ML12342A038

From kanterella
Jump to navigation Jump to search
Initial Exam 2012-301 Draft RO Written Exam
ML12342A038
Person / Time
Site: Hatch  Southern Nuclear icon.png
Issue date: 12/06/2012
From:
NRC/RGN-II
To:
Southern Co
References
Download: ML12342A038 (513)
v • d • e


Text

Controlled Copy I RO Questions 1 -

SE ERIAL HLT 7 NRC Exam If found unattended, IMMEDIATELY notify:

Charlie Edmund, Anthony Ball, Ray Rutan or Ed Jones at ext. 3123

HLT-07 SRO NRC EXAM

1. 201001K1.11 001 The Unit 2 reactor is operating at 100% power with the following indications present on panel 2H1 1-P603:

o Charging Water Pressure 1480 psig o Cooling Water dP 0 psid o Drive Water dP Upscale o Drive Water Flow 0 gpm o Cooling Water Flow 5 gpm With the above indications, which ONE of the choices below completes the following statements?

The CRD has been fully closed.

RWCU Seal Purge flow isolated?

A. Flow Control Valve, 2C1 1-FOO2A; is B. Flow Control Valve, 2C1 1-FOO2A; is NOT C. Drive Water Pressure Control Valve, 2C1 l-F003.

is D Drive Water Pressure Control Valve, 2C11-F003, is NOT I

HLT-07 SRO NRC EXAM

==

Description:==

The Drive Water Pressure Control Valve maintains drive water pressure approximately 260 psig above reactor pressure to allow for rod movement. Constant drive pressure is desired so that rod speed will remain constant. The Drive Water Pressure Control Valve is a manually controlled, motor-operated valve. Closing this valve fully increases Drive water dP and isolates Cooling water flow. With the conditions given, the F003 is fully closed. If the FOO2A is fully closed, very little drive water dP and cooling water flow will exist.

Rx Recirc Pump and RWCU Pump seal purge flow lines tap off between the Drive Water Filters and the CRD system Flow Element, which are up stream of the FOO2A & F003 valves.

The A distractor is plausible if the applicant confuses the response of the F003 valve closing with the FOO2A valve closing. The second part is plausible if the applicant confuses the interrelationship between RWCU and the CRD System. RWCU taps off upstream of the FCV FOO2A and will still be supplied seal water.

The B distractor is plausible if the applicant confuses the response of the F003 valve closing with the FOO2A valve closing. The second part is correct.

The C distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses the interrelationship between RWCU and the CRD System. RWCU taps off upstream of the FCV FOO2A and will still be supplied seal water.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

2

HLT-07 SRO NRC EXAM

References:

NONE K/A:

201001 Control Rod Drive Hydraulic System Ki. Knowledge of the physical connections andlor cause effect relationships between CONTROL ROD DRIVE HYDRAULIC SYSTEM and the following:

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1 .11 Reactor water cleanup pumps: Plant-Specific 2.8 2.8 LESSON PLAN/OBJECTIVE:

C11-CRD-LP-00101, Control Rod Drive System, EO 001.005.A.1O & EO 001.005.A.03 References used to develop this question:

H-26006 & H-26007, CRD P&]EDs Unit One CRD P&lDs H 16064 and H16065 3

C11-CRD-LP-OO1O1-09 Page 4 of 120 CONTROL ROD DRIVE SYSTEM Initial License (!J ENABLING OBJECTIVES

1. Given a simplified drawing or P&ID of the CRD Hydraulic System, LOCATE the following system interfaces: (001 .005.A.03) a) Condensate System b) CST c) RWCU System d) Reactor Recirculation System e) RBCCW f) Instrument Air (Non-Interruptible Essential Instrument Air) g) Equipment drains
2. STATE the reason the Condensate System is the preferred CRD suction source. (001 .005.A.08)
3. Given a simplified drawing or P&ID of the CRD Hydraulic System, IDENTIFY the following components: (001 .005.A.02) a) CRD Suction Filter b) CRD Pumps c) Drive Water Filters d) Y Strainers DOO4AJB e) Flow Element N003 f) Charging Water Line Restricting Orifices g) Charging Water Line Isolation Valve F034 h) Flow Control Valves FOO2A/B i) Drive Water Pressure Control Valve F003 j) Return Line Pressure Control Valve F005 k) Cooling Water Pressure Control Valve Fl27
1) Stabilizing Valves m) HCUs n) Scram Discharge Volume
4. Given a list of electrical buses, SELECT the correct power supply for the A or B CRD Pump. (001 .005.A.04)
5. Given a list, SELECT the CRD pump trips according to 34S0-Cl 1-005-1/2 Control Rod Drive Hydraulic System.(001.005 .A.09)

C11-CRD-LP-OO1O1-09 Page 5 of 120 CONTROL ROD DRIVE SYSTEM

6. From a list, SELECT the function of the following CRD system components: (00l.005.A.l0) a) Flow Control Valves b) Drive Water Pressure Control Valve c) Return Line Pressure Control Valve d) Cooling Water Pressure Control Valve e) Stabilizer Valves f) HCUs g) Directional Control Valves (EP12O, EP121, EP122 and EP123) h) Scram Valves i) Scram Pilot Solenoid Valve j) Scram Accumulator k) Scram Discharge Volume
7. Given a simplified drawing or P&ID of an HCU, IDENTIFY the following components:

(001 .002.A.0l) a) Scram Valves (EP126, EP127) b) Scram Solenoid Valves (EP117, EP118) c) Insert Riser Isolation Valve (EP1O1) d) Withdraw Riser Isolation Valve (EPIO2) e) Cooling Water Isolation Valve (EP1O4) f) Exhaust Header Isolation Valve (EP1O5) g) Drive Water Isolation Valve (EP1O3) h) Directional Control Valves (EP12O, EP121, EP122, EP123) i) Charging Header Isolation (EP1 13) j) Nitrogen Bottle k) Accumulator

8. Given a simplified drawing or P&ID of an HCU, TRACE the flowpaths for the following operations: (001 .OlO.A.05) a) Rod Insertion b) Rod Withdrawal c) Scram
9. STATE the function of the Speed Control Needle Valves located on Directional Control Valves EP12O and EP123. (001.018.A.04)
10. Given a simplified drawing or P&ID of the Scram Air Header and the Scram Discharge Volume, IDENTIFY the following components: (001.013 .A.08) a) SDV Drain Valves FOl 1 and HJ37 b) SDV Vent Valves FO1OAIB and FO35AIB c) SDV Vent and Drain Solenoid Valves F009 and F040 d) SDV Vent and Drain Test Valve F008
11. Given a list of HCU parameters, SELECT the parameters which will cause a CRD Accumulator Press Low or Level High annunciator. (OOl.0l6.A.0l) t

g(145#)

4Th n (1220- 1550#)

164 2N21 Condensate (Downstream Of Deinins)

CST 5 Orifices Installed Recirc to Prevent Runout 12 HgVac Seals During a SCRAM (10 Hg-30#)

Keep fi1l System Trouble NOTE: Low Pissui Setpoints in parentheses are normal values. 955 psig Setpoints NOT in parentheses are trip values. RWCU Hi Level:

Seals 60cc 12 Cooling Line Cooling Water Pressure Control Station C Driving Flow U Control Station (0 GPM Not Driving 3-5 GPM Insert 1-3 GPM WID)

CRD Hydraulic System Simplified (Unit 2)

C11-CRD-LP-OO1O1 FIGOI Page 100 of 120

0001:

0909)-H O 01. 0000001000 r40700010.1s 000 1001100) 11 0,01000040001

.030107, CII 404,00,1 61700000 5010(7 00e ------------

0301010 030.00 SCOlDS POlO 00 1 00110_c 0.404002 0. 10440 000000 100371 00 000.0 001Cr 0007.. 1040 P0000 0 0 0,010. -PD 00504 04 0 0 1100 030040 OCt00) (000.

1. 0. 5000105 00030050 3)40010 000 0(0430 0. (1)) 0 (37).

I 000070$ 00.00007031 010104 PC 001I) 0 1. 7005110 010 040103114000 00 O000 £00 0133 2I 511100 0000 4000

0. 1007001 1011)44103 0100 63 00 00670050 SOC -bl) 4 010 6 Ff005uOO 01)0001 CC P00 0000 001004161 304100 000 011703 p0,IS000 00.0$

O ProtOn 01000 SOS 0,674 0 YPOCOI toot :000. 771130 070010001 570740 007170710010 00 01 14)714000 00431 (0 107.

0000 1000

- I. 0100100 0406, 6.000110 0770 011 001) 410070 0 )ll64:o 000. 00 u471 1001 0-0 11 0 00001110110

0. 1100 P37) 30. 0 10. 5001000140100. 100 100 0000 10,0300 00 50000
7. 000 01,07-00110110100000 011411 00001000 10.640.

I $001170 001)5 40 0)7011 000 001 011001 000 001.11.

001171000000 0010 041 SS P701, 10.0 14 00000$ 01 00700 01 001100700411 000 0100 00 07.0 00010011.

1770 0007 0000400 11 01070$ 3-034 00 500000 01 710010 1 00314001010.7.7.5-01. 5 OP IS 700100547 031 0!774010, 0000OOVOO 40 070101, 0 6600111 070.7001 010000 0101701 10 0

70. 11)4147.0 $110000 001040700 00 01000 040 001 001 104 000.0011 070 000 5100 04115001 0000 70 0! 4 007 703 0035100
11. 674111 7)0040 43440 CII 04 10*1 100 0000700) 15010 0. 144704 0.1114 4 0,01 770, 0001000. 0-11353 4,54( 40) 00001 60.005 (10700 6. CII 50-T TO 7001107 00007 100 75) 1050005 00100 06)0 7,0000010>77100.00.501000031 0*10 0040,0 roes 11,01000 100 014.. 044 534:100 00110TO 54-17 00.10 :10111 007) 11001 711.5004 704 063. 004 1-10 006300405001000 500. 000) O 1 4 Id 7700500 IS,,) 0 7001,011Z001 0 40704 1113 1,: 00.70 20) 7 100 00010 0,03111 11,40 003 014,005--I 01010 0.00700 000)1 5050775550470,054 0010 041 0)17, 1.00:0 0:0000 4 0, 0 0301 0 0 0, 1,117 0010 010)0 000. 00 0.100200 00 II 300110) 154,45 0140 4, 0,010 0040 41 70 75003 0001011 106-14 rz 111713. bOll It.00 011 00 .201.

4747100041000 0440000 001510.0410 500 I

31445 00 7130), 10000. 301 04 000,010311CC 0, 0300 C01C13C34 0407 11 0400 0410. 01075 1 37 41 0000110 0530,700700 000 0000 4110) 013000013 001. ISO 0-2 303 11100101) 0406 00001,31010 01.1)00. 100 04 104 0 010 1051 401000-tO 000030000 0, 00100 07011 1004 00006

.10,50 511001106010 003 01101 00/100 01) 00011030 000071 734$. 1103 0.01$ 4001 4 0407/0040 0110 00000101ICS 0.11000 Or 0000004 445430,070.300 0.0550.103320.

00,0 (CE 30,0 40400 00004014035 500114 10 05661 FLOW 10 700 0401 000 7000L,00FC FF00 0(0 0.3.07 1107 114054404 7-. 60 104! 2 CRITICAL DOCUMENT 011-1010 SOUTHERN irSi.

COMPANY 00 40 1044011 007 011005 005104 0wE 7 4 ( 5 1 6 I 7 4 I 5 0 00 00 I 02

c%J co C

G)

E I

HLTO7 SRO NRC EXAM

2. 202001A1.02 001 Which ONE of the choices below completes the following statements concerning monitoring Unit 2 INDIVIDUAL jet pump flows and a potential problem associated with a jet pump failure during a Loss of Coolant Accident (LOCA)?

If jet pump #3 failure were to occur during plant operation, the INDIVIDUAL failed jet pump dP can be monitored on panel A catastrophic failure of jet pump #3 will be indicated by A. 2H11-P602; a sudden INCREASE in reactor power B. 2H11-P602; a sudden REDUCTION in jet pump differential pressure C. 2H11-P619; a sudden INCREASE in reactor power D& 2H11-P619; a sudden REDUCTION in jet pump differential pressure 4

HLT-07 SRO NRC EXAM

==

Description:==

Edwin, this was question 1 of 10 that you have already reviewed. Any discussed changes have been incorporated.

Indication of individual jet pump flow and total flow can be found in the Control Room on panels H11-P619, Hi i-P602, and Hi 1-P603. Panel Hi l-P619 has the indication of all twenty non-calibrated jet pump flow detectors (#3). Panel Hi 1-P602 has indication for all four calibrated pump flow detectors, total recirc loop A and total recirc loop B flow. Panel Hi 1 -P603 has a recorder B21-R613 for total core flow.

Jet Pump failure may be detected by one or more of the following:

A reduction in reactor power caused by a reduction in core flow, Recirc loop flows are not equal even though recirc pump speeds are the same, Individual jet pump differential pressure indication is lower than normal, Recirc pump flow differs from established speed flow characteristics.

The A distractor is plausible if the applicant confuses the 2Hi i-P602, jet pump calibrated flows (JP#5, 10, 15 & 20) with the P619 individual flows (JP#i 10 & JP#1 1 20). The second part is plausible if the applicant confuses jet pump dP in the non failed loop which would actually increase (due to less flow resistance) with the jet pump dP in the failed loop and thinking jet pump/core flow increasing is indicative of power increasing, resulting in a power increase.

The B distractor is plausible if the applicant confuses the 2H1 i-P602, jet pump calibrated flows (JP#5, 10, 15 & 20) with the P619 individual flows (JP#1 10 & JP#l 1 20). The second part is correct.

The C distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses jet pump dP in the non failed loop which would actually increase (due to less flow resistance) with the jet pump dP in the failed loop and thinking jet pump/core flow increasing is indicative of power increasing, resulting in a power increase.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

5

HLT-07 SRO NRC EXAM

References:

NONE K/A:

202001 Recirculation System Al. Ability to predict and/or monitor changes in parameters associated with operating the RECIRCULATION SYSTEM controls including: (CFR: 41.5 / 45.5)

A 1.02 Jet pump flow 3.4 3.4 LESSON PLAN/OBJECTIVE:

Bi 1-RXLNS-LP-04404, Reactor Vessel Instrumentation, EO 400.048.A.03 B3 1 -RRS-LP-0040 1, Reactor Recirculation System, EO 004.002.A. 10 References used to develop this question:

34SV-SUV-023-2, Jet Pump and Recirculation Flow Mismatch Operability B3 1-RRS-LP-00401, Reactor Recirc System (Indications of Jet Pump Failure)

OPS-1570, Loop A Average Jet Pump DIP vs. Core Flow 6

B11-RXINS-LP-04404 Page 2 of 92 REACTOR VESSEL INSTRUMENTATION Initial License (LT) ENABLING OBJECTIVES

1. DESCRIBE the following plant systems response to reactor water level column inaccuracies:

(200.002.A.08)

a. Reactor Protection System
b. Primary Containment Isolation System
c. Secondary Containment Isolation System
d. Emergency Core Cooling System.
2. NAME the four ranges of reactor vessel water level indication and IDENTIFY which of the five reference legs they tap off from. (200.002.A.09)
3. Given a list of reactor vessel water level indicators, CHOOSE the indicators that are most likely to experience reference leg flashing. (200.002.A.lO)
4. DESCRIBE the number, location, and range of the reactor vessel water level and pressure indicators found in the: (200.002.A.07)
a. Control Room
b. Reactor Building
c. Remote Shutdown Panel(s)
5. Given a Loss of Vital AC, IDENTIFY reactor water level instruments that are affected.

(200.020.A.02)

6. From a list, SELECT the power supply for Reactor Vessel Level and Pressure Instruments.

(200.020.A.06)

7. From a list of level setpoints, DETERMINE what automatic actions/alarms would occur.

(200.002.A.l 1)

8. From a list of pressure setpoints, DETERMINE what automatic actions/alarms would occur.

(200.002.A. 12)

9. DESCRIBE where Jet pump flow indicators and flow transmitters are found. (400.048A.03)
10. Given a list of statements, IDENTIFY where and DESCRIBE how core flow is measured.

(400.048.A.05)

11. Given a list of statements, IDENTIFY the statement which describes how core dip and core plate d/p are measured. (400.048.A.07)

B31-RRS-LP-00401 Page 8 of 132 REACTOR RECIRCULATION SYSTEM OBJECTIVES Initial License (LT) ENABLING OBJECTIVES From a list of statements, SELECT the one that best describes the purpose of the Reactor Recirculation Systems. (004.001 .A. 12, 004.001 .A. 14)

2. Given a P&ID or a simplified drawing of the Reactor Recirculation System, TRACE the flow path through the system and LABEL the following components: (004.001 .A.06)
a. Suction Valve
b. Recirc Pump
c. Discharge Valve
d. Jet Pumps
3. Given plant conditions and changes to those conditions, DETERMiNE the Recirc System automatic responses and interlocks, including: (004.001.A.07)
a. Pump suction and discharge valves
b. LOCA Condition
c. Pump Start
d. #1, #2, #3 and #4 speed limiters
4. From a list of statements, CHOOSE the one that best describes the function of: (004.001 .A. 11)
a. Recirc pump
b. RPT Breakers
c. Jet pumps
d. Adjustable Speed Drive
e. Adjustable Speed Drive Cooling Water System
5. Given a list of statements, CHOOSE the one that best describes how NPSH of Recirc Pumps is maintained at: (004.001 .A. 10)
a. Low power
b. High power
6. Given a list of statements, IDENTIFY the two problems associated with Jet Pump failures.

(004.002.A. 10)

7. Given a load list, DETERMINE the power supplies to the Reactor Recirculation System components: (004.00 1 .A.04)
a. Suction and Discharge Valves
b. Adjustable Speed Drive (ASD) Drive Motor
c. ASD Cooling Water Pumps

SNC PLANT E. I. HATCH I Pq 10 of 19 DOCUMENT TITLE: I IDOCUMENT NUMBER: VERSION No:

JET PUMP AND RECIRCULATION FLOW MISMATCH OPERABILITY I 34SV-SUV-023-2 7.13 I

ATTACHMENT I AU. Pg.

TITLE: JET PUMP I LOOP FLOW MISMATCH DATA 1 of 2 Surveillance Type (i):

  • Jet Pump & Loop Flow Mismatch Operability

. Jet Pump Operability

. Loop Flow Mismatch Operability Recirc Loop(s) operating (v):

IF in Single Loop operation (SLO), confirm Core Plate D/P bandwidth is < 1.5 psid, reduce Recirc pump speed UNTIL D/P bandwidth is < 1.5 psid JQ notify Engineering.

N/A, H in Dual Loop operation Record the following data (2H1 1-P603, 2H1 1-P602, or P/C):

KEY# PARAMETER VALUE INSTRUMENT(S)

(1) Core Thermal Power  % Power/Flow Log, NI Power/Flow Log, (2) Core Flow MlbIhr 2B21-R613 Black Pen X .77, sum of (7) and (8) below (3) Core Plate D/P Power/Flow Log, psid 2B21-R613 Red pen x 0.3 psid/%

(4) Recirc Suction Temperature °F 2B31-R650 (5) Recirc Pump A Driving Flow kgpm 2B31-R617 (6) Recirc Pump B Driving Flow kgpm 2B31-R613 (7) Loop A Jet Pump Flow Mlb!hr 2B21-R61 1A (8) Loop BJet Pump Flow Mlb/hr 2B21-R61IB (9) Recirc Pump A Speed 0/

/0 2B31-R661A-1 (10) Recirc Pump B Speed  % 2B31-R661B-1 Key numbers 16 thru 35 are NOT required when performing a LOOP FLOW MISMATCH NOTE:

OPERABILITY test, AND therefore; can be marked as N/A.

Jet Pump Differential_Pressures_(2H11-P619):

Jet Pump# #1 #2 #3 #4 #5 #6 #7 #8 #9 #10 Value (%PSID)

Key# (16) (17) (18) (19) (20) (21) (22) (23) (24) (25) 2B21-R608() B D F H K M P S U W Jet Pump# #11 #12 #13 #14 #15 #16 #17 #18 #19 #20 Value (%PSID)

Key# (26) (27) (28) (29) (30) (31) (32) (33) (34) (35) 2B21-R608() A C E G J L N R T V OPS-0623 Ver. N/A G16.30 MGR-0009 Ver. 4

SNCPLANTE.I.HATCH I Pgllofl9 DOCUMENT TITLE: I IDOCUMENT NUMBER: VERSION No:

JET PUMP AND RECIRCULATION FLOW MISMATCH OPERABILITY 34SV-SUV-023-2 7.13 ATTACHMENT I Att. Pg.

TITLE: JET PUMP I LOOP FLOW MISMATCH DATA 2 of 2

  • IF there is indication of a failure of one or more Jet Pumps, enter and execute 34A8-B21-004-0, Jet Pump Failure.
  • IF the M-ratio acceptance check fails above 25% RTP, initiate a Condition Report for Reactor Engineering review.
  • Attach copy of input and output data from P/C PRIOR to completed attachment. Mark this step N/A, IF performing Manual calculation.
  • Shift Supervisor notified of results: (all values WITHIN acceptance criteria)

( ) acceptable ( ) non-acceptable Non-acceptable Conditions:

Comments/Corrective Actions:

Completed and/or verified by:

I / I Print Name Initials Date Time

/ I /

Print Name Initials Date Time Results reviewed by: /

Shift Supervisor Date OPS-0623 Ver. N/A G16.30 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg lot 1 FORM TITLE:

LOOP A AVERAGE JET PUMP DIP vs. CORE FLOW UNIT 2 DATA FROM CYCLE 20 AND 21 THRU 27MARIl (TWO LOOP OPERATION) 3 20 30 50 60 70 80 90 CORE FLOW (MLB/HR)

OPS-1 570 Ver. 7.0 G1 6.030 34SV-S UV-023-2

B31-RRS-LP-60401 Page 39 of 132 REACTOR RECIRCULATION SYSTEM

4. The jet pump diffuser is a gradual conical section changing to a straight cylindrical section at the lower end. The diffuser is welded to the shroud support. The joint between the throat and the diffuser is a slip fit. A metal-to-metal spherical-to-conical (metal to metal) seal joint is used between the nozzle entry section and riser. Firm contact maintained by a clamp arrangement, which fits under ears on the riser and utilizes a bolt to provide a downward force on a pad on top of the nozzle entry section. The throat section is supported laterally by a bracket attached to the riser.
5. The consequences of a failed jet pump would be increasing the size of the blowdown area and hindering the capability to reflood the core during and after a LOCA. Figure 20A & 20B (LT 6)(LCT 6)(EN 8)

IrrrlrI 4.

6. Jet Pump failure may be detected by one or more of the following:

(LT *17)(LCT 16)

  • A reduction in reactor power caused by a reduction in core flow,
  • Recirc loop flows are not equal even though recirc pump speeds are the same,
  • Individual jet pump differential pressure indication is lower than normal,
  • Recirc pump flow differs from established speed flow characteristics.

HLT-07 SRO NRC EXAM

3. 202002K3.03 001 Unit 2 is operating at 100% RTP when a malfunction causes a Recirc runback signal to Speed Limiter #1 on BOTH Recirc Pumps.

Which ONE of the choices below completes the following statements?

While the Recirc Pumps are reducing speed, INDICATED reactor water level on 2C32-R606A, B and C, Narrow Range instruments will The FiNAL Recirc Pump speed will be A. decrease; 22%

B. decrease; 33%

C increase; 22%

D. increase; 33%

7

HLT-07 SRO NRC EXAM

==

Description:==

When the Recirc pump speed decreases, indicated reactor water level will increase due to formation of more voids in the core. This additional voided area will increase the backpres sure in the downcomer region, causing annulus level to rise. Since RWL is measured in the annulus region, indicated RWL increases. Also as the recirc pump reduces speed, the pumps will entrain less water from the downcomer region causing indicated RWL to increase. RWL decreasing is plausible because as voids increase in the core, water level in the core will actually decrease.

Speed Limiter #1 electronically limits the speed of the pumps to 22% to minimize the chance of pump cavitation.

The A distractor is plausible if the applicant confuses indicated RWL with water level in the core which will actually decrease. The second part is correct.

The B distractor is plausible if the applicant confuses indicated RWL with water level in the core which will actually decrease. The second part is plausible if the applicant confuses the 4 signals for #2 Speed Limiter with the 3 signals for #1 Speed Limiter.

The D distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses the 4 signals for #2 Speed Limiter with the 3 signals for #1 Speed Limiter.

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

8

HLT-07 SRO NRC EXAM

References:

NONE K/A:

202002 Recirculation Flow Control System K3. Knowledge of the effect that a loss or malfunction of the RECIRCULATION FLOW CONTROL SYSTEM will have on the following: (CFR: 41.7/45.4)

K3.03 Reactor water level 3.3 3.4 LESSON PLAN/OBJECTIVE:

B31-RRS-LP-00401, Reactor Recirculation System, EO 004.001.A.07 References used to develop this question:

34S0-B3 1-001-2, Reactor Recirculation System 9

B31-RRS-LP-00401 Page 8 of 132 REACTOR RECIRCULATION SYSTEM OBJECTIVES Initial License (LT) ENABLING OBJECTIVES From a list of statements, SELECT the one that best describes the purpose of the Reactor Recirculation Systems. (004.001 .A. 12, 004.001 .A. 14)

2. Given a P&ID or a simplified drawing of the Reactor Recirculation System, TRACE the flow path through the system and LABEL the following components: (004.001.A.06)
a. Suction Valve
b. Recirc Pump
c. Discharge Valve
d. Jet Pumps
3. Given plant conditions and changes to those conditions, DETERMINE the Recirc System automatic responses and interlocks, including: (004.001 .A.07)
a. Pump suction and discharge valves
b. LOCA Condition
c. Pump Start
d. #1, #2, #3 and #4 speed limiters
4. From a list of statements, CHOOSE the one that best describes the function of: (004.00 l.A.l1)
a. Recirc pump
b. RPT Breakers
c. Jet pumps
d. Adjustable Speed Drive
e. Adjustable Speed Drive Cooling Water System
5. Given a list of statements, CHOOSE the one that best describes how NPSH of Recirc Pumps is maintained at: (004.00l.A.l0)
a. Low power
b. High power
6. Given a list of statements, IDENTIFY the two problems associated with Jet Pump failures.

(004.002.A. 10)

7. Given a load list, DETERMINE the power supplies to the Reactor Recirculation System components: (004.00 1 .A.04)
a. Suction and Discharge Valves
b. Adjustable Speed Drive (ASD) Drive Motor
c. ASD Cooling Water Pumps

SOUTHERN NUCLEAR PLANT E. I. HATCH PAGE 16 OF 222 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REACTOR RECIRCULATION SYSTEM 34S0-B31-001-2 42.0 5.2.18

  1. I SL to 22% OR 370 RPM Recirc discharge valve <90% open OR Total F!W flow < 20% (15 sec delay)

OR RWL (Master FW controller output) <20 inches (Normally Median Level)

AND Total steam flow decreases by 60% of previous 6 minute value

  1. 2SL to 33% OR 554 RPM RWL < 32 inches OR Steam Flow> 65%

AND Either RFP < 20% rated flow AND Either REP has a trip signal from TMR

  1. 3 SL to 61% Q1O25 RPM CBP median suction pressure low [40 psig, 10 sec delay]

OR RFP median suction pressure low [225 psig, 5 sec delay]

  1. 4 SL Variable 100% to 33% OR 1680 RPM to 554 RPM 6.7 % speed decrease per ONE inch decrease in RWL (from 30 to 20 inches).

Variable runback from 100% to 33% speed caused by median RWL corresponding to a RWL decrease from 30 inches to 20 inches Loss of Vital AC All SLs are inhibited/disabled Loss of 2R25S101 A Recirc #1 Speed Limiter is inhibited/disabled Loss of 2R25S102 B Recirc #1 Speed Limiter is inhibited/disabled MGR-0001 Ver. 4

HLT-07 SRO NRC EXAM

4. 203000G2.4.34 001 A LOSP has occurred on BOTH units. The following conditions currently exist on Unit 1:

o Reactor water level -111 slowly decreasing o Reactor pressure 90 psig slowly decreasing o MAJOR fire Unit 2 130! Control Building near 600V Bus 2C The lA Diesel Generator (DIG) is the ONLY DG supplying emergency power.

NO other D/Gs (Unit 1 or Unit 2) can be started from the Main Control Room.

Which ONE of the choices below completes the following statement concerning Unit 1 RHR Loop A LPCI Injection?

To supply power to lEl l-FO15A, Inboard Injection valve, the IE1 1-FO15A can be ELECTRICALLY aligned for LPCI injection if A. lR43-SOO1C, Diesel Generator lC, is locally started B. 2R43-SOO1C, Diesel Generator 2C, is locally started C Alternate Power is aligned from 1R24-S0l 1, Reactor Bldg MCC-IC D. Alternate Power is aligned from lR24-S012, Reactor Bldg MCC-1B

Description:

Edwin, this was question 2 of 10 that you have already reviewed. Any discussed changes have been incorporated.

Reactor Building 600 VAC MCC 2E-A, 1R24-SO18A supplies power to 1E1 l-FO15A.

lR24-SO18A is powered from Unit 2 600V Bus 2C.

Reactor Building 600 VAC MCC 2E-B, 1R24-SO18B supplies power to lEl l-FO15B.

lR24-SO18B is powered from Unit 2 600V Bus 2D.

Equipment has been installed (similar to the RSDP) that will remove control and motive power to valves El l-FO15AIB from the control room and reestablish motive and control power through local components located on the Reactor Building 130 elevation. 34AB-X43-00l-1/2 has been revised to operate local equipment to transfer control and motive power from R24-SOl 8A/B to R24-S0l 1/SO 12 respectively (ONLY if power has been lost from the LPCI Valve MCCs was fire related). Two new transfer switches for each valve (one disconnect for motive power and one keylock for control power) have been installed on the 130! elevation of the Reactor Building to manually transfer power from the primary source to the alternate source.

10

HLT-07 SRO NRC EXAM Alternate Feed for lEl 1-FO15A is from Ui Rx Bldg MCC 1R24-SOl 1, which is powered from Ui 600V Bus 1C via 4160V Bus 1E via the 1A DG. Alternate Feed for 1E1 l-FO15B is from Ui Rx Bldg MCC 1R24-S012, which is powered from Ui 600V Bus 1D via 4160V Bus 1G via the 1CDG.

The A distractor is plausible if the applicant confuses the normal power supply (600V 2C) and thinks that starting the 1C DG will provide power to the appropriate 1R24-SO18AJB. Confusing the power supplies to 1R24-SO18A/B makes this distractor plausible since the alternate supply can be aligned from (1R24-S012 for IE1 l-FO15B verses 1R24-SOl 1 for 1E1 1-FO15A).

1R24-S012 supplies alternate power to lEl i-FO15B.

The B distractor is plausible if the applicant confuses the normal power supply (600V 2C) with 600V 2D since U2 buses supply power to Ul components. Confusing the power supplies to lR24-SO18A[B makes this distractor plausible.

The D distractor is plausible if the applicant remembers that the LPCI Injection valve, 1E1 l-FO15A has an alternate power supply (1R24-SOl 1) that can be aligned locally but confuses the alternate supply coming from (lR24-S012 for lEl l-FO15B verses 1R24-SO1 1 for lEl 1-FO15A). 1R24-S0l2 supplies alternate power to 1E1 1-FO15B.

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

11

HLT-07 SRO NRC EXAM

References:

NONE K/A:

203000 RHRJLPCI: Injection Mode (Plant Specific) 2.4.34 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects. (CFR: 41.10/43.5 / 45.13) 4.2 4.1 Changed K/A to the above K/A on 2/7/2012 per Chief Examiner Edwin Lea 2.4.4 1 Knowledge of the emergency action level thresholds and classifications.

(CFR:41.10/43.5/45.11) 2.9 4.6 LESSON PLAN/OBJECTIVE:

El 1-RHR-LP-00701, Residual Heat Removal System, EO 006.007.A.0l References used to develop this question:

34S0-Ell-0l0-2, RHR System 34AB-X43-00l-l, Fire Procedure 34AB-X43-00l-2, Fire Procedure 12

E1l-RIIR-LP-00701 Page 4 of 130 RESIDUAL HEAT REMOVAL SYSTEM Initial License (LT) ENABLING OBJECTIVES

1. Given a simplified drawing of the RHR system, TRACE the flowpath for RHR in LPCI mode.

(006.00 1 .a.01)

2. Given a simplified drawing of the RHR system, TRACE the flowpath of Drywell Sprays.

(007.00 1 .a.Ol)

3. Given a simplified diagram of the RHR system, TRACE the flowpath for Torus Spray.

(007.003.a.01)

4. Given a simplified drawing of the RHR system, TRACE the flowpath for Torus Cooling.

(007.005.a.01)

5. Given a simplified drawing of the RHR system, TRACE the flowpath for RHR operating in the Shutdown Cooling Mode. (007.007.a.04, 007.024.b.0 1)
6. Given plant conditions, EVALUATE those conditions to DETERMINE if RWL is adequate for RHR Shutdown Cooling flow. (007.007.a.02, 007.024.b.02)
7. Given a simplified drawing of the RHR system, TRACE the flowpath for RHR in fuel pool cooling assist mode. (007.016.a.01)
8. Give a simplified drawing of the RHR system, TRACE the flowpath for draining the RPV to the Torus. (006.01 1.a.01)
9. Given a simplified drawing of the RHR system, TRACE the flowpath for draining the Torus to Radwaste with RHR Loop B. (013.004.a.01)
10. From a list of statements, SELECT the statement that describes the consequences on the RHR system if the R}IR Pump Minimum Flow bypass Valve fails closed. (006.005.a.04)
11. From a list of statements, SELECT the statement that describes the maximum flow rate for RHR in Shutdown Cooling and the reason for the flow limit. (007.007.a.03, 007.024.b.03)
12. From a list of power supplies, SELECT the power supply for the RHR Pumps. (006.00 l.a.02)
13. Given a Plant Hatch load list, DETERMINE power supplies to RHR valves and components.

(006.007.a.0 1)

14. Given a simplified drawing of RHR, IDENTIFY the following system interfaces: (006.00 1.a.04)
a. Reactor Recirc System
b. Fuel Pool Cooling System
c. Core Spray Jockey Pump System
d. Radwaste
e. Core Spray System

SNC PLANT E. I. HATCH I Pci 177 of 306 DOCUMENT TITLE: I DOCUMENT NUMBER: Version No:

RESIDUAL HEAT REMOVAL SYSTEM 34SO-E11-010-2 38.0 ATTACHMENT 2 Att. Pg.

TITLE: RHR SYSTEM ELECTRICAL LINEUP 2 of 11 NUMBER DESCRIPTION POSITION CHECKED VERIFIED 2R24-SO11 6001208V MCC 2C ESS Div. I 13ORHR14 RCKD IN Frame 5B RHR Outboard Injection Valve 2E1 1-FO17A CLOSED RCKD IN Frame 5A RHR Torus Suction Valve 2E1 1-FOO4A CLOSED RCKD IN Frame 4B RHR Shutdown Cooling Valve 2E1 l-FOO6A CLOSED RCKD IN Frame 6A RHR Torus Suction Valve 2E1 1FOO4C CLOSED RCKD IN Frame 68 RHR Shutdown Cooling Valve 2E1 1-FOO6C CLOSED RCKD IN Frame 6C RHR Shutdown Cooling Valve 2E1 1-F009 CLOSED RCKD IN Frame 7A RHR Crosstie Valve 2E1 1-FOl 0 LOCKED RCKD IN Frame 7B RHR Heat Exchanger To Torus 2E1 1-FOl 1A CLOSED RCKD IN Frame 8A RHR Heat Exchanger Outlet Valve 2E1 l-FOO3A CLOSED RCKD IN Frame 8D RHR Containment Spray Inboard Valve 2E11-F021A CLOSED RCKD IN Frame 9A Reactor Head Spray Valve 2E11-F022 OPENED

  • RCKID IN Frame 9B RHR Full Flow Test Line Valve 2E1 l-F024A CLOSED
  • RCKID IN Frame 9C RHR Heat Exchanger to RCIC Valve 2E11-F026A OPEN RCKD IN Frame 1OA RHR Torus Spray Valve 2E11-F027A CLOSED RHR Torus Spray or Test Valve 2E11-F028A RCKD IN Frame lOB CLOSED RCKD IN Frame 1 OC RHR to Radwaste Valve 2E1 l-F040 CLOSED Valve CLOSED and breaker off. Posted 2H1 1-P601/P602 and 2R24-S01 I OPS-0276 Ver. 13 G16.030 MGR-0009 Ver. 5

SNC PLANT E. I. HATCH I Pg 178 of 306 DOCUMENT TITLE: I DOCUMENT NUMBER: Version No:

RESIDUAL HEAT REMOVAL SYSTEM 34SO-E11-010-2 38.0 ATTACHMENT 2 Att. Pg.

TITLE: RHR SYSTEM ELECTRICAL LINEUP 3 of 11 NUMBER DESCRIPTION NORMAL CHECKED VERIFIED POSITION Frame I IA RCKD IN RHR Heat Exchanger Inlet Valve 2E1 1F047A CLOSED Frame 1 1 B RCKD IN RHR Heat Exchanger Bypass Valve 2E1 1-F048A CLOSED Frame 11C RCKD IN RHRSW Control Valve MOV2E11-F068A CLOSED Frame 12A RCKD IN RHRSW Crosstie Valve 2E11-F073A CLOSED Frame 12B RCKD IN RHRSW to RHR Crosstie Valve 2E1 1-F075A CLOSED Alternate Power for 2E11-FO15A, RHR Inboard RCKD IN Frame 14B Injection Valve CLOSED Frame 15C RCKD IN RHR Heat Exchanger Vent Valve 2E11-F1O4A CLOSED Frame 16C RCKD IN RHR Containment Spray Outboard Valve 2E11-FO16A CLOSED Frame 17A RHR Heat Exchanger Vent Valve 2E11-F1O3A RCKD IN CLOSED RCKD IN Frame 17C Service Water Crosstie Valve 2E11-F119A CLOSED Frame 21A RCKD IN Panel 2R25S10i Fused Disconnect Switch CLOSED 2R24-SO18A600VMCC2 EESS Div. 1, 13ORHR17 RCKD IN Frame 3A RHR Minimum Flow Valve 2E11-FOO7A CLOSED RCKD IN Frame 3C RHR Inboard Injection Valve 2E11-FO15A CLOSED 2R26-M 119 Transfer Switch, 1 3ORH Ri 7 2R26-M1 19 Alternate Power Source for 2Ei 1 -FOi 5A FIRE LOCKED PRIMARY KEY LOCKED 2E1 I -S8E Control Switch for 2E1 I -FOl 5A PRIMARY OPS-0276Ver. 13 G16.030 MGR-0009 Ver. 5

SNC PLANT E. I. HATCH I Pg 49 of 104 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

FIRE PROCEDURE 34AB-X43-001-1 11.1 ATTACHMENT 1 Attachment Page TITLE: SAFE SHUTDOWN ACTIONS 7 OF 30 4.9.8 In the event of a loss of power to I R24-S01 8A or I R24-S01 8B, perform the following:

4.9.8.1 IF RHR Loop A is required AND power cannot be restored to MCC 1 R24-SOI8A, perform the following:

4.9.8.1.1 UNLOCK PLACE transfer switch, I R26-M1 19, in the ALTERNATE position, located at 13ORLRO8.

4.9.8.1.2 PLACE key lock control switch, I El l-S8E, in the ALTERNATE position, panel 1H21-P120, located at 13ORLRO8.

4.9.8.1.3 PLACE valve switch 1EI1-S8C, to OPEN position, panel 1H21-P120.

to open valve, IEI1-FO15A.

4.9.8.1.4 UNLOCK AND CLOSE 1E11-FO18A, RHR Pump A Minimum Flow isolation Valve, as required.

4.9.8.2 IF RHR Loop B is required power cannot be restored to MCC 1R24-SO18B, perform the following:

4.9.8.2.1 UNLOCK AND PLACE transfer switch I R26-M 121, in the ALTERNATE position, located at I3ORLRO5.

4.9.8.2.2 Place key lock control switch, 1EI 1-S8F, in the ALTERNATE position, panel 1H21-P122, located at 13ORLRO5 49.8.2.3 PLACE valve switch, 1E11-S8D to OPEN position, panel 1H21-P122, to open valve 1EI1-FOI5B.

4.9.8.2.4 UNLOCK AND CLOSE 1E1 I-FOI8B, RHR Pump B Minimum Flow Isolation Valve, as required.

MGR-0009 Ver. 5

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 10 OF 103 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

FIRE PROCEDURE 34AB-X43-001 -2 11.23 5.12 SWITCHGEAR ACCESS HALLWAY & 600V SWITCHGEAR ROOM 2C NOTE: This subsection is to be peormed ONLY a major fire is out of control in this area.

5.12.1 N a reactor SCRAM has occurred, perform safe shutdown actions listed in Attachment 1, Step 5.12. El 5.12.2 Deenergize the affected switchgear. El 5.12.3 Place standby equipment from other 600V switchgear in service as required. LI 5.12.4 IF UNIT 1 RHR Loop A is required JQ power CANNOT be restored to 1R24-SO18A MCC, perform the following:

5.12.4.1 Unlock AND place I R26-M1 19, transfer switch, in the ALTERNATE position, located at I3ORLRO8. LI 5.12.4.2 Place IEII-S8E, key lock control switch, IN. THE Alternate position, panel 1H21-P120, located at 13ORLRO8 (key 1EII-53). El 5.12.4.3 Place IEI1-S8C, valve switch, TO OPEN position, to open 1E11-FO15A, panel 1H21-P120. El 5.12.4.4 Unlock AND close 1 El 1-FOI8A, RHR Pump A MINIMUM Flow Isolation Valve, as required. El MGR-0001 Ver. 4

HLT-07 SRO NRC EXAM

5. 204000K4.01 001 Unit 2 is at operating at RTP when a malfunction occurs causing the following conditions:

o RWCU HX room temperature and pump room temperatures peaked at 125°F o RWCU differential flow reached 60 gpm for 30 seconds then went to 0 gpm (indicated) o Non-Regenerative HX (NRHX) outlet temperature peaked at 125°F o 2B RWCU pump cooling water temperature peaked at 145°F o The 2B RWCU pump tripped Based on these conditions, which ONE of the choices below is the reason the2B RWCU pump tripped?

The 2W RWCU pump tripped because of high A. differential flow B. pump room temperature C pump cooling water temperature D. Non-Regenerative HX (NRHX) outlet temperature

Description:

JAW 34S0-G31-003-2, Reactor Water Cleanup System, 2G3 1-FOOl AND 2G31-F004, RWCU Inboard Isolation AND RWCU Outboard Isolation, close on the following signals:

Low Reactor water level, -35 inches High differential flow, 56 gpm for 42.5 seconds High RWCU area ventilation differential temperature RWCU Pump Room 60°F RWCU Hx Room 60°F RWCU Phase Separator Room 60°F High RWCU area ambient temperature RWCU Pump Room 140°F RWCU Hx Room 140°F RWCU Phase Separator Room 140°F 2G3l-F004, RWCU Outboard Isolation, closes on the following signals:

Actuation of SBLC Non-Regenerative Heat Exchanger outlet temperature 140°F 2G3l-COO1A AND/OR 2G31-COO1B, RWCU Pumps, trip under the following conditions:

High cooling water temperature, 140°F Low flow for the A RWCU pump, <60 gpm, is bypassed for 15 sec. following pump start.

Low flow for the B RWCU pump, <60 gpm, is bypassed for 30 sec. following taking the 13

HLT-07 SRO NRC EXAM Pump Control Switch to START.

2G3 1-FOOl, RWCU Inboard Isolation, OR 2G3 1 -F004, RWCU Outboard Isolation, NOT full open.

The A distractor is plausible since this will lead to a pump eventually tripping and if the applicant realizes diffferential flow is high and confuses the time delay value thinking this caused the pump to trip.

The B distractor is plausible since this will lead to a pump eventually tripping and if the applicant realizes pump room temperature is high and confuses the value thinking this caused the pump to trip.

The D distractor is plausible since this will lead to a pump eventually tripping and if the applicant realizes NRHX outlet temperature is high and confuses the value thinking this caused the pump to trip.

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

14

HLT-07 SRO NRC EXAM

References:

NONE K/A:

204000 Reactor Water Cleanup System K4. Knowledge of REACTOR WATER CLEANUP SYSTEM design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)

K4.01 Pump protection 2.5 2.5 LESSON PLAN/OBJECTIVE:

G3 1 RWCU-LP-003O 1, Reactor Water Cleanup, EO 003 .002.A. 10 References used to develop this question:

34S0-G31-003-2, Reactor Water Cleanup System 34AR-602-415-2, RWCU Pump High Temp Trip 15

G31-RWCU-LP-00301 Page 4 of 80 REACTOR WATER CLEANUP

5. Given a list of statements, SELECT the statement representing the purpose of the following RWCU System components. (003.002.A.04)
a. RWCU Inboard/Outboard Isolation Valves F00l/F004
b. Regenerative Heat Exchanger
c. Non-Regenerative Heat Exchanger
d. RWCU Demineralizer Bypass Valve
e. Filter Demineralizers
f. Holding Pump
g. Filter Dernineralizer Effluent Filter
h. Return Isolation Valve (P042)
i. RWCU Blowdown Flow Control Valve (F033)
j. RWCU Dump Isolation Valves (F034, F035)
6. Given initial system and plant conditions and any changes to those conditions, PREDICT RWCU System responses including: (003.002.A.l0)
a. Isolations of2G3 1-FOOl, 2G31-F004, and 2G31-F033
b. Interlocks associated with low flow/loss of power demineralizer isolation
c. RWCU Pump Trips
7. Given a P&ID or a simplified drawing of the RWCU System, LOCATE the following system interfaces; (003.002.A.03)
a. Reactor Recirculation System
b. Reactor Vessel
c. High Pressure Coolant Injection System
d. Reactor Core Isolation Cooling System
e. Radwaste System (RWCU Drains)
f. Condensate System
g. Control Rod Drive Hydraulics System
h. Reactor Building Closed Cooling Water System
8. DETERMINE the power supplies to RWCU Pumps and motor operated valves. (003.002.A.02)
9. Given a list of plant conditions, DETERMINE the plant condition which will directly result in the auto trip of the RWCU System. (003.022.A.0l)
10. Given a list of plant parameters, SELECT those that would directly result in a RWCU Isolation.

(003.013.A.03)

SOUTHERN NUCLEAR PLANT E. I. HATCH PAGE 6 OF 165 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REACTOR WATER CLEANUP SYSTEM 34S0-G31-003-2 38.3 5.1.12 WHEN performing precoat, use ONLY fresh new bags of pre-mixed resin.

5.1.13 The following valves operate backwards, i.e. the handwheel is turned clockwise to open the valves.

2G31 -F065A, 2G31 -F065B, 2G31 -FO9OA, 2G31 -F0908, 2G31-FO9IA, 2G31-F091 B, 2G31-F126A, 2G31-F1 26B, and 2G31-F1 27B.

5.1.14 Actions such as placing Vessel Valves controller to Manual, will Trip the associated HOLD Pump, which will require resetting by tapping the OFF button, then AUTO/MAN.

5.2 LIMITATIONS 5.2.1 2G31 -FOOl AND 2G31-F004, RWCU Inboard Isolation AND RWCU Outboard Isolation, close on the following signals:

5.2.1.1 Low Reactor water level, -35 inches.

5.2.1.2 High differential flow, 56 gpm for 42.5 sec.

5.2.1.3 High RWCU area ventilation differential temperature.

  • RWCU Pump Room 6OF (Annunciated at 5OCF)
  • RWCU Hx Room 6OCF (Annunciated at 5OCF)
  • RWCU Phase Separator Room 60F (Annunciated at 500 F) 5.2.1.4 High RWCU area ambient temperature.
  • RWCU Pump Room I4OCF (Annunciated at I3OCF)
  • RWCU Hx Room 14OCF (Annunciated at 13OCF)
  • RWCU Phase Separator Room 140F (Annunciated at 13 0F) 5.2.2 2G31-F004, RWCU Outboard Isolation, closes on the following signals:
  • Actuation of SBLC
  • Non-Regenerative Heat Exchanger outlet temperature 140F.

5.2.3 2G31 -F033, RWCU Blowdown Flow Control Valve, closes under the following conditions:

  • Low pressure upstream of 2G31-F033, RWCU Blowdown Flow Control Valve, 5 psig decreasing.
  • High pressure downstream of 2G31-D001, 140 psig increasing.

MGR-0001 Rev4

SOUTHERN NUCLEAR PLANT E. I. HATCH P AGE 7 OF 165 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REACTOR WATER CLEANUP SYSTEM 34SO-G31-003-2 38.3 5.2.4 2G31 -COOl A AND/OR 2G31 -COOl B, RWCU Pumps, trip under the following conditions:

  • High cooling water temperature, 140cF.
  • Low flow for the A RWCU pump, <60 gpm, is bypassed for 15 sec. following pump start. Low flow for the B RWCU pump, <60 gpm, is bypassed for 30 sec. following taking the Pump Control Switch to START.
  • 2G31-F001, RWCU Inboard Isolation, OR2G3I-F004, RWCU Outboard Isolation, r1QIfull open.

5.2.5 For 2G31-COO1B, RWCU Sealless Pump, a ten minute cooling period must be allowed between starts. Failure to comply with this limitation will result in an elevated motor winding temperature which could damage the winding insulation. Three starts per hour are allowed.

5.2.6 2G31-COOIB, RWCU Sealless Pump, motor may be damagediF the motor temperature exceeds 140cF. IF the motor cavity exceeds 18OCF as indicated bylocaltemperature indicatorRWCU Pump RBCCW OutletTemp, 2G31-NOO2B, doNOTstartORoperatethe pump. Notify Engineering for corrective action per vendor manual S-60342.

5.2.7 Pump casing cooldown rate is limited to 140cF/hr. (2.33cF/min). Using 2G31-F148B, throttle purge flow to greater than .8 gpm (- 48 GPH) but less than OR equal to 1.5 gpm

( 90 GPH) as necessary to control cooldown, until pump casing temperature is <325F.

After pump casing to CRD seal purge A temp is 200°AF, increase purge flow to maximum, (- 3 gpm). This limit is imposed for normal pump shutdown AND is NOT to be applied following a pump trip OR isolation. It may be necessary to reduce purge flow to less than .8 gpm in order to prevent exceeding pump casing cooldown rate.

5.2.8 2G31-F078, Slow Backwash Water Supply Valve, open stops are pre-set at a position to limit flow rate to 40-50 gpm. Although the valve is throttled, it is normal for it to indicate FULL OPEN.

5.2.9 The 2A RWCU Demineralizer has manual isolations whereas the 28 RWCU does NOT.

5.2.10 The 2G31-COO1A RWCU Pump has been analyzed by the vendor to operate without RBCCW cooling water under the following conditions:

  • The moderator being pumped will be < 180 F, AND
  • The pump bearing casing temperature will be < 170 F.

These temperatures will be monitored at a frequency established by the U2 SS.

5.2.11 RWCU pump casing temperature readings obtained during heatup/cooldown must be confirmed to be accurate by use of a backup portable heat gun or other measurement device.

MGR-0001 Rev4

1.0 IDENTIFICATION

ALARM PANEL 602-4 RWCU PUMP

  • HIGH TEMP TRIP DEVICE: SETPOINT:

1G31-NOO2A(1G31-NOO2B) I4OCF

2.0 CONDITION

3.0 CLASSIFICATION

Temperature on Cleanup Recirc Pump 1G31-COO1A OR EQUIPMENT STATUS temperature on RBCCW outlet from Cleanup Recirc Pump 4.0 LOCATION:

1G31-0001B heat exchanger exceeds the setpoint. 1H11-P602 Panel 602-4 5.0 OPERATOR ACTIONS:

5.1 On panel 1H21-P002, monitor 1G31-NOO2A(1G31-NOO2B), Temperature Indicator, to determine the affected RWCU Pump. El 5.2 On panel IH1I-P602, confirm STOPPED OR STOP both 1G31-COOIA and 1G31-COO1B, Cleanup Recirc Pumps. LI 5.3 IF both Cleanup Recirc Pumps are affected, CLOSE 1G31-F001 AND 1G31-F004, Inboard and Outboard RxWaterCleanupVlvs. El 5.4 On panel 1G31-P001, remove the RWCU FIDs from service as follows:

5.4.1 PLACE 1G31-R023A(1G31-R023B), Vessel Flow Control, in SEAL and NULL the controller. El 5.4.2 PLACE Vessel Flow Control, 1G31-R023A(B) in MANUAL and adjust the controller to MINIMUM. El 5.5 Confirm RBCCW is in service and 1G31-COO1A(1G31-COO1B), Cleanup Recirc Pumps, cooling water and seal purge valves are aligned per Attachment 2 of 34SO-G31-003-1, Reactor Water Cleanup System. El 5.6 IF the above action does NOT correct the condition, vent the affected RWCU Pump Seal Heat Exchanger. El 5.7 IF the temperature is still high, determine and correct the cause of the high temperature. El 5.8 Return the system to service per 34SO-G31-003-1, Reactor Water Cleanup System. El

6.0 CAUSES

6.1 Blockage of RBCCW to the RWCU Pumps 6.2 Loss of seal purge 6.3 Fouling_of the_RWCU_Pump_Seal_Heat_Exchanger

7.0 REFERENCES

80 TECH. SPECS.ITRM/ODCMIFHA:

7.1 H-17177, RWCU System Elem, Sht2 7.2 H-16188, RWCU P&ID, Sheet 1 8.1 TRM T3.4.1 RCS Chemistry 7.3 H-16009, RBCCW System P&ID 7.4 57CP-CAL-065-1,_Fenwall_56100-1/551_Temp_Sw_Cal 34AR-602-41 5-1 Ver. 0.1 MGR-0048 Ver. 5.0 AG-MGR-75-1 101

HLT-07 SRO NRC EXAM

6. 205000K5.02 001 Unit 2 is shutdown with the following conditions:

o A loop of RHR is in Shutdown Cooling (SDC) o 2E1 1-F008, SDC Suction Vlv, is OPEN o 2E1 1-F009, SDC Suction Vlv, is OPEN o 2E1 l-FO15A, RHR Inbd Inj Vlv, is OPEN o Reactor pressure is 125 psig and steady o Subsequently, 2E1 1-F048A, Hx Bypass valve, fails full OPEN and cannot be re-closed o Reactor pressure starts increasing at 1 psig/minute Without any operator actions and based on the above conditions, which ONE of the choices below completes the following statements?

When Reactor pressure reaches 140 psig, RHR Valves, will be CLOSED.

The RHR pumps receive a trip signal as a result of these valve closures.

A 2E1 1-F008 and 2E1 1-F009 (ONLY);

will B. 2E1 1-F008 and 2E1 1-F009 (ONLY);

will NOT C. 2E11-F008, 2E11-F009 and 2Ell-FO15A; will D. 2E1 1-F008, 2Ei 1-F009 and 2E1 1-FO15A; will NOT 16

HLT-07 SRO NRC EXAM

==

Description:==

SDC Inboard/Outboard isolation valves, El l-F009 & F008, isolate if either of the following occur:

Reactor pressure exceeds 138 psig Reactor water level drops below 3 FO15AJB will auto close if a PCIS Group II signal is received (3 RWL or 1.85 psig Drywell pressure) while in shutdown cooling lineup. FO15A/B will automatically reopen if a LOCA signal is present provided that the F008 or F009 is fully closed OR if reactor pressure is greater than 138 psig. FO15AJB will not isolate on high reactor pressure.

Suction valve line up must have either the respective El l-F004 valve open OR the respective F006 valve open AND the El l-F008 and El l-F009 valves open. Without a suction source, the RHR pumps will trip.

The B distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses the F008/F009/F006 suction valve trip logic with the RHR P065 logic, which does not cause the RHR pump to trip if it closes.

The C distractor is plausible if the applicant confuses the reactor pressure high response and the low level response for the FO15AIB and thinks all three valves close. The second part is correct.

The D distractor is plausible if the applicant confuses the reactor pressure high response and the low level response for the FO15AIB and thinks all three valves close. The second part is plausible if the applicant confuses the F008/F009[F006 suction valve trip logic with the RHR F065 logic, which does not cause the RHR pump to trip if it closes.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

17

HLT-07 SRO NRC EXAM

References:

NONE K/A:

205000 Shutdown Cooling System (RHR Shutdown Cooling Mode)

K5. Knowledge of the operational implications of the following concepts as they apply to SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE):

(CFR: 41.5 I 45.3)

K5.02 Valve operation 2.8 2.9 LESSON PLAN/OBJECTIVE:

El l-RHR-LP-00701, Residual Heat Remova System, EO 006.008.A.03 References used to develop this guestion:

34S0-El 1-010-2, Residual Heat Removal System 18

E11-RIIR-LP-00701 Page 5 of 130 RESIDUAL HEAT REMOVAL SYSTEM

15. From a list of statements, IDENTIFY the two statements which best describe the reason for having ajockey pump system maintaining the RHR piping filled. (006.001.a.03)
16. From a list, IDENTIFY the two ?lant Conditions which will cause an isolation of RHR shutdown cooling inboard/outboard isolation MOVs 2E1 l-F009/F008. (006.008.a.03)
17. From a list of statements, SELECT the statement that describes the interlock between the RHR SDC suction MOVs and their respective RHR pump torus suction MOV. (006.008 .a.0 1)
18. Given a list of RHR valve positions, DETERMINE if 2E1 1-F006 can be opened from the control room. (007.007.a.0l)
19. From a list of statements, SELECT the statement that describes the interlock between the Torus Spray/Test Line Isolation Valve and the SDC Suction MOV. (007.003.a.04)
20. Given plant conditions, EVALUATE those conditions to determine whether RHR should have automatically initiated in LPCI mode. (006.005.a.01)
21. DETERMINE the starting sequence for the RHR pumps given a LOCA and: (006.005.a.03)
a. No power loss
b. LOSP
22. Given that a LOCA signal has been received, with a subsequent loss of one division of 125 VDC logic power to the RHR system, DETERMINE what actions would be required in order to restart the RHR pump if it was manually secured in the present condition (006.007.a.02)
23. Given a list of plant conditions, IDENTIFY the three conditions which will cause a trip of an RHR pump. (006.008.a.02)
24. Given a simplified drawing of the RHR system, IDENTIFY those valves which would automatically operate or receive a signal to operate, upon the RHR system receiving a LPCI initiation signal while RHR is in: (006.005.a.02)
a. Normal STDBY
b. Shutdown Cooling Mode
c. Torus Cooling
25. From a list of statements, SELECT the statement which describes the interlock between the RHR Heat Exchanger Bypass Valve and a LOCA signal. (006.005.a.05)
26. Given plant conditions, DETERMINE if RWL interlocks must be overridden to operate 2E11-F028 Torus Spray/Test Line Isolation and F027 Torus Spray Line Isolation. (007.003.a.03)
27. Given plant conditions, DETERMINE if RWL interlocks must be overridden to open 2E11-F024A/B RHR Test Discharge Isolation MOV. (007.005.a.03)
28. From a list of statements, SELECT the statement which describes the interlock between the RHR SDC Suction MOV and the Outboard Spray Isolation Valve. (007.00 1.a.04)

SNC PLANT E. I. HATCH I Pg 228 of 306 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

RESIDUAL HEAT REMOVAL SYSTEM 34SO-Ell-010-2 38.0 ATTACHMENT 9 Att. Pg.

TITLE: RHR SYSTEM GENERAL INFORMATION 1 of 5 1.0 RHR valve interlocks/logic 1.1 2E11-F009 & F008 MDV isolation signals (RPV pressure> 138 PSIG QE +3 inches low RWL) 1.1.1 Auto closes above MOVs.

1.1.2 RSDP operation BYPASSES above isolation signals AND prevents MOV closure.

1.1.3 SDC pump(s) TRIP on MDV closure AND pumps can only be started AFTER EITHER its associated 2E1 1 -F004 MOV has been opened OR 2E1 1 -F006, 2E1 1 -F008 fjQ 2E11-F009 MOVs have been opened.

1.2 2E11-F006AD MOVs 1.2.1 Can NOT be opened unless the following MOVs are CLOSED:

  • Associated 2E1 1-FOO4A-D MOV
  • Associated 2E1 1-FO28NB MDV
  • Associated 2E1 1-FOI6AIB MDV 1.2.2 MUST be CLOSED PRIOR to opening following MOVs:
  • Associated 2E1 1-FOO4A-D MDV
  • Associated 2E1 1-F028A/B MDV
  • Associated 2E1 1 -FOl 6NB MOV 1.2.3 RSDP operation allows 2E1 1-FOO6B MOV to be opened ONLY WHEN 2E1 1-FOO4B AND El 1-F024B MOVs have been closed. AFTER 2E1 l-FOO6B has been opened, 2E1 1-F024B can then be reopened.

1.3 2E11-FOI7A/2E11-FO17B MOVsNormally Open) 1.3.1 WHEN RPV pressure > 425 PSIG, ONLY can OPEN the MDV IF associated 2E11-FO15A/2E11-FO15B MDV is CLOSED.

1.3.2 LOCA signal AND RPV pressure <425 (449) PSIG:

  • MDV closure prevented UNTIL 5 minutes has elapsed
  • Auto opens MOVs IF CLOSED MGR-0009 Ver. 5

SNC PLANT E. I. HATCH I Pg 229 of 306 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

RESIDUAL HEAT REMOVAL SYSTEM 34S0-E11-010-2 38.0 ATTACHMENL9.. Att. Pg.

TITLE: RHR SYSTEM GENERAL INFORMATION 2 of 5 1.4 2E1 1-F01 5A12E1 1-FOl 5B MOVs Normally Closed) 1.4.1 LOCA signal AND RPV pressure <425 PSIG:

  • MOV closure prevented
  • FOI5AIB will auto close if a PCIS Group II signal is received (3 RWL or 1.85 psig Drywell pressure) while in shutdown cooling lineup. FOI5AJB will automatically reopen if a LOCA signal is present provided that:

- P008 or F009 is fully closed OR

- If reactor pressure is greater than 138 PSIG 1.4.2 WHEN RPV pressure > 425 PSIG, ONLY can OPEN the MOV IF associated 2E11-FO17N2E1I-FO17B MOV is CLOSED.

2.0 RHR pump interlocks!loiic 2.1 Automatic features 2.1.1 Auto start on LOCA signal (1.85 PSIG D/W pressure OR -101 inches low RWL) 2.1.1.1 IF normal OR alternate power is available to their Emergency switchgear(s), pump(s) will start immediately.

2.1.1.2 IF (LOSP) occurs, the following will occur:

2.1.1.2.1 All auto AND manually started pumps TRIP (load shed) 2.1.1.2.2 After EDG(s) energize their associated Emergency switchgear(s), only pump(s) with LOCA signal auto start in the following sequence:

  • RHR pump C auto starts immediately after associated EDG output breaker closes
  • All other RHR pumps auto start 12 seconds after their associated EDG output breaker closes MGR-0009 Ver. 5

HLT-07 SRO NRC EXAM

7. 206000K5.06 001 The Unit 2 Shift Supervisor has established operation at the Remote Shutdown Panel due to Control Room evacuation.

Currently Unit 2 reactor water level is -30 inches and reactor pressure is 1000 psig.

The HPCI system is operating with the following parameters:

o HPCI turbine speed = 3000 rpm o HPCI pump discharge pressure = 780 psig An operator makes an adjustment locally, which results in the HPCI turbine speed increasing to 5050 rpm.

After the operator makes the adjustment, which ONE of the choices below completes the following statement?

The operator is using a to monitor HPCI turbine speed and the HPCI turbine have received a trip signal.

A. locally mounted gauge; should B. locally mounted gauge; should NOT C strobe tachometer; should D. strobe tachometer; should NOT 19

HLT-07 SRO NRC EXAM

==

Description:==

JAW 31RS-E41-O0l-2, HPCI Operation From Outside Control Room, states Use the special strobe tachometer in the EOP gang box to monitor HPCI RPM WHILE operating the HPCI turbine manually. With HPCI speed above 5000 rpm the HPCI turbine should have tripped.

The A distractor is plausible if the applicant remembers that the local panel has various HPCI parameters indicated and thinks speed is one of them. Turbine speed is not indicated on a locally mounted gauge. The second part is correct.

The B distractor is plausible if the applicant remembers that the local panel has various HPCJ parameters indicated and thinks speed is one of them. Turbine speed is not indicated on a locally mounted gauge. The second part is plausible if the applicant confuses the overspeed trip setpoint or remembers that a lead is lifted which disables all HPCJ trips except overspeed hut thinks the overspeed is disabled also.

The D distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses the overspeed trip setpoint or remembers that a lead is lifted which disables all HPCI trips except overspeed but thinks the overspeed is disabled also.

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

20

HLT-07 SRO NRC EXAM

References:

NONE K/A:

206000 High Pressure Coolant Injection System K5. Knowledge of the operational implications of the following concepts as they apply to HIGH PRESSURE COOLANT INJECTION SYSTEM: (CFR: 41.5 /45.3)

K5.06 Turbine speed measurement: BWR-2,3,4 2.6* 2.6 LESSON PLAN/OBJECTIVE:

E41-HPCT-LP-00501, High Pressure Coolant Injection System, EO 005.021.A.O1 &

005.005.A.06 References used to develop this ciuestion:

31R5-E41-OO1-2, HPCI Operation From Outside Control Room 34S0-E41-OO1-2, High Pressure Coolant Injection System 21

E41-IIPCI-LP-00501 Page 5 of 92 HIGH PRESSURE COOLANT INJECTION SYSTEM

19. Given HPCI is injecting into the RPV, STATE the system response to a loss of the following power supplies: (005.005.a.07)
a. 25OVDC MCC 2B (R24-5022)
b. 125 VDC CAB 2B (R25-S002)

System is in STANDBY per 34S0-E41-00l -1/2, High Pressure Coolant Injection (HPCI)

System. (05.00 1 .a.07)

(005.004.a.03)

  • 29. Given Plant conditions, EVALUATE the conditions and DETERMINE if the HPCI turbine should have tripped. (005.005.a.06)
  • 30. Given Plant conditions, START HPCI locally per 3 1RS-E41-OOl-l/2, HPCI Operations From Outside the Control Room. (005.021.a.01)
  • 31. Given plant conditions, DETERMINE whether the HPCI system trip signals, except overspeed, should be disabled per 31RS-E41-00l-l/2, HPCI Operation from outside the Control Room.

(005.021 .a.03)

  • 32. Given plant conditions, SHUTDOWN HPCI locally per 3IRS-E41-001-1/2, HPCI Operation from Outside the Control Room. (005.022.a.0l)

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 6 OF 17 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

HPCI OPERATION FROM OUTSIDE CONTROL ROOM 31RS-E41-001-2 2.4 The following step disables all trip signals except the overspeed trip. Careful monitoring of the following HPCI operating parameters must be performed to attempt to maintain normal HPCI operation:

PARAMETER INDICATOR DESIRED RANGE NOTE:

Turbine Speed hand held strobe tachometer 2000-5000 rpm Turbine Exhaust Pressure 2E41-R005 on 2H21-P014 < 140 psig Pump Suction Pressure 2E41-R004 on 2H21-P014 < 10 inches Hg vacuum 4.1.13 LIFT lead at terminal D-4 in terminal box 2E41-C002-3 on the HPCI turbine pedestal.

4.1.14 At 087RGR25 slowly THROTTLE OPEN 2E41-F001, HPCI Turbine Steam Inlet MOV.

4.1.15 IF HPCI turbine speed changes are necessary, manually throttle 2E41-F001 OPEN and CLOSED as required.

4.1.16 At 106RGR24, CLOSE 2E41-F012, HPCI Mm Flow Isol MOV.

4.1 .17 At 2R24-S022, Frame 4A, 2E41 -C002-1, Barometric CNDSR Condensate Pump, perform the following steps:

4.1 .17.1 Confirm OPEN/OPEN the breaker.

4.1.17.2 OPEN the following links in the top compartment of the breaker frame:

-4A1 -4A4

-4A2 -4A5

-4A3 -4A6 4.1.17.3 INSTALL jumper on the INTERNAL side of terminal point 4A3 to the INTERNAL side of 4A1 in the top compartment of the breaker frame.

4.1.17.4 CYCLE the breaker as necessary to operate 2E41-C002-1, Vacuum Tank Condensate Pump to maintain barometric condenser level < 14 3/4 inches, indicated on 2E41-N760.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 5 OF 73 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

HIGH PRESSURE COOLANT INJECTION (HPCI) SYSTEM 34SO-E41-001-2 23.3 5.2 LIMITATIONS NOTES All turbine trips, except the mechanical overspeed, can be reset from the Control Room.

The mechanical overspeed will reset automatically at < 3000 rpm.

5.2.1 The following signals will cause a HPCI turbine trip:

5.2.1.1 Manual push-button on panel 2H11-P601 5.2.1.2 High Turbine Exhaust Pressure ( 140 psig) 5.2.1.3 Low HPCI Pump Suction Pressure (10 Inches Hg VAC) 5.2.1.4 Mechanical Overspeed [5000 rpm (125%)]

5.2.1.5 HPCI LogicAORLogicB Isolation Signal 5.2.1.6 High Reactor Vessel Water Level [ 51.7 Inches (Level 8)]

5.2.1.7 Local Manual Trip To manually trip turbine, lift knurled knob. Release knob and turbine will auto reset.

NOTE:

Posted @ Turbine 5.2.2 The HPCI System will automatically isolate upon receipt of any of the following signals:

5.2.2.1 HPCI Steam Line High Differential Pressure (Steam line break):

Instrument Setpoint, 202 H 0 increasing OR -100 H 2 0 decreasing.

2 (Technical Specification Limit <303%).

5.2.2.2 HPCI Steam Supply Low Pressure: Instrument Setpoint 134 psig (2E41-N058A, 2E41-N058B, 2E41-N058C, & 2E41-N058D). (Technical Specification Limit> 100 psig).

5.2.2.3 Turbine Exhaust Diaphragm High Pressure: Instrument Setpoint 10 psig.

(Technical Specification Limit < 20 psig).

5.2.2.4 HPCI Emergency Area Cooler Temperature High of 165CF (Technical Specification Limit 169cF).

5.2.2.5 HPCI Pipe Penetration Room Temperature High of 165CF (Technical Specification Limit < 169F).

MGR-0001 Ver. 4

HLT-07 SRO NRC EXAM

8. 209001K1.12 001 A transient has occurred on Unit 2.

At 1000 the following conditions exist:

o Drywell pressure 0.7 psig o Reactor water level -95 inches decreasing 2 inches/minute o Reactor pressure 500 psig decreasing 2 psig/minute o 2A Core Spray pump is MANUALLY started Which ONE of the choices below completes the following statements?

At 1001, the number of CS/RHR Pump Room Coolers operating is At 1005, CS/RT-IR Pump Room Coolers will be operating.

A. zero (0);

ONLY two (2)

B. zero (0);

ALL four (4)

C. two(2);

ONLY two (2)

D two (2);

ALL four (4) 22

HLT-07 SRO NRC EXAM

==

Description:==

In AUTO the RCIC room coolers BOO4AJB will automatically start on any one of the following:

RCIC initiation (F045 open)

High RCIC diagonal temperature (A: 95°F, B: 100°F)

In AUTO the HPCI room coolers BOO5AIB will automatically start on any one of the following:

HPCI initiation (FOOl open)

High HPCI diagonal temperature (A: 95°F. B: 100°F)

In AUTO the NE or SE Core Spray and RHR diagonal coolers BOO3AIB or BO02AJB will both automatically start on any one of the following:

Core spray pump COO1A I COOl B running RHR pump in diagonal running High diagonal temperature (A: 135°F, B: 140°F)

In AUTO the CRD diagonal coolers (BOO1AIB) will both automatically start on High CRD room temperature (A: 95°F, B: 100°F).

The A distractor is plausible if the applicant confuses the CS/RHR cooler operation with the CRD cooler operation which does not start when a CRD pump is started. The applicant would then select zero (0) coolers are operating at 1001. The second part is plausible if the applicant realizes that an initiation signal will exist (-101 ) at 1005 but only thinks there are only two coolers for the Core Spray system similiar to CRD.

The B distractor is plausible if the applicant confuses the CS/RHR cooler operation with the CRD cooler operation which does not start when a CRD pump is started. The applicant would then select zero (0) coolers are operating at 1001. The second part is correct.

The C distractor is plausible since the first part is correct. The second part is plausible if the applicant realizes that an initiation signal will exist (-101 ) at 1005 but only thinks there are only two coolers for the Core Spray system similiar to CRD.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

23

HLT-07 SRO NRC EXAM

References:

NONE K/A:

209001 Low Pressure Core Spray System Ki. Knowledge of the physical connections and/or cause effect relationships between LOW PRESSURE CORE SPRAY SYSTEM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.12 ECCS room coolers 2.9 3.1 LESSON PLAN/OBJECTIVE:

T41-SC HVAC-LP-01303, Secondary Containment HVAC Systems, EO 037.003.A.06 References used to develop this question:

3450-E21-OO1-2, Core Spray System 24

T41-SC HVAC-LP-01303 Page 8 of 86 Secondary Containment HVAC Systems

11. Given a list of statements, IDENTIFY the statement which best describes the plant response to an isolation signal being received by the following ventilation zones: (037.011 .A.12, 037.0 12.A.09, 037.022.A.12, 037.023.A.09)
a. Unit 2 Reactor Zone
b. Unit 1 Reactor Zone
c. Unit 2 Refueling Zone
d. Unit 1 Refueling Zone
12. Given a list of statements, IDENTIFY the statement which best describes the normal lineup of the Equipment Area Coolers. (037.003.A.05, 037.004.A.02, 037.005.A.04)
13. Given a list of statements, IDENTIFY the statement which best describes the trips and automatic start signals associated with the Safeguard Equipment Cooling (SEC) coolers.

(037.003 .A.06, 037.004.A.03, 037.005 .A.05)

14. Given a list of statements, IDENTIFY the statement which best describes the significance of receiving the SEC Auto Initiation Signal Present annunciator on P650. (037.002.A.0l)
15. Given a list of statements, IDENTIFY the statement which best describes how to reset the SEC coolers once they have started due to an automatic actuation signal. (037.002.A.03)
16. Given plant conditions involving Secondary Containment HVAC, DETERMINE if a Technical Specification Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (300.01 0.C.0 1)
17. Given plant conditions involving Secondary Containment HVAC, DETERMINE if a Technical Requirements Manual (TRM) Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (300.01 0.C.02)
18. Given plant conditions involving Secondary Containment HVAC, DETERMINE the Required Action(s) and Completion Time(s) in accordance with Technical Specifications for any combination of INOPERABLE systems, structures or components and the bases for the action(s). (SRO Only) (300.006.C.02)

Initial License (LT) LEARNING OBJECTIVES

1. Given appropriate references and SECONDARY CONTAINMENT system configuration, ANALYZE the effects of a SECONDARY CONTAINMENT system component malfunction.

(H-OP-90000.00 1)

2. Given appropriate references and SECONDARY CONTAINMENT system configuration, DIAGNOSE the SECONDARY CONTAINMENT system response to component misalignment. (H-OP-90000.002)

T41-SC HVAC-LP-01303 Page 43 of 861 Secondary Containment HVAC Systems I

F. Operation of the Equipment Area Coolers During normal operation of both Units 1 and 2, one SEC cooler is running in the CRD diagonal cooling the running CRD pump. One SEC cooler is running in the RCIC and HPCI rooms. All other SEC coolers are secured.

2. In addition to these coolers, Unit 2 normally operates eight of its nine remaining Reactor Building coolers.
a. All Reactor Building coolers except B009 have a two-position switch -

OFF/RUN. B009 has a three-position switch OFF/AUTO/RIJN.

b. The standby cooler (11009) in the main steam chase normally remains in AUTO and will automatically start if the steam chase temperature reaches 155°F. This action will also alarm an annunciator in the Control Room.
3. The SEC coolers in Unit 2 use a three position switch OFF/AUTO/RUN. All SEC coolers except the running CRD, RCIC, and HPCI coolers are normally in AUTO. Unit 1 SEC coolers use a four position switch OFF/STBY/AUTO/RIJN.

The STBY position is to be used during testing only.

4. In AUTO the CRD diagonal coolers (BOO lA/B) will both automatically start on High CRD room temperature (A: 95°F, B: 100°F).
5. In AUTO the southeast diagonal Core Spray and RHR coolers, BOO2AIB, will both automatically start on any one of the following:
a. Core spray pump COO lB (Unit 1 COO1A) running.
b. RHR pump COO2B (Unit 1 - COO2A) running.
c. RHR pump COO2D (Unit 1 - COO2C) running.
d. High SE diagonal temperature (A: 135°F, B: 140°F).
6. In AUTO the NE Core Spray and R.}IR diagonal coolers BOO3A/B will both automatically start on any one of the following:
a. Core spray pump COO1A (Unit 1 COO1B) running.
b. RHR pump COO2A (Unit 1 - COO2B) running.
c. RHR pump COO2C (Unit 1 - COO2D) running.
d. High NE diagonal temperature (A: 135°F, B: 140°F).

SOUTHERN NUCLEAR PLANT E.l. HATCH PAGE 130 OF 130 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

CORE SPRAY SYSTEM 34SO-E21-001-2 22.17 ATTACHMENT 13 ATTACHMENT PAGE:

TITLE: CORE SPRAY MANUAL STARTUP 1 OF 1 CORE SPRAY MANUAL STARTUP IF a valid Core Spray initiation signal is received AND Core Spray fails to inject, OR IF Core Spray is required to be started by the EOPs, THEN, PERFORM the following actions:

1. Start 2E21-COOIA(2E21-COO1B), Core Spray Pump, panel 2H11-P601 El
2. Confirm Core Spray Pump discharge pressure is > 265 PSIG LI
3. WHEN reactor pressure is less than or equal to 425 PSIG, fully open 2E21 -FOO5A(2E21 -FOO5B), lnbd Discharge Valve. El
4. WHEN Core Spray System flow increases to greater than 950 GPM, confirm 2E21-FO3IA(2E21-FO31B), Mm Flow Valve closes. LI
5. Throttle 2E21-FOO5A(2E21-FOO5B), lnbd Discharge Valve, as directed by the SS LI
6. Confirm Core Spray and RHR Room Coolers automatically start, panel 2H1 1-P654 OR 2H1 1-P657. El
7. Refer to 34S0-E21-001-2 Core Spray Manual Startup for additional required actions. El (Posted 2H1 1-P601)

MGR-0009 Ver. 3

HLT-07 SRO NRC EXAM

9. 209001K5.O1 001 A LOCA is in progress. The 2A Core Spray pump is injecting at 1000 gpm to maintain RPV water level above the top of active fuel.

Which ONE of the following would be an indication that the 2A Core Spray pump is experiencing cavitation?

A. The motor amps are steadily increasing.

B. The pump trips due to low suction pressure.

C The minimum flow valve intermittently opens and closes.

D. The pump discharge pressure steadily increases as flow decreases.

25

HLT-07 SRO NRC EXAM

==

Description:==

This question evaluates the applicants knowledge of how CS pump cavitation will affect plant indications.

Erratic flow is one consequence of cavitation. The opening (700 gpm) and closing (950 gpm) setpoints of E2l-FO31AJB is sufficiently close to the 1000 gpm being pumped through the system that it is possible, with sufficient cavitation occurring, for flow to swing 300+ gpm which will to cause F03 1 to open and close.

When cavitation occurs, motor amps will fluctuate, but on average, will actually decrease due to less water being pumped.

Motor winding temperature is an equilibrium condition based on heat removal systems (diagonal ventilation system used to cool the motor) and heat producing components (motot bearing friction, windage etc. and, principally, 1 R heating of the windings). Assuming the heat removal 2

system remains unchanged, overall heating of the windings due to 1 R will actually decrease, 2

over time, due to the reduced volume being pumped caused by cavitation.

Unlike some other ECCS systems, the Core Spray pumps do not have a low suction pressure trip.

The A distractor is plausible if the applicant confuses which way motor amps will change when a pump is experiencing cavitation.

The B distractor is plausible if the applicant remembers that some ECCS pumps have a low suction pressure trip and confuses this with Core Spray.

The D distractor is plausible if the applicant confuses flow changes on a cavitating pump with flow changes on a pump that is not expfriencing cavitation. Normally as flow decreases on a Core Spray (constant speed centrifugal) pump, discharge pressure follows the pump curve and increases towards shutoff head.

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

26

HLT-07 SRO NRC EXAM

References:

NONE K/A:

209001 Low Pressure Core Spray System K5. Knowledge of the operational implications of the following concepts as they apply to LOW PRESSURE CORE SPRAY SYSTEM: (CFR: 41.5 /45.3)

K5.O1 Indications of pump cavitation 2.6 2.7 LESSON PLAN/OBJECTIVE:

E21-CS-LP-00801, Core Spray System, EO 008.002.A.09 LT-LP-00073-04C, Pumps (indications of pump cavitation)

References used to develop this question:

Monticello 2007 NRC Exam Q#4 LT-LP-00073-04C, Pumps (indications of pump cavitation) 27

E21-CS-LP-00801 Page 4 of 68 CORE SPRAY SYSTEM

10. STATE the interlocks associated with the following Core Spray Valves: (008.002.a.05)
a. Minimum Flow Valve F03 1
b. Test By-Pass Valve F015
c. Outboard Isolation Valve F004
d. Inboard Isolation Valve F005
11. Given plant conditions and a Core Spray initiation, DETERMINE which Core Spray Valves will change position per 34S0-E21-00l-l/2, Core Spray System. (008.006.a.02)
12. Given a loss of 125 VDC power to a Core Spray Pump Logic Circuit, DETERMINE the effects of that power loss on Core Spray Operation. (008.0 14.a.03)
13. Given Core Spray is injecting to the RPV, STATE the action(s) necessary to throttle Core Spray flow per 34S0-E21-001-1/2, Core Spray System. (008.018.a.01)
14. From a list, SELECT the actions necessary to manually start the Core Spray System per 34SO-E2 1-001-1/2, Core Spray System. (008.002.a.0 1)
15. Given the Core Spray System is lined up to the CST, STATE the precaution for manual startup in 34S0-E21-001-1/2, Core Spray System. (008.002.a.04)
16. STATE 3 methods of determining whether Core Spray is injecting to the Reactor using:

(008.002.a.09)

a. Discharge Pressure
b. System Flow
c. RPV Level
  • 17. Given Plant conditions and the EOPs, EVALUATE these conditions and DETERMINE if the Core Spray System should be secured. (008.004.a.02)

(008.004.a.0 1)

a. Have auto initiated
b. Be injecting
  • 20. Given Core Spray has spuriously initiated, from a list, SELECT the appropriate actions as directed by 34AB-E 10-001 1/2, Inadvertent Initiation of ECCS/RCIC, to secure or prevent injection of Core Spray. (200.027.b.01)

BCO2Sr4_Pumps May 2011 Indications of cavitation are:

CAPACITY

  • Fluctuating pump discharge pressure A pump is designed to move a specific volume of fluid. The amount moved for a given set of
  • Fluctuating pumj flow rate conditions is plotted on a pump operational curve. The rated volumetric flow rate for a
  • Fluctuating pump motor current specific set of parameters is the capacity of the pump. Pump capacities are normally expressed
  • Excessive pump noise (pump sounds like it is in gallons per minute (gpm) or cubic feet per pumping rocks).

second (cfs).

If cavitation occurs, follow plant procedures and manufacturer and vendor guidelines. These steps CAVITATION may include the following, not necessarily in this order.

Cavitation is the formation of vapor bubbles in the low pressure region of the component and the 1. Verify the suction valve is open.

subsequent collapse of the vapor bubbles in the high pressure region of the component. 2. Reduce the speed of the pump.

It is commonly associated with centrifugal 3. Reduce the temperature of the fluid entering pumps, where the definition is modified to say the pump.

Cavitation is the formation of vapor bubbles in eye of the pump and the subsequent collapse of 4. Increase pressure on the suction of the pump.

the vapor bubbles in the impeller or in the Pressurize the head (expansion) tank or raise volute. When the pressure at the eye of the the height/volume of liquid in the head tank.

impeller is less than saturation pressure, some of the liquid water changes phase from a liquid to a 5. Throttle closed the discharge valve of a gas, and vapor bubbles form. The bubbles centrifugal pump.

collapse when the bubbles enter an area where the pressure is greater than the saturation pressure. This is usually in the vanes of the impeller.

Cavitation damages pumps due to the erosion and pitting that occurs when the vapor bubbles collapse. Cavitation should be avoided, prevented, and corrected to prevent damage to pump impellers and internals. It causes the impeller to erode, reducing its useful life. It can also cause excessive vibration, which could damage bearings, wearing rings, and seals.

Cavitation can be detected by its distinctive sound. The pump sounds like it is pumping rocks. This noise is produced when the vapor bubbles collapse.

BWR / COMPONENTS / CHAPTER 2 2 of 84 2011 GENERAL PHYSICS CORPORATION

/ PUMPS REV 4 GF@gpworldwide.com www.gpworldwide.com

HLT-07 SRO NRC EXAM

10. 211000K4.02 001 Unit 2 is operating at 100% RTP with the 2B SBLC pump inoperable.

o The 2B SBLC pump breaker is Danger Tagged OPEN The crew has just been tasked with performing 34SV-C41-002-2, Standby Liquid Control Monthly Test, for the 2A SBLC pump.

Which ONE of the choices below completes the following statement concerning the Squib Valve Continuity lights on the 2H1 1-P603 panel and the RWCU System?

After the operator locally starts the 2A SBLC Pump, A. two (2) Squib Valve Continuity lights on the 2Hl l-P603 panel will be ILLUMINATED and the RWCU System WILL isolate B. two (2) Squib Valve Continuity lights on the 2H1 l-P603 panel will be ILLUMINATED and the RWCU System WILL NOT isolate C. ONLY one (1) Squib Valve Continuity light on the 2H1 l-P603 panel will be ILLUMINATED and the RWCU System WILL isolate D ONLY one (1) Squib Valve Continuity light on the 2H1 1-P603 panel will be ILLUMINATED and the RWCU System WILL NOT isolate 28

HLT-07 SRO NRC EXAM

==

Description:==

Turning the Control Room keylock switch to either the Start Sys A or the Start Sys B position will cause the selected pump to start, both squib valves fire, loss of both continuity lights, and the Reactor Water Cleanup System isolates by its outboard isolation suction valve closing. (if the selected pump does not start, the keylock switch is turned to the other position in order to start the other pump). Manual initiation (Local) of SBLC will start the selected pump but will not fire either squib valve, not lose the Control Room Continuity lights nor isolate RWCU.

The A distractor is plausible if the applicant remembers that both Control Room Continuity lights are normally illuminated but confuses that RWCU will isolate on SBLC initiation from the control room while starting a pump locally.

The B distractor is plausible if the applicant remembers that both Control Room Continuity lights are normally illuminated and that RWCU will not isolate when attempting a SBLC initiation from the local panel.

The C distractor is plausible if the applicant remembers that with the 2B SBLC pump breaker tagged off there will not be power for the 2B Continuity circuit but confuses that RWCU will isolate on SBLC initiation from the control room while starting a pump locally.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

29

HLT-07 SRO NRC EXAM

References:

NONE K/A:

211000 Standby Liquid Control System K4. Knowledge of STANDBY LIQUID CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)

K4.02 Component and system testing 3.0 3.2 LESSON PLAN/OBJECTIVE:

C41-SBLC-LP-O1001, Standby Liquid Control, EO 011.002.A.04 References used to develop this guestion:

34S0-C41-003-2, Standby Liquid Control System 30

C41-SBLC-LP-Ol1O1 Page 2 of 58 STANDBY LIQUID CONTROL Initial License LD ENABLING OBJECTIVES

1. Given a P&ID or a simplified diagram of the Standby Liquid Control System, TRACE the flow path through the system for injection into the RPV. (01l.001.A.01)
2. Given a P&ID or a simplified diagram of the SBLC system, IDENTIFY the following components. (01 l.00l.A.02)
a. SBLC storage tank
b. SBLC pumps
c. SBLC squib valves
d. Valves FOOl, FOO3A/B and F008
e. SBLC Test Tank
3. Given a P&ID or a simplified diagram of the SBLC system LOCATE the following system interfaces. (01 l.001.A.03)
a. Demin Water system
b. Service Air systems
4. Given plant conditions, DETERMINE how the SBLC system will respond to a loss of Instrument/Service air. (011 .002.A.05)
5. Given a Plant Hatch load list, DETERMINE the power supplies for the following SBLC system components: (011.00 1.A.04)
a. SBLC pumps
b. SBLC squib valves
6. Given a list of SBLC system equipment and components, IDENTIFY those which have local controls. (011 .002.A.02)
7. STATE the two interlocks associated with the SBLC pumps and RWCU outboard isolation valve. (011 .002.A.04)
8. STATE the three positive indications that SBLC is injecting into the Reactor Core.

(011 .002.A.03)

11. Given plant conditions requiring shutdown of the SBLC systems, DETERMINE steps necessary to shutdown SBLC system per 34S0-C41-003-l/2, Standby Liquid Control System.

(01 1.003.A.01)

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 8 OF 55 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

STANDBY LIQUID CONTROL SYSTEM 34S0-C41-003-2 7.2 SYSTEM STARTUP AND OPERATION 7.2.1 Manual Initiation Control Room tZous THIS SUBSECTION IS PERFORMED ONLY AS DIRECTED BY PLANT HATCH EMERGENCY OPERATING PROCEDURES.

7.2.1.1 Unlock AND place the SBLC Pump Select Switch, panel 2H11-P603, in the START Sys A OR START Sys B position. LI 7.2.1.2 Confirm the following:

  • 1 106A AND I 106B Squib VIv Ready indicating lights EXTINGUISHED LI
  • Selected 2C41-COOIA or2C4l-COO1B, SBLC Pump, has started LI 7.2.1.3 IF the selected pump fails to start, repeat steps 7.2.1.1 through 7.2.1.2 for the other pump. LI 7.2.1.4 Confirm OR close 2G31-F004, Rx Water Cleanup Vlv, panel 2H1 1-P601. LI 7.2.1.5 IF the SBLC system fails to initiate, perform the SBLC PUMP CONTROL SWITCH OVERRIDE section of 31 EO-EOP-1 00-2. LI 7.2.1.5.1 If above step was successful, continue at 7.2.1.7:

otherwise, proceed to the following step. LI 7.2.1.6 IF the Standby Liquid Control system fails to initiate per 31 EO-EOP-1 00-2, perform local initiation per the Manual Initiation Local subsection of this procedure. LI MGR-0001 Ver. 4

HLT-07 SRO NRC EXAM

11. 212000A2.08 001 Unit 2 is in Mode 4 with RHR Shutdown Cooling in service.

An event occurs causing RWL to decrease and results in the following:

o 2C32-R606A-C GEMAC (+) 2 inches (LOWEST level reached) o Subsequently, the NPO recovers RWL to (+) 20 inches Which ONE of the choices below completes the following statements?

During this event, the RPS RWL Scram Setpoint EXCEEDED.

34AB-El 1-001-2, Loss Of Shutdown Cooling, REQUIRED to be entered.

A. was; is NOT B was; is C. was NOT; is NOT D. was NOT; is 31

HLT-07 SRO NRC EXAM

==

Description:==

With the A, B & C GEMACs decreasing to +2 inches, a RPS Scram setpoint will have occurred at +3 inches. This will require 34AB-El 1-001-2, Loss Of Shutdown Cooling, to be entered since +3 inches is the setpoint for the RHR SDC and injection valves to close. When these valves close the RHR pump in SDC trips, resulting in the entry to the abnormal. JAW TS Table 3.3.1.1-1 Reactor Protection System Instrumentation, Function 4 Reactor Vessel Water Level -Low, Level 3, the RPS Scram setpoint must be zero (0) inches.

The A distractor is plausible since the first part is correct. The second part is plausible if the applicant does not remember that 3 inches is the point where the R}IR valves isolate.

The C distractor is plausible if the applicant remembers that the TS RPS Scram setpoint is zero (0) inches and confuses this with the actual plant setpoint. if the applicant thinks that RWL did not drop to 0 inches then a RPS Scram setpoint was not exceeded. The second part is plausible if the applicant confuses TS vs. Actual plant setpoint and thinks SDC did not isolate, therefore is still in service with no need to enter the abnormal procedure.

The D distractor is plausible if the applicant remembers that the TS RPS Scram setpoint is zero (0) inches and confuses this with the actual plant setpoint. if the applicant thinks that RWL did not drop to 0 inches then a RPS Scram setpoint was not exceeded. The second part is plausible if the applicant thinks that since RWL decreased, there will be less pump head to the RHR pump thus causing an increase in reactor coolant temperature which is an entry condition to this abnormal procedure.

A. Incorrect See description above.

B. Correct See description above.

C. Incorrect See description above.

D. Incorrect See description above.

32

HLT-07 SRO NRC EXAM

References:

NONE K/A:

212000 Reactor Protection System A2. Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.08 Low reactor level 4.1* 4.2*

LESSON PLAN/OBJECTIVE:

C32-RWLC-LP-00202, Reactor Water Level Control, EO 002.020.A.06 C7 1 -RPS-LP-0 1001, Reactor Protection System, EO 300.008 .A.02 T23-PC-LP-0 1301, Primary Containment, EO 013.047. A.05 References used to develop this question:

34AB-C71-001-2, Scram Procedure 34AB-E1 1-001-02, Loss of Shutdown Cooling 33

C32-RWLC-LP-00202 Page 2 of 64 REACTOR WATER LEVEL CONTROL Initial License LLD ENABLING OBJECTIVES From a list of statements, SELECT the statement that identifies why the W Level Instrument is the preferred level input for the RWLC system. (002.0 19.A.04)

2. From a list of statements, ANALYZE the statements to determine which identifies how the dP mode of RWLC system functions. (002.026.A.02)
3. From a list of parameters, SELECT the parameters that input to the Three Element Reactor Water Level Control circuit. (002.020.A.0l)
4. From a list of statements, ANALYZE the statements to determine which identifies the initial and final response of RFPT speed and Reactor water level for any one of the following failures at maximum operating power condition. (002.020.A.05)
a. One Reactor feedwater flow or MSL flow detector failure
b. Loss of Vital AC
c. Loss of Instrument Bus A
d. Inadvertent HPCI Injection
5. From a list of statements, ANALYZE the statements to determine which one identifies how the following will respond to an upscale or downscale failure of the Reactor Vessel Water Level input to RWLC. (Assume rated conditions) (002.020.A.06)
a. Actual vessel water level
b. RPS Trip System
c. Reactor Feed Pump Turbine
d. Main Turbine
e. HPCI and RCIC
f. PCIS Group Isolation
g. Reactor Recirculation Pumps
h. RHR System
i. Core Spray System
j. Diesel Generators
k. Control Rod Drive System
6. From a list of plant conditions, DETERMINE the condition(s) which would result in the control of the RFPT automatically swapping to the Speed Setter Mode. (002.027.C.Ol)
7. Given a list of statements, ANALYZE the statements to determine which identifies the actions to be taken if RFPT control automatically swaps to the Speed Setter Mode and why those actions are necessary. (002.027.C.02, 002 .027.A.03)

C71-RPS-LP-O1001 Page 6 of 112 REACTOR PROTECTION SYSTEM

  • 30. In accordance with 34G0-OPS-013-1/2, Normal Plant Shutdown, DISCUSS the notes and precautions concerning the transfer of the Mode Switch from RUN to START UP/HOT STANDBY. (010.019.a.04)
31. Given plant conditions which require tripping a channel of RPS, SUMMARIZE the steps necessary to place an RPS channel in the Tripped condition according to Tech Specs.

(010.012.a.0l)

  • 32. Given plant conditions involving the RPS System, DETERMINE if a Technical Specification Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABiLITY and associated NOTES) (3 00.0 06.a. 19)
  • 33 Given plant conditions involving the RPS System, DETERMINE the Required Action(s) and Completion Time(s) in accordance with Technical Specifications for any combination of INOPERABLE systems, structures or components and the bases for the action(s). (SRO Only)

(300.010.a.l4)

  • 34* Given plant conditions involving the RPS System, DETERMINE the Required Actions for Completion Times 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in accordance with Technical Specifications for any combination of INOPERABLE systems, structures or components. (300.0 11.a.07)
35. Given a failed IRIvIIAPRM, DESCRIBE the steps necessary to bypass the failed instrument per 34AR-603-219-l/2, APRM Upscale, or 34AR-603-203-1/2, IRM Bus A Upscale Trip or 1NOP. (010.010.a.01)
36. STATE the possible effects on the Plant, if a spurious half scram is repeatedly tripping and being reset over a short period of time. (0l0.0ll.a.02)
  • 37 Given plant conditions which resulted in a Reactor Scram, using plant procedures and Tech Specs, DETERMINE the cause of the Reactor Scram. (300.008.a.02)
  • 38. Given that a loss of an RPS bus has occurred, SUMMARIZE the actions to correctly respond to these conditions per 34AB-C71-002-1/2, Loss of RPS. (200.102.c.01)
39. Given a list, IDENTIFY the statement that describes the purpose of pulling the scram solenoid fuses during an ATWS. (0l0.016.a.01)
40. In accordance with 31RS-OPS-001-1/2, Shutdown from Outside the Control Room, LOCATE the breakers used to de-energize the APRMs and INITIATE a scram. (0l0.017.a.0l)
41. In accordance with 31RS-OPS-001-l/2, Shutdown from Outside the Control Room, SUMMARIZE the steps necessary to trip the Scram Discharge Volume High Level switches.

(010.018.a.02)

42. DESCRIBE the reason for the 10-second time delay between receiving a reactor scram signal and resetting the signal. (0l0.0l0.A.03)
  • 43* Given plant conditions involving the RPS System, DETERMINE if a Technical Requirements Manual (TRM) Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (300.01 0.C.02)

T23-PC-LP-01301 Page 8 of 160 PRIMARY CONTAINMENT

28. Given a list of statements, IDENTIFY the statement which best describes the plant conditions which will generate the following PCIS isolation signals: (013.045.A.05, 013.046.A.05, 013.047.A.05)
a. Group 1 isolation
b. Group 2 isolation
c. Group 3 isolation
d. Group 4 isolation
e. Group 5 isolation
f. Group 6 isolation
29. Given a list of statements, IDENTIFY the statement which best describes the steps required to reset the following PCIS isolation signals: (013.045.A.06, 013.046.A.06, 0l3.047.A.06, 013.067.A.0l, 040.006.A.01)
a. Group 1 isolation
b. Group 2 isolation
c. Group 3 isolation
d. Group 4 isolation
e. Group 5 isolation
f. Group 6 isolation
30. Given a list of statements, IDENTIFY the statement which best describes the primary containment characteristics during a DBA LOCA, including the process by which energy is absorbed and dissipated by the primary containment. (200.004.A.03)
31. Given a list of statements, IDENTIFY the statement which best describes the consequences of having the torus to drywell vacuum breakers fail either open or closed during a DBA LOCA.

(200.004.A.02)

NORMAL OPERATIONS

(013.006.A.0l)

(013.007.A.0l)

SOUTHERN NUCLEAR PLANT E.I. HATCH PAGE 19 OF 31 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

SCRAM PROCEDURE 34AB-C71-OO1-2 1t2 ATTACHMENT j.. ATTACHMENT PAGE:

TITLE: PRIMARY CONTAINMENT ISOLATION CONFIRMATION 9 OF 9 6M GROUP 6 ISOLATION TNOTE. Some of the following conditions can be indicative of a RHR pipe break in the Reactor U Building. Refer To 34AB-T22-OO1 -2 as applicable.

6.1 EITHER of the following conditions cause isolation:

. Reactor Water Level Low (level 3) (+3)

. Reactor Pressure High (138 psig) 6.2 Confirm the following valves have CLOSED:

NOTES Position indication can also be found on SPDS diagnostic AND on panel 2H1 1-P601 vertical display except as noted.

2H1 1 -P602 2E1 1-F009 SDC Suction VIv El 2H11-P601 2E1 1-F008 SDC Suction VIv El MGR-0009 Rev. 5.0

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH .

20F33 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF SHUTDOWN COOLING 34AB-E11-001-2 6.10 4.0 SUBSEQUENT OPERATOR ACTIONS 4.1 IF, Reactor pressure increases to greater than 138 psig, El OR Reactor water level decreases to less than 3 inches, LI OR Drywell pressure increases to greater than 1.85 psig: El 4.1.1 Confirm tripped trip RHR Pumps operating in Shutdown Cooling. LI 4.1.2 Confirm closed close the following valves:

  • 2E1 1-F009, SDC Suction Valve LI
  • 2E11-FOI5B, RHR lnbd ml Valve 4.2 IF an actual high Drywell pressure condition (1 .85 psig) exists, exit this procedure, AND perform the Manual LPCI Initiation WHILE in Shutdown Cooling subsection of 34S0-E1 1-010-2, Residual Heat Removal System. El NOTES Attachments 1, 2, and 3 may be used as appropriate to determine time for bulk Reactor coolant to reach saturation temperature.

COOLANT HEATUP MAY OCCUR IN THE CORE AREA WITH LITTLE OR NO INDICATED TEMPERATURE INCREASE AT THE RECIRCULATION PUMP SUCTION OR RWCU INLET. VESSEL PRESSURIZATION COULD RESULT EVEN WITH REACTOR HEAD AUTION VENTS OPEN. VESSEL BULK COOLANT TEMPERATURE CAN BE ASSUMED TO BE UP TO 20°F HIGHER THAN INDICATED ON RWCU INLET, PROVIDED THAT NATURAL CIRCULATION IS OCCURRING. H NATURAL CIRCULATION IS NOT OCCURRING, THEN RWCU INLET TEMPERATURE WILL NOT BE INDICATIVE OF VESSEL BULK COOLANT TEMPERATURE.

Critical 4.3 IF 212°F, THEN determine estimated time for bulk Reactor coolant to reach saturation temperature. LI MGR-0001 Ver. 3

HLT-07 SRO NRC EXAM

12. 212000K3.03 001 Unit 2 is operating at 100% RTP when a loss of RPS Bus 2A occurs.

Which ONE of the choices below describes the affect of losing RPS Bus 2A on the LPRMs?

A. ONLY Half of the LPRMs inputs associated with APRMs 2B & 2D are still VALID.

B. ONLY Half of the LPRMs inputs associated with APRMs 2A & 2C are still VALID.

C. NONE of the LPRMs inputs associated with APRMs 2B & 2D are still VALID.

D ALL of the LPRMs inputs associated with APRMs 2A & 2C are still VALID.

34

HLT-07 SRO NRC EXAM

==

Description:==

Edwin, this was question 3 of 10 that you have already reviewed. Any discussed changes have been incorporated.

Two independently controlled High Voltage Power Supply (HVPS) Modules are used per APRM chassis. One module provides the normal supply of high voltage and powers all LPRM detectors connected to the APRM chassis. The second module serves as a backup power supply and provides power to a bypassed LPRM detector selected for currentlvoltage curve plotting, If the self-test detects failure of the normal power supply, the backup power supply automatically switches to supply high voltage to the LPRM detectors and a self-test alarm is issued. In this event, the APRM is incapable of performing current/voltage plotting until two fully functional HVPS modules are available.

Each APRM chassis receives power from one low-voltage power supply (LVPS) module connected to 120-VAC RPS Bus A and one LVPS module connected to 120-VAC RPS Bus B.

Each APRMs two-out-of-four voter logic module receives power from the RPS bus associated with the APRM channels trip outputs, as well as from the APRM chassis. Since only one RPS Bus is lost (A), all of the LPRMs will still receive power to provide valid indication of reactor power.

The A distractor is plausible if the applicant confuses the power supply to the APRMs and does not remember they have dual power supplies and thinks half of the LPRMs inputs are lost to APRMs 2B & 2D. Also plausible since two of the 2/4 Logic Modules have also lost power and thinks two APRMs have also lost power, thus LPRMs have lost power.

The B distractor is plausible if the applicant confuses the power supply to the APRMs and does not remember they have dual power supplies and thinks half of the LPRMs inputs are lost to APRMs 2A & 2C. Also plausible since two of the 2/4 Logic Modules have also lost power and thinks two APRMs have also lost power, thus LPRMs have lost power.

The C distractor is plausible if the applicant confuses the power supply to the APRMs and does not remember they have dual power supplies and thinks all of the LPRMs inputs are lost to APRMs 2B & 2D. Also plausible since two of the 2/4 Logic Modules have also lost power and thinks two APRMs have also lost power, thus LPRMs have lost power.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

35

HLT-07 SRO NRC EXAM

References:

NONE K/A:

212000 Reactor Protection System K3. Knowledge of the effect that a loss or malfunction of the REACTOR PROTECTION SYSTEM will have on following: (CFR: 41.7 / 45.4)

K3.03 Local power range monitoring system: Plant-Specific 3.3 3.4 LESSON PLAN/OBJECTIVE:

C51-PRNM-LP-01203, Power Range Neutron Monitoring System, EO 012.003.D.12 References used to develop this question:

Unit Two FSAR, Chapter 7.6.2.2.3, LPRM System (dual power supplies)

Unit Two FSAR, Figure 7.6-7 Sheet 1, APRM/RBM Power Distribution Unit Two FSAR, Figure 7.6-7 Sheet 3, ARPM/RPS Interface Block Diagram 36

C51-PRNIvI-LP-01203 Page 3 of 86 POWER RANGE NEUTRON MONITORING SYSTEM

5. STATE the number of LPRMs assigned to a specific APRM. (012.003.d.09)
6. Given Plant conditions, ANALYZE these conditions to DETERMINE if a Recirculation Flow unit generated Rod Block should have occurred. (012.003.f.0l)
7. STATE the LPRM detector assignments to the RBM channels to include number and level for a control rod selected near the center of the core. (012.003.e.05)
8. DESCRIBE the RBM nulling sequence to include: Initiation signal to null, recorder readings during and after the null sequence, and the function of the APRM reference signal.

(012.003 .e.06)

9. DETERMINE the power supplies to the following: (012.003.d.12)
a. APRM Drawer
b. RBM Drawer
c. APRM/RBM Recorders
10. Given plant conditions, ANALYZE those conditions to DETERMINE if an APRM generated Rod Block should have occurred. (012.003.d.02)
11. Given plant conditions, ANALYZE those conditions to DETERMINE if a RBM channel generated Rod Block should have occurred. (012.003.e.01)
12. Given plant conditions, ANALYZE those conditions to DETERMINE if a Two-out-of-Four Logic System generated reactor scram should have occurred. (012.003.d.0l)
13. Given a list of plant conditions, IDENTIFY which would result in an APRM INOP trip.

(012.003.d.04)

14. Given a value of Core Power and the actual APRM reading, STATE if any APRMs need gain adjustment. (400.009.a.02)
15. Given initial unadjusted LPRM readings at BOC, DETERMINE how those readings would change over core life. (400.062.a.02)
16. Given plant conditions, ANALYZE those conditions to DETERMINE the required actions if power oscillations occur with OPRMs INOPERABLE. (012.003.f.04)

HN P-2-FSAR-7 G. Environmental Considerations The wiring, cables, and connectors located in the primary containment are designed for the same environmental conditions as the SRMs. (See section 3.11.)

7,6Z23 Local Power Range Monitor System A. Equipment Design B. Identification The LPRM system consists of fission chamber detectors, signal-conditioning equipment, and trip functions. The LPRM system also provides outputs to the APRM, OPRM, RBM, and process computer.

C. Power Supply The high-voltage power supply (HVPS) modules provide variable 0 to 200-V-dc to power the LPRM detectors. The HVPS current rating is 120 mA. The 386SX computer module controls the HVPS output voltage and current via the data bus and a digital-to-analog (DIA) converter on the broadcaster module.

Two independently controlled HVPS modules are used per APRM chassis. One module provides the normal supply of high voltage and powers all LPRM detectors connected to the APRM chassis. The second module serves as a backup power supply and provides power to a bypassed LPRM detector selected for current/voltage curve plotting. If the self-test detects failure of the normal power supply, the backup power supply automatically switches to supply high voltage to the LPRM detectors and a self-test alarm is issued. In this event, the APRM is incapable of performing current/voltage plotting until two fully functional HVPS modules are available.

D. Physical Arrangement The LPRMs include 31 LPRM detector strings having detectors located at different axial heights in the core. Each string contains four fission chambers. These assemblies are distributed to monitor four horizontal planes throughout the core.

Figure 7.6-6 shows the LPRM detector radial layout scheme that provides a detector assembly at every fourth intersection of the water channels around the fuel bundles. Every location has either an actual detector assembly or a symmetrically equivalent assembly in some other quadrant.

Five LPRM assemblies located at core positions 04-37, 28-29, 36-05, 44-37, and 44-45, contain electrochemical corrosion potential (ECP) sensors in addition to the standard four LPRM detectors. The ECP sensors measure the stress 7.6-15 REV 29 9/11

(Th C) 0 0

4 0) 0 I

= c z

=

I

- C, C,

C,

-t 1

-U 4,

0 C

m

-D I 0

Cl) m m

ni C,)

0 m

-n C 0, (0 0

z 0

HLT-07 SRO NRC EXAM

13. 215002K6.04 001 Unit 2 is operating at 100% power with a centrally located control rod selected.

o The operator bypasses APRM channel A due to bad CPU Which ONE of the choices below identifies the effect this action will have on the RBM A channel?

RBM A will automatically use the A. C APRM channel to monitor the neutron flux in the locality of the selected rod B C APRM channel to determine one of the three upscale setpoints C. D APRM channel to monitor the neutron flux in the locality of the selected rod D. D APRM channel to determine one of the three upscale setpoints 37

HLT-07 SRO NRC EXAM

==

Description:==

Each RBM channel designates a hierarchy of normal and alternate APRM channels to use as their reference APRM channel. The alternate channels are used in hierarchical order when the preferred channels are not available. The primary reference APRM for RBM A t is APRM A with first alternate as C APRM and the second alternate is D APRM. The primary reference APRM for RBM B is APRM B with first alternate as D APRM and the second alternate is C APRM. When the RBM loses its primary reference APRM it automatically selects the first alternate APRM. The RBM uses its reference APRM for core flux and uses LPRMs for local flux.

The A distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses the APRM with the RBM and thinks it will monitor the local flux around the control rod, which is what the RBM does.

The C distractor is plausible if the applicant confuses the the primary/alternate APRMs. The second part is plausible if the applicant confuses the APRM with the RBM and thinks it will monitor the local flux around the control rod, which is what the RBM does.

The D distractor is plausible if the applicant confuses the the primary/alternate APRMs. The second part is correct.

A. Incorrect See description above.

B. Correct See description above.

C. Incorrect See description above.

D. Incorrect See description above.

38

HLT-07 SRO NRC EXAM

References:

NONE K/A:

215002 Rod Block Monitor System K6. Knowledge of the effect that a loss or malfunction of the following will have on the ROD BLOCK MONITOR SYSTEM: (CFR: 41.7 I 45.7)

K6.04 APRM reference channel: BWR3,4,5 2.8 3.0 LESSON PLAN/OBJECTIVE:

C51-PRNM-LP-01203, Power Range Neutron Monitoring System, EO 012.003f.05 References used to develop this ciuestion:

C5 1 -PRNM-LP-0 1203 Figure 16, RBM Functional Block Diagram Modified from HLT Database HLT-4 NRC Exam Q#13 ORIGINAL QUESTION (HLT-4 NRC Exam Q#13)

Unit 2 is operating at 100% power with a centrally located control rod selected.

o The operator bypasses APRM channel B Which ONE of the following identifies the effect this action will have on the RBM B channel?

The B RBM will automatically use the A. C APRM channel to monitor the neutron flux in the locality of the selected rod B. C APRM channel to determine one of the three upscale setpoints C. D APRM channel to monitor the neutron flux in the locality of the selected rod D. V D APRM channel to determine one of the three upscale setpoints 39

C5 1 -PRNM-LP-0 1203 Page 3 of 86 POWER RANGE NEUTRON MONITORING SYSTEM

5. STATE the number of LPRMs assigned to a specific APRM. (012.003.d.09)
6. Given Plant conditions, ANALYZE these conditions to DETERMINE if a Recirculation Flow unit generated Rod Block should have occurred. (012.003.f.01)
7. STATE the LPRM detector assignments to the RBM channels to include number and level for a control rod selected near the center of the core. (012.003.e.05)
8. DESCRIBE the RBM nulling sequence to include: Initiation signal to null, recorder readings during and after the null sequence, and the function of the APRM reference signal.

(0 12.003 .e.06)

9. DETERMINE the power supplies to the following: (012.003.d.12)
a. APRM Drawer
b. RBM Drawer
c. APRM/RBM Recorders
10. Given plant conditions, ANALYZE those conditions to DETERMINE if an APRM generated Rod Block should have occurred. (012.003 .d.02)
11. Given plant conditions, ANALYZE those conditions to DETERMINE if a RBM channel generated Rod Block should have occurred. (012.003.e.Ol)
12. Given plant conditions, ANALYZE those conditions to DETERMINE if a Two-out-of-Four Logic System generated reactor scram should have occurred. (012.003.d.01)
13. Given a list of plant conditions, IDENTIFY which would result in an APRM INOP trip.

(012.003 .d.04)

14. Given a value of Core Power and the actual APRM reading, STATE if any APRMs need gain adjustment. (400.009.a.02)
15. Given initial unadjusted LPRM readings at BOC, DETERMINE how those readings would change over core life. (400.062.a.02)
16. Given plant conditions, ANALYZE those conditions to DETERMINE the required actions if power oscillations occur with OPRMs INOPERABLE. (012.003.f.04)

n 4 ROD DISPLAY REF APRM %STP RBM ODAs LPRM BARGRAPHS REFERENCE RMCS APRM A LEVEL 124 LPRM SELECT FLUX MATRIX BCC,D SIGNALS  ! LEVELS -, %STP TRIP REFERENCE LEVEL SELECTOR NLIMAC KEYLOCK SWITCH IN IN0P 95%

%STP REF RBM CRITICAL APRM REFAPRM SELF-TEST SEL <LPSP FAULT\

+ SELF TEST h INOPERATIVE TRIP DOWNSCALE TRIP UPSCALE TR1J REFERENCE APRM BYPASS 4 4,

RBM CHANNEL BYPASSED RBM ROD INHIBIT 4r DURING NULL SEQUENCE OR IF INAUTOBYPASS TO RMCS FOR ROD WITHDRAWAL BLOCKS Rod Block Monitor Functional Block Diagram C51-PRNM-LP-01 203 Fig 16 Page 79 of 86

HLT-07 SRO NRC EXAM

14. 215003A1.06 001 Unit 1 is in Startup with JRM lA indicating 50 on Range 4. ALL other IRMs are reading Mid-scale on their respective range.

The NPO momentarily positions TRM lA Range switch to Range 3 and then back to Range 4.

Which ONE of the choices below completes the following statements?

With IRM lA on Range 3, and 603-238, ROD OUT BLOCK, will be in the alarm condition.

Once IRM lA is returned to Range 4, the IRM lA will be EXTTh1GUISHED.

A. annunciators 603-221, IRM UPSCALE, 603-203, IRM BUS A UPSCALE TRIP OR INOP; Drawer UPSC amber light on 1H1 1-P606 B annunciators 603-221, IRM UPSCALE, 603-203, IRM BUS A UPSCALE TRIP OR INOP; UPSC amber light on the benchboard (horizontal portion) of 1H1 1-P603 C. ONLY annunciators 603-221, IRM UPSCALE; UPSC amber light on the benchboard (horizontal portion) of lHl l-P603 D. ONLY annunciators 603-221, 1kM UPSCALE; Drawer UPSC amber light on 1H1 l-P606 40

HLT-07 SRO NRC EXAM

==

Description:==

JAW 34AR-603-221, IRM UPSCALE, the setpoint is 80/125 full scale.

JAW 34AR-603-203, IRM BUS UPSCALE TRIP OR INOP, the setpoint is 115/125 full scale.

fRIvI lA will be reading 50 on Range 4 and will be reading 50 on Range 3. This will be above the Range 3 trip value of 36.8/40 of Full Scale.

Panel Hi i-P603 IRM Indications, including Lights (RESET when the recorder decreases below the setpoint, in this case due to ranging the IRM):

UPSCALE TRIP: Lighted (RED) when respective IRM (115/125 or 36.8/40 of Full Scale) upscale scram trip circuit has tripped.

UPSCALE ALARM: Lighted (AMBER) when respective IRM (80/125 or 25.6/40 of full scale) upscale rod block trip circuit has tripped.

Panel Hi i-P606 Lights and Controls Reset Switch allows resetting the seal-in trip lights on the drawer front. Moving this switch out of the center position causes a reset. This switch is spring-returned to center position.

Lights (IRM Indicators) includes UPSCALE ALARM: Lighted (AMBER) when respective IRM (80/125 or 25 .6/40 of full scale) upscale rod block trip circuit has tripped.

The A distractor is plausible since the first part is correct. The second part is plausible if the applicant thinks that the front panel WM lights will remain illuminated and confuses this with the back panel indication.

The C distractor is plausible if the applicant confuses the setpoints between the upscale alarm and the upscale trip alarm (80/125 vs 115/125). The second part is correct.

The D distractor is plausible if the applicant confuses the setpoints between the upscale alarm and the upscale trip alarm (80/125 vs 115/125). The second part is plausible if the applicant thinks that the front panel JRM lights will remain illuminated and confuses this with the back panel indication.

A. Incorrect See description above.

B. Correct See description above.

C. Incorrect See description above.

D. Incorrect See description above.

References:

41

HLT-07 SRO NRC EXAM NONE K/A:

215003 Intermediate Range Monitor (IRM) System Al. Ability to predict and/or monitor changes in parameters associated with operating the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM controls including:

(CFR: 41.5 / 45.5)

A1.06 Lights and alarms 3.3 3.2 LESSON PLAN/OBJECTIVE:

C5 1 -IRM-LP-0 1202, Intermediate Range Monitors, EQ 012.003 .H.0 1 ReReferences used to develop this question:

34AR-603-22 1, IRM Upscale 34AR-603-203, IRM BUS A Upscale Trip Or Inop ORIGINAL QUESTION (HLT 6 NRC Exam Q#l1)

Unit 1 is in Startup with IRIVI 1A indicating 95 on range 4. ALL other IRMs are reading between 15 and 75 on range 4.

With IRM 1 A indicating 95 on range 4, annunciator(s) will be in the alarm condition.

Once IRM 1A is placed on range 5, the will be illuminated.

A. 603-221, IRM UPSCALE and 603-203, IRM BUS A UPSCALE TRIP OR INOP; IRM 1A Drawer UPSC amber light on IH1 1-P606 B. 603-221, IRM UPSCALE and 603-203, IRM BUS A UPSCALE TRIP OR 1NOP; IRM 1A benchboard UPSC amber light on 1H1 1-P603 C. 603-221, IRM UPSCALE, ONLY; IRM 1A benchboard UPSC amber light on 1H1 l-P603 D.V 603-221, IRM UPSCALE, ONLY; 42

C51-IRM-LP-01202 Page 3 of 69 INTERMEDIATE RANGE MONITORS

10. Given the applicable electrical prints, DETERMINE the power supplies to the following:

(012.003.C.15)

a. IRMs
b. IRM Recorders
11. Given a list of IRM components, a simplified IRM schematic and Plant conditions where an IRM has failed the overlap check between ranges 6 and 7, SELECT the component which has the highest probability of causing this failure. (012.0l0.A.02)
12. From a list, SELECT the statement which best describes the operation of the Range Switches.

(012.010.A.03)

13. From a list, SELECT the statement which describes the location of 1kM system control switches, recorders and indicators in the main control room. (012.003.H.01)
  • 14. Given plant conditions, reactor startup in progress, and 34G0-OPS-001-1/2, Plant Startup, DETERMINE, (012.006.A.01)
a. When 575 V-C5 1-004-1/2, IRM Instrument Functional Test, is required to be performed.
b. How the Operator verifies detector movement.
c. When the IRM detectors should be fully withdrawn.
  • 15. Given plant conditions, reactor startup in progress, and 34G0-OPS-001-1/2, Plant Startup, ANALYZE plant conditions and determine when the reactor is critical. (012.011.A.0l)
  • 16. Given 34G0-OPS-001-l/2, Plant Startup, and 34G0-OPS-013-l/2, Normal Plant Shutdown, DETERMINE when the Operator should range the IRMs to maintain the desired IRIVI scale indication during reactor startup/shutdown. (012 .007.A.0 1)
  • 17. Given 34G0-OPS-00l-l/2, Plant Startup, with a completed Table 1, DETERMINE if correct overlap has been demonstrated between IRM ranges 6 and 7. (012.010.A.0l)
18. Given that a Plant Operator is attempting to withdraw the IRMs, but the IRMs are remaining full in, DISCUSS why he would be directed to depress the DRIVE IN button and then attempt IRM withdrawal. (012.006.A.02)
  • 19. Given plant conditions and 34G0-OPS-013-1/2, Normal Plant Shutdown, DETERMINE, (012.005.B.0l)
a. When 57SV-C51-004-l/2, IRM Instrument Functional Test, is required to be performed.
b. When the Operator should start inserting the IRM detectors.
  • 20. Given plant conditions involving IRMs, DETERMINE if a Technical Specification Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination of APPLICABILITY and associatedNOTES) (300.006.A.41)

1.0 IDENTIFICATION

ALARM PANEL 603-2 IRM BUS A UPSCALE TRIP OR INOP DEVICE: SETPOINT:

2C51-K6O1A, 2C51-K6O1C, Upscale, 115/125 of full scale, 2C51-K6O1E, or 2C51-K6OIG Inoperative switch NOT in OPERATE, power supply voltage low, OR circuit boards NOT in circuit

2.0 CONDITION

3.0 CLASSIFICATION

One OR more of A, C, E or G intermediate range monitors are upscale EQUIPMENT STATUS OR inoperative AND the reactor mode switch is QI in RUN. 4.0 LOCATION:

2H11-P603 Panel 603-2 5.0 OPERATOR ACTIONS:

5.1 At Panel 2H1 1-P603, confirm that one OR more of A, C, E or G IRM Upscale Trip OR Inop lights are ILLUMINATED. LI 5.2 IF the IRM recorder indicates greater than 115/125 of full scale, the IRM is Upscale. LI 5.3 IF the IRM is Upscale, ADJUST the IRM range until the IRM indicates less than 115/125 of full scale AND IF necessary, INSERT in sequence control rods. LI 5.4 Confirm the following:

  • the Channel A Scram Group lights are EXTINGUISHED LI
  • annunciator 603-1 09, REACTOR NEUTRON MONITORING SYS TRIP, is ALARMED LI
  • the white Rod Out light is EXTINGUISHED LI
  • annunciator 603-238, ROD OUT BLOCK, is ALARMED LI 5.5 At Panel 2H1 1-P606, determine IF the IRM Upscale Trip OR Inop light is ILLUMINATED. LI 5.6 IF the IRM is Inop, confirm that the Mode Switch is in the OPERATE position. LI 5.7 IF the IRM is failed, notify the Shift Supervisor and IF possible, BYPASS the IRM. LI 5.8 IF a scram occurs, enter 34AB-C71-001-2, Scram Procedure. LI 5.9 After the cause is corrected, at Panel 2H1 1-P606, rotate reset knob to the right to reset. LI

6.0 CAUSES

6.1 IRM upscale 6.3 IRM INOP IRM NOT ranged correctly IRM Mode switch NOT in OPERATE 6.2 IRM failure IRM high voltage power supply voltage low IRM circuit board NOT in circuit

7.0 REFERENCES

8.0 TECH. SPECS.ITRMIODCMIFHA:

7.1 H-27527 thru H-27536, Start-Up Range Neutron 8.1 TS 3.3.1.1 Item 1 RPS Instr IRMs Mon. Sys. 2C51A Elementary Diagram 8.2 TRM T3.3.2 Item 2 Control Rod 7.2 H-27605 thru H-27619, RPS 2C71 Elem Diag Block Instr IRMs 7.3 H-27499_thru_H-2751 5,_RMCS_2C1 1A_Elem_Diag 34AR-603-203-2 Ver. 3 MGR-0048 Ver. 5.0 AG-MGR-75-1 101

1.0 IDENTIFICATION

ALARM PANEL 603-2 IRM

  • z::::: UPSCALE DEVICE: SETPOINT:

2C51-K601A, 2C51-K6OIB, 2C51-K601C, 2C51-K601D, 80/1 25 of full scale 2C51-K601 E, 2C51-K601 F, 2C51-K601 G or 2C51-K601 H

2.0 CONDITION

3.0 CLASSIFICATION

One or more of the intermediate range monitors are upscale and

. EQUIPMENT STATUS the reactor mode switch is not in RUN. 4 0 LOCATION*

2H1 1-P603 Panel 603-2 5.0 OPERATOR ACTIONS:

5.1 Confirm one or more of the IRM Upscale Alarm lights are ILLUMINATED, 2H1 1-P603. LI 5.2 Confirm the IRM indicates greater than 80/125 of full scale. LI 5.3 ADJUST the IRM range until the IRM indicates less than 80/125 of full scale. LI 5.4 Confirm the following:

  • the white Rod Out light is EXTINGUISHED LI
  • annunciator 603-238, ROD OUT BLOCK is ALARMED LI 5.5 IF reactor power is rapidly increasing, INSERT control rods in sequence to slow the power increase. LI NOTE If the IRM level increases to 115/125 of full scale, a reactor SCRAM will occur.

5.6 IF the IRM is failed, notify the Shift Supervisor and IF possible BYPASS the IRM. LI

6.0 CAUSES

6.1 IRM upscale IRM not ranged correctly 6.2 IRM failure

7.0 REFERENCES

8.0 TECH. SPECS.ITRM/ODCM/FHA:

7.1 H27527 thru H-27536, Start-up Range Neutron Mon. 8.1 TS 3.3.1.1, Item 1, RPS Inst IRMs Sys. 2C51A Elem 8.2 TRM T3.3.2, Item 2, Control Rod 7.2 H-27605 thru H-27619, RPS 2C71 Elem Block Instrumentation IRMs 7.3 H-27499 thru H-27515, RMCS 2C1IA Elem 34AR-603-22 1-2 Ver. 2 MGR-0048 Rev. 5.0 AG-MGR-75-1 101

HLT-07 SRO NRC EXAM

15. 215004K6.02 001 Unit 1 is starting up.

o Reactor Water Level (+) 37 Inches O The reactor is critical o IRMs are all on-scale on Range 3 A loss of 24/48 VDC t lB 1R25-S016, occurs.

Which ONE of the choices below will occur as a result of this power loss?

Source Range Monitor (SRM)___________ indicator will fail A. iC; as is B. iC; downscale C. iD; as is D iD; downscale 44

HLT-07 SRO NRC EXAM

==

Description:==

Source Range Monitors (SRM) Channels A & C are powered from 24/48 VDC Cabinet 2A, 2R25-S015. SRM Channels B & D are powered from 24/48 VDC Cabinet 2B, 2R25-S016. The SRM Recorder is powered from Vital AC, 2R25-S063.

System response to loss of power:

SRM Indicators Fail downscale on loss of 24/48 VDC Source Range Monitor Recorders Fail as is on loss of Vital AC, R25-S063 The A distractor is plausible if the applicant does not remember/confuses the power supply to the SRMs. The second part is plausible if the applicant confuses the response of the SRM recorder to the SRM indicator response to loss of power. Also would be correct if asking about the SRM recorder.

The B distractor is plausible if the applicant does not remember/confuses the power supply to the SRMs. The second part is correct.

The C distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses the response of the SRM recorder to the SRM indicator response to loss of power. Also would be correct if asking about the SRM recorder.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

References:

NONE K/A:

215004 Source Range Monitor (SRM) System K6. Knowledge of the effect that a loss or malfunction of the following will have on the SOURCE RANGE MONITOR (SRM) SYSTEM: (CFR: 41.7 / 45.7)

K6.02 24/48 volt D.C. power 3.1 3.3 45

HLT-07 SRO NRC EXAM LESSON PLAN/OBJECTIVE:

C51-SRM-LP-Ol2Ol, Source Range Monitors, EO 012.003.A.09 References used to develop this question:

34AB-R22-OOl-1, Loss of DC Buses Modified from HLT Database Q# 295004AK2.03-002 ORIGINAL QUESTION (Q#295004AK2.03-002)

Unit 1 is starting up.

o Reactor Water Level (+)37 Inches o The reactor is critical o IRMs are all on-scale on Range 3 A loss of 24/48 VDC 1W occurs.

Which ONE of the following will occur as a result of this power loss?

A. Division U Low-Low Set (LLS) relay logic will de-energize B. lA Intermediate Range Monitor (WM) indication will fail downscale C. Division II Core Spray Loss of Coolant Accident (LOCA) relay logic will de-energize D.V iB Source Range Monitor (SRM) indication will fail downscale 46

C51-SRM-LP-01201 Page 2 of 52 SOURCE RANGE MONITORS Initial License (LT) ENABLING OBJECTIVES Given a simplified schematic diagram of an SRM detector, TRACE the signal flow path from the fission chamber to the trip units/meters. (012.003.A.01)

2. Given a simplified block diagram of the source range monitors, LABEL the following components: (012.003.A.03)
a. Fission chamber
b. Pulse Preamplifier
c. Pulse Height Discriminator
d. Logarithmic Integrator
e. DC Amplifier
f. Period Circuits
g. High Voltage Power Supply
h. All trip circuits
3. Given a list of statements, SELECT the statement representing the purpose of the following SRM system components: (012.003 .A. 04)
a. Fission Chamber
b. Pulse Preamplifier
c. Pulse Height Discriminator
d. Logarithmic Integrator
e. DC Amplifier
f. Period Circuits
g. High Voltage Power Supply
h. All Trip Circuits
i. SRM Drive Mechanism
4. DESCRIBE the basic process by which our SRM fission chambers detect neutrons.

(012.003 .A.07)

5. DESCRIBE the process of gamma compensation utilized in the SRM, to include the circuitry which performs the compensation and the process it utilizes. (012.003.A.08)
6. Given the applicable electrical prints, DETERMINE the power supplies to the following:

(012.003.A.09)

a. Source Range Monitors
b. Source Range Recorders

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 46 OF 73 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF DC BUSES 34AB-R22-001-1 2.9 XIII. LOSS OF 24148V DC CABINET IB, 1R25-S016 1.0 CONDITIONS 1.1 ANNUNCIATORS 1.1.1 24/48V BATT VOLTS LOW 651-137 1.1.2 24/48V BATT CHGR MALFUNCTION, 651-1 38 1.1.3 IRM BUS B UPSCALE TRIP OR INOP, 603-212 1.1.4 RADWASTE EFFL RADIATION HIGH, 601-401 1.1.5 SERVICE WATER EFFLUENT RADIATION HIGH, 601-407 1.1.6 RBCCW RADIATION HIGH, 601-413 1.1.7 POSTTREATMENT O/G RADIATION HI-H 1-HI/INOP, 601-405 1.1.8 POSTTREATMENT OFF GAS RADIATION HIGH, 601-417 1.1.9 POSTTREATMENT OFF GAS RADIATION DNSC, 601-423 1.1.10 LIQUID PROCESS RADIATION DNSC!INOP, 601-419 1.1.11 SBGTAUTO SIGNAL PRESENT, 657-019 1.1.12 SBGT AUTO SIGNAL PRESENT, 657-019 (Unit 2) 1.1.13 CR OUTSIDE AIR INLET RADIATION DOWNSCALE TRIP, 654-048 1.1.14 OFF GAS VENT RADIATION DNSCIINOP, 601-424 1.1.15 OFF GAS VENT RADIATION HIGH-HIGH, 601-412 1.1.16 OFF GAS VENT RADIATION HIGH, 601-418 1.1.17 The following annunciators will alarm on the loss of 1 R25-S016 only if the Reactor Mode Switch is NOT in RUN:

1.1.17.1 REACTOR NEUTRON MONITORING SYS TRIP, 603-109 1.1.17.2 REACTOR AUTO SCRAM SYSTEM B TRIP, 603-118 1.1.17.3 ROD OUT BLOCK, 603-238 (in STARTUP or REFUEL only) 1.2 Loss of power to equipment listed on Attachment 15 MGR-0001 Ver. 3

SNC PLANT E. I. HATCH I Pg 71 of 73 DOCUMENT TITLE: DOCUMENT NUMBER: Ver No:

LOSS OF DC BUSES 34AB-R22-OO1-1 2.9 ATTACHMENT 15 Att. Pg.

TITLE: 24/48V DC CABINET IB, 1R25S016 I of 1 BREAKER FUNCTION

1. Battery Chargers 1C, lCD, ID
2. Spare
3. Process Rad Monitor Board, 1HI1-P604 (1D11-K605, K606, K615B, K604, K600B)
4. Spare
5. Spare
6. Start-Up Nuetron Monitoring System SRMs & IRMs, 1H11-P606 (SRMs B, D and IRMs B, D, F, H)
7. Start-up Nuet Trip Aux Units, IC5IA-Z2B &Z2D, IH11-P606 (Radiation Monitor relays for secondary containment isolation and SBGT start)
8. Process Rad Record Board, 1H1I-P645 (1Z41-trip auxiliary units for MCRECS Pressurization Mode initiation from MCR vent intake radiation monitors)
9. - 30. Spare MGR-0009 Ver. 4

HLT-07 SRO NRC EXAM

16. 215005A1.01 001 Unit 2 is at 9% RTP when the following occurs:

o LPRM 44-21A fails UPSCALE (Inputs into APRM 2D) o Subsequently, Operations bypasses the failed LPRM The following represents the status of APRM 2D after LPRM 44-21A is bypassed:

o Number of Level D LPRM Inputs ... 6 o Number of Level C LPRM Inputs ... 4 o Number of Level B LPRM Inputs ... 5 o Number of Level A LPRM Inputs ... 2 Which ONE of the choices below completes the following statements?

When LPRM 44-21A failed upscale, APRM 2D power indication After LPRM 4421A is bypassed, APRM 2D is inoperable because there are too few A. remained the same; total LPRM inputs B. remained the same; LPRM inputs per level C. increased; total LPRM inputs D increased; LPRM inputs per level 47

HLT-07 SRO NRC EXAM

==

Description:==

Each APRM channel receives only the subset of LPRM data and Total Recirculation Flow data associated with that APRM channel. The LPRM detector assignments have been made so that the mathematical average of the individual LPRM detector signals is reasonably representative of the average flux in the reactor. Each APRM is assigned one of four LPRM detectors in each LPRM string (31 total) and the quantity of LPRM detectors for each level in the reactor is relatively equal for all levels for each APRM.

The failed upscale LPRM will result in reactor power indication to increase. If the LPRM had failed downscale or indicated a lower value than previously, the APRM indication would be lower.

APRM operability requirements require at least 17 operable LPRM inputs AND at least 3 operable LPRMs per axial level. IF LPRM operability falls below these requirements, a rod block will be generated from the associated APRM. With the D axial level LPRMs for APRM 2D NOT sufficient, the APRM will be considered mop.

The A distractor is plausible if the applicant confuses how the RBM processes LPRM inputs with how the APRM processes LPRM inputs. In this case, the RBM would not use the A level LPRM inputs in its normalization process, therefore the APRM 2D power indication would remain the same. The second part is plausible if the applicant confuses the LPRM requirements and thinks the rod block will occur when the total drops to 17 and does not think about having sufficient inputs per level.

The B distractor is plausible if the applicant confuses how the RBM processes LPRM inputs with how the APRM processes LPRM inputs. In this case, the RBM would not use the A level LPRM inputs in its normalization process, therefore the APRM T 2D power indication would remain the same. The second part is correct.

The C distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses the LPRM requirements and thinks the rod block will occur when the total drops to 17 and does not think about having sufficient inputs per level.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

References:

NONE 48

HLT-07 SRO NRC EXAM K/A:

215005 Average Power Range Monitor/Local Power Range Monitor System Al. Ability to predict and/or monitor changes in parameters associated with operating the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM controls including: (CFR: 41.5 /45.5)

A1.01 Reactor power indication 4.0 4.0 LESSON PLAN/OBJECTIVE:

C51-PRNM-LP-01 203, Power Range Neutron Monitoring System, EU 012.003.D.09 &

EU 012.003.D.04 References used to develop this question:

34SV-C51-003-2, LPRM Operational Status Modified from HLT-5 NRC Exam Q#17 ORIGINAL QUESTION (HLT-5 NRC Exam Q#17)

Unit 2 is starting up with the Reactor Mode Switch in the START & HOT STBY position when the following occurs:

o I&C has determined a GE STh (Service Information Letter) causes several APRM D LPRMs to be inoperable o Operations bypasses the mop LPRMS The following is the present status of APRM D:

o Indicated Power Level 9%

o Number of Level D LPRM Inputs 5 ..

o Number of Level C LPRM Inputs 4 ...

o Number of Level B LPRM Inputs 4 ...

o Number of Level A LPRM Inputs 3 ...

Which ONE of the following identifies the ROD OUT BLOCK (603-23 8) alarm status based on the current status of APRM D?

A. APRM D is causing a ROD OUT BLOCK alarm because there are 49

HLT-07 SRO NRC EXAM B.V APRM !DI is causing a ROD OUT BLOCK alarm because there are too few total LPRM inputs.

C. APRM D is causing a ROD OUT BLOCK alarm because power is too high.

D. APRM t D is NOT causing a ROD OUT BLOCK alarm.

50

C5 1 -PRNIvI-LP-0 1203 Page 3 of 86 POWER RANGE NEUTRON MONITORING SYSTEM

5. STATE the number of LPRMs assigned to a specific APRM. (012.003.d.09)
6. Given Plant conditions, ANALYZE these conditions to DETERMINE if a Recirculation Flow unit generated Rod Block should have occurred. (012.003.f.0l)
7. STATE the LPRM detector assignments to the RBM channels to include number and level for a control rod selected near the center of the core. (012.003.e.05)
8. DESCRIBE the RBM nulling sequence to include: Initiation signal to null, recorder readings during and after the null sequence, and the function of the APRM reference signal.

(0 12.003.e.06)

9. DETERMINE the power supplies to the following: (012.003.d. 12)
a. APRM Drawer
b. RBM Drawer
c. APRMIRBM Recorders
10. Given plant conditions, ANALYZE those conditions to DETERMINE if an APRM generated Rod Block should have occurred. (012.003.d.02)
11. Given plant conditions, ANALYZE those conditions to DETERMINE if a RBM channel generated Rod Block should have occurred. (0l2.003.e.01)
12. Given plant conditions, ANALYZE those conditions to DETERMINE if a Two-out-of-Four Logic System generated reactor scram should have occurred. (012.003.d.01)
13. Given a list of plant conditions IDENTIFY which would result in an APRM INOP trip.

(012.003.d.04)

14. Given a value of Core Power and the actual APRM reading, STATE if any APRMs need gain adjustment. (400.009.a.02)
15. Given initial unadjusted LPRM readings at BOC, DETERMINE how those readings would change over core life. (400.062.a.02)
16. Given plant conditions, ANALYZE those conditions to DETERMINE the required actions if power oscillations occur with OPRMs INOPERABLE. (012.003.f.04)

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH . 30F25 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LPRM OPERATIONAL STATUS 34SV-C51-003-2 4.8 4.3.6 LPRMs may be bypassed while the APRM instrument is in a operable status.

This is performed from the BYPASS SELECTIONS display which is accessible by password control in the OPER-SET mode.

LPRM status can also be displayed on the APRM ODAs on 2H1 1-P603.

5.0 PRECAUTIONS/LIMITATIONS 5.1 PRECAUTIONS Observe the safety rules outlined in the Southern Nuclear Safety and Health Manual.

52 LIMITATIONS 5.2.1 APRM operability requirements require:

at least 17 operable LPRM inputs AND at least 3 operable LPRMs per axial level.

IE LPRM operability falls below these requirements, a rod block will be generated from the associated APRM.

5.2.2 OPRM operability requirements are:

at least 14 operable cells with at least one operable LPRM per cell, AND at least 17 total operable LPRMs.

(See Attachment 2 for OPRM cell assignments).

5.2.3 A RBM requires> 50 % of the LPRM inputs to be operable.

The RBM channel uses:

8 LPRM inputs for a 4 LPRM string rod, 6 LPRM inputs for a 3 LPRM string rod, and 4 LPRM inputs for a 2 LPRM string rod.

(the inputs can be determined by selecting a control rod and looking at the ODA for the RBM).

6.0 PREREQUISITES N/A Not applicable to this procedure MGR-0001 Ver. 3

HLT-07 SRO NRC EXAM

17. 215005A4.03 001 Unit 1 is operating at 100% RTP when 603-210-1, APRM/OPRM TRIP, alarm comes in and then immediately clears due to a momentary failure of one (1) APRM.

Which ONE of the choices below completes the following statements?

With the above condition, a RPS Half Scram signal have been received.

If the APRM is bypassed, the TOTAL number of 2/4 Logic modules (Voters) that will have the associated APRMs bypass light ILLUMINATED is A. would; two (2)

B. would; four (4)

C. would NOT; two (2)

D would NOT; four (4) 51

HLT-07 SRO NRC EXAM

==

Description:==

The safety section of each 2/4 Logic Module receives the safety trip status from EVERY APRM instrument and the channel bypass status from the other 2/4 Logic Modules. Each 2/4 Logic Module logically determines whether its safety outputs should be in a trip state. Each of the 2 Out Of 4 Logic Modules in lHl 1 P608 should indicate a trip input from the affected APRMIOPRM instrument. One APRM in a tripped condition will not cause a half scram condition but one 2/4 Logic Modules in a tripped condtion will cause a half scram condition.

The 2/4 Logic Module safety section monitors the state of manual bypass for each APRMIOPRM channel and incorporates it into its trip signal processing. The APRM instruments generate trip signals regardless of the associated APRMJOPRM channels bypass states. Each 2/4 Logic Module generates a dynamic digital signal which is transmitted to one of the four sections of the APRM/OPRM Manual Bypass Switch. A channel is bypassed if the digital signal is passed through the switch and received by the 2/4 Logic Module. The state of the manual bypass channel monitored is transmitted to the other three 2/4 Logic Modules (all four 2/4 Voters receive the bypass signal). This will be indicated by the 2/4 Logic Modules blue bypass light for the associated APRM being illuminated.

The A distractor is plausible if the applicant confuses that one 2/4 Logic Module in a tripped condtion will cause a half scram condition with how one APRM in a tripped condition functions (no half scram). The second part is plausible if the applicant confuses how the bypass switch works and thinks that since the title of the 2/4 Logic Module has a 2 in it that only (2) 2/4 Logic Modules receive the signal that the APRM is bypassed.

The B distractor is plausible if the applicant confuses that one 2/4 Logic Module in a tripped condtion will cause a half scram condition with how one APRM in a tripped condition functions (no half scram). The second part is correct.

The C distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses how the bypass switch works and thinks that since the title of the 2/4 Logic Module has a 2 in it that only (2) 2/4 Logic Modules receive the signal that the APRM is bypassed.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

52

HLT-07 SRO NRC EXAM

References:

NONE K/A:

215005 Average Power Range Monitor/Local Power Range Monitor System A4. Ability to manually operate and/or monitor in the control room:

(CFR: 41.7 / 45.5 to 45.8)

A4.03 APRM back panel switches, meters and indicating lights. . . 3.2 3.3 LESSON PLAN/OBJECTIVE:

C51-PRNM-LP-01203, Power Range Neutron Monitoring System, EO 012.003.D.08 &

EO 012.003.D.O1 References used to develop this question:

34AR-603-210-1, APRM/OPRM TRIP 34AR-603- 117-1, Reactor Auto Scram System A Trip 53

C51-PRNIvI-LP-01203 Page 2 of 86 POWER RANGE NEUTRON MONITORING SYSTEM Initial License LD ENABLING OBJECTIVES STATE the purpose of the: (012.003.d.05)

a. LPRM System
b. APRM System
c. OPRM System
d. Recirculation flow Units
e. RBM
f. Two-outof-Four Logic System
2. Given simplified diagrams of the following neutron instrumentation systems, TRACE the signal flow path through the system (detector to trip unit). (012.003 .d.06)
a. LPRM System
b. APRM System
c. RBM System
3. Given a list of statements, SELECT the statement representing the function of the following components. (0l2.003.d.08)
a. LPRM
1) Fission Chamber Detector
2) Gain Adjustment
3) LPRM Status
b. APRM
1) Averaging Circuit
2) Count Circuit
3) APRM Bypass Switch
c. RBM
1) Averaging Card
2) RBM Bypass Switch
4. STATE the functional interface between the following systems and other plant systems:

(0l2.003.d.16)

a. LPRMs
b. APRMs
c. OPRM5
d. RBM
e. Recirculation flow units
f. Twoout-of-Four Logic Module

C5 1 -PRNM-LP-0 1203 Page 3 of 86 POWER RANGE NEUTRON MONITORING SYSTEM

5. STATE the number of LPRMs assigned to a specific APRM. (012.003.d.09)
6. Given Plant conditions, ANALYZE these conditions to DETERMINE if a Recirculation Flow unit generated Rod Block should have occurred. (012.003.f.01)
7. STATE the LPRM detector assignments to the RBM channels to include number and level for a control rod selected near the center of the core. (012.003.e.05)
8. DESCRIBE the RBM nulling sequence to include: Initiation signal to null, recorder readings during and after the null sequence, and the function of the APRM reference signal.

(012.003 .e.06)

9. DETERMINE the power supplies to the following: (012.003 .d. 12)
a. APRM Drawer
b. RBM Drawer
c. APRM!RBM Recorders
10. Given plant conditions, ANALYZE those conditions to DETERMINE if an APRM generated Rod Block should have occurred. (012.003.d.02)
11. Given plant conditions, ANALYZE those conditions to DETERMINE if a RBM channel generated Rod Block should have occurred. (012.003.e.01) 12 Given plant conditions, ANALYZE those conditions to DETERMINE if a Two-out-of-Four Logic System generated reactor scram should have occurred (012 003 d 01)
13. Given a list of plant conditions, IDENTIFY which would result in an APRM INOP trip.

(012.003.d.04)

14. Given a value of Core Power and the actual APRM reading, STATE if any APRMs need gain adjustment. (400.009.a.02)
15. Given initial unadjusted LPRM readings at BOC, DETERMINE how those readings would change over core life. (400.062.a.02)
16. Given plant conditions, ANALYZE those conditions to DETERMINE the required actions if power oscillations occur with OPRMs INOPERABLE. (012.003.f.04)

1.0 IDENTIFICATION

ALARM PANEL 603-2 DEVICE:

i:

APRM/OPRM Instrument 1C51-K615A(1C51-K615B)

(1C51-K615C)( 1CS1-K615D)

SETPOINT:

APRMIOPRM TRIP

1) Neutron Flux High Trip (117% in RUN, 13% not in run)
2) STP High Trip (0.57W + 53% 0.57 LIW) clamped at 112.5%
3) Inop Trip (instrument mode switch not in operate, critical self test via 2 Out Of 4 Logic Module fault detected, loss of power) 1C51-K617A(1C51-K617B) 4) OPRM Trip See Setpoint Index, COLR, or ODAs OPRM Trip is (1C51-K617C)( 1C51-K617D) enabled only when reactor power is above 25% 1L recirculation flow is below 60%.

2.0 CONDITION

I 3.0 CLASSIFICATION:

One or more of the APRM/OPRM Monitors have an upscale trip, OPRM EQUIPMENT STATUS trip, or are inoperative. 4.0 LOCATION:

1 H 11 -P603 Panel 603-2 5.0 OPERATOR ACTIONS:

NOTES:

  • IF power is lost to APRM D, Recirc flow indications 1B31-R617 & R613, will be lost.
  • IF power is lost to APRM A, Recirc flow input to recorder 1B31-R614, will be lost.

5.1 IF the OPRM system is mop, AND the alarm is due to an OPRM trip OR periodic APRM oscillations, enter 34AB-C51-001-1. El 5.2 Confirm on the APRM ODAs on 1HII-P603 AND/OR the APRM Numacs on 1HI1-P608 that one OR more of the APRM/OPRM channels indicates a trip or mop condition. Each of the 2 Out Of 4 Logic Modules in 1HI1-P608 should indicate a trip input from the affected APRM/OPRM instrument.

Also, an OPRM trip condition will display the message Instability Detected.

5.3 IF more than one APRM/OPRM instrument indicates an APRM tripped or mop condition, OR an OPRM tripped condition, confirm that a full reactor scram has occurred AND enter 34AB-C71-001-1, Scram Procedure.

5.4 IF the annunciator is due to an STP upscale trip OR mop trip, confirm the following at IHII-P603:

  • the white Rod Out light is EXTINGUISHED El
  • annunciator 603-238 ROD OUT BLOCK is ALARMED El 5.5 Confirm the power and flow are within the analyzed region of operation defined on the power versus core flow map per 34GO-OPS-005-1. El 5.5.1 IF operating outside of the region, notify the Shift Supervisor and STA initiate corrective action within 15 minutes.

5.5.2 IF operating in the Region of Potential Instabilities, take action to exit per the STAs direction.

5.6 IF the APRM!OPRM instrument does not have a trip OR is failed, notify the Shift Supeniisor and STA AND H sufficient APRM/OPRM instruments are operable, BYPASS the APRM/OPRM.

6.0 CAUSES

6.1 APRM Upscale Trip (fixed neutron flux or flow biased thermal power) 6.3 APRM INOP 6.2 OPRM Trip (any of three algorithms)

7.0 REFERENCES

I 8.0 TECH. SPECS.ITRMIODCMIFHA:

7.1 H-44697thru H-44717, PRNM System 1C51B Elem.Diag.

7.2 S-53778, PRNM System Requirement Specification 8.1 TS 3.3.1.1-1, Item 2, RPS InstAPRM 7.3 H-17783 thru H-17798, RPS 1C71 Elementary Diagram I 8.2 TRM 3.3.2, Item 3, Control Rod Block 7.4 H-17853 thru H-17839, RMCS 1 Cl 1A Elementary Diagram Instrumentation APRM 34AR-603-21 0-1 VER. 4.4 MGR-0048 Ver. 5.0 AG-MGR-75-1 101

1.0 IDENTIFICATION

ALARM ANEL 603-1 REACTOR

AUTO SCRAM SYSTEM A TRIP DEVICE: SETPOINT:

1C71-K14A, 1 C71-K14C, I C71-K14E, 1 C71-K1 4G Relay De-Energized

2.0 CONDITION

3.0 CLASSIFICATION

EQUIPMENT STATUS RPS Channel A has tripped causing a half-scram. 4.0 LOCATION:

1H11-P603 Panel 603-1 5.0 OPERATOR ACTIONS:

5.1 Confirm the Scram Group A solenoid lights for trip system A, are EXTINGUISHED, 1H11-P603. LI 5.2 IF a half scram has occurred:

5.2.1 Determine the cause of the half-scram. LI 5.2.2 Attempt to correct OR bypass the cause of the trip.

5.2.3 IF successful, RESET the RPS Channel A. LI IF surveillance testing is in progress which initiates half scram signals, it is acceptable NOTE to delay the performance of the OD-7 Option 2 UNTIL the applicable surveillance is complete. The following step is NOT applicable IF all control rods are inserted.

5.2.4 Using the Process Computer, run an OD-7 Option 2 AND confirm control rod movement has NOT occurred.

IF control rod movement has occurred, enter 34AB-C1 1-004-1, Mispositioned Control Rods.

5.3 IF a full scram occurs, enter 34AB-C71 -001-1, Scram Procedure.

6.0 CAUSES

6.1 Auto scram relays are de-energized by:

6.1.1 Turbine Stop Valve closure 6.1.7 Reactor low water level 6.1.2 Turbine Control Valve fast closure 6.1.8 Neutron Monitoring System 6.1.3 Scram Discharge Volume high level (PRNM 2 out of 4 Logic Modules will still 6.1.4 MSIV closure require two APRM instrument trip inputs 6.t5 Containment high pressure unless placed into the 1 out of 4 logic mode.)

6.1.6 Reactor high pressure

7.0 REFERENCES

8.0 TECH. SPECS ITRM!ODCMIFHA:

H-17783 thru H-17798, RPS Elementary TS 3.3.11 RPS 34AR-603-1 17-1 VER. 6 MGR-0048 VER 5.0 AG-MGR-75-1101

HLT-07 SRO NRC EXAM

18. 216000A4.01 001 A transient has occurred on Unit 2 resulting in Reactor Water Level (RWL) decreasing.

A RWL Milestone is established at -175 inches.

Subsequently, an Emergency Depress is performed.

A Normal RWL band of (+) 32 to (+) 42 inches is established.

Which ONE of the choices below completes the following statement for determining ACCURATE RWL?

To determine when the -175 inch Milestone is reached, the NPO will be monitoring RWL at panel 21111- to determine when RWL is in its Normal band.

A P601 and then transition to the 2H1 1-P603 panel B. P601 and continue monitoring at the 2111 1-P601 panel C. P602 and then transition to the 2111 1-P603 panel D. P602 and continue monitoring at the 2111 1-P602 panel 54

HLT-07 SRO NRC EXAM

==

Description:==

Reference legs D004 A/B are used for Normal (Narrow) Range (0 to 60) and Post Accident Flooding (Fuel Zone) Range (-317 to -17) instrumentation. The Narrow Range uses tap Nil A/B for variable leg pressure. The Fuel Zone Range uses taps off the Jet Pump (#5 and 15) for variable leg pressure. Reference legs D004 A/B feed level transmitters that are used for Control Room/local indications, trip functions and input to Feedwater Level Control system. The Control Room indications off these reference legs are the level indicators R606A, B and C and level recorder R608 on panel H11-P603. On panel H11-P601, level indication/recorder R623A and R623B Fuel Zone. The Fuel Zone recorders have the ability to display a Fuel Zone water level indication that is AUTOMATICALLY corrected for Reactor Pressure and Drywell temperature. The R623A/B recorders will also provide Reactor Pressure and Fuel Zone uncompensated level indications.

Reference leg D002 is used for the Floodup Range (0 to 400). There are two level transmitters, B2l-N027 and C32-N010 off of this reference leg. Both are used for indication in the Control Room on panel H11-P602. Both use a Dixson Digital Bargraph for indicators.

Level transmitter B21-N027 feeds level indication R605 and has a 0 to 400 scale. Level transmitter C32-N0i0 feeds level indicator R655 and has a 0 to 200 scale.

The B distractor is plausible since the first part is correct. The second part is plausible since the wide range instrument R623A & R23B are on this panel and if the applicant does not remember that after a rapid depress is performed these instruments cannot be used for valid RWL indication.

The C distractor is plausible if the applicant confuses the RWL indicators/recorders and thinks RWL at -200 inches can be monitored on this panel. The second part is correct.

The D distractor is plausible if the applicant confuses the RWL indicators/recorders and thinks RWL at -200 inches can be monitored on this panel. The second part is plausible since after an Emergency Depress is performed the RWL instruments on this panel will indicate valid RWL.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

55

HLT-07 SRO NRC EXAM

References:

NONE K/A:

216000 Nuclear Boiler Instrumentation A4. Ability to manually operate and/or monitor in the control room:

(CFR: 41.7 /45.5 to 45.8)

A4.O1 Recorders 3.3 3.1 LESSON PLAN/OBJECTIVE:

B1 1-RXINS-LP-04404, Reactor Vessel Instrumentation, EO 200.002.A.07 EOP-CAU-LP-20305, EO 201.065.A.09 References used to develop this question:

34AB-B21-002-2, RPV Water Level Corrections 56

Bfl-RXINS-LP-04404 Page 2 of 92 REACTOR VESSEL INSTRUMENTATION Initial License (LT) ENABLING OBJECTIVES

1. DESCRIBE the following plant systems response to reactor water level column inaccuracies:

(200.002.A.08)

a. Reactor Protection System
b. Primary Containment Isolation System
c. Secondary Containment Isolation System
d. Emergency Core Cooling System.
2. NAME the four ranges of reactor vessel water level indication and IDENTIFY which of the five reference legs they tap off from. (200.002.A.09)
3. Given a list of reactor vessel water level indicators, CHOOSE the indicators that are most likely to experience reference leg flashing. (200.002.A.1O)
4. DESCRIBE the number, location, and range of the reactor vessel water level and pressure indicators found in the: (200.002.A.07)
a. Control Room
b. Reactor Building
c. Remote Shutdown Panel(s)
5. Given a Loss of Vital AC, IDENTIFY reactor water level instruments that are affected.

(200.020.A.02)

6. From a list, SELECT the power supply for Reactor Vessel Level and Pressure Instruments.

(200.020.A.06)

7. From a list of level setpoints, DETERMINE what automatic actions/alarms would occur.

(200.002.A.1 1)

8. From a list of pressure setpoints, DETERMINE what automatic actions/alarms would occur.

(200.002.A. 12)

9. DESCRIBE where Jet pump flow indicators and flow transmitters are found. (400.048.A.03)
10. Given a list of statements, IDENTIFY where and DESCRIBE how core flow is measured.

(400.048.A.05)

11. Given a list of statements, IDENTIFY the statement which describes how core dip and core plate d/p are measured. (400.048.A.07)

EOP-CAU-LP-20305 Page 2 of 25 EOP CAUTIONS Initial License (LT) ENABLING OBJECTIVES

1. Given a list of RWL instrument responses, IDENTIFY the response that would indicate boiling in an instrument leg. (201 .065.A.01, 201 .068.A.01, 201 .090.A.01)
2. Given Caution One and an RWL instrument, DETERIVITNE which Drywell RTDs monitor local temperatures near the instrument. (201 .065.A.02, 201 .068.A.02, 201 .090.A.02, 201 .099.A.0 1)
3. Defme Maximum Run Temperature (MRT). (201.065.A.04, 201.068.A.04, 201.090.A.04)
4. Define Minimum Indicated Level (MIL). (201.065.A.03, 201.068.A.03, 201.090.A.03)
5. Given an RWL indicator reading below MIL for a given MRT, IDENTIFY from a list the possible condition that could exist which prevents this instrument from being used.

(201.065.A.05, 201.068.A.05, 201.090.A.05)

6. Given Caution One and plant conditions, DETERMINE if a given RWL indicator is indicating above its MIL for the existing MRT. (201.065.A.06, 201.068.A.06, 201.090.A.06, 201.099.A.02)
7. Given Caution One and plant conditions, DETERMINE which RWL indicators may be used.

(201 .065.A.07, 201 .068.A.07, 201 .090.A.07, 201 .099.A.03)

8. Given the EOPs, 34AB-B21-002, and plant conditions, CALCULATE the corrected RWL.

(201 .099.A.04)

9. Given Caution Two, DETERMINE which RWL instruments cannot be used during rapid depressurization of the RPV below 500 psig. (201.065.A.09, 201.068.A.09)
10. Given a list, IDENTIFY the statement that describes the potential effect that rapidly depressurizing the RPV below 500 psig has on heated reference leg water level instruments.

(201.065.A.08, 201.068.A.08)

11. Given Caution Three, IDENTIFY the effect of operating pumps in the unsafe region of the NPSH or Vortex Limit graphs. (201.065.A.32, 201.068.A.38)
12. Given a list, IDENTIFY the effect that high Torus pressure might have on RCIC operation as stated in Caution Four. (201.065.A.13, 20L066.A.10, 201.068.A.20, 201.069.A.08, 201.090.A.17)
13. Given a list, IDENTIFY the effect that high Torus water temperature has on HPCI and RCIC operation as stated in Caution five. (201.090.A.29)
14. Given a list, IDENTIFY the statement that describes the plant response to rapid injection of water into the RPV during an ATWS. (201.068.A.22, 201.088.A.15, 201.090.A.19, 20l.091.A.08)
15. Given a list, IDENTIFY the effect that lowering primary containment pressure has on available NPSH as stated in Caution seven. (201.065.A.33, 201.088.A.25, 201.090.A.30)

Objectives marked by a RED (*) are required during RO-305 and SR-305 of the Initial License program.

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 2 OF 13 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

RPV WATER LEVEL CORRECTIONS 34AB-B21-002-2 6.13 2.0 AUTOMATIC ACTIONS NONE 3.0 IMMEDIATE OPERATOR ACTIONS NONE 4.0 SUBSEQUENT OPERATOR ACTIONS

. Recommended RWL instruments to use based on RWL range:

IF Reactor Water Level is: THEN recommended RWL instruments are:

> 60 inches SPDS, R605,R655 0 to 60 inches SPDS, R6O6NBIC, R6O4NB, R623A1B WR NOTES:

0 to -100 inches SPDS, R623A1B WR, R6O4AIB

< -100 inches R623A1B FZ

. A valid SPDS primary display level indication or the 2B21-R623A1B Fuel Zone compensated level indication may be used as the corrected water level indication in lieu of performing these manual corrections.

4.1 Observe Emergency Operating Procedure Cautions I jQ 2 contained in Attachment 1. Any instrument JI meeting the requirements of Cautions 1 AND 2 may ] be used for RPV water level indication.

4.2 Correct the range of instrumentation being used as follows:

NOTE: RTD Groups AND indication locations are in Attachment 1.

4.2.1 FUEL ZONE (-17 to -317 inches)

INSTRUMENT RTD GROUP

. 2B21-R623B, Fuel Zone Uncompensated 2 4.2.1.1 Determine corrected Fuel Zone Uncompensated Level per Attachment 3.

MGR-0001 Rev4

SNC PLANT E. I. HATCH I Pg 7 of 13 DOCUMENT TITLE: I DOCUMENT NUMBER: I Ver No:

RPV WATER LEVEL CORRECTIONS I 34AB-B21-002-2 I 6.13 ATTACHMENTJ Aff.Pg.

TITLE: DRYWELL RTD GROUPS AND CAUTION 1 AND 2 3 of 3 CAUTION I (CONTINUED)

2. FOR THE FOLLOWING TABLE, THE WATER LEVEL INSTRUMENT READS ABOVE MINIMUM INDICATED LEVEL FOR THE ASSOCIATED MAXIMUM RUN TEMPERATURE MEASURED BY THE HIGHEST TEMPERATURE IN THE ASSOCIATED RTD GROUP.
a. NARROW RANGE MINIMUM (0 TO +60 IN.) INDICATED MAXIMUM RUN LEVEL (IN.) TEMPERATURE (CF)

INSTRUMENT RTD GROUP 0 UPTO273 2C32-R606A & C 1 6 274 TO 350 2C32-R606B 2 9 351 TO 399 27 400 OR ABOVE

b. WIDE RANGE MINIMUM

(-150 TO +60 IN) INDICATED MAXIMUM RUN LEVEL (IN.) TEMPERATURE (CF)

INSTRUMENT RTD GROUP -150 UPTO197 2B21-R604A & R623A 1 -130 198TO350 2B21-R604B & R623B 2 - 122.5 351 TO 399

- 90 400 OR ABOVE

c. FLOODUP RANGE MINIMUM INDICATED MAXIMUM RUN INSTRUMENT RTD GROUP LEVEL (IN.) TEMPERATURE (CF) 2B21-R605 1 0 UPTO19O (0 TO +400 IN.) 16 191TO250 2C32-R655 2 46 251 TO 350 (0 TO +200 IN.) 60 351 TO 399 102 400 OR ABOVE
d. FUELZONE

(-317 TO -17 IN.) MINIMUM INDICATED MAXIMUM RUN INSTRUMENT RTD GROUP LEVEL (IN.) TEMPERATURE (CF) 2B21-R623B 2 -317 UPTO28O 2B21-R623A 1 - 299 281 AND ABOVE Ii L

2B21 -LI-R604A (2B21 -LI-R604B) AND 2B21 -LR-R623A (2B2 1 -LR-R623B)

(WIDE RANGE SIGNALS)

CAUTION: CANNOT BE USED TO DETERMINE RPV WATER LEVEL DURING RAPID RPV DEPRESSURIZATION BELOW 500 PSIG.

I MGR-0009 Rev 5

HLT-07 SRO NRC EXAM

19. 2170001(2.04 001 Which ONE of the choices below is the power supply to RCIC Barometric Condenser Vacuum Pump, 2E5 l-C002-2?

A. 1251250V DC SWGR 2C, 2R22-S018 B. 125/250V DC SWGR 2D, 2R22-S019 C 250V DC MCC 2A, Reactor Building Feeder, 2R24-S021 D. 125V DC DIST, CAB. 2D, ESS DIV I, 2R25-S 129

Description:

RCIC Barometric Condenser Vacuum pump (E51-C002-2) removes non-condensable from the Barometric Condenser and discharges them to the torus. It will automatically start on RCIC system initiation. The Vacuum Pump maintains 10 Hg vacuum on the condenser and removes non-condensables. The pump is operated by a 3 hp DC motor which receives power from 250 VDC Panel R24-S021. The RCIC Barometric Condenser Vacuum Pump is located on the Barometric Condenser.

The A distractor is plausible if the applicant confuses 125/250V DC SWGR 2C, 2R22-S018, with the correct power supply and is also plausible since it is a DC bus providing power to other pumps.

The B distractor is plausible if the applicant confuses 125/250V DC SWGR 2D, 2R22-S0l9, with the correct power supply and is also plausible since it is a DC bus providing power to other pumps.

The D distractor is plausible if the applicant confuses I 25V DC DIST. CAB. 2D, ESS DIV I, 2R25-S 129, with the correct power supply and is also plausible since it is a DC bus providing power to other components.

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

57

HLT-07 SRO NRC EXAM

References:

NONE KJA:

217000 Reactor Core Isolation Cooling System (RCIC)

K2. Knowledge of electrical power supplies to the following: (CFR: 41.7)

K2.04 Gland seal compressor (vacuum pump) 2.6* 2.6*

LESSON PLAN/OBJECTIVE:

E5 1-RCIC-LP-03901, Reactor Core Isolation Cooling (RCIC), EO lOO.036.C.16 References used to develop this question:

34AB-R22-OO12, Loss Of DC Buses 58

E51-RCIC-LP-03901-06 Page 5 of 95 REACTOR CORE ISOLATION COOLING (RCIC)

22. In accordance with procedure 34SV-E51-002-l/2, RCIC Pump Operability, IIJENTIFY the SO actions performed during the surveillance including local actions performed, test data taken or monitored. (1 00.036.C. 14)
23. Given the cuent status of RCIC, EVALUATE how RCIC operation would be affected when a RCIC system valve is repositioned. (100.036.C.15) (RCIC operation includes injection mode, test mode, and pressure control mode.)
24. State the systems that interface with RCIC. (100.03 6.C.16)
25. For each system that interfaces with RCIC, describe the purpose of the interface. (l00.036.C.17)
26. Given a drawing, label the following RCIC components. (100.036.C.18)
a. Steam Supply Drain Pot
b. Steam Exhaust Drain Pot
c. Exhaust Rupture Disc
d. Barometric Condenser
e. Exhaust Line Vacuum Breakers
f. Turbine
27. Given a RCIC P&ID, and a change in the standby status of a valve in the steam path or water flow path, RECOGNIZE the affect on the RCIC operations. (100.036.C.19)
28. Describe the effects the RCIC system will have on the Torus when the RCIC turbine is operating.

(100.036.C.20) 29 Identify two reasons why the RCIC system must be kept filled during standby configuration.

(100.036.C.21)

30. Describe the potential problem that exist when the CST suction valves and the Suppression Pool suction valves are left open at the same time. (100.036.C.22)
31. Identify the steps to perform a RCIC system normal fill and vent. (l00.036.C.23)
32. Select the method used to determine that the RCIC Diagonal Pump Room Coolers are operating properly during a pump operability test. (100.03 6.C.24)
33. Given a list of valves, SELECT which RCIC valves change position with an Auto Initiation Signal present and the manual isolation pushbutton depressed. (039.012.A.04)
  • 35 STATE which RCIC valve change position on RCIC while in Full Flow Test and a manual isolation push-button is depressed. (039.012.A.Ol)
  • 37 STATE the required actions necessary to reset an automatic RCIC isolation. (039.0 14.A.01)

SOUTHERN NUCLEAR I I PAGE600F74 PLANTE.I. HATCH I I DOCUMENT TITLE: DOCUMENT NUMBER: I VERSION NO:

LOSS OF DC BUSES 34AB-R22-OOl-2 I 3i3 TITLE:

ATTACHMENT 5 250V DC RX BLDG ESSEN. MCC 2k 2R24-S021 I ATTACHMENT PAGE:

j IOF1 FRAME NO. DESCRIPTION IA 2E51-F046, Turbine Cooling Water Valve (MOV) lB 2E51-F524, RCIC Trip and Throttle Valve (MDV) 2A 2E51-F008, RCIC Steam Supply Line Isolation Valve (MDV) 2B 2E51-FO1O, CST Suction Valve (MDV) 3A 2E51-F012, RCIC Pump Discharge Valve (MDV) 3B 2E51-F013, RCIC Pump Discharge Valve (MDV) 4A 2E51-C002-1, Barometric Condenser Condensate Pump 4B 2E51-C002-2, RCIC Barometric Condenser Vacuum Pump 5A Feeder Cable From 2R22-S016, Frame 2B 6B 2E51-F019, RCIC Minimum Flow Valve (MDV) 8A 2E51-F022, RCIC Test Line to CST Valve (MDV) 8B 2E51-F029, RCIC Torus Outboard Suction Valve (MDV) 9A 2E51-F031, RCIC Torus Inboard Suction Valve (MDV) 9B 2E51-F045, Steam to Turbine Valve (MDV)

MGR-0009 Ver. 4

HLT-07 SRO NRC EXAM

20. 218000K1.04 001 Unit 2 has experienced a Loss of Offsite Power (LOSP).

The following conditions existed at 15:00:

o Reactor All rods in o RPV Pressure 860 psig controlled by LLS o RWL -97 inches, decreasing at 2 inches/minute o Drywell Pressure 1.0 psig, increasing at 0.3 psi/minute o ADS Inhibit Switches Normal position Given these trends, which ONE of the following predicts the EARLIEST time that the ADS valves will have automatically OPENED?

A. 15:03 B 15:06 C. 15:13 D. 15:15

Description:

Refer to logic drawing provided as a reference for developing this test item.

1. With a high Drywell pressure signal present (1.85 psig), the following must occur to initiate ADS.
a. Low Reactor water level (Level 3) at +3.0
b. Low Reactor water level (Level 1) at -101
c. 102.5 second timer timed out
d. CS Pump discharge pressure of 152 psig or RHR Pump discharge pressure of 127 psig
e. Once the 102.5 second timer has timed out, if RHR or CS Pump discharge pressure is available, all 7 ADS Valves open
2. Initiation of ADS without high Drywell pressure will occur if the following conditions exist simultaneously:
a. Low Reactor water level (Level 3) at +3.0
b. Low Reactor water level (Level 1) at -101
c. High Drywell Pressure Bypass Timer timed out 11 minutes
d. 102.5 second timer timed out. Without the high Drywell pressure signal, the 102.5 second timer will not initiate until the High Drywell Pressure Bypass Timer times out and 2.a and 2.b are present.

59

HLT-07 SRO NRC EXAM

e. CS Pump discharge pressure of 152 psig or RHR Pump discharge pressure of 127 psig.
f. Once the 102.5 second timer is timed out, if RHR or CS Pump discharge pressure is available, all 7 ADS Valves open.

At 1502, RWL is -101 and at 1503 DW pressure is 1.9 psig, which starts the 102.5 second timer. At 1505 the 102.5 second timer has timed out and the ADS valves have automatically opened.

The A distractor is plausible if the applicant assumes the ADS permissives are met in 4 minutes (based on RPV level rate & 102.5 second timer timing out) causing the ADS valves to auto open.

The C distractor is plausible if the applicant assumes the ADS permissives are met in 11 minutes (based on RPV level rate timing out in 11 minutes) causing the ADS valves to auto open.

The D distractor is plausible if the applicant assumes the ADS permissives are met in 13 minutes (based on RPV level rate timing out in 11 minutes then the 102.5 second timer) causing the ADS valves to auto open.

A. Incorrect See description above.

B. Correct See description above.

C. Incorrect See description above.

D. Incorrect See description above.

References:

NONE K/A:

218000 Automatic Depressurization System K1. Knowledge of the physical connections and/or cause effect relationships between AUTOMATIC DEPRESSURIZATION SYSTEM and the following:

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

Kl.04 DrywelL/containment pressure: Plant-Specific 3.9 4.2 LESSON PLAN/OBJECTIVE:

60

HLT-07 SRO NRC EXAM B2 1 -ADS-LP-03 801 Automatic Depressurization System (ADS) EO 038 .004.a.02 References used to develop this question:

B21-ADS-03801, Automatic Depressurization System (ADS), Fig 2 & Fig 4 34S0-B21-001-2, ADS and LLS System Modified from HLT-5 NRC Exam Q#20 ORIGINAL QUESTION (HLT-5 NRC Exam Q#20)

Unit 2 has experienced a Loss of Offsite Power (LOSP).

The following conditions existed at 15:00:

o Reactor All rods in o RPV Pressure 860 psig controlled by LLS o RWL -93 inches, decreasing at 2 inches/minute o Drywell Pressure 0.6 psig, increasing at 0.05 psi/minute o ADS Inhibit Switches Normal position Given these trends, which ONE of the following predicts the EARLIEST time that the ADS valves will automatically open?

A. 15:04 B. 15:06 C. 15:15 D.V 15:17 61

B21-ADS-LP-03801 Page 2 of 36 AUTOMATIC DEPRESSURIZATION SYSTEM (ADS)

Initial License (LT) ENABLING OBJECTIVES

1. STATE the purpose of the Automatic Depressurization System per Unit 2 FSAR. (038.002.a.03)
2. List the (7) SRVs used for the ADS function. (038.001 .a.01)
3. Given a list, CHOOSE the correct electrical power source for the following ADS components:

(038.001 .a.02)

a. A logic circuit
b. B logic circuit
4. STATE the function of the following ADS components: (038.001.a.04)
a. 102.5 second timer
b. High Drywell Pressure Bypass Timer
c. ADS valve control switches
d. LLS valve control switches
e. Logic reset pushbuttons
f. Drywell High press reset pushbuttons
g. Low level reset pushbuttons
h. ADS inhibit key switches
5. Given a set of plant conditions, EVALUATE those conditions and DETERMINE if ADS should have initiated. (038.004.a.02)
6. Using a simplified drawing of the ADS Logic circuitry, STATE the expected system response on a LOCA with loss of all high pressure injection and no Operator action. (038.004.a.03)
7. Given that an ADS valve is open with an initiation signal present, LIST the expected indications observed in the Control Room per 34SO-B21-001-l/2 Automatic Depressurization (ADS) &

Low-Low Set (LLS) System. (038.002.a.02)

8. STATE the significance in terms of ADS operation, of the following annunciators:

(038.004.a.04)

a. AUTO BLOWDOWN TIMERS INITIATED
b. AUTO BLOWDOWN CS or RHR PRESSURE PERMISSIVE
9. Given ADS has spuriously initiated, from a list, SELECT the appropriate actions directed by 34AB-E10-00l-1/2; Inadvertent Initiation of ECCS/RCIC, to inhibit or prevent ADS Blowdown. (038.005 .a.0 1 & 200.027.a.02)
10. Given a fire exist in the Main Control Room; DETERMINE the Operator actions to prevent a spurious ADS initiation per 34AB-X43-001-1/2, Fire Procedure.

(038.007.a.02)

SNCPLANTE.I.HATCH I Pg34of36 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

AUTOMATIC DEPRESSURIZATION (ADS) AND 34S0-B21-001-2 13.13 LOW-LOW_SET (LLS) SYSTEMS ATTACHMENT 5 Att. Pg.

TITLE: ADS LOGIC DIAGRAM AND SYSTEM INFORMATION 1 of 2

(+) ..-I F-i j 1.85k DAN K754A K2A K3A K756A K752A I_I (seal in - I TDC -101lNI (sealin)

K3A (seal in) 102.5 SEC) 3RWL J

S3 A logic reset P0 2R25-S001 high (P602) 125 VDC 2A DAN reset (P602) 2021C-57A open to inhibit K2A ADS K752A K9I KI2A K4A K52A K6A K752A K5A K53A K7A K8A KI3A K5A timer initiated alarm X9A and KIOA CLOSE on EITHER 12A same as K6A for valve ADS LOGIC CHANNEL A CS. press OR 152 K1lAsameasl(7A FOI3M RHRpress@ 12Z#

The High Drywell Pressure Bypass Timer The High Dryweil Pressure Bypass Timer ensures fuel clad temperatures are maintained <1500cF during steam line breaks outside Primary Containment, between the containment wail and an outboard MSIV. This transient would NOT generate a high Drywell pressure signal. WHEN the High Drywell Pressure Bypass Timer times out, a set of contacts close which bypass the high Drywell pressure ADS permissive contacts. This allows ADS initiation, provided all of the other required permissives are present. The timer is set for 11 minutes.

  • The timer is initiated on Low Reactor Water Level (Level 1) at -101.
  • The High Drywell Pressure Bypass Timer can be reset with the ADS Low Water Level Reset Pushbuttons on 2H11-P602 panel only IF Reactor water level recovers to above -101.
  • Once initiated the timer will time out regardless of level recovery IF the Inhibit Switches being placed in INHIBIT, unless manually reset with Rx level above -101 inches.

MGR-0009 Ver. 4

(+)

F-i P927 1.85# 1.85#

K754A DIW-II A 756 F]=K K752A -:;;

(SEAL IN -

  • mc I 311 K8A 102.5 SEC) 1 RWL JSEAL IN)

_I I

-- K3A (SEAL IN)

S3 HIGH 2R25-S00 1 K2AT - D/W A LOGIC RESET 125 VDC2A RESET PUS HBUTTON (P602) (P602)

K2A J

K3A 2B21C-57A - 2B21C-57A OPEN TO INHIBIT I ADS KI OA 00 K752A (SEAL IN TD C 102.5 SEC)T IK9A K4$

4 Ki 2A A 4 K 7 52A 4 K5A K 53A K7A K 4 8A(13A 2A

(->-* IIF-2 II.

K5A TIMER INITIATED ALARM K9A & K1OA close on: K4A SRV Green light indication K6A & K7A SRV Solenoids Core SgaY Press at 152 psig KS2A & K53A SPD S Indications

-tJ c) RHR Press at 127 psig (0 K12A& K13A FO13M Solenoid C

0

()

C) ADS ID GIC C HANNEL A

I CLOSES ON REACTOR I i

I LOW WATER LEVEL

+

K753A I

I 2H11-P925 T

2E21A-ADS LOGIC K361A (LEVEL-lOl)

T K370C (LEVEL-lOl)

)WER SUPPLY a a J 2H11-P602 1 211h1-P602 1 I NORMAL NORMAL S6A S6A I

I K753A K755A ADS LOW WATER LEVEL TIMERS B21-ADS-03801 Fig4 Page 35 of 36

HLT-07 SRO NRC EXAM

21. 223001A3.04 001 Which ONE of the choices below is the expected Unit 2 PEAK Drywell pressure following a Design Bases LOCA and where this PEAK value can be monitored?

A. Approximately 28 psig; 2H1 1-P650 panel B. Approximately 28 psig; 2H1 1-P602 panel C Approximately 47 psig; 2H1 l-P650 panel D. Approximately 47 psig; 2H1 1-P602 panel 62

HLT-07 SRO NRC EXAM

==

Description:==

From a DBA LOCA water and steam from the reactor vessel is released into the drywell, thus pressurizing the drywell. As the drywell becomes pressurized, water and steam is blown down to the torus, where it is condensed. At the same time, the non-condensables that make up the normal drywell atmosphere are also blown to the torus, resulting in an increasing pressure in the torus. The water temperature of the torus heats up as the steam that is being released underwater is condensed. The pressure in the drywell reaches a peak of 46.9 psig approximately 3 to 5 seconds after the accident, after which the rate of blowdown decreases and the steam condenses faster than the blowdown rate. The peak pressure of 46.9 psig is below the design pressure of 56 psig. The buildup of non-condensables in the torus continues until there are no non-condensables in the drywell. Torus pressure will peak at approximately 28 psig. The drywell pressure decreases as it blows down to the torus and eventually stabilizes at out approximately 33 psig.

There are three ranges of indication:

Narrow range (-5 to 5 psig) has indication on panels ill l-P602 and P654 in the control room.

Intermediate range (-10 to 90 psig) has indication on panels H11-P650, P654 and P657 in the control room.

Wide range (0 to 250 psig) has indication on panels Hl l-P650 and P657 in the control room.

The A distractor is plausible if the applicant confuses the expected peak Torus pressure with the peak Drywell pressure. The second part is correct.

The B distractor is plausible if the applicant confuses the expected peak Torus pressure with the peak Drywell pressure. The second part is plausible if the applicant confuses the Narrow range Drywell pressure indication (P602) with the Intermediate range Drywell pressure indication (P650).

The D distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses the Narrow range Drywell pressure indication (P602) with the Intermediate range Drywell pressure indication (P650).

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

63

HLT-07 SRO NRC EXAM

References:

NONE K/A:

223001 Primary Containment System and Auxiliaries A3. Ability to monitor automatic operations of the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES including: (CFR: 41.7 /45.7)

A3.04 Containment/drywell response during LOCA 4.2* 4*3*

LESSON PLAN/OBJECTIVE:

T23-PC-LP-0 1301, Primary Containment, EO 200.004.A.03 D11-CAMS-LP-05101, Containment Atmosphere Monitoring System (CAMS),

EO 201 .076.A.33 References used to develop this question:

U2 FSAR Chapter 6.2.3.1.2, Containment Response to LOCA 34SV-SUV-0 19-2, Surveillance Checks 64

T23-PC-LP-01301 Page 8 of 160 PRIMARY CONTAINMENT

28. Given a list of statements, IDENTIFY the statement which best describes the plant conditions which will generate the following PCIS isolation signals: (013.045.A.05, 013.046.A.05 01 3.047.A.05)
a. Group 1 isolation
b. Group 2 isolation
c. Group 3 isolation
d. Group 4 isolation
e. Group 5 isolation
f. Group 6 isolation
29. Given a list of statements, IDENTIFY the statement which best describes the steps require d to reset the following PCIS isolation signals: (013.045.A.06, 013.046.A.06, 013.047.A.06 013.067.A.01, 040.006.A.01)
a. Group 1 isolation
b. Group 2 isolation
c. Group 3 isolation
d. Group 4 isolation
e. Group 5 isolation
f. Group 6 isolation
30. Given a list of statements, IDENTIFY the statement which best describes the primary containment characteristics during a D13A LOCA, including the process by which energy is absorbed and dissipated by theprimary containment. (200.004.A.03)
31. Given a list of statements, IDENTIFY the statement which best describes the consequences of having the torus to drywell vacuum breakers fail either open or closed during a DBA LOCA (200.004.A.02)

NORMAL OPERATIONS

(013.006.A.0l)

-l12, Containment Atmospheric Control and Dilution System, DETERMINE the sequence of actions necessary to inert the Primary Containment.

(013.007.A.0l)

Dhl-CAMS-LP-05101 Page 2 of 46 CONTAINMENT ATMOSPHERE MONITORING SYSTEM (CAMS)

Initial License (LT) ENABLING OBJECTIVES

1. LIST the four panels where Primary Containment hydrogen or oxygen level can be read.

(013.060.A.02)

2. LIST the four conditions that cause an automatic isolation of the Primary Containment atmosphere H./0 2 analyzer. (013.051 .A.02)
3. When the Priniary Containment Atmosphere 2 /0 analyzer override switches are placed in H

bypass, DESCRIBE what happens automatically to the analyzers. (013 051.AM3)

4. LIST the two panels in the Control Room that have multipoint recorders for Primary Containment temperatures for Unit 2 or Unit 1. (201 .073.A.19)
5. NAME the two panels in the Control Room where Torus temperature is found. (201 .074.A. 15)
6. IDENTIFY the correct ranges and panel locations in the Control Room, for all three ranges of Drywell pressure indications. (201.076.A.33)
7. IDENTIFY the correct ranges and panel locations in the Control Room, for both wide and narrow range Torus water level indications. (201.075.A.14)
8. In accordance with 34S0-P33-001-2 Primary Containment Atmosphere H 0 Analyzer 2

System, PLACE H 0 analyzers in service during the following conditions:

2 (013.051.A.04 and 013.060.A.01)

a. LOCA Signal present
b. LOCA Signal not present.
9. NAME the two situations when the Primary Containment Atmosphere 2 /0 analyzer must be H

placed in service. (013.060.A.03)

Objectives marked by a RED (*) are required during RO-305 and SR-305 of the Initial License program.

HNP-2-FSAR-6 The long-term containment response results in the most limiting suppression pool temperature and the discussion is provided in paragraph 6.2.3.1 .2.2.

The peak drywell pressure for expanded operating domain (EOD) operation is also calculated 9

using break flowrates and enthalpies calculated by a more mechanistic blowdown model6 that provides a more realistic prediction of the DBA-LOCA blowdown for operating conditions in EOD, including consideration of final feedwater temperature reduction at selected operating conditions.

6.2.3.1.2.1 Short-term Containment Pressure and Temperature Response to DBA LOCA. The short-term response was performed in accordance with Regulatory Guide 1.49 and reference 40 using the M3CPT computer code. The analysis was performed at various reactor operating conditions associated with plant performance improvements, such as maximum extended load line limit (MELLL) and final feedwater temperature reduction (FFWTR). The analyses used blowdown flowrates based on the blowdown model built into M3CPT and the LAMB blowdown model. In using the LAMB blowdown model, the blowdown is calculated first using the LAMB code; then the LAMB flow versus time is used as input to M3CPT.

M3CPT calculations based on LAMB blowdown flow time histories are performed for all reactor operating points considered, whereas the M3CPT blowdown model is used only for the 100%

power and the rated core flow point. The short-term response performed with M3CPT blowdown is the basis for the peak containment pressure of 50.78 psig for Unit 1 and 47.22 psig for Unit 2.

In addition, LAMB blowdown flow obtained for Unit 1 is also used for the Unit 2 analysis. The use of Unit 1 blowdown flow for Unit 2 is considered conservative because the blowdown flow for Unit 1 is higher at the same reactor power/flow point due to higher degree of subcooling for Unit I in the downcomer region.

The DBA-LOCA event (a double-ended break of a recirculation suction line) is assumed to occur at the following operating points:

  • Rated core flow and power level of 2818 MWt (100.5% of the current power level of 2804 MWt).
  • Minimum cOre flow with FFWTR at 2818 MWt (maximum subcooled condition).

Comparison of responses at the conditions defined above shows the impact of subcooling on the containment response to DBA-LOCA. It is generally considered that containment loads will be higher with higher subcooling because a higher subcooling will result in a higher blowdown flow for a given pressure.

The containment peak pressure was analyzed using the M3CPT built-in blowdown flow model for 100.5% of the current rated thermal power conditions, and the pressure and temperature 6.2-26 REV27 10/09

SOUTHERN NUCLEAR PLANT E. I. HATCH PAGE 24 OF 77 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

SURVEILLANCE CHECKS 345V-SUV-019-2 37.1 OPER T/SOR 7.2 PANEL - INSTRUMENT I TECH SPEC. NOTE FREQ NIGHT I DAY COND OPER LIM 2H11-P603 Control Rod Position Determination Position able to be 7.2.1 B,F 1,2 a letermined for each (SR3.1.3.1) control rod Confirm the following valves are open:

- 2C1 1 -FOl OA, Scram Discharge Volume Vent Valve

- 2C1 1 -FOl OB, Scram Discharge Volume Vent Valve

- 2C11-F011, Scram Discharge Volume Drain Valve 722 B 1,2 d(Mon) All open

- 2C1 1-F035A, Scram Discharge Volume Vent Valve

- 2C1 1-F035B, Scram Discharge Volume Vent Valve

- 2C1 1-F037, Scram Discharge Volume Drain Valve (SR 3.1.8.1) 2H1 1 -P650 2P41-R61 1, PSW INTK STRUCTURE WATER LEVEL 7.2.3 1,2,3,4,5 a >61.7feet

- 2P41-R612, RIVER ELEV.

(SR_3.7.2.1)

If river level in previous step is < 61.9 feet, record level 7.2.4 from 1P41-R700, Intake well level stick. 1,2,3,4,5 a > 61 .7 feet (SR 3.7.2.1) 2H1 1-P650 2P1 1-R601, Cond Storage Tank Level 725 (SR_3.5.2.2.b.) 6 a > 15 2H1 1-P657 2T48-R601A, Wide Range DAN Pressure (Red) 7.2.6 U 1,2,3 a 1.92PSlG 2H1 1-P650 2T48-R601 B, Wide Range DAN Pressure (Red)

Confirm Items in 7.2.6 are within 2 PSIG.

727 B 1,2,3 a (SR_3.3.3.1.1_for 3.3.3.1-1(4.c.))

2H1 1-P657: 2T48-R6O1A Wide Range DW Radiation (Green) 7.2.8 U 1,2,3 a <138 R/HR 2H1 1-P650: 2T48-R6O1 B Wide Range DW Radiation (Blue)

Confirm Max Minus Mm < 10 for Items in 7.2.8 7.2.9 (SR3.3.6.1.lfor3.3.6.1-1(2.c.)) B 1,2,3 a (SR33.3.1 .1_for_3.3.3.1-1 (5))

Record 2P41 -R373A(2P4 1 -R373B) 2P42-ROO2A(2P42-ROO2B)

PSW/RBCCW dP from 3400-OPS-030-2 7.2.10 2H1 1-P650 Confirm annun. 650-238, HX PSW/RBCCW B 6 c DIFF PRESS LOW is (is not)

ILLUMINATED if dp is < (>)7 PSID.

(TS 5.5.4.f., ODCM Table 2-2(4))

Initials Calculations verified Date________ Time Night! Day GI 6.030 MGR-0001 Ver. 3

HLT-07 SRO NRC EXAM

22. 223002K3.10 001 Unit 2 is at 100% power and the 2A Reactor Protection System (RPS) MIG Set trips.

Which ONE of the choices below completes the following statement?

After RPS M/G Set 2A trips, the operating Reactor Water Cleanup pump A will AUTOMATICALLY trip due to suction valve closure B. will AUTOMATICALLY trip due to low suction pressure C. must be MANUALLY tripped due to suction valve closure D. must be MANUALLY tripped due to low suction pressure 65

HLT-07 SRO NRC EXAM

==

Description:==

When RPS 2A trips, the power supply to PCIS logic is lost. This causes a half isolation of Groups 2 & 5 (RWCU). One RWCU isolation valve will close (FOOl) causing the RWCU pump to trip. Other systems (HPCI, RCIC, CRD,CBPs) have a low suction pressure trip and RWCU trips the pump before a low suction pressure condition is developed.

The B distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses a low suction pressure trip with the suction valve closing. Also plausible since other systems (HPCI, RCIC, CRD,CBP5) have a low suction pressure trip and the applicant confusing this with RWCU.

The C distractor is plausible if the applicant thinks the RWCU pump responds similar to Core Spray and RHR pumps which require manual tripping when a suction valve closes (2E21-FOO1IFO19 & 2E11-F065s).

The D distractor is plausible if the applicant confuses a low suction pressure with the suction valve closing. Also plausible since other systems like RHR (via F065, Torus Suction valve, closing) will develop a low suction pressure condition but does not have a trip, therefore, manual tripping of the pump would be required.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

66

HLT-07 SRO NRC EXAM

References:

NONE K/A:

223002 Primary Containment Isolation System/Nuclear Steam Supply Shut-Off K3. Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF will have on following:

(CFR: 41.7 /45.4)

K3.10 Reactor water cleanup 9 3.1 LESSON PLAN/OBJECTIVE:

C7 1 -RPS-LP-0 1001, Reactor Protection System, EU 200.1 02.a.0 1 References used to develop this question:

34AB-C71-002-2, Loss Of RPS, Attachment 3 34S0-G3 1003-2, Reactor Water Cleanup System (Pump trips) 67

C71-RPS-LP-O1001 Page 5 of 112 REACTOR PROTECTION SYSTEM

16. Given a loss of an RPS Bus, DESCRIBE the effects on the following Systems:

(O10.002.a.O1, 5 00.102.a.01)

a. RPS
b. Primary Containment Isolation
c. SBGT
d. Secondary HVAC
17. Given a list of statements, IDENTIFY the steps necessary to align alternate power from Essential Bus A or B to RPS Bus A or B. (010.025.a.02)
  • 18. Given plant conditions, DESCRIBE the steps necessary to startup the RPS System in accordance with 34S0-C71-001-1/2, 120 VAC RPS Power Supply System. (010.001.a.05)
19. Given plant conditions, DESCRIBE the steps necessary to shutdown the RPS System per 34S0-C71-0O1-1!2, 120 VAC RPS Power Supply System. (010.003.a.01)
  • 20. Given plant conditions, SUMMARIZE the steps required to transfer RPS buses to the alternate power supply per 34S0-C71-001-l/2, 120 VAC RPS Power Supply System. (010.002.a.04)
  • 21. Given that a transfer to alternate RPS power is necessary, DESCRIBE the conditions required to prevent Group II or Group V PCIS Isolation from occurring per 34S0-C7 1-001-1/2, 120 VAC RPS Power Supply System. (010.002.a.02)
22. Given the scram discharge volume Vent and Drain Valves are shut; DISCUSS the precaution regarding a possible reactor scram noted in 34SV-Cl 1-001-1/2, SDV Isolation Valve Functional Test. (0l0.007.a.Ol)
23. Given a list of statements, IDENTIFY the statement which describes the purpose of performing the RPS Channel Test Switch Functional Test. (010.004.a.0l)
24. In accordance with 34SV-C7l-003-1/2, Reactor Mode Switch in Shutdown Functional Test, SUMMARIZE the steps necessary to perform the Mode Switch Functional Test. (0 10.005.a.01)
25. Given a list of statements, SELECT the statement which best describes the purpose of performing the Mode Switch in Shutdown Functional Test. (0l0.005.a.02)
26. In accordance with 34SV-C71-004-1/2, Reactor Manual Scram Functional Test, SUMMARIZE the steps necessary to perform the Manual Scram Functional Test.

(010.006.a.0 1)

27. Given a list of statements, SELECT the statement that best describes the purpose of performing the Manual Scram Functional Test. (010.006.a.02)
  • 28. Given plant conditions, DETERMINE if conditions are met to change Mode Switch position from RUN to START UP/HOT STANDBY. (010.019.a.01)
29. Given plant conditions and the Mode Switch is changed from RUN to START UP/HOT STANDBY, DETERMINE if a Rod Block or Protective Action should have occurred.

(010.01 9.a.02)

SOUTHERN NUCLEAR DOCUMENT TYPE:

PLANT E. I. HATCH ABNORMAL OPERATING PROCEDURE PAGE 1 OF 12 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF RPS 34AB-C71-002-2 49 EXPIRATION APPROVALS: EFFECTIVE DATE: DEPARTMENT MANAGER J. I. Hammonds DATE 04-07-97 DATE:

N/A SSM/PM N/A DATE N/A 12-14-11 tO CONDITIONS 1.1 RPS Bus A or B has been de-energized.

1.2 Top row of annunciators ILLUMINATE on the 2H1 1-P603-1.

1.3 ANNUNCIATORS

  • 603-117, REACTOR AUTO-SCRAM SYSTEM A TRIP, OR
  • 603-118, REACTOR AUTO-SCRAM SYSTEM B TRIP
  • 603-208, GROUP 1 SYSTEM A TRIP, OR
  • 603-209, GROUP I SYSTEM B TRIP 1.4 For additional Annunciators see Attachment 1.

1.5 For additional Conditions see Attachment 2.

2.0 AUTOMATIC ACTIONS See Attachment 3.

3.0 IMMEDIATE OPERATOR ACTIONS None MGR-0002 Rev. 8

SOUTHERN NUCLEAR PLANTE.I.HATCH PA GE3 OF 12 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF RPS 34AB-C71-002-2 4.9 4.3.6 OPEN the Fission Product Monitoring isolation valves in the affected channel at panel 2H1 1-P700:

AffGcted RPS BUS Fission Product Monitor Valves RPS Bus A Inboard: 2D1 1-F050 LI 2D11-F051 LI 2D11-F071 LI RPS Bus B Outboard: 2D11-F052 LI 2D11-F053 LI 2D11-F072 LI 4.3.7 Restore Reactor Water Clean up per 34S0-G31-003-2. LI 4.3.8 Restore Control Room ventilation to the desired mode of operation per 34SO-Z41-001-1. LI 4.3.9 At 2H1 1-P607, TIP Control Panel, momentarily place 2C51D-S1, PCIS Reset and Test Switch, to reset.

4.3.10 If required, reset trips on ATTS instruments at the following panels:

  • 2H11-P924 LI 4.4 Evaluate equipment lost on Attachment 3 and enter appropriate Tech Spec LCOs. LI

5.0 REFERENCES

5.1 Technical Specifications 5.1.1 3.3 Instrumentation 5.1.2 3.6 Containment Systems 5.2 Technical Requirements Manual 5.3 Plant Drawings 5.3.1 H-23369, Plant Hatch Unit Two Single Line Diagram for 120/208V Essential A.C. System MGR-0001 Rev. 4.0

SOUTHERN NUCLEAR PLANT El. HATCH PAGE 10 OF 12 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF RPS 34AB-C71-002-2 4.9 ATTACHMENT 3 ATTACHMENT PAGE:

TITLE: LOSS OF .RPS BUS AUTOMATIC ACTIONS 2 OF 4 2E1 1-Fl 22A RHR Check Vlv 2E1 1-FO5OA Bypass Vlv 2E11-FO15A RHR Inbd Injection Valve (IF in shutdown cooling) 2E1 1-F040 RHR to Radwaste 2E11-F079A RHR Sample Line 2E11-F079B RHR Sample Line 2E11-F009 SDC Suction Vlv 2B21-SV-Fl 11 Post Acc Rx CInt CNMT ATMOS Sample Inbd Isol Closure of 2G31F001, Group 5 isolation valve TIP withdrawal AND isolation Closure of Group I inboard isolation valves:

2B21-F016 MSL Drain Vlv 2B31-F019 Rx Water Sample Vlv Control Room Ventilation shifts to pressurization mode 1Z41-CO12A AND lZ4l-CO12B, Fans START lZ4l-COllAandlZ4l-COIIB, Fan TRIP lZ4l-F0ll, 1Z42-F012, 1Z41-F019, 1Z41-F020 CLOSE 2N33-COO1A and 2N33-COOIB, Steam Packing Exhausters trip 2N22-C002, Mechanical Vacuum Pump trips AND isolates MGR-0009 Ver. 4

SOUTHERN NUCLEAR PLANTE.I. HATCH PAGE 7 OF 165 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REACTOR WATER CLEANUP SYSTEM 34S0-G31-003-2 38.3 5.2.4 2G31 -COO1A AND/OR 2G31 -COOl B, RWCU Pumps, trip under the following conditions:

  • High cooling water temperature, 140F.
  • Low flow for the A RWCU pump, <60 gpm, is bypassed for 15 sec. following pump start. Low flow for the B RWCU pump, <60 gpm, is bypassed for 30 sec. following taking the Pump Control Switch to START.
  • 2G31-F0O1, RWCU Inboard Isolation, OR2G31-F004, RWCU Outboard Isolation, NOT full open.

5.2.5 For2G3l-C0O1B, RWCU Sealless Pump, a ten minute cooling period must be allowed between starts. Failure to comply with this limitation will result in an elevated motor winding temperature which could damage the winding insulation. Three starts per hour are allowed.

5.2.6 2G31-COO1B, RWCU Sealless Pump, motor may be damaged jf the motor temperature exceeds 1 4OcF. IF the motor cavity exceeds 1 80F as indicated by bc al temperature indicator RWCU Pump RBCCW Outlet Temp, 2G31-NOO2B, do NOT start OR operate the pump. Notify Engineering for corrective action per vendor manual S-60342.

5.2.7 Pump casing cooldown rate is limited to 140F/hr. (2.33F/min). Using 2G31-F148B, throttle purge flow to greater than .8 gpm ( 48 GPH) but less than Q equal to 1.5 gpm

( 90 GPH) as necessary to control cooldown, until pump casing temperature is <325F.

After pump casing to CRD seal purge A temp is 200°AF, increase purge flow to maximum, ( 3 gpm). This limit is imposed for normal pump shutdown AND is NOT to be applied following a pump trip OR isolation. It may be necessary to reduce purge flow to less than 8 gpm in order to prevent exceeding pump casing cooldown rate.

5.2.8 2G31-F078, Slow Backwash Water Supply Valve, open stops are pre-set at a position to limit flow rate to 40-50 gpm. Although the valve is throttled, it is normal for it to indicate FULL OPEN.

5.2.9 The 2A RWCU Demineralizer has manual isolations whereas the 2B RWCU does NOT.

5.2.10 The 2G31-COO1A RWCU Pump has been analyzed by the vendor to operate without RBCCW cooling water under the following conditions:

  • The moderator being pumped will be < 180 F, AND
  • The pump bearing casing temperature will be < 170 F.

These temperatures will be monitored at a frequency established by the U2 SS.

5.2.11 RWCU pump casing temperature readings obtained during heatup/cooldown must be confirmed to be accurate by use of a backup portable heat gun or other measurement device.

MGR-0001 Rev4

HLT-07 SRO NRC EXAM

23. 233000K2.02 001 Unit 1 is in Mode 4 with iB and iD Residual Heat Removal (RHR) pumps running in Shutdown Cooling with Fuel Pool Cooling Assist in service.

o A condition develops on Emergency Bus 416OVAC iF which causes it to de-energize Which ONE of the following predicts the status of the lB and iD RI-IR pumps?

A. lB RHR Pump has stopped ID RHR Pump has stopped B iB RHR Pump is running iD RHR Pump has stopped C. iB RHR Pump has stopped iD RHR Pump is running D. iB RHR Pump is running iD RHR Pump is running 68

HLT-07 SRO NRC EXAM

==

Description:==

The iF bus supplies power to the iC and iD RHR pumps. When this bus is lost, those pumps will no longer be operating. The KA for this question requires the candidate to know the power supplies to the RHR pumps when RHR is in SDC mode of operation. Plausibility for distracters is based on whether or not the candidate actually does remember the power supply lists, and whether the candidate confuses the various pumps supplied by the buses (i.e. RHRSW pump iD is powered by 4160 VAC bus lG while RHR pump iD is powered from 4160 VAC bus iF).

The A distractor is plausible if the applicant confuses the power supplies of RHR pumps with RHRSW pumps.

The C distractor is plausible if the applicant confuses the power supplies of RHR pumps with RHRSW pumps.

The D distractor is plausible if the applicant confuses the power supplies of RHR pumps with RHRSW pumps.

A. Incorrect See description above.

B. Correct See description above.

C. Incorrect See description above.

D. Incorrect See description above.

69

HLT-07 SRO NRC EXAM

References:

NONE K/A:

233000 Fuel Pool Cooling and Clean-up K2. Knowledge of electrical power supplies to the following: (CFR: 41.7)

K2.02 RHR pumps 2.8* 2.9*

LESSON PLAN/OBJECTIVE:

El l-RHR-LP-0070l, Residual Heat Removal System, EO 006.001 .a.02 References used to develop this question:

HLT Database 233000K2.02-00l 34S0-El 1-010-1, RHR System (Att. 2, 1R22-S006/S007) 70

E11-RIIR-LP-00701 Page 4 of 130 RESIDUAL NEAT REMOVAL SYSTEM Initial License (LT) ENABLING OBJECTIVES

1. Given a simplified drawing of the RHR system, TRACE the flowpath for RHR in LPCI mode.

(006.00 1 .a.01)

2. Given a simplified drawing of the RHE. system, TRACE the flowpath of Drywell Sprays.

(007.001.a.0l)

3. Given a simplified diagram of the RHR system, TRACE the flowpath for Torus Spray.

(007.003.a.01)

4. Given a simplified drawing of the RHR system, TRACE the flowpath for Torus Cooling.

(007.005.a.01)

5. Given a simplified drawing of the RHR system, TRACE the flowpath for RFIR operating in the Shutdown Cooling Mode. (007.007.a.04, 007.024.b.01)
6. Given plant conditions, EVALUATE those conditions to DETERMINE if RWL is adequate for RHR Shutdown Cooling flow. (007.007. a.02, 007 .024.b.02)
7. Given a simplified drawing of the RHR system, TRACE the flowpath for RHR in fuel pool cooling assist mode. (007.0 l6.a.01)
8. Give a simplified drawing of the RHR system, TRACE the flowpath for draining the RPV to the Torus. (006.01 1.a.01)
9. Given a simplified drawing of the RHR system, TRACE the flowpath for draining the Torus to Radwaste with RHR Loop B. (013.004.a.01)
10. From a list of statements, SELECT the statement that describes the consequences on the RHR system if the RFIR Pump Minimum Flow bypass Valve fails closed. (006.005.a.04)
11. From a list of statements, SELECT the statement that describes the maximum flow rate for RHR in Shutdown Cooling and the reason for the flow limit. (007.007.a.03, 007.024.b.03)
12. From a list of power supplies, SELECT the power supply for the RHR Pumps. (006.001 .a.02)
13. Given a Plant Hatch load list, DETERMINE power supplies to RHR valves and components.

(006.007.a.01)

14. Given a simplified drawing of RHR, IDENTIFY the following system interfaces: (006.OOl.a.04)
a. Reactor Recite System
b. Fuel Pool Cooling System
c. Core Spray Jockey Pump System
d. Radwaste
e. Core Spray System

SOUTHERN NUCLEAR PLANT E.I. HATCH PAGE 159 OF 293 DOCUMENT TITLE: I DOCUMENT NUMBER: I VERSION NO:

RESIDUAL HEAT REMOVAL SYSTEM I 34S0-E1 1-010-1 I 38.0 ATTACHMENT 2 LTTACHMENT PAGE:

TITLE: RHR SYSTEM ELECTRICAL LINEUP 6 OF 8 I NORMAL NUMBER DESCRIPTION I I CHECKED VERIFIED I POSITION I 4160V Sta SerSwgr 1G 1R22-S007 RHR Service Water Pump 1 B 1 El 1-COO1B RACKED IN OPEN Frame 2 30 Amp DC Control Power Breaker CLOSED 20 Amp Space Heater Breaker CLOSED RHR Service Water Pump 1 D 1 El 1 -COOl D RACKED IN OPEN Frame 3 30 Amp DC Control Power Breaker CLOSED 20 Amp Space Heater Breaker CLOSED RHR Pump lB lEll-COO2B RACKED IN OPEN Frame 7 30 Amp DC Control Power Breaker CLOSED 20 Amp Space Heater Breaker CLOSED 4160V Sta Serv Swgr 1E 1R22-S005 RHR Service Water Pump 1A 1 El 1-COO1A RACKED IN OPEN Frame 2 30 Amp DC Control Power Breaker CLOSED 20 Amp Space Heater Breaker CLOSED RHR Pump 1A 1E1I-COO2A RACKED IN OPEN Frame 7 30 Amp DC Control Power Breaker CLOSED 20 Amp Space Heater Breaker CLOSED 4160 Stat Serv Swgr 1 F I R22-S006 RACKED IN RHR Service Water Pump IC 1E11-COO1C OPEN Frame 3 30 Amp DC Control Power Breaker CLOSED 20 Amp Space Heater Breaker CLOSED RHR Pump 1D 1E11-COO2D RACKED IN OPEN Frame 7 30 Amp DC Control Power Breaker CLOSED 20 Amp Space Heater Breaker CLOSED OPS-0257 Ver. 12.2 G16.030 MGR-0009 Ver. 5

HLT-07 SRO NRC EXAM

24. 234000K5.02 001 Unit 2 is in a refueling outage with the following conditions.

o Rx Mode Switch Locked in the REFUEL position o Control rods One control rod is fully WITHDRAWN o Main Fuel Grapple Lowered, but verified to be able to clear the cattle chute by 3 ft.

(Empty grapple move) o NO rod is selected The Refuel Bridge is over the Spent Fuel Pool and is moving toward the core to pick up the fuel bundle at position 17-42.

Which ONE of the following choices completes the following statements concerning how the refueling interlocks will respond?

With these conditions, the Refuel Bridge move to the intended core location.

When the Refuel Bridge stops, if the Grapple Raise/Lower switch is placed in the LOWER position, the Main Grapple lower.

A. will; will B. will; will NOT C. will NOT; will D will NOT; will NOT 71

HLT-07 SRO NRC EXAM

==

Description:==

There are three types of refueling interlocks:

o interlocks that prevent control rod motion, by causing rod blocks o interlocks that prevent the refueling platform from moving toward (over) the core, by interrupting power to the motor that moves the refueling platform o interlocks that prevent hoist operation by interrupting power to the motor that moves the hoist.

The refueling interlocks will not allow fuel to be moved in or near the core unless all control rods are fully inserted. The refueling interlocks prevent the operation of loaded refueling equipment over the core when any control rod is withdrawn. The refueling interlocks also prevent the withdrawal of any control rod when fuel is loaded on refueling equipment and operating over the core. Fuel loaded is determined by a contact that opens when anything heavy enough to be fuel is loaded onto a hoist. In addition, when the reactor mode switch is in REFUEL, only one rod can be withdrawn. The selection of a second rod will initiate a rod block. The Refuel Bridge will NOT move to the intended core location and the grapple will NOT be able to be moved.

The A distractor is plausible if the applicant confuses or does not remember the Refueling Interlocks and thinks the bridge will continue to move over the core. The second part is plausible if the applicant confuses or does not remember the Refueling Interlocks and thinks the grapple can be lowered and latched onto a bundle.

The B distractor is plausible if the applicant confuses or does not remember the Refueling Interlocks and thinks the bridge will continue to move over the core. The second part is correct.

The C distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses or does not remember the Refueling Interlocks and thinks the grapple can be lowered and latched onto a bundle.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

72

HLT-07 SRO NRC EXAM

References:

NONE K/A:

234000 Fuel Handling Equipment K5. Knowledge of the operational implications of the following concepts as they apply to FUEL HANDLING EQUIPMENT: (CFR: 41.5/45.3)

K5.02 tFuel handling equipment interlocks. .. . 3.1 3.7 LESSON PLAN/OBJECTIVE:

F15-RF-LP-04502, Refueling. EU 045.018.A.O1 & EU 045.019.A.02 References used to develop this question:

34SV-F15-OO1-1, Refueling Interlocks and Hoist Limit Checks 73

F15-RF-LP-04502 Page 2 of 93

- Refueling Initial License (LT) ENABLING OBJECTIVES

1. Given plant conditions involving refueling operations, SELECT the conditions which will initiate the following: (045.019.a.0l)
a. Slack cable light
b. Grapple normal up light
c. Hoist loaded light
d. Hoist jam light
2. Given plant conditions requiring refueling operations, IDENT[FY the conditions that will initiate the following: (045.01 8.a.01)
a. Refueling Rod Block
b. Refueling Bridge Movement Block
3. Given plant conditions requiring refueling operations, IDENTIFY the conditions that prevent the operation of the following hoists: (045.019.a.02)
a. Fuel Grapple
b. Frame Mounted Hoist
c. Monorail Mounted Hoist
d. Service Platform Jib Crane
4. DESCRIBE the required location of Refueling Floor Personnel when a control rod is withdrawn with the vessel head removed AND with fuel in the vessel. (045.0 18.a.02)
5. Given plant conditions, DETERMINE if the prerequisites for moving fuel to or from the core are complete. (300.048.a.03)
6. Given plant conditions, DETERMINE the required frequency for performance of 34SV-F 15-001 1/2, Refueling Interlocks and Hoist Limit Checks. (045.018. a.03)
7. Given plant conditions, ANALYZE conditions to RECOGNIZE and IDENTIFY any administrative, equipment status, or conditions that indicate core alteration must be ceased.

(300.048 .a.0 1)

8. Given plant conditions, VERIFY fuel movement during core alterations or refueling.

(400.028.a.01)

9. Given plant conditions and annunciators that are indicative of irradiated fuel damage during fuel handling, DETERI\4INE the proper operator actions. (200.036.a.01)
10. Given plant conditions requiring that core alteration must cease and fuel is loaded on the fuel grapple, DESCRIBE the actions necessary to correctly cease fuel movement. (300.048.a.04)

SOUTHERN NUCLEAR PAGE PLANT E, I. HATCH .

36OF75 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REFUELING INTERLOCKS AND HOIST LIMIT CHECKS 345V-F15-OO1-1 15.0 7.3.2.6.2.3 Raise the fuel grapple to the GRAPPLE NORMAL UP position.

7.3.2.6.2.4 Move Refueling Platform I Fuel Grapple Hoist over test weight marked SU2701485WT (OR weighted dummy fuel bundle).

7.3.2.6.2.4.1 IF weighted dummy fuel bundle is to be grappled, THEN record its Fuel Pool location:

(also record this Fuel Pool location in step 7.3.2.6.3.2, IF applicable).

7.3.2.6.2.4.2 Lower the Fuel Grapple Hoist.

7.3.2.6.2.4.3 Grapple test weight marked SU270/485WT (OR weighted dummy fuel bundle).

7.3.2.6.2.4.4 Raise the test weight / bundle to the GRAPPLE NORMAL UP position.

WHEN THE REFUELING PLATFORM HOISTS ARE LOADED, DO NOT MOVE CAUTION:

THE PTFORM OVER THE ACTIVE FUEL AREA IN THE REACTOR VESSEL.

7.3.2.6.2.5 Move the Refueling Platform to location defined as NEAR OR OVER THE CORE.

7.3.2.6.2.6 Confirm the selected peripheral rod will NOT withdraw by attempting to withdraw it ONE notch.

7.3.2.6.2.7 Return the Refueling Platform to location defined as NOT NEAR THE CORE.

7.3.2.6.3 WITHDRAW the selected peripheral rod, ONE notch.

MOVE THE REFUELING PLATFORM SLOWLY WHEN APPROACHING THE CAUTION: AREA NEAR THE CORE BECAUSE THE STOP WILL BE ABRUPT WHEN THE BRIDGE POWER IS INTERRUPTED.

7.3.2.6.3.1 Confirm the Refueling Platform STOPS before reaching location defined as NEAR OR OVER THE CORE by moving platform SLOWLY towards that location.

MGR-0001 Rev4

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 37 OF 75 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REFUELING INTERLOCKS AND HOIST LIMIT CHECKS 34SV-F1 5-001-1 15.0 7.3.2.6.3.2 Return the test weight I bundle to its previous storage location in Fuel Pool I Transfer Canal AND Record Fuel Pool location: IF applicable.

7.3.2.6.3.2.1 Lower test weight! bundle UNTIL seated in place.

7.3.2.6.3.2.2 Disengage test weight! bundle from the grapple.

7.3.2.6.3.2.3 Fully RAISE the grapple.

7.3.2.64 Fully INSERT the selected withdrawn control rod.

7.3.2.6.4.1 Move the Refueling Platform to location defined as NEAR OR OVER THE CORE.

7.3.2.6.5 WITHDRAW the selected peripheral control rod, ONE notch.

7.3.2.6.5.1 Confirm the refueling grapple can NOT be lowered by attempting to lower the grapple.

7.3.2.6.6 Fully INSERT the selected withdrawn control rod.

7.3.2.6.6.1 Return the Refueling Platform to location defined as NOT NEAR THE CORE.

7.3.2.6.6.2 Proceed to the next applicable subsection of REFUELING INTERLOCKS FUNCTIONAL TEST OR to the RESTORATION subsection.

7.3.2.7 IF it is desired to SIMULATE control rod withdrawal (Method 2) for the purpose of testing the Refueling Grapple Interlock, THEN proceed to next step:

OTHERWISE, all Method 2 steps in this subsection are NOT applicable.

MGR-0001 Rev4

HLT-07 SRO NRC EXAM

25. 239002A3.02 001 Unit 2 experienced a Group One from 100% power.

Eight (8) SRV amber lights are illuminated and B SRV is currently open with reactor pressure at 875 psig.

Which ONE of the choices below is the LOWEST reactor pressure which must have been reached to cause these conditions?

A. 1080 psig B. ll2Opsig C 1130 psig D. 1140 psig 74

HLT-07 SRO NRC EXAM

==

Description:==

Electrical Opening Setpoints:

(a) Electrical Setpoint 1120 psig: 2B21-FO13B 2B21-FO13D 2B21-FO13F 2B21-FO13G (b) Electrical Setpoint 1130 psig: 2B21F013A 2B21-FO13C 2B21-FO13K 2B21-FO13M (c) Electrical Setpoint 1140 psig: 2B21-FO13E 2B21-FO13H 2B21-FO13L Reactor pressure had to get to at least 1130 psig to cause eight (8) amber lights to illuminate, If pressure gets to 1140 psig, then the remaining three SRVs would open. Applicants get confused on which pressure setpoint opens the SRVs. 1074 psig is the reactor scram setpoint.

The A distractor is plausible if the applicant remembers that 1074 psig is the scram setpoint and remembers this is part of the LLS logic and confuses the number of SRVs which would open if reactor pressure is above the scram setpoint of 1074 psig.

The B distractor is plausible if the applicant remembers that 4 SRVs open at this pressure which will arm LLS and thinks 4 more SRVs open for a total of eight.

The D distractor is plausible if the applicant confuses the number of SRVs that will open and thinks reactor pressure has to exceed 1140 psig before eight SRVs will open.

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

75

HLT-07 SRO NRC EXAM

References:

NONE K/A:

239002 Relief/Safety Valves A3. Ability to monitor automatic operations of the RELIEF/SAFETY VALVES includ ing:

(CFR: 41.7 /45.7)

A3.02 SRV operation on high reactor pressure 4*3* 4*3*

LESSON PLAN/OBJECTIVE:

B21-SLLS-LP-01401, Main Steam & Low Low Set, EU 014.003.A.06 References used to develop this question:

34S0-B21-OO1-2, Automatic Depressurization (ADS) And Low-Low Set (LLS) Systems 76

B21-SLLS-LP-01401-08 Page 3 of 115 MAIN STEAM AN]) LO LO SET

8. Given a simplified drawing or P&ID of the Main Steam Line Drawing, IDENT IFY the following valves: (0l4.012.B.01)
a. 2B21-F016
b. 2B21-F019 2B21-F020
d. 2B21-F038 2B21-F021
9. Given Plant conditions, EVALUATE the conditions, and DETERMINE if any SRV should be open. (014.003.A.06) (200.009.A.04)
10. STATE the specific condition that will cause the niber indicator for a SRV to illuminate.

(014.003.A.02)

11. From a list, SELECT (5) FIVE Main Control Room indications that would indicate an SRV is open. (014.003.A.01)
12. Given plant conditions, EVALUATE those conditions and DETERMINE if a GP I PCIS isolation should have occurred. (014.007 .A.0 1)
  • 13. Given Plant data obtained from 34SV-B21-002-1/2, Main Steam Isolati on Valve Trip Test, EVALUATE the data and DETERMINE the operability of the MSIVs. (014.0 01.A.01)
14. Given RPV pressure is above SRVs setpoints and no SRVs are open, from a list, SELECT the correct operator actions to be taken per 34AB-B21-003-1/2; Failure of Safety

/Relief Valves.

(200.009.A.01)

15. Given RPV Ptessure is below the SRV reset points and one or more SRVs are open, STATE the correct operator action which should be taken per 34AB-B2 1-003-1/2; Failu re of Safety/Relief Valves. (200M09.A.02)
17. Given plant conditions involving the Main Steam/LLS System, DETERMIN E if a Technical Specification Limiting Condition for Operation has been exceeded. (implic it in this objective is a determinatioti ofAPPLICABiLITY and associated NOTES) (300.011 .A. 14)
18. Given plant conditions involving the Main Steam!LLS System, DETERMIN E the Required Action(s) and Completion Time(s) in accordance with Technical Specification s for any combination of INOPERABLE systems, structures or components and the bases for the action(s). (SRO Only) (300.006.A.30)
19. Given plant conditions resulting in ATTS alarm conditions and applicable proced ures, DETERMINE the proper actions that should be taken per: (055.001 .A. 13)
a. LLS Logic A/C (B/D) Armed (34AR-602-123(223)-2/1 / 34AR-602-135(235)

-2/1)

b. LLS Logic A/C (B/D) Power Loss (34AR-602-1 17(217)-2!1 / 34AR-602-12 9(229)-2/l)
c. ECCS/RPS Division 1(11) Trouble (34AR-602-l 10-2 (602-124-1) / 34AR-602-13 0-1/2)

SOUTHERN NUCLEAR PLANT E. I. HATCH PAGE 13 OF 36 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

AUTOMATIC DEPRESSURIZATION (ADS) AND 34S0-B21-001-2 13.13 LOW-LOW SET (LLS) SYSTEMS 7.2.3 Autoni.aic Initiation of LLS LLS logic requires the following TWO initiating conditions:

1) Reactor pressure> 1074 psig.
2) Any of the following safety relief valves has opened has a tailpipe press 85 psig:

(a) Efectrical Setpoint 1120 psig: 2B21-FOI3B 2B21-FOI3D 2B21-FO13F 2B21 -FOl 3G (b) E[ectrical Setpoint 1130 psig: 2B21-FOI3A 2B21-FO13C 2B21 -FOl 3K 2B21 -FOl 3M (c) Electrical Setpoint 1140 psig: 2B21-FO13E 2B21-FO13H 2B21-FOI3L (d) Mechanical Setpoint 1150 psig for ALL SRVs.

Once initieted, LLS will control reactor pressure by cycling 2B21-FO13B, D, F, G at the NOTES: following pressures:

VALVE OPEN CLOSE 2B21-FO13D 1036 psig 890 psig 2B21-FO13F 1027 psig 881 psig 2B21-F013G 1012 psig 866 psig 2B21-FO13B 997 psig 851 psig

  • The RED indicator light for SRV shows ONLY that the actuating solenoid is energiz ed. It is J!QI po[tive indication of valve position.

The RED light will be illuminated only WHEN the electrical signal is present to open the SRV.

  • The GREEN indicator light for SRV will NOT extinguish during LLS Actuation.
  • The YELLOW indicator light for SRV indicates the Tailpipe Pressure switch has reache d its setpoint.
  • Typical tailpipe peak temperatures for normal SRV opening will range from 300 CF to 420 CF.
  • In the event of a reactor vessel flooding incident, the temperature/pressure associa ted with a water environment may NOT actuate the 85 psig pressure switches as norma lly expected with an SRV open demand present. Tailpipe temperatures can range from 220 CF 420 to CF.

Operator action for failure OR apparent failure of SRVs should be to enter 34AB-B21-003-2, Failure Of Safety/Relief Valves.

MGR-0001 Ver. 3

Controlled Copy sj5J Sc ExajJ If found unattended, IMMEDIATELY notify:

Charlie Edmund, Anthony Ball, Ray Rutan or Ed Jones at ext. 3123

HLT-07 SRO NRC EXAM

26. 241000A2.23 001 Unit 2 is operating at 23% reactor power.

The Main Turbine/Generator has just been tied to the grid.

Subsequently, 650-136, Vibration Alarm, is received.

Investigation reveals that Turbine Eccentricity is 3 mils and Vibration is 14 mils and has been at these values for the last 2 minutes.

JAW 34S0-N30-001-2, Main Turbine Operation, which ONE of the choices below completes the following statements?

The is required to be used for monitoring Eccentricity on the Main Turbine.

To CLOSE the Main Turbine Stop Valves, the NPO will on 2111 1-P650 panel.

A. 2N32-R609, TURBINE METAL EXPANSION/TEMP recorder; select the Close Valves button on the Control -Speed Screen B. 2N32-R609, TURBINE METAL EXPANSION/TEMP recorder; depress the Turbine Trip pushbuttons C. Human Machine Interface (HMI) Screen; select the Close Valves button on the Control -Speed Screen D Human Machine Interface (HMI) Screen; depress the Turbine Trip pushbuttons 77

HLT-07 SRO NRC EXAM

==

Description:==

34S0-N30-OOl-2, Main Turbine Operation section 7.1.5, Turbine Roll And Initial Loading, states the requirement for monitoring Eccentricity is on HMI screens Aux Trends Eccent Speed. Using 2N32-R609, TURBiNE METAL EXPANSION/TEMP recorder is plausible because it does monitor turbine vibration and is located on the 2Hl l-P650 panel.

Step 5.2.11 states: IF the following vibration limits are exceeded, the turbine must be shut down:

5.2.11.1 At speeds < 800 rpm, the maximum vibration allowed is 8 mils.

5.2.11.2 At speeds between 800 rpm and 1400 rpm, the vibration limit for any bearing is 10 mils for 2 minutes. The maximum vibration allowed is 14 mils.

5.2.11.3 At speeds> 1400 rpm, the vibration limit for any bearing is 12 mils, unless approved by OPS Management and System Engineer. The turbine must be shutdown WHEN directed by OPS Management OR System Engineer.

Since the vibration/eccentricity values have been in for 2 minutes, the Main Turbine has failed to auto trip and the NPO is required to depress both Turbine Trip pushbuttons. CLOSE VALVES is disabled when either PCB is closed and the generator is tied to the grid.

The A distractor is plausible if the applicant remembers that this recorder does monitor vibration/eccentricity and is located on the P650 panel. The second part is plausible if the applicant confuses that close valves will not function with the generator PCBs closed and does not realize the need to trip due to high vibration.

The B distractor is plausible if the applicant remembers that this recorder does monitor vibration/eccentricity and is located on the P650 panel. The second part is correct.

The C distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses that Close Valves will not function with the generator PCBs closed and does not realize the need to trip due to high vibration.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

References:

78

HLE-07 SRO NRC EXAM NONE K/A:

241000 Reactor/Turbine Pressure Regulating System A2. Ability to (a) predict the impacts of the following on the REACTO1VTURBINE PRESSURE REGULATING SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 /45.6)

A2.23 Turbine high eccentricity 2.6 2.6 LESSON PLAN/OBJECTIVE:

N30-MTA-LP-01701, Main Turbine, E0 017.002.A.O1 References used to develop this question:

34AR-650- 136-2, Vibration Alarm (Main Turbine) 34S0-N30-OO1-2, Main Turbine Operation Modified from HLT Database HLT 5 NRC Exam Q#25 ORIGINAL QUESTION (HLT-5 NRC Exam Q#25)

Unit 2 is operating at 23% reactor power. Operators are in the process of starting up the Main Turbine. The Main Turbine is on turning gear.

Which ONE of the choices below completes the following statements?

JAW 34SO-N30-OO1-2, Main Turbine Operation section 7.1.5, Turbine Roll And Initial Loading , the is required to be used for monitoring Eccentricity on the Main Turbine.

The first speed selected for the Initial Turbine walkdown is Speed Cmd RPM A.V Human Machine Interface (HMI) Screen; LOW (100)

B. Human Machine Interface (HMI) Screen; MED (800)

C. 2N32-R609, TURBINE METAL 1 EXPANSION/TEMP recorder; MED (800) 79

HLT-07 SRO NRC EXAM D. 2N32-R609, TURBINE METAL EXPANSION/TEMP recorder; LOW (100) 80

-MTA-LP-01701-08 Page 4 of 105 MAIN TURBINE

14. GIVEN plant conditions, PREDICT the response of the MAIN TURBINE VIBRA TION MONITORING PANEL on a loss of 125 VDC, 1/2R25-S003. (017.004.D.01)
15. DETERMINE the steps required to engage the Main Turbine turning gear per 34SO-N30-OQ1-1/2, Main Turbine Operation. (017.003.A.01)
  • 17. DETERMINE the Heat Up/C ooldown rate for:
a. 1st Stage shell lower inner surface temperatures
b. Hot Reheat steam to the LP turbines PER 34SO-N30-001-1/2, Main Turbine Operation. (017.002.A.02)
19. DETERMINE the steps required to roll the Main Turbine from 0 - 1800 RPM per 34SO-N30-0O1 1/2, Main Turbine Operation. (017.01 5.A.0 1)
  • 23. DETERMINE the steps required to perform the following tests:
a. Power Load Unbalance Test
b. Turning Gear Oil Pump Auto Start Test
c. Emergency Bearing Oil Pump Auto Start Test
d. Motor Suction Pump Auto Start Test PER 341T-N30-001-1/2, Main Turbine and Auxiliaries Weekly Test respectively.

(017.010.A.0I)

SOUTHERN NUCLEAR PLANT E. I. HATCH PAGE 9 OF 119 DOCUMENT TITLE. DOCUMENT NUMBER: VERSION NO:

MAIN TURBINE OPERATION 34SO-N30-001-2 26.2 5.2.11 IF the fellowing vibration limits are exceeded, the turbine must be shut down:

5.2.11.1 At speeds <800rpm, the maximum vibration allowed is 8 mils.

5.2.11.2 At speeds between 800 rpm and 1400 rpm, the vibration limit for any bearing is 10 mils for 2 minutes.

The maximum vibration allowed is 14 mils.

5.2.11.3 At weeds 1400 rpm, the vibration limit for any bearing is 12 mils, unless approved by OPS Management and System Engineer.

The turbine must be shutdown WHEN directed by OPS Manag ement Q, System Engineer.

5.2.12 The maximum operating temperature limit for any bearing is 250cF.

5.2.13 Motoring of the turbine-generator is NOT allowed.

5.2.14 Do NOT open the generator output breakers before the turbine steam valves are closed, and generator KW output is zero, except on a generator protect ion trip.

5.2.15 The following electrical signals will trip the turbine:

5.2.15.1 High MSR level (10 sec time delay) 3 inches below MSR 5.2.15.2 High vibration (2 of 2 probes for one or more bearings) 12 mils 5.2.15.3 High reactor level 54 inches 5.2.15.4 Primary Overspeed Trip 1989 rpm (110.5%)

5.2.15.5 Emergency Overspeed Trip 2007 rpm (111.5%)

5.2.15.6 Low Main Shaft Oil Pump discharge pressure when above 1300 rpm (.15 sec time delay) 105 psig 5.2.15.7 High thrust bearing wear (.045 sec time delay) 35 mils 5.2.15.8 Low bearing header pressure (.045 sec time delay) 21 psig 5.2.15.9 Low hydraulic fluid (EHC) pump discharge pressure

(.5 sec time delay) 1100 psig 5.2.15.10 Low condenser vacuum (1 sec time delay) 22.3 inches Hg MGR-0001 Rev. 4.0

1.0 IDENTIFICATION

VIBRATION ALARM DEVICE:

2N31-R414 thru 2N31-R432 10 mils (12 mils on 2 of 2 probes on any bearing trips turbine) 2N31-K401, 2N31-K402 Power supply problem

2.0 CONDITION

MAIN TURBINE: 13.0 CLASSIFICATION:

Vibration exceeds 10 mils on any bearing, OR EQUIPMENT STATUS rotor OR shell e$insion in red band, OR 14.0 LOCATION:

a turbine trip on high vibration exceeding 12 mils, OR trouble at 2N31-P003. 2H1 1-P650 Panel 650-1 5.0 OPERATOR ACTIONS:

I 5.1 Determine the source of the annunciator by checking 2N32-K4001A OR 2N32-K4001 B, Turbine Eccentricity and Vibration Screen on Mark VI.

LI 5.2 If none of the devrces listed in step 5.1 indicate that an alarm should have occurred, check for the following at local panel 2N31-P003:

5.2.1 Supplies oK LED not lit on 2N31-K401 and/or 2N31-K402, power supplies.

(Power supply is 2R25-S003, Bkr 31)

LI 5.2.2 OK LED no lit on any of 2N31-R414 through 2N31-R432, Dual Vibration monitors.

LI 5.2.3 ALERT (10 mils) or DANGER (12 mils) LEDs lit on any of Dual Vibration Monitors 2N31-R414 thru 2N31-R432.

LI 5.3 IF vibration on any bearing exceeds 10 mils, perform the following:

5.3.1 Confirm OR set 2P41-R610, Main Turbine Lube Oil Temp controller, to maintain 110°-I 20F.

LI 5.3.2 Notify Engineering to perform an evaluation.

LI 5.3.3 IF water is in the turbine, as indicated by a sharp decrease, indicated on I Control

LI 5.3.4 Contact System Engineering for further instruction.

LI 5.3.5 The Main Turbine High Vibration Trips may be disabled per 34SO-N30-001-2.

LI 5.4 When vibration decreases to below 12(10) mils, reset the DANGER (ALERT) LEDs on 2N31-P003 by depressing the System Monitor Reset pushbuttons.

LI 5.5 IF the Turbine trips, do NOT reset it UNTIL the vibration has been investigated.

LI 5.6 At the direction of the SS, restore the Turbine Vibration Trip to NORMAL per 34SO-N30-001-2.

6.0 CAUSES

LI 6.1 Unbalanced rotof 6.3 Loss of Power 6.5 Coupling misalignment 6.2 Rubbing 6.4 Water in the turbine

7.0 REFERENCES

8.0 TECH. SPECS.ITRM/ODCMIFHA:

7.1 H-52202 l S-21 679, GEK-42234, sect 4542K63-OO1A 7.2 S-32647 7.5 GE Letter Oct 17, 1990 TC #237 N/A Not applicable to this procedure 7.3 S-21678, Turbine Section 34AR-650-1 36-2 10.4 MGR-0048 Ver. 5 AG-MGR-75-1 101

HLT-07 SRO NRC EXAM

27. 259002A4.07 001 Unit 2 is at 60% power removing the B RFPT from service with the following conditions:

o A RFPT is in Auto, controlled from the FW Master Controller 2C32-R600 o B RFPT is in Manual, as indicated below on M/A station, 2C32-R6O1B o NO other switches have been repositioned REP B MIA STATION asT rnIcnnTaIsr With the B RFPT controller in Manual

, which ONE of the choices below completes the t

following statements?

The B RFPT will be used to lower B RFPT flow.

Depressing the B RFPT PF key will A. speed setter switch; lower the B RFPT setpoint by 4 inches B. speed setter switch; change the Green bar display to Controller Output C. output lever; lower the B RFPT setpoint by 4 inches D output lever; change the Green bar display to Controller Output 81

HLT-07 SRO NRC EXAM

==

Description:==

The ouput lever ([<1 pushbutton) is used in MANUAL to control the speedlflow of the RFPT while the Speed setter switch will be used when control has been transferred from M/A to SS.

The PF key will change the display from controller input to controller output.

The A distractor is plausible if the applicant confuses when the Speed Switch will function with when the output level will change speed. The second part is plausible if the applicant confuses the PF key function on the Master Controller with the individual MJA controller.

The B distractor is plausible if the applicant confuses when the Speed Switch will function with when the output level will change speed. The second part is correct.

The C distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses the PF key function on the Master Controller with the individual M/A controller.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

References:

NONE K/A:

259002 Reactor Water Level Control System A4. Ability to manually operate and/or monitor in the control room:

(CFR: 41.7/45.5 to 45.8)

A4.07 All individual component controllers when transferring from automatic to manual mode..

3.8 3.6 LESSON PLAN/OBJECTIVE:

N21-CNDFW-LP-00201, Condensate & Feedwater System, EO 002.004.A.08 82

HLT-07 SRO NRC EXAM References used to develop this question:

34S0-N21-007-2, Condensate and Feedwater System (2C32-R6O1A PF Lamp)

Modified from HLT-5 NRC Exam Q#26 ORIGINAL QUESTION (HLT-5 NRC Exam Q#26)

Unit 1 is at 60% power with the following conditions:

o A RFPT is in Auto, controlled from the FW Master Controller 1C32-R600 o B RFPT is in Manual, controlled from M/A station 1C32-R6O1B, as indicated below Which one of the choices below completes the following two statements LAW 3450-N21-007-1 Condensate & Feedwater System?

Prior to placing the B RFPT controller in Auto, its OUTPUT signal is required to be matched with its INPUT signal (from Master Controller) using the When the B RFPT controller is in Auto, then the will be used to balance flows between the RFPTs.

A. output lever output lever B.V output lever set point keys C. set point keys output lever D. set point keys set point keys 83

N21-CNDFW-LP-00201 Page 4 of 154 CONDENSATE & FEEDWATER SYSTEM Initial License (LT) ENABLING OBJECTIVES

1. Given a list of statements, SELECT the statement that best describes the purpose of the Condensate and Reactor Feedwater System. (002.004.C.01)
2. From a list of pant conditions, SELECT the conditions that must exist prior to placing REP Seal Water in service. (002.004.A.03)
3. Evaluate the effect of the HP or LP Stop valve for the RFPT at various Plant conditions.

(002.004.A.07)

4. Given Plant conditions, DETERMINE whether the Speed Setter or M/A Station is controlling RFI T speed. (002.004.A.08) 3
5. Given the status of IP11-R751, IDENTIFY the parameter being utilized for hotwell level control per 34S0-N21 -007-1. (026.037.A.03)
6. EVALUATE the system response from toggling the IPI1-R751 controller PF key during normal system operation per 34S0-N21-007-1. (026.037.A.06)
7. Given plant conditions, DETERMINE the system response to a loss of 1R25-S064 (Instrument Bus IA) per 34S0-N21-007-1 and 34AB-R25-002-1. (026.037.A.04)
8. Given the status of 1P11-R751, DETERMINE the system response to a processor failure per 34S0-N21-007-1. (026.037.A.05)
9. Given a list of conditions, SELECT the conditions that should result in an auto start of the Condensate Pumps. (026.003.A.04)
10. From a list of conditions, SELECT the conditions which result in a Condensate Pump trip. (026.037A01)
11. Given a list of conditions, SELECT the conditions that should result in an auto start of the Condensate Booster Pumps. (026.003.A.05)
12. From a list of conditions, SELECT the conditions which result in a Condensate Booster pump trip. (026.038.A.01)
13. Given a list of plant conditions, SELECT the indications which would indicate a RFPT is in the tripped condition. (002.006.A.03)
14. Given a list of statements, SELECT the statement that best describes the operation of the Condensate System in Short Cycle Cleanup. (026.030.A.01)
15. Given a list of statements, SELECT the statement that best describes the operation of the Condensate System in Long Cycle Cleanup. (026.029.A.01)

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 75 OF 273 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

CONDENSATE AND FEEDWATER SYSTEM 3450-N21-007-2 50.0 7.1.11 Second Reactor Feed Pump Startup U9

  • The steps of this procedure describe the startup of Reactor Feed Pump 2B. Startup of Reactor Feed Pump 2A is the same UNLESS otherwise noted. REP 2A component numbers will be enclosed in parentheses.

NOTES:

  • Start condensate booster pump(s) as necessary to maintain Reactor Feed Pump suction pressure above the trip setpoint of 207 psig.
  • IF REP has been above approximately 3 mlbm/hr flow in the past six hours, RFPT is considered to be HOT.

Critical 7.1.11.1 IF REP has NOT been above approximately 3.0 x 106 Ibm / hr in the past 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, OR IF RFPT has been at zero speed for greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, LI confirm RFPT 2B(2A) is rotating on Turning Gear OR windmilling AND LI has been for a period of 1-2 hours. LI 7.1.11.2 Confirm RFPT 28 (2A) is in standby per Placinci Reactor Feed Pump Turbines In Standby, subsection. LI 7.1.11.3 Take RFPT 28 (2A) Speed Setter switch to RAISE NQ release.

7.1.11.3. 1 Repeat UNTIL RFPT 28 (2A) turning gear Engaged light EXTINGUISHES OR a speed increase is observed. LI 7.1.11.4 Place RFPT 2B (2A) Turning Gear motor switch to AUTO AND lock. LI 7.1.115 Immediately trip RFPT 28 (2A) AND confirm:

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 76 OF 273 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

CONDENSATE AND FEEDWATER SYSTEM 34SO-N21-007-2 50.0 7.1.11.6 Take RFPT 2B (2A) Reset-Trip control switch to RESET, UNTIL the following occurs:

  • RFPT 2B (2A) HP Stop Valve, Red, Open light ILLUMINATED, Green, Closed light EXTINGUISHED El
  • RFPT 2B (2A) LP Stop Valve, Red, Open light ILLUMINATED, Green, Closed light EXTINGUISHED El
  • RFPT 2B (2A) TRIP Annunciator 650-325 (326) CLEARS El 7.1.11.7 Slowly raise RFPT 2B (2A) Speed Setter to increase turbine speed AND maintain at less than OR equal to 1000 rpm UNTIL lube oil temperature is at least 1 10F as indicated by 2P41-R606 (R602), RFPT 2A (2B) Oil Temp controller.

7.1.11.8 Close the following RFPT 2B (2A) drains, panel 2H21-P244:

7.1.11.10 Confirm the RFP B (A) M/A Station output signal is tracking actual RFPT 2B(2A) speed. El 7.1.11.11 Place the RFPT 2B (2A) TMR Mode Switch to M/A.,

confirm M/A Station (Green) light ILLUMINATES. El MGR-0001 Ver. 4

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 77 OF 273 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

CONDENSATE AND FEEDWATER SYSTEM 34S0-N21-007-2 50.0

. WHEN Feedwater flow is established to Reactor Vessel from RFP being started, the running RFP speed will decrease AUTOMATICALLY to maintain Reactor Water Level Setpoint.

NOTES:

. WHEN 2N21-F117B (2N21-F117A), RFP2B(2A), Mm FlowAOV, CLOSES, an increase in feedwater flow will occur causing a reactor water level increase.

7.1.11.12 Slowly change RFPT 2B (2A) speed UNTIL the REP being started NQ the running REP flow MATCH. LI

. On 2C32-R6OIA, REP MIA Station, WHEN the lamp is LIT, the vertical bar graph display the digital display both indicate the controller output.

NOTES: DEPRESSING the PF key will switch OFF the lamp, the controller input will THEN be displayed on both the vertical bar graph AND the digital display.

. To monitor the input signal, use the inservice REP M/A Station with the FE lamp OFF.

7.1.11.13 Match the input AND the output of Pump B (A) M/A Station, by performing the following:

7.1.11.13.1 Depress the PE key read the controller output (FE lamp LIT). LI 7.1.11.13.2 THEN depress the PE key so the input to the controller is displayed.

(FE lamp is OFF) LI 7.1.11 .13.3 Carefully adjust the manual output lever UNTIL the input output are MATCHED, panel 2H1 1-P603. LI 7.1.11.14 Place REP B (A) M/AStation in automatic by depressing the A pushbutton UNTIL it ILLUMINATES 7.1.11.15 As required, adjust REP A (B) Speed Control Bias Setting to maintain RFPT 2ANQ 2B speed WITHIN 100 rpm. LI MGR-0001 Ver. 4

HLT-07 SRO NRC EXAM

28. 261000A1.01 001 Unit 2 is operating at 100% RTP.

At 1000, an event occurs resulting in Unit 2 Drywell pressure increasing to and stabilizing at 2.5 psig.

Which ONE of the choices below completes the following statements?

At 1015, with NO operator actions, the TOTAL amount of Unit 2 SBGT System flow going to the Main Stack is expected to be between SCFM.

Unit 2 SBGT flow can be monitored on panels 2H1 1-P654 &

A. 3000 and 4000; 2H1 1-P657 B. 3000 and 4000; 2H1 1-P700 0 6000 and 8000; 2H1 1-P657 D. 6000 and 8000; 2H1 1-P700

Description:

Any of the following signals for Unit 1 or Unit 2 will initiate all four SBGT Trains.

Unit 1 or 2 Reactor Zone exhaust high radiation:

Unit 1: 18 mrenilhr on 1D1 l-K609 A-D OR Unit 2: 18 mrem/hr on 2D1 l-K609 A-D Unit 1 or 2 Refueling Zone exhaust high radiation:

Unit 1 18 mrem/hr on 1D11-K611-A-D OR Unit 2 18 mrem/hr on 2D1 lK61 1 A-D, OR 6.9 mremlhr on 2D1 1-K634 A-D, OR 5.7 mremlhr on 2D1 1-K635 A-D High Drywell pressure (1.85#)

Low Reactor water level (-35).

IF low flow conditions exist concurrently with an automatic initiation signal, THEN SBGT Fan 84

HLT-07 SRO NRC EXAM 1T46-COO1A After 6 minutes of low flow 1T46-COO1B After 4 minutes of low flow Once the train is shutdown, it is in a standby logic condition. IF the operating train tripped WHILE an automatic initiation signal exists, THEN the shutdown (STBY) train will automatically start.

Normal SBGT flow is 3000 to 4000 SCFm and is monitored on Panels 2H11-P654 & P657.

The A distractor is plausible if the applicant confuses that Unit 1 SBGT will shutdown on a low flow condition after a time delay with U2 and thinks that one of the U2 SBGT fans has automatically shutdown resulting in only 3000 and 4000 SCFM from one U2 SBGT fan. The second part is correct.

The B distractor is plausible if the applicant confuses that Unit 1 SBGT will shutdown on a low flow condition after a time delay with U2 and thinks that one of the U2 SBGT fans has automatically shutdown resulting in only 3000 and 4000 SCFM from one U2 SBGT fan. The second part is plausible if the applicant remembers that the 2H1 l-P700 panel is where SBGT dP is indicated and confuses this with SBGT flow.

The D distractor is plausible since the first part is correct. The second part is plausible if the applicant remembers that the 2H1 i-P700 panel is where SBGT filter dP is indicated and confuses this with SBGT flow.

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

85

HLT-07 SRO NRC EXAM

References:

NONE K/A:

261000 Standby Gas Treatment System Al. Ability to predict and/or monitor changes in parameters associated with operating the STANDBY GAS TREATMENT SYSTEM controls including: (CFR: 41.5/45.5)

A1.O1 System flow 9 3.1 LESSON PLAN/OBJECTIVE:

T46-SBGT-LP-03001, Standby Gas Treatment System, EO 030.006.A.O1 References used to develop this question:

34S0-T46-OO1-1, Standby Gas Treatment System 34S0-T46-OO1-2, Standby Gas Treatment System PDMS 2T41-R618 on P657 86

T46-SBGT-LP-03001 Page 4 of 59 STANDBY GAS TREATMENT SYSTEM

13. EXPLAIN the Standby Gas Treatment (SBGT) Filter Train and RB/RF Ventilation Systems responses to each of the following initiations: (030.00 1.A.03)
a. Unit 1 Reactor Building Exhaust Ventilation High Radiation, and
b. Unit 2 High Drywell Pressure.
14. In accordance with 34S0-T46-001-l/2, Standby Gas Treatment System, and plant conditions, DETERMINE if Unit 1 andlor Unit 2 SEOT fans should have auto started. (030.006.A.Ol)
  • 15. In accordance With plant procedures and SBGT system parameters, DETERMINE if the SBGT system heaters should be operating. (030.002.A.09)
16. In accordance with plant procedures and SBGT system parameters, DETERMINE if the SBGT fans should have automatically tripped. (030.002.A.l0)
  • 17 In accordance with 34S0-T46-001--1/2, Standby Gas Treatment System, DESCRIBE the startup of the Standby Gas Treatment (SBGT) System for any mode of operation by describing the sequence of major component/subsystem startup. (030.002.A.07)
  • 18. Given the following Standby Gas Treatment (SBGT) System annunciators/alarms, EXPLAIN the significance of each: (030.001.A.09)
a. RIB Inside to Outside Air Diff Press Low
b. Refuel Floor Outside Air Diff Press Low
c. SBGT Standby System Running
d. SBGT Fltr Diff Press High
e. SBGT Fan 1A (B) Shutdown or Low Flow
f. SBGT Train 1A or lB Failure/Shutdown
19. In accordance with plant procedures, DESCRIBE the basic process for starting up the Standby Gas Treatment (SBGT) System from outside the Control Room. (030.007.A.01)
  • 20. In accordance with plant procedures, DESCRIBE the actions necessary to operate SBGT Dampers manually from outside the Control Room. (030.009.A.01)
  • 21. In accordance with the below listed surveillance procedure(s), DESCRIBE the purpose(s) of the surveillance and the reason(s) for included notes and cautions. (030.004.A.0l)
a. 34SV-T46-001-l/2, Standby Gas Treatment System Operability,
b. 34SV-T46-002-1/2, Standby Gas Treatment System Damper Operability, and
c. 34SV-T46-0031/2, Standby Gas Treatment Ventilation and Operability.

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 11 OF 30 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

STANDBY GAS TREATMENT SYSTEM 34SO-T46-001-1 20.13 7.2 SYSTEM ST...ARTUP AND OPERATION 7.2.1 Autonilic Startup NUS The SBGT System will automatically initiate on the following signals:

1. Unit On& OR Two high drywell pressure, 1 .85 psig
2. Unit One OR Two low vessel level, -35 inches NOTE: 3. Unit One OR Two Refueling Floor high radiation (See 6401-CAL-002-0, Process Monitors ARMs Ventilation Monitors Setpoints, for setpoint.)
4. Unit One OR Two Reactor Building high radiation (See 64C1-CAL-002-0, Process Monitors ARMs Ventilation Monitors Setpoints, for setpoint.)

. jf low ffow conditions exist concurrently with an automatic initiation signal, THEN SBGT Fan 1A OR I B will automatically shutdown.

. 1T4-C001A After 6 minutes

. IT4-C001B -After 4 minutes

  • Once the train is shutdown, it is in a standby logic condition.

NOTE: IF the operating train tripped WHILE an automatic initiation signal exists, THEN the shutdown (STBY) train will automatically start.

  • Placing the SHUTDOWN fan control switch to STANDBY is NOT necessary, but can be performed per step 7.2.1.3, W deemed necessary.
  • Therefore, IF a low flow interlock condition exists, THEN die SBGT Fan will be SHUTDOWN AND its associated 1T46 dampers, CLOSE.

7.2.1 i Upon automatic initiation, corrfirm the following actions:

7.2.1.1.1 1T46-COOIA, SBGT Fan, STARTS. LI 7.2.1.1.2 SBGT Filter 1T46-DOO1A Heater is ON.

7.2.1.1.3 1T46-FOO2A, SBGT Fan 1A Disch, OPENS.

7.2.1.1.4 1T46-FOOIA, SBGT Filter Inlet, OPENS. LI 7.2.1.1.5 1T41-FO4OA, Refuel FIr Isol Dmprto SBGT, OPENS.

7.2.1.1.6 1T41-F032A, Rx Bldg Isol Dmprto SBGT, OPENS.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANTE.I.HATCH 11OF28 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

STANDBY GAS TREATMENT SYSTEM 34S0-T46-OO1-2 14.11 7.2 SYSTEM STARTUP AND OPERATION 7.2.1 Automatic Startup The SBGT System will automatically initiate on the following signals:

. Unit One OR Two high drywell pressure, 1.85 psig

. Unit One OR Two low vessel level, -35 inches NOTE:

. Unit One OR Two Refueling Floor high radiation (See 64C1-CAL-002-O, ARMs, Vhtilation Monitors and Process Monitors Setpoints.)

. Unit One OR Two Reactor Building high radiation (See 64Cl-CAL-002-O, ARMs, Ventilation Monitors and Process Monitors Setpoints.)

7.2.1.1 Upon AUTOMATIC initiation, confirm the following actions:

72.1.1.1 2T46-DOOIA and 2T46-DOO1B, SBGT A and B Fan/Filters, START. LI 7.2.1.t2 SBGTAandBHTRisON. El 7.2.1.1 .3 2T46-FOO1A and 2T46-FOO1 B, SBGT A and B Fltr Inlets From Rx Bldg, OPEN El AND 2T46-FOO3A and 2T46-FOO3B, SBGT A AND B Fltr Inlets From Refuel FIr, OPEN. LI 7.2.1.1.4 2T46-FOO2A and 2T46-FOQ2B, SBGT A AND B Fltr Disch dampers, OPEN. El 7.2.1.2 Standby Gas Treatment System Flow increases to 3.0-4.0 KCFM, as indicated on 2T41-R618 and 2U41-R600, SBGT A and B Flow To Main Stack. LI 7.2.1.3 Cotifirm Unit One Standby Gas Treatment System actions per the Automatic Startup subsection of 34SO-T46-001-1.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 12 OF 28 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

STANDBY GAS TREATMENT SYSTEM 34SO-T46-001-2 14.11

  • Normally only one train of SBGT is required for design operation.

. Operation with both trains of SBGT will increase offsite release rates AND deplete both NOTE: charcoal filter trains.

. Operate both trains of SBGT only WHEN required per another procedure OR IF required to maintain adequate negative pressure.

7.2.1.4 Jf. bcsth trains of SBGT have started are operating properly &II operation of both trains is IIQI required for maintaining> 0.20 inch water vacuum in LYrit I AND/OR Unit 2 Secondary Containment, THEN place one train of SBGT in STANDBY as follows:

7.2.1.4.1 Place either SBGT A or SBGT B Filter 2T46-DOO1A or 2T46D001 B, control switch, n the OFF position. El 7.2.1.4.2 WHEN 2T46-DOO1A or 2T46-D001 B, SBGT A(B) Filter, flow has decreased to zero, as indicated on 2T41-R618 /2U41-R600, SBGTA(B) Flow to Main Stack, place 2T46-DOOIA or 2T46-D001 B, SBGT A(B) Fan/Filter, control switch,

[n the STBY position. LI MGR-0001 Ver. 3

04/24/201211:0 E. I. Hatch Nude nt Paç 1 Component Module: 2 IN 2T41R618 301312) [Summary**]

Component Unit CT Component ID Sub ID Status 2 IN Instrument 2T4lR6l8 As Built Description Description DCD User ID Time Stamp REF FIR SUPPLY FAN (COO2A/B) FLOW x2foltz 06/03/2005 11:02 Remark: x2foltz 06/03/2005 11:02 Detail Type Value Description DCD User ID Time Stamp Instrument Type INSTRUMENT SETPOINTS Instrument Setpoint Data x2foltz 06/03/2005 11:02 Document Type Document Sheet Unit Fm/To Coordinate DCD User ID Time Stamp Quantity Quantity Type Quantity DCD User ID Time Stamr, Note NT Seq Note DCD User ID Time Stamp Sub-Details Detail Type Detail Value Column Label Column Value DCD User ID Time Stamp IT INSTRUMENT SETPOINTCalculation x2foltz 06/03/2005 11:02 IT INSTRUMENT SETPOINTProcedure x2foltz 06/03/2005 11:02 IT INSTRUMENT SETPOINTPr0ce5s 13000.0 CFM x2toltz 06/03/2005 11:02 IT INSTRUMENT 5ETPOINTSignal x2foltz 06/03/2005 11:02 IT INSTRUMENT SETPOINTDirection DEC x2foltz 06/03/2005 11:02 IT INSTRUMENT SETPOINTTolerance Process 200.0 x2foltz 06/03/2005 11:02 IT INSTRUMENT 5ETPOINTTolerance Signal x2foltz 06/03/2005 11:02 IT INSTRUMENT SETPOINTCODE x2toltz 06/03/2005 11:02 IT INSTRUMENT SETPOINTInput From PROCESS x2foltz 06/03/2005 11:02 IT INSTRUMENT SETPOINTOutput To Nh P657 x2foltz 06/03/2005 11:02 IT INSTRUMENT SETPOINTHead Correction x2foltz 06/03/2005 11:02 For internal E. I. Hatch Nuclear Plant use only.

HLT-07 SRO NRC EXAM

29. 261000K3.02 001 Unit 2 is operating at 100% RTP with the following conditions:

o Unit 2 Refueling Hatch installed o 2A Standby Gas Treatment (SBGT) Fan is Danger Tagged out for maintenance Subsequently, the following occurs at the listed times:

10:00 A RWCU System break in the Unit 2 Reactor Building 10:05 2D1 l-K609A-D, Rx. Bldg. Contaminated Area Radiation increase to 20 mr/hr 10:10 The Supply breaker for 2R24-S012, 600V MCC, trips OPEN With NO operator action, which ONE of the choices below predicts how the Off-site Radioactive Release Rates will be affected?

At 10:08, the Reactor Building Stack release rate will be than at 10:04.

At 10:15, the Main Stack release rate will be at 10:08.

A. higher; approximately the same as B. higher; lower than C. lower; approximately the same as D lower; lower than 87

HLLO7 SRO NRC EXAM

==

Description:==

With the K609s exceeding their trip setpoints, at 10:05, the Rx Bldg & RF supply and exhaust fans will shutdown and isolate. SBGT fan 2B will auto start and draw a vacuum on the U2 Rx Bldg. With the U2 RF hatch installed, the Ui SBGT fans will have NO affect on U2 Rx. Bldg.

pressure. At 10:10, when 2R24S012 de-energizes, the 2B SBGT fan will stop resulting in SBGT flow decreasing to zero.

At 10:08, ALL Rx. Bldg. and RJF Ventilation will be secured and isolated. SBGT will be discharging into the Main Stack. SBGT flow is less than the combined Rx. Bldg. & R/F flow, therefore at 10:08, the Rx. Bldg. Stack release rate will be lower than before the isolation, since the Rx. Bldg. flow has stopped discharging into the Rx. Bldg. Stack.

The A distractor is plausible if the applicant does not remember the isolation setpoint and thinks Normal Ventilation is still in service discharging the untreated Rx. Bldg. atmosphere to the Rx. Bldg. Stack, causing the release rate to be higher. The second part is plausible if the applicant thinks that 2B SBGT fan is still running and maintaining the previously established flow to the Main Stack.

The B distractor is plausible if the applicant does not remember the isolation setpoint and thinks Normal Ventilation is still in service discharging the untreated Rx. Bldg. atmosphere to the Rx. Bldg. Stack, causing the release rate to be higher. The second part is correct.

The C distractor is plausible since the first part is correct. The second part is plausible if the applicant thinks that 2B SBGT fan is still running and maintaining the previously established flow to the Main Stack.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

References:

NONE K/A:

261009 Standby Gas Treatment System K3. Knowledge of the effect that a loss or malfunction of the STANDBY GAS 88

HLT-07 SRO NRC EXAM TREATMENT SYSTEM will have on following: (CFR: 41.7/45.6)

K3.02 Off-site release rate 3.6 3.9 LESSON PLAN/OBJECTIVE:

T41-SC HVAC-LP-01303, Secondary Containment HVAC Systems EU 037.01 l.A. 12 EOP-SCRR-LP-20325, Secondary Containment Radioactive Release Control, EU 201.080.A.02 References used to develop this cjuestion:

3450-T46-001-2, Standby Gas Treatment System (Att. 2, 2R24-S0l2) 34AB-T22-003-2, Secondary Containment Control (SBGT Auto Starts)

Unit Two FSAR 6.2.4.1.1 SBGT Safety Design Bases (Limits release)

Modified from HLT-4 NRC Exam 2009-301 Q#58 ORIGINAL QUESTION (HLT-4 NRC Exam Q#58)

Unit 2 is operating at 100% power.

o 2T41-COO7B, Rx Bldg Vent Exhaust Fan, is tagged out o 2T41-COO7A, Rx Bldg Vent Exhaust Fan, Trips With no operator action, which ONE of the following predicts how the Rx. Bldg differential pressure (DP) and Rx. Bldg monitored radioactive release rate will be affected?

The Rx. Bldg negative pressure will be (1) before the fan tripped.

The Rx. Bldg stack release rate will be (2) before the fan tripped A. (l)thesameas (2) higher than B. (1) the same as (2) lower than C. (l)less than (2) higher than D.V (1) less than (2) lower than 89

T41-SC HVAC-LP-01303 Page 8 of 86 Secondary Containment HVAC Systems

11. Given a list of statements, IDENTIFY the statement which best describes the plant response to an isolation signal being received by the following ventilation zones: (037.01 1.A.12, 037.012.A.09, 037.022.AJ2, 037.023.A.09)
a. Unit 2 1eactor Zone
b. Unit 1 Reactor Zone
c. Unit 2 1efueling Zone
d. Unit 1 Refueling Zone
12. Given a list of statements, IDENTIFY the statement which best describes the normal lineup of the Equipment Area Coolers. (037.003.A.05, 037.004.A.02, 037.005.A.04)
13. Given a list of statements, IDENTIFY the statement which best describes the trips and automatic start signals associated with the Safeguard Equipment Cooling (SEC) coolers.

(037.003.A.06, 037.004.A.03, 037.005.A.05)

14. Given a list of statements, IDENTIFY the statement which best describes the significance of receiving the SEC Auto Initiation Signal Present annunciator on P650. (037.002.A.0l)
15. Given a list of statements, IDENTIFY the statement which best describes how to reset the SEC coolers once they have started due to an automatic actuation signal. (037.002.A.03)
16. Given plant conditions involving Secondary Containment HVAC, DETERMINE if a Technical Specification Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (300.01 0.C.0 1)
17. Given plant conditions involving Secondary Containment HVAC, DETERMINE if a Technical Requirements Manual (TRM) Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (300.010.C.02)
18. Given plant conditions involving Secondary Containment HVAC, DETERMINE the Required Action(s) and Completion Time(s) in accordance with Technical Specifications for any combination of INOPERABLE systems, structures or components and the bases for the action(s). (SRO Only) (300.006.C.02)

Initial License JJ LEARNING OBJECTIVES

1. Given appropriate references and SECONDARY CONTAINMENT system configuration, ANALYZE the effects of a SECONDARY CONTAINMENT system component malfunction.

(H-OP-90000.00 1)

2. Given appropriate references and SECONDARY CONTAINMENT system configuration, DIAGNOSE the SECONDARY CONTAINMENT system response to component misalignment. (H-OP-90000.002)

EOP-SCRR-LP-20325 Page 6 of 45 SECONDA1Y CONTAINMENT I RADIOACTIVITY RELEASE CONTROL License Continuing Training (LCT) ENABLING OBJECTIVES

1. Given plant conditions, RECOGNIZE an EOP entry condition(s) and ENTER the appropriate EOP flow chart. (201.093.A.01)
2. Given a list, IDENTIFY the statement that describes the purpose of confirming reactor building HVAC isolation and SBGT initiation when reactor building exhaust exceeds 9.5 mR/hr. (201.080.A.02)
3. Given a list, IDENTIFY the statement that describes the purpose of confirming refueling floor and reactor building HVAC isolations and SBGT system initiations when refueling floor exhaust radiation levels exceed their Maximum Normal Operating Levels.

(201 .080.B.05)

4. Given a list, IDENTIFY the statement that describes the purpose of using HVAC systems as the first method of attempting to control secondary containment area temperatures.

(201 .079.A.05, 201.081 .A.0l, 201.081 .B.01)

5. Given plant conditions, including a refueling floor HVAC isolation, IDENTIFY the signal that caused the isolation to occur. (201.081 .A.03)
6. Given plant conditions, including a secondary containment HVAC isolation, IDENTIFY the signal that caused the isolation to occur. (201.081.B.03)
7. Given 3lEO-EOP-014-2 SC Secondary Containment Control, DETERMiNE the following values: (201.079.A.16 & 201.080.B.02)
a. Refueling Floor Exhaust Maximum Normal Operating Radiation Levels.
b. Reactor Building Exhaust Maximum Normal Operating Radiation Level.
8. Given 3 1EO-EOP-014-2, SC Secondary Containment Control, and process radiation monitor readings, DETERMiNE if the following HVAC exhaust radiation levels are above their Maximum N***ormal Operating Values. (20l.079.A.18 & 201.080.B.04)
a. Refueling Floor exhaust.
b. Reactor Building exhaust.
9. Given an area and 31E0-EOP-014-2 SC Secondary Containment Control, DETERMINE the following values for that area: (201.079.A.02)
a. Maximum Normal Operating Temperature.
b. Maximum Safe Operating Temperature.
10. Given area temperatures and 3 1EO-EOP-0 14-2, SC Secondary Containment Control, DETERMINE if area temperatures are above their Maximum Normal Operating Values.

(201 .079.A.04)

SNC PLANT E. I. HATCH Pg 21 of 28 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

STANDBY GAS TREATMENT SYSTEM 3450-T46-OO1-2 14.11 ATTACHMENT Att. Pg.

TITLE: SBGT SYSTEM ELECTRICAL LINEUP 2 of 2 NUMBER DESCRIPTION CHECKED VERIFIED 120/240V Vital AC Power Cab 2A 2R25-S063 13OTGT12 Breaker 15 OFFGAS Post Treatment Sample (Logic Supply B) CLOSED 120/208V Distr. Cab 2B lnstr. Bus 2A 2R25-S064 13OTGT13 CNTL Room, Rx Bldg. Drywell/Torus Ventilation Breaker 19 CLOSED (Logic/Instrument Supply A) 600/208V MCC -2C -ESS Div 1 2R24-SO11 I3ORFR14 Frame 4DR Standby Gas Filter Train 2T46-DOO1A CLOSED 6001208V MCC 2B - ESS Div 2 2R24-S012 130RFR24 Frame 1 5BR Standby Gas Filter Train 2T46-DOO1 B CLOSED 125 VDC Cabinet 2A 2R25-SOO1 13OTDT13 Breaker 32 SBGT Heat Detector and Water Spray (Division I) CLOSED 125 VDC Cabinet 2B 2R25-S002 Breaker 32 SBGT Heat Detector & Water Spray (Division II) CLOSED 120/208V Distr. Cab 2C lnstr. Bus 2B R25-S065 13OTGT12 Control Room, Rx Bldg. Drywell/Torus Ventilation (ESF Div Breaker 15 CLOSED II Instrument Supply)

OPS-0401 Ver. 0.2 G16.030 MGR-0009 Ver. 4

H NP-2-FSAR-6 6.2.4.1 Design Bases 6.2.4.1.1 Safety Design Bases The SGTS is designed to:

  • Limit the release of radioactivity to the environment following a DRA or leakage of radioactivity into the secondary containment or fuel handling area.
  • Ensure leakage into the secondary containment from the outside by maintaining at least a 0.20-in, water negative pressure.
  • Ensure all discharge from the primary or secondary containment or fuel handling area is through an elevated release; i.e., the main stack.
  • Meet the following requirements of an ESF system:

- Quality group classifications (B and C) in accordance with Regulatory Guide (RG) 1.26 (Revision 1, September 1974).

- Seismic Category I requirements of RG 1.29 (Revision 1, August 1973).

- Sufficient redundancy and separation to meet single-failure criteria and the requirements of IEEE-279-1971.

- Ability to obtain power from the essential ac power system upon loss of normal ac power (offsite power).

- Capability for periodic testing and inspection of principal system components.

- Applicable quality assurance requirements.

6.2.4.1.2 Power Generation Design Basis The SGTS also serves as a vent path for the manually initiated normal operation purging or venting of the primary containment. In the event any radioactivity exists in the primary containment at the time of purging or venting, the SGTS filters reduce the amount of radioactivity released.

6.2-45 REV 27 10/09

SOUTHERN NUCLEAR PLANTE. I. HATCH P A GE 2 0 F 3 5 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

SECONDARY CONTAINMENT CONTROL 34AB-T22-003-2 3.13 2O AUTOMATIC ACTIONS 2.1 HPCI, RCIC AND RWCU ISOLATE on high ambient/differential temperatures.

See Attachments 2, 3, 4, and 9.

2.2 High radiation in the Refueling Floor OR Reactor Building Exhaust causes various group 2 isolation valves to close along with isolation of the Unit 1 and Unit 2 Refueling Floor ffl Reactor Building vent fans Ni2 dampers. See Attachments 6, 7, and 10.

2.3 High levels in the Reactor Building Floor Drain Sumps cause the Instrument Sump Isolation valves to CLOSE. See Attachment 8.

3.0 IMMEDIATE OPERATOR ACTIONS NONE MGR-0001 Rev. 4.0

SOUTHERN NUCLEAR PLANT E. I. HATCH PAGE 3 OF 35 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

SECONDARY CONTAINMENT CONTROL 34AB-T22-003-2 3.13 4.0 SUBSEQUENT OPERATOR ACTIONS 4.1 Monitor Secondary Containment temperatures, pressures, radiation levels, AND sump levels.

4.2 IF at any time while performing this procedure, any of the following Secondary Containment parameters exceeds its Maximum Normal Operating value in any area, enter 31EO-EOP-014-2, SC/RR Secondary Containment/Radioactivity Release Control:

  • area ambient temperature (Attachment 2, 3, or 9)
  • area differential temperature (Attachment 2, 3, or 9)
  • differential pressure (Attachment 5)
  • area radiation (Attachment 6 or 10)
  • HVAC exhaust radiation (Attachment 6 or 10)
  • area water level (Attachment 8) 4.3 IF an ambient temperature AND/OR differential temperature alarm is received, perform the following:

4.3.1 At panel 2H1 1-P614, on 2G31-R604 or 2G31-R608, determine which sensor/area is in alarm.

432 Monitor Reactor Building N.P Refueling Floor to outside air differential pressures on 2T46-R604A & 2T46-R604B, instruments, panel 2H1 1-P700.

4.3.3 Operate available area coolers.

4.3.4 IF Secondary Containment HVAC Exhaust Radiation level is below the Secondary Containment HVAC isolation setpoints (see Attachment 6 or 10),

operate available secondary containment HVAC.

4.4 IF a Secondary Containment process radiation monitor alarm is received, perform the following:

4.4.1 At panels 2H1 1-P606, 2H1 1-P645, OR SPDS, determine/monitor actual radiation levels, including 2D11-R619, Stack Radiation Monitor.

4.4.2 IF Secondary Containment HVAC Exhaust Radiation level exceeds the Secondary Containment HVAC isolation setpoint (see Attachment 6 or 10), perform the following:

4.4.2.1 Confirm OR manually initiate isolation of Secondary Containment HVAC per Attachment 7.

4A.2.2 Confirm initiation of OR manually initiate SBGT per 34S0-T46-001-1/2, Standby Gas Treatment System.

MGR-0001 Rev. 4.0

SOUTHERN NUCLEAR PLANT E.I. HATCH PAGE 13 OF 35 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

SECONDARY CONTAINMENT CONTROL 34AB-T22-003-2 3.13 ATTACHMENT 6 ATTACHMENT PAGE:

TITLE: SECONDARY CONTAINMENT OPERATING RADIATION LEVELS 1 OF 4 MAX NORMAL HVAC EXHAUST RADIATION ANNUNCIATORS OPERATING ON2H11-P601 VALUE mR/hr HI-HI REACTOR BUILDING (RX) RADIATION ANNUNCIATOR

- RX BLDG POT CONTAM AREA RADIATION (2D11-K609 A-D) 18 REFUELING FLOOR

- REFUELING FLOOR VENT EXHAUST RADIATION (2D11-K611 A-D) 18

- REFUELING FLOOR VENT EXHAUST RADIATION (2D11-K634A-D) 6.9

- REFUELING FLOOR VENT EXHAUST RADIATION (2D1 1-K635 A-D) 5.7 MAX NORMAL MAX SAFE AREA RADIATION MONITORS OPERATING OPERATING ON 2H11-P600, 2D21-P600 VALUE VALUE mR/hr mR/hr REFUEL FLOOR AREA

1. Reactor head laydown area (2D21-K6O1A) 50 1000
2. Dryer separator pool (2D21-K6O1E) 50 1000
3. Spent Fuel Pool & New Fuel Storage (2D21-K6OIM) 50 1000
4. Reactor Vessel Refueling Floor (2D21-K611K) 50 1000
5. Reactor Vessel Refueling Floor (2D21-K611L) 50 1000 203 ELEVATION AREA (EAST)
6. CRD repair area (2D21-K6O1T) 50 1000 203 ELEVATION AREA (WEST)
7. HVAC Room West El. 203 (2D21-K600D) 50 100 MGR-0009 Rev, 5.0

HLT-07 SRO NRC EXAM

30. 262001G2.4.47 001 A LOSP has occurred on Unit 1.

The operator notices the following indications on panel 1H1 l-P652 for 416OVAC 1E bus and its associated Diesel Generator (DG):

o LOSP 86 lockout TRIPPED o Normal and Alternate supply breakers OPEN o Auto Start System Operative light EXTINGUISHED BLUE LIGHTS WHITE LIGHTS (EXTINGUISHED) (ILLUMINATED)

[ VOLT SELECT

] VOLT SELECT r VOLTMETER SELECT STARTUP AUX XFMR Which ONE of the choices below completes the following statement?

The DG associated with 416OVAC 1E bus is A running, but is NOT tied to 4160 VAC 1E bus B. running and IS tied to 4160 VAC 1E bus C. NOT running, but can be manually started from the control room D. NOT running and can NOT be started from the control room 90

HLT-07 SRO NRC EXAM

==

Description:==

Edwin, this was question 4 of 10 that you have already reviewed. Any discussed changes have been incorporated.

In order for an DG to auto start and auto tie the following must occur:

1. LOSP 86 lockout tripped
2. Proper DG frequency & voltage
3. NO fault on bus
4. Normal Supply breaker open
5. Alternate Supply breaker open In this question all conditions are met for the DG to start and tie to the 4160 VAC lE. The DG is running as indicated by the White 1A DG pot lights being illuminated. The DG is NOT tied to 4160 VAC 1E as indicated by the extinquished blue 4160 VAC 1E bus pot lights but can be manually tied since no other alarms are indicating a fault on the bus.

The B distractor is plausible since the first part is correct. The second part is plausible if the applicant diagnosis that the emergency bus is energized rather than de-energized based on confusing the available indications.

The C distractor is plausible if the applicant diagnosis that the DG is not running rather than running based on confusing the available indications. The second part is plausible since the applicant thinks since a DG is not running that it could be started from the control room or locally.

The D distractor is plausible if the applicant diagnosis that the DG is not running rather than running based on confusing the available indications. The second part is plausible since the applicant thinks a DG is not running that it could not be started from the control room and must be started locally.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

91

HLT-07 SRO NRC EXAM

References:

NONE K/A:

262001 A.C. Electrical Distribution 2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. (CFR: 41.10/43.5 /45.12). 4.2 4.2 LESSON PLAN/OBJECTIVE:

R22-ELECT-LP-02702, 4160 VAC, EO 027 .009.A.03 References used to develop this luestion:

34AB-R22-002- 1, Loss of 4160 VAC Emergency Bus 34AB-R43-00 1-1, Diesel Generator Recovery 92

R22-ELECT-LP-02702 Page 5 of 97 4160 VAC

  • 16. Given plant conditions and 34S0-E22-001-1/2, 4160 VAC System, PREDICT the 4160 VAC System response to the following: (027.009.A.03) a) Main Generator trip b) Fau1t on a 4160 VAC Bus

ç) Loss of ncirmal power to a 4160 VAC Bus.

  • 17 Given 34S0-R22-001-1/2, 4160 VAC System and 34AB-R22-002-1J2, Loss of 41 60 V Emergency Bus, DETERMiNE the major steps for the following 4160 VAC breaker operations: (027.008.A.04, 027.011 .A.0 1) a) Transfer 4160 Volt Bus power supplies.

b) Energize a de-energized 4160 VAC Bus.

c) De-energize an energized 4160 VAC Bus.

d) Restore off-site power to an emergency Bus.

18. Given plant couditions involving the 4160 VAC System, DETERMINE if a Technical Specification Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (300.006.A. 11)
  • 19. Given plant conditions involving the 4160 VAC System, DETERMINE the Required Action(s) and Completion Time(s) in accordance with Technical Specifications for any combination of INOPERABLE systems, structures or components and the bases for the action(s). (SRO Only) (300.006.A.l0)
  • 20. Given 31 GO-OPS-02 1-0, Manipulation of Controls and Equipment, DETERMINE:

(300.049.A.01) a) Whose permission is required to reset lockout relays?

b) Whose permission is required to reset relay targets?

c) Who may reset lockout relays and relay targets?

  • 21. Given the following 4160 VAC Electrical Distribution System annunciators and Annunciator Response Procedures, DETERMINE the significance of each:

(200.017.A.02) a) Station Service supply breaker tripped.

b) Loss of off-site power.

SOUTHERN NUCLEAR DOCUMENT TYPE:

PLANT E. I. HATCH ABNORMAL OPERATING PROCEDURE PAGE 1 OF 9 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF 4160V EMERGENCY BUS 34AB-R22-002-1 1.7 EXPIRATION APPROVALS: EFFECTIVE DATE: DEPARTMENT MANAGER C. R. Dedrickson DATE 10-28-94 DATE:

N/A SSM J PM 8-12-11 N/A DATE N/A 1.0 CONDITIONS 1.1 41 60V bus 1, 1 F, OR I G has been de-energized, as indicated by tripped breaker indications AND initially extinguished bus pot lights on 1 HI 1-P652.

1.2 ANNUNCIATORS LOSS OF OiFSITE POWER (4160V IE), 652-1 02 LOSS OF OFFSITE POWER (4160V IF), 652-202 LOSS OF OFFSITE POWER (4160V IG), 652-302 4160V BUS IE VOLTAGE LOW, 652-1 22 4160V BUS: IF VOLTAGE LOW, 652-222 4160V BUS IG VOLTAGE LOW, 652-322 1 .2.1 See Attachment 1 for additional annunciators.

2.0 AUTOMATIC ACTIONS 2.1 Diesel Genetator for affected bus has auto-started.

2.2 Bus re-energized by alternate supply (SAT 1 C) OR Diesel Generator output breakers closed.

2.3 Loads on the 600V busses are locked out following an undervoltage condition, UNTIL the 00V non-essential load lockout is reset: See Attachment 2.

2.4 On complete loss of offsite power with a LOCA signal present, loads to 41 60V E, F and G will restart at times indicated in Attachment 3.

2.5 Diesel MCC lB (1R24-S026) will automatically align to 4160V 2F on loss of 4160V IF.

3.0 IMMEDIATE OPERATOR ACTIONS 3.1 Confirm Diesel Generator for affected bus has auto-started.

MGR-0002 Rev 8.1

SOUTHERN NUCLEAR DOCUMENT TYPE: PAGE PLANT E. I. HATCH ABNORMAL OPERATING PROCEDURE I OF 59 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR RECOVERY 34AB-R43-001-1 2.0 EXPIRATION APPROVALS: EFFECTIVE DATE: DEPARTMENT MGR SAB for Roger Varnadore DATE 06/21/11 DATE:

N/A SSM / PM N/A DATE N/A 06/21/11 1.0 CONDITIONS 1.1 A LOSP or LOCA signal has been received and Diesel Generators IA, lB or IC have failed to automatically start.

1.2 ANNUNCIATORS

  • FUEL OIL LEVEL HIGH/LOW OR TANK WATER DETECTOR, 652-101(201, 301)
  • LOSS OF OFF-SITE POWER, 652-1 02 (202, 302)
  • START FAILURE, 652-103 (203, 303)
  • ENGINE CVERSPEED, 652-1 04 (204, 304)
  • GEN/STA SVC XFMR IC (ID) DIFF AUX TRIPPED, 652-109 (209, 309)
  • STARTING AIR PRESS LOW, 652-114 (214, 314)
  • 4160V BUS 1E (iF, IG) OR 600V BUS 1C (1D) D.C. OFF, 652-115 (215, 315)
  • BATTERY VOLTS LOW OR FUSE TROUBLE, 652-119 (219, 319)
  • BATTERY CHARGER MALFUNCTION, 652-1 20 (220, 320)
  • DIESEL iA (1B, 1C) AUTO-START LOCK-OUT/CONT. AT ENGINE, 652-105 (205, 305)
  • 4160V BUS IE (iF, IG) UNDERVOLTAGE, 652-1 22 (222, 322)
  • FUEL OIL PRESS LOW, 652-1 27 (227, 327)
  • EMERGENCY ENGINE SHUT-DOWN, 652-129 (229, 329)

MGR-0002 Ver. 8.1

HLT-07 SRO NRC EXAM

31. 262002A3.01 001 Unit 2 is at 100% power when the following occurs:

o Loss of Off-Site Power (LOSP) o 2A EDG fails to start Which ONE of the below choices completes the following statements?

After the LOSP, the Unit 2 Vital AC Bus will INITIALLY receive power from Three (3) hours later and without any operator actions, the Vital AC Bus automatically transfer to ANOTHER power source.

A. the Vital AC batteries; will B the Vital AC batteries; will NOT C. 600VBus2D; will D. 600VBus2D; will NOT 93

HLT-07 SRO NRC EXAM

==

Description:==

The Vital AC bus has three different power supplies:

The normal power supply is from 600 VAC Essential Bus D (R23-S004) through the Vital AC Battery Charger. The backup DC power supply is from the 240 VDC Vital AC Batteries. The alternate power supply is from 600 VAC Essential Bus (R23S003), through Vital AC 600 to 120 VAC Essential Transformer A.

The LOSP and failure of EDG 2A cause 2R23-S003, 600 VAC Bus 2C, to be de-energized. This bus supplies the Alternate power for Vital AC. Vital AC will not transfer to its alernate power supply, 600 VAC Bus 2C, until the non-essential loads are restored. If alternate power is not available from 600 VAC bus C and there was a drop in the inverter input voltage, the inverter would not attempt to transfer due to a sensing circuit which senses a loss of alternate power and blocks the transfer.

The A distractor is plausiblesince the first part is correct. The second part is plausible if the applicant believes that Vital AC will automatically swap to Alternate (600V 2C) without any operator actions. Alternate, 600V 2C, is de-energized due to EDG 2A failure and Vital AC will not transfer to a de-energized alternate (another) supply.

The C distractor is plausible if the applicant believes that Vital AC will be powered by 600V 2D but confuses which supply is NormallAlternate/Backup. The second part is plausible if the applicant believes that Vital AC will automatically swap to Alternate (600V 2C) without any operator actions. Alternate, 600V 2C, is de-energized due to EDG 2A failure and Vital AC will not transfer to a de-energized alternate (another) supply.

The D distractor is plausible if the applicant believes that Vital AC will be powered by 600V 2D but confuses which supply is Normal/Alternate/Backup. The second part is correct.

A. Incorrect See description above.

B. Correct See description above.

C. Incorrect See description above.

D. Incorrect See description above.

References:

NONE K/A:

94

HLT-07 SRO NRC EXAM 262002 Uninterruptable Power Supply (A.C.JD.C.)

A3. Ability to monitor automatic operations of the UNINTERRUPTABLE POWER SUPPLY (A.C.ID.C.) including: (CFR: 41.7 / 45.7)

A3.0l Transfer from preferred to alternate source 2.8 3.1 LESSON PLAN/OBJECTIVE:

R25-ELECT-LP-02705, Vital AC Electrical System, EO 200.020.A.05 References used to develop this question:

34SO-R25-002-2, 120/240 Volt Vital AC System 34AR-651-150-2, Vital AC Sys On Alt Supply Modified from HLT Database HLT-5 NRC Exam 2009302-031 Q#3 1 ORIGINAL QUESTION (HLT-5 NRC Exam Q#31)

Unit 2 is at 100% power when a Loss of Off-Site Power occurs and the 2C EDG fails to start.

Which ONE of the below choices completes the following statements?

The Vital AC Bus is currently receiving its power from The Vital AC Bus will transfer to its Alternate source ONLY after the non-essential loads from have been re-energized.

A. the Vital AC batteries; 600 V Bus 2D B.V the Vital AC batteries; 600VBus2C C. 600VBus2C; 600 V Bus 2D D. 600VBus2D; 600VBus2C 95

R25-ELECT-LP-02705 Page 2 of 38 VITAL AC ELECTRICAL SYSTEM Initial License (LT) ENABLING OBJECTIVES

1. STATE the purpose of the Vital AC System. (027.030.C.0l)
2. Given a simplified drawing of the following systems, TRACE the electrical flowpath through the 240 VDC Vital AC Battery System. (027.039.E.06)
3. LIST the three power supplies for the Vital AC Bus. (027.030.C.02)
4. DESCRIBE the location of the following components of the Vital AC System: (027.030.C.03)
a. Vital AC Static Inverter (R44-S00l)
b. Vital AC Battery
c. Vital AC Bus (R25-S063)
d. Essential Transformer A
5. DESCRIBE the function of the following Vital AC switches: (027.033.B.Ol)
a. Static Bypass Switch
b. Manual Bypass Switch
c. Test Toggle Switch
6. Given a set of conditions, SELECT the correct power supply (if any) to the Vital AC Bus.

(200.020.A.05)

7. Given a list of statements, CHOOSE the one that correctly describes the normal lineup for the Vital AC Static Inverter. (200.020.A.04)
  • 8. In accordance with 34S0-R25-002-l/2, 120/240 Volt Vital AC Systems, STATE the operator actions required to perform the following operations:
a. STARTUP the Vital AC Inverter. (027.030.B.0l)
b. SHUTDOWN the Vital AC Inverter. (027.033.A.02)
c. TRANSFER Vital AC power supplies. (027.03 l.A.0l)
  • 9* In accordance with 34AB-R25-00l-l/2, Loss of Vital AC Bus, DETERMINE: (200.020.A.07)
a. The reason for the automatic action(s).
b. What is accomplished by the automatic action(s).
c. Significance of the notes and cautions.

Objectives marked by a RED (*) are required during RO-305 and SR-305 of the Initial License program.

SNC PLANT E. I. HATCH I Pg 14 of 14 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

120/240 VOLT VITAL AC SYSTEM 34SO-R25-002-2 5.0 ATTACHMENT 1 AU. Pg.

TITLE: VITAL AC ELECTRICAL LINEUP 2 of 2 PERFORM THE BREAKER AND SWITCH ALIGNMENTS ON 2R1 1-S043 AND 2R44-S001 CAUTIONS BEFORE CLOSING THE 600V BUS SUPPLY BREAKERS.

NUMBER DESCRIPTION POSITION CHECKED VERIFIED 2R11-S043 Vital AC Alternate Source Transformer II2TETI2 CB Vital AC Alt. Xfmr Breaker OFF Vital AC Alt. Xfmr Bypass Switch NORMAL 2R44-SOO1 Vital AC CYBEREX UPS II2TETI2 Manual Bypass Switch BYPASS Return Mode Switch AUTO Test Switch CENTER DC Filter Charge Switch OFF Rectifier AC Input breaker OFF Inverter DC Input breaker OFF Equalize Timer Switch AUTO 2R23-S004 600V BUS 2D I3OTETI4 8M Vital AC INVERTER BATTERY CHARGER 2R44-S001 CLOSED 2R23-S003 600V BUS 2C I3OTETI4 STATIC INVERTER 2R44-S001 Via Vital AC XFMR 5M CLOSED 2R11-S043 OPS-0861 Ver. 1.1 G16.030 MGR-0009 Ver. 4

1.0 IDENTIFICATlON ALARM PANEL VITAL AC SYS ON ALT SUPPLY zzzz DEVICE: SETPOINT:

2R44-SO01. See Causes

2.0 CONDITION

3.0 CLASSIFICATION

The Vital AC Static Transfer Switch has transferred to the EQUIPMENT STATUS Alternate Source 4.0 LOCATION:

2H11-P651 Panel I 5.0 OPERATOR ACTIONS:

5.1 IF the Return Mode Switch is in AUTO, the load will automatically transfer back to the inverter when satisfactory operation of the inverter and rectifier is indicated. LI 5.2 IF the transfer occurred due to inverter high temperature or a blown fuse, the Inverter D.C. input breaker will trip. LI 5.3 IF the Return MOde Switch is in AUTO and it is desired to keep the load on the alternate source, PLACE the Manual Bypass Switch in BYPASS. LI 5.4 WHEN the cause of the transfer has been corrected, TRANSFER Vital AC to the Inverter per 3450-R25-002-2. El

6.0 CAUSES

6.1 Inverter voltage greater than 130.0 VAC 6.2 Inverter voltage less than 101.0 VAC 6.3 Inverter temperature greater than 123CF 6.4 Inverter output current limit 750 amps 6.5 Inverter input fuse blown 6.6 SCR failure

7.0 REFERENCES

8.0 TECH. SPECS.ITRM/ODCM/FHA:

7.1 S-60489, Installation-Operating Service Manual, 75KVA, 1 Phase, Uninterruptible Power Supply N/A Not Applicable to this procedure 7.2 H-23406, Inverter and Charger Alarms Sys 2R20P, 2R24A, B or E Elementary Diagram 7.3 H-23635, 120V Vital AC System 2R20P Single Line Diag.

34AR-651 -1 50-2 Ver. 1 MGR-0048 Ver. 5.0 AG-MGR-75-1 101

600V AC 2R1 1-SO4 120V AC 2R44-SOO1 INVERTER 240 VDC VITAL AC Battery System R25-ELECT-02705 Fig 04 Page 38 of 38

HLT-07 SRO NRC EXAM

32. 263000A2.02 001 Unit 2 is operating at 100% power, with the 2A and 2B 1251250VDC Station Service Batteries on equalize charge, when the Control Building ventilation is lost.

Which ONE of the following predicts a consequence of losing the Control Building ventilation and also identifies a required action JAW 34AB-T41-00l-2, Loss Of ECCS, MCREC Or Area Ventilation Systems?

A Hydrogen concentration will rise in the battery rooms; Start Emergency Exhaust Fans 2Z41-C014 and 2Z41-C015 B. Hydrogen concentration will rise in the battery rooms; Open DC breakers to minimize loads on 2R22-5016 and 2R22-S017 C. Battery Chargers will trip on high temperature; Start Emergency Exhaust Fans 2Z41-C014 and 2Z41-C015 D. Battery Chargers will trip on high temperature; Open DC breakers to minimize loads on 2R22S016 and 2R22-5017 96

HLT-07 SRO NRC EXAM

==

Description:==

With a loss of CR ventilation and battery chargers in service, then the Emergency Exhaust Fans must be started to prevent hydrogen buildup. Step 4.8.2 of 34AB-H1 1-001-2 states, IF the battery chargers are in operation start 2Z41-C0l4 AND 2Z41-C015, Emergency Exhaust Fans, 2Hl 1-P657 and 2H1 1-P654. The battery chargers cause an alarm/trip on high temperatures but have internal fans that keep them cool.

The B distractor is plausible since the first part is correct. The second part is plausible if the applicant thinks that reducing the loads on the system will lower the heat generated by the battery chargers or does not remember 34AB-H1 1-001-2 actions.

The C distractor is plausible if the applicant remembers the chargers will trip on high temperature but does not remember each charger has an internal fan. The second part is correct.

The D distractor is plausible if the applicant remembers the chargers will trip on high temperature but does not remember each charger has an internal fan. The second part is plausible if the applicant thinks that reducing the loads on the system will lower the heat generated by the battery chargers or does not remember 34AB-H1 1-001-2 actions.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

97

HLT-07 SRO NRC EXAM

References:

NONE K/A:

263000 D.C. Electrical Distribution A2. Ability to (a) predict the impacts of the following on the D.C. ELECTRICAL DISTRIBUTION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 I 45.6)

A2.02 Loss of ventilation during charging 2.6 2.9 LESSON PLAN/OBJECTIVE:

Z4 1 -CBHVAC-LP-03703, Control Building HVAC, LO H-OP-90000.OO 1 R42-ELECT-LP-02704, DC Electrical References used to develop this guestion:

34AB-T41-OOl-2, Loss Of ECCS, MCREC Or Area Ventilation System(s) 34AR-65 1-126-2, 1 251250V Batt Chgr Malfunction HLT DATABASE HiT 4 NRC Exam Q#32 98

Z41-CBIIVAC-LP-03703 Page 2 of 37 CONTROL BUILIMNG HVAC System Operator (SO) ENABLING OBJECTIVES

  • 1. STATE the purpose of the Control Building HVAC System. (037.036.A.0l)
  • 2 Given a list of areas, DETERMiNE the areas served by the Control Building Miscellaneous HVAC SystemS (037.036.A.02)
3. Given 34S0-Z4 1-004-0, CONTROL BUILDING VENTILATION SYSTEM and winter or summer conditions, DESCRIBE the manual valve alignment required to ensure proper operation of the Station Service Battery room heating coil. (037.036.A.04)
4. Given a list of area coolers, DETERMINE which area coolers are supplied by the Control Building Chilled Water System. (037.037.A.01)
5. Given a list of conditions, DETERMINE the condition(s) that would require the Health Physics HVAC System to be operated in the Emergency Mode. (037.038.A 01)
6. Given procedure 34SO-Z4 1-006-0, Health Physics HVAC System Operation, and plant conditions, STARTUP the Health Physics HVAC System for Normal and Emergency Mode of operation. (037M38.A.02)

System Operator (SO) LEARNING OBJECTIVES

1. Given appropriate references and CONTROL BUILDING HVAC system configuration, ANALYZE the effects of a CONTROL BUILDING HVAC system component malfunction.

cH-OP-90000.001)

2. Given appropriate references and CONTROL BUILDING HVAC system configuration, DIAGNOSE the CONTROL BUILDING HVAC system response to component misalignment.

(H-OP-90000.002)

3. Given appropriate references and CONTROL BUILDING HVAC system configuration, ANALYZE and IDENTIFY the CONTROL BUILDING HVA C system response to component manipulation. (H-OP-90000.003)
4. Given plant conditions and CONTROL BUILDING HVAC system configuration, EVALUATE CONTROL BUILDiNG HVAC system response to changing plant conditions.

(H-OP-90000.004)

  • Not Selected for Continuing training

R42-ELECT-LP-02704 Page 16 of 95 DC ELECTRICAL DISTRIBUTION 250 \kfts DC p NP 125 \ b its DC 1 125 \b its DC 125/250 VDC Battery Configuration A ventilation system in each battery room prevents a buildup of combustible gases and helps ensure operation during emergency conditions. Fire dampers are installed in the ventilation ducts to prevent a fire from spreading from one battery room to another. The batteries are mounted in racks secured to pads located five feet above the floor. Both the batteries and racks are designed to Class 1E requirements.

There are three battery chargers for each division (a total of six), two of which are normally in service, with the third is in standby. Each battery charger is rated to produce 400 amps at 129 VDC and is capable of recharging the battery from a minimum charged condition (1.75 VDC per cell) to a frilly charged condition (2.25 VDC per cell) in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while continuing to supply normal steady state loads.

Battery chargers A, B, and C are located in the same room next to R22-S016 (125/250 VDC Switchgear A). These battery chargers receive power from 600 VAC Essential Bus C, R23-S003, and supply normal power to 125/250 VDC Switchgear A (R22-S016).

(SO *3a)

Battery chargers D, E, and F are located in the same room next to R22-S017 (125/250 VDC Switchgear B). These battery chargers receive power from 600 VAC Essential Bus D, R23-S004, and supply normal power to 125/250 VDC Switchgear B (R22-S017).

NOTE: Both 125/250 VDC Switchgear rooms are located in the Control Building on 130 elevation.

Each battery charger is rated to produce 400 amps at 125 VDC and is capable of recharging the battery from a minimum charged condition (1.75 VDC per cell) to a fully charged condition (2.1 VDC per cell) in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while supplying normal steady state loads.

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 3 OF 6 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF ECCS, MCREC OR AREA VENTILATION 34AB-T41-001-2 3.6 SYSTEM(S) 4.0 SUBSEQUENT OPERATOR ACTIONS 4.1 Attempt to restore affected ventilation system to operation per applicable system operating procedures AND annunciator response procedures.

if both S.E.C. Room Coolers in a diagonal are lost, the availability of the RHR/CS pumps in NOTE: that diagonal can be assured for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, IF no more than one pump is operated at a time and the RHRICS pump is the only significant source of heat addition to the diagonal.

4.2 IF ventilation cannot be restored to an area, perform the following:

4.2.1 IF possible, SHUTDOWN operating equipment in the area OR reduce load on the operating equipment.

H THE INSTALLATION OF TEMPORARY VENTILATION REQUIRES OPENING DOORS CAUTION: OR BLOCKING BARRIERS, ENSURE THAT APPLICABLE ACTIONS OF THE FIRE HAZARD ANALYSIS OPERATING REQUIREMENTS ARE TAKEN, N NECESSARY.

4.2.2 K possible I.E desired, PLACE temporary ventilation systems in service to cool the affected areas and equipment.

4.2.3 IF supplemental cooling is desired to support running RHR/CS Pump and/or to assure pump availability, refer to the System Operation with Plant Service Water (PSW)

Unavailable to Reactor Building subsection of 34SO-T41-001-2.

4.3 IF a ventilation system required to be operable to maintain Tech Spec equipment operable is NQI operable, take action per applicable Technical Specifications.

4.4 IF Reactor Building Ventilation is lost, PLACE OR CONFIRM the Standby Gas Treatment System in operation per 34S0-T46-001-2, Standby Gas Treatment System, with suction from the Reactor Building to maintain at least .20 inch H0 vacuum.

2 MGR-0001 Rev 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 4 OF 6 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF ECCS, MCREC OR AREA VENTILATION 34AB-T41-001-2 3.6 SYSTEM(S) 4.5 H U-2 Refuel Floor Ventilation is lost:

4.5.1 Confirm U-I Refuel Floor Ventilation is in service per 34S0-T41-006-i, Refueling Floor Ventilation System.

4.5.2 N U-i Refuel Floor Ventilation CANNOT be placed in service, START Unit 1 AND/OR Unit 2 Standby Gas Treatment System per 3450-T46-001-112, Standby Gas Treatment System, with suction from the Refueling Floor to maintain at least .20 inch H0 vacuum.

2 4.6 IF ventilation is lost in the Intake Structure, CONFIRM/OPEN roof ventilators dampers per 34S0-X41 -002-0, Intake Structure Ventilation System.

4.6.1 If necessary, refer to 52GM-MNT-021-0, Intake Structure Auxiliary Cooling, to provide increased ventilation via natural draft 4.7 IF ventilation is lost to portions of the Diesel Generator Building, proceed as follows:

4.7.1 CONFIRM OPEN affected roof ventilator dampers per 345O-X41-00i-2, Diesel Generator Building Ventilation System.

4.7.2 H EDG battery room ventilation is lost, perform the following:

  • Request Engineering assistance for exhausting hydrogen from battery room.
  • Notify HP/Chemistry to monitor hydrogen concentration in battery room during the ventilation shutdown.

4.8 IF Control Building Ventilation is lost, perform the following actions within 30 minutes:

4.8.1 Block open the following doors AND establish fire watches as required:

  • C-58, Control Building 130 Elevation Stairwell Door
  • C-85, Control Building 180 Elevation Stairwell Door
  • C-52, Control Building 130 Elevation Entry Door
  • Control Building 130 Elevation door to the CB Supply Fan room
  • Reactor Building 228 Elevation Ductwork Doors (3), on the CB Exhaust fan outlet ductwork (Door key required from Security to gain access to Turb Bldg roof) 4.8.2 Nthe battery chargers are in operation start 2Z4i-C014 AND 2Z41-C015, Emergency Exhaust Fans, 2H11-P657 and 2Hil-P654.

MGR-000i Rev3

1.0 IDENTIFICATION

ALARM PANEL 651-1 1 251250V MALFUNCTION DEVICE: SETPOINT:

See Section 6.0 See Section 6.0

2.0 CONDITION

3.0 CLASSIFICATION

Any 125 V Station Service Battery Chargers have: Loss of Ac EQUIPMENT STATUS power supply, Low DC output voltage, High DC output voltage, 4.0 LOCATION:

high chgrtemp, blocked/clogged filter, loss of both charger fans. 2H11-P651 Panel 651-1 5.0 OPERATOR ACTIONS 5.1 On 2H1 1-P655 confirm normal voltages on 125/250V Battery 2A AND 28 DC Volts, 2R42-R600 and 2R42-R603. [1

[ f NOTE Zero (0) charger current indicates loss of output from the associated charger 5.2 On 2H1 1-P655, check 125/250V Battery 2A AND 2B Pos Amps AND 125/250 Neg Amps to help determine which charger is affected. LI Battery Chargers 2R42-S026, S027, AND S028 are located at 13OTFT13. Battery NOTE Chargers 2R42-S029, S030, AND S031 are located at I 3OTBT1 3.

5.3 At each battery charger, confirm the following indications:

  • DC Volt meter indicates above 125 VDC LI

. AC Power on light is ILLUMINATED LI

. Float/equalize indicating light is ILLUMINATED LI

  • All other indicating lights are EXTINGUISHED LI 5.4 flZ any in service battery charger indications are NOT per Step 5.3, PLACE the standby charger in service per 34S0-R42-001-2. LI 5.5 IF any fan fail alarm is lit, perform the following:

5.5.1 Initiate a CR stating which fan fail alarm is lit, which fan is selected, and which fan is actually running or if no fan is running. Fan operation can be confirmed by checking for fan exhaust thru the top of the charger. Li

6.0 CAUSES

6.1 Any 125 V Station Service Battery Charger, 2R42-S026, S027, S028, S029, S030, OR S031, has one of the following conditions:

. Loss of AC voltage

. DC output voltage is> 150V (Reset = 143V)

. DC output voltage is < 128V (Reset = 132V)

. Internal charger temperature is> 83C

. Charger filter is clogged OR blocked at > 75% flow area

. Both left AND right charger fans have failed

. Alarm by-pass switch on the standby charger is in the NORMAL position

7.0 REFERENCES

8.0 TECH. SPECS./TRMIODCM/FHA:

7.1 H-23390, Single Line Diag. 125/250 VDC Station Service 8.1 3.8.4 DC Sources-Operating 7.2 H-23406, Inverter & Charger Alarms Sys Elem Diags 8.2 3.8.5 DC Sources-Shutdown 7.3 S-60943, Cyberex Vendor Manual for 400-Amp Charger 34AR-651 -1 26-2 Ver. 2.1 MGR-0048 Ver. 5.0 AG-MGR-75-I 101

HLT-07 SRO NRC EXAM

33. 264000K4.06 001 Which ONE of the following knobs can be used to completely shut off the fuel supply to a running Plant Hatch Emergency Diesel Generator?

TYPE UG6

\ I 4

1 v.

M.LJMrr A LOAD LIMIT B. SYN INDICATOR C. SPEED DROOP D. SYNCHRONIZER 99

HLT-07 SRO NRC EXAM

==

Description:==

Load Limit controls the amount of fuel the DG can have. A setting of 0 is equal to 0% fuel.

Syn Indicator indicates the number of complete turns of the Synchronizer knob.

Speed Droop changes how quickly the DG recovers to 900 rpm after load is changed.

Synchronizer adjusts speed/load of the DG. locally and a motor, attached to the governor, allows remote control from the control room.

The B distractor is plausible if the applicant confuses the function of the SYN INDICATOR switch with the function of the LOAD LIMIT switch.

The C distractor is plausible if the applicant confuses the function of the SPEED DROP switch with the function of the LOAD LIMIT switch.

The D distractor is plausible if the applicant confuses the function of the SYNCHRONIZER switch with the function of the LOAD LIMIT switch.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

100

HLT-07 SRO NRC EXAM

References:

NONE K/A:

264000 Emergency Generators (Diesel/Jet)

K4. Knowledge of EMERGENCY GENERATORS (DIESEL/JET) design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)

K4.06 Governor control 2.6 2.7 LESSON PLAN/OBJECTIVE:

R43-EDG-LP-02801, Emergency Diesel Generators 028.022.a.03 Section V, Control features and interlocks References used to develop this question:

HLT DATABASE HLT 4 NRC Exam Q#33 SX28733 Woodward Vendor Manual (Load Limit Knob) 101

R43-EDG-LP-02801 Page 4 of 1131 EMERGENCY DIESEL GENERATORS I

Initial License (LT) ENABLING OBJECTIVES

1. Given a list, STATE the maximum acceptable time for a DIG to start and come up to rated speed and voltage. (028.025.A.03)
2. Given a list, SELECT the function of the following DIG support systems: (028.023.A.03)
a. Diesel starting air
b. Diesel lube oil
c. Diesel cooling water
d. Diesel fuel oil
3. Given the DIG is running and loaded, DETERMINE the D/G kilowatt continuous and 7-day limits according to 34S0-R43-OOl-l/2, Diesel Generator Standby AC System. (028.020.A.01)
4. IDENTIFY how diesel voltage and frequency can be adjusted from the control room or locally.

(028.027.A.01)

5. Given a simplified drawing or P&ID, TRACE the ventilation flowpaths through the following areas: (028.023.A.07)
a. Diesel Room
b. Switchgear Room
c. Diesel Battery Room
d. Oil Storage Room
6. Given a list, SELECT the Diesel Building Ventilation System response to the following:

(028.023.A.08)

a. Fire
b. High ambieiit temperatures
7. Given a list of cooling water supplies, SELECT the D/G sources of cooling water.

(028.023 .A.05)

  • 8. Following a control room evacuation, DETERMINE if additional Load may be added to an operating dieseL (028.028.A.01)
  • 9 Given Plant parameters, DETERMINE if the D/Gs should have auto-started per 34S0-R43-001-1/2, Diesel Generator Standby AC System. (028.025.A.0l)
  • 10. Given D/G parameters, DETERMINE if the D/G should have tripped per 34S0-R43-00l-l/2, Diesel Generator Standby AC System. (028.023.A.02)
11. Given a list, SELECT the available means to locally shutdown a DIG. (028.022.A.03)

R43-EDG-LP-02801 Page 46 of 113 EMERGENCY DIESEL GENERATORS NOTE: If the engine trips on over-speed, the fuel rack must be reset before the engine can be restarted. The resetting lever is located on the engine. Per the diesel vendor, a proper reset could be achieved if the lever is moved slowly (counterclockwise) but fully to the reset position and then repeated. The SO should then depress the Reset Pushbutton and confirm that the GENERATOR OVERSPEED annunciator has cleared.

c. Engine Start Failure Less then 250-RPM and 6 psig lube oil pressure within 7 sec of attempted start.
d. Differential Lockout
2. The following additional six conditions will trip the DIG if the TEST circuit is energized (Diesel Mode Select switch in TEST). (LT *10) (LCT 2)
a. Lube Oil Temperature High (23 0°F)
b. Jacket Water Temperature High (205°F)
c. Jacket Water Pressure Low
1) (Ul)10 psig decreasing (1J2) 9 psig decreasing
2) Armed 20 sec after DIG startup
d. Crankcase Pressure High
1) 0.5 H O

2

2) Armed 20 sec after DIG startup
e. Generator Reverse Power Trips the Differential Lockout
f. Local trip pushbutton or remote stop switch.

Armed when

1) DIG output breaker is open and
2) No automatic initiation signal is present.
3. Two additional methods for securing the DIG in the DIG room: (LT 11)

R43-EDG-LP-02801 Page 47 of 113 EMERGENCY DIESEL GENERATORS

a. Push Emergency Stop pushbutton on the diesel generator to trip the fuel racks.
b. Turn Load Limit Knob on the D/G Governor to 0 provides an electrical signal to shut off the fuel supply.
4. DIG Shutdown System.
a. Governor Shutdown Solenoid
1) The Governor Shutdown Solenoid is actuated by any D/G trip (manual or automatic).
2) When actuated, the solenoid plunger moves the load limit strap down to shutdown the diesel (shuts off fuel to the injectors).

The fuel racks must be reset prior to restarting the diesel.

b. The Shutdown Relay (SDR) initiates D/G trips on any trip signal.
c. The Governor Shutdown Solenoid (GSS) will trip the D/G on the following, if the D/G output breaker is open:
1) SDR trip
2) Differential Lockout Relay trip
  • On Differential Current
3) Local Pushbutton or Remote Control Switch to STOP This will not stop the D/G if either of the following conditions exist:

A) D/G Output Breaker closed B) D/G initiation signal exists D. Interlocks Interlocks for the diesel generator output breaker

a. With diesel generator mode switch in NORMAL the D/G output breaker will close provided the following exist: (LT 12)
1) Emergency 4160V bus undervoltage (86 relay)
2) Proper D/G frequency

W 000WA P13 and the smaH dash pot compensation (35) pistons, SYNCHRONIZER and the oil sump.

The synchronizer is the speed adjusting control, and is used to change engine speed for a single NOTE unit. On engines paralleled with other units, it is used to change engine load.

Compensation must be properly adjusted to the particular engine and load to The upper knob called SYNCHRONIZER on provide stable operation (see Section 4, Compensation Adjustments). most models, and SPEED SETTING KNOB on late models, is the control knob.

LOAD LIMIT CONTROL The lower knob, SYN. INDICATOR has no function of its own but has an indicator disc which The purpose of the load limit control is to shows the number of revolutions of the synchro hydraulically and mechanically limit the load that nizer (speed setting) control knob.

can be placed on the engine by restricting the travel of the governor output shaft in the increase fuel direction, and consequently the amount of SPEED DROOP fuel supplied to the engine.

The load limit control may also be used for Speed droop or simply droop is one method of shutting down the engine by turning it to zero. creating stability in a governor. Droop is also used to divide and balance load between units driving the same shaft or paralleled in the electrical system.

Droop is the decrease in speed taking place when Do not manually force prime mover the governor output shaft moves from the minimum linkage to increase fuel without first to the maximum fuel position in response to a load turning the load limit control knob to maximum position (10). Failure to do so increase, expressed as a percentage of rated may cause dar.iage and/or failure of speed.

governor internal parts.

If instead of a decrease in speed, an increase takes place, the governor is showing a negative The load limit control consists of an indicator disc droop. Negative droop will cause instability in a (7) geared to a load limit rack (8). The control governor.

knob is also attached to the load limit cam (16).

Not enough droop can cause instability in the Load is limited mechanically by positioning the form of hunting, surging or difficulty in response load limit knob (cam 16). When the load indicator to a load change. Too much droop can result in reaches the preset point, the pilot valve plunger slow governor response in picking up or dropping (39) is lifted, stopping any further increase in fuel.

off a load.

Turning the load limit control to zero to shut down the engine turns the cam (16) forcing the load limit Using an example where the governor speed is (shutdown) lever (20) and shutdown strap (17) 1500 rpm at no load and 1450 rpm at full load, down. As the right end of the load limit (shutdown) droop can be calculated with the formula:

lever (20) is forced downward, it pivots about its fulcrum and lifts the pilot valve plunger (39),

No load speed Full load speed X100 releasing oil from under the power piston (9).

%Droop =

Full load speed Pressure oil acting on top of the power piston (9) forces it downward, rotating the governor output shaft (6) to minimum fuel and causing the prime 1500 rpm 1450 rpm 100

%Droop 1450 rpm mover to shutdown.

12

HLT-07 SRO NRC EXAM

34. 268000A4.01 001 JAW 34SV-SUV-019-2, Surveillance Checks. and TS 3.4.4. RCS Operational LEAKAGE, which ONE of the choices below identifies the integrator normally used in determining UNIDENTIFIED Leakage rate and its location?

A. Drywell Floor Drain Sump Integrator, 2G1 1-K601, located on panel 2H11-P602.

B Drywell Floor Drain Sump Integrator, 2G1 l-K601, located on panel 2H11-P613.

C. Drywell Equipment Drain Sump Integrator. 201 1-K603, located on panel 2H1 1-P602.

D. Drywell Equipment Drain Sump Integrator, 201 1-K603, located on panel 2H1 l-P613.

102

HLT-07 SRO NRC EXAM

==

Description:==

One sump is the Drywell Equipment Drain Sump (DWEDS) which is used to collect water from identified or known sources (i.e. designed seal leakage). The other sump is the Drywell Floor Drain Sump (DWFDS) which is used to collect water from unknown or unidentified sources.

Drywell inleakage is an essential part of primary containment integrity determination. Technical Specifications has determined the leakage limits that are acceptable without declaring primary containment inoperable. The inleakage has to be checked periodically and the leakage values confirmed less than those allowed by Technical Specifications.

Drywell sump inleakage determination is performed once per eight hours, using sections 7.24.1 and 7.24.2 of 34SV-SUV-019.1/2, Surveillance Checks. The integrator readings are recorded at 0000, 0800 and 1600 each day. The operator then calculates the inleakage into each sump by dividing the total inleakage since the last reading by the number of minutes since the last reading. The calculations determine the inleakage.

The Drywell Floor Drain Sump Integrator, 2G1 1-K601, is located in the Main Control Room on panel 2H11-P613.

The A distractor is plausible since the first part is correct. The second part is plausible since other DW Floor & Equipment Drain switches/flow recorders/indications are located on this panel and the applicant can confuse them with where the DW Floor Drain intergrator is located.

The C distractor is plausible if the applicant does not understand that DW Floor Drain is Unidentified Leakage. The second part is plausible since other DW Floor & Equipment Drain switches/flow recorders/indications are located on this panel and the applicant can confuse them with where the DW Floor Drain intergrator is located.

The D distractor is plausible if the applicant does not understand that DW Floor Drain is Unidentified Leakage. The second part is correct.

A. Incorrect See description above.

B. Correct See description above.

C. Incorrect See description above.

D. Incorrect See description above.

103

HLT-07 SRO NRC EXAM

References:

NONE K/A:

268000 Radwaste A4. Ability to manually operate and/or monitor in the control room:

(CFR: 41.7/ 45.5 to 45.8)

A4.01 Sump integrators 3.4 3.6 LESSON PLAN/OBJECTIVE:

T23-PC-LP0 1301, Primary Containment, EO 040.004.A.02 References used to develop this question:

34SV-SUV-0 19-2, Surveillance Checks 34S0-G1 1-009-2, Drywell And Reactor Building Sumps Systems GRAND GULF 2007 NRC EXAM Q#61 104

T23-PC-LP-01301 Page 9 of 160 PRIMARY CONTAINMENT

(013.061.A.0l)

(013.062.A.01)

  • 36 Per 34S0-T48-002-1/2, Containment Atmospheric Control and Dilution System, DETERMINE the sequence of actions necessary to makeup Nitrogen to the Containment by the normal method. (013.008.A.01)
  • 37 Per 3450-T48-002-1/2, Containment Atmospheric Control and Dilution System, DETERMINE the sequence of actions necessary to makeup Nitrogen to the Containment from Unit 1 Nitrogen Storage Tank. (013.009.A.0l)
  • 38. Per 34S0-T48-002-1/2, Containment Atmospheric Control and Dilution System, DETERMINE the sequence of actions necessary to makeup Nitrogen to the Containment from Unit 2 Nitrogen Storage Tank. (013.OlO.A.01)
  • 40. Per 34G0-OPS-028-2, Drywell Closeout, DETERMINE the actions necessary to perform a Drywell Closeout Inspection. (013.021 .A.0 1)
  • 42. DETERMINE the major steps to startup the DW ventilation system per the normal operating procedure 34S0-T47-00 1- 1/2. (043 .002.A.0 1)
  • 43 With Suppression Pool level slightly below normal, PERFORM a low volume fill of the suppression chamber per 34SO-E21-00l-1/2, Core Spray System to maintain suppression chamber level within the limits of Technical Specifications. (013.001 .A.03)
  • 44* With Suppression Pool level extremely low, PERFORM a high volume fill of the suppression chamber per 34G0-OPS-087-l/2, Suppression Chamber Fill and Drain to maintain suppression chamber level within the limits of Technical Specifications. (013.002.A.04)
  • 45* Given plant conditions DETERMINE the leakage rates into the Drywell Equipment and Floor Drain Sumps per 34SV-SUV-01 9-1/2, Surveillance Checks. (040 .004.A.02)

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 8 OF 62 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DRYVVELL AND REACTOR BUILDING SUMPS SYSTEMS 34S0-G11-009-2 10.8 7.13 Placing The Drywell Floor Drain Sump In Automatic L:is 7.1.3.1 Confirm 2A71-S22A and C and 2A71-S22B and D, MSIV LOGIC switches, are in the NORMAL position, panels 2H11-P609 AND 2H11-P611. LI 7.1.3.2 Depress Drywell Sumps Outbd Isol VIvs reset pushbutton, panel 2H1 1-P601. [1 7.1.3.3 Depress Drywell Sumps Inbd Isol Vlvs reset pushbutton, panel 2H1 1-P602.

7.1.3.4 Open 2G11-F003, Drywell Floor Drain Vlv, panel 2H11-P602.

7.1.3.5 Open 2G11-F004, Drywell Floor Drain VIv, panel 2H11-P601. LI 7.1.3.6 Depress 2G1 1-S400, Drywell Sumps Pump Out and Fill timers, reset pushbutton, AND confirm the DRYWELL FLOOR DRAINS SUMP LEAK HIGH (602-408) annunciator is NOT lit, panel 2H1 1-P602.

7.1.3.7 Place 2G11-COO1A, Drywell Floor Drain Pump, control switch in AUTO, panel 2H11-P602.

7.1.3.8 Place 2G11-CO0IB, Drywell Floor Drain Pump, control switch in AUTO, panel 2H11-P602. LI

  • The Drywell Floor Drain Sump is now in Automatic. The pumps will cycle on under high sump level OR off under low sump level conditions. To ensure even wear, the Number One Radwaste Logic Controller alternates the pump which starts in response to a high sump level condition. Under high high sump level conditions, both pumps auto start AND run UNTIL a low sump level is reached.
  • The sump pumps and discharge valves are normally controlled from the Number One Radwaste Logic Controller. IF the Number One Logic Controller is placed in manual NOTES: override, THEN the pumps control is transferred to the Number Three Radwaste Logic Controller. All functional operation will remain as previously described in this note.
  • Under conditions of high drywell pressure, greater than OR equal to 1 .85 psig, OR low vessel level less than OR equal to 3 inches, the Drywell Floor Drain Sump Isolation Valves, 2G1 1-F003 AND 2G11-F004 will close and remain closed, regardless of operator action, UNTIL the initiating condition has cleared. The isolation can THEN be reset. The sump pumps CANNOT be started WHEN EITHER 2G1 1-F003 OR 2G1 1-F004, Discharge Isolation Valve, is closed, OR a common discharge line high radiation trip present.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PLANT E. I. HATCH PAGE 59 OF 77 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

SURVEILLANCE CHECKS 34SV.-SUV-O1 9-2 37.1 T24 PANEL-INSTRUMENT/TECHSPEC. NOTE FREQ 0000 0800 1600 2H11-P613-2G11-K603, DIW Equipment Drain Leakage Actual Time 1 H.1 Present Reading 2 H.1 Yesterdays Reading Time 3 H.1 Yesterdays Reading 4 H.1 7.24.1 1,2,3 Difference (2) - (4) 5 H.1 Difference Conversion (5) x 20 6 H.1 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> elapsed minutes 7 H.1 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> leakage (6)1(7) 8 LL, H.2 7.24.2 Confirm pump shuts off automatically on decreasing level of sump B 1,2,3 2H11-P613-2G11K601, DAN Floor Drain Leakage Actual Time 9 H.1 Present Reading 10 H.1 Yesterdays Reading Time 11 H.1 Yesterdays Reading 12 H.1 Difference(10)-(12) 13 H.1 Difference Conversion (13) x 20 14 H.1 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> elapsed minutes 15 H.1 24hourleakage(14)/15 16 Q,H.2 5gpm 7.24.3 . 1,2,3 Previous 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reading 17 H.1 Previous 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reading time 18 H.1 8hourdifference(10)-(17) 19 H.1 Difference conversion (19) x 20 20 H.1 8hourelapsedminutes 21 H.1 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> leakage (20)1(21) 22 Q,H.2 <5 gpm Yesterdays 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> leakage 23 Total Leakage (8) + (16) 24 < 30 gpm Differential Floor Drain Leakage (22)-(23) 25 Q < 2 gpm (SR 3.4.4.1) in Mode 1 7.24.4 Confirm pump shuts off automatically on decreasing level of sump B 1,2,3 Initials Calculations verified I I Date__________ Time 000 I 0800 I 1600 G16.030 MGR-0001 Ver. 3

HLT-07 SRO NRC EXAM

35. 286000G2.4.9 001 Unit 1 experiences a LOCA and LOSP with ALL Unit 1 Emergency Diesel Generators failing to start and cannot be manually started.

The following conditions exist on Unit 1:

o Reactor water level -160 inches, decreasing 2 inches/minute o Reactor pressure 25 psig, slowly lowering 31E0-EOP-1 10-1, Alternate RPV Water Level Control, is in progress.

With the above conditions, which ONE of the choices below completes the following statements concerning Fire Water Injection into the Residual Heat Removal (RHR) System?

JAW 31 EO-EOP- 110-1, Alternate RPV Water Level Control, the Fire System can be used for RPV injection by aligning it The MAXIMUM number of Fire Pumps that can be started is A INDIRECTLY into the RHR System via the Condensate Transfer System piping; two (2)

B. INDIRECTLY into the RHR System via the Condensate Transfer System piping; three (3)

C. DIRECTLY into the RHR System piping; two (2)

D. DIRECTLY into the RHR System piping:

three (3) 105

HLT-07 SRO NRC EXAM

==

Description:==

If fire main header pressure decreases to 110 psig, a pressure switch starts the Motor Driven Electric Fire pump, 1 X43-C00 1 A. If fire main header pressure continues to decrease to 100 psig, a pressure switch starts Engine Driven Fire Pump, 1X43-COO2A. If fire main header pressure continues to decrease to 90 psig, a pressure switch starts Engine Driven Fire Pump, 1X43-COO2B. The power supply to the Electric Fire pump is 416OVAC lE, which has lost power in this question, therefore, leaving only the two (2) Engine Driven Fire Pumps able to be started for injection.

JAW 31E0-EOP-1 10-1, Alternate RPV Water Level Control, the Fire System can be used for RPV injection by aligning it indirectly into the RHR System via the Condensate Transfer System piping. Fire System is used in other procedures for makeup water and is directly aligned to the system.

The B distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses the Fire Pumps and their power supplies and thinks three will be available, Electric (no power) and 2 Diesel Fire pumps. Also plausible since it would be correct if 41 60V lE were available from DG 1A.

The C distractor is plausible if the applicant confuses/does not remember the Fire System does not align directly into the RHR System piping. Also plausible since the Fire System is used in other procedures for makeup water and is directly aligned to those systems. The second part is correct.

The D distractor is plausible if the applicant confuses/does not remember the Fire System does not align directly into the RIHIR System piping. Also plausible since the Fire System is used in other procedures for makeup water and is directly aligned to those systems. The second part is plausible if the applicant confuses the Fire Pumps and their power supplies and thinks three will be available, Electric (no power) and 2 Diesel Fire pumps. Also plausible since it would be correct if 41 60V I E were available from DG 1 A.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

106

HLT-07 SRO NRC EXAM

References:

NONE K/A:

286000 Fire Protection System 2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

(CFR: 41.10/43.5/45.13) 3.8 4.2 LESSON PLAN/OBJECTIVE:

EOP- 11 0-LP-20321. EOP 110: Alternate RPV Water Level Control, EU 036.022.A.02 X43-FPS-LP-0360 1, Fire Protection, EU 036.020.B.13 References used to develop this question:

31 EU-EOP- 110-1, Alternate RPV Water Level Control 34S0-X43-001-1, Fire Pumps Operating Procedure 107

EOP-110-LP-20321 Page 3 of 32 EOP 110: ALTERNATE RPV WATER LEVEL CONTROL Initial License (LT) ENABLING OBJECTIVES

1. Given 3 1EO-EOP-1 10-1/2, Alternate RPV Water Level Control, and the status of both RHR loops, DETERMINE if RHRSW should be used as a source of injection. (034.012.A.02)
2. Given 31E0-EOP-l 10-1/2, Alternate RPV Water Level Control, and the applicable system drawing(s), IDENTIFY the flow path used to restore and maintain the RPV water level with the following systems:
a. RHRSW (034.012.A.01)
b. Condensate Transfer via the LPCI path (201.094.A.01)
c. Fire System via the LPCI path (036.022.A.01)
d. RHR Crosstie method (034.013.A.01)
e. Condensate Transfer via the Shutdown Cooling flush path (201.095.A.0l)
f. Fire System via the Shutdown Cooling flush path (036.023.A.01)
g. Jockey pumps (008 .017.A.05)
h. SBLC from the test tank (01 l.015.A.01)
i. SBLC from the storage tank (01 1.014.A.01)
3. Given 3 lEO-FOP-i 10-1/2, Alternate RPV Water Level Control, and the status of both RHR loops, DETERMINE if Condensate Transfer Pumps should be used as a source of injection.

(201 .094.A.02)

4. Given 31E0-EOP-1 10-1/2, Alternate RPV Water Level Control, and the status of both R}IR loops, DETERMINE if the Fire System should be used as a source of injection. (036.022.A.02)
5. Given 3 1EO-EOP-l 10-1/2, Alternate RPV Water Level Control, and a list, IDENTIFY the statement that describes the effect that lining up the Fire System for injection into the RPV will have on the Condensate Transfer System. (036.022.A.04 & 036.023.A.03)
6. Locate Fire Hydrant 11. (201.097.A.01)
7. Given plant conditions and 31E0-EOP-llO-1/2, Alternate RPV Water Level Control, EVALUATE plant conditions and determine if the RHR Crosstie Method would be a useful method of controlling RPV water level. (034.0 13.A.02)
8. Given 31 EO-EOP- 110-1/2, Alternate RPV Water Level Control, and the status of CS and RHR, DETERMINE ifjockey pump injection into the RPV will occur. (008.0 17.A.06)
9. Given a list, IDENTIFY the statement that describes why the jockey pump minimum flow valves are NOT closed until RPV pressure is below 100 psig when using the jockey pumps as an injection source. (008.0 l7.A.04)
10. Given plant conditions and 31E0-EOP-llO-1/2, Alternate RPV Water Level Control, DETERMINE if SBLC can be used to restore and maintain RPV water level. (011.015 .A.02 &

011 .014.A.02)

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 10 OF 42 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

ALTERNATE RPV WATER LEVEL CONTROL 31 EO-EOP-1 10-1 2.6 3.1.3 FIRE SYSTEM VIA CONDENSATE TRANSFER CROSSTIE

  • This section crossties the plant fire main with the Unit 2 Condensate Transfer System. Fire main pressure is then used to inject to the RPV via the Unit 2 Condensate Transfer System to the Unit 1 Condensate Transfer System and then to the Unit I RHR flushing connections. The Condensate Transfer Pumps for both Units will NOT be available during the performance of this subsection.
  • This section can only be performed using Unit 2 Condensate Transfer Loop B due NOTES: to design differences between 2P11-F027A and 2P11-F027B.
  • Keys for the following locally operated valves will be required during the performance of this section.
  • 2P11-F026B THIS SECTION IS NOT TO BE PERFORMED ON ANY OPERATING UNIT 1 CAUTION:

RHR LOOP.

3.1 .3.1 IF any Fire Protection deluge system has actuated, AND there is no fire in the area, THEN isolate the deluge system per Attachment 2.

3.1.3.2 ConfIrm STOPPED/STOP both Units 1P1I-COOIA and 1PI1-COOIB and 2P1 1-COO1A and 2P1 1-COOl B, Condensate Transfer Pumps.

3.1.3.3 CLOSE:

VALVE DESCRIPTION LOCATION 2P11-F024A Pump 2P11-COOIA Discharge U-2 CTP enclosure 2P1 1-F024B Pump 2P11-COO1B Discharge U-2 CTP enclosure 2P1 1-F025B Pump 2P1 1-COOIB Suction U-2 CTP enclosure 3.1.3.4 Unlock and CLOSE 2P1 1-F026B, Minimum Flow B, U-2 CTP enclosure G16.030 MGR-0001 Ver. 3

X43-FPS-LP-03601 Page 3 of 145 FIRE PROTECTION

10. Given that a fire has started, DESCRIBE the starting sequence for the three fire pumps and the pressure when they start. (036.020.B.13)
11. From a list of statements, SELECT the statement that best describes the automatic operation of the CO 2 (Cardox) system per 34S0-X43-005-0, Diesel Generator Building Carbon Dioxide System. (036,020.B.05)
12. Given that a fire has occurred in the Diesel Generator Bldg., the 5 ton CO 2 has failed to actuate automatically and a copy of plant procedures, MANUALLY INITIATE the 5 ton unit per 34S0-X43-005-0, Diesel Generator Building Carbon Dioxide System using either the manual-electric or manual-manual method. (200.024.A.03)
13. Given a fire has occurred in the Cable Spreading or Computer Room, MANUALLY OPERATE the 13 ton CO 2 system per 34SO-Z43-002-0, Turbine and Control Building Carbon Dioxide System. (200.024.A.04)
14. From a list of statements, SELECT the statement which describes the operation of the CO2 hose reels per 34S0-X43-005-0, Diesel Generator Building Carbon Dioxide System.

(036.020.B.06)

15. STATE (3) personnel hazards associated with use of CO 2 or Halon 1301 systems as described in plant procedures. (036.020.B.09)
  • 16. Given a list of characteristics of the four types of Fire Sprinkler Systems, SELECT the characteristics pertaining to each of the following systems: (036.020.A.02)
a. Deluge
b. Wet Pipe Sprinklers
c. Dry Pipe Sprinklers
d. Preaction Sprinklers
17. Given an SO reports a small stream of water is flowing from a Deluge System ball drip valve, STATE the significance of this finding. (036.020.A.06)
  • 18. Given a list of plant areas IDENTIFY those areas that are protected by the following CO 2

Systems: (200.024.A.05)

a. 5 Ton Cardox
b. l3TonCardox
19. Given an alarmltrouble printout from the CXL computer, Master Panel, or Slave Panel and Plant Procedures DETERMINE the following: (036.02 1.A.02)
a. Affected Unit
b. Affected Building
c. Elevation
d. Type of System Giving Alarm
e. Zone Sequence Number

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH .

40F60 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

FIRE PUMPS OPERATING PROCEDURE 34S0-X43-001-.1 5.0 5.2 LIMITATIONS 5.2.1 Each fuel oil storage tank shall contain at least 275 gallons of fuel oil.

5.2.2 Each Fire Protection Storage Tank shall contain at least 270,000 gallons (28 ft) of water.

5.2.3 WHEN in standby, engine driven pumps cooling system temperature must be maintained at greater than 120cF.

5.2.4 Operation with lubricating oil pressure <15 psig or engine coolant Temperature >200cF is NOT permitted.

5.2.5 WHEN fire protection storage tank level is less than 10, the fire protection system jockey pump must be shutdown the motor driven and engine driven fire pumps must be monitored for cavitation.

5.2.6 WHEN fire protection storage tank level is less than 3, the motor driven and engine driven fire pumps must be shutdown.

5.2.7 IF cavitation of any pump is observed, it is to be shutdown immediately.

5.2.8 Except WHEN an emergency exists, observe proper limitations on Electric Fire Pump motor starts.

  • Two starts in succession from ambient temperature QfEE
  • one start from rated temperature are allowed.
  • For subsequent starts, allow 30 minutes running time or 60 minutes idle time.

5.2.9 1Y42-F309N, Demin Bypass Vlv, will AUTO open on Low header pressure in the Fire system as sensed by either of the 3 pressure switches which auto-start the Fire pumps, provided the valve is in the NORMAL position.

Hthe valve is placed in TEST, the 1Y42-F309N can be manually cycled with the local control switch; however, it will NOT automatically OPEN on Low header pressure.

H in NORMAL, the 1Y42-F309N is interlocked open following a Low header pressure condition UNTIL header pressure increases above the reset setpoints, AND the RESET pushbutton is depressed.

5.2.10 Fire Pumps auto start setpoints:

1X43-C001 Electrical Fire Pump 110 psig 1X43-COO2A Diesel Fire Pump 100 psig 1X43-COO2B Diesel Fire Pump 90 psig 1X43-C003 Jockey Fire Pump 125 psig MGR-0001 Rev4

SNC PLANT E. I. HATCH I Pg 24 of 60 DOCUMENT TITLE: DOCUMENT NUMBER: Ver No:

FIRE PUMPS OPERATING PROCEDURE 34SO-X43-001-1 5.0 ATTACHMENT1 Att.Pg.

TITLE: FIRE PUMPS ELECTRICAL LINEUP 4 of 4 NUMBER DESCRIPTION CHECKED Engine Driven Fire Pump 3 Area (CONTD) 1X43-R302 Fire Protection Pump #3 Controller Control Switch OFF 1X43-R302 Fire Protection Pump #3 Controller Battery 1 switch ON 1X43-R302 Fire Protection Pump #3 Controller Battery 2 switch ON 1X41-NOO1A Unit Heater A thermostat 50F 1X41-NOO1B Unit Heater B thermostat 50F 1X41-NOO1C Unit Heater C thermostat 5OCF 1R22-S005 4160V Bus IE, Diesel Building, Switchgear Room IE RACKED FR 6 Electric Fire Pump, 1X43-C001 30 amp control power breaker CLOSED 1R25-S004, 125 VDC Cabinet 1D, Diesel Building Switchgear Room 1E 4160 V Swgr Bus 1E FDR ACBs (Frames 2-10) & UNDV brkr 1 ON RELAYING 1R25-S064 120/208V Cab 1A Instrument Bus 1A 13OTGT11 brkr 21 Auxiliary Boiler Instrumentation (1Y43-N001, 1Y43-N002) ON 1R24-S028 600/208V Start-Up Boiler MCC-1A, Aux Boiler FR 2B Fire Pump House 600V Cabinet 1A (1R25-5051) ON OPS-0828 Ver 2.1 G16.035 MGR-0009 Ver 5

HLT-07 SRO NRC EXAM

36. 295001AK2.01 001 Unit 2 is operating at 73% power with both Recirc Pumps operating at 70% speed when the 2A Recirc Pump trips.

With NO operator action, which ONE of the choices below completes the following statement concerning the Jet Pump Loop A Total flow?

2B2 1 -R6 11 A, Total A Flow, will drop to a MINIMUM of A. zero (0) Mlbmlhr and then stabilize there B. 2.0 Mlbmlhr and then stabilize there 0 zero (0) Mlbmlhr and then increase to approximately 6.0 Mlbmlhr D. 2.0 Mlbmlhr and then increase to approximately 6.0 Mlbmlhr 108

HLT-07 SRO NRC EXAM

==

Description:==

When the 2A Recirc pump tripped, 2B21-R6l 1A, Total A Flow, will drop to 0 Mlbmlhr and then increase to an indicated value of 6.0 Mlbmlhr. As the flow indication decreases to zero the indicated flow is positive. IF the running Recirc pump speed is 70%, reverse flow will be indicated through the idle loop. During single loop operation, when the speed of the running pump decreases below approximately 35% speed, positive flow through the idle pump loop due to natural circulation overcomes the negative flow due to reverse flow. This will add plausibility to the remaining distractors.

The A distractor is plausible if the applicant confuses this flow with Drive Flow which does drop to zero and stabilize there.

The B distractor is plausible if the applicant confuses/mis-interprets the Recirc response and thinks flow drops to 2.0 Mlbmlhr and stabilizes there. 2.0 Mlbrnlhr is where flow indication will stabilize from a pump trip at a lower power/speed.

The D distractor is plausible if the applicant confuses/mis-interprets the Recirc response and thinks flow drops to 2.0 Mlbmlhr and then increases to 6.0 Mlbmlhr. 6.0 Mlbmlhr is where flow indication will stabilize from a pump trip at this power/speed.

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

109

HLT-07 SRO NRC EXAM

References:

NONE K/A:

295001 Partial or Complete Loss of Forced Core Flow Circulation AK2. Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION and the following: (CFR: 41.7 /45.8)

AK2.O1 Recirculation system 3.6 3.7 Changed K/A to the above K/A on 2/7/2012 per Chief Examiner Edwin Lea AK2. Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION and the following: (CFR: 41.7 /45.8)

AK2.05 LPCI ioop select logic: Plant-Specific 3.2 3.6 LESSON PLAN/OBJECTIVE:

B3 1 -RRS-LP-0040 1, Reactor Recirculation System, EO 200.037 .A.0 1 References used to develop this question:

34AB-B31-001-2, Reactor Recirculation Pump(s) Trip, Or Recirc Loops Flow Mismatch, Or ASD Power Cell Failure 345O-B31-001-2, Reactor Recirculation System (50% Speed vs. reverse flow) 110

B31-RRS-LP-00401 Page 10 of 132 REACTOR RECIRCULATION SYSTEM

  • 17. Given a list of Recirc System controls and indications, DESCRIBE the operation of each.

(004.002.A.03)

  • 18. Given Plant conditions, DETERMINE if the plant is operating in the region of potential instabilities per 34G0-OPS-005- 1/2 Power Changes. (004.003 .A.0 1)
  • 19. Given a set of parameters, EVALUATE those parameters to determine if one or both Recirc pumps have tripped. (200.037.A.01)
  • 20. Given Plant conditions, DETERMINE the proper actions for resetting the RPT breakers.

(004.015.A.01)

  • 21. Given a list of statements, DETERMINE which is an indication of an RPT Breaker Trip.

(004.002.A.09)

  • 22. Given that a Recirc Runback exists, DETERMINE the proper actions for resetting the Recirc Runback per 34S0-B3 1-001-1 /2,Reactor Recirculation System. (004.01 7.C.0 1)
  • 23. CONFIRM a proper start of a Recirc pump per procedure 34S0-B3 1-001 1/2 Reactor Recirculation System. (004.002 .A.04)
  • 24. STATE the reasons for shutting the Recirc pump discharge valve upon securing the pump.

(004.003.A.07)

  • 25. STATE the maximum heat-up/cool down rate for a Recirc loop. (004.003.A.05)
  • 26. STATE the reasons for re-opening the Recirc pump discharge valve within 4-5 minutes of securing the pump. (004.003.A.08)

(200.037.A.02)

  • 28. Given Plant conditions, DETERMINE the proper actions for maintaining an idle Recirc loop at the proper temperature per 34S0-B31-001-1/2 Reactor Recirculation System.

(004.002.A. 12)

  • 29. Given plant conditions involving the Reactor Recirculation System, DETERMINE if a Technical Specification Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (300.006.A.23)
  • 30. DETERMINE which instruments supply signals to Reactor Recirculation System per Technical Specifications. (300.009.A.1 1)

B31-RRS-LP-00401 Page 75 of 132 REACTOR RECIRCULATION SYSTEM

f. Core Flow Determination at Low Recirc Speeds
  • During single loop operation (sensed by the ASD Not Running status provided by the Recirc NXG computer), when the speed of the running pump decreases below approximately 35% speed, positive flow through the idle pump loop due to natural circulation overcomes the negative flow due to reverse flow.
  • The total core flow summing circuitry will continue to subtract this positive idle loop flow from the running loop flow and give a misleading LOW core flow indication.
  • Total Core Flow can be calculated by adding the JET PUMP LOOP A AND the JET PUMP LOOP B flows.

Running I Idle Loop Recirc Pump Speed I Jet Pump Flow Direction

<35% Positive 35%toSO% NoFlow

>50% j Reverse

2. Recirc Pump Restarts 34S0-B31-OO1-1/2
a. Must meet the idle loop temperature requirements plus be within winding temperature limitations of the pump motor.

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 2 OF 15 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REACTOR RECIRCULATION PUMP(S) TRIP, OR RECIRC LOOPS 34AB-B31-001-2 10.0 FLOW MISMATCH, OR ASD POWER CELL FAILURE I. LOSS OF A SINGLE REACTOR RECIRCULATION PUMP A OR B 1.0 CONDITIONS 1.1 ANNUNCIATOR ALARMS 1.1.1 ASDATRIP WARNING (602-101) OR 1.1.2 ASD B TRIP WARNING (602-201) 1.1.3 ASD A FATAL FAULT (602-1 02) OR 1.1.4 ASD B FATAL FAULT (602-202) 1.1.5 RECIRC LOOP A OUT OF SERVICE (602-127) OR 1.1.6 RECIRC LOOP B OUT OF SERVICE (602-227) 1.1.7 ASD A LOCKOUT (602-1 03) OR 1.1.8 ASD B LOCKOUT (602-203) 1.1.9 ASD A AUX LOCKOUT (602-1 09) OR 1.1.10 ASD B AUX LOCKOUT (602-209) 1.1.11 ASD A COOLING FAULT (602-1 26) OR 1.1.12 ASD B COOLING FAULT (602-126) 1.2 Recirculation pump differential pressure on the affected ioop decreases to zero as indicated on 2B31-R612A0R2B31-R612B, Pump Diff Press, panel 2H11-P602.

1.3 Loop flow in the affected recirculation loop decreases to zero, THEN increases slightly as reverse flow is established as indicated on 2B21-R61 1A Q 2B21-R61 1 B, Jet Pump Total Flow, panel 2H1 1-P602.

1.4 Total core flow decreases, as indicated on 2B21 -R61 3, Core Plate Dp/Rx Core Flow, black pen, panel 2H1 1-P603.

1.5 Reactor core plate differential pressure decreases, as indicated on 2B21-R613, Core Plate Dp/Rx Core Flow, red pen, panel 2H1 1-P603.

1.6 Reactor power decreases.

MGR-0001 Rev4

SOUTHERN NUCLEAR PAGE PLANT F. I. HATCH 4 OF 15 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REACTOR RECIRCULATION PUMP(S) TRIP, OR RECIRC LOOPS 34AB-B31-001-2 10.0 FLOW MISMATCH, OR ASD POWER CELL FAILURE 4.0 SUBSEQUENT OPERATOR ACTIONS 4.1 Hthe OPRM System is INOPERABLE, enter 34AB-C51-001-2, AND perform concurrently with this procedure.

During single loop operation, WHEN the speed of the running pump decreases below approximately 35% speed, positive flow through the idle pump loop due to natural circulation overcomes the negative flow due to reverse flow. The total core flow summing circuitry will continue to subtract this positive idle loop flow from the running loop flow AND give a misleading LOW core flow indication.

NOTE: Total core flow can be calculated by adding the JET PUMP LOOP A the JET PUMP LOOP B flows.

  • SIL 517 describes increased noise in nuclear instrumentation from single loop operations. Plant Hatch data analysis taken from plant single loop operation shows that operators should expect the following peak to peak instrument oscillations:

RWL 5.2 in, APRM Flux 5.3%, Core flow 2.1 Mlb/hr, and Rx Press 1 .3 PSIG.

Critical 4.2 Reduce the operating Reactor Recirculation Pump flow to < 100% rated flow for one pump (38.5 Mlbm/hr, 45,200 gpm), IF possible.

4.3 Do NOT enter the Region of Potential Instabilities of Attachment I of 34G0-OPS-005-2, UNLESS required for protection of plant equipment whose failure could adversely impact plant safety.

4.4 N the Region of Potential Instabilities shown on Attachment I of 34GO-OPS-005-2 is entered at any time during the performance of this procedure, take actions to exit per the STAs direction, within any limitations needed for equipment protection.

MGR-0001 Rev 4

SOUTHERN NUCLEAR PLANT E. I. HATCH PAGE 99 OF 222 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REACTOR RECIRCULATION SYSTEM 34SO-B31-001-2 42.0 WITH THE IDLE LOOP TEMPERATURE MORE THAN 75°F DIFFERENT FROM REACTOR COOLANT TEMPERATURE, THE IDLE PUMPS DISCHARGE VALVE, CAUTlON 2B31-FO3IAOR2B31-FO31B, MUST BE CLOSED BEFORE MAKING ANY SPEED CHANGES ON THE OPERATING RECIRC PUMP.

7.3.5.3.1 At the direction of the Shift Supervisor, perform one of the following.

7.3.5.3.1.1 Reduce the speed of the operating pump to less than 30% speed in accordance with 34G0-OPS-005-2, Power Changes, AND warm the idle loop in accordance with step 7.3.5.2. LI WITH THE IDLE LOOP TEMPERATURE MORE THAN 75°F DIFFERENT FROM REACTOR COOLANT TEMPERATURE, THE IDLE PUMPS DISCHARGE VALVE, 2B31-FO3IA OR 2B31-F031 B, MUST BE CLOSED BEFORE MAKING ANY SPEED CHANGES ON THE OPERATING RECIRC PUMP.

7.3.5.3.1.2 Slowly increase the speed of the operating pump to 50% speed in accordance with 34GO-OPS-005-2, Power Changes, AND warm the idle loop in accordance with step 7.3.5.4. LI 7.3.5.4 !E the speed of the operating Recirc pump is 50% or greater, adequate reverse flow may be available to allow flow through the idle recirc loop.

This can be determined by observing 2B21-R608A through W, Jet Pump Differential Pressure Instruments, indicating a differential pressure on the jet pumps associated with the idle recirc loop, panel 2H11-P619. LI 7.3.5.4.1 H the A pump is the idle Recirc Pump, THEN slowly throttle open 2B31-FO31A, Discharge Valve, as required to establish a slow heat up of the idle recirc loop. LI 7.3.5.4.2 IF the B pump is the idle Recirc Pump, THEN slowly throttle open 2B31-FO31B, Discharge Valve, as required to establish a slow heat up of the idle recirc loop.

7.3.5.4.3 WHEN the idle recirc loop temperature begins increasing, 7.3.5.4.3.1 Hthe A pump is the idle Recirc Pump immediately close 2B31 -F031 A, Discharge Valve. LI 7.3.5.4.3.2 H the B pump is the idle Recirc Pump immediately close 2B31-FO3IB, Discharge Valve. LI MGR-0001 Ver. 4

HLT-07 SRO NRC EXAM

37. 295003AK1.03 001 Unit 2 is operating at 100% RTP when the following occurs on 4160V Bus 2G:

o Normal Supply Breaker trips OPEN o Alternate Supply Breaker automatically CLOSES Which ONE of the choices below completes the following statements?

The associated Diesel Generator is expected to Three (3) minutes later, PSW flow to the Turbine Building is expected to A. remain in standby; have isolated B. remain in standby; remain in service C. have automatically started; have isolated D have automatically started; remain in service

Description:

652-302, states; 5.3 At 2H1 l-P652 Panel, confirm ALL the following:

5.3.1 EDG 2C is operating.

5.3.2 4160V Bus 2G LOSP Lockout Relay is locked out.

5.3.3 4160V Emergency Bus 2G is energized.

5.3.4 600V Emergency Bus 2D is energized.

652-3 17, states; 5.1 At Panel 2H11-P652, determine which breaker indicates OPEN.

5.2 IF 4160V Bus 2G was energized thru the normal supply breaker, confirm the alternate supply breaker is CLOSED AND 4160V Bus 2G is ENERGIZED.

Auto starts for DG:

1. Loss of power to respective 4160V bus; less than 87.5% nominal bus voltage for greater than five seconds, or less than 77.5% nominal bus voltage for greater than two seconds.
2. Reactor Water Level Low; -101
3. High Drywell Pressure; 1.85 psig NOTE: The signal that closes the alternate breaker also starts the D/G.

111

HLT-07 SRO NRC EXAM NOTE: On loss of power to the respective bus the DIG output breaker will close after the bus undervoltage 86 relay trips and diesel comes up to speed. (An additional one-second time delay is provided to allow the alternate power supply breaker an opportunity to close.) On Reactor Water Level Low or High Drywell Pressure the output breaker remains open.

The Turbine Building Isolation Valves (two valves in series in each division header) isolate on any of the following:

a. A signal from either Core Spray System LOCA logic Div 1 or Div 2.

(-101 RWL or 1.85 psig DIW pressure)

b. LOSP The valves will close after power is restored to the valves. During the power loss, the valves remain in the open position because they are MOVs and have no power to reposition.
c. Condenser Room Flooding Level switches are located in the Cond Bay floor with a small dam around each switch. The switches will pick up if water level increases to 3 above the floor level.

The A distractor is plausible if the applicant confuses/does not remember the start signal for the DG and thinks voltage has to lower to the point where the LOSP 86 Lockout Relay trips before the DG will start. The second part is plausible if the applicant confuses the Normal breaker trip/lockout alarm (652-3 17, which does not isolate PSW) with the LOSP alarm (652-302, which does isolate PSW) and thinks the PSW valves have isolated.

The B distractor is plausible if the applicant confuses/does not remember the start signal for the DG and thinks voltage has to lower to the point where the LOSP 86 Lockout Relay trips before the DG will start. The second part is correct.

The C distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses the Normal breaker trip/lockout alarm (652-317, which does not isolate PSW) with the LOSP alarm (652-302, which does isolate PSW) and thinks the PSW valves have isolated.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

112

HLT-07 SRO NRC EXAM

References:

NONE K/A:

295003 Partial or Complete Loss of A.C. Power AK1. Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: (CFR: 41.8 to 41.10)

AK1.03 Under voltage/degraded voltage effects on electrical loads 9 3.2 LESSON PLAN/OBJECTIVE:

R22-ELECT-LP-02702, 4160 VAC, EO 027.009.A.03 P4L-PSW-LP-03301, Plant Service Water System, EO 033.015.B.0l References used to develop this question:

34S0-R22-001-2, 4160 VAC System 34S0-R43-001-2, Diesel Generator Standby AC System 34S0-P41-001-2, Plant Service Water System 113

R22-ELECT-LP-02702 Page 5 of 97 4160 VAC

  • 16. Given plant conditions and 34S0-R22-00l-1/2, 4160 VAC System, PREDICT the 4160 VAC System response to the following: (027.009.A.03) a) Main Generator trip b) Faultona4l6OVACBus Loss ofnormal power to a 4160 VAC Bus.
  • 17. Given 34S0-R22-00l-1/2, 4160 VAC System and 34AB-R22-002-l/2, Loss of4160 V Emergency Bus, DETERMINE the major steps for the following 4160 VAC breaker operations: (027M08.A.04, 027.011 .A.01) a) Transfer 4160 Volt Bus power supplies.

b) Energize a de-energized 4160 VAC Bus.

c) De-energize an energized 4160 VAC Bus.

d) Restore off-site power to an emergency Bus.

18. Given plant conditions involving the 4160 VAC System, DETERMTNE if a Technical Specification Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (300.006.A. 11)
  • 19. Given plant conditions involving the 4160 VAC System, DETERMINE the Required Action(s) and Completion Time(s) in accordance with Technical Specifications for any combination of INOPERABLE systems, structures or components and the bases for the action(s). (SRO Only) (300.006.A.lO)
  • 20. Given 3lGO-OPS-021-0, Manipulation of Controls and Equipment, DETERMINE:

(300.049.A.0l) a) Whose permission is required to reset lockout relays?

b) Whose permission is required to reset relay targets?

c) Who may reset lockout relays and relay targets?

  • 21. Given the following 4160 VAC Electrical Distribution System annunciators and Annunciator Response Procedures, DETERMINE the significance of each:

(200M17.A.02) a) Station Service supply breaker tripped.

b) Loss of off-site power.

P41-PSW-LP-03301 Page 4 of 97 PLANT SERVICE WATER SYSTEM

16. Given plant conditions, DETERMINE if an auto start of the Standby Diesel Service Water pump should have occurred. (033.003 .A. 04)
  • 17. Given plant conditions, DETERMINE if an auto closure of the PSWTurb Bldg isolation valves should have occurred. (033.01 5.B.0l)
18. Given conditions requiring startup of the PSW system, DETERMINE the major steps to perform system startup per 34S0-P41-001-1/2, Plant Service Water System. (033.00 1.A.02, 033.002.A.04)
19. Given that two pumps are operating in a PSW division, SHUTDOWN one pump per 34S0-P41-00l-1/2, Plant Service Water System. (033.004.A.02)
  • 20. Given plant conditions, DETERMINE if the PSW system is properly aligned to supply the lB DG with cooling water per 34S0-P4l-005-2, Standby Diesel Service Water System, (047.006.A.03)
  • 21. In accordance with 34AB-X43-002-0, Fire Protection System Failures, DETERMINE the major steps to crosstie PSW and fire protection water. (200.024.B.Ol)
  • 22. DESCRIBE the method to use PSW for emergency fuel pooi makeup. (033.0l6.B.01)
  • 23. Given plant conditions, DETERMINE the actions necessary to mitigate the effects of a partial or complete loss of PSW per 34AB-P4l-OOl-1/2, Loss of Plant Service Water. (200.0 l3.A.06)
  • 24. Given a loss of PSW with fire water available, DETERMINE the necessary actions in accordance with 34AB-P4 1-001 1/2 to align alternate cooling to the condensate pumps andlor MVP. (200.013.G.01)
  • 25. Given plant conditions, and that the PSW Tm-b Bldg isolation valves have auto isolated, DETERMINE if the auto closure of the PSW Turb Bldg isolation valves can be overridden per 34AB-P41-OOl-l/2, Loss of Plant Service Water. (033.0l5.B.03)
  • 26. Given plant conditions, PREDICT the effect on plant operations as a result of a loss of PSW to the following systems: (200.0 13.A.05)
a. Turbine/Generator Auxiliaries
b. Drywell Chillers (Unit 2)
c. Diesel Generators
  • 27. Deleted (200.052.K.0l)
  • 28. Given plant conditions involving PSW system, DETERMINE if a Technical Specification Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (300.01 0.C.0 1)

SOUTHERN NUCLEAR PLANT E. I. HATCH I PAGE7OF 112 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

4160 VAC SYSTEM 34SO-R22-001-2 18.1 5.1.9 Use the applicable attachment, Attachment 7, 8,9, 10, to obtain the incident energy level for the bus to be racked, and refer to NMP-SH-003, Electrical Work Practices, for the protective gear requirements based on this energy level.

5.2 LIMITATIONS 5.2.1 Undervoltage (Tech. Specs.: Loss Of Voltage (2800 for 6.5 Secs.),

Degraded Voltage (3280 for 21 .5 Secs.)) on 41 60V Bus 2E, 2F OR 2G, will cause the normal supply breaker to trip, the Diesel Generator 2A, B, OR 2C to start afternate (Startup) supply breaker to close.

5.2.2 IF alternate power supply undervoltage occurs, the 86 relay drops out, the normal supply breaker locks out, the alternate supply breaker trips locks out fjQ load is dropped from bus. WHEN Diesel Generator voltage is up to 90% rated, the Diesel Generator breaker closes to energize the bus. Loads are sequentially placed on the bus by automatic timers. The 600V buses 2C AND 2D are energized WHEN the Diesel Generator breakers close.

5.2.3 IF one of the Emergency Diesel Generator Mode switches is in TEST, the MANUAL closure of start-up Transformer supply breakers to 4160V Busses 2A, 2B, 2C AND 2D is prevented. (AUTO fast transfer will still occur.)

5.2.4 IF any Emergency 41 60V bus (2E, 2F, OR 2G) is on alternate supply (powered from SAT 2C),

4160 VAC buses 2A, 2B, 2C, AND 2D will NOT Auto transfer to alternate supply.

However 20 AND 2D can be manually transferred to Alternate Supply (SAT 2D).

5.2.5 IF either Station Service 4160V bus 2C Q 2D is on alternate supply, THEN Reactor power must be maintained 2558 Mwth due to loading considerations on Start-Up Aux Transf 2D.

5.2.6 IF any Emergency 41 60V bus (2E, 2F OR 2G) undervoltage condition occurs Q the alternate breaker closes in for 4160V bus 2E, 2F OR 2G, THEN the alternate supply breaker to 41 60V buses 2A AND 2B will trip OPEN.

MGR-0001 Ver. 4

SOUTHERN NUCLEAR PLANTE.I.HATCH I I PAGE200FI12 DOCUMENT TITLE DOCUMENT NUMBER: VERSION NO:

4160 VAC SYSTEM 345O-R22-001-2 18.1 7.3 INFREQUENT OPERATIONS 7.3.1 Restoration Of Power To 4160v Emergency Bus (2E, 2F, 2G) From Normal Power Supply (SAT 2D) 7.3.1.1 Restoration Of Power To 4160V Emergency Bus 2E, From Normal Power Supply (SAT 2D) lEJ

. This procedure is used to restore power to a de-energized emergency bus WHEN normal power has been restored AND the bus is NOT being supplied from alternate NOTES: power (SAT 2C) Q the appropriate diesel generator.

. All operations are performed from panel 2H1 1-P652 UNLESS otherwise indicated.

7.3.1.1.1 Confirm offsite power has been restored to SAT 2D per 34SO-S22-001-1, 230KV Substation Switching, OR 34S0-S22-001-2, 500KV Substation Switching. El 7.3.1.1.2 IF restoration of power to 4160V emergency bus 2E is to be performed, place 2P41-S33, PSW Override switch, in the OVERRIDE position, to prevent closing 2P41-F316A I 2P41-F316D, PSW to Turbine Building Isol VIvs.

7.3.1.1.3 IF desired, confirm diesel generator 2A is tripped AND tagged out. LI 73.1 .1.4 IF desired, confirm Alternate (Startup) Supply Breaker, ACB 135544 is racked out. LI 7.3.1.1.5 IF SAT 2C is DE-ENERGIZED, place 2A Diesel Test SAT 20 OUT OF SVC. Interlock switch to TEST. El FOTE: IF LOCA signal is present, D/G will NOT go into test mode.

7.3.1.1.6 Place Diesel Generator 2A Mode Select Switch in Test. El 7.3.1.1.7 Place Sync Switch (SSW) ACB 135554 in ON. LI MGR-0001 Ver. 4

SOUTHERN NUCLEAR PLANTE. I. HATCH P A G E 4OF8 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF 4160V EMERGENCY BUS 34AB-R22-002-2 1.11 4.5 HZ 41 60V Bus 2E OR 2G is lost, enter the fo[lowing applicable procedures AND perform concurrently with this procedure:

  • 34AB-T47-001-2, Complete Loss of Drywell Cooling 4.6 IF the Fuel Pool Cooling System is unavailable, enter 34AB-G41-001-2, Loss of Fuel Pool Cooling.

4.7 WHEN SAT 2C OR 2D becomes available:

4.7.1 bus is deenergized, energiZG the bus from the available SAT in accordance with 34SO-R22-001-2.

4.7.2 IF bus is energized from associated Diesel Generator, transfer the bus to available SAT in accordance with 34S0-R43-001-2.

4.8 IF there is a fault on bus, investigate and correct IF possible, THEN restore the bus in accordance with 34SO-R22-001-2.

4.9 OPEN 2P41-F316A, 2P41-F316B, 2P41-F316C, OR2P41-F316D to restore PSW flow to Turbine Building per 34AB-P41001-2, Loss of Plant Service Water.

4.10 If applicable, monitor running diesel generator oil levels make arrangements in advance to add oil, W needed.

4.11 If applicable, monitor running diesel generator fuel quantity AND make arrangements in advance to transfer AND/OR add fuel, HZ needed.

5.0 REFERENCES

5.1 Technical Specifications 3.8, Electrical Power Systems 5.2 Plant Drawings

  • H-23357, Plant Hatch Unit Two Single Line Diagram for 4160V Bus 2E and 2F
  • H-23358, Plant Hatch Unit Two Single Line Diagram for 4160V Bus 2G MGR-0001 Ver. 2

SNC PLANT E. I. HATCH I IPQ 118 of 130 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 25.0 ATTACHMENT 7 Aft. Pg.

TITLE: D/G TEST, LOSP, LOCA, LOGIC DIAGRAMS AND D/G GENERAL INFORMATION lof6 Figure 1 1A, IC, 2A, 2C D/G TEST Logic Mode Select Switch Aux Relay& ..LOSP Lock Out Relay 43A1X 1. Allows closure of normal supply breaker if DG supply breaker is closed and no LOSP is sensed.

2. Locks out emergency start ckt. A.
3. Prevents auto closure of DG supply breaker.
4. Prevents auto closure of alternate supply breaker.
5. Allows auto or manual closure of alternate supply when DG breaker is closed if no LOSP is sensed.

43A1X1 1. Locks out emergency start ckt. B.

2. Locks out anti-parallel relay trip.
3. Allows manual closure of DG supply breaker.
4. Arms additional DG trips (0TH, 0TH, CPL, CCPH).

43A1X2 1. Prevents load shed when DG supply breaker is closed.

43A1X3 1. Cannot manually close SUT supplies for 4160V A, B, C, & D busses if any DG is in TEST mode.

2. Alternate supply for 41 60V A, B, C, & D buses will not fast transfer if alternate supply to emergency bus is closed.

G16.030 MGR-0009 Ver. 4

HLT-07 SRO NRC EXAM

38. 295004G2.4.11 001 Unit 2 was operating at 100% power.

o A loss of 2R25-S002, 125 VDC Cabinet 2B, occurs o Subsequently, a malfunction occurs causing 2B31-FO31B, Recirc 2B Discharge Valve, to inadvertently travel full CLOSED With the above conditions, which ONE of the choices below completes the following statements?

As 2B31-FO3IB CLOSES, the 2B ASD A. will perform a tRamp Shutdown to zero rpm and then the 4160V Input Breaker must be manually OPENED from the Control Room B. will perform a Ramp Shutdown to zero rpm and then the 4160V Input Breaker will trip OPEN C 4160V Input Breaker will remain closed UNTIL it is manually OPENED LOCALLY D. 4160V Input Breaker will IMMEDIATELY trip OPEN 114

HLT-07 SRO NRC EXAM

==

Description:==

JAW section 1V 125 VDC Cabinet 2B, of 34AB-R22-001-2, Loss of DC, 2R22-S002, the following Caution exists:

REACTOR RECIRC PUMP 2A & 2B WILL BE RUNNING WITHOUT INPUT BREAKER AUTOMATIC TRIP CAPABILITY. ASD 2A & 2B CAUTION: SHUTDOWN PUSHBUTTONS MUST BE USED TO SHUTDOWN REACTOR RECIRC PUMP 2A & 2B AND THEN THE INPUT BREAKERS SHOULD BE OPENED LOCALLY PER 34SO-R22-001-2.

When the discharge valve is <90% open, a direct trip to the ASD exists. However, with the loss of 2R25S002, the ASD Jiiput Breaker will remain closed and require local operation to open the breaker.

lAW 34S0-B31-OO1-2, not all Fatal Faults will trip the 4160V Jnput BKR, some Fatal Faults activate a Ramp Shutdown where the Drive will Ramp to minimum speed (22% or 370 rpm),

then go to 0 rpm and drive flow coasts to 0 gprn. The 2B31-SOO2A or 2B31-SOO2B, ASD A or ASD B control switch, on 2H1 l-P602, will have to be taken to Stop to trip the Input BKR.

The A distractor is plausible if the applicant confuses a ramp shutdown with a trip of the input breaker and also does not consider the impact of a loss of DC on both the Input Breaker. This loss of DC will require the Input Breaker to be open locally since no control exists for Control Room manipulation.

The B distractor is plausible if the applicant confuses a ramp shutdown with a trip of the input breaker and also does not consider the impact of a loss of DC on both the Input Breaker. In this case, the applicant will think the ASD ramps down and then the Input Breaker will open.

The D distractor is plausible since it would be correct if the loss of DC bus did not exist.

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

115

HLT-07 SRO NRC EXAM

References:

NONE K/A:

295004 Partial or Complete Loss of D.C. Power 2.4.11 Knowledge of abnormal condition procedures.

(CFR:41.10/43.5/45.13) 4.0 4.2 LESSON PLAN/OBJECTIVE:

R42-ELECT-LP-02704. DC Electrical Distribution, EO 200.018.A.03 References used to develop this Question:

34AB-R22-O01-2, Loss of DC 34S0-B3 1-001-2, Reactor Recirculation System 116

R42-ELECT-LP-02704 Page 3 of 95 DC ELECTRICAL DISTRIBUTION

11. Per 34SV-SUV-0 13-0, Weekly Breaker Alignment Check, and initial plant conditions, DETERMINE if Offsite Power Supply requirements have been satisfied. (027.039.A.06)
12. Per 34AB-R22-001 -1/2, Loss of DC Buses, DETERMINE: (200.01 8.A.03)
a. The reason for the automatic action(s).
b. What is accomplished by the automatic action(s).
c. Significance of the notes and cautions.
  • 13. Given a loss of one 125/250 VDC Bus, per 34AB-R22-001-1/2, Loss of DC Buses, and System Operating Procedures, PREDICT the limitations on plant operation, limited to FWLC, available injection systems, turbine operation, 4160 VAC power supplies available, (normal and emergency), RPV Water Level Instrumentation, and Reactor Pressure Instrumentation.

(200.01 8.A.05)

  • 14 Given a loss of one 24/48 VDC Bus, per 34AB-R22-001-1/2, Loss of DC Buses, and System Operating Procedures, DESCRIBE the effect on Nuclear Instrumentation and any limitations these effects would have on plant start-up. (200.0 18.A.06)
  • 15. Given initial plant conditions and a copy of Technical Specifications, DETERMINE when a Breaker Alignment Surveillance is required. (027.039.A.03)
16. Given plant conditions involving the DC Electrical Distribution System, DETERMINE if a Technical Specification Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (3 00.006.A.07)
17. Given plant conditions involving the DC Electrical Distribution System, DETERMINE the Required Action(s) and Completion Time(s) in accordance with Technical Specifications for any combination of INOPERABLE systems, structures or components and the bases for the action(s).

(SRO Only) (300.006.A.08)

  • 18. Given plant conditions involving the DC Electrical Distribution System, DETERMINE the Required Actions for Completion Times < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in accordance with Technical Specifications for any combination of INOPERABLE systems, structures or components. (300.006.A.09)
  • 19. Given plant conditions involving the DC Electrical Distribution System, DETERMINE if a Technical Requirements Manual (TRM) Limiting Condition for Operation has been exceeded.

(implicit in this objective is a determination ofAPPLICABILITY and associated NOTES)

(300.006.A. 10)

  • 20. Given plant conditions involving the DC Electrical Distribution System, DETERMINE the Required Action(s) and Completion Time(s) in accordance with the Technical Requirements Manual (TRM) for any combination of INOPERABLE systems, structures or components and the bases for the action(s). (SRO Only) (300.006.A. 11)

Objectives marked by a RED (*) are required during RO-305 and SR-305 of the Initial License program.

SOUTHERN NUCLEAR PLANT E. I. HATCH PAGE 12 OF 222 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REACTOR RECIRCULATION SYSTEM 34S0-B31-001-2 42.0 5.2.2 The ASD will trip (input breaker opens) upon receipt of any of the following FATAL FAULT signals:

5.2.2.1 ASD Input Line Undervoltage, <55% / 2288V 5.2.2.2 ASD Input Line Overvoltage, >120% I 4992V 5.2.2.3 Rectifier Low Level / Input One Cycle (excessive reactive current flow in xfmr) 5.2.2.4 Recirc. Pump discharge valve <90% OPEN

  • Bypassed for ASD start
  • Must be off the CLOSED seat WITHIN 3 seconds of the pump start
  • Must be> 90% OPEN WITHIN 96 seconds of the pump start 5.2.2.5 Recirc. Pump suction valve < 90% OPEN 5.2.2.6 ATWS
  • Low reactor water level of -60
  • High reactor pressure of 1170 psig 5.2.2.7 Secondary Winding Low Level / Excessive Drive Losses, >3 Power Cells bypassed I faulted 5.2.2.8 Loss of Both NXG Computer Control Power Supply A & B 5.2.2.9 ASD lockout relay tripped 5.2.2.10 RPT breaker OPEN 5.2.2.10.1 H EOC RPT is operable, breakers 2C71-CB3A, CB3B, CB4A, AND CB4B will trip on the following:
  • Turbine stop valve < 90% OPEN at> 27.6% reactor power 5.2.2.10.2 ONLY breakers 2C71-CB3A AND CB3B will trip on the following:
  • Recirc motor time delayed overcurrent
  • Recirc motor instantaneous differential overcurrent 5.2.2.11 Manual trip by using the Emergency Stop Pushbutton locally at ASD 5.2.2.12 Transformer Winding Overtemp (Transformer High Temp >176F, AND Transformer Very High Temp >194F, AND a one minute time delay)

MGR-0001 Ver. 4

SOUTHERN NUCLEAR PLANT E. I. HATCH PAGE 14 OF 74 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF DC BUSES 34AB-R22-001-2 113 IV. LOSS OF 125V DC CABINET 2B, 2R25-S002 1.0 CONDITIONS 1.1 Loss of DC control power indicating lights on Panel 2H11-P651.

1.2 ANNUNCIATORS

  • HPCI SYSTEM INVERTER CIRCUIT FAILURE, 601-1 20
  • HPCI LOGIC BUS POWER FAILURE, 601-126
  • AUTO BLOWDOWN CONTROL POWER FAILURE, 602-317
  • RCIC LOGIC OR TORUS LVL LOGIC POWER FAILURE, 602-325
  • COND STOR TANK LEVEL LOW, 601-1 33
  • RHR RELAY LOGIC B POWER FAILURE, 601-207
  • 4160V BUS 2G OR 600V BUS 2D DC OFF, 652-315
  • STA SVC SWGR DC OFF, 651-143 1.3 Loss of power to equipment listed on Attachment 8.

2.0 AUTOMATIC ACTIONS None 3.0 IMMEDIATE OPERATOR ACTIONS None MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGEI5OF74 PLANTE. I. HATCH I DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF DC BUSES 34AB-R22-OO1-2 3.13 4.0 SUBSEQUENT OPERATOR ACTIONS REACTOR RECIRC PUMP 2A & 2B WILL BE RUNNING WITHOUT INPUT BREAKER AUTOMATIC TRIP CAPABILITY. ASD 2A & 2B SHUTDOWN PUSHBUTTONS MUST BE USED TO SHUTDOWN REACTOR RECIRC PUMP 2A & 2B AND THEN THE INPUT BREAKERS SHOULD BE OPENED LOCALLY PER 34S0-R22-OO1-2.

4.1 IF a turbine trip with power 27.6% RTP has occurred AND Rx Recirc 2A should have tripped, OR IF required to trip Rx Recirc 2A, perform the following:

4.1.1 Perform Abnormal Recirc Pump 2A Shutdown per 34S0-B31-OO1-2. El 4.1.2 Locally trip Reactor Recirc Pump 2A Input Breaker at 2R22-S002 per 34S0-R22-OO1-2. El 4.1.3 Enter 34AB-B31-OO1-2, Reactor Recirculation Pump(s) Trip, or Recirc Loops Flow Mismatch, or ASD Power Cell Failure. LI 4.2 IF a turbine trip with power 27.6% RTP has occurred INJ2 Rx Recirc 2B should have tripped, OR IF required to trip Rx Recirc 2B, perform the following:

4.2.1 Perform Abnormal Recirc Pump 2B Shutdown per 34SO-B31-OO1-2. El 4.2.2 Locally trip Reactor Recirc Pump 2B Input Breaker at 2R22-S002 per 34SO-R22-OO1-2. El 4.2.3 Enter 34AB-B31-OO1-2, Reactor Recirculation Pump(s) Trip, or Recirc Loops Flow Mismatch, or ASD Power Cell Failure. El 4.3 Attempt to restore power to Cabinet 2B, 2R25-S002 as follows:

4.3.1 Check 125!250V DC Swgr 2B, 2R22-S017 Frame iT feeder breaker. El 4.3.1.1 IF CLOSED, Refer To Section II of this procedure. El 4.3.1.2 IF OPEN, manually close. El MGR-0001 Ver. 3

HLT-07 SRO NRC EXAM

39. 295005AK3.06 001 Unit 2 was operating at 100% RTP when the following occurred:

o Startup Aux. Transformer (SAT) 2D de-energized o Reactor scrammed o Main Generator tripped Which ONE of the choices below describes the expected status of the 4160 VAC Station Service Buses 2A and 2B and why? (Assume NO operator actions have been taken).

4160V buses 2A and 2B are A de-energized to prevent overloading SAT 2C B. de-energized to prevent paralleling out of synch power supplies C. energized from SAT 2C to keep the Main Condenser available D. energized from SAT 2C to keep forced Core flow available 117

HLT-07 SRO NRC EXAM

==

Description:==

When SAT 2D de-energizes, the Emergency Buses transfer to the alternate SAT 2C supply.

When the Main Turbine tripped, if any of the 4160 VAC Emergency Busses are tied to the 2C SAT, a fast transfer of house loads is prohibited and the SAT supply breakers to 4160 2A and B receive a trip signal. Manual transfer to SAT 2D is allowed for the 4160 VAC Busses 2C and 2D only, and is required if the generator is no longer available. The purpose of this interlock is to minimize the loading of the 2C SAT. SAT 2C is smaller than 2D transformer and is not sized to carry the emergency loads and the 2A and 2B 4160 VAC bus loads at the same time.

The B distractor is plausible if the applicant understands the 2A & 2B will be de-energized but confuses the reason why and thinks parrelling out of synch power supplies is applicable. If the applicant confuses that the Emergency Bus is powered by the DG (which has auto started), the Emergency bus would be Out of synch with Station Service buses (2A & 2B) therefore adding plausibility about parrelling the Emergency Bus/DG with 4160V buses.

The C distractor is plausible if the applicant confuses the expected status of the buses and thinks they are energized from their respective UAT and does not realize that the UAT is de-energized. Circ Water pump reason is plausible since at least 1 Circ Water pump in service would be needed for a normal cooldown to the condenser.

The D distractor is plausible if the applicant confuses the expected status of the buses and thinks they are energized from their respective SAT. Recirc pump reason is plausible since at least 1 Recirc pump in service would be needed to prevent thermal stratification after a scram.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

118

HLT-07 SRO NRC EXAM

References:

NONE K/A:

295005 Main Turbine Generator Trip AK3. Knowledge of the reasons for the following responses as they apply to MAIN TURBINE GENERATOR TRIP: (CFR: 41.5/ 45.6)

AK3.06 Realignment of electrical distribution 3.3 3.3 LESSON PLAN/OBJECTIVE:

R22-ELECT-LP-02702, 4160 VAC, EO 200.0 17.A.03 References used to develop this question:

34SO-R22-001-2, 4160 VAC System Unit Two Tech Spec Bases 3.8.1, AC Sources Operating (Overloading SAT 2C) 119

R22-ELECT-LP-02702 Page 4 of 97 4160 VAC

  • 8. Given the following list of control room controls for the 4160 VAC electrical distribution system, STATE the function of each: (027.0 10.A.02)
a. Interlock cutout switches.
b. LOSP lockout relay.
  • 9* Given plant procedures, DETERMINE the proper operator actions to RACKOUT a 4160 VAC Breaker. (027.012.A.01)
10. Given plant procedures, DETERMINE the proper operator actions to RACKIN a 4160 VAC Breaker. (027.0 15.A.02)
11. Given plant conditions, correctly DETERfVIINE if 4160 VAC busses are energized or de-energized. (200.0 17.A.03)
12. Given a simplified drawing of the 4160 VAC electrical distribution system, correctly IDENTIFY the following system interfaces: (027.01 0.A.0 I)
a. Startup Transformers
b. Unit Auxiliary Transformers
c. 600 Volt Distribution Buses
  • 13. Given plant procedures, DESCRIBE the actions necessary to override the loss of control power trip to selected loads from 4160 VAC Emergency Bus F. (027.065.A.02)

(027.009.A.01) a) Will auto transfer when the Main Generator PCBs open.

b) May be operated from the Control Room to energize a 4160 VAC Bus from its alternate source.

15. Given 34S0-R22-00 1-112, 4160 VAC System, a 4160 VAC Emergency Bus is being supplied from its alternate source, and a plant transient occurs, DETERMINE which 4160 VAC breakers: (027.009.A.02) a) Will auto transfer when the Main Generator PCBs open.

b) May be operated from the Control Room to energize a 4160 VAC Bus from its alternate source.

R22-ELECT-LP-02702 Page 27 of 97 4160 VAC This operation normally occurs following a manual or auto turbine trip. It ensures that all loads remain energized.

d. There are several conditions in the electrical distribution system which will lockout (prevent) a fast transfer. Those conditions are as follows:
1) If the fast transfer does not occur within 0.2 seconds, the automatic fast transfer is locked out, requiring a manual transfer to re-energize the 4160 VAC Station Service Busses 2A, B, C, and D.
2) If any of the 4160 VAC Emergency Busses are tied to the 2C SAT, a fast transfer ofhouse loads is prohibited and the SAT supply breakers to 4160 2A and B receive a trip signal.

a) Manual transfer to SAT 2D is allowed for the 4160 VAC Busses 2C and 2D only, and is required if the generator is no longer available.

b) The purpose of this interlock is to minimize the loading of the 2C SAT. SAT 2C is smaller than 2D transformer and is not sized to carry the emergency loads 4 the 2A and 2B 4160 VAC bus loads at the same time.

3) The SAT supply breakers (alternate supply) to the station service 4160 VAC buses A, B, C, and D cannot be manually closed with their control switch if any diesel generator on the associated unit is in the test mode.

Having the diesel in test does NOT prevent fast transfer.

4) If t C SUT voltage is <105 volts, sensed by Undervoltage Relays lS32Kl24 & K125 and 2S32K772-l & K772-2 NOTE: The purpose of this interlock is to prevent the DIG from undergoing a load transient when reenergizing a Bus with loads attached (high starting currents). An auto fast transfer is allowed because there are no starting currents involved.
2. Trips of station service bus supply breakers (SO-8) (LT-6)
a. Supply breakers will trip on an overcurrent for that line.
b. Supply breakers will trip on a supply transformer fault.
c. Condenser Bay flooding trips the normal and alternate supply breakers to 4160 VAC buses 2A(1A) and 2B(1B). This immediately de-energizes the buses to ensure condenser circulating water pumps are tripped. (They are assumed to be the source of the flooding.)

SOUTHERN NUCLEAR PLANT E. I. HATCH PAGE 7 OF 112 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

4160 VAC SYSTEM 34SO-R22-001-2 18.1 5.1.9 Use the applicable attachment, Attachment 7,8,9, 10, to obtain the incident energy level for the bus to be racked, and refer to NMP-SH-003, Electrical Work Practices, for the protective gear requirements based on this energy level.

5.2 LIMITATIONS 5.2.1 Undervoltage (Tech. Specs.: Loss Of Voltage (2800 for 6.5 Secs.),

Degraded Voltage (3280 for 21 .5 Secs.)) on 41 60V Bus 2E, 2F OR 2G, will cause the normal supply breaker to trip, the Diesel Generator 2A, B, OR 2C to start AND alternate (Startup) supply breaker to close.

5.2.2 IF alternate power supply undervoltage occurs, the 86 relay drops out, the normal supply breaker locks out, the alternate supply breaker trips AND locks out AND load is dropped from bus. WHEN Diesel Generator voltage is up to 90% rated, the Diesel Generator breaker closes to energize the bus. Loads are sequentially placed on the bus by automatic timers. The 600V buses 2C AND 2D are energized WHEN the Diesel Generator breakers close.

5.2.3 IF one of the Emergency Diesel Generator Mode switches is in TEST, the MANUAL closure of start-up Transformer supply breakers to 4160V Busses 2A, 2B, 20 AND 2D is prevented. (AUTO fast transfer will still occur.)

5.2.4 IF any Emergency 41 60V bus (2E, 2F, OR 2G) is on alternate supply (powered from SAT 2C),

4160 VAC buses 2A, 2B, 20, AND 2D will NOT Auto transfer to alternate supply.

However 2C AND 2D can be manually transferred to Alternate Supply (SAT 2D).

5.2.5 IF either Station Service 4160V bus 2C OR 2D is on alternate supply, THEN Reactor power must be maintained <2558 Mwth due to loading considerations on Start-Up Aux Transf 2D.

5.2.6 IF any Emergency 41 60V bus (2E, 2F OR 2G) undervoltage condition occurs Qi, the alternate breaker closes in for 41 60V bus 2E, 2F OR 2G, THEN the alternate supply breaker to 41 60V buses 2A AND 2B will trip OPEN.

MGR-0001 Ver. 4

AC Sources Operating B 3.8.1 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.1 AC Sources Operating BASES BACKGROUND The Unit 2 Class 1 E AC Electrical Power Distribution System AC sources consist of the offsite power sources (preferred power sources, normal and alternate), and the onsite standby power sources (diesel generators (DG5) 2A, 2C, and 1B). As required by 10 CFR 50, Appendix A, GDC 17 (Ref. 1), the design of the AC electrical power system provides independence and redundancy to ensure an available source of power to the Engineered Safety Feature (ESF) systems.

The Class 1E AC distribution system is divided into redundant load groups, so loss of any one group does not prevent the minimum safety functions from being performed. Each load group has connections to two preferred offsite power supplies and a single DG.

Offsite power is supplied to the 230 kV and 500 kV switchyards from the transmission network by eight transmission lines. From the 230 kV switchyards, two electrically and physically separated circuits provide AC power, through startup auxiliary transformers 2C and 2D, to 4.16 kV ESF buses 2E, 2F, and 2G. A detailed description of the offsite power network and circuits to the onsite Class 1 E ESF buses is found in the FSAR, Sections 8.2 and 8.3 (Ref. 2).

An offsite circuit consists of all breakers, transformers, switches, interrupting devices, cabling, and controls required to transmit power from the offsite transmission network to the onsite Class 1 E ESF bus or buses.

Startup auxiliary transformer (SAT) 2D provides the normal source of power to the ESF buses 2E, 2F, and 2G. If any 4.16 kV ESF bus loses power, an automatic transfer from SAT 2D to SAT 2C occurs.

At this time, 4.16 kV buses 2A and 2B and supply breakers from SAT 2C also trip open, disconnecting all nonessential loads from SAT 2C to preclude overloading of the transformer.

SATs 2C and 2D are sized to accommodate the simultaneous starting of all required ESF loads on receipt of an accident signal without the need for load sequencing. However, ESF loads are sequenced when the associated 4.16 kV ESF bus is supplied from SAT 2C.

A description of the Unit 1 offsite power sources is provided in the Bases for Unit 1 LCO 3.8.1, AC Sources Operating.

(continued)

HATCH UNIT 2 B 3.8-1 REVISION 1

HLLO7 SRO NRC EXAM

40. 295006AA1.01 001 Unit 2 is in a refueling outage.

o Reactor Mode Switch position REFUEL o SRM Shorting Links REMOVED A SRM detector failure results in the SRM 2A indication reading 5x10 5 CPS.

What is the expected status of the 2111 1-P603 panel WHITE RPS Scram Group A and B lights?

RPS Scram Group A WHITE lights are expected to be RPS Scram Group B WHITE lights are expected to be A EXTINGUISHED; EXTINGUISHED B. EXTINGUISHED; ILLUMINATED C. ILLUMINATED; EXTINGUISHED D. ILLUMINATED; ILLUMINATED 120

HLT-07 SRO NRC EXAM

==

Description:==

The UPSCALE TRIP (High High scram) is used only during refueling operations (setpoint 3 x 105 cps). With the shorting links removed, a single trip signal from any nuclear instrument channel (8 TRMs or 4 SRMs) or a single 2/4 voter module will cause a full reactor scram. With the shorting links installed, the SRMs will not cause a trip of RPS.

The 2A SRM has exceeded the trip setpoint and will cause RPS bus A & B to trip causing ALL RPS Scram Group Lights to extinguish.

The B distractor is plausible if the applicant thinks that the C SRM input into the A side of RPS and does not remember the shorting link effect.

The C distractor is plausible if the applicant thinks that the C SRM input into the B side of RPS and does not remember the shorting link effect.

The D distractor is plausible if the applicant thinks that SRMs can not give a RPS trip and does not remember the Upscale trip setpoint. This distractor would be correct if the shorting links were installed.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

References:

NONE K/A:

295006 SCRAM AA1. Ability to operate and/or monitor the following as they apply to SCRAM:

(CFR: 41.7 / 45.6)

AA1.01 RPS 4.2* 4.2*

LESSON PLAN/OBJECTIVE:

C7 1 -RPS-LP-0 1001, Reactor Protection System, EO 300.008 .A.02 121

HLT-07 SRO NRC EXAM LM-SKIV1-Lt-ULLUI, Source Kange MOmtors, to UILUU3.A.1 I References used to develop this question:

34AR-603-204-2, SRM Upscale or limperative Modified from 2011 HLT-6 NRC Exam Q#13 ORIGINAL QUESTION (HLT 6 NRC Exam Q# 13)

Unit 2 is in a refueling outage.

o Reactor Mode Switch position REFUEL o SRM Shorting Links REMOVED SOURCE RANGE MONITOR LEVEL A B C 0 2C51.R600A 2C51-R600B 2C51.R600C 2C51-R6000 24188V0C CAB 2A 24143V0C CAB 28 23J4SVDC CAB 2A 24MSVOC CAB 28 6

io c io- c iO6 c 0 0 0 0 io- 10- u 10i 10 u E E. N N N 10 10 10

=

T T T T s ES S S

-E 3

io p p p p E jQ2 E £ 2 E io2 R R 10- R R

E E E E g 100_ i00 g oo. g N - N N N 101_ 0 o 10 D 0 r r A SRM detector failure results in the indication as shown in this figure.

With NO operator action, based on these conditions, what is the expected status of RPS Channel A and B?

RPS Channel A expected to be tripped.

RPS Channel B expected to be tripped.

A. is NOT; is NOT B. is NOT; is 122

HLT-07 SRO NRC EXAM C.v is; is D. is; is NOT 123

C71-RPS-LP-OlOO1 Page 6 of 112 REACTOR PROTECTION SYSTEM

  • 30. In accordance with 34G0-OPS-013-1/2, Normal Plant Shutdown, DISCUSS the notes and precautions concerning the transfer of the Mode Switch from RUN to START UP/HOT STANDBY. (0l0.019.a.04)
31. Given plant conditions which require tripping a channel of RPS, SUMMARIZE the steps necessary to place an RPS channel in the Tripped condition according to Tech Specs.

(010.012.a.Ol)

  • 32. Given plant conditions involving the RPS System, DETERMINE if a Technical Specification Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (300 .006.a. 19)
  • 33* Given plant conditions involving the RPS System, DETERMINE the Required Action(s) and Completion Time(s) in accordance with Technical Specifications for any combination of INOPERABLE systems, structures or components and the bases for the action(s). (SRO Only)

(300.010.a.14)

  • 34* Given plant conditions involving the RPS System, DETERMINE the Required Actions for Completion Times 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in accordance with Technical Specifications for any combination of INOPERABLE systems, structures or components. (300.011 .a.07)
35. Given a failed IRM/APRM, DESCRIBE the steps necessary to bypass the failed instrument per 34AR-603-2l9-1/2, APRM Upscale, or 34AR-603-203-1/2, IRM Bus A Upscale Trip or INOP. (010.OlO.a.01)
36. STATE the possible effects on the Plant, if a spurious half scram is repeatedly tripping and being reset over a short period of time. (010.0 ll.a.02)
  • 37 Given plant conditions which resulted in a Reactor Scram, using plant procedures and Tech Specs, DETERMINE the cause of the Reactor Scram. (300.008.a.02)
  • 38. Given that a loss of an RPS bus has occurred, SUMMARIZE the actions to correctly respond to these conditions per 34AWC71-002-l/2, Loss of RPS. (200.102.c.Ol)
39. Given a list, IDENTIFY the statement that describes the purpose of pulling the scram solenoid fuses during an ATWS. (010.0 16.a.01)
40. In accordance with 3 1RS-OPS-001-1/2, Shutdown from Outside the Control Room, LOCATE the breakers used to de-energize the APRMs and INITIATE a scram. (010.017.a.01)
41. In accordance with 31RS-OPS-OOl-l/2, Shutdown from Outside the Control Room, SUMMARIZE the steps necessary to trip the Scram Discharge Volume High Level switches.

(010.01 8.a.02)

42. DESCRIBE the reason for the 10-second time delay between receiving a reactor scram signal and resetting the signal. (010.0 l0.A.03)
  • 43* Given plant conditions involving the RPS System, DETERMINE if a Technical Requirements Manual (TRM) Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (300.010. C.02)

C51-SRM-LP-01201 Page 3 of 52 SOURCE RANGE MONITORS

7. DESCRIBE the operation of the following SRM controls and indications (012.003.A.06)
a. SRM Drive in/Drive out controls
b. SRM Meters/Recorders (cps) P603/P606
c. SRM Period Indication P603/P606
d. SRM Bypass Switch
e. SRM upscale/downscale indication
8. Given plant conditions, ANALYZE these conditions to DETERMINE if an SRM generated Rod Block should have occurred. (012.003.A.l0)
  • 9* Given plant conditions, ANALYZE these conditions to DETERMINE if an SRM generated reactor scram should have occurred. (012.003.A.l 1)
10. Given plant conditions, ANALYZE these conditions to DETERMINE if an SRM bypass should have occurred. (012.003.A.12)
11. Given plant conditions, DETERMINE correct SRM response during a reactor startup per 34G0-OPS-00 1-1/2, Plant Startup. (400.027.A.02)
12. Given plant conditions with reactor startup in progress, ANALYZE plant conditions and determine when the reactor is critical per 34G0-OPS-001-l/2, Plant StartupH. (0l2.0l1.A.01)
13. Given plant conditions with reactor startup in progress, DETERMINE when correct SRM/IRM overlap has been demonstrated per 34G0-OPS-00l-l/2, Plant Startup. (012.009.A.01)
14. Given plant conditions with reactor startup in progress, DETERMINE correct operator action if proper SRM/IRM overlap is not confirmed per 34G0-OPS-00l-l/2, Plant Startup.

(01 2.009.A.02)

15. Given plant conditions with reactor startup in progress, DETERMINE the following per 34G0-OPS-00l-l/2, Plant Startup: (012.006.B.01)
a. When the operator should start withdrawing the SRM detector.
b. Required SRM cps to be maintained during withdraw.
c. How many SRM detectors can be withdrawn at a time.
d. When SRM detectors should be fully withdrawn.
  • 16. Given plant conditions, DETERMINE the following per 34G0-OPS-013-l/2, Normal Plant Shutdown: (012.005.A.0l)
a. When the operator should start inserting the SRM detectors.
b. Reason for positioning SRM detector to maintain 200 to 7 x bE 4 cps.
c. When SRM detectors should be fully inserted.

1.0 IDENTIFICATION

ALARM PANEL 603-2 z:::z: SRM UPSCALE OR INOPERATIVE DEVICE: SETPOINT:

2C51-K600A, 2C51-K600B, Upscale 7 X lO

- cps 4

2C51-K600C, or 2C51-K600D Inoperative Switch not in OPERATE, Power supply voltage low or circuit boards not in circuit.

2.0 CONDITION

3.0 CLASSIFICATION

EQUIPMENT STATUS One or more of the source range monitors are upscale or inoperative and the reactor mode switch is not in RUN.

2H1 1-P603 Panel 603-2 5.0 OPERATOR ACTIONS:

5.1 Confirm one or more of the SRM Upscale or mop lights are ILLUMINATED, 2H11-P603. El 5.2 H a SRM count meter indicates greater than 7 X lO cps, the SRM is Upscale.

4 El 5.3 IF the SRM is Upscale AND the SRM-IRM overlap has been checked AND the SRM is not fully withdrawn, WITHDRAW the SRM detector as required to maintain an indicated level between 200 and 7 X 1 O cps.

4 El 5.4 IF the IRMs are not on range 8 or above, confirm the following:

  • the white Rod Out light is EXTINGUISHED El
  • annunciator 603-238, ROD OUT BLOCK is ALARMED El 5.5 IF necessary, INSERT control rods in sequence to decrease reactor power. El 5.6 Determined if the SRM Upscale Alarm light or the Inop light is ILLUMINATED, 2H1 1-P606. El 5.7 H the SRM is Inop, confirm the selector Switch is in the OPERATE position. El NOTE If the RPS shorting links are removed and SRM level increases to 3 X 10 cps, the reactor will scram.

5.8 IF the SRM is failed, notify the Shift Supervisor and IF possible, BYPASS the SRM. El 5.9 IF core alterations are in progress, ensure compliance with Technical Specification 3.3.1.2, SRM Instrumentation. El

6.0 CAUSES

6.1 SRM upscale 6.3 SRM mop SRM detector not positioned correctly SRM Selector switch not in OPERATE 6.2 Failure of SRM SRM high voltage power supply voltage low SRM circuit board not in circuit

7.0 REFERENCES

8.0 TECH. SPECS./TRM/ODCMIFHA:

7.1 H-27527 thru H-27536, Start Up RNM. Sys. 2C51A Elem 8.1 TRM T3.3.2 Control Rod Block lnstr 7.2 57SV-C51-007-2, SRM Calibration 8.2 TS 3.3.1.2 Source Range Monitors 34AR-603-204-2 Ver. I MGR-0048 Ver. 5.0 AG-MGR-75-1 101

HLT-07 SRO NRC EXAM 4L 295009AA1.03 001 Unit 1 is operating at 60% RTP with the following conditions:

o lB RFPT in service with iA RFPT on its turning gear o Both Recirculation Pump speeds are 65%

o lB RFP trips due to the running Condensate Booster pumps tripping o RWL lowered to -65 before the operator recovered RWL to +/-35 with HPCI Which ONE of the choices below describes the FINAL condition of BOTH Recirculation Pumps?

A. Running at #1 speed limiter B. Running at #2 speed limiter C. Running at #3 speed limiter D Tripped

Description:

Speed Limiter #1 electronically limits the speed of the pumps to 22% to minimize the chance of pump cavitation if either of the following exist:

The Pump Discharge Valve is not FULL OPEN (< 90%)

OR Total feedwater flow is less than 20% (15 second time delay)

OR Total Steam Flow decrease of 60% of previous 6 minute value AND RWL < 20 (Median)

Speed Limiter #2 electronically limits the speed of the recirc pumps to 33% if the following conditions exist:

Either Feedwater Pump < 20% rated flow AND either RFP has a trip signal from TMR AND Reactor water level is less than 32 inches (low level alarm setpoint) OR Steam Flow is greater than 65%

Speed Limiter #3 electronically limits the speed of the recirc ASD to 61% if the following conditions exist:

CBP median suction pressure is <40 psig (U-i is 25 psig) after a 10 second time delay RFP median suction pressure is <225 psig (U-i is 165 psig) after a 5 second time delay Speed Limiter #4 Runback to <100% to 33%

Reactor Water Level (Median) < 30 and 20 or greater, Recirc speed will runback 6.7% speed I inch of level differential from 30 124

HLT-07 SRO NRC EXAM The ATWS Recirc pump trip from low reactor water level (-60 inches) or high reactor pressure (1170 psig) is sensed from the ART logic which comes from the Analog Transmitter Trip System (ATTS).

The A distractor is plausible since it would be correct with the loss of the feeding RFP and the other idling, Total Feedwater Flow reduces to < 20% causing Speed Limiter 1 to be in effect.

Since RWL reduced to <-60 inches, Speed Limiter #1 is not the final condition.

The B distractor is plausible if the applicant confuses the Speed Limiters and thinks since a RFPT has tripped that the conditions for Speed Limiter #2 exists, which it does. Since RWL reduced to <-60 inches, Speed Limiter #2 is not the final condition.

The C distractor is plausible if the applicant confuses the Speed Limiters and thinks since the Condensate Booster pumps have tripped that the conditions for Speed Limiter #3 exists, which it does. Since RWL reduced to <-60 inches, Speed Limiter #3 is not the final condition.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

125

HLT-07 SRO NRC EXAM

References:

NONE K/A:

295009 Low Reactor Water Level AA1. Ability to operate and/or monitor the following as they apply to LOW REACTOR WATER LEVEL: (CFR: 41.7 /45.6)

AA1.03 Recirculation system: Plant-Specific 3.0 3.1 LESSON PLAN/OBJECTIVE:

B31-RRS-LP-00401, Reactor Recirculation System, EO 004.001.A.07 References used to develop this question:

34S0-B31-001-1, Reactor Recirculation System (ATWS and speed limiters)

Modified from HLT DATABASE Q# 202001-007 ORIGINAL QUESTION HLT Datatbase Q#202001-007 Unit 1 is operating at 50% RTP with the following conditions:

o lB RFP in service with 1A RFP on its turning gear o Both Recirculation Pump speeds are 55%

o lB RFP trips due to low suction pressure because both running CBPs trip o RWL lowered to -50 before the operator recovered RWL to +35 with HPCJJRCIC Which ONE of the following describes the FINAL condition of both Recirculation Pumps?

A.v Running at 22% speed due to #1 speed limiter B. Running at 33% speed due to #2 speed limiter C. Running at 61% speed due to #3 speed limiter D. Tripped due to ATWS 126

B31-RRS-LP-00401 Page 8 of 132 REACTOR RECIRCULATION SYSTEM OBJECTIVES Initial License fjJ ENABLING OBJECTIVES From a list of statements, SELECT the one that best describes the purpose of the Reactor Recirculation Systems. (004.001 .A. 12, 004.001 .A. 14)

2. Given a P&ID or a simplified drawing of the Reactor Recirculation System, TRACE the flow path through the system and LABEL the following components: (004.00 1 .A.06)
a. Suction Valve
b. Recirc Pump
c. Discharge Valve
d. Jet Pumps
3. Given plant conditions and changes to those conditions, DETERMiNE the Recirc System automatic responses and interlocks, including: (004.001 .A.07)
a. Pump suction and discharge valves
b. LOCA Condition
c. Pump Start
d. #1, #2, #3 and #4 speed limiters
4. From a list of statements, CHOOSE the one that best describes the function of: (004.00l.A.ll)
a. Recirc pump
b. RPT Breakers
c. Jet pumps
d. Adjustable Speed Drive
e. Adjustable Speed Drive Cooling Water System
5. Given a list of statements, CHOOSE the one that best describes how NPSH of Recirc Pumps is maintained at: (004.00l.A.lO)
a. Low power
b. High power
6. Given a list of statements, IDENTIFY the two problems associated with Jet Pump failures.

(004.002.A. 10)

7. Given a load list, DETERMINE the power supplies to the Reactor Recirculation System components: (004.001 .A.04)
a. Suction and Discharge Valves
b. Adjustable Speed Drive (ASD) Drive Motor
c. ASD Cooling Water Pumps

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 11 OF 208 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REACTOR RECIRCULATION SYSTEM 34S0-B31-.001-1 41.1 5.2 LIMITATIONS 5.2.1 Pump Motor Wind ings Starting Limitations Pump motor A & B winding temperatures are indicated by 1B31-R601, Recirc Pump Temp NOTE*

on panel 1H11-P614.

recorder, 5.2.1.1 With winding temperature indicated:

1. Ambient is defined as less than Q equal to drywell temperature in the pump motor area.
2. Rated temperature is defined as ambient plus the design temperature rise of 63CC (1 13F).

NOTES: 3. Recirc Pump A motor area temperature is indicated on 1T47-TR-R61 1, Drywell/Torus Temps recorder, Point No. 11, panel IHII-P657.

4. Recirc Pump B motor area temperature is indicated on 1T47-TR-R612, Drywell/Torus Temps recorder, Point No. 12, panel 1H1I-P654.
  • With the initial winding temperature at ambient temperature, the motor can be started two times in succession.

. With the winding temperature greater than ambient but less than OR equal to rated temperature, the motor can be started once.

5.2.1.2 Without winding temperature indicated, the winding can be assumed to have returned to rated temperature after 45 minutes unenergized OR after running for 15 minutes.

NOT ALL FATAL FAULTS WILL TRIP THE 41 60V INPUT BKR, SOME FATAL FAULTS ACTIVATE A RAMP SHUTDOWN WHERE THE DRIVE WILL CAUTION: RAMP TO MINIMUM SPEED (22% OR 370 RPM), THEN GO TOO RPM AND DRIVE FLOW COASTS TOO GPM.

THE 1B31-SOO2AOR 1B31-SOO2B, ASD AORASD B CONTROL SWITCH, ON I HI 1-P602, WILL HAVE TO BE TAKEN TO STOP TO TRIP THE INPUT BKR.

5.2.2 The ASD will trip (input breaker opens) upon receipt of any of the following FATAL FAULT signals:

MGR-0001 Ver. 4

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 12 OF 208 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REACTOR RECIRCULATION SYSTEM 345O-B31-001-1 41.1 5.2.2.1 ASD Input Line Undervoltage, <55% I 2288V 5.2.2.2 ASD Input Line Overvoltage, >120%! 4992V 5.2.2.3 Rectifier Low Level / Input One Cycle (excessive reactive current flow in xfmr) 5.2.2.4 Recirc. Pump discharge valve <90% OPEN

  • Bypassed for ASD start
  • Must be off the CLOSED seat WITHIN 3 seconds of the pump start
  • Must be > 90% OPEN WITHIN 96 seconds of the pump start 5.2.2.5 Recirc. Pump suction valve < 90% OPEN 5.2.2.6 ATWS
  • Low reactor water level of -60°
  • High reactor pressure of 1170 psig 5.2.2.7 Secondary Winding Low Level! Excessive Drive Losses, >3 Power Cells bypassed!

faulted 5.2.2.8 Loss of Both NXG Computer Control Power Supply A & B 5.2.2.9 ASD lockout relay tripped 5.2.2.10 RPT breaker OPEN 5.2.2.10.1 H EOC RPT is operable, breakers 1C71-CB3A, 1C71-CB4A, 1C71-CB3B, AND 1C71-CB4B will trip on the following:

  • Turbine stop valve < 90% OPEN at> 27.6% reactor power 5.2.2.10.2 ONLY breakers 1C71-CB3A AND CB3B will trip on the following:
  • Recirc motor time delayed overcurrent
  • Recirc motor instantaneous differential overcurrent 5.2.2.11 Manual trip by using the Emergency Stop Pushbutton locally at ASD 5.2.2.12 Transformer Winding Overtemp (Transformer High Temp >176cF, AND Transformer Very High Temp >1 94cF, AND a one minute time delay)

MGR-0001 Ver. 4

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 16 OF 208 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REACTOR RECIRCULATION SYSTEM 345O-B31-001-1 41.1 5.2.18 Recirc Speed Limiter Summary

  1. 1 SL to 22%

Recirc discharge valve <90% open OR Total FIW flow < 20% (15 sec delay)

OR RWL (Master FW controller output) <20 inches (Normally Median Level)

AND Total steam flow decreases by 60% of previous 6 minute value

  1. 2SLto 33%

Either RFP <20% rated flow AND Either REP has a trip signal from TMR AND RWL < 32 inches OR Steam Flow> 65%

  1. 3SLto6l%

CBP median suction pressure low [25 psig, 10 sec delay]

OR REP median suction pressure low [165 psig, 5 sec delay]

  1. 4 SL Variable 100% to 33%

6.7 % speed decrease per ONE inch decrease in RWL (from 30 to 20 inches) 5.2.19 The auto closure of 1B31-FO3IA & 1B31-FO31B, Recirc Pump Discharge Valves, is initiated, WHEN reactor pressure decreases to 370 psig AND a LOCA signal (from either Div I OR Div II isolation logic) is present simultaneously.

The valve CANNOT be opened IF a LOCA signal is present, regardless of reactor pressure.

MGR-0001 Ver. 4

HLT-07 SRO NRC EXAM

42. 295010G2.4.18 001 The following conditions exist on Unit 2:

o Drywell temperature = 300°F o Drywell pressure = 5 psig o Reactor = shutdown (all rods in) o RPV water level = 32 and steady I GRAPH 8 I L[1I DRYWELL SPRAY [N ITIAT1ON L IM IT 600 r

500 1 V/1%W /

400 /#A £f DW NSAFE TEMP //

(°F) 300 rri ,,,,

Is ISAFE 20 100 0 10 20 30 40 DRYWELL PRESSURE (psig)

With the above plant conditions, which ONE of the choices below completes the following statements?

If Drywell Sprays are initiated, Primary Containment could The Unit 2 Primary Containment Design Temperature been exceeded.

A. fail from exceeding the capacity of the Torus-to-Drywell vacuum breakers; has B. fail from exceeding the capacity of the Torus-to-Drywell vacuum breakers; has NOT C. be de-inerted from opening the Reactor Bldg. to Torus vacuum breakers; has D be de-inerted from opening the Reactor Bldg. to Torus vacuum breakers; has NOT 127

HLT-07 SRO NRC EXAM

42. 295010G2.4.18 001

Description:

To prevent large, rapid pressure reductions when Drywell sprays are initiated, the DSIL Curve prevents spray initiation if it will result in an evaporative cooling pressure drop below either:

The Drywell-to-Torus design differential pressure or, The high Drywell pressure scram setpoint Keeping the evaporate cooling pressure drop above the scram setpoint gives the operator time to secure sprays before the containment negative design pressure is exceeded or the Reactor Building-to-Torus vacuum breakers open.

At higher Drywell pressures, the rate of pressure reduction can be beyond the capacity of the Torus-to- Drywell vacuum breakers. Differential pressures between the Drywell and suppression chamber may exceed design, causing failure of boundary between the Drywell and the Torus.

At lower Drywell pressures, the Drywell to Torus differential pressure is not limiting. At these pressures, the concerns become: Reducing Drywell pressure below its negative design before the operator can secure sprays. Popping open Reactor Building to Torus vacuum breakers, which could de-inert the containment, before the operator can secure sprays.

The purpose of drywell temperature control (DW/T) is to maintain drywell temperatures below the design temperature of the primary containment or the temperature at which equipment is qualified to operate (U-2 is 340°F and U-i is 280°F). In this question the Unit 1 value has been exceeded not U2.

The A distractor is plausible if the applicant confuses the DSIL curve and would be correct if t

asking about the other side of DSIL (higher temp & pressure). The second part is plausible if the applicant confuses the Unit 2 DW Design Temperature (340°F) with the Unit 1 DW Design Temperature (280°F).

The B distractor is plausible if the applicant confuses the DSIL curve and would be correct if 128

HLT-07 SRO NRC EXAM

References:

NONE K/A:

295010 High Drywell Pressure 2.4.18 Knowledge of the specific bases for EOPs.

(CFR:41.10/43.1/45.13) 3.3 4.0 LESSON PLAN/OBJECTIVE:

T23-PC-LP-0 1301, Primary Containment, EO 013.021. A.03 EOP-CURVES-LP-20306, EOP Curves & Limits, EO 201.076.A.15 References used to develop this iuestion:

U 1 Graph 8, Drywell Spray Initiation Limit U2 Graph 8, Drywell Spray Initiation Limit 31E0-OPS-001-0, EOP General Information (Attachment 4, Graph 8, U2 DSTL) 129

T23-PC-LP-01301 Page 5 of 160 PRIMARY CONTAINMENT Initial License (LT) ENABLING OBJECTIVES

1. Given a list of statements, IDENTIFY the statement which best describes the purpose of primary containment. (013.021.A.02, 013.033.A.02)
2. Given a list of statements, IDENTIFY the statement which best describes the major components of the primary containment. (013.021 .A.04, 013.033 .A.04)
3. Given a list of statements, IDENTIFY the statement which best describes the basic construction of the drywell, including the design pressures and temperature. (013.021.A.03, 013.033.A.03)
4. Given a list of statements, IDENTIFY the statement which best describes the purpose of the reinforced concrete that encloses the drywell. (013.021.A.05, 013.033.A.05)
5. Given a list of statements, IDENTIFY the statement which best describes the purpose of the steel bulkhead plate that surrounds the reactor vessel below the drywell head.

(013.021.A.06, 013.033.A.06)

6. Given a list of statements, IDENTIFY the statement which best describes the purpose of the jet deflectors that are installed at the entrance of each vent pipe. (013.021 .A.07, 013.033 .A.07)
7. Given a list of statements, IDENTIFY the statement which best describes the basic construction of the pressure suppression chamber (torus). (013.022.A.01)
8. Given a list of statements, IDENTIFY the statement which best describes the purpose of the drywell/torus differential pressure control system. (013.017.A.01)
9. Given a list of statements, IDENTIFY the statement which best describes the purpose of the vacuum relief system. (01 3.023.A.01, 013.024.A.01)
10. Given a list of statements, IDENTIFY the statement which best describes basic operation of the vacuum relief system. (013.023 .A.02, 013 .024.A.02)
11. Given a list of statements, IDENTIFY the statement which best describes basic operation of all drywell penetrations. (013.039.A.02)
12. Given a list of statements, IDENTIFY the statement which best describes the purpose of the drywell drain sumps (drywell leak detection system). (040.001 .A.0 1, 040.002 .A.0 1)
13. Given a list of statements, IDENTIFY the statement which best describes basic operation of the drywell drain sumps, including the operation of the pump out and fill timers.

(040.003.A.01, 040.004.A.0l)

14. Given a list of statements, IDENTIFY the statement which best describes the purpose of the containment ventilation and purge system. (013 .007.A.02, 013.01 2.A.0 1, 013.01 9.A.0 1, 013.020.A.01)

EOP-CURVES-LP-20306 Page 3 of 77 EOP CURVES & LIMITS Initial License (LT) ENABLING OBJECTIVES

1. DEFINE Large Oscillation Threshold (LOT), and what operator actions are REQUIRED if they occur. (201.071.A.21)
2. DEFINE Boron Injection Initiation Temperature. (201.071 .A.09, 201 .074.A.06)
3. Given an ATWS and the Boron Injection Initiation Temperature Curve, DETERMINE the Boron Injection Initiation Temperature. (201.071 .A. 10)
4. DEFINE Hot Shutdown Boron Weight. (201.092.A.01)
5. DEFINE Cold Shutdown Boron Weight. (201.070.A.03, 201.071.A.17, 201.085.A.19)
6. Given a list, IDENTIFY the statement that describes the failure mode that the Heat Capacity Temperature Limit protects against. (201 .074.A.10)
7. Given plant conditions and the Heat Capacity Temperature Limit Curve, DETERMINE the current operating point on the Curve within 2.5 degrees, 5 inches, and 10 psig.

(201 .075.A.06)

8. Given a list, IDENTIFY the failure mode that the Primary Containment Pressure Limit is designed to prevent. (201 .076.A.24)
9. Given a list, IDENTIFY the statement that describes the purpose of the Primary Containment Pressure Limit. (201 .065.A.30, 201 .068.A. 14, 201 .075.B. 17, 201.083 .A.20, 201 .086.A.09, 201 .087.A.06, 201 .088.A.05, 201 .090.A.09)
10. DEFINE Maximum Pressure Suppression Primary Containment Water Level (MPSPCWL). (201.075.B.21)
11. Given a list, IDENTIFY the two conditions that the Drywell Spray Initiation Limit is designed to prevent. (201 .072.A.28, 201.073 .A.07, 201 .076.A. 15)
12. Given plant conditions and the Drywell Spray Initiation Limit Curve, DETERMINE whether or not Drywell sprays can be initiated. (201.072.A.29, 201.073.A.08, 201.076.A. 16)
13. DEFINE the Minimum Drywell Spray Flow (MDSF). (800.003.B.02)
14. Given a list, IDENTIFY the primary containment failure mode that the Suppression Chamber Spray Initiation Pressure is designed to prevent. (201 .076.A.05)
15. Given a list, RECOGNIZE the primary containment failure modes that the Pressure Suppression Pressure Curve is designed to prevent. (201.076.A.21)
16. Given plant conditions and the PSP Curve, DETERMINE the current operating point on the Curve within 2.5 inches and 1.0 psig. (201.076.A.22)
17. Given a list, RECOGNIZE the component failure modes that the SRV Tail Pipe Level Limit is designed to prevent. (201.075.B.05)

SNC PLANT E. I. HATCH Pg 16 of 23 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

EOP GENERAL INFORMATION 31E0-OPS-001-0 1.7 ATTACHMENT 4 Attachment Page TITLE: UNIT TWO PC CHART EOP GRAPHS GENERAL INFORMATION 7 of 10 Graph 8: Drywell Spray Initiation Limit (DSIL) ri DRYWELL SPRAY INITIATION L1IvIIT DW TEMP

(°F) 0 10 20 30 40 DRYWELL PRESSURE (psig)

A) Is defined to be the highest Drywell temperature at which initiation of Drywell sprays will not result in a rapid pressure drop to below either:

a. The Drywell below torus design differential pressure capability, or
b. The high Drywell pressure scram setpoint.

MGR-0009 Ver. 5

DRYWELL TEMPERATURE (F) ci, c

C

C N

a I I I. II SI I N

H I II C en II Ia I III N N Cl)

%I*.Ir____

uII_____ 0 cu iiir______

h________ C e

C I)

C) liii Ii 00 C

C c

II_IiLIIII C

C C

C C

C C

C C

C en L Q2

HLT-07 SRO NRC EXAM

43. 295013AK2.01 001 Unit 1 is operating at 90% power.

o A Safety Relief Valve (SRV) inadvertently opened causing Suppression Pool water temperature to increase o Suppression Pool water temperature reaches 102°F before operators are able to close the SRV Which ONE of the choices below describes the required Residual Heat Removal (RHR)

Suppression Pool Cooling alignment?

A. Place ONLY one loop of RHR in Suppression Pool cooling, and the RHR heat exchanger IS required to be isolated prior to starting the RHR pump.

B. Place ONLY one loop of RHR in Suppression Pool cooling, and the RHR heat exchanger IS NOT required to be isolated prior to starting the RHR pump.

C. Place ALL available RHR loops in Suppression Pool cooling, and the RHR heat exchanger IS required to he isolated prior to starting the RHR pumps.

D Place ALL available RHR loops in Suppression Pool cooling, and the RHR heat exchanger IS NOT required to be isolated prior to starting the RHR pumps.

130

HLT-07 SRO NRC EXAM

==

Description:==

Between 95°F and 100°F, 34AB-T23-003-1 requires one ioop of RHR to be placed in service.

When below 100F the RHR HX is required to be isolated. Above 100°F, all available RHR pumps except for pumps required for adequate core cooling. Above 100°F, the RHR heat exchanger is not required to be isolated prior to starting RHR pumps since it is being placed in service to support the EOPs.

The A distractor is plausible if the applicant does not realize operation is being directed by the EOPs which requires all available to be placed into service. The second part is plausible if the applicant does not recall the difference between RHR alignment requirements above and below 100°F.

The B distractor is plausible if the applicant does not realize operation is being directed by the EOPs which requires all available to be placed into service. The second part is correct.

The C distractor is plausible since the first part is correct. The second part is plausible if the applicant does not recall the difference between RHR alignment requirements above and below 100°F.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

References:

NONE K/A:

295013 High Suppression Pool Temperature AK2. Knowledge of the interrelations between HIGH SUPPRESSION POOL TEMPERATURE and the following: (CFR: 41.7 /45.8)

AK2.01 Suppression pool cooling 3.6 3.7 LESSON PLAN/OBJECTIVE:

HLT-07 SRO NRC EXAM T23-PC-LP-0 1301, Primary Containment, EO 200.040.A.0 1 References used to develop this question:

31E0-EOP-012-l, Primary Containment Control EOP flowchart 34AB-T23-003-l, Torus Temperature Above 95°F 34S0-El 1-010-1, RHR System Modified from HLT-4 NRC Exam Q#43 ORIGINAL QUESTION (HLT-4 NRC Exam Q#43)

Unit 1 is operating at 90% power.

o A Safety Relief Valve (SRV) inadvertently opened causing Suppression Pool water temperature to increase o Suppression Pool water temperature reaches 97°F before operators are able to close the SRV LAW 34AB-T23-003-l, Torus Temperature Above 95°F, which ONE of the following describes the required Residual Heat Removal (RHR) Suppression Pool Cooling alignment?

A.V Place only one ioop of RHR in Suppression Pool cooling, and the RHR heat exchanger is required to be isolated prior to starting the RHR pump.

B. Place only one ioop of RHR in Suppression Pool cooling, and the RHR heat exchanger is NOT required to be isolated prior to starting the RHR pump.

C. Place all available RHR loops in Suppression Pool cooling, and the RHR heat exchanger is required to be isolated prior to starting the RHR pumps.

D. Place all available RHR loops in Suppression Pool cooling, and the RHR heat exchanger is NOT required to be isolated prior to starting the RHR pump.

132

T T23-PC-LP-01361 Page 10 of 160 I PRIMARY CONTAINMENT ABNORMAL, EMERGENCY, AND SURVEILLANCE OPERATIONS

  • 47 Per 34AB-T47-001-l/2, Complete Loss of Drywell Cooling, and given plant conditions, RECOGNIZE and RESPOND to a loss of Drywell Cooling. (200.032.A.01)
  • 48. Per 34AB-T23-003-1/2, Torus Temperature above 95°F, and given plant conditions, RECOGNIZE and RESPOND to Torus Temperature above 95°F. (200.040.A.01)
  • 49 Per 34SO-P70-001-l/2, Drywell Pneumatic System, and given plant conditions, SUPPLY nitrogen to drywell pneumatic system via the emergency Nitrogen supply (N2 Bottles).

(042.005 .A.04)

  • 50. Per 34SO-P33-00l-l/2, LINEUP and STARTUP the H 2 and 02 Analyzers for various plant conditions. (051 .002.A.0l)
  • 51 Per 34SO-P33-001-l/2, Primary Containment Atmosphere H202 Analyzer System, PLACE H2O2 Analyzers in service during the following conditions: (013.051 .A.0 1, 013.060.A.01)
a. LOCA Signal present
b. LOCA Signal not present
52. Per 34SO-P33-00 1-1, 34S0-P33-002- 1 and/or 34AB-C7 1-001-1, DETERMINE if valve, 1 P33-F605 should be open or closed. (013.046.A.07)
  • 53 Per 34AB-C71-00l-l/2, Scram Procedure, DETERMINE if PCIS Groups 1, 2, and 6 isolations should have occurred and confirm the associated group isolation valves have closed as required.

(013.039.A.01, 013.045.A.01, 013.046.A.01, 013.047.A.0l)

  • 54 Per 34AB-C71-001-1/2, Scram Procedure, DETERMINE if the Tip Shear Valves are required to be fired. (0l3.036.A.01, 400.045.C.01) 55 Per 31EO-EOP-1l 1-1/2, Emergency Opening of the MSIVs, LOCATE the terminal points used to override the high flow isolation on P70-F004 and 70 F005. (042.007.A.04)
  • 56. Given plant conditions involving the Primary Containment System, DETERMINE if a Technical Specification Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (300 .006.A.22)
  • 57* Given plant conditions involving the Primary Containment System, DETERMINE the Required Actions for Completion Times < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in accordance with Technical Specifications for any combination of INOPERABLE systems, structures or components. (300.009.A.06)

SOUTHERN NUCLEAR I I PAGE2OF6 PLANT E. I. HATCH DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION TORUS TEMPERATURE ABOVE 95°F 34AB-T23-003-IS NO:

2 ED 2 4.0 SUBSEQUENT OPERATOR ACTIONS NOTE U SPDS and recorder 1T48-R647 are inoperable, refer to 34SV-SUV-019-1S, Attachment 2, Torus Temperature Monitoring, for determining the torus water temperature.

4.1 Confirm the high temperature by OBSERVING the suppression pool bulk temperature on the SPDS primary display.

4.2 IF the SPDS is inoperable, confirm the high temperature by performing calculations per Attachment 2 of this procedure.

4.3 IL Suppression Pool bulk average temperature exceeds 95°F, PLACE RHR in Suppression Pool cooling per 34SO-E1 1-010-IS, Residual Heat Removal.

4.4 IF testing is in progress that adds heat to the Suppression Pool, record Suppression Pool bulk average temperature on Attachment I every 5 minutes.

4.5 IF Suppression Pool bulk average temperature exceeds 100°F, START ALL available RHR loops in Suppression Pool cooling AND enter 31E0-EOP-012-IS, PC-I Primary Containment Control, AND 31 EO-EOP-01 3-IS, PC-2 Primary Containment Control, AND perform concurrently with this procedure.

4.6 IF Suppression Pool bulk average temperature is> 100°F but < 110°F, AND any OPERABLE IRM channel is> 25/40 on Range 7, AND no testing that adds heat to the Suppression Pool is in progress, THEN verify temperature < 110°F once per hour (record on Attachment 1), AND restore to < 100°F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.7 j.[ Suppression Pool bulk average temperature is> 105°F AND any operable IRM channel is > 25/40 on Range 7, AND testing is in progress that adds heat to the Suppression Pool, THEN suspend all testing immediately.

4.8 11 Suppression Pool bulk average temperature exceeds 110°F, enter 34AB-C71-001 -15, Scram Procedure, AND SCRAM the reactor, AND verify temperature <120°F once per 30 minutes.

4.9 H Suppression Pool bulk average temperature exceeds 120°F, enter 34GO-OPS-01 3-iS, Normal Plant Shutdown, AND reduce reactor pressure to < 200 PSIG within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.10 Complywith all applicable Required ActionsforTechnical Specification LCO 3.6.2.1.

MGR-0001 Rev 3

SOUTHERN NUCLEAR PLANT E.I. HATCH PAGE 282 OF 293 DOCUMENT TITLE: DOCUMENT NUMBER:* VERSION NO:

RESIDUAL HEAT REMOVAL SYSTEM 34SO-E1 1-010-1 38.0 ATTACHMENT 14 ATT. PAGE:

TITLE: SUPPRESSION POOL COOLING INITIATION 1 OF I SUPPRESSION POOL COOLING INITIATION

1. PLACE RHRSW in operation per 345O-E11-010-1. LI
2. IF RWL <2/3 core height, (-193 inches),

PLACE the Cnmt Spray Vlv Cntl 2/3 Core Ht Permis keylock in the MANUAL OVERRD. LI

3. IF required by EOPs AND LOCA signal present, PLACE Cnmt Spray VIv CntI switch in the MANUAL position. LI
4. OPEN 1EI1-F048A(IEI1-F048B). El
5. UNLESS SPC is required per the EOPs, CLOSE 1EII-F047A (1EI1-F047B). LI
6. OPEN 1 Eli -FOO3A (1 El 1-FOO3B). LI
7. IF power is being provided by EDG, CHECK EDG loading prior to start of RHR pump(s). El
8. IF the RHR loop was already in operation in a mode other than Suppression Pool Cooling, As required, CLOSE OR THROTTLE the appropriate valve(s) to prevent an unintentional injection:

MODE VALVES LPCI 1E11-FOI7A(1E11-FOI7B)OR1E11-FO15A(IEI1-FO15B) El Drywell Spray 1E1 i-FOI6A (IE1 i-FOI6B)OR iEii-FO2IA (1E11-FO21B) LI Suppression Pool Spray 1EI1-F027A(1E11-F027B) LI

9. START RHR Loop A(B) pump(s). El
10. OPEN 1E11-F028A(1E11-F028B). El
  • RHR system IA OR lB flow ranges between 4,000 gpm to 6,500 gpm with one pump or 9,000 gpm to 13,000 gpm with two pumps are to be avoided, NOTES: unless required for EOP actions.

. IF IEII-F048A(1EI1-F048B), 1AOR lB Hx Bypass Vlv, is NOT full OPEN, RHR Flow is limited to 9,000 gpm, unless required for EOP actions.

11. THROTTLE OPEN lEl i-F024A (1 El l-F024B). LI
12. OPEN 1EII-F047A(lEll-F047B). LI
13. UNLESS it is desired to provide mixing of the torus rather than cooling, ensure RHR A OR RHR B loop flow is 9,000, THEN CLOSE 1 El 1-F048A (1 Eli -F048B).
14. THROTTLE 1EI1-F068A(1E1I-F068B)to maintain>20 PSID HxA(B)dp.
15. REFERto34SO-Ei1-OlO-i. LI 34SO-El 1-010-i (Posted lHli-P601)

MGR-0009 Ver. 5

{

C )

TOts water level DrvveIl lemperature A above 150 in. abovelbo?

( P0wa%conSinwonlhwgnnconcenEabon Twaswalnvlev&

belOw 146 in. l.

Twaswalortowpewlore aboVe 100 F

\ above 1.5%

J j

Y WHILE PERFORMING THE FOLLOWING L IF PRIMARY CONTWNMENT FLOODING IS OR HAS BEEN REQUIRED THEN evil the EOPs and enter the Severe Accident GuideJ B If fuel failure is suspeclnd consult with I I P1001 Chemistry prior to discharging waler I c

HLT-07 SRO NRC EXAM

44. 295016AA2.07 001 The control room has been abandoned and 31RSOPS-OOl-2, Shutdown From Outside Control Room, is being implemented.

The SS has ordered Torus Sprays to be placed in service.

Which ONE of the choices below completes the following statements?

Torus pressure can be determined using Torus Sprays can be initiated from the Remote Shutdown Panel, 2C82-POOl, ONLY in the A SPDS in the TSC; t Loop B

B. SPDS in the TSC; A Loop C. a permanently mounted gauge on the 87 level of the Reactor Building; B Loop D. a permanently mounted gauge on the 87 level of the Reactor Building; A Loop 133

HLT-07 SRO NRC EXAM

==

Description:==

Torus pressure is not monitored at the RSDP, nor a permanently mounted gauge on the 87 elevation of the Reactor Building. Procedure 31E0-EOP-105, Primary Containment Water Level Determination, utilizes 2 pressure gauges to determine Torus pressure and DW pressure, since Torus pressure is not indicated locally or at the RSDP. Torus pressure will be monitored by the STA in the TSC using SPDS.

lAW EOPs, with Primary Containment pressure above 1.85 psig, Torus Sprays shoud be initiated.

The B distractor is plausible since the first part is correct. The second is plausible if the applicant confuses which Loop of RHR can be aligned to Torus sprays from the Remote Shutdown Panel. Also would be plausible if applicant confuses Ui CRD pump operation, which can be started from a Remote Shutdown panel.

The C distractor is plausible if the applicant confuse the RSDP indications along with Torus pressure locally on the 87 elevation and thinks Torus pressure will be monitored locally or does not remember 31E0-EOP-l05-2 actions. The second part is correct.

The D distractor is plausible if the applicant confuse the RSDP indications along with Torus pressure locally on the 87 elevation and thinks Torus pressure will be monitored locally or does not remember 31E0-EOP-l05-2 actions. The second is plausible if the applicant confuses which Loop of RHR can be aligned to Torus sprays from the Remote Shutdown Panel. Also would be plausible if applicant confuses Ui CRD pump operation, which can be started from a Remote Shutdown panel.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

134

HLT-07 SRO NRC EXAM

References:

NONE K/A:

295016 Control Room Abandonment AA2. Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT: (CFR: 41.10/43.5/45.13)

AA2.07 Suppression chamber pressure 3.2 3.4 LESSON PLAN/OBJECTIVE:

C82-RSDP-LP-05201, Remote Shutdown Panel (RSDP), EO 052.001.A.06 References used to develop this question:

31RS-OPS-001-2, Shutdown From Outside Control Room 135

C82-RSDP-LP-05201 Page 4 of 94 REMOTE SHUTDOWN PANEL (RSDP)

Initial License (LT) ENABLING OBJECTIVES

1. IDENTIFY the purpose of the Remote Shutdown Panels (RSDPs). (052.00 1 .A.02)
2. Given plant conditions, IDENTIFY the conditions which would make it necessary to evacuate the Main Control Room. (200.0 10.A.01)
3. LIST the systems and/or components operated at the RSDP. (052.00 1 .A.03)
4. LIST the instrumentation which is available at the RSDP. (052.00 1.A.06)
5. IDENTIFY the plant location of the Ui and U2 RSDPs. (052.001.A.0l)
6. Given a list of electrical buses, IDENTIFY the correct power supplies for the RSDPs.

(052.001 .A.07)

7. STATE how a Control Room operator can DETERMINE if an Emergency Transfer switch has been taken to the Emergency position. (001 .023.A.06)
8. IDENTIFY the effects of placing a Remote Transfer Switch on the RSDP to the Emergency position on system operation from the Main Control Room. (052.001 .A.09, 001.023 .A.07, 007.01 9.A.06)
9. IDENTIFY which SRVs are controlled from RSDPs. (014.018.A.0l)
10. IDENTIFY which CRD pump can be controlled from the U2 Remote Shutdown Panel.

(001 .023.A.0l)

11. IDENTIFY which CRD pump(s) can be controlled from the Ul Remote Shutdown Panel.

(001 .023.B.02)

12. DESCRIBE how transfer of control to the RSDP would affect the following for RHR:

(007.01 9.A.04)

a. RI-JR valve interlocks
b. RHR pump A, B, C, D, auto starts
c. RHRpumpA, B, C, D, trips
d. RHR mm flow valve operation

SNC PLANT E. I. HATCH I Pg 35 of 48 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

SHUTDOWN FROM OUTSIDE CONTROL ROOM 3IRS-OPS-001-2 6.19 ATTACHMENT Attachment Page TITLE: TORUS COOLING FROM THE REMOTE SHUTDOWN PANEL 3 of 5 1.0 At panel 2C82-P001, PLACE the following transfer switches in the EMERG position:

2E1 1-F009, SDC Suction Vlv.

2E1 1-FOO6A, S/D CIg Vlv.

2E1 1-COOl B, Serv Wtr Pump 2E11-FOI5B, lnbd lnjVlv.

  • 2C82-S1 0, 2E1 1 -FOO4B, Torus Suction Vlv.

2E11-FOO6B, Shutdown ClgVlv.

2E1 1-FOO6D, Shutdown CIg Vlv.

2E1 1-F023, Rx Head Spray lsol Vlv.

2E1 1-F027B, Torus Spray Vlv.

2E11-F048B, Hx Bypass Vlv.

2E1 1-F073B, Serv Wtr crosstie Vlv.

2E1 1-COOl D, Serv Wtr Pump

2E1 1-FOI6B, Cnmt Spray Outbd Vlv.

2.0 PLACE RHR Service Water in operation by performing the following steps:

2.1 At 600!208V MCC 2B ESS Div 2 2R24-S012, Frame hG, OPEN the breaker for 2E1 I -F068B, RHRSW Control Valve.

2.2 At 106RJR24, manually OPEN 2E11-F068B to approximately 40% OPEN.

2.3 At panel 2C82-P001, START 2E11-COOIBOR 2E11-COOID, RHRSW Pump.

2.4 At 106RJR24 WHILE in communication with an operator at panel 2H21-P173, ADJUST 2E11-F0688, RHRSW Control Valve, to obtain a flow rate of 4400 GPM as indicated on 2E1 1-R071, RHR Heat Exchanger Service Water Flow.

Gi 6.030 MGR-0009 Rev. 5.0

SNC PLANT E. I. HATCH I Pg 36 of 48 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

SHUTDOWN FROM OUTSIDE CONTROL ROOM 31RS-OPS-001-2 6.19 ATTACHMENT 6 Attachment Page TITLE: TORUS COOLING FROM THE REMOTE SHUTDOWN PANEL 4 of 5 NOTES* e The RHR Hx is initially isolated to prevent damage to the Hx from the hydraulic shock created by starting the RHR Pump.

IF THE ONLY OPERATING PUMP IN AN RHR LOOP IN SUPPRESSION POOL COOLING WATER TRIPS, IT IS POSSIBLE THAT THE LPCI INJECTION LINE AND THE DRWJELL SPRAY LINE MAY DRAIN TO THE TORUS.

3.0 At panel 2C82-P001, Perform the following:

  • IF Suppression Pool temperature is < 100°F, Confirm CLOSED/CLOSE 2E1 1-F047B, Hx Inlet Vlv.

4.0 At panel 2C82-P001, confirm 2E11-F048B, Hx Bypass Vlv, is OPEN.

5.0 At panel 2C82-P001, OPEN 2E1 1-F028B, RHR Torus Spray or Test Vlv.

6.0 Confirm 2E1 1-FOO7B, Mm Flow Vlv, is OPEN.

7.0 At panel 2C82-P001, START 2E1 1-COO2B, RHR Pump 2B.

8.0 At panel 2C82-P001, THROTTLE OPEN 2E1 I -F024B, Full Flow Test Line, to obtain a flow rate of less than or equal to 7700 GPM as indicated on 2C82-R004, RHR Flow, on panel 2C82-P001.

8.1 Confirm 2E11-FOO7B, Mm Flow Vlv, CLOSES.

RHR Service Water flow must be maintained at less than 4400 GPM with one RHR OTE: Service Water Pump operating, OR less than 8800 GPM with two RHR Service Water Pumps operating in a single RHR Service Water Loop.

9.0 At panel 2C82-P001, Confirm OPEN/OPEN the following valves:

100 At panel 2C82-P001, CLOSE 2E1 1-F048B, Hx Bypass Vlv.

Gi 6.030 MGR-0009 Rev. 5.0

SNC PLANT E. I. HATCH I Pg 37 of 48 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

SHUTDOWN FROM OUTSIDE CONTROL ROOM 3IRS-OPS-001-2 6.19 ATTACHMENT 6 Attachment Page TITLE: TORUS COOLING FROM THE REMOTE SHUTDOWN PANEL 5 of 5 11.0 At 1 06RJR24, ADJUST 2E1 I -F068B, RHRSW Control Valve, to maintain greater then or equal to 20 PSID as indicated on 2E1 1-DPIS-NOO3B, at 87RLR24, on panel 2H21-P021.

12.0 IF Torus Spray is desired, THEN at panel 2C82-P001, OPEN 2E1 1-F027B, Torus Spray Vlv.

13.0 WHEN Suppression Pool Cooling is no longer required, SHUT DOWN RHR from Suppression Pool Cooling by performing the following steps:

13.1 At panel 2C82-P001, CLOSE 2E11-F027B, Torus Spray Vlv.

13.2 At panel 2082-POOl, CLOSE 2E11-F024B Full Flow Test Line.

13.3 At panel 2C82-P001, SHUT DOWN 2E11-COO2B, RHR Pump 2B.

13.4 At panel 2C82-P001, CLOSE 2E1 1-F028B, RHR Torus Spray or Test Vlv.

13.5 At panel 2C82-P001, OPEN 2E11-F048B, Hx Bypass Vlv.

NOTE: RHR Service Water may be left in operation to support Shutdown Cooling.

G16.030 MGR-0009 Rev. 5.0

HLT-07 SRO NRC EXAM

45. 295018AK3.06 001 Unit 2 is operating at 5% power when two (2) RBCCW pumps fail and will NOT run.

The following exists:

o PSW/RBCCW Hx dP increases o The SS directs the SO to LOWER PSW/RBCCW Hx dP to 5 psid ABOVE the alarm setpoint by throttling 2P41-F491, PSW Outlet Valve From RBCCW Hx JAW 34S0-P42-00l-2, Reactor Building Closed Cooling Water System, which ONE of the following completes these statements?

To LOWER the PSW/RBCCW Hx dP, the SO will throttle 2P41-F491 in the direction.

The setpoint for 650-238, HX PSW/RBCCW DIFF PRESS LOW, is A. CLOSE; 7 psid B. CLOSE; 20 psid 0 OPEN; 7 psid D. OPEN; 20 psid 136

HLT-07 SRO NRC EXAM

==

Description:==

With 2 RBCCW pumps failing, RBCCW flow/pressure in the Hx will be less, resulting in a high PSW[RBCCW Hx dP. PSW outlet pressure minus the RBCCW inlet pressure are the values used in determing this dP with PSW maintained at a higher pressure than RBCCW. In some systems RBCCW is the higher pressure.

LAW 34S0-P42-OOl-2 section 7.3.8, Adjusting heat exchanger PSW/RBCCW Differential Pressure, step 7.3.8.3 states To lower PSW/RBCCW dP, throttle OPEN 2P41-F491 Outlet Valve From RBCCW HX. Opening this valve will allow more PSW flow through the Hx and lower PSW outlet pressure. To raise PSW/RBCCW dP, throttle CLOSE 2P41-F491 Outlet Valve From RBCCW HX.

JAW 650238, HX PSW/RBCCW DIFE PRESS LOW, the alarm setpoint is 7 psid. 20 psid is the RHRSW/RHR required dP.

The A distractor is plausible if the applicant confuses RBCCW is the higher pressure and thinks closing the PSW outlet valve will increase PSW outlet pressure bringing the dP between PSW/RBCCW to a lower value. The second part is correct.

The B distractor is plausible if the applicant confuses RBCCW is the higher pressure and thinks closing the PSW outlet valve will increase PSW outlet pressure bringing the dP between PSW[RBCCW to a lower value. The second part is plausible if the applicant confuses this dP with the RHRSW/RHR dP and thinks 20 psid is correct.

The D distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses this dP with the RHRSW/RHR dp and thinks 20 psid is correct.

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

137

HLT-07 SRO NRC EXAM

References:

NONE K/A:

295018 Partial or Complete Loss of Component Cooling Water AK3. Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: (CFR: 41.5/45.6)

AK3.06 Increasing cooling water flow to heat exchangers 3 3.3 LESSON PLAN/OBJECTIVE:

P42-RBCCW-LP--00901, EO 009.006.C.02 References used to develop this iuestion:

34S0-P42-001-2, Reactor Building Closed Cooling Water System 650-23 8, HX PSW/RBCCW Diff Press Low 138

P42-RBCCW-LP-00901 Page 4 of 50 IEACTOR BUILDING CLOSED COOLING WATER System Operator (SO) ENABLING OBJECTIVES

  • 1. Given a list of purposes, SELECT the one that identifies the purpose of the RBCCW System.

(009.002.A.0 1)

2. Given a P & ID or simplified drawing of the RBCCW System, TRACE the flowpath through the system identifying loads served. (009.001 .A.02)
  • 3 STATE the source of cooling water to the RBCCW Heat Exchangers. (009.002.A. 10)
  • 4 Given a list of design features, DETERMINE the 2 design features that ensure RBCCW Pumps NPSH is maintained. (009.002.A.05)
5. From a list of statements, SELECT the statement(s) that identify the purpose of the RBCCW System Surge Tank. (009.002.A.1 1)
6. STATE the source of Fill and Makeup water for the RBCCW System. (009.001.A.01)
  • 7 EXPLAIN why laboratory supervision must be notified prior to draining all or any part of the RBCCW system, per 34S0-P42-001-1/2, RBCCW System. (009.002.A.06)
  • 8. Given a list of functions, SELECT the function for the Process Radiation Monitor in the RBCCW System. (009.006.C.03) 9 Given a list of components, IDENTIFY the components supplied cooling water from the Reactor Building Closed Cooling Water System. (009.002.A.09)
10. Given a list of statements, SELECT the one that identifies why PSW to RBCCW differential pressure is important to be maintained within the design range as given in 34G0-OPS-030-1/2, Daily Inside Rounds.

(009.006.C.02) t

11. DESCRIBE the operation of the RBCCW timer/counter system. (009.002.A.04)
  • 12. Given a loss of RBCCW and 34AB-P42-001-1/2, Loss of All RBCCW, DETERMINE the effects of this on the following systems and on continued Reactor operations. (009.002.A.07)
  • RX Recirculation
  • Fuel Pool Cooling & Cleanup

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 7 OF 56 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REACTOR BUILDING CLOSED COOLING WATER SYSTEM 34SO-P42-001-2 16i THROTTLING 2P42-F038 WILL AFFECT RWCU AND FPC TEMPERATURES, TEMPERATURES CAN BE MONITORED AT:

. 2G31-N008, NON-REGENERATIVE HX OUTLET TEMPERATURE, RWCU SYSTEM CAUTION: TEMPERATURE SHOULD BE MAINTAINED <125 °F, RWCU WILL ISOLATE AT 140.

. 2T41-R620, MULTIPOINT TEMPERATURES RECORDER, POINT 16 (2G41-N002),

FUEL POOL COOLING HEAT EXCHANGER OUTLET TEMPERATURE.

TEMPERATURE LIMITATIONS ARE SET FORTH IN 34SO-G41-003-2.

7.1.10 ll RWCU system is in service 4j the reactor is at rated temperature, throttle 2P42-F038, RBCCW Outlet Throttle Valve, for 2G31-B002, RWCU Non-regenerative Hx,

  • to maintain < 125°F on 2G31-N008, non-regenerative Hx outlet temperature, LI
  • while maintaining RBCCW flow 480 gpm on 2P42-R034 in the RWCU HX Room. LI Critical 7.1.11 Ensure the in-service RBCCW heat exchanger differential pressure is maintained greater than the Ap alarm setpoint (7 psid) per the Adjusting Heat Exchanger PSW/RBCCW Differential Pressure subsection of this procedure. LI MGR-0001 Ver. 4

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 11 OF56 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REACTOR BUILDING CLOSED COOLING WATER SYSTEM 34SO-P42-OO1-2 16.1 7.3.2 Swapping RBCCW Heat Exchangers, Placing A Into Service And Removing B

. PSW/RBCCW Ap for a RBCCW HX can be determined by subtracting 2P42-ROO2A from 2P41-R373A (RBCCW inlet HX pressure from PSW outlet HX pressure).

NOTES

  • There is a potential for system vibration/cavitation IF 2P41-F491 is throttled excessively.
  • To raise PSW/RBCCW Ap, throttle CLOSE 2P41-F491.
  • To lower PSW/RBCCW Ap, throttle OPEN 2P41-F491.

7.3.2.1 Notify the following prior to perlorming any draining / venting:

  • Lab Supervision or alternate El
  • Unit 2 Radwaste operator El Critical 7.3.2.2 During performance of this subsection, slowly throttle 2P41-F491, PSW Outlet From The RBCCW HX, as necessary to maintain d/p greater than alarm setpoint (7 psid). El 7.3.2.3 Line up Plant Service Water to 2P42-BOO1A, Heat Exchanger as follows:
  • Open 2P41-F439A. El Do NOT drain/vent RBCCW water to Radwaste without first contacting Lab Supervision AND NOTE: the Unit 2 Radwaste Operator. The Lab is notified, so they can begin monitoring for molybdates in Waste Sample Tanks.
  • MONITOR RBCCW SURGE TANK LEVEL AND SYSTEM PRESSURE WHILE OPENING VALVES

7.3.2.4 Slowly open OR confirm open 2P42-FOO2A, Heat Exchanger Inlet Valve. El MGR-0001 Ver. 4

1.0 IDENTIFICATION

ALARM PANEL 2H11-P650-2 HX PSWIRBCCW 01FF PRESS zz:zz::: zzz LOW I

DEVICE: SETPOINT:

2P42-dPS-N065A or 7 PSID decreasing when respective 2P42-dPS-N065B RBCCW Hx inlet valve is open

2.0 CONDITION

3.0 CLASSIFICATION

The differential pressure between Plant Service Water leaving and EQUIPMENT STATUS Reactor Building Closed Cooling Water entering an operating 4.0 LOCATION:

RBCCW Heat Exchanger has decreased below 7 PSID. 2H1 1-P650 Panel 650-2 5.0 OPERATOR ACTIONS:

5.1 Confirm the differential pressure is low by comparing:

2P41-R373A(2P41-R373B), Hx 2A (2B) PSW Out Press indicator, and 2P42-ROO2A(2P42-ROO2B),

RBCCW Hx A(B) Inlet Press indicator, on each operating heat exchanger. LI 5.2 IF Service Water Outlet pressure is < 7 PSIG greater than RBCCW Hx Inlet pressure, perform the following: LI 5.2.1 IF PSW Div I OR II is<85 PSIG, as indicated on 2P41-R6OIAOR2P4I-R6OIB, 2H11-P650, perform the following:

5.2.1.1 START the PSW Pump in standby, IF not running. LI 5.2.1.2 THROTTLE 2P41-F491, PSW Outlet From The RBCCW HX valve, Hr necessary. LI 5.2.1.3 Locate and repair any leakage in the Plant Service Water System. LI 5.2.2 IF RBCCW heat exchanger inlet pressure is high, confirm the valve lineup at the heat exchanger is in accordance with 34S0-P42-001-2, Reactor Building Closed Cooling Water (RBCCW) System. LI 5.2.3 IF a leak is indicated between Plant Service Water and RBCCW at the heat exchanger, perform the following:

5.2.3.1 IF only one heat exchanger is in operation, place the standby heat exchanger in service and remove the leaking heat exchanger from service per 34S0-P42-001-2, RBCCW System, until the leak is repaired. LI 5.2.3.2 IF both heat exchangers are in service, reduce loads to the capacity of one and remove and repair the leaking heat exchanger. LI 5.3 IF the differential pressure is > 7 PSID, notify the Instrument Shop to confirm 2P42-N065A and 2P42-N065B, Differential Pressure Instruments, are operable, and IF NOT, confirm 2D1 1-RM-K605, Service Water Effluent Monitor, is operable in accordance with Unit 2 Technical Specifications Table 3.3.6.9-1. LI

6.0 CAUSES

6.1 Excessive flow OR piping rupture in the Plant Service Water System 6.2 Tube leak in the operating RBCCW Heat Exchanger 6.3 Instrument_malfunction

7.0 REFERENCES

8.0 TECH. SPECS.ITRMIODCMIFHA:

7.1 57CP-CAL-036-2, Static 0-Ring pressure Switch 7.2 H-27750, RBCCW System 2P42, Elem Diag, Sht 1 ODCM, Chapter 2 7.3 H-26054, RBCCW Sys P&ID, Sht. I 7.4 34S0-P41-001-2,_Plant_Service_Water_System 34AR-650-238-2 Ver4 MGR-0048 VER 5.0 AG-MGR-75-1101

HLT-07 SRO NRC EXAM

46. 295019AA2.02 001 Unit 2 is operating at 100% power with 2P52-F565, Rx Bldg Inst N2 To Non-mt Air El 185 Isol Vlv

, tagged in the closed position when the following occurs:

t o Unit 2 experiences a loss of ALL Unit 2 Station Service Air Compressors o Instrument and Service Air pressure drops to zero (0) psig Which ONE of the choices below completes BOTH statements for the following ventilation dampers as their respective air pressure decreases to zero (0) psig?

2T46-F00l A, SBGT A Fltr Inlet From Rx Bldg, valve will 2T41-F044A, Rx Bldg Inboard Isol Dmprs Inaccessible Areas Exhaust Fans Disch, valve will_______

A. travel open; remain open B travel open; travel closed C. remain closed; remain open D. remain closed; travel closed 139

HLT-07 SRO NRC EXAM

==

Description:==

2T46-FOO1A, SBGT A Fltr Inlet From Rx Bldg, valve is normally closed and will travel open when a loss of air condition exists. 2T41-F044A, Rx Bldg Inboard Isol Dmprs Inaccessible Areas Exhaust Fans Disch, valve is normally open and will travel closed when a loss of air condition exists.

The A distractor is plausible since the first part is correct. The second part is plausible if the applicant does not realize that the F044A will isolate to fulfill it safety function on a loss of air.

The C distractor is plausible if the applicant does not realize that the safety function 2T46-FOO1A is to be open, therefore will fail open on a loss of air/loss of solenoid power. The second part is plausible if the applicant does not realize that the F044A will isolate to fulfill it safety function on a loss of air.

The D distractor is plausible if the applicant does not realize that the safety function 2T46-FOO1A is to be open, therefore will fail open on a loss of air/loss of solenoid power. The second part is correct.

A. Incorrect See description above.

B. Correct See description above.

C. Incorrect See description above.

D. Incorrect See description above.

140

HLT-07 SRO NRC EXAM

References:

NONE K/A:

295019 Partial or Complete Loss of Instrument Air AA2. Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: (CFR: 41.10 / 43.5 /45.13)

AA2.02 Status of safety-related instmment air system loads (seeAK2.1-AK2.19) 3.6 3.7 AK2.08 Plant ventilation 2.8 2.9 LESSON PLAN/OBJECTIVE:

T46-SBGT-LP-03001, Standby Gas Treatment System, EO 030.007.A.02 P51-P52-P70-Plant Air-LP-03501, Plant Air lesson Plan EO 200.025.a.03 References used to develop this question:

34AB-P51-001-2, Loss of Instrument and Service Air System or Water Intrusion Into The Service Air System 141

T46-SBGT-LP-03001 Page 3 of 59 STANDBY GAS TREATMENT SYSTEM

5. STATE the purpose(s) of the following Standby Gas Treatment (SBGT) System components.

(030.002.A.0l)

a. Drywell/Torus Excess Flow Isolation Dampers F076 and F077 (T48-F392 and F393, Unit 1)
b. Demister (Moisture Separator)
c. Electric heater
d. Prefilter
e. HEPA filter
f. Carbon filters
g. SBGTS Discharge Isolation Valves FOO2A/B
h. Decay Heat Removal Line
i. SBGT Fan/Heater Reset Pushbutton
j. SBGT Fan/Heater Auto Start Reset Pushbutton
6. DESCRIBE the interrelationships between the Standby Gas Treatment (SBGT) System and other systems, including the following: (030.002.A.06)
a. Electrical power sources for the blowers, and
b. Instrument and/or service air systems
7. STATE how the SBGT System dampers fail on loss of instrument air or the loss of an instrument bus. (030.007.A.02)
8. EXPLAIN the Standby Gas Treatment (SBGT) System response to a Carbon bed temperature

>325°F. (030.0O1.A.04)

9. LIST the following Standby Gas Treatment (SBGT) System automatic features, and their setpoints: (030.001.A.01)
a. Initiations (4),
b. Discharge Damper Interlock,
c. Fan Trips (3),
d. Heater Trips (4)
e. Control switch STBY function
10. DESCRIBE the principles of operation of the Standby Gas Treatment (SBGT) System Carbon Filters including the processes of adsorption and radioactive decay. (030.00 1.A.02)
11. Given a drawing of the Control Room, IDENTIFY the Standby Gas Treatment (SBGT) System related control panels. (03 0.002.A.02)
12. Given a list of the Control Room controls for the Standby Gas Treatment (SBGT) System, STATE the function of each. (030.001.A.06)

P51-P52-P70-PLANT AIR-LP-03501 Page 4 of 112 PLANT AIR SYSTEMS

6. Given a list of statements, SELECT the statement which best describes the possible adverse effects of operating Instrument Air with pre or after filters bypassed per 34S0-P51-002-l Instrument and Service Air Systems. (035.005.a.02)
7. Given a list of statements, SELECT the statement that best describes the function of the Instrument Air Dryers. (035.004.a.01)
8. Given a list of statements, SELECT the statement that best describes the possible detrimental effects of shutting down both Instrument Air Dryers per 34S0-P5 1-002-1 Instrument and Service Air Systems. (035 .006.a.0 1)
9. Given a simplified drawing or P&ID, TRACE the system flowpath when the air dryers are shutdown. (035.006.a.02)
10. STATE the purpose of the Instrument Air Accumulators. (035.001.a.05)
11. Given a simplified drawing or P&ID, TRACE the Air System flowpath for cross connecting Unit 1 and Unit 2 Service Air Systems. (035.01 1.a.01)
12. Given a list of electrical buses, SELECT the bus that supplies power to the following loads:

(035.001 .a.03)

a. Station Service Air Compressors
b. Low Pressure Blower
c. Closed Cooling Water Pumps
13. Given a list, IDENTIFY the panel locations necessary to operate the air compressors and dryers per 34S0-P51-002-2/1 Instrument and Service Air Systems. (035.001.a.15)
14. LIST five plant responses to a loss of Instrument/Service Air as stated in 34AB-P51-001-2/1 Loss of Instrument and Service Air System, (200.025.a.01)
  • 15. Given a list of plant conditions, DETERMiNE if the reactor must be manually scrammed per 34AB-P51-00l-2!1 Loss of Instrument and Service Air System. (200.025.a.02)
16. Given a list, IDENTIFY the firewatch requirements if fire sprinklers are isolated during a loss of air per 34AB-P51-001-2/1 Loss of Instrument and Service Air System. (200.025.a.04)
17. Given a list of statements, SELECT the statement that best describes the response to a loss of air for the following systems: (200.025.a.05)
a. CRD
b. Reactor Water Level Control
c. SJAEs
d. Outboard MSIV s

T

18. Given a simplified drawing or P&ID of the Instrument/Service Air System and the location of a leak, PREDICT the air system response and final status. (200.025.a.03)

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 2 OF 20 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF INSTRUMENT AND SERVICE AIR SYSTEM OR 34AB-P51-001-2 4.8 WATER INTRUSION INTO THE SERVICE AIR SYSTEM LOSS OF INSTRUMENT AND SERVICE AIR SYSTEM 1.0 CONDITIONS 1.1 ANNUNCIATORS 1.1.1 CONTROL BLDG SERVICE AIR PRESS LOW (700-222) 1.1.2 AIR CMPSR 2A TRIPPED/SHUTDOWN (700-221) 1.1.3 AIR CMPSR 2B TRIPPED/SHUTDOWN (700-227) 1.1.4 AIR CMPSR 2C TRIPPED/SHUTDOWN (700-233) 1.1.5 REACTOR LEVEL CONTROL VALVE LOCKED (603-1 23) 1.1.6 SCRAM VLV PILOT AIR HDR PRESS HIGH/LOW (603-131) 1.1.7 CONTROL BLDG SERVICE AIR SHUTOFF (700-234) 1.1.8 INST AIR DRYER SYS PRESS LOW (700-219) 1.2 For Additional Annunciators Received See Attachment 2 1 .3 Decreasing Instrument and Service Air Pressure 1.4 CRD Flow decreasing (CRD High Temperature) - Flow Control Valve fails closed 1.5 Feedwater Startup Level Control Valve fails as is 1.6 Decreasing Condenser vacuum - SJAE steam supply valves fail closed 1.7 Ventilation System failures dampers fail closed 1.8 Feedwater Heater Level Control problems 1 .9 For a List of Additional Conditions See Attachment 3 MGR-0001 Rev3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 5 OF 20 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF INSTRUMENT AND SERVICE AIR SYSTEM OR 34AB-P51-001-2 4.8 WATER INTRUSION INTO THE SERVICE AIR SYSTEM 4.8 IF unable to restore system air pressure, TRANSFER operation of Refueling Floor Fuel Pool Seals to nitrogen bottle pressure supply per 34SO-P51-002-2, Instrument And Service Air Systems. LI 4.9 Manually initiate any automatic actions which did NOT occur. LI 4.10 Confirm CLOSED/CLOSE 2N21-F125, S/U Level Control Isol valve, on panel 2H11-P650. LI 4.11 Maintain Reactor water level between 15 and 45 inches utilizing 2N21-F1 10, S/U Level Control Bypass, AND/OR LI varying RFPT speed. LI 4.12 Confirm CLOSED/CLOSE 2P51-F017, Service Air Isolation valve, on panel 2H11-P650, AND LI notify Health Physics. LI 4.13 Confirm CLOSED/CLOSE 2P52-F015, Turb Bldg Inst Air After Fltrs, 2P52-D102N2P52-D1 02B, to RW Bldg Isol valve, on panel 2H11-P700. LI 4.14 IF Control Air header Pressure is below 45 PSIG AND NOT increasing, THEN perform the following:

4.14.1 START one train of 2T46-DOO1A or2T46-DOOIB, SBGT, with suction from the Reactor Building fQ Refuel Floor. LI 4.14.2 TRIP the following equipment:

MGR-0001 Rev3

SNC PLANT E. I. HATCH I Pg 13 of 20 DOCUMENT TITLE: DOCUMENT NUMBER: Ver No:

LOSS OF INSTRUMENT AND SERVICE AIR SYSTEM OR 34AB-P51-001-2 4.8 WATER_INTRUSION_INTO THE_SERVICE AIR SYSTEM ATTACHMENT 2 Att. Pg.

TITLE: ADDITIONAL ANNUNCIATORS AND EQUIPMENT 2 of 4 Non-Interruptible Essential Instrument Air supplies:

1. Refuel Floor Suction to Standby Gas treatment System (SBGT) (FOO3NB)
2. Reactor Building Suction to SBGT (FOO1A/B)
3. Hydrogen oxygen analyzer isolations
4. SBGT Train Discharge Isolation (FOO2AIB)
5. Plant Service Water (PSW) Isolation of Reactor Building Non-ECCS Equipment (F066, F067)
6. Residual Heat Removal (RHR) Suction Isolation (FO65AIB/C/D)
7. Core Spray (CS) Suction Isolation (FOI9AJB)
8. Drywell/Torus pressure sensing line isolations
9. High Pressure Coolant Injection System (HPCI) suction from Torus (F051)
10. Scram Air Supply (75 psig)
11. Control Rod Drive System (CRD) valves (30 psig)
12. Reactor Building/Torus Butterfly Vacuum Breakers (2T48-F31 1, F3 10)
13. Torus water level transmitter isolations
14. HPCI drain Air Operated Valves (AOVs)
15. Reactor Building floor drain isolations
16. Feedwater Line AOV Check Valves (FO76NB, F077A/B)
17. Outboard Main Steam Isolation Valves (MSIVs) 2B21-FO28NB!C!D)
18. Reactor Core Isolation Cooling (RCIC) Suction from Torus (E51-F003)
19. RCIC drains AOVs
20. Backup supply to Drywell Pneumatic System MGR-0009 Rev4

H C Cl)

G) H 0 Z I 0 0 0

m oc 0

0 (0 >0Z Z H

CD ci p]

0 Cn--1 H ZHr Hfl *1 0 mc..

Z C

> o I 0Z I 0 ZH 0 >

Z ci H H 0

_C1) 0 CD Zn m

rn 0o fll(J)

DH I C,)

m CDm Non-Essential C Ci)

Instrument Air HO rn;o Drywell Pneumatics ci 0

4C Drywell Pneumatics jZ 0

Non-Interruptable Essential Instrument Air 0

CD -

0:Z 0 -n 0

0

HLT-07 SRO NRC EXAM

47. 295020AK1.05 001 Unit 2 is operating at 100% power.

o Nitrogen (N) is being added to the Unit 2 Drywell (DW) from the Unit 1 N 7 2 Storage Tank 34S0-T48-002-2, Containment Atmosphere Control and Dilution System section 7.3.1, Alternate Primary Containment Nitrogen Makeup From CAD loop A, Unit 1 or Unit 2 N 2 Storage Tank is being used to add the N .

2 o 2T48-F113, Nitrogen to DW isolation valve is OPEN o 2T48-F1 14, Nitrogen to DW isolation valve is OPEN o DW venting using Standby Gas Treatment is in progress o A fault in 2C71-P001, RPS Power Dist Panel results in a loss of the2A Reactor Protection System (RPS) bus o It will take 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to restore the 2K RPS bus If the operator stationed at the 2111 1-P657 panel does NOT take any action, which ONE of the following describes the operational implications for the DW?

A 2 N addition to the DW will continue and a loss of DW cooling will eventually occur due to high DW pressure.

B. N 2 addition to the DW will continue and DW cooling will remain in operation indefinitely.

C. 2 N addition to the DW will automatically isolate and DW cooling will remain in operation indefinitely.

D. 2 N addition to the DW will automatically isolate and a simultaneous loss of DW cooling will occur.

142

HLT-07 SRO NRC EXAM

==

Description:==

The procedure (34S0-T48-002-2) explains that 2T48-Fl 13, 2T48-Fl 14 will not close on a group 2 isolation and that an operator must be stationed to close the valves in the event that an isolation signal/condition actually occurs.

As a result of the loss of RPS, a group 2 signal will be generated; however, nitrogen will continue to be added to the DW. The group 2 signal will isolate the DW vent line up. As a result, DW pressure will increase and eventually result in a DW LOCA pressure signal (assuming no operator actions are taken, which the question stem states).

Typically nitrogen addition to the DW automatically isolates on a Group 2 isolation signal; however, when performing Alternate Nitrogen makeup (as in this question), it does not auto isolate.

The B distractor is plausible since the first part is correct in that nitrogen addition will continue. The second part is plausible if the applicant assumes that DW venting remains in service, however, DW cooling will be lost eventually when a DW pressure LOCA signal occurs.

The C distractor is plausible if the applicant assumes nitrogen addition auto isolates due to the Gr 2 signal generated by the loss of RPS. The second part is plausible if the applicant assumes that Nitrogen addition will auto isolate and stop adding pressure to the DW, therefore, DW cooling will not shutdown from a DW pressure LOCA signal.

The D distractor is plausible if the applicant assumes nitrogen addition auto isolates due to the Gr 2 signal generated by the loss of RPS. The second part is plausible if the applicant assumes that DW cooling is lost as a result of a group 2 signal which is generated due to the RPS loss.

DW cooling loss would actually occur if a Gr 2 signal due to DW pressure (1.85 psig) occurs.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

143

HLT-07 SRO NRC EXAM

References:

NONE K/A:

295020 Inadvertent Containment Isolation AK1. Knowledge of the operational implications of the following concepts as they apply to INADVERTENT CONTAINMENT ISOLATION: (CFR: 41.8 to 41.10)

AK1.05 Loss of drywell/containment cooling 3 3.6 LESSON PLAN/OBJECTIVE:

T23-PC-LP-01301, Primary Containment, EO 013.009.A.01 References used to develop this question:

34S0-T48-002-2, Containment Atmospheric Control and Dilution Systems 34AB-C71-002-2, Loss of RPS Bank question from HLT-4 NRC Exam Q# 048 144

T23-PC-LP-01301 Page 9 of 160 PRIMARY CONTAINMENT

(013.061.A.01)

(013.062.A.01)

  • 36. Per 34S0-T48-002-1/2, Containment Atmospheric Control and Dilution System, DETERMINE the sequence of actions necessary to makeup Nitrogen to the Containment by the normal method. (013.008.A.0l)
  • 37 Per 34S0-T48-002-112, Containment Atmospheric Control and Dilution System, DETERMINE the sequence of actions necessary to makeup Nitrogen to the Containment from Unit 1 Nitrogen Storage Tank. (013.009.A.01)
  • 38. Per 34S&-T48-002-1/2, Containment Atmospheric Control and Dilution System, DETERMINE the sequence of actions necessary to makeup Nitrogen to the Containment from Unit 2 Nitrogen Storage Tank. (013.0l0.A.01)
  • 40. Per 34G0-OPS-028-2, Drywell Closeout, DETERMINE the actions necessary to perform a Drywell Closeout Inspection. (013.021 .A.0 1)
  • 42. DETERMINE the major steps to startup the DW ventilation system per the normal operating procedure 34S0-T47-00 1-1/2. (043 .002.A.0 1)
  • 43 With Suppression Pool level slightly below normal, PERFORM a low volume fill of the suppression chamber per 34S0-E21-00l-l/2, Core Spray System to maintain suppression chamber level within the limits of Technical Specifications. (013.001 .A.03)
  • 44 With Suppression Pool level extremely low, PERFORM a high volume fill of the suppression chamber per 34G0-OPS-087-l/2, Suppression Chamber Fill and Drain to maintain suppression chamber level within the limits of Technical Specifications. (013.002.A.04)
  • 45 Given plant conditions DETERMINE the leakage rates into the Drywell Equipment and Floor Drain Sumps per 34SV-SUV-019-l/2, Surveillance Checks. (040.004.A.02)

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 28 OF 62 DOCUMENT TITLE:

ATMOSPHERIC CONTROL AND DOCUMENT NUMBER: VERSION NO:

DILUTION SYSTEMS - -

7.3 INFREQUENT OPERATIONS 7.3.1 Alternate Primary Containment Nitrogen Makeup From CAD Loop A, Unit 2 Or Unit I N 2 Storage Tank I CONTINUOUS 1 This method of makeup to the Primary Containment may be used H the following are met:

  • Obtain Shift Supervisor permission to open the CAD Nitrogen Supply valves.
  • Station an operator at the valve control switches to close the CAD valves H a Group 2 NOTES: isolation signal is received.
  • 2T48-F1 13, 2T48-F1 14, 2T48-F1 15, & 2T48-F1 16 Do NOT receive a signal to close Jr a Group 2 isolation signal is received. The time the valves are allowed to be OPEN shall be minimized. Jr the valves are required to be OPEN for more than four hours the SS AND SOS shall be notified in order to receive their approval.

7.3.1.1 ll Nitrogen is to be supplied from Unit I Nitrogen Storage Tank, perform the following:

7.3.1.1.1 Notify the Unit 1 Shift Supervisor AND Unit 1 Plant Operator that Nitrogen Makeup to Unit 2 is being performed from Unit 1 Nitrogen Storage Tank. El 7.3.1.1.2 Open the following valves:

. 1T48-FO13B, Nitrogen To Unit 2 Outboard Isol Vlv (IHII-P654) LI 7.3.1.2 Station an operator at panel 2H11-P657, WHILE valves 2T48-F113, 2T48-F114, 2T48-F115, AND/OR 2T48-F116 are OPEN. LI MGR-0001 Rev. 4.0

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 29 OF 62 DOCUMENT TITLE:

CONTAINMENT ATMOSPHERIC CONTROL AND DOCUMENT NUMBER: VERSION NO:

DILUTION SYSTEMS 34SO-T48-002-2 22.0 2T48-F1 I 3, 2T48-F1 14, 2T48-F1 15, AND 2T48-F1 16 MUST BE CLOSED IMMEDIATELY CAUTION:

i[. A GROUP 2 ISOLATION SIGNAL IS RECEIVED.

7.3.1.3 Open makeup valves from the desired path (2H11-P657):

CAD LOOP A DRYWELL MAKEUP VALVES 2T48-F113 El 2T48-F114 LI TORUS MAKEUP VALVES 2T48-F115 LI 2T48-F116 LI 7.3.1.4 Throttle open 2T48-F1 I 2A, using 2T48-R613A, Drwl/Torus Flow Cntl for FII2A, to the desired Nitrogen flow rate. LI 7.3.1.5 Monitor Drywell pressure on 2T48-R607A AND 2T48-R607B, Recorders. LI 7.3.1.6 WHEN Drywell makeup is complete, THEN close 2T48-F1 12A.

7.3.1.7 Notify the Unit 1 Shift Supervisor AND Plant Operator, that Unit 2 use of Unit 1 Nitrogen Storage Tank has been discontinued. LI 7.3.1.8 Notify the Shift Supervisor.

THEN restore the CAD System with independent verification per Attachment 1. LI MGR-0001 Rev. 4.0

SOUTHERN NUCLEAR PLANTE.I.HATCH I PAGE9OF12 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF RPS 34AB-C71-002-2 4.9 ATTACHMENT 3 ATTACHMENT PAGE:

TITLE: LOSS_OF_RPS_BUS AUTOMATIC ACTIONS 1 OF 4 AUTOMATIC ACTIONS LOSS OF RPS BUS A WILL RESULT IN THE FOLLOWING ACTIONS:

Half Scram on RPS Channel A The following Group 2 isolation valves close:

2T48-F339 Torus Vent & Relief VIv 2T48-F341 Drywell Vent & Relief Vlv 2T48-FII8A 2 Makeup to Drywell Vlv N

2T48-F1 18B 2 Makeup to Torus Vlv N

2T48-F333A Torus Vent Isol Vlv 2T48-F333B Torus Vent Isol VIv 2T48-F335A Drywell Vent Isol Vlv 2T48-F335B Drywell Vent Isol VIv 2G11-F003 DrwI Sumps Floor Drain VIv 2G11-F019 DrwI Sumps Equip Drain Vlv 2D11-F050 Pri CNMT Fis Prod Mon Inbd Isol 2D11-F051 Pri CNMT Fis Prod Mon Inbd lsol 2D11-F071 Fis Prod Mon Sample Line lsol 2G51-FO11 Torus Water Cleanup lnbd Isol 2G51-F017 Torus Water Makeup Outbd Isol 2E41-SV-F122 Post Acc Rx CIntCNMTATMOS Sample lnbd Isol 2T48-F209 DrwI to Torus DP Sys Inbd Isol 2T48-F21 1 Drwl to Torus DP Sys lnbd Isol 2T48-F307 Drywell Air Purge Vlv 2T48-F309 Torus Air Purge VIv 2T48-F318 Torus Vent Vlv 2T48-F31 9 Drywell Vent VIv 2P33-F002 Pri CNMT ATMOS H Anly 0

2 Ch B Inbd Isol 2P33-F003 Pri CNMT ATMOS 0 Anly 2

H Ch A lnbd lsol 2P33-F004 Pri CNMT ATMOS 0 Anly 2

H Ch A Return Line Inbd Isol 2P33-F006 Pri CNMT ATMOS O Anly 2

H Ch B lnbd Isol 2P33-F007 Pri CNMT ATMOS 0 Anly 2

H Ch A Inbd Isol 2P33-F005 Pri CNMT ATMOS 0 Anly 2

H Ch B Return Line Inbd Isol 2P70-F002 Drywell Pneumatic Inboard Suction Isolation MGR-0009 Ver. 4

HLT-07 SRO NRC EXAM

48. 295021G2.2.40 001 Unit 2 is shutdown with the following conditons:

O Rx pressure 34 psig o 2A Rx Recirculation pump Off o 2B Rx Recirculation pump Off Which ONE of the choices below completes BOTH the following statement?

JAW Tech Spec 3.4.7, Residual Heat Removal (RHR) Shutdown Cooling System Hot -

Shutdown the MINMUM number of RHR Shutdown Cooling (SDC) subsystems required to be operable, (without requiring entry into a Required Action Statement) (RAS), is Also, lAW with Tech Spec 3.4.7 and with current plant conditions, RHR SDC subsystem is required to be in operation.

A. one; one B. one; neither C two; one D. two; neither 145

HLT-07 SRO NRC EXAM

==

Description:==

3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System Hot Shutdown LCO 3.4.7 states; Two Shutdown cooling subsystems shall be OPERABLE and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation.

APPLICABILITY: MODE 3 with reactor steam dome pressure less than the RHR low pressure permissive pressure.

Given plant conditions of <135 psig Rx pressure and no recirc pump operating, 2 subsystems are required, and one Shutdown Cooling System is required to be in operation.

With no recirc pumps running one RHR SDC subsystem would be required to be in service. In mode 5 only one RHR subsystem is required JAW TS 3.9.7.

The A distractor is plausible if the applicant confuses Mode 5 requirement of only one SDC subsytem with the current condition requirements. The second part is correct.

The B distractor is plausible if the applicant confuses Mode 5 requirement of only one SDC subsytem with the current condition requirements. The second part is plausible if the applicant confuses neither subsystem would be required to be in service if a recirc pump was in service with the current condition requirements.

The D distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses neither subsystem would be required to be in service if a recirc pump was in service with the current condition requirements.

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

References:

NONE K/A:

295021 Loss of Shutdown Cooling 2.2.40 Ability to apply Technical Specifications for a system.

146

HLTM7 SRO NRC EXAM (CFR: 41.10/43.2/ 43.5 /45.3) 3.4 4.7 LESSON PLAN/OBJECTIVE:

E11-RHR-LP-00701, RHR System, EO 400.067.a.27 LT-LP-30005, Technical Specifications, EO 300.006. A.26 References used to develop this guestion:

Tech Spec 3.4.7, Residual Heat Removal (RHR) Shutdown Cooling System Hot Shutdown Modified from HLT-5 NRC Exam Q#49 ORIGINAL QUESTION (HLT-5 NRC Exam Q#49)

Unit 2 is shutdown with the following conditons:

o Rx pressure 134 psig o 2A Rx Recirculation pump Running o 2B Rx Recirculation pump Off Which ONE of the choices below completes BOTH the following statement?

lAW Tech Spec 3.4.7, Residual Heat Removal (RHR) Shutdown Cooling System Hot -

Shutdown the MINMUM number of RHR Shutdown Cooling (SDC) subsystems required to be operable, (without requiring entry into a Required Action Statement) (RAS), is Also, JAW with Tech Spec 3.4.7 and with current plant conditions, RHR SDC subsystem is required to be in operation.

A. one; one B. one; neither C. two; one D.V two; neither 147

E11-RHR-LP-00701 Page 8 of 130 RESIDUAL HEAT REMOVAL SYSTEM

  • 54 Given appropriate references, DETERMINE proper operator actions to prove RHR valve operability during a Refueling outage, per 34SV-E1 1-002-1/2, RHR Valve Operability.

(100.035.b.01)

  • 56 Given plant conditions involving the RHR System, DETERMINE if a Technical Specification Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (300.006 .a. 34)
  • 57 Given plant conditions involving the RHR System, DETERMINE the Required Actions for Completion Times 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in accordance with Technical Specifications for any combination of INOPERABLE systems, structures or components. (300.010.a.27)
  • 5g Given plant conditions involving the RHR System, DETERMINE the Required Action(s) and Completion Time(s) in accordance with Technical Specifications for any combination of INOPERABLE systems, structures or components and the bases for the action(s). (SRO Only)

(400.067.a.27)

  • 59 Given plant conditions involving the RHR System, DETERMINE the Required Action(s) and Completion Time(s) in accordance with the Technical Requirements Manual (TRM) for any combination of INOPERABLE systems, structures or components and the bases for the action(s).

(SRO Only) (300.011 .a.20)

  • 60. Given plant conditions, DESCRIBE the actions necessary to start RHR Loop B in Torus Cooling and/or Spray from the RSDP per 3 1RS-OPS-001-l/2, Shutdown from Outside the Control Room.

(007.020.a.0l)

  • 61. Given plant conditions, DESCRIBE the actions necessary to start RHR Loop B in Shutdown Cooling Mode from the RSDP per 31RS-OPS-00l-l/2, Shutdown from Outside the Control Room. (007.021.a.0l)
  • 62. Given plant conditions and that RHR Loop B is being operated from the RSDP, DESCRIBE the actions necessary to shutdown RHR Loop B per 31RS-OPS-00l-1/2, Shutdown from Outside the Control Room. (007.026.a.0l)

Objectives marked by a RED (*) are required during RO-305 and SR-305 of the Initial License program.

LT-LP-30005 Page 3 of 86 TECHNICAL SPECIFICATIONS

26. IDENTIFY the rules for extending SR frequencies and repeating required action completion times (SR 3.0.2). (300.006.A.24) (SRO Only)
27. IDENTIFY the rules for a condition where an SR is discovered to have been missed (SR 3.0.3).

(300.006.A.25) (SRO Only)

28. STATE the actions required for a safety limit violation per Technical Specifications.

(3 00.003 .A.03)

29. Given a copy of Technical Specifications and a set of plant conditions, DETERMINE the actions required if the conditions do not comply with the LCO. (300.0 l0.A.06) (SRO Only)
30. IDENTIFY the Technical Specification knowledge requirements, in accordance with NUREG 1021 and Attachment 1 of this Lesson Plan, for a Reactor Operator Candidate. (300.006.A.26)
31. IDENTIFY the Technical Specification knowledge requirements, in accordance with NUREG 1021 and Attachment 1 of this Lesson Plan, for a Senior Reactor Operator Candidate.

(300.006.A.27) (SRO Only)

Not Selected for Requal

RHR Shutdown Cooling System - Hot Shutdown 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown LCO 3.4.7 Two RHR shutdown cooling subsystems shall be OPERABLE and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation.

1. Both RHR shutdown cooling subsystems and recirculation pumps may be removed from operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period.
2. One RHR shutdown cooling subsystem may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for performance of Surveillances.

APPLICABILITY: MODE 3 with reactor steam dome pressure less than the RHR low pressure permissive pressure.

ACTIONS NOTE------

Separate Condition entry is allowed for each RHR shutdown cooling subsystem.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or two RHR shutdown A.1 Initiate action to restore Immediately cooling subsystems RHR shutdown cooling inoperable, subsystem(s) to OPERABLE status.

AND (continued)

HATCH UNIT 2 3.4-13 Amendment No. 210 I

HLT-07 SRO NRC EXAM

49. 295023AA1.04 001 An event occurs on Unit 2 causing the Fuel Pool water level to decrease.

The following Area Radiation Monitors (ARM) red Trip Lights illuminate at Control Room ARM Panel 2D21-P600:

o 2D21-K6O1A, Reactor head laydown area o 2D21-K6O1M, Spent Fuel Pool & New Storage Which ONE of the following predicts how the Main Control Room Environmental Control (MCREC) system is affected and operation of the ARMs?

With the above conditions, the MCREC system will After Fuel Pool water level is restored AND the above ARMs are indicating below their Trip setpoints, the ARM red TRIP lights reset.

A. remain in the Normal Mode; must be manually B. remain in the Normal Mode; will automatically C align to the Pressurization Mode; must be manually D. align to the Pressurization Mode; will automatically 148

HLT-07 SRO NRC EXAM

==

Description:==

The lowering fuel pool level, by itself, would cause general area radiation levels to increase. In this case, the grappled bundle is uncovered which will produce high radiation levels (per LT-LP-10015). The ARM trip lights indicate that the ARMs have exceeded the setpoint (15 mr/hr) which causes the red trip lights to illuminate and the MCREC System to align to the Pressurization Mode. Once rad levels are below the ARM setpoint, the trip will automatically reset but the red trip lights must be manually reset at P600.

The A distractor is plausible if the applicant does not know or confuses the MCREC System response to the above ARMs. The second part is correct.

The B distractor is plausible if the applicant does not know or confuses the MCREC System response to the above ARMs. The second part is plausible if the applicant confuses the operation of the ARMs with other radiation monitors which will reset (clear) when radiation levels drop below their trip setpoint.

The D distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses the operation of the ARMs with other radiation monitors which will reset (clear) when radiation levels drop below their trip setpoint.

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

References:

NONE K/A:

295023 Refueling Accidents AA1. Ability to operate andlor monitor the following as they apply to REFUELING ACCIDENTS: (CFR: 41.7 / 45.6)

AA1 .04 Radiation monitoring equipment 3.4 3.7 LESSON PLAN/OBJECTIVE:

149

HLT-07 SRO NRC EXAM Dl l-PRM-LP-10007, Process Radiation Monitors, EQ 200.030.a.l 1 Z41-MCREC-LP-03701, Control Room Environmental Control, EO 037.008.A.02 References used to develop this question:

34AR-601-312-2, Refueling Floor Area Radiation High Modified from HLT-4 NRC Exam Q#50 ORIGINAL QUESTION (HLT-4 NRC Exam Q#50)

An irradiated fuel bundle is on the Unit 2 Refueling Bridge Main Grapple, which is in the FULL-UP position, and can NOT be lowered due to an equipment malfunction.

o The Fuel Pool Transfer Canal seals deflate which causes Fuel Pool water level to decrease to its lowest possible level o Only the 2D21-K6O1A and 2D2l-K6O1M Area Radiation Monitors (ARM) red TRIP lights illuminate and these two ARMs will not reset Which ONE of the following predicts the Fuel Pool level and how the Main Control Room Environmental Control (MCREC) system is affected.

Fuel Pool Water level will (1) the top of the Fuel Bundle and the MCREC system will (2)

A. (1) remain above (2) remain in Normal Mode B. (1) remain above (2) align to the Pressurization Mode C. (l)gobelow (2) remain in Normal Mode D.V (l)gobelow (2) align to the Pressurization Mode 150

Dhl-PRM-LP-10007 Page 3 of 87 PROCESS RADIATION MONITORS

8. Given plant conditions, DETERMINE if an automatic initiation signal has occurred for the following Kaman systems. (200.030.a.08)
a. Main Stack Effluent Accident Range Gas Monitor
b. Reactor Building Vent Effluent Accident Range Gas Monitor
9. IDENTIFY the type of radiation detector used for the following systems: (200.030.a.09)
a. PSW
b. RBCCW
c. Liquid Radwaste
10. Given plant conditions, EVALUATE those conditions and DETERMINE the automatic actions that occur when any of the following monitors reach their Hi-Hi setpoint: (200.030.a.10)
a. Unit One Reactor Building Ventilation Exhaust
b. Unit One Refuel Floor Ventilation Exhaust
c. Unit Two Reactor Building Ventilation Exhaust
d. Unit Two Refuel Floor Ventilation Exhaust
11. From a list, IDENTIFY the function of each position of the selector switch on an Area Radiation Monitor indicator trip unit. (200.030.a.1 1)
12. STATE the reason for monitoring radiation in the RBCCW system, PSW system, and the Liquid Radwaste system. (200.030.a. 13)
13. Given system conditions, EVALUATE and DETERMINE if Liquid Radwaste should have tripped or isolated. (200.030.a. 14)
14. Given plant conditions involving Process Radiation Monitors, DETERMINE if a Technical Specification Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (300.01 0.C.0 1)
15. Given plant conditions involving Process Radiation Monitors, DETERMINE if a Technical Requirements Manual (TRM) Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (300.01 0.C.02)
16. Given plant conditions involving Process Radiation Monitors, DETERMINE the Required Actions for Completion Times < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in accordance with Technical Specifications for any combination of INOPERABLE systems, structures or components. (300.006.C.0l)
17. Given plant conditions involving Process Radiation Monitors, DETERMINE the Required Action(s) and Completion Time(s) in accordance with Technical Specifications for any combination of INOPERABLE systems, structures or components and the bases for the action(s).

(SRO Only)

(300.006.C.02)

Z41-MCREC-LP-03701-06 Page 2 of 60 MAIN CONTROL ROOM ENVIRONMENTAL CONTROL Initial License (LT) ENABLING OBJECTIVES

1. STATE the purpose of the Control Room HVAC System. (037.006.A.01)
2. Per 34S0-Z41-001-1 Control Room Ventilation System, DETERMINE the required steps to place the Control Room HVAC System in Normal Mode while ensuring that general notes are met. (037.007.A.0l)
3. Given that the exceeding of an auto initiation setpoint is imminent, DETERMINE the required steps to place the Control Room HVAC System in the Pressurization Mode per 34S0-Z4l-001-l Control Room Ventilation System. (037.008.A.01)
4. Given a set of plant conditions, DETERMINE the mode of operation in which the Control Room HVAC System should be operating. (037.008.A.02)
5. Given that the auto initiation signal, of MCRECS in Pressurization Mode, has been exceeded DETERMINE the required actions to ensure MCR ventilation has switched to Pressurization Mode per 3450-Z41-00l-l Control Room Ventilation System. (037.017.0.01)
6. Per 34S0-Z41-001-1 Control Room Ventilation System, DETERMINE the required steps to place the Control Room HVAC System in the Isolation Mode while ensuring that general notes are met. (037.009.A.0l)
7. Per 34S0-Z41-00l-l Control Room Ventilation System, DETERMINE the required steps to place the Control Room HVAC System in the Purge Mode while ensuring that general notes are met. (037.0l0.A.0l)
8. Per 34S0-Z4 1-001-1, Control Room Ventilation System, DETERMINE the steps required to perform the initial lineups for MCREC System. (037.006.A.07)

ABNORMAL \ SURV \ TECH SPECS

  • 9 Given plant conditions involving the Control Room HVAC System, DETERMINE if a Technical Specification Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (300.01 0.C .01)
10. Given plant conditions involving the Control Room HVAC System, DETERMINE if a Technical Requirements Manual (TRM) Limiting Condition for Operation has been exceeded.

(implicit in this objective is a determination ofAPPLICABILITY and associated NOTES)

(300.0 10.C.02)

11. Given plant conditions involving the Control Room HVAC System, DETERMINE the Required Actions for Completion Times 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in accordance with Technical Specifications for any combination of INOPERABLE systems, structures or components. (300.006.C.01)

10 IDENTIFICATION:

ALARM PANEL 601-3 REFUELING FLOOR Zr ZZZ*

AREA RADIATION HIGH DEVICE: SETPOINT:

2D21-K601-A,E,M See 64C1-CAL-002-0 2D21-K61 1-K,L

2.0 CONDITION

3.0 CLASSIFICATION

One Q more of the Refueling Floor Area Radiation Monitors has E*STATUS 0

exceeded its setpoint.

5.0 OPERATOR ACTIONS:

I 2H11-P601 Panel 601-3 5.1 Check for ILLUMINATED red upscale light AND high reading on individual indicator/trip units on panel 2D21-P600. LI 5.11 Attemptto resetTrip Unit. LI 5.2 H2D21-K601 A.QM Tripped, THEN confirm Control Room ventilation shifted to pressurization (Mode II) per 34S0-Z41 -001 -1, Control Room Ventilation System. LI 5.3 IF Trip Unit will NOT reset, perform the following:

5.3.1 Confirm monitor reading by checking 2D21-R600A, Area Radn Monitor Recorder, panel 2H1 1-P600 (see channel below) LI 5.3.2 Contact Health Physics to confirm radiological condition in the affected area, evacuate area IF required. LI 5.3.3 Periodically monitor panel 2D21-P600 to ensure no other channels on the Refueling Floor are indicating high OR in alarm. LI 5.3.4 Enter 34AB-T22-003-2, Secondary Containment Control. LI 5.3.5 IF a radioactive release outside the Reactor Building is occurring, OR has occurred, enter 34AB-D11-001-2, Radioactivity Release Control. LI

6.0 CAUSES

6.1 High radiation causing a trip of one OR more of the following indicator trip units MONITOR CHANNEL LOCATION SENSOR TRIP UNIT I Reactor Head Layout Area 2D21-N002-A K601-A 5 Dryer/Separator Pool 2D21-N002-E K601-E 10 Spent Fuel/Fuel Pool Areas 2D21-N002-M K601-M 21 Reactor Vessel Refueling Area 2D21-N012-K K611-K 22 Reactor Vessel Refueling Area 2D21-N012-L K61 1-L 6.2 Malfunction of trip unit

7.0 REFERENCES

8.0 TECH. SPECS./TRM/ODCM/FHA:

7.1 H-27805 thru H-27810 and H-27992, Area Radiation Monitoring Sys 2D21 Elementary Diagram 8.1 TRM Table T3.3.7-1 34AR-60 1-31 2-2 Ver. 4.3 MGR-0048 Ver. 5.0 AG-MGR-75-1 101

HLT-07 SRO NRC EXAM

50. 295024EA2.06 001 A steam line break inside containment has occurred on Unit 1.

o Drywell pressure is steady at 10.5 psig o Neither Drywell or Torus sprays have been initiated Which ONE of the following describes the effect of the steam line break on Tonis water temperature?

A. The saturation temperature of the Torus water will be lower than at normal operating conditions because of the non-condensable gases.

B The Torus water temperature will heat up throughout the Torus due to the design of the downcomers.

C. The Torus water temperature will only heat up directly under the area of the DW leak due to the energy being distributed directly to the torus water in that area.

D. The Torus water average temperature indication is unreliable until suppression pool cooling is established.

151

HLT-07 SRO NRC EXAM

==

Description:==

The steam will enter the torus via a ring header and downcomers. The ring header helps to ensure steam distribution is approximately equal throughout the torus. Plausibility for local area heating of the torus is SRVs leaking.

The A distractor is plausible if the applicant confuses that the saturation temperature will be higher due to the higher pressure and also confuses the fundamental principles of Psat-Tsat relationship.

The C t distractor is plausible if the applicant confuses how a steam leak will be distributed to the Torus and does not realize it will be evenly dispersed through the downcomers. Also plausible if the applicant does not understand the purpose/design of the ring header!

downcomers.

The D distractor is plausible if the applicant is not familiar with how water temperature is measured and monitored in the suppression chamber. The temperature monitors of the torus still work and the average temp is just an average of all the monitors.

A. Incorrect See description above.

B. Correct See description above.

C. Incorrect See description above.

D. Incorrect See description above.

152

HLT-07 SRO NRC EXAM

References:

NONE K/A:

295024 High Drywell Pressure EA2. Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE: (CFR: 41.10 /43.5 /45.13)

EA2.06 Suppression pooi temperature 4.1 4.1 LESSON PLAN/OBJECTIVE:

Steam Tables T23-PC-LP-0 1301, Primary Containment, EQ 200.004.A.03 References used to develop this iuestion:

Bank Question from HLT-4 NRC Exam 2009301-051 153

T23-PC-LP-01301 Page 8 of 160 PRIMARY CONTAINMENT

28. Given a list of statements, IDENTIFY the statement which best describes the plant conditions which will generate the following PCIS isolation signals: (013.045.A.05, 013.046.A.05, 013.047.A,05)
a. Group 1 isolation
b. Group 2 isolation
c. Group 3 isolation
d. Group 4 isolation
e. Group 5 isolation
f. Group 6 isolation
29. Given a list of statements, IDENTIFY the statement which best describes the steps required to reset the following PCIS isolation signals: (013.045.A.06, 013.046.A.06, 013.047.A.06, 013.067.A.01, 040.006.A.Ol)
a. Group 1 isolation
b. Group 2 isolation
c. Group 3 isolation
d. Group 4 isolation
e. Group 5 isolation
f. Group 6 isolation
30. Given a list of statements, IDENTIFY the statement which best describes the primary containment characteristics during a DBA LOCA, including the process by which energy is absorbed and dissipated by the primary containment. (200.004.A.03)
31. Given a list of statements, IDENTIFY the statement which best describes the consequences of having the torus to drywell vacuum breakers fail either open or closed during a DBA LOCA.

(200.004.A.02)

NORMAL OPERATIONS

(0l3.006.A.0l)

(013.007.A.0l)

SNCPLANTE.I.HATCHI I Pq139of208 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION No:

REACTOR RECIRCULATION SYSTEM 34SO-B31 -001-1 41.1 ATTACHMENT 1 Attachment Page TITLE: SATURATION PRESSURE/TEMPERATURE 1 of 3 Abs. Press. °F Abs. Press. Abs. Press.

Lb./Sq. In. Lb/Sq. In. Lb/Sq. In.

p t p t p 0.20 53.14 7.5 179.94 35 259.28 0.25 59.30 8.0 182.86 36 260.95 0.30 64.47 8.5 185.64 37 262.57 0.35 68.93 9.0 188.28 38 264.16 0.40 72.86 9.5 190.80 39 265.72 0.45 76.38 0.50 79.58 10 193.21 40 267.25 0.60 85.21 11 197.75 41 268.74 0.70 90.08 12 201.96 42 270.21 0.80 94.38 13 205.88 43 271.64 0.90 98.24 14 209.56 44 273.05 14.696 212.00 1.0 104.74 15 213.03 45 274.44 1.2 107.92 16 216.32 46 275.80 1.4 113.26 17 219.44 47 277.13 1.6 117.99 18 222.41 48 278.45 1.8 122.23 19 225.24 49 279.74 2.0 126.08 20 227.96 50 281.01 2.2 129.62 21 230.57 51 282.26 2.4 132.89 22 233.07 52 283.49 2.6 135.94 23 235.49 53 284.70 2.8 138.79 24 237.82 54 285.90 3.0 141.48 25 240.07 55 287.07 3,5 147.57 26 242.25 56 288.23 4.0 152.97 27 244.36 57 289.37 4.5 157.83 28 246.41 58 290.50 29 248.40 59 291.61 5.0 162.24 30 250.33 60 292.71 5.5 166.30 31 252.22 61 293.79 6.0 170.06 32 254.05 62 294.85 6.5 173.56 33 255.84 63 295.90 7.0 176.85 34 257.58 64 296.94 MGR-0009 Ver. 5

Controlled Copy

[RO Questions 51 - 75 f Sc MATERIAL If found unattended, IMMEDIATELY notify:

Charlie Edmund, Anthony Ball, Ray Rutan or Ed Jones at ext. 3123

i 0)

C N -

N N C N N N N c 0 N C*C*CC*C NCCNJ CNCCCC.N tr Q 00 C N N 00 C C C C C C C C N D N $ C C C N C O

0) C c C C N N z N N N c c c c r c C C C C C C C C C C C C C C C C C C C C C N N N s, CCCCCCCICC C C C C C C C C C N N N c C C C C C C C C C C C C C C C C C C N N c CNN C C* Q\C C CC C CCCNNC N N N N N N N N N Cfl C C N (D () ( (Z c) cZ c Q N N N N N N N N N N N N N N N N - .C N c N 00 C C N cr D N 00 O C N c r t .0 N 00 C C N c - .0 N 00 0 C N c .0 N 00 0 0 r .0 .0 .0 .0 .0 .0 .0 .0 .0 N N N N N N N N N N 00000000000000000000 C Q C 0 C 0 i

t N c .0 00 N .0 N C N N .0 C C C C C C N c N .0 N N 00 r .0 N r C N N N r N C C C C C C C C C C C C C C C C C C C C C N C C C r C C C C f N c N N c N N N ICICjCICICICICICICICICICICICICICICICICICICICICICICjCICICICICICICjCCIClCjCICICIC C CICICICICCICICICI I IIININIc I00I.0IIvIt tr c

II ICICICICICICICICICI_I_I_I_I_I_I_I-I_I_I_ININIcICIC I.0I.0I.0I.0I.0I.0I.0i00ICIQICIQ N TITT

HLT-07 SRO NRC EXAM

51. 295025G2.1.27 001 Which ONE of the choices below completes the following statements?

JAW TS 3.4.3, Safety/Relief Valves (S/RVs), the SRV Safety function requires a MINIMUM of SRVs to be operable.

if reactor pressure reaches 1200 psig, the Reactor Coolant System DESIGN pressure have been exceeded.

A. 10; will B 10; will NOT C.5; will D5; will NOT 154

HLT-07 SRO NRC EXAM

==

Description:==

lAW 3.4.3 Safety/Relief Valves (S/RVs) The safety function of 10 of 11 S/RVs shall be OPERABLE. 5 SRVs is the minimum number of SRVs for an Emergency Depress.

JAW TS 2.1.2, The RCS pressure SL has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section ifi of the ASME, Boiler and Pressure Vessel Code, 1968 Edition, including Addenda through the Summer of 1970 (Ref. 5), which permits a maximum pressure transient of 110%, 1375 psig, of design pressure 1250 psig. The SL of 1325 psig, as measured in the reactor steam dome, is equivalent to 1375 psig at the lowest elevation of the RCS. Therefore, the design pressure limit has not been exceeded.

The A distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses design pressure and think it has been exceeded.

The C distractor is plausible if the applicant confuses the minimum number of SRVs for an Emergency Depress with the number of SRVs required meet the SRV Safety function. The second part is plausible if the applicant confuses design pressure and think it has been exceeded.

The D distractor is plausible if the applicant confuses the minimum number of SRVs for an Emergency Depress with the number of SRVs required meet the SRV Safety function. The second part is correct.

A. Incorrect See description above.

B. Correct See description above.

C. Incorrect See description above.

D. Incorrect See description above.

155

HLT-07 SRO NRC EXAM

References:

NONE K/A:

295025 High Reactor Pressure 2.1.27 Knowledge of system purpose and/or function. (CFR: 41.7) ...... 3.9 4.0 LESSON PLAN/OBJECTIVE:

LT-LP-30005, Technical Specifications, EO 300.003 .A.04 B21-SLLS-LP-01401, Main Steam & Low Low Set, EO 300.011.A.14 References used to develop this question:

TS 2.1.2, Reactor Coolant System Safety Limit Unit One & Unit Two Tech Spec 3.4.3, Safety/Relief Valves Unit One and Unit Two Tech Spec 2.1, Safety Limits 156

B21-SLLS-LP-01401-O8 Page 3 of 115 MAiN STEAM AND LO LO SET

8. Given a simplified drawing or P&ID of the Main Steam Line Drawing, IDENTIFY the following valves: (014.012.B.0l)
a. 2B21-F016
b. 2B21-F019
c. 2B21-F020
d. 2B21-F038
e. 2B21-F021
9. Given Plant conditions, EVALUATE the conditions, and DETERMINE if any SRV should be open. (0l4.003.A.06) (200.009.A.04)
10. STATE the specific condition that will cause the rnbe indicator for a SRV to illuminate.

(0l4.003.A02)

11. From a list, SELECT (5) FIVE Main Control Room indications that would indicate an SRV is open. (014.003.A.01)
12. Given plant conditions, EVALUATE those conditions and DETERMINE if a GP I PCIS isolation should have occurred. (014.007.A.01)
  • 14. Given RPV pressure is above SRVs setpoints and no SRVs are open, from a list, SELECT the correct operator actions to be taken per 34AB-B21-003-1/2; Failure of Safety/Relief Valves.

(200.009.A.01)

  • 15. Given RPV Pressure is below the SRV reset points and one or more SRVs are open, STATE the correct operator action which should be taken per 34AB-B2 1-003-1/2; Failure of Safety/Relief Valves. (200.009.A.02)
  • 16. Give plant conditions; DETERMINE the actions that must be taken per 34AB-B2 1-001 1/2, -

Main Steam Line Radiation. (200.098.A.01)

  • 17. Given plant conditions involving the Main Steam/LLS System, DETERMINE if a Technical Specification Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (300.011 .A. 14)
  • 18. Given plant conditions involving the Main Steam/LLS System, DETERMINE the Required Action(s) and Completion Time(s) in accordance with Technical Specifications for any combination of INOPERABLE systems, structures or components and the bases for the action(s). (SRO Only) (300.006.A.30)
  • 19. Given plant conditions resulting in ATTS alarm conditions and applicable procedures, DETERMINE the proper actions that should be taken per: (055.00l.A.13)
a. LLS Logic AJC (B/D) Armed (34AR-602-123(223)-2/l / 34AR-602-135(235)-2/l)
b. LLS Logic A/C (B/D) Power Loss (34AR-602-117(217)-2/1 / 34AR-602-129(229)-2/l)
c. ECCS/RPS Division 1(11) Trouble (34AR-602-l 10-2 (602-124-1) / 34AR-602-130-l/2)

LT-LP-3 0005 Page 2 of 86 TECHNICAL SPECIFICATIONS

10. IDENTIFY the required completion times for multiple, sequential, inoperabilities and sequential restorations. (300.006.A. 10)
11. STATE two (2) times when a required action need not be completed. (300.006.A.11)
12. IDENTIFY how a specification tells us it permits a condition to have multiple completion times.

(300.006.A. 12)

13. STATE the meaning of immediate as applied to Technical Specifications. (300.006.A.13)
14. LIST the four safety limits for the designated unit per Technical Specifications. (300.003 .A.0 1)
15. Given a safety limit, STATE the basis for that safety limit per Technical Specifications.

(300.003.A.04)

16. DEFINE a Limiting Condition for Operation (LCO 3.0.1). (300.006.A.14)
17. STATE the purpose of the required actions and completion times (LCO 3.0.2). (300.006.A.l5)
18. STATE the required action when there is not a set of corresponding conditions for inoperable Tech Spec equipment (LCO 3.0.3). (300.006.A.16) (SRO Only)
19. Given a start time and conditions requiring LCO 3.0.3 entry, CALCULATE the time when MODE 4 must be entered from various starting MODES. (300.006.A.17) (SRO Only)
20. IDENTIFY the rules for entering the applicability of an LCO (LCO 3.0.4) and, given Technical Specification, identify examples of otherwise specified. (300.006.A.18) (SRO Only)
21. STATE conditions permitting inoperable equipment to be placed in service (LCO 3.0.5).

(300.006.A.19) (SRO Only)

22. Use Technical Specification section 5.5.10 to IDENTIFY support and supported systems.

(300.006.A.20) (SRO Only)

23. Use TRM T12.0 to IDENTIFY Loss of Safety Function due to inoperabilities. (300.006.A.21)

(SRO Only)

24. Given support system inoperabilities, Technical Specifications, and TRM T12.0, DETERMINE the action required and the allowed time for completion when a system is inoperable solely because a Tech Spec support system is inoperable (LCO 3.0.6). (300.006.A.22) (SRO Only)
25. STATE the purpose of SRs(SR 3.0.1). (300.006.A.23)

S/RVs 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety/Relief Valves (S/RVs)

LCO 3.4.3 The safety function of 10 of 11 S/RVs shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Two or more S/RVs A.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> inoperable.

AND A.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> HATCH UNIT 1 3.4-5 Amendment No. 266

S/RVs 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety/Relief Valves (S/RVs)

LCO 3.4.3 The safety function of 10 of 11 S/RVs shall be OPERABLE.

APPLICABILITY: MODES 1,2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Two or more S/RVs A.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> inoperable.

AND A.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> HATCH UNIT 2 3.4-5 Amendment No. 210

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure <785 psig or core flow

<10% rated core flow:

THERMAL POWER shall be 24% RTP.

2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10% rated core flow:

MCPR shall be 1.07 for two recirculation loop operation or 1.09 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System (RCS) Pressure SL Reactor steam dome pressure shall be 1325 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

(continued)

HATCH UNIT 1 2.0-1 Amendment No. 264

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1 i With the reactor steam dome pressure < 785 psig or core flow

<10% rated core flow:

THERMAL POWER shall be 24% RTP.

2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10% rated core flow:

MCPR shall be 1.08 for two recirculation loop operation or 1.10 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System (RCS) Pressure SL Reactor steam dome pressure shall be 1325 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

HATCH UNIT 2 2.0-1 Amendment No. 208

HLT-07 SRO NRC EXAM

52. 295026EK3.04 001 An ATWS is in progress on Unit 2 with 31E0-EOP-01 1-2, RCA RPV Control (ATWS),

flowchart in progress.

Which ONE of the choices below completes the following statements?

JAW EOP definitions the REASON SBLC is injected before exceeding the B11T Curve (Graph 5) is to ensure that Hot Shutdown Boron Weight is injected before exceeding the At 10% RTP, the LOWEST listed Torus temperature at which the BITT Curve (Graph 5) will be EXCEEDED is Reference Provided A. Primary Containment Pressure Limit; 120°F B. Primary Containment Pressure Limit; 125°F; C. Heat Capacity Temperature Limit; 120°F D Heat Capacity Temperature Limit; 125°F; 157

HLT-07 SRO NRC EXAM

==

Description:==

The Boron Injection Initiation Temperature is defined to be the greater of either:

The highest Torus temperature at which initiation of boron injection will result in injection of the Hot Shutdown Boron Weight before Torus temperature exceeds the Heat Capacity Temperature Limit or, The Torus temperature at which a Reactor scram is required by plant Technical Specifications.

The Boron Injection Initiation Temperature defines the point BEFORE which boron injected should be started, so that the Reactor will be shutdown prior to reaching the Heat Capacity Temperature Limit (HCTL).

Plotting 125°F on Graph 5 places the plant in the Unsafe region. This results in the HCTL being exceeded before the hot shutdown boron weight being injected, if asking this question on Ul 120°F would be the correct temperature.

The A distractor is plausible if the applicant does not know or remember the bases for injecting SBLC. The second part is plausible if the applicant plots the wrong %RTP or Torus temperature on Graph 5 and would be correct if on Ui.

The B distractor is plausible if the applicant does not know or remember the bases for injecting SBLC. The second part is correct.

The C distractor is plausible since the first part is correct. The second part is plausible if the applicant plots the wrong %RTP or Torus temperature on Graph 5 and would be correct if on Ul.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

158

HLT-07 SRO NRC EXAM

References:

Unit 2 BuT Curve (Graph 5)

K/A:

295026 Suppression Pool High Water Temperature EK3. Knowledge of the reasons for the following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: (CFR: 41.5 I 45.6)

EK3.04 tSBLC injection 3.7 4.P LESSON PLAN/OBJECTIVE:

EOP-CURVES-LP-20306, EOP Curves & Limits, EO 201.071 .A.09 References used to develop this iuestion:

31E0-EOP-011-2, RCA RPV Control (ATWS) 31 EO-EOP-0 12-2, Primary Containment Control, EOP Flowchart (PC)

EOP Graph 2, Heat Capacity Temperature Limit EOP PSTG, Appendix B 31E0-OPS-00l-0, EOP General Information (Attachment 4, Graph 5, U2 B11T) 159

EOP-CURVES-LP-20306 Page 3 of 77 EOP CURVES & LIMITS Initial License (LT) ENABLING OBJECTIVES

1. DEFINE Large Oscillation Threshold (LOT), and what operator actions are REQUIRED if they occur. (201.071.A.21)
2. DEFiNE Boron Injection Initiation Temperature. (201.071.A.09, 201.074.A.06)
3. Given an ATWS and the Boron Injection Initiation Temperature Curve, DETERMINE the Boron Injection Initiation Temperature. (201.071.A.10)
4. DEFINE Hot Shutdown Boron Weight. (201.092.A.01)
5. DEFINE Cold Shutdown Boron Weight. (201.070.A.03, 201 .071.A.17, 201.085.A.19)
6. Given a list, IDENTIFY the statement that describes the failure mode that the Heat Capacity Temperature Limit protects against. (201 .074.A. 10)
7. Given plant conditions and the Heat Capacity Temperature Limit Curve, DETERMINE the current operating point on the Curve within 2.5 degrees, 5 inches, and 10 psig.

(201.075 .A.06)

8. Given a list, IDENTIFY the failure mode that the Primary Containment Pressure Limit is designed to prevent. (201.076.A.24)
9. Given a list, IDENTIFY the statement that describes the purpose of the Primary Containment Pressure Limit. (201.065 .A.30, 201 .068.A. 14, 201.075 .B. 17, 201.083 .A.20, 201.086.A.09, 201.087.A.06, 201 .088.A.05, 201.090.A.09)
10. DEFINE Maximum Pressure Suppression Primary Containment Water Level (MPSPCWL). (201 .075.B.21)
11. Given a list, IDENTIFY the two conditions that the Drywell Spray Initiation Limit is designed to prevent. (201.072.A.28, 20L073.AM7, 20L076A.15)
12. Given plant conditions and the Drywell Spray Initiation Limit Curve, DETERMINE whether or not Drywell sprays can be initiated. (201 .072.A.29, 201 .071A.08, 201 .076.A. 16)
13. DEFINE the Minimum Drywell Spray Flow (MDSF). (800.003.B.02)
14. Given a list, IDENTIFY the primary containment failure mode that the Suppression Chamber Spray Initiation Pressure is designed to prevent. (201.076.A.05)
15. Given a list, RECOGNIZE the primary containment failure modes that the Pressure Suppression Pressure Curve is designed to prevent. (201.076.A.21)
16. Given plant conditions and the PSP Curve, DETERMINE the current operating point on the Curve within 2.5 inches and 1.0 psig. (201.076.A.22)
17. Given a list, RECOGNIZE the component failure modes that the SRV Tail Pipe Level Limit is designed to prevent. (201 .075.B.05)

SNC PLANT E. I. HATCH I Pg 13 of 23 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

EOP GENERAL INFORMATION 31EO-OPS-001-0 17 ATTACHMENT 4 Affachment Page TITLE: UNIT TWO PC CHART EOP GRAPHS GENERAL INFORMATION 4 of 10 Graph 5: Boron Injection Initiation Temperature (BuT)

BORON INJECTION IMTIATION TEMPERATURE 1 70 TORUS WATER TEMP

(°F) 0 2 4 6 8 10 12 14 16 18 20 REACTOR POWER (%)

A) Is defined to be the greater of either:

a. The highest Torus temperature at which initiation of boron injection will result in injection of the Hot Shutdown Boron Weight before Torus temperature exceeds the Heat Capacity Temperature Limit or,
b. The Torus temperature at which a Reactor scram is required by plant Technical Specifications.

B) The Boron Injection Initiation Temperature curve is comprised of three segments.

a. For low Reactor power (levels less than decay heat), Torus heatup during boron injection is independent of Reactor power. Thus Segment One reflects a constant Boron Injection Initiation Temperature of 165.6cF from 0 to 2.2% power.
b. At higher powers, Torus heatup during boron injection is proportional to Reactor power; thus Segment Two reflects a decreasing Boron Injection Initiation Temperature with increasing Reactor power.
c. For high Reactor power, the highest Torus temperature at which initiation of boron injection will result in injection of the Hot Shutdown Boron Weight before Torus temperature exceeds the Heat Capacity Temperature Limit is less than the Torus temperature at which a Reactor scram is required by Plant Technical Specifications. Thus Segment Three reflects a constant Boron Injection Initiation Temperature, specifically the Torus temperature (1 10F), at which a scram is required by plant Technical Specifications.

MGR0009 Ver. 5

BORON INJECTION INITIATION TEMPERATURE 170-TORUS 150 WATER TEMP 140 160

________ rdr,Ar,d/PJr,A

(°F) 130 :__SAL 120 110 100 0 2 4 6 8 10 12 14 16 18 20 REACTOR POWER (%)

NOTE: May use SPDS Emergency Displays in place of this Graph.

PERFORM CONCURRENTLY Reset ARI and insert control rods per 31 EO-EOP-1 03-2 E

..xlc.

I Initiate SBLC per 34S0-C41-003-2 AND CONFIRM 2G31-F004 isolation

N r 1

  • rlj

r 1 0 I -a 0

-a 0

CANNOT be maintained below 100°F WAIT UNTIL dell temperature CANNOT be maintained Operate ALL available torus cooling below 150°F per 34S0-E1 1-010-2 EXCEPT RHR pumps torus water level required for adequate core cooling by CANNOT be maintained continuous operation in LPCI mode C between 146 in. and 150 in.

torus water temperature reaches BIIT curve limit (Graph 5)

Operate ALL available drywell cooIin o Drywell cooling fans per PERFORM CONCURRENTLY 34SO-T47-001 -2 o Drywell chillers per 34SO-P64-RC[AJ point A If necessary defeat isolation interlock BELOW 150 31 EO-EOP-1 00-2 WAIT UNTIL torus water temperature f

WHILE PERFORM NG THE FOLL AND reactor pressure IE primary containment water level IHJ4 termir CANNOT be maintained below and torus pressure CANNOT be to the Heat Capacity Temperature Limit maintained below Primary Containment Pressure Limit (Graph 13)

D EMERGENCY DEPRESS IS REQUIRED rr WI-411 P PFRFflRMN( THP POll

CI) i r 1

  • I I C

E rt

= = = = = rJ) fli fl c3 riD C

z

n N OT be mntaine UNTIL dell temperature CANNOT be maintained Operate ALL available torus cooling below 150°F per 34SO-E1 1-010-2 EXCEPT RHR pumps torus water level required for adequate core cooling by AIT CANNOT be maintained continuous operation in LPCI mode C between 146 in. and 150 in.

BEFORE torus water temperature reaches BIIT curve limit (Graph 5)

Operate ALL available drywell cooIin o Drywell cooling fans per F PERFORM CONC 34SO-T47-001 -2 ENTLZ> o Drywell chillers per 34SO-P64-RC[A] point A If necessary defeat isolation interlock ABOVE 150 31 EO-EOP-1 00-2 WAIT UNTIL torus water temperature WHILE PERFORMING THE FOLLJ LI?ymT1t !E primary containment water level and torus pressure CANNOT be maintained below Primary Containment Pressure Limit (Graph 13)

I I

termir to the D

EMERGENCY DEPRESS IS REQUIRED WF-iV P PPRFORMIN( ThP P01 I

HLT-07 SRO NRC EXAM

53. 295028EA1.03 001 Unit 2 was at 100% power when a Steam Line break occurred inside the Drywell (DW).

The following conditions now exist:

o Drywell Pressure 8 psig o Bulk Average Drywell Temperature 265°F Which ONE of the choices below completes the following statement?

lAW 3 1EO-EOP- 100-2, Miscellaneous Emergency Overrides, the2A DW Chiller A. is NOT allowed to be restarted, because this DW temperature is above the allowed winding temperature for restart of the DW Cooling Fan Motors B is NOT allowed to be restarted, because at this DW temperature the potential for a rupture in the DW coolers exist C. is allowed to be restarted. The operator must first place the LOCA override switch to BYPASS and then reset the 86 Lockout relay at the DW Chiller breaker D. is allowed to be restarted. The operator must first reset the 86 Lockout relay at the DW Chiller breaker and then place the LOCA override switch to BYPASS 160

HLT-07 SRO NRC EXAM

==

Description:==

31E0-EOP-lOO prohibits performing action to restart DW chillers and coolers when DW temp

>250°F if the LOCA signal is due to a LOCA (i.e. can only do it if the loss of DW cooling is due to a loss of DW cooling rather than a leak if temp is above 250°F). In this question, DW temperature is above 250°F AND the increase in DW pressure is indicative of a break in the DW not just a loss of DW Chillers, so the actions may NOT be performed.

The sequence of component manipulation is critical if allowed to restart the chiller. The LOCA override/bypass must be performed first, or the lock out relay will not reset.

The A distractor is plausible if the applicant remembers 250°F and confuses the winding temps for the DW Cooling fans with the winding temps for other DW component motors and thinks the fans/chiller can not be started above this temperature.

The C distractor is plausible if the applicant confuses the temperature limits for restarting the DW chillers during a LOCA.

The D distractor is plausible if the applicant confuses the temperature limits for restarting the DW chillers during a LOCA and then confuses the proper sequence to restart the DW chiller.

A. Incorrect See description above.

B. Correct See description above.

C. Incorrect See description above.

D. Incorrect See description above.

References:

NONE K/A:

295028 High Drywell Temperature EA1. Ability to operate andlor monitor the following as they apply to HIGH DRYWELL TEMPERATURE: (CFR: 41.7 / 45.6)

EA1.03 Drywell cooling system 3.9 3.9 LESSON PLAN/OBJECTIVE:

161

HLT-07 SRO NRC EXAM P64-PCCCW-LP-0 1304, Primary Containment Cooling and Chilled Water System, EO 013.059.A.06 References used to develop this question:

31E0-EOP-100-2, Miscellaneous Emergency Overrides Modified from HLT-4 NRC Exam Q#54 ORIGINAL QUESTION (HLT-4 NRC Exam Q#54 Unit 2 was at 100% power when a pipe break occurred inside the Drywell (DW).

The following conditions now exist:

o Drywell Pressure 8 psig o Bulk Average Drywell Temperature 245°F Which ONE of the following completes the following statement?

JAW 31E0-EOP-lOO-2, Miscellaneous Emergency Overrides, the2A DW Chiller A. is NOT allowed to be restarted because overriding the high DW pressure isolation has the potential for increased off-site release rates B. is NOT allowed to be restarted because at this DW temperature the potential for a rupture in the DW coolers exist C.V is allowed to be restarted. The operator must first place the LOCA override switch to BYPASS and then reset the 86 Lockout relay at the DW Chiller breaker D. is allowed to be restarted. The operator must first reset the 86 Lockout relay at the DW Chiller breaker and then place the LOCA override switch to BYPASS 162

P64-PCCCW-LP-01304-06 Page 4 of 69 PRIMARY CONTAINMENT COOLING AN]) CHILLED WATER SYSTEM

17. Given a list, IDENTIFY the statement that describes the purpose of the Primary Containment Chiller LOCA Trip Override. (013.059.A.0l)
18. DESCRIBE the response of the Primary Containment Chillers to a LOCA signal with a simultaneous LOSP and STATE the three (3) actions required to restart a Primary Containment Chiller. (013.059.B.06)
19. Given a list, IDENTIFY the statement that describes the effects of overriding the Primary Containment Chiller LOSP Trip on system performance. (013.059.B.02)
20. Given a list, IDENTIFY the statement that describes the purpose of the Primary Containment Chiller LOSP Trip Override. (013.059.B.01)
21. Following a Control Room Evacuation, STATE the three (3) actions required to restart a Primary Containment Chiller with a LOCAILOSP signal present. (013.068.A.01)
22. DESCRIBE the response of the Primary Containment Cooling System to each of the following events: (013.034.A.03)
a. Loss of the 600 VAC C or D essential bus
b. Loss of PSW
c. Loss of the 4160 VAC E or G emergency bus
23. STATE the required condition of all Drywell Cooling Fans prior to initiating Drywell sprays.

(013 .059.C.05)

  • 24. DETERMINE the major steps necessary to start the Primary Containment Chilled Water System. (050.002.a.01)
  • 25. DETERMINE which steps are necessary to align a Primary Containment (DIW) Chiller from a standby condition. (200.032.a.03)
  • 26. DETERMINE which actions must be performed to restore a Primary Containment (D/W) Chiller after a local chiller trip has occulTed. (200.032.a.02)

Cooling. (200.032.a.0l)

  • 28. DETERMINE which actions must be performed to restart a Primary Containment (D/W) Chiller during a LOCA or a LOCAJLOSP. (013.059.a.06)

Chiller trips during a LOCAILOSP. (201.058.c.0l)

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 5OF23 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

MISCELLANEOUS EMERGENCY OVERRIDES 31 EO-EOP-1 00-2 7.2 3.6 DRYWELL COOLER OVERRIDE

=

DO NOT PERFORM THIS SECTION :

. BULK AVERAGE DRYWELL TEMPERATURE HAS BEEN 250 °F.

CAUTION: AND

. A LOCA HAS OCCURRED (e.g., LOSS OF DRYWELL COOLING DID CAUSE THE LOCA SIGNAL).

3.6.1 Prior to re-starting the D/W cooler, the following actions must be performed:

3.6.1.1 Verify the chilled water expansion tank is within normal level, (no high/low alarms on 2H 11 -P700 panel or verify locally) 3.6.1.2 Verify D/W temperature is <250cF, in the vicinity of 2T47-BOO7A I 2T47-BOO7B. SPDS points NOOIA and NOlO can be used for 2T47-BOO7A and NOOl K and N002 can be used for 2T47-BOO7B. These points can be read directly from the SPDS Diagnostic Screen for Drywell temperature.

3.6.1.3 Place the switch for 2P64-COO8A OR 2P64-COO8B, Chilled Water pump, associated with the chiller to be started, to RUN, I

(this will re-establish flow to the DIW coolers to minimize heat loading on the chiller to prevent it from tripping once the chiller is started.)

3.6.1.4 Verify Chilled Water return temperature is <IOOCF, if necessary perform the following to maintain return temperature <100F:

3.6.1.4.1 Trip the breakers to the DIW cooling fans as necessary to open the D/W cooler valves, which will allow chilled water flow to lower the return temperature. (Reference Attachment 2) 3.6.1.4.2 Re-close the breakers to the D/W cooling fans, as necessary, to restart the DM1 cooling fans.

3.6.2 To override LOCA signal for drywell cooling fans:

3.6.2.1 Place drywell cooling fans key locked LOCA override switches to BYPASS:

DESCRIPTION PANEL Sys A Fans LOCA Override 2H1 1-P657 Sys B Fans LOCA Override 2H1 1-P654 3.6.2.2 Operate all available DM1 cooling fans, per 34S0-T47-001-2.

G16.30 MGR-0001 Ver. 4

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 7 OF 23 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

MISCELLANEOUS EMERGENCY OVERRIDES 31 EO-EOP-1 00-2 7.2 3.7 DRYWELL CHILLER OVERRIDE FiU1 DO NOT PERFORM THIS SECTION ll:

  • BULK AVERAGE DRYWELL TEMPERATURE HAS BEEN > 250° F.

CAUTION:

. A LOCA HAS OCCURRED (e.g., LOSS OF DRYWELL COOLING DID NQI CAUSE THE LOCA SIGNAL).

3.7.1 Prior to re-starting the DIW chiller, the following actions must be performed:

3.7.1.1 Verify the chilled water expansion tank is within normal level, (no high/low alarms on 2H1 1-P700 panel or verify locally) 3.7.1.2 Verify D/W temperature is <250cF, in the vicinity of 2T47-BOO7A I 2T47-BOO7B.

SPDS points NOO1A and NOlO can be used for 2T47-BOO7A and NOOl K and N002 can be used for 2T47-BOO7B. These points can be read directly from the SPDS Diagnostic Screen for Drywell temperature.

3.7.1.3 Place the switch for 2P64-COO8A OR 2P64-COO8B, Chilled Water pump, associated with the chiller to be started, to RUN, (this will re-establish flow to the D/W coolers to minimize heat loading on the chiller to prevent it from tripping once the chiller is started.)

3.7.1.4 Verify Chilled Water return temperature is <IOOCF, if necessary perform the following to maintain return temperature <100F:

3.7.1.4.1 Trip the breakers to the D/W cooling fans as necessary to open the D/W cooler valves, which will allow chilled water flow to lower the return temperature. (Reference Attachment 2) 3.7.1.4.2 Re-close the breakers to the DJW cooling fans, as necessary, to restart the DJW cooling fans.

3.7.2 Na LOCA signal is present OR has occurred, THEN perform the following:

3.7.2.1 Place 2P64-S3, LOCA Override Switch, to BYPASS, panel 2H1 1-P700.

3.7.2.2 Reset 86 lockout relay on Drywell Chiller breakers on 41 60V busses 2E AND 2G (2R22-S005 Fr. 11 AND 2R22-S007 Fr. 11).

3.7.2.3 Operate Drywell Chillers per 34S0-P64-001-2.

G16.30 MGR-0001 Ver. 4

HLT-07 SRO NRC EXAM

54. 295030EK1.02 001 Unit 2 is operating at 100% power when a Loss of Coolant Accident occurs.

The High Pressure Coolant Injection (HPCI) system is being used to control RPV water level.

o HPCI flow rate 4,000 gpm o RWL -60 and increasing at 20 per minute o Torus level 135 inches o Torus temperature 190°F o Torus Pressure 4 psig With the above conditions, SELECT the HIGHEST listed flowrate that maintains HPCI operation in the ACCEPTABLE Region of the HPCI NPSH limits?

Reference Provided A. 3,750 gpm B. 3,250 gpm 0 2,750 gpm D. 2,250 gpm 163

HLT07 SRO NRC EXAM

==

Description:==

The correct answer to this question is dependent on analyzing the correct graph. Graph selection is determined by whether suppression pool water level is At or Above 146 inches, or Below 146 inches. Common misconception among candidates on how to use the graphs, hard copy, due to having the safe region changing as torus pressure changes. Graph 1 7B must be used to obtain the correct operating point for HPCL The A distractor is plausible if the applicant refers to Graph 17A instead of Graph 17B which would place this flow in the safe region of the graph.

The B distractor is plausible if the applicant refers to Graph 17A instead of Graph 17B which would place this flow in the safe region of the graph.

The D distractor is plausible since this value is in the safe region of the graph just not the highest.

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

164

HLT-07 SRO NRC EXAM

References:

Unit 2 EOP Graph 17A, HPCI Pump NPSH Limit, (Torus Water Level At or Above 146)

Unit 2 EOP Graph 17B, HPCI Pump NPSH Limit, (Torus Water Level Below 146)

K/A:

295030 Low Suppression Pool Water Level EK1. Knowledge of the operational implications of the following concepts as they apply to LOW SUPPRESSION POOL WATER LEVEL: (CFR: 41.8 to 41.10)

EK1.O2PurnpNPSH 3.5 3.8 LESSON PLAN/OBJECTIVE:

EOP-CURVES-LP-20306, EOP Curves and Limits, EO 201.065.A.23 References used to develop this guestion:

Bank from HLT Database Q#295030EK1.02001 165

EOP-CURVES-LP-20306 Page 5 of 77 EOP CURVES & LIMITS

35. Given NPSH Graphs, and plant conditions, EVALUATE plant conditions and DETE1MINE if adequate NPSH exists for a running RCIC/ECCS pump. (201.065 .A.23, 201.068.A.25, 201.090.A.22)
36. Given NPSH Graphs and plant conditions, RECOGNIZE action that could be taken to prevent RCIC/ECCS pump from operating below its NPSH Limit. (201.065.A.27, 201.068.A.29, 201.090.A.26)
37. Given a list, IDENTIFY the statement that describes a centrifugal pumps response to operation below its Vortex Limit. (201 .065.A.24, 201 .068.A.26, 201 .090.A.23)
38. DETERMINE if RCIC/ECCS pump Vortex Limit will INCREASE OR DECREASE for a given change in RCIC/ECCS flow rate. (201.065.A.25, 201.068.A27, 201.090.A.24)
39. Given Vortex Graphs and plant conditions, EVALUATE plant conditions and DETERMINE if RCIC/ECCS pump is operating above or below its Vortex Limit.

(201.065 .A.26, 201 .068.A.28, 201 .090.A.25)

40. Given Vortex Graphs and plant conditions, RECOGNIZE actions that could be taken to prevent a RCIC/ECCS pump from operating below its Vortex Limit. (201.065.A.28, 201.068.A.30, 201.090.A.27)
41. DETERMINE the minimum number of SRVs required when Emergency Depressurizing the RPV (Minimum Number of SRVs required for Emergency Depressurization)

(201.085.A.17)

42. Given a list, IDENTIFY the statement that defines the term Minimum SRV Re-opening Pressure. (201 .085.A. 18)
43. DEFINE Maximum Normal Operating Temperature. (201.079.A.01)
44. DEFINE Maximum Normal Operating Water Level. (201.078.A.0l)
45. DEFINE Maximum Normal Operating Radiation Level. (201.077.A.01)
46. DEFINE Maximum Safe Operating Temperature. (201.079.A.09)
47. DEFINE Maximum Safe Operating Water Level. (201.078.A.09)
48. DEFINE Maximum Safe Operating Radiation Level. (201.077.A.08)

Objectives marked by a RED (*) are required during RO-305 and SR-305 of the Initial License program.

(%)

fl GRAPH 17A UNIT 2 HPCI PUMP NPSH LIMIT (Torus Water Level At or Above 146)

Torus Temp

(°F) 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 HPCI FLOW (gpm)

NOTE: May use SPDS Emergency Displays in place of this Graph.

Suppression Chamber Pressure.

    • Safe operating region is below the applicable pressure line.

GRAPH 17B UNIT 2 HPCI PUMP NPSH LIMIT (Torus Water Level Below 146)

UNSAFE 220- or Above 10 psigi Torus Temp 210-(°F) 200-190-Below5psig*

180-

    • SAFE 170-160-(3 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 HPCI FLOW (gpm)

NOTE: May use SPDS Emergency Displays in place of this Graph.

Suppression Chamber Pressure.

Safe operating region is below the applicable pressure line.

HLT-07 SRO NRC EXAM

55. 29503 1EA1.02 001 Unit 2 was operating at 100% power, when an event occurred resulting in the following:

o Reactor water level: -45 (lowest level reached) and increasing 45 inches/minute o Reactor pressure: 900 psig and slowly decreasing o Drywell pressure: 0.8 psig and steady Which ONE of the choices below completes the following statement?

Based on these conditions, with NO additional operator actions, one (1) minute later will have indication of flow to the reactor vessel.

A. ONLY HPCI and RCIC B. ONLY the Reactor Feedpumps C HPCI, RCIC and the Reactor Feedpumps D. ONLY RCIC and the Reactor Feedpumps 166

HLT-07 SRO NRC EXAM

==

Description:==

HPCI and RCIC automatically start on -35 RWL. Since RWL did not drop below -101 which would have closed the MSIVs and stopped the Reactor Feedpumps, the Reactor Feedpumps are still injecting.

The A distractor is plausible if the applicant confuses the RWL setpoints and the actions that occur and thinks the MISVs have closed and the Reactor Feedpumps have stopped injecting.

The B distractor is plausible if the applicant confuses the RWL setpoints and the actions that occur and thinks HPCI & RCIC have not reached their auto start signal. Also plausible if the applicant thinks -60 is the auto start setpoint which is the setpoint for the Recirc pumps tripping.

The D distractor is plausible if the applicant confuses the RWL setpoints and the actions that occur and thinks RCIC did not receive an auto signal. Also plausible since HPCI & RCIC have different auto start signals.

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

References:

NONE K/A:

295031 Reactor Low Water Level EA1. Ability to operate and/or monitor the following as they apply to REACTOR LOW WATER LEVEL: (CFR: 41.7 I 45.6)

EA1.02 High pressure (feedwater) coolant injection: Plant-Specific 45* 4.5 LESSON PLAN/OBJECTIVE:

E41-HPCI-LP-00501, High Pressure Coolant Injection System, EO 005.005.A.02 E51-RCIC-LP-03901, Reactor Core Isolation Cooling (RCIC), EO 039.004.A.01 167

HLT-07 SRO NRC EXAM T23-PC-LP-0 1301, Primary Containment, EQ 013.045 .A.05 References used to develop this question:

34AB-C7 1-001-2, Scram Procedure Modified from HLT-4 NRC Exam Q#56 ORIGINAL QUESTION (HLT-4 NRC Exam Q#56)

Unit 2 was operating at 100% power, when a Loss of Coolant Accident (LOCA) occurred.

o Drywell pressure: 3.7 psig o Reactor pressure has decreased to 285 psig Which ONE of the following predicts the expected control panel flow indication JAW 31EO-EOP-010-2, RC Flowchart?

Based on these conditions, with no additional operator actions, will have indication of flow to the reactor vessel.

A. NEiTHER Core Spray nor RHR B. BOTH Core Spray and RHR C. ONLY Core Spray D. ONLY RHR 168

T23-PC-LP-01301 Page 8 of 160 PRIMARY CONTAINMENT

28. Given a list of statements, IDENTIFY the statement which best describes the plant conditions which will generate the following PCIS isolation signals: (013.045.A.05, 0l3.046.A.05, 0l3.047.A.05)
a. Group 1 isolation
b. Group 2 isolation
c. Group 3 isolation
d. Group 4 isolation
e. Group 5 isolation
f. Group 6 isolation
29. Given a list of statements, IDENTIFY the statement which best describes the steps required to reset the following PCIS isolation signals: (013.045 .A.06, 013 .046.A.06, 013 .047.A.06, 013.067.A.0l, 040.006.A.Ol)
a. Group 1 isolation
b. Group 2 isolation
c. Group 3 isolation
d. Group 4 isolation
e. Group 5 isolation
f. Group 6 isolation
30. Given a list of statements, IDENTIFY the statement which best describes the primary containment characteristics during a DBA LOCA, including the process by which energy is absorbed and dissipated by the primary containment. (200.004.A.03)
31. Given a list of statements, IDENTIFY the statement which best describes the consequences of having the torus to drywell vacuum breakers fail either open or closed during a DBA LOCA.

(200.004.A.02)

?ORMAL OPERATIONS

(0l3.006.A.0l)

(0l3.007.A.0l)

I E5l-RCICLP-O39Ol-O6 Page 3 of 95 L REACTOR CORE ISOLATION COOLING (RCIC)

Initial License (LT) ENABLING OBJECTIVES

1. State the purpose of the RCIC system. (lOO.036.C.02)
2. STATE the RCIC initiation signal and setpoint. (039.004.A.0l, l00.036.C.OI, lO0.036.C.12)
3. LOCATE the following system interfaces on a RCIC P&ID or RCIC simplified drawing:

(039.001.A.04, l00.036.C.03)

a. Main Steam
b. Feed Water
c. Suppression Pool
d. Radwaste
e. Main Condenser
4. IDENTIFY the function for each of the three modes that RCIC may be operated in.

(1 00.036.C.04)

5. Given a highlighted diagram showing a lineup of the RCIC System, DETERMINE which mode it is aligned to function in. (lOO.036.C.05)
6. Given a P&ID or simplified drawing of the RCIC System, LABEL the following RCIC valves:

(039.001 .A.06)

a. Inboard Steam Isolation Valve (F007)
b. Outboard Steam Isolation Valve (F008)
c. Steam Supply Valve (F045)
d. Trip and Throttle Valve
e. Governor Valve
f. Minimum Flow Valve (FO 19)
g. Outboard Injection Valve (FO 12)
h. Inboard Injection Valve (F0l3)
i. Outboard Test Return to CST (F022)
j. Inboard Test Return to CST (E41-F0l 1)
k. CST Suction (FOlO)
1. Inboard Suppression Pool Suction (F031)
m. Outboard Suppression Pool Suction (F029)
7. Given a RCIC P&ID or a simplified drawing, TRACE the flowpath of RCIC from the CST suction to the discharge into the Reactor Vessel via Feedwater line. (039.002.A.03, 1 00.036.C.06)
8. Given a RCIC P&ID or a simplified drawing, TRACE the flowpath of RCIC in the pressure control mode. (039.002.A.02, 100.036.C.06)

E41-HPCI-LP-00501 Page 5 of 92 HIGH PRESSURE COOLANT INJECTION SYSTEM

19. Given HPCI is injecting into the RPV, STATE the system response to a loss of the following power supplies: (005.005.a.07)
a. 25OVDC MCC 2B (R24-S022)
b. I25VDC CAB 2B (R25-S002)

System is in STANDBY per 34S0-E41-00l-l/2, High Pressure Coolant Injection (HPCI)

System. (05.001.a.07)

  • 21. Given plant conditions, DETERMINE whether HPCI should have auto started per 34S0-E41-0014/2, High Pressure Coolant Ihjection System. (005.005.a.02)
  • 25. DETERMINE the actions necessary to startup HPCI to restore and maintain RWL in a specified band per 3450-E41-00I-l/2, High Pressure Coolant Injection System. (005.024.a.02)

(005.004.a.03)

  • 29. Given Plant conditions, EVALUATE the conditions and DETERMINE if the HPCI turbine should have tripped. (005.005.a.06)
  • 30. Given Plant conditions, START HPCI locally per 31RS-E41-00l-1/2, HPCI Operations From Outside the Control Room. (005.02 1.a.01)
  • 31. Given plant conditions, DETERMINE whether the HPCI system trip signals, except overspeed, should be disabled per 31RS-E41-001-1/2, HPCI Operation from outside the Control Room.

(005.02 1 .a.03)

  • 32. Given plant conditions, SHUTDOWN HPCI locally per 31RS-E41-001-1/2, HPCI Operation from Outside the Control Room. (005.022.a.01)

SOUTHERN NUCLEAR PLANT El. HATCH PAGE 11 OF 31 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

SCRAM PROCEDURE 34AB-C71-001-2 11.2 ATTACHMENT 1 ATTACHMENT PAGE:

TITLE: PRIMARY CONTAINMENT ISOLATION CONFIRMATION 1 OF 9 1.0 GROUP I ISOLATION NOTE Some of the following conditions can be indicative of a pipe break in the Reactor Building.

Refer To 34AB-T22-001-2 as applicable.

1.1 Any of the following conditions cause isolation:

. Reactor Water Level Low Low Low (Level I )(-1 01.0)

. Main Steam Line Radiation High (3 Times Normal)

. Main Steam Line Flow High (169 psid)

. Main Steam Tunnel Temp. High (190 degrees F)

. Turbine Building Temp at Steam Lines (196 degrees F)

. Main Steam Line Pressure Low (855 psig in RUN)

. Condenser Vacuum Low (10 Hg) 1.2 Confirm the following valves have CLOSED:

NOTE Position indication can also be found on SPDS diagnostic AND on panel 2H11-P601 vertical display except as noted.

2H1 1 -P602 Verified Closed Reopened 2B21-F022A *M5IV 2B21-F022B *MSIV LI 2B21-F022C *MSIV LI LI 2B21-F022D *MSIV LI 2B21-F016 *MSL Drain Valve LI LI 2B31-F019 RxWater Sample Valve LI LI 2H1 1 -P601 2B21-F028A *MSIV LI LI 2B21-F028B *MSIV LI LI 2B21-F028C *M5IV LI LI 2B21-F028D *M5IV LI LI 2B21-F019 *MSL Drain Valve LI LI 2B31F020 Rx Water Sample Valve LI LI Will NOT close on Main Steam Line Radiation High (3 times normal).

MGR-0009 Rev. 5.0

SOUTHERN NUCLEAR PLANT El. HATCH PAGE 16 OF 31 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

SCRAM PROCEDURE 34AB-C71-O01-2 11.2 ATTACHMENT 1 ATTACHMENT PAGE:

TITLE: PRIMARY CONTAINMENT ISOLATION CONFIRMATION 6 OF 9 3.0 GROUP 3 ISOLATION NOTE- The following conditions can be indicative of a HPCI pipe break in the Reactor Building.

Refer To 34AB-T22-0O1-2 as applicable.

3.1 Any of the following conditions cause isolation:

. HPCI Turbine Exhaust Diaphragm Press High (10 psig)

. HPCI Steam Line Flow High (202 in. H 0 dp or -1 00 in. H 2 0 dp) 2

. HPCI Steam Line Pressure Low (134 psig)

. HPCI Equipment Room Temp. High (165 degrees F)

. Supp. Chamb. Area Air Temp. High (165 degrees F)

. Supp. Chamb. Area Duff. Air Temp High (36 degrees F)

. Emergency Area Cooler Temp. High (165 degrees F) 3.2 Confirm the following valves have closed:

NOTE- Position indication can also be found on SPDS diagnostic NP. on panel 2H11-P601 vertical display except as noted.

2H11-P601 2E41-F003 Outbd Steam Isol VIv LI 2E41-F002 Inbd Steam Isol Vlv LI (a)(b) 2E41-F041 Torus Outbd SuctVlv LI (a)(b) 2E41-F042 Torus lnbd Suct Vlv LI (a) No indication on SPDS diagnostic (b) No indication on 2H11-P601 vertical display 3.3 Place the following control switches to CLOSE:

2E41-F002 Inbd Steam Isol Vlv LI 2E41-.F003 Outbd Steam Isol Vlv LI MGR-0009 Rev. 5.0

SOUTHERN NUCLEAR PLANT El. HATCH PAGE 17 OF 31 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

SCRAM PROCEDURE 34AB-C7 1-001-2 11.2 ATTACHMENT I ATTACHMENT PAGE:

TITLE: PRIMARY CONTAINMENT ISOLATION CONFIRMATION 7 OF 9 4O GROUP 4 ISOLATION NOTE The following conditions can be indicative of a RCIC pipe break in the Reactor Building.

Refer To 34AB-T22-001-2 as applicable.

4.1 Any of the following conditions cause isolation:

. Supp. Chamber Area Air Temp. High (165 degrees F)

. Supp. Chamb. Area Duff. Air Temp. High (36 degrees F)

. RCIC Turbine Exhaust Diaphragm Press. High (10 psig)

. RCIC Steam Line Flow High (143 in. H O dp or -1 00 in. H 2 O dp) 2

. RCIC Steam Line Pressure Low (95 psig)

. RCIC Equipment Room Temp. High (165 degrees F) 4.2 Confirm the following valves have CLOSED:

NOTE* Position indication can also be found on SPDS diagnostic AND on panel 2H1 1-P601 vertical display except as noted.

2H1 1 -P602 2E51-F008 Steam Supply Line Isol Vlv 2E51-F007 Steam Supply Isol Vlv 4.3 Place the following control switches to CLOSE:

2E51-F007 Steam Supply Isol Vlv LI 2E51-F008 Steam Supply Line Isol Vlv MGR-0009 Rev. 5.0

HLT-07 SRO NRC EXAM

56. 295032EA1.05 001 Following a transient on Unit 2, current plant conditions are:

o Reactor water level is -20 inches, rising 5 inches/minute o Reactor pressure is 750 psig, slowly lowering Subsequently the following conditions occur:

o RCIC Room Area Cooler Temperature is 160°F o HPCI Room Area Cooler Temperature is 175°F o RCIC is injecting at 400 gpm o HPCI is injecting at 600 gpm Which ONE of the choices below completes the following statements?

With the above conditions, REQUIRED to be MANUALLY isolated.

A ONLY the HPCI System is B. ONLY the RCIC System is C. BOTH HPCI and RCIC Systems are D. NEITHER HPCI nor RCIC Systems are 169

HLT-07 SRO NRC EXAM

==

Description:==

The HPCI System will isolate on the following:

High Temperature Leak Detection HPCI area cooler outlet 165°F Suppression chamber area ambient temperature 165°F 14 mm. TD (Ui 13.5 mi TD)

Suppression chamber area differential temperature 36°F 14 mm. TD (UI 15 mm. TD)

The RCIC System will isolate on the following High Temperature Leak Detection:

RCIC area cooler outlet 165°F Suppression chamber area ambient temperature 165°F 29 mm. TD (Ui 28.5 mm. TD)

Suppression chamber area differential temperature 36°F 29 mm. TD (Ui 28.5 mi TD)

The B distractor is plausible if the applicant confuses the isolation values for HPCI & RCIC and thinks only RCIC has isolated.

The C distractor is plausible if the applicant confuses the isolation values for HPCI & RCIC and thinks both systems have isolated.

The D distractor is plausible if the applicant confuses the isolation values for HPCI & RCIC and thinks neither systems have isolated.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

170

HLT-07 SRO NRC EXAM

References:

NONE K/A:

295032 High Secondary Containment Area Temperature EA1. Ability to operate andlor monitor the following as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: (CFR: 41.7 / 45.6)

EA1 .05 Affected systems so as to isolate damaged portions . . . . 3.7 3.9 LESSON PLAN/OBJECTIVE:

E41-HPCI-LP-00501, High Pressure Coolant Injection System, EU 005.012.A.04 EOP-SCRR-LP-20325, Secondary Containment I Radioactivity Release Control, EQ 201.081.B.01 References used to develop this iuestion:

31E0-EOP-014-2, SC RR, EOP flowchart 3450-E41-001-2, High Pressure Coolant Injection (HPCI) System 171

EOP-SCRR-LP-20325 Page 2 of 45 SECOIDARY CONTAINMENT I RADIOACTIVITY RELEASE CONTROL Initial License (LT) ENABLING OBJECTIVES

1. Given plant conditions, RECOGNIZE an FOP entry condition(s) and ENTER the appropriate EOP flow chart. (201.093.A.01)
2. Given a list, IDENTIFY the statement that describes the purpose of confirming reactor building HVAC isolation and SBGT initiation when reactor building exhaust exceeds 9.5 mRlhr. (201.080.A.02)
3. Given a list, IDENTIFY the statement that describes the purpose of confirming refueling floor and reactor building HVAC isolations and SBGT system initiations when refueling floor exhaust radiation levels exceed their Maximum Normal Operating Levels.

(201 .080.B.05)

4. Given a list, IDENTIFY the statement that describes the purpose of using HVAC systems as the first method of attempting to control secondary containment area temperatures.

(201 .079.A.05, 201.081 .A.01, 201.081 .B.01)

5. Given plant conditions, including a refueling floor HVAC isolation, IDENTIFY the signal that caused the isolation to occur. (201.08 l.A.03)
6. Given plant conditions, including a secondary containment HVAC isolation, IDENTIFY the signal that caused the isolation to occur. (201.08 l.B.03)
7. Given 31E0-EOP-014-2 SC Secondary Containment Control, DETERMINE the following values: (20l.079.A.l6 & 201.080.B.02)
a. Refueling Floor Exhaust Maximum Normal Operating Radiation Levels.
b. Reactor Building Exhaust Maximum Normal Operating Radiation Level.
8. Given 3 1EO-EOP-014-2, SC Secondary Containment Control, and process radiation monitor readings, DETERMINE if the following HVAC exhaust radiation levels are above their Maximum Normal Operating Values. (20l.079.A.18 & 201.080.B.04)
a. Refueling Floor exhaust.
b. Reactor Building exhaust.
9. Given an area and 3lEO-EOP-014-2 SC Secondary Containment Control, DETERMINE the following values for that area: (201 .079.A.02)
a. Maximum Normal Operating Temperature.
b. Maximum Safe Operating Temperature.
10. Given area temperatures and 31EO-EOP-0l4-2, SC Secondary Containment Control, DETERMINE if area temperatures are above their Maximum Normal Operating Values.

(201 .079.A.04)

E41-HPCI-LP-00501 Page 6 of 92 HIGH PRESSURE COOLANT INJECTION SYSTEM

  • 33 Given plant conditions involving the HPCI System, DETERMINE if a Technical Specification Limiting Condition for Operation has been exceeded. (implicit in this objective is a determination ofAPPLICABILITY and associated NOTES) (3 00.006.a. 13)
  • 34 Given plant conditions involving the HPCI System, DETERMINE the Required Actions for Completion Times 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in accordance with Technical Specifications for any combination of INOPERABLE systems, structures or components. (300.009.a.05)
  • 35 Given plant conditions involving the HPCI System, DETERMINE the Required Action(s) and Completion Time(s) in accordance with the Technical Requirements Manual (TRM) for any combination of INOPERABLE systems, structures or components and the bases for the action(s).

(SRO Only) (300.01 1.a.06)

  • 36. Given plant conditions, DETERMINE the actions necessary to be taken upon inadvertent initiation of an ECCS system per 34AB-El 0-001-1/2, Inadvertent Initiation of ECCS SystemJRCIC. (200.027.c.02)
  • 37 DETERMINE the conditions that must be met prior to shutting down the HPCI system after an auto start per 3IGO-OPS-021-0 Manipulation of Controls and Equipment. (200.027.c.01)
  • 38. Given the High Pressure Coolant Injection (HPCI) System is running and injecting into the RPV and either of the two following conditions, verify that an automatic change of the suction sources for HPCI occurs: (005.011.a.01)
a. CST low level, and
b. Suppression Pool high level.
  • 39 DETERMINE the sequence of steps necessary to swap HPCI suction from the Torus to the CST.

(005.01 1.a.02)

  • 40. DETERMINE the actions that are required to be taken for HPCI isolation per 34AR-601-l 15-2 (123-1) or 34AR-601-121-2 (129-1), HPCI Isolation Trip Logic A/B Initiated. (005.012.a.03)
  • 41. Given plant conditions, DETERMINE whether HPCI should be isolated per 34AR-601-l 15-2 (123-1) or 34AR-601-121-2 (129-fl, HPCI Isolation Trip Logic AJB Initiated. (005.012.a.04)

System. (005.024.a.03)

Objectives marked by a RED (*) are required during RO-305 and SR-305 of the Initial License program.

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 5 OF 73 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

HIGH PRESSURE COOLANT INJECTION (HPCI) SYSTEM 345O-E41-001-2 23.3 5.2 LIMITATIONS NOTE* All turbine trips, except the mechanical overspeed, can be reset from the Control Room.

The mechanical overspeed will reset automatically at < 3000 rpm.

5.2.1 The following signals will cause a HPCI turbine trip:

5.2.1.1 Manual push-button on panel 2H11-P601 5.2.1.2 High Turbine Exhaust Pressure ( 140 psig) 5.2.1.3 Low HPCI Pump Suction Pressure (10 Inches Hg VAC) 5.2.1.4 Mechanical Overspeed [5000 rpm (125%)j 5.2i.5 HPCI Logic A OR Logic B Isolation Signal 5.2.1.6 High Reactor Vessel Water Level [ 51.7 Inches (Level 8)]

5.2.1.7 Local Manual Trip To manually trip turbine, lift knurled knob. Release knob and turbine will auto reset.

NOTE:

Posted @ Turbine 5.2.2 The HPCI System will automatically isolate upon receipt of any of the following signals:

5.2.2.1 HPCI Steam Line High Differential Pressure (Steam line break):

Instrument Setpoint, 202 H 0 increasing OR -100 H 2 0 decreasing.

2 (Technical Specification Limit <303%).

5.2.2.2 HPCI Steam Supply Low Pressure: Instrument Setpoint 134 psig (2E41-N058A, 2E41-N058B, 2E41-N058C, & 2E41-N058D). (Technical Specification Limit> 100 psig).

5.2.2.3 Turbine Exhaust Diaphragm High Pressure: Instrument Setpoint 10 psig.

(Technical Specification Limit < 20 psig).

5.2.2.4 HPCI EmergencyArea Cooler Temperature High of 165F (Technical Specification Limit 169F).

5.2.2.5 HPCI Pipe Penetration Room Temperature High of 165F (Technical Specification Limit 1 69F).

MGR-0001 Ver. 4

SECONDARY CONTAINMENT Table 4 OPERATING TEMPERATURES Max Normal Max Safe SECONDARY CONTAINMENT TEMP Operating Operating G Value Value on 2H11-P614, 2G31-R604

°F 158 ELEVATION AREA (RWCU) 101 B pump room (2G31-NO16B) 150 215 102 B pump room DIFF (2G31-N023B12G31-N022B) 67 99 103 Hx Room (2G31-NOI6D) 150 215 104 Hx Room DIFF (2G31-N023D12G31-N022D) 67 99 185 ELEVATION AREA (RWCU) 105 Valve Nest (2G31-NOI6F) 150 215 106 Valve Nest DIFF (2G31-N023F/2G31-N022F) 67 99 SOUTHEAST DIAGONAL AREA 107 RHR/CS B (2E11-NOO9B) 150 190 108 RHRICS B DIFF (2E11-NO30BI2E11-N029B) 40 74 HPCI ROOM AREA 109 Emer Area CIr (2E41-NO3OB) 167.5 245 RCIC ROOM AREA 111 Emer Area Clr(2E51-N023B) 167.5 310 H

TORUS ROOM AREA

HLT-07 SRO NRC EXAM

57. 295035EA2.02 001 Unit 2 is operating at 100% power with the following:

o 2T41-COO1A, Rx. Bldg Supply Fan, in service with 5,300 cfm flow o 2T41-COO7A, Rx Bldg Vent Exhaust Fan, in service with 6,500 cfm flow o 2T41-C007B, Rx Bldg Vent Exhaust Fan, is tagged out Subsequently, 2T41-COO7A, Rx Bldg Vent Exhaust Fan, Trips.

With NO operator action, which ONE of the choices below completes the following statements?

After 2T41-COO7A trips, the Rx. Bldg SUPPLY Fan flow will After 2T41-COO7A trips, the Offsite release rate from the Rx. Bldg Stack will A. remain at 5,300 cfm; decrease to zero (0) cpm because NO Exhaust fans are discharging to the Rx. Bldg. Stack B. remain at 5,300 cfm; decrease but remain greater than zero (0) cpm since some Exhaust fans are still discharging to the Rx. Bldg. Stack C. decrease to zero (0) cfm; decrease to zero (0) cpm because NO Exhaust fans are discharging to the Rx. Bldg. Stack D decrease to zero (0) cfm; decrease but remain greater than zero (0) cpm since some Exhaust fans are still discharging to the Rx. Bldg. Stack 172

HLT-07 SRO NRC EXAM

==

Description:==

The trip of the Rx Bldg Vent fan leaves no running exhaust fan and causes the trip of the supply fans. This causes building negative pressure to decrease (towards 0 DP) and, with no vent exhaust out the Rx Bldg stack, the release rate will lower.

JAW 34S0-T41-005-2, Reactor Building Ventilation System, step 5.3 states 2T41-COO1A OR 2T41-COO1B, Rx Bldg Supply Fan CANNOT be started UNLESS a 2T41-COO7A OR 2T41-COO7B, Rx Bldg Vent Exhaust Fan, is running.

The A distractor is plausible if the applicant confuses or does not know the interlock between the Exhaust Fans and the Supply Fans and thinks the Rx. Bldg. Supply fan flow will remain at 5,300 cfm. The second part is plausible if the applicant thinks that with no Rx. Bldg. Exhaust fans discharging into the Rx. Bldg. Stack that the cpm decreases to zero (0) and does not realize that other fans are discharging into the Rx. Bldg. Stack.

The Bdistractor is plausible if the applicant confuses or does not know the interlock between the Exhaust Fans and the Supply Fans and thinks the Rx. Bldg. Supply fan flow will remain at 5,300 cfm. The second part is correct.

The C distractor is plausible since the first part is correct. The second part is plausible if the applicant thinks that with no Rx. Bldg. Exhaust fans discharging into the Rx. Bldg. Stack that the cpm decreases to zero (0) and does not realize that other fans are discharging into the Rx. Bldg.

Stack.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

173

HLT-07 SRO NRC EXAM

References:

NO?.E K/A:

295035 Secondary Containment High Differential Pressure EA2. Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE: (CFR: 41.8 to 41.10)

EA2.02 tOff-site release rate: Plant-Specific 2.8* 4.1 LESSON PLAN/OBJECTIVE:

T41-SC HVAC-LP-01303, Secondary Containment HVAC Systems EU 037.01 1.A.10 References used to develop this question:

34S0-T41-005-2, Reactor Building Ventilation System 174

T41-SC HVAC-LP-01303 Page 7 of 86 Secondary Containment HVAC Systems

6. Given a list of plant locations, IDENTIFY the locations which best describes the in-plant locations of the following Secondary Containment HVAC system components: (013.040.A.Ol)
a. Reactor Zone Supply Ventilation System
b. Recirculation Cooling Unit (2T41-B017)
c. Reactor Zone Exhaust Ventilation System
d. AOV 1T41-F027 (Accessible to Inaccessible Area Bypass)
e. Refueling Zone Supply Ventilation System
f. Refueling Zone Exhaust Ventilation System
g. Equipment Area Cooling System
7. Given a list of systems, SELECT the system which best describes which systems provide cooling water to the Equipment Area Coolers. (037.004.A.0l, 037.005.A.03)
8. Given a list of statements, IDENTIFY the statement which best describes how a slight negative pressure is maintained in the following areas during normal operating conditions.

(037.011 .A.09, 037.012.A.06, 037.022.A.09, 037.023.A.06)

a. Unit 2 Reactor Zone
b. Unit I Reactor Zone
c. Unit 2 Refueling Zone
d. Unit 1 Refueling Zone
9. Given a list of statements, IDENTIFY the statement which best describes the trips and interlocks associated with the following ventilation zones: (037.Oll.A.10, 037.012.A.07, 037.022.A.10, 037.023.A.07)
a. Unit 2 Reactor Zone
b. Unit 1 Reactor Zone
c. Unit 2 Refueling Zone
d. Unit 1 Refueling Zone
10. Given a list of plant conditions, SELECT the plant conditions which will generate isolation signals for the following ventilation zones: (037.01 1.A.1 1, 037.012.A.08, 037.022.A.1 1, 037.023.A.08)
a. Unit 2 Reactor Zone
b. Unit 1 Reactor Zone
c. Unit 2 Refueling Zone
d. Unit 1 Refueling Zone

SOUTHERN NUCLEAR PAGE PLANT F. I. HATCH 3 OF 26 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REACTOR BUILDING VENTILATION SYSTEM 34S0-T41-0O5-2 8.9 5.0 PRECAUTIONS I LIMITATIONS 5.1 2T41-COO1A AND 2T41-COOIB, Rx Bldg Supply Fans, are interlocked to prevent simultaneous operation.

5.2 2T41-COO7A and 2T41-COO7B, Rx Bldg Vent Exhaust Fans, are interlocked to prevent simultaneous operation.

5.3 2T41-COO1A OR 2T41-COO1 B, Rx Bldg Supply Fan CANNOT be started UNLESS a 2T41-COO7AOR2T41-COO7B, Rx Bldg Vent Exhaust Fan, is running.

5.4 Removing the ventilation from service will cause the Reactor Bldg temperatures to INCREASE.

Therefore, ensure contingencies are in place WHEN removing Rx Bldg ventilation from service.

Increased monitoring of building/area temperatures will be required to ensure that acceptable proximity to temperature isolation setpoints is maintained.

5.5 In the event that the Rx Bldg Vent Flow Recorder 2T41-R618 is unavailable, the supply AND exhaust fan flows may be determined by requesting l&C to obtain the dp across the respective flow transmitter in inches of water jQ use the following equation to obtain actual flow.

FAN FLOW TRANSMITTER EQUATION 2T41 -COO 1 A / 2T41 -COO 1 B, RX BLDG SUPPLY FANS 2T41 -N026 ((flio,ooo SCFM O.480J 2T41-COO7A / 2T41-COO7B, dp RX BLDG EXHAUST FANS 2T41-N007 x1O,OOO SCFM O.689J MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANTE.I.HATCH 8OF26 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REACTOR BUILDING VENTILATION SYSTEM 34S0-T41-005-2 8.9 7.2 SYSTEM SHUTDOWN FONTOU9 Area coolers do NOT have to be shutdown during containment testing. Individual coolers may NOTE*

be shutdown for maintenance OR environmental control.

PRIOR TO SHUTDOWN, REFER TO STEP 5.3 OF PRECAUTIONS I LIMITATIONS CAUTION SECTION OF THIS PROCEDURE.

7.2.1 To SHUTDOWN desired area coolers, place the following switches in STOP:

7.2.1.1 2T41-B007, Rx Bldg Cooler Working Floor EL 130 North LI 7.2.1.2 2T41B008, Rx Bldg Cooler Working Floor El 130 South LI 7.2.1.3 2T41-B026, Rx Bldg Primary Cooler Mn Steam Pipe Chase LI 7.2.1.4 2T41-B010, Rx Bldg Cooler Equip Room EL 164 LI 7.2.1.5 2T41-B011, Rx Bldg Cooler Operating Floor EL 185 LI 7.2.1.6 2T41-B014, Rx Bldg Cooler RWCU PmpIHx EL 158 LI 7.2.1.7 2T41-B015, Rx Bldg Cooler Operating Floor EL 185 LI 7.2.1.8 2T41-B016, Rx Bldg Cooler Working Floor EL 203 LI 72.2 Place in OFF 2T41-B017, Fan 2 or Fan 1, Rx Bldg Recirc. LI 7.2.3 Place in OFF 2T41-B017, Fan I or Fan 2, Rx Bldg Recirc. LI 7.2.4 Place in OFF 2T41-B009, Rx Bldg Secondary Cooler Mn. Steam Pipe Chase. LI MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 9 OF 26 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REACTOR BUILDING VENTILATION SYSTEM 34SO-T41-OO52 8.9

[NOTE: The fan in STBY must be placed in OFF before attempting to SHUT DOWN the operating fan.

1 7.2.5 To shutdown Supply and Exhaust Fans, perform the following:

7.2.5.1 Place in OFF the STANDBY:

  • 2T41-COO1A, Rx Bldg Supply Fan LI 7.2.5.2 Place in OFF the OPERATING:
  • 2T41-COO7B, Rx Bldg Vent Exhaust Fan LI 7.2.5.4 Place in OFF:

HLT-07 SRO NRC EXAM

58. 295037EK2.01 001 Unit 2 was at 84% RTP when it is discovered that the Alternate Rod Insertion (ART) System will NOT perform its intended function.

Subsequently, an auto-scram signal was received.

The OATC inserted a manual scram and performed RC-1.

The following conditions currently exist:

o Reactor power is 35% RTP o The 8 white RPS Scram Group lights on 2H1 l-P603 are illuminated o Full core display blue lights are ALL extinguished With the above conditions and lAW 31E0-EOP-103-2, EOP Control Rod Insertion Methods, INDEPENDENTLY EVALUATE each of the following control rod insertion methods to determine if they will be an effective method for control rod insertion?

Placing the RPS Test Trip Logic Switches to TRIP be effective.

Venting the Scram Air Header be effective.

A will; will B. will:

will NOT C. will NOT; will D. will NOT; will NOT 175

HLT-07 SRO NRC EXAM

==

Description:==

JAW 31E0-EOP-103-2, EOP Control Rod Insertion Methods, the required action would be to de-energize the scram solenoids since the white RPS Scram Group lights are still illuminated.

Once this action is complete and if the blue Full core display lights are still extinguished, venting the Scram Air Header will result in the blue lights to illuminate as each control rod scrams in.

Since both the white and blue lights are illuminated both methods would be an effective method to insert control rods. Plausibilty of the distractors depends upon knowledge of how the systems work and comprehending the function of the aforementioned lights.

The B distractor is plausible since the first part is correct. The second part is plausible if the applicant remembers that this is a step in EOP- 103 but confuses when it is performed and believes it would not be effective in inserting rods.

The C distractor is plausible if the applicant remembers that this is a step in EOP-l03 but confuses when this is performed and believes it would not be effective in inserting rods. The second part is correct.

The D distractor is plausible if the applicant remembers that this is a step in EOP-103 but confuses when this is performed and believes it would not be effective in inserting rods. The second part is plausible if the applicant remembers that this is a step in EOP-103 but confuses when it is performed and believes it would not be effective in inserting rods.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

176

HLT-07 SRO NRC EXAM

References:

NONE K/A:

295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown EK2. Knowledge of the interrelations between SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN and the following:

(CFR: 41.7/45.8)

EK2.01 RPS 4.2* 43*

LESSON PLAN/OBJECTIVE:

C7 1 -RPS-LP-0 1001, EO 01 0.022.A.02 EOP103-LP-20314. EOP-103: EOP Control Rod Insertion Methods, EO 010.022.A.02 References used to develop this question:

31E0-EOP-103-2, EOP Control Rod Insertion Methods 177

C71-RPS-LP-OlOO1 Page 4 of 112 REACTOR PROTECTION SYSTEM Initial License (LT) ENABLING OBJECTIVES

1. Given a list, IDENTIFY the statement that describes the control rod drive system response if the scram air header was vented. (0l0.015.a.02)
2. Given a list of definitions, SELECT the item which best describes the term one-out-of-two taken-twice. (010.011 .a.04) 3 Given a list, SELECT the statement(s) which best describe the conditions necessary to Open or Close the following valves: (0l0.013.a.05)
a. Scram Discharge Volume Vent and Drain and Backup Vent and Drain Valves.
b. Scram Air Header Pilot and Backup Scram Valves.
4. Given a list of power supplies, SELECT the power supply for the Back-Up Scram Air Header Valves. (0l0.00l.a.02)
5. Given a simplified drawing or P&ID of the Scram Air Header/Discharge Volume, IDENTIFY the valve positions and flowpaths for non-scram and scram conditions. (010.0 13.a.03)
6. Given a list, SELECT the statement which best describes the reason the Backup SDV Vent and Drain Valves are shut per 31RS-OPS-001-I/2 Shutdown Outside Control Room.

(0 10.013 .a.04)

7. From a list of possible loads, SELECT the loads supplied by each RPS Bus. (010.001.a.03)
8. In accordance with the procedure, SUMMARIZE the steps necessary to reset a half scram per 34AR-603-1 17(11 8)-1/2 Reactor Auto Scram System A(B) Trips. (010.011 .a.0 1)
9. Given a list, IDENTIFY the statement that describes the effect on RPS of pulling the scram solenoid fhses. (010.0 16.a.02)
10. Given a list, SELECT the Scrams or Rod Blocks which are bypassed with the Mode switch in RUN. (010.019.a.03)
11. Given a list, IDENTIFY the statement that describes the effect on RPS of placing the RPS Trip Test Logic Switches in TRiP. (010.022.a.02)
12. Given a list. IDENTIFY the statement that describes the effect on individual CRDMs of placing the individual control rod scram test switches in SCRAM. (010.020.a.02) 13 Given a simplified drawing or P&ID, TRACE the normal power supply and the alternate power supply for the RPS System. (010.001 .a.04, 01 0.025.a.0 1)
14. Given a list of power supplies, SELECT the Normal and Alternate power supplies for each RPS bus. (010.001.a.01)
15. IDENTIFY the back panels associated with the RPS System. (010.001.a.06)

EOP-103-LP-20314 Page 3 of 38 EOP 103: EOP CONTROL ROD INSERTION METHODS

16. Given a list, IDENTIFY the statement that describes the effects of defeating the ART logic trips on the CRD system. (0l0023.A.02)
17. Given 31E0-EOP-103-2, EOP Control Rod Insertion Methods, IDENTIFY the location of the RPS Test Trip Logic Switches. (0l0.020.A.04, 0l0.021.A.03 and 0l0.022.A.03)
18. Given a list, IDENTIFY the statement that describes the effect on RPS of placing the RPS Test Trip Logic Switches in NORMAL. (0l0.021.A.02 and 010.020.A.03)
19. Given a list, IDENTIFY the statement that describes the effects of overriding all of the automatic scram signals during an ATWS. (010.021.A.09 and 010.020.A.10)
20. Given a list, IDENTIFY the statement that describes the purpose of overriding all of the automatic scram signals during an ATWS. (010.02l.A.08)
21. Given a list, IDENTIFY the statement that describes the purpose of opening 2C 11 -F034 prior to repeating a manual scram if reactor pressure is less than 800 psig. (010.021 .A.05)
22. Given a list, IDENTIFY the statement that describes the purpose of placing the RPS Test Trip Logic Switches in TRIP during an ATWS. (0l0.022.A.01)
23. Given a list, IDENTIFY the statement that describes the effects on RPS of placing the RIS Test Trip Logic Switches in TRIP. (010.022.A.02)
24. Given a list, IDENTIFY the statement that describes the purpose of pulling the scram solenoid fuses during an ATWS. (010.016.A.01)
25. Given 31E0-EOP-l03-2, EOP Control Rod Insertion Methods, IDENTIFY the location of the scram solenoid fuses. (OlO.016.A.03)
26. Given a list, IDENTIFY the statement that describes the effect on RPS of pulling the scram solenoid fuses. (010.016.A.02)
27. Given 31E0-EOP-103-2, EOP Control Rod Insertion Methods, and a list of possible rod movements, RECOGNIZE the rod movement that would occur if scram solenoid fuses 2C7 1-Fl 8A-H were pulled. (010.01 6.A.06)
28. Given a list, IDENTIFY the statement that describes the purpose of opening 2Cll-F034 prior to individually scramming control rods if reactor pressure is less than 800 psig. (0l0.020.A.06)
29. Given 31EO-EOP-103-2, EOP Control Rod Insertion Methods, IDENTIFY the location of the individual control rod scram test switches. (OlO.020.A.l4)
30. Given a list, IDENTIFY the statement that describes the purpose of bypassing the RWM when inserting control rods with Reactor Manual Control during an ATWS. (001.031 .A.03)
31. Given 3 IEO-EOP-103-2, EOP Control Rod Insertion Methods, 34GO-OPS-065-0, Control Rod Movement, and plant conditions, DETERMINE which control rods can be inserted.

(400.072.A.02)

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 2 OF 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

EOP CONTROL ROD INSERTION METHODS 31 EO-EOP-1 03-2 60 3.0 OPERATOR ACTIONS ENTRY CONDITiON per section 2.0 YES G1&030 MGR-0001 Rev4

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 6 OF 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

EOP CONTROL ROD INSERTION METHODS 31E0-EOP-103-2 6.0 3.4 DE-ENERGIZING SCRAM SOLENOIDS 3.4.1 Place the following keylock RPS Test Trip Logic switches to TRIP position:

SWITCH DESCRIPTION LOCATION 2C71-S2A RPS Test Al Trip Logic 2H11-P609 2C71-S2C RPS Test A2 Trip Logic 2H1 l-P609 2C71-S2B RPS Test Bi Trip Logic 2H1 l-P6l 1 2C71-S2D RPS Test B2 Trip Logic 2H1 l-P61 1 Fuse pullers may be found in the Unit 1 AND Unit 2 key cabinets located behind the NOTE STA desk.

Critical 3.4.2 IF scram solenoids are still energized, THEN remove the following fuses:

LOCATION FUSES PANEL CCF03 2C7l-F18A 2Hl l-P609B CC-F04 2C71-FI8E 2H11-P609B CC-F05 2C7l-FI8C 2Hll-P609B CCF06 2C71-F18G 2H1 l-P609B CC-F03 2C71-F18B 2H11-P611B CCF04 2C7l-F18F 2Hll-P611B CC-F05 2C71-FI8D 2H11-P611B CC-F06 2C71-F18H 2Hll-P611B 3.4.3 Perform SYSTEM RESTORATION per Affachment 1 Section 3.0, WHEN directed by the Shift Supervisor.

G16.030 MGR-000l Rev4

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 9 OF 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

EOP CONTROL ROD INSERTION METHODS 31 EO-EOP-1 03-2 6.0 3.6 VENTING SCRAM AIR HEADER 3.6.1 Close 2C11-F095, Scram Air Header Isolation Valve, at 130RAR22.

Pliers may be obtained from the EOP gang box on the 130 elevation to aid in the cap NOTE*

removal.

3.6.2 IF the line is capped, THEN remove cap down stream of 2C1 1-R013-TV1, Pressure Instrumentation Vent Valve.

3.6.3 Confirm open 2C11-R013-IV1, Pressure Instrumentation Isolation Valve, at 130RAR22.

3.6.4 Open 2C11-R013-TV1, Pressure Instrumentation Vent Valve, at 130RAR22.

3.6.5 Perform SYSTEM RESTORATION per Attachment 1 Section 5.0, WHEN directed by the Shift Supervisor.

G16.030 MGR-0001 Rev 4

HLT-07 SRO NRC EXAM

59. 295038EK1.02 001 An event has ocurred at Plant Hatch with a potential radiation release to the public.

Which ONE of the choices below completes the following statements?

The NPO will determine Projected Offsite dose using If an emergency is declared due to the high Offsite release rate, the INITIAL emergency notification to State and Local Officials is REQUIRED within a MAXIMUM of from when the emergency is declared.

A. 73EP-EIP-015-0, Offsite Dose Assessment; 15 minutes B. 73EP-EIP-015-0, Offsite Dose Assessment; 30 minutes C 73EP-EIP-0 18-0, Prompt Offsite Dose Assessment; 15 minutes D. 73EP-EIP-01 8-0, Prompt Offsite Dose Assessment; 30 minutes 178

HLT-07 SRO NRC EXAM

==

Description:==

73EP-EIP-018-0, Prompt Offsite Dose Assessment provides the initial method used for prompt dose assessment to determine the dose rate at the site boundary for use in Emergency Classifications based on gaseous effluent and release determination. 73EP-E1P-0l5-0, Offsite Dose Assessment, provides dose assessments during abnormal (emergency) conditions. Various ARPs & Abnormal procedures require 73EP-EIP-Ol 8-0 to be performed for the initial projected dose.

NOTE: The current calculated daily average site dose rate is E-03 mR/hr.

Step 7.4.21 of 73EP-EIP-018-0 states IF the peak TEDE dose rate (mRlhr) value is an order of magnitude (10 times) higher than the current calculated daily average AND an emergency has been declared, THEN notify the Emergency Director that a radioactive release is in progress.

NMP-EP- ill, Emergency Notifications, Checklist 1 step 6.0 a. states Repeat the announcement approximately every thirty (30) minutes during the first (2) hours of the declared emergency and track time of announcement below . Also Limitation 6.1.1 states Initial notifications of applicable State and Local Agencies shall be accomplished as soon as practicable and within 15 minutes of the declaration of an emergency, an upgrade to a higher emergency classification level, or the approval of protective actions recommendations.

The A distractor is plausible if the applicant remembers that this is a procedure listed on the PCG EOP flowchart and thinks this is the one used. The second part is plausible since it is correct.

The B distractor is plausible if the applicant remembers that this is a procedure listed on the PCG EOP flowchart and thinks this is the one used. The second part is plausible if the applicant confuses/remembers the 30 minute requirement for followup notification times.

The C distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses/remembers the 30 minute requirement for followup notification times.

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

179

HLT-07 SRO NRC EXAM

References:

NONE K/A:

295038 High Off-Site Release Rate EK1. Knowledge of the operational implications of the following concepts as they apply to HIGH OFF-SITE RELEASE RATE: (CFR: 41.8 to 41.10)

EK1 .02 IProtection of the general public 4.2* 4*4*

LESSON PLAN/OBJECTIVE:

References used to develop this question:

73EP-EIP-018-0, Prompt Offsite Dose Assessment 73EP-EJP-015-0, Offsite Dose Assessment NMP-EP- 111, Emergency Notifications 180

SOUTHERN NUCLEAR I DOCUMENT TYPE: PAGE PLANT E. I. HATCH EMERGENCY PREPAREDNESS 1 OF 11 DOCUMENT TITLE: I DOCUMENT NUMBER: VERSION NO:

PROMPT OFFSITE DOSE ASSESSMENT 73EP-EIP-018-0 8.7 EXPIRATION I APPROVALS: EFFECTIVE DATE: I DEPARTMENT MGR J.C. Lewis DATE 03/23/05 DATE:

N/A PM / SSM I ED 09/22/11 C.R. Dedrickson DATE 03/24/05 1.0 OBJECTIVE This procedure provides the initial method used for prompt dose assessment to determine the dose rate at the site boundary for use in Emergency Classifications based on gaseous effluent and release determination. This procedure also provides the projection of offsite TEDE and CDE Thyroid doses for use in determination of Emergency Classifications based on dose projections and initial protective action recommendations (PARs).

TABLE OF CONTENTS Section 7.0 PROCEDURE 5 7.2 SYSTEM START-UP 5 7.3 DATA ACQUISITION 5 7.4 DETERMINATION OF OFFSITE DOSE RATES AND DOSE PROJECTIONS 7 7.5 REVIEWS AND RECORDS 11 2.0 APPLICABILITY This procedure is applicable to initial determinations of offsite dose and dose rates based upon estimated noble gas release from the Main Stack and Unit 1 and Unit 2 Reactor Building Vents.

This procedure is performed as required.

3.0 REFERENCES

3.1 MANUALS & PROCEDURES 3.1.1 1OAC-MGR-006-0, Hatch Emergency Plan 3.1.2 2OAC-ADM-002-0, Quality Assurance Records Administration 3.1.3 NMP-EP-110, Emergency Classification Determination and Initial Action 3.1.4 73EP-EIP-005-0, On-Shift Operations Personnel Emergency Duties 3.1.5 NMP-EP-112, Protective Action Recommendations 3.1.6 31 EO-EOP-01 3-1/2, Primary Containment Control 3.1.7 73EP-EIP-01 5-0, Offsite Dose Assessment MGR-0002 Rev 8

SOUTHERN NUCLEAR DOCUMENT TYPE: PAGE PLANT E. I. HATCH EMERGENCY PREPAREDNESS PROCEDURE 1 OF 30 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

OFFSITE DOSE ASSESSMENT 73EP-EIP-015-0 7.6 EXPI RATION APPROVALS: EFFECTIVE DATE: DEPARTMENT MGR J. C. Lewis DATE 5/4/2005 DATE:

N/A PM! SSM 09/22/11 D. Madison DATE 5/4/2005 1.0 OBJECTIVE To provide dose assessments during abnormal (emergency) conditions.

TABLE OF CONTENTS Section 7.0 PROCEDURE 5 7.1 SYSTEM STARTUP 5 7.2 DATA ACQUISITION 6 7.3 SELECTION OF THE MIDAS DOSE ASSESSMENT OPTIONS 8 7.4 RECAP A PREVIOUS MIDAS PROJECTION 9 7.5 PLANTHATCHSTATE OPTION 10 7.6 QUICK DOSE PROJECTION (MENU A) 12 7.7 ENHANCED DOSE PROJECTION (MENU B) 14 7.8 EVENT TREE NUREG-1228 (MENU C) 16 7.9 DEFAULT CLASS 9 ACCIDENTS (MENU D) 18 7.10 BACK DOSE CALCULATION (MENU E-W) 19 7.11 TOTAL DOSE FOR ACCIDENT (MENU F) 21 7.12 EVALUATION OF MIDAS OUTPUT 23 7.13 RUNNING NEXT TIME STEP OR EXITING MIDAS 24 7.14 DOCUMENTATION AND RECORDS 25 Attachments 1 DESCRIPTION OF MIDAS MENU OPTIONS 26 2 ISOTOPIC MIX SELECTION GUIDE 27 3 DEFAULT CLASS 9 ACCIDENT INFORMATION 29 2.0 APPLICABILITY This procedure is applicable to the SNC Plant Hatch dose assessment activities for assessing offsite radiological releases during emergency conditions. It may NOT be used for assessing normal or routine operating releases. This procedure will be performed as necessary.

MGR-0002 Rev 8.1

Southern Nuclear Operating Company II Emergency I NMP-EP-111 I

SOUTHERNA COMPANY Implementing Emergency Notifications Version 6.0 I Procedure Page 6of48 I 60 PROCEDURE 6.1 Precautions and Limitations 6.1.1 Initial notifications of applicable State and Local Agencies shall be accomplished as soon as practicable and within 15 minutes of the declaration of an emergency, an upgrade to a higher emergency classification level, or the approval of protective actions recommendations.

6.1.2 Initial notification of the NRC shall be completed as soon as possible after notifications to the state and county agencies and within an hour of the declaration of an emergency. Follow-up notifications of the NRC shall be made promptly after any further degradation in the plant conditions, any change from one emergency class to another, or for the termination of an emergency. NRC notifications are typically performed utilizing the Federal Telephone System (FTS). The Emergency Notification System (ENS) line is normally utilized. An open line is maintained for the duration of the event at the request of the NRC communicator receiving the initial notification.

6.1.3 For security based emergencies, notifications to the NRC should be performed within 15 minutes of discovery of an imminent threat or attack against the plant to ensure proper mobilization of federal resources.

CAUTION AN INITIAL NOTIFICATION OF AN UPGRADE IN EMERGENCY CLASSIFICATION SHOULD TAKE PRECEDENCE OVER A FOLLOW-UP MESSAGE OF A LOWER RANKING EMERGENCY. (I.E., AN INITIAL SITE AREA EMERGENCY NOTIFICATION TAKES PRECEDENCE OVER AN ALERT FOLLOW-UP NOTIFICATION.)

NOTE The following guidance shall be used for making emergency notifications during rapidly changing events in which the emergency classification changes before the lesser notification is made to the offsite response organizations (OROs) and the NRC. This information is consistent with RIS 2007-02, Clarification Of NRC Guidance For Emergency Notifications During Quickly Changing Events 6.1.4 If the plant condition degrades and a higher emergency classification is declared before the notifications are confirmed for the lesser emergency declaration, then a notification reflecting the higher emergency classification should be made. This notification should be made within 15 minutes of the lesser emergency declaration.

This should be performed IF the notification can be made within 15 minutes of the lesser (first) classification.

Southern Nuclear Operating Company Emergency I Emergency Classification Determination i NMP-EP-1 10 SOUTHERN Implementing Version 3.0 COMPANY and Initial Action i Procedure I Page 15 of 24 Checklist 2 Emergency Plan Initiation (page 3 of 8)

Initial Actions listed in order of priority. Take actions in parallel if resources Completed by are available. Utilize Figure 1 to delegate roles and responsibilities.

8. Ensure transmission of the EN form. Receipt of Emergency Notification must be verified within 15 minutes of emergency declaration, in accordance with NMP-EP-1 11, Emergency Notifications, Checklist 2, Emergency Communicator Electronic Method, or checklist 3, Emergency Communicator, Manual method, as appropriate.
9. Designate an individual to complete Control Room assembly and accountability in accordance with site procedures.
10. Designate an individual to make NRC notifications as soon as possible but no later than one hour following the declaration of an emergency in accordance with NMP-EP-1 11, Checklist 4. Ensure an open line with the NRC is maintained, IF requested.
11. Direct appropriate personnel to perform dose assessment per procedure.

Farley Hatch Vogtle FNP-O-EIP-9.1 73EPElP-015 91304-C FNP-0-EIP-9.3 73EP-EIP-018 FNP-0-EIP-9.5

12. Direct ERDS activation to transmit data to the NRC within one hour of the declaration of the emergency (ALERT and higher) using the appropriate site specific procedure below:

Farley Hatch Vogtle NMP-EP-1 1 1001 73EP-EIP-063 (Aft 1) 9111 1-C

HLT-07 SRO NRC EXAM

60. 300000K1.02 001 Unit 2 is at 100% power with the following conditions:

o A Service Air Header break has occured o The break is greater than the capacity of the Service Air Compressors Which ONE of the choices below completes the following statements?

As Service air pressure decreases, 2P51-F017 Turbine Building Service Air Isolation Valve, will isolate at If 2P51-F017 fails to isolate, CRD Flow Control Valves, 2C1 1-FOO2A & B lose pneumatic supply.

A. 70 psig; will B 70 psig; will NOT C. 50 psig; will D. 50 psig; will NOT 181

HLT-07 SRO NRC EXAM

==

Description:==

34S0-P51-002-2, Instrument And Service Air Systems, step 5.2.2 states:

The following will occur on system decreasing air pressure:

5.2.2.3 70 psig decreasing, 2P51-F017, TB Service Air Isolation Valve, CLOSES.

5.2.2.4 50 psig decreasing, 2P52-F015, Non-Essential Inst. Air Isolation Valve, CLOSES.

At 80 psig, the 2P52-F565, Rx. Bldg Inst. N2 to Non-mt Air El. 185 Isol Vlv, OPENS to supply nitrogen to the Non-Interruptible Air Header. This header also supplys backup nitrogen to the CRD Master Piping Area, therefore maintaining a pneumatic supply to the CRD FCVs.

The A distractor is plausible since the first part is correct. The second part is plausible if the applicant does not remember the opening setpoint for the 2P52-F565 or confuses whether or not the CRD Master Piping Area header receives backup nitrogen.

The C distractor is plausible if the applicant does not remember or confuses the closing setpoint. The second part is plausible if the applicant does not remember the opening setpoint for the 2P52-F565 or confuses whether or not the CRD Master Piping Area header receives backup nitrogen.

The D distractor is plausible if the applicant does not remember or confuses the closing setpoint. The second part is correct.

A. Incorrect See description above.

B. Correct See description above.

C. Incorrect See description above.

D. Incorrect See description above.

References:

NONE K/A:

300000 Instrument Air System (lAS)

Ki Knowledge of the connections and / or cause effect relationships between INSTRUMENT AIR SYSTEM and the following: (CFR: 41.2 to 41.9/45.7 to 45.8) 182

HLT-07 SRO NRC EXAM lçI.U2 Service air .1 LESSON PLAN/OBJECTIVE:

P51-P52-P70-PLANT AIR-LP-03501, Plant Air Systems, EO 200.025.A.0l References used to develop this guestion:

34AB-P51-00l-2, Loss of Instrument and Service Air Modified from HLT Database Q#LT-200025-065 ORIGINAL QUESTION (HLT Database Q#LT-200025-065)

Unit 2 is operating at 100% RTP when a complete (100%) rupture on the Non-Essential Instrument Air Header, down stream of 2P52-F015, occurs.

This line breaks r1 FOHA

[014

{>1 i:

O13R HXi Fill HI

)

1565 If NO operator action is taken, which ONE of the following choices correctly states how the plant will respond to this pipe break?

The Service Air Header isolation valve (F017) will and the Outboard MSIVs 183

HLT-07 SRO NRC EXAM will A.v close and remain closed; remain open B. close and remain closed; drift closed C. continously cycle open and closed; remain open D. continously cycle open and closed; drift closed 184

P51-P52-P70-PLANT AIR-LP-03501 Page 4 of 112 PLANT AIR SYSTEMS

6. Given a list of statements, SELECT the statement which best describes the possible adverse effects of operating Instrument Air with pre or after filters bypassed per 34S0-P5 1-002-1 Instrument and Service Air Systems. (035.005.a.02)
7. Given a list of statements, SELECT the statement that best describes the function of the Instrument Air Dryers. (035 .004.a.0 1)
8. Given a list of statements, SELECT the statement that best describes the possible detrimental effects of shutting down both Instrument Air Dryers per 34S0-P5 1-002-1 Instrument and Service Air Systems. (035.006.a.01)
9. Given a simplified drawing or P&ID, TRACE the system flowpath when the air dryers are shutdown. (035.006.a.02)
10. STATE the purpose of the Instrument Air Accumulators. (035.001.a.05)
11. Given a simplified drawing or P&ID, TRACE the Air System flowpath for cross connecting Unit 1 and Unit 2 Service Air Systems. (035.01 l.a.01)
12. Given a list of electrical buses, SELECT the bus that supplies power to the following loads:

(035.00 1 .a.03)

a. Station Service Air Compressors
b. Low Pressure Blower
c. Closed Cooling Water Pumps
13. Given a list, IDENTIFY the panel locations necessary to operate the air compressors and dryers per 34S0-P51-002-2/1 Instrument and Service Air Systems. (035.001.a.15)
14. LIST five plant responses to a loss of Instrument/Service Air as stated in 34AB-P51-001-2/1 Loss of Instrument and Service Air System. (200.025.a.01)
15. Given a list of plant conditions, DETERMINE if the reactor must be manually scrammed per 34AB-P51-001-2/1 Loss of Instrument and Service Air System. (200.025.a.02)
16. Given a list, IDENTIFY the firewatch requirements if fire sprinklers are isolated during a loss of air per 34AB-P51-001-2/1 Loss of Instrument and Service Air System. (200.025.a.04)
17. Given a list of statements, SELECT the statement that best describes the response to a loss of air for the following systems: (200.025.a05)
a. CRD
b. Reactor Water Level Control
c. SJAEs
d. Outboard MSIVs
18. Given a simplified drawing or P&ID of the Instrument/Service Air System and the location of a leak, PREDICT the air system response and final status. (200.025.a.03)

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 2 OF 20 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF INSTRUMENT AND SERVICE AIR SYSTEM OR 34AB-P51-001-2 4.8 WATER INTRUSION INTO THE SERVICE AIR SYSTEM LOSS OF INSTRUMENT AND SERVICE AIR SYSTEM 1.0 CONDITIONS 1.1 ANNUNCIATORS 1.1.1 CONTROL BLDG SERVICE AIR PRESS LOW (700-222) 1.1.2 AIR CMPSR 2A TRIPPED/SHUTDOWN (700-221) 1.1.3 AIR CMPSR 2B TRIPPED/SHUTDOWN (700-227) 1.1.4 AIR CMPSR 2C TRIPPED/SHUTDOWN (700-233) 1.1.5 REACTOR LEVEL CONTROL VALVE LOCKED (603-123) 1.1.6 SCRAM VLV PILOT AIR HDR PRESS HIGH/LOW (603-131) 1.1.7 CONTROL BLDG SERVICE AIR SHUTOFF (700-234) 1.1.8 INST AIR DRYER SYS PRESS LOW (700-219) 1.2 For Additional Annunciators Received See Attachment 2 1.3 Decreasing Instrument and Service Air Pressure 1.4 CRD Flow decreasing (CRD High Temperature) - Flow Control Valve fails closed 1.5 Feedwater Startup Level Control Valve fails as is 1.6 Decreasing Condenser vacuum - SJAE steam supply valves fail closed 1 .7 Ventilation System failures dampers fail closed 1.8 Feedwater Heater Level Control problems 1.9 For a List of Additional Conditions See Attachment 3 MGR-0001 Rev 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 3 OF 20 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF INSTRUMENT AND SERVICE AIR SYSTEM OR 34AB-P51-001-2 4.8 WATER INTRUSION INTO THE SERVICE AIR SYSTEM 2.0 AUTOMATIC ACTIONS 2.1 At 80 PSIG Non-Interruptible Air Pressure, 2P52-F565, Rx Bldg Inst N 2 To Non-Int Air El. 185 Isol Vlv, OPENS to supply this header from the N 2 Inerting System.

2.2 At 75 PSIG Turbine Building Instrument Air Pressure, standby prefilter and afterfilter are automatically put into service.

2.3 At 70 PSIG Service Air Pressure, 2P51-F017, Service Air Isolation valve, CLOSES, isolating Service Air System.

2.4 At 50 PSIG Instrument Air Pressure, 2P52-F01 5, Turb Bldg Inst Air After Fltrs, Dl O2AIB To RW Bldg Isol valve, CLOSES, isolating Non-Essential Instrument Air.

2.5 At 50 PSIG Control Air Pressure, 2N21-F1ll, Startup LCV, LOCKS UP in its existing position.

3.0 IMMEDIATE OPERATOR ACTIONS NONE MGR-0001 Rev3

SNCPLANTE. I. HATCH I Pgl3of2O DOCUMENT TITLE: DOCUMENT NUMBER: Ver No:

LOSS OF INSTRUMENT AND SERVICE AIR SYSTEM OR 34AB-P51-001-2 4.8 WATER INTRUSION INTO THE SERVICE AIR SYSTEM ATTACHMENT 2 Aft. Pg.

TITLE: ADDITIONAL ANNUNCIATORS AND EQUIPMENT 2 of 4 Non-Interruptible Essential Instrument Air supplies:

1. Refuel Floor Suction to Standby Gas treatment System (SBGT) (FOO3AIB)
2. Reactor Building Suction to SBGT (FO0 lA/B)
3. Hydrogen oxygen analyzer isolations
4. SBGT Train Discharge Isolation (FOO2AIB)
5. Plant Service Water (PSW) Isolation of Reactor Building Non-ECCS Equipment (F066, F067)
6. Residual Heat Removal (RHR) Suction Isolation (FO65AJBICID)
7. Core Spray (CS) Suction Isolation (FOl 9A/B)
8. Drywell/Torus pressure sensing line isolations
9. High Pressure Coolant Injection System (HPCI) suction from Torus (F051)
10. Scram Air Supply (75 psig)
11. Control Rod Drive System (CRD) valves (30 psig)
12. Reactor Building/Torus Butterfly Vacuum Breakers (2T48-F311, F310)
13. Torus water level transmitter isolations
14. HPCI drain Air Operated Valves (AOVs)
15. Reactor Building floor drain isolations
16. Feedwater Line AOV Check Valves (FO76AJB, FO77AJB)
17. Outboard Main Steam Isolation Valves (MSIVs) 2B21-FO28AIB/C/D)
18. Reactor Core Isolation Cooling (RCIC) Suction from Torus (E51-F003)
19. RCIC drains AOVs
20. Backup supply to Drywell Pneumatic System MGR-0009 Rev 4

HLT-07 SRO NRC EXAM

61. 300000K6.04 001 Unit 2 is operating at 100% RTP with the 2P51-COO1C, Station Service Air Compressor (SSAC), in service and 2P51-COO1B, SSAC, in the Automatic Mode.

o 2P51-COO1C. SSAC, trips.

If NO operator action is taken, which ONE of the choices below completes the following statements?

The setpoint at which 2P51-COO1B, SSAC, will Automatically Start & Load is lithe 2P51-COO1B, SSAC fails to Start & Load, the Outboard MSIVs (2B21-F028A-D),

are expected to A 100 psig; remain open B. 100 psig; drift closed C. 107 psig; remain open D. 107 psig; drift closed 185

HLT-07 SRO NRC EXAM

==

Description:==

Plant Hatch does not have a valve labeled Service Air Refusal Valve nor is it referenced in our procedures or lesson plans. After our conversation on 04182012 with Chief Examiner (Edwin Lea), it was agreed upon to write a discriminating question concerning the SSAC load &

unload valves which would meet the intent of this K/A.

Automatic Initiations and Isolations:

The standby Service Air Compressor (A or B) starts at 100 psig and C starts at 107 psig The Nitrogen Backup Valve(s) (F565) open automatically at 80 psig decreasing to supply Nitrogen pressure to the Non-Interruptible Essential Air Header.

The Air Compressor Discharge Valves FO1OA/BJC (F200A/B/C) close if compressor discharge pressure decreases to 80 psig.

Also at 70 psig, the Service Air Isolation valve F017 closes, isolating the Service Air System.

In this question, when 2C SSAC trips the header pressure will decrease causing the 2P51-COO1B to automatically start and load at 100 psig. If the the SSAC fails to Start & Load, system header pressure will decrease to the point where Instrument Air will be supplied backup Nitrogen. The Outboard MSLVs will remain open since there header will be supplied this backup Nitrogen.

The B distractor is plausible since the first part is correct. The second part is plausible since it would be correct if the backup Nitrogen valve did not operate.

The C distractor is plausible if the applicant does not remember the start and loading signal for the SSAC and would be correct if asking about the 2P51-COO1C SSAC. The second part is correct.

The D distractor is plausible if the applicant does not remember the start and loading signal for the SSAC and would be correct if asking about the 2P51-COO1C SSAC. The second part is plausible since it would be correct if the backup Nitrogen valve did not operate.

A. Correct: See description above.

B. Incorrect: See description above.

C. Incorrect: See description above.

D. Incorrect: See description above.

186

HLT-07 SRO NRC EXAM

References:

NONE K/A:

300000 Instrument Air System (lAS)

K6 Knowledge of the effect that a loss or malfunction of the following will have on the INSTRUMENT AIR SYSTEM: (CFR: 41.7/45.7)

K6.04 Service air refusal valve 2.6 2.5 LESSON PLAN/OBJECTIVE:

P51P52-P70-PLANT AIR-LP-03501, Plant Air Systems, EO 200.025.A.05 References used to develop this question:

34S0-P51-002-2, Instrument And Service Air Systems 187

P51-P52-P70-PLANT AIR-LP-03501 Page 4 of 112 PLANT AIR SYSTEMS

6. Given a list of statements, SELECT the statement which best describes the possible adverse effects of operating Instrument Air with pre or afier filters bypassed per 34S0-P5 1-002-1 Instrument and Service Air Systems. (035.005.a.02)
7. Given a list of statements, SELECT the statement that best describes the function of the Instrument Air Dryers. (035.004.a.01)
8. Given a list of statements, SELECT the statement that best describes the possible detrimental effects of shutting down both Instrument Air Dryers per 3450-PS 1-002-1 Instrument and Service Air Systems. (035.006.a.0l)
9. Given a simplified drawing or P&ID, TRACE the system flowpath when the air dryers are shutdown. (035.006.a.02)
10. STATE the purpose of the Instrument Air Accumulators. (035.00 l.a.05)
11. Given a simplified drawing or P&ID, TRACE the Air System flowpath for cross connecting Unit I and Unit 2 Service Air Systems. (035.01 l.a.01)
12. Given a list of electrical buses, SELECT the bus that supplies power to the following loads:

(035.001 .a.03)

a. Station Service Air Compressors
b. Low Pressure Blower
c. Closed Cooling Water Pumps
13. Given a list, IDENTIFY the panel locations necessary to operate the air compressors and dryers per 34S0-P51-002-2/1 Instrument and Service Air Systems. (035.001.a.15)
14. LIST five plant responses to a loss of Instrument/Service Air as stated in 34AB-P51-001 -2/1 Loss of Instrument and Service Air System. (200.025.a.01)
  • 15. Given a list of plant conditions, DETERMINE if the reactor must be manually scrammed per 34AB-P51-001-2/1 Loss of Instrument and Service Air System. (200.025.a.02)
16. Given a list, IDENTIFY the firewatch requirements if fire sprinklers are isolated during a loss of air per 34AB-P51-001-2/1 Loss of Instrument and Service Air System. (200.025.a.04)
17. Given a list of statements, SELECT the statement that best describes the response to a loss of air for the following systems: (200.025.a.05)
a. CRD
b. Reactor Water Level Control
c. SJAEs
d. Outboard MSIVs
18. Given a simplified drawing or P&ID of the Instrument/Service Air System and the location of a leak, PREDICT the air system response and final status. (200.025.a.03)

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 3 OF 20 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF INSTRUMENT AND SERVICE AIR SYSTEM OR 34AB-P51-001-2 4.8 WATER INTRUSION INTO THE SERVICE AIR SYSTEM 2O AUTOMATIC ACTIONS 2.1 At 80 PSIG Non-Interruptible Air Pressure, 2P52-F565, Rx Bldg Inst N 2 To Non-Int Air El. 185 Isol Vlv, OPENS to supply this header from the N 2 Inerting System.

2.2 At 75 PSIG Turbine Building Instrument Air Pressure, standby prefilter and afferfilter are automatically put into service.

23 At 70 PSIG Service Air Pressure, 2P51-F017, Service Air Isolation valve, CLOSES, isolating Service Air System.

2.4 At 50 PSIG Instrument Air Pressure, 2P52-F01 5, Turb Bldg Inst Air After Fltrs, Dl O2AIB To RW Bldg Isol valve, CLOSES, isolating Non-Essential Instrument Air.

2.5 At 50 PSIG Control Air Pressure, 2N21-F111, Startup LCV, LOCKS UP in its existing position.

3M IMMEDIATE OPERATOR ACTIONS NONE MGR-000l Rev 3

SNC PLANT E. I. HATCH Pg 13 of 20 DOCUMENT TITLE: DOCUMENT NUMBER: Ver No:

LOSS OF INSTRUMENT AND SERVICE AIR SYSTEM OR 34AB-P51-001-2 4.8 WATER INTRUSION INTO THE SERVICE AIR SYSTEM ATTACHMENT 2 Aff. Pg.

TITLE: ADDITIONAL ANNUNCIATORS AND EQUIPMENT 2 of 4 Non-Interruptible Essential Instrument Air supplies:

1. Refuel Floor Suction to Standby Gas treatment System (SBGT) (FOO3A/B)
2. Reactor Building Suction to SBGT (F00 lA/B)
3. Hydrogen oxygen analyzer isolations
4. SBGT Train Discharge Isolation (FOO2AJB)
5. Plant Service Water (PSW) Isolation of Reactor Building Non-ECCS Equipment (F066, F067)
6. Residual Heat Removal (RHR) Suction Isolation (FO65AJB/CID)
7. Core Spray (CS) Suction Isolation (FO19AIB)
8. Drywell/Torus pressure sensing line isolations
9. High Pressure Coolant Injection System (HPCI) suction from Torus (F051)
10. Scram Air Supply (75 psig)
11. Control Rod Drive System (CRD) valves (30 psig)
12. Reactor Building/Torus Butterfly Vacuum Breakers (2T48-F311, F310)
13. Torus water level transmitter isolations
14. HPCI drain Air Operated Valves (AOV5)
15. Reactor Building floor drain isolations
16. Feedwater Line AOV Check Valves (FO76AIB, FO77NB)
17. Outboard Main Steam Isolation Valves (MSIVs) 2B21-FO28NB/C/D)
18. Reactor Core Isolation Cooling (RCIC) Suction from Torus (E51-F003)
19. RCIC drains AOVs
20. Backup supply to Drywell Pneumatic System MGR-0009 Rev 4

HLT-07 SRO NRC EXAM

62. 400000K2.02 001 Unit 1 is at 100% power when the following annunciators alarm:

o 600V BUS 1D BREAKER TRIPPED, 652-3 18 o 600V BUS 1D UNDERVOLTAGE, 652-323 Which ONE of the following lists two components that have lost power?

A. RBCCW Pump A; RBCCW Drywell Outlet Isolation Valve, 1P42-F052 B. RBCCWPumpA; RWCU Inboard Suction Isolation Valve, 1 G3 1-FOOl C RBCCW Pump B; RBCCW Drywell Outlet Isolation Valve, 1P42-F052 D. RBCCW Pump B; RWCU Inboard Suction Isolation Valve, 1G3 1-FOOl 188

HLT-07 SRO NRC EXAM

==

Description:==

RBCCW pump A is powered from Bus C, RBCCW pump B is powered from Bus D. RBCCW F052 valve is powered from MCC S012 which gets power from Bus D.

RWCU FOOl valve is powered from MCC SOil which gets power from Bus C, F004 is powered from 600 Bus D thru SO 12.

The A distractor is plausible if the applicant confuses the power supplies of the RBCCW pumps. The second part is correct.

The B distractor is plausible if the applicant confuses the power supplies of the RBCCW pumps. The second part if the appicant confuses the power supply and thinks lG3 1-FOOl comes off S012 instead of SOl 1.

The D distractor is plausible since the first part is correct. The second part if the applcant confuses the power supply and thinks 1G3 1-FOOl comes off S012 instead of SO 11.

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

References:

NONE K/A:

400000 Component Cooling Water System (CCWS)

K2. Knowledge of electrical power supplies to the following: (CFR: 41.7)

K2.02 CCW valves 2.9 2.9 LESSON PLAN/OBJECTIVE:

P42-RBCCW-LP-00901, Reactor Building Closed Cooling Water, EO 0O9.002.A.02 R23-ELECT-LP-O2703, 600 I 480 / 208 VAC Electrical G3 1 -RWCU-LP-0030 1, Reactor Water Cleanup 189

HLT-07 SRO NRC EXAM References used to develop this cjuestion:

34S0-P42-001-1, RBCCW System (Att. 1, 1R24-S012) 34S0-G31-003-1, RWCU System (Att. 1, 1R24-S01 1)

Modified from HLT-5 NRC Exam Q#62 ORIGINAL QUESTION (HLT-5 NRC Exam Q#62)

Unit 1 is at 100% power when the following annunciators alarm:

o 600V BUS 1D BREAKER TRIPPED, 652-3 18 o 600V BUS 1D UNDERVOLTAGE, 652-323 Which ONE of the following lists two components that have lost power?

A. Station Service Air Compressor A; RBCCW Drywell Inlet Isolation Valve (P42-F05 1)

B. Station Service Air Compressor A; RWCU Inboard Suction Isolation Valve (G3 1-FOOl)

C.v Station Service Air Compressor B; RBCCW Drywell Inlet Isolation Valve (P42-F051)

D. Station Service Air Compressor B; RWCU Inboard Suction Isolation Valve (G31-F00l) 190

P42-RBCCW-LP-00901 Page 2 of 50 REACTOR BUILDING CLOSED COOLING WATER Initial License LT) ENABLING OBJECTIVES

1. Given a P&ID or simplified drawing of the RBCCW system, TRACE the flowpath through the system. (009.001 .A.02)
2. Given a list of major plant components, IDENTIFY those components which are supplied with cooling water from RBCCW. (009.003.A.02)
3. STATE the power supplies for each of the three RBCCW pumps. (009.002.A.02)
4. STATE the source of fill and makeup water for the RBCCW system. (009.00 l.A.0l)
5. IDENTIFY the location of the following RBCCW system instruments and controls.

(009.002.A.03)

a. Pump control switches
b. Drywell isolation valves
c. RBCCW suction temperature indication
6. DESCRIBE how RBCCW surge tank level is automatically maintained using demineralized water for makeup. (009.001 .A.03)
7. Given a set of plant conditions, DETERMINE if a specific RBCCW pump should have auto started. (200.01 4.A.06)
8. Given a list of plant conditions, SELECT the condition which would cause a trip of a running RBCCW pump. (200.014.A.05)
9. Given a P&ID or simplified drawing of the RBCCW system, IDENTIFY the correct valve lineup for an inservice and standby PSW/RBCCW heat exchanger. (009.004.A.01)

Objectives marked by a RED (*) are required during RO-305 and SR-305 of the Initial License program.

P42-RBCCW-LP-00901 Page 43 of 50 REACTOR BUILDING CLOSED COOLING WATER TABLE 2 POWER SUPPLIES COMPONENT POWER SUPPLY RBCCW Pump A 600 VAC Bus C (R23-5003)

RBCCW Pump B 600 VAC Bus D (R23-S004)

RBCCW Pump C 600 VAC Bus C (R23-S003)

RBCCW Metering Pump 208V MCC 2R25-S 105 (1R25-Sl 19,120/208 VAC Dist Cab in 1R24-S015)

RBCCW Agitator Motor 208V MCC 2R25-S 105 (1R25-S1 19 120/208 VAC Dist Cab in 1R24-S015)

Drywell Inlet Isolation Vlv (F05 1) 600 VAC MCC R24-S012 Drywell Outlet Isolation Vlv (P052) 600 VAC MCC R24-S012 Surge Tank Makeup Valve (F054) 2R25-S102 in 2R24-S012 (1R25-S1 16 in 1R24-S012)

RBCCW Local Control Panel 1R25-S 122 120 VAC MISC POWER PANEL on 1R24-S037 Panel 1H21-P659 (RBCCW) 1R25-S1 16 120/208 Distribution Cabinet in 600VAC MCC lB 13ORFRO2 Reactor Building Instrument Enclosure 1H2 1- 1R25-S1 16 120/208 Distribution Cabinet in P151 600VAC MCC lB 13ORFRO2 RBCCW Inlet Hx A & B Valves Position 1R25-S064 Instrument Bus 1A Indication RBCCW Heat Exchanger Outlet temperature 1R25-S065 120/208V Cabinet IC Instrument switch 1P42-R600 Bus lB

SOUTHERN NUCLEAR PLANT El. HATCH DOCUMENT TITLE:

I PAGE38OF55 DOCUMENT NUMBER: VERSION NO:

REACTOR BUILDING CLOSED COOLING WATER SYSTEM 34S0-P42-001-l 11.13 ATTACHMENT I ATTACHMENT PAGE:

TITLE: RBCCW ELECTRICAL LINEUP 2 OF 3 CONTROL SWITCHES FOR RBCCW PUMPS IA, lB & IC MUST BE IN PULL TO LOCK PRIOR TO RACKING IN THE ASSOCIATED BREAKER.

NUMBER DESCRIPTION CHECKED VERIFIED 1R23-S003 600 VAC BUS 1C I3OTETI1 Frame 7B RBCCW Pump 1P42-C001 A RACKED IN Frame 8B RBCCW Pump 1P42-C001 C RACKED IN 1R23-S004 600VAC BUS 1D 130TCT1O Frame 7B RBCCW Pump 1P42-C0O1B RACKED IN 1R25-S122 120 VAC MISC POWER PANEL on 1R24-S037 203RHR03 BRKR 2 RBCCW Local Control Panel CLOSED 1R25-5116 120/208 Distribution Cabinet in 600VAC MCC lB 13ORFRO2 BRKR 9 Panel lH2l-P659 (RBCCW) CLOSED BRKR 22 Reactor Building Instrument Enclosure 1H21-P151 CLOSED 1R24-S012 600VAC MCC lB 13ORFRO2 Frame 9A RBCCW Drywell Isolation Valve I P42-F051 CLOSED Frame 14C RBCCW Drywell Isolation Valve 1P42-F052 CLOSED 1R25-S064 Instrument Bus 1A I3OTGT11 BRKR 11 RBCCW Inlet Hx A & B Valves Position Indication CLOSED OPS-0656 Ver. 0 G16.030 MGR-0009 Ver. 5

G31-RWCU-LP-00301 Page 29 of 80 REACTOR WATER CLEANUP H. RBCCW RBCCW supplies cooling water to RWCU Pumps 1B and 2A bearings and seals, and the 1A and 2B pump motor heat exchangers.

Electrical (LT 8)

1. AC:
a. The Unit 2 RWCU Pumps A/B are powered from Unit 2 Turbine Building 600/208 VAC MCC 2R23-S01 1 (600 2AA).
b. The Unit 1 A RWCU Pump is powered from Unit 1 Turbine Building 600/208 VAC MCC R23S01 1 (600 1AA). IB is powered from Unit I Reactor Building 600/208 VAC MCC R24-S0l 5.
c. RWCU Inboard Suction Isolation Valve (FOOl) is powered from 2R24-S01 1 (Unit 1 froinR24-S0ll).
2. DC:
  • RWCU Outboard Suction Isolation Valve (F004) is powered from Reactor Building VDC MCC 2B Essential Division II 2R24-S022 (Unit 1 from R24-S022).
3. Review Table 2 for additional system power supplies.

J. Instrument and/or Service Air Systems

1. The Service Air system supplies air for backwashing the RWCU demins.
2. Interruptible Instrument Air supplies the AOVs and instrumentation.

K. Liquid Sampling System

1. The Liquid Sampling System continuously monitors Rx water chemistry. It takes samples from:
a. Combined demineralizer inlet
b. Demineralizer A outlet
c. Demineralizer B outlet
d. Reactor Recirculation System if aligned via control switch on 2H 11 -P602 (Unit 1 on local panel H2l-P023).

SNC PLANT E. I. HATCH I Pg 109 of 186 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

REACTOR WATER CLEANUP SYSTEM 34SO-G31-003-1 35.0 ATTACHMENT 1 Att. Pg.

TITLE: RWCU ELECTRICAL LINEUP 2 of 3 DESCRIPTION CHECKED VERIFIED 1 C71 -P001 1 3OTGT1 2 RWCU INBOARD VALVE, 1G31-FOO1, ISOLATION LOGIC CB5A C LOSED POWER SUPPLY RWCU OUTBOARD VALVE, 1G31-F004, ISOLATION CB5B CLOSED LOGIC POWER SUPPLY 1R25-S119 (120/208 VAC POWER PANEL ON 1R24-S015) 185RHR03 17 RWCU DEMIN PRECOAT TANK AGITATOR 1G31-D003 CLOSED 34 RWCU DEMIN BYPASS VALVE 1G31-F044 CLOSED 1R25-S122 (120/208 VAC POWER PANEL ON 1R24-S037) 203RHR03 32 CONTROL POWER TO RWCU DEMIN PANEL 1G31-P005 CLOSED 34 POWER TO 1G31-P005 AIR CONDITIONER CLOSED MCC 1R24-S011 600V MCC 1C ESS DIV 113ORHR13 14C RWCU INBOARD ISOLATION VALVE 1G31 -FOOl 1R24-S014 600V MCC 1D 13ORHR13 2B RESTRICTING ORIFICE BY-PASS VALVE, 1G31-F031 RKED IN CLOSED 2C DISCHARGE TO MAIN CONDENSER 1G31-F034 RKED IN CLOSED RKED IN 3A DRAIN TO WASTE COLLECTOR VALVE 1 G31 -F035 CLOSED 3B CLEANUP RETURN ISOLATION VALVE 1G31-F042 RKED IN CLOSED 1R24-S015 600V MCC iF 185RHR03 1C FP & RWCU DEMINERALIZER BLOWER 1 P51-C003 RKED IN CLOSED 2A RWCU DEMINERALIZER PRECOAT PUMP 1G31-C003 RKED IN CLOSED REACTOR CLEAN-UP RECIRCULATING PUMP 3B OPEN 1G31-COO1B 5B RWCU HOLDING PUMP 1G31-COO2A RKED IN CLOSED 5C RWCU HOLDING PUMP 1G31-COO2B RKED IN CLOSED 1P33-P102 PROCESS SAMPLING STATION RKED IN 2DL CHILLER MOTOR CLOSED OPS-0216 Ver. 6 G16.030 MGR-0009 Ver. 5

HLT-07 SRO NRC EXAM

63. 500000EK3.07 001 Unit 1 has experienced an accident that results in these Primary Containment parameters:

o Hydrogen concentration .... 8%

o Oxygen concentration 7%

o Drywell (DW) pressure .... 14 psig and slowly increasing o Torus level >300 inches With the above conditions and lAW 31E0-PCG-001-1, Primary Containment Gas Control chart, which ONE of the choices below completes the following statements?

The operator is required to vent the the containment atmosphere.

Placing Torus Sprays in service allowed.

A. DW to reduce the flammability of; is B DW to reduce the flammability of; is NOT C. Torus to reduce the flammability and scrub; is D. Torus to reduce the flammability and scrub; is NOT 191

HLT-07 SRO NRC EXAM

==

Description:==

With the concentrations of H2 and 02 given in this question, the EOPs direct the operators to:

o Vent the Suppression Chamber if Suppression Pool water level is below 300 o Vent the DW if Suppression Pool water level is > 300 o Spray the DW and Suppression Chamber Venting the DW or Torus will reduce the flammability of the containment atmosphere.

Scrubbing only occurs when the DW is sprayed or when reducing DW pressure via Torus venting through the water in the Torus.

lAW the PC EOP flowchart Torus level must be below 285 inches to spray the Torus The A distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses/does not remember the Torus level limit of 285 inches which does not allow Torus Sprays to be placed in service.

The D distractor is plausible if the applicant confuses which area (DW or Torus) will be vented based on Torus level and would be correct if Torus level was <300 inches. The second part is correct.

The C distractor is plausible if the applicant confuses which area (DW or Torus) will be vented based on Torus level and would be correct if Torus level was <300 inches. The second part is plausible if the applicant confuses/does not remember the Torus level limit of 285 inches which does not allow Torus Sprays to be placed in service.

A. Incorrect See description above.

B. Correct See description above.

C. Incorrect See description above.

D. Incorrect See description above.

192

HLT-07 SRO NRC EXAM K/A:

500000 High Containment Hydrogen Concentration EK3. Knowledge of the reasons for the following responses as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATIONS: (CFR: 41.5/45.6)

EK3.07 Operation of drywell vent 3.1 3.7 LESSON PLAN/OBJECTIVE:

EOP-PC-LP-203 10, Primary Containment Control, EO 201 .072.A. 15 & EO 201 .072.A.25 References used to develop this question:

31EO-PCG-001-1, Primary Containment Gas Control 31 EO-EOP-0 12-1, PC Primary Containment Control 193

EOP-PC-LP-20310 Page 5 of 82 PRIMARY CONTAINMENT CONTROL (PC)

34. Given plant conditions and access to plant procedures, EVALUATE plant conditions and DETERMINE if torus temperature and RPV pressure can be maintained below the HCTL.

(201.074.A.13)

35. Given a list, IDENTIFY the statement that describes the reason for emergency depressurizing the RPV if torus temperature and reactor pressure cannot be maintained below the HCTL.

(201 .074.A. 14)

36. Given a list, IDENTIFY the statement that describes the purpose of attempting to maintain primary containment pressure below 1.85 psig. (201.076.A.0l)
37. Given a list, IDENTIFY the statement that describes the purpose of using primary containment venting systems and SBGT as the first method of attempting to maintain primary containment pressure below 1.85 psig. (20L076.A.02)
38. Given a list, IDENTIFY the statement that describes the types of information an operator would use to determine if primary containment pressure can be maintained below a specified value.

(201 .076.A.03)

39. Given plant conditions and access to plant procedures, EVALUATE plant conditions and DETERMINE if primary containment pressure can be maintained below 1.85 psig.

(201 .076.A.04)

40. Given a list, IDENTIFY the statement that describes the two possible results of continuing to spray the suppression chamber when suppression chamber pressure is below 0 psig.

(201.072.A.23 and 201.076.A.09)

41. Given a list, IDENTIFY the statement that describes the purpose of NOT initiating suppression chamber sprays if torus level is above 285 inches. (201 .072.A.25 and 201 .076.A. 11)
42. Given a list, IDENTIFY the two possible consequences of continuing to spray the drywell when drywell pressure is below 0 psig. (20l.072.A.32, 20l.073.A.13, 201.076.A.19)
43. Given a list, IDENTIFY the statement that describes the purpose of securing recirculation pumps and drywell cooling fans prior to initiating drywell sprays. (201 .072.A.30, 201 .073.A.09, 201 .076.A.l7)

EOP-PC-LP-20310 Page 7 of 82 PRIMARY CONTAINMENT CONTROL (PC)

55. Given plant conditions and access to plant procedures, EVALUATE plant conditions and DETERMINE if drywell temperature can be maintained below 150°F. (201.073.A.03)
56. Given a list, IDENTIFY the statement that describes the reason for attempting to maintain drywell temperature below 340°F. (201.073.A.17)
57. Given a list, IDENTIFY the statement that describes the reason for entering RC [A] RPV Control before drywell temperature reaches 340°F. (201.073.A.16)
58. Given drywell temperature approaching 340°F and 31E0-EOP-012-2, PC Primary Containment Control, EVALUATE plant conditions and enter 31 EO-EOP-0 10(011 )-2, RC

[A] RPV Control at Point A before drywell temperature reaches 340°F. (201.073.A.05)

59. Given plant conditions and access to plant procedures, EVALUATE plant conditions and DETERMINE if drywell temperature can be maintained below 340°F. (201.073.A.15)
60. Given a list, IDENTIFY the statement that describes the reason for emergency depressurizing the RPV if drywell temperature cannot be maintained below 340°F. (201 .073.A.18)
61. IDENTIFY the three sources of hydrogen and the three sources of oxygen inside the primary containment. (201 .072.A.01)
62. Given 3 1EO-EOP-012-2, PC Primary Containment Control, IDENTIFY the two methods of determining hydrogen and oxygefl concentrations stated on the flow chart. (201.072.A.02)
63. Given plant conditions and. 3 1EO-PCG-00 1-2, Primary Containment Gas Control, EVALUATE hydrogen and oxygen concentrations in the primary containment and DETERMINE if actions for deflagration conditions should be taken. (201 .072.A. 15)
64. Given a list, RECOGNIZE the statement that describes the possible consequence of having a deflagration condition inside the primary containment. (201 .072.A. 16)
65. Given a list, IDENTIFY the statement that describes the undesirable result of continuing to vent and purge the primary containment if offsite release rate reaches 0.O57mRfhr. (201 .072.A.04)
66. Given a list, IDENTIFY the statement that describes the possible consequences of attempting to purge the primary containment without a vent path when hydrogen concentrations are high.

(201 .072.A. 10)

I Ic 7 1

I NOTE4 Plant Chemistry must supply hydrogen and oxygen concentrations for drywell and torus if monitors are unavailable 3%

0 Place hydrogen and oxygen analyzers in service per 34S0-P33-OO1-1. IF necessary, defeat isolation interlocks. Monitor and control containment pressure 0

below 1.85 psig using:

o containment pressure control system

.... WAIT UNTIL per 34SO-T48-002-1 o SBGT per 34S0-T46-OO1-1 dryweil Qtorus hydrogen concentration reaches 1.5% WAIT UNTIL OR cannot be determined primary containment pressure CANNOT be maintained

> below 1.85 psig Goto 31EO-PCG-OO1-1 point Q

jected Projected Peak TEDE IF torus water level is below 018-0 285 in.

(p, mR/hr in the TSC or EOF THEN initiate torus sprays per 34S0-E11-010-1, irrespective of adequate core cooling WHILE PERFORMING THE FOLLOWIN being vented venting through torus 31E0-EOP-104-1 D

4 DURRENTLY

> GO TO point P on this sheet talof I Projected E + P Total EDE > NO Is at or above_ torus water level

_ (T) mR/hr <300 in.

NO Vent torus per 31 EO-EOP-1 04-1 irrespective of offsite radioactivity release rate If necessary, defeat isolation interlocks

NO Is torus being vented 3

YES Vent drywell per 31 EO-EOP-1 04-1 irrespective of offsite radioactivity release rate If necessary, defeat isolation interlocks Vent torus per 31E0-EOP-104-1.

If necessary, defeat isolation interlocks

HLT-07 SRO NRC EXAM

64. 600000AK1.02 001 A fire is reported on Unit 1, in the lB Diesel Generator Room.

Which ONE of the choices below is the fire suppression system that is in this room?

A Automatically actuated Carbon Dioxide System with a manual backup B. Automatically actuated Halon System with a manual backup C. ONLY a manually actuated Carbon Dioxide System D. ONLY a manually actuated Halon System

Description:

The iF switchgear is located in the Diesel Generator Building. The lB DG is protected with an automatic CO2 system with a manual backup, while the switchgear is protected with a manually activated CO 2 hose reel. Halon is used in 4 areas of the plant but not in switchgear areas (i.e.

Class C, electrical fires).

The B distractor is plausible if the applicant confuses between the switchgear room and the DG room and the fire suppression agent for this area.

The C distractor is plausible if the applicant confuses between the switchgear room and the DG room and the fire suppression agent for this area.

The D distractor is plausible if the applicant confuses between the switchgear room and the DG room and the fire suppression agent for this area.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

194

HLT07 SRO NRC EXAM

References:

NONE K/A:

600000 Plant Fire On Site AK1 Knowledge of the operation applications of the following concepts as they apply to Plant Fire On Site:

AK1.02 Fire Fighting 2.9 3.1 LESSON PLAN/OBJECTIVE:

X43-FPS-LP-03601, Fire Protection, EO 036.020.B.11 References used to develop this question:

34S0-X43-005-O, DG Bldg C02 System (Section 7.2, Flooding A DG Room with C02) 34S0-X43-OO1-1, Fire Procedure (Section 6.4, DG Engine Rooms) 195

X43-FPS-LP-03601 Page 2 of 145 FIRE PROTECTION Initial License (LT) ENABLING OBJECTIVES

1. Given a list of statements, SELECT the statement that best describes the operation of the following types of Fire Suppression systems: (036.020.B.Ol)
a. Deluge
b. Fixed Water Spray
c. Wet Pipe Sprinklers
d. Dry Pipe Sprinklers
e. Preaction Sprinklers
f. CO (Cardox)
g. Halon 1301
2. Given a simplified drawing of a deluge system, IDENTIFY the following components:

(036.020.B.03)

a. Latch
b. Clapper Valve
c. Riser and Isolation valve
d. Air Diaphragm
e. Ball Drip Valve
f. Supervisory Water Line
3. Given a drawing of a deluge system, TRACE the operational flowpath for the system.

(036.020.B.04)

4. Given that a small stream of water is flowing from a deluge system ball drip valve, STATE the significance of this condition. (036.020.B.02)
5. For the areas protected by the CO 2 (Cardox) system, STATE if the system can be initiated automatically, semi automatically, or manually. (036.020.B.l 1)
6. IDENTIFY the particular odor used to identify the presence of CO

. (036.020.B. 12) 2

7. Given a list of plant areas, SELECT the areas protected by a Halon 1301 system (036.020.B.07)
8. STATE the location of the Halon 1301 system abort switches. (036.020.B.08)
9. Given a list of statements, SELECT the statement(s) describing the function of the following Fire Detection systems: (036.021 .A.03)
a. CXL system
b. XL3 Master panel
c. XL3 Slave panels
d. Protectowire system
e. Alison Control Detection system

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 6 OF 16 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR BUILDING CARBON DIOXIDE SYSTEM 34SO-X43-005-0 0.9 72 FLOODING A DIESEL GENERATOR ROOM WITH CO 2 IF AUTOMATIC OPERATION FAILS

. The Diesel Generator Building CO 2 system has enough capacity to provide at least 2 discharges of approximately 1 minute each.

NOTES:

. The following two steps are performed in the hallway outside the Diesel Generator Room containing the fire.

7.2.1 IF the RED ready light is ILLUMINATED, THEN depress AND hold appropriate START pushbutton switch UNTIL the RED light is EXTINGUISHED AND IMMEDIATELY exit the area. El 7.2.2 fl the RED ready light was NOT ILLUMINATED, OR IF CO 2 discharge does NOT occur, THEN place in OPEN applicable Diesel Generator Room CARDOX Pilot Control Valve. El 7.2.3 On the Diesel Generator Building West Wall, place in OPEN 1X43-P007, Master Pilot for Diesel Generator Bldg.

7.2.4 AFTER one minute, place in CLOSED 1X43-P007, Master Pilot for Diesel Generator Bldg. LI MGR-0001 Ver, 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH . 160F104 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

FIRE PROCEDURE 34AB-X43-001-1 11.1 6A DG ENGINE ROOMS 6.4.1 SHUT DOWN the affected diesel generator, AND DE-ENERGIZE associated electrical equipment as requested by fire brigade or to prevent spurious equipment operation.

Fire damage to control cables may cause spurious operation of the diesel fuel oil transfer NOTE: pumps which could result in a fuel oil spill. Monitor fuel oil transfer pump operation and trip pumps as necessary.

6.4.2 Turn the breaker OFF to DE-ENERGIZE the Diesel Fuel pumps in the affected room.

ROOM PUMPS DE-ENERGIZE AT DG BLDG DG 1A Diesel 1A Fuel Pumps 1Y52-COO1A 1R24-S025, frame 6D 1E SWGR RM 1Y52-C1O1A 1R24-S027, frame 6D 1G SWGR RM DG lB Diesel lB Fuel Pumps 1Y52-C001 B 1 R24-S025, frame 7D 1 E SWGR RM 1Y52-C1O1B 1R24-S027, frame 7D 1G SWGR RM DG 10 Diesel 10 Fuel Pumps 1 Y52-0001 C 1 R24-S025, frame 7F I E SWGR RM 1Y52-C1O1C 1R24-S027, frame 7F 1G SWGR RM 6.4.3 IF necessary to remove 002 OR smoke from the Diesel Generator Rooms, PERFORM the following, as directed by the Fire Brigade Leader:

6.4.3.1 Provide an air supply path from outside air.

6.4.3.2 START 002 purge operation using 002 purge switch located at appropriate panel:

DIG PANEL LOCATION 1A 1 H21-P260 Switchgear Room 1 E lB 1H21-P261 Switchgear Room iF 10 1H21-P262 Switchgear Room 1G MGR-0001 Ver. 4

HLT-07 SRO NRC EXAM

65. 700000AK2.07 001 BOTH units are operating at 100% power.

The Northern Control Center has notified the control room crews that 230 KV Bus voltage cannot be maintained above the normal minimum voltage.

Current 416OVAC Emergency Bus voltages are 3800 VAC.

Which ONE of the choices below completes the following statements?

To RAISE reactive load (VARS) from the HMI Screen, the operator will select lAW 34AB-Sl 1-001-0, Operation With Degraded System Voltage, if after 30 minutes the 416OVAC Emergency Bus voltages have not changed, a TOTAL of__________ 416OVAC Emergency Buses will be transferred to their respective Diesel Generator.

A EX2 100 then REGULATOR ADJUST and depress RAISE; two (2)

B. EX2100 then REGULATOR ADJUST and depress RAISE; four (4)

C. PSI-LOAD then LOAD SET and depress RAISE; two (2)

D. PSI-LOAD then LOAD SET and depress RAISE; four (4) 196

HLT-07 SRO NRC EXAM

==

Description:==

JAW 34S0-N40-001-112, Main Generator:

7.1.4.4 IF the System Operator at GCC requests Generator voltage to be adjusted, perform one of the following:

7.1.4.4.2 To use the Mark VI HMI EX 2100 Screen:

From 2N32-K4001A OR 2N32-K4001B Static Exciter 2N51-P4001 by selecting

    • Control** **EX2100**

Under EX2100 VOLTAGE ADJUST select **RAISE** to raise voltage, OR Under EX2 100 VOLTAGE ADJUST select **LOWER** to lower voltage, as requested by the GCC System Operator.

34AB-S 11-001-2, Operation With Degraded System Voltage, states that if voltages are

<3825VAC for 30 minutes, then the 1A and 2A Diesel Generators will be supplying their respective emergency buses.

The B distractor is plausible since the first part is correct. The second part is plausible if the applicant remembers that 2 EDGs are supplying power and confuses this and thinks 2 EDGs for each unit will be started for a total of 4.

The C distractor is plausible if the applicant confuses the operation of voltage adjustments with the operation of load set and thinks that raising load set will raise reactive load. The second part is correct.

The D distractor is plausible if the applicant confuses the operation of voltage adjustments with the operation of load set and thinks that raising load set will raise reactive load. The second part is plausible if the applicant remembers that 2 EDGs are supplying power and confuses this and thinks 2 EDGs for each unit will be started for a total of 4.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

References:

NONE K/A:

700000 Generator Voltage and Electric Grid Disturbances 197

HLT-07 SRO NRC EXAM AK2. Knowledge of the interrelations between GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES and the following:

(CFR: 41.4, 41.5, 41.7, 41.10 / 45.8)

AK2.07 Turbine/generator control 3.6 3.7 LESSON PLAN/OBJECTIVE:

S11-LP-02706, Basic Grid Operating Concepts, EO 200.116.A.04 References used to develop this question:

34SO-N40-001-l/2, Main Generator 34AB-S 11-001-0, Operation With Degraded System Voltage Modified from HLT-5 NRC Exam Q#65 ORIGINAL QUESTION (IILT-5 NRC Exam Q#65)

Both units are operating at 100% power. On panel 1H1 1-P653, the 230 KV voltmeter 1S40-R600 indicates 228 KV and is fluctuating +2 Ky.

The NCC has notified the control room crews that 230 KV Bus voltage cannot be maintained above the normal minimum voltage.

Which ONE of the following identifies how reactive load can be raised AND a required action lAW 34AB-Sl 1-001-0, Operation With Degraded System Voltage?

A.V Placing the REGULATOR VOLTAGE ADJUST control switch to the RAISE position; Initiate a one hour Required Action Statement.

B. Selecting LOAD SET and then RAISE at the HMI Screen; Initiate a one hour Required Action Statement.

C. Placing the REGULATOR VOLTAGE ADJUST control switch to the RAISE position; Transfer Station Service Busses to their alternate supply.

D. Selecting LOAD SET and then RAISE at the HMI Screen; Transfer Station Service Busses to their alternate supply.

198

S11-LP-02706-02 Page 1 of 102 BASIC GRID OPERATING CONCEPTS OBJECTIVES TERMINAL OBJECTIVES 200.1 16.A RECOGNIZE and RESPOND to a degraded voltage condition per 34AB-S1 1-001-0, Operation with a Degraded System Voltage.

ENABLING OBJECTIVES

1. Given plant conditions and 34AB-S 11-001-0, DETERMINE if the minimum voltage requirements are being met for the Plant E. I. Hatch: (200.1 16.A.0l)
a. 23 0KV switchyard
b. 416OVAC Emergency Buses
2. Given a list, select the definition of the following terms as they are used in association with operation of the grid: (200.11 6.A.02)
a. Contingency
b. System Tools
3. Given plant and grid conditions, PREDICT plant response to a degraded grid condition if no Operator actions are taken. (200.l16.A.03)
4. Given plant conditions and 34AB-S 11-001-0, DETERMINE the required Operator actions when the following reports are made to the Control Room: (200.11 6.A.04)
a. The grid Security Tools are unavailable for >8 hours,
b. The offsite system is one contingency away from being unable to maintain normal minimum voltage to the 230KV bus
c. The offsite system cannot be maintained above the normal minimum voltage.
d. The 4160 VAC emergencybuses cannot be maintained above 3825 VAC.
5. Given a list of options, DETERMINE why a shutdown is required if voltage cannot be maintained above 233 KV in the 230KV switchyard or above 3825 VAC on the 4160 VAC Emergency buses (200.1 l6.A.06)
6. Given plant and grid conditions, DETERMINE the significance of being One contingency away from being able to maintain normal minimum voltage on the 230KV bus. (200.116.A.05)

Not Selected for Requal

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH .

30F6 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

OPERATION WITH DEGRADED SYSTEM VOLTAGE 34AB-S1 1-001-0 3.0 4.4 Upon notification from GCC that the 230KV bus voltage CANNOT be maintained above normal minimum voltage, OR H the 4160 VAC bus voltages CANNOT be maintained above 3825VAC (Tech Spec),

the following action will be taken:

4.4.1 Initiate a One Hour RAS to restore the 416OVAC Bus voltages to acceptable levels (greater than Q, equal to 3825VAC).

4.4.2 Notify the following:

  • Manager of Operations LI
  • On-call Hatch Project Duty Manager LI Sustained low voltage conditions may cause the 4160 VAC emergency busses to trip (LOSP NOTE: degraded voltage relaying). The following section transfers one bus to its associated diesel.

The 1 E(2E) bus was chosen because of the plant impact of losing loads on 600 VAC bus 1 C(2C).

4.4.3 H the 4160 VAC Bus voltages are jQ RESTORED to acceptable levels WITHIN 30 minutes, perform the following to maintain 41 60V I E emergency bus voltage.

(Two handed operations will be necessary):

4.4.3.1 Start the I R43-SOO1A DIG, using the start switch, panel 1 Hi i-P652. LI 4.4.3.1.1 Override 1 P41-F31 OA AND 1 P41-F31OD, per 34AB-P4i-OOi-1.

4.4.3.1 .2 Open AND hold the following control switches for 1 R22-S005, 41 60V I E Bus UNTIL the emergency supply breaker closes:

  • ACB 135712, Normal Supply, 4160V Bus 1E LI
  • ACB 135711, Alternate Supply, 4160 V Bus 1 E. LI 4.4.3.1.3 Load 1A D/G as necessary AND perform applicable abnormal procedures for:
  • loss of 4160 V emergency busses
  • loss of 600V emergency busses LI
  • loss of essential busses LI
  • loss of instrument busses
  • loss of RPS busses LI 4.4.3.1.4 Reset 4160V bus iE LOSP LOCKOUT (86) Relay.

4.4.3.1.5 Place the Overrides for 1 P4i-F31OA AND 1 P41-F31 OD in NORMAL.

MGR-0001 Ver. 4

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 4 OF 6 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

OPERATION WITH DEGRADED SYSTEM VOLTAGE 34AB-S1 1-001-0 3.0 4.4.3.2 Start the 2R43-SOOIA DIG, using the start switch, panel 2H1 1-P652. LI 4.4.3.2.1 Override 2P41-F31 6A AND 2P41-F316D per 34AB-P41-001-2. LI 4.4.3.2.2 Open AND hold the following control switches for 2R22-S005, 41 60V 2E Bus UNTIL the emergency supply breaker closes:

  • ACB 135554 Unit 2, Normal Supply, 4160V Bus 2E LI
  • ACB 135544 Unit 2, Alternate Supply, 4160V Bus 2E. LI 4.4.3.2.3 Load 2A D/G as necessary perform applicable abnormal procedures for:
  • loss of 4l60Vemergency busses LI
  • loss of 600V emergency busses LI
  • loss of essential busses LI
  • loss of instrument busses LI
  • loss of RPS busses LI 4.4.3.2.4 Reset 41 60V bus 2E LOSP LOCKOUT (86) Relay. LI 4.4.3.2.5 Place the Overrides for 2P4 1 -F31 6A AND 2P41 -F31 6D in NORMAL. LI 4.4.4 IF the 416OVAC Bus voltages are NOT restored to acceptable levels WITHIN one hour, an orderly plant SHUTDOWN will be initiated with the intent of reaching MODE 2 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, MODE 3 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> AND MODE 4 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.

Refer To NMP-EP-110, Emergency Classification Determination and Initial Action. LI 4.5 IF ANY of the conditions in 4.4 are met, it may be necessary to ENTER into and perform the following:

  • Enter 34G0-OPS-013-2, Normal Plant Shutdown LI 4.6 During a degraded system voltage condition, it may become necessary to enter several procedures, as well as those listed in 4.5.

Follow the ARPs AND take the actions necessary to mitigate any transient. LI 4.7 See attachment 1 for list of essential equipment affected by degraded voltage. LI MGR-0001 Ver. 4

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 16 OF 62 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

MAIN GENERATOR OPERATION 34SO-N40-001-1 14.0

. Once the Main Generator is tied, the load will automatically increase to 2% ( 20 MWe) to prevent motoring the generator. Because of this feature, there is no need to immediately increase the Load Set setpoint above 0 (zero) WHEN INITIALLY tying the generator.

NOTES:

  • Load set is in percent (%) NOT MWe; 1% 6.5 Mwe
  • The number 1, 2, and 3 control valves will open first controlling load. WHEN No. 1, 2, and 3 reach approximately 60% open, No. 4 control valve will begin opening AND THEN it will become the load control valve.

. THE HEATUPORCOOLDOWN RATE FORTHE ISTSTAGE INNERSURFACE TEMPERATURE MUST BE < 150F/HR.

. THE HEATUP OR COOLDOWN RATE FOR THE HOT REHEAT STEAM TO THE LP TURBINES MUST < 125FIHR.

CAUTIONS:

  • EXTENDED OPERATION BELOW 120 MWE CAN RESULT IN THE FIRST STAGE SHELL TEMPERATURE DECREASING BELOW 250F AND THE LP PREWARMING TEMPERATURE DECREASING BELOW 120F. SEE ATTACHMENT 2 OF 34SO-N30-001-2 FOR THE APPROXIMATE TEMPERATURE AT DIFFERENT LOAD CONDITIONS.

7.1.3.2.6 Set Ramp Rate to> 0 but 1%! mm by performing the following:

  • Enter Control I El
  • Enter psi loadi LI
  • In Load Set box, enter IRamp Rate7
  • Enter an amount between 0 (zero) and 1. LI 7.1.3.2.7 While increasing load, adhere to the temperature limitations in the CAUTION above.

Increase load to 20% (120 MWe) by performing the following:

  • In Load Set box, enter I Load
  • Enter 20.

7.1.3.3 Adjust generator reactive load for zero megavars, as indicated on the Generator Megavar meter using the REGULATOR ADJUST control switch.

MGR-0001 Rev 4

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH . 200F77 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

MAIN GENERATOR OPERATION 34SO-N40-001-2 17.1 7.1.3.2.7 While increasing load, adhere to the temperature limitations in the CAUTION above.

Increase load to 20% (1 20 MWe) by performing the following:

  • In Load Set box, enter I Load I LI
  • Enter 20. LI I NOTE: Subsequent load increases are QI to exceed 10 MWe/minute.
  • Additional load increases to be done in accordance with 34GO-OPS-001-2 and 34GO-OPS-005-2. LI 7.1.3.3 Adjust generator reactive load, as indicated on the Generator Megavar meter for zero megavars, using the REGULATOR ADJUSTcontrol switch.

7.1.3.4 Place the SSW for the breaker just closed, PCB 179740 OR PCB 179750, in the OFF position. LI 7.1.3.5 Place the Auto SSW in MAN position. LI At the direction of the SS the second output breaker 179740 OR 179750 may be left open to support work activity OR testing.

NOTE: Notify the System Operator IF the output breaker will NOT be closed at this time.

WHEN it is decided to close the second output breaker return to this step AND follow these instruction.

7.1.3.6 Place the SSW for the open breaker, PCR 179740 OR 179750, in the ON position. LI 7.1.3.7 Close the open breaker, PCB 179740 OR PCB 179750. LI 7.1.3.8 Place the SSW for the breaker just closed, PCB 179740 OR PCB 179750, in the OFF position.

7.1.3.9 Notify the System Operator at GCC that the generator is on the line. LI 7.1.3.10 WHEN generator output is at 120 MWe, verify Control

  • psi-load] I Load Set I Ramp Rate Is between 0 (zero) AND 1. LI MGR-0001 Ver. 4

HLT-O7 SRO NRC EXAM

66. G2.1.2 001 Unit 2 experiences a Loss of ALL High pressure feed.

o LPCI and Core Spray (CS) automatically initiate o RWL is being controlled by the OATC using LPCI and CS RWL is recovered to its normal band and the OATC is ready to depress the auto initiation signal reset pushbutton for RHR.

JAW 31G0-OPS-021-O, Manipulation Of Controls And Equipment, which ONE of the choices below completes the following statement?

The DOES NOT have the AUTHORITY to procedurally authorize resetting the LPCI initiation signal under these conditions.

NOTE: Operator At The Controls = OATC; Balance Of Plant = BOP A. Shift Manager B OATC on Unit 2 C. BOP NPO on Unit 2 D. Extra NPO on Unit 2 199

HLT-07 SRO NRC EXAM

==

Description:==

3lGO-OPS-02l-O, Manipulation Of Controls And Equipment, states the following limitation:

5.2.2 Resetting an auto start signal requires the authorization of an Reactor Operator (RO) or Senior Reactor Operator (SRO).

The authorizing individual will NOT be the individual performing the actual manipulations on the control boards.

Since the OATC is going to be the one resetting the LPCI signal, they cannot be the one to authorize the reset. All of the other individuals can provide this authorization.

The A distractor is plausible if the applicant does not remember the procedure requirment and thinks this individual can not authorize the reset.

The C distractor is plausible if the applicant does not remember the procedure requirment and thinks this individual can not authorize the reset.

The D distractor is plausible if the applicant does not remember the procedure requirment and thinks this individual can not authorize the reset.

A. Incorrect See description above.

B. Correct See description above.

C. Incorrect See description above.

D. Incorrect See description above.

200

HLT-07 SRO NRC EXAM

References:

NONE K/A:

2.1 Conduct of Operations 2.1.2 Knowledge of operator responsibilities during all modes of plant operation.

(CFR:41.l0/45.13) 4.1 4.4 LESSON PLAN/OBJECTIVE:

El 1-RHR-LP-00701, Residual Heat Removal System, EU 006.004.A.0l References used to develop this question:

31G0-OPS-021-0, Manipulation Of Controls And Equipment 201

E11-RHR-LP-00701 Page 6 of 130 RESIDUAL HEAT REMOVAL SYSTEM

29. From a list of statements, SELECT the statement which describes the interlock between RWL and the Inboard Spray Isolation Valves. (007.001.a.03)
30. From a list of statements, SELECT the statement that describes the interlocks associated with the LPCI Outboard Injection Valve. (006.002.a.01)
31. From a list of statements, SELECT the statement that describes the interlocks associated with the LPCI Inboard Injection Valve. (006.002.a.02)
32. Given a list of plant conditions, IDENTIFY the condition which will cause an isolation of 2E1 1-FO15A/B LPCI Inboard Injection Valve if in SDC. (006.008.a.04)
33. From a list of statements, SELECT the statement that describes the actions required to start a RHR pump after the pump was secured with an initiation signal present. (006.007.a.05)
34. Given an RHR pump has been manually secured with a LOCA signal present, STATE the system response if a loss of 125 VDC Logic Power occurs after the RHR pump has been secured.

(006.007.a.04)

35. From a list of statements, SELECT the statement which describes the equipment operating location and indication in the following conditions: (007.019.a.01)
a. The emergency transfer switch at the RSDP is in NORMAL.
b. The emergency transfer switch at the RSDP is in EMERGENCY
36. Given a list, IDENTIFY the three interlocks that prevent draining of the RPV to the Torus which are still in effect when control is transferred to the RSDP. (007.01 9.a.02)
37. Given the RHR system is in Shutdown Cooling, IDENTIFY the system response to decreasing reactor water level, including any manual actions required to place the system in the LPCI mode.

(006.008.a.05)

  • 39 STATE the conditions which must be met prior to securing LPCI following an Auto-Initiation per 34S0-El 1-010-1/2, ResidualHeat Removal System. (006.004.a.0l)
  • 40. Given 34S0-E1 1-010-1, IDENTIFY IEI I-F053A/B valve position. (007.007.a.08)
  • 41. With Control of RHR transferred to the RSDP, while in the STBY lineup; DESCRIBE how RHR/LPCI mode would respond to a given plant condition. (007.019.a.03)

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 3 OF 14 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

MANIPULATION OF CONTROLS AND EQUIPMENT 31GO-OPS-021-0 24 5.0 PRECAUTIONSILIMITATIONS 5.1 PRECAUTIONS N/A Not applicable to this procedure 5.2 LIMITATIONS 5.2.1 With the exception of routine Recirc Flow adjustments and weekly control rod exercises provided for in 34GO-OPS-022-0, MAINTAINING RATED THERMAL POWER and 34SV-C1 1-003-1/34SV-C1 1-003-2, WEEKLY CONTROL ROD EXERCISE, ALL planned reactivity changes MUST be reviewed and approved by the Shift Manager(SM).

Additionally, the Shift Supervisor (SS) must be made aware of all planned reactivity changes, PRIOR to their performance, with the above noted exceptions 52.2 Resetting an auto start signal requires the authorization of an Reactor Operator (RO) or Senior Reactor Operator (SRO).

The authorizing individual will NOT be the individual performing the actual manipulations on the control boards.

5.2.3 Control of plant components and systems is maintained by Operations per NMP-AD-003, Equipment Clearance and Tagging. The types of controls established shall cover, as a minimum, the following:

5.2.3.1 Control of removal from service and return to service of Plant components and systems.

5.2.3.2 Control of Temporary Modifications per 4OAC-ENG-01 8-0, Temporary Modification Control.

5.2.3.3 Control of keys for keyed bypasses, Reactor Mode Switch, DIG Mode Switches, reactor safety systems, and locked valves.

5.2.3.4 Control of startup and shutdown of safety related systems.

5.2.3.5 Control of the position/condition of all plant components. This includes control of components that must be changed from their normal position by use of Danger or Caution Tags per NMP-AD-003, Equipment Clearance and Tagging.

MGR-0001 Ver. 4

HLT-07 SRO NRC EXAM

67. G2.1.21 001 Southern Nuclear Operating Company I Nuclear NMP-OS-007 SOUTHERNK... Management Conduct of Operations Version 8X3 Procedure Page 1 of 51 Peer Team ChampionlProcedure Owner: David R\/inevard/ Fleet Operations Manaaer/ Corporate (Print: Name / Title I Site)

Approved By: Original Signed by David Vineyard / 12/08/2011 (Peer Team Champion/Procedure Owners Approval Signature) (Approval Date)

Effective Dates: NA 12/14/2011 1 2/1 420 1 1 1 2/I 4i20I 1 Corporate FNP HNP /EG P PROCEDURE LEVEL OF USE CLASSIFICATION PER NMP-AP-003 CATEGORY SECTIONS Continuous: NONE

Reference:

NONE Information: ALL Based on the above cover sheet, which ONE of the choices below completes the following statements?

JAW NMP-AP-001, Development and Control of Southern Nuclear Procedures, on 12/10/2011 this version of NMP-OS-007 the CORRECT version to be used by Operations.

For the indicated PROCEDURE LEVEL OF USE, placekeeping is A. was; required B. was; allowed, but NOT required C. was NOT; required D was NOT; allowed, but NOT required 202

HLT-07 SRO NRC EXAM

==

Description:==

Edwin, this was question 5 of 10 that you have already reviewed. Any discussed changes have been incorporated.

NMP-AP-001, Development and Control of Southern Nuclear Procedures, step 3.6 states Effective Date The date upon which a procedure will take effect and shall be used to perform work.

Step 3.20.6 states Procedure Owner Review A final review by the procedure owner. The purpose of the procedure owner review is for the procedure owner to verify the fleet procedure promotes simple, standard processes that are consistent with industry best practices and ensure adequate expectations, communications plans, implementation plans, and measures are in place.

The procedure owners signature on the appropriate Procedure Approval Form indicates compliance with these requirements.

NMP-AP-003, Procedure and Work Instruction Use and Adherence, step 5.7.1 states Place-keeping is used when a Continuous Use or Reference Use document is used to perform an activity. It may also be used with other procedures and documents. Since this procedure level of use is Information, then placekeeping is allowed but not required.

The A distractor is plausible if the applicant does not remember the meaning or confuses the requirements of this procedure. With two approval dates the applicant could confuse which procedure version is required to be used. The second part is plausible if the applicant does not remember/confuses the requirments of NMP-AP-003, Procedure and Work Instruction Use and Adherence, which requires placekeeping for Continuous & Reference Use procedures but not for Information Use. Placekeeping is allowed but not required.

The B distractor is plausible if the applicant does not remember the meaning or confuses the requirements of this procedure. With two approval dates the applicant could confuse which procedure version is required to be used. The second part is correct.

The C distractor is plausible since the first part is correct. The second part is plausible if the applicant does not remember/confuses the requirments of NMP-AP-003, Procedure and Work Instruction Use and Adherence, which requires placekeeping for Continuous & Reference Use procedures but not for Information Use. Placekeeping is allowed but not required.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

203

HLT-07 SRO NRC EXAM

References:

NONE K/A:

2.1 Conduct of Operations 2.1.21 Ability to verify the controlled procedure copy.

(CFR: 41.10/45.10/45.13) 35* 3.6*

LESSON PLAN/OBJECTIVE:

LT-LP--30007, Shift Operations And Evolutions, EO 500.003.A.02 References used to develop this question:

NMP-AP-001, Development and Control of Southern Nuclear Procedures NMP-AP-003, Procedure and Work Instruction Use and Adherence 204

LT-LP-30007-07 Page 1 of 58 SHIFT OPERATIONS AND EVOLUTIONS OBJECTIVES TERMINAL OBJECTIVES 500.003.A Given that a specific operational evolution is to be performed and given the following:

a. Applicable Control Room system indication
b. Applicable Tech Specs and references STATE whether the operational evolution can be performed without violating Tech Specs, station procedures and good engineering practices.

500.003.B Given that a specified operational evolution is to be performed, STATE what actions must be taken by the SRO to ensure that the operational evolution is successfully completed.

ENABLING OBJECTIVES

1. Given plant status, EVALUATE plant conditions and determine if continued operation is permitted.

STATE factors to be considered in making this determination. (500.003.A.03)

2. Given plant procedures and a specified operational evolution, DETERMTNE which procedure(s) and data package(s) you would utilize to ensure administrative and regulatory compliance.

(500.003.A.Ol)

3. Given a complete description of an operational evolution and the current plant parameters, STATE what systems directly interface with the component or system which you will be operating.

(500.003.B.02)

4. Given the procedure that you would use to implement an operational evolution, PERFORM a pre evolution brief completing all appropriate steps of NMP-GM-005-GLO3. (500.003.A.02)
5. Given conditions requiring a procedure change in the field, STATE the requirements that must be meet for a procedure Pen & Ink change to be processed. (500.003.A.04)
6. Given procedure DI-OPS-07-0685, Operations Night Orders or 31 GO-OPS-Ol 5-0, Operating Orders STATE how short-term information, such as night orders or operating orders are managed.

(500.003.A.05)

7. Given procedure 8OAC-SEC-OOl-O, Plant Security and Entry Control, IDENTIFY the facility requirements for controlling access to various locations on company property. (500.003.B.03)

Southern Nuclear Operating Company Nuclear I Development and Control of Southern NMP-AP-001 SOUTHERNA& Management Version 13.0 i COMPANY Procedure Nuclear Procedures i Page 4 of 32 1.0 Purpose The purpose of this procedure is to establish the process for developing, revising, and approving Southern Nuclear corporate procedures, site procedures, Nuclear Management procedures, and vendor/contractor procedures utilized at Southern Nuclear Facilities.

2.0 Applicability 2.1 Throughout this series of procedures, the reference to procedure will include procedures, instructions, guidelines and forms. This procedure establishes the process for developing, revising, and approving Southern Nuclear corporate procedures, site procedures, Nuclear Management procedures (NMPs), and vendor/contractor procedures utilized at Southern Nuclear Facilities.

2.2 This procedure is not applicable to the development, maintenance, and approval of procedures for Nuclear Development (ND).

2.3 The use of a form within the NMP-AP-001 series is acceptable as long as the procedure to which it is attached is not revised after the issue date of the new version to that form. This allows changes that are in the approval process to continue in that process uninterrupted unless otherwise required by the Procedures Corporate Functional Area Manager (CFAM).

2.4 This procedure also governs the development and control of procedures associated with developing, revising, and approving Operational Readiness procedures.

3.0 Definitions 3.1 Administrative Procedures The procedures that provide rules, instructions, policies, and practices that affect fleet employees. They describe organization, staff responsibilities, and qualifications and establish an administrative controls procedural program.

3.2 Corporate Procedures The procedures for those activities that are subject to the requirements of the SNC Quality Assurance Topical Report (QATR) performed by departments within the Corporate Southern Nuclear (CS) organization. Refer to NMP-AP-001-002, Review and Approval of Corporate Procedures.

3.3 Change of Intent A change in what the procedure does and how it does it. Refer to Attachment 3, Change of Intent.

3.4 Document Management System (DMS) The system used by Southern Nuclear sites to control content management of procedures / version support documents.

3.5 Editorial Change In order to be considered an editorial change, the proposed change must fit one of the examples listed in Attachment 1, Editorial Changes. The editorial change process is restricted to those examples, and NO other changes can be made using this process.

3.6 Effective Date . The date upon which a procedure will take effect and shall be used to perform work.

Printed April 23, 2012 at 12:23

Southern Nuclear Operating Company Nuclear I I NMP-AP 003 Procedure and Work Instruction SOUTHERN Management Version .0 OMPAN Procedure Use and Adherence Page 14of21 5.7 Place-keeping 5.7.1 Place-keeping is used when a Continuous Use or Reference Use document is used to perform an activity. It may also be used with other procedures and documents.

5.7.2 Place-keeping shall be used for notes, cautions and warnings for both Continuous and Reference Use procedures.

5.7.3 Place-keeping in Continuous Use documents shall be performed for all steps, sub-steps, and actions, including list bullets, prerequisites, and precautions and limitations.

NOTE Step 5.7.4 is applicable to notes, cautions and warnings.

5.7,4 For Continuous Use documents, each step is marked as completed before proceeding to the next step, except for certain circumstances as follows:

  • The approved document provides instructions for performing a step (or series of related steps) concurrently.
  • Concurrent actions are necessary for the document.
  • ALARA, equipment, or personal safety may be compromised if step-by-step documentation of the procedure is performed during execution.
  • jf place-keeping boxes, brackets or signoffs are t4QI present, perform circle/slash described in section 5.7.7.

5.7.5 When any of the exceptions to place-keeping listed above are met, the following shall apply:

  • The series of steps is clearly identified prior to performance.
  • All personnel involved in the evolution (document user, verifier or peer-checker, supervisor, and so forth) review and understand all steps in the series prior to performance.
  • The document user reviews and place-keeps the steps after completion to ensure the steps were performed correctly.

5.7.6 If the document requires that an individual initials or signs steps, the signing or initialing of steps constitutes place-keeping.

Printed April25, 2012 at 12:04

HLT-07 SRO NRC EXAM

68. G2.1.6 001 Which ONE of the following identifies the required communication protocol during transients lAW DI-OPS-59-896, Operations Management Expectations, Attachment 5, Strategies For Successful Transient Mitigation?

An annunciator check required prior to the SS giving a crew brief.

If during a Crew Brief, alarms 603-205 & 603-206, Reactor Vessel Level 2 Div I & Div II Trip, are received, the NPO expected to call a Time Out.

A is; is B. is; is NOT C. is NOT; is D. is NOT; is NOT 205

HLT-07 SRO NRC EXAM

==

Description:==

DI-OPS-59-0896, Operations Management Expectations states in Attachment 5 section 4.1 for Briefs and Updates that Briefs are a tool used by the Shift Supervisor (or Shift Manager in rare cases) to bring all crew members to the same level of knowledge regarding plant status.

Step 4.1.1 for Briefs states prior to initiating a crew brief, all front panel annunciators will be acknowledged The SRO announces: Attention Crew, Annunciator Check.

Step 4.1.1.1.4 for Briefs states Panel and parameter monitoring is expected to continue in the normal manner during the brief. It is expected that front panel alarms be acknowledged during briefs. The RO is expected to call a Time Out and notify the SS for alarms related to critical or key parameters. The SS will prioritize the alarm and either continue or end the brief as appropriate. 603-205 & 603-206 are key parameter alarms.

The B distractor is plausible since the first part is correct The second part is plausible if the applicant remembers that all unexcused crew members must participate in the brief and since all of the front panel alarms must be acknowledged prior to the brief that any alarms that come in later will be addressed after the brief. Also plausible if the applicant thinks the SS will terminate the brief without any information from the NPO.

The C distractor is plausible if the applicant confuses the Crew Brief requirements for a preceding annuciator check with Crew Updates which do not require an annunciator check. The second part is correct.

The D distractor is plausible if the applicant confuses the Crew Brief requirements for a preceding annuciator check with Crew Updates which do not require an annunciator check. The second part is plausible if the applicant remembers that all unexcused crew members must participate in the brief and since all of the front panel alarms must be acknowledged prior to the brief that any alarms that come in later will be addressed after the brief. Also plausible if the applicant thinks the SS will terminate the brief without any information from the NPO.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

206

HLT-07 SRO NRC EXAM

References:

NONE K/A:

2.1 Conduct of Operations 2.1.6 Ability to manage the control room crew during plant transients.

(CFR: 41.10 / 43.5 / 45.12 /45.13) 3.8* 4.8 LESSON PLAN/OBJECTIVE:

References used to develop this question:

DI-OPS-59-896, Operations Management Expectations, Attachment 5, Strategies For Successful Transient Mitigation Modified from HLT-5 NRC Exam Q#66 ORIGINAL QUESTION (HLT5 NRC EXAM Q#66)

Which ONE of the following identifies the required communication protocol during transients in accordance with DI-OPS-59-896, Operations Management Expectations, Attachment 5, Strategies For Successful Transient Mitigation?

A. A crew update is required to be preceded by an annunciator check. A crew update is a tool used primarily by the Shift Supervisor.

B. A crew update is required to be preceded by an annunciator check. A crew update is a short transfer of information from anyone on the crew to the rest of the crew.

C.V A crew brief is required to be preceded by an annunciator check. A crew brief is a tool used primarily by the Shift Supervisor.

D. A crew brief is required to be preceded by an annunciator check. A crew brief is a short transfer of information from anyone on the crew to the rest of the crew.

207

SNC PLANT E. I. HATCH I Pg 45 of 53 DOCUMENT TITLE: DOCUMENT NUMBER: Ver No:

OPERATIONS MANAGEMENT EXPECTATIONS Dl-OPS-59-0896 68.0 ATTACHMENT Att. Pg.

TITLE: STRATEGIES FOR SUCCESSFUL TRANSIENT MITIGATION 4 of 12 4 BRIEFS AND UPDATES 4.1 Brief Guidelines

4.1.1 Brief

a tool used by the SS (or SM in rare cases) to share information with the crew. The goal is to bring all crew members to the same level of knowledge regarding plant status and completed/expected crew actions. A good brief shares important information at the right time, in a concise fashion.

4.1.1.1 Briefs are usually less than 3 minutes in duration and are conducted as follows:

4.1.1.1.1 Prior to initiating a crew brief, all front panel annunciators will be acknowledged.

Back panel annunciators can be acknowledged if time allows, but this is not required. The SRO announces: ATTENTION CREW, ANNUNCIATOR CHECK.

4.1.1.1.2 SS/SM announces ATTENTION CREW, BRIEF. Each crew member will signify readiness by raising their hand.

4.1.1.1.3 If an individual is to be excluded from a brief, the SS/SM will verbally excuse the individual. The SS/SM is responsible for briefing the excluded individual as soon as practical.

4.1.1.1.4 Panel and parameter monitoring is expected to continue in the normal manner during the brief. It is expected that front panel alarms be acknowledged during briefs. The RO is expected to call a Time Out and notify the SS for alarms related to critical or key parameters. The SS will prioritize the alarm and either continue or end the brief as appropriate. The SS must authorize actions during a brief.

4.1.1.1.5 Brief Template

  • Announce ATTENTION CREW, BRIEF.
  • Current Plant Status (How did we get here?)
  • Person in charge of Critical or Key Parameter report. (Where are we now?)
  • Priorities (What is currently most important.)
  • Contingencies (What happens if...)
  • Goals (Where we are headed)
  • Ask the crew members for any additional input or questions.
  • Announce END OF BRIEF MGR-0009 Rev 5

HLT-07 SRO NRC EXAM

69. G2.2.23 001 Which ONE of the choices below identifies a Unit 1 Tech Spec LCO condition that will require entry into a Required Action Statement (RAS)? (NOT A TRACKING RAS)

A. Torus water level is 149.5 inches in Mode 1.

B. RCIC is mop in Mode 3 with reactor steam dome pressure at 100 psig.

C. Reactor steam dome pressure is 1052 psig in Mode 1.

D Drywell pressure is 1.80 psig in Mode 2.

Description:

JAW TS LCO 3.6.1.4, Drywell Pressure, drywell pressure must be < 1.75 psig in Mode 2.

This will require the SRO to initiate a Real RAS, not a Tracking.

The A distractor is plausible and would be a correct answer if Torus level was >150 or <146 JAW TS LCO 3.6.2.2, therefore this is a tracking RAS.

The B distractor is plausible and would be a correct answer if reactor steam dome pressure was

> 150 psig JAW TS LCO 3.5.3, Condition A, therefore this is a tracking RAS.

The C distractor is plausible and would be correct if reactor steam dome pressure was> 1058 psig JAW TS LCO 3.4.10, therefore this is a tracking RAS.

A. Incorrect See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Correct See description above.

208

HLT-07 SRO NRC EXAM

References:

NONE K/A:

2.2 Equipment Control 2.2.23 Ability to track Technical Specification limiting conditions for operations.

(CFR: 41.10 /43.2 / 45.13) 3.1 4.6 LESSON PLAN/OBJECTIVE:

LT-LP-30005, Technical Specifications, EO 300.006.A. 26 References used to develop this question:

Unit 1 Tech Specs LCO 3.4.10, 3.5.3, 3.6.1.4 & 3.6.2.2 209

LT-LP-30005 Page 3 of 86 TECHNICAL SPECIFICATIONS

26. IDENTIFY the rules for extending SR frequencies and repeating required action completion times (SR 3.0.2). (300.006.A.24) (SRO Only)
27. IDENTIFY the rules for a condition where an SR is discovered to have been missed (SR 3.0.3).

(300.006.A.25) (SRO Only)

28. STATE the actions required for a safety limit violation per Technical Specifications.

(300.003.A.03)

29. Given a copy of Technical Specifications and a set of plant conditions, DETERMINE the actions required if the conditions do not comply with the LCO. (300.0 10.A.06) (SRO Only)
30. IDENTIFY the Technical Specification knowledge requirements, in accordance with NUREG 1021 and Attachment 1 of this Lesson Plan, for a Reactor Qperator Candidate. (300.006.A.26)
31. IDENTIFY the Technical Specification knowledge requirements, in accordance with NUREG 1021 and Attachment 1 of this Lesson Plan, for a Senior Reactor Operator Candidate.

(300.006.A.27) (SRO Only)

Not Selected for Requal

Reactor Steam Dome Pressure 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Reactor Steam Dome Pressure LCD 3.4.10 The reactor steam dome pressure shall be S 1058 psig.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor steam dome A.1 Restore reactor steam 15 minutes pressure not within limit, dome pressure to within limit.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.

SURVEILLANCE_REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify reactor steam dome pressure is In accordance with 5 1058 psig. the Surveillance Frequency Control Program HATCH UNIT 1 3.4-25 Amendment No. 266

RCIC System 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.3 RCIC System LCO 3.5.3 The RCIC System shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3 with reactor steam dome pressure> 150 psig.

ACTIONS NOTE LCO 3.0.4.b is not applicable to RCIC.

CONDITION REQUIRED ACTION COMPLETION TIME A. RCIC System inoperable. A.1 Verify by administrative 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> means high pressure coolant injection (HPCI)

System is OPERABLE.

AND A.2 Restore RCIC System 14 days to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met. AND B.2 Reduce reactorsteam 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> dome pressure to 150 psig.

HATCH UNIT 1 3.5-9 Amendment No. 246

Drywell Pressure 3.6.1.4 3.6 CONTAINMENT SYSTEMS 3.6.1.4 Drywell Pressure LCO 3.6.1.4 Drywell pressure shall be 1.75 psig.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell pressure not within A.1 Restore drywell 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> limit, pressure to within limit.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE_REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.4.1 Verify drywell pressure is within limit. In accordance with the Surveillance Frequency Control Program HATCH UNIT 1 36-13 Amendment No. 266

Suppression Pool Water Level 3.6.2.2 3.6 CONTAINMENT SYSTEMS 3.6.2.2 Suppression Pool Water Level LCO 3.6.2.2 Suppression pool water level shall be 146 inches and 150 inches.

APPLICABILITY: MODES 1,2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Suppression pool water A.1 Restore suppression 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> level not within limits, pool water level to within limits.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE_REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.2.1 Verify suppression pool water level is within limits. In accordance with the Surveillance Frequency Control Program HATCH UNIT 1 3.6-24 Amendment No. 266

HLT-07 SRO NRC EXAM

70. G2.2.37 001 An event has occurred on Unit 2 requiring entry into 31E0-EOP-113-2, Terminating And Preventing Injection Into The RPV.

JAW step 3.1.1.1, HPCJ is OFF with 2E41-C002-3, HPCI Aux Oil pump, in the PULLTO-LOCK position.

Which ONE of the choices below completes the following statement concerning HPCI operability and Turbine Control Valve position?

JAW Tech Spec 3.5.1, HPCI is considered A. operable and the Turbine Control Valve will be fully closed B. operable and the Turbine Control Valve will be fully open C inoperable and the Turbine Control Valve will be fully closed D. inoperable and the Turbine Control Valve will be fully open 210

HLT-07 SRO NRC EXAM

==

Description:==

A system, subsystem, division, component, or device shall be OPERABILiTY OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s). With HPCI Aux Oil pump is considered mop.

HPCI Turbine Control Valve (TCV) requires oil pressure supplied to the valve for the valve to open. If oil pressure is removed the valve will travel closed. This is right the opposite for the RCIC TCV which on a loss of oil pressure the valve will travel open.

The A distractor is plausible if the applicant confuses operable with available and thinks since a procedure directed the action, that HPCI would not be mop just available. Also plausible if the applicant remembers that at certain times systems can be controlled by a stationed operator with manual actions so equipment can be considered available/operable. The second part is correct.

The B distractor is plausible if the applicant confuses operable with available and thinks since a procedure directed the action, that HPCI would not be mop just available. Also plausible if the applicant remembers that at certain times systems can be controlled by a stationed operator with manual actions so equipment can be considered available/operable. The second part is plausible if the appicant confuses the HPCI Turbine Control Valve operation with RCIC TCV.

The D distractor is plausible since the first part is correct. The second part is plausible if the appicant confuses the HPCI Turbine Control Valve operation with RCIC TCV.

A. Incorrect See description above.

B. Incorrect See description above.

C. Correct See description above.

D. Incorrect See description above.

211

HLT-07 SRO NRC EXAM

References:

NONE K/A:

2.2 Equipment Control 2.2.37 Ability to determine operability and/or availability of safety related equipment.

(CFR:41.7143.5!45.12) 3.6 4.6 LESSON PLAN/OBJECTIVE:

LT-LP-30005., Technical Specifications, EO 300.006.A.08 E41-HPCI-LP-00501, High Pressure Coolant Injection System, EO 005.005.A.02 References used to develop this question:

TS 1.1 Definitions 3450-E41-OOl-2, High Pressure Coolant Injection (HPCI) System 212

LT-LP-3 0005 Page 1 of 86 TECHNICAL SPECIFICATIONS OBJECTIVES TERMINAL OBJECTIVES 300.003.A Given plant conditions, IDENTIFY a Safety Limit Violation.

300.006.A Given plant conditions affecting a system covered in Tech Specs, DECLARE the system operable or inoperable per Technical Specifications.

300.0 10.A Given plant conditions effecting Tech Spec system, IDENTIFY the approaching or exceeding of a Limiting Condition for Operation (LCO) per Technical Specifications.

ENABLING OBJECTIVES

1. IDENTIFY how Technical Specifications and Technical Requirements Manual relate to the HNP license. (300.006.A.01)
2. IDENTIFY the four criteria for identifying equipment/parameters required to be included in Technical Specifications. (300.006.A.02)
3. IDENTIFY the purpose of the Technical Requirements Manual. (300.006.A.03)
4. IDENTIFY the authority required to modify the Technical Specifications and the Technical Requirements Manual. (300.006.A.04)
5. IDENTIFY the purpose of the Technical Specification Bases. (300.006.A.05)
6. IDENTIFY the five sections in the Technical Specifications and their relation to Technical Requirements Manual sections. (300.006.A.06)
7. IDENTIFY the purpose of definitions and state how they are identified in the specifications.

(300.006.A.07)

8. Given plant conditions, APPLY the following definitions: (300.006.A.08)
a. ACTIONS
b. OPERABLE
c. CORE ALTERATIONS
d. LEAKAGE
e. MODE (1,2,3,4,5)
9. IDENTIFY a correct set of required actions from an example specification containing multiple logical connectors. (300.006.A.09)

E41-HPCI-LP-00501 Page 5 of 92 HIGH PRESSURE COOLANT INJECTION SYSTEM

19. Given HPCI is injecting into the RPV, STATE the system response to a loss of the following power supplies: (005.005.a.07)
a. 25OVDC MCC 2B (R24-S022)
b. 125VDC CAB 2B (R25-S002)

System is in STANDBY per 34S0-E41-001-1/2, High Pressure Coolant Injection (HPCI)

System. (05.001 .a.07)

(005. 004 a. 03)

  • 29. Given Plant conditions, EVALUATE the conditions and DETERMINE if the HPCI turbine should have tripped. (005.005.a.06)
  • 30. Given Plant conditions, START HPCI locally per 3 1RS-E41-00l-1/2, HPCI Operations From Outside the Control Room. (005.02 l.a.01)
  • 31. Given plant conditions, DETERMINE whether the HPCI system trip signals, except overspeed, should be disabled per 31RS-E41-00I-l/2, HPCI Operation from outside the Control Room.

(005.021 .a.03)

  • 32. Given plant conditions, SHUTDOWN HPCI locally per 31RS-E41-001-1!2, HPCI Operation from Outside the Control Room. (005.022.a.01)

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 8 OF 73 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

HIGH PRESSURE COOLANT INJECTION (HPCI) SYSTEM 34S0-E41-001-2 23.3 7O PROCEDURE 7i SYSTEM STANDBY

  • it performing this section to support valve stroking and timing, with the permission of the Shift Supervisor jQ the 1ST Engineer steps may be marked NA, provided the system is aligned NOTES: to assure valve testing is in compliance with ASME testing requirements.
  • UNLESS indicated otherwise, controls and indications used in this subsection are on panel 2H11-P601 AND 2H11-P602.

7.1.1 WHEN performing this subsection to place the HPCI System in STANDBY, complete Attachment 1 to document independent verification. LI 7.1.2 Confirm Suppression Pool Level is> 146 inches AND < 150 inches, by 2T48-R607A (2H1 1-P602) Q 2T48-R6078 (2H1 1-P654). LI IF performing this section to support valve surveillance restoration, the following step may be NOTE: marked NA on Attachment 1, PROVIDED the HPCI Pump suction pressure did NOT drop below 14 psig during valve surveillance.

Critical 7.1.3 Confirm the HPCI system piping is filled and vented per the Filling and Venting subsection of this procedure. LI 7.1.4 Confirm 2E41-C002-2, Barom Cndsr Vacuum Pump, control switch is in AUTO. LI 7.1.5 Confirm 2E41-C002-1, Barom Cndsr Cond Pump, control switch is in AUTO. LI 7.1.6 Confirm 2E41-C002-3, Aux Oil Pump, control switch is in AUTO. LI 7.1.7 Confirm Turbine Test selector switch is in NORMAL. LI 7.1.8 Confirm Test Power selector switch is OFF. LI 7.1.9 Confirm 2E41-R612, HPCI Flow Control, is in AUTO, with setpoint set at 4250 gpm. LI 7.1.10 Confirm Turbine Control Valve is CLOSED. LI MGR-0001 Ver. 4

Definitions 1.1 1.1 Definitions (continued)

MINIMUM The MCPR shall be the smallest critical power ratio (CPR) that CRITICAL POWER exists in the core for each class of fuel. The CPR is that power RATIO (MCPR) in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE - A system, subsystem, division, component, or device shall be OPERABILITY OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a. Described in Section 136, Startup and Power Test Program, of the FSAR;
b. Authorized under the provisions of 10 CFR 50.59; or
c. Otherwise approved by the Nuclear Regulatory Commission.

RATED THERMAL RTP shall be a total reactor core heat transfer rate to the reactor POWER (RTP) coolant of 2804 MWt.

REACTOR The RPS RESPONSE TIME shall be that time interval from when the PROTECTION monitored parameter exceeds its RPS trip setpoint at the channel SYSTEM (RPS) sensor until de-energization of the scram pilot valve solenoids. The RESPONSE TIME response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

(continued)

HATCH UNIT 1 1.1-4 Amendment No. 238

HLT-07 SRO NRC EXAM

71. G2.2.44 OO Unit 1 RHR Loop B has been placed in the Shutdown Cooling (SDC) Mode of operation with the following pump alignment:

RHR Pump lB Running in SDC Mode RHRSW Pump lB .... Running The following alarms/conditions are received:

(601-3 14) RHR SYSTEM I ACTIVATED (601-234) RHR SYSTEM II ACTIVATED Reactor Water level is -120 inches LAW 34S0-E1 1-010-1, Residual Heat Removal System, which ONE of the choices below completes the following statements concerning the status of RHR Loop B Suction Valves?

With the above conditions, 1EI 1-FOO6B, Shutdown Cooling Vlv, After 1E1 1-FOO6B is closed, the operator will A. must be manually CLOSED; verify 1E1 1-FOO4B, Torus Suction Vlv, has automatically opened B must be manually CLOSED; manually open lEl l-FOO4B, Torus Suction Vlv C. will automatically CLOSE; verify 1E1 l-FOO4B, Torus Suction Vlv, has automatically opened D. will automatically CLOSE; manually open lEl 1-FOO4B, Torus Suction Vlv 213

HLT-07 SRO NRC EXAM

==

Description:==

With RWL dropping through +3, the SDC Suction Isolations Valves F008 and F009 will close.

This will cause RHR Pump lB to trip. As RWL continues to decrease to -120, at -101, RHR FO15B reopens ready for injection to the vessel. However, the RHR pumps will not inject due to a suction lineup trip. No suction path exists for the pumps unless the operator manually closes the F006 (SDC) valves and the manually opens the F004 (Torus) valves.

The A distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses the suction & injection valves interlocks or does not remember the procedure actions that must be taken in order to inject with this RHR Loop. Also plausible since other ECCS have suction valves that auto open with an initiation signal present. With RWL below

-101 an initiation signal is present.

The C distractor is plausible if the applicant confuses the suction valve logic and thinks the RHR suction valves (F008, F009 & F006s) will automatically close. Also would be plausible that all SDC suction valves will realign for LPC1 to occur. The second part is plausible if the applicant confuses the suction & injection valves interlocks or does not remember the procedure actions that must be taken in order to inject with this RHR Loop. Also plausible since other ECCS have suction valves that auto open with an initiation signal present. With RWL below

-101 an initiation signal is present.

The D distractor is plausible if the applicant confuses the suction valve logic and thinks the RHR suction valves (F008, F009 & F006s) will automatically close. Also would be plausible that all SDC suction valves will realign for LPCI to occur. The second part is correct.

A. Incorrect See description above.

B. Correct See description above.

C. Incorrect See description above.

D. Incorrect See description above.

214

HLT-07 SRO NRC EXAM

References:

NONE K/A:

2.2 Equipment Control 2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. (CFR: 41.5 /43.5 /45.12) 4.2 4.4 LESSON PLAN/OBJECTIVE:

El 1-RHR-LP-00701, Residual Heat Removal System, EO 006.005.A.02 References used to develop this question:

34S0-Ell-OlO-1, RHR System 215

E11-R[fR-LP-00701 Page 5 of 130 RESIIUAL REAT REMOVAL SYSTEM

15. From a list of statements, IDENTIFY the two statements which best describe the reason for having a jockey pump system maintaining the RHR piping filled. (006.001 .a.03)
16. From a list, IDENTIFY the two Plant Conditions which will cause an isolation of RHR shutdown cooling inboard/outboard isolation MOVs 2E1 l-F009/F008. (006.008.a.03)
17. From a list of statements, SELECT the statement that describes the interlock between the RHR SDC suction MOVs and their respective RHR pump torus suction MOV. (006.008.a.0l)
18. Given a list of RHR valve positions, DETERMINE if 2E11-F006 can be opened from the control room. (007.007.a.01)
19. From a list of statements, SELECT the statement that describes the interlock between the Torus Spray/Test Line Isolation Valve and the SDC Suction MOV. (007.003.a.04)
20. Given plant conditions, EVALUATE those conditions to determine whether RHR should have automatically initiated in LPCI mode. (006.005.a.0l)
21. DETERMINE the starting sequence for the RHR pumps given a LOCA and: (006.005.a.03)
a. No power loss
b. LOSP
22. Given that a LOCA signal has been received, with a subsequent loss of one division of 125 VDC logic power to the RHR system, DETERMINE what actions would be required in order to restart the RHR pump if it was manually secured in the present condition (006.007.a.02)
23. Given a list of plant conditions, IDENTIFY the three conditions which will cause a trip of an RHR pump. (006.008.a.02)
24. Given a simplified thawing of the RHR system, IDENTIFY those valves which would automatically operate or receive a signal to operate, upon the RHR system receiving a LPCI initiation signal while RHR is in: (006.005.a.02)
a. Normal STDBY
b. Shutdown Cooling Mode
c. Torus Cooling
25. From a list of statements, SELECT the statement which describes the interlock between the RHR Heat Exchanger Bypass Valve and a LOCA signal. (006.005.a.05)
26. Given plant conditions, DETERMINE if RWL interlocks must be overridden to operate 2E 11 -F028 Torus Spray/Test Line Isolation and F027 Torus Spray Line Isolation. (007.003.a.03)
27. Given plant conditions, DETERMINE if RWL interlocks must be overridden to open 2E 11-FO24AIB RHR Test Discharge Isolation MOV. (007.005.a.03)
28. From a list of statements, SELECT the statement which describes the interlock between the RHR SDC Suction MOV and the Outboard Spray Isolation Valve. (007.001 .a.04)

SOUTHERN NUCLEAR PLANT E. I. HATCH I I PAGE 123 0F293 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

RESIDUAL HEAT REMOVAL SYSTEM 34SO-El 1 -01 0-1 38.0 7.4.11 Manual LPCI Initiation While In Shutdown Cooling LUS The RHR System will automatically initiate in the LPCI mode on the following signals:

NOTE 1) Reactor level greater than OR equal to -113 inches (actual setpoint = -101 inches)

OR

2) High drywell pressure less than Q, equal to 1.92 psig (actual setpoint = 1.85 psig) 7.4.111 IF a valid LPCI initiation signal is received AND RHR is in the Shutdown Cooling Mode of operation, perform the following actions:

7.4.11.2 IF running, trip all RHRSW pumps. LI 7.4.11.3 jf running, tripRHRpumpsA&C(B&D). LI 7.4.11.4 Confirm closed ORclose 1E11-F008. LI 7.4.11.5 Confirm closed OR close IE11-F009. LI 7.4.11.6 Confirm closed OR close 1 El l-FOO6A & 1 El l-FOO6C(l El l-FOO6B & I El l-FOO6D). LI OPENING 1 El 1 -FOO4A & 1 El l-FOO4C (1 El l-FOO4B & 1 El l-FOO4D) WITH REACTOR PRESSURE 100 PSIG OR GREATER COULD RESULT IN EXCEEDING THE MAXIMUM TORQUE RATING OF THE VALVE OPERATOR. SUCH OPENING OF THE VALVES CAUTION: MUST BE FOLLOWED BY AN INSPECTION OF THE VALVE OPERATOR TO DETERMINE HANY DAMAGE OCCURRED ANDTHE OPERATOR IS SUITABLE FOR FURTHER SERVICE. INITIATE A CONDITION REPORT TO ENSURE AN INSPECTION IS PERFORMED.

7.4.11.7 Open lEll-F065A& 1E11-F065C(1E11-F065B & lEll-F065D).

7.4.11.8 Open lEll-FOO4A& lEll-FOO4C(1E1I-FOO4B & lEll-FOO4D).

MGR-000l Ver. 4

HLT-07 SRO NRC EXAM

72. G2.3.13 001 An INITIAL Drywell (DW) entry is required on Unit 1.

With Unit 1 at 8% RTP, which ONE of the choices below completes the following statements lAW 31G0-OPS-005-O, Primary Containment Entry?

The MAXIMUM reactor power allowed for this type of DW entry (INITIAL) is A Two (2) man Backup Team required to be in place for this entry.

A 1O%RTP; is B. lO%RTP; is NOT C. 15%RTP; is D. 15%RTP; is NOT 216

HLT-07 SRO NRC EXAM

==

Description:==

31G0-OPS-005-0 states that for an Initial DW entry, the maximum RTP is 10% (step 7.2.3.1) and a backup team is in place for this entry (step 7.2.7).

The B distractor is plausible since the first part is correct. The second part is plausible if the t

applicant thinks that since reactor power is <10% RTP that the backup team is not require d.

The C distractor is plausible if the applicant remembers the 15% Tech Spec limit of requiri ng the 02 level in the DW to be less than 4.0 volume %. The second part is correct.

The D distractor is plausible if the applicant remembers the 15% Tech Spec limit of requiri ng the 02 level in the DW to be less than 4.0 volume %. The second part is plausible if the applicant thinks that since reactor power is <10% RTP that the backup team is not require d.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

217

HLT-07 SRO NRC EXAM

References:

NONE K/A:

2.3 Radiation Control 2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

(CFR: 41.12143.4/45.9/45.10) 4 3.8 LESSON PLAN/OBJECTIVE:

LT-LP-30004 ,Administrative Procedures, EO 013.033 .A. 13 References used to develop this question:

31G0-OPS-005-0, Primary Containment Entry 218

Page 3 of 67 LT-LP-30004-12 ADMINISTRATIVE PROCEDURES

20. STATE the actions that must be performed by personnel who manipulate a locked valve.

(300.021 .b.03)

21. STATE the actions that must be performed by personnel performing an independent verification.

(300.022.a.04)

22. LIST the action that must be performed by personnel upon fmding a locked valve mispositioned or a valve-locking device inoperable. (300.022.a.05)
23. LIST the three (3) steps that must be performed to confirm locking device operability.

(300.022.a.03)

24. IDENTIFY the minimum required compositions of the Hatch Nuclear Fire Service (HNFS) per procedure 4OAC-ENG-008-0, Fire Protection Program. (SRO ONLY) (300.00 l.c.0l)
25. IDENTIFY the personnel authorized to act as a Fire Brigade leader during a fire per procedure 4OAC-ENG-008-0, Fire Protection Program. (SRO ONLY) (300.001 .c.02)
26. Given procedure 4OAC-ENG-Ol 8-0, Temporary Modification Control, IDENTIFY the conditions requiring a temporary modification sheet. (300.017.a.02)
27. Given procedure 4OAC-ENG-0l 8-0, Temporary Modification Control, STATE the conditions that must be met for a lifted wire or jumper request to be issued on a safety system that is required to be operable by technical specifications. (SRO ONLY) (300.0 17.a.03)
28. STATE who controls and issues Grand Master Keys per 8OAC-SEC-002-0, Key and Annunciated Door Control. (300.041.c.Ol)
29. Given specific plant conditions and system status, DETERMINE if a system or component can be removed from service for preventative maintenance per 9OAC-OAM-002-0. (300.011 .b.02)
30. Given plant status and an Order from the Load Dispatcher, IDENTIFY the correct actions necessary to complete the Switching Order per 34SO-S22-001-0. (300.0l3.b.01)
31. STATE what a short-term turnover is and when it is allowed per NMP-OS-007-0. (300.035.B.15)
32. IDENTIFY the information that must be transferred from the OATC to another NPO during a Short-term turnover, per NMP-OS-007-0. (300.035 .B. 16)
33. STATE what items are required to be recorded in shift logs and/or sub-logs as required by procedure 31 GO-OP S-007-0. (300.035 .B. 17)
34. Given 31 GO-OPS-005-0, STATE what requirements, special requirements, precautions and limitations, prerequisites that are necessary for a Primary Containment Entry. (013.033.A.13)
35. Given an example of a completed OPS-1958, DETERMINE if the form was completed per 3IGO-OPS-007-0 (300.035.B.18)
36. Given a specific crew manning alTangement, EVALUATE whether Safe Shutdown and Fire Brigade requirements are met. (300.035.B.19)

SOUTHERN NUCLEAR DOCUMENT TYPE:

PLANT E. I. HATCH PAGE GENERAL OPERATING PROCEDURE 1 OF 18 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

PRIMARY CONTAINMENT ENTRY 31 GO-OPS-005-0 12.17 EXPIRATION APPROVALS:

EFFECTIVE DATE: DEPARTMENT MGR G. L. Johnson DATE 11/02/00 DATE:

N/A SSM / PM N/A DATE N/A 10-28-2011 1.0 OBJECTIVE This procedure provides a barrier to protect personnel from the hazards of unexpected radiation exposure, deficient atmosphere, over exposure to high temperature, and includ e maximum acceptable Drywell to Reactor Bldg. t P for primary containment entry. This procedure establishes the method that will be used whenever an entry into the primary containment is made.

TABLE OF CONTENTS Section Page 2.0 APPLICABILITY 1

3.0 REFERENCES

2 4.0 REQUIREMENTS 3

5.0 PRECAUTIONS/LIMITATIONS 5

6.0 PREREQUISITES 6

7.0 PROCEDURE 10 7.1 INITIAL ENTRY SAMPLING 10 7.2 INITIAL OR ABNORMAL CONTAINMENT ENTRY 11 7.3 NORMAL CONTAINMENT ENTRY 14 7.4 PERSONNEL AIRLOCK OPERATION 15 7.5 OVERRIDE OF ELECTRICAL INTERLOCK FOR THE INTERIOR DRYWELL AIRLOCK DOOR 17 2.0 APPLICABILITY This procedure is applicable to all entries into the primary containment includ ing:

  • NORMAL ENTRY - Any containment entry after the initial entry where reactor power is less than 10% and containment atmospheric conditions are acceptable as per subsection 6.3.
  • INITIAL ENTRY - The first containment entry made subsequent to containment being de-inerted or drywell pneumatics on nitrogen supply.
  • ABNORMAL ENTRY Any containment entry made with reactor power greater than 10% OR the containment atmospheric conditions unknown or NOT acceptable as per subsection 6.3.

MGR-0002 Rev 8.1

SOUTHERN NUCLEAR PAGE PLANTE.I.HATCH 3OF18 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

PRIMARY CONTAINMENT ENTRY 31GO-OPS-005-O 12.17 4.0 REQUIREMENTS 4.1 PERSONNEL REQUIREMENT 4.1.1 At least two persons are required when initial or abnormal primary contain ment entries are made; one of whom must be a qualified Health Physics Technician. A Health Physics Technician may NOT be required IF Health Physics Technicians are posted in primary containment as during a refueling outage.

4.1.2 A two person backup team must be available for an Initial OR Abnormal Entry.

This team will be equipped as specified in this procedure.

4.1.3 An Entry Supervisor and an Attendant are required when making a Permit Requir ed Confined Space entry per, NMP-SHOO5 Confined Space Procedure.

4.1.4 An Entry Supervisor is required when making a Confined Space entry per NMP-S H-005 Confined Space Procedure.

4.1.5 Additional personnel may be required as per NMP-SH-005 Confined Space Proced ure.

4.2 MATERIALS AND EQUIPMENT 4.2.1 Protective clothing 4.2.2 Self contained breathing apparatus (SCBA) (as required) 4.2.3 Portable air samplers and air monitoring equipment (as required) 42.4 Portable radiation monitors (as required) 4.2.5 Communication devices, IF required 4.2.6 Key # MC-52, for 1 H21-P331, Panel Keylock Selector Switch Key, located in Operations Micellaneous key cabinet 4.2.7 Key VHR-3, for Ui Drywell Airlock chain, located in HP High Radiation Key Cabinet 4.2.8 Key # 2RB-20, for 2H21-P331, Panel Keylock Selector Switch Key, located in Operations U2 Door Key Cabinet 4.2.9 Key VHR-3A, for U2 Drywell Airlock chain, located in HP High Radiation Key Cabinet MGR-0001 Rev4

HLT-07 SRO NRC EXAM

73. G2.3.14 001 Unit 1 was at 35% power when the Hydrogen Injection System was placed in service in AUTOMATIC EXTERNAL mode JAW 34SO-P73-00 1-1, Hydrogen and Oxygen injection and Control for HWC section 7.1.2, Placing 1P73-R025, Hydrogen Controller, in EXTERNAL.

o Power is raised from 35% power to 100% power o At 100% power hydrogen flow rate indicates 40 SCFM Which ONE of the following answers both of these statements?

JAW 34SO-P73-00l-l, Hydrogen and Oxygen injection and Control for HWC, hydrogen injection flow rate is the normal 100% power flow rate.

Radiation levels in the Condenser Bay will stabilize expected normal full power radiation levels.

A. above; at B above; above C. below; at D. below; below 219

HLT-07 SRO NRC EXAM

Description:

In the Automatic External Mode the hydrogen injection system is load following.

It is normally placed in service at just above 30% Rx power. As Rx power is increased, the hydrog en flow rate is increased to the maximum amount that the controller is set for. The amount of hydrog en has varied over time from as low as 8 SCFM to as high as 45 SCFM. The plant is curren tly operating at 10 SCFM, but this amount is determined by the Chemistry Dept. and is not specifically listed in a procedure. There is a system shutdown on hydrogen flow of 25 SCFM.

There was an event at Plant Hatch: Hydrogen flow was manually lowered in Interna l mode, the controller was then placed in External mode and hydrogen gas flow ramped back up to the original setpoint. This resulted in a radiation exposure to several individuals that was above the expected dose for the task. As hydrogen injection rate is increased, radiation levels increas e

around any area where Main Steam is piped.

The A distractor is plausible since the first part is correct (refer to description above)

. The second part is not correct; radiation levels are expected to increase to above norma l levels and is plausible because flow is higher than normal, but the applicant must understand the operation of the Automatic-External mode and the effects of H2 injection on Radiation levels.

The C distractor is plausible since H2 flow rate is NOT below normal. The second part is not correct; radiation levels are expected to increase to above normal levels and is plausib le, if applicant does not understand the operations of the Automatic-External mode and the effects of H2 injection on Radiation levels or if candidate incorrectly recalls that normal flow is above 40 SCFM.

The D distractor is not correct since H2 flow rate is above normal. The second part is not correct; radiation levels are expected to increase to above normal levels and is plausib le, if applicant does not understand the operations of the Automatic-External mode and the effects of H2 injection on Radiation levels or if applicant incorrectly recalls that normal flow is above 40 SCFM.

A. Incorrect See description above.

B. Correct See description above.

C. Incorrect See description above.

D. Incorrect See description above.

220

HLT-07 SRO NRC EXAM

References:

NONE K/A:

2.3 Radiation Control 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. (CFR: 41.12 / 43.4/4 5.10). 3.4 3.8 LESSON PLAN/OBJECTIVE:

P73-HWCI-LP-07301, Reactor Water Chemical Injection System, EO 026.03 4.A.05 and 026.034.A.04 Reference(s) used to develop this question:

34S0-P73-0014, Hydrogen and Oxygen injection and Control for HWC Bank Question from 2009 HLT-4 NRC Exam Q#72 221

P73 -HWCI-LP-0730 1 Page 2 of 105 REACTOR WATER CHEMICAL INJECTION SYSTEMS Initial License (LT) ENABLING OBJECTIVES

1. From a list of statements, SELECT the statement that best describ es the function of the following system components: (026.033.A.05)
a. Hydrogen Supply Tanks
b. Excess Flow Check Valve, P73-F002
c. Hydrogen Main Isolation Valve, P73-F004
d. Hydrogen Flow Control Valves, P73-F 109A, B, C
e. Flow Controllers, P73-R025, R026
f. Hydrogen Injection Isolation Valves, P73-Fl 12A, B, C
g. Oxygen Storage Tank
h. Oxygen Flow Control Valves, P73-F2 11 A, B
i. Hydrogen Area Monitors, P73-N027, N028
j. Recirculation Water Monitoring System
k. Off Gas Oxygen Sample Panel
2. From a list of statements, SELECT the statement that best describes the operation of the following Hydrogen Water Chemistry Injection System components:

(026.034.A.0 1)

a. Excess Flow Check Valve, P73-F002
b. Hydrogen Flow Control Valves, P73-F 109A, B, C
c. Hydrogen Main Isolation Valve, P73-F004
3. Given the following list of local controls for the HWCI System, IDENT IFY the function of each. (026.034.A.03)
a. Booster Pump Hydrogen Flow Control Valves, P73-F1O9A, B, C
b. Booster Pump Hydrogen Isolation Valve, P73-F 11 2A, B, C
c. Oxygen Flow Control Valves, P73-F21IA/B
d. Oxygen Flow Controller, P73-R026
e. Hydrogen Flow Controller, P73-R025
4. From a list of statements, SELECT the statement that best describ es the significance of each of the following HWCI System annunciators/alarms: (026.034.A.0 4)
a. High Off Gas % Oxygen
b. Low Off Gas % Oxygen
c. High Hydrogen Flow
d. High and High High Hydrogen Area Monitor
e. Purge Alarm
f. Local Shutdown Demand
g. Low Hydrogen Pressure
h. High Oxygen Pressure
i. Low Oxygen Pressure
j. Low % Oxygen Monitor Sample Flow
5. From a list of statements, SELECT the statement that best describ es the HWCI System response to the following: (026.034.A.02)
a. High Hydrogen Supply Pressure
b. Low Hydrogen Injection Pressure

P73-HWCI-LP-07301 Page 3 of 105 REACTOR WATER CHEMICAL INJECTION SYSTEMS

6. DESCRIBE the HWCI System response when a Reactor scram, power change, or isolation of the Reactor occurs as stated in 34S0-P73-001-l/2, Hydrogen and Oxygen Injection for HWC. (026.035.A.01)
7. DESCRIBE the actions for manual shutdown of the HWCI System per 34S0-P73-00l-l/2, Hydrogen and Oxygen Injection for HWC. (026.03 5.A.02)
  • 8. STATE the purpose of the Hydrogen Water Chemistry (HWCI) Injection system.

(026.033.A.01)

  • 9 DESCRIBE the effects on radiation levels within the plant as well as areas involved when Hydrogen is being injected into the Condensate/Feedwater System as stated in 34S0-P73-001-1/2. (026.034.A.05)
  • 10. Given a P&ID or a simplified drawing of the HWCI System, TRACE the Hydrogen and Oxygen flow path(s) through the system. (026.033.A.02)
  • 11. Given a P&ID or a simplified drawing of the HWCI System, LABEL the following components: (026.033.A.03)
a. Hydrogen storage tank
b. Hydrogen Main Isolation Valve
c. Hydrogen Flow Control Valves
d. Excess Flow Check Valve
e. Oxygen storage tank
f. Flow controllers
g. Hydrogen Injection Isolation Valves
h. Oxygen Flow Control Valves
  • 12. From a list, SELECT the consequence of returning the Hydrogen Injection Flow Controller to External when reducing Hydrogen flow without reducing power. (026.034.A 08)
  • 13. Given the list of systems below that interface with the HWCI System, DESCRIBE the reason/function of the interface and general physical location of the interface. (026.033.A.04)
a. Instrument Air System
b. Condensate System
c. Off Gas System
  • 14. DESCRIBE the actions for manual startup of the HWCI System as given in 34S0-P73-00l-l/2. (026.034.A.06)
  • 15. DESCRIBE the proper operator actions for changing system flow rates. (026.034.A.0 7)
  • 16. Given a copy of Technical Specifications and a set of plant conditions, DETERMIN E the actions required if the conditions do not comply with the LCO. (300.0 10.A.06)

SOUTHERN NUCLEAR PLANT E. I. HATCH PAGE 5 OF 64 DOCUMENT TITLE:

DOCUMENT NUMBER: VERSION NO:

HYDROGEN WATER CHEMISTRY INJECTION 34SO-P73-0O1-1 15.6 5.1.7 For the Hydrogen and Oxygen flow controllers, the EXTERNAL mode functions with the setpoint generated from an external source (for example, the oxygen controller receives input from the hydrogen controller setpoint. The hydrogen controller receives input from percent of Total Feedwater Flow).

5.1.8 The Automatic! External mode of HWC operation is also referred to as Load Following.

With hydrogen controller, 1 P73-R025, in Automati c I External, hydrogen flowrate will be automatically adjusted to a pre-programmed setpo int determined from percent (%) of rated feedwater flow. Placing the hydrogen controlle r in EXTERNAL will result in a change in hydrogen flowrate, unless the actual flowrate is equa l to the setpoint. (i.e., IF the hydrogen controller is operating in INTERNAL mode and adjusted to 8 SCFM with the Reactor operating at 100% feedwater flow (100% RTP), AND the controller is subsequently placed to EXTERNAL, THEN the hydrogen flowrate will auto matically ramp to the hydrogen flowrate value corresponding to 100% feedwater flow.)

5.2 LIMITAT)ONS 5.2.1 Any of the following conditions will automatically trip the Hydrogen Injection System:

5.2.1.1 Reactor Mode Switch is NOT in Run Mode.

5.2.1.2 Low Offgas Percent Oxygen - Less than QE equal to 5% Oxygen.

5.2.1.3 High-High Hydrogen area Monitors Greater than OR equal to 50% Combustible Mixture of Hydrogen.

5.2.1.4 High Hydrogen Flow Greater than OR equa l to 25 SCFM.

5.2.1.5 High Hydrogen Pressure -250 PSIG, increasing 5.2.1.6 Offgas Train/Recombiner Train Trip Train Isolati on Valves Closed. (See Attachment 5) 5.2.2 Any of the following conditions will automatically shut down the Oxygen and Hydrogen Gas Supply System:

5.2.2.1 Low Instrument Gas Pressure - 10 PSIG, decreasing.

5.2.2.2 High Pressure Pipeline Low Temperature -20F,-

decreasing.

5.2.3 Any of the following conditions will automatically shut down 1P73-COO1A (1P73-COOIB),

Pump, from the Gas Supply System:

5.2.3.1 Pump relief valve open -

-40F, decreasing 5.2.3.2 High pump pressure -

2300 PSIG, increasing 5.2.3.3 Pump fails to start -

Three attempts 5.2.3.4 Motor purge low pressure -

5 PSIG, decreasing MGR-0001 Ver. 4

HLT-07 SRO NRC EXAM

74. G2.4.26 001 Which ONE of the choices below completes the following statements concerning fire a in the Service Building Document Storage Room?

The Service Building Document Storage Room is equipped with a fire suppression system.

JAW 3OAC-OPS-003-O, Plant Operations, a MINIMUM of Fire Brigade Members must be onsite to respond to this fire.

A Halon; five (5)

B. Halon; three (3)

C. 2C0 five (5)

D. 2C0 three (3) 222

HLT-07 SRO NRC EXAM

Description:

JAW 34S0-X43-003-1N, Document Control Fire Protection System, Detection of combustion gases by any one of the eight ionization detectors in Document Control will result in system initiation. After a 2 minute delay Halon will be discharged in Document Contro l if no preventive action is taken.

3OAC-OPS-003-O. Plant Operations, step 8.1.3.4 states A Fire Brigade of a least 5 members shall be on site at all times. 4OAC-ENG-008-O, Fire Protection Program, step 8.1.1.1.2 states The Fire Brigade SHALL consist of a minimum of five qualified persons on each shift. A minimum of three of these persons must have competent knowledge of safety-related systems and components. One of these will be the HNFS Fire Brigade Leader.

The B distractor is plausible since the first part is correct. The second part is plausible if the applicant confuses three with being the number of persons that must have competent knowledge of safety-related systems and components.

The C distractor is plausible if the applicant confuses the Halon System with the CO 2 System and thinks CO2 is covering the Document Storage Room Area. The second part is correct The D distractor is plausible if the applicant confuses the Halon System with the CO 2 System and thinks CO2 is covering the Document Storage Room Area. The second part is plausib le if the applicant confuses three with being the number of persons that must have competent knowledge of safety-related systems and components.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

223

HLT-07 SRO NRC EXAM

References:

NONE K/A:

2.4 Emergency Procedures I Plan 2.4.26 Knowledge of facility protection requirements, including fire brigad e and portable fire fighting equipment usage. (CFR: 41.10/43.5 /45.12) 3.1 3.6 LESSON PLAN/OBJECTIVE:

X43-FPS-LP-0360 1, Fire Protection, EO 036 .020.B .01 References used to develop this question:

34SO-X43-003-1N, Document Control Fire Protection System 3OAC-OPS-003-0, Plant Operations (Section 8.1.3, General Manning Requir ements)

Modified from HLT-5 NRC Exam Q#74 ORIGINAL QUESTION (HLT-5 NRC Exam Q#74)

Which ONE of the choices below completes the following two statements?

The Cable Spreading Room is equipped with a fire suppression system.

The Cable Spreading Room fire suppression system initiate without operator action.

A. Dry pipe sprinkler; will B. Dry pipe sprinkler; will NOT C. CO2; will D./ C02; will NOT 224

X43-FPS-LP-03601 Page 2 of 145 FIRE PROTECTION Initial License (LT) ENABLING OBJECTIVES

1. Given a list of statements, SELECT the statement that best describes the operation of the following types of Fire Suppression systems: (036.020.B.0l)
a. Deluge
b. Fixed Water Spray
c. Wet Pipe Sprinklers
d. Dry Pipe Sprinklers
e. Preaction Sprinklers
f. 2 (Cardox)

CO

g. Halon 1301
2. Given a simplified drawing of a deluge system, IDENTIFY the following compo nents:

(036.020.B.03)

a. Latch
b. Clapper Valve
c. Riser and Isolation valve
d. Air Diaphragm
e. Ball Drip Valve
f. Supervisory Water Line
3. Given a drawing of a deluge system, TRACE the operational fiowpath for the system.

(036.020.B.04)

4. Given that a small stream of water is flowing from a deluge system ball drip valve, STATE the significance of this condition. (03 6.020.B.02)
5. For the areas protected by the CO 2 (Cardox) system, STATE if the system can be initiated automatically, semi automatically, or manually. (036.020.B. 11)
6. IDENTIFY the particular odor used to identify the presence of CO

. (036.020.B. 12) 2

7. Given a list of plant areas, SELECT the areas protected by a Halon 1301 system (036.020.B.07)
8. STATE the location of the Halon 1301 system abort switches. (036.020.B.08)
9. Given a list of statements, SELECT the statement(s) describing the function of the follow ing Fire Detection systems: (036.021 .A.03)
a. CXL system
b. XL3 Master panel
c. XL3 Slave panels
d. Protectowire system
e. Alison Control Detection system

SOUTHERN NUCLEAR DOCUMENT TYPE:

PLANT E. I. HATCH PAGE SYSTEM OPERATING PROCEDURE 1 OF 7 DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION DOCUMENT CONTROL FIRE PROTECTION 34SO-X43-003-1N NO:

SYSTEM 1 ED 1 EXPIRATION APPROVALS:

EFFECTIVE DATE: DEPARTMENT MGR D. R. Madison DATE 11/29/89 DATE:

N/A NPGM/POAGM/PSAGM N/A 08/16/00 DATE N/A 1.0 OBJECTIVE This procedure provides instructions for manual, automatic and emergency initiation and system restoration for the Document Control Halon Fire Protection System.

TABLE OF CONTENTS Section Page 2.0 APPLICABILITY 1

3.0 REFERENCES

2 4.0 REQUIREMENTS 2

5.0 PRECAUTIONS/LIMITATIONS 3

6.0 PREREQUISITES 3

7.0 PROCEDURE 4

7.1 SYSTEM INITIATION 4

7.1.1 AUTOMATIC INITIATION 4

7.1.2 MANUAL INITIATION 7.1.3 EMERGENCY INITIATION 4 7.2 SYSTEM RESPONSE (Automatic and Manual Initiation) 5 6

7.3 SYSTEM RESTORATION 7

2.0 APPLICABILITY TIis procedure applies to the Document Control Halon 1301 System MGR-0002 Rev 8

SOUTHERN NUCLEAR PLANT E. I. HATCH PAGE 50F8 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

PLANT OPERATIONS 30AC-OPS-003-O 26.7 8.1.3 General Manning Requirements The following shall apply to manning of the plant in general.

8.1.3.1 At least 3 NPOs will be on site at all times [SROs may substitute provided they are NOT in a supervisory role and efforts are being made to find replacements as noted in 8.1.2.2].

8.1.3.2 A total of 3 SOs for the 2 units is required at all times. At least 1 of the require SOs d shall be assigned to each reactor.

8.1.3.3 At least one STA shall be on site and available WITHIN 10 minutes to the MCR at all times WHEN EITHER reactor is in operational condition 1, 2, or 3.

8 1 34 A Fire Bngade of a least 5 members shall be on site at all times 8.1.3.5 An individual qualified to implement Radiation Protection Procedures shall be onsite WHEN fuel is in EITHER reactor. This position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

8.1.3.6 Watch-Standing proficiency is governed by Attachment 2 and personnel assign ed these positions must complete a full shift to receive credit for the shift.

8.2 MEDICAL STATUS CHANGE 8.2.1 Operations personnel (licensed or non-licensed) will notify the SS/ SM and Safety and Health when that individuals medical status has changed in accordance with 2OAC-ADM-005-0, Mandatory Physical Examinations, 27SH-SFT-00 1-0, Physical Examination Proced ure, and Dl-TRN-55-0601, Operations Personnel Qualification Instructions.

MGR-0001 Rev 4.0

HLT-07 SRO NRC EXAM

75. G2.4.32 001 Unit 2 is operating at 100% RTP with a NPO testing the NSSS panel alarms (P603, P602, & P60 1).

o NONE of the NSSS panel alarms worked lAW 34AB-Hl 1-001-2, Loss Of Power To Annunciators In Main Control Room, which ONE of the choices below completes the following statement?

The OATC will dispatch another operator to to investigate the NORMAL power supply breaker and to to investigate the ALTERNATE power supply breaker.

A Instrument Bus 2B, 2R25-S065; 125 VDC Bus 2A, 2R25-S001 B. Instrument Bus 2B, 2R25-S065; 125 VDC Bus 2B, 2R25-S002 C. Vital AC, 2R25-S063; 125VDC Bus 2A, 2R25-S00l D. Vital AC, 2R25-S063; 125 VDC Bus 2B, 2R25-S002 225

HLT-07 SRO NRC EXAM

Description:

The NSSS panel alarms are powered from the following:

Instrument Bus 2B, 2R25-S065 NORMAL Supply 125 VDC Bus 2A, 2R25-SOO1 ALTERNATE Supply The BOP panel alarms are powered from the following:

125 VDC Bus 2B, 2R25-S002 NORMAL Supply Instrument Bus 2A, 2R25-S064 ALTERNATE Supply The B distractor is plausible since the first part is correct. The second part is plausib le if the applicant remembers the alternate is a DC bus but confuses which one.

The C distractor is plausible if the applicant remembers Vital AC powers various compo nents and confuses it as the Normal power supply for the NSSS alarms. The second part is correct The D distractor is plausible if the applicant remembers Vital AC powers various compo nents and confuses it as the Normal power supply for the NSSS alarms. The second part is plausible if the applicant remembers the alternate is a DC bus but confuses which one.

A. Correct See description above.

B. Incorrect See description above.

C. Incorrect See description above.

D. Incorrect See description above.

226

HLT-07 SRO NRC EXAM

References:

NONE K/A:

2.4 Emergency Procedures / Plan 2.4.32 Knowledge of operator response to loss of all annunciators.

(CFR:41.10/43.5/45.13) 3.6 4.0 LESSON PLAN/OBJECTIVE:

LT-LP-20201, Introduction To Abnormal Procedures, LO LT-20201

.019 References used to develop this question:

34AB-Hl 1-001-2, Loss Of Power To Annunciators In Main Control Room 227

Page 2 of 26 LT-LP-20201-11 iNTRODUCTION TO ABNORMAL PRO CEDURES

14. Per 34AB-P5 1-001-2, Loss of Instrument and Service Air or Water Intrusion Into The Air System, STATE the two conditions requ Service iring a manual SCRAM. (LT-20201 .009)
15. STATE the time limit for opening the gene rator output breakers after a turbine trip with of 125/250 \/DC Switchgear 2A, 2R22-S0 a loss 16 or 125 VDC Cabinet 2A, 2R25-S001, EXPLAIN the reason for this limit, as state and d in 34AB-R22-OOl-2. (LT-2020l.007)
16. Given plant conditions, DETERMINE if entry into 34AB-C71-00l-2, Scram Proc required per 34AB-T23-003-2, Tonis Tem edure is perature Above 95°F. (LT-20201 .011)
7. Given applicable procedure and plant condition s, DETERMINE the proper operator actio course of actions. (LT-20201 .019) n or Not Selected for Requal

SOUTHERN NUCLEAR PAGE 2 OF 2 PLANT E. I. HATCH DOCUMENT TITLE: DOCUMENT NUMBER: REVISION NO:

LOSS OF POWER TO ANNUNCIATORS IN MAIN 34AB-H11-001-2 1.4 CONTROL ROOM 4.0 SUBSEQUENT OPERATOR ACTIONS 4.1 power is lost to NSSS Annunciators (Panels 2H11-P601, P602, P603),

dispatch an Operator to Panel 2H 11 -P630 on the 130 elevation of the Control Building to perform a visual inspection.

4.1.1 Dispatch Operator to check normal and alternate power supplies for 2H11-P630.

4.1.1.1 Normal power supply is Instrument Bus 2B (2R25-S065) BRKR #19 on the 130 elevation Control Building RPS MG Set Room.

4.1.1.2 Alternate power supply is 125VDC Bus 2A (2R25-S001) BRKR #25 on the 130 elevation Control Building 2C 600V MCC room.

4.1.2 Notify Maintenance to investigate the loss of the normal and/or alternate power supplies to Panel 2H1 1-P630.

4.2 IF power is lost to BOP Annunciators (Panels 2H1 1-P650, 2H1 1-P651, 2H1 1-P656, 2H1 1-P654, 2H 11 -P657, 2H 11 -P659, 2H 11 -P700 or 2N62-P600) OR the Diesel Generator Annunciators (2H1 1-P652-1, 2H1 1-P6522, 2H1 1-P6523),

dispatch an Operator to Panel 2H21-P237 on the 112 elevation of the Control Building to perform a visual inspection.

4.2.1 Dispatch Operator to check normal AND alternate power supplies for 2H21-P237.

4.2.1.1 NORMAL power supply is 125VDC Bus 2B (2R25-S002) BRKR. #13, on the 130 elevation Control Building, South Hallway.

4.2.1.2 ALTERNATE power supply is Instrument Bus 2A (2R25-S064) BRKR. #12, on the 130 elevation Control Building, NSSS Annunciator Logic room.

4.12 Notify Maintenance to investigate the loss of the normal AND/OR alternate power supplies to Panel 2H21-P237.

4.3 Increase Operator attention (awareness) on those operating boards whose annunciators are affected.

4.4 For a failure OR loss of Plant Annunciators that monitor parameters which cannot be monitored continuously by increased operator attention, to ensure that the plant is, and will remain in a safe condition, refer to NMP-EP-110-GLO2 for classification of emergency.

MGR-0001 Rev 2

Retrieved from "https://kanterella.com/w/index.php?title=ML12342A038&oldid=2638310"