in accordance with AOP 3552 "Malfunction of the Rod Drive System".
unaffected rods in Control The RO reports that he placed the lift coil disconnect switches for the the switch for the dropped Bank "C" Group 1 in the ROD DISCONNECTED position, and left rod in the ROD CONNECTED position. Which of the following actions should be taken? A. Continue to recover the dropped rod in accordance with the procedure. ROD DISCONNECTED. B. Direct the RO to also place all Control Bank "C" Group 2 rods in DISCONNECTED and C. Direct the RO to place the dropped rod disconnect switch in ROD return the group 1 rods to ROD CONNECTED. switches in ROD D. Direct the RO to place the Group 2 rods and the dropped rod disconnect DISCONNECTED and return the Group 1 rods to ROD CONNECTED. Proposed Answer: B Explanation (Optional): Step in procedure states, "EXCEPT for the dropped rod, place all the lift coil disconnect the Group 2 rods also must be switched. B switches for the affected bank to - ROD DISCONNECTED". A incorrect unaffected rods placed in ROD correct, group 2 rods to ROD DISCONNECTED. C incorrect, step states remain in DISCONNECTED. DISCONNECTED. D incorrect, group 1 rods, except dropped rod to Technical Reference(s): AOP 3552, Attachment A, step 5. None Proposed references to be provided to applicants during examination: 3552. Learning Objective: MC-03901 Describe the major action categories contained within AOP Question Source: Bank # 75456 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
3 MALFUNCTION OF THE AOP 3552 Page 6 of 10 ROD DRIVE SYSTEM Rev. 3 Attachment A Misaligned Rod NOTE
- A ROD CONTROL URGENT FAILURE (MB4C 4-8) alarm will occur during recovery unless the affected rod is in Shutdown Bank C, D, or E.
- If the affected rod is in a Control Bank, a ROD CONTROL BANKS LIMIT LO (MB4C 3-9) alarm and ROD CONTROL BANKS LIMIT LO LO (MB4C 4-9) alarm may occur during recovery and remain in alarm until the P/A converter is reset. Therefore, response to these alarms is not appropriate during this period.
- 5. Establish Conditions For Rod Alignment
- a. Verify fuse check of affected a. WHEN fuse check complete, rod - COMPLETE THEN Proceed to step 5.b.
- b. Record affected group step counter position
- c. Align control rod disconnect switches:
- 1) Unlock and Open control rod disconnect switch box (Box 3RDS-HDSBOX1, CAT 60, Key #18 in CO key locker)
- 2) Place each rod disconnect switch for the affected bank, except the misaligned rod, to the ROD DISCONNECTED position
- d. Place control rod bank SEL switch to affected bank position
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 4 Tier # I Group # I K/A # 011.EKI.0I Importance Rating 4.4 Proposed Question: Current conditions:
- The crew is responding to a large break LOCA per E-1, "Loss of Reactor or Secondary Coolant".
- RCS pressure is stable at 200 psia.
- All ECCS pumps are running.
What currently is the primary method of decay heat removal? A. Heat transfer between the RCS and the S/Gs due to subcooled natural circulation flow. B. Condensation of reflux boiling in the S/Gs, flowing back to the vessel via the hot legs. C. Injection of ECCS water from the RWST and the removal of steam/water out of the break. D. Injection of water from the CTMT sump and the removal of steam/water out of the break. Proposed Answer: C Explanation (Optional):
"A" and "B" incorrect, since during a large break LOCA, the SGs are at higher temperatures than the RCS steam/water mixture. "A" and "B" are plausible since natural circulation is a vital cooling mechanism during small break LOCAs. "C" is correct, since during large break LOCAs, the break is large enough to remove all of the decay heat from the core. "D" is wrong, since the switchover to cold leg recirculation is not performed until the RHR pumps have tripped.
Technical Reference(s): E- I background (Attach if not previously provided) ES-1.3, step 2. Proposed references to be provided to applicants during examination: None Learning MC-04912 For a Large Break LOCA, EXPLAIN Core Cooling during the (As available) Objective: 4 major stages of the Event... Question Source: Modified Bank # 70054 Parent attached Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.5, 41.8, and 41.10 Comments: 4 1 of 46 NUREG-1021, Revision 8, Supplement 1
Form ES-401-6 ES-401 Written Examination Question Worksheet Original 70054 Given the following conditions:
- A LOCA has occurred.
and Safety Injection", and E-1,
- Appropriate actions in accordance with E-0, "Reactor Trip "Loss of Reactor or Secondary Coolant", have been completed.
- ECCS is operating in cold leg recirculation mode.
- RCS pressure is stable at 200 psia.
method of decay heat removal? Which of the following statements describes the primary A. The condensation of reflux boiling in the S/Gs. circulation flow B. Heat transfer between the RCS and the S/Gs due to natural circulation flow C. Heat transfer between the RCS and the S/Gs due to forced the removal of steam/water out from D. The injection of water from the containment sump and the break. Answer: D 41 of 46 NUREG-1 021, Revision 8, Supplement I
'1 Breaks <.3/8" equivalent diameter hole Breaks in this range are considered to be leaks, rather than small LOCA's, since the normal charging system can maintain reactor coolant inventory so that RCS pressure and pressurizer level do not decrease. Very slight system depressurization may occur but no automatic trip or safety injection signal would be generated. The core will remain fully covered provided that the steam generators are available to remove energy, and makeup flow is the RCS continuously delivered to the RCS. If charging flow is not available, 2.
transient behavior would be similar to the response described for Category If the leak is within Technical Specification limits or it can be isolated, the plant could remain in power operation. If the leak is above Technical to a Specification limits and cannot be isolated, then the plant should go cold shutdown condition utilizing the normal shutdown procedures. During RCS cooldown the charging system should maintain pressurizer level and the limits. depressurization should be controlled to conform to the normal cooldown Breaks 3/8" < diameter <- 1", minimum safety injection, or Category 1 breaks above with no charging flow assumed For these break sizes the normal makeup system cannot maintain level and safety pressure. The RCS will depressurize and an automatic reactor trip and sink injection signal will be generated. Provided that a secondary side heat to the exists, the RCS will reach an equilibrium pressure which corresponds pumped pressure at which the liquid phase break flow equals the high pressure safety injection flow. It has been verified that this equilibrium pressure effect condition will be established for plants with charging/SI pumps. This is described here by the presentation of a specific plant analysis for break sizes within this range. A general description of system behavior applicable to the sample transient is provided first, then specific comments concerning the sample analysis are provided. E-1 5 HP-Rev. 1 6998B:lb
Early in the transient a loss of subcooled liquid in the RCS occurs which results in a moderate depressurization to the pressure which corresponds to saturation pressure in the core and hot legs. At this point the upper head, upper plenum, hot legs, and core begin to experience some slight voiding, but more than enough liquid flow exists through the core to keep it covered and cooled. During this period of voiding, however, RCS depressurization occurs at a much slower rate than during the time when the entire system was subcooled. Eventually the RCS depressurizes to the point of the reactor trip signal. Immediately following reactor trip, the RCS rapidly depressurizes, since only a fraction of the heat previous to trip is now being transferred to the primary fluid. Due to this rapid depressurization follbwing reactor trip, a safety injection signal is quickly generated. Within a few minutes of the reactor trip time, an equilibrium pressure is established which is above the steam generator pressure. The fluid conditions in the RCS at the time of equilibrium pressure establishment may be characterized by slight voiding in the core and upper plenum and hot legs, and saturated or slightly subcooled liquid in the cold legs. Core heat is removed through the steam generators by continuous single or two-phase natural circulation. The primary mixture level in the steam generators does not drain for breaks of this size, and the core remains covered throughout the entire transient provided that SI is not interrupted. Once equilibrium pressure is established there is no further net loss of liquid volume in the RCS. The natural circulation heat removal mode continues until the time that the break can remove all the decay heat ( 1 day for a 1" break). Prior to this time, auxiliary feedwater is required to maintain the heat sink. Since the equilibrium pressure established is determined by means of a volume balance of SI flow and break flow, the 0P and AT from primary to secondary side, together with the cold safety Injection water, may provide a total heat sink greater than the decay heat generated and a cooling of the primary fluid can occur. E-1 6 HP-Rev. 1 6998B:lb
would correspond to an event which is approaching the design basis assump-tions. It should be noted that more probable delivery rates for the safety injection system (e.g., maximum safeguards or minimum safeguards with no spilling line) will yield less core uncovery or no core uncovery. For break locations other than at the cold leg, little or no core uncovery is calculated. For breaks in the crossover leg there could be an uncovery similar to the uncovery experienced for the cold leg break when steam was vented through the crossover legs. Since for crossover leg breaks safety injection is injected into all RCS cold legs and steam does not have to pass through the broken loop RCP, there is no subsequent core uncovery. For breaks in the RCS hot leg or pressurizer vapor space, steam is vented earlier then for other locations (immediately for vapor space breaks) so that essentially no core uncovery is experienced. The method used for long-term plant recovery for breaks in this category depends upon the RCS pressure at the time that the operator determines if further RCS cooldown and depressurization is required (step 13 in the E-1 guideline) . If at this time the RCS pressure is greater than the low-head S: pumps shutoff head pressure, ES-1.2, POST LOCA COOLDOWN AND DEPRESSURIZATION. is used for long-term plant recovery. Even if the RCS pressure is less than the shutoff head pressure of the low-head SI pumps but it cannot be verifiec that the low-head SI pumps are injecting flow into the RCS, the operator should transfer to ES-1.2 for long-term plant recovery. The operator would stay in E-1 if the RCS pressure is less than the shutoff head pressure of tne low-head SI pumps and flow into the RCS from the low-head SI pumps has been verified. Large Break LOCA, 1 FT2 < Area < Double-Ended, Minimum Safeguards A large break LOCA is the design basis for many aspects of the NSSS design. Some of the major design considerations impacted by the large break LOCA are peak core power, containment sizing, and loop/vessel internal forces. In order to describe the large break LOCA hydraulic transient a typical SAR safety analysis for a 4-loop plant will be utilized. The phenomena descrioec here are similar for all Westinghouse plants. E-1 27 HP-Rev. 1 6998B:lb
A large break LOCA (one square foot total area up to the double-ended break) has four characteristic stages: blowdown, refill, reflood, and long-term recirculation. Blowdown starts with the assumed initiation of the LOCA and ends when the reactor coolant system pressure (initially 2250 psig) falls to essentially that of the containment atmosphere. Refill starts at the end of blowdown and ends when the addition of emergency core cooling water fills the bottom of the reactor vessel and reaches the elevation of the bottom of the fuel rods. Reflood is defined as the time from the end of refill until the reactor vessel has been filled with water to the extent that core temperature rise has been terminated and core temperatures subsequently have been reduced to their long-term steady-state levels associated with dissipating decay heat. These time divisions are established mainly for analytical convenience. As contrasted with the large break, the blowdown phase of the small break occurs over a longer period and does not result in reduction of the effective water level in the reactor vessel below the bottom of the core. Thus, for the small break LOCA there are only three characteristic stages, i.e., a gradual blowdown in which the decrease in water level is checked before the bottom of the core is uncovered (and before pressure equilibrium between reactor and containment is reached), reflood, and long-term recirculation. Figure 11 illustrates the primary system pressure transient occurring during the blowdown phase of the LOCA for a double-ended cold leg guillotine break. The primary pressure rapidly drops from an initial value of 2250 psig to a low value of 40-50 psig by the end of blowdown (-1/2 minute). Also shown on Figure 11 is the break flow transient occurring during blowdown. The break flow starts at a very high value (critical flow, -70,000 lbm/sec) and is reduced to zero by the end of blowdown. Figure 11 also includes a plot of the SI accumulators mass flow rate. Note that accumulator flow is initiated approximately 16 seconds after the break occurs. This corresponds to the time when the RCS pressure has decreased to 600 psig, which corresponds to a minimum accumulator pressure set point. E-1 28 HP-Rev. 1 6998B:lb
The containment pressure transient is shown on Figure 12. As shown, the containment pressure reaches a peak value early in the transient during the blowdown phase of the transient. A safety injection signal will be initiated on a containment High-l pressure signal in a matter of seconds after the break and containment spray may be initiated on a containment High-3 pressure signal depending on the magnitude of the break and the specific plant's containment design. The important hydraulic transient parameters during the reflood phase are downcomer water level (ZD), core water level (ZC), and the core inlet flooding rate (VIN) as shown in Figure 12. During refill the ECCS cooling water from the SI accumulators and safety Injection pumps enters the top of the reactor vessel downcomer annulus and starts to fill the reactor vessel lower plenum, which is filled after 45 seconds. This is commonly called bottom of core (BOC) recovery time. After BOC occurs, the downcomer annulus starts to fill rapidly and thus provides a. static head for pushing cooling water into the core. A core inlet flooding rate (inches/second) is established and the water starts to move up into the core, thus providing the mechanism for core cooling during reflood. Table 1 presents a time sequence of events for the double-ended cold leg guillotine break. After successful initial operation of the ECCS, the reactor core is once again covered with borated water. This water has enough boron concentration to maintain the core in a shutdown condition. Decay heat is removed by a continuous supply of water from the ECCS. This supply initially comes from the refueling water storage tank (RWST). When the RWST level reaches the switchover setpoint the ECCS pumps are transferred into the recirculation mode (using ES-1.3, TRANSFER TO COLD LEG RECIRCULATION) wherein water is drawn from the containment sump and is cooled in the residual heat removal heat exchangers. Thus, long-term cooling of the core is maintained by the ECCS in sump recirculation mode. The core is maintained in a shutdown state by borated water. E-l 30 HP-Rev. lA 6998B:lb
EOP 35 ES-1.3 Page 3 of 14 q-TRANSFER TO COLD LEG RECIRCULATION Rev. 010 STP ACTION/EXPECTED RESPONS RESPONSE NOT OBTAINED E CQ A U T I 0 N
- SI recirculation flow to RCS must be maintained at all times.
- If offsite power is lost after SI reset, manual action to restart safeguards equipment may be required.
- High radiation levels may be experienced in the Auxiliary and ESF buildings following the transfer to cold leg recirculation. I NOTE Functional Response procedures should not be implemented until completion of step 5. I
.1. RESET ESF Actuation Signals, If Required
- SI
- CDA
- LOP
- CIA
- CIB
- 2. STOP Both RHR Pumps And Place Control Switches In PULL-TO-LOCK
Form ES-401-6 ES-401 Written Examination Question Worksheet Question # 5 (SRO) Examination Outline Cross-reference: I Tier # Group# I K/A # W/E04.EA2.01 Importance Rating 4.3 Proposed Question: With the plant at 100% power, the following sequence of events occurs:
- 1. The reactor trips and safety injection actuates.
to and stabilizes at 1600 psia.
- 2. Over the next 10 minutes, RCS pressure decreases Outside Containment".
- 3. The crew is responding using ECA- 1.2, "LOCA valve the crew closes is the "A" RHR pump
- 4. While attempting to isolate the break, the final cold leg injection valve 3SIL*MV8809A.
starts increasing.
- 5. After 3SIL*MV8809A closes, RCS pressure from ECA- 1.2?
Which procedure will the crew transition to A. E- 1 "Loss of Reactor or Secondary Coolant". B. ES-1.1 "SI Termination". C. ES-1.2 "Post LOCA Cooldown and Depressurization". Recirculation". D. ECA- 1.1 "Loss of Emergency Coolant Proposed Answer: A Explanation (Optional): "A" is correct, and the break is isolated since RCS pressure increases. JUSTIFICATION: It can be determined that subsequently to ES- 1.1. break isolated, the crew transitions to E- I and "B" is wrong, but plausible, since with the "C" is wrong because there is this procedure will be used to terminate SI.
"B" is plausible since with the leak isolated, ES-1.2 is designed to mitigate plausible since the break size is small, and no entry into ES-1.2 from ECA-1.2. "C" is to ECA- .1.
the break is isolated, the crew would transition small break LOCAs. "D" is wrong, since if ECA-1.2, steps 4 and 5 (Attach if not previously provided) Technical Reference(s): during examination: None Proposed references to be provided to applicants (As available) to other procedures Learning 03878, Discuss conditions which require transition Objective: from EOP 35 ECA-1.2. Question Source: Modified Bank # 75669 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.10 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
Form ES-401-6 ES-401 Written Examination Question Worksheet Original question 75669 sequence of events occurs: With the plant at 100% power, the following
- 1. The reactor trips and safety injection actuates.
to and stabilizes at 1600 psia.
- 2. Over the next 10 minutes, RCS pressure decreases Outside Containment."
- 3. The crew is responding using ECA-1.2, "LOCA valve the crew closes in attempt to isolate the
- 4. While attempting to isolate the break, the final (3SIL*MV8809A).
leak is RHR pump "A" cold leg injection valve remains stable.
- 5. After 3SIL*MV8809A closes, RCS pressure from ECA-1.2?
Which procedure will the crew transition to A. E-1 "Loss of Reactor or Secondary Coolant". B. ES-1.1 "SI Termination". C. ES-1.2 "Post LOCA Cooldown and Depressurization". D. ECA- 1.1 "Loss of Emergency Coolant Recirculation". Answer: D 41 of 46 NUREG-1021, Revision 8, Supplement 1
EOP 35 ECA-1.2 Page 6 of 7 LOCA OUTSIDE CONTAINMENT Rev. 007 ACTION/EXPECTED RESPONS STEP RESPONSE NOT OBTAINED
- 4. Try To Identify And Isolate Break
- a. Tmrn the power lockout switch to ON for the following valves (MB2R):
- RHR pump A (3SIL*MV8809A)
- RHRpump B (3SIL MV8809B)
- SI injection (3SIH*MV8835)
- b. CLOSE o of the following:
- RHR pump A cold leg injection valve (3SIL*MV8809A)
- RHR pump B cold leg injection valve (3SIL*MV8809B)
- SI cold leg injection valve (3SIH*MV8835)
- c. Check RCS pressure - c. Perform the following:
INCREASING
- 1) OPEN valve closed in step 4.b.
- 2) IF all lines have been checked, THEN Proceed to step 4.d.
- 3) Return to step 4.b.
)
LOCA OUTSIDE EOP 35 ECA-1.2 Page7of77 CONTAINMENT Rev. 007 I
- 4. (continued)
- d. Turn the power lockout switch to OFF for the following valves (MB2R):
- RHR pump A (3SIL MV8809A)
- RHR pump B (3SIL*MV8809B)
- SI injection (3SIH*MV8835)
- 5. Check If Break Is Isolated
- a. Check RCS pressure - a. Go to ECA-1.1, Loss of INCREASING Emergency Coolant Recirculation.
- b. Go to E-1, Loss of Reactor or Secondary Coolant
-FINAL-
Form ES-401-6 ES-401 Written Examination Question Worksheet Question # 6 (SRO) Examination Outline Cross-reference: Tier # I Group # K/A # W/EO 1.EK 1.03 Importance Rating 3.5 Proposed Question: Initial conditions: SG.
- The plant tripped due to a fault in the "A"
- The crew has isolated the SG per E-2 "Faulted Steam Generator Isolation".
- The crew has just transitioned to ES-i.1 "SI Termination".
The following sequence of events occurs:
- 1. RCS Pressure starts decreasing.
go into alarm.
- 2. Numerous Auxiliary Building radiation monitors
- 3. The US desires to enter ES-0.0 "Rediagnosis".
take? Based on current conditions, what action will the crew actuated. A. The crew will NOT enter ES-0.0, since SIS has already been exited. B. The crew will NOT enter ES-0.0, since E-0 has already to ECA-1.2 "LOCA Outside Containment". C. The crew will perform ES-0.0 and then transition E-2 "Faulted Steam Generator Isolation". D. The crew will perform ES-0.0 and then transition to Proposed Answer: C E-0 ("B" wrong). Explanation (Optional): Rediagnosis can only be used if SI is actuated ("A" wrong) after exiting and the faulted SG has already been isolated ("D" wrong) ES-0.0 will send the crew to an E-I series procedure since a SGTR is not occurring ("C" correct). OP 3272, section 1.2, pg 6 (Attach if not previously provided) Technical Reference(s): ES-0.0 Rediagnosis examination: None Proposed references to be provided to applicants during can be used. (As available) Learning MC-0445 1 Discuss the conditions under which ES-0.0 ES-0.0. Objective: MC-06275 Describe the major action categories within Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.8,41.10 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
ES-0.0, "Rediagnosis," is entered solely based on Operator judgement if SI is actuated. This is a unique procedure which may be used as an aid to I determine the correct path through the EOP Network after exiting E-0, "PReactnr Trin nr Safety Injection," with SI actuated.
- - - - - - - J, -- - -- - -. , -- J OP 3272 Level of Use l . A- Rev. 008 Information .. ... 6of 44
(9 ESPONSE NOT OBTA I NOl This procedure should only be use( I if SI has been actuated.
- 1. Check At Least One SG Is Intact
- a. Check pressures in all SGs - a. IF a controlled cooldown is in ANY STABLE OR progress, INCREASING THEN Proceed to step 2.
IF a controlled cooldown is NOT in progress, THEN Perform the applicable action:
- IF main steam lines are NOT isolated, THEN Go to E-2, Faulted Steam Generator Isolation.
OR
- IF main steam lines are isolated, THEN Go to ECA-2.1, Uncontrolled Depressurization of All Steam Generators.
SP ACTION/EXPECTED RESPONS RESPONSE NOT OBTAINED
- 2. Check If All SGs Are Intact I
- a. Check pressures in all SGs a. Verify each faulted SG isolated
- NO SG PRESSURE
- Main steam line DECREASING IN AN UNCONTROLLED
- Main feed line MANNER
- SG blowdown line
- NO SG COMPLETELY DEPRESSURIZED
- SG blowdown sample line I
- SG chemical feed line
- Auxiliary feed line
- Steam supply to TD AFW pump
- Main steam line drains upstream of MSIVs and TD AFW pump I
- SG atmospheric dump lines IF the faulted SGs are NOT isolated, THEN Go to E-2, Faulted Steam Generator Isolation.
RESPONSE NOT OBTAINED ACTION/EXPECTED RESPONSH STEP I n ' AP Rxpintured Go to one of the following: 1-I ai~ r- - Ai. t'Inef'K U1 ;OJ kuxst;>as
- E-1, Loss of Reactor or
- ANY SO LEVEL Secondary Coolant INCREASING IN AN UNCONTROLLED MANNER OR
- ANY SG WITH HIGH
- ECA-1.1, Loss of Emergency RADIATION (by trend Coolant Recirculation history, alarm status, or sample results)
- ECA-1.2, LOCA Outside Containment
- 4. Go To One Of The Following:
- E-3, Steam Generator Tube Rupture OR
- ECA-3.1, SGTR With Loss of Reactor Coolant OR
- ECA-3.2, SGTR With Loss of Reactor Coolant Saturated Recovery Desired OR
- ECA-3.3, SGTR Without Pressurizer Pressure Control
-FINAL-
Written Examination Question Worksheet Form ES-401-6 ES-401 Question # 7 (SRO) Examination Outline Cross-reference: Tier # I Group # I K/A # 015/17.AK1.02 Importance Rating 4.1 Proposed Question: Current Conditions:
- The plant is at 30% power.
3.
- "RCP HI RANGE LKG FLOW HI" annunciator is lit on Main Board
* "B" RCP #1 seal leakoff flow is 6.8 gallons per minute. 0
- Both "B" RCP #1 seal inlet temperature indicators read 113 F and increasing.
What actions are the crew directed to take? valve after the pump A. Trip the reactor, stop the "B" RCP, and close its number 1 seal leakoff has been stopped for between 3 to 5 minutes. service within 8 hours. B. Commence an orderly plant shutdown, and remove the "B" RCP from the pump for continued C. Continue to monitor the "B" RCP, and request Engineering to evaluate operation. initiate action to D. Transition to AOP 3554 "RCP Trip or Stopping an RCP at Power" and perform an immediate shutdown of the "B" RCP. Proposed Answer: D Explanation (Optional): With RCP #1 seal leakoff flow >6 gpm AND #1 seal inlet temperatures with power less than increasing, MB3B 2-10 (Leakoff Flow high) directs the crew to AOP 3554 is wrong). "A" is 37% (P-8), since the RCP can be stopped without tripping the unit ("A" requires tripping the plausible since this would be the correct action if power was >P-8. OP 3554 trip ("D" is correct). "B" RCP and closing the seal leakoff valve at least 3 minutes after the pump temperature is stable. is wrong, but plausible, since these actions would be required if seal inlet flow was < 6 gpm with "C" is wrong, and plausible, since these actions are required if seal leakoff seal inlet temperatures stable. Technical Reference(s): OP3353.MB3B 2-10 (Attach if not previously provided) None Proposed references to be provided to applicants during examination: Learning Objective: MC-03905 IDENTIFY plant conditions that require entry into AOP-3554. Question Source: Bank # 64317 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.10 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
Setpoint: Greater than 5.7 gpm 2-10 RCP Hi RANGE LKG FLOW HI LI AUTOMATIC FUNCTIONS
- 1. None CORRECTIVE ACTIONS
- 1. CHECK the following to confirm alarm and determine affected RCP:
flow
- 3CHS-FR158 and 3CHS-FR160, high range RCP No. 1 seal leakoff recorders (MB3)
- CHS-F161*, RCP A No. 1 seal leakoff flow computer point
- CHS-F16O*, RCP B No. 1 seal leakoff flow computer point
- CHS-F159*, RCP C No. 1 seal leakoff flow computer point
- CHS-F158*, RCP D No. 1 seal leakoff flow computer point
- 2. DISPLAY "RCP Status" NSSS, picture 15.
- 3. VERIFY leakage flow high indication by observing the following indications:
- Seal injection flow
- Affected RCP #1 seal inlet temperatures
- VCT level
- Charging header flow
- Pressurizer level flow
- 3CHS-FR158 and 3CHS-FR160, high range RCP No. 1 seal leakoff recorders (MB3)
A 3CHS-PI 124, excess lID Hx outlet pressure OP 3353.MB3B Level of UseRe.062 ContinuousRe.06 2
RCP, and Go To I
- 4. Using Table 1, EVALUATE plant conditions for the affected indicated Step.
OP 3353.MB3B Level of Use Rev. 006-02 Continuous III, WI
2-10
- 5. PERFORM the following:
5.1 TRIP reactor. 5.2 STOP affected reactor coolant pumps. 5.3 WHEN affected RCP has been tripped for between 3 and 5 minutes, CLOSE affected RCP No. 1 seal leakoff isolation valve. 5.4 Go To E-0, "Reactor Trip or Safety Injection."
- 6. Go To AOP 3554, "RCP Trip or Stopping an RCP at Power," and INITIATE actions to perform an immediate RCP shutdown.
- 7. PERFORM the following to remove affected RCP from service within 8 hours:
7.1 IF reactor power is greater than 25%, Refer To OP 3204, "At Power Operation," and COMMENCE an orderly plant shutdown while continuing with this step. 7.2 IF reactor power is less than or equal to 25%, Refer To OP 3206, "Plant Shutdown," and COMMENCE an orderly plant shutdown while continuing with this step. 7.3 IF, at any time, RCP No. 1 seal parameters degrade, IMPLEMENT steps as specified in Table 1. 7.4 WHEN in MODE 3, Go To OP 3301D, "Reactor Coolant Pump Operation," and STOP the affected reactor coolant pump.
- 8. PERFORM the following:
8.1 NOTIFY Duty Officer of alarm condition. 8.2 IF "VCT TEMP HI" (MB3A 5-10) is lit, Refer To OP 3353.MB3A 5-10, "VCT TEMP HI." 8.3 REQUEST Engineering Department evaluate continued pump operation. 8.4 IF at any time, affected RCP no. 1 seal parameters degrade, IMPLEMENT steps as specified in Table 1.
- 9. IF total seal return flow from all four RCPs exceeds 16 gpm, Refer To 3TRM-7.4, Section I, "Fire Related Safe Shutdown Components," and PERFORM ACTION 2.
Level of Use ,.- OP 3353.MB3B Continuous Rev. 006-02 40 of 97
2-10 SUPPORTING INFORMATION
- 1. Initiating Device (1 of 4) Setpoint 1.1 FY/161 1.1 > 5.7 gpm 1.2 FY/160 1.2 > 5.7 gpm 1.3 FY/159 1.3 > 5.7 gpm 1.4 FY/158 1.4 > 5.7 gpm
- 2. Computer Points 2.1 CHS-F161*, RCP A No. 1 seal leakoff flow 2.2 CHS-F160*, RCP B No. 1 seal leakoff flow 2.3 CHS-F159*, RCP C No. 1 seal leakoff flow 2.4 CHS-F158*, RCP D No. 1 seal leakoff flow 2.5 CHS-T172, RCP 1 No. 1 seal inlet temperature 2.6 CHS-T173, RCP 1 No. 1 seal inlet temperature 2.7 CHS-T170, RCP 2 No. 1 seal inlet temperature 2.8 CHS-T171, RCP 2 No. 1 seal inlet temperature 2.9 CHS-T168, RCP 3 No. 1 seal inlet temperature 2.10 CHS-T169, RCP 3 No. 1 seal inlet temperature 2.11 CHS-T166, RCP 4 No. 1 seal inlet temperature 2.12 CHS-T167, RCP 4 No. 1 seal inlet temperature
- 3. Possible Causes 3.1 Loss of seal injection 3.2 Seal injection water high temperature 3.3 RCP No. 1 seal damage 3.4 RCP No. 1 seal ring mispositioned
- 4. Procedures 4.1 E-0, "Reactor Trip or Safety Injection" 4.2 OP 3204, 'At Power Operation" 4.3 OP 3206, "Plant Shutdown" 4.4 AOP 3554, "RCP Trip or Stopping and RCP at Power" 4.5 OP 3353.MB3A, "Main Board 3A Annunciator Response"
- 5. Control Room Drawings 5.1 ESK 10HB 5.2 LSK 26-2.6B Level of Use Rev.3353.MBoB O 006 -02 Continuous 4 f9
Written Examination Question Worksheet Form ES-401-6 ES-401 Question # 8 Examination Outline Cross-reference: Tier# 1I Group # I K/A # W/EIO.EK1.02 Importance Rating 3.6 Proposed Question: in Vessel (without What strategy is used in ES-0.4, "Natural Circulation with Steam Void RVLMS)," to prevent excessive void growth? RCS temperature is held A. During RCS depressurization steps where void growth is possible, constant, and charging and letdown flows are equalized. pressure is held constant, B. During RCS cooldown steps where void growth is possible, RCS and charging and letdown flows are equalized. temperature. C. RCS Cooldown rate is limited to 50'F/hr based on WR cold leg exit TCs. D. RCS subcooling is limited to a minimum of 132TF based on core Proposed Answer: A Explanation (Optional): when RCS pressure drops less than saturation "A" is correct, since void growth occurs during depressurization steps all PZR level increases are due to void growth, pressure for head temperature. With charging and letdown matched, steps, PZR level is maintained stable via and PZR level it kept < 91%. "B" is wrong, since during cooldown 0 rate limit in ES-0.4 is 80 F/hr. "C" is plausible, charging to make up for shrinkage. "C" is wrong, since cooldown to 50'F/hr to prevent void formation. "D" is since ES-0.2 "Natural Circulation Cooldown" limits cooldown rate steps in ES-0.4. "D" is plausible, since wrong, since subcooling is maintained via discrete cooldown depressurization 0 ES-0.2 limits subcooling to 132 F to prevent void formation. Technical Reference(s): ES-0.4, steps 12-15 None Proposed references to be provided to applicants during examination: Natural Circulation Cooldown Learning MC-05943 Describe the major action categories within EOP 35 ES-0.4, Objective: with Steam Voids in Vessel (w/o RVLMS). Question Source: Bank # 67599 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.8, 41.10 Comments: 41 Of 46 NUREG-1021, Revision 8, Supplement 1
NATURAL CIRCULATION EOP 35 ES-0.4 Page 12 of 18 COOLDOWN WITH STEAM VOID Rev. 011 IN VESSEL (WITHOUT RVLMS) STP ACTION/EXPECTED RESPONS RESPONSE NOT OBTAINED I NOTE To continue overall system depressurization, it may be necessary to cycle PZR level (pressure) to enhance upper head cooling.
- 12. Check PZR Level - Perform the following:
LESS THAN 90% I
- a. Increase RCS pressure 100 psi using PZR heaters.
- b. Return to step 11.
- 13. Decrease RCS Hot Leg WR Temperatures To 400'F
-a. Maintain cooldown rate in RCS cold legs -
LESS THAN 800F/hr
- b. Maintain RCS pressure -
STABLE
- c. Maintain RCS temperature and pressure in the acceptable operation region of Technical Specification Figure 3.4-3, RCS Cooldown Limitations
- d. Maintain stable PZR level using charging
__e. Check RCS hot leg e. Return to step 13.a. WR temperatures - LESS THAN 400 0F
- f. Stop RCS cooldown
NATURAL CIRCULATION EOP 35 ES-0.4 Page 13 of 18 COOLDOWN WITH STEAM VOID Rev. 011 IN VESSEL (WITIHOUT RVLMS) ACTIOIEXPCTEDRESPNSEHRESPONSE NOT OBTAINEDl
- 14. Equalize Charging And Letdown Flows
- a. Place charging controls in manual
- b. Adjust charging and seal injection flows to equal letdown and seal leakoff flows
- 15. Depressurize RCS
-a. Verify letdown - a. Perform the following:
IN SERVICE
- 1) OPEN one PZR PORV.
- 2) Depressurize the RCS until one of the following is satisfied:
- RCS pressure is LESS THAN 650 psia OR
- PZR level is GREATER THAN 90%
- 3) CLOSE the PORV.
- 4) Proceed to step 16.
- b. Unlock and OPEN auxiliary spray valve (3RCS*AV8145)
_ c. CLOSE charging header loop isolation valves
- 3CHS*AV8146
- 3CHS*AV8147
NATURAL CIRCULATION EOP 35 ES-0.4 Page 14 of 18 COOLDOWN WITH STEAM VOID Rev. 011 IN VESSEL (WITHOUT RVLMS) ACTIO/EXPETED RSPON RESPONSE NOT OBTAINEDl
- 15. (continued)
- d. Throttle the charging flow control valve to adjust and maintain auxiliary spray flow
- e. Check REGEN HX e. Proceed to step 15.g. and, LETDOWN TEMP HI (395 F) (MB3A 5-4) IF the annunciator actuates, annunciator - LIT THEN Perform step 15.f.
- f. OPEN one charging header loop isolation valve
- g. Depressurize RCS until one of the following conditions is satisfied:
- RCS pressure -
LESS THAN 650 psia OR
- PZR level-GREATER THAN 90%
- h. Stop RCS depressurization
- 1) OPEN one charging header loop isolation valve
- 2) CLOSE auxiliary spray valve (3RCS*AV8145)
- 16. Check PZR Level - Perform the following:
LESS THAN 90Yol
- a. Increase RCS pressure 100 psi using the PZR heaters.
- b. Return to step 15.
Written Examination Question Worksheet Form ES-401-6 ES-401 Question # 9 Examination Outline Cross-reference: Tier # I Group # K/A # 024.AK2.04 Importance Rating 2.5 Proposed Question: With the plant at 100% power, the following sequence of events occurs:
- 1. An unexplained positive reactivity addition event occurs.
- 2. The crew enters AOP 3566 "Immediate Boration".
- 3. The RO reports that neither boric acid pump will start.
- 4. The RO opens both gravity feed boration valves.
Return) is 80
- 5. The RO reports that net charging flow (Charging + Seal Injection - RCP Seal gpm.
What action is the crew required to take? A. Open emergency boration valve 3CHS*MV8104. B. Decrease net charging flow to less than 75 gpm. than 100 gpm. C. Throttle open the charging line flow control valve to increase flow to greater SI valve. D. Open both RWST to charging pump suction valves and one charging header Proposed Answer: B Explanation (Optional): to < 75 gpm to prevent the loss "B" is correct since when neither boric acid pump will start, charging flow is limited since this action is taken if the of charging pump suction due to limited suction source. "A" is wrong, but plausible, since flow is aligned to the crew had been able to start a boric acid pump. "C" and "D" are wrong, but plausible, RWST and increased to > 100 gpm if boration flow is < 33 gpm. Technical Reference(s): AOP 3566, steps I and 3. (Attach if not previously provided) N one Proposed references to be provided to applicants during examination: (As available) Learning MC-03961, Describe the major action categories contained within Objective: AOP-3566, Immediate Boration. Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7, 41.8 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
H RESPONSE NOT OBTAINEDll STP ACTION/EXPECTED RESPONS Of 1.Initiate Immediate Boration RCS
- a. START one charging pump.
- a. Check one charging pump RUNNING
- b. Perform the following:
- b. Align boration path: one gravity o 1) OPEN at leastvalve.
one boric feed boration
- 1) START at least acid transfer pump flow to
- 2) Limit net chargingTHAN the RCS to LESS
- 2) OPEN emergency 75 gpm (charging + seal boration valve 4 injection - RCP seal (3CHS*MV810 ) return).
- 3) CLOSE at least one VCT outlet isolation valve.
- c. Position valves as required.
- c. Check normal charging flow IF the normal charging flow
-path aligned: path can NOT be aligned, N
- 1) Charging flow control Proceed to step 2.
valve - THROTTLED OR OPEN
- 2) Charging header loop isolation valve4 6 (3CHS*AV814 7 or 3CHS*AV81 ) - OPEN
- 3) Charging header isolation 06 valves (3CHS*MV81 and 3CHS*MV8105) -
OPEN
- d. Place charging flow control valve in MANUAL
- e. Proceed to step 3.
E RESPONSE NOT OBTAINEDl STE ACTION/EXPECTED RESPONS
- 3. Verify Boration Flowl
- a. Check the PZR PORVs and
- a. Check PZR pressure - block valves are open.
LESS THAN 2350 psia IF the PORVs or block valves are closed, THEN OPEN as necessary until PZR pressure is LESS THAN 2150 psia.
- b. Perform the following:
- b. Adjust boration flow to the RCS -EQUAL TO OR 1) OPEN at least one RWST GREATER THAN 33 gpO to charging pump suction isolation valve.
- 2) IF suction flow path from a BAT is open, THEN Close the flow path.
- 3) CLOSE at least one VCT outlet isolation valve.
- 4) Adjust charging flow to obtain EQUAL TO OR GREATER THAN 100 gpm to the RCS using one of the following:
- Normal charging header flowpath
- Charging header SI flowpath (3SIH*MV8801A or 3SIH*MV8801B)
Written Examination Question Worksheet Form ES-401-6 ES-401 Examination Outline Cross-reference: Question # 10 Tier # I Group # I K/A # 026.AA2.05 Importance Rating 2.5 Proposed Question: on MB 1. With the plant at 100% power, an "RPCCW SPLY FLOW HI" annunciator is received the RO The crew enters AOP 3561 "Loss of Reactor Plant Component Cooling Water", and reports RPCCW flow rates and surge tank levels are as follows:
"A" Train "B" Train Safety Header: 0 gpm 2000 gpm Non-Safety Header: 2000 gpm 2000 gpm CTMT Header: 700 gpm 1450 gpm and increasing Surge Tank Level 92.5% and decreasing (Both trains) source of the Have RPCCW heat exchanger flow limits been exceeded, and what is a potential leak?
is the "C" A. The heat exchanger flow limits have been exceeded. A potential source of leakage RCP Thermal Barrier Heat Exchanger. is the Seal B. The heat exchanger flow limits have been exceeded. A potential source of leakage Water Heat Exchanger. is the C. The heat exchanger flow limits have not been exceeded. A potential source of leakage "C" RCP Thermal Barrier Heat Exchanger. is the D. The heat exchanger flow limits have not been exceeded. A potential source of leakage Seal Water Heat Exchanger. Proposed Answer: D is below the limit of 8100 Explanation (Optional): The "B" RPCCW heat exchanger flow rate is 5450 gpm, which exchanger and the "D" RCP are cooled from the "B" Train gpm ("A" and "B" wrong). Both the seal water heat is greater than RPCCW pressure, which would result in surge RPCCW CTMT Header, but thermal barrier pressure tank level increasing ("C" wrong, "D" correct). Technical Reference(s): OP3353.MB IC, 4-7 (Attach if not previously provided) P&ID 121A and B Proposed references to be provided to applicants during examination: None MC-04 154 Describe the operation of the RPCCW system under the (As available) Learning Objective: following:... CCP System Leak... 41 of 46 NUREG-1021, Revision 8, Supplement 1
Written Examination Question Worksheet Form ES-401-6 ES-401 Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.8 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
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10 Setpoint: Safety: greater than 7,435 gpm 4-7 Non-safety A: greater than 6,550 gpm Non-safety B: greater than 5,375 gpm Containment A: greater than 1,964 gpm RPCCW Containment B: greater than 1,423 gpm HEADER SPLY FLOW Hi AUTOMATIC FUNCTIONS
- 1. None CORRECTIVE ACTIONS
- 1. CHECK the following and DETERMINE source of excessive RPCCW demand (MBI1):
- 3CCP-FI 11A, "RPCCW SAFETY HDR FLOW"
- 3CCP-FI 11B, "RPCCW SAFETY HDR FLOW"
- 3CCP-FI 12A, "RPCCW NON-SAFETY HDR FLOW"
- 3CCP-FI 12B, "RPCCW NON-SAFETY HDR FLOW"
- 3CCP-FI 15A, "RPCCW CTMT SPLY FLOW"
- 3CCP-FI 15B,"RPCCW CTMT SPLY FLOW"
- 3CCP-LI 20A, "RPCCW SURGE TK LVL A'
- 3CCP-LI 20B, "RPCCW SURGE TK LVL B"
- 2. IF total RPCCW flow is greater than 8,100 gpm, REDUCE total RPCCW flow to less than 8,100 gpm.
- 3. IF any RPCCW header flow has increased OR RPCCW surge tank level has decreased, SEND Operator to check for RPCCW piping rupture.
- 4. IF RPCCW is ruptured, Go To AOP 3561, "Loss of Reactor Plant Component Cooling Water."
- 5. Refer To T/S 3.7.3, "Reactor Plant Component Cooling Water System," and DETERMINE Limiting Condition for Operation.
OP 3353.MBIC Level of Use BVI NRe.0 5 0 C onti uous92 of 115
4-7 SUPPORTING INFORMATION Initiating Device (1 of 6) Setpoint 1. 80-3CCP-FS11A (HI) 1.1 > 7,435 gpm 1.1 80-3CCP-FS12A (HI) 1.2 > 6,550 gpm 1.2 80-3CCP-FS1SA (HI) 1.3 > 1,964 gpm 1.3 80-3CCP-FS11B (HI) 1.4 > 7,435 gpm 1.4 80-3CCP-FS12B (HI) 1.5 > 5,375 gpm 1.5 80-3CCP-FS1SB (HI) 1.6 > 1,423 gpm 1.6
- 2. Computer Points 2.1 CCP-F11A 2.2 CCP-F12A 2.3 CCP-F15A 2.4 SP-CCP23-7 2.5 SP-CCP23-8 2.6 SP-CCP23-9 2.7 SP-CCP23-10 2.8 CCP-F11B 2.9 CCP-F12B 2.10 CCP-F15B 2.11 SP-CCP23-11 2.12 SP-CCP23-12 OP 3353.MB1C 1 e.050 Level of Use Continuous
4-7
- 3. Technical Specifications System" 3.1 T/S 3.7.3, "Reactor Plant Component Cooling Water
- 4. Procedures Water" 4.1 AOP 3561, "Loss of Reactor Plant Component Cooling
- 5. Control Room Drawings 5.1 ESK 1OGM 5.2 LSK 9-1D
~O 33M1 Level of Use 005 -00 CotiuosRev.
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # I11 Tier # I Group # K/A # 029.EKI .03 Importance Rating 3.8 Proposed Question: The crew is progressing through EOP FR-S. 1 "Response To Nuclear Power Generation/ATWS" and at step 6 "Verify Boration Flow" the RO has been directed to check pressurizer pressure less than 2350 psia. The RO reports PZR pressure is 2360 and stable. What action will FR-S. 1 direct the crew to take, and why? A. Start a second charging pump if only one is running, in order to increase the negative reactivity added by boron injection B. Open PORVs as necessary until PZR pressure is less than 2150 psia, in order to increase the negative reactivity added by boron injection C. Verify both PORVs are open, in order to minimize the pressure transient on the RCS. D. Verify all three PZR safety valves are open, in order to minimize the pressure transient on the RCS. Proposed Answer: B Explanation (Optional): The Westinghouse background states: "The check on RCS pressure is intended to alert the operator to a condition which would reduce charging or SI pump injection into the RCS and, therefore, boration ("C" and "D" wrong). The PZR PORV setpoint is chosen as that pressure at which flow into the RCS is insufficient." FR-S. I directs the operators to open PORVs as necessary to lower PZR pressure to less than 2150 psia ("B" correct, A wrong). "C" and "D" are plausible, since RCS overpressure is the limiting event for the worst case ATWS. Technical Reference(s): WOG Bkgd Doc for FR-S. I step 4 (Attach if not previously provided) FR-S. I, step 6. Proposed references to be provided to applicants during examination: None Learning MC-04626 Discuss the basis of major procedure steps and/or sequence of steps (As available) Objective: in EOP 35 FR-S.1. Question Source: Bank #75459 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.8, 41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
RESPONSE TO NUCLEAR EOP 35 FR-S.1 Page 7 of 14 POWER GENERATION/ATWS Rev. 016 RSPOSEHRESPNSENOTOBTAINEDl ACTINIEPECED
- 6. Verify Boration Flow
- a. Check PZR pressure - a. Check the PZR PORVs and LESS THAN 2350 psia block valves are open.
IF the PORVs or block valves are closed, THEN OPEN as necessary until PZR pressure is LESS THAN 2150 psia.
- b. Check normal charging flow b. Perform the applicable action:
path - ALIGNED
- IF SI actuated, THEN
- 1) Restore from immediate boration, if in service.
- 2) Proceed to step 7.
- IF a safety grade boration flow path is established, THEN
- 1) Using GA-14, Establish reactor vessel head vent letdown to the PRT.
- 2) Proceed to step 7.
_ c. Place charging flow control valve in MANUAL
STEP DESCRIPTION TABLE FOR FR-S.1 Step 4 STEP: Initiate Emergency Boration of RCS PURPOSE: To add negative reactivity to bring the reactor core subcritical BASIS: next most After control rod trip and rod insertion functions, boration is the direct manner of adding negative reactivity to the core. The intended SI boration path here is the most direct one available, not requiring miniflow lines are initiation, but using normal charging pump(s). Pump pressure. assumed to be open to protect the pumps in the event of high RCS and Several plant specific means are usually available for rapid boration boration should be specified here in order of preference. Methods of rapid actuation. include emergency boration, injecting the BIT, and safety injection pumps. If It should be noted that SI actuation will trip the main feedwater for safety this is undesirable, the operator can manually align the system should be injection. However, the RWST valves to the suction of the SI pumps injection is opened first before opening up the BIT valves. If a safety the BIT and already in progress but is having no effect on nuclear flux, then due to blockage or RWST are not performing their intended function, perhaps leakage. In this case some other alignment using the BATs and/or non-safeguards charging pump(s) is required. The check on RCS pressure is intended to alert the operator to a condition therefore, which would reduce charging or SI pump injection into the RCS and, at which boration. The PRZR PORV pressure setpoint is chosen as that pressure flow into the RCS is insufficient. The contingent action is a rapid depressurization to a pressure which would allow increased injection flow. When primary pressure drops 200 psi below the PORV pressure setpoint, the the PORVs should be closed. The operator must verify successful closure of PORVs, closing the isolation valves, if necessary. FR-S.1 79 HP-Rev. 1C HFRS1
Form ES-40 1-6 ES-40 1 Written Examination Question Worksheet Question # 12 Examination Outline Cross-reference: Tier # 1 1 Group # K/A # 040.AK2.02 Importance Rating 2.6 Proposed Question: PARAMETER: CURRENT VALUE: TREND: 58% Increasing Reactor power RCS pressure 2225 PSIA Decreasing Auctioneered high Tavg 569 0F Decreasing Turbine power 595 MWE Decreasing S/G NR levels 52% Increasing Steam pressure 1030 PSIG Decreasing 15 PSIA Increasing Containment pressure progress? Based on the plant conditions, which of the following events is in A. Small RCS LOCA B. Steamline break C. RCS dilution event D. Steam generator tube rupture Proposed Answer: B Explanation (Optional): Steam pressure and electric Reactor power is increasing, indicating positive reactivity event ("A" and "D"wrong). load is decreasing, indicating loss of steam to the turbine ("B" correct, "C" wrong). FSAR Chapter 15.1.3 (Attach if not previously provided) Technical Reference(s): None Proposed references to be provided to applicants during examination: increased (As Learning Objective: MC-04881 DESCRIBE the major parameter changes associated with available) heat removal by the Secondary System. Question Source: Bank # 64268 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
MNPS-3 FSAR Since the power level rises during the excessive feedwater flow incident, the fuel tempera- . , tures also rise until after reactor trip occurs. The core heat flux lags behind the neutron not flux response due to the fuel rod thermal time constant, hence the peak value does flux trip set point). exceed 118 percent of its nominal value (i.e., the assumed high neutron The peak fuel temperature will thus remain well below the fuel melting temperature. The transient results show that DNB does not occur at any time during the excessive the feedwater flow incident; thus, the ability of the primary coolant to remove heat from fuel rod is not reduced. The fuel cladding temperature therefore does not rise significantly above its initial value during the transient. The calculated sequence of events for this accident is shown in Table 15.1-1. 15.1.2.3 Conclusions 1il9s' The results of the analysis show that the DNBRs encountered for excessive feedwater (t tlaf 4 addition events are at all times above the limiting values. Therefore, the DNBR design s'iVal basis as described in Section 4.4 is met. 15.1 .2.4 Radiological Consequences release (9?53)1 Since no fuel damage is postulated for this transient a specific radiological I calculation was not performed. 15.1.3 EXCESSIVE INCREASE IN SECONDARY STEAM FLOW 15.1 .3.1 Identification of Causes and Accident Description is An excessive increase in secondary system steam flow (excessive load increase incident) the defined as a rapid increase in steam flow that causes a power mismatch between is reactor core power and the steam generator load demand. The reactor control system increase or a 5 percent per minute ramp designed to accommodate a 10 percent step load load increase in the range of 15 to 100 percent of full power. Any loading rate in excess Steam of these values may cause a reactor trip actuated by the reactor protection system. flow increases greater than 10 percent are analyzed in Sections 15.1.4 and 15.1.5. This accident could result from either an administrative violation such as excessive loading by the operator or an equipment malfunction in the steam dump control or turbine speed control. During power operation, steam dump to the condenser is controlled by reactor coolant condition signals, i.e., high reactor coolant temperature indicates a need for steam dump. A single controller malfunction does not cause steam dump; an interlock is provided which trip has blocks the opening of the valves unless a large turbine load decrease or a reactor occurred. Protection against an excessive load increase accident is provided by the following reactor protection system signals:
- 1. overpower AT 15S1.MP3 15.1-6 May 1998
MNIPS-3 FSAR
- 2. overtemperature AT
- 3. power range high neutron flux #'aAw)
- 4. low pressurizer pressure.
An excessive load increase incident is considered to be an ANS Condition II event, fault of moderate frequency. See Section 15.0.1 for a discussion of Condition II events. 15.1.3.2 Analysis of Effects and Consequences Method of Analysis This accident is analyzed using the LOFTRAN Code (WCAP-7907-P-A). The code simu- l91;1 lates the neutron kinetics, RCS, pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, steam generator safety valves, and feedwater system. The code computes pertinent plant variables including temperatures, pressures, and power level. Four cases are analyzed to demonstrate the plant behavior following a 10 percent step load increase from rated load. These cases are as follows.
- 1. Reactor control in manual with minimum moderator reactivity feedback.
- 2. Reactor control in manual with maximum moderator reactivity feedback.
- 3. Reactor control in automatic with minimum moderator reactivity feedback.
- 4. Reactor control in automatic with maximum moderator reactivity feedback.
Based on results from safety analysis performed for complete core of standard fuel, it has been shown that four-loop operation is more limiting than three-loop operation. For this reason, cases with three loops in operation were not explicitly reanalyzed for the conver-sion to V5H fuel. Results for four and three-loop operation with standard fuel are available for comparison in Appendix 15B. For the minimum moderator feedback cases, the core has the least negative moderator temperature coefficient of reactivity and therefore the least inherent transient capability. For the maximum moderator feedback cases, the moderator temperature coefficient of reactivity has its highest absolute value. This results in the largest amount of reactivity feedback due to changes in coolant temperature. For the cases with automatic rod control, no credit was taken for AT trips on overtemperature or overpower in order to demonstrate the inherent transient capability of the plant. Under actual operating condi-tions, such a trip may occur, after which the plant would quickly stabilize. A conservative limit on the turbine throttle valve opening is assumed, and all cases are studied without credit being taken for pressurizer heaters. Initial operating conditions are assumed at values consistent with steady state full-power four-loop operation. Plant characteristics and initial conditions are further discussed in Section 15.0.3. 15S1.MP3 15.1-7 May 19981
MNPS-3 FSAR Normal reactor control systems and engineered safety systems are not required to )
- function. The reactor protection system is assumed to be operable; however, a reactor trip 114ia,l is not encountered for any case due to the error allowances assumed in the set points. No single active failure will prevent the reactor protection system from performing its intended function.
The cases which assume automatic rod control are analyzed to ensure that the worst case is presented. The automatic function is not required. Results Figures 15.1-3 through 15.1-6 illustrate the transient with the reactor in the manual control mode. As expected, for the minimum moderator feedback case there is a slight power increase, and the average core temperature shows a large decrease. This results in a DNBR which increases above its initial value. For the maximum moderator feedback, manually controlled case there is a much larger increase in reactor power due to the moderator feedback. A reduction in DNBR is experienced but DNBR remains above the safety analysis limit. Figures 15.1-7 through 15.1-10 illustrate the transient assuming the reactor is in the automatic control mode. Both the minimum and maximum moderator feedback cases show that core power-increases, thereby increasing the coolant average temperature and pressurizer pressure above their initial value. For both of these cases, the minimum DNBR (g-,) 1 remains above the safety analysis limit. For all cases, the plant rapidly reaches a stabilized condition at the higher power level. Normal plant operating procedures would then be followed to reduce power. Note that due to the measurement errors assumed in the set points, it is possible that reactor trip could actually occur for the automatic control cases. The plant would then reach a stabilized condition following the trip. The excessive load increase incident is an overpower transient for which the fuel tempera-tures rise. Reactor trip does not occur for any of the cases analyzed, and the plant reaches a new equilibrium condition at a higher power level corresponding to the increase. in steam flow. Since DNB does not occur at any time during the excessive load increase transients, the ability of the primary coolant to rem-ove heat from the fuel rod is not reduced. Thus, the fuel cladding temperature does not rise significantly above its initial value during the transient. The calculated sequence of events for the excessive load increase incident is shown in Table 15.1-1. 15.1.3.3 Conclusions The analysis presented above shows that for a 10 percent step load increase, the DNBR remains above the safety analysis limit; the design basis for DNBR as described in Section 4.4 is met. The plant reaches a stabilized condition rapidly following the load increase. I1l5S1l.MP3 15.1-8 May 1998
MNPS-3 FSAR 15.1.3.4 Radiological Consequences Since no fuel damage is postulated for this transient a specific radiological release l(q-sS9 calculation was not performed. I 15.1 .4 Inadvertent Opening of a Steam Generator Relief or Safety Valve Causing a Depressurization of the Main Steam System 15.1.4.1 Identification of Causes and Accident Description The most severe core conditions resulting from an accidental depressurization of the main steam system are associated with an inadvertent opening, with failure to close, of the largest of any single steam dump, relief, or safety valve. The analyses performed assum-ing a rupture of a main steam line are given in Section 15.1.5. The steam release as a consequence of this accident results in an initial increase in steam flow which decreases during the accident as the steam pressure falls. The energy removal from the RCS causes a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in an insertion of positive reactivity. The analysis is performed to demonstrate that the following criterion is satisfied: Assuming a stuck rod cluster control assembly, with off-site power available, and assum-ing a single failure in the engineered safety features system there will be no consequential damage to the core or reactor coolant system after reactor trip for a steam release equivalent to the spurious opening, with failure to close, of the largest of any single steam dump, relief, or safety valve. Accidental depressurization of the secondary system is classified as an ANS Condition II event. See Section 15.0.1 for a discussion of Condition II events. The following systems provide the necessary protection against an accidental depressuri-zation of the main steam system due to the opening of a steam generator relief or safety valve.
- 1. Safety injection system actuation from any of the following:
- a. two out of four pressurizer pressure signals;
- b. two out of three low steamline pressure signals in a loop.
- 2. The overpower reactor trips (neutron flux and AT) and the reactor trip occurring in conjunction with receipt of the safety injection signal
- 3. Redundant isolation of the main feedwater lines Sustained high feedwater flow would cause additional cooldown. Therefore, in addition to the normal control action which closes the main feedwater valves following reactor trip, a safety injection signal rapidly closes all ISS1.MP3 15.1-9 May 19981
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 13 Tier # 1 Group # 1 K/A # W/EO8.GEN.2.4.18 Importance Rating 3.6 Proposed Question: While responding to a pressurized thermal shock (PTS) condition in accordance with FR-P. 1, the Operator is directed to check if ECCS can be terminated. What is the bases for terminating ECCS in this condition? A. SI flow may have contributed to the RCS cooldown, and may cause excessive cycling of the pressurizer PORVs. B. SI flow may have contributed to thermal stresses in the reactor vessel Thot & Tcold nozzles, and may cause excessive cycling of the pressurizer PORVs. C. SI flow may have contributed to the RCS cooldown, or may prevent a subsequent RCS pressure reduction. D. SI flow may have contributed to thermal stresses in the reactor vessel Thot & Tcold nozzles, or may prevent a subsequent RCS pressure reduction. Proposed Answer: C Explanation (Optional): SI flow may have contributed to the RCS cooldown, and may prevent a subsequent RCS pressure reduction ("C" correct). Excessive PORV cycling is not the major concern ("A" and "B" wrong), and the vessel downcomer is the area of greatest concern during a PTS event ("B" and "D" wrong). Technical Reference(s): WOG Bkgd FR-P. 1, step 6 (Attach if not previously provided) Proposed references to be provided to applicants during examination: None Learning Objective: MC-04553 Discuss the basis of major procedure steps and/or sequence of steps (As available) in EOP 35 FR-P.1. Question Source: Bank # 65033 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
Step 6 STEP DESCRI PTION TABLE FOR FR-P.1 STEP: Check If SI Can Be Terminated been, established which indicate PURPOSE: To determine if conditions have that full SI flow is no longer required BASIS: may be restored to within acceptable Following SI actuation, RCS conditions The combination of a minimum limits for SI termination to be allowed. in the vessel to cover the core subcooling and sufficient liquid level criteria in this guideline than represents less restrictive SI termination may an imminent PTS condition, SI flow those present in the ORGs since, for in contributed to the RCS cooldown or may prevent a subsequent reduction have RCS pressure. subcooled conditions and the RVLIS The subcooling criterion will ensure ensures the existence of an adequate vessel inventory such that indication in the document SI TERMINATION/REINITIATION core cooling is ensured. Refer to Executive Volume. Generic Issues section of the are not satisfied, then SI is required If either of the termination criteria be terminated. Most likely the cold to ensure core cooling and should not is due to SI water mixing effects and leg/downcomer low temperature condition an RCP restart is attempted. a the SBLOCA transient may result in Of the transients considered in PTS, In (SI) flow cannot be terminated. condition whereby Safety Injection OG-1lO and OG-117 titled "Evaluation Westinghouse Owners Group (WOG) reports "Justification of Manual RCP Trip for of Alternate RCP Trip Criteria" and a range of SBLOCAs were identified J Small Break LOCA Events" respectively, untimely RCP restart could result where continued RCP operation or conversely loss of additional inventory could in increased RCS inventory loss. The transients which could in turn ultimately result in deeper core uncovery in excess of the plant's design basis result in fuel cladding temperatures at an from a SBLOCA standpoint, RCP restart FSAR analysis result. Therefore, degraded core cooling scenario. inopportune time could result in a of Significant Flaw Extension, In WCAP-10319 titled "A Generic Assessment from Pressurized Thermal Shock of Reactor Including Stagnant Loop Conditions, Plants", numerous transient analyses Vessels on Westinghouse Nuclear Power The results those of SBLOCA have been analyzed without RCP restart. including 27 HP-Rev. IC FR-P.1 HFRP1
- - - r STEP DESCRIPTION TABLE FOR VR-P.1 Step D BASIS:
of the stagnant loop evaluation demonstrate that the total expected frequency of significant flaw extension in a typical W PWR reactor vessel due to PTS, including the contributions from stagnant loop SBLOCA transients, does not exceed the NRC required Up., screening value of 270'F for axial flaws. the Therefore, based on analyses results, RCP restart is not required to meet NRC PTS risk goal for a typical W plant. Therefore, an additional support condition, RCS subcooling, in addition to no plant specific minimum support conditions is recommended to assure that potential RCS inventory aggravation will occur due to RCP restart. safety An analysis of the effect of an RCP restart has been made to ensure the of this action relative to vessel integrity. For conservatism in the analysis the assumption was made that a small preexisting flaw had grown and arrested at 75 percent of wall thickness before RCP start. Starting an RCP was shown not to result in any further flaw propagation and loss of vessel integrity. For a case where a flaw has not grown prior to RCP start, the subsequent heat-up of the downcomer region will decrease the possibility of flaw initiation. Therefore, in order to mix the cold incoming SI water and the warm reactor coolant water and thereby decrease the likelihood of a PTS condition, an RCP restart is attempted. Whether an RCP is started or not, the next step performed (Step 24), if SI is still required, provides guidance on subsequent cooldown restrictions. ACTIONS: o Determine if RCS subcooling (based on core exit TCs) is greater than (R.12) 0 F [(R.13) 0 F for adverse containment] o Determine if RVLIS full range indication indicates greater than (K.02) if no RCP running o Determine if RVLIS dynamic head range indicates greater than (L.08) if one RCP running o Determine if RYLIS dynamic head range indicates greater than (L.07) if two RCPs running o Determine if RVLIS dynamic head range indicdtes greater than (L.06) if three RCPs running o Determine if RVLIS dynamic head range indicates greater than (L.05) if four RCPs running o Determine if RCS subcooling (based on core exit TCs) is greater than (R.01)0 F [(R.02)0 F for adverse containment] o Determine if no RCP is running o Attempt to start one RCP FR-P.1 28 HP-Rev. IC HFRP1
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 14 Tier # 1 Group # 1 K/A # W/E08.EKI.3 Importance Rating 4.0 Proposed Question: A large steam break occurs inside CTMT on the "B" and "C" SGs. Current conditions are as follows:
- "A" and "D" SG narrow range levels are 15%
- AFW flow is 100 GPM to the "A" and "D" SGs
- The "A" and "D" SG "STEAMLINE PRESSURE LO" and "LEVEL LO-LO" annunciators are lit on Main Board 5.
- The crew is in FR-P. 1 "Response to Imminent Pressurized Thermal Shock Condition"
- RCS temperature and pressure are stable
- Only the control group of pressurizer heaters energized The crew has determined a 1 hour soak is required.
Which of the following evolutions could be performed by the crew in the next hour? A. Energize additional pressurizer heaters. B. Place auxiliary spray in service. C. Increase AFW flow to 300GPM each to the "A" and "D" SGs. D. Recirc RHR to establish boron concentration and place it in service. Proposed Answer: B Explanation (Optional):
"B" is correct, since soak requirements do not prohibit lowering RCS pressure. "A" is wrong since pressure can not be raised, and "C" and "D" are wrong, since temperature can not be lowered.
Technical Reference(s): FR-P. 1, step 23 (Attach if not previously provided) Proposed references to be provided to applicants during examination: None Learning Objective: MC-04552 Describe the major action categories within EOP 35 FR-P.l. (As available) Question Source: 1997 Millstone 3 NRC SRO exam #88 Question History: 1997 Millstone 3 NRC SRO exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.8, 41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
EOP 35 FR-PE1 Page20f 21 RESPONSE TO IMMINENT Rev. 013 PRESSURIZED THERMAL SHOCK CONDITION NOT OTIE~ ok RESP~~~~RESPONSESOS
~NOTE it is expected that RCS If ECCS flow can NOT be terminated, temperature will tend to decrease.
- 23. Determine If RCS Temperature Soak Is Required
- a. Go to procedure and step in
- a. Check cooldown rate in effect.
any RCS cold leg - 0 GREATER THAN 100 F IN ANY 60 minute PERIOD
- b. Perform the following soak for 1 hr:
- 1) Maintain RCS temperature GREATER THAN OR EQUAL TO current temperature
- 2) Maintain RCS pressure LESS THAN OR EQUAL TO current pressure
- 3) Perform the actions of other procedures in effect which do NOT decrease RCS temperature or increase RCS pressure
- c. Return to step 23.b.
- c. Check RCS 1 hr soak -
COMPLETED
- d. Maintain RCS pressure and cold leg temperature within the limits shown on Attachment A (Attachment B ADVERSE CTMT)
'D THERMAL Rev. 013 DMON ACTIONIEXPECTED RE, RESPON E NOT OBT-4 itinued) Maintain cooldown rate in RCS cold legs - LESS THAN 50"F IN ANY 60 minute PERIOD
- o To Procedure And Step In ffect
-FINAL-
Form ES-401-6 ES-401 Written Examination Question Worksheet Question # 15 (SRO) Examination Outline Cross-reference: 1 Tier # Group# I K/A # PLANT SPECIFIC Importance Rating N/A Proposed Question: sequence of events occurs: With the unit at full power, the following system orders a load reduction from Millstone Station due to a transmission
- CONVEX emergency. MWe) in the next 15 Unit 3 is directed to reduce total generation by 760 MWe (to 440 minutes.
and commences reducing generator
- The Crew enters AOP 3575, "Rapid DownpoWer,"
load accordingly. from the BAT The Crew has just initiated the required boration for this load reduction tanks. is 75 gpm.
- Direct Boric Acid flow as read on 3CHS-FI183A and of Boric Acid should be added to the RCS to support this load reduction How many gallons how long should the boration last?
Acid. A. Borate approximately 550 gal. of Boric Boration should last for 7.3 minutes. Acid. B. Borate approximately 550 gal. of Boric Boration should last for 15 minutes. Acid. C. Borate approximately 950 gal. of Boric Boration should last for 12.7 minutes. Acid. D. Borate approximately 950 gal. of Boric Boration should last for 15 minutes. C Proposed Answer: Explanation (Optional): at 5%/omin. Step 5 has the CONVEX requested load reductions shall be performed is: Per Step I of AOP 3575 all total amount of boration per Step 5.h borate using a BAT pump via MV8104 from the BAT tanks. The operator Change (A%/o)] x [ 15(gal/%/o) I Total Boration (gal) = [Total Power 0 MWe x 100% ] x [ 15 gal/% ] = 950 gal. of Boric Acid. Per Step 5.i the Total Boration (gal) = [(760 MWe)/120 Time = [ 950 gal. ] / [ last for: Time = [Total Boration (gal)] / [Direct Boric Acid Flowrate (gpm)]. is the value boration should since this min. "C" is correct. All others are incorrect. 550 gallons is plausible in power. 15 75 gpm ] = 12.7 rather than the change if the calculation is performed using the final power level of 440 MWe, obtained to complete the downpower. time in which the crew is requested minutes is plausible since this is the (Attach if not previously provided) Technical Reference(s): AOP 3575, steps I and 5 None applicants during examination: Proposed references to be provided to 1 41 of 46 NUREG-1021, Revision 8, Supplement
Form ES-401-6 ES-401 Written Examination Question Worksheet contained within AOP-3575. (As available) Learning Objective: MC-03982 Describe the major action categories Bank # 70168 Question Source: Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.10 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
RESPONS WR PONSENOTOBTAMMNED-CIN/EXPECTED NOTE
- A CONVEX requested emergency generation reduction should be completed within 15 minutes of notification.
be
- If a unit shutdown is required, the target power level should between 20% and 25% reactor power.
- If at any time ROD CONTROL BANKS LIMIT LO-LO (MB4C 4-9) annunciator is received, DO NOT go to step 9.
AOP 3566, Immediate Boration. Immediately perform Determine Power Reduction Rate (%O/min)
- a. Check desired power a. Perform the following for an reduction rate - immediate turbine unloading:
LESS THAN OR EQUAL TO 5%/min 1) Place rod control SEL switch in AUTO.
- 2) Rotate LOAD LIMIT SET to reduce turbine load the desired amount up to a 480 MWe reduction.
- 3) Proceed to step 5.
- b. Check power reduction - b. Proceed to step Ld.
CONVEX REQUESTED
- c. Perform load reduction at 5%o/min and Proceed to step 2.
ACTION/EXPECTED RESPONS RESPONSE NOT OBTAINED STEP I
- 1. (continued)
- d. Determine power reduction rate using Table Poe Totaloe
_ .geA Cag(% Toa I - - - 50%- 60% - 70%I 80% lo-/, 20%1 3014% 70%l
- . -~
l - I 3 I 3 5 5 15 l 3 5 1 3 3 3 0.5 I 3 3 30 3 3 3 0.5 0.5 1 3 ! 45 1 3 1 1 0.5 0.5 .1 1 3 0 60 3 75 0.5 0.5 1 I Time to 90 1 Reduce 0.5 0.05 1 1 Power 0.5 (mm) 105 0.05 0.5 1 1 0.5 120 0.5 0.5 0.5 1 0.5 0.5 135 0.5 _ 0.5 0.5 0.5 1 150 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 160 0.5 Check Rod Control - IN AUTO Place rod control SEL switch in
- 2. AUTO and Proceed to step 3.
IF automatic rod control NOT available, THEN Move rods in MAN as necessary to maintain Tavg within +5oF of Tref.
SE ACTION/EXPECTED RESPONSREPNE NO OBAIE NOTE If at any time the power reduction rate or target power level must I be changed, Return to step 1.
- 4. Verify Power Reduction Rate
- a. Check power reduction rate - a. Perform the applicable action:
5%/MIN
- IF power reduction rate is 3%/min, THEN Proceed to step 5.
- IF power reduction rate is LESS THAN 3%,/min, THEN Proceed to NOTE prior step 6.
- b. Check power reduction - b. Proceed to step 5.
REQUIRED TO STABILIZE PLANT
- c. Proceed to NOTE prior step 7.
- 5. Initiate Rapid Boration
- a. Verify RCS makeup system in a. Perform the following:
- AUTO 1) Select AUTO on the reactor coolant makeup select switch.
- 2) Select START on the reactor coolant makeup start switch.
- b. START one boric acid transfer b. Proceed to step 5.d.
pump
ACTIONEXPECTED RESPONS PONSE NOT OBTAINED I
- 5. (continued)
- c. OPEN emergency boration valve (3CHS*MV8104)
- d. Verify direct boric acid flow d. Perform the following to (3CHS-FI 183A) - initiate gravity boration:
VNICATED
- 1) Place the charging line flow control valve in MAN.
- 2) OPEN at least one gravity feed boration valve.
- 3) CLOSE at least one VCT outlet isolation valve.
- 4) Limit net charging flow to the RCS to LESS THAN 75 gpm (charging + seal injection - RCP seal return).
- 5) Adjust charging line flow control valve as required.
IF gravity boration flow can NOT be established, THEN TRIP the reactor and Go to E-0, Reactor Trip or Safety Injection.
- 6) Proceed to step 5.f.
- e. OPEN charging line flow control valve, to match boric acid flow (3CHS-FI 183A)
- f. Record time boration started Time
---- i I
- 5. (continued)
- g. Energize all PZR heaters
- h. Determine required boric acid addition by multiplying total power change (Ao by 15(galI%) = .gal
- i. Determine required time to borate by dividing required gallons of boric acid by the direct boric acid flowrate (net chargingflow rate if using gravity boration) min I
- j. Check turbine load decrease j. Proceed to NOTE prior to
- IN PROGRESS OR step 7.
COMPLETED
- k. Proceed to NOTE prior to step 8.
Form ES-40 1-6 ES-40 1 Written Examination Question Worksheet Question # 16 Examination Outline Cross-reference: Tier # I Group# I K/A # 055.EA2.03 Importance Rating 4.7 Proposed Question: occurs. The crew has entered ECA-With the plant operating at 100%, a total loss of offsite power 0.0, "Loss of All AC Power". After 15 minutes, the SBO diesel has been started. emergency bus 34C from the SBO Which of the following will allow the operator to energize diesel? only. A. The operator resets the station LOP signal at the sequencer B. The operator resets the LOP signal at Main Board 2 only. the station LOP signal at the sequencer. C. The operator resets the LOP signal at MB2 and resets at MB8R and resets the LOP signal at D. The operator presses the UV block bypass pushbutton MB2. Proposed Answer: D Explanation (Optional): Once the LOP occurs, if the RSST does not energize the bus after 1.8 seconds, the RSST supply breaker and the bus tie breaker are locked out for 6 minutes. Per the reset at MB8R and either off-site or the stem, 15 minutes has elapsed, so the LOP lockout can be Also, the LOP signal must be reset at SBO could be placed on the bus. (A, B, and C wrong). MB2 or the sequencer ("D" correct). LSK 24-3K, 24.4A (Attach if not previously provided) Technical Reference(s): None Proposed references to be provided to applicants during examination: MC-03 851 Describe the major action categories (As available) Learning Objective: within EOP 35 ECA-0.0. Question Source: Bank # 73167 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.7 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
SOURCE CONDITION ACTION( CONTROL RESULTANT RESUI.TANT (, OR (
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- 2.
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- 2. BUS 5140UNDERVOLTAGE PROTECTIONI SHOWN. ____
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ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 17 Tier # 1 Group # 1 K/A # 057.GEN.2.4.11 Importance Rating 3.6 Proposed Question: With the plant at 100% power, VIAC 2 deenergizes, and the crew enters AOP 3564 "Loss of One Protective System Channel". How does AOP 3564 address the bistable tripping requirements of Tech Spec LCO 3.3.2 "ESF Actuation System Instrumentation" for all of the instruments that have lost power, if required? A. The bistables are directed to be tripped in AOP 3564 "Loss of One Protective System Channel", step 2 while checking if the major control systems are operating normally in auto. B. The bistables are directed to be tripped via AOP 3571 "Instrument Failure Response" attachments, referenced by AOP 3564, step 11, after verifying the VIAC is still deenergized. C. The bistables are not required to be tripped, since Tech Spec LCO 3.8.3.1, referenced by AOP 3564 "Loss of One Protective System Channel", step 6, includes all ACTIONs required for a loss of VIAC. D. The bistables are not required to be tripped, since Tech Spec LCO 3.8.3.2, referenced by AOP 3564 "Loss of One Protective System Channel", step 6, includes all ACTIONs required for a loss of VIAC. Proposed Answer: B Explanation (Optional): "B" is correct, since AOP 3564, step 11 directs the crew to perform the applicable AOP 3571 attachments, which trip bistables, after verifying the VIAC is still deenergized. "A", "C", and "D" are plausible, since these are actual steps in AOP 3564 that deal with Tech Specs and control systems. Technical Reference(s): AOP 3564, steps 2, 6, and 11. (Attach if not previously provided) Proposed references to be provided to applicants during examination: Tech Spec sections 3/4 Learning MC-03955 Describe the major action categories contained within AOP 3564 (As available) Objective: "Loss of one protective system channel". MC-03958 Given a plant condition requiring the use of AOP 3564... identify applicable Tech Spec action requirements. Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 55.43.2 and 43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement I
RESPONSE NOT OBTAIN C A U T I O N,f A(B) being Loss of VIAC- 1(2) resu[Its in diesel generator sequencer condition, de-energized. If an ES]F actuation takes place during this the following items will rtot occur automatically. LOP)
- The A(B) diesel will not start (except busses
- Loads will not be sti*ipped from the emergency
- A(B) train loads wiliI not start
- 1. Determine Plant Status
- a. Check reactor status - a. Proceed to step 2.
TRIPPED
- b. Go to E-0, Reactor Trip or Safety Injection
- 2. Check The Following Perform the following:
Control Systems:
- a. Place the affected controller in
- Verify rod control - MANUAL.
OPERATING NORMALLY IN AUTO b. Stabilize plant parameters.
- Verify SG level - c. Defeat the failed channel OPERATING NORMALLY input.
IN AUTO
- d. Return the affected controller
- Verify PZR level - to AUTO.
OPERATING NORMALLY IN AUTO
- Verify PZR pressure -
OPERATING NORMALLY IN AUTO
ACTION/EXPECTED RESPONS RESPONSE NOT OBTAINED SE
- 6. Refer To The Following Technical Specifications For Applicable Actions:
- If in Mode 2 and LESS THAN 10-10 IR amps, 3.3.1 Reactor Trip System Instrumentation:
Action 4 (FU 6.a) Action 8 (FU 17.a)
- 3.3.1 Reactor Trip System Instrumentation Action 8 (FU 17.b)
- 3.3.2 ESF Actuation System Instrumentation:
Action 16 (FU 7.d) Action 17 (FU 2.c & 3.b.3) Action 18 (FU 7.e) Action 20 (FU 8.a & 8.b)*
*entry into TIS 3.0.3 required
- If in Mode 5 or 6, 3.3.2 ESF Actuation System Instrumentation Action 26 (FU 3.c)
- If in Modes 4, 5 or 6, 3.4.9.3 Overpressure Protection Systems
- 3.8.3.1 and 3.8.3.2, Onsite Power Distribution
ACTION/EXPECTED RESPONS RESPONSE NOT OBTAINED STEP
- 9. Verify Cold Overpressure Protection System Operation
- a. Check PORVs required for a. Proceed to step 10.
cold overpressure protection
-b. Using OP 33011, 'Arming the Cold Overpressure Protection System," Arm COPS
- 10. Verify MB Annunciators And Immediately report any Parameters - AS EXPECTED unexpected or unexplained conditions to the Shift Manager.
- 11. Verify Affected Protection Perform the following:
Channel - ENERGIZED
- 1) Perform follow-up actions for failed instrumentation using AOP 3571, Instrument Failure Response, applicable attachments.
- 2) Review the loads that are de-energized and determine the overall effect on the plant.
- 12. Continue With Normal Plant Evolutions Using Applicable Plant Procedures
-FINAL-
Form ES-401-6 ES-401 Written Examination Question Worksheet Question # 18 (SRO) Examination Outline Cross-reference: 1 Tier # Group # I K/A # PLANT SPECIFIC Importance Rating N/A Proposed Question: sequence of events occurs: With the plant initially at 100% power, the following ECA-0.0 "Loss of All AC Power".
- 1. All AC power is lost, and the crew enters before offsite power is restored.
- 2. CONVEX reports that it will be several hours cooldown the RCS.
- 3. The crew starts depressurizing all SGs to hours.
"B" EDGs will not be available for several
- 4. Maintenance reports that both the "A" and and used to restore power to 34Aand 34C
- 5. After 25 minutes, the SBO diesel is started All AC Power, Recovery with the SBO Diesel".
- 6. The crew transitions to ECA-0.3 "Loss of accordance with ECA-0.3?
Which of the following will the crew do in psig while performing the actions of ECA-0.3. A. Continue depressurizing all SGs to 260 Cooldown" upon completion of ECA-0.3. B. Transition to ES-0.2 "Natural Circulation Power, Recovery with SI Required" if pressurizer C. Transition to ECA-0.2 "Loss of All AC 0 drops less than 32 F. level drops to less than 16%, or RCS subcooling Heat Sink", if the Turbine Driven Auxiliary D. Transition to FR-H. 1 "Loss of Secondary range in all Steam Generators. Feedwater Pump trips with less than 8% narrow B Proposed Answer: crew will monitor status trees (Optional): Due to significant loading limitations with the SBO diesel, the Explanation need exists for an SIS only in ECA -0.3 ("D" wrong), and will not transition to ECA-0.2 when the for information SI Required" to ECA-0.3 uses the strategy of ECA-0. 1 Loss of All AC Power, Recovery Without ("C" wrong). "B" correct). "A" is plausible, the RCP seals by cooling down the RCS via natural circulation ("A" wrong, cooldown to ECA-0.2 is done from is a strategy in ECA-0.0. "C" is plausible since transitioning since SG depressurization ECA-0. 1 when the under these conditions. "D" is plausible since status tree monitoring is performed in ECA-0. 1 equipment running. emergency bus is restored with the proper (Attach if not previously provided) Technical Reference(s): ECA-0.3, NOTE prior to step 1, ECA-0.3 steps 1, 7, 22 and 25. during examination: None Proposed references to be provided to applicants (As available) transition to other Learning MC-03868 Discuss conditions which require Objective: procedures from ECA-0.3. Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement I
ACTION/EXPECTED RESPONS RESPONSE NOT OBTAINED I I STP
< ~C AUT I0 N permit manual
- If an SI signal is actuated, it should be reset to loading of equipment on an AC emergency bus.
Valve Building
- Minimize personnel stay times in the Main Steam by when ventilation is inoperable. Cooling can be obtained doors level opening the 72' outside door (376) and 3 ground (370, 371 and 308). These include SLCRS boundaries.
THAN
- If Main Steam Valve Building temperature is GREATERutilized must be OR EQUAL TO 100'F, determine if cool vests Stress by referring to MP-19-SF-REF01514, "Heat Management."
NOTE only. The CSF Status Trees should be monitored for information to prior DO NOT implement Functional Response procedures or an restoring power to an AC emergency bus from offsite emergency diesel generator.
- 1. Stabilize SG Pressures
- a. Adjust the following: a. Stabilize SG pressure by locally Operating the
- SG atmospheric steam atmospheric steam dump dump valves isolation and bypass valves:
- Atmospheric steam dump OR isolation valves:
3MSS*MOV18A
- SG atmospheric dump 3MSS*MOV18B bypass valves 3MSS*MOV18C 3MSS*MOV18D
- Atmospheric steam dump bypass valves:
3MSS*MOV74A 3MSS*MOV74B 3MSS*MOV74C 3MSS*MOV74D
LOSS OF ALL AC POWER - EOP 35 ECA-0.3 Page 9 of 22 RECOVERY WITH THE SBO DIESEL Rev. 009
.RESPONSE NOT OBTAINED (
- 6. (continued)
- g. Check RCP seal supply g. WHEN isolation valves - CLOSED RCP seal supply isolation valves are closed, THEN Proceed to step 6.h.
- h. Verify total SBO diesel load h. Perform the following:
(SBO D/G control panel) - LESS THAN OR EQUAL 1) Determine nonessential TO 1450 Kw loads to be shed.
- 2) Remove loads as necessary.
- 3) WHEN Total load on the SBO diesel is less than required, THEN Proceed to step 6.i.
__ i. START one charging pump
- 7. Verify SI Flow Not Required
- a. Check PZR level - a. Perform the following:
GREATER THAN 16% (50% ADVERSE CTMT) 1) OPEN one charging pump cold leg injection valve (3SIH*MV8801A or 3SIH*MV8801B).
- 2) WHEN PZR level is GREATER THAN 16% (50% ADVERSE CTMT)
THEN CLOSE charging pump cold leg injection valve.
LOSS OF ALL AC POWER - EOP 35 ECA-0.3 Page 18 of 22 RECOVERY WITH THE SBO DIESEL Rev. 009 STEP ACTION/EXPECTED RESPONSH RESPONSE NOT OBTAINED
- 21. Start Spent Fuel Pool Cooling
- a. Verify RPCCW pump - AT a. Proceed to step 22. and, LEAST ONE RUNNING WHEN One RPCCW is started, THEN Perform step 21.b.
- b. Using Attachment A and GA-5, Startup spent fuel pool cooling
- 22. Start MD AFW Pump, If Desired
- a. Verify DWST level - a. Using GA-4, Shift AFW GREATER THAN 80,000 gal pump suction to the CST, and Fill the DWST IF MD AFW pump suction can NOT be shifted to the CST, THEN Proceed to step 23., and, WHEN DWST is filled to GREATER THAN 80,000 gal, THEN Perform step 22.b.
- b. Using Attachment A, Start the MD AFW pump
LOSS OF ALL AC POWER - EOP 35 ECA-0.3 Page 22 of 22 RECOVERY WITH THE SBO DIESEL Rev. 009
- 23. (continued)
- g. If required,Using Attachment A, START one MCC/Rod Control Area A/C Unit (3HVR*ACU1A or 3HVR*ACU1B)
- h. If required, Using Attachment A and GA-21, Startup fuel building ventilation
- i. If required, Using Attachment A and OP 3315B, "Main Steam Valve Building Ventilation," Perform the applicable action:
- Place normal MSVB ventilation in service
- Remove normal MSVB ventilation from service
- 24. Consult ADTS On The Following:
- a. Determine if bus loading may be reduced based on plant conditions
- b. Determine additional equipment to be started to assist with plant control
- 25. Go To ES-0.2, Natural Circulation Cooldown And Perform Plant Cooldown
-FINAL-
Written Examination Question Worksheet Form ES-401-6 ES-401 Question # 19 Examination Outline Cross-reference: Tier# I Group # I KIA # 062.GEN.2.4.4 Importance Rating 4.3 Proposed Question: Which of the following conditions meets the entry conditions for AOP 3560 "Loss of Service Water"? than normal flow and A. In MODE 1, RPCCW heat exchanger flow meter indicates higher RPCCW surge tank level indicates 85% and decreasing. no pressure and the B. In MODE 3, Service Water discharge pressure instrument indicates RPCCW heat exchanger flow meter indicates no flow. and the Service Water C. In MODE 4, temperature control is lost to the RHR heat exchangers closed. dilution water supply isolation valves for the hypochlorite pumps fail and Service Water flow to D. In MODE 5, RHR heat exchanger outlet temperature is increasing the operating train's RPCCW heat exchanger indicates 8000 gpm. Proposed Answer: B Explanation (Optional): If the higher than normal Service Water flow A is wrong: This is an entry condition for AOP 3561, Loss of RPCCW. Listed as entry conditions for AOP 3560, Loss of is due to a piping break, this is dealt with using ARPs. B is correct: 3562, Loss of Instrument Air. D is wrong: This Service Water C is wrong: These are indications for entry into AOP gpm. is a procedural direction to AOP 3560 but the criteria is less than 6200 AOP 3560 entry conditions (Attach if not previously provided) Technical Reference(s): None Proposed references to be provided to applicants during examination: (As available) Learning Objective: MC-03926 Identify conditions that require entry into AOP 3560, Loss of Service Water Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.10 55.43.2 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
Peas.' n fl AOP 3560 cmv. Re 36007 LOSS OF SERVICE WATERRev. 007 A. PURPOpSaE or provides the actions necessary to respond to a partial
- 1. This procedure total loss of service water.
B. ENIY CONDITIONS of a loss of service water:
- 1. The following are indications alarm
- a. SERVICE WTR PP AUTO TRIP/OVERCURRENT (MB1C 5-3) indicator (3SWP-P1 26A and/or
- b. Service water discharge pressure heat exchanger flow meters 3SWP-PI 26B) and RPCCW 43B) indicate no pressure or (3SWP-FI 43A and/or 3SWP-FI flow.
Response Procedures direct the operator to this Alarm of both
- 2. The following to respond to service water piping ruptures or to a loss procedure SW pumps in a single train: alarm OP 3353.MB1C 1-lA, RPCCW HX SW FLOW HI/LO a.
1-1B, CTMT RECIRC CLR SW FLOW HI/LO
- b. OP 3353.MB1C alarm 3353.MB1C 2-1B, DG A COOLER SW FLOW LO alarm
- c. OP COOLER SW FLOW LO alarm
- d. OP 3353.MB1C 3-1B, DG B alarm 3353.MB1C 4-3, SERVICE WTR PUMP DIS PRES LO
- e. OP 3-3, CONTROL BLDG CHLR CNDSR A SW
- f. OP 3353.VP1A FLOW LO alarm 3-3, CONTROL BLDG CHLR CNDSR B SW
- g. OP 3353.VP1C FLOW LO alarm
- 3. This procedure is entered from:
Plant Component Cooling Water"
- a. AOP 3561, "Loss of Reactor Cooling and/or RCS Inventory",
- b. EOP 3505, "Loss of Shutdown
- c. EOP 3506, "Loss of All Charging Pumps"
Form ES-401-6 ES-401 Written Examination Question Worksheet I20 Question # Examination Outline Cross-reference: 1 Tier # Group # K/A # 067.GEN.2.4.7 Importance Rating 3.8 Proposed Question: sequence of events occurs: With the plant at 100% power, the following
- 1. A fire breaks out in the Auxiliary Building.
- 2. The crew enters EOP 3509 "Fire Emergency". Fire".
El. 24'6" South Floor Area, 43'6", & 66'6"
- 3. The crew enters EOP 3509.2 "Aux Bldg. trips the reactor and verifies the turbine is
- 4. The fire progresses to the point where the crew tripped.
and why? will EOP 3509.2 direct the crew to take regarding the secondary plant, What action dumps in the steam pressure mode to maintain RCS temperature at no load A. Place the steam Tave. cooldown. interlock in order to commence a plant B. Bypass the steam dump Lo-Lo Tave of the primary to prevent an uncontrolled cooldown C. Actuate Main Steam Isolation (MSI) plant. the secondary MSIVs and MSIV bypass valves to establish a known configuration on D. Close plant. Proposed Answer: D on the secondary "D" is correct since the potential impact of a fire on all electrical components Explanation (Optional): directs closing the of the MSIVs has not been determined. "C" is plausible, since the procedure pressure mode and side downstream since steam dumps are placed in the steam Byp. Valves. "A" and "B" are plausible, MSIVs and in non-fire situations. used to cool down the plant on reactor trips (Attach if not previously provided) Reference(s): EOP 3509.2, step 20 Technical 3509.2 basis doc, step 20 during examination: None Proposed references to be provided to applicants (As.savailable) EOP 3509.2 procedure steps Learning MC-06184 Discuss the basis of major Objective: and/or sequence of steps 60684 (Parent attached) Question Source: Modified Bank # Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 55.43.5 Comments: 1 41 of 46 NUREG-1021, Revision 8, Supplement
Form ES-401-6 ES-401 Written Examination Question Worksheet Original question 60684 The plant is operating at 100% power. enters EOP 3509 and 3509.2. A fire breaks out in the auxiliary building. The crew trip, does EOP 3509.2 then CLOSE MSIVs Why, after tripping the reactor and verifying turbine and MSIV bypass valves? reactivity. A. Ensure the RCS will heat up, thereby inserting negative B. Preclude blowing rupture discs on the LP turbines. C. Prevent rotor bowing of the main turbine. D. Establish a known configuration on the secondary plant. Answer: D 41 of 46 NUREG-1021, Revision 8, Supplement 1
AUX. BLDG. EL. 24' 6",SOUTH EOP 3509.2 Page 22 of 48 FLOOR AREA, 43'66" & 66'66" FIRE Rev. 002
- I,- ,i STCTION/EXCPECTED RESPONS RESPONSE NOT OBTAINED l
- 18. Verify 'lTrbineTrip
- a. Check all turbine stop a. TRIP the turbine.
valves - CLOSED
- 19. Verify Power To AC Emergency Busses
- a. Check busses 34C and 34D - a. Try to Restore power to at AT LEAST ONE least one AC emergency bus.
ENERGIZED IF power can NOT be restored to at least one AC emergency bus, THEN Go to ECA-0.0, Loss of All AC Power, step 3.
- b. Check busses 34C and 34D - b. Try to Restore power to BOTH ENERGIZED de-energized AC emergency busses.
- 20. CLOSE The MSIVs And MSIV Perform the following:
Bypass Valves
- At 3BYS*PNL15F Remove fuse blocks 6, 7, 9, 11, 39, 40, 41 and 42.
(Control Bldg. 47')
- At 3BYS*PNL16F Remove fuse blocks 13, 15, 18, 20, 39, 40,41 and 42.
(Control Bldg. 47')
- 21. STOP All RCPs Locally Trip RCP circuit breakers:
- RCP 1: BUS 35A (35A4-2)
- RCP 2: BUS 35B (35B2-2)
- RCP 3: BUS-35C (35C5-2)
- RCP 4: BUS 35D (35D1-2)
d~o AUX. BLDG. EL. 24' 6", SOUTH EOP 3509.2 Page 9 of 26 FLOOR AREA, 43' 6" & 66' 6" FIRE Rev. 001 STEP ACTION/EX(PECTEDRSPN RESPONSE NOT OBTAINEDl
- 18. Verify Tbrbine Trip BASIS: The turbine is tripped automatically with the reactor trip to prevent an uncontrolled cooldown of the RCS due to steam flow to the turbine. Steam demand will be controlled by the code safety valves and the atmospheric steam dump valves. If the turbine stop valves cannot be verified as closed, the operators are directed to manually trip the turbine.
- 19. Verify Power To AC Emergency Busses BASIS: The steps that follow, operate or verify the operation of components powered from the AC emergency busses. Before attempting to start or verify their operation, it is necessary to ensure that AC power is available to these components so that the failure of a component to start is not attributed to fire damage where it is only due to the unavailability of power.
that Since a reactor trip is initiated at the start of this procedure, the potential exists this off-site power could be lost due to the removal of the load generating capacity of by unit. It is assumed that the plant emergency busses 34C and 34D will be energized either off-site power or EDG operation. A fire in this area will not cause a loss of off-site power or prevent the starting of either EDG. If both busses 34C and 34D are deenergized, to the plant is outside of the analyzed condition for BTP 9.5-1, and ECA-0.0 is initiated restore power to at least one of the emergency AC busses. Once this is accomplished, ECA-0.0 will return the operators to this procedure.
- 20. CLOSE The MSIVs And MSIV Bypass Valves BASIS: This step is performed whether the turbine is tripped or not to establish the integrity of the secondary system boundary. The potential impact of the fire on all electrical components on the secondary side downstream of the MSIVs has not been determined. Bypotential closing the MSIVs and their bypass valves, all of the downstream secondary valves and uncontrolled steam release paths are isolated. All potential steam release paths between the steam generators and MSIVs have been evaluated for the potential effects of fires, and compensatory actions developed where required. By closing the MSIVs and their bypass valves, a known plant configuration on the secondary side is established.
- 21. STOP All RCPs BASIS: The RCPs are tripped to establish the conditions for a natural circulation cooldown. The fire evaluation performed for BTP 9.5-1 has looked at the RCP supporting systems only to ensure that seal cooling is provided and a seal LOCA is prevented. All of the other equipment needed to support RCP operation has not been evaluated. The shutdown method that has been shown to be available with a fire is based on all RCPs being secured. This step establishes this required configuration.
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 21 Tier # I Group # 1 K/A # 068.AK2.03 Importance Rating 3.1 Proposed Question: The crew has evacuated the control room due to a fire in the Instrument Rack Room. What actions will be taken by the operators to ensure the PORV's (3RCS*PCV455A and 3RCS*PCV456), will not inadvertently open? A. Control of 3RCS*PCV455A is isolated from the Control Room at the FTSP and control power is removed from 3RCS*PCV456. B. Control of 3RCS*PCV456 is isolated from the Control Room at the FTSP and control power is removed from 3RCS*PCV455A. C. The block valves are closed with power removed from both PORVs. D. Control power is removed from both PORVs. Proposed Answer: A Explanation (Optional): PORV 455A has a switch on the FTSP, 456 does not. Step 2 of aft A has the operator remove control power for PORV 456 at 3BYS*PNL23F, circuit 3. A correct. Technical Reference(s): LSK 1.2B, ESK-7DW, (Attach if not previously provided) EOP 3509.1, Att. A Proposed references to be provided to applicants during examination: None Learning MC-06266 Discuss the basis of major precautions, procedure steps/or sequence of steps (As Objective: contained within the EOP 3509 series procedures. available) Question Source: Bank# 68334 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
RESULTANT MONITOR CON;ITION 1O COTROLAC CONTROACTION SOURCMONITOR 3RC V404A SW2RCSW PCV465A RDRVALVES EENV.
. L RRCO6*6V4E5A VI4SU OR PCV455A S 3RCS;SR GEA(NTT D j E;jI lllD V'IG l nl I)F111iE
-SC/E _R~
~
A LOSS OFA/ APORV SHA2 DiSR A EM Rh~iCIO DOPEI I P N D ILROL MB43 PWI < ePESRZRJ _ _ _ _ _ __LE LOCAL LLOICIARA CONTROL DPWNI/-- RSUIE , ( GENRA L 2 .2 I ER OERAL
- g. *l tx C L 22-OO4SE.
7 m A AGE'VCNTO SE14DG.NO 1279 LKWER.2 i L} MAINTAIND MAIAIN M
PRESSURIZER POWER RELIEF VALVE [3RCS*PCV455A] (ZO)CKT 3RCSA04 Iift 1+ ,LI 4 I INTLK CKT THIS DWG. TB510-4 3RPS*RAKLOGA SVI4I INTLK TSPA7A 9 3T5-5 0 3TV 3RCS4AMV8000A o-] F-0 7 0COM RCS-Z455A= BU-6R50 O ESK-6TD Ltl[- I SPARE 10 6 COMRCS-Z455A* 20 43_ INTLK( 3T-85V15 d,-V17 6- 0 BV-BR50 3T5-6 X 01 ANNRCS81 T8394-1l IT17-1 ?ASP 02 ESK-6TD ANN RCS77 SPARE 2 33L-3RCS*PCV455A ESK-IOHJ IC-3RCS*PCV455A 9T To ARM) (E/C
,i [-E INTLK CKT 0 AW-BG20 ANN RCS95 2? t24 THIS DWG. TB3 -5V16 i-SVI8 i-4 ,- ESK-IOHJ 4 2 GREENLT ASP INTLK 3TSPA 0 BQ-BG20 4T3-2 0 4BO1V~ ANN RCS81 3T5 __ -S 1 2 26 SPARE I FTS lAPPENDIX R1 i ,- SPARE 25 T tw 0 ESK-6TD 22 2 ANN RCS89 THIS DWG 2 1 FTSP 3 A2INTLK 7 ,,-8 ECON.CKT ESK-BOHJ 26T 43A- 3T 3RCS*MV8OOOA SVIS 4+ THIS DWG. IT5- I ° ESK-STD TB510-50 3RPS.RAKLOGA NUCLEAR SAFETY RELATED IT5-7 ITS-8 0 AT-BS20 33R-3RCS*PCV455A N4T3-4 0 ANN RCS77 ID-3RCS*PCV455A N/
QA CAT I, II NOTES 3 & 4 ESK-IOHJ PZR PORV SSPS INTLK 74-3RCS*PCV455A 4-3RCS*PCV455A 0 AW-BG20 CKT NO. 3RCSA04 I P.A.' OLRINGCONSTRUCTION REVISION4S CONT SW DIAG ESK SPARE i NOTE! A 43-3RCS*PCV455A 14 RAL REVISIONS MM IOC.WENT 19 THAIS I CY 30 IA-3RCS*PCV455A IB ;WHEN AS-BUILTARE PROHIBITED. NOTES: AX 3H NOTED. IB-3RCS*PCV455A AR2 3G IT, 2B
- 1. ALL EQUIPMENTLOCATEDAT MB4 UNLESSOTHERWISE IC-3RCSwPCV455A 2T. 3B, 4T. 4B AS2 3G
- 2. REF. LOGIC DIAG LSK-25-l.2 (284041 1D-3RCSOPCV455A AS2 3G 1I, 2T. 26, 36, 4T
- 3. LIMIT SWITCHIS GARRETP/N SERIES 3760060-1 43A-3RCS*PCV455A FN 3AB 35-77 REF. MANUFACTURER DWG.2221.315-001-001
- 4. CONTACTSARE SHOWNFOR VALVECLOSEPOSITION.
i NOTE: A
- 5. REF WNESDWGS&W FILE NO. 2450.000-001-092 I u
AND 2450.000-001-093. 1 IIINUAL REVISIONS TOTHISDOCUMENT
- 6. TYPICAL DIALCO PUSH-TO-TEST LIGHT (3 )
31 12 REDRAWN FROM 25212-32001 SH 7DW REV 13 L WHEN AS-BUILT AREPROHIBITED.
- - - - - - - - - -I I
t
CONTROL ROOM, CABLE EOP 3509.1 Page 1 of 4 SPREADING AREA OR Rev. 005 Attachment A INSTRUMENT RACK ROOM FIRE Primary Side PEO Actions on a Control Room Evacuation
- 1. Verify Reactor Ripped
- a. Check Reactor Trip and a. TRIP the reactor trip and Bypass Breakers - OPEN bypass breakers.
- b. Notify ASP operator (X6224) the reactor trip and bypass breakers are open
EOP 3509.1 Page 2 of 4 CONTROL ROOM, CABLE Attachment A SPREADING ARE A OR Rev. 005 INSTRUMENT RACK ROO)M SIRE Primary Side PEO Actions on a Control Room Evacuation
- 2. Deenergize Train B Equipment (West MCC[Rod Control)
- a. At 3BYS*PNL23F (AB 24' 6" west), Remove the following fuse blocks:
- 1) Circuit 3 to deenergize PZR PORV (3RCS*PCV456)
- 2) Circuit 19 to deenergize SG atmospheric relief valve 3MSS*PV20A
- 3) Circuit 21 to deenergize SG atmospheric relief valve 3MSS*PV20C
- 4) Circuit 40 to deenergize PZR heater group B (3RCS*H1B)
- b. TRIP breaker 32V4-2 to deenergize PZR heater B
- c. Deenergize RCP B and D seal injection isolation valves
- Place breaker 32-2W-F4F to OFF
- Place breaker 32-2W-R3F to OFF
- d. Place breaker 32-2W-R5F to OFF tQ-deenergize.
charging pump minflow isolation to RWST (3CHS*MV8512B)
EOP3 509.1 Page 3 of 4 CONTROL ROOM, CABLE SPREADING AREA OR Rev. oC6S Attachment A RACK ROOM FIRE I - INSTRUMENT Primary Side PEO Actions on a Control Room Evacuation
- 3. Deenergize Train A Equipment (East MCC/Rod Control)
- a. At 3BYS*PNL22F (AB 24' 6" east) Remove the following fuse blocks:
- circuit 17 for 3MSS*PV20B
- circuit 19 for 3MSS*PV20D
- b. Deenergize RCP A and C seal injection isolation valves
- Place breaker 32-2R-F4F to OFF
- Place breaker 32-2R-F4J to OFF
- c. Place breaker 32-2R-RlM to OFF to deenergize charging pump minflow isolation to RWST (3CHS*MV8511A)
- 4. Verify Charging Valves Aligned
- a. Check the charging pump a. Open valve(s).
minflow isolations to RWST (3CHS*MV8511A and 3CHS*MV8512B) - OPEN
-M5t Check the RCP seal supply -. Open vaNe(sy.
isolation valves - OPEN 3CHS*MV8109A 3CHS*MV8109B 3CHS*MV8109C 3CHS*MV8109D
CONTROL ROOM, CABLE EOP 3509.1 Page 4 of 4 Rev. 005 Attachment A SPREADING AREA OR INSTRUMENT RACK ROOM FIRE j Primary Side PEO Actions on a Control Room Evacuation
- 5. Block Open Auxiliary Building North Doors to Outside on El. 24'-6" Door A-24-1 Door A-24-2
- 6. De-energize TD AFW Pump Steam Supply Isolation Valves
- For 3MSS*MOV17A, Place breaker 32-4U-R5H to OFF
- For 3MSS*MOV17B, Place breaker 32-4T-R6E to OFF
- For 3MSS*MOV17D, Place breaker 32-4T-R6H to OFF
- 7. Block Open TD AFW Pump Cubicle Doors Door SF-24-1 Door SF-24-2
- 8. Verify TD AFW Pump Steam Open valve(s).
Supply Isolation Valves - OPEN 3MSS*MOV17A 3MSS*MOV17B 3MSS*MOV17D
- 9. Perform The Following:
- a. Establish communication with ASP operator
- b- Repam bn1tften -
complete
- c. Provide support as required
-FINAL-
Form ES-401-6 ES-401 Written Examination Question Worksheet Question # 22 (SRO) Examination Outline Cross-reference: Tier # 1 Group# I K/A # W/E14.EA2.1 Importance Rating 3.8 Proposed Question: The following sequence of event occurs:
- 1. A LOCA has occurred.
- 2. The crew is performing actions in E-1 LOSS OF REACTOR OR SECONDARY COOLANT.
from GREEN to
- 3. The RO reports that the "Containment" status tree has turned ORANGE.
paths exist.
- 4. The STA informs the crew that no other RED or ORANGE
- 5. Containment pressure is 24 psia and slowly increasing.
- 6. Containment radiation is 100 R/hr and slowly increasing.
- 7. Containment sump level is 15 feet and slowly increasing.
the ORANGE path? What action, if any, should be taken by the crew to address the status tree is invalid. A. No action is required, since the crew has verified that Pressure". B. Transition to FR-Z. 1 "Response to High Containment C. Transition to FR-Z.2 "Response to Containment Flooding". Radiation Level". D. Transition to FR-Z.3 "Response to High Containment Proposed Answer: B pressure of 23 psia ("B" correct, "A" wrong), or CTMT Explanation (Optional): CTMT orange paths are from CTMT ("D" is above the setpoint of 1OR/hr, but is only a yellow path high sump level 15.75 feet ("C" wrong). CTMT radiation wrong). CTMT CSF Status Tree (Attach if not previously provided) Technical Reference(s): None Proposed references to be provided to applicants during examination: into (As available) Learning MC-04666 Identify plant conditions which require entry Objective: EOP35 FR-Z.1. Modified Bank # 71298 Parent attached Question Source: Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.5 Comments: 41 Of 46 NUREG-1021, Revision 8, Supplement 1
Formn ES-40 1-6 ES-40 I Written Examination Question Worksheet Original question 71298 The following sequence of event occurs:
- 1. A LOCA has occurred.
OF REACTOR OR SECONDARY
- 2. The crew is performing actions in E-1 LOSS COOLANT.
tree has turned from GREEN to
- 3. The RO reports that the "Containment" status ORANGE.
or ORANGE paths exist.
- 4. The STA informs the crew that no other RED increasing.
- 5. Containment pressure is 19 psia and slowly slowly increasing.
- 6. Containment radiation is 100 R/hr and increasing.
- 7. Containment sump level is 16 feet and slowly crew to address the ORANGE path?
What action, if any, should be taken by the verified that the status tree is invalid. A. No action is required, since the crew has Containment Pressure". B. Transition to FR-Z.1 "Response to High Flooding". C. Transition to FR-Z.2 "Response to Containment Containment Radiation Level". Transition to FR-Z.3 "Response to High Answer: C 41 of 46 NUREG-1021, Revision 8, Supplement I
( ( I APPROVEC IBY
\
N
-~~~iI7~.IfAT lSA INM N I --,-
IA Rg^I I%u 9l9 LUULf SORC MTG NO. I rnp - o,4 -b{4 II t -1I TITLE /-TAINMAINENT Vw. Containment pressure I_ f-DCATrD TAMAAN6 .nn ia GO TO I I- _ FR-Z.1 Containment pressure -
- GO TO GREATER THAN 23 psia
- m - -- - _ _ _ _ _ _ , _ _ _ _ _ _ FR-Z.1 I
Containment I pressure-I rIIAAtI & Ia I I Containment sump level - GREATER THAN 15.75 feet c GOTO is M M M M M M M M M M M _ _ _ FR-Z.2 I Containment pressure - LESS THAN 23 psia I Containment radiation - GREATER THAN 10 R/hr GO TO FR-Z.3 Containment sump level - LESS THAN 15.75 feet Containment radiation - LESS THAN 10 R/hr CSF SAT I EOP 35 F-0.5 Rev. 003 Page 1 of 1
Written Examination Question Worksheet Form ES-401-6 ES-40 1 Question # 23 Examination Outline Cross-reference: Tier # I Group# I K/A # W/E06.EK1.2 Importance Rating 4.1 Proposed Question: due to an ORANGE The crew has entered EOP 35 FR-C.2 "Response to Degraded Core Cooling" path on the "Core Cooling" status tree. to 140 psig, the While the crew is performing the depressurization of all intact steam generators INTEGRITY critical safety function status turns RED. What actions will be taken by the crew in response to these conditions? A. Complete the current step in FR-C.2, then address FR-P. 1. B. Immediately transition to FR-P. 1, and when completed, return to FR-C.2. C. Perform appropriate actions of FR-P.1 concurrently with FR-C.2. D. Complete FR-C.2 in its entirety and then address FR-P. 1. Proposed Answer: D Explanation (Optional):Caution prior to step 10 tells operators to remain in C.2, it is expected that during the core cooling problems depressurization, accumulators will inject and the integrity tree may turn RED. To preclude P. 1. prior to addressing that may be aggravated due to FR-P. I actions later, C.2 should be completed FR-C.2, CAUTION prior to step II (Attach if not previously provided) Technical Reference(s): Proposed references to be provided to applicants during examination: None MC-0453 1 Discuss conditions which require transition to other procedures (As available) Learning Objective: from EOP 35 FR-C.2 Question Source: Bank# 65051 Question History: Millstone 2000 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
RESPONSE NOT OBTAINEDI STEP ACTION/EXPECTED RESPONS I before 2 Function Status-Tree. This procedure must be completed Pressurized Thermal to Imminent transitioning to FR-P1, Response [ Shock. NOTE during a controlled
- To allow steam dump operation to continue interlock is bypassed cooldown, ensure the Low-Low Tavg at 5530F.
SI is blocked when PZR
- Ensure that Low Steam Line Pressure pressure is LESS THAN 2000 psia.
SI signal is blocked, MSI
- After the Low Steamline Pressure setpoint is exceeded.
will occur if the high steam pressure rate
- 11. Depressurize All Intact SGs to LESS THAN 140 psig
- a. Maintain cooldown rate in RCS cold legs-LESS THAN 800F/h
Written Examination Question Worksheet Form ES-401-6 ES-401 Question # 24 (SRO) Examination Outline Cross-reference: Tier # 1 1 Group # K/A # 076.AK1.06 Importance Rating 2.6 Proposed Question: The plant has just shutdown for a refueling outage when the following sequence of events occurs: the COLD
- 1. Prior to commencing the cooldown, the crew performs an early boration to SHUTDOWN boron concentration.
- 2. RCS activity increases higher than was expected.
- 3. The crew enters AOP 3553 "High RCS Activity".
- 4. The RO verifies letdown flow is 75 gpm.
- 5. The Primary Rounds PEO verifies that the letdown demineralizer is in service.
and Reactor
- 6. The Primary Rounds PEO verifies that proper DP exists on both the Letdown Coolant filters.
is acceptable.
- 7. Chemistry reports that the Purification Demineralizer decontamination factor What operational implications exist based on the above conditions?
and the crew will A. Access to the Auxiliary Building may have to be immediately restricted, consult with Reactor Engineering about increasing letdown flow. the crew will consult B. Access to the ESF Building may have to be immediately restricted, and with Reactor Engineering about increasing letdown flow. the crew will C. Access to the Auxiliary Building may have to be immediately restricted, and place the standby purification demineralizer (mixed bed) in service. crew will place the D. Access to the ESF Building may have to be immediately restricted, and the standby purification demineralizer (mixed bed) in service. Proposed Answer: A Explanation (Optional): The early boration results in increase in radiation levels in the Auxiliary Bldg and Bldg radiation levels will be an issue Containment ("B" and "D" wrong). "B" and "D" are plausible, since the ESF to consult with RE about increasing letdown after RHR is placed in service. AOP 3553 step 7 directs the operators placed in service if improper letdown flow exists, or flow ("A" correct). The standby purification demin will only be if purification demineralizer DF is >25 ("C" and "D" wrong). Technical Reference(s): AOP 3553 (Attach if not previously provided) OP 3208, step 4.1.3 ALARA Note None Proposed references to be provided to applicants during examination: MC-05893 Discuss the basis of major procedure steps or sequence of (As available) Learning Objective: steps in AOP 3553. Question Source: New 41 of 46 NUREG-1021, Revision 8, Supplement 1
Written Examination Question Worksheet Form ES-401-6 ES-401 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.8, 41.10 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
EOP Review and Approval Form (Sheet I of 1) DOCUMENT NO. AOP 3553 REV. NO. TITLE High Reactor Coolant System Activity 6 PREPARED BY DEPARTMENT 54+ COperations, U3 DOCUMENT REVIEW If V noUnit or Review Type Sign and Date Print Comments Department f /,Z;f {2 / Me vte Ad w ' . /Z 2AI fatŽ~ TLZ$ v- v Safety Evaluation Attached? .YES 2 NO , l Environmental Review . ....... . ... . YES D NO E PORC APPROVAL ; go ) APPROVAL DATE 9 9 MEETING NUMBER i3Iell' gem CbJ EFFECTIVE DATE / 2. / / e/q 7 \ Printed: 09-27-99 12:15 pm OP3265,Att.5Rev.6 (08-19-97) Station Admin Procedures Group OSCAR Report
HIGH REACTOR COOLANT AOP 3553 Page 1 SYSTEM ACTIVITY Rev. 6 Section Eff. Rev Procedure Page 1-5 6
HIGH REACTOR COOLANT AOP 3553 Page 2 SYSTEM ACTIVITY Rev. 6 A. PURPOSE This procedure provides the necessary actions to handle an unexpected increase in Reactor Coolant System activity. B. ENTRY CONDITIONS
- 1. Any of the following are indication of high RCS activity:
- High activity based on chemistry samples.
- Failed fuel monitor Alert or Alarm.
- 2. AOP 3573, Radiation Monitor Response, when a radiological condition warrants transition to this procedure.
HIGH REACTOR COOLANT AOP 3553 Page 3 SYSTEM ACTIVITY Rev. 6 STEPACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED I i CQ A U T I 0 N Consideration (ALARA) must be given to restricting or limiting access to the Auxiliary Building (and ESF Building if RHR System is in operation) if the radiation levels are higher than normal.
- 1. Verify One Of The Following Continue plant operations using Conditions Exists: normal plant procedures.
- RCS failed fuel monitor (3CHS-RE69) -
IN ALARM, ALERT, OR INCREASING TREND
- Chemistry Department reports high RCS activity following sampling
- 2. Perform The Following:
- a. Inform Health Physics that the Auxiliary Building (and ESF Building if RHR System is in operation) radiation levels will be increasing
- b. Notify Reactor Engineering and request operational recommendations
- c. Request Chemistry obtain RCS sample using HP coverage to determine:
- Dose rate at one foot
- Dose Equivalent I-131
- 100/E
HIGH REACTOR COOLANT AOP 3553 Page 4 SYSTEM ACTIVITY Rev. 6 STEP ACTION/EXPECTED RESPONS RESPONSE NOT OBTAINED'
- 3. Determine Additional Requirements
- a. Check applicable actions using Technical Specification 3.4.8
- b. Check applicable actions using EPIP 4400, Event Assessment, Classification and Reportability
- 4. Check The Purification Demineralizer
- a. Verify letdown a. Using OP 3304A, "Charging demineralizer - and Letdown," Place IN SERVICE purification demineralizer (mixed bed) in operation.
- b. Verify letdown flow rate -
BETWEEN 70 AND 90 gpm
- 5. Request Chemistry Check Using OP 3304A, "Charging and Purification Demineralizer Letdown," Place standby Decontamination Factor - purification demineralizer (mixed GREATER THAN 25 bed) in operation.
HIGH REACTOR COOLANT AOP 3553 Page 5 SYSTEM ACTIVITY Rev. 6 l -I RESPONSE NOT OBTAINED I
- 6. Verify Letdown Filter And Reactor Coolant Filter Operation
- a. Locally Check letdown filter a. Using the following sections in differential pressure - OP 3304A, Charging and LESS THAN 20 psid Letdown," Replace the letdown filter:
- 1) Removing letdown filter from service
- 2) Placing letdown filter in service.
- b. Locally Check reactor coolant b. Using the following sections in filter differential pressure - OP 3304A, Charging and LESS THAN 20 psid Letdown," Replace the reactor coolant filter:
- 1) Removing reactor coolant filter from service
- 2) Placing reactor coolant filter in service.
- 7. Adjust Letdown Flow
- a. Consult with Reactor Engineer on the advisability of increasing letdown flow
- b. Using the section in OP 3304A, "Charging and Letdown," for increasing and decreasing charging and letdown flows, Adjust letdown flow, if desired
-FINAL-
ALARA &V The resulting pH change from an early boration may result in an increase in area dose rates in the Containment and the Auxiliary Building. NOTE
- 1. The acid reducing phase of controlled radio-cobalt dissolution is optimized by borating the RCS as quickly as possible while at, or near, operating temperature.
- 2. When required to refill the VCT during a cooldown to refueling, it is preferred to adjust the cooldown rate or increase the direct boration rate rather than initiate a blended makeup or suction shift to the RWST 4.1.3 PERFORM the following to borate the RCS to the highest concentration recorded in step 4.1.2 while continuing with this procedure:
- a. IF a controlled radio-cobalt dissolution is to be accomplished OR a high boration rate desired, PERFORM the following:
- 1) Refer To OP 3304C, "Primary Makeup and Chemical Addition," and INITIATE direct boration.
- 2) SET 3CHS-FK110, "BORIC ACID BLEND FLOW CONT," to provide a blended flow rate for the highest RCS boron concentration to be obtained.
- 3) Go To step 4.1.3.c.
- b. Refer To OP 3304C, "Primary Makeup and Chemical Addition," and ALIGN for boration.
- c. OPERATE pressurizer heaters and sprays to reduce boron concentration difference between Pzr and RCS.
- d. REQUEST Chemistry sample RCS for boron concentration every 2 hours when borating.
4.1.4 IF closing the MSIVs is desired, PERFORM the following: Level of Use RvOP 3208 Rev. 020-03 Continuous 17 of 103 I
Written Examination Question Worksheet Form ES-401-6 ES-401 Examination Outline Cross-reference: Question # 25 Tier # I Group# 2 K/A # 007.EK3.01 Importance Rating 4.6 Proposed Question: Trip The plant trips due to a loss of offsite power. The crew is currently in ES-0. 1, "Reactor Response." Why will the crew start the SBO diesel? A. To power Station Blackout Diesel auxiliaries. B. To power selected "A" Train non-emergency loads , such as the plant process computer. C. To power various non-emergency loads from either train, such as a Fuel Pool Purification Pump. D. To power PGS Pumps from either train, allowing for RCS makeup. Proposed Answer: A Explanation (Optional): (8) hours. SBO diesel installed to cope with loss of all AC power event, postulated to last eight results in a station black out. If the SBO diesel battery is depleted, With the loss of offsite power, loss of the EDGs to supply a source of AC power for an eight (8) hour period. the SBO diesel cannot perform the intended function busses have been lost, and the SBO diesel will be loaded later "B", "C", and "D" are plausible, since non-emergency for an extended period of time allows buildup of combustibles on in ES-0.2, since running the SBO diesel unloaded in the SBO diesel exhaust. Technical Reference(s): ES-0. 1 step 8 deviation document (Attach if not previously provided) Proposed references to be provided to applicants during examination: None MC-055 12 DISCUSS the basis of major procedure steps &/or (As available) Learning Objective: sequence of steps in EOP 35 ES-0. 1. Question Source: Bank # 60749 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.8, 41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
REACTOR TRIP RESPONSE EOP 35 ES-0.1 Page 23 of 48 Rev. 019 STEP DEVIATION DOCUMENTATION EOP STEP: CAUTION prior to step 8. NOTE If the SBO diesel auxiliaries are not repowered within an hour, the SBO diesel may be unavailable for starting. ERG STEP: None JUSTIFICATION OF DIFFERENCES
- 1. Caution was added to identify the consequences of not supplying the auxiliaries for a prolonged time period.
REACTOR TRIP RESPONSE EOP 35 ES-0.1 Page 24 of 48 Rev. 019 STEP DEVIATION DOCUMENTATION EOP STEP:
- 8. Check Power To SBO Diesel Auxiliaries
- a. Verify any SBO bus tie a. Proceed to step 8.c.
breaker - CLOSED TO AN ENERGIZED BUS
- Bus 34A: 34A1 -2
- Bus 34B: 34B1-2
- Bus 24E: A505 (Unit 2)
__b. Proceed to step 9.
- c. Check busses 34A AND 34B c. Perform the following:
- BOTH DE-ENERGIZED
- 1) Using GA-25, Align the SBO diesel to an energized bus (34A or 34B).
- 2) IF the SBO diesel can NOT be aligned to an energized bus, THEN Proceed step 8.d.
- 3) Proceed to step 9.
- d. OPEN all SBO bus tie d. Locally Open breaker(s).
breakers
- Bus 34A: 34A1-2
- Bus 34B: 34B1-2
- Bus 24E: A505 (Unit 2)
- e. Using Attachment A, locally Start the SBO diesel
- f. Verify local start of SBO f. Proceed to step 9. and, diesel (Using Attachment A) WHEN
- COMPLETED Attachment A completed, THEN Perform step 8.g.
REACTOR TRIP RESPONSE EOP 35 ES-0.1 Page 25 of 48 Rev. 019 STEP DEVIATION DOCUMENTATION
- g. Locally Close SBO diesel output breaker ERG STEP: None JUSTIFICATION OF DIFFERENCES
- 1. Commitment RCR-40874 made in NU Letter B16726 states: "procedure EOP 35 ES-0.1 will be revised to ensure SBO-DG availability during the postulated eight hour LOOP"
- 2. Step ensures that the SBO diesel will remain available in the event there is a station blackout at some time in the future. If offsite power is not available, the SBO diesel auxiliaries will have to be powered from the SBO diesel to ensure it is available if both EDGs become unavailable.
- 3. Added steps to account for the ability to crosstie the SBO diesel output to Unit 2 bus 24E as described in DCR M3-99-039.
- 4. Added RNO in step 8.c. to proceed to the next step if the SBO auxiliaries cannot be restored from one of its normal sources. If neither 34A nor 34B is available to supply SBO auxiliaries, the SBO must be started to supply its own auxiliaries to maintain SBO availability.
Written Examination Question Worksheet Form ES-401-6 ES-401 Question # 26 Examination Outline Cross-reference: Tier # 1 Group # 2 K/A # 008.AKI .01 Importance Rating 3.7 Proposed Question: Current conditions: 0 The crew is recovering from a faulted SG outside CTMT, upstream of the MSIVs. SI has been terminated. The crew is maintaining the plant in a stable condition per ES- l. l "SI Termination". S S S PZR pressure is 2300 psia and stable. S PZR level is 60% and stable. PRT pressure is 30 psia. PZR safety valves. RCS pressure starts to decrease rapidly due to a significant leak on one of the of Safety Based on the event in progress, which parameter will require the manual reinitiation Injection, and what will the PZR Safety valve tailpipe temperature be indicating? A. Low PZR level; 6550 F tailpipe temperature. 0 B. Low RCS subcooling; 655 F tailpipe temperature. C. Low PZR level; 250'F tailpipe temperature. D. Low RCS subcooling; 250'F tailpipe temperature. Proposed Answer: D Explanation (Optional): On a vapor space break, PZR level is not a valid indication of RCS inventory, since pressure of a two phase mixture will will quickly drop to saturation in the vessel and the hot legs ("D" correct), and formation increase until the PZR is full pressurizer level to force flow up the surge line and into the pressurizer. This will cause not change as it passes through a ("A" and "C" wrong). The enthalpy of the saturated fluid in the vapor space does to the pressure in the PRT ("A" and "B" wrong). safety valve, resulting in a temperature indication corresponding the PZR will decrease. "A" and "B" are plausible, "A" and "C" are plausible, since for most small break LOCAs, 0 F is saturation temperature for PZR pressure. since 655 Westinghouse MITCORE text pg 16-45 and 46 (Attach if not previously providedL) Technical Reference(s): E-l series foldout page. Steam Tables Steam tables Proposed references to be provided to applicants during examination: Learning MC-04914 OUTLINE the unique characteristics of a Pressurizer (As available) Objective: Vapor Space LOCA. nml.n.titn Sunirrce New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.8, 41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
failure of a PORV or pressurizer safety valve are bounded by this analysis. Appendix K assumptions were employed for both of these cases including minimum safety injection and auxiliary feed water, and loss of off-site power at the time of reactor trip. For the case where one PORV is stuck open (figures 33 through 35) the Reactor Coolant System depressurizes to 1000 psia by 2500 seconds, at which time it begins to repressurize until it becomes stable at approximately 1060 psia, remaining there for the remainder of the transient. It should be noted that the RCS pressure for this small LOCA has stabilized at a pressure less than the steam generator safety valve setpoint indicating that by this time the break is adequate to remove the decay heat. Initially the pressurizer mixture level falls as the pressure is decreasing in the pressurizer (because of the increasing vaporization rate), allowing steam to be formed which flows out the break. The sudden pressuire drop is reflected throughout the RCS; as void formation begins, the hot legs and upper core reach saturation. This increase in void fraction in the RCS causes the pressurizer pressure curve to level off, and the pressurizer begins to fill as two-phase mixture begins to flow into the pressurizer through the surge line. At about 500 seconds, the mixture level reaches the top of the pressurizer (figure 34) and a two-phase mixture begins to flow through the break. For the remainder of the transient, the break flow is alternately vapor and liquid. The pressurizer mixture quality remains very low during the entire transient. While the two-phase mixture begins to flow into the pressurizer via the surge line, the reactor coolant liquid inventory is being depleted and replaced by vapor. The upper head is drained at about. 1200 seconds and the core mixture level reaches the top of the hot legs, figure 35. As more mass flows into the pressurizer, the steam generator cold-side mixture level begins to drain at 5500 seconds and continues to fall until it becomes stable at about 20 feet. Both the loop seal and the downcomer remain full throughout the transient. The cold-side mixture level stabilizes when the. safety injection flow matches the break flow, and the system inventory is maintained. At the end of the transient, all of the pressures, temperatures, and mixture levels have reached equilibrium. There is no net loss or gain of mass in the 0869S:4 16-45
system, and the break is removing decay heat. The core remains covered during the transient with virtually no voiding. Since the plant has reached a stable condition, the core will continue to remain fully covered. The larger vapor space break of 2.5 inches in diameter causes system pressure to decrease faster, figure 36. The system also stabilizes much sooner and the key events happen earlier inItime. After the break occurs, water from the Reactor Coolant System flows to the pressurizer. By about 100 seconds, the mixture level reaches the top of the pressurizer (figure 37), after which the pressurizer stays essentially full. The water level in the RCS continues to fall, draining the upper head and part of the steam generator cold side, until an equilibrium level is reached. The plot of the core mixture level, figure 38, shows that the level remains at the hot leg elevation and the core does not uncover for this break. No voiding is present in the loop seal or downcomer at any time during the transient. After about 3000 seconds the safety injection flow is replenishing the mass lost from the break, maintaining RCS liquid inventory. Pressurizer vapor space breaks of sizes encompassed by PORVs or by combinations of PORVs are not limiting breaks in terms of peak clad temperatures. No core uncovery and very slight core voiding are expected in these transients. Perhaps more important are the indications of the plant's response to these events. r fter the POR is stuck oen he pressurizer is essentially full and primary pressure is less than the steam generator safety valve setpoint. This response is much different than the short-term response for a cold leg small break LOCA. The operator must analyze all available indications to determine the nature and extent of the transient. This is one example of pressurizer level not being a valid indication of core inventory. Other indications, such as core exit thermo-couples, must be used. I Th Dess ccdent_(OBA) The design basis accident is defined as the double-ended rupture of the cold-leg piping with a simultaneous loss of off-site power. The most limiting active failure is also assumed, taken to be the failure of a diesel generator 0869S:4 16-46
LOSS OF REACTOR OR SECONDARY EOP 35 E-1 COOLANT Rev. 018 E-1 FOLDOUT PAGE
- 1. RCP TRIP CRITERIA Trip all RCPs if BOTH conditions listed below occur:
- a. At least one charging OR SI pump is running AND
- b. RCS pressure is LESS THAN 1500 psia (1800 psia ADVERSE CITT)
- 2. SI REINITIATION CRITERIA If EITHER condition below occurs, start ECCS pumps as necessary:
0
- RCS subcooling based on core exit TCs is LESS THAN 32 F (1150 F ADVERSE CTmT)
OR
- PZR level can not be maintained GREATER THAN 16%
(50% ADVERSE CTEM1)
- 3. SECONDARY INTEGRITY CRITERIA If EITHER condition below occurs on any unisolated SG, Go to E-2, Faulted SG Isolation:
- Any SG pressure decreasing in an uncontrolled manner OR
- Any SG completely depressurized
- 4. E-3 TRANSITION CRITERIA If EITHER condition below occurs, start ECCS pumps as necessary and Go to E-3, Steam Generator Tube Rupture
- Any SG level increases in an uncontrolled manner
- Any SG has abnormal radiation
- 5. COLD LEG RECIRCULATION SWITCHOVER CRITERION If RWST level decreases to LESS THAN 520,000 gal, Go to ES-1.3, Transfer to Cold Leg Recirculation.
- 6. AFW SUPPLY SWITCHOVER CRITERION If DWST level decreases to LESS THAN 80,000 gal, AFW pump suction must be shifted to the CST and the DWST filled using GA-4.
- 7. CONTROL BLDG VENTILATION REALIGNMENT CRITERION (CWB)
With CBI actuated for 1 hour, establish Control Bldg outside filtered air using GA-18.
Form ES-401-6 ES-401 Written Examination Question Worksheet Question # 27 (SRO) Examination Outline Cross-reference: Tier # 1 Group # 2 K/A # 009.EKI .02 Importance Rating 4.2 Proposed Question: A small break LOCA has occurred, with current conditions as follows: And Depressurization".
- The crew is performing the actions of ES-1.2 "Post LOCA Cooldown
- ECCS pumps have been stopped, and normal charging is aligned.
subcooling.
- The crew has just completed depressurizing the RCS to minimize
- PZR pressure is 400 psia and decreasing slowly.
0
- Core Exit Thermocouples read 416 F and decreasing slowly.
0
- RCS hot leg temperatures are 410 F and decreasing slowly.
- Pressurizer Level is 20% and decreasing slowly.
- CTMT temperature is 1350 and increasing slowly.
crew to take next? Based on these indications, what actions does ES-1.2 direct the A. Isolate the low pressure SI accumulators. B. Operate ECCS pumps as necessary and stop both EDGs. C. Reinitiate SI and use normal spray to refill the PZR. D. Energize all PZR heaters to restore subcooling. Proposed Answer: B Explanation (Optional): ECCS pumps as necessary. The crew will skip Since subcooling is < 32 F, ES- 1.2 step 21 directs the crew to operate 0 and stop both EDGs, which are running the accumulator isolation in step 22 since subcooling is not adequate, plausible, since this is the action the crew would unloaded due to the SIS signal ("A" wrong, "B" correct). "A" is reinitiation of SI will result in a higher pressure take if subcooling and PZR level were adequate. "C" is wrong since "C" is plausible, since SI would restore subcooling than necessary, and will realign systems to the SI configuration. were too low. "D" is wrong, since heaters are not and PZR level, and spray would be used per step 11 if PZR level since step 19 uses heaters to restore saturation used to repressurize the RCS on loss of subcooling. "D" is plausible to establish a bubble in the pressurizer, they are conditions in the PZR. while pressurizer heaters are energized ES-1.2 steps 11, 19-23 (Attach if not previously provided) Technical Reference(s): Steam tables
-Steam tables Proposed references to be provided to applicants during examination:
kAs availdulu) Learning MC-05530 Discuss the basis of major procedure steps and/or Objective: sequence of steps in EOP 35 ES- 1.2. Modified Bank # 72380 Parent attached Question Source: Question Cognitive Level: Comprehension or Analysis 41 of 46 NUREG-1021, Revision 8, Supplement 1
Written Examination Question Worksheet Form ES-401-6 ES-401 10 CFR Part 55 Content: 55.41.8, 41.10 55.43.5 Comments: Original Question 72380 of ES1.2, POST LOCA A small break LOCA has occurred. The crew is performing the actions COOLDOWN AND DEPRESSURIZATION. is depressurizing the ECCS pumps have been stopped. Normal Charging is aligned. The crew RCS. When the depressurization is stopped, the following conditions exist: 0
- RCS Subcooling is 28 F and DECREASING
- Pressurizer Level is 20% and DECREASING Based on these indications, what actions should be taken?
16%. A. ISOLATE Letdown. Check to ensure Pressurizer Level stabilizes above B. Manually start ECCS pumps as necessary to regain subcooling. has actuated. C. REINITIATE Safety Injection and verify all safeguards equipment D. INCREASE RCS pressure using pressurizer heaters to regain subcooling. Answer: B 41 of 46 NUREG-1021, Revision 8, Supplement 1
POST LOCA COOLDOWN EOP 35 ES-1.2 Page 10 of 31 AND DEPRESSURIZATION Rev. 013 TIONEXPETEDESPOSEHRESPONSE NOT OBTAINEDl
- 9. Check If ECCS Is In Service Proceed to step 17.
- Verify SI pumps -
ANY RUNNING OR
- Verify charging pump cold leg injection valves -
EITHER VALVE OPEN OR
- Verify RHR pumps -
ANY RUNNING IN SI MODE
- 10. Place All PZR Heater Switches To OFF Position NOTE Voiding may occur in the RCS during depressurization which may result in a rapidly increasing PZR level.
- 11. Depressurize RCS To Refill PZR
- a. Check normal PZR spray - a. Perform the applicable action:
AVAILABLE
- IF a PZR PORV is available, THEN Depressurize RCS using one PZR PORV and Proceed to step 11.d.
- IF a PZR PORV is NOT available, THEN Proceed to step 11.c.
__b. Depressurize RCS using normal PZR spray and Proceed to step 1l.d.
POST LOCA COOLDOWN EOP 35 ES-1.2 Page 11 of 31 AND DEPRESSURIZATION Rev. 013 I SE ACIO/XPCEDRSPN RESPONSE NOT OBTAINED I
- 11. (continued)
- c. Use auxiliary spray:
- 1) Verify at least one SIH 1) Proceed to CAUTION pump - RUNNING prior to step 12.
- 2) Verify at least one 2) Proceed to CAUTION charging pump - prior to step 12.
RUNNING
- 3) CLOSE charging header loop isolation valves
- 3CHS*AV8146
- 3CHS*AV8147
- 4) Fully Open charging line flow control valve.
- 5) OPEN charging header isolation valves
- 3CHS*MV8106
- 3CHS*MV8105
- 6) Unlock and OPEN auxiliary spray valve (3RCS*AV8145)
- 7) CLOSE both charging pump cold leg injection valves
- 3SIH*MV8801A
- 3SIH*MV8801B
- d. Verify PZR level - d. Proceed to CAUTION prior to GREATER THAN 25% step 12. and, (50% ADVERSE CTMT) WHEN Level is GREATER THAN 25%
(50% ADVERSE CTMT), PEN Perform step 11.e.
EOP 335 ES-1.2 Page 12 of 31 POST LOCA COOLDOWN AND DEPRESSURIZATION Rev. 0 13
- 11. (continued)
- e. Check normal PZR spray e. Perform the applicable action:
valve - OPEN FOR DEPRESSURIZATION
- IF PZR PORV OPEN, THEN CLOSE PZR PORV and Proceed to CAUTION prior to step 12.
- IF auxiliary spray valve open, THEN,
- 1) OPEN both charging pump cold leg injection valves.
- 2) CLOSE auxiliary spray valve (3RCS*AV8145).
- 3) CLOSE charging header isolation valves.
- 4) Proceed to CAUTION prior to step 12.
- f. Close normal PZR spray valve f. Stop RCP 1 and RCP 2.
POST LOCA COOLDOWN EOP 35 ES - 1.2 Page 22 of 31 AND DEPRESSURIZATION Rev. 013 i STECED RESPONS RESPONSE NOT OBTAINED l 11
- 18. (continued)
- h. Verify RCPs -AT LEAST h. Using Attachment A, Verify ONE RUNNING natural circulation and Proceed to NOTE prior to step 19.
IF natural circulation can NOT be verified, THEN Increase dumping steam from intact SGs. NOTE Voiding may occur in the RCS during depressurization which may result in a rapidly increasing PZR level.
- 19. Depressurize RCS to Minimize RCS Subcooling
- a. Control PZR heaters as necessary
- b. Check normal PZR spray - b. Perform the applicable action:
AVAILABLE
- IF a PZR PORV is available, THEN Depressurize RCS using one PZR PORV and Proceed to step 19.e.
- IF a PZR PORV is NOT available, THEN Proceed to step 19.d.
- c. Depressurize RCS using normal PZR spray and Proceed to step 19.e.
POST LOCA COOLDOWN EOP 35 ES-1.2 Page 23 of 31 AND DEPRESSURIZATION Rev. 013 I I STPCTDRESPONS RESPONSE NOT OBTAINED I
- 19. (continued)
- d. Use auxiliary spray:
- 1) Fully OPEN charging flow controller
- 2) Unlock and OPEN auxiliary spray valve (3RCS*AV8145)
- 3) CLOSE charging header loop isolation valves
- 3CHS*AV8146
- 3CHS*AV8147
- 4) Throttle charging flow controller to adjust and maintain auxiliary spray flow
- 5) Check REGEN HX 5) Proceed to step 19.e. and, LETDOWN TEMP HI (395°F) (MB3A 5-4) IF the annunciator annunciator - LIT actuates, THEN Perform step 19.d.6).
- 6) OPEN one charging header loop isolation valve
POST LOCA COOLDOWN EOP 35 ES-1.2 Page 24 of 31 AND DEPRESSURIZATION Rev. 013 STCTION/EXPECTED RESPON RESPONSE NOT OBTAINED I
- 19. (continued)
- e. Check if EITHER of the e. Return to step 19.a.
following conditions are satisfied:
- PZR level -
GREATER THAN 73% (63% ADVERSE CTMT) OR
- RCS subcooling based on core exit TCs -
LESS THAN 42cF (125 OF ADVERSE CTMT)
- f. Check normal PZR spray f. Perform the applicable action:
valve - OPEN FOR DEPRESSURIZATION
- IF PZR PORV OPEN, THEN CLOSE PZR PORV and Proceed to step 20.
- IF auxiliary spray valve open, THEN,
- 1) Fully OPEN charging flow controller.
- 2) OPEN either charging header loop isolation valve (3CHS*AV8146 or 3CHS*AV8147).
- 3) CLOSE auxiliary spray valve (3RCS*AV8145).
- 4) Proceed to step 20.
- g. Close normal PZR spray valve g. Stop RCP 1 and RCP 2.
POST LOCA COOLDOWN EOP 35 ES-1.2 Page 25 of 31 AND DEPRESSURIZATION Rev. 013 SON/EXPECTED RESPONS RESPONSE NOT OBTAINED I
- 20. Verify Adequate Shutdown Margin
- a. Request Chemistry sample the RCS
- b. Perform shutdown margin using OP 3209B, Shutdown Margin
- 21. Verify ECCS Flow Not Required
- a. Check RCS subcooling based a. Align valves and START I on core exit TCs - ECCS pumps as necessary and GREATER THAN 32 0F Proceed to step 22.
(115FADVERSE C`TMT) ___b. Check PZR level - b. Align valves and START I GREATER THAN 16% ECCS pumps as necessary and (50% ADVERSE CTMT) Return to NOTE prior to step 11.
- 22. Check If SI Accumulators Should Be Isolated
- a. Check RCS subcooling based a. Perform the applicable action:
on core exit TCs - GREATER THAN 32 0F
- IF at least two RCS hot leg (115°F ADVERSE CTMT) WR temperatures are LESS THAN 380 0F THEN Proceed to step 22.c.
- IF at least two hot leg WR temperatures are NOT LESS THAN 3800 F THEN Proceed to step 23.
- b. Verify PZR level - b. Return to NOTE prior to GREATER THAN 16% step 11.
(50% ADVERSE C`TMT) __c. Using GA-7, Isolate SI accumulators I
EOP 35 ES-1.2 Page 26 of 31 POST LOCA COOLDOWN 1 AND DEPRE'3SURIZATION Rev. 013
-1 RI T RESPONSE NOT OBTAII
-- E - -FACTION/EXPECTED - I 01
- 23. Check If Diesel Generators Should Be Stopped
- a. Check emergency diesel a. Perform the following:
generators - BOTH RUNNING UNLOADED 1) For the loaded EDG(s), locally Perform the following at the associated diesel generator sequencer panel: a) RESET station LOP, b) Press SIS,CDA TO RECIRC pushbutton c) Place the automatic tester toggle switch to RESET, then Place to ON
- 2) IF one EDG is running unloaded, THEN Proceed to step 23.b.
- 3) Proceed to step 24.
- b. STOP unloaded emergency diesel generator(s)
- c. Locally Perform the following for unloaded emergency diesel generator(s):
- For EDG A, Place 3EGS*PNL1A control switch on MCC 32-1T-3H to START
- For EDG B, Place 3EGS*PNL1B control switch on MCC 32-1U-3H to START
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 28 Tier # 1 Group # 2 K/A # W/E03.EK1 .1 Importance Rating 4.0 Proposed Question: PLANT CONDITIONS:
- A LOCA occurred, and the crew is in ES-1.2, "Post-LOCA Cooldown and Depressurization."
- One charging pump has been stopped
- All other high head ECCS pumps are running
- No RCPs are running
- Pressurizer level is 30%
Sufficient subcooling exists and the operator stops the "A" Safety Injection pump. Immediately after stopping the pump, RCS pressure begins to decrease. What action should the crew take immediately in response to the decreasing RCS pressure? A. Manually reinitiate Safety Injection. B. Restart the "A" Safety Injection pump to restore RCS pressure to its previous value. C. Restart the "A" CHS pump to restore RCS pressure to its previous value. D. Monitor RCS subcooling and PZR level to ensure they stabilize above SI reinitiation values. Proposed Answer: D Explanation (Optional): "D" iscorrect, since after stopping any SI pump, RCS pressure should be allowed to stabilize or increase before stopping another SI pump. The basis for this is that after the SI pump is stopped, RCS pressure may decrease rapidly to a new equilibrium value where the reduced SI flow again matches leakage from the RCS. The criteria for stopping the next SI pump has been calculated assuming steady state conditions. Hence, to ensure that these criteria are appropriate, RCS pressure and subcooling should be allowed to stabilize or increase before stopping additional SI pumps. Technical Reference(s): EOP 35 ES- 1.2 Foldout Page, (Attach if not previously provided) ERG ES- 1.2 basis doc, step 13 Proposed references to be provided to applicants during examination: None Learning MC-05530 Discuss the basis of major procedure steps and/or sequence of steps (As available) Objective: in EOP 35 ES-1.2. Question Source: Bank # 63960 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.5, 41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
POST LOCA COOLDOWN AND EOP 35 ES-1.2 DEPRESSURIZATION Rev. 013 ES-1.2 FOLDOUT PAGE
- 1. SI REINTIATION CRITERIA If EITHER condition below occurs, start ECCS pumps as necessary:
- RCS subcooling based on core exit TCs is LESS THAN 0 (115F ADVERSE CTMT) 32 F OR
- PZR level can not be maintained GREATER THAN 16%
(50% ADVERSE CTMT)
- 2. SECONDARY INTEGRITY CRITERIA If EITHER condition below occurs on any unisolated SG, Go to E-2, Faulted SG Isolation:
- Any SG pressure decreasing in an uncontrolled manner OR
- Any SG completely depressurized
- 3. E-3 TRANSITION CRITERIA If EITHER condition below occurs, start ECCS pumps as necessary and Go to E-3, Steam Generator Tube Rupture
- Any SG level increases in an uncontrolled manner OR
- Any SG has abnormal radiation
- 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION If RWST level decreases to LESS THAN 520,000 gal, Go to ES-1.3, Transfer to Cold Leg Recirculation.
- 5. SUPPLY SWITCHOVERLW CRITERION If DWST level decreases to LESS THAN 80,000 gal, AFW pump suction must be shifted to the CST and the DWST filled using GA-4.
- 6. CONTROL BLDG VENTILATION REALIGNMENT CRITERION (CBI)
With CBI actuated for 1 hour, establish Control Bldg outside filtered air using GA-18.
STEP DESCRIPTION TABLE FOR ES-1.2 Step 13 - NOTE I NOTE: After stopping any SI pump, RCS pressure should be allowed to stabilize or increase before stopping another SI pump. PURPOSE: To remind the operator that RCS pressure should be allowed to stabilize before stopping another SI pump BASIS: After an SI pump is stopped, RCS pressure may decrease rapidly to a new equilibrium value where the reduced SI flow again matches leakage from the RCS. The criteria for stopping the next SI pump has been calculated assuming steady-state conditions. Hence, to ensure that these criteria are appropriate, RCS pressure and subcooling should be allowed to stabilize or increase before additional SI pumps. ACTIONS: N/A INSTRUMENTATION: N/A CONTROL/EOUIPMENT: N/A KNOWLEDGE: RCS pressure may continue to decrease slowly-as the reactor coolant temperature is reduced. However, if subcooling is increasing, the SI ) reduction criteria are appropriate and the SI flow can be further reduced when such criteria are satisfied. PLANT-SPECIFIC INFORMATION: N/A ES-1.2 93 HP-Rev. 1C HES 12
STEP DESCRIPTION TABLE FOR ES-1.2S Step 13 STEP: Check If One Charging/SI Pump Should Be Stopped PURPOSE: To determine if conditions have been established which indicate SI flow can be reduced by stopping one charging/SI pump BASIS: With SI in service, RCS pressure will tend toward an equilibrium value where SI flow matches leakage from the RCS. For subcooled conditions, the amount of leakage from the RCS is directly related to the capacity of the operating SI pumps, as shown in Figure 2. Hence, in order to minimize the loss of coolant from the primary system, SI flow must be reduced. On the other hand, some SI flow is necessary to maintain coolant inventory and pressurize the RCS sufficiently to promote primary-to-secondary heat transfer. A conflict arises between keeping the SI pumps running to maintain adequate coolant inventory and reducing SI flow to minimize leakage from the RCS. A program REDUCE has been developed for calculating various pressure/temperature relationships for stopping or realigning SI pumps which ensures that the reduced SI flow will be sufficient to maintain adequate coolant inventory. Application of this program on a plant specific basis for calculating such criteria and determining the best method of reducing SI flow is described in the document SI REDUCTION SEQUENCE EVALUATION in the Generic Issues section of the Executive Volume. The results of the above efforts are expressed in terms of subcooling requirements for stopping or realigning the different SI pumps to reduce SI flow. The required subcooling depends on the number and type of operating SI pumps, the status of the RCPs, and the method by which SI flow is reduced. For the reference high-pressure SI plant, the preferred sequence for reducing SI flow begins by stopping one of two operating charging/SI pumps. This step checks that RCS subcooling and pressurizer level are sufficient to perform this action. When this step is completed with all criteria satisfied, a maximum of one charging/SI pump should be running. Note that for other plants, the first step in reducing SI flow may be different, e.g., isolate BIT.In that case, this step and the subcooling requirements will be different. If the RCS subcooling criterion is not satisfied, but the RCS hot leg temperatures are less than the saturation temperature corresponding to the low-head (RHR) SI pump head at minimum pump recirculation flow, the charging/SI pump can be stopped if a low-head SI pump is running or can be started. Starting a low-head SI pump for this case ensures that RCS subcooling will be maintained after the charging/SI pump is stopped. ES-1.2 95 HP-Rev. IC HES12
Form ES-401-6 ES-401 Written Examination Question Worksheet Question # 29 (SRO) Examination Outline Cross-reference: 1 Tier # Group # 2 K/A # W/El .EA1.1 Importance Rating 4.0 Proposed Question: of events: A LOCA occurs, resulting in the following sequence
- 1. The crew is NOT successful in establishing cold leg recirc.
- 2. The crew enters ECA-l.1 "Loss of Containment Recirculation".
- 3. One train of ECCS flow is established.
and is establishing the
- 4. The crew does NOT meet the conditions to terminate SI altogether, ECCS Flow Requirements".
minimum ECCS flow based on ECA-1.1 Att. A "Minimum to establish the minimum ECCS How does ECA- 1.1 direct the crew to operate the ECCS pumps flow? and one RHR pump in the injection A. Maintain a minimum of one Charging pump, one SI pump mode. one Charging pump and SI B. Start and stop the RHR pumps as necessary, while maintaining pump running. and/or SI pumps as necessary. C. Maintain one Charging pump running, then start/stop Charging one SI pump and RHR pump D. Establish the normal charging flowpath while maintaining running. Proposed Answer: C a maximum Explanation (Optional): Step 12 of ECA- 1.1 has the operators manually stop ECCS pumps to establish ECA- 1.I has the operators evaluate if SI can be of one CHS, on SI, and one RHR pump running. Step 15 of 15 RNO has the operators determine the then step terminated. If conditions are such that SI can not be terminated step 15 RNO the minimum flow is determined minimum ECCS flow required using Att. A of the procedure. Once This is to conserve RWST water. The SI pumps as necessary". has the operators "START or STOP charging and/or pump running for seal injection ("C" correct). "A" NOTE prior to step 15 requires the crew to maintain one Charging guillotine shear with maximum initial decay is wrong, since one train will handle an accident up to a double ended charging and SI pumps not RHR pumps. "D" is heat. "B" is wrong, since Step 15, RNO directs the operator to utilize to terminate SI are met. wrong, since charging is left in the SI lineup unless the conditions ECA- 1.1, Steps 12 & 15 (Attach if not previously provided) Technical Reference(s): None Proposed references to be provided to applicants during examination: MC-03871 Describe the major action categories within (As available) Learning Objective: EOP 35 ECA-l.1 n'Aeetinn Source: Bank # 71075 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
91 ACTION/EXPECTED RESPONS RESPONSE NOT OBTAINED NOTE The charging and SI Fpumps should be stopped on alternate ECCS trains when possible.
- 12. Establish One Train Of ECCS Flow
- a. Check charging pumps - a. START or STOP a charging ONLY ONE RUNNING pump to establish only one running.
- b. Check SI pumps - b. START or STOP an SI pump ONLY ONE RUNNING to establish only one running.
- c. Check RCS pressure - c. STOP RHR pumps and LESS THAN 300 psia Proceed to step 13.
(500 psia ADVERSE CTMT)
- d. Check RHR pumps - d. IF two RHR pumps are ONLY ONE RUNNING IN running, SI MODE THEN STOP one RHR pump.
- 13. Verify No Backflow From RWST To Sump
- a. Locally, Check recirculation a. CLOSE valve.
spray pump supply and return isolation valve from RWST (3RSS-V25) - CLOSED
LOSS OF EMERGENCY EOP 35 ECA-1.1 Page 13 of 26 COOLANT RECIRCULATION Rev. 012 ACTION/EXPECTED RESPONS STP RESPONSE NOT OBTAINEDI NOTE Maintain one charging pump operating to supply seal injection flow to the RCPs.
- 15. Check If SI Can Be Terminated
- a. Check RVLMS plenum a. Proceed to NOTE prior to level - step 20.
GREATER THAN 19%
- b. Check RCS subcooling based b. Establish minimum ECCS flow on core exit TCs - to remove decay heat:
GREATER THAN 82 0F (165 0F ADVERSE CTMT) 1) Using Attachment A, Determine minimum ECCS flow required.
- 2) START or STOP charging and/or SI pumps as necessary to establish flow equal to or greater than the minimum flow required to match decay heat.
- 3) Proceed to NOTE prior to step 20.
- 16. Stop ECCS Pumps And Place In Standby
- a. STOP RHR pumps and Place in AUTO
__b. STOP SI pumps and Place in AUTO
- c. STOP all but one charging pump and Place in AUTO
Written Examination Question Worksheet Form ES-401-6 ES-401 Question # 30 (SRO) Examination Outline Cross-reference: Tier # I Group # 2 K/A # 025.GEN.2.4.24 Importance Rating 3.7 Proposed Question: INITIAL CONDITIONS:
- Plant is in MODE 5, solid plant operations, on the "A" Train of RHR restored
* "B" Electrical Distribution Train outage is in progress, and cannot be immediately
- RCS temperature is 150'F
- RCS pressure is 150 psia
- All RCS loops are full
- All Steam Generators are in Wet-Layup
- The "A" Charging Pump is running
- No RCPs are running The "A" RPCCW Pump fails due to a motor bearing failure.
What action will the crew take to remove decay heat? the RWST. A. Open both PORVs, fill the RCS using one Charging Pump from B. Open both PORVs, fill the RCS using one SI Pump from the RWST. pressure to >170 psia, and C. Throttle open the charging line flow control valve to raise RCS open the steam generator atmospheric dump valves. across its #1 seal, and D. Start one Reactor Coolant Pump, check proper differential pressure open the steam generator atmospheric dump valves. Proposed Answer: C steam generators are available. The Explanation (Optional): "C" is correct, since the RCS is already full and is increased to ensure subcooled natural procedure has conditions established for natural circulation, RCS pressure "A" and "B" are wrong, since bleed and feed circulation cooling, and the steam generators are used to dump steam. pump is the preferred feed source, and the SI is only used if natural circulation cooling is unsuccessful. The charging is only used if a RCP is already running. Pump is the backup source of feed. "D" is wrong, since forced cooling (Attach if not previously provided) Technical Reference(s): EOP 3505 Attachment B, Steps 9-14 None Proposed references to be provided to applicants during examination: (As available) Learning Objective: MC-04351 Describe major action categories within EOP 3505 Modified Bank # 64291 Parent attached Question Source: Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.10 55.43.5 41 of 46 NUREG-1021, Revision 8, Supplement 1
Written Examination Question Worksheet Form ES-401-6 ES-401 Comments: Original question 64291 INITIAL CONDITIONS:
- Plant is in MODE 5 on the "A" Train of RHR
- "B" Electrical Distribution Train is deenergized for maintenance, and cannot be immediately restored
- All RCS loops are full
- All Steam Generators are in Wet-Layup
- No RCPs are running The "A" RHR Pump fails due to a motor bearing failure.
What action will the crew take to remove decay heat? A. Open both PORVs, fill the RCS using one Charging Pump from the RWST. B. Open both PORVs, fill the RCS using one SI Pump from the RWST. dump C. Establish conditions for natural circulation and open the steam generator atmospheric valves. D. Open 1 PORV, fill RCS using one charging pump from the RWST. Answer: C 41 of 46 NUREG-1021, Revision 8, Supplement 1
3 2 EOP 3505 Page 6 of 16 LOSS OF SHUTDOWN COOLING Attachment B AND/OR RCS INVENTORY Rev. 9 I Loss Of Shutdown Cooling And/Or RCS Inventory During Mode 5 I
- 9. Check If RCP(s) Should Be Stopped
- a. Check RCP status - AT a. Proceed to step 11.
LEAST ONE RUNNING
- b. Verify only one RCP - b. PERFORM the applicable RUNNING action:
- IF a PZR steam bubble established, THEN STOP all but RCP 2.
- IF PZR is solid, THEN STOP any RCP.
- c. Check the following for the c. Perform the following:
running RCP
- 1) STOP the RCP.
- No. 1 seal differential pressure - GREATER THAN 210 psid 2) Proceed to step 11.
- No. 1 seal leakoff flow -
GREATER THAN OR EQUAL TO 0.2 gpm
- d. Maintain RCS pressure d. Perform the applicable action:
BETWEEN 310 psia and 375 psia
- IF RCS pressure can NOT be maintained GREATER THAN 310 psia, THEN STOP running RCP and Proceed to step 11.
- IF RCS pressure can NOT be maintained LESS THAN 375 psia, THEN Use PZR PORV to maintain RCS pressure and Proceed to step 10.
Loss Of Shutdown Cooling And/Or RCS Inventory During Mode 5
- 10. Verify Forced Circulation Decay Heat Removal
- a. Place the available SG atmospheric dump controller in MANUAL
- b. OPEN the available SG b. OPEN SG atmospheric dump atmospheric dump valve bypass valve (MB or Locally).
- c. Control feed flow to maintain c. START any available MD the available SG NR level - AFW pump.
BETWEEN 8% and 50% IF a MD AFW pump NOT available, THEN Using OP 3321, "Main Feedwater," Align the Feedwater System for startup to feed the SGs with a condensate pump.
- d. Check RCS temperature - d. WHEN GREATER THAN RCS temperature AVAILABLE SG GREATER THAN SATURATION available SG saturation TEMPERATURE temperature, THEN Proceed to step 10.e.
- e. Check forced circulation e. IF cooling can NOT be cooling verified, THEN
- RCS subcooling based on core exit TCs -
- 1) STOP running RCP.
GREATER THAN 320 F
- RCS hot leg WR 2) Proceed to step 13.
temperatures - STABLE OR DECREASING
- Core exit TCs - STABLE OR DECREASING
- f. Continue attempts to establish RHR cooling
_g. Proceed to step 16.
Loss Of Shutdown Cooling And/Or lRCS InventorDuigMd 51
- 11. Establish Conditions For Natural Circulation
- a. Check one charging pump - a. Perform the following:
RUNNING
- 1) CLOSE the charging line flow control valve.
IF the charging line flow control valve can NOT be closed, THEN CLOSE the charging header isolation valve (3CHS*MV8106).
- 2) START the charging pump aligned for service.
IF the aligned charging pump can NOT be started, THEN
- 1) Rack-up available charging pump breaker.
- 2) Using OP 3304A, "Charging and Letdown,"
Perform starting/shifting charging pumps for the available charging pump.
- b. Check RCS is intact b. Restore items required for RCS integrity.
- Using OPS Form 3260A-5, "Containment Boundary or RCS Integrity Work Log," or the plant status board, Verify - NO RCS INTEGRITY EXCEPTIONS
Loss Of Shutdown Cooling And/Or RCS Inventory During Mode 5
- c. Check the following c. WHEN conditions established: The conditions are established, THEN
- One charging pump - Proceed to step 11.d.
RUNNING IF the conditions can NOT be
- RCS - INTACT established, THEN Proceed to step 13.
- d. Check RCS pressure - d. Proceed to step 11.g.
LESS THAN 170 psia
- e. Throttle Open charging line flow control valve to increase RCS pressure
- f. Check RCS pressure -
- e. Using OP 3304A, "Charging and Letdown," Align Charging System for Safety Grade Cold Shutdown.
- f. WHEN I
GREATER'THAN 170 psia RCS pressure is GREATER THAN 170 psia, THEN Proceed to step 11.g. IF RCS pressure GREATER THAN 170 psia can NOT be established, THEN Proceed to step 13.
LOSS OF SHUTDOWN COOLING EOP 3505 Page 10 of 16 AND/OR RCS INVENTORY Rev. 9 Attachment B
. _ I Loss Of Shutdown Cooling And/Or RCS Inventory During Mode 5
- g. Adjust the charging line flow g. IF RCS pressure is increasing, control valve and the letdown THEN flow control valve as necessary Reduce RCP seal injection to maintain RCS pressure flow to 8 gpm per pump.
BETWEEN 170 and 330 psia IF RCS pressure continues to increase, THEN Using OP 3304A, "Charging and Letdown," Align Charging System for Safety Grade Cold Shutdown. IF RCS pressure can NOTQ be I maintained LESS THAN 330 psia, THEN Control pressure using one PZR PORV.
- h. Place the available SG(s) atmospheric dump controller in MANUAL
- i. OPEN the available SG(s) i. OPEN SG atmospheric dump atmospheric dump valve bypass valve (MB or Locally).
- j. Control feed flow to maintain j. START any available MD available SG(s) NR level - AFW pump.
BETWEEN 8% and 50% IF a MD AFW pump NOT available, THEN Using OP 3321, "Main Feedwater," Align the Feedwater System for startup to feed the SGs with a condensate pump. IF feed flow can NOT be established, THEN Proceed to step 13.
Loss Of Shutdown Cooling And/Or RCS Inventory During Mode 5 NOTE temperature Natural circulation flow will proceed when RCS saturation increases to approximately 50F greater than the With the available temperature of the available secondary water. temperature SG(s) maintained at atmospheric pressure,0 RCS should stablize in a range of 250'F to 270 F
- 12. Verify Natural Circulation Decay Heat Removal
- a. Check RCS temperature - a. WHEN APPROXIMATELY 50°F RCS to secondary AT GREATER THAN established, AVAILABLE SG THEN SATURATION Proceed to step 12.b.
TEMPERATURE
- b. Check natural circulation b. IF natural circulation can NOT be verified,
- RCS subcooling based on THEN core exit TCs - 0 Proceed to step 13.
GREATER THAN 32 F
- RCS hot leg WR temperatures - STABLE OR DECREASING
- Core exit TCs - STABLE OR DECREASING
- RCS cold leg temperature(s) (in available loops)- AT SATURATION TEMPERATURE FOR SG PRESSURE
- c. Continue attempts to establish RHR cooling
- d. Proceed to step 16.
LOSS OF SHUTDOWN COOLING EOP 3505 Page 12 of 16 AND/OR RCS INVENTORY Rev. 9 Attachment B Loss Of Shutdown Cooling And/Or RCS Inventory During Mode 5
- 13. Check RCS Loops - CLOSE the loop isolation valves INTACT OR LOOP ISOLATION for non-intact loops.
VALVES CLOSED F CQ A Q T I 0 Ni Evacuate all unneccessary personnel from the Containment if there is a possibility of rupturing the PRT rupture disk during feed and bleed operations. NOTE
- The PZR PORVs may not open until RCS pressure is between I 150 and 200 psia.
- During feed and bleed, a charging pump should be used, if available.
- 14. Establish Feed And Bleed
- a. OPEN both PZR PORV block valves
- b. OPEN both PZR PORVs
- c. Feed the RCS using one c. Feed the RCS using one safety I charging pump from RWST injection pump from the RWST IE a safety injection pump is NOT available, THEN Feed the RCS using a RHR pump from the RWST.
LOSS OF SHUTDOWN COOLING EOP 3505 Page 13 of 16 AND/OR RCS INVENTORY Rev. 9 Attachment B i Loss Of Shutdown Cooling And/Or RCS Inventory During Mode 5
- d. Maintain feed and bleed cooling
- e. Continue attempts to restore:
- RHR cooling
- RCS integrity
- 15. Evaluate The Incident Classification Using EPIP 4400, "Event Assessment, Classification, And Reportability," EPIP Form 4400-3 Unit 3 EAL Table
Written Examination Question Worksheet Form ES-401-6 ES-401 Question # 31 Examination Outline Cross-reference: Tier # I Group # 2 K/A # 032.AK1.01 Importance Rating 3.1 Proposed Question: Initial conditions:
- A reactor startup is in progress
- Control banks are being withdrawn
- Source Range counts: N31 = 1500 cps N32 = 1600 cps Range A faulty power supply causes significant instrument power voltage fluctuations to Source channel N3 1.
the plant? Which of the following describes the effect of the power supply voltage fluctuations on rate. The A. Instrument voltage changes will result in proportional changes in indicated count reactor will trip on high source range flux. rate. The B. Instrument voltage changes will result in proportional changes in indicated count reactor will not trip since the source range high flux trip is already blocked. the C. Instrument voltage changes will not result in changes in indicated count rate, since detector operates in the ion chamber region. The crew will continue the startup. the D. Instrument voltage changes will not result in changes in indicated count rate, since in. detector operates in the ion chamber region. The crew will stop the startup and drive rods Proposed Answer: A region of the gas Explanation (Optional): "C" and "D" are wrong since the source ranges operate in the proportional when counts exceed 1O' cps on 1/2 channels. amplification curve. The Source Range high flux trip will actuate source ranges are blocked Since the coincidence is 1/2, the reactor will trip. "A" is correct, and "B" is wrong, since range, which is above 104 cps. "B" is plausible since above P-6, which comes in at 10O- amps in the intermediate "C" and "D" are plausible since the intermediate range detectors source ranges will be blocked during the startup. operate in the ion chamber region. Technical OP 3360, Precautions 3.3 and 3.4 (Attach if not previously provided) Reference(s): OP 3202, step 4.28 NISO15C, pg 13 and 14 Proposed references to be provided to applicants during examination: None MC-05229 For the following conditions, determine the effects on the NIS system (As available) Learning Objective: and on interrelated systems: Source range instrument failure below P-6... Question Source: Bank # 75606 Question History: Previous NRC Exam 41 of 46 NUREG-1021, Revision 8, Supplement 1
Written Examination Question Worksheet Form ES-401-6 ES-401 Question Cognitive Level: Comprehension or Analysis IO CFR Part 55 Content: 55.41.6, 41.7,41.8 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
Plant calorimetric is based on steam flow at 92% RTP (Rated Thermal Power) (increasing) and feed flow at 88% RTP (decreasing). Switch over is automatic at these power levels. SP 31002, "Plant Calorimetric," gives directions for manual switch over. If any of the calorimetric input qualities are bad, the calorimetric calculation is based on the nuclear instrumentation average. If any of the nuclear instrumentation qualities are bad, the calorimetric calculation will then be bad.
- 2. PREREQUISITES 2.1 General 2.1.1 Vital instrument 120 VAC busses VIAC- 1, VLAC-2, VIAC-3, and VIAC-4 are energized to provide control and instrumentation power.
2.1.2 Detectors and bistables meet calibration requirements.
- 3. PRECAUTIONS 3.1 Exercise extreme caution at all times around electrical components of the Nuclear Instrumentation System. There is no safety interlock to de-energize the equipment when a drawer assembly is pulled out of a cabinet recess on its slide rails.
3.2 It is mandatory that the cabinet door panels for each cabinet be closed during equipment operation unless authorized by the Shift Manager or Unit Supervisor. 3.3 Removal of instrument or control power on the source and intermediate range channels can cause a reactor trip when below 10% reactor power. 3.4 The source range high flux level reactor trip must be blocked and the high voltage removed from the source range detector when the "P-6, SOURCE RANGE TRIP BLOCK PERMISSIVE" (MB4D 3-1) light on the permissive status panel is illuminated. 3.4.1 Blocking is done manually at MB4 and can be verified blocked by the illumination of "SOURCE RANGE RX TRIP BLOCK TR A' (MB4D 1-1) and "SOURCE RANGE RX TRIP BLOCK TR B" (MB4D 2-1) and also on the source range drawers in the nuclear instrumentation racks. Level of Use O 360-Continuous Rev. 007-03 1 3 of 21
3.4.2 In the event that high voltage is applied to source range detectors at power, the source range channels should be de-energized promptly. Signals from the two source range detectors should be recorded on Nuclear Recorder NR45 immediately after the source range high voltage is unblocked.
- Data should be acquired over a time period sufficient to demonstrate normal operation of the source range detectors.
- Normal operation can be verified by observing the decay of the neutron count rate of both channels.
- Any erratic indications such as disagreement between signal decay and excessive noise should be investigated.
3.5 Do not reset the source range "RX RESET/BLOCK" switches, located on MB4, above 4 X 10-10 amperes indication on either intermediate range channel, as this could cause a reactor trip. 3.6 Alignment and calibration of equipment shall be performed prior to CD startup after a sustained shutdown and at operating conditions as specified in the Nuclear Instrumentation System instruction manual. 3.7 If the AFD Monitor Alarm (Computer Program 3R5) becomes () INOPERABLE, monitoring of AFD must be commenced within one hour in accordance with SP 3602A.3, in order to ensure the requirements of Technical Specification 4.2.1.1.1b are met. OP 3360 Level of Use l _*-AL- Rev. 007-03 Continuous ITl r 4 of 21 I
-.3I 4.23.8 Go To OP 3207, "Reactor Shutdown," and PERFORM maintaining HOT STANDBY (MODE 3).
/4.24 IF using Gamma-Metrics Channel 1 or 2 for audio indication, PLACE the associated "COUNTS DIVIDER" switch in "OFF."
/4.25 STOP performance of Attachment 2, "Shutdown Rods Insertion Limit Verification Log."
4.26 Refer To SP 3601G.3, "TAVG Monitoring" and PERFORM the following: 4.26.1 STOP logging Tavg as specified for Tavg monitoring during reactor startup. 4.26.2 IF annunciator MB4C 6-5, "T REF/AUCT T AVE DEVIATION" is lit, PERFORM Tavg monitoring with Tref/Auct TAVE deviation alarm not reset.
/4.27 ESTABLISH a stable startup rate not to exceed 0.75 dpm.
NOTE
- 1. The reactor will trip at 105 cps on either source range channel.
- 2. The "P-6, SOURCE RANGE TRIP BLOCK PERMISSIVE," is enabled when either intermediate range channel increases to approximately 10-10 amps.
- 3. The first decade of the intermediate range overlaps with the source range from approximately 5 X 103 cps to 5 x 104 cps.
4.28 WHEN power level increases into the intermediate range, PERFORM the following: -I--- 4.28.1 CHECK for consistent startup rate indication between source range and intermediate range nuclear instrumentation. 4.28.2 WHEN at least one decade of overlap between source and intermediate range nuclear instrumentation verified, TRANSFER 3NME-NR45, "NIS," recorder to the intermediate range. 4.28.3 WHEN status light MB4D 3-1 "P-6, SOURCE RANGE TRIP BLOCK PERMISSIVE," is lit, PRESS the following pushbuttons:
- NMS*N33A, "SR"
- NMS*N33B, "SR" OP 3202 Level of Use STOP THINK ACT REVIEW Rev. 017-01 Continuous 21 of 33
4.28.4 CHECK the following windows lit: --- Z
- Permissive light MB4D 1-1, "SOURCE RANGE RX TRIP BLOCK TR A"
- Permissive light MB4D 2- 1, "SOURCE RANGE RX TRIP BLOCK TR B"
- Annunciator MB4C 4-1, "SR LOSS OF DET VOLTAGE" 4.29 WHEN reactor power approaches 10-8 amps in the intermediate range, PERFORM the following:
4.29.1 INSERT control rods to stabilize reactor power at approximately 10-8 amps in the intermediate range. 4.29.2 Refer To OPS Form 3209A-1, "Estimated Critical Conditions," and RECORD 'Actual Critical Data." 4.30 PERFORM the following for the shutdown margin monitors: 4.30.1 At 3NME*SMMI, "SHUTDOWN MONITOR," PRESS "1/M." 4.30.2 CHECK annunciator MB4C 2-2, "SHUTDOWN MARGIN MONITOR CHANNEL 1," not lit. 4.30.3 At 3NME*SMM2, "SHUTDOWN MONITOR," PRESS "1/M." 4.30.4 CHECK annunciator MB4C 2-3, "SHUTDOWN MARGIN MONITOR CHANNEL 2," not lit. _4.31 SELECT speed of NME-NR45, "NIS," recorder to "LOW" (MB4). 4.32 PERFORM the following:
- IF power increase desired, Go To OP 3203, "Plant Startup"
- IF shutdown desired, Go To OP 3207, "Reactor Shutdown"
- End of Section 4. -
OP 3202 Level of Use STCf: THINK ACT' REVIEW Rev. 017-01 Continuous 22of33
'I~
I/ Lesson
Title:
EXCORE NUCLEAR INSTRUMENTATION Page 13 of 91 Revision: 3 ID Number: NIS015C I TEXT ACTIVITIES/NOTES channels. Covering approximately three decades of the highest range and overlapped completely by the intermediate range instrumentation is the power range (PR) instrumentation. It provides a linear indication for normal power operation. The Uncompensated Ion Chamber is used here, since no gamma compensation is required (in this range, the bulk of the signal is from neutrons, and gamma flux is proportional to power) and the higher flux levels do not require a very sensitive detector to produce an indication. Redundancy in the PR instrumentation is provided by four separate channels, each containing an upper and lower detector. The upper and lower detectors provide for the determination of relative axial power. The detectors are spaced equally around the vessel oriented at the 450, 1350, 2250, and 3150 locations adjacent to the "truncated corners" of the core. The two SR/IR instrument wells at 0° and 1800, and the Gamma-Metric locations at 900 and 2700 are equally spaced between the PR detectors adjacent to the "flats" of the core. These detector wells are located in the neutron shield tank.
- 2. SOURCE RANGE INSTRUMENTATION The source range instrumentation is designed to provide Refer to Figure NIS -
independent monitoring of leakage neutron flux during 004 and 005 shutdown conditions, the initial phase of reactor startup and final phase of reactor shutdown. The SR channels utilize RO, SRO& STA MC Boron Trifluoride gas (BF3) proportional detectors which, - 05219, 05221, along with appropriate circuitry, provide relatively high 05223 sensitivity to small incident gamma and neutron radiation. The neutrons that are detected escape the reactor core as fast neutrons and are thermalized by a polyethylene cover surrounding the detector. The incident neutrons react with the boron in the detector to produce positively charged particles (B'0 + n --> Li 7 + He4). The positively charged particles (Li7 and He4) have significantly greater ionizing potential than does an incoming gamma. As a result, the pulses produced from neutron interactions have significantly greater amplitude than gamma pulses. By electronically discriminating (removing) the smaller gamma pulses, the source range output is strictly due to neutrons detected.
Lesson
Title:
EXCORE NUCLEAR INSTRUMENTATION Page 14 of 91 Revision: 3 ID Number: NIS015C I TEXT ACTIVITIES/NOTES The high-voltage power supply for each source range provides high-voltage inputs to the following:
- The preamplifier, which provides coupling between the high-voltage and the SR detector.
- The detector voltage meter located on the source range drawer. It provides indication of the detector high voltage in the range of 0 to 2500 volts DC.
- A bistable, which actuates a "SR LOSS OF DET VOLTAGE" annunciator on MB4, warning the operator that instrument power has been lost to the SR channel.
This condition will result in a reactor trip when power is less than P-6. This annunciator is normally lit when above P-6. The SR detector high voltage is removed by the operators when power is > P-6 to prevent burning out the detector at higher power levels. When >P-6, power is indicating in the intermediate range. In addition to the high-voltage power supply, the source range utilizes a remote preamplifier to achieve a high signal-to-noise (S/N) ratio over long lengths of field cable. The preamplifier assembly is located as near the detector as possible but outside the containment. The fixed-gain preamplifier optimizes the signal-to-noise ratio and also furnishes high voltage coupling to the detector. The pulse amplifier/discriminator assembly further amplifies the preamplifier pulses prior to coupling them to a bistable pulse shaper. Additionally, the pulse amplifier discriminates against noise and gamma pulses by providing an adjustable reference bias against which the input pulse is compared. This amplifier also provides an isolated output to drive audio circuits for generating an audible signal proportional to the Refer to Figure NIS - count rate. The Source Range channel used is selected at 015 the Audio Count Rate drawer. The pulses from the pulse amplifier are supplied to the pulse shaper which shapes them into a square wave.
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 32 Tier # 1 Group # 2 K/A # 033.AA2.11 Importance Rating 3.4 Proposed Question: Twenty minutes after a reactor trip, the RO reports the following indications: Intermediate Range channel N-35 reads 5 x IO-' amps, with 0 DPM startup rate. Intermediate Range channel N-36 reads 1 x 10-" amps, with 0 DPM startup rate. What is the cause of the above condition, and what action will be taken? A. N-35 has lost compensating voltage. The crew will actuate both source range switches on MB4, since actual reactor power indication has been lost on MB4. B. N-35 is overcompensated. The crew will actuate both source range switches on MB4, since actual reactor power indication has been lost on MB4. C. N-36 has lost compensating voltage. The crew will wait for the Source Ranges to automatically energize to prevent reenergizing the source ranges prematurely. D. N-36 is overcompensated. The crew will wait for the Source Ranges to automatically energize to prevent reenergizing the source ranges prematurely. Proposed Answer: A Explanation (Optional): With an IR is undercompensated, gamma current will be seen by the instrument, and indicated flux level for that channel will plateau at a point above P-6. N-35 exhibits this, and N-36 is at the bottom of its indicating range ("A" correct, "B" wrong). N-36 reads low, which could be an indication of overcompensation ("C" wrong), but it has the been >12 minutes since the trip, and power should be low. This is backed up by the fact that N-35 does not have characteristic - 1/3 DPM post trip SUR, which would be the case if N-36 was overcompensated. Technical Reference(s): Functional Dwg # 3 & 4 (Attach if not previously provided) AOP 3571, Att. E Proposed references to be provided to applicants during examination: None MC-05229 For the following conditions, determine the effects on the NIS system and (As available) Learning Objective: on interrelated systems:... Under-compensated intermediate range detector during reactor shut down operations... Question Source: Bank # 75472 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.6 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement I
POWER RANGE HIGH NEUTRON CT TRI P FLUX RATE REACTOR TRIP a' l_' TV I I , I I I Ei~Zh I 0- - - -- I
~- -- - - I I___
I I Ir I Il I m MANUAL TRI P (MAIN CONTROL BOARD) WNT~~~~ARYlETRMOA} S CiOjIT. (NOTE S Nfi "FSAR FIGURE" RESET REACTOR TRIP REACTOR TRIP (SHEET 2) [ZRPS
' -~---:CAD NOTE: ; "WA TOTHISI RtEISIONS VIENAS-BUILTAKEPRH~A 4 3 6 5
INTERMEDIATE RANGE REACTOR TRIP POWER RANGE REACTOR SOURCE RANGE REACTOR TRIP I c _- I C HIGH NEUTRON FLUX HIGH NEUTRON FLUX HIGH N REAC1 REACTOR TRIP (LOW SETPOINT)
<SH I SHEET 2) REACTOR TRIP (SHEET 2)
NOTES: CClSIST OF TVE DNTMCS ON THE CONTRDL TO THIS CIL LIGHTS AR CONNECTED THE REUCANT NUAL BL= C NRCLS TRAIN. 7 TvO PEI*4ISSIE STATLUS 5 I. uPPLISID BYoTHLR8 . BOARO FOR EACHRANGE.ONE FOR EACH INOIVIDJAL FOR EACH TRAIN.
- 2. I/N 33A IS IN LOGIC TRAIN A. & S5t0PuLE Y V-05 I/N 338 IS IN LOGIC TRAIN B.
- 3. IfN 38A IS IN LOGIC TRAIN A.
IflN 388 IS IN LOGIC TRAIN B.
- i. I/N 47A IS IN LOGIC TRAIN A.
I/N 479 IS IN LOGIC TRAIN 8. ItVIVIDJM. FOR EACHTRAIN.
-- 5 TWOD CD4TER IWIPTS ARE COV*ID TO THIS CIRCUIT.
ISTOF FMI EL. WOEFAR'CONTROLS IN TIE C R~. 6 IANAL RESET CONW NTE C TR L OC . ONEa~mTRO'. FOR EAC I NST J4M OW A 9 8 7 11I 10
Rpi E POWER RANGE I II 3a i I H1 514C FROM I/N 34A IR BYPASs r I I(SN H C T 3) 17 1 P-S (SHEET 160 C-l HIGI NEUTRON .Ld. ROD STOP
- CK, UTOMA,'Tke A MANUAL. D T14DRANNM) (5t4EET 9)
DOvi RANGE
;I C -2 vtRPOWER ROD STOP LOCK AUTOMATIC 1 MANUAL 00 WITHORA.WAL)
"FSAR FIGURE" (SHEEr S)
[3RPS * -SYS ] 12179-2472.011-001-004 NOTES: I.THE BYPASS SIGNALS ARE MADE UP BY MEANS OF TWO THREE-POSTIoT)
"dITCHES ON A N15 RACK. SWITCH I/N 45A BYPASSES EITHER. NC-41 L NC.'43L. SWITCH I/N458 BYPASSES EITHER NC-42L OR NC-+4L.
-ME TWO P-. 5157TA3LE NO. NC-35D AND NC-34.D ARE'ENERG.IED TO ACTUATE" 5UCH THAT A LOGIrC I SIGNAL IS DEFINED TO BE PRESENT WHEN THE B1STABLE OUTPUT V3LTAGE IS ON.
6 5 4 3 25212-39001 SH4005
i IV -L-a a Hs 1199UGIOI I, I POWER RANGE INTERMEDIATE RANGE FRON% JIM 35 iR BYPASS (SHEET 5'j P- 6 (SHEET 3) HIGH NEI (sHEETr 5 4 0 (SHEET 3 ) tCr RE0UtA0AJ,4r ROOC I POWER RANGE POWEP
--- N 4 T I
f I>Xs 7. P-8 (SHE~r S) I OVERPOWE. (BLOCK AUn (SKI NOTEl I.TH SB CF
- 2. T)
A IINT:CAD I WVA I1VISIWNSTO THIS DIOCUMENT I-AINCORP DCN DM13-00-0682-98 Mk
- 114NAS-BUlLTAREPROHIBITB~ED.
1 P.A# N.- DAYE REVISIONIS BY C = PP. II 10 9 I 8 7 6
Intermediate Range Nuclear Instrument Channel Failure The following annunciators are symptoms of an IR NI failure: IR HI FLUX ROD STOP MB4C 6-7 IR LOSS OF DET VOLTAGE MB4C 4-2 IR 1 LOSS OF COMPENSATION VOLTAGE MB4C5-2 IR 2 LOSS OF COMPENSATION VOLTAGE MB4C 6-2
- 1. If the intermediate range NI failure occurs during a reactor startup, determine the plant status and follow the applicable instructions:
- If power is below P- 6, Restore the IR NI to OPERABLE status prior to exceeding the P-6 setpoint.
- If power is above P- 6 AND below 5%, Restore the IR NI to OPERABLE status prior to exceeding 5% power (Mode 1).
- If power is GREATER THAN 5%, AND LESS THAN 10%, Restore the IR NI to OPERABLE status prior to exceeding 10% power.
- If power is above 10%, Place the level trip switch on the IR drawer in BYPASS AND, if not already actuated, actuate both intermediate range block pushbuttons.
- 2. If the intermediate range NI failure occurs during a reactor shutdown OR the channel has symptoms of undercompensation, when reactor power is believed to be in or near the source range, Actuate both source range reset switches.
- 3. When conditions have stabilized, Observe MB annunciators and parameters and immediately Report any unexpected or unexplained conditions to the Shift Manager.
AOP 3571 Page 2 of 2 Rev. 007 Attachment E Intermediate Range Nuclear Instrument Channel Failure NOTE No bistables are required to be tripped for this failure.
- 4. Refer to Tech. Specs. 3.3.1 and 3.3.3.5. See Table 3.3-1, Action 3 and Action 8.
- 5. Request I&C Department perform corrective maintenance on failed instrument.
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 33 Tier # 1 Group # 2 K/A # PLANT SPECIFIC Importance Rating N/A Proposed Question: Initial conditions:
- A hurricane warning is in effect for southeastern Connecticut.
- 100 mph winds expected in the next 6 hours.
- The crew has entered AOP 3569 "Severe Weather Conditions".
- A plant shutdown is being conducted.
What is the expected final condition of the plant when the actions specified in AOP 3569 are completed? A. RCS pressure at 2250 psia, with accumulator isolation valves open and PZR level at 28%. B. RCS pressure at 2250 psia, with accumulator isolation valves closed and PZR level at 60%. C. RCS pressure at 710 psia, with accumulator isolation valves open and PZR level at 60%. D. RCS pressure at 710 psia, with accumulator isolation valves closed and PZR level at 28%. Proposed Answer: C Explanation (Optional): With winds in excess of 90 mph, a loss of all AC power is a plausible event, and the crew is directed to cooldown the plant to less than 400'F to minimize potential for damage to the RCP seals; lower RCS pressure to above 700 psia ("A" and "B" wrong) to minimize RCS inventory loss through the RCP seals, with the accumulators available to inject ("B" and "D" wrong); and PZR level at 60% ("A" and "D" wrong) to maintain extra inventory available in the pressurizer ("C" correct). Technical Reference(s): AOP 3569, step 11 (Attach if not previously provided) AOP 3569, Attachment A OP 3208, Attachment I Proposed references to be provided to applicants during examination: None Learning MC-03971 Describe the major action categories contained within AOP (As available) Objective: 3569, Severe Weather Conditions. Question Source: Bank # 64019 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
---- I SEVERE WEATHER AOP 3569 Page 8 of 14 CONDITIONS Rev. 014 I
STEP ACTION/EXPECTED RESPONSI RESPONSE NOT OBTAINEDi
>~~ CA Q T I 0 N If offsite power is currently threatened or the grid is currently experiencing transients, the diesel generators shall NOT be placed in parallel with the grid.
NOTE
- The following step provides for the performance of actions prior to conditions occurring that would preclude their safe performance.
- Emergency Action Levels (EPIP Form 4400-3 Unit 3 EAL Table) specify the following event classifications:
UNUSUAL EVENT (GREATER THAN 75 mph onsite sustained wind speed) ALERT (GREATER THAN 90 mph onsite sustained wind speed)
- 11. Evaluate Weather Situation And Perform Applicable Actions:
- IF the wind speed is expected to exceed 75 mph in the next 8 hours, THEN Use the Emergency Action Levels (EPIP Form 4400-3 Unit 3 EAL Table) to aid in event classification and implementation of the Emergency Plan
- IF 90 mph sustained winds are expected within the next 8 hours, THEN Load test each EDG to GREATER THAN 49%
load for GREATER THAN 15 minutes
SEVERE WEATHER AOP 3569 Page 9 of 14 CONDITIONS Rev. 014
- 11. (continued)
IF the wind speed is expected to exceed 90 mph in the next 6 hours, THEN Perform an orderly plant shutdown using OP 3204, Power Operation, and establish the following conditions:
- RCS cold leg WR temperature LESS THAN 400 0F
- RCS pressure GREATER THAN 700 psia
- SI accumulator isolation valves open
- PZR level AT 60%
- IF sustained wind speed is expected to exceed 90 mph within the next 4 hours, THEN Consider an early ALERT notification to allow access time for emergency response personnel
SEVERE WEATHER AOP 3569 Page 1 of 2 CONDITIONS Rev. 014 Attachment A Guidance For Severe Weather Conditions
- 1. General Information Normally there will be ample warning of approaching storm conditions.
of This advance warning will allow time for the plant to be placed in a state to aid in standby readiness. CONVEX will provide the following information determining when this procedure is to be implemented.
- a. Position of storm, Latitude/Longitude, when storm center is 400 to 500 miles from the facility.
- b. Direction of storm movement.
- c. Speed at center of storm.
- d. Wind speed out X number of miles from center of storm.
- e. Forecast of storm pressure - reporting begins when the storm is headed in the general direction of some land mass.
- 2. Electrical Power and Equipment Concerns The switchyard and transmission lines cannot withstand wind speeds in excess of 90 mph. For this reason, it is best to shutdown to HOT STANDBY and run one EDG supplying one emergency bus. Use the EDG which the corresponds to the bus with the operating charging pump. Consider using EDG/charging pump combination on the electrical train with the most of the available safety equipment. RCP seal failure could occur as a result loss of all AC power, complicating the accident. Reducing temperature to LESS THAN 400'F prior to a loss of all AC makes seal failure unlikely.
SEVERE WEATHER AOP 3569 Page 2 of 2 CONDITIONS Rev. 014 Attachment A Guidance For Severe Weather Conditions High winds with little or no rain can create a hazardous condition when arcing of high voltage electrical equipment occurs due to salt spray being carried by the wind. To avoid damage to the switchyard equipment should arcing occur, the Control Room personnel should shift AC load and de-energize the equipment. If both the Reserve and Main Transformers require de-energizing and insufficient time is available for an orderly shut down of the plant, the reactor should be tripped. Visual surveillance of the switchyard, RSSTs, NSSTs, and main transformers for evidence of arcing is recommended.
- 3. Intake Structure Concerns Normal surveillance of the intake structure provides indication of potential problems with the screenwash and travelling screen system. Rough sea conditions make the intake structure vulnerable due to the seaweed, sand, and debris which could enter the structure and cause the trip of circulating water pumps due to high DP across the travelling screens.
- 4. Plant Cooldown Considerations When a cooldown is performed, it is important to maintain RCS pressure GREATER THAN 700 psia so that the SI accumulators do not inject. The accumulator isolation valves must remain open so that the accumulators are available in the event of a loss of all AC power. Maintain pressurizer level at approximately 60% to provide an additional margin to core uncovery. The end state of just under 400'F provides a stable condition consistent with ECA 0.0, Loss of All AC Power.
Attachment 1 RCS Cooldown Curves (Sheet 1 of 1) 2600
... _., S~w VIve 1-ift'-, ---'---------------1-- -
1; -,". -- I - ,------- I ........ C -- -4 - - ___4 - - - -I 11-
---- ------ j tI I i i :
1 ! ; I i q I 2400
-PRV pen F t 2200 S. -.... .. I i Temeraur 4'-
-t Nnal 2000
;. .1.. Operating 1.-j . 1-I, .
!5I! P-i-I- gI - ,1 1800
--1--t--
320 0 FATS----- 1600 I _______.. Cal 100°F i) U -1t [
.,. Subcoolin -- 1 iKr__lW L..) 1400 Ii N 50WF U1)
- 01) Subcooling
, i- -- , Limt-,
C,) .... j_. _ C-) 1200 _ N T...... c-. -rPS- --- 1OOC ..- , _ p C 80C Minimum Pressure for RCP Operation 60( 400 I-
-I-I 200 RCPs off I i Head4 Bolt -- -1 w, -- _+ ..
Lim it 1~RCSeal LeakoffIso oLlmit -ated +t 500 600 0 100 200 300 400 RCS Cold Leg Temperature (Degrees F) OP 3208 Level of Use Ai Rev. 020-03 Continuous 99 of 103
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 34 (SRO) Tier # 1 Group # 2 K/A # 038.EKI.02 Importance Rating 3.5 Proposed Question: Current Plant Conditions:
- A SGTR has occurred
- After cooldown and depressurization, ECCS flow was terminated
- Normal charging and letdown have been established
- Pressurizer level is 61% and slowly increasing
- Ruptured SG level is 65% and increasing What are all of the required actions to be taken per E-3, step 27 "Control RCS Pressure And Charging Flow To Minimize RCS-to-Secondary Leakage"?
A. Increase charging flow AND maintain RCS and ruptured SG pressures equal. B. Turn on the pressurizer heaters. C. Maintain RCS and ruptured SG pressures equal. D. Depressurize the RCS AND decrease charging flow. Proposed Answer: D Explanation (Optional): indicative of RCS Step 27 table. Ruptured SG level is increasing with pressurizer level between 50% and 73%. This Required action is to depressurize the RCS, to secondary leakage occurring, with excess inventory in the RCS. decrease charging flow ("A" wrong, "D" correct). stopping primary to secondary leakage ("B" and "C" wrong) and that may be taken at step 27, depending on conditions.
"A", "B", and "C" are plausible since they all are actions Technical Reference(s): E-3, Step 27 (Attach if not previously provided)
WOG E-3 Bkgd doc, step 30 None Proposed references to be provided to applicants during examination: MC-04371 Describe the major action categories within EOP 35 E-3. (As available) Learning Objective: Question Source: Bank # 60617 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.8, 41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
5Lk STEAM GENERATOR TUBE RUPTURE EOP 35 E-3 Rev. 017 Page 27 of 39 J SDRESPONSE RESPONSE NOT OBTAINED l Q A UT I0 N the ruptured SGs pressures must be maintained LESS the ruptured SGs atmospheric dump valve controller setpoint. 1%
- 27. Control RCS Pressure And Charging Flow To Minimize RCS-To-Secondary Leakage
-_a. Perform appropriate actions from table RUPTURED SGs LEVEL INCREASING DECREASING OFFSCALE HIGH PZR LEVEL l_____ _______
Increase charging Increase charging flow flow LESS THAN 25% AND Increase AND (50% ADVERSE CTMT) Depressurize charging flow Maintain RCS and RCS using ruptured SGs step 27.b. pressures equal BETWEEN 25% and 50% Depressurize Turn ON Maintain RCS and RCS using PZR ruptured SGs step 27.b. heaters pressures equal BETWEEN 50% and 73% Depressurize lbrn ON Maintain RCS and (50% and 63% ADVERSE step 27.b. PZR ruptured SGs, AN1M1)heaters pressures equal CTfT) Decrease charging flow Maintain RCS and GREATRXTHADV3ETurn GREATVERSTEA CT3% Decrease PZR ON ruptured SGs (3ADES ') charging flow heaters pressures equal
STEAM GENERATOR EOP 35 E-3 Page 28 of 39 TUBE RUPTURE Rev. 017
-~ I I1 ACTIN/EPECED RSPOSH ESPONSE NOT OBTAINED I
- 27. (continued)
- b. Check table requires - b. Proceed to NOTE prior to step DEPRESSURIZATION 28. and, IF depressurization is required, THEN Perform step 27.c.
- c. Depressurize using normal c. Perform the applicable action:
PZR spray as determined in
- IF normal letdown is in step 27.a. service, THEN Proceed to step 27.e.
- IF normal letdown is NOT in service, THEN Use one PZR PORV as necessary and Proceed to NOTE prior to step 28.
- d. Proceed to NOTE prior to step 28.
- e. Establish auxiliary spray e. Return to step 27.c.
- 1) Unlock and OPEN auxiliary spray valve (3RCS*AV8145)
- 2) CLOSE charging header loop isolation valves:
- 3CHS*AV8146
- 3CHS*AV8147
- 3) Throttle charging flow controller to adjust and maintain auxiliary spray
STEAM GENERATOR EOP 35 E-3 Page 29 of 39 TUBE RUPTURE Rev. 017 ACTION/EXPECTED RESPONS SP RESPONSE NOT OBTAINED
- 27. (continued)
- f. Check either of the following: f. Proceed to NOTE prior to step 28. and,
- Stopping depressurization WHEN
- DESIRED Either condition is met, THEN OR Perform step 27.g.
- Annunciator REGEN HX LETDOWN TEMP HI (395-F) (MB3A 5-4)
-LIT
__g. Establish charging
- 1) OPEN one charging header loop isolation valve
- 3CHS*AV8146
- 3CHS*AV8147
- 2) CLOSE auxiliary spray valve (3RCS*AV8145)
- 3) Adjust charging flow controller as desired
3L1 STEP DESCRIPTION TABLE FOR E-3 Step 30 - CAUTION CAUTION: RCS and ruptured SG(s) pressures must be maintained less than the ruptured SG(s) PORV setpoint. PURPOSE: To alert the operator that both the ruptured SG pressure and RCS pressure must be maintained less than the SG PORV setpoint to prevent lifting of the PORV and code safety valve BASIS: Since flow can occur between the RCS and the secondary side of the ruptured steam generator via the failed tube, the pressure in the primary system will affect the pressure in the ruptured steam generator and vice versa. If RCS pressure is greater than the SG PORV setpoint, primary-to-secondary leakage will compress the steam bubble in the ruptured steam generator and increase pressure until the PORV eventually lifts. Therefore, the RCS and ruptured steam generator pressures should be maintained less than the SG PORV setpoint. ACTIONS: '--'Check RCS and ruptured SG pressures INSTRUMENTATION: 0 Wide range RCS pressure indication 0 SG pressure indication CONTROL/EOUIPMENT: N/A KNOWLEDGE: N/A PLANT-SPECIFIC INFORMATION: N/A E-3 136 HP-Rev. 1C HE3
STEP DESCRIPTION TABLE FOR E-3S Step 30 STEP: Control RCS Pressure And Charging Flow To Minimize RCS-To-Secondary Leakage PURPOSE: To control RCS pressure and charging flow to maintain an indicated pressurizer level while minimizing primary-to-secondary leakage. BASIS: In order to explain the basis for the guidance provided in this step, consider again equilibrium conditions between leakage through the failed SG tube and charging flow, as shown in Figure 30. For primary system pressures greater than the ruptured steam generator pressure (PSG)' primary-to-secondary leakage will occur so that excess charging flow, i.e., greater than letdown and coolant shrinkage, is necessary to maintain pressurizer inventory. Conversely, for letdown flows greater than charging flow, the equilibrium RCS pressure is less than the ruptured steam generator pressure and secondary-to-primary leakage will occur. The ideal conditions, shown by Point B, occur when charging flow exactly compensates for letdown and coolant shrinkage so that RCS pressure and the ruptured steam generator pressure equalize. For these conditions both the pressurizer and ruptured steam generator inventories will remain constant. Obviously fluctuations about these ideal conditions will occur due to variations in ruptured steam generator pressure, cooldown rates, and letdown flows. Consequently, the operator must continuously adjust RCS pressure and charging flow to control pressurizer and ruptured steam generator inventories. This step provides guidance for performing these actions in the form of a table. Figure 30 can be divided into four different regions which are characterized by pressurizer and ruptured steam generator level behavior. For primary pressures greater than the ruptured steam generator pressure, leakage into the steam generator will increase steam generator water level (LSG). Alternatively, water level will decrease for RCS pressures less than the ruptured steam generator pressure. Similarly, pressurizer level (LpRzR) will increase for RCS pressures less than equilibrium. This leads to the four regions illustrated in Figure 30. The steps one performs to stabilize the plant at the ideal, equilibrium conditions depend on the pressurizer inventory and ruptured steam generator water level behavior. For example, if pressurizer level is low, region II or region III must be entered to increase pressurizer level. This requires one to increase charging flow or decrease RCS pressure, as shown in Figure 30. The further into these regions, the more rapidly pressurizer level will increase. Of course, if pressurizer level is high, the opposite response would be necessary. However, the ruptured steam generator water level must also be considered. E-3 137 HP-Rev. 1C HE3
-- STEP DESCRIPTION TABLE FOR E-3 Step 30 If the steam generator water level is increasing, RCS pressure must be reduced to stop primary-to-secondary leakage. If the steam generator water level is decreasing, primary pressure should be increased by energizing pressurizer heaters to minimize leakage into the RCS.
Note that in some cases, actions which address pressurizer level conflict with those which address steam generator level. For example, if steam generator level is increasing one must decrease RCS pressure. Since this will also increase pressurizer level, the pressurizer could fill with water if level is initially high. However, by reducing charging flow, pressurizer level will decrease. Since this will also decrease RCS pressure if heaters are not energized, steam generator level will also stabilize. Hence, for this situation the preferred action is to reduce charging flow. In other situations, controlling either RCS pressure or charging flow or a combination of both will work equally well. In general, steam generator level provides a more direct and more accurate indication of leakage between the primary and secondary than RCS and steam generator pressure indications. However, if level is offscale so that no trend can be observed, one must rely on RCS and ruptured steam generator pressures as indirect indications of leakage. While instrument inaccuracies may lead to some primary-to-secondary leakage, such leakage will be contained provided RCS and ruptured steam generator pressures are maintained below the steam generator PORV setpressure. Normal pressurizer spray is the preferred means of controlling RCS pressure. If it is not available, auxiliary spray should not be us-d if letdown is in service. A pressurizer PORV should be used instead of auxiliary spray if letdown is not in service to minimize thermal stresses on the spray nozzle. ACTIONS: o Monitor ruptured SG narrow range level and pressure o Monitor RCS pressure and PRZR level o Adjust charging flow control valve and start charging pumps o Energize PRZR heaters o Spray PRZR with normal or auxiliary spray o Operate one PRZR PORV o Establish letdown and excess letdown E-3 138 HP-Rev. IC HE3
( W ( w W ( Im rn M I Figure 30. ILLUSTRATION OF RCS AND RUPTURED SG PRESSURE EQUILIBRIUM RCS Pressure 0, (rO Excess Letdown Flow - Excess Charging Flow I-
.--- a0 I
ED n-
STEP DESCRIPTION TABLE FOR E-3 Step 30 INSTRUMENTATION: o SG narrow range level indication o SG pressure indication o Charging flow indication o PRZR heaters status indication o PRZR spray valve position indication o PRZR auxiliary spray valve position indications o PRZR PORV and block valve position indications o Position indications for letdown and excess letdown valves o Wide range RCS pressure indication o PRZR level indication CONTROL/EOUIPMENT: o Charging flow control valve o PRZR heater controls o Normal PRZR spray valves o Auxiliary PRZR spray valves o Valves for letdown and excess letdown o PRZK PORVs o Plant specific controls for letdown isolation and orifice isolation valves KNOWLEDGE: o Use of table provided in this step o "Charging" flow refers to total flow into the RCS from the charging pumps. If the PRZR is solid, charging flow will also control RCS pressure. In that case, flow must be controlled to maintain RCS subcooling greater than instrument uncertainties and control ruptured steam generator inventory. o The operator is expected to continue with subsequent steps while performing actions prescribed in this step. Reference to this table should be made throughout the remainder of this guideline as necessary to control RCS and ruptured steam generator inventories until an alternate cooldown guideline is implemented. o Steam generator level increasing or decreasing refers to conditions which may lead to overfill or to uncovering the U-tubes before corrective actions can be taken. For tube failures in multiple steam generators, actions should be based on the steam generator most likely to overfill. E-3 139 HP-Rev. 1C HE3
STEP DESCRIPTION TABLE FOR E-3 Step 30 PLANT-SPECIFIC INFORMATION: o This step contains detailed guidance for controlling RCS pressure and charging flow to maintain reactor coolant inventory and minimize leakage between the primary and secondary. It is possible to simplify or replace this table on a plant specific basis in combination with operator training. o (D.08) PRZR level at the upper tap, including allowances for normal channel accuracy, minus 20% for operating margin. Upper tap refers to the top of the indication range. o (D.09) PRZR level at the upper tap, including allowances for normal channel accuracy, post-accident transmitter errors, and reference leg process errors, minus 20% for operating margin, not less than 50%. A lower limit of 50% is imposed to provide margin to draining the pressurizer for control of RCS inventory. Upper tap refers to the top of the indication range. o (D.06) PRZR level above the top of the heaters, including allowances for normal channel accuracy and reference leg process errors. o (D.07) PRZR level above the top of the heaters, including allowances for normal channel accuracy, post-accident transmitter errors, and reference leg process errors, not to exceed 50%. An upper limit of 50% is imposed to provide margin to filling the pressurizer for RCS inventory control. E-3 140 HP-Rev. 1C HE3
Written Examination Question Worksheet Form ES-401-6 ES-401 Question # 35 Examination Outline Cross-reference: Tier # I Group# 2 K/A # 054.AK3.03 Importance Rating 4.1 Proposed Question: Conditions five A complete loss of main feedwater has occurred, resulting in a reactor trip. minutes after the trip are as follows:
*All equipment operated as designed. *No operator actions have been taken, other than E-0 immediate actions. *The crew has just entered EOP 35 ES-0. 1 "Reactor Trip Response". *PZR level is 25% and slowly decreasing. *SGnarrow range levels are off-scale low. *SteamGenerator pressures are approximately 990 psig and slowly decreasing. *Tave is 5450 F and slowly decreasing. *RCS pressure is 2020 psia and slowly decreasing.
What action does ES-0. 1 direct the crew to take, and why? A. Throttle AFW flow to between 530 and 600 gpm to stop the cooldown. hammer in the feed ring. B. Throttle AFW flow to between 530 and 600 gpm to prevent water C. Maintain full AFW flow to limit peak RCS pressure on a loss of feed event. D. Maintain full AFW flow to cover the SG U-Tubes. Proposed Answer: A not required to be performing an Explanation (Optional): The balance of plant operator may, at any time when flow if minimum heat sink requirements are satisfied. This immediate action or sequenced steps, throttle AFW ES-0. step I I has the crew throttle AFW if a cooldown is in includes throttling flow to minimize RCS cooldown. at full flow on the trip. "C" is plausible, since full AFW progress. "A" is wrong since AFW has already initiated peak RCS pressure. "D" is plausible, since keeping the tubes flow is assumed for a loss of feed ATWS to limit wetted is a concern if heat sink has been lost, and flow is being restored. Technical Reference(s): ES-0. I step I (Attach if not previously provided) OP3272 Attachment 3 None Proposed references to be provided to applicants during examination: MC-055 12 DISCUSS the basis of major procedure steps (As available) Learning Objective: &/or sequence of steps in EOP 35 ES-0.1. Question Source: Bank # 70200 Question Cognitive Level: Comprehension or Analysis 10CFR Part 55 Content: 55.41.5, 41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
RESPONSE NOT OBTAINED I F VW" CAUTION If SI actuation occurs during this procedure, immediately Go to E-0, Reactor Trip or Safety Injection. 4Qwwwwww- - - - - - - - - -
- 1. Check RCS Temperature
- a. Verify RCS cold leg WR a. Perform the applicable action:
temperature - BETWEEN 5500 FAND
- IF temperature is 5600 F GREATER THAN 560 0F, THEN Proceed to step 1.c. I
- IF temperature is LESS THAN 55 0 cF, THEN Proceed to step i.e.
- b. Proceed to step 2.
. r OBTAINED
- 1. (continued)
*c. Dump steam to condenser c. Dump steam to atmosphere using SG atmospheric dump
- 1) Verify the following: valves or the SG atmospheric dump bypass valves (MB or
- MSIVs - AT LEAST local y).
ONE OPEN
- Annunciator "CONDENSER AVAIL FOR STM DUMP C-9" (MB4D 5-6) - LIT
- 2) Adjust steam pressure controller to obtain zero output in MANUAL
- 3) Transfer condenser steam dumdps to Steam Pressure Mode
- 4) Place both condenser steam dump interlock selectors - ON
- 5) Adjust steam pressure controller to dump steam to condenser
- d. Proceed to step 2.
- e. Maintain total feed flow BETWEEN 530 and 600 gpm until NR level is GREATER THAN 8% in at least one SG
- f. CLOSE SG atmospheric dump and dump bypass valves
- g. STOP the MD FW pump and Place control switch in PULL-TO -LOCK
REACTOR TRIP RESPONSE EOP 35 ES-0.1 Page 5 of 21 Rev. 019 ACTIN/EXECTE RESONSHRESPONSE NOT OBTAINED I
- 1. (continued)
.h. TRIP the TD FW pumps
- i. Check SG code safety valves i. Consult Duty Officer to closed determine if safety valve(s) not closed should be gagged using
- Flow switches (MB5) - GA-17. I NOT LIT
- Local observation of safety valves (MSVB roof) -
NO STEAM OBSERVED
- j. Verify RCS cold leg WR j. Place both condenser steam temperature - dump interlock selector STABLE OR INCREASING switches to OFF.
IF RCS cooldown continues, THEN CLOSE the MSIVs and MSIV bypass valves.
- k. Verify RCS cold leg WR k. Perform the following:
temperature - GREATER THAN 5300 F 1) CLOSE the MSIVs and MSIV bypass valves.
- 2) Using AOP 3566, "Immediate Boration,"
Immediate borate.
Attachment 3 Special Considerations (Sheet 2 of 10) Adverse Containment For parameters that can be affected by adverse containment conditions, Adverse Containment values are provided in parenthesis following the normal containment value and should be utilized by the operator when either specified adverse containment criterion is exceeded. After attaining an adverse value, the following rules apply when either containment parameter subsequently decreases below the specified value:
- When Ctmt temperature decreases to less than or equal to 180'F, Adverse Ctmt values no longer apply.
5
- When Ctmt radiation decreases to less than or equal to 10 R/hr, Adverse Ctmt values continue to apply.
IsolatinglThrottling AFW Flow The balance of plant operator may, at any time when not required to be performing an immediate action or sequenced steps, throttle AFW flow if -' x minimum heat sink requirements are satisfied. This includes the following:
- Isolating AFW flow to a faulted SG
- Throttling flow to minimize RCS cooldown (flow to all SGs should be throttled evenly including any ruptured SG)
However, the SM/US should direct isolating AFW flow to a ruptured SG when the minimum WR or NR ruptured SG level specified in E-3 is satisfied. Isolating Ruptured SG Steam Line An operator may not isolate the steam line to a ruptured steam generator until directed by the procedure. This is noi an identified safe condition unless specifically directed by the SGTR procedure. Checking for Faulted SGs When checking for a faulted SG, use the steam pressure recorders on MB2 for proper trending accuracy.
.... OP 3272 Level of. Use.. l
.... . 26.of.44 Rev. 008 fo o. ,F.
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 36 Tier # 1 Group # 2 K/A # W/EO5.EK3.1 Importance Rating 3.8 Proposed Question: The crew is responding to a loss of Heat Sink and has just entered FR-H. 1 "Response to Loss of Secondary Heat Sink". Why will the crew be directed to trip the RCPs during this event? A. Allow more time to restore feedwater to the S/Gs prior to the need to bleed and feed, since RCP operation adds a significant amount of heat to the RCS. B. Reserve the Reactor Coolant Pumps for future use, since the RCS will be will become highly voided in the event that bleed and feed conditions are required. C. Prevent damage to the Reactor Coolant Pumps during the upcoming depressurization of the secondary plant, required to allow feeding the SGs with a condensate pump. D. Minimize RCS inventory loss in the event of a small break LOCA, reducing the amount of core uncovery should the RCPs trip later in the event. Proposed Answer: A Explanation (Optional): "A" is correct, since RCP operation results in heat addition to the RCS water. By tripping the RCPs, the effectiveness of the remaining water inventory on the SG's is extended, which extends the time at which operator action to initiate bleed and feed must occur. "B" is plausible, since this is a basis for tripping RCPs in FR-C. 1. "C" is plausible, since this is a basis for tripping an RCP if FR-C.2. "D" is plausible, since this is the basis for tripping RCPs on the E-0 foldout page. Technical Reference(s): FR-H. 1 WOG Bkgd Doc, step 4 (Attach if not previously provided) Proposed references to be provided to applicants during examination: None Learning MC-04535 Discuss the basis of major procedure steps and/or sequence of (As available) Objective: steps in EOP 35 FR-H. 1 through H.5 Question Source: Bank # 63975 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.5, 41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
STEP DESCRIPTION TABLE FOR FR-H. S tep 4 STEP: Stop All RCPs PURPOSE: To stop RCPs in order to extend the time to restore feed flow-to the SGs BASIS: RCP operation results in heat addition to the RCS water. By tripping the RCPs, the effectiveness of the remaining water inventory in the SGs is extended, which extends the time at which the operator action to initiate bleed and feed must occur. This extension of time is additional time for the operator to restore feedwater flow to the SGs. Additional information is provided in subsection 2.5, Reactor Coolant PumD Operation, of this background document. ACTIONS: Stop all RCPs INSTRUMENTATION: RCP status indication CONTROL/EOUIPMENT: Switches for RCPs KNOWLEDGE: Stopping all RCPs will result in an interim plant transient on RCS pressure and temperature as natural circulation flow"conditions are established in the RCS. An example of this is shown in Figures 6 and 7 where RCS pressure and temperature rise and reestablish new steady state conditions prior to steam generator dryout occurring. If rising RCS pressure and hot leg temperatures are the criteria for initiation of bleed and feed heat removal, the operator must evaluate whether these conditions are caused by an RCP trip or by a loss of secondary heat sink in order to determine if bleed and feed heat removal is to be established. FR-H.1 71 HP-Rev. IC HFRH1
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 37 Tier # I Group # 2 K/A # 058.AK3.02 Importance Rating 4.2 Proposed Question: With the plant at 100% power, Battery Bus 1 (30 1A-1) deenergizes, and the crew enters AOP 3563 "Loss of DC Bus Power". Why does AOP 3563 direct the crew to trip the reactor and enter E-0 "Reactor Trip or Safety Injection"? A. SG levels all shrink out of the narrow range, since all 4 MSIVs close. B. Condenser vacuum will be lost, since the SJAE steam supplies isolate. C. The reactor will trip, since power is lost to 2 of the 3 ETS pressure switches, resulting in SSPS generating a low ETS pressure reactor trip. D. A feedwater transient is in progress, since the MSR Drain Tank and Feed Heater emergency level control valves have failed open. Proposed Answer: A Explanation (Optional): "A" is correct, since these components are powered from Batteries I and 2. "B" is wrong since these valves are powered from Battery 5. "C" is wrong, since 2 of the 3 ETS pressure switches are powered from Battery 5. "D" is wrong, since battery 6 powers these valves. Technical AOP 3563, step 1 (Attach if not Reference(s): AOP 3563, Att A, pg 5, Att B, pg 5, Att E, pg 4, 6 and 7, Att F, pg 2 previously provided) Proposed references to be provided to applicants during examination: None Learning Objective: MC-03947 Discuss the basis of major precautions, procedure steps or (As available) sequence of steps. Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.5, 41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
P ACTION/EXPECTED RESPONSRNE NOTE If DC Bus 301A- 1, 301B- 1, 301C-1 or 301D-1 is the affected bus, the reactor must be manually tripped and the actions in E-0, Reactor Trip or Safety Injection shall be performed. Attachments A, B, E or F respectively, of this procedure, provide additional guidance which may be used concurrently with E-0.
- 1. Perform The Applicable Actions Based On Abnormal MB Annunciators and Indications
- Check Bus 301A-1 - Perform the following:
ENERGIZED
- 1) Trip the reactor.
- 2) Go to E-0, Reactor Trip or Safety Injection.
- Check Bus 301B-1 - Perform the following:
ENERGIZED
- 1) Trip the reactor.
- 2) Go to E-0, Reactor Trip or Safety Injection.
- CheckBus 301A Use Attachment C.
ENERGIZED
- CheckBus 301B Use Attachment D.
ENERGIZED
STEP ACTION/EXPECTED RESPONS RESPONSE NOT OBTAINED
- Check Bus 301C-1 - Perform the following:
ENERGIZED
- 1) Trip the reactor.
- 2) Go to E-0, Reactor Trip or Safety Injection.
- Check Bus 301D-1 - Perform the following:
ENERGIZED
- 1) Trip the reactor.
- 2) Go to E-0, Reactor Trip or Safety Injection.
- 2. Continue With Normal Plant Evolutions Using Applicable Plant Procedures
-FINAL-
31 LOSS OF DC BUS POWER AOP 3563 Page 5 of 8 Rev. 5 Attachment A Loss of DC Bus 301A- 1 (Battery Bus 1) 301A-1A2 (3BYS*PNL2F)
- SIH test lines isolate, cannot fill accumulators from SI pumps
- Accumulator nitrogen supply isolates (3SIL*CV8968)
-Main steam valve building exhaust isolates (3HVV*AOD50B closes, FNIB trips)
-RHR test lines isolate (3SIL *CV8890A and 3SIL *CV8890B close)
-Charging pump cooling cross-connect valves isolate (3CCE*AOV30A and 3CCE*AOV26A close)
-Letdown isolates (3CHS*CV8160closes)
- Normal RCS makeup isolates (3CHS*FCV11OB and 3CHS*FCV111B close)
- Ctmt atmosphere moniter isolates (3CMS*CTV20 and 3CMS*CTV23 close)
-PRT nitrogen supply isolates (3GSN*CTV105 closes)
- Ctmt drain isolates (3DAS*CTV24 closes)
- PRT and accumulator sample lines isolate
- 301A-1A3 (3BYS-PNL13F)
- Auxiliary feed valves fail open (SG A 3FWA *HV3L4 and 3FWA *HV32A, SG D 3FWA *HV31D and 3FWA *HV132D, SOG B 3FWA *HV36B, SG C 3FWA *HV36C)
- Auxiliary feed water cross-connect isolates (3FWA*AOV62A closes)
- Terry turbine steam supply valves open (3MSS*AOV31A, 3MSS*AOV31B, and 3MSS*AOV31D)
- DWST. auxiliary feed suction opens (3FWA*AOV61A)
- Reactor vessel flange leakoff opens (3RCS*AV8032)
- Charging pump cooling temperature control valve opens (3CCE*TV37B)
-Ctmt purge inlet damper closes (3HVR*AOD174B)
- 301A-1A4 (3BYS*PNL15F)
- Steam generator feedwater isolates (feedwaterregulatingvalves 3FWS*FCV510, 3FWS*FCV520, 3FWS*FCV530, and 3FWS*FCV540 close)
-Steam generator feedwater isolates (feedwaterregulatingbypass valves 3FWS*LV550, 3FWS*LV560, 3FWS*LV570, and 3FWS *LV580 close)
- Main steam isolation valves close
- Main steam isolation bypass valves close
- MSIV and Terry tubine upstream drain valves close
- DWST heater recirculation isolates (3FWA *AOV/25 closes)
LOSS OF DC BUS POWER AOP 3563 Page 5 of 9 Rev. 5 Attachment B Loss of DC Bus 301B- 1 (Battery Bus 2) 301B-1A2 (3BYS*PNL4F)
- SIH test line isolates, cannot fill accumulators from SI pumps
- Accumulator nitrogen supply isolates (3SIL*CV8880)
- Main steam valve building exhaust isolates (3HVV*AOD50A closes, FNIA tnris)
- Charging pump cooling cross-connect valves isolate (3CCE *AOV30B and 3CCE*AOV26B close and temperaturecontrolvalve 3CCE*TV37B opens)
- Letdown isolates (3CHS*CV8J52 closes)
- Ctmt drain isolates (3DAS*CTV25 closes)
- Excess letdown diverts to the VCTICDTr(3cHs*AV8143failsto VCTposition)
- Auxiliary feed flow control valves open (SGA 3FWA *HV36A, SG B 3FWA *HV31B, 3FWA *HV32B, SG C 3FWA *HV32C, 3FWA *HV31 C, SG D 3FWA *HV36D)
- Terry turbine steam supply valves open (3MSS*AOV31A, 3MSS*AOV31B, 3MSS*AOV31D)
- Excess letdown isolates (3RCS*AV8153 closes)
- MD AFW pump B DWST suction to opens (3FWA*AOV61B)
- Auxiliary feed water cross-connect isolates (3FWA *AOV62B closes)
- RCP B thermal barrier return isolates (3CCP*AOV178B closes)
- Chilled water to Ctmt isolates (3CDS*CTV91B closes)
- 301B-1A3 (3BYS-PNL14F)
- None
- 301B-1A4 (3BYS*PNL16F)
- Feedwater isolation trip valves close (3FWS*CTV41A, 3FWS*CTV41B, 3FWS*CTV41C, and 3FWS*CTV41D close)
- Main steam isolation valves close
- Main steam isolation bypass valves close
- MSIV and Terry turbine upstream drain valves close
- DWST heater recirculation isolates (3FWA *AOV26 closes)
LOSS OF DC BUS POWER AOP 3563 Page 4 of 8 Rev. 5 Attachment E Loss of DC Bus 301C- 1 (Battery Bus 5) 301C- 1B (continued) Loss of power to auxiliary relay rack 3 (3RPS-RAK4UXC)
- BTRS letdown reheater heat exchanger TCV isolates (3CHS*TCV381A closes)
- BTRS letdown reheater heat exchanger divert valve (3CHS*TCV381B) opens
- BTRS letdown chiller heat exchanger isolates (3CHS*AT17002A and 3CHS*AV7002B closes)
- BTRS outlet valve isolates (3CHS*7022 closes)
- BTRS demineralizer top/bottom divert valve repositions (3CHS*AV7057fails to bottom of demineralizer)
- BTRS chiller heat exchanger divert valve (3CHS*AV7040) opens
- BTRS chiller heat exchanger bypass valve (3CHS*AV7041) opens
- BTRS demineralizer dilution outlet valve (3CHS*AV7045) opens
- BTRS demineralizer boration outlet valve isolates (3CHS*Av7046 closes)
- Condenser steam dumps fail closed
- 301C-lAl (BYS-PNL-5F)
- CRDM shroud cooler inlet (3CDS-AOV53A) opens
- All DGS pumps stop
- Accumulator fill valve isolates (3SIL*AV8878A closes)
- Reactor plant gaseous drains isolate (3DGS*CTV24 closes)
- Shutdown air compressor (3IAS-AOV95) opens
- Neutron shield tank surge tank LCV isolates (3PGS-LV49 closes)
- PRT vent isolates (3RCS-PCV469 closes)
- Various check valve leakage and accumulator test lines valves close
- SJAEs steam supply isolates (3ASS-AOV22A and 3ASS-AOV22B close)
- Automatic CO2 actuation not functional
- Main transformer primary protection not functional
51 LOSS OF DC BUS POWER AOP 3563 Page 6 of 8 Rev. s Attachment E Loss of DC Bus 301C- 1 (Battery Bus 5) 301C-1A4 (BYS-PNL11F)
- Condenser air removal pump suction isolates (3ARC-AOV22A closes)
- Various steam line trap bypass valves open
- Condenser water box vacuum priming AOVs close
- Panel 3CES*PNLBD10 deenergizes
- Auto open feature for DTM non-return MOVs not functional 301C-1A6 (BYS-PNL33F)
- Auxiliary steam to boric acid batch tank isolates (3ASS-TCV100 closes)
- Auxiliary condensate pump discharge isolates (3CNA-AOV56 closes)
- Failed fuel detector PGS purge isolates (3CHS*AOV64 closes)
- Spent fuel pool HVAC hot water heating valve (3HVH-TV61) modulates
- Degasifier shuts down and letdown aligns to VCT
- Panel 3CES*PNLBR20 deenergizes (Automatic VCTmakeupnotfunctional)
- Loss of control power for busses 32C and 32M
- GWS panel deenergizes
- Rod drive power supply test cabinet deenergizes
- Reactor trip signal to channel III from No. 3 turbine stop valve
- ETS lo pressure to SSPS (3TMB-PS15OC deenergizes)
- 301C-1A7 (BYS-PNL34F)
- LWS evaporator shuts down
- Boron test tank isolates
- 301C-1B1 (BYS-PNL6F)
- CRDM shroud cooler inlet (3CDS-AOV53B) opens
- Various check valve leakage and accumulator test lines valves close
- RWST recirculation cooler TCV (3CDS-TV26) opens
- Accumulator fill isolation isolates (3SIL*AV8878B)
- RPCCW surge tank makeup valve isolates (3CCP-LV20closes)
- SI pump cooling surge tank makeup isolates (3CCP*LV61 closes)
- Charging pump cooling surge tank makeup isolates (3CCP*LV91 closes)
- Spent fuel pool makeup isolates (3SFC-LV44)
- Chilled water surge tank makeup valve isolates (3CCP-LV74 closes)
- Containment vacuum air ejector suction isolates (3CDS*AOV23 closes)
- Main generator breaker primary open/closed power deenergizes
- Main generator breaker backup protection not functional
LOSS OF DC BUS POWER AOP 3563 Page 7 of 8 Rev. 5 Attachment E Loss of DC Bus 301C-1 (Battery Bus 5) 301C-1B2 (BYS-PNL8F)
- Various post accident sample valves close
- Charging to loops 2, 3, and 4 fill lines isolates (3RCS*AV8036B, 3RCS*AV8036C, 3RCS*AV8036D close)
- BTRS demineralizers 3B and 3D (3CHS*AV7010B and 3CHS*Al.7010D) open
- Loop 2 and Loop 4 drain valves isolate (3RCS*AV8037B and 3RCS*AV8037D close)
- BTRS isolates (3CHS*AV7054 closes)
- Failed fuel detector inlet (3CHS*AV68) opens
- PZR spray valve (3RCS*455B) closes
- RCP B and RCP D #1 seal leakoff isolation valves open
- Steam generator blowdown flow control valves close
- House DC electrical power lost to EHC System
- RSST trip tone not functional
- Generator line protector not functional
- 301C-1B3 (BYS-PNL1OF)
- B TDFW pump recirculation valve (3FWR-FV21B) opens
- Condensate minimum flow recirculation valve (3CNM-FV48) opens
- Fish trough water spray isolates (3SWT-AOV33 closes)
- Condenser steam dump bank C isolates
- Reheater drain tank B level control valve (3DSR-AOV2OB) opens
- Cooling to CAR fan A, CAR fan B, and neutron shield tank, isolates (3CDS-AOV45B, 3CDS-AOV45C, 3CDS-AOV46B, 3CDS-AOV46C close)
- Hot water heating to auxiliary building, fuel building, waste building, ESF building isolates (3HVH-AOV135A, 3HVH-AOV135B, 3HVH-136A, and 3HIVH-136B close)
- Panel 3CES*PNLBS30 deenergizes
- Reactor Trip signal to channel I from No. 1 turbine stop valve
- ETS lo pressure to SSPS (3TMB-PS15OA deenergizes)
- Panel 3CES*PNLBD1P deenergizes
- Automatic open function for ESS non-return valves deenergized
- Condenser water box vacuum primary AOVs close
Loss of DC Bus 301D-1 (Battery Bus 6) OTHER EFFECTS FROM LOSS OF DC BUS 301D-1
- Control power for Busses 32D, 32L, and 32Q not functional
- The main generator is affected as follows:
- Auto and manual voltage control is not functional
- Control power for the exciter and field breakers is not functional (require local tripping)
- Auxiliary boilers auto shut down
- SJAE discharge valves close resulting in a loss of condenser vacuum
- VCT divert valve fails (3CHS*112A) to VCT and does not reposition on high VCT level
- Automatic control of PZR backup heater group A is not functional
- Control power to RSST B supply breaker not functional
- Local panels deenergize causing the following systems to shutdown:
- Condensate demineralizer liquid waste
- Hot water heating
- Chlorination for service water
- Water treatment
- Auxiliary boiler
- Condensate demineralization
- Bus duct cooling
- Moisture separator drain tank emergency level control valves fail open (3DSM-LV2OA2 and 3DSM-LV2OB2)
- Auxiliary condensate divert valve fails to the auxiliary building sump position (3CNA-AOV48)
- Transformer backup protection not functional
- Main transformer MODs not functional
Form ES-401-6 ES-401 Written Examination Question Worksheet Question # 38 Examination Outline Cross-reference: Tier # 1 Group # 2 K/A # 061 .GEN.2.4.48 Importance Rating 3.8 Proposed Question: Tank area) With the plant at 100% power, area radiation monitor RMS16-1 (VCT and Boric Acid recently following evolutions have goes into alarm. The RO reviews the rough log and notes the been conducted:
- A Liquid Waste Discharge was commenced.
- The Degassifier was shutdown.
- The Boron Evaporator was started up.
- A Solid Waste System resin transfer was commenced.
Which of the above activities was the likely cause of the alarm? A. The Liquid Waste discharge. B. The shutdown of the Degassifier. C. The startup of the Boron Evaporator. D. The resin transfer. Proposed Answer: B Explanation (Optional): VCT. With the degassifier shutdown, a RMS16-1 The degassifier degasses the letdown stream prior to entry into the the VCT ("B" correct). "A", "C", and "D" are plausible, alarm can be anticipated as radioactive gasses accumulate in the auxiliary building. since they involve movement of radioactive material through AOP 3573, Att. B, pg 3 (Attach if not previously provided) Technical Reference(s): None Proposed references to be provided to applicants during examination: and limitations (As available) Learning MC-05469 Describe the major administrative or procedural precautions the basis for each. Objective: placed on the operation of the Radiation Monitoring System, and Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.11 55.43.4 and 43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
l 3 Page 3 of 5 i ALARM I RADIATION MONITOR
RESPONSE
AOP 3573 Rev. 012 Attachment B I Area Radiation Monitors Monitor Area Monitored Automatic/Subsequent Actions R MS1 6-1 VCT and boric acid tanks Note: If degasifier shutdown with the unit at power (FloorplanAUX43) then ALERT and ALARM can be anticipated.
- 1. Check trend history and alarm status of monitor CHS69 R MS 17-1 Entrance to test tank pumps (FloorplanWSTE04)
RMS18-1 Outside sample sink (Floorplan WSTE04) RMS19-1 Solid waste storage (FloorplanWSTE24) RMS20-1 Sample room (FloorplanAUX43) R MS21 -1 Radioactive chemistry laboratory (Service Bldg 24') RMS22-1 Control Room Check trend history and alarm status of monitors HVC16A, HVC16B and HVC91. RMS24-1 Liquid radwaste area (FloorplanWSTE24) RMS25-1 Muck bay (Floorplan WSTE24) RMS28-1 Fuel Building pipe rack (FloorplanFUELI1) An asterisk (*) indicates the monitor is safety related or Class 1E.
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 39 Tier # 1 Group # 2 K/A # W/E 16.EA2.2 Importance Rating 3.3 Proposed Question: The reactor has tripped, and the crew is progressing through the EOP network. The crew enters EOP 35 FR-Z.3 "Response To High Containment Radiation Level". Per FR-Z.3 guidance, which two actions are specified for ADTS consideration for use in lowering CTMT radiation levels? A. Use of CTMT Air Filtration System and CTMT Spray Pumps. B. Use of CTMT Vacuum System and CTMT Purge System. C. Use of CTMT Air Filtration System and CTMT Vacuum System. D. Use of CTMT Purge System and CTMT Spray Pumps. Proposed Answer: A Explanation (Optional): FR-Z.3 consists of one step plus a transition step. The one step samples CTMT atmosphere, considers the use of CTMT Air Filtration, and considers use of CTMT Spray System ("A" correct). CTMT Purge System is plausible, since it would remove activity from CTMT, and CTMT Vacuum System is plausible, since it is used as a backup to the Hydrogen Recombiner System to remove excess Hydrogen from CTMT during an accident. Technical Reference(s): FR-Z.3, step I (Attach if not previously provided) Proposed references to be provided to applicants during examination: None Learning Objective: Describe the major action categories within EOP 35 FR-Z.3 (As available) Question Source: Bank # 74362 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
RESPONSE TO HIGH EOP 35 FR-Z.3 Page 3of3 CONTAINMENT RADIATION LEVEL Rev. 005 STEP ACTION/EX(PECTED RESRESPONSE NOT OBTAINED I NOTE The following items should be considered when determining if a containment air filtration fan should be placed in service:
- Containment radiation levels
- Containment atmosphere I-131 levels
- 1. Request ADTS Support For The Following Evolutions:
- a. Obtaining containment atmosphere l-131 sample I
- b. Obtaining concurrence to start one containment air filtration fan using OP 3313D, "Containment Air Filtration"
- c. Obtaining additional recommendations for the further course of action (including the operation of the containment spraypumps)
- 2. Go To Procedure And Step In Effect
-FINAL-
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 40 Tier # I Group# 2 K/A # 065.AA2.05 Importance Rating 4.1 Proposed Question: With the plant at 100% power, a leak in the instrument air system occurs, and the following sequence of events occurs: 1400 The RO reports that instrument air pressure is decreasing at a moderate rate. 1401 The crew enters AOP 3562 "Loss of Instrument Air" 1412 Letdown isolates 1414 PZR spray valves close 1415 Feed Reg Valves close 1417 Reactor Plant Chilled Water CTMT header isolates At what time did AOP 3562 require the crew to shutdown the reactor via manual reactor trip? A. 1412 B. 1414 C. 1415 D. 1417 Proposed Answer: C Explanation (Optional): The crew is directed to trip the reactor and go to E-0 when instrument air pressure is they are decreasing rapidly or when feedwater control is lost ("C" correct). "A", "B", and "D" are plausible since actions that will occur on a loss of air that have adverse effects on the plant. Technical Reference(s): AOP 3562, step I (Attach if not previously provided) Proposed references to be provided to applicants during examination: None MC-03941 Discuss conditions which require transition to other procedures. (As available) Learning Objective: Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
D STE ACTION/EX(PECTED RESPONSREPNEOTBAIED NOTE The actions specified in this procedure may be performed concurrently with E-0, Reactor Trip or Safety Injection.
- 1. Verify Plant Status
- a. Check instrument air pressure a. Proceed to step 2. and, rapidly decreasing OR loss of feedwater control IF instrument air pressure decreases rapidly OER feedwater control is lost, THEN TRIP the reactor and Go to E-0, Reactor Trip or Safety Injection
- b. TRIP the reactor and Go to E-0, Reactor Trip or Safety Injectior
- 2. Check Instrument Air System Alignment
- a. Verify both instrument air a. Locally Place both instrument compressors - RUNNING air compressor control switches to CS (continuous service).
- b. Check instrument b. Perform the following:
air pressure - STABLE OR INCREASING 1) Using Attachment A, locally Start air compressors and Perform filter and dryer checks.
- 2) Proceed to step 2.d.
*c. Proceed to step 12.
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 41 Tier # 1 Group # 3 KIA # 028.AAI .08 Importance Rating 3.6 Proposed Question: With reactor power at 100% and the pressurizer level control selector switch in the CHAN I-II position, 3RCS-LT460 fails low. The following indications and annunciators are received:
- PRESSURIZER LEVEL DEVIATION
- PZR LVL LO HTR OFF AND LTDOWN SECURE
- The PZR Level recorder on MB4 drops to 0%.
The operators respond promptly per AOP 3571 "Instrument Failure Response", and take manual control of 3CHS*FCV121. Both functional PZR level indicators increase to 71%. What will occur when the operators select "CHAN I-III" on the pressurizer level control switch (3RCS-LS459D)? A. The PZR LVL LO HTR OFF AND LTDOWN SECURE annunciator will clear, and the PZR Level recorder on MB4 will return to proper indication. B. The PZR LVL LO HTR OFF AND LTDOWN SECURE annunciator will clear, and all backup heaters in "AUTO" will automatically energize. C. The PRESSURIZER LEVEL DEVIATION annunciator will clear, and the PZR Level recorder on MB4 will return to proper indication. D. The PRESSURIZER LEVEL DEVIATION annunciator will clear, and all backup heaters in "AUTO" will automatically energize. Proposed Answer: B Explanation (Optional): The failed backup channel initially brought in the level deviation annunciator, and isolated letdown, since both the controlling and backup channels provide letdown isolate and heater trip protection. Actual level increased due to the letdown isolation. When actual level increased to 5% above program level of 61.5%, an actual level deviation occurred ("C" and "D" wrong), calling for heaters to energize. When the operators select CHAN I-III, the low level input to the PZR LVL LO HTR OFF AND LTDOWN SECURE annunciator will be removed, clearing the annunciator. Since actual level is high, the heaters will energize ("B" correct). The PZR level recorder will not return to proper operation until its separate recorder select switch is taken to an unaffected channel ("A" and "C" wrong). Technical Reference(s): Functional Dwg # 11 (Attach if not previously provided) Process Dwg #I I OP 3353.MB4A, 4-1 and 5-1 Prnnne,-d references to he provided to anplicants during examination: None
. . ---. -___ __ __ r- - . ___ - -- -, -- - - -
41 of 46 NUREG-1021, Revision 8, Supplement 1
Form ES-401-6 ES-401 Written Examination Question Worksheet and Level Control (As available) Learning MC-0534 1 Describe the operation of the Pressurizer Pressure Objective: System under Normal, Abnormal, and Emergency Operating conditions. Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
PRESSURIZER LEVEL CHANNELS AUCTIONCERED TAVG. (SMEET 9) I i I II I I i ADVUSTABLE I I Cu4ARclmq
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I/_4~ AUTO- MANUAL (Is CONTROL (HOT* P) STATLON (EOCAL) MODULATE M\ODULATE CHARGING PrRAY VALV£ SPRAY VALVE FLOW I -Z CONTROL PC.V-45 5C (BY ennl4w-q) I s "FSAR FIGURE" 12179-2472.01 1-001-Ol IQ.A T--_ I I 7ITI I I I H I I
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(NOTE I ADJUSTABLE ("O~ts) ll oi-- - PRIL%%IRli t CPO51Teou 1 NOIIMALLY( ~r' . 1 - 74FEENCI. SETPOI NT l SCLECTED) WITH IN ((NoTE GI a___ - CONTROLLER ( qEF.) I(NOTEM I i a.S I SPRAY
-- 2i lCONT RO Lt II II II I I I I I I I I I I I AuTo- MANUAL CONTROL I
STATION I (coNrRoL T ROOM) I I. ALL CIICUITS CH THIS SHXT ARENUTIEOJLANT. Z. LOCAL OCL.. OMIIDS ALL OR SIW4ALS. TO TO LOCALOVMI E ACTUATSALARMIN CONTRORC. MODULATE
- 3. PERM BISTAkE NO. l-4S55; AND LEVEL B1SrABLES N. - TURN-iON VARIABL .E 5PRAY VALVE' LBSX9., Lb-45W.0L1.-460 0 ARE ENEAIZE T0 ACUATE ALL HEATt
- 4. WM INDICATIONIN 1ADM11CL tl AwD ow Tww AUXILIAQ'Y SHTDowN PAWeL(A6PJ BZACI.-UP- CONTRO >L- PCV-455B
- 5. LIOWF %A4vtO M ep pn
,4 c04wo.AqcL eoom Vol %t." HEATERS SIGIAL L (-o ,AS.)
SW '-WM 'M SWI146A-. -MIA" VT I'M r ALLY I (SHE£T IE) (SNEEr Iz
<.7T4WSE. COWTROLS ONTHa Coh-tL SaaO ARoD SLUPPL. qLv g-W5S Br T
- 7. LOCAL CoNTROL AT THE AUXILIARYSHUTDOWON. PA439LOWRRIDEb THIL FROM Ts"E O..rJokL ROM AONDACTrYTES AIJW ALARM ANw-JC iNORq I W rMI SIGjA.Lb COIJTROL Roo,.t
- 8. A REMOTE/LOCAL. CONTROL TRANSFER SwrITCm it LOCATED ON A LOCAL TRANFER SWrrTCt PANEL (TSP).
hi I1 I: I WAL CAD t VISIMS10 THIS00AAIATi I I S. MENAS-euILTAKOIIDIIED . _ ___ I - I I I. 10 I 9 a I I 7
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-' REACTORTRIP
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I - - - - - - - - - - - - - - - - - - _- - L7s q~4-210, 4- 1 . lT I I I R ft I* R LYI459)w LY 459Y j I I CQMR L-/459 10 LY/ 459 I ACOMP SH.Z7l CC)AMr- 9H _I
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l "T CAL O l CONTROL C.F-4 PWL. CONTROL B t . 111I i9 109 II I to I 9 S 7
Setpoint: 5% deviation from program or I l 4-1 I1 greater than 70% level PRESSURIZER LEVEL DEVIATION AUTOMATIC FUNCTIONS
- 1. Pressurizer backup heaters in "AUTO" energize on high level deviation.
CORRECTIVE ACTIONS
- 1. IF in MODE 3, 4, 5, or 6 AND alarm is expected due to planned shutdown evolutions, no further action is required.
I
- 2. CONFIRM pressurizer level deviation from program on the following indicators (MB4):
2.1 3RCS*LI 459A, "PZR LEVEL' 2.2 3RCS*LI 460A, "PZR LEVEL' 2.3 3RCS*LI 461, "PZR LEVEL' 2.4 3RCS-LI 462, "PZR LEVEL' "COLD CAL'
- 3. IF alarm is due to instrument failure, Go To AOP 3571, "Instrument Failure Response."
- 4. CHECK 3CHS-FI 121A, "CHG LINE FLOW" "CHARGING FLOW" (normally 55 to 100 gpm) (MB3).
- 5. CHECK 3CHS-FI 132, "LETDOWN" "FLOW" (normally 75 to 120 gpm) (MB3).
- 6. IF charging or letdown flow is not in normal range, Go To OP 3304A, "Charging and Letdown," and adjust charging and letdown flows.
- 7. IF in MODE 1, or 2, Refer To T/S 3.4.3.1, "Pressurizer, Startup and Power Operation" for Limiting Condition for Operation. 10 SUPPORTING INFORMATION
- 1. Initiating Device Setpoint 1.1 3RCS*LT459, Pzr level 1.1 5% deviation from program or greater than 70% level 1.2 3RCS*LT460, Pzr level 1.2 5% deviation from program 0 or greater than 70% level 1.3 3RCS*LT461, Pzr level 1.3 5% deviation from program or greater than 70% level Level of Use i 4- -- -I OP 3353.MB4A Rev. 002-05 Continuous IT- 71rr IT 31 of 64 I
LI4-1J
- 2. Computer Points 2.1 RCS-L459E (HI) 2.2 RCS-L460C (HI) 2.3 RCS-L459F(LO)
- 3. Possible Causes 3.1 Charging versus letdown mismatch 3.2 Turbine runback 3.3 Reactor power versus turbine load mismatch 3.4 Instrument failure 3.5 Planned shutdown evolution
- 4. Technical Specifications 4.1 T/S 3.4.3.1, "Pressurizer, Startup and Power Operation" I0
- 5. Procedures 5.1 AOP 3571, "Instrument Failure Response" 5.2 OP 3304A, "Charging and Letdown"
- 6. Control Room Drawings 6.1 ESK 10HG 6.2 LSK 25-1.2C e of Ue AOP 3353.MEB4A Level of Use Rev. 002-05 Continuous 32 of 64 t
Setpoint: 22% 5-1 PZR LVL LO HTR OFF AND LTDOWN SECURE AUTOMATIC FUNCTIONS
- 1. Pressurizer heaters deenergize.
- 2. Letdown isolation valves close (MB3):
- 3RCS*LCV459, "LID ISOL'
- 3RCS*LCV460, "L/D ISOL'
- 3. Letdown orifice isolation valves close (MB3):
- 3CHS*AV8149A, "L/D ORIFICE ISOL"
- 3CHS*AV8149B, "L/D ORIFICE ISOL"
- 3CHS*AV8149C, "LAD ORIFICE ISOL' CORRECTIVE ACTIONS
- 1. CONFIRM pressurizer level low on the following indicators (MB4):
- 3RCS*LI 459A, "PZR LVL'
- 3RCS*LI 460A, "PZR LVL'
- 3RCS*LI 461, "PZR LVE'
- 3RCS-LI 462, "PZR LVy' "COLD CAL'
- 2. IF alarm is due to instrument failure, Go To AOP 3571, "Instrument Failure Response."
- 3. IF pressurizer level decrease occurred due to charging versus letdown mismatch, Refer To OP 3304A, "Charging and Letdown," and ADJUST charging and letdown as necessary.
- 4. IF pressurizer level decrease occurred with no cooldown in progress, Go To AOP 3555, "Reactor Coolant Leak."
- 5. IF pressurizer level decrease is due to cooldown, PERFORM the following:
5.1 STOP cooldown. 5.2 RESTORE pressurizer level. Level f UseOP 3353.MB4A Continuous I Rev. 44 002-05 of 64
5-1
- 6. WHEN pressurizer level is greater than 22%, RESET pressurizer heaters and PLACE in 'AUTO" (MB4).
- 7. WHEN RCS leak is isolated AMD pressurizer level is restored, Refer To OP 3304A, "Charging and Letdown," and RESTORE normal charging and letdown.
- 8. IF in MODE 1, or 2, Refer To T/S 3.4.3.1, "Pressurizer, Startup and Power I0 Operation" for Limiting Condition for Operation.
- 9. IF cooldown was stopped, RESUME cooldown.
- 10. NOTIFY Engineering Department of letdown isolation and restoration.
SUPPORTING INFORMATION
- 1. Initiating Device Setpoint 1.1 3RCS*LT459, Pzr level 1.1 22%
1.2 3RCS*LT460, Pzr level 1.2 22% 1.3 3RCS*LT461, Pzr level 1.3 22%
- 2. Computer Points 2.1 3RCS-L459C
- 3. Possible Causes 3.1 RCS leak 3.2 Charging versus letdown mismatch 3.3 High RCS cooldown rate 3.4 Instrument failure
- 4. Technical Specifications 4.1 T/S 3.4.3.1, "Pressurizer, Startup and Power Operation" IT 4.2 T/S 3.4.6.2, "Operational Leakage" 4.3 T/S 3.3.1, "Reactor Trip System Instrumentation" Level of Use -q A
W, I I OP 3353.MB4A Rev. 002-05 Continuous 45 of 64 F r I
- p. 1
5-1
- 5. Procedures 5.1 AOP 3571, "Instrument Failure Response" 5.2 E-O, "Reactor Trip or Safety Injection" 5.3 AOP 3555, "Reactor Coolant Leak" 5.4 OP 3304A, "Charging and Letdown"
- 6. Control Room Drawings 6.1 ESK 10HG 6.2 ESK 7JA 6.3 LSK 25-1.2C I
OP 3353.MB4A Level of Use WI Rev. 002-05 Continuous 46 of 64
~om b- A1-F orm h6-4u I-0 PqS-401 Written Examination Question Worksheet Examination Outline Cross-reference: Question # 42 Tier# 1 Group # 3 K/A # WIE13.EKI.3 Importance Rating 3.2 Proposed Question:
The reactor has tripped and the crew performing actions per ES-0. 1 "Reactor Trip Response" when the following sequence of events occurs: to a
- 1. The crew enters EOP 35 FR-H.2 "Response To Steam Generator Overpressure" due yellow path on the "heat sink" status tree.
- 2. The crew is preparing to dump steam from the affected "A" SG.
- 3. The BOP reports "A" SG narrow range level is 89% and increasing slowly.
Should the crew dump steam from the "A" SG, and why? likely A. No, since releasing steam may cause an uncontrolled radiation release, as the SG is ruptured. B. No, since releasing steam may result in two phase flow and water hammer, potentially damaging pipes and valves C. Yes, since releasing steam is necessary to lower SG pressure. D. Yes, since releasing steam is necessary to lower SG level. Proposed Answer: B level is at the upper SG Explanation (Optional): "B" is correct, and "C" and "D" are wrong, since 87% narrow range resulting in water level tap, and may be indicative of a full SG, where releasing steam may also release water, tube rupture is suspected, the crew will be in E-3. "A" is plausible, since a SGTR hammer. "A" is wrong since if a respond to an overpressure will raise SG level. "C" and "D" are plausible, since releasing steam is the normal way to event in FR-H.2, and removing inventory will reduce both pressure and mass. Technical Reference(s): FR-H.2 WOG Bkgd doc, Caution prior to step 4. (Attach if not previously provided) Proposed references to be provided to applicants during examination: None MC-05976 Discuss the basis of major procedure steps and/or sequence of (As available) Learning Objective: steps in EOP FR-H.2 Question Source: Bank # 74634 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.8, 41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement I
STEP DESCRIPTION TABLE FOR FR-H.2 Step 4 - CAUTION CAUTION: If affected SG narrow range level increases to greater than (M.08)% I [(M.09)% for adverse containment], steam should not be released from I the affected SG(s). PURPOSE: To warn the operator not to release steam from the affected SG if level increases to greater than (M.08)% [(M.09)% for adverse containment] to prevent potential damage to piping, valves, or turbines BASIS: If SG level increases to greater than (M.08)% [(M.09)% for adverse containment], steam release from the SG may result in two-phase flow and/or water hammer, which may potentially damage piping, valves, or turbines. ACTIONS: Determine if the affected SG narrow range level increases to greater than (M.08)% [(M.09)% for adverse containment] INSTRUMENTATION: SG narrow range level CONTROL/EOUIPMENT: N/A KNOWLEDGE: N/A PLANT-SPECIFIC INFORMATION: o (M.08) SG level at the upper tap, including allowances for normal channel I accuracy. o (M.09) SG level at the upper tap, including allowances for normal channel I accuracy, post-accident transmitter errors, and reference leg process errors. FR-H.2 12 HP-Rev. 1C HFRH2
Written Examination Question Worksheet Form ES-401-6 ES-401 Examination Outline Cross-reference: Question # 43 Tier # 1 Group # 3 K/A # W/E15.EK1.1 Importance Rating 3.0 Proposed Question: and The crew has entered EOP 35 FR-Z.2 "Response To Ctmt Flooding" and is trying to identify isolate sources of water to the CTMT sump per FR-Z.2 step 1. What water sources will the crew check per FR-Z.2 step 1, and why? A. Sources such as RPCCW and Auxiliary Feed Water, since water level is above that expected from emergency stored water sources, and may damage critical plant components. B. Sources such as RPCCW and Auxiliary Feed Water, since water level is above that expected from emergency stored water sources, and may lower TSP concentration below acceptable limits. level C. Emergency stored water sources such as the RWST and ECCS accumulators, since water has reached the point where it may damage critical plant components. D. Emergency stored water sources such as the RWST and ECCS accumulators, since water level has reached the point where it may lower TSP concentration below acceptable limits. Proposed Answer: A Explanation (Optional): "A" is correct, and "C" and "D" wrong since level above the design basis flood level in level is the potential CTMT is indicative of an unexpected water source. "B" is wrong since the concern with high damage of critical plant components. Technical Reference(s): FR-Z.2 WOG Bkgd Doc, Step I (Attach if not previously provided) Proposed references to be provided to applicants during examination: None Learning MC-05993 DISCUSS the basis of major procedure steps and/or (As available) Objective: sequence of steps in EOP 35 FR-Z.2. Question Source: Bank # 71936 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.8 and 41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
STEP DESCRIPTION TABLE FOR FR-Z.2 STEP 1 STEP: Try To Identify Unexpected Source Of Water To Sump PURPOSE: To identify unexpected source of water in sump BASIS: This step instructs the operator to try to identify the unexpected source of the water in the containment sump. Containment flooding is a concern since critical plant components necessary for plant recovery may be damaged and rendered inoperable. A water level greater than the design basis flood level provides an indication that water volumes other than those represented by the emergency stored water sources (e.g., RWST, accumulators, etc.) have been introduced into the containment sump. Typical sources which penetrate containment are service water, component cooling water, primary makeup water and demineralized water. All possible plant specific sources which penetrate containment should be included in this step. These systems provide large water flow rates to components inside the containment and a major leak or break in one of these lines could introduce large quantities of water into the sump. Identification and isolation of any broken or leaking water line inside containment is essential to maintaining the water level below the design basis flood level. ACTIONS: Try to identify unexpected sources of water to the sump INSTRUMENTATION: Plant specific instrumentation to identify unexpected sources of water to the sump CONTROL/EQUIPMENT: N/A KNOWLEDGE: N/A PLANT-SPECIFIC INFORMATION: Sources of water which supply components inside containment FR-Z. 2 7 HP-Rev. 1 0022V: 1
Written Examination Question Worksheet Form ES-401-6 ES-401 Question # 44 Examination Outline Cross-reference: Tier # 2 Group # I K/A # 001.K5.97 Importance Rating 3.6 Proposed Question: PLANT CONDITIONS:
- Rods are in "AUTOMATIC"
- RCS Tave is 587 0F
- Tref is 583 0F How fast should the control rods be moving?
A. 32 spin B. 36 spm C. 40 spm D. 48spm Proposed Answer: C Explanation (Optional): 0 Rods move at a minimum of 8 and a maximum of 72 spin. An error of 2 F will cause the minimum (8 spm) motion. 0 or 40 spm ("C" correct). All Speed ramps up linearly from 3 to 57F error. At 4 F, rod speed will be half way, speeds. distractors are plausible since they are within the band of minimum and maximum Functional sheet 9 (Attach if not previously provided) Technical Reference(s): None Proposed references to be provided to applicants during examination: (As Learning MC-05475 Describe the function and location of the following Rod Control System available) Objective: components... Reactor Control Unit.. Temperature Error Summer ...Speed Controller... Question Source: Bank # 68652 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
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- 5. ALARM I AND Aas No? stamum~0kT. S. ALARIA I AN D ALARM z MUST HAVE
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- 2. KC-r MW'1 VA.U'V %NVERS&-L-IN TWO 0%03CRETra. S-T-ps Wtrrb BREA.K PcItws G. *THESE CONTROLS ON THtE CONTROL BOARD ARE S1I'PLED LkbkE Op K..&,< VA&W TUaB1g LarAD. BY OT14ERS, AT 30W 0so . AAk4D GO To say%
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ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 45 Tier # 2 Group # 1 K/A # 003.K1.01 Importance Rating 2.8 Proposed Question: The plant has been operating at 100% power steady state for several weeks, when the Control Room crew determines that the "A" Reactor Coolant Pump upper oil reservoir level has just started to slowly and steadily increase. What is the cause and effect of the increasing level? A. RPCCW is leaking into the reservoir. The pump will lose lubrication. B. Thermal expansion of the oil due to reaching equilibrium thermal conditions. Pump lubrication will be maintained. C. Oil reservoir automatic level control valve is malfunctioning. Pump lubrication will be maintained. D. Wear of the upper radial bearing is reducing oil flow, backing up oil into the reservoir. The pump will lose lubrication. Proposed Answer: A Explanation (Optional):
"B" is "A" is correct. OP3353.MB4B 4-2A has operators check RPCCW surge tank for indications of in-leakage.
equilibrium. "C" is wrong, there is no LCV for oil reservoir. "D" is wrong, wrong, oil would not expand if it is at radial bearing wear will cause oil flow to increase due to increased mechanical clearances. Technical Reference(s): OP3353.MB4B 4-2A (Attach if not previously provided) Proposed references to be provided to applicants during examination: None MC-05288 Describe the function and location of the ... Upper and Lower (As available) Learning Objective: Oil Reservoirs Question Source: Bank # 70675 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.3 and 41.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
Setpoint: Greater than 1.25 inches above sight glass centerline l-l RCP A UPR OIL RSVR LVL Hi AUTOMATIC FUNCTIONS
- 1. None CORRECTIVE ACTIONS
- 1. CHECK RCS-L475A, RCP A upper oil reservoir level computer point, to confirm alarm.
- 2. MONITOR the following RCP A computer points:
- RCS-T479A, RCP A upper thrust bearing temperature
- RCS-T479B, RCP A lower thrust bearing temperature
- RCS-T483A, RCP A upper radial bearing temperature
- 3. IF at any time any RCP A thrust bearing or radial bearing temperature computer point is greater than 195°F, PERFORM the following:
3.1 IF reactor power is greater than P-8 (37%), PERFORM the following: 3.1.1 TRIP reactor. 3.1.2 STOP RCP A. 3.1.3 Go To E-0, "Reactor Trip or Safety Injection." 3.2 Refer To AOP 3554, "RCP Trip or Stopping an RCP at Power," and REMOVE RCP A from service.
- 4. CHECK 3CCP-LI 20A, RPCCW surge tank level, for indication of RPCCW leakage.
hA ALARA RA The following step requires containment entry.
-J
- 5. IF directed by SM/US, CHECK for RPCCW to RCP A lube oil leak.
- 6. Refer To Technical Specifications, and DETERMINE Limiting Condition for Operation.
LC oftinUose lOP 3353.MB4B Continuous ThJ~JRev. 004-02 66oWI12
lI4-2AHl SUPPORTING INFORMATION
- 1. Initiating Device Setpoint 1.1 3RCS-LS475A 1.1 >1.25 inches above sight glass centerline
- 2. Computer Points 2.1 RCS-MA75A 2.2 RCS-T479A 2.3 RCS-T479B 2.4 RCS-T483A
- 3. Technical Specifications 3.1 T/S 3.4.1.1, "Reactor Coolant System, Reactor Coolant Loops and Coolant Circulation, STARTUP and POWER OPERATION" 3.2 T/S 3.4.1.2, "Reactor Coolant System, Reactor Coolant Loops and Coolant Circulation, HOT STANDBY" 3.3 T/S 3.4.1.3, "Reactor Coolant System, Reactor Coolant Loops and Coolant Circulation, HOT SHUTDOWN"
- 4. Procedures 4.1 E-0, "Reactor Trip or Safety Injection" 4.2 AOP 3554, "RCP Trip or Stopping an RCP at Power"
- 5. Control Room Drawings 5.1 ESK 10HH 5.2 LSK 25-1.1D C ofnuouse L l OP 3353.MB4B ContiuousRev. 004-02
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 46 Tier # 2 1 Group # K/A # 004 .K2.06 Importance Rating 2.7 Proposed Question: PLANT CONDITIONS:
- 100% power
- All systems are in AUTOMATIC
- LT-459 is selected for Pressurizer level control
- PT-456 is selected for Pressurizer pressure control A loss of VIAC-I occurs.
What impact will the loss of VIAC 1 have on PZR pressure and level control? A. The PZR master pressure controller will have to be taken to MANUAL only. B. The charging line flow control valve/PZR level controller will have to be taken to MANUAL only. C. The PZR pressure controller will have to be taken to MANUAL, and the "A" PZR PORV will no longer function in AUTOMATIC. D. The charging line flow control valve/PZR level controller will have to be taken to MANUAL and letdown will have to be restored. Proposed Answer: D Explanation (Optional): signal ("B" "D" is correct, since LT459 is powered from VIAC 1. It also feeds the low PZR level letdown isolation 2 ("A" wrong). PORVs wrong). PZR pressure is not affected because its controlling channel is powered by VIAC to open ("C" wrong). are still available, since they require two channels of pressure to indicate high in order Technical Reference(s): Functional Dwg # 6, 11, and 18 (Attach if not previously provided) Proposed references to be provided to applicants during examination: None MC-05342 Given a failure, partial or complete, of the Pressurizer (As available) Learning Objective: Pressure and Level Control System, determine the effects on the system and on interrelated systems. Modified Bank # 1997 Millstone NRC SRO # 34 Parent attached Question Source: Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
ES-401 Written Examination Question Worksheet Form ES-401-6 Original Millstone 3 1997 NRC question SRO-34: PLANT CONDITIONS:
- 100% power
- All systems in AUTOMATIC
- LT-459 selected for control of Pressurizer level control selected to LT-459
- PT-456 selected for control of Pressurizer Pressure
- Instrument for "A" and "C" steam generators selected to Channel I
- Instruments for "B" and "D" steam generators selected to Channel II A loss of VIAC-2 occurs.
Which of the following lists controllers which should be taken to MANUAL as a result of the VIAC-2 failure? A. Rod control Pressurizer Pressure Pressurizer Level B. Rod Control Pressurizer Pressure Master main feed pump controller C. Pressurizer Pressure Pressurizer Level Feed Regulating Valves for "A" & "C" SGs D. Pressurizer Pressure Master main feed pump controller Feed Regulating Valves for "B" & "D" SGs ANSWER: D 41 of 46 NUREG-1021, Revision 8, Supplement 1
Lot PPESSL.11IZER L.O~W PPESSuL~a PPESSU21IZEM P2ESSu~- t NOTES: SAF RY TNJECTrlO 1. THEREDINDNTNAWALBLOCKCCNTR1L CONSISTS OF Mro CONTROLS CNTHECONTROL. ROAID.CONEFOREACHTWAIN. ( S" E-- X ) 2. TWOC1fUTER IWNUTS ARECONNECTE TO THIS CIRCUIT. INDIVIOUALFOREACHTRAIN.
- 3. TWOPERMISSIVES STATUSLIGHTSARE CDRACCTIED TO THIS CIRCUIT. INOIVIDUALFOREAO4TRAIN.
- 4. SVPPLiED BY OTHIRS.
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- 3. PPWSMK IIISTABLE ND. P11-455r. AND LEVEL flISrA8LES NO. flJRF-OlV VARIABLE
- SPRAY VALVE L"-59C, LlB-4SSEL&-4W D ARE ENERIZE TO ACUATE
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- 7. LOC.AL CC*JTROL AT THELAUXKILIARY SAIUT~oow,41 PAUEL FRQATM 0 aT64o"L PRmcv£&.0crwi6WAu OOVeRRIDE5 THE 6I*IQ.L'b ALARMA Amwmc.A~om Ib.J FMG COIJTROL ROOAA.
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- 1. PROTEC7TIOM GkR?.DE WtOE RkNGE RCS TEPERNTURS SLG.PM%-
6rkBA\ TPRJllIA "" WEIKE PV, TE~C.LOI SErS. I Z. A.NfO4N`ATLIO IN TIAE WBPKL?CON4TROL R~OOM'LS RLEQUIPEb TO7 BE VISZLIE 'T0 T`IAS OPE-0TOP, XT T*AE t'AAI4 I COWM1OL RJOOPA.
- 3. ROTECT'LON GRKbm wILbe Pk46E P.Cs PkESSURF S'LGB4k;.
vmot& A. 7v.0xLB -9 WELPCIE. PROTECTL.ON SEVT.
- 4. STATUJS 1G.4TS WJISr T>E PROVLEM TZOR. ENMPOkV AND EA.C% PORN~ BLOCK( 'JA.VE. AZT T¶.4 WTAINB& COMTP.DL I SOPIRD TO 1,'Lc.kTE WIEN -r'.4. VoA.LVEIS U~,LL2YCLOSED OR.. PULLY Oppv4.
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ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 47 Tier # 2 Group # 1 K/A # 004.K6.07 Importance Rating 2.8 Proposed Question: The plant is at 100% power when the RO notices letdown flow oscillating due to flashing downstream of the letdown orifices. What could have caused the flashing to occur? A. The Regenerative Heat Exchanger has developed a tube leak. B. 3CHS*PCV131, Letdown Pressure Control Valve, has failed closed. C. RPCCW flow to the Letdown Heat Exchanger has increased. D. 3CHS*FCV121, Charging Line Flow Control Valve, has failed closed. Proposed Answer: D Explanation (Optional): Flashing will occur downstream of the orifices if the letdown stream is not adequately cooled in the regenerative heat exchanger, or if pressure drops excessively. "A" is wrong since a tube leak results in colder charging water leaking into the letdown line. "B" is wrong since this raises pressure downstream of the orifices. "C" is wrong since this cools the letdown stream. "D" is correct since this removes cooling to the regenerative heat exchanger, reducing cooling to the letdown steam. Technical Reference(s): P&ID 104A (Attach if not previously provided) Proposed references to be provided to applicants during examination: None Learning MC-04201 Describe the major administrative or procedural precautions and (As available) Objective: limitations placed on the operation of the Chemical and Volume Control Question Source: Bank # 68582 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
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Form ES-401-6 ES-401 Written Examination Question Worksheet Question # 48 (SRO) Examination Outline Cross-reference: Tier # 2 1 Group # K/A # 013.GEN.2.2.22 Importance Rating 4.1 Proposed Question: Initial conditions:
- The plant is in MODE 0.
- Fuel movement is in progress in the fuel pool.
- Rigging is in progress in the Control Building 64' level.
their load and damaged "Control The control room receives a report that the riggers lost control of RO reports that Room Makeup Air Supply" radiation monitor 3HVC*RE16B. The 3HVC*RE16B is NOT functioning. What ACTION, if any, is required? A. No ACTION is required, since the plant is in MODE 0. B. No ACTION is required, since 3HVC*RE16A is available. C. Restore 3HVC*RE16B to OPERABLE within 7 days. D. Immediately suspend fuel movement. Proposed Answer: C Explanation (Optional): fuel handling analysis. "C" is correct, and This is a new change to Technical Specifications based on a re-evaluated of operable channels 1 less than the total in "A" and "B" are wrong, since the ACTION applies with the number since the requirement previously was less than the MODES 1-6, or with fuel movement in progress. "B" is plausible, the applicable modes were "All", and Technical minimum operable channels. "A" is plausible, since previously 2, 3, 4, 5, and 6 and during fuel movement within Specifications do not define MODE 0. Now, it reads MODES 1, is the action required for loss of both channels. containment or the spent fuel pool". "D" is plausible since this Tech Spec 3.3.2 ESFAS Inst. ACTION 18 (Attach if not previously provided) Technical Reference(s): Tech Spec Table 3.3-3 Tech Specs Sections 3/4 Proposed references to be provided to applicants during examination: provided reference (As available) Learning MC-04765 Given a plant condition or equipment malfunction, use Specification applicability and required actions. Objective: materials to evaluate Technical Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.2 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
TABLE-3 .1 (Continued) ( ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION o =
-J rn-MINIMUM 0 APPLICABLE z TOTAL NO. CHANNELS CHANNELS OF CHANNELS TO TRIP OPERABLE MODES ACTION CD FUNCTIONAL UNIT -I 8-I 6. Auxiliary Feedwater (Continued)
(Aj
- f. Containment Depres- See Item 2. above for all CDA functions and requirements.
surization Actuation (CDA) Start Motor-Driven Pumps
- 7. Control Building Isolation 1 2
- 19 1
- a. Manual Actuation 2 1 2 1, 2, 3, 4 19 W b. Manual Safety 2 Injection Actuation NA 2 1, 2, 3, 4 14 1 W
C. Automatic Actuation 2 1 Logic and Actuation Relays 3 2 2 1, 2, 3 16
- d. Containment Pressure--
High-1 1 2/intake 18 1
- e. Control Building Inlet 2/intake .CD Ventilation Radiation C
- 8. Loss of Power g '
3/bus 1, 2, 3, 4 20 -%a 02 a. 4 kV Bus Under- 4/bus 2/bus 0 voltage-Loss of Voltage ro V-4 b. 4 kV Bus Undervoltage- 0 2/bus 3/bus 1, 2, 3, 4 20 C) Grid Degraded Voltage 4/bus 0 (A
February 20, 2002 TABLE 3.3-3 (Continued) TABLE NOTATIONS
# The Steamline Isolation Logic and Safety Injection Logic for this trip function may be blocked in this MODE below the P-11 (Pressurizer Pressure Interlock) Setpoint.
- MODES 1, 2, 3, 4, 5, and 6. *1 During fuel movement within containment or the spent fuel pool.
Trip function automatically blocked above P-1i and may be blocked below P-11 when Safety Injection on low steam line pressure is not blocked. t During core alterations or movement of irradiated fuel within the containment. The provisions of Specification 3.0.3 are not applicable. ACTION STATEMENTS ACTION 14 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 6 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours; however, one channel may be bypassed for up to 4 hours for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE. ACTION 15 - (not used). ACTION 16 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour. ACTION 17 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 4 hours for surveillance testing per Specification 4.3.2.1. ACTION 18 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 7 days. After 7 days, or if no channels are OPERABLE, immediately suspend CORE ALTERATIONS and fuel movement, if applicable, and be in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. ACTION 19 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. MILLSTONE - UNIT 3 3/4 3-24 Amendment No. 97, 70, ii, X79, 0725 203
Written Examination Question Worksheet Form ES-401-6 ES-401 Question # 49 Examination Outline Cross-reference: Tier # 2 Group# I K/A # 014.K1.01 Importance Rating 3.6 Proposed Question: misaligned higher than While operating at 90% power, a control bank "D" Group 1 rod becomes the rest of its group. and is currently The crew has entered AOP 3552, "Malfunction of the Rod Drive System" aligning the affected rod to the rest of the bank. to insert the affected rod Prior to moving the affected bank, the procedure requires the operators until the next lower DRPI LED changes state. Which of the following describes the reason for inserting the control rod? when the affected rod is A. To reset the logic cabinet master cycler to ensure proper rod stepping aligned to the group. affected bank. B. To accurately determine the actual rod position prior to realigning the compliance with C. To accurately determine how far the affected rod is misaligned to verify Technical Specifications. of actual position of the D. To ensure when the affected rod is realigned, it will be within 3 steps bank. Proposed Answer: B 1/2 the distance between the coils. Explanation (Optional): The coils are placed 6 steps apart with the LED positioned band is 2 steps above and 3 steps below the As the rod steps through the coil, the LED above the coil will light. The affected rod can then be inserted to the indicated position. With the actual position o the bank rod now located, the same height as the rest of the rods in the group. ("B" correct) AOP 3552 Basis Doc, Att. A, step 6 (Attach if not previously provided) Technical Reference(s): None Proposed references to be provided to applicants during examination: the (As available) Learning MC-05484 Describe the operation of the Rod Position Indication System under Emergency conditions... Stuck, Misaligned, or Objective: following Normal, Abnormal, and Dropped Rod (including recovery operations)... Question Source: Bank # 72414 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.2, 41.6 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
NOTE If the affected rod is in a Control Bank, a ROD CONTROL BANKS LIMIT LO (MB4C 3-9) alarm and ROD CONTROL BANKS LIMIT LO LO (MB4C 4-9) alarm may occur during recovery and remain in alarm until the P/A converter is reset. Therefore, response to these alarms is not appropriate during this period. If the misaligned rod is in group I of a control bank and the rod is being inserted to provide alignment, these alarms will come in if the alignment process moves the affected rod in group 1 as seen by the P/A converter below the present insertion limit. The annunciators will remain in alarm until the pulse-to-analog converter is reset thereby, providing the proper bank position to the RIL computer. If the insertion limit was exceeded prior to this point, it is expected that the operator would have responded to these alarms in the normal fashion. It is acceptable to proceed with these alarms present for the following reasons: I) The SHUTDOWN MARGIN has previously been verified satisfactory. 2) The rods will remain in "bank select" or "manual" during this entire period, which precludes any automatic movement in the inward direction. 3) The affected bank is not moved during the procedure, only the misaligned rod. 4) The power level will remain constant or be decreased by direction of the procedure, but never increased during the alignment limit process. The above listed conditions make it highly unlikely that a rod insertion would be approached during the period when these alarms are present. STEP 5: Establish Conditions For Rod Alignment step After l&C completes the check on the affected rod lift coil fuses, the affected group counter position is recorded because this will be the stopping point during alignment. All lift coils in the affected bank (except the misaligned rod) are open circuited so only the affected rod will move on demand. The rod control selector switch is placed in the affected bank position so only the affected rod moves and not rods in another bank due to bank overlap. STEP 6: Align Rod The initial substeps determine the position of the misaligned rod so the affected group step counter can be reset to the located position. The operator must first determine whether the affected rod is higher or lower than its associated bank by using DRPI. If the affected rod is higher than its bank, the rod is inserted until the next DRPI indicator lights. Since the range of a DRPI LED is 6 steps (two steps above and three steps below the indicated position), this action places the affected rod two step higher than the lit LED's indicated position. With the position of the misaligned rod now located, the affected group step counter can be reset to that position and the rod then inserted to the same height as the rest of the rods in its group as recorded in the initial substep. In a like manner, if the affected rod is misaligned lower than its bank, the affected rod's position is determined by withdrawing the rod until the next higher DRPI LED lights. This indicates the affected rod is 3 steps less than the lit LED's indicated position. The rod may then be withdrawn to the recorded height after the group step counter is reset to the located position. 9-JO)A k. A( ~+ep Lf
Reactor Engineering has determined the rod speed during alignment should be limited to 3 steps/hour (as in fuel preconditioning) if the rod has been misaligned for greater than 16 hours, otherwise, rod speed during alignment is unrestricted. If the affected rod does not move, its trippability must be determined after resetting the group step counter in the RNO. A misaligned rod is not inoperable until it is shown to be immovable or misaligned for greater than 1 hour. Once the rod is shown to be immovable, Technical Specification ACTION requirement 3.1.3.1 a., or b .2. or b .3. is in effect, each requiring the rod to be declared inoperable. Further clarification of the proper ACTION requirement will be made after trippablility is determined in Attachment D. STEP 7: Restore Rod Control System After rod alignment, the rod control system is returned to the normal condition by reconnecting the lift coils, and clearing the urgent failure alarm. The bank selector switch is returned to manual to ensure proper rod sequencing. If the misaligned rod was in group I of a Control Bank, the operator is directed to Attachment E to reset the pulse-to-analog converter bank demand position to correct indication. If the misaligned rod was in a Control Bank or in Shutdown Bank A or B, I&C is directed to check the master cycler count (0 through 6) at the proper count for the correct "next group" to step in the affected bank. This will ensure that the correct group will initially step upon initiation of rod movement. The LO -LO insertion limit alarm is checked clear following the P/A converter restoration since the alarm may have been brought in by the alignment process. STEP 8: Perform Follow-up Actions After rod alignment, the OPERABIUTY of the affected rod is verified by performing the SURVEILLANCE REQUIREMENTS for Technical Specificafion 3.1.3.1 and 3.1.3.5 as applicable. After notifying Reactor Engineering of the realignment, the operator is directed to procedure step 6 to verify the entire Rod Control System is functioning properly. Form ES-40 1-6 ES-40 1 Written Examination Question Worksheet Question # 50 Examination Outline Cross-reference: Tier # 2 Group # I K/A # 015.K1.08 Importance Rating 2.9 Proposed Question: a reactor During the Three Mile Island event, with core uncovery in progress, the crew started coolant pump. 3 rapidly from 5x104 cps to 5x10 Immediately after starting the RCP, source range counts dropped cps. Why did source range counts drop? A. Molten core material relocated to the bottom of the core. B. Borated water added negative reactivity to the core. C. A slug of cold water filled the vessel downcomer. D. Increased voiding added negative reactivity to the core. Proposed Answer: C Explanation (Optional): shielding excore NIS from neutrons.
"C" is correct, since crossover leg water refilled the vessel downcomer, due to less neutron attenuation. The crew had Previously, source range counts had increased as the core voided, the SR detectors are near the bottom of the core, concerns that a restart accident was in progress. "A" is wrong, since since the core was already adequately shutdown.
and counts increased as the core relocated. "B" and "D" are wrong, Westinghouse MITCORE Text, pg 9.12.K (Attach if not previously provided) Technical Reference(s): Westinghouse MITCORE Text, Figure 9. 10 None Proposed references to be provided to applicants during examination: in the core (As available) Learning MC-04954 Describe the effects on Reactor kinetics of coolant voiding to voiding. Objective: region, and relate the excore nuclear instrumentation system response Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.2,3,5,6,7, and 9 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement I
by H. A sharp reduction in the rate of increase is believed to be caused the reduction of feedwater addition to the A OTSG (approximately at 124 minutes). This reduces condensation in the A loop, leaving the open relief valve as the only pathway for boil-off and the removal of core coolant. The increase in the detector count rate slows, which is believed to correspond to a reduction in the rate of the core uncovery. I. The signal level continues to increase slightly as core uncovery proceeds at a slower rate, approaching a near-equilibrium level. It is believed that the maximum count rate coincides with loop A refill to the reactor vessel inlet level. At 142 minutes after turbine trip, the operator shuts the electromatic relief block valve. It is observed that the maximum count rate does not coincide with shuttting the block valve. Increased make-up to the core (about 36 gpm) pro-duces gradual recovery. J. Over this period the count rate decreases as level in the core rises. The increased core mixture level is facilitated by relief block valve closure. K. The operator starts reactor coolant pump 2B, sending a slug of cold water into the downcomer and essentially filling it. L. Loop flow data indicates that the pump worked effectively for a very brief period. This is corroborated by the abrupt turnaround in the source range detector trace, as flow ceases and excess downcomer fluid moves into the core and is boiled off; equilibrium levels are reestablished. M. High pressure injection flow is initiated at 200 minutes, 8 minutes after the electromatic relief valve is opened by the operator. Coolant passes into the downcomer, filling it. Detector count rates drop sharply. N. Continued addition of high pressure injection flow begins to quench the core. It is conjectured that the coolant first rewets the outer region of the core, bypassing the hot center. 9.12 1693S:4 Rev 0 8/84
18346-3 TMI SOURCE RANGE DETECTOR RECORD Counts/Second (Log Decades) 10 6 1 , I I 1o 4 10 3 L-o- 60 120 180 240 Turbine Time After Turbine Trip (Minutes) Trip Figure 10
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 51 Tier # 2 Group # 1I_ K/A # 015.K6.05 Importance Rating 2.6 Proposed Question: Refueling is in progress with Westinghouse NIS providing audible SR counts indication to both the Control Room and Containment with the following NIS switch positions selected:
- The "Channel Selector Switch" on the "Audio Count Rate" drawer is in the "N-3 1" position.
- The "W/GM1/GM2" selector switch on "A" Train Shutdown Margin Monitor Drawer is in the "W" position.
- The "Amplifier Selector" switch on the rear of the "Audio Count Rate" drawer is in the "Normal" position.
- The "Audio Level Control" knobs on both SMM Drawers are at "Minimum".
- The "Audio Multiplier Switch" on the "Audio Count Rate" drawer is in the "10" position.
The "Audio Multiplier Switch" on the "Audio Count Rate" NIS drawer is inadvertently taken from the " 10" position to the "Off' position. How will this error affect audio countrate indication to both CTMT and the control room? A. Audible countrates will be maintained to both locations. B. Audible countrate will be lost to Containment only. C. Audible countrate will be lost to the Control Room only. D. Audible countrates will be lost to both the Control Room and CTMT. Proposed Answer: D Explanation (Optional): CTMT counts are lost, since the "W/GM1/GM2" selector switch receives the Westinghouse signal downstream of the "Audio Multiplier Switch" for the CTMT speaker ( "A" and "C" are wrong). Control Room counts are lost, since the "Audio Multiplier Switch" supplies the control room speaker as well ("D" correct and "B" is wrong). Technical Reference(s): Figure NIS015T-010 (Attach if not previously provided) Proposed references to be provided to applicants during examination: None Learning MC-0522 1,Describe the function... of the following... Audio Count Rate/ (As available) Objective: Timer-Scaler... Containment Speakers...Gamma-metrics Drawer... Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
( ( ( WESTINGHOUSE GAMMA- METRICS OFF I l
]I I
Al I 0 A2 1 l (FIGURE) NMS015T-010 AUDIO COUNTRATE CIRCUIT 94000397 REVISION 2
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 52 Tier # 2 Group# I K/A # 017.A4.02 Importance Rating 4.1 Proposed Question: The plant is experiencing an inadequate core cooling condition, and the crew has entered EOP 35 FR-C. 1 "Response to Inadequate Core Cooling". The crew is currently checking if RCPs should be started to provide forced cooling to the core. What is the minimum temperature at which the crew will be directed by procedure to start RCPs? A. Core Exit Thermocouples > 7180 F. B. Core Exit Thermocouples > 1200'F. C. RCS Hot Leg temperature > 718'F. D. RCS Hot Leg temperature > 1200'F. Proposed Answer: B Explanation (Optional): "B" is correct, since CETCs > 1200'F is the criterion ("B" correct, "A" wrong). Hot leg temperatures should not be used, since they may react significantly slower than CETCs ("C" and "D" wrong). Technical Reference(s): FR-C. 1, step 17 (Attach if not previously provided) Proposed references to be provided to applicants during examination: None Learning Objective: MC-04524 Describe the major action categories within EOP 35 FR-C. I (As available) Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.7, 41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
RESPONSE TO INADEQUATE EOP 35 FR-C.1 Page 16 of 18 CORE COOLING Rev. 013 Jl P ACTION/EXPECTED RRESPONSE NOT OBTAINED NOTE Normal conditions are desired but NOT required for starting the RCPs.
- 17. Check If RCPs Should Be Started
- a. Check core exit TCs - a. Proceed to step 18.
GREATER THAN 12000 F
- b. Check if an idle RCS cooling b. Perform the following:
loop is available
- 1) OPEN all PZR PORVs and
- Check idle loop SG NR block valves.
level THAN 8%
- GREATER I
- 2) IF core exit TCs remain (42% ADVERSE CTMT) GREATER THAN 12000 F, THEN
- Check RCP in associated idle loop - AVAILABLE a) OPEN reactor head vent isolation valves:
I 3RCS*SV8095A 3RCS*SV8095B 3RCS*SV8096A 3RCS*SV8096B b) Open the reactor head vent to PRT isolation valves: 3RCS*HCV442A 3RCS*HCV442B
- 3) Proceed to step 18.
- c. START RCP in one idle RCS cooling loop
- d. Return to step 17.a.
Form ES-401-6 ES-401 Written Examination Question Worksheet Question # 53 Examination Outline Cross-reference: Tier # 2 Group # 1 K/A # 022.GEN.2.1.32 3.8. Importance Rating Proposed Question: Water (CDS) pumps With the plant at 100% power, one of the two running Reactor Plant Chilled well. trips, resulting in one of the running chillers to trip as heats up, which system limit will be the As the CDS cooling water for CTMT ventilation systems first concern for the crew? limit. A. Reaching a CTMT temperature Technical Specification limit. B. Reaching a CTMT pressure Technical Specification Coolant Pumps. C. Reaching a high temperature limit on the Reactor Rod Drive Mechanisms. D. Reaching a high temperature limit on the Control Proposed Answer: B Explanation (Optional): CTMT, so as CDS heats up, CTMT temperature increases, "A" is wrong, since CDS supplies the CAR coolers in the Tech much less margin than temperature before reaching which also raises CTMT pressure. CTMT pressure has immediate since RCPs and CRDMs will not experience Spec limit (B correct, "A" wrong). "C" and "D" are wrong an event exhausting from these heat loads. This is similar to problems since the associated coolers cool the hot air at Millstone 3 on 8/2/99. P&ID 122A & 122B (Attach if not previously provided) Technical Reference(s): Millstone 3 CR M3-99-2843 None Proposed references to be provided to applicants during examination: plant chilled (As available) Learning MC-04189 Given a failure, partial or complete, of the reactor and on interrelated systems. Objective: water system, determine effects on the system Bank # 73616 Question Source: Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.10 55.43.2 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
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07'14;99 O7/2/9 9 Approval Date Effective Date AR No. CR Form 6 1 1 54 St
&9 Initiation CR M3-99-2843 Section 1: To be completed by initiator (please type or print)
O~r-anizaationa identifvin- condition: I Discover) date: 812/99 lAffected Ulnit(s): I Systenli f.' MP30OPS lDiscovcrv time: 2252 IO] 20] 3MC I Condition description (including how condition was discovcred. or-anization creating condition, what activitv was il progress when event was discovered): spray painted. It was iThe B CDS Chiller tripped. It was found that when the chiller tripped it was actively being substantially covered by plastic and taped up. There was plastic over the control panel. Continuation Sheet E Component ID. 3CDS-CHL1B Source Document: Method of Discovery: Event (RP 4, Att I)
- 2. Immediate corrective action taken area found that painting was Both containment vacuum pumps were immediately started. An operator dispatched to the The chiller had tripped on high motor temperature. It was decided to taking place (this was not known to the SM and US).
the A train of CCP. While swapping the C chiller tripped, apparently due to swap the from the B chiller to the A chiller on chiller restarted automatically. The A Chiller was started and tripped on low oil the flow transient on CCP. The C on the CDS system both Service Water booster pumps to the MCC and Rod Control area AC unit pressure. To reduce loads on restart, 3CCP-were started and CDS to it isolated. To prevent a flow transient on CCP from tripping the A CDS chiller TV188A was bypassed to allow manual control. After 30 minutes the A CDS chiller twas restarted. in containment Containment pressure increased from 13.75 to 13.89 psia, the0 Tech Spec limit is 14.0 psia. Temperature increased from 111.40F to 113.4WF, the Tech Spec limit is 120 F. trouble in the past and The System Engineer was contacted and came in. The high motor temperature module has caused of PPC data showed that the motor bearing temperatures for the B chiller were may be the cause of the trip. A review normal prior to the trip indicating that the motor was most likely not overheating. It was meggered The lube oil pump for the B chiller was spray painted while running and the windings were found painted. covering the control panels which and found acceptable. All three chillers had been prepared for painting which included impeded operator action on the panels. TR# AWO// Continuation Sheet E
- 3. Recommended corrective action Determine if the B chiller tripped due to a problem with the temperature module or motor.
Place better controls on painting in the plant on operating equipment. Determine the cause for the low oil pressure trip on the A chiller. need to he addressed. Problems with the chillers tripping on flow transients have been identified in other CRs and do not Modifications are scheduled to address these. the Alarm Develop a procedure section to swap the MCC and Rod Control Area ACUs from CDS to SWP. Add this to Response procedure. Continuation Sheet Q
- 4. Initiator Requests Follow-up: Z Y [] N Initiator Name: Barrett Nichols Time: 0340 Phone No.: 0582 Initiator's Signature: ,5 C Date: 813/99 Cost Control Center 881 nutmber.
If continuationsheets (RP 4-1, Page 7) are required, identify the section being continued by section Form RP4-1 Rev. 7 Chg 3 Page I of 7 Sheet I
ES-40 1 Written Examination Question Worksheet Form ES-40 1-6 Examination Outline Cross-reference: Question # 54 Tier # 2 Group # 1 K/A # 026.K3.02 Importance Rating 4.3 Proposed Question: The plant has tripped due to a LOCA, and the following sequence of events occurs: 0800: The crew enters E-0 0812: CDA actuates. 0813: The RO reports that neither QSS pump can be started. 0816: The crew enters FR-Z.1 "Response to High CTMT Pressure". 0818: The STA reports CTMT wide range sump level is 1 foot. How is the operation of the RSS system affected by the QSS system failure? A. All RSS pumps will all be placed in Pull-To-Lock, since inadequate sump level exists. RSS Pump "C" or "D" will be aligned to take a suction on the RWST to spray CTMT, since "A" and "B" RSS pumps have a recirc path and are preferred for supplying cold leg recirc. B. All RSS pumps will all be placed in Pull-To-Lock, since inadequate sump level exists. RSS Pump "A" or "B" will be aligned to take a suction on the RWST to spray CTMT, since "A" and "B" RSS pumps have a recirc path, and are preferred for supplying spray. C. All RSS pumps will be manually started after 11 minutes have elapsed, since the CDA signal did not fully actuate. RSS Pump "C" or "D" will be aligned to take a suction on the RWST to spray CTMT, since "A" and "B" RSS pumps have a recirc path and are preferred for supplying cold leg recirc. D. All RSS pumps will be manually started after 11 minutes have elapsed, since the CDA signal did not fully actuate. RSS Pump "A" or "B" will be aligned to take a suction on the RWST to spray CTMT, since "A" and "B" RSS pumps have a recirc path and are preferred for supplying spray. Proposed Answer: A Explanation (Optional): Without QSS, the steam in CTMT will not condense as rapidly into the CTMT sump, and RWST water will not be added as rapidly to the CTMT sump, so RSS pumps are placed in PTL, since inadequate sump level exists ("A" correct, "C" and "D" wrong). "C" and "D" are preferred for spray, since "A" and "B" are preferred for recirc ("A" correct, "B" wrong). Technical Reference(s): P&ID 112C (Attach if not previously provided) EOP35 FR-Z.1, steps 7 and 8 D--r -nrbf- to he nrnvidpd toapplicants, dnrinp examination: None 41 of 46 NUREG-1021, Revision 8, Supplement 1
ES-401 Written Examination Question Worksheet Form ES-401-6 Leaming MC-04668 Discuss the basis of major procedure steps and/or sequence (As available) Objective: of s temp in EOP 35 FR-Z. 1. Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7, 41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
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RESPONSE TO HIGH EOP 35 ER-Z.1 Page 5 of 10 CONTAINMENT PRESSURE Rev. 011
- r. I STEP ACTION/EXPECTED RESPONS E RESPONSE NOT OBTAINED
- 7. Verify Recirculation Spray System Operation
- a. Check annunciator CTMT a. Proceed to step 9. and, RECIRC PUMP AUTO WHEN START SIGNAL The annunciator actuates (MB2B 1-8) - LIT QM 11 minutes elapse since CDA initiation, THEN Return to step 7.b.
- b. Check recirculation spray b. Proceed to step 7.d.
pumps - RUNNING
- c. Proceed to step 7.f.
- d. Check Ctmt WR sump level - d. Proceed to step 9. and, GREATER THAN 1.5 feet WHEN Ctmt WR sump level is GREATER THAN 1.5 feet THEN Return to step 7.e.
RESPONSE TO MGH EOP 35 FR-Z.1 Page 6 of 10 CONTAINMENT PRESSURE Rev. 011
- I S/EXPECTED RESPONS RESPONSE NOT OBTAINED l
- 7. (continued)
- e. START recirculation spray e. Perform the applicable action:
pumps
- E starting RSS pump A OR C desired, THEN
- 1) At bus 34C, using GE75 key, Place the T OCAL/RMOTE switch in LOCAL for desired RSS pump.
For RSS pump A breaker 34C19-2 For RSS pump C breaker 34C20-2
- 2) Place the local breaker control switch in START for the desired RSS pump.
IF starting RSS pump B OR D desired, THEN
- 1) At bus 34D, Place the LOCAL/REMOTE switch in LOCAL for desired RSS pump.
For RSS pump B breaker 34D18-2 For RSS pump D breaker 34D19-2
- 2) Place the local breaker control switch in START for the desired RSS pump.
IF no RSS pump is running, THEN Proceed to step 9.
RESPONSE TO HGH EOP 35 FR-Z.1 Page 7 of 10 CONTAINMENT PRESSURE Rev. 011 ACTION/EXPECTED RESPONS P RESPONSENOTOBTAINED l
- 7. (continued)
- f. Check recirculation spray f. OPEN valves.
pump suction isolation valves - OPEN
- g. Verify recirculation spray g. OPEN valves.
pump discharge isolation valves - OPEN
- h. Check recirculation spray - h. Proceed to NOTE prior to FLOW INDICATED IN AT step 8.
LEAST ONE TRAIN
- i. Proceed to step 9.
NOTE
- The Locked Valve Key is required for performance of some of the local actions in the following step.
- The preferred priority for selecting a recirculation spray pump is as follows:
- 1. Pump C or D
- 2. PumpAorB
- 8. Check If Spray Using An RSS Pump From The RWST Should Be Established
- a. Check the following: a. Proceed to step 9.
- RWST level - GREATER THAN 100,000 gal
- Quench Spray pumps -
NONE RUNNING
- b. Place all recirculation spray pumps in PULL TO LOCK
RESPONSE TO HIGH EOP 35 FR-Z.1 Page 8 of 10 CONTAINMENT PRESSURE Rev. 011
-, , I STP ACTION/EXPECTED RESPONS RESPONSE NOT OBTAINED l
- 8. (continued)
- c. RESET SI
- d. Determine the RSS pump to d. IE no recirculation spray be placed in service pumps are available, THEN Proceed to step 9.
- e. For the selected RSS pump, START the associated RECIRC SPRAY ACU (3HVQ*ACU2A or 3HVQ*ACU2B)
- f. OPEN the RWST recirculation suction valves (MB1)
- 3QSS*AOV27
- 3QSS*AOV28
- g. Using Attachment B, locally align the selected RSS pump
- h. Check the selected h. CLOSE valve.
recirculation spray pump RHR isolation - CLOSED
- For pump A-3kS3-*MV8837A
- For pump B -
3RSS*MV8837B
- For pump C-3RSS*MV8838A
- For pump D-3RSS*MV8838B
Written Examination Question Worksheet Form ES-40 1-6 ES-40 1 Question # 55 Examination Outline Cross-reference: Tier # 2 Group# I K/A # 026.A3.02 Importance Rating 4.2 Proposed Question: (RSS) A CDA occurred and the crew is progressing through the BOP network. The CTMT Recirc in the desired lineup. pumps have just started, and the RO is verifying that equipment is water inlet isolation valves What should be the position of the RPCCW heat exchanger service recirc coolers (3SWP*MOV50A/B), and the service water inlet valves to the containment (3SWP*MOV54A/B/C/D)? and the service water A. Both the RPCCW heat exchanger service water inlet isolation valves inlet valves to the containment recirc coolers should be OPEN. should be OPEN and the B. The RPCCW heat exchanger service water inlet isolation valves be CLOSED. service water inlet valves to the containment recirc coolers should should be CLOSED and the C. The RPCCW heat exchanger service water inlet isolation valves OPEN. service water inlet valves to the containment recirc coolers should be valves and the service water D. Both the RPCCW heat exchanger service water inlet isolation inlet valves to the containment recirc coolers should be CLOSED. Proposed Answer: C Explanation (Optional): "C" is correct, since on a CDA, the RPCCW heat exchanger service water inlet isolation water pumps. The service water inlet valves valves receive a CLOSE signal, to prevent excessive flow for the service cooling to the CTMT sump water being to the containment recirc coolers receive an OPEN signal to provide RSS heat exchangers. "B" and "D" are plausible recirculated. "A" is plausible, since this also supplies cooling to the but since the RSS pumps are running, at least since there is a 3 minute time delay prior to 3SWP-MOV54C opening, 11 minutes has passes since the CDA actuated. LSK I G and 27-1 IL (Attach if not previously provided) Technical Reference(s): P&ID 133B None Proposed references to be provided to applicants during examination: the following (As Learning MC-057 18 Describe the operation of the Service Water System under Actuation available) Objective: normal, abnormal, and emergency conditions... Ctmt Depressurization Question Source: Bank # 69683 Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
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- 1. CONTROL FORREACTOR PLANTCOMPONENT COOLINGHEATEXCHANGER SUPPLY MANAAIL RAAOSIONS TOTHISOoCUItNT I VNENAS-IAJLTAREPROHIBITED. _I _
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- 4. BOTHLIGHTSAREONWHENVALVEIS IN AN INTERMEDIATE - - -
- 6. SSWP-FT43A ANDASSOCIATED INSTRUMENTATION SHOWN, I I SSWP-FT43BANDASSOCIATED INSTRUMENTATION SIMILAR. COOLING WATER SPRINGRETURN I ICOTTOCROA~-A-I57OREACTOR I2j~ PLANTCOMPONENT-
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WI12 I42 CONTAIINMENT RECIRCULATIONCOOLER SERVICE tATER SUPPLYVALVE (NOTE 7) NOTES:
- 1. FLOWPATHANDESFATRAIN DES11NATION VALVE TRAIN ASSOCIATEDCOOLER VALVEFUNICTIO 3SI4P*MVSrA A SRS5XEIA CONTAINMENT RECIRCULATION 3SWPMIOV54C A SRSSEIC COOLERSUPPLY YALYES
- 2. REFERTO LSK-0-S8 (28020 S.2): NOTE 6.1
- 3. REFERTO LSX-O-Stt28020 SH. 2): NOTE6.6
- 4. BOTHLIGHTSARE ONWIHEN VALVEIS IN AN INTERiEDIATE POSITION.
- 5. SAFEGUARD SIGNALIS tLOCKEDFROMACTUATION DURINGA NO-NOEQUIPMENT TEST.
(FOR DETAILSSEELSK-O-3AQ28020 SH.11 :NOTE 1.8 ) SWPE
- 6. LOCATED ON162 IN THERECIRCULATION CHANGEOVER ARRAY. (SRWP.fOVSIA
- 7. CONTROL FORCONTAINMENT RECIRCULATION COOLER SUPPLYVALVESSP30NVOVS4A SHOWN.
CONTROL FOR VALVES 3SWP*MOV56CI SIMILAREXCEPTTHAT3SWP*#OV5SIC STATUSLIGHT IS ORANGEANODOESPOHAVE CONTROL OR LIGHTSFROMRECIRCULATION CHANGEOVERARRAY ON W42. CONTROL FOR3SWtP*VSIIl AND54D SHOWN ONLSX-27-IfE (28457 SH.S)
- 8. TRANSFER SWITCHIS A KEYLOCK SWITCH WITH KEYREMOVAL IN BOTHPOSITIONS.
- 9. TIME DELAY APPLICABLE TO 3SWP*MOV54C ONLY.
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ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 56 Tier # 2 Group # I K/A # 056.A22.05 Importance Rating 2.5 Proposed Question: While operating at 100% power, the unit experiences a catastrophic failure of the 6B Feedwater Heater extraction steam bellows. The following alarms and indications are promptly received:
- CONDENSATE CONDUCT HI - MB6A 3-8
- HOTWELL SODIUM HI - MB6A 4-8
- COND DEMIN SYSTEM TROUBLE - MB6A 1-6
- COMMON INFLUENT CONDUCTIVITY HIGH - CD 5-1
- COMMON EFFLUENT CONDUCTIVITY HIGH - CD 5-2
- Unit Electrical output has decreased by 5 MWe Which of the following is the most probable result of this event, and what is the correct mitigation strategy?
A. The bellows failure has created debris and ruptured the main condenser structure. The crew will use AOP 3559 "Loss of Main Condenser Vacuum" to shutdown the plant when backpressure is greater than 5 " Hg Abs. B. The bellows failure has created debris and caused a chemistry excursion. The crew will use OP 3319C "Condensate Demineralizer Mixed Bed System" to swap clogged demineralizers. C. The bellows failure has created debris and caused condenser tube leakage. The crew will use AOP 3557 "Secondary Chemistry" to shutdown the plant if recommended by Chemistry. D. The bellows failure has created debris and disrupted the extraction steam flow. The crew will use AOP 3567 "Operation with One Heater String Isolated" to isolate the B Low Pressure Heater string and return the unit to full power. Proposed Answer: C Explanation (Optional): Event is based on industry operating experience. Each distracter contains a contains a plausible result and remedy, but "A" is wrong, since significant conductivity problems exist. "B" and "D" are wrong, due to conductivity alarms and Mw loss. "C" is correct, since AOP 3557 contains entry conditions (alarms) and shutdown recommendation. Technical AOP 3557 Secondary Chemistry, Entry Conditions, and steps 1-5 (Attach if not Reference(s): Operational Experience, O&MR 434, Extraction Steam Bellows Failure previously provided) Proposed references to be provided to applicants during examination: None 41 of 46 NUREG-1021, Revision 8, Supplement 1
ES-401 Written Examination Question Worksheet Form ES-401-6 Learning Objective: MC-05899 Identify conditions that require entry into AOP 3557. (As available) Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.5 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
A. PURPOSE This procedure provides the actions necessary to respond to abnormal chemistry conditions in the Feedwater System, Condensate System, Main Steam System, steam generators, or condenser hotwell including a suspected main condenser tube leak. B. ENTRY CONDITIONS Any of the following are indications that secondary chemistry is not within the proper specifications.
- 1. CONDENSATE CONDUCT HI MB6A 3-8
- 2. HOTWELL SODIUM HI MB6A 4-8
- 3. TURB PLANT SAMPLE PNL TROUBLE MB6B 1-4
- 4. RX PLANT SAMPLE SYS TROUBLE MB1B 2-4
- 5. COND DEMIN SYSTEM TROUBLE annunciator alarm (MB6A 1-6)
- COMMON INFLUENT CONDUCTIVITY HIGH annunciator alarm (CD 5-1)
- COMMON EFFLUENT CONDUCTIVITY HIGH annunciator alarm (CD 5-2)
- DEMIN EFFLUENT SODIUM HIGH annunciator alarm (CD 1-7)
- DEMIN EFFLUENT CHLORIDE HIGH annunciator alarm (CD 2-7)
- 6. Increasing condensate conductivity 3SST-CR134 (MB6)
- 7. Increasing hotwell sodium 3SST-AR138 (MB6)
- 8. Reports from the Chemistry Department that any secondary chemistry sample is out of specification.
ACTIN/EPECED ESPNSE RESPONSE NOT OBTAIE
- 1. Check Local Sample Panel Alarms
- a. Verify the following: a. Dispatch an operator to locally acknowledge and investigate
- Annunciator TURB alarm.
PLANT SAMPLE PNL TRDUBLE (MB6B 1-4)
- NOT LIT
- Annunciator RX PLANT SAMPLE SYS TROUBLE (MB1B 2-4)
- NOT LIT
- 2. Verify Main Condenser libes Are Intact
- a. Check any of the following a. Proceed to step 6.
conditions exist:
- Condensate conductivity indicator (3SST-CR134) on MB6 - INCREASING O)R
- Hotwell sodium concentration indicator (3SST-AR138) on MB6
- INCREASING OR
- Annunciator CONDENSATE CONDUCT HI (MB6A 3-8)
- LIT OR
- Annunciator HOTWELL SODIUM HI (MB6A 4-8) - LIT
STEP ACTION/EXPECTED RESPONS RESPONSE NOT OBTAINED l
- 3. Isolate Condenser Drawoff
- a. Locally Close the condenser level control valve outlet isolation (3CNS-V12)
- 4. Notify Chemistry
- a. Check condenser a. Proceed to step 4.c.
backpressure - INCREASED FOR TURBINE WARMING
- b. Inform Chemistry of increased condenser backpressure
- c. Request Chemistry sample the hotwells and condensate pump discharge (CPD) for impurities
- d. Request Chemistry recommend plant actions using CP 3802B, "Secondary Chemistry Control"
ACTION/EXPECTED RESPONSRA Q A U TI 0 N DO NOT throttle open 3CNM-MOV78, demin bed bypass valve, if demineralizer differential pressure exceeds 60 psi. Instead, reduce power as necessary to maintain the proper condensate flow through the polishing demineralizers. NOTE Chemistry recommendations on plant shutdown may be amended as secondary chemistry conditions change.
- 5. Perform Corrective Actions
- a. Check Chemistry recommends a. Proceed to step 5.c.
- A PLANT SHUTDOvVN
\ _,,.
- b. Using OP 3204, "At Power Operation," Shutdown the plant within at least the next 12 hours
- c. Check Chemistry c. Perform the following:
recommends - ISOLATING A 1) Perform the actions CIRCULATING WATER recommended by BOX Chemistry.
- 2) Proceed to step 8.
SP ACTION/EXPECTEDRO R ESPONSE NOT OBTAINED I
- d. Check both circulating water d. Perform the following:
boxes on selected condenser
- IN SERVICE 1) Using OP 3204, "At Power Operation," Commence an orderly plant shutdown
- 2) WHEN The main generator breaker is open, THEN Proceed to step 5.e.
- e. Using OP 3250.25A, "Draining and Restoring Circulating Water Boxes,"
Isolate and Drain the selected Circulating Water System water box
- f. Request Chemistry sample hotwells for impurities
- g. Proceed to step 8.
Forced Outage Due to an Extraction Steam Expansion Bellows Failure Page 1 of 2 9k Event
Title:
Forced Outage Due to an Extraction Steam Expansion Bellows Failure Event On March 8, 1995, while Unit 2 was at 100 percent power, a condenser tube leak was Summary: identified in the B condenser. While investigating, operators noted that the number 6B heater level control valve (LCV) was found to be wide open with the bypass valve cycling. A review of past data showed that the number 6B extraction pressure had dropped about 20 percent and a drop of 20 MWe at the same time of the condenser tube leak. The leak caused the pressure to drop in the 6B heater which caused in level control problems. A decision was made to continue operating until a plant shutdown plan could be developed. Operations personnel noted the position of the 6B LCV on a periodic basis. On April 3, 1995, 6B LCV had fully opened and a loud roar could be heard in the B condenser. The plant was shutdown to inspect the condenser. The 6B extraction line bellows had blown out and the 7B bellows was damaged. Additional damage was limited to cracks in the B condenser as well as all other condensers. The cause of the bellows failure was from fatigue cracking resulting from flow-induced vibration. Further review identified that pipe support and configuration was satisfactory and that bellows failure was a possibility, though time to failure could not be anticipated. Industry operating experience identified that several other plants had experience similar problems, all resulting from fatigue. A replacement schedule for extraction bellows was initiated. This event is not significant as this problem was isolated to the condenser and there was no major equipment damage. However, this event is noteworthy due to secondary plant damage that required a plant shutdown. Event Number: 328-950308-1 Event Date: 03/08/1995 INPO Change Date: 07/13/1995 Unit: 328, SEQUOYAH 2 NSSS Vendor: WESTINGHOUSE ELECTRIC NSSS Type: PWR - Pressurized Water Reactor Country: USA INPO Significance: Noteworthy Event Initial Plant Condition: Steady State Power, 100% Power Event Descriptor: Major Equipment Damage Causal Factor: Other Identified Component/System Failures: Component System EFailure Consequence Equipment Causal Factor PIPE XAE 1-Was Caused by the Event-Aging/Deterioration Keywords: BELLOWS, EXTRACTION STEAM, FATIGUE CRACKING, RUPTURE Primary Source OPERATING EXPERIENCE ENTRY (NUCLEAR NETWORK), Document: OE7271, 05/12/1995 Related Document: SIGNIFICANT EVENT NOTIFICATION, 0099, 05/27/1993 http://www.inpo.org/DataBases/Events/ 9 5 /3 28_950308_1.htm 03/20/2002
Forced Outage Due to an Extraction Steam Expansion Bellows Failure Page 2 of 2 LIMITED DISTRIBUTION use. Copyright 2000 by the Institute of Nuclear Power Operations. Not for sale nor for commercial is expressly prohibited. Reproduction of this report without the prior written consent of INPO may Unauthorized reproduction is a violation of applicable law. Each INPO member and participant or reproduce this document for its business use. This document should not be otherwise transferred delivered to any third party, and its contents should not be made public, without the prior agreement of INPO. All other rights reserved. NOTICE Power This information was prepared in connection with work sponsored by the Institute of Nuclear on Operations (INPO). Neither INPO, INPO members, INPO participants, nor any person acting to the behalf of them: (a) makes any warranty or representation, expressed or implied, with respect the use accuracy, completeness, or usefulness of the information contained in this document, or that on of any information, apparatus, method, or process disclosed in this document may not infringe privately owned rights; or (b) assumes any liabilities with respect to the use of, or for damages resulting from the use of any information, apparatus, method, or process disclosed in this document. http://www.inpo.org/DataBases/Events/ 9 5 /3 28_950308_1.htm 03/20/2002
C01802C Operational Experience Rev. 0 CONTENT ACTIVITIES I. INTRODUCTION A. Review objectives B. Distribute OE handout I I. OW&MR 4 34 - ET O&MR434-
/TICT~:K R AC I 0 iOS ,,E b'
d-
~MC-06286.2~AF s
, _ d MC068 idgU dXAMC-06729 A. Review event history
- 1. The industry has a number of extraction steam bellows failure reports dating back to 1987. The failures were attributed to various causes including fabrication errors, design weaknesses, fatigue, and end of life (EOL). The number of failures reported in 2000 (four) represented a significant increase from the past, and the commonalties associated with the failures suggest that improved preventive maintenance measures could help prevent these types of failures.
- 2. This topic review is to help operators become knowledgeable of the operational indications of extraction steam bellows failures.
- 3. Extraction steam bellows failure reports indicate that failures during operation can manifest themselves by a loss of plant efficiency (MWe loss), FW heater extraction steam pressure drops, FW heater level control valve position changes, and chemistry excursions because of debris or condenser tube leaks.
- 4. DAMAGE TO ADJACENT COMPONENTS
- a. On August 31, 2000, Diablo Canyon Power Plant Unit 2 experienced a three MWe drop in main generator output.
A review of plant parameters indicated that the most likely cause of the MW drop was a fault in either the 4a or 5a feedwater heater (FWH) extraction steam bellows. Further investigations indicated a rupture in the 4a extraction steam bellows. On September 4, 2000, indications of a condenser tube leak arose. The unit power was reduced to 50 percent and taken off line on September 5, 2000. The FWH 4a east extraction steam bellows had failed catastrophically. Page 3
C01802C Operational Experience Rev. 0 CONTENT ACTIVITIES Adjacent bellows were damaged, presumably from the shrapnel generated by the initial failure of the bellows, and steam impingement. The debris impacted condenser tubes, causing damage to some of the tubes.
- b. On January 10, 2000, a decrease in megawatt output on Farley Nuclear Plant Unit 1 was noticed. From January 11 to January 23, 2000, the 3A FWH shell pressure trended downward with periodic step change decreases of one psi to seven psi. The extraction steam bellows to the 3A FWH had failed. On March 4, 2000, the unit was shut down for a refueling outage and the failed bellows were replaced.
- Additionally, three top row condenser tubes were found dented.
- c. On January 5, 2000, at Palo Verde Unit 2, a large sodium spike in the 1B condenser hotwell was caused by a condenser tube leak. Debris from 11 extraction steam expansion bellows assemblies had fallen onto the 1B upper tube bundle causing the tube leak.
- d. At the Waterford 3 Nuclear Power Plant one of the extraction steam bellows assemblies from the low-pressure turbine C to the 3C FWH failed. This resulted in a loss of plant efficiency that was first noticed following a plant startup in October 1999, and resulted in additional loss of generation.
- Condenser tube leakage was noted in April 2000. In June 2000, the plant was shut down and the failed assembly was replaced. Approximately 200,000 MWe-hours of generation were lost because of the failure.
- e. On March 8, 1995, Sequoyah Nuclear Plant Unit 2 found the 6B FWH level control valve wide open with the bypass valve cycling. A review of extraction steam pressures showed a drop of approximately 20 percent in the 6B extraction pressure and a drop in electrical output of five MWe at the same time a condenser tube leak occurred.
Inspections of the B condenser found the 6B north extraction steam line bellows blown out. B. LOST GENERATION
- 1. Extraction steam bellows failures are many times Page 4
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 57 Tier # 2 Group # 1 K/A # 059.K3.03 Importance Rating 3.7 Proposed Question: INITIAL CONDITIONS:
- Power is being increased from 90% to 100% in accordance with OP-3204 "At Power Operation".
- Both turbine driven feedwater pumps are operating.
- Two condensate pumps are in service.
- 6 condensate demineralizers are in service.
A malfunction has caused the recirc valve for the "B" TDMFW pump to travel to the full open position. Without operator action, what impact might this failure have on plant conditions? A. SG Lo-Lo level resulting in a reactor trip. B. Lowering demineralizer d/p resulting in degrading SG chemistry. C. Standby condensate pump start, preventing a reactor trip. D. Auto start of the MDFW pump, preventing a reactor trip. Proposed Answer: A Explanation (Optional): Millstone 3 OE has shown that the maximum number of condensate demineralizers should be available to provide adequate feed pump suction flow when increasing feed flow, since condensate pump flow is limited by the number of demins in service. The recirc valve opening will cause condensate flow to increase by 5,400 gpm. Without operator action to start the standby condensate pump or open the demin bypass valve, the added condensate flow due to recirc flow back to the condenser will cause demineralizer d/p to increase ("B" incorrect), and suction pressure to the feed pumps to decrease. The feed pump speed will increase, attempting to restore Feed Reg Valve DP, further lowering feed pump suction pressure. After 30 seconds the feed pumps will trip on low suction pressure ("A" correct). The MDFW pump may auto start on low discharge pressure (less than 955 psig), but this will aggravate the low suction pressure condition ("D" incorrect). Technical Reference(s): OP 3353.MB5C, 3-4 and 3-5 (Attach if not previously provided) OP 3321, step 4.3.3, 4.3.35 Proposed
. references to be .provided to applicants
.. during.. examination: None 41 of 46 NUREG-1021, Revision 8, Supplement 1
ES-401 Written Examination Question Worksheet Form ES-401-6 Learning MC-04660 DESCRIBE the operation of the following Main Feedwater & Steam (As available) Objective: Generator Water Level Control Systems Controls & Interlocks... Main Feed Pump Recirculation Valves (FWR-FV2 1A/B, FV20)... Turbine Driven Main Feed Pump Master Speed Controller (FWS-SK509A)... Low Main Feed Pump Suction Pressure Interlock... Low Main Feed Pump Discharge Pressure Interlock... Question Source: Bank # 69604 Question History: Previous NRC Exam Question Cognitive Level: - Comprehension or Analysis 10 CFR Part 55 Content: 55.41.5, 41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
61 Setpoint: Less than 267 psig I 3-4 TDFW PP B SUCTION PRESSURE LO I. AUTOMATIC FUNCTIONS
- 1. 3FWS-P2B, TDFW pump B, trips when suction pressure is less than 242 psig following a 30 second time delay.
CORRECTIVE ACTIONS
- 1. IF TDFW pump B trips, Go To MB5C 5-3, "TDFW PP B TRIP."
- 2. IF CND- PDI 33, condensate pump discharge pressure is low, START standby condensate pump (MB6).
- 3. IF annunciator MB6A 2-7, "COND DEMIN DP HIGH," is lit, Go To MB6A 2-7.
- 4. CHECK feedwater pump recirculation valves operating properly
- 5. IF directed by SM/US, REQUEST Generation Electrical Services department check contact resistance on suction pressure agastat relays.
SUPPORTING INFORMATION
- 1. Initiating Device Process Setpoint 1.1 3CNM-PS76B (alarm) 1.1 < 267 psig 1.2 3CNM-PS74B (trip) 1.2 < 242 psig
- 2. Procedures 2.1 OP 3319C, "Condensate Demineralizer Mixed Bed System"
- 3. Control Room Drawings 3.1 LSK 7-5A OP 3353.MB5C Level of Use ST OP HTHINK
; .~{-
A
. ,T ;. R V I.
Rev. 003 - 05 Continuous 33 of 54
Setpoint: 5,180 gpm decreasing e NOTE Alarm is expected on feedwater pump startup or during low power operation. AUTOMATIC FUNCTIONS
- 1. None CORRECTIVE ACTIONS
- 1. CHECK 3CNM-FI 72B, TDFW pump B suction flow (MB5), to confirm alarm.
- 2. VERIFY 3CNM-PI 33, condensate discharge pressure (MB6), normal.
- 3. IF directed by SM/US, START standby condensate pump (MB6).
- 4. SEND Operator to verify 3CNM-FIC71B, SGFP recirculation to condenser (auxiliary condensate panel), operating in "AUTO."
- 5. THROTTLE open 3FWR-FV21B, SGFP recirculation to condenser (local).
SUPPORTING INFORMATION
- 1. Initiating Device Setpoint 1.1 3CNM-FS71B 1.1 5,180 gpm decreasing I1.H
;k OP 3353.MB5C Level of Use .00, .for<. .A
-`ACT' i ' rC Tr . .
REW~
- V E E
A -... . .; .... Rev. 003-05 Continuous 34 of 54
. 51 4.3 Starting TDFW Pump A NOTE
- 1. This Section addresses starting TDFW pump A from both cold and warm conditions.
- 2. This Section is performed when directed by one of the following:
- Section 4.5, "Shifting from Auxiliary Feedwater to Main Feedwater"
- Section 4.7, "Shifting from MDFWP to TDFWP A'
- Section 4.12, "Shifting from TDFWP B to TDFWP A' 4.3.1 VERIFY initial conditions as follows:
- a. TDFW pump A is tripped and has been operating on turning gear for at least two hours.
- b. To prevent damage due to poor oil circulation, TML-T163 lube oil temperature, shall be greater than 80'F
- c. TDFW pump A eccentricity is less than 1 mil using Foxboro DCS (TFIN25A under TagSrc).
4.3.2 VERIFY at least two condensate pumps are running. 4.3.3 In anticipation for additional condensate flow from TDFW pump A recirc flow, Refer To OP 3319C, "Condensate Demineralizer Mixed Bed System," and PLACE additional demineralizers in service as necessary. 4.3.4 VERIFY 3TFM-V8, TDFW pump turbine 1A steam inlet isolation open. 4.3.5 VERIFY 3FWS-V986, flushing recirculation manual isolation, locked closed. 4.3.6 OPEN 3FWS-V47, TDFW pump 2A bypass warmup isolation. 4.3.7 IF this is the first feed pump to be operated, CLOSE and LOCK 3SST-V914, first point FW heater RO bypass (prevents overpressurizing the secondary sample sink). Level of Use O 3321 Continuous R~ e.050
N7 CAUTION V7 To minimize pump impeller damage at higher speeds, turbine driven feedwater pump suction flow setpoint must be adjusted to 5,400 gpm prior to exceeding 3,500 rpm. NOTE Actual suction flow may be less than 5,400 gpm when the turbine driven main feedwater pump is running at 3,500 rpm. 4.3.35 To raise turbine speed, PERFORM the following:
- a. STATION an operator at the Auxiliary Condensate Panel (AC) and establish communications with Control Room.
- b. Simultaneously PERFORM the following:
- ADJUST FWS-SK46A, "PP A SPEED CNTL," to raise turbine driven feedwater pump speed to approximately 3,400 rpm (MB5).
- Slowly INCREASE 3CNM-FIC71C, 'A TDFW PP SUCT FLOW TO RECIRC VV 3FWR-FV21A,"
setpoint to 5,400 gpm and VERIFY in 'AUTO" (AC). V7 CAUTION V To minimize feedwater system transient following opening 3FWS-MOV23C, "PP A DIS" the desired discharge pressure on 3FWS-PI 21C is slightly lower than feedwater header pressure as indicated on 3FWS-PI 32, "1ST PT HTR OUT" "HDR PRES." 4.3.36 To place TDFW pump A in service, PERFORM the following:
- a. ADJUST FWS-SK46A, "PP A SPEED CNTL," to establish desired discharge pressure using 3FWS-PI 21C.
- b. OPEN 3FWS-MOV23C, "TDFW A' "PP A DIS."
4.3.37 CLOSE 3FWS-V47, TDFW pump 2A bypass warmup isolation. OP 3321 Level of Use Rev. 015-03 Continuous 25 of 122
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 58 (SRO) Tier # 2 Group # K/A # )61.GEN.2.2.24 Proposed Question: Importance Rating 3.8 INITIAL CONDITIONS:
- The plant is at 100% power.
- The Turbine Driven Auxiliary Feedwater Pump (Terry Turbine) was removed from service one hour ago for preventative maintenance.
- All other equipment is in service.
The following sequence of events occurs:
- 1. Maintenance reports that both MDAFW pumps had improper lube oi I added during the last lube oil change, and the SM declares both pumps INOPERABLE.
- 2. Operations and Maintenance Dept immediately commence efforts to restore the AFW pumps to OPERABLE.
For this condition, what ACTION is required by Technical Specification s concerning reaching HOT STANDBY? A. Be in at least HOT STANDBY within 6 hours. B. Within 1 hour initiate action to be in a MODE where the specific ation does no apply, and be in at least HOT STANDBY within the next 6 hours. C. Restore at least one AFW pump to OPERABLE within 72 hours or be in at least HOT STANDBY within the next 6 hours. D. Shutdown to HOT STANDBY is not required. Proposed Answer: D Explanation (Optional): "D" is correct, per LCO 3.7.1.2 ACTION c. This requires immediate action to restore at least one AFW pump to OPERABLE, which is already in progress. "A" is wrong, but Iplausible, since 3.7.1.2 ACTION b applies with two INOP AFW pumps. "B" is wrong, but plausible since thiss applies if LCO 3.0.3 is in effect. "C" is wrong, but plausible since LCO 3.7.1.2 ACTION a applies with a single INOP AFW pump. Technical Reference(s): Tech Spec 3.7.1.2 (Attach if not Ipreviously provided) Proposed references to be provided to applicants during examination: Tech Spec 'Sections 3/4 Learning MC-04837 The crew operates the plant in compliance with all applic,able plant (As available) Objective: procedures and technical specifications Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.2 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
January 3, 1995 AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary7-feedwater pumps and associated flow paths shall be OPERABLE with:
- a. Two motor-driven auxiliary feedwater pumps, each capable of being powered from separate emergency busses, and
- b. One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.
APPLICAB1LITY: MODES 1, 2, and 3. ACTION:
- a. With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. With two auxiliary feedwater pumps inoperable, be STANDBY within 6 hours and in HOT SHUTDOWN withinin the at least HOT following 6 hours.
- c. With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.
an OPERATIONAL MOOE pursuant to Specification 3.0.4 isnotEntry into permitted with three auxiliary feedwater pumps inoperable. SURVEILLANCE REQUIRDEJETS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:
- a. At least once per 31 days by:
- 1) Verifying that each non-automatic valve in the flow path that isnot locked, sealed, or otherwise secured in position is in its correct position; and
- 2) Verifying that each auxiliary feedwater control and isolation valve in the flow path is in the fully open position when above 10% RATED THERMAL POWER.
KILLSTONE - UNT 3 3/4 7^4 ANENDKENT NO. 7,1OO
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 59 (SRO) Tier # 2 Group # 1 K/A # 063.GEN.2.2.22 Importance Rating 4.1 Proposed Question: With the plant at 100% power and Battery Charger 8 out of service for maintenance, the following sequence of events occurs: An electrical fault occurs on Battery Bus #3
- 1. The "Battery 3 Trouble" annunciator lights on MB8.
- 2. The BOP reports Battery Bus #3 voltage on MB8 indicates zero.
- 3. The PEO reports that the Battery supply breaker to Battery bus #3 has tripped open.
- 4. The PEO reports that the DC output breaker has tripped open from Charger 3BYS*CHGR-3.
For this condition, how long does the crew have to reach HOT STANDBY? A. Reenergize the DC bus from the battery within 2 hours or be in HOT STANDBY within the next 6 hours. B. Reenergize the DC bus from the battery within 8 hours or be in HOT STANDBY within the next 6 hours. C. Restore the Battery and the Charger to OPERABLE status within 2 hours or be in HOT STANDBY within the next 6 hours. D. Restore the Battery and the Charger to OPERABLE status within 24 hours or be in HOT STANDBY within the next 6 hours. Proposed Answer: A Explanation (Optional): "A" is correct, since the DC bus is not energized from its battery bank, making LCO 3.8.3.1 ACTION c apply. "B" is wrong, but plausible, since this is the action of LCO 3.8.3.1 ACTION a if an emergency bus is deenergized. "C" is wrong, but plausible since this is the action of LCO 3.8.2.1 ACTION a if Battery Bank 1 or 2 is inoperable. "D" " is wrong, but plausible since this is the action of LCO 3.8.2.1 ACTION b if Battery Bank 3 or 4 is inoperable, and this is less restrictive than LCO 3.8.3.1 ACTION a.. Technical Reference(s): LCO 3.8.2.1 (Attach if not previously provided) LCO 3.8.3.1 Proposed references to be provided to applicants during examination: Tec h Spec Sections 3/4 Learning MC-03951 Given a plant condition requiring the use of AOP-3563, (As available) Objective: identify applicable technical specification action requirements. Question Source: Bank # 68683 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.2 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
3/09/92 FtECTRUCAL POWER SYsTEMs 3/4.8.2 D.C. SOURCES LTMTTMNG CONDITTON FOR OPERATION 3.8.2.1 As a minimum, the following D.C. electrical sources shall be OPERABLE:
- a. 125-volt Battery Bank 301A-1, and an associated full capacity charger,
- b. 125-volt Battery Bank 3D1A-2, and an associated full capacity charger,
- c. 125-volt Battery Bank 301B-1 and an associated full capacity charger, and
- d. 125-volt Battery Bank 3015-2 and an associated full capacity charger.
APPLJCABTLMT: NODES I, 2,3, and 4. ACtTON:'
., 7 ,
- a. With either Battery Bank'30IA-1 or 301-5-1; nd/or one of the
. . t x --- ....'. .. required full capacity chargers inoperable, restore the inoperable
- . I... -; .
battery bank and/or full capacity charger to OPERABLE status within 2hoursayof be in at'least'IOT'STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With either Battery Bank 301A-2 or 3016-2 inoperable, and/or one of the required full capacity chargers inoperable, restore the tnoper-ible battery bank lmd/or. full, capacity charger to OPERABLE status wlftfirn' 24hours or be Inf it least..HOT STANDBY within the next hours and in COLD SHUTDOWN within the following 30 hours.
.1 . ..
SURVELLAYCE REOUTREMTS 4.8.2.1 Each'12S-volt battery bank and charger shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that:
- 1) The parameters in Table 4.S-2a meet the Category A limits, and
- 2) The total battery terminal voltage is greater than or equal to 129 volts on float charge.
MILLSTONE - UNIT 3 3t4 8-11 Amendment No. 64 I
3/09/92 ELECTRICAL POWER SYST 3/4.8.3 ONSITE POWER DISTRIBUTION OPERATING LIMITING CONDITION FOR OPEMATION 3.8.3.1 The following electrical busses shall be energized in the specified sanner:
- a. Train A A.C. Emergency Busses consisting of:
- 1) 4160-Volt Emergency Bus *34C, and
- 2) 480-Volt Emergency Bus 032R, 32S, 32T, and 32Y.
- b. Train B A.C. Emergency Busses consisting of:
- 1) 4160-Volt Emergency Bus #34D, and
- 2) 480-Volt Emergency Bus #32., 32V, 32W, and 32X.
- c. 120-Volt A.C. Vital Bus *VIAC-1 energized from its associated inverter connected to D.C. Bus *301A-1*,
- d. 120-Volt A.C. Vital Bus *VIAC-2 energized from its associated inverter connected to D.C. Bus 1301B-1*,
- e. 120-Volt A.C. Vital Bus IVIAC-3 energized from its associated inverter connected to D.C. Bus *301A-2*,
- f. 120-Volt A.C. Vital Bus fVIAC-4 energized from its associated inverter connected to D.C. Bus *301B-2*,
- g. 125-Volt D.C. Bus f301A-1 energized from Battery Bank f301A-1,
- h. 125-Volt D.C. Bus F3OIA-2 energized from Battery Bank #301A-2, I. 125-Volt D.C. Bus 1301B-1 energized from Battery Bank *301B-1, and J. 125-Volt D.C. Bus *3018-2 energized from Battery Bank #301B-2.
APPLICABILITY: NODES 1, 2, 3, and 4. ACTION:
- a. With one of the required trains of A.C. emergency busses not fully energized, reenergize the division within 8 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the folfowing 30 hours.
- b. With one A.C. vital bus either not energized from Its associated inverter, or with the inverter not connected to its associated D.C.
bus: (1) reenergize the A.C. vital bus within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN
*Two inverters may be disconnected from their D.C. bus for up to 24 hours as necessary, for the purpose of performing an equalizing charge on their associated battery bank provided: (1) their vital busses are energized, and .(2) the vital busses associated with the other battery bank arte energized from their associated inverters and connected to their associated D.C. bus.
MILLSTONE - UNIT 3 3/4 8-16 Amendment No. 64 8654
3/09/92 ELECTRICAL POWER SYSTEMS LTMITING CONDITTIM FOR OPERTION ACTTON (Continued) within the following 30 hours; and (2) reenergize the A.C. vital bus from its associated Inverter connected to its associated D.C. bus within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. With on* D.C.. bus not energized from its associated battery bank, reenergize the D.C. bus from its associated battery bank within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REOUTREMENTS 4.8.3.1 The specified busses shall be determined energized jn the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses. MILLSTONE - UNIT 3 3/4 8-17 knendment No. 64 oS."
Written Examination Question Worksheet Form ES-401-6 ES-401 Examination Outline Cross-reference: Question # 60 (SRO) Tier # 2 Group # I K/A # 068.KI.05 Importance Rating 2.6 Proposed Question: The SM is preparing to authorize the discharge of the "A" Low Level Waste Drain Tank (LLWDT). Current conditions are as follows:
- The plant is in MODE 5.
- An "A" train electrical outage is in progress.
- The "D" Main Circulating Water Pump has been disassembled for bearing replacement.
- The "B" main circulating pump has just tripped, and can not be restarted.
- Both the "A" and the "B" Low Level Waste Drain Tanks (LLWDTs) are full.
- The "A" LLWDT was recirculated for 40 minutes, as annotated on the discharge permit.
- The tide is coming in.
- All other plant conditions are normal, and all other actions related to the discharge are acceptable.
Which of the following actions should the SM take concerning the discharge of the "A" LLWDT? A. Authorize the discharge. All requirements are met. tides. B. Do not authorize the discharge. No provision exists for a discharge during incoming C. Do not authorize the discharge. Tank recirculation requirements are not met. D. Do not authorize the discharge. Insufficient dilution flow exists. Proposed Answer: D available due to the "A" Explanation (Optional): "A" is wrong, and "D" correct, since 3 of the 6 circ pumps are not Only one circ pump is running after the "B" pump tripped. train outage, and one is not available due to maintenance. and a service water pump. "B" is wrong, but plausible, since it is While shutdown, the requirement is 2 circ pumps tides, but needs of the station may preclude this. "C" is wrong since 40 desirable to perform discharges during ebb since the WTT minutes is the time required for recirculating the Low Level Waste Drain Tank. "C" is plausible requires a 175 minute recirc time. Technical Reference(s): OP 3335D, step 4.23 (Attach if not previously provided) OP 3335D, section 4.2.5 OP 3325A, Precaution 3.2 Proposed references to be provided to applicants during examination: None MC-04867 Describe the major administrative or procedural precautions and (As available) Learning Objective: limitations placed on the operation of the LWS system, and the basis for each. Question Source: Modified question 75648 Parent attached Question Cognitive Level: Comprehension or Analysis 41 of 46 NUREG-1021, Revision 8, Supplement 1
Written Examination Question Worksheet Form ES-401-6 ES-401 10 CFRPart 55 Content: 55.41.10 55.43.4, 43.5 Comments: Original question 75648 Current Conditions:
- Reactor power is 100%.
- A thermal backwash is in progress on the "A" Waterbox.
- The tide is coming in.
- Both the "A" and the "B" Waste Test Tanks (WTTs) are full.
- The SM is preparing to authorize the discharge of the "A" WTT.
- The "A" WTT was recirculated for 40 minutes, as annotated on the discharge permit.
acceptable.
- All other plant conditions are normal, and all other actions related to the discharge are WTT?
Which of the following actions should the SM take concerning the discharge of the "A" A. Authorize the discharge. All requirements are met. B. Do not authorize the discharge. Discharges are only allowed during ebb tides. C. Do not authorize the discharge. Tank recirculation requirements are not met. D. Do not authorize the discharge. Insufficient dilution flow exists. Answer: C 41 of 46 NUREG-1021, Revision 8, Supplement 1
4.23 Recirculating and Sampling Low Level Waste Drain Tank A NOTE
- 1. A LLWDT is normally removed from service for recirculation and discharge when its level reaches 80%.
- 2. Under normal conditions a LLWDT is removed from from service at 80% but can be safely filled to 90% in times of reduced discharges.
4.23.1 START 3LWS-P7A, LLWDT pump A. 4.23.2 CHECK closed 3LWS-V690, 3LWS-FLT 4A Inlet Isolation 4.23.3 THROTT'LE 3LWS-V175, LLWDTA recirculation orifice bypass, to maintain 3LWS-PI 68A, LLWDT pump discharge pressure, 50 to 55 psig. 4.23.4 NOTIFY Chemistry of the time LLWDT A went on recirc (minimum recirc time is 40 minutes). 4.23.5 IF discharging LLWDT A, WHEN satisfactory sample results are obtained, Go To Section 4.25. 4.23.6 WHEN recirculation is complete, PERFORM the following:
- a. STOP 3LWS-P7A, LLWDT pump A.
- b. CLOSE 3LWS-V175, LLWDT A recirculation orifice bypass.
- End of Section 4.23 -
OP 3335D Level of Use l A - Rev. 016-04 Continuous ICT 65 of 121
4.25 Discharging Low Level Waste Drain Tank A to Circulating Water Discharge Ibnnel NOTE It is desirable to perform radioactive liquid waste discharges during an ebb tide (high tide going out) but the needs of the station may preclude this. 4.25.1 Refer To OPS Form 3335D-015, "3LWS-TK4A, LLWDT A, Discharge Line-up," and PERFORM the following:
- a. PERFORM lineup and DOCUMENT.
- b. PERFORM Independent Verification and DOCUMENT.
NOTE If activity action limits are met or exceeded, chemistry will not issue a liquid discharge permit and will notify supervision, as directed by Chemistry Procedure 3809A, "Liquid Waste Discharges." 4.25.2 REQUEST SM/US permission to set up 3LWS-RE70 for discharge of 3LWS-TK4A, LLWDT A. 4.25.3 CHECK required dilution flow rate is met. 4.25.4 IF 3LWS-RE70 is OPERABLE, PERFORM the following:
- a. ENTER "LWS70 1" and PRESS "DATABASE" to display 3LWS70-1, Liquid Waste Effluent Monitor (3RMS-CNSL1).
- b. CHECK 3LWS70-1 unit on line and sample pump on (3RMS-CNSL1).
- c. PRESS "PURGE UNIT" to purge radiation monitor (3RMS-CNSL1) (purge lasts for 5 to 10 minutes).
- d. CHECK "ENDED PURGE" message is received approximately 5 to 10 minutes after pressing "PURGE UNIT" (3RMS-CNSL1).
- e. CHECK "UNREACHABLE" message not received (3RMS-CNSL1).
---Al_- AL AL_--- OP 3335D Level of Use Rev. 016-04 Continuous
)KIRr IT 67 of 121
- 3. PRECAUTIONS 3.1 As specified in NPDES Permit, Revision 2, the following temperature limits are applied to the Unit 3 discharge: (Ref. 6.1.2) 3.1.1 The maximum temperature increase at the Unit No. 3 discharge above the intake water temperature shall be 240 F The discharge temperature is measured at the Unit 3 outfall. If this temperature is not available, notify Chemistry that the unit discharge temperature instrumentation is out of service and the time that the probe became inoperable.
3.1.2 The differential temperature increase at the Unit No. 3 discharge above the intake water temperature under conditions of reduced cooling water flow may be increased to 30'F for a period not to exceed 24 hours. In the event the temperature differential exceeds 240F, the Department of Environmental Protection shall be notified in the monthly monitoring report. 3.1.3 The maximum temperature increase at the Quarry Cut above the intake water temperature shall be 32 0 E 3.1.4 The differential temperature increase at the Quarry Cut above the intake water temperature under unusual conditions may be increased to 44 0 F for a period not to exceed 24 hours. In the event the temperature differential exceeds 320F, the Department of Environmental Protection shall be immediately notified and a written report of the incident filed. 3.1.5 The maximum temperature allowed at the Quarry Cut is 105 0 F 3.1.6 The maximum temperature allowed at the Unit 3 Discharge is 980 E 3.2 As specified in NPDES Permit, at least two circulating water pumps must be in operation when any of the following evolutions are in progress: 3.2.1 Discharge of a tank containing radioactive liquid effluent (Discharge Serial No. 001C-2 and 001C-3) 3.2.2 Discharge of steam generator secondary side wet layup water (Discharge Serial No. 001C-1 (a)) 3.2.3 Discharge of water from a non-contaminated closed Cooling Water System (Discharge Serial No. 001C-9) Level of Use OP 3325A Continuous STOP THINK ACT REVIEWv Rev. 021-02 6 of 138
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 61 Tier # 2 Group # 1 K/A # 071.A4.14 Importance Rating 3.0 Proposed Question: With the plant at 100% power the following annunciators are received in the control room:
- Main Board 1B, 3-3, "RAD GASEOUS WASTE SYS TROUBLE".
- Main Board 6A, 2-10, "AUX COND CONDUCT HI".
A PEO is dispatched, and he reports that the local annunciator is Gaseous Waste Panel annunciator 2-6, "GAS WASTE FEED PREHTR OUT COND HI". What are all of the automatic responses that the operators should expect from the alarms? A. Degasifier Feed Preheater Steam Supply Valve (3ASS-TV31) closes. B. Degasifier Feed Preheater Steam Supply Valve (3ASS-TV3 1) closes, and Degasifier Preheater Condensate Divert Valve (3CNA-AOV46) diverts to the Auxiliary Building Sump. C. Degasified Stream Outlet Valve (3GWS-AOV54) transfers to Boron Recovery System. D. Degasified Stream Outlet Valve (3GWS-AOV54) transfers to Boron Recovery System, and Volume Control Tank to Degasifier Letdown Valve (3CHS-AOV71) transfers to the VCT. Proposed Answer: B Explanation (Optional): ("B" Upon receipt of high conductivity, 3CNA-AOV 46 diverts to the Aux Building Sump AND 3ASS-TV31 closes (such as correct, "A" wrong). "C" and "D" are wrong, and plausible, since numerous alarms and conditions degasifier high level) cause these valves to auto reposition to the positions listed above. Technical P&ID 109A and 135C (Attach if not previously provided) Reference(s): OP 3353.GW (2-6) OP3353.MB6A (2-10) Proposed references to be provided to applicants during examination: None MC-04733 Given a failure, partial or complete, of the GWS system, (As available) Learning Objective: determine the effects on the system and on interrelated systems. Question Source: Bank # 60456 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
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Setpoint: Greater than 45 gmho/cm 2-6 GAS WASTE FEED PREHTR OUT COND Hi AUTOMATIC FUNCTIONS
- 1. MB6A 2-10 annunciates Aux. Bldg. sump
- 2. 3CNA-AOV46, CNA flash tank 2 divert valve, diverts to
- 3. 3ASS-TV31, auxiliary steam preheater valve, closes CORRECTIVE ACTIONS
- 1. NOTIFY Chemistry Department of alarm condition.
(local), to
- 2. CHECK 3CNA-CIT45, gas waste feed preheater outlet conductivity confirm alarm.
outlet temperature
- 3. CHECK 3CNA-TI 54, gas waste feed preheater sample cooler (local), for proper operation.
and REMOVE degasifier
- 4. Refer To OP 3337, "Radioactive Gaseous Waste System,"
feed preheater from service. SUPPORTING INFORMATION
- 1. Initiating Device Setpoint 1.1 3CNA-CS45B 1.1 > 45 gmho/cm l ©
- 2. Procedures 2.1 OP 3337, "Radioactive Gaseous Waste"
- 3. Control Room Drawings 3.1 ESK 1OMH 3.2 LSK 31-2.1A 3.3 LSK 4-4B l......... C1OP 3353.GW Level of Use ~ N<Re.000 4
Continuous. Re7of 28 5
Setpoint: Greater than 45 micromho/cm 2-10 ] (9' AUX COND CONDUCT HI AUTOMATIC FUNCTIONS
- 1. IF 3CNA- CE42, waste evaporator reboiler conductivity, is high, the following occur:
- 3CNA-AOV43, condensate divert, diverts condensate to reactor plant drains
- 3ASS-AOV40, steam supply, closes
- 2. IF 3CNA- CE45, gas waste feed preheater outlet conductivity, is high, the following occur:
- 3CNA-AOV46, condensate divert, diverts condensate to aux building sump
- 3ASS-TV31, auxiliary steam preheater valve, closes
- 3. IF 3CNA-CE39, boron evaporator reboiler conductivity, is high, the following occur:
- 3CNA-AOV40, condensate divert, diverts condensate to aux building sump
- 3ASS-PV30, steam supply, closes CORRECTIVE ACTIONS
- 1. CHECK the following to confirm and locate high conductivity:
1.1 3CNA-CE42, waste evaporator reboiler conductivity, and 3CNA-TI 53, waste evaporator reboiler temperature. 1.2 3CNA- CE45, gas waste feed preheater outlet conductivity, and 3CNA-TI 54, gas waste feed preheater outlet temperature. 1.3 3CNA-CE39, boron evaporator reboiler conductivity, and 3CNA-TI 52, boron evaporator reboiler temperature.
- 2. IF 3CNA-CE42, waste evaporator reboiler conductivity, is high, Refer To OP 3335D, "Radioactive Liquid Waste System," and REMOVE waste evaporator reboiler from service.
- 3. IF 3CNA- CE45, gas waste feed preheater outlet conductivity, is high, Refer To OP 3337, "Radioactive Gaseous Waste System," and REMOVE degasifier feed preheater from service.
- 4. IF 3CNA-CE39, boron evaporator reboiler conductivity, is high, Refer To OP 3335C, "Boron Recovery System," and REMOVE boron evaporator reboiler from service.
Level of Use OP 3353.MB6A Continuous Rev. 005-01 42 of 84 I
U2-10 SUPPORTING INFORMATION
- 1. Initiating Devices 1.1 3CNA-CE42 1.1 >45 micromho/cm 1.2 3CNA-CE45 1.2 >45 micromho/cm 1.3 3CNA-CE39 1.3 >45 micromho/cm
- 2. Computer Points 2.1 CNA-COND
- 3. Procedures 3.1 OP 3335C, "Boron Recovery System" 3.2 OP 3335D, "Radioactive Liquid Waste System" 3.3 OP 3337, "Radioactive Gaseous Waste System" OP 3353.MB6A Level of Use Rev. 005-01 Continuous 43 of 84 I
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 62 (SRO) Tier # 2 Group #I K/A # F'LANT SPECIFIC Importance Rating I V/A Proposed Question: With the plant at 100% power, the following sequence of events occurs:
- 1. The N-16 ALERT annunciator on MB2 comes in for the "D" SG, wit h 40 gpd leakage.
- 2. The crew enters AOP 3576 "Steam Generator Tube Leak".
- 3. The leak is verified by an increasing trend on Air Ejector Rad Monitc ir3ARC-RE21.
- 4. The N-16 "Rate of Change" annunciator comes in.
What actions are directed by AOP 3576? A. Request Chemistry to perform SP3861 "Primary to Secondary Leak IRate Determination" at the specified frequency while continuing with normal plant operation S. B. Reduce leakage to within Tech Spec limits within 4 hours or be in at least HOT STANDBY within the next 6 hours. C. Go to OP 3204 "At Power Operation" or AOP 3575 "Rapid Downpo' wer" and be in MODE 3 within approximately 6 hours. D. Go to AOP 3575 "Rapid Downpower", reduce power to < 50% withii in1 hour, and be in MODE 3 within the next 2 hours. Proposed Answer: D Explanation (Optional): "A" is wrong, since this action is performed with the ARC-21 setpoints reset AND clear. The given leak rate (40 gpd and increasing) is less than the'. rS limit of 500 gpd through the affected SG ("B" wrong). "D" is correct, since this is required with tl Le "rate of increase" alarm.
"C" is wrong, since the crew is directed to commence a plant shutdown t ising OP 3204 or AOP 3575 to get to MODE 3 within approximately 6 hours with greater than 150 gl d leakage.
Technical Reference(s): AOP 3576, steps 7- 9 (Attach if'not previously provided) Proposed references to be provided to applicants during examination: Tech Spec sections 3/4 Learning MC-00 189 Discuss conditions which require transition to (As available) Objective: other procedures from AOP 3576. Question Source: Bank # 74488 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.10 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
EP ACTIONEXPECTED RESPONSE RESPONSE NOT OBTAINED
- 7. Check If Unit Shutdown Should Be Initiated
- a. Check any of the following: a. Proceed to step 8.
- Condenser air ejector radiation monitor -
IN ALERT OR ALARM OR
- Chemistry grab sample indicates primary to secondary leakage in any SG - GREATER THAN OR EQUAL TO 75 gpd OR
- Condenser air ejector radiation monitor correlation to leak rate (gpd) indicates primary to secondary leakage -
GREATER THAN OR EQUAL TO 75 gpd ORl
- N16 monitor in ALARM l
- b. Evaluate event using MP-26-EPI-FAP06-003, Unit 3 Emergency Action Levels (Barrier Failure)
- c. Check plant status - c. Proceed to step 10.
MODE 1 0R2
- d. Proceed to step 9.
ACTION/EXPECTED RESPONS RESPONSE NOT OBTAINED
- 8. Perform Continued Monitoring
- a. Request Chemistry perform SP 3861, "Primary to Secondary Leak Rate Determination," at the specified frequency
- b. Check air ejector radiation b. Proceed to step 8.e.
monitor ALERT and ALARM setpoints have been reset - BASED ON CURRENT CHEMISTRY LEAK RATE CALCULATION
- c. Check condenser air ejector c. Return to step 5.
radiation monitor ALERT/ALARMs - CLEAR
- d. Continue with normal plant evolutions using applicable plant procedures
- e. Check primary to secondary e. Return to step 7.
leakage - STABLIZED ( < 10% increase in 1 hour period)
- f. Request Chemistry provide new ALERT and ALARM setpoints for the condenser air ejector radiation monitor based on current leak rate calculation
- g. Check new ALERT and g. Return to step 5. and, ALARM setpoints - WHEN AVAILABLE FROM New ALERT and ALARM Chemistry setpoints are available, THEN Perform steps 8.h., 8.i., and 8j.
STEP ACTION/EXPECTED RESPONSH RESPONSE NOT OBTAINEDl
- 8. (continued)
- h. Using OP 3362, "Radiation Monitor System Display and Control System," Adjust radiation monitor ALERT and ALARM setpoints for the condenser air ejector radiation monitor
- i. Request Chemistry provide the following:
- Correlation of condenser air ejector radiation monitor readings to leak rate (gpd)
- Leak rate trend plots
__j. Return to step 7.
STEAM GENERATOR TtjdE AOP 3576 Page 12 of 31 LEAK Rev. 001 I STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINEDl NOTE
- The shutdown times provided are EPRI recommendations and should not be achieved at the cost of plant stability or intentional reactor trip.
- If a unit shutdown is initiated based on a condenser air ejector or N16 monitor ALERT or ALARM condition and subsequent Chemistry calculations indicate actual leakage does not meet I or exceed a shutdown criterion, then the shutdown may be suspended.
- The rate of increase limit does not apply to leak rate spikes followed by decreasing leak rates.
- Plant shutdown may be accomplished using AOP 3575, "Rapid Downpower," or appropriate General Operating Procedures as determined necessary.
- 9. Perform Unit Shutdown
- a. Verify a leakage rate of a. Proceed to step 9.c.
increase limit met:
- Check leakage increased in any SG by -
GREATER THAN OR EQUAL TO 15 gpd IN A 30 min PERIOD OR
- Check condenser air ejector radiation monitor RATE OF CHANGE ALARM - IN ALARM FOR 30 min OR
- Check N16 S/G Leak Detection Status (PPC)
HIGH RATE (75 GPD & 15 GPD RISE IN 30 MIN)
- IN ALARM
STEAM GENERATOR TUBE AOP 3576 Page 13 of 31 LEAK Rev. 001 _ ACIONEXPCTE SEF REPONE RSPOSE NOT OBTAINEDI
- 9. (continued)
- b. Proceed to step 9.d.
- c. Check charging pumps - c. Proceed to step 9.i.
TWO RUNNING
- d. Check power level - d. Proceed to step 9.g.
GREATER THAN 50%
- e. Initiate power reduction to be LESS THAN 50% within 1 hour (downpower rate of 3%/min or 5%/min recommended)
- f. Return to step 9.d.
- g. Initiate power reduction to be in MODE 3 within the next 2 hours
- h. Proceed to step 9.k.
- i. Check SG leakage: i. Return to NOTE prior to step 9. and,
- Condenser air ejector WHEN radiation monitor - IN Leakage in any SG has ALARM remained GREATER THAN OR EQUAL TO OR 75 gpd for 1 hr, THEN
- Chemistry grab sample indicates any SG leakage 1) Initiate power reduction to
- GREATER THAN OR be in MODE 3 within EQUAL TO 150 gpd 24 hours from the initial indication of exceeding OR 75 gpd (Alert, Alarm or sample).
- N16 S/G Leak Detection Status (PPC) 2) Proceed to step 9.k.
HIGH HIGH (>150 GPD) - IN ALARM
STEP ACTION/EXPECTED RESPONS RESPONSE NOT OBTAINED
- 9. (continued)
__j. Initiate power reduction to be in MODE 3 within 6 hours.
- k. Check plant status - k. Return to NOTE prior to MODE 3 step 9.
- 10. Check If Shutdown Banks Should Be Inserted
- a. Check reactor trip breakers - a. Proceed to step 11.
CLOSED
- b. Using OP 3302A, "Control Rod Drive," Perform Shutdown Bank insertion
- 11. Initiate Immediate Boration Of RCS
- a. START one boric acid transfer pump
__b. OPEN emergency boration b. Locally Open valve. valve (3CHS*MV8104)
- c. Place the charging flow control valve in MAN and Adjust to maintain net flow to the RCS - GREATER THAN OR EQUAL TO 65 gpm
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 63 (SRO) Tier # 2 Group # 2 K/A # 006.A2.08 Importance Rating 3.3 Proposed Question: With the plant at 100% power, the following sequence of events occurs:
- 1. An earthquake occurs, resulting in both a 34C bus differential, and a small break LOCA.
- 2. The crew transitions to ES-1.2 "Post LOCA Cooldown and Depressurization".
- 3. The crew is preparing to isolate accumulators per ES-1.2, step 22 "Check if Acci imulators Should Be Isolated".
What will be the impact of the loss of bus 34C on the performance of step 22? A. The crew will not need to isolate the "A" and "C" accumulators, since the isolati' on valves did not open on the SIS signal. B. The crew will not need to isolate the "A" and "C" accumulators, since the isolati' on valves failed closed. C. The crew will be unable to close the isolation valves for the "A" and "C" accurni ilators. The crew will mitigate this by venting nitrogen off of both of these accumulators. D. The crew will be unable to close the isolation valves for the "A" and "C" accumr flators. The crew will also not be able to vent nitrogen off of these accumulators. Proposed Answer: C Explanation (Optional): Accumulators must be isolated to prevent N2 injection into the RCS. "A" auid "C" accm isol valves are normally open MOVs powered from "A" Train 48OVAC, so they can not be closed on losss of power ("A" and "B" wrong). ES- 1.2 directs the crew to vent the unisolable accumulators, and this is possible, sirice the vent valves are in parallel for each accumulator, powered from opposite train 125VDC power ("C" correc t, "D" wrong). Technical ES-1.2, step 22. (Attach if not previously provided) Reference(s): EOP 35 GA-7 Isolating Accumulators P&ID 112B Proposed references to be provided to applicants during examination: None Learning MC-062809 Given a failure, partial or complete, of the Emergency Core Cooling (As available) Objective: System, determine the effects on the system and on interrelated systems. Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.5, 41.8, and 41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
[POST LOCA COOLDOWN AND DEPRESSURIZATION EOP 35 ES-1.2 Rev. 013 Page 25 of 31 ACTION/EXPECTED RESPONS SP RESPONSE NOT OBTAINED I
- 20. Verify Adequate Shutdown Margin
- a. Request Chemistry sample the RCS
- b. Perform shutdown margin using OP 3209B, Shutdown Margin
- 21. Verify ECCS Flow Not Required
- a. Check RCS subcooling based a. Align valves and START I on core exit TCs - ECCS pumps as necessary and GREATER THAN 32"F Proceed to step 22.
(115 0 F ADVERSE CTMT)
- b. Check PZR level - b. Align valves and START GREATER THAN 16% ECCS pumps as necessary and (50% ADVERSE CTMT) Return to NOTE prior to step 11.
- 22. Check If SI Accumulators Should Be Isolated
- a. Check RCS subcooling based a. Perform the applicable action:
on core exit TCs - GREATER THAN 32 0F
- IF at least two RCS hot leg (115 0F ADVERSE CTMT) WR temperatures are LESS THAN 380cF THEN Proceed to step 22.c.
- IF at least two hot leg WR temperatures are NOT LESS THAN 3800F THEN Proceed to step 23.
- b. Verify PZR level - b. Return to NOTE prior to GREATER THAN 16% step 11.
(50% ADVERSE CTMT)
- c. Using GA-7, Isolate SI accumulators I
- w STEP ACTION/EXPECTED RESPRESPONSE NOT OBTAINED I
- 1. Locally Unlock And Place The SI Accumulator Isolation Valve Breakers To ON
- 32-2R-F4M
- 32-2R-R5F
- 32-2W-F4M
- 32-2W-R3J
- 2. RESET SI, If Necessary
- 3. CLOSE All SI Accumulator Vent any unisolated Isolation Valves accumulator(s):
- 3SIL*MV8808A a. Verify SI accumulator nitrogen supply valves (3SIL*CV8880
- 3SIL*MV8808B an 3SIL*CV8968) closed on MB2.
- 3SIL*MV8808C b. For each accumulator requiring venting, OPEN one
- 3SIL*MV8808D from each pair of the following isolation valves:
- For tank A (3SIL*SV8875A or 3SIL*SV8875E)
- For tank B (3SIL*SV8875B or 3SIL*SV8875F)
- For tank C (3SIL*SV8875C or 3SIL*SV8875G)
- Fortank D (3SIL*SV8875D or 3SIL*SV8875H)
- c. OPEN one SI accumulator vent control valve (3SIL*HC943A or 3SIL*HC943B).
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ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 64 Tier # - 2 Group # 2 K/A # - 0 10.A4.03 Importance Rating 3.8 - Proposed Question: Initial conditions:
- 1. The reactor tripped and safety injection actuated.
- 2. The crew is in E-1 "Loss of Reactor or Secondary Coolant", checking if ECCS flow should be reduced.
PORV 3RCS-PCV455A fails OPEN. How will the crew diagnose which PORV is open; and what action does E-1 direct the crew to take to isolate flow through the PORV? A. The PORV is verified open by its tail piece temperature indication on MB4, and flow is isolated by going to CLOSE on the PORV control switch on MB4. B. The PORV is verified open by its tail piece temperature indication on MB4, and flow is isolated by going to "decrease" on the master pressure controller on MB4. C. The PORV is verified open by its red indicating light on MB4, and flow is isolated by going to CLOSE on the PORV control switch on MB4. D. The PORV is verified open by its red indicating light on MB4, and flow is isolated by going to "decrease" on the master pressure controller on MB4. Proposed Answer: C Explanation (Optional): PORV indication is via the red indicating light, since the tailpipe temperature indication is common to both PORVs ("A" and "B" wrong). PORV operation is via its control switch, and due to a plant modification, the master pressure controller no longer inputs to the PORV ("C" correct, "D" wrong). Technical Reference(s): E- 1, step 5. (Attach if not previously provided) P&ID 102C. Proposed references to be provided to applicants during examination: None Learning MC-05339 Demonstrate the ability to manually manipulate:... PORV Block (As available) Objective: Valve and PORV... Question Source: 2000 LOIT NRC Question 21 Question History: Millstone 3 2000 LOIT NRC Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
LOSS OF REACTOR OR EOP 35 E-1 Page 6 of 19 SECONDARY COOLANT Rev. 018 ACTION/EXPECTED RESPONS STP RESPONSE NOT OBTAINED I 2 CAUTION If any PZR PORV opens because of high PZR pressure, step 5.a. should be repeated after pressure decreases to LESS THAN 2350 psia.
- 5. Check PZR PORVs and Block Valves
- a. Verify PORVs - a. CLOSE the PORV(s) if PZR CLOSED pressure is LESS THAN 2350 psia.
IF any PORV can NOT be closed, THEN CLOSE its block valve.
- b. Verify block valves - b. OPEN one block valve AT LEAST ONE OPEN UNLESS it was closed to isolate an open PORV
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Written Examination Question Worksheet Form ES-401-6 ES-401 Question # 65 Examination Outline Cross-reference: Tier # 2 Group # 2 K/A # 01 1.A1.04 Importance Rating 3.3 Proposed Question: The unit has been reducing power from 100%. When the crew stabilizes power, Auctioneered High Tave stabilizes at 568 OF. At what level will the Pressurizer stabilize? A. 35% B. 40% C. 45% D. 50% Proposed Answer: B Explanation (Optional): Tave increases from 5571F to 587'F.
"B" is correct since PZR level varies with Tave from 28% to 61.5% linearly as 28% + 12.3% = 40.3%. All distractors are 568 0F is 11 F above no load Tave, so 11/30 = X/33.5. Solving, X = 12.3.
0 plausible since they are within the normal PZR level operating band. Technical Reference(s): Functional Dwg #11 (Attach if not previously provided) None Proposed references to be provided to applicants during examination: (As available) Learning MC-05341 Describe the operation of the Pressurizer Pressure and Level Control Objective: System under Normal, Abnormal, and Emergency Operating conditions. Modified Bank # 73123 (Parent attached) Question Source: Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
Written Examination Question Worksheet Form ES-401-6 ES-401 Original Question 73123 PLANT CONDITIONS: The unit is reducing power from 100% to take the unit off-line. Loop 3 Tave fails unobserved to a constant output of 572°F. Assume no operator action taken. Pressurizer level will stabilize at: A. 22% B. 28% C. 45% D. 89% Answer: C 41 of 46 NUREG-1021, Revision 8, Supplement 1
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- 3. PRELWREMISTAB.E ND. P0-4554 AND LEVEL 515rABLES N40.-
ALL HEATER WI L"59- 5CLb459E. iL&-46C)0 ARE ENIER'ZE TO ACUAT9E 4 .OPEN/S4UT INDICATIONI IN 0.UullO FCI AND ow THE AUXKILIAQY( S"JTDo~M PANeL (Asp) BACK-UP-COK)TROL- PCV-455E& 1 LIG.WT %qouLo am pqn s w-rIAr GO*AmTcp 0com Poa vU.&w4. HEATERS SIGNIAL (soVE. S) SPMW NV4.W.YI 1VV r0rOIAW.~0T PiLL.' C.IJ5tgc.. (SHEET It) (SNEEr 12) a-..THE.F C0hlTROL5 OIjTHl COI.)TP.@. oA%S SI&P&IUoO y@TrEr~S.
- 1. LOCAL Co"TROL. AX *Tmr AUXILIARY SOIUTCO0044 PA4AEL *vLERR%0t- THEL 5I1hb.Lb FROM T%9 OPWTXC. 0COA4 ANID CTU~TES Atj ALARm AII&UCIATOER II. T14%, COUTROL I ROOPAA 8.A avEMOTE/LOCAL CONTROL TRANSFER SWITrCH is LOCATED ONJA LOCAL TRANPER 3WrTCI4 PANEL (TSP).
- - - -- - -- - - - I llsTaws O 1.AICOsrAtrIMO.
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- K"MALSHE11615N TOTHISSllS AT :I I4 551 A -OU LrAAPRomIII. ILL..J.._______
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ES-40 I Written Examination Question Worksheet Form ES-40 1-6 Examination Outline Cross-reference: Question # 66 (SRO) Tier # 2 Group # 2 K/A # 012.GEN.2.1.12 Importance Rating 4.0 Proposed Question: Initial conditions:
- The plant is in MODE 3 after a refueling outage.
- The shutdown banks are withdrawn.
- Preparations are being made to perform a reactor startup.
Engineering reports to the Shift Manager that a modification was installed improperly on the reactor trip switch on Main Board 7, and that it may not function to trip the reactor. What ACTIONs are required by Technical Specifications? A. Shutdown banks may remain withdrawn, but entry into MODE 2 shall not be made until the trip switch is restored to OPERABLE. B. Restore the inoperable switch to OPERABLE within 48 hours or be in HOT STANDBY within the next 6 hours. C. Restore the inoperable switch to OPERABLE within 48 hours or open the Reactor Trip Breakers within the next hour. D. Within 1 hour take action to be in HOT STANDBY within the next 6 hours, and HOT SHUTDOWN within the following 6 hours. Proposed Answer: C Explanation (Optional): Per Table 3.3-1, Functional Unit 1, ACTION 11 applies ("C" correct). "B" is wrong, since be the ACTION ACTION 1 would be the correct if the plant was in MODE 1 or 2. "A" is plausible, since this would of LCO 3.0.3. if the trip breakers were open in MODE 3. "D" is plausible, since these are the ACTIONs Technical Reference(s): Tech Spec 3.03 and 3.04. (Attach if not previously provided) Tech Spec Table 3.3-1 Functional Unit 1, and ACTIONs 1 and 11. Proposed references to be provided to applicants during examination: Tec h Spec sections 3/4 MC-05499 Given a plant condition or equipment malfunction... (As available) Learning Objective: Evaluate Technical Specifications and determine required actions. Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.2 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
.L4 LIM1IINU CONDITIONS FOR OPERATION-AND"SURVEILLANCE REQUIREMENTS April 17, 2000 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met, except as provided in Specification 3.0.5. 3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals, except as provided in Specification 3.0.5. If the Limiting Condition for Operation is restored prior I to expiration of the specified time intervals, completion of the ACTION requirements is not required. 3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 hour action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:
- a. At least HOT STANDBY within the next 6 hours,
- b. At least HOT SHUTDOWN within the following 6 hours, and
- c. At least COLD SHUTDOWN within the subsequent 24 hours.
Where corrective measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications. This specification is not applicable in MODE 5 or 6. 3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made when the conditions for the Limiting Condition for Operation are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL MODE or specified condition may be made in accordance with ACTION requirements when conformance to them permit continued operation of the facility for an unlimited period of time. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual specifications. 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to Specifications 3.0.1 and 3.0.2 for the system returned to service under administrative controls to perform the testing required to demonstrate OPERABILITY. 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement. MILLSTONE - UNIT 3 3/4 0-1 Amendment No. 90, 97, 179
( TABLE 3,3-1 C-, REACTOR TRIP SYSTEM INSTRUMENTATION r-0 MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE 'I OF CHANNELS TO TRIP QPERABL MODES - FUNCTIONAL UNIT C 6-i
- 1. Manual Reactor Trip 2 I 2 1, 2 I 11
-I 2 1 2 3*, 4*, 5* W
- 2. Power Range, Neutron Flux 1, 2 2
- a. High Setpoint 4 2 3 4 2 3 1###, 2 2
- b. Low Setpoint 4 2 3 1, 2 2
- 3. Power Range, Neutron Flux High Positive Rate
- 4. Deleted I
2 1 2 1###, 2 3 W (JJ
- 5. Intermediate Range, Neutron Flux W 6. Source Range, Neutron Flux 2 2## 4 2 1 11 f4 a. Startup 2 1 2 3*, 4*, 5*
- b. Shutdown
- 7. Overtemperature AT 3 1, 2 6
- a. Four Loop Operation 4 2 6 3 2 2 1, 2
- b. Three Loop Operation
- 8. Overpower AT 3 1t 2 6
- a. Four Loop Operation 4 2 1,2 3 2 2
- b. Three Loop Operation %ar F
4 2 3 6 (1) -,. CL 9. Pressurizer Pressure--Low z 4 2 3 1, 2 6 (1) 4
- 10. Pressurizer Pressure--High 3 2 2 6
- 11. Pressurizer Water Level--High NA
OCT 21 1998 TABLE 3.3-1 (Continued) TABLE NOTATIONS
*When the Reactor Trip System breakers are in the closed position and the Control Rod Drive System is capable of rod withdrawal. **Above the P-7 (At Power) Setpoint.
- Above the P-9 (Reactor Trip/Turbine Trip Interlock) Setpoint.
##Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
Setpoint.
- EBiow the P-10 (Low Setpoint Power Range Neutron Flux Interlock) in (1) The applicable MODES and ACTION statements for these channels noted Table 3.3-3 are more restrictive and, therefore, applicable.
closed (2) Including any reactor trip bypass breakers that are racked in and for bypassing a reactor trip breaker. ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within the next 6 hours. ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 6 hours,
- b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours for surveillance testing of other channels per Specification 4.3.1.1, and
- c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER for four loop operation or 50% of RATED THERMAL POWER for three loop operation and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85 % of RATED THERMAL POWER for four loop operation or 60% of RATED THERMAL POWER for three loop operation within 4 hours; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours per Specification 4.2.4.2.
MILLSTONE - UNIT 3 3/4 3-5 Amendment No. 97, id, 9l, 164 0492
03/24/94 TABLE 3.3-I (Continued) ACTION STATEMENTS (Continuedl ACTION 9 - (Not used) I ACTION 10 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be In at least HOT STANDBY within 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE. ACTION 11 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or open the Reactor Trip System breakers within the next hour. ACTION 12 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 6 hours, and
- b. When the Minimum Channels OPERABLE requirement is met, the inoperable channel may be bypassed for up to 4 hours for surveillance testing of the Turbine Control Valves.
ACTION 13 - With one of the diverse trip features (undervoltage or shunt trip attachments) inoperable, restore it to OPERABLE status within 48 hours or declare the breaker inoperable and apply ACTION 10. The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status. ACTION 13A - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable Channel to OPERABLE status within 6 hours or be in at least HOT STANDBY within the next 6 hours; however, one channel may be bypassed for up to 4 hours for surveillance testing per Specification 4.3.1.1, provided the other channel is operable. VILLS E - UNIT 3 - 3/4 3-7 Amendment No. 70, 89, bosh
Form ES-401-6 ES-401 Written Examination Question Worksheet Question # 67 Examination Outline Cross-reference: Tier # 2 Group # 2 KIA # 016.A2.01 Importance Rating 3.1 Proposed Question: INITIAL CONDITIONS:
- Reactor Power is 80%.
- All control systems are in automatic.
- Tave is on program.
RCS Loop 1 narrow range Thot RTD (3RCS*TE41 IA) fails high. and one of the reasons why the What is one operator action required as a result of this failure, action is required? decreased. A. Manually increase Charging flow, since Charging flow has B. Trip the OTAT bistable, since the trip setpoint has increased. C. Reset the Steam dumps, since the steam dumps have ARMed. D. Cutout the Loop 1 AT input, since the RIL setpoint has increased. Proposed Answer: D raising Auctioneered Hi Tave. This raises PZR Explanation (Optional): Tave for the affected loop will increase, lowering the setpoint ("B" incorrect). In Tave level stpt, increasing charging ("A" incorrect). OTDT is penalized, ("C" is incorrect). RIL increases since Delta-T mode, steam dumps only arm on a plant trip, or load reject (PT506), increases due to the failure ("D" is correct). Tech Spec Table 2.2-1 (Attach if not previously provided) Technical Reference(s): Functional sheets 9, 10, and 11. AOP 3571, Att. A, step 1. to applicants during examination: None Proposed references to be provided system, (As available) Learning MC-05447 Given a failure, partial or complete, of the reactor coolant determine the effects on the system and on interrelated systems. Objective: Bank # 64331 Question Source: Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.5 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
( I TABLE 2.2-1l ( 01 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS e r-- (I,
-4 NOMINAL FUNCTIONAL UNIT TRIP SETPOINT ALLOWALE YALUE CD
- 1. Manual Reactor Trip N.A. N.A.
-4i 2. Power Range, Neutron Flux C,.,
- a. High Setpoint
- 1) Four Loops Operating 109% of RTP** 109.6% of RTP**
- 2) Three Loops Operating 80% of RTP** 80.6% of RTP**
- b. Low Setpoint 25% of RTP** 25.6% of RTP**
- 3. Power Range, Neutron Flux, 5% of RTP** with 5.6% of RTP** with High Positive Rate a time constant a time constant N > 2 seconds 2 seconds
- 4. Deleted
- 5. Intermediate Range, 25% of RTP** < 27.4% of RTP**
Neutron Flux ID 6. Source Range, Neutron Flux 1 X 10+ 5 cps < 1.06 x 10+5 cps
- 7. Overtemperature AT
- a. Four Loops Operating 1-4 1) Channels I, II See Note 1 See Note 2
- 2) Channels III, IV See Note 1 See Note 2
.f
**RTP - RATED THERMAL POWER ED-
(= I,. E 2.2-1 (Continued) ( REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS o :C 01 o .5-r,
-Il NOMINAL 0_ FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE m~
c b. Three Loops Operating
- 1) Channels I, II See Note 1 See Note 2 WA
- 2) Channels III, IV See Note 1 See Note 2
- 8. Overpower AT (Four Loops Operating) See Note 3 See Note 4
- 9. Pressurizer Pressure-Low 1900 psia 2 1897.6 psia
- 10. Pressurizer Pressure-High 2385 psia < 2387.4 psla
- 11. Pressurizer Water Level-High 89% of instrument < 89.3% of instrument span span
- 12. Reactor Coolant Flow-Low 90% of loop > 89.8% of loop design flow* design flow*
CD,
- 13. Steam Generator Water 18.1% of narrow > 17.8% of narrow Level Low-Low range instrument range instrument span span
- 14. General Warning Alarm N.A. N.A.
- 15. Low Shaft Speed - Reactor 92.4% of rated > 92.2% of rated Coolant Pumps speed speed
.11
*Minimum Measured Flow Per Loop = 1/4 of the RCS Flow Rate Limit as listed in Section 3.2.3.1.a (Four
'.0 Loops Operating); 1/3 of the RCS Flow Rate Limit as listed in Section 3.2.3.2.a (Three Loops Operating)
IrNULL. -+/- Gusmv t%1 u x.', ( REACTOR TRIP STST iSTRUMENTATION TRIP SETPOINTS f' NOMINAL oxz TRIP SETPOINT ALLOWABLE VALUE 01'. FUNCTIONAL UNIT I-C-,
-4 0
M 16. Turbine Trip Iv
- a. Low Fluid Oil Pressure 500 psig > 450 psig a
z- b. Turbine Stop Valve 1% open > 1% open WA Closure
- 17. Safety Injection Input N.A. N.A.
from ESF
- 18. Reactor Trip System Interlocks
- a. Intermediate Range 1 x 10.1 amp > 9.0 x 1011 amp Neutron Flux, P-6 rM b. Low Power Reactor Trips Block, P-7
- 1) P-10 input (Note 5) 11% of RTP** < 11.6% of RTP**
- 2) P-13 input 10% RTP** Turbine < 10.6% RTP** Turbine Impulse Pressure Impulse Pressure Equivalent Equivalent i c. Power Range Neutron M Flux, P-8 Ln 3 1) Four Loops Operating 37.5% of RTP** < 38.1% of RTP** co:
e 2) Three Loops Operating 37.5% of RTP** < 38.1% of RTiP** F IM **RTP = RATED THERMAL POWER le I'D
( ( ( TABLE 2.2-1 (Continued) I-QX REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS z r C NOMINAL 2 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE -4i WA
- d. Power Range Neutron 51% of RTP** < 51.6% of RTP**
Flux, P-9
- e. Power Range Neutron 9% of RTP** > 8.4% of RTP**
Flux, P-10 (Note 6)
- 19. Reactor Trip Breakers N.A. N.A.
- 20. Automatic Trip and Interlock N.A. N.A.
Logic
- 21. Three Loop Operation N.A. N.A Bypass Circuitry U,
a) co z i 04
**RTP = RATED THERMAL POWER
( Il a TABLE 2.2-l (Continued) TABLE NOTATIONS f NOTE 1: OVERTEMPERATURE AT
-4)
(AT . (1+lls) -K (K+2 4 s) (T - a') +K (p -p') f ( C=
'-4 Where: AT is measured Reactor Coolant System AT,
'F; ATo is loop specific indicated AT at RATED THERMAL POWER, 'F; (1.Ts) is the function generated by the lead-lag compensator on measured AT; Tr and r2 are' the time constants utilized in the lead-lag compensator for AT, T, a 8 sec, r2
- 3 sec;.
K, < 1.20 (Four Loops Operating);
- 1.20 (Three Loops Operating);
K2 0 O.02456/'F; (1+¶5 a) is the function generated by the lead-lag compensator for T.y;
/ r4 and rT are the time constants utilized in the lead-lag compensator for T.a 74 > 20 sec, r1 , 4 sec; T is measured Reactor Coolant System average a- temperature, OF; 3D T' is loop specific indicated Tavg at RATED THERMAL POWER, < 587.1F; K3 2 0.001311/psi P is measured pressurizer pressure, psia; P' is nominal pressurizer pressure,
> 2250 psia; s is the Laplace transform operator, sect; it
( ( ( 0 TABLE 2.2-1 (Continued) 0
< TABLE NOTATIONS (Continued) ml TABLE 2.2-1 (Continued)
TABLE NOTATIONS (Continued) NOTE 1: (Continued) and fl(AI) is a function of the indicated difference between top and bottom detectors of the power range neutron ion chambers; with nominal gains to be selected based on measured instrument response during plant startup tests calibrations such that: (1) For q, - qb between -26% and +3%, fl,(AI) > 0, where q, and qb are percent RATED THERMAL POWER in the upper and lower halves of the core, respectively and qt + qb is the total THERMAL POWER in percent RATED THERMAL POWER; (2) For each percent that the magnitude of q, - qb exceeds -26%, the AT Trip Setpoint shall be automatically reduced by > 3.55% of its value at RATED THERMAL POWER. 0 (3) For each percent that the magnitude of q, - qb exceeds +3%, the AT Trip Setpoint shall be automatically reduced by > 1.98% of its value at RATED THERMAL POWER. NOTE 2: The maximum channel as left trip setpoint shall not exceed its computed trip setpoint by more than the following: (1) 0.4% AT span for the AT channel (2) 0.4% AT span for the Tav channel (3) 0.4% AT span for the pressurizer pressure channel 3 . (4) 0.8% AT span for the f(AI) channel D+ vi
i ( '.( a( TABLE 2.2-i (Continued) TABLE NOTATIONS (Continued) 0n s NOTE 3: OVERPOWER AT z i T) (1+(tS) s))45 T 77S (T-
&To (1+-r s) (31+r 7 s)
Where: AT is measured Reactor Coolant System AT, *F; ATo is loop specific indicated AT at RATED THERMAL POWER, *F; (1+¶s8) (1+T 2s) is the function generated by the lead-lag compensator on measured AT;
\ .
rT and r2 are the time constants utilized in the lead-lag compensator for AT, TI Z 8 sec, r2 < 3 sec; K4 < 1.09; K5 k 0.02/F for increasing Tavg and K6 < 0 for decreasing T..; (r7 s) (1+r7 S) is the function generated by the rate-lag compensator for Tavg; 17 is the time constant utilized in the rate-lag compensator for Tavg, r7 > 10 sec; T is measured average Reactor Coolant System temperature,. F; C+ T' is loop specific indicated Tavg at RATED THERMAL POWER, < 587.1*F; Ks > 0.00180/'F when T > T' and K. < O/'F when T < T"; s is the Laplace transform operator, .sec1. vi
( ( o zTABLE 2.2-1 (Continued) TABLE NOTATIONS (Continued) z0 c z NOTE 4: The maximum channel as left trip setpoint shall not exceed its computed trip setpoint by more than 0.4% AT span for the AT channel and 0.4% AT span for the Tavg channel. NOTE 5: Setpoint is for increasing power. NOTE 6: Setpoint is for decreasing power. r.. ED In IN r9 k 0
-c
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_N 2 . _M SP
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PRESSURIZER LEVEL C4ANNMELS AUCTIONEERCD TAVG. (SuEET 9) Im I ADJUSTABLE I I I CHARCI&I INOLOAD r A.W I.P.
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CoV__tE Ž LETD0WN LINE -ISOLATIoN LETDOWN LINE ISOLATION STACO I 1 1'5 'e O LL VALVE CONTROL VALVE CONTROL UO I(COWTRL M^ULl AT -l.AIO ROO#) E SELECTOR SWITCH C AFUA SELECTOR SWITCH
- SEI LOSE _ BO (NTE AUTJ l LVSL e1 I.ANUAL A AUTO - MANUAL A Al I I~I.
ITNTOCtTRLVALRO CONTROL VAL4O OT FROL S ATONT A I ELE TORSW 'lC SE R WIT R o-MAUL AT- Z AUL AUTOTLOSE
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OPIEN SIGNALSIGNAL OPEN 12179-2472.01 1-001-011 [~1Ea ss.s TSAR FIGURE"
It 133ws I
-' - PRESSURIZER PRESSUJRX CHANN.ELS 33r z (NOTE G) Io'~ ADJUSTABLE
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- 3. PRSSR BISTA81.1 NO. PB-4554 ANIDLEVEL DNSTA.BLES VARIABLE SPRAY VALVE NO. - ALL L"-59C, L&-459E. 111.-460 D ARE ENERUZ.E TO ACUAT E HEATER -I
- 4. OEKSNUT4 INDICATIONIN C54fTRS. R0C4ANDON 0 TH AUXILIARY SH~JT~owW Mlift. (ASP) 5ACK-UP-CONTROL- PCV-455B.
3A. LlswJI %N4OvA EI PR~n`C IQ WrAV ac4rq'll HEATERS SIGNAL Rcom -R. W.D-e.H (ROTr S)
-wa~iV 1C NOCI. 6.1,4 rT I 0V PJILLT cGaLq~.. .SHEET Is) (SHEEr IE)
C-.T"HESE. COS'JTIFSL5 0, OTH C064TV,,01 SoAato AqtS sulPwLAE.O ay'eTYKPS.5
- 7. LOCAL CONJTROL AT TmF AILUARY SHUTDOiAJ PAJ1EL CWERRNOE)E VRWAA T49 02lMOL. THIM SlGb4A.L-R~A~&NDACrUA6TESkkij ALA.RAM AmIjw'jIA.Toq SW yrk COUTROL ROOM#.
- 8. A REmOr1E/LOCAL CONTROL. T2,NSrFgZ Swr-rCe4 in LOCATED ON A LOCAL TRANFER SWrTCi4 PANEL (TSP).
It I 0%in
- - - - - - -- - -- - - - KISrAOrS A CADTYSOG7 S."I III I . I NT:
; WAA AM IMEN ASEWILS ICAS TIlS DlOMWEI MGIIITEO J 4 JjI I
IV I 8 9 I
0u RCS Narrow Range Temperature Channel Failure The following annunciators are symptoms of an RTD failure: TAVE/AUCT TAVE DEVIATION MB4C 5-5 TREF/AUCT TAVE DEVIATION MB4C 6-5 AT/AUCT AT DEVIATION MB4C 4-5 OVERPOWER AT MB4C3-6 TAVE HI MB4C 5-6 OVERTEMP AT MB4C 4-6 LOOP 1,2,3,4 OVR TEMP AT MB4F 1,2,3,4 - 5 LOOP 1,2,3,4 OVR PWR AT MB4F 1,2,3,4 - 6 LOOP 1,2,3,4 OVR TEMP AT MB4F 1,2,3,4 - 7 LOOP 1,2,3,4 OVR PWR AT MB4F 1,2,3,4 - 8 LOOP 1,2,3,4 TAVE LO MB2D 1,2,3,4 - 7 LOOP 1,2,3,4 TAVE LO-LO MB2D 1,2,3,4 - 8
- 1. Defeat the failed channel input.
Loop Temp Cutout - AT 3RCS-TS411F Loop Temp Cutout - TAVG 3RCS-TS412T OTIOPAT Record Select 3RCS-TS411E
- 2. Check the following annunciators NOT LIT:
TREF/AUCT TAVE DEVIATION MB4C 6-5 TAVE HI MB4C 5-6
- 3. Restore TAVE - TREF error to within 1°F and return rod control to automatic.
- 4. Monitor PZR level until stable. If PZR level controller is in manual, Restore pressurizer level to program level and Place PZR level controller in automatic. I
- 5. When conditions have stabilized, Observe MB annunciators and parameters.
Immediately report any unexpected or unexplained conditions to the Shift Manager.
Written Examination Question Worksheet Form ES-401-6 ES-401 Question # 68 Examination Outline Cross-reference: Tier # 2 Group# 2 K/A # 027.A4.01 Importance Rating 3.3 Proposed Question: exist: The crew is responding to a small break LOCA, and the following conditions
- CDA actuated 14 minutes ago.
- CTMT pressure is 16.5 psia and decreasing.
RE05A) are LIT.
- Alarm lights for both CTMT high range rad monitors (3RMS*RE04A and
- The crew is at E-1, step 7 "Check if CTMT Spray Should Be Stopped".
to reduce
- The ADTS has determined that the crew will operate only the "A" QSS pump CTMT radiation levels.
- RWST level is 800,000 gallons.
required by the RO in Based on current plant conditions, what are the minimum actions physically containment? order to align the QSS pumps so that only the "A" QSS pump will be spraying and open the "A" A. Reset CDA, take the "A" QSS pump to "START" and back to "AUTO", QSS pump discharge spray valve. and close the "B" QSS B. Reset CDA, take the "B" QSS pump to "STOP" and back to "AUTO", pump discharge spray valve.
"AUTO", and open the C. Reset SIS and CDA, take the "A" QSS pump to "START" and back to "A" QSS pump discharge spray valve.
and close the D. Reset SIS and CDA, take the "B" QSS pump to "STOP" and back to "AUTO",
"B" QSS pump discharge spray valve.
Proposed Answer: B ("A" and Explanation (Optional): The QSS pumps are already running from the CDA signal with adequate RWST level to allow stopping of the pumps ("B" correct, "D" wrong).
"C" wrong). Only the CDA signal needs to be reset in order E- 1, steps 7 and 22. (Attach if not previously provided)
Technical Reference(s): LSK 24-9.4F and 27-12F None Proposed references to be provided to applicants during examination: system (As available) Learning MC-05 171 Describe operation of the following containment depressurization Objective: components controls and interlocks: Quench Spray System (QSS)... Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
LOSS OF REACTOR OR EOP 35 E-1 Page 8 of 19 SECONDARY COOLANT Rev. 018 XPECTED RESPONS RESPONSE NOT OBTAINED I CAUTION To ensure adequate ECCS flow, DO NOT stop any recirculation spray pumps used for core injection flow. NOTE The recirculation spray pumps are sequenced to automatically start 11 minutes after a CDA.
- 7. Check If Ctmt Spray Should Be Stopped
- a. Verify quench spray a. Proceed to CAUTION prior to pumps - RUNNING step 8.
___b. Verify Ctmt pressure - b. Proceed to CAUTION prior to LESS THAN 17.5 psia step 8. and, WHIEN Ctmt pressure is LESS THAN 17.5 psia, THEN Perform step 7.c.
- c. Using GA-8, Stop Ctmt spray II
LOSS OF REACTOR OR EOP 35 E-1 Page 19 of 19 SECONDARY COOLANT Rev. 018 ACTION/EXPECTED RESPONS STP RESPONSE NOT OBTAINEDI
- 20. (continued)
- c. Check the following cold leg c. IF one RHR pump cold leg injection valves - CLOSED injection valve is open as a result of only one charging or
- RHR pump A SI pump running, (3SIL*MV8809A) T1HEN Request the ADTS
- RHR pump B determine if the injection (3SIL*MV8809B) valve should remain open for hot leg recirculation.
- d. Check the following cold leg d. OPEN valve(s). I injection valves - OPEN
- SI pump A (3SIH*MV8821A)
- SI pump B (3SIH*MV8821B)
- SI injection (3SIH*MV8835)
- 21. At 9 hrs After Event Initiation, Go To ES-1.4, Transfer To Hot Leg Recirculation
- 22. Evaluate Long Term Plant Status
- a. Consult ADTS to determine if the Ctmt spray pumps should be started to reduce Ctmt radiation levels
-FINAL-
SOURCE 'MIONTIOR CONOITION , NTROL ACTION wsUITANY ; MONTIOR 4F MAUA START I NOTE:
<_ (e 6) l S*P3A
\~~STAQT / 1. THE SEQUENCED SAFEGUARDS SIGNAL IS COMPOSED OF THE
/ i \ FOLLOWING COMBINATION OFSAFEGUARDS SIGNALS f QUENCH SPRAY PUM ANDW r SIGNALS I TIME DELAY I
- I
\ ~~~,%rDISCHARGFLW =ID ICDA RECIRCMOOETHENLOPI 2
- 2. INDICATING UGHTS ARE GNAT THE SWITCHGEAR AND AT
\! ti2 Q5*PRA qEcO_> MAIN CONTROLBOARD. INDICATING LIGHTS AREONLY ONAT THE SWITCHGEAR WHEN CONTROL ISAT THESWITCH6EAR AN FLWLO,\SAT/_
D 3 SOURCE DEVICE COMMON TOALL LOADS ON BUS 34-C
- 4. THIS ISA KEY LOCK TRANSFER SWITCH WITH KEY REMOVAL IN THE REMOTE & LOCALPOSITMOWS
() LS"t4-9.4F SE=Q > (2a723 SKIS) J/WS\DAN rE36 s~~eARM6TEMPE8TURE
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( 5T3 BREAERS TnIC R )IU- D \uJ A JE[NS SPOITION .,_ N-II I -- I rII I [O THISIDOCUMENT I I I lI I I (i) SK-2-7A = 4UNCH PRAY PUMP A PRM 8 TED. _ I lor I THE AMILSTNE POINTCOPANY l W NORTHEAST UTILITIES
-4CAGPE3M
- I3QSS*P3A 3QSS *P7A 3QSS *P3A II TE MILLSTONENULEAR POWER STATION-UNITNOS LOGIC DIAGRAM AUTO AUTO l QUENCH SPRAY I 11 ::
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.1 I S&W DWG. NO. 12179-LSK-27-12F
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- -. MANUAL START (28723 S7H.2) BLOCKSIGNAL 30i554 P3A K> BUS34C LOSSOF POWER SIGNAL F j8 LSK-27-12 (28463 SI
* ' LOP SIGNALS SEUENCER SPRAYPUMP3055fP3A FOR QUENCN
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_5CA SIGNAL NOTES:I. LOGICFORSEQOUENCER A (TRAINA) SHOWN. NL(CLEAfl SAFETY PELATED LOGICFORSEQUENCER B (TRAIN B) SIMILAR. Q.A. CAT. I. III NIU A 1Oh.S-0-t TU.... I l.IP ArY I RlAI I RI.1N I 1_5 NC E&DCR'S ,ti 5 _ NORTHEASTNUCLEARENERGY!-iPANY _I NO EUES;AEU 8 / TC-02147&TE- 3725 _ NURTHEAST UTILITIES
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ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 69 Tier # 2 Group# 2 K/A # 028.GEN.2. 1.32 Importance Rating 3.8 Proposed Question: A LOCA has occurred, and the crew has started up the hydrogen monitoring system per FR-C. 1 "Response to Inadequate Core Cooling". Assuming a Hydrogen Recombiner is started up, why does OP 3313A "Hydrogen Recombiners" PRECAUTION 3.1 recommend a flow greater than 7.5" of H2 0 (3HCS*FI IA or 1B); and at which of the below CTMT hydrogen concentrations will the crew be directed by FR-C. 1 to start a hydrogen recombiner? A. Flow > 7.5 inches H20 ensures that the hydrogen in the gas stream does not ignite. The recombiner will be started up with 0.1% CTMT hydrogen concentration. B. Flow > 7.5 inches H20 ensures that the hydrogen in the gas stream does not ignite. The recombiner will be started up with 0.5% CTMT hydrogen concentration. C. Flow > 7.5 inches H2 0 ensures adequate equipment protection and performance. The recombiner will be started up with 3.0% CTMT hydrogen concentration. D. Flow > 7.5 inches H20 ensures adequate equipment protection and performance. The recombiner will be started up with 6.0% CTMT hydrogen concentration. Proposed Answer: C Explanation (Optional): wrong) and 5%
""C is correct since the recombiner is only operated if hydrogen concentration is between 0.2% ("A" no hydrogen burn is possible, or
("D" wrong). This range ensures Hydrogen concentration is in a range where either Flow greater than 7.5 inches H1O ensures the burn will be limited to prevent a significant rise in CTMT pressure. adequate recombiner equipment protection and performance ("C" correct, "A" and "B" wrong). Technical Reference(s): OP 3313A, PRECAUTION 3.1 (Attach if not previously provided) EOP 35 FR-C.1, step 6 WOG Bkgd Doc, FR-C.1, step 8 Proposed references to be provided to applicants during examination: None MC-04741 Describe the major administrative or procedural precautions and (As available) Learning Objective: limitations placed on the operation of the HCS system, and the basis for each. Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement I
p3-4Z L I 2.1.7 IF starting a Hydrogen Monitoring System (Section 4.7 or 4.8), the applicable Train valves are aligned as specified in the following:
- OPS Form 3313A-6, "Post Accident Sampling (Hydrogen Monitors)"
- OPS Form 3311C-1, "Post Accident Sampling System-Liquid/Air" 2.2 Documents 2.2.1 SP 3613A.1, "Hydrogen Recombiner System Low Power Surveillance" 2.2.2 SP3447C1 1-1, "Train A Containment Hydrogen Monitor Op Test and Channel Check" 2.2.3 SP3447C12- 1, "Train B Containment Hydrogen Monitor Op Test and Channel Check" 2.3 Measuring and Test Equipment 2.3.1 Type K (chromel/alumel) thermocouple reader.
- 3. PRECAUTIONS 3.1 A flow greater than 7.5" of H 2 0 (3HCS*FI 1A or 3HCS*FI 1B) ensures © adequate equipment protection and performance.
3.2 Hydrogen Recombiner operation has resulted in spurious bus 32T (U) ground detection alarms, and increased activity indications on 3HVZ*RE09A(B) and 3CMS*RE22A. CMS-22 (particulate) indication may exceed the alarm setpoint during recombiner operation. A. OP 3313A Level of Use Rev. 013-07 Continuous 5 of 47 I
RESPONSE TO INADEQUATE EOP 35 FR-C.1 Page 8 of 18 CORE COOLING Rev. 013 r--.-.-- ACTION/EXPECTED RESPONS STP RESPONSE NOT OBTAINED l NOTE This procedure should be continued while obtaining hydrogen samples in the next step.
- 6. Check Containment Hydrogen Concentration
- a. Using GA-22, Startup the hydrogen monitors
__b. Verify the hydrogen b. Perform the following: concentration - LESS THAN 5% 1) Consult the ADTS to determine if Ctmt purge system should be placed in service to reduce hydrogen concentration using GA-24.
- 2) Proceed to step 7.
- c. Verify the hydrogen c. Using GA-23, Startup the concentration - hydrogen recombiners.
LESS THAN 0.5%
- 7. Check AFW Suction Source
- a. Check DWST level - a. Proceed to CAUTION prior to LESS THAN 80,000 gal step 8. and, IF DWST level decreases to LESS THAN 80,000 gal, THEN Perform step 7.b.
- b. Using GA-4, Shift AFW pump suction to the CST and Fill the DWST
STEP DESCRIPTION TABLE FOR FR-C.1 Step 8 STEP: Check Containment Hydrogen Concentration PURPOSE: To check if an excessive containment hydrogen concentration is present BASIS: This step instructs the operator to obtain a current hydrogen concentration measurement. Depending upon the magnitude of the hydrogen concentration, the operator will either continue with guideline FR-C.1, turn on the hydrogen recombiners or notify the plant engineering staff to determine additional recovery actions before continuing with the guideline. When inadequate core cooling has occurred, the containment hydrogen concentration may be as much as 10 to 12 volume percent, depending on the amount of metal-water reactor (to produce hydrogen) that has occurred in the core. The hydrogen concentration is of concern since a flammable mixture can burn, if an ignition source is available, and cause a sudden rise in containment pressure which may challenge containment integrity. The operator is instructed to obtain a current containment hydrogen concentration measurement at this point in order to ascertain the potential flammability of the combustible gases in the containment. Note that in order to have the potential for flammable hydrogen concentrations, an inadequate core cooling situation must have already existed. Without an inadequate core cooling situation, sufficient hydrogen would not be expected to have been produced to cause potentially flammable mixtures. A determination is made of the flammability of the hydrogen mixture with respect to the possible containment pressure rise. If the containment mixture is between 0.5 volume percent and (T.05) volume percent in dry air, either no hydrogen burn is possible or a limited burn may occur which does not produce a significant pressure rise. In this case the operator is instructed to start the hydrogen recombiner system to slowly reduce containment hydrogen concentration. If the concentration is greater than (T.05) volume percent in dry air, the operator is instructed to immediately notify the plant engineering staff of the situation. All hydrogen measurements are referenced to concentrations in dry air even though the actual containment environment may contain significant steam concentrations. The reason for this is twofold: 1) most hydrogen measurement systems remove moisture from the sample thus approximating a dry air condition and 2) the indication of the potential of hydrogen flammability is conservative when based upon using hydrogen concentration in dry air. FR-C.1 24 HP-Rev. IC HFRC1
Written Examination Question Worksheet Form ES-401-6 ES-401 Question # 70 Examination Outline Cross-reference: Tier # 2 Group # 2 K/A # 034.K4.01 Importance Rating 3.4 Proposed Question: fuel Fuel movement is in progress in the fuel pool, and the fuel handler is preparing to lift a spent rod that is mechanically bound. from damage due to the Which of the following spent fuel bridge interlocks will protect the fuel binding? A. Slack Cable Limit. B. Fuel Underload Limit. C. Fuel Overload Limit. D. Hoist Full Up Limit. Proposed Answer: C Explanation (Optional): of a fuel rod occurs (2000 Ibs), stopping the "C" is correct, since Fuel Overload interlock will be exceeded if binding downward movement when the weight of the lifting evolution. "A" is wrong, since this interlock stops the crane's is wrong, since the Fuel Underload Limit stops fuel rod is completely supported by the storage rack (200 lbs). "B" wrong, since Hoist Full Up Limit stops upward movement as the weight of the load decreases (1500 lbs). "D" is depth of water shielding depth is maintained. motion at a maximum lift height so that the minimum required OP3303A, section 4.2.23 - 4.2.31 (Attach if not previously provided) Technical Reference(s): OP3303A, Att.l None Proposed references to be provided to applicants during examination: (As available) Learning MC-04541 Describe the operation of the following Fuel Handling System Objective: equipment, controls, and interlocks ... Spent fuel bridge crane and hoist... Modified Bank # 69811 Parent attached Question Source: Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
Written Examination Question Worksheet Form ES-401-6 ES-401 Original 69811 to be Which of the following spent fuel bridge interlocks requires the Bypass Enable key switch positioned to "ON" in order to override the interlock? A. Slack Cable. B. Fuel Underload. C. Hoist Overload. D. Traverse Travel Limit. Answer: C 41 of 46 NUREG-1021, Revision 8, Supplement 1
4.2.20 PRESS "TRAVERSE TRAVEL AVERT" and VERIFY blue light, not lit. 4.2.21 VERIFY a Fuel Building exhaust filter is in operation if necessary, and all T/S requirements are met, prior to continuing. NOTE If loads are not be handled by the hoist, step 4.2.22 is not required to be performed. 4.2.22 VERIFY overload interlock by performing SP 3603A.1, "Testing of the Spent Fuel Bridge Fuel Overload." NOTE
- 1. If Spent Fuel Bridge is not to be used to move fuel, steps 4.2.23 through 4.2.32 may be eliminated, and associated steps on OP 3303A-001 marked "N/A."
- 2. Dummy fuel assembly contains a dummy RCCA.
- 3. Dummy fuel assembly is normally stored in spent fuel pool location U-01.
4.2.23 Using the spent fuel handling tool, PERFORM the following:
- a. LATCH to dummy fuel assembly.
- b. PICK up dummy fuel assembly from fuel pool location U-01.
4.2.24 CHECK dummy fuel assembly engaged by observing load. 4.2.25 RAISE hoist to "FULL UP" as indicated by amber "HOIST HEIGHT LIMIT" light, lit. 4.2.26 CHECK indicated weight is between 1,700 and 1,900 lbs. 4.2.27 POSITION dummy fuel assembly over open water. 4.2.28 LOWER hoist until amber "LIMIT" light starts blinking.
- a. VERIFY crane mode switches to "JOG."
- b. ATTEMPT to move bridge and trolley and VERIFY bridge and trolley move in the "JOG" mode only.
4.2.29 RAISE dummy fuel assembly to "FULL UP" and POSITION it over its storage location, (U-01). OP 3303A Level of Use General or' ' Rev. 009-05 11 of 29
4.2.30 LOWER dummy fuel assembly until it just enters its storage location. 4.2.31 With the bottom of dummy fuel assembly already inserted into its storage location (U-01), LOWER dummy fuel assembly in "HIGH" speed.
- a. WHEN dummy fuel assembly is approximately 12 inches from "FULL DOWN," VERIFY hoist automatically shifts into "LOW" speed (3 fpm).
- b. WHEN fuel underload limit is reached, as indicated by the white "FUEL UNDERLOAD" light lit, VERIFY hoist stops.
- c. PRESS "FUEL UNDERLOAD AVERT," and VERIFY blue light, lit.
- d. REDUCE load until slack cable limit is reached, as indicated by white "SLACK CABLE LIMIT" light lit, and VERIFY hoist stops.
- e. UNLATCH spent fuel handling tool from dummy fuel assembly.
- f. RAISE hoist and VERIFY spent .fuel handling tool is unlatched from dummy.
4.2.32 CHECK for proper operation of slack cable limit.
- a. POSITION tool over storage bracket.
- b. REDUCE load until slack cable limit is reached, as indicated by white "SLACK CABLE LIMIT" light lit, and VERIFY hoist stops.
- c. PRESS "SLACK CABLE AVERT," and VERIFY blue light, lit.
- d. VERIFY hoist can be lowered.
- e. IF desired, UNLATCH tool from hoist.
- End of Section 4.2 -
Level of Use OP 3303A General Rev. 009-05 r 11 12 of 29
Attachment 1 Description of Interlocks (Sheet I of 1) The following interlocks are designed into the Spent Fuel Bridge Control System. An AVERT is used to override interlocks routinely encountered while handling loads. A BYPASS is used to override interlocks designed to prevent damage to fuel or equipment, and which should not be routinely encountered while handling loads. Geared Limit Switch Hoist Full Down - Prevents further lowering of the hoist (may be BYPASSED). Geared Limit Switch Hoist Near Full Down - Automatically shifts the hoist into the LOW speed range, if a fuel assembly is on the hook, when the bottom of the fuel assembly is approximately 12 inches from the bottom of a storage cell or the transfer cart (may be BYPASSED). Geared Limit Switch Hoist Full Up - Prevents further raising of the hoist (may be BYPASSED). Allows upender operation when hoist is above upender. Power Interrupt Switch (PIS) Full Up - Prevents all hoist motion (may be BYPASSED with keyswitch in Power Distribution Enclosure). Fuel Overload (Set at 2,000 pounds) (Note 1) - Prevents further raising of the hoist (may be BYPASSED). Hoist Overload LO Setpoint (Set at 2,400 pounds) - Prevents further raising of the hoist. This serves as a backup to the Fuel Overload (may be BYPASSED). Hoist Overload HI Setpoint (Set at 6,000 pounds) - Prevents further raising of the hoist. This prevents a load in excess of the hoist capacity from being lifted. Fuel on Hook (Set at 1,000 pounds) (Note 1) - This interlock (1) limits the bridge and trolley to the JOG mode unless the hoist is full up, and, (2) inputs to the hoist near full down limit to automatically shift the crane into the LOW speed range (may be BYPASSED). The presence of this interlock is indicated by a flashing amber Hoist Height Limit light when the hoist is not full up. Slack Cable (Set at 200 pounds) (Note 1) - Prevents further lowering of the hoist. (Slack Cable may be AVERTED. It will automatically reset when load is picked up again.) Fuel Underload (Set at 1,500 pounds) (Note 1) - Prevents further lowering of the hoist. (Fuel Underload may be AVERTED. It will automatically reset when weight is increased above its setpoint. Traverse Travel Limits - Prevents loads suspended from the hook from being driven into the pool walls. (Travel limits may be AVERTED. The AVERT must be manually reset when the travel limit has been cleared.) Note 1: Values given are nominal values and are subject to change by Reactor Engineering. Level of Use OP 3303A GeneralRev. General28 009-05 of 29
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 71 Tier # 2 Group # 2 K/A # 035.K3.02 Importance Rating 4.3 Proposed Question: INITIAL CONDITIONS:
- The crew is performing a plant cooldown in accordance with OP 3208 "Plant Cooldown".
- Tave is 540'F and slowly decreasing.
- PZR pressure is 2110 psia and slowly decreasing.
- S/G pressures are all approximately 950 psig and slowly decreasing.
A small steam break occurs, and the "SG B STEAMLINE PRESSURE LO" annunciator is received on MB5B prior to any further operator action. The "SG B PRESSURE RATE HI" annunciator is NOT received. What automatic function(s), if any, occur as a result of this alarm condition? A. No automatic actions occur. B. Safety Injection Actuation actuates only. C. Main Steam Line Isolation actuates only. D. Main Steam Line Isolation and Safety Injection Actuation occur. Proposed Answer: D Explanation (Optional): "D" is correct, since Low Steam Pressure SI and MSI are not procedurally blocked until less than 2000 psia [Can not be blocked unless < P- 11 (2000 psia)]. "A" is plausible, since the crew will block Low Steam Pressure SI and MSI <2000 psia, and "STEAM PRESSURE RATE HI" did not come in. "B" is plausible since SIS
< P-lI.
actuates. "C" is wrong, but plausible, since, per MB5B, 2-4, high steam rate will actuate MSI with SIS blocked Technical Reference(s): OP 3353.MB5B, 2-4 and 3-4 (Attach if not previously provided) Functional Dwgs #7 and 8 OP 3208, step 4.2.5 Proposed references to be provided to applicants during examination: None MC-05493 Describe the operation of the following RPS controls and interlocks... ESF (As Learning Actuation Signals... available) Objective: Modified Bank # 75626 (Note changes or attach pa rent) Question Source: Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 and 41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
ES-40 1 Written Examination Question Worksheet Form ES-40 1-6 Original question 75626 INITIAL CONDITIONS:
- The crew is performing a plant cooldown in accordance with OP 3208 "Plant Cooldown".
- Tave is 540'F and slowly decreasing.
- PZR pressure is 1900 psia and slowly decreasing.
- S/G pressures are all approximately 950 psig and slowly decreasing.
While attempting to increase the RCS cooldown rate, the BOP operator inadvertently throttles too far open on the Condenser Steam Dumps. The "SG B PRESSURE RATE HI" setpoint is exceeded, and the annunciator is received on MB5B. What automatic function(s), if any, occur as a result of this alarm condition? A. No automatic actions occur. B. Safety Injection Actuation actuates only. C. Main Steam Line Isolation actuates only. D. Main Steam Line Isolation and Safety Injection Actuation occur. Answer: C 41 of 46 NUREG-1021, Revision 8, Supplement 1
Setpoint: Variable 2-4 SG B PRESSURE RATE Hi AUTOMATIC FUNCTIONS steam low
- 1. Steam line isolation signal when less than P-11 (2 of 3) setpoint and main pressure safety injection signal blocked.
CORRECTIVE ACTIONS
- 1. IF reactor trips, Go To E-0, "Reactor Trip or Safety Injection."
- 2. CHECK for steam line rupture by observing 3MSS-PI 524A and 3MSS-PI 526A, steam generator pressures (MB5).
- 3. IF alarm is due to a failed instrument, Go To AOP 3571, "Instrument Failure."
SUPPORTING INFORMATION
- 1. Initiating Devices 1.1 3MSS*PT524 1.1 Variable 1.2 3MSS*PT525 1.2 Variable 1.3 3MSS*PT526 1.3 Variable
- 2. Procedures 2.1 E-0, "Reactor Trip or Safety Injection" 2.2 AOP 3571, "Instrument Failure" OP 3353.MB5B Level of Use 001 -02 I-- ConinuousRev. 18 of 46 lContinuous
Setpoint: Less than 658.6 psig 3 4 l SG B STEAMLINE PRESSURE LO I_ AUTOMATIC FUNCTIONS
- 1. If Pressurizer Pressure is less than P-11 AND SIS is not blocked, a safety injection and main steam line isolation will occur (2 of 3).
- 2. If Pressurizer Pressure is greater than P-II, a main steam line isolation and safety injection on low steam line pressure will occur (2 of 3).
CORRECTIVE ACTIONS I. IF reactor trips, Go To E-0, "Reactor Trip or Safety Injection."
- 2. CHECK for steam line rupture by observing 3MSS-PI 524A and 3MSS-PI 526A, steam generator pressures (MB5).
NOTE The computer can confirm only one pressure instrument as faulty.
- 3. IF alarm is due to failed instrument, Go To AOP 3571, "Instrument Failure."
SUPPORTING INFORMATION
- 1. Initiating Devices 1.1 3MSS*PT524 1.1 <658.6 psig 1.2 3MSS*PT525 1.2 <658.6 psig 1.3 3MSS*PT526 1.3 <658.6 psig
- 2. Procedures 2.1 E-0, "Reactor Trip or Safety Injection" 2.2 AOP 3571, "Instrument Failure" Level of Use OP 3353.MB5B Continuous Rev. 001 -02 26 of 46
II IP W. PUmps P-Il Ip LOGIC (SIEET 6)
- ET S) NOT REDUNDANT NOTES:
-10"'vi@ 1 . THE IRtEDOUANT ANUALBLOCKCONTROL CCHSISTS OF Two CONTROLS ON THE CONTROLBOARD ONE FOR EACH TRAIl I SUPPLIED BY OTHERS.
2'9 12s Z. TWO COMPUTER INPUTS ARE CONNECTED TO THIS CIRCUIT INFIVIDUAL FOR EACH TRAIN.
- 3. TWO PERMISSIVE STATUS LIGHTS ARE CONNECTED TO THIS CIRCUIT. INDIVIDUAL FOR EACH TRAIN.
- 4. POSITION DETEC7ION IS ACCOMPLISHED BY TWO POSITION SWITC.ES (INDIVIDUAL FOR EA,- TRAIN)PER STOP VALVE.
(NIMTE3) 5 LOOP BLOCI-5 FOR TMIS FUrCTOv4 TO ALLOW (NrOTE21 3 LOOP OPERAY1014 PROVIDED E5Y SEYLOCMv SWITCFIES 0l 5SPS CA,6I,4FT. I VEYLOCM PER LOOP PE.R TRAiE. 5,T EA Mtl R A~ 11-HI VIR4 I.E.IV UL- (K Qor- 5)
"FSAR FIGURE" OURING REVISIONS CONSTRUCIIN I P.A IQ A FEEOVWATER I50LSATIObM (ESHEST '3> f3RPS *5Y$ ]
12179 -2472.CII-001 -007 INOIE: I OANUAL CADro THIS 0SCLEKNT REVISIOHS EfN AS-UILd AREPRPNIBITED. 1 5 4 3
LOOP ISLATION VALVES I so SI .VES FULL CLrS A. s 'LOOP I LOOp 2 LOOP 3 LOOP 4 ' COLD LEG COLD LEG COLD LEG COLD LEG (NOTE 4) (NOTE 4) SECONDS
&P_- C Pf C- 0_PE STEAM GENERATOR LOW-LOW WATER LEVI (NOTE LOW STEANLWNE PRESSURE (LEAD-LAG CO.FENSATED)
(N4OTE 5) r 5TEAM GENE tar I LO 2 LOOP3 LOP 4 IT SAFETY ISOLATION I NOTE!. P
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I II aA4TROL ROOM NTAKE MANUAL ACTUATION FROM CONTROL BOARD 1 1 A Ut'm- I TEST
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CoNTROL BUILDI iK K0 ISoLATI0o10 COwOiS) I I o-4iA(TWO CNTRIOLS) (N*OTEi I) (TWOCONTROLS) (TWltC4TTOLS) . ONTIS I(N° IbS i I (7 17 7""'Y t :.I , ' L_____ ___ (NOTE 9) T T S T A 3> 3 3 t
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MANUAL MANUAL MANUAL RESET RESET RESET (NOTES ) (NOTES890) NOT1E5&9) f LOP Co A, LOP A LA EMIERGSWC11I' EMGEh[CN, by By' SENERA.:TOR EsERATEOR N LOAO 5EQI.)E R l l lLOAD 5EEQEUECl (NOTE I6) (NOTE15) OTWEwb 2179-LSK-24q9.* l7-SE-4 IuSCO,25212-287a3 O 212-?87_T 0 (NOTE 13) (NOTE 13) (NOTF15) EICRGEN.CY SAFET SERVICE WATER C TANEJ TAMET CONTAINMCNT CONTAINMENT COW~TRO. IINTAKE OUCTI (PUP START& A SPRAY ISOLATION ISOLATION ROP ISOLATION DIESEL INETIq PHSE B START-UP (PR BY (A)) SYSTEMISO.*) RCR.AP1 ACTUATION PHASEA RPiZA'TINI
- (NTE )0 (NOTE 7) R tk AN (NOTE j Ir. I'1)
(NOTE6) (NOTE6) (NOTE6) (NOTE6) (NOTE 6 (NOTE Q) (NOTE6) EMER6 A GElt EAG C7E1 i 1219-L.SK-24-9.4 II2179-LSK-Z4-9.4l REQUIREENTS (IF SEUENCING IS NECESSARY) ARESPECIFIED OUT OF 4 COINCIDENCE, MEMORIES, AND OOR" LOGIC ARE DUPLICATED WITHIN
~Tkms. EACH TRAIN. SEPARATE OUTPUT RELAYS ARE ALSO PROVIDED. TO MINIMIZE FALSE ALVEIN PARALLEL WITHTIE ASSOCIATED STEAM LINE STOPVALVE. CONTAINMENT SPRAY, ONE CUTOUT RELAY SHOULD START THE PUMPS WHILE ANOTHER VIN THECONTROlT ROO F AO EACHSTEAM LINE STOPVALVETO SHOULD OPEN THE SYSTEM VALVES.
IS RJLLY CLOSED ORFULLY OPEN. 17. THE CONTAINMENT SPRAY SYSTEM SHOULD BE INDEPENDENT OF THE SAFETY XYED ANOSEIUENCIED IF THEEMERGENCY DIESEL POWR CAPABILITY INJECTION SYSTEM. IF THE CONDITIONS OF NOTE 13 APPLY, THE SEQUENCE
)AD WITHALL SYSTEMS STARTING. THETIME DELAY(S), IF USED INTERLOCK SHOULD BE SUCH THAT SPRAY WILL START WITHIN THE REQUIRED 'M STARTING TIME REQUIREMENTS FOREACHSYSTEM. TIME INDEPENDENT OF THE SAFETY INJECTION SIGNAL STATUS.
0L BOARD.OPERATING EITHER CGMTROL WILL ACTUATE. 18. TWOMOMENTARY CONTROLS. ONE ON MAIN CONTROL BOARD 2 AND THE
)S FU4CTIONS ARENOTWITHIN THEFUNCTIONAL DESIGN SCOPE OF) OTHER ON MAIN VENTILATION PANEL VP.. OPERATING EITHER CONTROL IT lNLY 9Sft ONTHIS SHEETAS THEFUNCTIONS AREEBtILT IN THE WILL ACTUATE.
3RP*S* Y "FSAR FIGURE" JQAo Electric corporarion r - WuStIAghOUSS WesRtU house leStric Corporation 12179-2472.011-001-008 1.1r. c"aw" UTILimiv-. seRICIE Ce go., .mAOR1HEBST I 1
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a - 133HS l STEA m GENERAToJ PSSURZER CONTAINMENT PRESSURE
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IPRESSURE ToCONTROS) HIH STEAK PAME~~ H3E 9IN 9 2 9 (S.HEET 7) NtS , StHI-I X HI -2 N- - kL RESET P-4 D 61.00 REACTCP. TRIP (lw)TE a S 91 (SMEET2) SE.O" Ar A I -I7 II I. MANUALT_ (N*OTE8a 98 FEEDWATER REACTOR I \_ L 0l NtlRS L9N-E-S- ISOLATION (9SHEET 13) TRIP (SHEET2) By t ( N.E.5. BY I OTHERS 1D IOO UlP2L0 4 CDA LO IST l I1LTATIO ISTION ISOLATION IEMER6ENCX (NTE3NT (NOTE (NOTE3) (NOT E 3) ENUERAlTOR 1ENERA LOAD sEQUGEICEtI LOAD A 12179LSIl "Usco5 152152-2023 NuSCO, 25 l LOOP I 11 LOCP 2 11 LOOP 3 11 LOOP 4 i STEAtLINE STEASTINE ll STEANLINE STEAMILINE I 1 SW 11 SW l SWP 11 STOP I PUMPS I DIESE VALVE VAILVE l VALVE l VALKVE CSMkEE IS) (NDTIE NO NOE (NMTDT NOT (N JTTl NO
- 12) FvoLL 12) /^5FULL 12 PN 11 ^5FULL t(NOTE NOTES:
I. TWO IGNTARY CONTROLS ON THECONTROL 10. SAFETYINtECTI ON SE0UENCE REUYIRDE BOARD.OPERATING El7HER CONTROL WILL ACTUATE. BY tp NUCLEAR ENERGY SYSTD4S.
- 2. THE ANIJAL SPRAY ACTUATION CONSISTS OF FUR It0OENTARY CONTROLS. ACTUATION WILL OCCUR ONLYIF TW ALSO CLOSESTHE BYPASS ASSOCIATED CONTROLS AREOPERATED SIMULTANEOSLY. It. VALVE IN Pi
- 3. ONECONTROL PERLO ORTHECONTROL BOARD. 12. LIGHTS SHOULD BE PROVIDED IN TCEI INDICATE WHENTHE VALVE IS FULLY t
- 4. CONTAINMENT PRESSUIE BISTA8LES FORSPRAY ACTUATION AREENERGIZE TO ACTUATE (OTHERBISTAILES ARE 1. THEACTUATION MAYBIEDELAYED DE-EIMIZE TO ACTUATE), AND IS LESS THANTIE TOTAL LOAD WIVITH I S. CONTAINMENT AIR RECIRC FAN IS TRIPPED BY CDA SIGNAL. MAYNOT EXCEEDTHE NAltXII STARTII
- 6. COtOENTS AREALL INDIVIDUALLY SEALEDIN (LATCHED). SOTHAT LOSS OF THEACTUATION SIGNAL WILL NOT 14. TWOCONTROLS ON THE COTROL BOARD.
CAUSE THE£E CCPENTS TO RETURN TO THECONDITION HELD PRIOR TO THEADVENT OF THEACTUATION SIGNAL.
- 7. SERVICE WATER SYSTEN ISOLATION IS USEDONLYIF REUIRED. 15. SON ENGINEERED SAFEGUARDS FUNCTIC NUCLEAR ENERGY SYSTEHS BUTONLYSH B. THEREDUNDANT tANUAL RESETCONSISTS OF TWOMOENTARY CONTROLS ONTHECONTROL BOARD.ONEFOREACHTRAIN.
SU.PLIED EOUIPMENT.
- 9. SUPPLIED BY OTHERS.
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N CAUTION \117
- 1. After blocking SI, manual operator action may be required to actuate safety injection if required.
- 2. To avoid a low pressure SI, the following step must be completed prior to decreasing pressurizer pressure below 1892 psia.
4.2.5 WHEN pressurizer pressure reduced below P-11 setpoint (2000 psia on 2 of 3 channels) as indicated by MB4D 3-5, "PRESSURIZER PRESSURE LO INTERLOCK P-11," permissive light lit, PERFORM the following to block the associated SI signals:
- a. Under "MAIN STM PRES - SI," PERFORM the following:
- 1) VERIFY both switch collars in the "BLOCK NORMAL' position.
- 2) PRESS both pushbuttons.
- b. CHECK the following permissive lights lit:
- MB4D 1-4, "STM LINE ISOL SI BLOCK TR A"
- MB4D 2-4, "STM LINE ISOL SI BLOCK TR B"
- c. Under "PZR PRES - SI," PERFORM the following:
- 1) VERIFY both switch collars in the "BLOCK NORMAL' position.
- 2) PRESS both pushbuttons.
- d. CHECK the following permissive lights lit:
- MB4D 1-5, "PRESSURIZER SI BLOCK TR A'
- MB4D 2-5, "PRESSURIZER SI BLOCK TR B"
- e. IF pressurizer pressure increases above the P-Il setpoint, REPEAT step 4.2.5.
OP 3208 Level of Us Rev. 020-03 Continuou S 27 of 103 l
Written Examination Question Worksheet Form ES-401-6 ES-401 Examination Outline Cross-reference: Question# 72 Tier # 2 Group# 2 K/A # 039.A1.07 Importance Rating 2.6 Proposed Question: With the crew raising power per OP 3204 "At Power Operations", the BOP operator has been directed to manually place MSR Reheaters in service using OP 3317 "Reheat and Moisture Separator". observes the Over the course of raising the amount of steam supplied to the reheaters, the BOP following conditions: (3MSS-
- The maximum heatup rate for the "A" reheater tube bundle is 40'F in one hour T39A).
0
- The maximum heatup rate for the "B" reheater tube bundle is 22 F in 30 minutes (3MSS-T39B).
0 Foxboro
- The maximum LP turbine inlet steam heatup rate is 80 F in one hour as read on DCS.
side as
- Steam temperature entering the LP turbine is 60'F higher on one side than the other read on Foxboro DCS.
Which temperature limit has been exceeded? A. The maximum heatup rate for the "A" reheater tube bundle. B. The maximum heatup rate for the "B" reheater tube bundle. C. The maximum LP turbine inlet steam heatup rate. D. The maximum differential temperature entering the LP turbine. Proposed Answer: D Explanation (Optional): The temperature limits for tube bundle heatup rate include a recommended maximum heatup thermal stressing of the rate limit of 25 F per 30 minutes ("B" wrong) and 50'F per hour ("A" wrong)0to prevent 0 is 125 F per hour ("C" wrong) to prevent LP reheater tube region. The maximum heatup rate on the LP turbine inlet turbine must not vary by more than 500F turbine vibration. The steam temperature entering both sides of the LP ("D" correct) to prevent rotor rubbing. OP 3317, Precaution 3.1 - 3.3 Technical Reference(s): OP 3317, section 4.7 (Attach if not previously provided) Proposed references to be provided to applicants during examination: None (As Learning MC-04992 Describe major administrative or procedural precautions and limitations available) Objective: placed on operation of the moisture separator reheater system, and the basis for each. 41 of 46 NUREG-1021, Revision 8, Supplement 1
ES-401 Written Examination Question Worksheet Form ES-401-6 Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
-17N 2.1.2 Instrument air is available for pneumatic valves and controls as specified in OP 3332A, "Instrument Air System."
2.1.3 4,160 volt busses 34A and 34B are energized to supply power to the moisture separator drain pumps. 2.1.4 Main Steam System is in operation as specified in OP 3316A, "Main Steam," to provide reheater steam supply. 2.1.5 Turbine plant component cooling water is in operation as specified in OP 3330B, "Turbine Plant Component Cooling Water," and supplying the moisture separator drain pump(s). 2.2 Definitions 2.2.1 WARM - To minimize thermal shock by slowly admitting steam to components, allowing for uniform, even heating.
- 3. PRECAUTIONS 3.1 Steam temperature entering both sides of the low pressure (LP) turbine must be the same. If these temperatures vary by more than 50'F, rotor rubbing may occur. The steam supply valves to the reheater tube sections must be operated simultaneously.
3.2 Symmetry of operation should be monitored as load is changed by observing steam pressures, flows, and temperatures in the reheater tube bundles. 3.3 If heatup rate of 25 0 F per 15 min. is exceeded, excessive thermal stressing of reheater tube sections can occur. When started in "MANUAL," a maximum heatup rate of 250 F per 30 min. is recommended. 3.4 Full reheat temperature at low turbine loads results in over-heating of the LP turbine (reheat temperature and windage losses). Reheat steam high load valves and reheat steam low load valves should be checked for proper automatic operation. 3.5 Scavenging steam flow must be maintained at all times when the associated tube bundle is in service. Level of Use OP 3317 STOP THINK ACT REVIEW Rev. 014 6 of 68
3.6 In order to avoid steam cutting within the tube bundle, early detection of reheater tube leaks is important. Deviations of more than 5% above the operating curve for a particular reheater requires the reheaters be taken out of service until repairs are made. Leak detection readings must be monitored frequently. Leak detection data may be determined through Section 4.13. 3.7 To minimize thermal cycling during frequent startups and shutdowns, the turbine should be operated non-reheat by leaving the reheaters out of service until sustained loads above 65 percent of full load are anticipated. 3.8 Expansion joints on the suction side of the moisture separator drain tank pumps are not designed to withstand pump discharge pressure. To avoid over pressurizing the expansion joint when isolating a drain tank pump for maintenance, the discharge valve should always be closed prior to closing the suction valve. When closing the suction valve, any increase on drain tank pump local discharge pressure indication should be considered evidence of discharge valve back leakage and the suction valve reopened until the back leakage is corrected. OP 3317 Level of Use STOP THINK ACT REVIEW Rev. 014 Continuous 7 of 68
4.7 Manually Placing MSR Reheaters In Service During Power Ascension (30% to 65%) NZ7 CAUTION 7
- 1. The preferred method of MSR start is to place reheat in service at as close to 30% as possible and permit MSR heatup during turbine loading.
- 2. To prevent damage to the MSRs, heatup rates, temperature, and pressure curves (Attachments 1 and 2) must be followed. If these limits are not followed, adverse temperature gradients are induced in the shell side structural components and the reheat tube bundle resulting in structural damage and leaking tubes.
- 3. Excessive thermal stressing of reheater tube sections will result if a heatup rate of 50'F per hour is exceeded or if the MSR pressure and temperature curves (Attachment 1 and 2) are not adhered to.
- 4. If steam supply pressure or reheat steam temperature vs.
cross- around pressure are off of the "curve," return to the "curve" slowly rather than with a step change. NOTE
- 1. During plant start, reference legs for 3DSR-LT20A/3DSR-LT20B, reheater drain tank level transmitters, may flash and indicated level may go to 100%. This condition will clear as steam is admitted to the tank and the reference leg is filled by the condensing pot.
- 2. Section 4.15 provides guidance for operating MSR controls in Manual.
4.7.1 ESTABLISH the following initial conditions:
- a. CLOSE the following (MB6):
- 3MSS-MOV5OA, "REHEAT STM SPLY"
- 3MSS-MOV5OB, "REHEAT STM SPLY"
- b. Refer to Section 4.15 and VERIFY the following steam flow controllers in "MANUAL' and "SET" to 0% output (Foxboro DCS):
- 3CRS-MSR1A
- 3CRS-MSR1B OP 3317 Level of Use THINK ACT REVIEW . Rev. 014 STOP Continuous 25 of 68
- c. OPEN the following reheater steam valves to the condenser "SCAVENGE STM-VVS" (MB7):
- 3DSR-MOV29A
- 3DSR-MOV29B
- d. OPEN the following reheater steam supply drains "MSIV/REHEAT STM SPLY DRAIN" (MB7):
- 3DTM-MOV33A-33B
- 3DTM-AOV36A-36B
- e. VERIFY the following reheater drain tank high level drain valve controllers in 'AUTO" and set for 80%:
- 3DSR-LIC20A1
- 3DSR-LIC20B1
- f. VERIFY the following reheater drain tank normal level drain valve controllers in "MANUALU' and set to minimum output:
- 3DSR-LIC2OA2
- 3DSR-LIC2OB2
- g. VERIFY the following reheat drain tank level control valves, closed:
- 3DSR-LV2OA4
- 3DSR-LV2OA3
- 3DSR-LV20A2
- 3DSR-LV2OB4
- 3DSR-LV2OB3
- 3DSR-LV20B2 OP 3317 Level of Use THINK AT REVIEW Rev, 014 Continuous STOP 26 of 68
4.7.2 To warm reheater, PERFORM the following:
- a. Refer To Section 4.15 and ADJUST the following steam flow controllers to 2% output (Foxboro DCS):
- 3CRS-MSR1A
- 3CRS-MSR1B
'\7 CAUTION Excessive thermal stressing of the reheater tube bundle can occur if a heatup rate of 250 F per 15 min. is exceeded. A maximum heatup rate of 25 0F per 30 min. is recommended when placing the tube bundle in service manually.
- b. To begin warmup of the reheater tube bundles, simultaneously and slowly THROTTLE open the following reheater steam supply bypass valves:
- 3MSS-V892, A reheater steam supply bypass valve
- 3MSS-V891, B reheater steam supply bypass valve
- c. MONITOR heatup rate and ADJUST the following valves as needed to maintain the heatup rate less than of equal to 250F per 30 min.
- 3MSS-V892, A reheater steam supply bypass valve
- 3MSS-V891, B reheater steam supply bypass valve
- d. OBSERVE the following reheat steam supply pressures increasing (computer):
- MSS-P53A
- MSS-P53B OP 3317 Level of Use Rev. 014 STOP THINK ACT REVIEW Continuous 3; 1.$
27 of 68
- e. WHEN reheat steam supply pressures have stabilized, OPEN the following (MB6):
- 3MSS-MOV50A, "REHEAT STM SPLY"
- 3MSS-MOV5OB, "REHEAT STM SPLY"
- f. CLOSE the following:
- 3MSS-V892, A reheater steam supply bypass valve
- 3MSS-V891, B reheater steam supply bypass valve
- g. While tube bundles are being warmed, MAINTAIN power level steady for a minimum of 15 minutes.
4.7.3 To supply steam to the reheaters, PERFORM the following:
- a. VERIFY turbine load is approximately 30%.
- b. To ESTABLISH a heatup rate, not to exceed 25°F per 30 minutes, as indicated on 3MSS-TE39A and 3MSS-TE39B, slowly INCREASE output on the following steam flow controllers (Foxboro DCS):
- 3CRS-MSR1A
- 3CRS-MSR1B A. OP 3317 Level of Use IACT REVIEW Rev. 014 STOP THINK Continuous 28 of 68
- c. CONTINUE increasing output on the steam flow controllers until one of the following is met:
- The following "REHEAT STEAM SUPPLY" valves are full open (MB6):
- 3MSS-PDV37A, "LOW-LOAD VVS"
- 3MSS-PDV37B, "LOW-LOAD VVS"
- The following temperature conditions are met (plant process computer and Foxboro DCS):
- MSS-T39A, reheat steam supply temp, is approximately 50'F greater than 3CRS-T23A, MSR temp
- MSS-T39B, reheat steam supply temp, is approximately 50'F greater than 3CRS-T23B, MSR temp NOTE If the output of the steam flow controllers is increased further, 3MSS-PDV36A and 3MSS-PDV36B, high load valves, will begin to open.
- d. IF necessary, CONTINUE increasing output on the steam flow controllers until the following is met:
- MSS-T39A, reheat steam supply temp., is approximately 50'F greater than 3CRS-T23A, MSR temp
- MSS-T39B, reheat steam supply temp., is approximately 50'F greater than 3CRS-T23B, MSR temp 4.7.4 WHEN reheat steam supply temperatures indicate approximately 50°F greater than MSR temperatures, MAINTAIN plant power and reheat steam supply steady for at least 30 minutes.
OP 3317 Level of Use ACT REVIEW Rev. 014 STOP THINK Continuous 29 of 68
CAUTION 7
- 1. The reheater steam load valves must be operated simultaneously to prevent differential temperature between the two sides of the low pressure turbine reaching 50'F as read on the Foxboro Digital Control System.
- 2. Full reheat temperature (approximately 500'F) at low turbine steam flows will result in overheating of the low pressure turbine.
- 3. Do not exceed a 500 F per hour heatup rate in MSS-T39A and MSS-T39B, reheater tube bundles (plant process computer or Foxboro Digital Control System), or saturation temperature for 3MSS-PT57A and 3MSS-PT57B (Foxboro Digital Control System).
- 4. Do not exceed 1250F per hour temperature change on the LP turbine inlet as read on the Foxboro Digital Control System. Heatups in excess of this rate may lead to LP turbine vibration: [Ref. 6.2.6]
Reheater A steam to LP turbine:
- 3HRS-TE20A
- 3HRS-TE20B
- 3HRS-TE20C Reheater B steam to LP turbine:
- 3HRS-TE28A
- 3HRS-TE28B
- 3HRS-TE28C NOTE The heatup should be performed in a steady manner; step changes should be avoided.
4.7.5 To open the reheater steam supply valves, PERFORM the following:
- a. Refer To Attachments 1 and 2 and OBSERVE the following turbine cross-around pressures (computer or Foxboro DCS)
- CRS-P24A
- CRS-P24B OP 3317 Level of Use ACT REVIEW Rev. 014 STOP THINK Continuous - 30 of 68
- b. MONITOR the following parameters:
- MSS-T39A, MSR A reheater tube bundle temperature (computer or Foxboro Digital Control System)
- MSS-T39B, MSR B reheater tube bundle temperature (computer or Foxboro Digital Control System)
- 3MSS-PT57A, reheater A steam supply pressure (MB6 or Foxboro Digital Control System)
- 3MSS-PT57B, reheater B steam supply pressure (MB6 or Foxboro Digital Control System)
NOTE If manual control of the reheater steam supply valves is taken at a location other than the Foxboro Digital Control System operator station, constant communications with the Control Room must be established.
- c. Using the following reheat steam load flow controllers, slowly RAISE MSR steam supply pressures and temperatures, based on CRS-P24A and CRS-P24B, cross-around pressures (computer) found in Attachment 1 and 2, or "MSR Startup Pressure Curve" (Foxboro DCS):
- CRS-MSR1A
- CRS-MSR1B NOTE The procedure should be continued and whenever conditions are met, step 4.7.5.d. should be performed.
- d. WHEN reheat drain tank levels indicate greater than 13% on 3DSR-LI 20, "MSR DRAINS" "TK LVL" (MB6), VERIFY the following annunciators, not lit:
- MB6B 3-9, "MOIST SEP RHT DRN TK A LEVEL HI/LO"
- MB6B 4-9, "MOIST SEP RHT DRN TK B LEVEL HI/LO" OP 3317 Level of Use Rev. 014 STOP THINK ACT REVIEW Continuous 31 of 68
4.7.6 WHEN both the following conditions have existed for at least two hours, Go To step 4.7.7:
- Reheater steam supply pressure is approximately 300 psig or greater (MB6 or Foxboro DCS):
- 3MSS-PT57A, "REHEAT STEAM SPLY" "PRES"
- 3MSS-PT57B, "REHEAT STEAM SPLY" "PRES"
- Turbine load is 50% or greater 4.7.7 PERFORM the following:
- a. OPEN the following first point feedwater heater scavengin steam valves, "SCAVENGE STM-VVS" (MB7):
- 3DSR-MOV31A
- 3DSR-MOV31B
- 3DSR-MOV31C
- b. CLOSE the following reheater steam valves to the condenser, "SCAVENGE STM-VVS" (MB7):
- 3DSR-MOV29A
- 3DSR-MOV29B
- c. CLOSE the following reheater steam supply drains, "MSIV/REHEAT STM SPLY DRAIN" (MB7):
- 3DTM-MOV33A-33B
- 3DTM-AOV36A-36B Level of Use OP 3317 STOP THINK A REVIEW Rev. 014 Continuous 32 of 68
4.7.8 VERIFY output at 100% and IF desired, PLACE the following-reheater steam load flow controllers in "AUTO" (Foxboro DCS):
- 3CRS-MSR1A
- 3CRS-MSR1B 4.7.9 IF the reheat scavenging steam flow valve position is to be adjusted, Go To Section 4.14.
- End of Section 4.7 -
OP 3317 Level of Use THINK -`rACT REVIEW Rev. 014 STOP Continuous 33 of 68
Written Examination Question Worksheet Form ES-401-6 ES-401 Question # 73 Examination Outline Cross-reference: Tier # 2 Group# 2 K/A # 055.GEN.2.1.32 Importance Rating 3.8 Proposed Question: a slowly decreasing With the plant in HOT STANDBY, a PEO is dispatched to investigate condenser vacuum. The PEO and BOP operator report the following:
- The "A", "C", and "E" main circulating water pumps are running.
- Gland sealing steam pressure is 0.5 psig.
- There is excessive flow out of the air ejector atmospheric discharge.
- Air ejector suction manifold temperatures are normal.
What is the likely cause of the decreasing vacuum? A. Insufficient circ water flow to the condenser bays. B. Insufficient gland sealing steam pressure exists. C. Loss of steam supply to the SJAEs. D. The SJAE is backfiring. Proposed Answer: B Explanation (Optional): "A" is wrong since one circ pump per bay is adequate in HOT STANDBY, and excessive are accumulating in the main condenser. "B" is flow out of the air ejector exists, indicating non-condensable gasses to 5 psig, and the AOP checks for between 2 and 6 correct, since the normal gland sealing steam pressure band is 3 is wrong since OP 3329, Precaution 3.1 states that psig. "C" is wrong, since SJAE flow is greater than normal. "D" elevated SJAE suction manifold temperature, and/or indications of backfiring include decreasing condenser vacuum, excess flow exists. All of the distractors are plausible, reduced or no flow out of SJAE atmospheric discharge, and condenser vacuum. since they are conditions that would lead to a decreasing OP 3329, Precaution 3.1 (Attach if not previously provided) Technical Reference(s): OP 3323D, section 4.3 AOP 3559, steps 2, 3, and 4 None Proposed references to be provided to applicants during examination: and limitations (As Learning MC-04096 Describe the major administrative or procedural precautions available) air removal system ...: MC-06149 Describe the Objectives: placed on the operation of the condenser and limitations placed on the operation of major administrative or procedural precautions the gland seal and gland exhaust systems. Modified bank # 69789 (Parent attached) Question Source: Question Cognitive Level: Comprehension or Analysis 10 CFRPart 55 Content: 55.41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
Written Examination Question Worksheet Form ES-401-6 ES-401 Original question 69789 a decreasing condenser vacuum. The PEO With the plant in HOT STANDBY, a PEO is sent out to investigate reports the following:
- There is no flow out of the air ejector atmospheric discharge.
- There is elevated air ejector suction manifold temperatures.
What is the likely cause of the decreasing vacuum? A. The SJAE is backfiring. B. Loss of steam supply to the SJAEs. C. Loss of circ water flow to a condenser bay D. Loss of condenser vacuum breaker water seal. Answer: A 41 of 46 NUREG-1021, Revision 8, Supplement 1
2.1.6 Auxiliary steam is in operation as specified in OP 3331A,
'Auxiliary Boiler, Steam and Condensate," to operate steam jet air ejectors.
2.1.7 Hotwell sampling is in service as specified in OP 3311B, "Turbine Plant Sampling." 2.1.8 The main turbine and SGFP turbines are on turning gear with gland steam and exhaust in service as specified in OP 3323D, "Turbine Gland Steam and Exhaust." 2.2 Personnel 2.2.1 Chemistry Department personnel support is required for the following sections:
- Section 4.2 and 4.3 to monitor Oxygen concentration as required
- Section 4.6 to coordinate Nitrogen overpressure on main condenser
- Section 4.7 to ensure no activity in SJAE exhaust 2.3 Documents 2.3.1 OP 3311B, "Turbine Plant Sampling" 2.3.2 OP 3319C, "Condensate Demineralizer Mixed Bed System" 2.3.3 OP 3323A, "Main Turbine" 2.3.4 OP 3323D, "Turbine Gland Steam and Exhaust" 2.3.5 OP 3330B, "Turbine Plant Component Cooling Water" 2.3.6 OP 3331A, "Auxiliary Boiler, Steam and Condensate"
- 3. PRECAUTIONS 3.1 SJAE backfiring can occur when adjustments are made to steam jet air ejector lineup and can significantly affect condenser vacuum.
l : -- OP 3329 Level of UseOP32 REVIEW Rev. 10CHG2 ContinoUs Continuous3 STOP THINK ACT of 21
Indications of backfiring include:
- Decreasing condenser vacuum
- Elevated ejector suction manifold temperature
- Reduced or no flow out of ejector atmospheric discharge OP 3329 Level of Use STOP THINK ACT REVIEW Rev. 10CHG2 Continuous 4 of 21
4.3 Adjusting Gland Seal Pressure NOTE
- 1. This section may be used to adjust gland steam header pressure between 3 - 5 psig when main steam is supplying gland seals and the nominal setpoint of 3.5 psig is not providing satisfactory turbine sealing.
- 2. Increased seal supply pressure improves the sealing of the turbine.
However, supply pressure greater than 4 psig may cause steam packing unloading valve (3TME-PCV31) to open and reduce plant efficiency. 4.3.1 CHECK main steam supplying gland seals:
- 3TME-MOV33, "GLAND STM" "MS SPLY" (MB7) open
- 3TME-MOV35, "GLAND STM" "AUX SPLY" (MB7) closed 4.3.2 ADJUST "TG GLAND SEAL PRESSURE" setpoint between 3 - 5 psig. (DCS) 4.3.3 IF 3TME-PCV31, steam packing unloading valve opens, NOTIFY system engineer for guidance on supply pressure.
- End of Section 4.3 -
OP 3323D Level of Use REVIEW Rev. 009 STOP THINK :ACT-Continuous 9 of 14 I
I RERESPONSE NOT OBT I-
- e. Check condenser e. Perform the following:
backpressure:
- 1) IF turbine load is
- LESS THAN OR EQUAL GREATER THAN TO 5 inches Hg Absolute 360 Mwe, THEN Return to step 1.d.
- STABLE OR DECREASING 2) IF backpressure increases to GREATER THAN 5 inches Hg Absolute, THEN TRIP the reactor and Go to E-0, Reactor Trip or Safety Injection.
- 2. Check Circulating Water System Operation
- a. Verify circulating water a. Place both condenser steam pumps - ONE PER dump interlock selector CONDENSER RUNNING switches to OFF
- b. Verify water box outlet b. OPEN water box outlet isolation valves - OPEN isolation valves.
- c. Verify all circulating water c. Using OP 3325A, "Circulating pumps - RUNNING Water," Restore any available circulating water pump to service (normal or with seaweed accumulation).
IF a pump can NOT be started, THEN OPEN the associated water box inlet cross-connect valve, if desired.
ACTION/EPECTED RESPONS ONSE NOT OBTAINED
- d. Verify the traveling screen d. Locally Verify debris conveyor differential pressure - (3SWT-SSC2) operating and LESS THAN 12 inches H 2 0 NQT clogged.
IF conveyor clogged, THEN Open debris trough trap door.
- 3. Check Condenser Air Removal Alignment
- a. Verify steam jet air ejector a. OPEN valve.
auxiliary steam supply valve (3ASS-AOV22) - OPEN
- b. Using OP 3329, "Condenser Air Removal," locally Perform the following:
- 1) Verify both sets of steam 1) Place a second steam jet air jet air ejectors in service ejector in service.
- 2) Verify all first stage jets in 2) Place all first stage jets in service on each air ejector service, if desired.
- 3) Check for indications of air ejector backfiring
- c. Verify isolation dampers for c. OPEN dampers gaseous waste to Unit 1 stack (3GWS*AOD78A and 3GWS*AOD78B) - OPEN
- d. At Gas Waste Panel d. Locally START one process (3GWS-PNL-P6), Verify vent fan.
process vent fans (3GWS-FN1A or 3GWS-FN1B) - ONE RUNNING
LOSS OF CONDENSER AOP 3559 Page 6 VACUUM Rev. 6 SAPRESPONSE NOT OBTAINED
- e. Locally (Turbine Bldg 38' e. Locally (Turbine Bldg 38' southwest) Verify steam jet air southwest), using OP 3329, ejector exhaust valves "Condenser Air Removal,"
(3ARC-AOV36A and Align SJAE to exhaust to 3ARC-AOV36B) - OPEN Turbine Building.
- 4. Check Gland Seal Pressure - Adjust gland steam pressure to BETWEEN 2 and 6 psig maintain between 2 and 6 psig using the supply pressure regulator bypass valve (3TME- MOV34).
NOTE Locked valve key is required for some local operations.
- 5. Check Condensate Surge Tank Perform the following:
Level
- a. IF condensate storage tank level is GREATER
- GREATER THAN 18,000 gal THAN 200,000 gallons, THEN
- NOT DECREASING IN AN Locally Unlock and Open UNEXPECTED MANNER the condensate storage to surge tank cross-connect valve (3CNS-V29).
- b. Consult Duty Officer to determine if plant shutdown is necessary.
ES-40 I Written Examination Question Worksheet Form ES-40 1-6 Examination Outline Cross-reference: Question # 74 Tier # 2 Group# 2 K/A # 062.A3.01 Importance Rating 3.1 Proposed Question: With the plant at 100% power, and swing charger 301B-3 tagged out, an electrical fault iin MCC32-2W results in the MCC de-energizing. All systems respond as designed, and the PEO is checking the effects of the MCC loss o]n related equipment. What should inverter 4 output amps to VIAC-4 be indicating, and why? A. 0 amps, since VIAC-4 has lost power until the PEO manually aligns VIAC-4 to the aLlternate source via the manual bypass switch. B. 0 amps, since VIAC-4 has automatically aligned to the alternate source via the static switch. C. 60 amps, since inverter 4 is receiving power from battery charger 301B-2 via the DC bus. D. 60 amps, since inverter 4 is receiving power from the battery via the DC bus Proposed Answer: D Explanation (Optional): Voltage at the rectifier output drops to 0 volts (less than 132 VDC) due to the loss of MCC-32-2W, causir ig the blocking diode to lose the reverse bias, allowing the DC bus to supply power to the inverter without loss cof continuity ("D" correct). If the DC bus had not been available, the high speed static switch would automatically tranisfer to the alternate source upon loss of inverter output power ("B" wrong). When the inverter is out of service, the <alternate path may be manually selected via a manual bypass switch ("A" wrong). "C" is wrong since the bus 32-21W supplies both the rectifier and the charger. Technical Reference(s): EE-IBA (Attach if not previously pro vided) Proposed references to be provided to applicants during examination: None Learning MC-03305 ...describe the 125 VDC distribution system electrical alignment under the (As Objective: following conditions... Loss of normal AC power supply. available) Question Source: Bank # 68106 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
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Written Examination Question Worksheet Form ES-401-6 ES-401 Question # 75 Examination Outline Cross-reference: Tier # 2 Group# 2 K/A # 064.K2.02 Importance Rating 3.1 Proposed Question: The crew is responding to a loss of power using ECA-0.0 "Loss of All AC Power", and the following sequence of events occurs:
- 1. A PEO starts the "A" EDG using ECA-0.0, Attachment "F'.
- 2. The EDG Output Breaker automatically closes and re-energizes Bus 34C.
- 3. A PEO reports that the "B" EDG can not be started.
Lo-Lo"
- 4. After approximately 20 minutes of loaded operation the "DG A Day Tank Level annunciator comes in.
the Fuel Oil
- 5. The PEO reports that the "A" Day Tank Level is 185 gallons and neither of Transfer Pumps (3EGF*P1A & 3EGF*P1C) are running.
are unsuccessful.
- 6. Although power is available, attempts to start the Fuel Oil Transfer Pumps How will the crew fill the "A" EDG Fuel Oil Day Tank?
Fuel Oil A. Obtain maintenance department assistance and align hoses from the "B" train Transfer Pump. the Fuel Oil Transfer B. Place the 3EGS*PNLA control switch on MCC32-lT-3H to "start" and start Pumps from 3EGS*PNLA. operate 3EGF*PI B, C. Mechanically align the system, and use Kirk keys to electrically align and Fuel Oil Transfer Pump, from alternate power supply, 32-IT. 3EGF*P1D, D. Mechanically align the system, and use Kirk keys to electrically align and operate Fuel Oil Transfer Pump, from alternate power supply, 32-IT. Proposed Answer: D Explanation (Optional): "A" is wrong since power is not available for the "B" fuel oil transfer pump. "B" is wrong is on 32-IT, and it is since 3EGS*PNLA contains non-essential loads. This is plausible, since its control switch supply. "D" have an alternate power operated when recovering in ECA-0.1. "C" is wrong since 3EGF*PlB does not to alternate power supply and flowpath. is correct, since ARP MB8B, 5-3 refers to OP3346B to align 3EGF*PID Technical Reference(s): OP 3353.MB8B, 5-3 (Attach if not previously provided) FSAR Figure 8.3-6 OP 3346B, sections 4.6 and 4.8.5 None Proposed references to be provided to applicants during examination: Diesel (As available) Learning MC-04404 Given a failure (partial or complete) of the Emergency Generator System, determine the effects on the system and on interrelated systems. Objective: Question Source: Bank # 75629 41 of 46 NUREG-1021, Revision 8, Supplement 1
ES-401 Written Examination Question Worksheet Form ES-401-6 Question History: Previous NRC Exam Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
Setpoint: Less than 187 gallons 5 -3 DGA DAY TANK LEVEL LO-LO AUTOMATIC FUNCTIONS
- 1. None CORRECTIVE ACTIONS
- 1. CHECK 2A day tank level to confirm alarm.
- 2. VERIFY 3EGF*P1A and 3EGF*P1C, fuel transfer pumps, in 'AUTO."
- 3. CHECK 3EGF*P1A and 3EGF*P1C, fuel transfer pumps, running.
- 4. IF 3EGF*P1A and 3EGF*P1C, fuel transfer pumps, are not running, VERIFY power available at load center 32-1T
- 5. IF power is available to 3EGF*P1A and 3EGF*P1C, fuel transfer pumps, START 3EGF*PlA and 3EGF*P1C, fuel transfer pumps, as required to maintain day tank level 350 to 490 gallons (3EGS*PNLA).
- 6. IF power is lost to 3EGF*PlA and 3EGF*P1C, fuel transfer pumps, PERFORM one of the following:
6.1 RESTORE power to 32-iT 6.2 Refer To OP 3346B, "Diesel Fuel Oil," and ALIGN fuel transfer pumps to alternate power source.
- 7. Refer To the following Technical Specifications and DETERMINE Limiting Condition for Operation:
- T/S 3.8.1.1, "A.C. Sources - Operating"
- T/S 3.8.1.2, "A.C. Sources - Shutdown"
- 8. WHEN emergency diesel generator A is stopped, RAISE day tank level to 490 gallons.
SUPPORTING INFORMATION
- 1. Initiating Device Setpoint 1.1 3EGF-LS28A 1.1 < 187 gallons OP 3353.MB8B Level of Use STOP THIN' ACT R.1W Rev. 002-02 Continuous 93of 109 I
5-3
- 2. Computer Points 2.1 EGF-L28A
- 3. Possible Causes 3.1 Fuel transfer pump malfunction 3.2 Fuel Oil System leak 3.3 Loss of power to fuel transfer pump 3.4 Relay failure
- 4. Technical Specifications 4.1 T/S 3.8.1.1, "A.C. Sources - Operating" 4.2 T/S 3.8.1.2, "A.C. Sources - Shutdown"
- 5. Procedures 5.1 OP 3346B, "Diesel Fuel Oil" 5.2 OP 3353.EGPA, "Emergency Generator Panel A Annunciator Response"
- 6. Control Room Drawings 6.1 ESK 10JG 6.2 LSK 8-9B
- of U OP 3353.MB8B
- Level e STOP THINK Rev. 002-02 Continuous 94 of 109 I
LEGEND D BREAKER k COMBINATION STARTER OF KEY REMOVED FROM DEVICE IN POSITION SHOWN ?AIN B Fia KEY HELD IN LOCK IN DEVICE IN POSITION SHOWN L-O LOCKED OPEN, KEY REMOVABLE WHEN OPEN gL-O L-C LOCKED CLOSED, KEY REMOVABLE WHEN CLOSED KEY A-C HELD CAPTIVE IN 3EGF*TRSIA CBI WHEN OPEN; HELD CAPTIVE IN 32 -IT STR 2K WHEN CLOSED
%KEY A KEY B-C HELD CAPTIVE IN 3EGF*TRSIA CBI OR CB2 WHEN CLOSED KEYC-C HELD CAPTIVE IN 3EGF*TRSIA CB2 WHEN OPEN; HELD CAPTIVE IN 32-lU BKR 3M WHEN CLOSED
\ KEY A-D HELD CAPTIVE IN 3EGF*TRSIB CBI WHEN OPEN; I' HELD CAPTIVE IN 32-lU STR 4K WHEN CLOSED 3EGF*PIB KEY B-D HELD CAPTIVE IN 3EGF*TRSIB CBI OR C82 WHEN CLOSED
_-71 % I \ KEYC-C HELD CAPTIVE IN 3EGF*TRSIB CB2 WHEN OPEN; HELD CAPTIVE IN 32-IT BKR 3M WHEN CLOSED 3EFG*PIC NORMALLY ALIGNED WITH TRAIN A L-O L-( 3EFG*PID NORMALLY ALIGNED WITH TRAIN B
\3EGF*TRSIB FIGURE 8.3-6 EMERGENCY GENERATOR FUEL OIL TRANSFER PUMPS MILLSTONE NUCLEAR POWER STATION UNIT 3 FINALSAFETY ANALYSIS REPORT 15 SEPTEMBER 1985 AMENDMENT 15 SEPTEMBER 1985
I ( (TRAIN A) E 4M 2K 3K 2M 3M KEY C-D KEY A-C KEY KEY C-C r--
\ rJ L-C L-C L-0 3EGF*TRSIA 3EGF*PIC 3EGF
-5 4.6 Operation of 3EGF*P1D, Fuel Oil Transfer Pump, from Alternate Power Supply V7 CAUTION V This section defeats the automatic start and stop feature of 3EGF*PID, fuel oil transfer pump.
4.6.1 IF aligning 3EGF*P1D, fuel oil transfer pump, to alternate power, PERFORM the following:
- a. At motor control center "BUS 32- 1U" (B DIESEL),
PERFORM the following:
- 1) TURN 3EGF*PlD, "FEEDER A' (BUS 32-lU (4K)),
to "OFF."
- 2) TURN Kirk Key #16151 (BUS 32-IU (4K)), and REMOVE.
- b. At transfer switch panel "3EGF*TRS1B" (B DIESEL),
PERFORM the following:
- 1) INSERT Kirk Key #16151, "TRSIB (1DB)," and TURN.
- 2) TURN "TRANSFER SWITCH BREAKER A' to "OFF."
- 3) TURN Kirk Key #16158, "TRS1B (1DT)," and REMOVE.
- 4) INSERT Kirk Key #16158, "TRS1B (2CT)" and TURN.
- 5) TURN "TRANSFER SWITCH BREAKER B" to "ON."
- 6) TURN Kirk Key #16155, "TRS1B (2CB)," and REMOVE.
OP 3346B Level of Use ..... ... Contiuous tj ~Rev. 009-03 14 of 46
- c. At motor control center "BUS 32- iT" (A DIESEL),
INSERT Kirk Key #16155, "FEEDER B" (32-iT (3K)), and TURN. 4.6.2 IF starting 3EGF*PlD, fuel oil transfer pump, TURN "FEEDER B" (32-iT (3K)), to "ON." 4.6.3 IF stopping 3EGF*PlD, fuel oil transfer pump, TURN "FEEDER B" (32- IT (3K)), to "OFF." 4.6.4 IF restoring 3EGF*PlD, fuel oil transfer pump, to normal power, PERFORM the following:
- a. At motor control center "BUS 32- IT" (A DIESEL),
PERFORM the following:
- 1) VERIFY 3EGF*TRS1B, "FEEDER B" (32-iT (3K)),
in "OFF."
- 2) TURN Kirk Key #16155, "FEEDER B" (32-IT (3K)),
and REMOVE.
- b. At transfer switch panel "3EGF*TRS1B" (B DIESEL),
PERFORM the following:
- 1) INSERT Kirk Key #16155, "TRS1B (2CB)," and TURN
- 2) TURN "TRANSFER SWITCH BREAKER B," to
- 3) TURN Kirk Key #16158, "TRS1B (2CT)," and REMOVE.
- 4) INSERT Kirk Key #16158, "TRS1B (1DT)," and TURN.
- 5) TURN "TRANSFER SWITCH BREAKER A" to "ON."
- 6) TURN Kirk Key #16151, "TRS1B (1DB) "and REMOVE..
Levelonf Use lOP 3346B Contnuou Th~~X A~T FF~W Rev. 009-03 15 of 46
- c. At motor control center "BUS 32-lU" (B DIESEL),
PERFORM the following:
- 1) INSERT Kirk Key #16151 (BUS 32-lU (4K)),
and TURN.
- 2) TURN 3EGF*P1D, "FEEDER A' (BUS 32-1U (4K)),
to "ON."
- End of Section 4.6 -
- k OP 3346B Level of Use ........ Rev. 009-03 Continuous iK 16 of 46
- 1. CLOSE and LOCK 3EGF*V13, cross connect valve (A diesel).
4.8.5 IF using fuel oil transfer pump 3EGF*P1D (powered from alternate power) to fill fuel oil day tank 2A, PERFORM the following:
- a. Refer To Section 4.6 and ALIGN 3EGF*P1D to alternate power supply.
- b. OPERATE 3EGF*P1D, "FUEL OIL TRANSFER PUMP" (32- iT (3K)), as necessary to raise fuel oil day tank 2B level to approximately 490 gallons. 1g
- c. UNLOCK and OPEN 3EGF*V14, cross connect valve (B diesel).
- d. UNLOCK and CLOSE 3EGF*V12, common pump discharge isolation valve (B diesel).
- e. OPERATE 3EGF*PlD, "FUEL OIL TRANSFER PUMP" (32- iT (3K)), as necessary to MAINTAIN fuel oil day tank 2A level between 350 and 490 gallons.
- f. WHEN cross connect operation is no longer required, STOP 3EGF*PlD, "FUEL OIL TRANSFER PUMP" (32-IT (3K)).
- g. OPEN and LOCK 3EGF*V12, common pump discharge valve (B diesel).
- h. CLOSE and LOCK 3EGF*V14, cross connect valve (B diesel).
- i. Refer To Section 4.6 and RESTORE 3EGF*P1D to normal power supply.
- j. VERIFY 3EGF*P1D, "FUEL OIL TRANSFER PUMP" (3EGS*PNLB), in "AUTO."
- k. CLOSE and LOCK 3EGF*V13, cross connect valve (A diesel).
-End of Section 4.8 -
lLevelofUse lOP 3346B
'A T -7............ Rev. 009-03 22 of 46
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 76 Tier # 2 Group # 2 K/A # 073 .A2.0 1 Importance Rating 2.9 Proposed Question: With the plant at 100% power, the following sequence of events occurs: T=0: Control Room Ventilation Supply Radiation Monitor 3HVC*RE16A has a momentary loss of power. T+30 seconds: The RO verifies no actual radiation increase on 3HVC*RE16A and 16B. T+45 seconds: Based on US direction, the RO resets CBI on MB2. What action will physically be required in order to restore from this event? A. Close Control Room Air Bank Isolation Valve 3HVC*SOV74A. B. Place Emergency Filter Recirc Damper 3HVC*AOD1 19A in NORMAL. C. Close Outside Air Isolation Valve 3HVC*AOV25. D. Open Normal Supply Damper 3HVC*AOD 27A. Proposed Answer: D Explanation (Optional): The momentary loss of power causes a CBI signal, automatically closing the following valves and dampers: Kitchen Exhaust Isolation, Purge Exhaust Isolation, Outside Air Isolation ("C" wrong), and the Normal Supply Damper ("D" correct). After 60 seconds, the Air Bank Isolation opens, but the crew resets CBI prior to I minute elapsing, preventing the air bank from discharging ("A" wrong). The Recirc Damper is only placed in "Emergency" manually I hour after a required CBI ("B" is wrong). Technical Reference(s): OP 3314F, section 4.13 (Attach if not previouslv provided) Am __ raw l- v v6J Functional Dwg # 8 Proposed references to be provided to applicants during examination: None Learning MC-05472 Given a failure, of the Radiation Monitoring System (partial or (As available) Objective: complete), describe the effects on the system and on interrelated systems. Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.5 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
4.13 Operation of Control Building Isolation (CBI) System X t Tis Setion ontains EOP Related Material NOTE Both air banks are rated to provide adequate air flow for one hour. When one hour has elapsed, outside filtered air will be necessary to provide makeup air as specified in step 4.10.1 for Train A OR step 4.11.1 for Train B. 4.13.1 To manual initiate Control Building isolation, PERFORM the following:
- a. PLACE 3HVC-FN6, "KITCHEN EXH FAN" to "OFF" (VP1).
- b. IF Purge System is in service, Refer To step 4.12.2 and SHUTDOWN Control Building Purge System.
- c. PERFORM the following:
- CLOSE and DOG the following Control Building pressure boundary doors:
- CBwest47'6" (C-47-IA)
- CB east 64' 6" (C-64-1B)
- VERIFY the following Control Building pressure boundary doors, closed:
- CB west 47'6" (C-47-1)
- CB north 64'6" chiller room door, (C-64-4)
- CB north 64'6" chiller room door, (C-64-5)
- CB east 49'6" (C-49-1)
OP 3314F Level of Use ACT REViEW Rev. 019-01 STOP, THINK Continuous I -. 90 of 150 I -- --- ,
- d. PRESS "CBI" pushbutton on MB2 or VP1 and OBSERVE the following:
- 1) MB4D 3-6, "CONTROL BUILDING ISOLATION,"
lit.
- 2) The following close (VPI):
- a. "KITCHEN EXH AIR ISOL":
- 3HVC*AOV20
- 3HVC*AOV21
- b. "CNTRL BLDG PURGE EXH ISOL DMPR":
- 3HVC*AOV22
- 3HVC*AOV23
- c. "OUTSIDE AIR ]SOL':
- 3HVC*AOV25
- 3HVC*AOV26
- d. "NORMAL SPLY DMPR":
- 3HVC*AOD27A
- 3HVC*AOD27B
- e. WHEN 60 seconds have elapsed, OBSERVE the following "CNTL RM AIR BANK ISOL Vt"open (VP1):
- 3HVC*SOV74A
- 3HVC*SOV74B Level of Use O 3314F Continuous IN EYE Rv 090
~' 91 of 150
NOTE 3HVC*PCV68A and 3HVC*PCV68B, pressure control valves, will control outlet flow for 60 minutes.
- f. WHEN one hour has elapsed, START the Pressure Envelope Emergency Ventilation System as follows:
- 1) RESET CBI Train A and Train B (MB2).
- 2) Go To Section 4.10 for Train A, or Section 4.11 for Train B and START Emergency Ventilation System in full recirculated filtered air or recirculated outside filtered air, as determined by the SM/US.
4.13.2 To respond to an automatic actuation of Control Building isolation, PERFORM the following:
- a. To verify all pressure boundary doors, closed and sealed, PERFORM the following:
- CLOSE and DOG the following Control Building pressure boundary doors:
- CB west 47'6 " (C-47-1A)
- CB east 64'6" (C-64-IB)
- VERIFY the following Control Building pressure boundary doors, closed:
- CB west 47'6" (C-47-1)
- CB north 64'6" chiller room door, (C-64-4)
- CB north 64'6" chiller room door, (C-64-5)
- CB east 49'6" (C-49-1)
- b. PLACE 3HVC-FN6, "KITCHEN EXH FAN, to "OFF" (VP1):
- c. IF Purge System is in service Refer To step 4.12.2 and SHUTDOWN Control Building Purge System.
Level of Use ~O 34 l e U lK Rev. 019-01 ContinuousI2o ACT REVIW
'.-' 92of1505
- d. WHEN 60 seconds have elapsed, OBSERVE the following:
- 1) The following "CNTL RM AIR BANK ISOL VV," open:
- 3HVC*SOV74A
- 3HVC*SOV74B
- 2) Control Room pressure is maintained greater than or equal to 1/8 inch WG as indicated on 3HVC-PDI 113 "CNTL RM AP" (VP1).
NOTE
- 1. 3HVC*PCV68A and 3HVC*PCV68B, pressure control valves, will control outlet flow for 60 minutes.
- 2. If actual release of radiation to outside has occurred or will occur, the Emergency Ventilation System must be placed in the pressurization mode (recirculated outside filtered air).
- e. WHEN one hour has elapsed, START the Pressure Envelope Emergency Ventilation System as follows:
- 1) RESET CB1 Train A and Train B (MB2).
- 2) Go To Section 4.10 for Train A, or Section 4.11 for Train B, and START Emergency Ventilation System in full recirculated filtered air or recirculated outside filtered air, as determined by the SM/US.
4.13.3 To restoration from a Control Building isolation, PERFORM the following:
- a. RESET CBI Train A and Train B (MB2).
- b. CLOSE the following "CNTL RM AIR BANK ISOL V" (VP1):
- 3HVC*SOV74A
- 3HVC*SOV74B OP 3314F Level of Use Cnnuu -UK AC T. RE~EW Rev. 019-01 Continuous93 of 93oI5 150
- c. OPEN the following (VP1):
- 1) "KITCHEN EXH ISOE':
- 3HVC*AOV20
- 3HVC*AOV21
- 2) "OUTSIDE AIR ISOL:
- 3HVC*AOV25
- 3HVC*AOV26
- 3) "NORMAL SPLY DMPR":
- 31VC*AOD27A
- 3HVC*AOD27B
- d. Refer To Section 4.14 and RECHARGE air banks.
- e. RESTORE Control Building Ventilation System as determined by the SM/US.
- f. Refer To OP 3315E, "Technical Support Center Ventilation,"
and RESTORE Technical Support Center Ventilation System to "NORMAL."
-End of Section 4.13 -
OP 3314F Level of Use T~INK STOP AT Et~W Rev. 019-01 Continuous .. 94 of 150 94oI5
I (.0 MANUAL ACTUATION FROM CONTROL BOARIDAK ar I rESTA 5 TEST T I YPASS BYPASS A(lA~ ~ SNRAY ATIMD~ SONA CACTUATION ISOLATIGI PHASE A rOCe170 COTOLL tILo~ MDCNRLCW ONRL) (W O T T A __ MA_ __L 3+^D D+ZE 4 (34( ) I SIFEr (MOTE 9) IB OTHERS BY N.E.5. (.TE Is) MANUAL MANUAL. MANUAL a SETRST RlESE (NOT! Sl 9) (NOTES*S9) OTESS9-) f LOP COA LOP A LE 3FNEiERATO.R RGENE7AToR BY
~OAO 5EQi.EN"IkQ LAO a CE.RN SOWR ZI71 LSK-24-S4 a9OtSK-Z49 I4 (NO TE 15)I9.
lusco.25212-287Z3 l NUSCO252?873 Z l) IPE 1) (-20 It-OT 5_ 60 (O IO NOTE13* ERENEC St RVTE 6AT) R O4AT r CNTAMEINE OTAI5ENT CSkjTROL INTAKE UCT
<6v DIOSL ICtEI POTEi6 3), IPTE13 ITRISOLATION ISOLATIONd R A ISOLATION F-ERGE 1 (POTE 15-275 IEMERG.GEI1E (IU4O5E t-Z72 Ic.n I IL0OAOEQuEi.4Cr=tIO" SEQMEMCE I1ti?9-LSK.2449.4 1 ?,SZ-.
- REOUIREMENTS (IF SEQUENCINGIS NECESSARY)ARE SPECIFIED 16. THE 2 OUT OF 4 COINCIDENCE, MEMORIES, AND 'OR LOGIC ARE DUPLICATED WITHIN LEM IO LVE IN PAaLLEL NITH THEASSOCIATED STEAMLINE STOPVALVE.
EACH TRAIN. SEPARATE OUTPUT RELAYS CONTAINMENT SPRAY, ONE CUTOUT RELAY ARE ALSO PROVIDED. ro MINIMIZE FALSE SHOULD START THE PUMPS WHILE ANOTHER IN THE CONTROLROOMFt7 EACH SEAM LINE STOP VALVE TO SHOULD OPEN THE SYSTEM VALVES. is RJLLY CLOSEOOR FULLY OPEN. 17. THE CONTAINMENT SPRAY SYSTEM SHOULD %YED ANE)SEQUJENCED IF THEEMERGENCY BE INDEPENDENT OF THE SAFETY DIESEL PONR CAPABILITY INJECTION SYSTEM. IF THE CONDITIONS OF NOTE 13 APPLY. THE SEQUENCE )AD WITH ALL SYSTEG STARTINOG THE TIME OELAY(S). IF USED INTERLOCK SHOULD BE -SUCH THAT SPRAY WILL START WITHIN THE REQUIRED JMSTARTING TIlE IEQJIROME,4TS FOREACMSYSTEM. TIME INDEPENDENT OF THE SAFETY INJECTION SIGNAL STATUS. 0L BOARD. OPERATING EITHER C114TO1TL WILL ACTUATE. 18. TWO MOMENTARY CONTROLS. ONE ON MAIN CONTROL BOARD 2 AND THE s FI.CTIONS ARE NOT WITHIN THE FUNCTIONAL DESIGN SCOPEOF OTHER ON MAIN VENTILATION PANEL VPI. OPERATING EITHER CONTROL IT ONLY SH ON THIS SHEETAS THE FUNCTIONSARE BUILT IN THE WILL ACTUATE. 12179-247 2.01 1-001-008 aRP*.* 5yg "IFSAR FIGURE" Q.AI
; _,CAD .7- .
I~ I l ~ I . 1-1-4-- . I ! I IT I l I,~~~ ,,,.I ~ , . . W46 9"- AILI/vT0AI. '14U'!~ I'/ d, - ' .- I
-d1CDCN 0&43-00-0682-SBIN581~%
IV
- , I . .
I i i 4A l I I I s
- 1. . I - I NU. UI
[O104 BUILT I 4& JASIKORP PER DCNDCR-U3-S-0185-94 DM13-S-0292-92 R RUI ICJCND IVOEDJ Din"S.IO ChLs SHEET 8 I 1 I I .- I -./ I- ... RtYISIIRiS I - I -H I1- -OYIAP
-II 4~~
_ __ _ _ _ _ DONOT SCA E
- W9 0 . _51 _ H 0 I a 4 I 3 I 25212-39001 SH4009
cONTAWMENT PRESSURE AQ X I L I AP,,Y FEiOWATE i PUMPS CSHEET IS) NoT (tNoT FULL 12) SOPEN (NOTE
'ES:
I. TWOM ETARY CNTRIK.S ONTHE NTlOL BOARD. THEMNUAAL OPERATINGElTHER CONTROLWILL ACTUATE.
- 2. SPRAYACTUATIGN CO4SISTS OPF GP MOMENTARY CONTROLS. 10. SAFZlY 1ECTION SEQUECE REUJIRD ASSOIATED CITRQ.S AREOPERATED ACTUJATION WILL =CCI GLy IFrToo SYA NLEAR C ENMGYSYSTEMS.
SLITNEOuSL.LUI ONECONTROL PERLOO ONTHECNTROL BOASD. II* ALSOCLOSES THEBYPASS VALVE IN P 0 3. 12. LICKTS SHOJLDBE PROVIDED IN THE(
- 4. CONTAINENT PRSSJjm STABLES FORSPRAY ACTUATION OE41001ZE TO ACTUATE). ARE EPEIGZE TO ACTUATE (OTHER BISTAGLES AVE INDICATE WHEN THEVALVEIS FULLY I CONTAINMENT AIR RECIRC FAN IS TRIPPED TIEACTUATION MAY E OELAYED T. AND!
BY CDA SIGNAL. IS LESS THANTIE TOTAL LOADWITH A S. COPONETS AREALL 11o1VIlUALLY SEALEDIN (LATCHED) MAYNOTExCEED CAUSE TMESECGWENTS TO RETURN SOTffAT LOSSOF THEACTUATION SIGfNAL WILL NOT THE MAXIPJ4 STARTII TO THECGOITIGN ffLh PRIOR TO THEADVENT 14. TWOCONTROLSS 9, SERVICE WATERSYSTOI ISQLATION Is USEDONLYIF REQUIED. OF TIE ACTUATIOtNSIGNAL. GdTt DNTL BOARD. S. THEtDDATMAUAL RESETCt3NSISTS OF TWONCt 15. SOWENGINEERED SAFEQJRAsS FUICTliC ENTARY CONTROLS ONTIE CONTRIOL B NJCLEAR eNGI' SYSTE4s 9JT ONLY
- 9. SUPLI D BYT 0EtS. .ROKE FOREAGI TRAIN. SH SUFPL IED EGIJ I PiMNT.
I WOTEs I k.4 As-S Is 10
. 7 a
Form ES-401-6 ES-401 Written Examination Question Worksheet Question # 77 Examination Outline Cross-reference: 2 Tier # Group# 2 K/A # 075.A4.01 Importance Rating 3.2 Proposed Question: Initial Conditions:
- The plant is at 100% power.
and stopped the "A"
- The crew has just started the "C" Service Water Pump 3SWP*PlC Service Water Pump on the "A" Train. PEO still selected to "A-Lead/C-Follow", since the
- At bus 34C, the "Lead/Follow" switch is switch.
has NOT yet been dispatched to operate the "Lead" pump on the "B" Train.
- The "B" Service Water Pump is running, and selected as the
- All Service Water Pumps are in "AUTO".
operates as designed. The RO is monitoring the A loss of offsite power occurs, and all equipment loads on the emergency busses. Main Boards to verify that the proper equipment automatically on Main Board 1? What Service Water Pumps will the RO see starting Pumps will start. A. The "A", "B", "C", and "D" Service Water Pumps will start. B. Only the "A", "B", and "C" Service Water will start. C. Only the "A" and "B" Service Water Pumps will start. D. Only the "B" and "C" Service Water Pumps Proposed Answer: C on an LOP signal. "A" is (Optional): "C" is correct, since only the LEAD Service Water Pumps start Explanation pressure signal OR a CDA signal on the affected train. plausible, since both of the pumps start on a low discharge plausible, since the "A" and "B" pumps are the LEAD signal. "B" is The LOP signal blocks the low pressure start were running at the time of the LOP. pumps. "D" is plausible, since the "B" and "C" pumps 24-9.4J (Attach if not previously provided) Technical Reference(s): LSK 9-1 OA, 9-1 OH, 24-9.4A, examination: None Proposed references to be provided to applicants during Service Water System components (As available) Learning MC-05714 Describe the operation of the following Objective: controls and interlocks: Service Water Pumps.... Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement I
-A1i...
SOUR IMON ITOR CONDITION CONTROL AC RESULTANT MTO BRAERRCIDBUS 3 . (+~~ 8U aLobo I WPXPI8 , CONT.PWN. LOCAL AX TNO Ali SERVICE WATER 74 COPTROL'POWER- )4 1 PUMP (DPTA~>BE /: SWS N LOCAL CONTROL E3 SOURCE SIMILAR TO START
\3WP)PIR el SDURCER3d'ON rMANUAL START / "
LSK-9' I(Ai BLOCK 1281193SHB. 8)
~3SWPjPI8 - F. i
!lI Dl --
> v!
REMOTE SOURCE SIMILAR TO (NOTE4) SEQUENCED SAFEGUARD SOURCE'280ON f - SIONAL'TRAIN B -- I I LSK-9-IOH FOBLEAD PUMP J (28193 SH. ) SWMBI AD
-e iC W L3SWP PIS C LEAD SERVAICEWASTERNPL4 AND o-GROUP 33
< 3SWP#PIB \ DlCONNECTOR SIMILARTOL
-1 ESF STATUS SOURCE SIMILAR TO (NOTE5) SEQUENCED SAFEGUARD \PULL TO LCK / IL CONNECTOR 3 ON I YELLOW M8 LS-9-ION, 3SWP*P1D SOURCE36HN1SIONAL TRAINB SIMILAR O LSK-9-IOH ' FORFOLLOW PUMP (28193 SH. a ) -
(28193 SH.a)
.3SWP PID S kSL2 SERVICE WATERPUMP
\_ VSTOPPED, OR
- ;;v SOURCE SIMILAR TO / SWP PID 3SWP*U31B FOLLO - \!t/- __Iml
.SOURCE 27'blt RUNNINGSIGNAL \3SWPiIPID LEAD 34 D LSK-9-IOH BYPASS / sswt\
(28193 SH11 )I - C "_AM ),
/ SERVICEWATERPOMP I I -. f 1 3WP* PIS WATERPIIMP ......
J_ AND SERVICE P527.B .... NEADER- B DISCHARGE PRESSURELOUO STAR J (GSW*P*PI . __ V SUOURCE 0* SIMILARTO LS FEEG SOURCE 100ON B LSK-9-IONUSPWR EA F" (28193H. 8a) LSK-24l-7A (NOTE3) BUS3'ID BUSDIFF. 6HO LOCKOUT RELAY lAPPENI Rl I28712 SH1) RESE LSK-9-108 8 6E MOTOR NUCLEAR SAFETYRELATED OA. CATT. Tr.I o
,t 28193 SR 2 )~A EE SERVICEWATER PUMP (TRAINB) (NOTE I) I - I - _,_ I I I I NOTES:
- 1. CONTROL FORSERVICE WATERPUMP3SWP PIB SHOWN.3SWP*PID SIMILAR. 4. THESEQUENCED SAFEGUARD SIGNALIS COMPOSED OF THE . . . .
CONTROL FORPUMPS 3SWP*PIA0C SHOWN ON LSK-9 -IOH (28119358 . FOILOWING COMBINATIONOF SAFFIIARD SIGNAL: I I I I I I. 3SOVP~PIAA . LOPI 20 SEC. SW P~lI
,SWR* FIB. I
..MODE-THENLLPI 2. SEC ESCUTCH EON
.I.: II -
-.1-1I el THE MILLSTONEPOINTCOMPANY NORTHEAST
- -- T-T-I.i'ITLE MIlLSTONE NUCLEARPOWERSTATION-UNIT NO.3 UTILITIES
- 2. INDICATIIIGLIGHTSAT MAINCONTROL BOARDAND SWITCHGEAR AREONWHEN CONTROL IS FROMMAINCONTROL BOARD. INDICATINGLIGHTSAREONLYON 3SWP*PIB&PID 35WP*P18 35WPfP B -. . - Iti LOGICDIAGRAM I
AT SWITCHGEAR WHENCONTROL IS FROMSWITCHGEAR. S. SEQUENCED SAFEGUARD SIGNALTO SERVICqYWTER SYSTEM A FOLLOW PUMPIS DELAYED BEHIND -LED T REMOTE 1 - D . . .
- 3. SOURCE DEVICECOMMON TO ALL LOADSONBUS34D. SEQUENCED SAFEGUARD SIGNALTO 0 F I START LOCA
- I LEADPUMP OCL LWRR- I - PIL- DATo E~i
- O-* .00.J3 III 10 4
0m Zj- I .1 -I T. Lw_X E&ZR
,DWG3 ,.T.
EC
-rJur, !rA
- 1 I I II, j-"o 2521Z-281935SEL1 STONEAWEBSTERENGINEERING BOSTON MASS>.
CORPORATION I4NTAINWD SPRING REII MAINTAINED I - II ~- IP. W 0 VV nWtU-L'V V - Mf NW
'I 61j70 -
I & II - aRtt
-'..I 0-fIFlA P .
RESULTANT MONITOR COIITROL ACTI1ON ACTION CONTROL RESULTAHT MONI TOR ESCUTCHEOH NOTES: 3SWPA PIA&PIC 355WPIA 3SWP#PIA i I I II I I. CONTROL FORSERVICEWATER PUMP3SWP*PIA SHOWN, 3SMP* PIC SIMILAR. A - LEAD AUTO REtlOTE CONTROL FORPUMPSSSWP*PIN AND D SHOWONLSK-9 4OA. (28193 S1RI) C- FOLLOW STOPV STAR LOCAL l . . . . . PUMP I ESFATRAIN DESIGNATION THE MILLSTONEPOINTCOMPANY 1 jPTA 3SWP 35W?P PIC 3sWP*PIB A B A-FOLLOW C LEAD PULLTO LOCK o
-N-I-I I-N-I-I-I FIT 1 MILLSTONENUCLEAR M
NORTHEAST UTILITIES POWERSTATION -UNIT NO.3 3SWP PID LOGIC DIAGRAM MAINTAINED SPRINGRETURN MAINTAINED -N-I-I I-N-I-I-I SERVICE wATER SYSTE4
- 2. INDICATINGLIGHTSAT MAINCONTROL BOAROAND SWITCNGEARAREON WHER CONTROL IS FROM MAIN CONTROL BOARD. INDICATINGLIGHTSAREONLYON - IA-AT SWITCHGEAR WHEN CONTROL IS FROMSWITCHGEAR.LOCAL/REE TRASFERSW IS EYLOCKD-EMOVABLEINBOT POSITIONS. -- ------------ i I 131"qI INCOA IN 1 INCOR?
O A E&OCR*C-
.v MG TY
~Z .fr __ I , _ . _
- 3. SOURCE DEVICECOIltOI TO ALL DOADS ONBUS34C. I NOTE: CAD 1 12 02147.NEC-C782
,DaWING WDATE A. , i -.. IA25212-2 SKIS mln-t
--- -- r---- ----- r >ssI-----
- 4. FORADDITIONALSEQUENCED SAFEGUARD SIGNALINFORMATION,REFERTOLSR-9-IOA(28M1S3SHlINOTE4. 1 MANUAL REVISIONS I TOTHISDOCUMNT _& _I _S _ _ _. _ A f "" f" i
8STOSINA MAS.
- 5. SEQUENCED SAFEGUARD SIGNALTO FOLLOW PUMPIS DELAYED BEHINDSIGNALTO LEADPUHP. I EN AS--UILTAREP-O-IBIIE-. I I SW I DWG. NO. 12179- LSK-9-loH
i' ( f ALL S WITCHGEARLOADS C080IT IONS STRIPPED FROMEIERGENCY Bus I-oTo-- -- -- COA RECIRC. NODE ITEN LOP SIS RECIRC. MODE TOEN LOP BTDBEV - tDO OR CDA AND (NOTE 0)
*DP 650 660 SIS OR SIS AND LOP
-O--0 --- (:)--- - -'I[HI 2-EfE L- - -
- --(C) -()- - --(- --- - - --(:DS )-(R)_--
LAP ONLY
-(N- 7)NO- I -
(NOTE 7) vINCII Al (NoTE 91 (NOTE 91 ii -1 Is----+ Is 120 I205 2i2'255'30 L I p I 6- an I~. e I I A_ I _l l
~ l *_
jIDIESEL J DIESEL NO ID 200 2IIO GENERATORGENERATOR TIME IN SECONDS EOUIPMFENT MANUAL STARI IREAKERCLOSE START BLOCE (NOTE 3) S IGNA L REI40YED (NOTE A) LE 0 CHARGINGPUNPS(Hi PRE! is sI) LSK-2-S-i .E (2a723 SH 51 ($0 COHr. AIR RECIRC. FAN rHIA LStK-2i4-.ii (20723 SH 031 I. SEOUEICEDSTARTSARE INITIATED 51 THE O SE0UENCESTART (NOYE I) SEQUENCER FORSTEP LOADING OF EOERSEN Cy DUENCHSPRAYPRTIP (I)CORT. RECIRC. POMP PlO BSAIDURNING LOPCONSITIRNS. LS1-.2S-9.11F128723 SH A \ /LS K-2A-9.40 128723 ON TOT 2. MANUALSTATS ARE INITIATED ST THE OP] O MANUAL S7ART (NOTE Z) ERA7TO SAFErY INJECTION PONP
)LSE-24_9 5 G 2Y723 537 S I CONI. RECIRC. PUMPPIC LSK-24-1.i I (27304101EQAIPMONT SH AS PLANT CONDITIONSREOUIRE THENAFTE REiOYEA E8723 NV)
MANUAL STORY BL0CK DISMAL 1 ORtING LOP CONDITIONS. R THE RNR PUMP (LO PRESSSi) PRESSURIZERHEATER 3. WHEN THE DIESEL GENERATOR B- cAKER CLOSES 4 LSX-24-9 4n(2R723 SH I I) 12 ] LSK-24-9.4S 126723 SH 17) (OSECI)LL SAFETY RELATED MO.OR CON TROL CENTERS ON TOE EMERGENCYBUS WILL BE POWERED. SERVICE WATER PUMP(PIA PEL POOLCOOLINGPUMP ALL APPLICABLE SWIICHGEAR LOADS WILL BE STARTED LSK-24-9.'4J (2e723 S. A [1:3_ LSK.-24-9,SS28723SH 17) IN ACCORDANCEWITH SEQUENCE CHNRT.
- 4. THE MANUAL STARTSARE ADMINISTRATIOEL' SERVICE WATERPUMP(P11 NON CAT. I DIESEL GEN.tSSOC.EOUIP. CONTROLLED AND SOME REQUIREBLOCKING 51GORAL 0J LSK-24-9.4J (28723 S. S I) LSK-24-9.4U(26723 ST19 RESTS IN ADOITION TO THE AUTOMATICEl ;UINNENT ml MANUAL DTARY SLOCE SDGiAL REMOVED, REACTORPUNT CONPONE111 NSTRUHENT AIR CON?. CIS (TRAIH R( s. bHEHEYERA SIS OR COA OCWORS W1THWI0 A LOP ALL CIOLING PUMP (PIA )
6 IS .LSE-24-9.NP126723 SH*41 AREUIREDEQUIPMENTIS ENERCIZEDWITTD: JT SEOU£NCING LSR-2N-9.4E (28723 S. I, . DELAYS.EXCEPT CONTAINMENTRECIRC.PuOPALWAYSSEQUENCED. 60 REACTORPLANT COMPONENT COOLINGPUMP(PIC ) LSK-24-V.4K (20723 S. I 0) IO AUIILIAOt BUILDING FILTER
'IX(028 SISTEN EXHAUSI FAR aLSx-2i-_9.47t(2723 SH 01)
- 6. SEE LSK-24 9 4UE2672S SH lI.0V28723 A OH20DOW26723 SH 211A1 723 SH 221 VONRADDITIONAL EC0UbMENT CONTROLLED BY THE SEOUENCER NOT SHOWN ON LC)A^DING CHART NUCLEAR
.COT.
SAFETY RELATED I QRA.o CONTROLNUILVIN4 CHILLI AFUEL WILDIHG 7. TrE GHARGlit PUEPSAREtNOT TRIPPED F O TNHE WS LStX-24-S.4L. (28723 SH 1 IT IELHAUILDING AND DO NOT RECEIVEA START SIGHAL OPT *ERA LOP ONLY. IJ OAxST PAR THE CHARGINGPUMP THATMAS RURNINGWILL NESTART 0 O AUX FEED PuMP LSK-24-9.4H (28723 SH I 2) LSE-24-U.'IT128723 0.4081 a. WFM7PAWN IS 8fOTUFS CONTROl RODDRIMEMECHANISM AF TER CLOSURE OF THE DIESEL GENERA ITOR CIRCUIT BREAKER. _ ,8,0fl-P 04,kS1541N1100018 _ _ _ C GOOLINEGSYSTEUPFA 311VU-PR? THE SEQUENCERHAS A 1.0 SEC. TIME IDELAY BEFORE INITIA7INC
,SI-24..4 1T(28723 SH 23) EQUIPMENT STARTING. BECAUSE OF ,. I(1.0 SEC (APPROX.) INORTHEASTNUJLEAR E*OIlOCy cONpANy DAMPER OPENING TIME. THE ACTUAL ENERGIZATION OF THE
>4
, ,N NOOTCAUTILITIES MOTOR OCCURS APPROXIMATELY II SEC5 C. AFTER DC BREAKER (Ii IRS CODTINN D COTPUTERROOM CLOSURE.IF FAN A FAILS TO START FOR A LOP ONLY OR A _O I*L)0ETONEYOIGOE*R AIRCOHOITCIAINGDUHIT 3HYCP ACT2A NYSTOSTAYIIO UHII1NOY LSA 24-S-.6Z12723 SH 24) LOP COINCIDENT WITH A CDA/SIS. FAN B WILL START AT 30 SECS. AFTER OG BREAKER CLOUSURE. LOGICDIALRAU SEOUEHCER LOADINGCHARE EMERGENCY GENERATOR LOADSEQUENUCR TIIHL
- 9. STARTS ONLY IF IA PUMP FAILS.
10 I.-,81 7.4 2.1 0001811 R.
£1 A~l.
~~ D172F
, 2/ ~ ~Ul 1 1LA 1.w SEI
- NOIE:
I NANUAL CAD : 10 ISDOCUET I tEkLVISNIDNSTO 9 7IlK0( 56-2)80212 R., .Sp~ NX iJO 2-1
- c. .* 5 A
, I 25Z2-28723 SN.I STON1BREUTER PIONIREERING CORPONATION I EN AS-BUILTAREPEOHIBfTED. !R 'r NTOM, MASS IS EW DWG1. No. I 2179-LSK-24-9.4A
------ 1I
SOURCE, SOURCE IT I NY CVUlD!TION CONTROL ACTION RESNITAR RESULTAOT SW INhIBIT RSLATI i;ONITOR WPAP APIA' hBYPASRUNNING SGNAL SIMILARTO PIC D . ~FOR 35WP{ PIC\ 3SWPXPAPE3Ct OR LSE-24-U.48 AND - 1SEUNCE L LK- 9L-10H (82 \ = EUNCRI TS U N (28193 SHa I ANDi-03S(Vp4 PIA 3SWPPiA SEQUENCER S SAFECGUARD SIGNAL 28 LSK-9-IOH O3Rl -- IP (28193 SHMB) MANUALRIP29 LSK-T(IOA BLCK SH893.M) 3SWP*PlC SAFEGUARD SIGNAL SWP PICSIMILAR TO (PIA) ( 5PPC>ADxSEQUENCED NESCA ADMANUAL TRIP SSIMILARTO I-PIA) NYPT ASSERUNNIN SIGNAL 2 S--O LSK-24-9.4B LOSSOF EMERGENCY 31 .05S POWER (2873 5121 ELAYOPERATED BYASRN=)GiIA ~ LSK-9_IOH 1(2893 SH.8) AN ' 6 I \_>.(PIC) OR \W 9 -1 SEQUENCED MANUAL El SAFEGUARD TRIP BLOCK A_____ AND SIGNAL L F35SWP*PiA MANUAL START LSK-9-IOM
--- \ LSK-24-9.48 BLOCK (28193 SH.S 35WP*PIC
-F-T.D7. _ MANUAL START fSIMILAR TO(PIA)
RIP SIGNAL I BLOCK LLO'RPT LSK-24-9.4C /A: = -/C\ [1ll SIlONA L XE5w
& (ll28TD-723 SH.3 IGAPOWER
,,O SS OF SIGNALs
-!I---
L a LK-9-1 .0J.. LK2-.C;llSE4UENCED SAFEOUARD AND IA)
\ _ SINAL MODE ) Dl IIUI^UAL TRIPBLOCK SIGNAL IV \
LOPLEARIPT B8US'34C LOSS OF POWER SIGNAL SIMILAR TO (PIA) LSK-24-9.4B (EQUIPMENT Y RELATED III: V' EFl- F I AT-.- EA -.EIP-T.OS IT SlA A. LSK-24-9.4D DPPEMI a TC02NORTHEAST UPDATE NUCLEAR fNERGYCwMPANY _ # 1M1 NO CHANGE NOCHANGE DAT LOG I IAORAM NORTHEAST UTILITIES TIL JIILLDTONE NUCLEARPOWERSTATION -UNIT NO.3 NOTES: 1. LOGICFORSEQUENCER A (TRAINA) SHOWN. C EMERGENCY GENERATOR LOADSEQUENCE LOGICFORSEQUENCER B (TRAIN B) SIMILAR. SEQUENCER SIGNALSFORSERVICEWATERPUMPS 3SWP#VPIA AND 3SWP+PIC BR N CHANGE I4 -GENERAL l REDRAWN SAAB. l AHOPIIN T l25212-28723 DM8 T U#RL 550. S'PSISI A. JA~ T M HVA AMBOSTON, SN &WEBSTER ENGINEERING MASS. CORPORATION _ I I -SW DWG.NO. 12179-LSK 2 4-9.4J (
Form ES-401-6 ES-401 Written Examination Question Worksheet Question # 78 Examination Outline Cross-reference: Tier # 2 Group # 2 K/A # 079.2.4.11 Importance Rating 3.6 Proposed Question: and the crew enters AOP With the plant at 100% power, instrument air pressure starts decreasing, the plant. 3562 "Loss of Instrument Air". The crew is NOT required to trip crew to monitor for potential Which level/DP indicators will AOP 3562 direct or caution the inaccurate operation? A. Condenser Hotwell level and traveling screen DP. compartment DP. B. Spent Fuel Pool level and Turbine Lube Oil Bowser filter Oil / H2 DP. C. Low Level Waste Drain Tank level and Main Generator Seal D. PRT level and Emergency Diesel crankcase vacuum DP. Proposed Answer: A Explanation (Optional): level instruments/indicators are pneumatic "A" is correct, per AOP 3562 CAUTION prior to step 7 "All condenser... air", and prior to step 10 "Travelling screen differential and do not provide accurate indication on a loss of instrument accurate indication on a loss of instrument air". pressure instruments/indicators are pneumatic and do not provide AOP 3562, Steps 7 and 10 (Attach if not previously provided) Technical Reference(s): None Proposed references to be provided to applicants during examination: (As available) Learning MC-03939 Describe the major action categories contained within Objective: AOP 3562. Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
'vI I
PACTION/EXPECTED RRESPONSE NOT OBTAINED I
- 6. Verify Train A and B Chilled OPEN the Train A and Train B Water Ctmt Header Isolation RPCCW supply and return Valves - OPEN isolation valves to chilled water.
- Check inlet valves 3CDS*CTV38A 3CDS*CTV91A 3CDS*CTV38B 3CDS*CTV91B
- Check outlet valves 3CDS*CTV39A 3CDS*CTV40A 3CDS*CTV39B 3CDS*CTV40B CAUTION All condenser temperature and level instruments/indicators are pneumatic and do not provide accurate indication on a loss of instrument air.
D~ ~ ~~~~~~oal Close normal makeupxx~~xx~~xx~~xx~~x
- 7. Monitor Condenser Hotwell Level Locally Close normal makeup
- NORMAL isolation valve (3CNS-V2).
- 8. Monitor VCT Level - Perform the following:
NORMAL
- a. OPEN RWST to charging isolation valves.
- b. CLOSE VCT to charging isolation valves.
- 9. Control RCS Pressure
- a. Energize PZR heaters or use a. Use one PZR PORV to normal spray valves as depressurize if necessary.
necessary
I LOSS OF INSTRUMENT AIR AOP 3562 Page 8 Rev. 4 STEP ACTION/EXPECTED RESPONS REPNEOTBAID l CQ A U T I 0 Ni
- Traveling screen differential pressure instruments/indicators are pneumatic and do not provide accurate indication on a loss of instrument air.
- Traveling screen differential pressure circulating water pump trip relays are pneumatic and will not operate to trip the pump if the situation requires.
- 10. Increase Surveillance Of Intake Structures
- Locally Place control switches for traveling screens to SLOW
- 11. Verify RHR Alignment
- a. Check RHR Train A or B - a. Proceed to step 12.
IN COOLDOWN MODE
- b. Check the operating RHR b. OPEN the operating RHR HX RPCCW outlet flow heat exchanger "RPCCW DIS" control valve(s) - OPEN valve(s).
- 3CCP*FV66A
- 3CCP*FV66B
- c. Check an increase in RCS c. Proceed to step 12. and, cooldown rate - DESIRED IF increased RCS cooldown rate becomes necessary, THEN Perform steps 11.d. and 11.e.
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 79 Tier # 2 Group # 2 K/A # 086.K4.04 Importance Rating 3.5 Proposed Question: A deep seated fire started in the computer room, and the Halon system automatically actuated 10 minutes ago. The crew desires to enter the computer room in order to inspect for damage. How can the crew enter the computer room in accordance with the Halon related precautions of OP 3341 B, "Fire Protection Halon System"? A. The crew commences ventilating the computer room, and three men enter the area 15 minutes later. B. One man enters the computer room wearing a canister mask. C. Two men enter the computer room, with each of them wearing a self contained breathing apparatus. D. Three men enter the computer room with each of them wearing a filter type mask. Proposed Answer: C Explanation (Optional): seated "A" is wrong since the space should be kept sealed for 30 to 60 minutes to ensure extinguishment of a deep the presence of heat, Halon fire, and the fire must be completely extinguished prior to ventilating the area. In decomposes and forms acrid byproducts, posing a hazard to personnel. Therefore, the area must not be entered alone ("B" wrong), and SCBAs ("C" is correct and "D" wrong) must be worn, since filter type masks only remove particulates, and Halon is a gas. Technical Reference(s): OP 3341B, section 3 (Precautions) (Attach if not previously provided) Proposed references to be provided to applicants during examination: None (As Learning MC-04565 Describe the major administrative or procedural precautions and limitations available) Objective: placed on the operation of the Halon Fire Protection System, including the basis for each. Question Source: Bank # 74349 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.5, 41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
2.2 Definitions 2.2.1 IRR - Instrument Rack Room
- 3. PRECAUTIONS 3.1 If an individual is affected by Halon 1301, the individual must be immediately moved to fresh air and medical attention must be received.
3.2 For a deep seated fire, the space should be kept sealed for 30 to 60 minutes. 3.3 Fire must be completely extinguished prior to ventilating the area. 3.4 Halon 1301 and combustion products decompose and form a sharp acrid aroma. Upon entering a suspected halon filled room: 0 Self contained breathing apparatus must be used.
- A filter type or canister mask must not be used.
0 The area must not entered alone. 3.5 All evacuation alarms must be observed. 3.6 With the entire Fire Protection Halon System out of service, a fire watch must be stationed with portable fire extinguishers while cutting, welding, or burning is in progress. 3.7 Disassembly of system piping outside the Control Room results in a Control Room Boundary breach. The appropriate system isolation valve may be used to isolate and prevent the breach. k 11
'1q OP 3341B Level of Use .J. ".,"W A$,W Rev. 005-02 Continuous 3 of 16 I
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 80 Tier # 2 Group # 3 K/A # 008.A4.05 Importance Rating 2.5 Proposed Question: Initial Conditions:
- The plant is being cooled down on RHR to MODE 5 per the ACTI'ON of LCO 3.7.3.
- Only the "A" Train of RPCCW is available, and it is supplying the "A" RHR heat exchanger.
Current Conditions:
- The problem with the "B" Train of RPCCW has been corrected.
- The "A" RPCCW Train is supplying RHR.
- The RO is preparing to start the "B" RPCCW pump per OP 3330AI "Reactor Plant Component Cooling Water".
What is the minimum preferred load that should be available for startihig the "B" RPCCW Pump, and how will the other major heat loads be divided between the RPCC W Trains? A. 2000 gpm of load should be available, and the "B" RPCCW train N vill be used to supply both the fuel pool cooler and a CDS chiller. B. 2000 gpm of load should be available, and the "B" RPCCW train ivill supply the fuel pool cooler while the "A" Train will continue to supply the CDS chiller C. 3000 gpm of load should be available, and the "B" RPCCW train )Aill be used to supply both the fuel pool cooler and a CDS chiller. D. 3000 gpm of load should be available, and the "B" RPCCW train iwvill supply the fuel pool cooler while the "A" Train will continue to supply the CDS chiller Proposed Answer: A Explanation (Optional): The minimum preferred load for starting an RPCCW pump is 2000 gpm ("C" and " D" wrong). The option for sharing loads between RPCCW trains is one train supplying RHR while the other tr,ain supplies the other two major loads ("A" correct, "B" wrong). Technical Reference(s): OP 3330A, Note prior to step 4.2.1 (Attac]Eiif not previously provided) OP 3330A, step 4.16.4 OP 3330A, Attachment I Proposed references to be provided to applicants during examination: None . 41 of 46 NUREG-1021, Revision 8, Supplement 1
ES-40 1 Written Examination Question Worksheet Form ES-40 1-6 Learning MC-04154 Describe the operation of the Reactor Plant Component Cooling (As available) Objective: System under the following... Plant Cooldown... Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement I
4.2 Start Train B RPCCW System C This Section Contains EOP Related Material V7 CAUTION V
- 1. A review of the RPCCW System valve lineup status should be conducted prior to start to ensure enough loads will be placed on RPCCW pump B. A minimum load of 1,000 gpm should be available and a load of greater than or equal to 2,000 gpm is preferable.
of any new
- 2. Care must be taken to properly fill and vent all pipingservice components, branches, or loops that will be placed in by this section.
NOTE
- 1. If both trains are to be started at the same time, this section may be performed concurrent with Section 4.1 of this procedure.
- 2. This section assumes associated RPCCW heat exchanger B is in a fresh water layup alignment as specified in OP3326, "Service Water System."
4.2.1 IF the Train B RPCCW will be started with pump C and heat exchanger, Go To Section 4.4. 4.2.2 CHECK 3CCP*V91, RPCCW B pump suction isolation, open. 4.2.3 CHECK the following annunciators not lit:
- MB1CC 2-7A, "RPCCW SURGE TK LEVEL HI"
- MB1C 2-71B, "RPCCW SURGE TK LEVEL LO" 4.2.4 VERIFY 3CCP*TIC32B, "COMPONENT CLG WTR CLR" 0 0 "3CCP*E1B OUTLET TIC," set between 82 F and 88 F.
4.2.5 OPEN 3CCP*V6, RPCCW pump B discharge valve. Level of Use OP 3330A l ContinuousCoEnt n u o u 7Rev. 13 of015-04 86
during this 4.16.3 IF the Boron Recovery System must bebeoperated aligned as specified in ___1A_-_ the RPPCW Svstem should Option 1 of step 4.16.4 (Table - 1). Option 1 or 4.16.4 With the Boron Recovery System shutdown either System Option 2 may be used to supply the remaining RPCCW loads: V CAUTION V The RPCCW System flow to the RHR heat exchanger should be throttled as necessary to prevent exceeding 8,100 gpm on the applicable RPCCW System train. one of the 4.16.5 Refer To OP 3208, "Plant Cooldown," and OPERATE following as necessary (MB2): outlet
- 3CCP*FCV66A, RHR heat exchanger A cooling isolation outlet 3CCP*FCV66B RHR heat exchanger B cooling isolation
-End of Section 4.16 -
Level of Use **30A O
- Mr Rev. 015-04 ContinuousM
Attachment 1 Design Flow Rates Required by the Respective Individual Components Cooled by RPCCW (Sheet 1 of 1) GPM RPCCW NORMAL COMPONENTS REQUIRED BY RPCCW TRAIN DESIGN SUPPLY Reactor Coolant Pump (RCS) 286 A and B Letdown Heat Exchanger (CHS) 1,253* A Excess Letdown Heat Exchanger (CHS) 375 B Seal Water Heat Exchanger (CHS) 375 B Residual Heat Removal Hx (RHR) 6,600 A and/or B Boron Thermal Regeneration Chiller (BTRS) 414 A Fuel Pool Cooler (SFC) 1,800 A and/or B Mechanical Refrigeration Unit (CDS) 2,700 A and/or B Boron Recovery System (BRS) 1,615 B Gaseous Waste System (GWS) 514 B Auxiliary Condensate Cooler 111 A
*This load requires only 300 gpm on the average during plant cooldown.
OP 3330A Level of Use l A i a AB Rev. 015-04 Continuous No r 83 of 86
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 81 Tier # 2 Group # 3 K/A # 041.K4.05 Importance Rating 2.7 Proposed Question: With reactor power at 19% and a plant startup in progress per OP 3203 "Plant Startup", the following sequence of events occurs:
- 1. The BOP operator closes the main generator output breaker.
- 2. Just as the BOP operator starts picking up load on the turbine, the turbine first stage pressure starts increasing on transmitters 3MSS-PT505 and PT506.
What will be the response of the steam dump system to this operation? A. All steam dump valves will remain closed. B. The partially open steam dump valves will throttle closed. C. The partially open steam dump valves will throttle open. D. The 3 cooldown dumps trip fully open, cooling down the RCS to 5537F. Proposed Answer: B Explanation (Optional): Steam dumps are in the steam pressure mode per OP 3203, step 4.3.1, so they are already armed, and responding to Main Steam Pressure Transmitter 3MSS-PT507, which will be decreasing as the turbine draws steam ("B" correct).
"A" is plausible, since in the Tave Mode this would be true. "C" and "D" are plausible, since MSS-PT 505 and 506 are increasing. "D" is also wrong since the trip open feature is only functional in the Tave Mode. "D" is plausible, 0
since P-12 operates to close the dumps at 553 F Technical Reference(s): OP 3203, step 4.3.1 (Attach if not previously provided) Functional Dwg # 10 Proposed references to be provided to applicants during examination: None Learning Objective: MC-05302 Describe the operation of the steam dump system (As available) when in the steam pressure mode of operation... Question Source: Modified Bank # 75481 Parent Attached Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7, 41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
-- - _I
-aeh-_{
UVJ U ES-401 Written Examination Question worksheet Original Question 75481 With reactor power at 19% and a plant startup in progress per OP 3203 "Plant Startup", the following sequence of events occurs:
- 1. The BOP operator closes the main generator output breaker.
- 2. Turbine load has just been increased to close the steam dump valves.
- 3. Turbine first stage pressure transmitter 3MSS-PT506 fails high, with PT505 selected on MB7.
What will be the response of the steam dump system to this failure? A. All steam dump valves will remain in their present position. 0 B. The 3 cooldown dumps fully open only, cooling down the RCS to 553 F, whereafter the 3 cooldown dumps close. 0 all C. All steam dumps fully open, cooling down the RCS to 553 F, whereafter the steam dumps close. D. All steam dumps fully open, resulting in a low PZR pressure reactor trip and safety injection. Answer: A 41 of 46 NUREG-1021, Revision 8, Supplement 1
4.3 Power Increase to 25% NOTE This section is used to load the turbine generator and increase power to approximately 25% following a plant startup or for partial load increases below 25% power. 4.3.1 VERIFY the following conditions:
- Reactor power between 6% and 9%
- Condenser steam dumps in the following configuration:
- MSS-N07, "MODE SEE' switch, in "STM PRESS"
- 3MSS-PK507, "SG PRES CNTE' in "AUTO"
- Steam dump 3MSS-UI 500, "DEMAND" equal to or greater than 6%
- SG NR levels at approximately 50%
- SG bypass level control valves in "AUTO" 1
- One turbine driven FW pump operating with 3FWS-SK509A, "PP A & B MSTR SPEED CNTL' in "MANUAL" NOTE Low pressure turbine warming may already be in progress or completed.
-I--- 4.3.2 Refer To OP 3323A, "Main Turbine," and PREPARE for turbine startup. NOTE Steps 4.3.3 through 4.3.6 may be performed in parallel. 4.3.3 NOTIFY GES the turbine is being prepared for startup. I [!. Ref. 6.3.21 Level of Use M Sw y 4, OP 3203 Rev. 017-01 Continuous 18 of 48
M.EAVI 'RNES LW STEAM LIW.E PRESJSU;PE T AVC. FSs-5 LOOP I LOP'2 LOOPI O4 sLTpa'W.r STEAM~ DLOAR C.O*JTML MOOC 5ELECIC)R SW. IT.MH (NoTE 8) I~1( LiEOB.~~RE PesssBee PRESS URE IBUE I iLLER CANTwmtLBU COUAOLLB.Ep CjLLR I o~Lir.9I I I IcI v7I.L I TIs Ii I I PLANT II I CCff%0LLER I I I - I- I
- -1 I B I I
(&OT5
'I - -
I I I I I NES I I MOULTE 140X.LAE WCWLrE MD AIE I I T"E C<>PI TH LopImELc M OP A~OSMEIC4TIA PEZC TOPIC AiSI,4~q I I IEE 14V REIEFVLERiFVLE ME A I I 3MSP2A 3SSP2B 3S*V2C 3SIVO I I 1. SEA. %Z VDO 15*XI IUk IR'ai~.DM 4~e m i i I I I I i I i I I I I I ON STAMDUP ISPI ITE I I I I I MOFFISTEA DMP IS NOT E FM AVO&MIANDRST AV BYPAS.LL THE CNRL, OOP. N FOR EAC LOPRAIN. LO 3 7'ELO I I FROMOFFRETPRESURETAP ~ThO MM HESNGL FAILURPECRICTEMONp~t I II I PRMSS*PEV2OSABL O PMS-06 O 3MS*V2C 3MSS*PV00 Co. A.MTSHUMD BE PROVIED IN SLT0iI CMNRO ROCKFUOP VACLWV~bALVE VMIN4I.TIE. 0WHENTHALV~E F ~WIS CLOSM OR UTUY OP~R~EN.* w
.bLY O~fO I I TH STAMVLINE PESSURE SIGN&AMLY VE OIINtTE.OSeT BEDIF EHEN FROEMTkCATV*4.VH IS~USED FORTE SOLEA4IN'ETP'dSA UrE 70 PRLOrCrIEWE A4SLINE 1ET'SIUI SHO.VCOAN SEEGM MEETLTMASINGLED POITINR.E CRITERI4ON.A.Y6h
'TNTWO BEW'TJL-AAITB IIGI&I3RO BOATh i M SoUWI PDANThLCD EIAII O WLlS. .I
- 9. HECODESE DUMP VALVNO¶OPrES OPHENREUETALY AN OES UL BEOREBN 2 BEGINS TO OSPENBAK PES ULYBEOEEAND BEG TO ASS TOPEN. ML(I BPSE OR~WTVO PtS mycmHIES. I- ~
MODULATE THE FOLLOWING DUMP VALVES BANK I 3MSS-PV47A, B, C BANK 2 3MSS-PV48A, B. C BANK 3 3MSS-PV49A, B, C (NOTE 9) [3RPS* S IF " SAR FIGURE"I 12179-2472.01 1-00I-010 W gi1gouse Electric Corpuratton ~
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ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 82 Tier # 2 Group # 3 K/A # 045.K5.23 Importance Rating 2.8 Proposed Question: With the plant at 80% power, the crew is preparing to raise power to 100% on the turbine load limiter at 3% per hour. The RO/STA have gathered the following data:
- The expected change in reactivity due to power defect is -350 pcm.
- Rods are in auto with 175 pcm of integral rod worth remaining to be added by Control Bank "D"l.
- It is predicted that Xenon concentration will change from -2550 pmrn to -2700 pmrn over the course of the power change.
- RCS Boron concentration is 550 ppm.
- Burnup is 14000 MWD/MTU.
- RE desires to have rods fully out when 100% power is reached.
Using the curve and nomograph attached to this exam, approximately how much dilution water will have to be added to the RCS during the course of the up-power? A. 3,000 gallons B. 6,000 gallons C. 9,000 gallons D. 12,000 gallons Proposed Answer: B Explanation (Optional): Rods and boron dilution will have to overcome -350 pcm power defect, and -150 pcm. Xenon during the power increase. Rods will add +175 pcm leaving dilution to cover the remaining 325 pcm. Boron Worth is -6.77 pcm/ppm. 325/6.77 = 48 ppm dilution. This requires about 6000 gpm. Technical Reference(s): Curve RE-F-02 (Attach if not previously provided) 3304C, Att. 5 Proposed references to be provided to applicants during examination: Curve RE-F-02, nomograph 3304C, Att. 5 Learning MC-04202 Describe the operation of the Chemical and Volume Control (As available) Objective: System under normal, abnormal, and emergency operating conditions. Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
R(2 2521 2-ER-07 v. 19 (N MP3.08-01 Page I la Differential Boron Worth vs Burnup (HFP, ARO, Eq Xe) And (HZP, ARO, No Xenon) Cycle 8
-5.0 -5.5 E
X -6.0 U C. t -6.5 0 0 C
~! -7.5 S
-8.0
-8.5 6000 8000 10000 12000 14000 16000 18000 20000 0 2000 4000 Core Average Burnup (MWD/MTU) rjux) 7 Preparer/Datex...Y?411 I--
Reviewer:/DateA, Page 1 of 2 Approver/Dat (I 'I) I-V LT-L--,L
25212-ER-01-0X30, Rev. 19 RE-F-02 Page jp.. of ___ MP3-08-01 Differential Boron Worth During Cycle 8 (HFP, ARO, Eq Xe) And (HZP, ARO, No Xe) Core Average HFP HZP Burnup DBW DBW (MWD/MTU) (pcm/ppm) (pcm/ppm) 0 -5.94 -6.36 150 -5.92 -6.44 500 -6.04 -6.43 1000 -6.01 -6.45 2000 -6.00 -6.41 3000 -6.02 -6.44 4000 -6.02 -6.48 6000 -6.16 -6.61 8000 -6.25 -6.80 10000 -6.35 -6.97 12000 -6.59 -7.15 14000 -6.77 -7.32 16000 -6.99 -7.58 18000 -7.21 -7.81 19150 -7.36 -7.90 20500 -7.68 -8.07 Preparer/Date Reviewer/Dat Page 2 of 2 Approver/Dat
Attachment 5 Boron Dilution Nomograph (Sheet 1 of 1) V CAUTION V7 Refer to Attachment 2, "Table of Nomograph Correction Factors," for the applicable correction factor for plant conditions. VW 8. () 20 50 100 I-t4 I 200 100 ci I-- 500 IC 200 CD 1000 CO 3-
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-J 2000
-a 0-3- A 500 - =6 5000 100 a "-'
S 10.000 1000 ft x 20.000
- a. 1 2000 1000 7-7- 50,000 100.000 3000-It OP 3304C Level of Use l _ Rev. 020-04 Continuous I 72 of 78 I --
Form ES-401-6 ES-401 Written Examination Question Worksheet Question # 83 Examination Outline Cross-reference: Tier # 2 Group # 3 K/A # 078.K2.01 Importance Rating 2.9 Proposed Question: in a reactor trip and An electrical fault occurs in the "B" Train 4160 volt bus tie breaker, resulting a loss of both 34B and 34D. the performance of ES-0. 1 "Reactor Trip What will be the status of instrument air (lAS) during Response"? A. Both instrument air compressors are still available. but the "B" instrument air compressor has B. The "A" instrument air compressor will be running, been lost. Air compressor will maintain IAS system C. Both IAS compressors will be lost, but the Service pressure. the service air compressor have been lost. D. Both of the instrument air compressors and Proposed Answer: D Explanation (Optional): to the "B" via 34B ("A" and "B" wrong), and the power supply The power supply to the "A" IAS compressor is correct). "A" is service air compressor is via 34B ("C" wrong, "D" compressor is via 34D. The power supply to the all from the same train. "B" is plausible, since almost plausible, since both IAS compressors receive power is not labeled trains. "C" is plausible, since the SAS compressor equipment at Millstone 3 is powered from opposite with a train designator. Form OP 3332A-004 (Attach if not previously provided) Technical Reference(s): Form OP 3332C-3 examination: None Proposed references to be provided to applicants during plant air systems (As available) Learning MC-05321 Describe the operation of the following Air Compressors... Objective: components... Service Air Compressor... Instrument Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
Instrument Air System - Electrical Lineup I Performed Component ID Description Position Initial Date 32P5-2 3IAS-C1A, instrument air compressor, ON supply breaker 32U8-2 3IAS-C1B, instrument air compressor, ON supply breaker 32 -3T (2FT) 3IAS-DRY1, instrument air dryer, ON supply breaker ON 3BYS-PNL32F 3CCS-AOV191A, 192A, 193A, 194A, (ckt 07) 191B, 192B, 193B, 194B, domestic and cooling water valves, and ON 3IAS-AOV14, control power service to instrument air valve, control power DC control power available light (IS) VERIFY for annunciator power illuminated 3BYS-PNL5F 3IAS-AOV95, shutdown air ON (ckt 08) compressor header air valve Bus 32-1R (F5M) 3IAS*C2A, shutdown instrument air OFF compressor Bus 32-IW (F5M) 3IAS*C2B, shutdown instrument air OFF compressor 3SCA-PNL21N 3IAS-C1A, instrument air compressor ON (ckt 16) breaker auxiliary control circuit Performed by: OP 3332A-004 Rev. 004-01 Page 2 of 2
Service Air System - Electrical LineupP I Performed Power Supply Description Position Initial Date BUS 32K Bkr 10 3SAS-C1, service air compressor ON 125 VDC 3BYS-PNL32F l 3SAS-AOV14, service air to instrument air valve ON Ckt 07 l 3SAS-AOV33, service air supply valve ON 120 VAC 3SCA-PNL21N Ckt 08 3SASN03, service air compressor auxiliary circuit ON 125 VDC 3BYS-PNL-5-2 Ckt 06 3SASN01, service air compressor breaker control ON power ckt Performed by: OPS Form 3332C-3 Rev. 0 Page 2 of 2 I
Form ES-40 1-6 ES-40 1 Written Examination Question Worksheet Question # 84 Examination Outline Cross-reference: Tier # 3 Group# 1 K/A# GEN.2.1.1 Importance Rating 3.8 Proposed Question:
"Plant Startup".
The plant is being started up in accordance with OP 3203 3 of MP- 14-RXM-PRG "Reactivity According to the Reactivity Management Standards, Attachment must be observed during the startup? Management", which of the following is a standard which significant feed flow change. A. The operator in control of feedwater announces every notified of every Load Set/Load Limit adjustment. B. During turbine load changes the Unit Supervisor is rod motion to raise primary temperature. C. During an unexpected plant transient, the RO uses control not leave the controls to acknowledge an expected D. During an extended blended make-up, the RO will annunciator on Main Board 1. Proposed Answer: A Explanation (Optional):
"A" is correct, Att. 3, Standard 2.d does not have to be notified of every "B" is wrong, Att. 3, Standard 2.c states that the Unit Supervisor Load Set/Load Limit adjustment motion shall not be used to attempt to raise "C" is wrong, Att. 3, Standard 10 states that control rod primary temperature.
are granted for extended make-ups provided the "D" is wrong, Att. 3, Standard 8.a states that exceptions RO frequently monitors the operation. MP-14-RXM-PRG Reactivity Management (Attach if not previously provide d) Technical Reference(s): Standards, Attachment 3 None Proposed references to be provided to applicants during examination: in (As available) Learning MC-06342 Master Reactivity Management Principles as outlined Objective: the Reactivity Management Program Manual. Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
Attachment 3 Reactivity Management Standards (Sheet I of 3) Reactivity Management Standards that affect, or have the Reactivity Management is conservatively managing all evolutions of these assemblies when potential to affect, fuel assemblies and reactivity conditions they are located in the reactor or the spent fuel pool. any operation that affects Unit Supervisor permission is required prior to performing results shall be discussed with reactivity of the core. The intended actions and anticipated the Unit Supervisor. reactor core through This standard conveys the expectation that protecting the priority and attention. conservative reactivity management shall receive the highest delineated on MM 14, Expectations for reactivity management activities are also 7, and 8. Guideline 01, Work Observation Form, Attachments 6, The following standards are to be observed: with approved
- 1. All Reactivity manipulations shall be performed in accordance from Reactor Engineering.
procedures, including the use of reactivity thumb rules desired result. The RO shall monitor parameters necessary to verify the to the Control Room
- 2. All Operator initiated reactivity changes shall be announced include blended crew and acknowledged by an SRO. These manipulations which may affect operations, letdown demineralizer flow or temperature changes additions, steam RCS boron concentration, manual rod movements, feedwater crew is aware of
-demand changes, and Tave changes. The intent is to ensure the controlled. these changes and that reactivity changes are positively announcements
- a. During Unit power changes, the CO's are required to make and dilutions and concerning commencement of rod movement, borations, state the magnitude of such actions.
make-ups
- b. Unit Supervisor is to acknowledge and concur with all blended prior to their performance (including makeup to the RWST).
boration will be
- The RO shall state how many gallons of dilution or added prior to starting the addition.
will
- The US will repeat the announced intention. This verbalization for the allow the crew to verify that the intended volume is reasonable task.
The Unit
- c. Turbine load changes will be coordinated between the CO's.
start and Supervisor will be made aware of and acknowledge the initial does not final completion of Turbine load changes. The Unit Supervisor need to be aware of every Load Set/Load Limit adjustment. MP-14-RXM-PRG Rev. 000 43 of 45
Attachment 3 Reactivity Management Standards (Sheet 2 of 3) of feedwater shall also
- d. During startup evolutions, the operator in control feed flow changes.
make announcements with regard to significant or Rx Engineer shall calculate
- 3. During planned/scheduled ramps COs, STA, from the US.
borationlrod movement and have concurrence the COs during unit transients or
- 4. The SM/US shall minimize distractions to control.
evolutions that might detract from reactivity RO during the reactivity
- a. The US shall be positioned to monitor the activities, an additional manipulation. During planned startup or shutdown oversight of reactivity SRO may be assigned to allow for increased management tasks.
communications of control
- b. During reactor startup (pull to criticality) all by the SRO in charge of rod movement will be made to and acknowledged do not have to be relayed reactivity management. These communications of the unit and will be to the US. The US will remain in overall control (i.e., group at top of informed of all major rod group bank movements Mode changes, and reactor core, next group starts being withdrawn, etc.),
criticality. STA ) will ensure that they keep
- 5. The Control Room team (including SM, and reactivity i.e., any maintenance, each other up to date on all systems affecting dumps, or unusual lineups in CVCS, steam tagouts, degradation, unavailability reactivity effects shall be feedwater, or turbine control systems. Potential even during shutdown conditions.
considered when changing system alignment of demineralizers that
- 6. The Control Room team will control the manipulations could possibly affect reactivity.
of any evolution that could result
- 7. The Control Room team will maintain oversight RCS, i.e., VCT chemical additions.
in admission of any unborated fluids into the shall be established and peer For such evolutions, continuous communications checking utilized. attentive to the changes and
- 8. During reactivity changes the RO shall remain monitor the plant for the expected response.
in a position to be cognizant
- a. During blended make up, the RO shall remain power. To the extent of activities being performed which affect reactor controls for blended make possible the RO should remain in front of the to the RCS provided ups. Exceptions are granted for extended makeups that the RO frequently monitors the operation.
MP-14-RXM-PRG Rev. 000 44 of 45
Attachment 3 Reactivity Management Standards (Sheet 3 of 3) changes
- b. During physics testing, low power operations, and planned power reactor's response to the RO shall maintain attention specifically on the the evolution being performed verifying that the response is as expected.
- 9. The STA should help provide oversight of the overall reactivity management process. The STA shall provide feedback to the SM/US on any perceived reactivity issues.
shall not be
- 10. During an unexpected secondary plant transient, control rod motion will be used to attempt to raise primary temperature. Control of the transient regained by use of plant equipment such as
- Reducing turbine load
- Adjusting steam dump or atmospheric dump valves Closing steam drain valves
- Inserting control rods MP-14-RXM-PRG Rev. 000 45 of 45
Written Examination Question Worksheet Forrmr ES-401-6 ES-401 Examination Outline Cross-reference: Question # 85 (SRO) Tier # 3 Group # I K/A # GEN.2.1.7 Importance Rating 4.4 Proposed Question: The Plant is in Mode 3 with a cooldown in progress in accordance with OP 3208 "Plant at the Cooldown". Due to a Tech Spec ACTION requirement, the crew is attempting to cooldown logged over maximum allowed administrative cooldown rate limit. The following data has been the last hour: TIME RCS TEMP RCS PRESS 1500 5490F 2075 psia 1515 534 0F 1700 psia 1530 517 0F 1500 psia 1545 502 OF 1375 psia 1600 489 0F 1250 psia Which of the following actions should be taken at Time 1600? A. Maintain current cooldown rate, since it is at the administrative limit. B. Decrease the cooldown rate, since it exceeds the administrative limit, but not the Tech Spec limit. C. Stop the cooldown, since it exceeds both the administrative limit and the Tech Spec limit. D. Increase the cooldown rate, since it is below the administrative limit. Proposed Answer: D Explanation (Optional): to calculate the Step 4.2.3.d requires that the elapsed time between the current and previous readings be used The administrative cooldown rate and used to adjust the cooldown, and includes a check over the last hour as well. 60'F, and DT for limit is 75 0F in any one hour period, or 1.25 F per minute. Cooldown over the last hour was 0 the cooldown rate should be cooldown is currently 13°F in 15 minutes, or 0.87 °F/min. Based on plant conditions increased to approx. 1.25°F/min. Technical Reference(s): SP 3601G, step 4.2.3.d & e. (Attach if not previously provided) Form 3601G.2-1, pages 3 & 4 OP 3208, step 4.2.8 41 of 46 NUREG-1021, Revision 8, Supplement 1
Written Examination Question Worksheet Form ES-401-6 ES-401 Tech Spec sections 3/4 Proposed references to be provided to applicants during examination: (As available) Learning MC-04837 The crew operates the plant in compliance with all applicable Objective: plant procedures and technical specifications Question Source: Bank # 69844 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.5 Comments: 41 of 46 NUREG'1021, Revision 8, Supplement 1
- d. PERFORM the following to determine the current cooldown rate:
- 1) CALCULATE the temperature difference (0 F) between current RCS temperature and previous RCS temperature.
- 2) CALCULATE the time elapsed (min) between current and previous reading.
- 3) CALCULATE the current cooldown rate ( 0 F/min) by dividing the temperature difference by the time elapsed and RECORD on SP 3601G.2-001.
- 4) IF existing cooldown rate ( 0 F/min) exceeds limits required for soak time, SUSPEND cooldown for the required soak time.
- e. WHEN four logging intervals (1 hour) are completed, PERFORM the following, and every 15 minutes thereafter:
- 1) CALCULATE the temperature difference between the current RCS temperature (Tc) and the previous hour's RCS temperature (Tc) and RECORD the current hour's cooldown (0 F).
- 2) Refer To SP 3601G.2-001, "Acceptance Criteria," and DETERMINE the RCS cooldown limit (0 F) for the current plant conditions and RECORD on the form.
- 3) IF the hourly cooldown is greater than the cooldown limit, Refer To TIS 3.4.9.1 and DETERMINE ACTION required.
4.2.4 IF a hold point is established during cooldown OR the target temperature is attained, PERFORM the following:
- a. WHEN two additional logging intervals indicate RCS temperature has stabilized within the desired band, SUSPEND cooldown rate logging.
- b. IF continued cooldown desired, Go To step 4.2.2.
4.2.5 WHEN cooldown completed, SEND SP 3601G.2-001 to the SM for review.
-End of Section 4.2 -
Level of Use SP 3601G.2 General Rev. 008-01 10 of 17 I
RCS Heatup and Cooldown Rate Heatup Soak Time Determination Condition RATE ACTION ( 0 F/min) ATO
<0.66°F/min No soak required.
Any T.< 1600 F Stop heattkp. 2 0.66 0 F/min Determine soak time.* When soak complete, restart heatup.
< 1.33°F/min No soak required.
All T, > 160'F Stop heatup. 2 1.33°F/min Determine soak time.* When soak complete, restart heatup. Cooldown Soak Time Determination Condition RATE ACTION (OF/min) ATO
< 1.33°F/min No soak required.
All T > 180 0 F Stop heatup. c 1.33"F/min Determine soak time.* When soak complete, restart cooldown.
< 0.33 "F/min No soak required.
Any Tc < 180"F Stop heatup. 2 0.33 "F/min Determine soak time.* When soak complete, restart cooldown
- Soak Time is measured from the time heatup or cooldown is stopped.
Soak Time = Number of OF/min heatup or cooldown rate was exceeded multiplied by (minutes) the time elapsed (minutes) between current and previous reading SP 3601G.2-001 Rev. 005 Page 3 of 4
RCS Heatup and Cooldown Rate ACCEPTANCE CRITERIA PARAMETER CONDITION TIS ACCEPTANCE CRITERIA Within heatup limit line shown RCS Temperature and onT/S Figure 3.4-2 Pressure Within cooldown limit line shown Normal Cooldown on TIS Figure 3.4-3 Any Tc < 1600F
- 40'F in any 1 hour period All Tc > 160'F < 80'F in any 1 hour period All Tc 2 160 0F < 80'F in any 1 hour period Any T, < 160'F and RCS pressure between high & low < 20'F in any 1 hour period RCS Cooldown Limit limit lines on T/S Figure 3.4-3 Any Tc < 1600 F and RCS pressure maintained less than < 40'F in any 1 hour period the lower limit line on T/S Figure
.. 3.4-3 Performed by: Date:
Performed by: Date: Performed by: Date: Performed by: - Date: Performed by: Date: Performed by: . Date: Performed by: Date: Performed by: . Date: SP 3601G.2-001 Rev. 005 i Page 4 of 4
4.2.7 OBTAIN Key No. 7 (CO Key Locker) and PLACE the following switches in the "COLD" position while continuing with this procedure:
- Bus 35A "INST & PT CUBICLE" - 43PB
- Bus 35A 4-2 (3RCS-P1A) - 43PP
- Bus 35B "INST & PT CUBICLE" - 43PB
- Bus 35B 2-2 (3RCS-P1B) - 43PP
- Bus 35C "INST & PT CUBICLE" - 43PB
- Bus 35C 5-2 (3RCS-P1C) - 43PP
- Bus 35D "INST & PT CUBICLE" - 43PB
- Bus 35D 1-2 (3RCS-P1D) - 43PP V CAUTION V
- 1. Exercise extreme care when drawing steam to avoid apressure rate compensated low steam line pressure SI. Steam line will change more rapidly if fewer than four steam generators are supplying the main steam header.
- 2. RCS temperature must not be reduced below 520'F until main steam line low pressure SI is blocked (P-11).
4.2.8 IF the condenser is available AND using condenser steam dumps desired, INITIATE RCS cooldown as follows:
- a. IF the condenser steam dumps are not in service, Refer To OP3316A, "Main Steam," and in Section for Main Steam System Startup, PERFORM step to place the condenser steam dumps in service and remove the main steam pressure relieving valves from service.
- b. CHECK MSS-N07, "MODE SEE' switch, in "STM PRESS" mode (MB5).
- c. CHECK MB4D 6-7, "TURB BYPASS VV ARM FOR OPENING," permissive light lit.
- d. PLACE 3MSS-PK507, "SG PRESS CNTL," in "MANUAL' (MB5).
OP 3208 Level of Use Rev 020-03 Continuous 29 of 103
- e. To establish a cooldown rate not to exceed the administrative limit of 750 F in any one hour period, slowly INCREASE output on 3MSS-PK507, "SG PRESS CNTL' (MB5).
- f. WHEN Tavg decreases below the P- 12 setpoint as indicated by MB4D 5-1, "LO-LO TAVE BLOCK P-12," permissive light lit, PERFORM the following:
- 1) DECREASE output on 3MSS-PK507, "SG PRESS CNTL," to establish a "0" output (MB5).
- 2) Momentarily PLACE 3MSS-N05, "INTLK-TR A' and 3MSS-N06, "INTLK-TR B," in the "BYP INTLK" position (MB5).
- 3) CHECK MB4D 6-9, "TURB BYPASS TAVE INTLK BYPASSED," permissive light lit.
- g. To establish a cooldown rate not to exceed the administrative limit of 750 F in any one hour period, ADJUST output on 3MSS-PK507, "SG PRESS CNTL," as necessary (MB5).
- h. IF Tavg increases above 553°F, REPEAT step 4.2.8.
- i. IF desired, SET a low limit value for computer point alarm CVRH, "Hourly RCS HURICDR" as desired to warn of an approach to the cooldown rate limit (set as a negative value).
- j. Go To step 4.2.10.
CAUTION V
- 1. Exercise extreme care when drawing steam to avoid a rate compensated low steam line pressure SI. Steam line pressure will change more rapidly if fewer than four steam generators are supplying the main steam header.
- 2. RCS temperature must not be reduced below 520°F until main steam line low pressure SI is blocked (P-11).
4.2.9 INITIATE RCS cooldown using atmospheric relief valves or relief bypass valves as follows:
- a. IF the atmospheric relief valves are available, PERFORM the following:
Level of Use OP 3208 L Continuous 711 ICar30 11r Rev. 020-03 of103
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 86 (SRO) Tier # 3 Group # I K/A # GEN.2.1.12 Importance Rating 4.0 Proposed Question: With the MDMFP out of service, a rapid downpower is conducted to 50% power due to the pending loss of a turbine driven main feed pump. After completion of the downpower, the following sequence of events occurs:
- 1. The RO notices that two control bank "D" rods did not insert with the others.
- 2. After 1 hour, the crew has not been able to withdraw the remaining bank "D" rods back to within 12 steps of the misaligned rods.
- 3. I&C has determined that both rods are trippable.
What Technical Specification ACTION is required? A. The plant must be in HOT STANDBY within 6 hours of the discovery of the problem. B. Initiate action within 1 hour to be in HOT STANDBY within the next 6 hours. C. Power ops may continue if the rods are declared inoperable, SDM requirements are met, and power is reduced below 75% within the next hour. D. Power operation may continue if the inoperable rods are restored to OPERABLE status within 72 hours. Proposed Answer: A Explanation (Optional): "A" is correct per TS 3.1.3. I.d. "B" is wrong since 3.0.3 is less restrictive than 3.1.3. l.d, which covers the condition. "C" is wrong since these are ACTIONs for ONE rod inoperable in 3.1.3. I.b. "D" is wrong, since this ACTION is taken in 3.1.3.1 .c only if the crew was successful in realigning the bank to the rods. Technical Reference(s): Tech Spec 3.1.3.1. (Attach if not previously provided) Proposed references to be provided to applicants during examination: Tech Spec sections 3/4 Learning MC-03904Given a plant condition which requires the use of AOP 3552, (As available) Objective: identify applicable Technical Specification action requirements. Question Source: Bank # 74491 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.2 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT March 11, 1991 I TMTING CtONnlTTION FOR OPFRATIDN 3.1.3.1 All full-length shutdown and control rods shall be OPERABLE and positioned within +/-12 steps (indicated position) of their group step counter demand position. APPLICABILITY: MODES 1 and 2 ACTION:
- a. With one or more full-length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour and be in HOT STANDBY within 6 hours.
- b. With one full-length rod trippable but inoperable due to causes other than addressed by ACTION a., above, or misaligned from its group step counter demand height by more than +/-12 steps (indicated position), POWER OPERATION may continue provided that within 1 hour:
- 1. The rod is restored to OPERABLE status within the above alignment requirements, or
- 2. The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within
+12 steps of the inoperable rod while maintaining the rod sequence and insertion limits of Specification 3.1.3.6. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or
- 3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:
a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions; b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours; See Special Test Exceptions Specifications 3.10.2 and 3.10.3. MILLSTONE - UNIT 3 3/4 1-20 Amendment No. by, 60
qt 9 REACTIVITY CONTROL SYSTEMS December 27, 2000 LIMITING CONDITION FOR OPERATION ACTION (Continued) c) A power distribution map is obtained from the movable incore detectors and FQ(Z) and FN are verified to be within their limits within 72 hours; and d) With four loops operating, the THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER, or e) With three loops operating, the THERMAL POWER level is reduced to less than or equal to 50% of RATED THERMAL POWER within the next hour and within the following 4 hours the Neutron Flux High Trip Setpoint is reduced to less than or equal to 60% of RATED THERMAL POWER.
- c. With more than one rod trippable but inoperable due to causes other than addressed by ACTION a. above, POWER OPERATION may continue provided that:
- 1. Within 1 hour, the remainder of the rods in the bank(s) with the inoperable rods are aligned to within +12 steps of the inoperable rods while maintaining the rod sequence and insertion limits of Specification 3.1.3.6. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, and
- 2. The inoperable rods are restored to OPERABLE status within 72 hours.
- d. With more than one rod misaligned from its group step counter demand height by more than +12 steps (indicated position), be in HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours except during time intervals when the rod position deviation monitor is inoperable, then verify the group positions at least once per 4 hours. 4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 92 days. MILLSTONE - UNIT 3 3/4 1-21 Amendment No. P9, SP, 191
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 87 Tier # 3 Group # I K/A # GEN.22.1.13 Importance Rating 2.9 Proposed Question: Two Unit-3 Operations personnel have been assigned to escort 10 visiting people while giving them a tour of the Transformer and Switchgear areas of Unit-3. The following conditions exist:
- The only vital area that the visitors have been given authorization for entry is the switchgear area.
- The tour is progressing from the transformer area in the yard to the East Switchgear Room.
- Prior to entering the switchgear room, one of the escorts is paged, and is required to return to the control room.
What action is in accordance with SC-I "Access and Egress Control"? A. The one remaining escort may take escort responsibility for all 10 visitors and continue the tour into the switchgear area. B. The one remaining escort may take 5 visitors into the switchgear area, while the other escort takes the other 5 visitors with him into the control room. C. The one remaining escort may take escort responsibility for all 10 visitors and remain outside of the switchgear room until the second escort returns. D. Both escorts must take the visitors outside the protected area prior to the one escort leaving for the control room. Proposed Answer: C Explanation (Optional): An escort is required to maintain both observation and control of visitors. The injured escort can not walk, therefore can not maintain observation and control of the visitors. Escort/ visitor ratios are l/ol for non vital areas and 511 for vital areas. Since the tour was not within a vital area at the time of the accident. One person is allowed to escort the visitors as long as they do not enter a vital area. "I" is incorrect because of the 5/1 rule. "2" is incorrect because the injured person does not fulfill the definition requirements of an escort. "3" is incorrect because 5 visitors are left with the injured person who cannot be an escort. Technical Reference(s): SC-1 Sections 1.6.5 andl .10 (Attach if not previously provided) Proposed references to be provided to applicants during examination: None Learning Objective: GE-00 177 State your responsibilities in the protected area when: (As available) Escorting a visitor... Question Source: Bank # 64349 41 of 46 NUREG-1021, Revision 8, Supplement 1
For ES-41-b Written Examination Question Worksheet Form ES-4UI-6 ES-401 . Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
All personnel are subject to search for prohibited items prior to Protected Area entry. Possession of prohibited items may have serious repercussions involving local law enforcement agencies or corrective actions which could affect employment. Personnel who process into the Protected Area must pass through explosive and metal detectors. All hand-carried items (including covered coffee cups, lunch boxes, brief cases, hand bags and metal objects) are required to be searched via X- Ray machine. These items should be placed at the center of the X-Ray machine conveyor belt. Small items should be placed in the provided plastic trays prior to being put through the X-Ray machine. If you have an open liquid container, containers must be disposed of in the proper receptacle and will not be allowed through the X-Ray machine. Individuals who alarm the detectors may be required to submit to a hands on or hand-held metal detector search and should follow the instructions of the Security Officer at the Access Point. Failure to comply with these requirements will result in the denial of site access. After successfully passing through the detection equipment, Protected Area access will be accomplished as follows: 1.6.1 OBTAINhand-carried items off X-Ray machine conveyor belt. 1.6.2 IF area in front of the turnstile, indicated by the line on the floor is clear, INSERT key card into turnstile key reader and REMOVE. 1.6.3 PLACE hand into hand geometry reader. 1.6.4 WHEN green access light is displayed, ENTER Protected Area. NOTE The Escort Responsibility Card is given to escorts to provide instructions on escorting a visitor in the Protected Area. Escort 1.6.5 IF escorting a visitor into Protected Area, PERFORM the following:
- a. OBTAIN and READ Escort Responsibilities Card.
- b. WEAR and COMPLY with Escort Responsibilities Card at all times.
...... SC-1 l Level of Use l fc; Tft- ACT § Rv 00-0 Information 7 of 27
- c. ENSURE visitor-to-escort ratio does not exceed 10:1.
- d. INSTRUCT visitor to perform the following:
- 1) INSERT key card into designated visitor turnstile key reader.
- 2) PLACE hand into hand geometry reader.
- 3) MOVE aside.
- e. INSERT your keycard into designated visitor turnstile key reader.
- f. PLACE your hand into hand geometry reader and ENTER Protected Area.
- g. WIHEN Security Officer releases turnstile, INSTRUCT visitor to enter the Protected Area.
16.6 If escorting a visitor into the Protected Area in conjunction with a Special Event, PERFORM the following:
- a. PROCEED to designated security badge issue counter.
- b. IF escorting a Visitor into the Protected Area perform the following:
- 1) OBTAIN and READ Escort Responsibilities Card.
- 2) WEAR and COMPLY with Escort Responsibilities Card at all times.
- 3) ENSURE visitor to escort ratio does not exceed 10:1.
Security c. ENSURE completed 'Access Authorization Form" is on file. Personnel
- d. VERIFY proper identification.
SC-1 Level of Use l. tI ACT X< A.E Rev. 006-03 Information i 8 of 27
n7 NOTE Authorization to Security Alarm Stations require the approval of the Process Owner - Protective Services. I All Personnel l 1.9.2 1E not authorized for a Vital Area, or changes in Vital Area access must be made, PERFORM the following:
- a. OBTAIN "Vital Area Access Authorization Form" from Security and COMPLETE.
- b. IF necessary, OBTAIN name of personnel authorized to authorize Vital Area Access from Security.
- c. SEND completed "Vital Area Access Authorization" form to NAP Visitor Sign In or Processing Center.
I All Personnel l 1.9.3 IF authorized for a Vital Area, PERFORM the following:
- a. INSERT keycard in appropriate key reader.
- b. VERIFY green light is lit.
NOTE In the case of multiple person entry / exit, it is not necessary for the door to close between each person however, each person must wait for the keycard reader light to extinguish prior to inserting their keycard.
- c. ENTER or EXIT Vital Area.
1.10 EscortingVisitors Into and Out of Vital Areas Visitors requiring Vital Area access must be escorted by an individual who is picture-badged and authorized for entrance into the Vital Area. IEscort I: t 1.10.1 WLULN escorting visitors in a Vita] Area, ENSURE that the visitor-to-escort ratio does not exceed 5:1. SC-1 Level of Use l TikIE Rev. 006-03 Information 11 of 27
NOTE When keycards are not available, visitors may be logged in and out of Vital Areas by Security. 1.10.2 IF keycard is not available for visitor to enter Vital Area, NOTIFY Security at extension 4607. I 1.10.3 IF authorized for Vital Area, INSTRUCT visitor to insert their keycard into appropriate key reader and VERIFY red light is lit. 1.10.4 Immediately INSERT your keycard into appropriate key reader and VERIFY green light is lit. 1.10.5 WHEN light is green, ENTER Vital Area with visitor. 1.10.6 WHE.N escorting visitors out of Vital Areas, EXIT Vital Area with visitor and ENSURE visitor complies with all Vital Area exit requirements. 1.11 Material Entry To or Exit From Vital Areas When attempting to move property or material into a Vital Area, it may become necessary to contact Security for assistance. If the movement of material requires that the security door be opened for longer than the amount of time required for normal personnel access / egress, personnel involved in the movement should contact Security 48 hours in advance of movement to schedule posting of Security Officer. This eliminates unnecessary security alarms and allows for smoother passage of material into the area. IAll Personnel 1.11.1 IF property or material must enter or exit a Vital Area AND the movement would require that the security door be open for longer than normal, PERFORM the following:
- a. IF possible, NOTIFY security 48 hours in advance.
- b. REQUEST Security Officer be posted to facilitate movement of material.
Level of Use l . ACT ?CtW SC-1 Rev. 006-03 InformationI 12 of 27
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 88 Tier # 3 Group # I K/A # GEN.2.1.25 Importance Rating 3.1 Proposed Question: The plant has been at 100% power for several months with the MDMFP unavailable when the following sequence of events occurs: 8/6/02, 0330: A rapid downpower to 50% is conducted due to a failed seal on the "A" TDMFP. 8/8/02, 0830: The seal has been replaced, and an up-power is commenced at 3%/hr to 80%. 8/8/02, 1900: With power at 80%, a calorimetric is in progress when the new seal fails. A rapid downpower is again performed to reduce power to 50%. 8/8/02, 1906: Power is steady at 50%. 8/9/02, 2030: The seal is replaced, and the crew is preparing to raise power from 50% to 100%. Using Attachment 4 of OP 3204 "At Power Operation", attached to this exam, what is the maximum rate of power increase allowed during the up-power? A. 3%/hr until 72 cumulative hours of operation at 100% power has been attained. B. 10%/hr up to 80% power, and then 3%/hr from 80% to 100% power. C. 10%/hr up to 90% power, and then 3/o/hr from 90% to 100% power. D. 10%/hr up to 100% power. Proposed Answer: C Explanation (Optional): NOTE: This is a new table within OP 3204. "C" is correct, since power has been at 100% for at least 72 hours out of the 7 day operating period, and the bottom row of Attachment 4 is applicable. "A", "B", and "D" are plausible, since 3% and 10% are the increase rates available for selection from this attachment, and power had been returned temporarily to 80% during the first up-power. Technical Reference(s): OP 3204, Attachment 4 (Attach if not previously provided) Proposed references to be provided to applicants during examination: OP 3204, Attachment 4 Learning MC-03397 Describe the major action categories contained (As available) Objective: within the OP 3204 procedure. Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.10 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
Attacinment 4 Fuel Condition Load Increase Restrictions (sheet 1 of 1) Fuel Condition Category Definition Applicable Range of Maximum Rate of Power Control Rod Withdrawal Category Power Level Increase Restrictions Refueling Initial return to power 20% to 100% Power 3% per hour until 72 following refueling 1.Minimize control rod cumulative hours of motion OR operation at 100% power 2. Limit withdrawal to 3 Mid-cycle shutdown out of any 7 day operating steps per hour above 50% involving fuel handling in period has been attained. power (see Note) core. Partially Conditioned Fuel is conditioned up to 20% power to power level 10% per hour up to power power level "P" and rod No withdrawal restrictions 7,, level "E" up to rod position"N" position "N" "P"=Highest power level sustained for at least 72 cumulative hours out of Power level "P" to 100% 3% per hour until 72 Limit withdrawal to 3 steps any 7 day operating period power cumulative hours at 100% per hour above rod power out of any 7 day position "N" (see Note)
"N"=Highest rod position operating period has been sustained for at least 72 attained.
cumulative hours out of any 7 day operating period Extended Low Power Reduced power operation Power levels above the 3% per hour until 72 Operation Limit withdrawal to 3 steps for greater than 27 days highest power level sustained cumulative hours of for at least 72 cumulative operation at 100% power has per hour during power hours during the preceding been attained. ascension (see Note) 30 days Fully Conditioned 100% power sustained for at 20% to 90% power 10% per hour least 72 cumulative hours out No restrictions on control rod motion of any 7 day operating period 90% to 100% power 3% per hour Note - Control rod withdrawl may exceed 3 steps per hour for axial xenon oscillation control. Level of Use OP 3204 General STOP THINK ACT REVIEW Rev. 015-04 57of 57 57 of_
ES-40 I Written Examination Question Worksheet Form ES-40 1-6 Examination Outline Cross-reference: Question # 89 (SRO) Tier # 3 Group # I K/A # GEN.2.1.33 Importance Rating 4.0 Proposed Question: With the plant at 100% power, the following sequence of events occurs:
- 1. Charging flow had been increasing over a period of time and has stabilized.
- 2. PRT level is slowly increasing.
- 3. Engineering and operations personnel have determined BOTH pressurizer PORVs are leaking excessively, but well within the capacity of the running charging pump.
- 4. Leakage is determined to be approximately 2 gpm (equivalent) per valve.
What actions, if any, are required to be taken per technical specifications? A. No actions are required to be taken by technical specifications. B. Close the block valves for BOTH PORVs, with power maintained to the block valves. C. Restore the PORVs within 7 days, or depressurize and vent the RCS within the next 12 hours. D. Reduce RCS leakage to within limits within 4 hours or be in at least HOT STANDBY within 6 hours. Proposed Answer: B Explanation (Optional): PORV inoperability due to seat leakage does not prevent automatic or manual use. Therefore, the block valve may be closed but the action requires power to be maintained to the valve ("B" correct). The EOPs provide guidance to assure that the block valves would be opened early in the event, ensuring that the PORVs would be available to mitigate the event. "A" is wrong, since the PORVs are leaking excessively. "C" is wrong since the COPPS function of the PORVs per LCO 3.4.9.3 is not required in MODE 1. "D" is wrong, but plausible, since this action is required if the leakage exceeded identified leakage rates per LCO 3.4.6.2. Technical Reference(s): Tech Spec 3/4.4.4 (Attach if not previously provided) Tech Spec 3/4.4.6.2 Tech Spec 3/4.4.9.3 Proposed references to be provided to applicants during examination: Tech Spec section 3/4 Learning MC-06063 Determine applicable LCO action requirements for a given (As available) Objective: plant condition or event Question Source: Bank # 73762 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.2 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
06/65/98 15:15 NO. 637 [0 6/5/98 REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES LIMITING CONDITION FOR OPEATION 3.4.4. Both power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With one or both PORV(s) inoperable because leakage, within 1 hour either restore of excessive seat the PORV(s) to OPERABLE status or close the associated block valve(s) block valve(s); otherwise, be in at leastwith power maintained to the next 6 hours and in HOT SHUTDOWN within HOT STANDBY within the the following 6 hours.
- b. With one PORY inoperable due to causes leakage, within 1 hour either restore other than excessive seat the PORY to OPERABLE status or close the associated block valve and remove valve; restore the PORV to OPERABLE status power from the block 72 hours or be in HOT STANDBY within within the following the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- c. With both PORYs inoperable due to causes other than excessive leakage, within 1 hour either restore at least one PORY seat status or close to OPERABLE its associated block valve and remove block valve and power from the be in HOT STANDBY 'within the next 6 hours SHUTDOWN within the following 6 hours. and in HOT
- d. With one or both block valve(s) inoperable, the block valve(sy to OPERABLE status, within I hour restore or place its associated PORV(s) control switch to 'CLOSE.' Restore to OPERABLE status within the next hour at least one block valve if inoperable; restore any remaining inoperable both block valves are status within 72 hours; otherwise, block valve to operable be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- e. Entry into an OPERATIONAL MODE is permitted ACTION requirements. while subject to these Wtl I _
-TALr 1JTT 2 -
34 14-12 "' q-iz.AmMndment No.. 97.. U. fili
REACTOR COOLANT SYSTEM JAN 31 1986 OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to: -.
- a. No PRESSURE BOUNDARY LEAKAGE,
- b. 1 gpm UNIDENTIFIED LEAKAGE, C. 1 gpm total reactor-to-secondary leakage through all steam generators not isolated from the Reactor Coolant System and 500 gallons per day through any one steam generator not isolated from the Reactor Coolant System,
- d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System,
- e. 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2250 t 20 psia, and
- f. 0.5 gpm leakage per nominal inch of valve size up to a maximum of 5 gpm at a Reactor Coolant System pressure of 2250
- 20 psia from any Reactor Coolant System Pressure Isolatlion Valve specified in Table 3.4-1.
APPLICABILITY MODE$ 1, 2, 3, and 4.. ACTION:
- a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
. ElI MILLSTONE - UNIT 3 3/4 4-22
REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS August 27, 2001 LIMITING CONDITION FOR OPERATION 3.4.9.3 Cold Overpressure Protection shall be OPERABLE with a maximum of centrifugal charging pump* and no Safety Injection pumps capable of injecting one into the Reactor Coolant System-(RCS) and one of the following pressure capabilities: relief
- 1. One power operated relief valve (PORV) with a nominal lift setting established in Figure 3.4-4a and one PORV with a nominal lift setting established in Figure 3.4-4b with no. more than one isolated RCS loop, }
or
- 2. Two residual heat-removal (RHR) suction relief valves with setpoints
> 426.8 psig- and < 453.2 psig, or
- 3. One PORV with a-nominal lift setting established in Figure 3.4-4a or Figure 3.4-4b with no more than one isolated RCS loop and one RHR suction relief valve with a setpoint > 426.8 psig and < 453.2 psig, or
- 4. RCS depressurized with an RCS vent of'> 2.0 square inches.
APPLICABILITY: MODE 4 when any RCS cold leg temperature is < 226°F, MODE 5, and MODE 6 when the head is on the reactor vessel. ACTION:
- a. With two or more centrifugal charging pumps capable of injecting into the RCS, immediately initiate action to establish that a maximum of one centrifugal charging pump is capable of injecting into the RCS.
- b. With any Safety Injection pump capable of injecting into the RCS, immediately initiate action to establish that no Safety Injection pumps are capable of injecting into the RCS.
- c. With one required relief valve inoperable in MODE 4, restore the required relief valve to OPERABLE status within 7 days, or depressurize and vent the RCS through at least a 2.0 square inch vent within the next 12 hours.
*Two centrifugal charging pumps may be capable of injecting into the than one hour, during pump swap operations. However, at no time will RCS for less two charging "umps be simultaneously out of pull-to-lock during pump swap operations.
l -1ILLSTONE - UNIT 3 3/4 4-38 Amendment No. j$, X7, Up, 10, J07, 197
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 90 (SRO) Tier # 3 Group # 2 K/A # GEN.2.2.11 Importance Rating 3.4 Proposed Question: Which of the following is required in order to ensure strict control of temporary modification tags? A. The jumper tags must be hung by a licensed operator. B. Independent or Dual verification as required is performed when installing the tags. C. The on-shift SM, on-shift US, or designated SRO must approve the temporary modification prior to hanging the tags. D. Caution tags shall be used and the clearance number noted in the "Comments" section of the Jumper Device Control Sheet. Proposed Answer: B Explanation (Optional): B A is wrong since the installer does not have to be licensed (i.e. may be from the department requesting the jumper). the installation is correct since verification (dual or independent) is required. C is wrong since the SM alone approves when the jumper is installed. D is wrong since caution tags may not be required. Technical Reference(s): WC- 10 Attachment 4, Temporary Modification (Attach if not previously provided) Control Sheet Proposed references to be provided to applicants during examination: None MC-05104 Outline the process for Temporary Modification installation. (As available) Learning Objective: Question Source: Bank # 72387 Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 55.43.3 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
Attachment 4 Temporary Modification Control Sheet (Sheet 1 of 7) IStep A - Temporary Modification Index Number Revision NumberI REQUESTER I
Title:
Step D - Requested By: Print name and sign Date: Step B - Explain why Temporary Modification is necessary (Attach additional Sheets as necessary): Describe Proposed Temporary Modification and List all equipment and functions the Temporary Modification will affect (Attach additional Sheets as necessary): Step C - State Expected Removal Date: Step E - System Engineer: Print name and sign Date: WC10 Rev. 004-01 53 of 67
Attachment 4 Temporary Modification Control Sheet (Sheet 2 of 7) I Step A - Temporary Modification Index Number Revision Number I ENGINEER Step F - Method of removal: Step G - List operations critical drawings affected by this installation and if applicable, DCN numbers (Attach additional Sheets as necessary): Step H - List special actions, instructions and procedure changes required (Attach additional Sheets as necessary): Step K - Indicate plant operation conditions and modes that this temporary modification will be limited to (Attach additional Sheets as necessary): Step I -List all testing considerations after installation and removal of temporary modification (Attach additional Sheets as necessary): Step M - Temporary Modification may be lifted and reinstalled [ YES E] NO Design Engineering Screening Evaluation Step N. - List document references (Attach additional Sheets as necessary): WC10 Rev. 004-01 54 of 67
Attachment 4 Temporary Modification Control Sheet
- (Sheet 3 of 7)
IStep A - Temporary Modification Index Number Revision Number Independent Review Step P - Independent Reviewer: X Print names and sign Date Step R Attached: CRED E yes E1 no Design Engineering Screening Evaluation (DCM 3-2C) E1 yes El no Detailed Design Review number attached: 50.59 Screen (RAC124) E yes El no 50.59 Evaluation(RAC 12-5) El yes El no Independent Review (DCM 4-1A) El yes El no Other documentation included (ist) Engineer review: Step S - Engineer Print names and sign Date System Engineer: Step T - Walk down required YES or [ NO Step S - System Engineer . _: Print names and sign Date Engineering Approval: Step V - Engineering Manager E APPROVAL OR [ DISAPPROVAL Engineering Manager: Print names and sign Date WC 10 Rev. 004-01 55 of 67
Attachment 4 Temporary Modification Control Sheet (Sheet 4 of 7) l Step A- Temporary Modification Index Number Revision Number l SORC REVIEW AND APPROVAL Step W - SORC Review and Approval Meeting No. . Date SORC Established Removal Date: SORC Chairman: Print name and sign SORC approval of implementing procedures and procedure changes are complete Step X - Engineer: Print names and sign Date WP&OM WC 10
. Rev. 004-01 56 of 67
Attachment 4 Temporary Modification Control Sheet (Sheet 5 of 7) Step A - Temporary Modification: Index Number Revision Number Step AA Tag Number Location Description OPERATIONS Lustallation Step AB- Verification Method: ]Visual Confirmation LIFunctional Check Step AD LI Independent Verification or [Dual Verification Step AE - Approved Shift Manager: __________________ Print names and sign Date Step AF - Installed by: _____________________ Print names and sign Date Step AG- Verified by: __________ __________ Print names and sign Date Step AC - Comment: ENGINEER Affected operational critical drawings in the Control Room, Tech. Support Center, Work Control and the EOF are updated as required. If required walkdown performed Required procedure changes have been implemented. Step AHl Eng Walkdown Signature: Print names and sign Date.' Step AK Drawing Update Signature: Print names and sign. Date WC 10 Rev. 004-01 57 of 67
Attachment 4 Temporary Modification Control Sheet A. (Sheet 6 of 7) Step A - Temporary Modification Index Number Revision Numberl AUDIT Step AMV ...
- 1. _ ____ ___ ____ ____ ___ 13.
2 _ _ _ _ __ _ _ _ _ 14.
- 3. . 15.
- 4. 16.
- 5. . 17.
6.* 18. *
- 7. 19.
- 8. 20.
- 9. ._._--_ . 21.
- 10. . 22.
II. _23. :
- 12.
- 24. *
- SORC review required Step AN - Review comments (Attach additional sheets as necessary):
When decided to continue Temporary Modification installed beyond PERD, provide new PERD. Step AO - New PERD(s): Extension beyond one refueling cycle is authorized: Step AP -Station Director: Print name and sign Date WC 10* Rev. 004-01 58 of 67
Attachment 4 Temporary Modification Control Sheet (Sheet 7 of 7) Revision Number I Step A - Temporary Modification Index Number Removal Step AR - Requester: Print names and sign Date [] Independent Veirification OR []Dual Verification Step AS - Verification Type: E] Visual Confirm, ation OR E] Functional Check Step AT - Verification Method: Step AU - Approved Shift Manager: DateI Print names and sign Step AV - Removed By:: Date Print names and sign Step AW - Verified By: Date Print names and sign Comments ENGINEER Engineer Temporary Modification Removal Walkdown: Affected operational critical drawings updated as required. If required, walkdown performed. Required procedure changes have been implemented. Step AX - Walkdown Engineer: Date Print names and,<ign T--: V-A %7 TT,-A~t.*A t 0 ti-p 1-lilawlgS' upuatw h uxv Print names and sign Date Step AZ - Engineer: - Print names and sign Date WC 10 Rev. 004-01 59 of 67
ES-40 1 Written Examination Question Worksheet Form ES-40 1-6 Examination Outline Cross-reference: Question # 91 Tier # 3 Group # 2 K/A # GEN.2.2.25 Importance Rating 3.7 Proposed Question: The crew is responding to a dropped rod, and the QPTR is calculated to be 1.03. The SM refers to Tech Spec 3.2.4 QUADRANT POWER TILT RATIO. The SM determines that the crew must reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of QPTR in excess of 1 within 2 hours. What is the basis of the 2 hour allowance prior to the power reduction requirement? A. Allow for boration to regain Shutdown Margin. B. Heat flux hot channel factors could be violated after 2 hours. C. Axial peaking factors could be exceeded after 2 hours. D. Allow for identification and correction of a dropped or misaligned rod. Proposed Answer: D Explanation (Optional): "D" correct from Tech Spec bases: "the 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod." Technical Reference(s): Tech Spec basis 3/4.2.4 (Attach if not previously provided) Proposed references to be provided to applicants during examination: None Learning MC-05227 Describe the major administrative or procedural precautions and (As available) Objective: limitations placed on the operation of the NIS system, and the basis for each Question Source: Bank # 65077 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.43.2 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
March 11, 1991 POWER DISTRIBUTION LIMITS. BASES HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the accept-able region of operation defined in Specifications 3.2.3.1 and 3.2.3.2. 3/4.2.4 OUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and periodically during power operation. The limit of 1.02, at which corrective action -is required, provides DNB and linear heat generation rate protection with x-y plane power tilts. A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in F is depleted. A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt. The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc-tion of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on F. is reinstated by reducing the maximum allowed power by 3%for each percent of tilt in excess of 1. For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8. 3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR greater than the design limit throughout each analyzed transient. The indicated T value of 591.1@F (four loop avg MILLSTONE - UNIT 3 B 3/4 2-5 Amendment No. 77, 5S.5C 0028
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 92 (SRO) Tier # 3 Group # 2 K/A # GEN.2.2.27 Importance Rating 3.5 Proposed Question: You are the Refueling Senior Reactor Operator on-duty in Containment during Core Load. Which of the following conditions would allow Core Alterations to continue (consider each condition independently)? A. The RHR flow is deliberately suspended for 45 minutes after running for the previous 8 hours. B. Communications between the Refueling Floor and the Control Room interrupted for 3 minutes. C. Source Range audio countrate is lost in CTMT but available in the control room. D. The Refuel SRO is absent from the Refuel floor while replacing sources. Proposed Answer: A Explanation (Optional): "A" is correct since OP 321 OB allows RHR flow to be deliberately suspended for 1 hour in the last 8 hour period if no dilution in progress ( OP 321 OB, Refueling Operations 3.10). "B" is wrong, but plausible since the "temporary" nature may lead candidate to suppose the interruption is permissible (3.13). "C" is wrong since LCO 3.9.2 requires audible countrate in CTMT. "C" is plausible since audible countrate is still available in the control room. "D" is wrong, but plausible since the sources may appear not to be a significant core alteration. Technical Reference(s): OP 3210B Section 3 (Attach if not previouslv nrovided) Tech Spec LCO 3.9.2 Proposed references to be provided to applicants during examination: Tech Spec section 3/4 Learning MC-06495 Describe the stop work requirements with regards to fuel (As available) Objective: movement... Question Source: Modified Bank # 75651 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.43.6 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
ES-401 Written Examination Question Worksheet Form ES-401-6 Original 75651 You are the Refueling Senior Reactor Operator on-duty in Containment during Core Load. Which of the following conditions would allow Core Alterations to continue (consider each condition independently)? A. Communications between the Refueling Floor and the Control Room are temporarily (<5 minutes) interrupted. B. The RHR flow is deliberately suspended for 2 hours in the last 8 hour period. C. The RHR flow rates increase to 3,000 gpm per pump. D. The Refuel SRO is absent from the Refuel floor while replacing sources. Answer: C 41 of 46 NUREG-1021, Revision 8, Supplement I
2.2 Documents 2.2.1 AOP 3566, "Immediate Boration" 2.2.2 AOP 3572, "Failure of Refueling Cavity Seal" 2.2.3 EOP 3505, "Loss of Shutdown Cooling and/or RCS Inventory" 2.2.4 EOP 3502, "Fuel Handling Accident" 2.2.5 EN 31007, "Refueling Operations" 2.3 Responsibilities 2.3.1 Refer to Attachment 1 for responsibilities. 2.4 Personnel 2.4.1 One Senior Reactor Operator or Refueling Senior Reactor Operator, who has no other concurrent responsibilities, shall observe and directly supervise all CORE ALTERATIONS.
- 3. PRECAUTIONS 3.1 If a fuel assembly, RCCA, or source is believed to be damaged, fuel handling shall cease. The Reactor Engineer and the Control Room will be notified and an inspection and evaluation is to be made on the component.
3.2 A fuel assembly will come into contact with adjacent fuel assemblies and specific fixtures (such as storage racks) that are part of normal fuel handling. Fuel handling sequences that generate only side to side contact between adjacent fuel assemblies are preferable to those that may include corner to corner contact. Contact with hardware that is not part of normal fuel handling (such as light stanchions, fixtures, etc.) is not permitted. 3.3 A free standing fuel assembly is defined as one that does not have at least one side adjacent to another fuel assembly, fixture, or core structure. Free standing fuel assemblies shall not be permitted in the core. Under controlled conditions, such as required for irradiated fuel examination, a free standing fuel assembly is permissible outside the core. 3.4 A fuel assembly shall not be left unattended in a suspended position. 3.5 Any lateral loads applied to a fuel assembly for purpose of guidance, support, or clamping shall be applied only at the grid or nozzle locations. Level of Use *4 Contnuou"NOWRev. 007- 04 Rqk 7 ~8 of 28
3.6 Any axial loads applied to a fuel assembly for purposes of lifting, support, or guidance shall be applied only at nozzle locations and shall not be applied to the fuel assembly hold down springs mounted on the top nozzle. 3.7 The lateral deflection of any fuel assembly at the top nozzle shall be less than two inches. 3.8 The axial loads applied to a fuel assembly (except for support or lifting) shall always be less than 1,200 pounds. All movements of reactor components should be slow and deliberate watching for binding or a sudden increase in load which would indicate binding. 3.9 RHR flowrate should not exceed 4,000 gpm per pump and shall not be less than 2,800 gpm. See T/S 3.9.8.1 for exception to the low flow limit. 3.10 RHR flow is maintained throughout CORE ALTERATIONS except as provided by T/S 3.9.8, "Refueling Operations, Residual Heat Removal and Coolant Circulation," which allows the RHR flow to be suspended for up to 1 hour per 8 hour period provided no operations are permitted that could cause dilution of RCS Cb. 3.11 Health Physics procedures must be strictly adhered to at all times, paying particular attention to the following: 3.11.1 Radiation levels near the surface of the water in the refueling cavity must be monitored. 3.11.2 Radiation levels near the surface of the water in the spent fuel pool must be monitored as spent fuel assembles and RCC elements are received for storage. 3.11.3 Care must be exercised in monitoring for radiation and contamination hazards when handling of tools and equipment withdrawn from the refueling cavity or spent fuel pool water. 3.12 Due to the use of many non-routine system valve alignments during the refueling mode (e.g., residual heat removal, fuel pool cooling and purification, reactor vessel head removed, fuel transfer tube open to Containment, etc.), the following actions may apply: 3.12.1 Care should be exercised in reviewing scheduled surveillance testing procedures on reactor plant systems prior to approving them for performance, to preclude an inadvertent flooding of the containment sump. Level of Use I0B OP 32 ContiuousRev. Contiuous9 007-04 of 28
3.13 Where communications are required (for CORE ALTERATIONS, etc.), component movements must be stopped if communications are lost. Suspension of component movements should not preclude completion of movement of component to a safe, conservative position. 3.14 A minimum of 10.5 feet of water should be maintained above the top of each fuel assembly during all handling operations to ensure the gamma dose rate at the surface of the water is less than or equal to 2.5 mrem/hr. 3.15 Loads in excess of 2,200 pounds shall be prohibited from traveling over fuel assemblies in the storage pool as required by T/S 3.9.7, "Refueling Operations, Crane Travel - Spent Fuel Storage Areas." 3.16 Loads greater than the combination of a fuel assembly and RCCA, shall be prohibited from travel over irradiated fuel in the vessel with the head removed, without approval of the SM. 3.17 All Prerequisites have been completed and all Precautions have been read and understood by each shift. A Shift I B Shift l C Shift D Shift E Shift A OP 3210B Level of Use l Rev. 007-04 Continuous r *r 10 of 28
REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION February 20, 2002 LIMITING CONDITION FOR OPERATION 3.9.2 Two Source Range Neutron Flux Monitors shal l be. OPERABLE with continuous visual indication in the control room, and one with audible indication in the containment and control room. APPLICABILITY: MODE 6. ACTION:
- a. With one of the above required monitors inoperable immediately suspend I all operations involving CORE ALTERATIONS or positive reactivity changes.
- b. With both of the above required monitors. inoperable determine the boron concentration of the Reactor Coolant System within 4 hours and at least once per 12 hours thereafter.
SURVEILLANCE REQUIREMENTS 4.9.2 Each Source Range Neutron Flux Monitor shall be demonstrated OPERABLE by performance of:
- a. A CHANNEL CHECK and verification of audible counts at least once per 12 hours,
- b. A CHANNEL CALIBRATION at least once per 18 months.*
- Neutron detectors are excluded from CHANNEL CALIBRATION.
MILLSTONE - UNIT 3 3/4 9-2 Amendment No. J07, 203 0729
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 93 (SRO) Tier # 3 Group # 2 K/A # GEN.2.2.29 Importance Rating 3.8 Proposed Question: In accordance with MP-14-OPS-GDLO2 "Operations Standards" and OP 3210A "Refueling Preparations", which of the following evolutions requires that the Refueling SRO be present on the Refueling Floor? A. Moving the Upper Guide Structure from or into the reactor vessel B. Installing or removing the Fuel Transfer Tube blind flange C. Start filling the reactor vessel from the RWST using the RHR system D. Initial Reactor Vessel Stud detensioning Proposed Answer: A Explanation (Optional):
"A" is correct, Operations Standards, MP-14-OPS-GDLO2, Attachment 6, sheet 16 of 21 "B" is wrong, blind flange removal occurs at step 4.1.7 of OP 321 OA, Refueling Preparations, before step 4.1.17 which stations the Refueling SRO and is NOT identified as a CORE ALT. "C" is wrong, filling the reactor vessel from the RWST occurs at step 4.3.8 or 4.3.9 of OP 3210A, Refueling Preparations, and does not require stationing the Refueling SRO and is NOT identified as a CORE ALT. "D" is wrong, detensioning occurs at step 4.1.6 of OP 3210A, Refueling Preparations, before step 4.1.17 which stations the Refueling SRO and is NOT identified as a CORE ALT.
Technical Reference(s): Operations Standards, MP-14-OPS-GDLO2 (Attach if not previously provideed) OP 3210A, Refueling Preparations Proposed references to be provided to applicants during examination: None Learning MC-04544 Describe the following: (As available) Objective: A. Core Alterations and what specifically marks the start of Core Alterations B. Who has authority to direct and/or approve all core component movements C. Who has authority to terminate fuel handling operations due to a discrepancy or safety concern and which type of discrepancies warrant the termination of fuel handling operations D. When the reactor vessel head is considered to be tensioned Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.43.6 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
q3 Attachment 6 Qualification and Training (Sheet 16 of 21) Refueling SRO Training And Qualification (Sheet 1 of 4) Name:
- 1. Discussion: This guideline is intended to delineate the authority, responsibility and license requirements of the Refuel SRO while assigned to the refuel floor or when directly supervising CORE ALTERATIONS as defined by the Technical Specifications.
The main function of the Refuel SRO is to ensure safe CORE ALTERATIONS.
- 2. Refuel SRO Personnel ALL CORE ALTERATIONS shall be directly supervised by either a Licensed SRO (License Active per IOCFR55.53(e) or a SRO limited to fuel handling (per IOCFR55.53(f)(2); and shall be qualified per Nuclear Training Manual guidance for activating inactive SRO Licenses for Fuel Handling and shall have no other concurrent responsibilities during this operation.
- 3. Refuel SRO Authority
- Stop CORE ALTERATIONS when deemed necessary.
- Stop or defer any activity around the refuel floor which would jeopardize the safety of personnel or equipment.
- 4. Refuel SRO Responsibilities
- Be present on the Refuel Floor and responsible for maintaining OPS Procedures as required during the following CORE ALTERATIONS:
- Initial Reactor head lift
- Fuel shuffle
- Moving/replacing sources
- CEA shuffle in the reactor vessel
- Moving the upper guide structure from or into the reactor vessel
- Any other CORE ALTERATION as determined by Reactor Engineering
- Assist HP to ensure the Exclusion Area around the refuel pool is maintained per OA-8, "Maintenance and Housekeeping of Site Buildings and Facilities."
- Assist HP to ensure proper radiological practices are maintained around the refuel pool.
- Ensure general safety of personnel and equipment.
MP-14-OPS-GDLO2 Rev. 005 85 of 90
NOTE With any reactor vessel stud not fully tensioned, reactor is in MODE 6. 1* 4.1.6 Using the following, WHEN RCS temperature is less than 140 OF, REQUEST Maintenance Department detension reactor vessel head and prepare for head lift:
- MP 3790AA, "Preparation for Reactor Head Removal"
- MP 3790AC, "Reactor Vessel Head Removal"
- MP 3790AD, "Reactor Vessel Stud Detensioning Using 3, 4, 5 or 6 Tensioners"
- MP 3790AF, "Heated Junction Thermocouple Removal and Installation"
- MP 3790AQ, "Core Exit Thermocouple Greyloc Seal Removal and Installation"
- MP 3790AT, "Polar Crane Force Measuring System Installation and Removal"
- MP 3790AY, "Checkered Plate Removal, Storage, and Installation" NOTE SteTp 4.1.7 is performed when schedule allows, but must be performed prior to flooding the refueling cavity.
4.1.7 WHEN required, REQUEST Maintenance Department perform MP 3790AL, "Fuel Transfer Thbe Blind Flange Handling," to remove fuel transfer tube blind flange. -I--- 4.1.8 IF desired to not fill the north saddle area, PERFORM the following:
- a. OBTAIN concurrence from the Health Physics Department to lift reactor head prior to filling north saddle area.
- b. CLOSE 3SFC-V906, refueling cavity to aerated drain stop (MIDS 3' platform across from excess letdown heat exchanger room).
- c. Go To step 4.1.11.
Level of Use A OP 3210A Rev. 012-03 Continuous l IRz 10 of 33
4.1.17 STATION the Refueling SRO.
&*A ALARA A**
Radiation measurements should be taken while lifting vessel head to ensure safe radiation limits. NOTE Lifting reactor vessel head marks the start of CORE ALTERATIONS as defined in T/S 1.9, "Definitions, Core Alterations" and continues until it has been verified no control rods are moving with head removal. I CORE ALT :]' 4.1.18 REQUEST Maintenance Department perform MP 3790AC, "Reactor Vessel Head Removal," to lift reactor vessel head to approximately elev. 27' 6" (3' above flange). I CORE ALT ::I' a. Using a camera or flashlight or both, VERIFY all control rod drive shafts are free and not lifting.
- b. WHEN all control rod drive shafts are free and not lifting, NOTIFY Maintenance Department reactor vessel head removal requirements are complete.
V7 CAUTION \/ All personnel must be kept clear of vessel head during movement across refueling deck. NOTE While moving the head, it may be necessary to stop containment purge. 4.1.19 REQUEST Maintenance Department perform MP 3790AC, "Reactor Vessel Head Removal," to perform the following:
- LIFT reactor vessel head out of cavity
- MOVE reactor vessel head over mat access opening
- LOWER reactor vessel head onto storage stand OP 3210A Level of Use l *
- Rev. 012-03 Continuous 12 of 33
4.3.5 IF RHR pump B will be used for gravity fill, PERFORM the following: -I----- a. IF 3RHS*P1B, "PP B," is running, STOP 3RHS*P1B, "PP B." fr b. CLOSE one of the following valves (PLACE a circle around the valve closed):
- 3RHS*MV8702A, "B ISOL (OUT)"
- 3RHS*MV8702B, "B ISOL (IN)"
- 3RHS*MV8702C, "RCS/PP B SUCT ISOL" 1* c. PLACE 3RHS*FK619, "RHR HDR FLOW," in "MAN" and CLOSE (MB2).
4.3.6 STATION an individual at the refueling cavity wall and ESTABLISH communications with the Control Room. NOTE
- 1. Elev. 49' 10" corresponds to approximately centerline of skimmers.
- 2. Attachment 2 shows pressurizer level conversion to elevation.
4.3.7 Refer To Attachment 1 and COMMENCE recording level data. 4.3.8 IF using RHR Train A, COMMENCE filling reactor vessel as follows (MB2): i-I--- a. OPEN 3SIL*MV8809A, "PP A COLD LEG INJ."
- b. OPEN 3SIL*MV8812A, "RWSTIPP A SUCT ISOL."
1 c. THROTTLE 3RHS*HCV606, "HX A FLOW," to achieve any desired fill rate up to a fill rate of approximately 1,000 gpm. 1
- d. IF desired to stop filling cavity using gravity feed prior to achieving normal refueling level of approximately 49'10",
0 PERFORM the following: (MB2)
- 1) CLOSE 3RHS*HCV606, "HX A FLOW."
Lof Use l *
- i OP 3210A Rev. 012-03 Continuous I KI 16 of 33
- 2) CLOSE 3SIL*MV8812A, "RWST/PP A SUCT ISOV' (MB2).
- 3) OPEN the following valves (MB2):
- 3RHS*MV8701B, 'A ISOL (OUT)"
- 3RHS*MV8701A, 'A ISOL (IN)"
- 3RHS*MV8701C, "RCS/PP A SUCT ISOL"
- 4) Go To Step 4.3.14.
- e. IF Train A RHR pump will not be started, WHEN water level in refueling cavity approaches refueling level of approximately elev. 49' 10", PERFORM the following:
(MB2) 0D
- 1) CLOSE 3RHS*HCV606, "HX A FLOW"
- 2) CLOSE 3SIL*MV8812A, "RWST/PP A SUCT ISOL' (MB2).
/~~ 3) OPEN the following valves (MB2):
- 3RHS*MV8701B, "A ISOL (OUT)"
- 3RHS*MV8701A, 'A ISOL (IN)"
- 3RHS*MV8701C, "RCS/PP A SUCT ISOL'
- 4) Go To Step 4.3.14.
- f. WHEN water spills out over the reactor vessel flange and cavity is at the desired level, START 3RHS*P1A, "PP A."
/I--
- g. THROTTLE 3RHS*FK618, "RHR HDR FLOW" to achieve 2,000 gpm.
- h. ADJUST 3RHS*FK618, "RHR HDR FLOW," auto setpoint
/I- to 2,000 gpm.
- i. PLACE 3RHS*FK618, "RHR HDR FLOW," in "AUTO."
- j. CLOSE 3RHS*HCV606, "HX A FLOW" OP 3210A Level of Use l L Ak Rev. 012-03 Continuous 7 1 1 k I 17 of 33 I
- k. IF desired to stop Train A RHR pump prior to achieving normal refueling level of approximately 49'10", PERFORM the following: (MB2)
- 1) CLOSE 3RHS*FK618, "RHR HDR FLOW" Q
- 2) Go To Step 4.3.12.
4.3.9 IF using RHR Train B, COMMENCE filling reactor vessel as follows (MB2):
- a. OPEN 3SIL*MV8809B, "PP B COLD LEG INJ."
- b. OPEN 3SIL*MV8812B, "RWST/PP B SUCT ISOL."
- c. THROTTLE 3RHS*HCV607, "HX B FLOW," to achieve any desired fill rate up to a fill rate of approximately 1,000 gpm.
- d. IF desired to stop filling cavity using gravity feed prior to achieving normal refueling level of approximately 49'10",
PERFORM the following: (MB2)
- 1) CLOSE 3RHS*HCV607, "HX B FLOW"
- 2) CLOSE 3SIL*MV8812B, "RWST/PP B SUCT ISOL' (MB2).
- 3) OPEN the following valves (MB2):
- 3RHS*MV8702A, "B ISOL (OUT)" 0
- 3RHS*MV8702B, "B ISOL (IN)"
- 3RHS*MV8702C, "RCS/PP B SUCT ISOL'
- 4) Go To Step 4.3.14.
- e. IF Train B RHR pump will not be started, WHEN water level in refueling cavity approaches refueling level of approximately elev. 49' 10", PERFORM the following:
(MB2)
- 1) CLOSE 3RHS*HCV607, "HX B FLOW" Level of Use OP 3210A Continuous Rev. 012-03 7r 1pir18 of 33 I
- 2) CLOSE 3SIL*MV8812B, "RWST/PP B SUCT ISOV' (MB2).
- 3) OPEN the following valves (MB2):
- 3RHS*MV8702A, "B ISOL (OUT)" 0
- 3RHS*MV8702B, "B ISOL (IN)"
- 3RHS*MV8702C, "RCS/PP B SUCT ISOL'
- 4) Go To Step 4.3.14.
-I---- f. WHEN water spills out over the reactor vessel flange and cavity is at the desired level, START 3RHS*P1B, "PP B." 1 g. THROTTLE 3RHS*FK619, "RHR HDR FLOW," to achieve 2,000 gpm.
- h. ADJUST 3RHS*FK619, "RHR HDR FLOW," auto setpoint to 2,000 gpm.
- i. PLACE 3RHS*FK619, "RHR HDR FLOW," in 'AUTO."
- j. CLOSE 3RHS*HCV607, "HX B FLOW"
- k. IF desired to stop Train B RHR pump prior to achieving normal refueling level of approximately 49'10", PERFORM the following: (MB2)
- 1) CLOSE 3RHS*FK619, "RHR HDR FLOW" 0
- 2) Go To Step 4.3.13.
4.3.10 While RHR pump is running with flow rate less than 3,000 gpm, RECORD RHR flow rate hourly on OP 3270B-004, "RHR Pump Reduced Flow Operating Log". Level of Use l * -- A OP 3210A Continuous lm r N w i Rev. 012-03 19 of 33
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 94 Tier # 3 Group# 3 K/A # GEN.2.3.1 Importance Rating 3.0 Proposed Question: A maintenance employee at Millstone has the following exposure records for the current year: Total Effective Dose Equivalent (TEDE): 0.94 Rem Dose to the eyes: 18.3 Rem Which dose limits, if any, has this worker exceeded? A. No limits have been exceeded. B. Millstone administrative TEDE limit has been exceeded only. C. Federal TEDE limit has been exceeded. D. Federal dose limit to the eyes has been exceeded. Proposed Answer: D Explanation (Optional): of 5.0 TEDE is less than the Millstone Admin limit of 1 Rem/year TEDE ("B" wrong), and less than the federal limit 18.3 Rem Rem/year ("C" wrong). "D" is correct, since the dose limit for eyes is 15 Rem/year. "A" is plausible since is < limit for extremities (50 R). Technical Reference(s): Rad Worker Training Manual, Ch 3 (Attach if not previously provided) Proposed references to be provided to applicants during examination: None Learning GE-00059 State the Federal radiation limits... (As Objective: GE-00064 State the plant administrative guidelines for radiation dose. available) Question Source: Modified Bank # 75637 Parent attached Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.12 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
ES-40 1 Written Examination Question Worksheet Form ES-40 1-6 Original question 75637 A maintenance employee at Millstone has the following exposure records for the current year: Committed Effective Dose Equivalent (CEDE): 0.95 Rem Deep Dose Equivalent (DDE): 0.06 Rem Dose to the extremities: 18.3 Rem Which dose limits, if any, has this worker exceeded? A. No limits have been exceeded. B. Millstone administrative TEDE limit has been exceeded only. C. Federal TEDE limit has been exceeded. D. Federal dose limit to extremities has been exceeded. Answer: B 41 of 46 NUREG-1021, Revision 8, Supplement 1
I E I RADIATION WORKER TRAINING MANUAL - CHAPTER 3 EXPOSURE LIMITS I FEDERAL LIMITS The Nuclear Regulatory Commission has established federal limits for radiation exposure. Millstone Station sets additional limitations on exposure that are lower than the federal limits. Federal Radiation Exposure Limits are established in Title 10 of the Code of Federal Regulationsi Part 20 - known in short as 10 CFR 20, entitled, 'Standards For Protection Against Radiation." TEDE FEDERAL LIMIT - TEDE GE-00056 GE-00059 The Total Effective Dose Equivalent, or TEDE, is the worker's external exposure plus internal exposure. The external exposure, is to the head, trunk, upper arms, upper legs and gonads, and is called Deep Dose Equivalent or DDE. The Internal exposure is to all the organs, and is called Committed Effective Dose Equivalent or MD . CEDE. TEDE is limited to 5 rem per year. 5 remlyear I .ITEDE I-I limit = 5 rem per year FEDERAL LIMIT - SKIN GE-00059 The skin of the body is less radiosensitive than the whole body. Skin exposure is referred to as Shallow Dose Equivalent or SDE and is limited to 50 rem per year. SDE limit = 50 rem per year Exposure Limits Page 3-3 GE-00006, Rev. 5, 0711612001
RADIATION WORKER TRAINING MANUAL:- CHAPTER 3 EXPOSURE LIMITS FEDERAL LIMIT - EXTREMITIES GE-00059 The extremities of the body, defined as the elbow out to the finger tips, and the E knees out to the toes, are linited ,separately. Since the skin is the critical organ in the extremifies, they are also limited to 50 rem per year. This is also referred to as Shallow Dose Equivalent It should be noted that the head is not an extremity. SDE limit =50 rem per year. 50 remlyear SDE limit 50 rem per year Eye Dose (LDE) FEDERAL LIMIT - EYES GE-00059 Dose to the lens of the eye, is referred to as Lens Dose Equivalent or LDE. The LDE is limited to 15 rem Per year. The lens of the eye is not considered part of the whole body. 15 remlyear LDE limit = 15 rem per year I FEDERAL LIMIT - EMBRYO-FETUS GE-00059 A radiation worker who has voluntarily declared her pregnancy is entitled to a restricted limit. The limit for a declared pregnant radiation worker is 500 millirem for the entire term of the pregnancy. This is established to limit exposure to the embryo-fetus. Declared Pregnant = 500 millirem whole pregnancy Exposure Limits Page 3-4 GE-00006, Rev. 5, 07116/2001
RADIATION WORKER TRAINING MANUAL - CHAPTER 3 EXPOSURE LIMITS. MILLSTONE STATION ADMINISTRATIVE LIMITSGE-ooomt Millstone Station To minimize the risk from radiation exposure and to keep from exceeding the federal frnimts, of tiered limits which has established administrative limits. The administrative exposure limits are a set to exceed. These approvals are known as require sucessively higher levels of management approval be authorized. upgrades. Of course, at no time shall exposure above the federal occupational limits The Millstone Station Administrative Umits are:
- Initially: 1,000 millirem TEDE for the year.
- No more than 3 000 millirem TEDE per year from all licensees combined).
than or equal This is reduced to 1,000 millirem per year if your lifetime dose (in rem) is greater to your age (in years). millirem'total
- These dose assignments must never exceed the federal limit of 5,000 occupational dose per year.
Protection
- Currently, Millstone Station has no administrative limits. SDE or LDE. The Radiation Department evaluates dose limits in these areas on case by case basis.
Exposure limits are raised Exposure may be further limited by the station Radiation Protection Department as evenly as practical incrementally with the supervisor responsible for ensuring that exposure is spread will be further limited for those individuals whose lifetime dose (in among equally qualified workers. Exposure rem) is equal to or greater than his or her age (in years). GE-0066 you to exceed the You are responsible for your own exposure. If you believe that assigned work will cause the work begins. The work may need to be assigned to established limits inform your supervisor before obtained for you. Since the administrative limits have been another worker, or an exposure upgrade established by Millstone Station they- may be exceeded without violating federal law. Exposure Limits Page 3-6 GE-00006, Rev. 5, 0711612001
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 95 (SRO) Tier # 3 Group # 3 K/A # GEN.2.3 .9 Importance Rating 3.4 Proposed Question: Initial conditions:
- The plant has just been cooled down at the start of a refueling outage.
- The crew is preparing to start the Containment Purge System in the "Unfiltered" mode of operation.
- The plan is to open the containment access hatch shortly.
- Outside air temperature is 550 F.
What will be the desired Containment Purge System lineup with the Containment Access Hatch open? A. One supply HVU and one exhaust fan running to prevent excessively cooling down CTMT. B. One supply HVU and two exhaust fans running to keep air flow into CTMT through the access hatch. C. Two supply HVUs and one exhaust fan running to keep air flow out of CTMT through the access hatch. D. Two supply HVUs and two exhaust fans running to maximize air flow in CTMT. Proposed Answer: B Explanation (Optional): One supply HVU and two exhaust fans are desired to be running to keep air flow into CTMT through the access hatch ("B" correct, "C" and "D" wrong) "A" is wrong, since hot water heating is modulated to the HVUs to maintain 70'F outlet temperature. "A" is plausible, since there is a minimum desired CTMT temperature, and only one train is allowed in the filtered mode. "D" is plausible since operating with 2 trains of purge is allowed with the CTMT hatch closed. Technical Reference(s): OP 3313F, section 4.1 (Attach if not previously provided) Proposed references to be provided to applicants during examination: None Learning MC-0426 I Describe the major administrative or procedural precautions placed on (As available) Objective: the operation of the CTMT ventilation systems, and the basis for each. Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.43.4 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
- 4. INSTRUCTIONS 4.1 Start the Containment Purge System (Unfiltered) e3 WARNING This section is normally performed after the containment atmosphere has been cleaned enough to discharge directly to atmosphere without filtration, otherwise use Section 4.2.
4.1.1 IF the Containment must be filtered prior to discharge, Go To Section 4.2. 4.1.2 CHECK Containment is at atmospheric pressure. 4.1.3 REQUEST Chemistry Department to sample containment air and determine if chemistry specifications are satisfied. 4.1.4 To place the containment purge air system exhaust dampers in an unfiltered alignment, PERFORM the following (VP1):
- a. PRESS and HOLD normal pushbuttons for "CTMT PURGE" "EXH DMPRS."
- b. WHEN the following exhaust dampers open, RELEASE "C7MT PURGE" "EXH DMPRS":
- 3HVR*AOD32A
- 3HVR*AOD32B 4.1.5 To open the containment purge air containment isolation valves, PERFORM the following:
- a. OPEN instrument air isolations to the following valves (local):
- 3HVU*CTV32A
- 3HVU*CTV32B
- b. PLACE key switch for 3HVU*CTV32A and 3HVU*CTV32B, "PURGE VVS," to "OPEN" (VP1).
Level of Use rAA OP 3313F Continuous Rev. 009-01 4:7 5 of 21
- c. WHEN 3HVU*CrV32A and 3HVU*CTV32B, "PURGE VVS," are open, return switch to 'AUTO" (VP1).
- d. PLACE key switch for 3HVU*CTV33A and 3HVU*CTV33B, "PURGE VVS," to "OPEN" (VP1).
- e. WHEN 3HVU*CIV33A and 3HVU*CTV33B, "PURGE VVS," are open, return switch to 'AUTO" (VP1).
NOTE During outages with containment access open, it is desirable to run two exhaust fans and one supply fan to keep the flow of air into Containment through the open containment accesses. 4.1.6 IF it is desired to start containment purge Train A, PERFORM the following (VP1):
- a. OPEN the following "CTMT PURGE" "SPLY DMPRS" "HVUlA':
- 3HVR*AOD55A
- 3HVR*AOD174A
- b. START 3HVR-FN4A, "EXH FANS."
NOTE 3HVH-TV1 11A, containment purge air handling unit A hot water valve, modulates to maintain a preset temperature of 70F.
- c. START 3HVR-HVU1A, "SPLY HVU'S."
4.1.7 IF it is desired to start containment purge Train B, PERFORM the following (VP1):
- a. OPEN the following "CTMT PURGE" "SPLY DMPRS" "HVU1B":
- 3HVR*AOD55B
- 3HVR*AOD174B Level of Use AII OP 3313F Continuous Rev. 009-01 6of 21 6 of 21 I
- b. START 3HVR-FN4B, "EXH FANS."
NOTE 3HVH-TV1 11B, containment purge air handling unit B hot water valve, modulates to maintain a preset temperature of 700 F.
- c. START 3HVR-HVU1B, "SPLY HVU'S."
- End of Section 4.1 -
Level of Use -F Rv 00930 Continuous l 7
`
70*~^ 4, I r7 of 21
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 96 Tier # 3 Group # 4 KIA # GEN.2.4.2 Importance Rating 4.1 Proposed Question: The crew is performing a plant startup per OP 3203 "Plant Startup" with reactor power at 4%. Which condition will immediately result in an automatic reactor trip, requiring the crew to enter E-0 "Reactor Trip or Safety Injection"? A. RCS pressure decreases to 1850 psia. B. Pressurizer level increases to 90%. C. 3MSS-PT505 fails high. D. All four RCPs trip. Proposed Answer: A Explanation (Optional): Pressurizer High Level, Pressurizer Low Pressure, and all loop flow trips are automatically blocked below P-7 [Turbine (P- 13) and Reactor (P- 10) < 10% power]. (B and D incorrect). Reactor trip on turbine trip is blocked below P-9 (-50% power). Failure of PT-505 would enable P-7 but not cause a reactor trip. (C incorrect). As pressure decreases to 1892 psia, SI is actuated. The SI signal generates a Reactor Trip Signal. Technical Reference(s): Functional Dwgs 2, 4, 5, 6, and 16 (Attach if not previouslv provided) I. - Proposed references to be provided to applicants during examination: None Learning MC-05493 Describe the operation of the following RPS controls (As available) Objective: and interlocks... Reactor Trip Signals... Protective Interlocks... Question Source: Bank# 69344 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
ROD DRIVE SUPPLY ONE LINE DIAGRAM A 250 V AC ELS
) 52/RTB 5)2VY8
) . ( NOTE 4 )
X-G SET REACTORTRIP SWITEAR FCCCRIVE POWER SiPY ) 52/RTA ) 2/YA RODDRIVE POWERBUS (NOTE I) NOTE 5 LOGIC TRAIN A OPENORNOT CONNECTED P-4 ({NOTESI a 2) CL i - TO SAFETY INJECTION 5[ _BLOC LOGICtSHEET B) (NOTE 4) TO STEAM DUMP 5 OPENOR NOTCONNECTED CONTROL LOGIC (SNEET IC) 5 CLOSED TO FEEDOATER I"ATON0.I
,- E ED(SHEET 1)
~TO TURBINE. TRIP N
L (SHEET IC&)
-------- TRAIN B TO SOLID STATE LtGIC PROTECT ION SYSTEM ITR1A -- A JLOG GENERAL IC WARING ALARM I TO TUREI#ETRIP
- (sREq.T mo) lfl - CONNECTED To STEAM DUMP lONTROL C LOGIC (SWEET 10)
- N , ICLOSED A LL OPENOR NOT CONNECTED TO FEISOATER P. CISOATION~ LOGIC l l (aNOTES IA 2) n (NOTE 4) (SHEET 13)
N OPENOR NOT CONNECTED TO S TY INJECTION P-4 BLOCKLOGIC (SHEET 8) CY a., I *ELOGIC TRAIN B NOTE 5 NOTES: I. TRIPPING THE REACTORTRIP BREAKERS52/RTA AND 52/RTB REcODDANTLY DE-ENIERGIZESTHE ROD DRIVES. ALL FULL LWENTHtOWTPCL RODSAND S4JTtlUI RODS ARE THEPREYRELEASEDFOR GRAVITY INSERTION INTO THE REACTOROGE.
- 2. NORMALREACTOROPERATION IS TO BE WITH REACTORTRIP BREAKERS52/RTA AND 52VRTB IN SERVICE ANDBY-PASS BREARS 2BYA AND52/lrU WITHORIIN.
"JAING TEST, ONE SY-PASS WEAKER IS TO BE PUT IN SERVICE AND THEN THE RESPECTIVE REACTORTRIP BREKER IS tATED USING A SIILATW REACTORTRIP SIGIAL IN THE TRAIN UNDERTEST. THE REACTORWILL MDTBE TPIPPE£ BY THE SIMULATED SIGNAL SINGE TE BY-PASS 8EAER IS CONTROLED FR04 THE OTHERTRAIN. ONLY ONE REACTORTRIP 8FEAER IS TO BE TESTED AT A TIlE.
- 3. ALL CItRUITS ON THIS SMT ARE ICT REllJDANT BECAUSE TRAINS AE SH . "FSAR FIGURE" A. tP9WYODSED INlICATlION FO EACs TRIP W1EAKERAND EAH BYPASS BeAKER IN CtNTIM a.
5, ODEOTF.S AOOITION4AL CIQCUlTRY FOOL SWIlJ4T COIL BACICuKIl DURING REVISIONS I P.A.- CONSTRUCTION [3RPS~sYS]
^ L~i:wwingiose Efectrlc C CA D ._ _ _I_ _ _ _ __N _ _ _ I-_IjI REV~SOBS T IBI D~SUTILI ITIE YE ASRUL ME 5R4E I INCORP OCRDM130006o8 98 RMVj p _~ i S E[IN T HSOCWTI NA M BYE NCR ACNIIN BM-00629 ERM
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z ImHs 1 ve9cls 1 .1 TRAIN A REACTOR SHUNT TRIP SIGNALS MANUALREACTORTRIP SIGNA& (SZET 3) - MANUALSAFETY INJECTION SIGNAL (SHEET 8I) LOGIC TRAIN A REACTOR TRIP SIGNALS MIANUALTRIP SIGNAL (SHEET 3) (GH0 SM)RCIERANGE, HIGH FLUX (INTERL.OCKEDBY P-6 £ P-10) INITERMEtIATE RMNGE.HIGH FLUX CINTEFLOOCED By P-I0) NEUTRON FLUX TRIP SIGAAAS CISHEET 3) FLUX. LOWSETPOINT ( INTERLOCKED BY P-10) POR RANGE H10 FLUX: HI1GHSETPOINI N: HIGHWFLUX RATE OEATEMPERATURE &T OVERPGR OT PRtIWRY Ctt.AT SYSTEM TRIP SIlGtALS (SHEET 5) LON PRIKARY f LCA REACTORCOOLANTFLCU IN ANY I OF LLOOKP(INTERUOCKEO By P-B) COOAN FLOW 'LC REZACTOR COD-AT FLOV IN ANY 2 OF 4 LOOS ( INTERLOCXED BY P.T) P115JIZEt TRIP SIGNALS (SrE 6) (HIGH LRCP SHAFT PRESSUbE LOW SPEZED(INTERLOCKED LON PRESagRE (INTERLOCED BY P-7) BY P-7) NIG) LEVEL (INTEXDCKED BY P-7) STEMUd RATRIR TRIP SIGNA lSHEET 7) {LOU-LOW STEAMGEOERATOR WATERLEVEL NMANUAL SIGNAL TURBINE TRIP SIGtAL (SHEET16) ( LON TRIP FtlUID PRESsaRE OR ALL STOPVALVES CLOSED(INTERLOCKED BY P.9) SOLID STATE PROTECT104 SYSTEIA (GENERAL vARNING ALARt IN BOTH TRAINS 145p
-I mI s LIGNMsk3T LOGIC TRAIN B REACTOR TRIP SIGNALS KVIaUL TRIP SIGHAL (SHEET 3)
SO1JCE RANGE.HIGH FLUX lINTERLOCKED By P-6 6 P-l t) INTERtHEDIATERAfGE. HIGH FLUX (INTEROOEO Y P-D10) IEUTt" FLUX TRIP SIGNLS r N'GI FLUX. LON SETPOINT (INTERLOCKED BY P-I0) (SHET 3) tR RANGE t HIGH FLUX. HIGH SETPOINT HIGH FLUtXRATE OVERTEtPERATURE OT OVER aT PRIwmARY OANT SYSTEt TRIP SIWAAL (SHEET 5) LC5 PRTFLY r LOt REACTORCOOLANTFLOM IN ANY I OF 4 LOOPS ( INTERLOCKED CI HTFL t LOWtREACTORCOCLAbTFLOV IN ANY 2 OF 4 LOOPS (INTERLOCKED BY P-B) BY P.7). RCP SHAFT LtW SPEED (INTERLOCKED BY P-7) PRESSURIZER TRIP SIGtALS (HISW PRESSIRE --- D BY P-7) (SEET 6) (HIGPF LEVEL llINTERLOED BY P-T) STEA G0fEPTOR TRIP SIGOL L (SHEET T)f LOvdL.wSTEAMtGENERATCR WATER LEVEL SAfETY INJECT1t IN.CTI) SIGNtAL ~E~ S04A. (SHTIE 8) ) AUTOMATIC SIGNM. MAUUALSIL TfltINE TRIP SIGtiM csFr I6) (LOt TRIP FLUID PRESSURE OR ALILSrOP VALVESCLOSED (INTERLOOEED BY P.9) S~taUDSTATE PROTECTItNSt~rEFJ ( GENERAL KRANING ALAFHIN B0TH TRAINS SSPS N-I MISAlLI6MMEPtCT TRAIN 8 REAClOR SHUNT TRIP SIGNALS IAMJAL REACTORTRIP SIGNAL (SlEET 3) - MNUWAL SAFETY INJECTION SIGiAL (SHEET S) CA] It 10 9 a 7 76
R_ c POWER RANGE II A irrc I 1I NC 4 55 FROhA ilN 3( _R BYPASS (SHE ET 3) I I C-1
.imH NEOTRON TLVA ROD STOP CV AWUTOM&TIC A MANNU&L -O0
\4DRAWA) (S4wseT 9)
C -2 ERPCwR ROO STOP OC r AUTOMATIC I MANUAL )O wITmDRAwAL-) (SH EEr S)
"FSAR FIGURE"
[3RPS* SYS ] 12179-2472.011-001-004 NOTES: I-THE BYPA5S SIGNALS ARE MADE UP BY MEANS OF Two THREE.POSlTaoW
'ITCHES ON A NIS RACK. SWITCh I/N 45A BYPASSES EITHER. NC-41L NC&43L. SWITCt I/N*58 eYPA55E5 EI'fH&R 14C-*ZL OR NC-*4L.
'- TWO P4 515-TASLE U0. NC-35D AN4D NC-340 ARE'EMERGIIED TO ACTUATE" SUCH THAT A LOGIC I SIGNAL IS DEFINED TO BE PRESENT WHEN THE 515TABLE OUTPUT VOLTAGE IS ON.
5 4 3 25212-39001 SH4005
IPOWERoRANGE POWER RANGE INTERMEDIATE RANGE
/) (X z fI i ---
I I I I R BYPASS (SHEET B) I I Nu ~CUKDA"-r R HIGH NEt Roc (e-.ocv AwUTOc WITMORAWj POWER RANGE POWEP x P-a (SHEETr S) c-i OVERPOWE. (BLOCK AU-i ROD WIT (581 I NOTES CAD I MAMA TOTHIlS RMVISIOVG OcmNT V~I ~IIIITBED. SAE S-IULT ARE RO II I 10 9 9 I 8 7 6
REAZTOR COOLANT PUMP SHAFT LOW SPEED (BDOTE I) RCP I RCP I RCP 3 RCP 4 REACTOR TRIP ( SHEET 2) REACTOR C=OLABT LOOP LOW FTLOW% LOW T^A& (MOTE 1) LCOF 2 LOOP 3 LOOP 4 n4 .\ Tr m _ P-7 (5c
-mET4)
REACTOR TRIP (514EET 2) (SMEET Z) TO FEEDWATEI I$OLATION ($tMET 13)
"FSAR FIGURE" [Q~XA IRlEVISIONSDURING I
CONSTRUCTICNP.A.* [3Rps s'-ys I II II II II II 12179 2472.011 001-005_[_I II I II II I I I .
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- FUNt1ONAL DIAG, 5Ams 1RIM COOLANT SYSTEM TRIP SICNf4L' K_ _ = 5 INCORP 3CN DM3-00-0682-98 RMV DI _ De I: I.ji 1,3T£:
I CAD i l-I JASBUILTPEROCR-MJ3-S-0812-94 DL JhIV6DIE
/08D684 REVISIONS WAkINAL TOTHIS3OCaENr -95 & INCORPOCNDM3-S-0843-94 YHEN AS-BUILTAREPRCIBITED. ._. WWWteM s S FWE.T 5
_ P.A. AR. DATE
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AE A. REVISIONS
- Y CAP.CORPAPP. l0 CT-x sCkuz SU _ .. _
_ I 5_ _ _ U1 3'I 2 - 3_ 25 212- 39 001 -SH-4-0-0 l a 5 4 3 I
___ ewlr (LEA>/LAC4 C OMPE:NATE.) (LEAC/LA(4 COMPENSAT=D) LOOPL LOOP 2 LOOP3 LOOP4 LOOP& LOOP P. Loop 3 nooP 4
- t 7ea wa 'T-% I S @8 z RCAR TRIP RECTOR TRIP (SsfE "(DKr lZE) (SuFrcr2) l !VRTEVPrRkruRE
&T T SRPTme A Ar (LrEAU/LAr. CoftAPE-NATrtZ) (LFAb/LA(. COQN%;,tiSATc>))
TO START TURBINE RUN4BACK AKO BLOCK AUTOMATIC AND MtAULtAL ROO WITHCRAWAk-NOTr REOUNo0ANr LO-LO TAV6 (NOTE I) ).OW TA LOOP I LOOP I LOOP 3 LOOP 4 LOOP I LOOP 2 I Ft,l) --
- 1. LooP BOCCKS FOR TI4IS IJQC-rICrTz TO ALLOW 3LOOl CFER>ATtON PROVIOE 0 5tY VSYLOCV( -WITC.MFES Owt rSPS CkT51tET, I KE-YLOCV4 PER LOOP PER TrRAdl4.
CS(UEET FT) 4 TO FEEDWA-TM (ShEET I NT: MANUAL CAD REVISIONSTO THIS
-'-I EN AS-BUILTAREPi118 11 10 9 B 7 6
prES5W1ZEr LoW PCr'2SLI2 SAFETy =NjrTI'rl~h I. TW REDUNDANT (sb4-r 6) MANUAL 2= CTD1 COISIST$ OF TY* CON4TROLSCkdTHECG4TRC.BOARD.ONEFOREACI TRAIN.
- 2. TWOCOfPUTE INPUTlS ARECONNCTEDTO THIS CIRCUIT, WINDVIDUALFOREACHTRAIN.
- 3. TWOPERMdISSIVE STATUSLIGHTS.ARECONNECTEDMOThIS CIRCUIT. INCIVIWUALFOREACHTRAIN.
- 4. SA9PPLIEO BY OTb&IRS.
"FSAR FIGURE" FRPS w SYS I a 3 4 3
P-7 (SHEEr4) PRESSURIZER HI PRESSURE REACTOR TRIP (SHEETa) I a n mPf NPB PRESSURtIER HIGH PRES5URE t I~ I 11 . I PORVACTUATION (SHEETSI8 & I5S REACTOR TRIP (SHEETZ) PRESSURIZER HIGH WATER LEVEL 4 Lt _P FLEET *) I3 II TO 9 8 7
NOTES: I. THESE SIGNALS INDICATE THE CLOSING OF THE STOP VALVES. POSITION DETECTION IS ACCOMPLISHED BY I CONTACT PER STOP VALVE. TURBINE POWER MULTIPLIED THRU ISOLATING RELAYS TO PROVIDE (TURBINE IMPULSE CHAMBER PRESSURE) ONE INPUT EACH FOR TRAIN A & B PER VALVE. r
- 2. REDUNDANCY IS INDICATED IN REGARDS TO REQUI REHENTS ONLY.
ZI31 3. OPEN/SHUT INDICATION IN CONTROL ROOM.
- 4. GENERATOR MOTORING PROTECTION SHOULD NOT IN DEFEAT THE 30 SECOND DELAY.
- 5. THE REACTOR COOLANT PUMP BUS TRANSFER I PQB I SHCULD BE COMPLETED WITHIN SIX CYCLES IN ORDER TO INSURE COMPATIBILITY WITH THE UNDERFREQUENCY TRIP OF THE REACTOR COOLANT I PUMP CIRCUIT BREAKERS. THE TRANSFER TIME LIMIT MAY BE EXTENDED ANOTHER 4 CYCLES TO A TOTAL OF NOT GREATER THAN 10 CYCLES, IF THE SYSTEM DYNAMICS ARE SUCH THAT A RCP TRIP DOES NOT OCCUR, THE 10 CYCLE LIMIT IS FOR PUMP MOTOR PROTECTION.
- 6. THE 30 SECOND DELAY IS DEFEATED BY TURBINE MOTORING PROTECTION ON THE FOLLOWING
> 3%
>X, < Y5 % SIGNALS: 1. THRUST BEARING FAILURE. 2. LOW VACUUM. 3. HIGH VIBRATION.
- 7. THIS FUNCTION IS USED TO BYPASS THE REDUCED P-Is TEMPERATURE RETURN TO POWER CONTROL I CIRCUITRY TO ALLOW TURBINE TESTING AND/DR I TO DEFEAT THE C-16 ANNUNCIATOR SUCH AS
- 8. L BE RQUIREO DURING A PLANT COOLDOmN.
- 8. THE AMSAC SIQNAL IS NOT REDUNDANT ISOLATION P 13 DEVICES ARE REQUIRED BETWEEN THE NON-1E AM6AC CIRCUITS AND THE iE TURBINE TRIP TO P- 7 BLOCK AUTOMATIC CIRCUITS.
(SHEET 4) ROD WITHDRAWAL 9. TURBINE IMPULSE CHAMBER ANALOG PRESSURE I (s~iE-rSIGNAL FROM SENSORS PT505 AND PT50S ARE _, I IISOLATED FUNC;T BEFORE ROUTING TO AMSAC. DEPICTED RE
-DUNDANT FUNCTIONS PERFORMED BY AMSAC SOFTWARE.
N- NOT REDUNDANT TURBINE RUNIBACK_ VIA LOAD REFERENCE NOT - D MANUAL BYPASS FOR TURBINE TESTING/ PLANT COOLDOWN (MOMENTARY) (NOTE 7) C-16 C-3 LOW TAVG C-4 OVERTEMPERTURE OVERPOWER (SHEET 9) AT (2/4) L T (2/4) E POWFR (SHEET 4) (SHEET 4) ImPLULSE -RESSURE)
- 9) I I
1 0-420 SEC. ON DE-ENG O-1 rI-60 SEC. OFF FL 1F
., -1 t-1-5 SEC.
iAC ON 7, STOP BLOCK REFERENCE REDUCE LOAD REFERENCE TURBINE INCREASE AT 133% PER MINUTE LOADING NOT REDUNDANT /
"FSAR FIGURE" 1
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91 133HSI 1.-Q1>9 STEAM GEN. AMSAC P.4 HI-HIl LEVEL ORSLATAONI REACTOR TRIP CTRAIM B) (SWEET 13) (SFE *) W." %-"L a S. REATO TRI (TRAIN B)(SNEET2) (NOTE 8) (riuim ) tsmr.LTa) (TRAIN A)(SHEETZ) SY WrNr.M. my v _______ TO RZACTrOP rP I - (CS~"T Z) I REDUNDANT I
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 97 Tier # 3 Group # 4 K/A # GEN.2.4.6 Importance Rating 4.0 Proposed Question: A small break LOCA has occurred and no charging or SIH pumps are running. The operating crew has transitioned to EOP 35 FR-C. 1 "Response to Inadequate Core Cooling". Which of the following lists the recovery strategies in the correct sequence for the condition? A. Start ECCS, depressurize secondary, depressurize RCS, start RCPs. B. Start ECCS, depressurize secondary, start RCPs, depressurize RCS. C. Depressurize secondary, start ECCS, depressurize RCS, start RCPs. D. Depressurize secondary, start ECCS, start RCPs, depressurize RCS. Proposed Answer: B Explanation (Optional): FR-C. I step 2: Verify ECCS flow, step 10 depressurizes all intact SGs to 140 psig, step 17 starts RCPs. Step 17 RNO Opens PZR PORV's if RCPs are running. Technical Reference(s): FR-C. 1 steps 2, 10, 17, 17 RNO (Attach if not previously provided) FR-C. I Bkgd Doc, "Major action categories (pg 7), and step 18 Proposed references to be provided to applicants during examination: None Learning MC-04934 PRIORITIZE the operator-initiated recovery techniques that (As available) Objective: mitigate the consequences of a loss of core cooling Question Source: Bank # 65036 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement I
RESPONSE TO INADEQUATE EOP 35 FR-C.1 Page 3 of 18 CORE COOLING Rev. 013 m--n STEP ACTION/EXPECTED RESPONSEI RESPONSE NOT OBTAINED t ~C A UTI0 N
- If the RWST level decreases to LESS THAN 520,000 gal, Go to ES- 1.3, "Transfer to Cold Leg Recirculation," to align the ECCS system.
- DO NOT operate the RHR pumps with RHR heat exchanger inlet temperatures GREATER THAN 195 0 F without RPCCW flow to the heat exchangers.
- 1. Verify ECCS Valve Alignment - Align valves as necessary.
PROPER EMERGENCY ALIGNMENT
- 2. Verify ECCS Flow Perform the following:
- Check charging pumps - a. START pumps and Align FLOW INDICATED valves as necessary.
- Check SI pumps- b. Consult ADTS for FLOW INDICATED recommendations on aligning any available injection source
- Check RHR pumps - to the RCS.
FLOW INDICATED
I1 RESPONSE TO INADEQUATE EOP 35 FR-C.1 Page 12 of 18 CORE COOLING Rev. 013 t HATIO/EXECTD RSPOSEHRESPONSE NOT OBTAIN]ED I NOTE
- During the following steps, partial uncovery of SG tubes is acceptable.
- To allow steam dump operation to continue during a controlled cooldown, ensure the Low-Low Tkvg interlock is bypassed at 553 0F
- Ensure Low Steam Line Pressure SI is blocked wher. rLZR pressure is LESS THAN 2000 psia.
- After the Low Steam Line Pressure SI signal is blocked, MSI will occur if the high steam pressure rate setpoint is exceeded.
- 10. Depressurize All Intact SGs To 140 psig
- a. Dump steam to condenser a. Dump steam to atmosphere from intact SGs at maximum from intact SGs at maximum rate rate using atmospheric steam
- 1) Verify the following: dump valves or atmospheric
- Intact SG(s) MSIV(s) dump bypass valves (MB or locally).
- OPEN
- Annunciator CONDENSER AVAIL FOR STM DUMP C-9 (MB4D 5-6) -
LIT
- 2) Adjust steam pressure controller to obtain zero output in MANUAL
- 3) Transfer condenser steam dumps to Steam Pressure Mode
- 4) Place both condenser steam dump interlock selectors - ON
- 5) Adjust steam pressure controller to dump steam to condenser at maximum rate
RESPONSE TO INADEQUATE EOP 35 FR-C.1 Page 13 of 18 CORE COOLING Rev. 013 m--- ---
- 10. ( ACTIONd)SPON EXPECTED RESPONSE NOT OBTAINEDed
- 10. (continued)
- b. Check all SG pressures - b. Perform the applicable action:
LESS THAN 140 psig I
- IE SG pressures are decreasing, THEN Return to CAUTION prior to step 8.
- IF SG pressures are stable OR increasing, THEN I Proceed to NOTE prior to step 17.
- c. Check RCS hot leg c. Perform the applicable action:
WR temperatures - AT LEAST TWO HOT
- IF RCS hot leg LEG WR TEMPERATURES WR temperatures are LESS THAN 380 0F decreasing, THEN Return to CAUTION prior to step 8.
- IF RCS hot leg WR temperatures are stable QR increasing, THEN I Proceed to NOTE prior to step 17.
- d. Stop SG depressurization
RESPONSE TO INADEQUATE EOP 35 FR-C.1 Page 16 of 18 CORE COOLING Rev. 013 STEP ACTION/EXPECTED RESPONS RESPONSE NOT OBTAINED NOTE Normal conditions are desired but NOT required for starting the RCPs.
- 17. Check If RCPs Should Be Started
- a. Check core exit TCs - a. Proceed to step 18.
GREATER THAN 1200 0F
- b. Check if an idle RCS cooling b. Perform the following:
loop is available
- 1) OPEN all PZR PORVs and
- Check idle loop SG NR block valves.
level - GREATER THAN 8% 2) IF core exit TCs remain (42% ADVERSE CTMT) GREATER THAN 12000F, I THEN
- Check RCP in associated idle loop - AVAILABLE a) OPEN reactor head vent isolation valves:
3RCS*SV8095A 3RCS*SV8095B 3RCS*SV8096A 3RCS*SV8096B b) Open the reactor head vent to PRT isolation valves: 3RCS*HCV442A 3RCS*HCV442B
- 3) Proceed to step 18.
- c. START RCP in one idle RCS cooling loop
- d. Return to step 17.a.
MAJOR ACTION CATEGORIES IN FR-C.1 ) o Establish Safety Injection Flow To the RCS o Rapidly Depressurize SGs to Depressurize RCS o Start RCPs and Open All RCS Vent Paths to Containment FR-C.1 7 HP-Rev. 1C HFRC1
STEP DESCRIPTION TABLE FOR FR-C.1 Step 18 STEP: Check If RCPs Should Be Started PURPOSE: To ensure core exit TC temperatures are greater than 1200 0 F before restarting RCPs BASIS: The operator will enter this step if:
- a. He is unable to depressurize the SGs; or
- b. SG depressurization was not effective in restoring adequate core cooling; or
- c. Secondary heat sink is lost The actions of Step 18 may provide temporary core cooling until some form of makeup flow to the RCS is established or one of the above items is restored.
To temporarily restore core cooling, the operator is instructed to start RCPs one at a time until core exit TCs are less than 1200 0F. The RCPs should force two phase flow through the core, temporarily keeping it cool. Even single phase forced steam flow will cool the core for some time provided the RCPs can be kept running and a heat sink is available. Starting the RCPs in this step when the core exit temperatures are greater than 12000 F will result in the clearing of the water inventory in the RCS intermediate leg (loop seal) and permit the circulation of hot gases from the overheated core to circulate through the steam generators. If the water level in the steam generators is very low at the time the RCPs are started, high steam generator tube temperatures would occur, leading to possible creep failure of the steam generator tubes. Therefore, RCPs are only started in this step if there is sufficient water level in their associated steam generator to protect the steam generator tubes from creep rupture. If RCP restart is not effective in decreasing core exit TC temperatures below 1200 0F, then the PRZR PORVs should be opened. Opening the PRZR PORVs may help reduce RCS pressure enough to cause low-head safety injection. If core exit TCs remain above 12000 F after all PRZR PORVs and block valves are open, the operator is instructed to open all other RCS vent paths to containment to reduce RCS pressure. The pressurizer PORVs require instrument air for long-term operation, however, FR-C.1 50 HP-Rev. 1C HFRC1
STEP DESCRIPTION TABLE FOR FR-C.1 Step .18 instrument air may not be available to the pressurizer PORVs if the event sequence included: a) initiation of a Phase A containment isolation signal, or b) a coincident loss of instrument air. For example, small LOCA event sequences may result in initiation of safety injection, initiation of containment isolation Phase A, subsequent repressurization of the RCS to the pressurizer PORV setpoint and cycling of the PORVs. If the instrument air supply was lost to the pressurizer PORVs, a large volume air receiver located inside containment can provide for limited operation (i.e., number of cycles) of the pressurizer PORVs. Should FR-C.1 subsequently be implemented, by the time that the operator would perform Step 18, the pressurizer PORVs may have lost their ability to open. Hence the operators may not be able to open the PORVs and maintain them open to rapidly depressurize the RCS. To address this possibility, the following actions are performed in the RNO column: o Reset SI signal - The action to reset automatic actuation logic is taken so that safeguards equipment that receive the SI signal may be realigned or reset. o Reset containment isolation Phase A - The action to reset automatic actuation logic is taken so that equipment (e.g., isolation valves) that receive a Phase A signal can be realigned. No valve will reposition upon actuation of the reset, but subsequent control actions will open the valves. Until the cause of the automatic actuation is determined or corrected, Phase A containment isolation valves should remain closed unless required to be opened to establish necessary process streams such as instrument air. o Start one air compressor and establish instrument air to containment - The actions to provide a sustained source of instrument air to containment is taken to support operation of air-operated equipment inside containment such as the pressurizer PORVs. The instrument air system for the ERG Reference Plant includes an air receiver inside containment to allow limited equipment operation, however, the line from the air compressor (located outside containment) to the air receiver is isolated with Phase A isolation. In addition to opening the containment isolation valves, a compressor may also have to be started (with attendant electrical considerations) to establish a sustained source of instrument air to equipment inside containment. FR-C.1 51 HP-Rev. 1C HFRC1
STEP DESCRIPTION TABLE FOR FR-C.1 Step 18 ACTIONS: 0 Determine if core exit TCs are greater than 12000 F 0 Determine if an idle RCS cooling loop is available I 0 Determine if narrow range SG level is greater than (M.02)% I adverse containment] [(M.03)% for I 0 Determine if RCP in associated loop is available and not oF I perating I 0 Reset SI 0 Reset Containment Isolation Phase A I 0 Establish instrument air to containment I 0 Start one air compressor and establish instrument air to cc)ntainment I 0 Open all PRZR PORVs and block valves I 0 Open all other RCS vent paths to containment 0 Start RCP in one idle RCS cooling loop 0 Determine if core exit TCs remain greater than 12000 F INSTRUMENTATION: o Core exit TC temperature indication o RCP status indication o RCP support condition status indication o SG narrow range level indication o SI signal indication I o Containment Phase A indication I o Containment isolation valve position indications I o Air pressure indications I o Air compressor status indications I o PRZR PORV position indication I o PRZR PORV block valve position indication o Plant specific instrumentation to determine open/closed status of other RCS vent paths CONTROL/EOUIPMENT: o RCP switches o RCP support equipment controls o PRZR PORV switches o PRZR PORV block valve switches o Plant specific controls to open other RCS vent paths o SI reset switch o Containment Phase A reset switch I o Containment isolation valves switches I o Air compressor control switch I I I FR-C.1 52 HFRC1 HP-Rev. iC
STEP DESCRIPTION TABLE FOR FR-C.1 Step 18 KNOWLEDGE: Understanding of RCP behavior under forced single phase steam and two-phase flow conditions PLANT-SPECIFIC INFORMATION: o Other RCS vent paths to containment o (M.02) SG level just in the narrow range including allowances for normal channel accuracy and reference leg process errors. o (M.03) SG level just in the narrow range including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%. FR-C.1 53 HP-Rev. 1C HFRC1
ES-40 1 Written Examination Question Worksheet Form ES-40 1-6 Examination Outline Cross-reference: Question # 98 (SRO) Tier # 3 Group# 4 K/A # GEN.2.4.28 Importance Rating 3.3 Proposed Question: A hostile force gains access to the Protected Area and commits several acts of sabotage before taking refuge in the old Unit 1 Gas Turbine building where they are currently surrounded by security. The Unit 3 Shift Manager decides to implement a precautionary dismissal in response to this security event. Which of the following items, if not completed, would delay the precautionary dismissal? A. Notification of local law enforcement agencies is not completed. B. Personnel accountability is not completed. C. SERO activation is not completed. D. A security assessment has not been completed. Proposed Answer: D Explanation (Optional): "A" wrong: a good practice but not a procedural requirement. "B" wrong: personnel accountability occurs after the dismissal. (step 2.1.6) "C" wrong SERO activation should be considered but will not delay dismissal (step 2.1.1 .c). "D" correct, security assessment must be completed (step 2.1.1 .a) Technical Reference(s): MP-26-EPI-FAP08, Evacuation and Assembly (Attach if not previously provided) Proposed references to be provided to applicants during examination: None Learning Objective: MC-05571 Describe the Shift Manager's responsibilities during a (As available) security event. Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
- 2. INSTRUCTIONS 2.1 Precautionary Dismissal 2.1.1 Assess the nature, probable cause, and duration of the hazard and perform the following:
- a. IF event is security related and a Security assessment has not been completed, delay the dismissal until the assessment is completed by Security.
- b. IF event is not security related OR a Security assessment had been completed, provide the SSS/MOS with all available information.
- c. Consider the status of SERO activation prior to the dismissal of personnel.
2.1.2 Contact SSS/MOS and MRCA to discuss the following:
- Additional personnel assigned to the NAP and SAP to assist in the egress of large numbers of personnel as necessary.
- Estimated time to pre-position personnel to support the dismissal.
- Existence of any local area or site access restrictions. l
- Need to sweep areas outside the protected area.
2.1.3 Notify the following of planned actions and announcements:
- a. The DSEO and the ADTS if the SERO is in the process of activation.
- b. The unaffected unit control room.
2.1.4 Perform the following:
- a. Activate the outside speakers.
- b. Select station public address system (priority page or 8 10).
- c. Announce the following:
Attention all personnel. Attention all personnel. All non-SERO employees, contractors and visitors leave the site at this time. d, Repeat the announcement.
- e. Log the time the announcement was completed.
MP-26-EPI-FAP08 Rev. 000-04 7 of 21
2.1.5 IF the public address system is inoperable, consider using the following as alternatives for personnel notification: a Security sweeps using bull horns Q
- HP personnel
- O&M radios 2.1.6 WHEN the precautionary dismissal has been completed, DIRECT SSS/MOS to perform accountability. 10 MP-26-EPI-FAP08 Rev. 000-04 58 f21`
)
2.2 Sheltering 2.2.1 IF the event involves a situation where an evacuation may not be possible (i.e., Security-related, weather-related, fire, toxic gas), perform the following:
- a. Consider the following and determine the nature of the constraint:
- Not enough time to conduct an evacuation (weather-related, rad release).
- Short-lived hazard (chemical, toxic gas)
- Radiological release
- Evacuation would threaten the safety of the evacuees
- Intrusion by a hostile force
- b. IF SERO is staffed, contact the following to discuss course of action:
- For radiological-related, MRCA
- For security-related, MOS
- EOF DSEO
- c. Inform unaffected unit of the event and sheltering actions planned.
2.2.2 Refer To Attachment 3, "Examples of On-Site Protective Actions and Announcements," and prepare announcement. 2.2.3 Ensure outside speakers are activated. 2.2.4 Select the station public address system (priority page or 810) and announce the sheltering instructions. 2.2.5 Log the time of announcement. MP-26-EPI-FAP08 Rev. 000-04 9 of 21
2.3 Evacuation NOTE Evacuation is automatically conducted at a Site Area Emergency or General Emergency unless constraints exist. Other situations which involve the 4 01 evacuation of personnel from occupied localized areas onsite must be controlled on a case by case basis. 2.3.1 Assess the nature, probable cause, and duration of the hazard. 7/ C A U TIO N\v Movement of personnel should consider potential on-site and off-site constraints.
- a. IF the evacuation is constrained (i.e. security related, weather related, fire or toxic gases) consider delaying evacuation until an assessment has been completed.
2.3.2 Direct the SSSIMOS to perform the following:
- a. Inform Waterford Dispatch of time and purpose of any planned on-site siren activation.
- b. Establish and maintain traffic control with the Waterford and Connecticut State Police departments.
2.3.3 Perform the announcement over the public address system as follows:
- a. Activate the outside speakers.
- b. Sound the Evacuation Alarm for 30 seconds.
- c. Select station public address system (priority page or 8110).
- d. Announce the following:
Attention all personnel, Attention all personnel, All non-SERO employees, contractors, and visitors evacuate the site at this time. Security initiate accountability. 0
- e. Repeat the announcement.
- f. Log the time of the announcement.
MP-26-EPI-FAP08 Rev. 000-04 10 of 21 . .
2.3.4 IF public address system is inoperable, consider using the following as alternatives for personnel notification:
- Security sweeps using bull horns
- HP personnel
- O&M radios 2.3.5 Direct the SSS/MOS to perform the following:
- a. Coordinate security patrols to sweep the open areas, outdoors, and buildings outside the Protected Area to ensure the message has been received.
- b. Verify personnel are moving as instructed and report back on the status.
- c. Provide accountability results within 30 minutes if not previously conducted.
MP-26-EPI-FAP08 Rev. 000-04 . 11 of 21
2.4 Accountability 2.4.1 Upon direction of the ADTS or declaration of a Site Area Emergency or General l( Emergency, direct CAS to implement accountability procedures. 2.4.2 Within 15 to 25 minutes after station announcement, ensure CAS has run an area summary report or similar printout to account for personnel in the protected area. 2.4.3 Within 30 minutes of the announcement to conduct accountability, perform the following:
- a. Obtain the missing persons report.
- b. Determine the approximate number of personnel who are unaccounted for by badge or telephone call.
- c. Notify the ADTS of the results.
2.4.4 IF personnel are unaccounted for in the Protected Area, provide the ADTS with the following:
- Name of missing individual
- Last known location of missing individual
- Special access requirements for intended search and rescue route NOTE Announcement by name in 45 minutes fulfills the initial accountability commitment.
2.4.5 Announce the names of unaccounted personnel over station PA system. 2.4.6 Coordinate with the MOSC to initiate the dispatch of Search and Rescue Teams to locate any unaccounted for personnel. 2.4.7 Maintain continuous accountability of personnel within the protected area until directed otherwise by the ADTS. MP-26-EPI-FAPO8 Rev. 000-04 12 of 21
2.5 Assembly 2.5.1 Dispatch the ERC to the Simulator Foyer Assembly Area. 2.5.2 Direct the ERC (in the Simulator Foyer) and the MOSC (in the OSC Assembly Area) to establish a roster of personnel which contains the following information:
- a. Name
- b. SERO position
- c. Home or point of contact number 2.5.3 IF any minimum staffing positions are not filled, perform the following:
- a. Obtain qualified personnel from the OSC Assembly Area or Simulator Foyer Assembly Area.
- b. IF qualified personnel are not available from the Assembly Areas, Refer To MP-26-EPA-REF08B, "Millstone Emergency Plan Resource Book,"
and notify an individual for each unstaffed position. J(
- Refer To EPI-FAPIS-Ol 1, "Fitness for Duty Questionnaire," and determine if notified personnel are fit for duty.
- IF notified personnel are determined fit for duty, request personnel to report to the EOF.
- Upon arrival, coordinate access for the responders into the Protected Area with Security as necessary.
2.5.4 Determine if any of the emergency facilities require the assistance of additional personnel and coordinate their movement as necessary. 2.5.5 IF any augmented positions are not filled, perform the following:
- a. Obtain qualified personnel from the OSC Assembly Area or Simulator Foyer Assembly Area.
- b. IF qualified personnel are not available from the Assembly Areas, Refer To MP-26-EPA-REF08B, "Millstone Emergency Plan Resource Book," l 0 and notify an individual for each unstaffed position.
- Refer To EPI-FAP15-01 1, "Fitness for Duty Questionnaire," and determine if notified personnel are fit for duty.
- IF notified personnel are determined fit for duty, request personnel to report to the EOF.
- Upon arrival, coordinate access for the responders into the Protected Area with Security as necessary.
2.5.6 Begin a first relief roster and schedule for the Emergency Response Facilities from the assembled personnel. MP-26-EPI-FAP08 Rev. 000-04 13 of 21
2.5.7 Release personnel from the Assembly Areas as soon as possible as follows:
- a. Ensure personnel released from the Assembly Areas are directed to Stand-by their point of contact for further information and instructions.
- b. Coordinate the release of personnel with the MOS (Security is in contact with local law enforcement for egress and access logistics).
- c. Inform the DSEO when all personnel have been released from the Assembly Areas.
2.5.8 Discuss establishing a staging area for personnel and resources outside the 10 mile EPZ with DSEO as conditions warrant. MP-26-EPI-FAP08 Rev. 000.04 14 of 21
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 99 (SRO) Tier # 3 Group # 4 K/A # GEN.2.4.36 Importance Rating 2.8 Proposed Question: A reactor trip has occurred involving a radiation release, an ALERT, C-I has been declared, and SERO has been activated. What actions are the responsibility of the chemistry technician once he arrives in the control room? A. Refer to EPI-FAP06, "Classifications and PARs", and determine if a PAR is required. B. Access OFIS to obtain and provide data to the TIC on the status of the offsite dose release. C. Perform the initial dose assessment using "IDA". D. Conduct in plant surveys and sample analysis Proposed Answer: C Explanation (Optional): "A" is wrong since the CR DSEO is responsible for PARs. "B" is wrong, since the CRDC will be accessing OFIS and communication with the TIC. "C" is correct since this is the responsibility of the Chemistry Technician. "D" is wrong, since this is the responsibility of RMTI. Technical Reference(s): MP-26-EPI-FAPO 1, section 1.4.4 (Attach if not previously provided) MP-26-EPI-FAP1O, section 2.3 and Att. 2 Proposed references to be provided to applicants during examination: None Learning MC-02534, The Shift Manager and Unit Supervisor will perform all administrative (As Objective: actions necessary to protect the public in accordance with emergency plan procedures. available) Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
- d. Subsequent Actions Following invocation of 50.54(x) and notification of the NRC, actions are taken as soon as practical to restore the plant to full compliance with Technical Specifications and all conditions of license.
1.4.3 Radiological Monitoring Team #1 During initial SERO activation, RMT #1 provides Control Room health physics support and conducts in-plant surveys and sample analysis. Upon full SERO activation, the MRCA assumes control of the RMT #1 members. An RMT #1 member will report to the MCRO for the duration of the event. 1.4.4 Initial Dose Assessment The Initial Dose Assessment (IDA) computerized method provides the capability to perform a dose projection using effluent release information and real-time meteorology. For the purposes of calculating a total integrated TEDE, a default release duration of 2 hours may be assumed. This assumption corresponds to a period within which SERO activation will occur and a more refined dose assessment can then be performed. This assessment is performed by a Chemistry Technician after a radiological release has occurred and all required actions critical to mitigating the plant event are completed or determined to be of a severity less than the need for performing an initial dose assessment. This is acceptable because initial EALs and PARs will be based upon plant conditions. IDA is used only as a supplement to the initial recommendations. Input provided to the CR-DSEO may be used to validate the initial protective action recommendation or classification. Event classification, off-site agency notifications, and protective action recommendations made by the CR-DSEO should not be delayed by awaiting the results of this dose assessment. 1.4.5 OFIS OFIS provides critical plant parameters to allow communication of plant data for analysis of plant conditions. OFIS may be accessed from LAN PCs. 1.4.6 Definitions and abbreviation are contained in Attachment 1. 1.4.7 Responsibilities are contained in Attachment 2. MP-26-EPI-FAPOI Rev. 001-01 5 of 12
2.3 Control Room IDA Dose Calculations NOTE A back-up computer is located in the TSC if the Control Room PC is not available. 2.3.1 Select IDA icon from the designated Control Room PC. 2.3.2 Refer To EPI-FAP10-001, "IDA - Data Input Information," Part 1, and enter the following on the "Accident Description" screen:
- Unit affected
- Accident type
- Fuel damage state
- IF applicable, containment sprays "YES" (on) or "NO" (off)
NOTE
- 1. If reactor is still critical, the reactor shutdown date and time should be left blank.
- 2. If a fuel drop accident, most recent refueling date and time must be estimated by the CR-DSEO and entered.
- Current (now) and reactor shutdown date and time
- Release duration (2 hour default unless instructed otherwise by the CR-DSEO) 2.3.3 Select "Next."
2.3.4 Refer To EPI-FAPI0-001, "IDA - Data Input Information," Part 2, and enter all of the following on the "Meteorology" screen:
- Wind speeds from the 033', 142', and 374' elevations
- Wind directions from the 033', 142', and 374' elevations
- Delta temperatures from the 142' and 374' elevations NOTE
- 1. If the unmonitored ground release pathway is selected, no other release pathway can be selected.
- 2. IDA can accept up to two NON GROUND release pathways.
- 3. If multiple NON GROUND release pathways are chosen, only the two LOWEST elevation pathways are entered.
2.3.5 Select "Next." __ MP-26-EPI-FAPI0 Rev. 001-01 9 of 27
2.3.6 Refer To EPI-FAPlO-001, "IDA - Data Input Information," Part 3, and enter the following on the "Release Pathways" screen:
- Active release pathways
- Filters operating, if applicable
- Number of safeties releasing, if applicable
- Flow rates using default values or OFIS 2.3.7 Select "Next."
NOTE Plant monitor data is zeroed if unmonitored field team data is entered. 2.3.8 Refer To EPI-FAP10-001, "IDA - Data Input Information," Part 3, and enter the following on the "Monitor" screen:
- Applicable radiation monitor readings
- Applicable field team reading (If unmonitored release) 2.3.9 Select "Finish."
2.3.10 Press "Printer" icon and select "All." 2.3.11 Press "OK" to print output. 2.3.12 IF printer is not available, Refer To EPI-FAP10-003, "Doses for Protective Action Recommendation," and manually record data. 2.3.13 Attach EPI-FAP10-001, "IDA - Data Input Information," to printed output or to EPI-FAP1O-003, "Doses for Protective Action Recommendation." 2.3.14 Refer To EPI-FAP10-001, "IDA - Data Input Information," and perform verification of input data from Output Summary. 2.3.15 Submit results to the CR-DSEO. 2.3.16 IF warranted by changing conditions, repeat Sections 2.3 and notify CR-DSEO of changes. MP-26-EPI-FAPIO Rev. 001-01 10 of 27
I Attachment 2 Responsibilities (Sheet I of 1) Manager of Radiological Dose Assessment (MRDA) - Responsible for determining when the Emergency Operations Facility will assume offsite dose assessment responsibilities from the Control Room and for performing IDA dose calculations as necessary. Radiological Assessment Engineer (RAE) - Responsible for performing the appropriate calculations. On-Shift Chemistry Technician - Responsible for performing initial dose assessment if available until relieved by the MRDA. MP-26-EPI-FAPI0 Rev. 001 -01 22 of 27
ES-401 Written Examination Question Worksheet Form ES-401-6 Examination Outline Cross-reference: Question # 100 Tier # 3 Group # 4 K/A # PLANT SPECIFIC Importance Rating N/A Proposed Question: The crew is responding to a tube rupture, and the following sequence of events occurs: 0930 The crew enters E-3 "Steam Generator Tube Rupture". 0933 The RHR pumps are both stopped and placed in AUTO. 0939 The SIH pumps are both stopped and placed in AUTO. 0940 The "B" CHS pump is stopped and placed in AUTO. 0942 The normal charging flowpath is aligned. 0943 Both charging cold leg injection valves are closed. In accordance with OP 3272 "EOP User's Guide", at what time is the foldout page "SI Reinitiation Criteria" initially in effect? A. 0939 B. 0940 C. 0942 D. 0943 Proposed Answer: D Explanation (Optional): SI reinitiation criteria are not in effect until SI has been terminated. SI is considered terminated when ALL of the following have occurred: RHR Pumps stopped and in AUTO, SI pumps stopped and in AUTO, one charging pump running and aligned to the normal charging flowpath, and both charging pump cold leg injection valves closed ("D" correct). All distractors are plausible, since all of the actions are in accordance with E-3, and are involved with terminating SI. Technical Reference(s): OP 3272, section 1.5 (Attach if not previously provided) E-3, steps 12, 19-21 Proposed references to be provided to applicants during examination: None Learning MC-04461 Explain the usage of foldout pages within the Emergency (As available) Objective: Operating Procedure network Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 55.43.5 Comments: 41 of 46 NUREG-1021, Revision 8, Supplement 1
too 1.5 Using the Foldout Page The Foldout Page is used in the EOP Network to remind the Operator of certain parameters or conditions which must be continuously monitored throughout the performance of a particular Optimal Recovery Procedure. It provides a mechanism to address potentially unexpected plant responses that may occur at any time during the performance of a specific ORP. If any of the parameters or conditions specified on the Foldout Page are satisfied, the operating crew performs the action(s) provided on the Foldout Page. Although the items on a Foldout Page are applicable at all times once it is required to be opened, the operating crew must ensure completion of all immediate actions prior to the implementation of any Foldout Page item. Also, some Foldout Page action items can not be performed until the overriding condition requiring the action(s) becomes effective. For example, Control Building ventilation realignment can not become necessary if CBI has not occurred. Likewise, SI reinitiation criteria are not in effect until SI has been terminated. SI is considered terminated when all of the following have occurred:
- RHR pumps stopped and in AUTO, or running in the cooldown mode
- SI pumps stopped and in AUTO
- Only one charging pump running and aligned to the normal charging flow path
- Both charging pump cold leg injection valves closed Foldout Pages are used to aid optimal recovery in the absence of challenges to the plant safety state. They are, therefore, only used for the Optimal Recovery Procedures to detect changing symptoms and to provide procedure transitions in response to new symptoms as they appear. The Foldout Page is continuously monitored while in the Optimal Recovery Procedure to which it applies. Since the Status Trees are continuously monitored while in a RED or ORANGE path FRP (see Section 1.7, Monitoring Status Trees), there is no need for Foldout Pages for the Function Restoration Procedures. However, since the ORPs have priority over Yellow path Functional Restoration Procedures, Foldout Page items of the ORP in effect remain applicable when implementing a Yellow path FRP OP 3272 THNX F~FW Rev. 008 10 of 44
STEAM GENERATOR EOP 35 E-3 Page 15 of 39 TUBE RUPTURE Rev. 017 STACTION/EXPECTED RESPONS RESPONSE NOT OBTAINED I CAUTION To provide adequate ECCS flow, RCS pressure should be monitored to ensure that the RHR pumps are manually restarted if pressure decreases in an uncontrolled manner to LESS THAN 300 psia (500 psia ADVERSE CTMT).
- 12. Check If RHR Pumps Should Be Stopped
- a. Verify RCS pressure - a. Proceed to step 13.
GREATER THAN 300 psia (500 psia ADVERSE CTMT)
- b. STOP RHR pumps and Place in AUTO
- 13. Check If Cooldown Should Be Stopped
- a. Check Cooldown - a. Proceed to step 13.d.
IN PROGRESS ____b. Check core exit TCs - b. WHEN LESS THAN REQUIRED Core exit TCs are less than TEMPERATURE required temperature, THEN Proceed to step 13.c.
- c. Stop RCS cooldown
- d. Maintain core exit TCs -
LESS THAN REQUIRED TEMPERATURE
STEAM GENERATOR EOP 35 E-3 Page 22 of 39 TUBE RUPTURE Rev. 017 ACTION/EXPECTED RESPONS STP RESPONSE NOT OBTAINED CAUTION Voiding in the upper head region shall NOT preclude SI termination. SI MUST be terminated when termination criteria are satisfied to prevent overfilling of the ruptured SGs. S \\\\\\\\\\\\\\\\\\\\\\\N\\\
- 19. Check If ECCS Flow Should Be Terminated
- a. Verify RCS subcooling based a. Go to ECA-3.1, SGTR With on core exit TCs - Loss of Reactor Coolant -
GREATER THAN 32 0F Subcooled Recovery Desired. (115F ADVERSE CTMT)
- b. Verify secondary heat sink: b. Go to ECA-3.1, SGTR With Loss of Reactor Coolant -
- Total feed flow to SGs - Subcooled Recovery Desired.
GREATER THAN 530 gpm AVAILABLE OR
- NR level in at least one intact SG -
GREATER THAN 8% (42% ADVERSE CIMT)
- c. RCS pressure - c. Go to ECA 3.1, SGTR With STABLE OR INCREASING Loss of Reactor Coolant -
Subcooled Recovery Desired.
- d. PZR level - d. Return to CAUTION prior to GREATER THAN 16% step 5.
(50% ADVERSE CTMT)
- 20. Stop ECCS Pumps
- STOP SI pumps and Place in AUTO
- STOP all but one charging pump and Place in AUTO
STEAM GENERATOR EOP 35 E-3 Page 23 of 39 TUBE RUPTURE Rev. 017 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINEDl
- 21. Establish Normal Charging Flow Path
- a. Fully Open charging line flow control valve
- b. Verify charging header loop b. Re-position valves to isolation valves establish only one open.
(3CHS*AV8146 or 3CHS*AV8147) - ONE OPEN __c. OPEN charging header isolation valves
- 3CHS*MV8106
- 3CHS*MV8105
- d. CLOSE the charging pump miniflow isolations to the RWST
- 3CHS*MV8511A
- 3CHS*MV8511B
- e. CLOSE both charging pump cold leg injection valves
- 3SIH*MV8801A
- 3SIH*MV8801B
- f. OPEN the charging pump recirculation isolation valves
- 3CHS*MV8111A
- 3CHS*MV8111B
- 3CHS*MV8111C
- 3CHS*MV8110
- 22. Control Charging Flow To Maintain PZR Level}}