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| g g UNITED STATES
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| : g. It NUCLEAR REGULATORY COMMISSION
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| {, .
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| REGION IV J
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| % 611 RYAN PLAZA DRIVE, SUITE 400
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| $9 e[ AR LINGTON, TE XAS 760118064 l
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| May 20,1998 NOTE TO: NRC Document Control Desk Mail Stop O 5-D-24 1
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| i FROM: Laura Hurley, Licensing Assistant Operations Branch, Region IV ,
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| l
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| | |
| ==SUBJECT:==
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| OPERATOR LICENSING EXAMINATIONS ADMINISTERED ON FEBRUARY l 17,1998, AT WOLF CREEK GENERATING STATION, UNIT 1 1
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| DOCKET #50-482 On February 17,1998, Operator Licensing Examinations were administered at the referenced facil.ity. Attached you will find the following information for processing through NUDOCS and distribution to the NRC staff, including the NRC PDR: .
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| l ltem #1 - Facility submitted outline, the initial exam submittal designated for distribution under RIDS Code A070.
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| Item #2 - Examination Report wit;. 2he as given written examination attached, designated for distribution under RIDS Code IE42.
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| If you have any questions, please contact Laura Hurley, Licensing Assistant, Operations Branch, Region IV at (817) 860-8253.
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| '.'!0007 9805260319 990520 PDR ADOCK 05000482 V PDR I
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| | |
| ,2o
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| (
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| Wolf Creek Generating Station Written Examination Outline Examination Date: February 17,1998 Quality Assurance Checklist Explanations item la. The outline matches the distribution shown in ES-401.
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| Item lb. The distribution of questions X/As are balanced in the outline. All areas are being sampled.
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| Item Ic. No system or evolution is sampled more than two times. This matches the distribution numbers found in ES-401 for PWR RO and SRO. Topical areas not represented in Wolf Creek design or Westinghouse design are not noted in the outline. >
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| ltem Id. The August 1997 examination outline formed the basis for this examinations outline.
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| Questions used for the most part are new questions or those from examination banks not used in the candidates previous training or testing.
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| Item 2. This section is not prepared for this examination. The company is requesting the operational examination be waived.
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| Item 3. This section is not prepared for this examination. The company is requesting the operational examination be waived.
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| Item 4a. PRA information was evaluated to determine if the test outline should be altered. The ES-401 outline criteria provides for an adequate representation of the high risk evruts. ;
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| No changes were made in the NUREG outline topics.
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| Item 4b. 10CFR55.41 and .43 requirements have been evaluated. The examination will shrre approximatdy 75% of the test between the RO and SRO candidates. The 25% ofihe questions that are SRO only will measure 10CFR5U3 topical areas.
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| Item 4c. No examination questions used have importance factors less than 2.5. j ltem 4d. This criteria was not evaluated. Only a written examination is being prepared.
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| L Item 4e. The examination was reviewed. It is a balanced examination matching the ES-401 outline system and evolution topic areas.
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| Item 4f. The examinations being prepared assess at the appropriate job level.
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| l l -070 .
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| - - - - - - - - - - - - - - _ - - - - - _ JA
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| | |
| m RO 13cility: WCGS Date of Exam: February 17,1998 Exam Level: RO Tier Group K/A Category Pc ints l Point Total A2 A3 A4 G K1 K2 K3 K4 l K5 K6 ' Al
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| : 1. 1 013 016 004 001 002 010 16 005 009 003 014 012 011 006 015 007 Emergency & 2 -W 018 W 023 027 17 032 029 020 026 076 017 024 028 019 031 021 Abnormal 3 033 3
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| )'
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| 034 Plant Tier 4 7 8 5 9 3 36 Evolutions Totals
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| : 2. 1 080 079 042 045 050 078 041 036 049 043 037 23 082 047 035 038 039 083 081 040 046 084 044 048 Plant 2 066 053 063 056 052 054 059 055 060 061 20 064 058 065 051 057 062 086 067 087 085 Systems 3 069 073 071 070 8 068l l 072 088 089 Tier 6 2 4 7 3 4 5 6 l5 6 3 51 Totals l
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| : 3. Generic Knowledge and Cat 1 Cat 2 Cat 3 Cat 4 Abilities 90 95 74 75 13 91 96 98 100 92 97 99 93 94 l
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| I
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| | |
| Note:
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| * Attempt to distribute topics among all K/A categories; select at least one topic from every K/A category within each tier.
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| * Actual point totals must match those specified in the table.
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| * Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant specific priorities.
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| . Systems / Evolutions within each group are identified on the associated outline.
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| * The shaded areas are not applicable to the category /Fier.
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| Facility: WCGS Generic Knowledge and Abilities Outline (Tier 3) Exam Level: RO l l Category K/A # Topic Imp. Points l 2.1.1 Knowledge of conduct of operations 3.7 1 90 requirements.
| |
| Conduct of 2.1.2 Knowledge of operator responsibilities 3.0 1 Operations 91 during all modes of plant operation.
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| 2.1.3 Knowledge of shift turnover practices. 3.0 1 92 2.1.16 Ability to operate plant phone, paging 2.9 1 93 system, and two-way radio.
| |
| 2.1.25 Ability to obtain and interpret station 2.8/3.1 1 94 reference materials such as graphs, monographs, and tables which contain performance data.
| |
| Total 5 2.2.12 Knowledge of surveillance procedures. 3.0 1
| |
| ]
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| 95 Equipment 2.2.13 Knowledge of tagging and clearance 3.6 1 Control 96 procedures.
| |
| 2.2.26 Knowledge of refueling administrative 2.5 1 97 requirements.
| |
| Total 3 2.3.1 Knowledge of 10CFR: 20 and related 2.6 1 98 facility radiation control requirements.)
| |
| Radiation 2.3.2 Knowledge of facility ALARA program. 2.5 1 74 Control 2.3.10 Ability to perform procedures to reduce 2.9 1 99 excessive levels of radiation and guard against personnel exposure.
| |
| Total 3 I
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| i i
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| Category K/A # Topic Imp. Points !
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| Emergency 2.4.23 Knowledge of the bases for prioritizing 3.1 1 Procedures and 75 emergency procedure implementation Plan during emergency operations.
| |
| 2.4.21 Knowledge of the parameters and logic 3.7 1 100 used to assess the status of safety functions including: 1. Reactivity control 2. Core cooling and heat removal 3. Reactor coolant system l
| |
| integrity 4. Containment conditions 5.
| |
| I Radioactivity release control.
| |
| Total 2 Tier 3 Generic Knowledge and Abilities (RO-13) 13
| |
| | |
| j SRO Facility: WCUS Date of lixam: February 17,1998 Lixam Level: Instant 5110 Tier Group K/A Category Pc ints l Point Total Kl K2 K3 K4 K5 K6 ' Al A2 A3 A4 U
| |
| : l. 1 03U~ W 7F 101 028 025 022 003 017 103 016 009 011 006 102 013 014 012 007 015 018 008 Emergency & 2 ~U32-~D29- W 16 024 005 020 026 104 031 106 021 105 107 034 Abnormal 3 W 3
| |
| , I'lant lier TT ~T T 43 l Evolutions Totals i
| |
| TW W' T 048 064 045 057 040 050 047 039 044 Plant 2 066 053 063 056 052 054 059 055 062 061 17 065 051 067 060 110 071 070 069 4 Systems 3 073 l 068 l l l l 072 Tier 4 1 5 l 6 l 2 l3 4 5 3 4l3 40 Totals l l l l
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| : 3. Generic Knowledge and Cat I Cat 2 Cat 3 Cat 4 Abilities 1I1 116 074 075 17 112 117 120 123 113 118 121 124 114 119 122 125 115 l
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| l
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| l Note:
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| * Attempt to distribute topics among all K/A categories; select at least one topic from every K/A category I within each tier.
| |
| * Actual point totals must match those specified in the table.
| |
| * Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant specific priorities.
| |
| * Systems / Evolutions within each group are identified on the associated outline.
| |
| . The shaded areas are not applicable to the category / Tier. l l
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| . 6 l Facility: WCGS Date of Exam : 2-17-98 Exam Level: SRO Category K/A # Topic Imp. Points 2.1.4 Knowledge of shill staffing 3.4 )
| |
| 111 requirements.
| |
| 2.1.10 Knowledge of conditions and 3.9 112 limitations in the facility license. l Conduct of 2.1.11 Knowledge ofless than one hour 3.8 113 technical specification action statements for systems. ,
| |
| /
| |
| Operations 2.1.13 Knowledge of facility requirements 2.9 115 for controlling vital / controlled access.
| |
| 2.1.12 Ability to apply technical 4.0 114 specifications for a system.
| |
| Total 5 ;
| |
| 2.2.17 Knowledge of the process for 3.5 116 managing maintenance activities during power operations.
| |
| 2.2.22 Knowledge oflimiting conditions 4.1 117 for operations and safety limits.
| |
| Equipment 2.2.27 Knowledge of the refueling process. 3.5
| |
| 'l19 Control 2.2.25 Knowledge of bases in Technical 3.7 118 Specifications for limiting conditions for operations and safety limits.
| |
| Total 4 2.3.2 2.9 74 Knowledge of facility ALARA program.
| |
| Radiation 2.3.4 Knowledge of radiation exposure 3.1 120 limits and contamination control, including permissible levels in excess of those authorized.
| |
| Control 2.3.6 Knowledge of the requirements for 3.1 122 reviewing and approving release permits.
| |
| 2.3.7 Knowledge of the process for 3.3 ,
| |
| 121 preparing a radiation work permit.
| |
| Total 4 l
| |
| l l
| |
| t 19 l
| |
| | |
| i-e D 2.4.18 Knowledge of the specific bases for 3.6 123 EOPs Emergency 2.4.05 Knowledge of the organization of the 3.6 125 operating procedures network for normal, abnormal, and emergency I evolutions l Procedures 2.4.23 Knowledge of the bases for 3.8 075 prioritizing emergency procedure implementation during emergency conditions and Plan 2.4.44 Knowledge of emergency plan 4.0 l
| |
| ; 124 protective action recommendations.
| |
| Total 4 l
| |
| Tier 3 Generic Knowledge and Abilities (SRO-17) 17 1
| |
| 1 20-i I
| |
| | |
| .T To: Howard Bundy(NRC)
| |
| From: George Smith (WCGS)
| |
| Date: 1-15-98
| |
| | |
| ==Subject:==
| |
| Examination submittal Enclosed is the submitted written examination to be conducted for two SRO candidates and two RO candidates the 17th of February. The examination package consists of the following items:
| |
| Examination Key Student Examination version Examination data sheets with references Examination cross-reference As we have planned, I will be at your ofEce 'the 22nd of January to review the test with you. If you have
| |
| - any questions before then, please contact me at (316)364-8831 x5117.
| |
| Thank you.
| |
| l W
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| |
| | |
| RO Facility: WCGS Date of Exam: February 17,1998 Exam Level: RO Tier Group K/A Category Points l Point Total Kl K2 K3 K4 K5 K6 Al A2 A A
| |
| : 1. 1 013 016 004 001 002 010 16 005 009 003 014 012 011 006 015 007 Emergency & T-W 018 022 3 027 T 032 029 020 026 076 017 024 028 019 031 021 Abnormal T-- 077 033 T 034 Plant Tier TT 8 5 9 3 36 Evolutions Totals
| |
| : 2. 1 080 079 042 045 050 078 041 036 049 043 037 23 082 047 035 038 039 083 081 040 046 084 044 048 Plant 2 066 053 063 056 052 054 059 055 060 061 20 064 058 065 051 057 062 086 067 087 085 Systems 3 069 073l068l l 071 l l070 8 072 088 089 Tier Totals 6 2 4 l7 l3 l l 4 l5 6 5 l6 l3 51 l l l
| |
| : 3. Generic Knowledge and Cat 1 Cat 2 l Cat 3 Cat 4 l Abilities 90 95 74 75 13 l 91 96 98 100 92 97 99 93 94
| |
| | |
| l d
| |
| Note:
| |
| * Attemg to distribute topics among all K/A categories; select at least one topic from every K/A category within each tier.
| |
| * Actual point totals must match those specified in the table.
| |
| * Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant specific priorities.
| |
| * Systems / Evolutions within each group are identified on the associated outline.
| |
| l e The shaded areas are not applicable to t!'e category / Tier. 1 l
| |
| I l
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| l Facility: WCGS Generic Knowledge and Abilities Outline (Tier 'i) Exam Level: RO i Category K/A # Topic Imp. Points 2.1.1 Knowledge of conduct of operations 3.7 1 l 90 requirements.
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| Conduct of 2.1.2 Knowledge of operator responsibilities 3.0 1 Operations 91 during all modes of plant operation.
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| 2.1.3 Knowledge of shif turnover practices. 3.0 1 92 2.1.10 Knowledge of conditions and limitations 2.7 1
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| ! 97 in the facility license l 2.1.16 Ability to operate plant phone, paging 2.9 1 l 93 system, and two-way radio. I l 2.1.25 Ability to obtain and interpret station 2.8/3.1 1
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| ; 94 reference materials such as graphs, monographs, and tables which contain 1 performance data, f Total 6 )
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| 2.2.12 Knowledge of surveillance procedures. 3.0 I 95 l
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| Equipment 2.2.13 Knowledge of tagging and clearance 3.6 1 ,
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| Control 96 procedures. l l Total 2 l 2.3.1 Knowledge of 10CFR: 20 and related 2.6 1 98 facility radiation control requirements.)
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| : Radiation 2.3.2 Knowledge of facility ALARA program. 2.5 1 74 Control 2.3.10 Ability to perform procedures to reduce 2.9 1 l 99 excessive levels of radiation and guard l against personnel exposure.
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| Total 3 i
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| i
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| Category K/A # Topic imp. Points Emergency 2.4.23 Knowledge of the bases for prioritizing 3.1 1 l Procedures and 75 emergency procedure implementation Plan during emergency operations. .
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| l 2.4.21 Knowledge of the parameters and logic 3.7 1 {
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| 100 used to assess the status of safety functions including: 1. Reactivity control 2. Core cooling and heat i L removal 3. Reactor coolant system l integrity 4. Containment conditions 5.
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| ; Radioactivity release control. i l Total 2 Tier 3 Generic Knowledge and Abilities (RO-13) 13 l
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| l l
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| I l
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| I I
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| ' .. SRO Facility: WCGS Date of Exam: February 17,1998 Exam Level: Instant SRO Tier Group K/A Category Points l Point Total K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G
| |
| : 1. 1 030 004 001 002 010 24 101 028 025 022 003 017 103 016 009 011 006 102 013 014 012 007 Emergency & 2 W 019 023 W 16 024 005 020 026 104 031 106 021 105 107 034 Abnonnal 3 109 108 033 3 Plant Tier 7 8 8 6 10 4 43 Evolutions Totals
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| : 2. 1 046 042 058 035 038 041 036 049 043 037 19
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| , 048 064 045 057 040 1
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| 050 047 039 044 Plant 2 066 053 063 056 052 054 059 055 062 061 17 065 051 067 060 110 071 070 Systems 3 069 073 068 4 072 i
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| Tier Totals 4 1 5 6 2 l3 4 5 3 4 3 40 l
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| : 3. Generic Knowledge and Cat 1 Cat 2 l Cat 3 Cat 4 Abilities 111 116 074 075 17 112 117 120 123 113 118 121 124 114 119 122 125 115
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| l l
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| Note:
| |
| * Attempt to distribute topics among all K/A categories; select at least one topic from every K/A category within each tier.
| |
| * Actual point totals must match those specified in the table.
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| * Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant specific priorities.
| |
| * Systems / Evolutions within each group are identified on the associated outline.
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| * The shaded areas are not applicable to the category / Tier.
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| 0 4 4 7 7 E 0 0 0B 0 0 0 K 7
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| 1 t l l; l
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| ! ., 1-20-98
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| -- Facility: WCGS Date of Exam : 2-17-98 Exam Level: SRO j Category K/A # Topic Imp. Points !
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| l 2.1.4 Knowledge of shift staffing 3.4 111 requirements.
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| Conduct of 2.1.10 Knowledge of conditions and 3.9 i 112 limitations in the facility license.
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| Operations 2.1.11 Knowledge ofless than one hour 3.8 113 technical specification action statements for systems.
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| l 2.1.12 Ability to apply technical 4.0 114 specifications for a system.
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| Total 4 l 2.2.14 Knowledry of the process for 3.0 115 making enfiguration changes.
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| 2.2.17 Knowledge of the process for 3.5 l 116 managing maintenance activities during power operations. l 2.2.22 Knowledge oflimiting conditions 4.1 i l 117 for operations and safety limits.
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| l Equipment 2.2.27 Knowledge of the refueling process. 3.5 1 1 119 Control 2.2.25 Knowledge of bases in Technical 3.7 118 Specifications for limiting conditions for operations and safety limits.
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| Total 5 2.3.2 2.9 74 Knowledge of facility ALARA program.
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| Radiation 2.3.4 Knowledge of radiation exposure 3.1 120 limits and contamination control, including permissible levels in excess of those authorized.
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| Control 2.3.6 Knowledge of the requirements for 3.1 122 reviewing and approving release permits.
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| 2.3.7 Knowledge of the process for 3.3 121 preparing a radiation work permit.
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| Total 4 l
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| 19
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| l
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| ! _. 2.4.18 Knowledge of the specific bases for 3.6 123 EOPs Emergency 2.4.05 Knowledge of the organization of the 3.6 l
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| ; 125 operating procedures network for normal, abnormal, and emergency 4 evolutions f
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| Procedures 2.4.23 Knowledge of the bases for 3.8 075 prioritizing emergency procedure implementation during emergency conditions and Plan 2.4.44 Knowledge of emergency plan 4.0 124 protective action recommendations.
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| Total 4 Tier 3 Generic Knowledge and Abilities (SRO-17) 17 l
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| l l
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| l 20
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| ES-401 Site-Specific Written Form ES-401-7 Examination Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific Written Examination _
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| Applicant Information Name: Region: IV Date: 2-17-98 Facility / Unit WCGS License Level RO Reactor Type W Start Time: Finish Time:
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| Instructions l
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| Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade l
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| of at least 80.00 percent. Examination papers will be collected four hours after the examination starts.
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| Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
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| Applicant's Signature -
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| Results Examination Value Points Applicant's Score Points Applicant's Grade Percent NUREG-1021 Interim Rev. 8, January 1997
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| | |
| Initial License Examination Answer Sheet Name: Date: 2-17-98 I 1 ABCD 26 ABCD 51 ABCD 76 ABCD 2 ABCD 27 ABCD 52 ABCD 77 ABCD 3 ABCD 28 ABCD 53 ABCD 78 ABCD 4 ABCD 29 ABCD 54 ABCD 79 ABCD 5 ABCD 30 ABCD 55 ABCD 80 ABCD 6 ABCD 31 ABCD 56 ABCD 81 ABCD 7 ABCD 32 ABCD 57 ABCD 82 ABCD 8 ABCD 33 ABCD 58 ABCD 83 ABCD l 9 ABCD 34 ABCD 59 ABCD 84 ABCD 10 ABCD 35 ABCD 60 ABCD 85 ABCD l
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| 11 ABCD 36 ABCD 61 ABCD 86 ABCD i 12 ABCD 37 ABCD 62 ABCD 87 ABCD 13 ABCD 38 ABCD 63 ABCD 88 ABCD l
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| 14 ABCD 39 ABCD 64 ABCD 89 ABCD l 15 ABCD 40 ABCD 65 ABCD 90 ABCD l
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| l 16 ABCD 41 ABCD 66 ABCD 91 ABCD l
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| l 17 ABCD 42 ABCD 67 ABCD 92 ABCD 18 ABCD 43 ABCD 68 ABCD 93 ABCD 19 ABCD 44 ABCD 69 ABCD 94 ABCD 20 ABCD 45 ABCD 70 ABCD 95 ABCD 21 ABCD 46 ABCD 71 ABCD 96 ABCD 22 ABCD 47 ABCD 72 ABCD 97 ABCD 23 ABCD 48 ABCD 73 ABCD 98 ABCD 24 ABCD 49 ABCD 74 ABCD 99 ABCD '
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| 25 ABCD 50 ABCD 75 ABCD 100 ABCD
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| Question Number 001 A The Control Room is performing STS SF-001 " Control and Shutdown Rod Operability Verification" on Control Bank "B".
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| While returning Bank "B" to its fully withdrawn position a Rod Control Urgent alarm (79A) is received and all rod movement :dops.
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| From this information it can be determined that there is a problem in:7 A. The Logic Cabinet.
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| B. The Power Cabinet.
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| C. The Reactor Control Unit.
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| D. One of the Rod Withdrawal Blocks.
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| Question 002 He following plant conditions exist:
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| * Rx power 75%
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| + ANN 00-041 A, SEAL INJ TO RCP FLOW LO is alarming
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| * ANN 00-071C, RCP B THRM BAR CCW FLOW is alarming
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| * RCP "B" No. I seal and bearing water temperature is 235'F Which of the following is the correct action to take?
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| A. Trip RCP "B" and perform Attachment D," Restoration of RCP Seal Cooling".
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| l B. Reduce power to less than 48% and then secure RCP "B".
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| C. Immediately restore seal cooling to RCP "B" using Attachment D," Restoration of RCP Seal Cooling".
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| l D. Trip the reactor and then trip" RCP "B".
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| l l
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| i
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| ~
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| i Question 003 i L
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| l The plant has stabilized at approximately 75% power following a grid disturbance.
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| NB02 is de-energized and OFN NB-030," Loss of AC Emergency bus NB01 (NB02)" is being
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| ! performed.
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| * Ann. 00-081C " Rod Bank LOLO Limit" is Lit.
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| .,.f.,
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| o Control bank 'D' is at 100 steps.
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| * CVCS is in service with the NCP running.
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| * All other equipment is functioning as designed. CVCS is in service with the NCP running.
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| You Have lined up for ' Emergency Boration' and MCB indication BG FI-183A,"EMERG BORATE FLOW", reads zero; Your actions are:
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| A. No action required because BG FI-183A has lost power but boration flow is present.
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| B. Align charging pump suction to the RWST using ' Red Train' RWST & VCT valves (BN HIS-112D & BG HIS-112B).
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| C. Emergency boration is not required due to the plant transient causing rod insertion.
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| D. Commence boration using the reactor make up system.
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| | |
| Question 004A The plant has experienced a complete loss of component cooling water.
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| Which is the preferred source of cooling water to 'A' CCP and 'A' SI pumps?
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| A. GB system, Central Chilled Water B. BL system, Reactor Makeup Water System.
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| C. EA system, Essential Service Water D. KD system, Domestic Potable Water
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| | |
| w Question 005 The Pressurizer master pressure controller has failed to zero output in auto.
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| No operator actions have been performed.
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| RCS pressure will:
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| A. Decrease to PORV interlock setpoint. ,
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| B. Increase to Reactor trip setpoint.
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| C. Stabilize at Pragram setpoint.
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| D. Cycle at the PORV setpoint.
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| I r
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| i l
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| Question 006A l
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| ; While performing diagnostic steps in E-0, the following conditions are observed:
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| l -
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| * RCS T.,, - 536 'F e RCS Press - 1800 psig
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| * PZR LVL - 4%
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| l
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| * . MSIVs are Open
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| * CTMT Temp. - 240 F e - GE RE-92 pre-event was normal e SG A Press /WR Level - 915 psig/35%
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| * SG B Press /WR Level - 915 psig/40%
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| ;
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| * SG C Press /WR Level - ' 915 psig/45%
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| * SG D Press /WR Level - 915 psig/40%
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| e CTMT radiation Levels are normal e Aux. bldg. radiation levels are normal All plant equipment is functioning as designed.
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| ; Which one of the following is the most likely cause of this event?
| |
| A. LOCA inside containment.
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| B. Feedline break outside containment.
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| C. Steam generator tube rupture.
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| D. Steam line break inside containment.
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| l
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| Question 007 l
| |
| l A plant start up is in progress and turbine load is 28%.
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| I Which of the following conditions would require the turbine to be tripped?
| |
| l j A. ANN 00-116B 'COND A VAC LO' is alarming and turbine exhaust pressure is 6 inches
| |
| ; HgA.
| |
| l l B. ALR 00-117C ' STANDBY EHC PMP START' is alarming and EHC fluid pressure is 1500 i psig C. ALR 00-017D 'PG XFMR UV' is alarming and PG-11 bus voltage is zero.
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| D. Turbine bearing oil pressure is 20 psig t
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| Question 008 I
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| i l Initial conditions
| |
| * Train B outage in progress ;
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| l
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| * 'B' CCP tagged out j
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| * 'B' ESW pump tagged out 1
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| * 'B' SI pump tagged out j A station blackout occurs. Power becomes available from the West Bus.
| |
| Which electrical lineups should be placed in service?
| |
| A. NB01 through the alternate feeder breaker.
| |
| B. NB02 through the normal feeder breaker.
| |
| C. NB01 through the alternate feeder breaker and NB02 though the normal feeder breaker.
| |
| D. Neither NB bus, energize PA02 from the West Bus.
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| i l
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| l i
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| l
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| I' Question 009 Step 3.b of EMG C-0," Loss of All AC Power," checks RCS Letdown to Regenerative heat exchanger valves BG HIS-459 and 460 closed.
| |
| Why is Letdown isolated early in EMG C-07 l A. To ensure Containment Integrity is maintained.
| |
| : i. B. To prevent flashing and two-phase flow in the letdown heat exchanger and upstream piping.
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| l l C. To prevent a dilution when hotter letdown flow enters the mixed bed demineralizers.
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| D. To ensure that RCS inventory loss is minimized.
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| i l
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| l I
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| l
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| r - , .
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| Question 010 The plant is at 100% power.
| |
| Due to an inverter problem NN02 was de-energized.
| |
| All required actions of OFN NN-021 " Loss of Vital 120 VAC Instrumentation Bus", have been completed and 'NN02' is energized from its 'SOLA' transformer, i
| |
| Which of the following describes what must be done to allow continued operation?
| |
| A. Perform STS NB-005," Breaker Alignment Verification".
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| B. Restore NN02 to its inverter.
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| C. Ensure all 'A' train components are operable.
| |
| D. Ensure Turbine Driven Aux. Feed pump is operable, i
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| i 1
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| I
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| Question Question 011 A l
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| ! The plant is at 100% power normal line-up with 'A' train components inservice. A problem develops in the 'A' train ESFAS cabinet which causes the following valves to re-position:
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| * EF HV-23 'A' Train SW/ESW cross-connect e EF HV-41 'A' Train SW/ESW cross-connect j With no operator action, which of the following would be the expected plant response?
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| t A. VCT Outlet Temperature t , Seal Return Temperature + , Charging Temperature 4.
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| B. VCT Outlet Temperature t , Seal Return Temperature 4 , Charging Temperature t.
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| C. VCT Outlet Temperature + , Seal Return Temperature 4 , Charging Temperature t.
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| t I
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| D. VCT Outlet Temperature t , Seal Return Temperature t , Charging Temperature t.
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| i l
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| \
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| f u____________________ _ _ . _ . _ _ . _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ .- . . _ _ _ _ _ _ _ _ . _ _____ ._ _ ...
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| Question 012 Per OFN KC-016 " Fire Response", a continuous fire watch is required following a fire onsite if: 3 1
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| A. ' Die fire lasts for longer than 30 minutes.
| |
| 1 B. Offsite assistance was required.
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| C. The fire has been out for 9 hours.
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| D. The fire area is hot to the touch.
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| )
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| 6 l
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| t I
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| ..W
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| _ _ _ _ . _ _ _ _ _ . _ . . _ _ _ . _ _ _ _ _ _ _ _ _ _ U
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| l l
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| l l
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| l Question 013 A The plant was stable in Mode 3 at 488 *F and 1900 psig.
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| A LOCA has occurred and the following conditions now exist:
| |
| * RCS pressure - 1800 psig e CTMT pressure - 4 psig e S/G pressure - 595 psig Which of the following signals should have actuated?
| |
| A. Main Steam Line Isolation.
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| B. Steam Line Safety Injection.
| |
| C. CTMT pressure Safety injection.
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| D. RCS Low pressure Safety injection.
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| Question 014 l
| |
| A loss of coolant accident coincident with a loss of NB02 occurred several hours ago.
| |
| All " Red" train equipment is operating as required.
| |
| The crew has entered EMG ES-12, " Transfer to Cold Leg Recirculation" and are performing step 10
| |
| " Align CCP and SI pump suctions to RHR pump discharge."
| |
| EM HIS-8807A,"CVCS to Si Pump Suction Valve" will not open.
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| Recirculation flow from the CTMT sumps to SIP "A" is:
| |
| A. Not available unless EM HV-8807B is open locally.
| |
| B. Not available unless EJ HV-8804B is opened locally.
| |
| C. Available from "B" train flowpath to suction of St Pump "B".
| |
| D. Available from "A" train flowpath to suction of SIP "A".
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| | |
| Question 015 The following conditions exist:
| |
| e 100 % power
| |
| * SJ RE-01,"CVCS Letdown Radiation Monitor," is inoperable e Containment Purge in progress e 1.3 gpm leakage from BG LCV-459 packing if fuel failure should occur, which of the following monitors would be the last to show an increased radiation iesding?
| |
| A. GT RE-21B Unit Vent Monitor.
| |
| B. GT RE-33 CTMT Purge Monitor.
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| C. GT RE-31 CTMT Atmosphere Monitor.
| |
| D. GT RE-59 CTMT High Range Area Monitor.
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| l I
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| l l
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| ? l Question 016. l 1
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| , 1
| |
| : l. J The plant was operating at full power when offsite power was lost, all equipment operated as required and no Si occurred.
| |
| Power will be restored in 12 hours. 1 Plant staff determines a cooldown is required, the crew should conduct a natural circulation cooldown using EMG ES-04," Natural Circulation Cooldown" at:
| |
| A. < 100 *F per hour and maintain 125 *F subcooling.
| |
| B. < 50 F per hour and maintain 125 'F subcooling.
| |
| C. < 100 F per hour and maintain 75 *F subcooling.
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| D. < 50 F per hour and maintain 75 *F subcooling.
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| | |
| Question 017 Current plant conditions are:
| |
| * Control rods are moving outward e Power 50 % and stable
| |
| . T.,,573 *F initially stable, but increasing slowly e Turbine load 625 MWe and stable e Average loop AT is 50 % and stable e Rods are in AUTO Under these conditions the RO actions are to:
| |
| A. Select Rod Control to MANUAL to stop the rod motion.
| |
| B. Select Rod Control to Group D to stop rod motion.
| |
| C. Monitor rod motion to ensure proper system response.
| |
| D. Monitor rod motion to verify that power remains less than 100%.
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| l 1
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| l l
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| \
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| l L_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - - . - J
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| Question 018 If two or more rods in the same group drop into the core OFN SF-011 " Realignment of Dropped.
| |
| Misaligned Rods (s) and Rod Control Malfunctions" requires the plant be shutdown.
| |
| The reason for this is:
| |
| A. Changes in Xenon concentration would make recovery hazardous.
| |
| B. Technical Specifications requires the plant be in HSB in 6 hours.
| |
| C. Inadequate shutdown margin remains in event of a reactor trip.
| |
| D. Two rods could not be recovered in the one hour allowed.
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| l l
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| l 4
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| r
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| _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . ._ _ _ _ _ _ _ . . _ _ _ J
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| Question 019 A Reactor trip from 100% power with no Safety Injection has occurred.
| |
| 1 The transient that caused the trip was initiated by the loss of the "C" Condensate Pump.
| |
| The Main Feed Water Pumps tripped due to:
| |
| 1 A. The trip of the 'C' Cond. sate Pump.
| |
| B. The resultant AFAS(M) actuation signal.
| |
| C. P-4, Reactor Trip Permissive.
| |
| D. He resultant Feed Water Isolation Signal.
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| l l
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| l t
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| i i
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| l _
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| 1 l
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| Question 020A The following conditions exist:
| |
| * Pzr. pressure - 1985 psig
| |
| * Pzr temperature - 636 *F e RCS temperature - 364 *F e PORV BB PV-455 A indicates open l Which parameter set is correct?
| |
| PRT Press. PRT Temp.
| |
| A. 40 psig 267 F B. 25 psig 267*F I C. 25 psig 636 *F j D. 145 psig 364 *F l
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| l l
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| 1 l
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| l
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| Question 021 RCS leakage has increased to 0.9 gpm. The increase is coming from a weld where the BB FI-426 sensing line penetrates the loop.
| |
| Which of the following is required?
| |
| A. Plant operations can continue, identified leakage is less than 10 gpm.
| |
| B. Plant operations can continue, total leakage is less than I gpm.
| |
| C. Commence a plant S/D as this leakage could affect the leakage detection systems.
| |
| D. Commence a plant S/D as this is pressure boundary leakage.
| |
| | |
| l Question 022 l
| |
| A plant event has occurred e RCS Wide Range pressure - 1500 psig and decreasing
| |
| * RCS Wide Range Temperature - 557*F and stable
| |
| * Main Steam Header Pressure - 1092 psig and stable l
| |
| * S/G "A" WR level- 58% and increasing e S/G "B" WR level- 56% and increasing
| |
| * S/G "C" WR level- 57% and increasing
| |
| * S/G "D" WR level- 59% and increasing
| |
| * Containment Pressure - 7 psig and decreasing l
| |
| l Based on the above Main Control Board indications what has occurred?
| |
| l l A. A feedline break.
| |
| I B. A steamline break.
| |
| C. A reactor coolant system break.
| |
| D. A steam generator tube rupture.
| |
| 1 l
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| l l
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| Question 023 Initial Conditions:
| |
| * Mode one,100% power, NOT/NOP e NCP tagged out for maintenance
| |
| * 'A' CCP is in service The 'A' CCP trips and cannot be restarted. The only operator action is to start the 'B' CCP reestablishing seal injection flow at a minimum of 32 gpm.
| |
| With no further operator action which of the following adverse events would occur?
| |
| A. Reactor Trip on high pressurizer level.
| |
| B. Reactor trip on low pressurizer pressure.
| |
| C. Loss of flow through the CVCS demineralizers resulting in increasing RCS activity.
| |
| D. Overheating of the letdown flow resulting in bypassing the CVCS demineralizers on high temperature.
| |
| l l
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| [
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| I 1
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| i f
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| 4
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| ,#4 Question 024 The following plant conditions exist:
| |
| * The plant has been shutdown for 10 days after 100 days of full power operation
| |
| * Plant is in Mode 5
| |
| * The RCS is drained to mid loop for RCP seal work
| |
| * The RCS is vented to atmosphere
| |
| * Heat removal is being provided by RHR Using the Attached pages, which of the following is the closest time it will take for the core to begin to uncover if RHR is lost.
| |
| A. 14 minutes.
| |
| B. 32 minutes.
| |
| C. 125 minutes.
| |
| l D. 263 minutes.
| |
| l l
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| l 1
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| i 1
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| I I
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| I
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| ]
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| Question 025 A reactor trip has occurred from full power.
| |
| The crew is performing the immediate actions of EMG E-0 " Reactor Trip Or Safety injection" The following items are observed:
| |
| * 'A' trip breaker is oper.
| |
| * 'B' trip breaker is closed
| |
| * 3 rods are stuck out = mid core
| |
| . SE NI-35D SUR -l/3 DPM e SE NI-36D SUR -1/3 DPM e SE NI-35B 4 x 10-5 amps
| |
| * SE NI-36B 5 x 10-5 amps Which of the following actions must be performed?
| |
| A. Manually trip the reactor.
| |
| B. Attempt to manually drive the stuck rods in.
| |
| C. Continue with step 2 in EMG E-0 " Reactor Trip Or Safety Injection".
| |
| I D. Transition to EMG FR-SI " Response to Nuclear Power Generation /ATWT".
| |
| l l
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| l
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| l l
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| Question 026 l
| |
| l l The crew is responding to a Safety Injection caused by a SGTR on the S/G "B".
| |
| Assume all signals actuated as required and none have been reset.
| |
| Which of the following radiation monitors will continue to give radiation readings consistent with the accident?
| |
| l A. SJ RE-2, S/G sample monitor.
| |
| B. GE RE-92, Condenser air discharge monitor.
| |
| C. BM RE-25, S/G blowdown radiation monitor. ,
| |
| l D. FC RE-381, TDAFW pump turbine exhaust monitor.
| |
| l l
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| 1 l
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| I l l
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| ! 4 i
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| i 1
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| . i
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| m Question 027A EMG E-3 " Steam Generator Tube Rupture", is being performed in response to a tube rupture on S/G "A".
| |
| l The cooldown hasjust been completed but the target temperature value selected by the crew was higher than that stipulated in the procedure.
| |
| This error could result in which one of the following conditions?
| |
| A. Filling the Pressurizer solid during the subsequent depressurization.
| |
| l B. ' Decrease the time for termination of the primary to secondary leakage.
| |
| C. Decrease in pressure of the ruptured S/G with increased leakage from the RCS.
| |
| [ D. Loss of RCS subcooling before RCS and ruptured S/G pressures are equalized.
| |
| )
| |
| i
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| l l
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| l l
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| , Question 028 l
| |
| l l A plant start up was in progress per GEN 00-003, " Hot Standby to Minimum Load," when an
| |
| ; inadvertent Safety injection occurred.
| |
| All equipment functioned as designed except the "B" train trip breaker did not open.
| |
| The initiating signal has been cleared and the SI reset switches have been depressed.
| |
| l l If a valid auto Si signal is received, it will actuate:
| |
| l l A. Both trains.
| |
| t B. "A" train only.
| |
| l C. "B" train only.
| |
| D. Neither train.
| |
| l 1
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| l ..
| |
| Question 029A A small break LOCA occurred at full power. The crew is performing EMG ES-Il " Post LOCA Cooldown and Depressurization", and are at step 21 " check if one CCP should be stopped".
| |
| Plant conditions were:
| |
| * SI pumps running - 2
| |
| * RCPs running- 1
| |
| * Subcooling - 90 *F '
| |
| * PZR level- 25% .
| |
| The crew secured 'A' CCP.
| |
| The following plant parameters were reported after the CCP was secured.
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| * RCS temperature-426 F e RCS pressure - 415 psig
| |
| * PZR level- 10% and decreasing '
| |
| Based on this information what can be stated about the LOCA?
| |
| A. It has not changed.
| |
| B. It has decreased in size.
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| C. It has increased in size.
| |
| D. It has been isolated, l
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| l
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| .__ . _ _ _ _ _ _ ________________ _ ____ a
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| | |
| Question 030 De plant experienced a LOCA in the Aux Building and the crew is performing EMG C-12, "LOCA Outside Containment". The RO has closed EJ HV-8809A "RHR to accumulator i injection loops 1 & 2". RCS pressure is reported to be stable.
| |
| What does this tell you about the break?
| |
| A. He break was on the accumulator injection line and has been isolated.
| |
| B. The break was on the RHR injection line and has been isolated.
| |
| C. The break location is not known but has been isolated.
| |
| D. The break location is not known and is not isolated.
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| f-Question 031 l
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| Initial Conditions:
| |
| e The crew is performing ' Bleed & Feed" per EMG FR-H1 " Response To Loss Of Secondary Heat Sink",
| |
| e S/G "A" level - 9% WR i
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| e S/G "B" level - 8% WR e S/G "C" level - 10% WR e S/G "D" level - 6% WR e CTMT pressure - 3.5 psig -
| |
| e RCS Thot - 553*F and decreasing "A" MDAFW pump is restored to service.
| |
| Which of the following is the expected course'of action?
| |
| A. Continue bleed & feed until Thot < 550 F.
| |
| B. Commence feeding all S/Gs at 40,000 lbm/hr.
| |
| C. Commence feeding" S/Gs "B" & "C at 40,000 lbm/hr.
| |
| D. Feed S/G "C" as necessary to restore narrow range level.
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| Question 032 l
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| ! The purpose of Attachment C " Establishing Minimum ECCS Flow to Remove Decay Heat", of EMG C-11 " Loss of Emergency Recirculation", is to provide minimum ECCS flow required to:
| |
| . A. Restore emergency coolant recirculation capability.
| |
| B. Prevent heat up and minimize RWST depletion.
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| C. Cool down and minimize RWST depletion.
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| D. Commence a cool down.
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| Question 033 At EOL a loss of power has occurred. EMG C-0," Loss of All AC Power,"is in progress and the intact steam generators are being depressurized to 260 psig.
| |
| The Reactor Operator reports a positive startup rate on the source and intermediate range channels.
| |
| l What action is required?
| |
| A. Control the S/G ARVs or TD AFWP to stop the depressurization and allow the RCS to heat up.
| |
| B. Control the S/G ARVs or TD AFWP to stabilize temperature until Xenon builds in.
| |
| C. Slow down the cooldown rate to 50 *F/hr to allow Xenon to maintain reactor suberitical.
| |
| D. Increase the cooldown rate so ECCS accumulators can inject before point of adding heat is reached.
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| | |
| Question 034A Initial Conditions:
| |
| * The ur.it is at 100% reactor power.
| |
| * KA HV-29 " Instrument Air to Containment," fails closed.
| |
| With no operator action 24 hours after the event pressurizer level will:
| |
| A. Stabilize at program level.
| |
| B. Decrease to a lower level then stabilize.
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| C. Increase to a new higher level then stabilize.
| |
| D. Continue to increase until the reactor trip setpoint is reached.
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| Question 035
| |
| ' Which of the following conectly describes the reasons for Rod Insertion Limits?
| |
| A. Ensures acceptable power distribution limits and minimum shutdown margin are maintained.
| |
| B. Ensures the reactivity transients associated with postulated accident conditions are controllable within acceptable limits.
| |
| C. Ensures that negative reactivity control is available during all modes ofoperation.
| |
| l D. Ensures the radial flux difference is limited to maintain the minimum DNBR in the core at or above the safety analysis DNBR limits.
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| Question 036 l During full power operation a spurious Phase A Isolation occurs. No operator action is taken.
| |
| Which of the following describes the effect upon Loop D RCP seal operation?
| |
| i A. No. I seal leakoff flow is stopped.
| |
| l B. No. 2 seal leakoff flow rate increases.
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| i 1 C. No. 3 seal standpipe makeup rate increases. '
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| D. Seal injection flow is stopped.
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| Question 037 CCW flow to the letdown heat exchanger is reduced by half.
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| The resultant plant response is?
| |
| (assume no protective actuation's occur and that temperature limits are not exceeded or alarm)
| |
| A. RCS Tavg will decrease.
| |
| B. RCS Tavg will increase.
| |
| C. Reactor power will decrease.
| |
| D. Turbine MWe output will increase.
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| t ,,
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| Question 038 The Cation Demineralized saturated at MOL is placed in service at EOL during coastdown.
| |
| l Assume rod control is automatic and control rods at: parked at 232 steps.
| |
| l Which one of the following will occur? ,
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| j A. The control rods will step inward.
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| B. The control rods will step outward. l C. Primary coolant pH will decrease.
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| l l D. No control rod motion will occur i
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| ! Question 039 1
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| I i
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| i: The plant is operating at full power with 'A' train CCW supplying the service loop. An event has 1 occurred and the following is observed:
| |
| * 'A' & 'C' CCW pumps are running e BB HIS-13,14,15, & 16 are open i e EG HIS-62 is closed '
| |
| What has occurred?
| |
| A. RCP thermal barrier leak.
| |
| B. Loss of battery charger NK21.
| |
| C. Loss of NG03.
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| D. Loss ofNN01.
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| l Question 040A l
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| During a normal controlled cooldown we can block the Low Steamline Pressure SI signal when RCS pressure is Excess cooldown protection is provided by .
| |
| l A. < 1920 psig ; Hi-I containment pressure.
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| B. < 1920 psig ; Negative Steamline Pressure rate.
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| C. < 1830 psig ; Hi-I containment pressure.
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| D. < 1830 psig ; Negative Steamline Pressure rate.
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| .d
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| Question 041 The following condithns exist:
| |
| * Reactor power is 10%
| |
| e Xenon is at equilibrium
| |
| * Group D rod position is 223 steps
| |
| * Rods are in MANUAL A power reduction to 50% is performed using boration (in progress) and maintaining Tavg ni accordance with program.
| |
| With no rod motion, which cf the following is the expected response of Axial Flux Differene,(AFD) to the power reduction?
| |
| A. Remains the same.
| |
| B. Becomes more positive.
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| \
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| C. Becomes more negative.
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| D. Deviates between channels.
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| | |
| Question 042 l
| |
| Given the following:
| |
| * Reactor power is 3 %
| |
| e Reactor / Plant Startup in progress.
| |
| * NI-35 has failed high.
| |
| * NI-36 is operable.
| |
| The crew should:
| |
| A. Verify reactor trip.
| |
| B. Maintain power less than 10% to prevent a reactor trip.
| |
| C. Maintain power less than 25% to prevent a reactor trip.
| |
| D. Maintain power less than 48% to prevent a reactor trip.
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| l Question 043 The following conditions exist:
| |
| * A LOCA has occured e The operators are performing EMG E-1, " Loss of Reactor or Secondary Coolant".
| |
| * "D" RCP is running.
| |
| Under which of the following conditions would an inadequate Core Cooling condition be diagnosed requiring entry into EMG FR-Cl, " Response to Inadequate Cere Cooling"?
| |
| A. Five core exit thermocouple indicate greater than 712 'F.
| |
| B. Any core exit thermocouple indicates greater than 712 *F.
| |
| C. Five core exit thermocouple indicate greate than 1200 *F.
| |
| D. Any core exit thermocouple indicates greater than 1200 *F.
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| i Question 044 ,
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| j l
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| Which one of the functions listed below is NOT one of the functions of the Containment Cooling System?
| |
| l l
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| l A. Maintains temperature less than 120 F during normal plant operations.
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| l
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| ! B. Provides a means of cooling the containment atmosphere to reduce pressure.
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| C. Provides cooling to the nuclear instrumentation system.
| |
| D. Maintains an air cleaning train to reduce particulate and iodine in the CTMT atmosphere.
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| l Question 045 i
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| 'Ihe crew manually trips the reactor, which of the following will generate a direct feedwater isolation signal (FWIS)?
| |
| A. Opening of the reactor trip breakers.
| |
| B. Shutting the main turbine stop valves.
| |
| C. . Tavg n i one loop decreasing below 564 F.
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| D. Level in one S/G decreasing to 23.5%.
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| Question 046B l
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| A plant fire has caused the loss of all Train "A" vital AC and forced the evacuation of the Control Room.
| |
| The plant is being controlled from the ASP, On a low Condensate Storage Tank level the suction to the AFW pumps will:
| |
| A. Automatically transfer both train suctions to ESW.
| |
| B. Automatically transfer only "A " train MDAFW pump suction.
| |
| C. Automatically transfer only"B " train MDAFW pump suction.
| |
| D. Not automatically transfer the MDAFW suction valves to ESW.
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| i I
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| l t . ---__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
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| ( Question 047 The following condition exist:
| |
| * The unit is in mode one at 8% power
| |
| . NB01 normal supply breaker trips and the bus is repowered from NE01 i What is the expected lineup of the auxiliary feedwater system following the transient?
| |
| "A" AFWP "B" AFWP TD AFWP A. running running not running B. running not running running C. not running not running running '
| |
| D. not running running not running 1
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| 1
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| Question 048 Reactor vessel flange leakoff accumulates in which of the following?
| |
| : A. Pressurizer Relief Tank.
| |
| B. Instrument Tunnel sump.
| |
| 1 C. Containment Normal sumps.
| |
| D. Reactor Coolant Drain Tank.
| |
| | |
| Question 049 f.
| |
| l Discharge of a Waste Gas Decay tank (WGDT) is in progress.
| |
| : l. In the event of a High alarm on SP-056A for GH RE-10A,"Radwaste Building Vent Monitor", the j following will occur:
| |
| ! A. The suction to GH RE-10B is isolated.
| |
| B. The Waste Gas Decay Tank discharge is terminated.
| |
| C. Operation of GH RE-10B is unaffected at the High alarm level.
| |
| D. Suction of GH RE-108 is switched to the room in which it's located.
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| I Question 050 i
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| i
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| ! While performing the Emergency Operating Procedures, a step is encountered which states, i
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| 1 l
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| . " Maintain PZR level between 22%[50%) and 70%[55%)." i i
| |
| Containment pressure spiked to 11 psig before dropping to 4.5 psig. Containment radiation levels rose to i
| |
| 3E+5 R/Hr before dropping to SE+3 R/Hr. He TSC reports the integrated dose is SE+5 R.
| |
| What is the minimum and maximum indicated PZR levels that must be maintained?
| |
| l
| |
| ! A. Minimum of 22%; Maximum of 70%. !
| |
| B. Minimum of 22%; Maximum of 55%.
| |
| C. Minimum of 50%; Maximum of 70%.
| |
| 1 D. Minimum of 50%; Maximum of 55%. i l
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| __-_____-___m
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| l Question 051 The following plant conditions exist:
| |
| * Reactor power- 100%
| |
| * Loop 4 AT - off-scale low l Loop 4Tavg - 624 'F l 1
| |
| Which of the following would cause these indications?
| |
| Which of the following conditions in Loop 4 would cause these indications? {
| |
| A. A T hot 1
| |
| .TD failing low.
| |
| B. A T ot h RTD failing high.
| |
| C. A Tcoid RTD failing low.
| |
| D. A Tcold RTD failing high.
| |
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| i L-__________ .
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| j
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| i l Question 052 l
| |
| ! A start up is in progress, with the following conditions:
| |
| * Steam dumps are in AUTO and group i is open = 75%
| |
| * Turbine load 200 MWe
| |
| * SE NI-41 20%
| |
| * SE NI-42 19%
| |
| * SE NI-43 21%
| |
| * SE NI-44 20%
| |
| * Loop i AT 23%
| |
| l
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| * Loop 2 AT 24%
| |
| * Loop 3 AT24%
| |
| * Loop 4 AT 25%
| |
| Actual reactor power is approximately:
| |
| A.7.5%
| |
| B. 16%
| |
| C. 20%
| |
| D 24%
| |
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| Question 053D The following sequence of events occurs:
| |
| * A small break LOCA has occurred
| |
| . A Safety Injection Signal has started all ECCS pumps
| |
| . The Emergency Diesel generators are running normally
| |
| * Per EMG E-1 the RHR pumps were secured a 45 minutes after the SIS, offsite power is lost
| |
| . The SIS has been reset Which one of the following describes the expected response of the RHR pumps to the loss of offsite power?
| |
| A. The RHR pumps restart when their train's D/G breaker closes.
| |
| B. The LOCA Sequencer will restart the RHR pumps.
| |
| C. The Shutdown Sequencer will restart the RHR pumps.
| |
| D. The RHR pumps must be started by operator action.
| |
| (
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| l Question 054 The pressurizer pressure control system is lined up such that BB PT-455 and BB PT-458 are selected on BB PS-455F.
| |
| BB PT-458 fails to 2500 psig. No actions have been performed.
| |
| i Which of the below describes the plant response?
| |
| l l A. Pressure will stabilize at Program setpoint.
| |
| B. Reactor will trip on Low Pressurizer Pressure.
| |
| l C. Reactor will trip on High Pressurizer Pressure.
| |
| D. Pressure will cycle at the PORV interlock setpoint.
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| I.
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| _ _ - _ _ - _ - _ _ _ - - - - _ - _ - _ - - _ _ _ _ _ _ ~
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| | |
| Question 055 l Annunciator 042A "CHG LINE FLOW HILO" alarms.
| |
| i The following condition exist:
| |
| * Mode 1,100% power.
| |
| . "A" CCP is in service.
| |
| * Charging header flow indicates 170 gpm.
| |
| * PZR level is'60% and increasing.
| |
| CCP Discharge Flow Control BG FK-121 is in ALTTO and indicates 90% open.
| |
| BG FK-121 will not respond n manual.
| |
| Which one of the following actions should the Control Room take?
| |
| A. Shift to the Normal Charging Pump.
| |
| l B. Shift to the standby CCP.
| |
| C. Initiate Excess Letdown. l i
| |
| D. Take local control of BG FCV-121.
| |
| 1 i
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| _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ . _ _ _ _ _ _ .J
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| l l -
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| f Question 056 l The following conditions exist:
| |
| * Reactor power is 6 %.
| |
| * RCS pressure is 1900 psig.
| |
| Which one of the following events will cause a reactor trip?
| |
| A. The reactor should already be tripped.
| |
| l B. One Turbine Impulse Pressure Channel fails high.
| |
| C. Loss of power to PA01 or PA02.
| |
| D. One Power Range Channel fails high.
| |
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| L Question 057 1
| |
| While realigning a Control Bank D rod that was misaligned high, the Auto-Manual switch at the Pulse to Analog converter was not held in MANUAL.
| |
| The rod bank LoLo limit alarm would:
| |
| A. Come in sooner than required.
| |
| B. Come in later than required.
| |
| C. Not alarm.
| |
| D. Lock in.
| |
| l l
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| | |
| A Quest ion 058 A valid Containment Spray Signal (CSAS) was received and has not been reset.
| |
| When can the Spray Additive Tank isolation valves be closed by the operator?
| |
| A. Spray Additive Tank low level setpoint is reached.
| |
| B. Spray Additive Tank low-low level setpoint is reached.
| |
| C. RWST low level setpoint is reached.
| |
| D. RWST low-low level setpoint is reached.
| |
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| l l m Question 059 l
| |
| Containment Mini-Purge was in progress when a failure in GT RE-22 isolated the purge. The channel is still reading above the Hi-Hi alarm setpoint.
| |
| What is required to restart the purge?
| |
| A. GT RE-22 must be restored to operable.
| |
| B. Either bypass GT RE-22 or reset CPIS.
| |
| C. Bypass GT RE-22 then reset CPIS.
| |
| D. Verify GT RE-31 & 32 operable then reset CPIS.
| |
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| l Question 060 .
| |
| l
| |
| ! Spent Fuel Pool Cooling Pump "A" is in service when a leak causes SFP level to rapidly decrease.
| |
| l Which of the following would be the first indication an indication of decreasing level in the SFP7 A. Amber light on Spent Fuel Pool Cooling Pump "A" handswitch.
| |
| l B. Annunciator 75F, SFP Cooling Pump"A" Low Flow.
| |
| C. Annunciator 75E, SFP Cooling Pump "A" Trip.
| |
| l D. Annunciator 75D, SFP Temperature Hi.
| |
| l a
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| l Question 061 l
| |
| ! )
| |
| l l A power ascension from 60 to 100 % power hasjust been completed when Annunciator 061B, t
| |
| " PROCESS RAD HI", alarms. l l The following conditions exist:
| |
| e i GT RE-92, Condenser Air Removal monitor, is in alert, the increase occurred over the past
| |
| : hour.
| |
| l e SJ RE-01, CVCS Letdown Monitor, increased =10% during the power up ramp.
| |
| ' l e BM RE-25 and SJ RE-02, SG Blowdown Monitors, have increased over the past hour but have not alarmed.
| |
| * SG levels and feedflows have remained constant.
| |
| * Pressurizer level, charging and letdown show no change.
| |
| The probable cause of the increased reading on GT RE-92 is:
| |
| 1 A. S/G Tube Rupture.
| |
| B. S/G Tube leak C. Crud burst D. Monitor failure I
| |
| 1 l
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| I J
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| I i
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| m Question 062 The following plant conditions exist:
| |
| * Mode 1,100% Reactor Power
| |
| * STS AB-201 A," Main Steam Isolation and Bypass Valve Inservice Valve Test," is in progress e Yellow Train Exercise Actuate pushbutton for AB HV-11 ("D" S/G MSIV)is depressed A Main Steam Isolation Signal is received aRer the actuate pushbutton is depressed.
| |
| AB HV-1I will close:
| |
| A. Along with MSIV's for the other S/Gs.
| |
| B. Only if manually closed by the operator.
| |
| C. Only aRer the operator releases the test pushbutton.
| |
| D. When "D" S/G MSIV returns to its full open position.
| |
| 1 J
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| Question 063 De following conditions exist:
| |
| * CCW train B inoperable requiring plant shutdown
| |
| * 555 'F Mode 3,NOP
| |
| * A fault occurs in the No.1 ESF transformer Which of the following will occur?
| |
| A. CCW pumps "A" and "C" will auto-start.
| |
| B. All Aux. Feedwater Pumps will auto-start.
| |
| C. ESW trains will isolate from Service Water.
| |
| D. Cooling will be lost to Reactor Cooling Pumps.
| |
| l 1
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| l l !
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| l l-L___.____________.________________. _ _ _ _ _ _ _ . . _ . _ _ _ )
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| | |
| 4 Question 064 EDG NE01 is supplying bus NB01 when NK01 is lost. Which of the following describes the effect that this event will have on the EDG7 Emergency Diesel Generator will:
| |
| A. Continue to run until fuel in the day tank is used up.
| |
| B. Continue to run but cannot be controlled from the MCB.
| |
| C. Trip on mechanical overspeed since speed control is lost.
| |
| D. Trip since the shutdown solenoid valve is de-energized.
| |
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| ~
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| \
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| Question 065 Fuel oil level is maintained in the standpipe above the Day Tank for the Emergency Diesel Generators to ;
| |
| ensure:
| |
| l A. A positive pressure is applied throughout the engine fuel oil system.
| |
| B. A positive suction head on the shaft driven fuel oil pump. ;
| |
| 1 C An hours worth of fuel for the diesel at rated load + 10%.
| |
| , D. Adequate fuel is supplied during periods of high demand.
| |
| t
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| {
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| l l
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| l Question 066A l
| |
| The following plant conditions exist:
| |
| * Mode 1, NOP, NOT.
| |
| l e Containment Purge in progress. ,
| |
| l Annunciators 061B " PROCESS RAD HI" and 061 A " PROCESS RAD HI-HI" alarm.
| |
| t At SP 056 the Fuel Building Exhaust monitor, GG RE-27 is in red.
| |
| l Which of the following automatic actions would be expected to occur in this situation?
| |
| A. Containment Purge Isolation Signal (CPIS)
| |
| B. Control Room Ventilation Isolation Signal (CRVIS).
| |
| C. Containment Isolation Signal, Phase A (CISA). )
| |
| D. There are no automatic actions associated with this monitor.
| |
| I i
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| l l
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| \
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| Question 067
| |
| ^
| |
| l The following conditions exist: ,
| |
| e Plant is stable at 28% power {
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| l e Lake temperature is 45 'F .
| |
| * Circ Water pumps "A" & "C" are running.
| |
| e "B"Cire Water pump is tagged out e Condenser pressure is 2.5" HgA.
| |
| j "A" Cire Water pump trips, condenser pressure stabilizes at 6" HgA.
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| l
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| ' Assume all automatic actions occur, which of the following is required?
| |
| i A. No action is required unless condenser vacuum decrease funher.
| |
| B. Isolate four of six circulating water flow paths within 2 minutes.
| |
| C. Manually runback the turbine until condenser pressure < 4" HgA.
| |
| ! D. Manually trip turbine and break condenser vacuum when <l200 RPM.
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| 1 Question 068A l Before placing RHR in service in EMG ES-11 " Post LOCA Cooldown and Depressurization," the operator is directed to ensure the RWST switchover SI signal is reset.
| |
| Why is resetting the RWST switchover SI signal required?
| |
| A. RHR suction cannot be aligned to the Hot Leg.
| |
| l B. RCS could drain if RWST level dropped below 36%.
| |
| C. RWST cannot be isolated from the RHR system.
| |
| D. RHR outlet temperature cannot be controlled.
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| l 1
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| Question 069 l
| |
| l Under " Precautions / Limitation" in SYS EJ 120 "Startup of a RHR Train", Step 4.2 states: ,
| |
| l-
| |
| "CCW supply to the RHR Heat Exchangers should not be throttled during cooldown of the RCS" The reason for this precaution is to prevent:
| |
| A. Void formation in the RMR System.
| |
| B. Void formation in the CCW System.
| |
| C. Overheating RHR Pump seal.
| |
| D. Low flow cavitation in the CCW pump.
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| 1 j
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| _ _ - _ _ - _ _ _ - _ - _ _ _ _ - - _ - _ _ _ _ - - - _ _ _ _ _ _ - - _ _ - N
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| l Question 070 l
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| l j The plant has experienced a large LOCA, Hydrogen Recombiner "A" is being placed in service.
| |
| l Given the following and using the attached pages from SYS GS-120, determine the pressure factor (Cp).
| |
| , e Pre LOCA Containment temperature - 100 *F .
| |
| I e Post LOCA Containment temperature - 120 'F l
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| * Pre LOCA pressure 18" H2O e Post LOCA pressure 20 psia l
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| l A. 1.26 i
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| B. 1.24 C. 1.20 l D. 1.16 1
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| i
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| | |
| l' Question 07)
| |
| CTMT Closure is set. The following conditions exist.
| |
| * SFP level on LI-29A indicates 0" e Refueling pool level is 2045' 11" l
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| l After opening EC-V995, the operator reports that the Transfer Canal level is decreasing.
| |
| The Control Room notes that the SFP Pool level decreased to - 2".
| |
| The Instrument Sump level has not changed.
| |
| l The probable cause of the level change is:
| |
| A. Levels in the Refueling Pool and SFP equalizing.
| |
| B. Leakage around the cavity seal ring.
| |
| C. Negative pressure in the CTMT.
| |
| D. Positive pressure in the CTMT.
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| i l
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| l-
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| _________________________j
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| i r
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| Question 072 l
| |
| ! The unit is operating at 38% power with all systems in automatic. A subsequent problem in the turbine controls results in a rapid turbine load reduction to 8%.
| |
| How will the steam dump system respond to this turbine control system failure? Steam dumps will:
| |
| A. Have an open demand signal, but will not arm. Steam dumps will remain closed.
| |
| l B. Open and modulate closed to lower Tavg to no-load Tavg-C. Arm, but will not have an demand signal. Steam dumps will remain closed.
| |
| D. Open and modulate closed to lower Tavg ot a value a few degrees above no-load Tavg-i
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| | |
| 1 l
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| Question 073 The following conditions exist:
| |
| e Mode 3, NOP e 557 'F being maintained by the steam dumps in AUTO e Stan Up feed pump in service Which one of the following describes the effect that a loss ofInstrument Air pressure will have on RCS temperature and S/G level control?
| |
| A. RCS temperature will increase, S/G levels are unaffected.
| |
| B. RCS temperature will decrease, S/G level will increase.
| |
| l C. RCS temperature will increase, S/G level will decrease.
| |
| l D. RCS temperature and S/G levels will decrease.
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| l l
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| L_--_-_--_-_----- -
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| Question 074 While touring the Radwaste Bldg. you notice water dripping from a valve body which has its bonnet removed.
| |
| The work area is separated from the surrounding area by HP boundary tape and is posted as a Contaminated Area..
| |
| Which of the following actions should NOT be performed?
| |
| A. Immediately notify HP and the Control Room.
| |
| B. Confine and cover the spill.
| |
| C. Leave the area.
| |
| D. Monitor yourself for contamination.
| |
| l l
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| l l Question 075 l
| |
| A steam break due to a failed open safety valve on S/G "A" has occurred.
| |
| , i
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| ( EMG E-2," Faulted S/G isolation" has been completed on S/G "A".
| |
| l S/G "A" has not blown dry and RCS and S/G are still decreasing.
| |
| l The crew has returned to EMG E-1, " Loss of Reactor or Secondary Coolant" and have reached the step j that has them check RCS and S/G pressures.
| |
| What action is required?
| |
| l 1
| |
| A. Stay in EMG E-1 and retum to step 1.
| |
| B. Stay in EMG E-1 and continue with the procedure.
| |
| C. Return to EMG E-2 until S/G "A" is depressurized.
| |
| l D. Return to EMG E-2 and check for additional faulted S/Gs.
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| t l
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| Question 076 1
| |
| The plant had been operating at full power for 102 days when a reactor trip occured as the result of a loss of power to NK04.
| |
| The bus is faulted and cannot be reenergized. -
| |
| At what temperature will Tavg stabilize?
| |
| A. 550 'F. -
| |
| B. 557 'F.
| |
| 1 C. 561 "F.
| |
| D. 567 'F.
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| 6 l
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| Question 077 A fuel handling accident has occurred inside the Fuel Building . Fuel Building Radiation Monitors, GG RIC 27 and GG RIC-28, have gone into HiHi alarm.
| |
| What is the expected Fuel Building Ventilation alignment as a result of this accident?
| |
| A. Emergency Exhaust Fans RUNNING, Fuel Building Supply Fans RUNNING, Aux / Fuel Emergency Exhaust Trains IN SERVICE.
| |
| B. Emergency Exhaust Fans RUNNING, Fuel Building Supply Fans OFF, Aux / Fuel Emergency Exhaust Trains IN SERVICE.
| |
| C. One Emergency Exhaust Fans OFF, One Emergency Exhaust Fan RUNNING in recirculation, Fuel Building Supply Fans OFF, One train Aux / Fuel Emergency Exhaust Train IN SERVICE.
| |
| A. Emergency Exhaust Fans OFF, Fuel Building Supply Fans RUNNING, Aux / Fuel Emergency Exhaust Trains BYPASSED.
| |
| l l
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| i Question 078 i
| |
| The plant is returning to full power and has been stabilized at 50%.
| |
| The following conditions exist
| |
| . Power Range channel N 51%
| |
| * Power Range channel N 48%
| |
| * Power Range channel N 50%
| |
| * Power Range channel N 50%
| |
| * Control rods are in AUTOMATIC Power Range N-42 lower detector fails high.
| |
| Which of the following will occur? l q
| |
| I A. Control rods will step in until rate of change between nuclear power and turbine power is zero. ]
| |
| B. Control rods will step out until rate of change between nuclear power and turbine power is zero.
| |
| C. Control rods will continue to step in until the bank selector switch is placed to MANUAL.
| |
| D. No rod movement will occur as power range channel N-42 was the lowest reading channel.
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| I l
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| ^
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| l
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| ! Question 079
| |
| , The plant was stabilized at 20% power when the Reactor Operator noticed window "RCP A UF" lit on SB-069, " Partial Trip Status Permissive / Block Panel".
| |
| i l Flow indication is normal for all four reactor loops.
| |
| l Which of the following will cause a reactor trip?
| |
| ! A. Failure of RCP "A" UV relay. .
| |
| l B. Failure of RCP "B" UF relay.
| |
| l l C. Failure of RCP "C" UF relay, i
| |
| D. Failure of RCP "D" UV relay.
| |
| i l-l l
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| l
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| [
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| . _ _ _ _ _ _ _ _ _ O
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| | |
| Question 080 Which of the following ESF actuation signals is ONLY initiated from a measured parameter or manually?
| |
| A. CRVIS, B. CPIS.
| |
| C, CSAS.
| |
| D. AFAS.
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| I i
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| 1 l
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| L t
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| Question 081 1 l 1 l
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| I 1
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| The station has suffered a loss of off-site power from 100% power. l l
| |
| * The emergency diesels have started and are supplying the safety-related busses l
| |
| * No Si occurred, all systems performed as designed l
| |
| * Prior to the loss of power all fans were running in fast speed Which one of the following describes the status of the containment cooling system?
| |
| l l A. Containment Fan Coolers and H2Mixing fans are running in slow speed.
| |
| B. Containment Fan Coolers and H2Mixing fans are mnning in fast speed.
| |
| C. Containment Fan Coolers are running in fast speed, H2 Mixing fans are stopped. l D. Containment Fan Coolers are running in slow speed, H2 Mixing fans are stopped.
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| l 1
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| ! 1 1- ,
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| {
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| j
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| . Question 082 1
| |
| 1 A plant startup is in progress, the plant is stable at 80% power, the following conditions exist: l e Main Feed Pump speed control is in manual e Main Feed Regulating Valves are 85 - 90 % open
| |
| ; A Condensate Pump's minimum flow valve failing open will have which of the following consequences?
| |
| A. Reject valve AD LV-79A will open.
| |
| , B. Main Feed Regulating Valves will open.
| |
| ! l C. Main Feed Regulating Valves will close.
| |
| l D. The Main Feed Pumps will trip.
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| l l
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| _ . . _ _ _ . . . _ . _ . _ _ . . _ _ _ _ i--
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| | |
| Question 083 The following conditions exist:
| |
| * Mode 3 with Tavg at no-load values using S/G ARVs
| |
| * Reactor Trip breakers open .
| |
| * MSIVs closed S/G "A" was overfilled to the Hi-Hi level causing a FWIS.
| |
| To reset FWIS:
| |
| A. P-14 must clear, the Reactor Trip Breakers do not need to be closed.
| |
| B. P-14 must clear and the Reactor Trip Breakers must be closed.
| |
| C. FWIS can be reset with P-14 present and Reactor Trip Breakers open.
| |
| D. The Reactor Trip breakers must be closed, P-14 need not be cleared.
| |
| l L________________________
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| l' l
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| l w
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| (-
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| l Question 084 l
| |
| ! 1 i-Annunciator 062B, AREA RAD Hi alarms, to determine which monitor is affected, you should:
| |
| A. At SD055A and SD055B, check for any monitor red light lit.
| |
| B. At SD055A and SD055B, check for any monitor alert yellow lit.
| |
| C. Using NPIS check for any monitor display that is flashing red.
| |
| I D. Using NPIS check for any monitor display that is flashing yellow.
| |
| )
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| ! l I
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| l.
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| l, l
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| l J
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| l l
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| Question 085 l
| |
| The plant is at normal full power operation with all systems in normal alignment.
| |
| , Feed header pressure instrument AE PT-508 fails off scale high.
| |
| l Which of the following is the expected plant response?
| |
| A. Feedwater pump speed increases B. Feedwater pump speed decreases r
| |
| l C. Feedwater pumps trip due to high discharge pressure l
| |
| D. Indicated feed flow increase due to increased density compensation l
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| Question 086
| |
| . The plant has been stabilized following an inadvertent Safety injection.
| |
| Prior to the event the Instrument Air Compressors were sequenced B/C/A.
| |
| The status of the air system after the event is:
| |
| A. Compressor"C" running, KA PV-1I open.
| |
| ! B. Compressor "B" running, KA PV-11 open.
| |
| C. Compressor "C" running, KA PV-11 closed.
| |
| l D. Compressor "B" running, KA PV-11 closed.
| |
| 1 l
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| )
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| l Question 087 i
| |
| I l An automatic pre- action sprinkler system " trouble" alarm would indicate:
| |
| ! A. An open sprinkler head B. A deluge valve actuation l B. An alarm check valve operation C. A fire detector in alarm condition i
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| l I
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| l f
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| 6 .
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| Question 088 l
| |
| I I
| |
| A plant start-up with a positive moderator temperature coefUcient is in progress, the following conditions exist:
| |
| e The plant is stable at 35% Reactor Power
| |
| !
| |
| * Rods are in manual Tavg is 571 *F, Tref si 568 "F e Steam Dumps are in auto and selected to steam pressure mode e Cooldown valves are 90% open How would RCS temperature respond if the turbine were to trip?
| |
| l A. Tavg would increase.
| |
| B. Tavg would stabilize at 557 *F.
| |
| C. Tavg would stabilize to 568 *F.
| |
| l D. Tavg would stabilize 571 F.
| |
| 1 l
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| Question 089 The plant is at full power, with "B" train components in service.
| |
| A ground fault develops on XNB01.
| |
| All components function as designed.
| |
| With no operator actions the AT on containment coolers will:
| |
| A. Decrease on coolers "A" and "C".
| |
| l B. Increase on coolers "A" and "C".
| |
| C. Decrease on coolers "B" and "D".
| |
| 1 i D. Increase on coolers "B" and "D".
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| . _ _ _ . _ _ __. __ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ I
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| | |
| Question 090 l
| |
| While perfonning an independent verification of components positioned earlier, the position of a locked I throttled valve needs verified. j Verification can be accomplished by:
| |
| A. Contacting the original positioner and asking where the valve was positioned. 1 B. Closing the valve then repositioning it to the required throttled position.
| |
| C. Verifying the throttle position on the valve tag agrees with the checklist position.
| |
| D. Having the original positioner reposition the valve and observing the throttle setting.
| |
| i .
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| :. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ a
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| | |
| Question 091 Which of the following would be used to identify temporary equipment supplied by a welding outlet?
| |
| A. Operator Aid.
| |
| B. Magnetic information pad.
| |
| C. Information Tag.
| |
| D. Operator Note.
| |
| l i
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| 1
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| - - - - - - - - - - - - - - - - - _ _ _ - - _ - . _ - - _ _ . . - - - _ - - - - - - _ _ - - . - - - - _ _ _ . _ ._ J
| |
| | |
| ' Question 092 The BOP returns to the control room after conducting a plant tour lasting an hour.
| |
| Which of the following actions is expect of the BOP prior to resuming control room activities?
| |
| A. Review the control room logs and walk the control boards down with the RO.
| |
| B. Report to the SO, review annunciator status, and conduct a control board walkdown.
| |
| C. Report to the SO and have the Shift Engineer give you a turnover briefing.
| |
| D. Report to the SS, review the logs and books as listed in the turnover sheet, and walk the control boards down with the Shift Engineer.
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| . _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ . . . . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ a
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| | |
| Question 093 1
| |
| The Site Operator has left the protected area on rounds.
| |
| He called in from Security to say that his radio is not working.
| |
| In which of the following locations would you NOT be able to page him using the Gaitronics?
| |
| A. Circulating Water Screen House.
| |
| B. Waste Water Treatment Facility.
| |
| C. Make Up Screen House.
| |
| D. Fuel Oil Pump House.
| |
| 1 l l t i l
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| Question 094 The plant was stable at 100 % power when the relief valve on the Generator H2 system, CC PCV-8, failed open.
| |
| By the time the reliefis isolated generator gas pressure is 45 psig.
| |
| The following conditions exist:
| |
| * Generator output - 1225 MW.
| |
| * 350 Megavars out.
| |
| Using the attached Generator Performance Curve, reduce generator loading to:.
| |
| A. I125 Megawatts and 250 Megavars out.
| |
| B. I125 Megawatts and 200 Megavars out.
| |
| C. I100 Megawatts and 325 Megavars out.
| |
| D. I100 Megawatts and 250 Megavars out.
| |
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| l i
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| Question 095 i What does a test performer do with an STS when a test deficiency exists that results in the surveillance being unsatisfactory?
| |
| A. Submit the failed package for nonnal review process.
| |
| B. It should be discarded and a new test package prepared.
| |
| I C. It should be retained until the test is satisfactorily completed.
| |
| D. Send the completed STS directly to the Surveillance Coordinator.
| |
| l l
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| l-1 L____________________-.____-----
| |
| | |
| Question 096 A Local Control Acceptor has placed Local Control Tags on the breaker and handswitch for a Turbine Building fan and requests you verify the tags.
| |
| You should:
| |
| A. Verify the tags are on the correct component and initial the field copy of the Local Control form.
| |
| B. Verify the tags are on the correct component and initial the original copy of the Local Control form.
| |
| C. Verify the tags are on the proper components but do not initial the field or original copies of the Local Control Form.
| |
| D. Do not perform the verification, verification is the responsibility of the Local Control Acceptors work group.
| |
| | |
| 1 l
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| l Question 097
| |
| 'Which one of the following describes the MINIMUM compliment oflicensed personnel required to be 1 present in the Control Room by Technical Specifications when the plant is in Mode 47 A. One SRO and one RO. ,
| |
| B. One SRO and two ROs.
| |
| C. A SS or SO, an STA and one RO.
| |
| D. One RO at the controls and one RO.
| |
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| I
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| - _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ . -__ l
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| l Question 098 l
| |
| l 1
| |
| After performing a valve lineup verification in a potential hot particle area, you are required to:
| |
| A. The nearest frisking station and perform a whole body fris'K.
| |
| B. The nearest frisking station and perform a hands and feet frisk.
| |
| C. The nearest frisking station and perform a hands, feet and face frisk.
| |
| D. The personnel contamination monitor at access control and perform a whole body frisk..
| |
| i j
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| 1 1
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| [
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| I 1
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| Question 099 An operator has the following Total Effective Dose Equivalent (TEDE) at the middle of the fourth quarter:
| |
| * Current quarterly TEDE - 280 mrem
| |
| * Current yearly TEDE - 880 mrem -
| |
| With no extension, which of the following is the maximum additional TEDE exposure he can receive without exceeding WCGS administrative control limits? ,
| |
| i A. I120 mrem.
| |
| B. I720 mrem.
| |
| C. 2120 mrem.
| |
| D. 4120 mrem.
| |
| i l
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| .j
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| I Question 100 A reactor trip and safety injection has occurred. The operators have performed their EMG E-0 immediate actions.
| |
| The Supervising Operator is verifying the immediate action steps when the STA reports a red path on Heat Sink.
| |
| ! The crew will:
| |
| 1
| |
| { A. Transition to EMG FR-H1," Response to Loss of Secondary Heat Sink" when E-0 immediate action j venfication complete.
| |
| B. Immediately transition to EMG FR H1," Response to Loss of Secondary Heat Sink".
| |
| C. Stay in E-0, " Reactor Trip or Safety Injection" until automatic actuation signals have been verified.
| |
| D. Stay in E-0," Reactor Trip or Safety Injection" until directed to transition or CSFST monitoring is directed.
| |
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| CHIEF EXAMINER RO EXAM DIFFICULTY ANALYIS - WOLF CREEK 2/17/98 Legend-L - Level of knowlege required A - Knowledge at the comprehension / analysis level K - Fundamental knowledge only C "X" indicates a chief examiner comment exists D - Overall difficulty of question on a scale of 1 to 5, with 5 being the most difficult
| |
| )
| |
| Question L/D C7 Question L/D C? Question L/D C7 1 N3 35 K/2 X 69 N3 2 N3 36 A/3 70 A/2 ,
| |
| 3 N3 37 N3 71 N3 4 K/2 38 N3 72 N2 5 A/3 39 A/4 X 73 N3 6 N2 40 K/2 74 K/2 7 A/3 41 N3 75 A/3 8 A/3 42 A/2 76 N3 9 K/2 43 K/3 77 N2 10 A/3 X 44 K/2 X 78 N3 11 A/3 X 45 K/2 X 79 N3 12 K/2 X 46 N3 80 K/3 13 A/4 X 47 N3 81 N3 14 N4 X 48 K/2 82 A/2 15 K/1 X 49 K/3 83 N3 16 A/4 X 50 N3 84 K/2 17 A/2 X 51 N3 85 N3 18 K/2 X 52 A/2 86 N3 19 A/2 X 53 K/3 87 VJ3 20 A/3 54 A/3 X 88 N4 21 K/2 55 N3 89 N3 22 N2 56 A/3 90 K/3 23 N3 57 N3 91 K/2 24 A/2 58 K/2 X 92 K/3 25 A/2 X 59 N3 X 93 K/2 26 K/2 60 K/4 94 K/2 27 K/3 61 N3 95 K/2 X 28 N4 62 K/2 96 K/3 29 N4 63 N3 97 K/3 X 30 A/3 64 N3 98 K/2 31 A/2 65 K/4 99 K/2 1 32 K/2 66 N2 100 N4 33 K/3 67 K/2 X l 34 A/4 68 N3 A = 62 D.,, = 2.72 l i
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| CHIEF EXAMINER RO WRITTEN EXAM COMMENTS - WOLF CREEK 2/17/98 Question No. Comment 10 The answer has a specific determiner in that it is the only choice referring to the inverter and NN-02, which were discussed in the stem. Resolution:
| |
| Eliminated reference to NN-02 in answer.
| |
| l l
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| l 11 Because of the structure of the distractors, the difficulty of this question is i minimal. The examinee only has to know that the seal water temperature increases to eliminate the distractors. Resolution: Changed one of the distractors.
| |
| 12 Why is distractor A not correct? Resolution: Replaced distractor.
| |
| 13- KA does not match outline, i.e. EK1.2 vice AK1.01. Resolution: Changed outline.
| |
| 14 KA importance rating in the outline is incorrect, i.e. 4.1/4.7 vice 3.9/4.1.
| |
| Also, from the references provided, it is not clear that Distractors C and D are plausible. Resolution: Changed reference sheet to match outline.
| |
| Distractor C and D are plausible. Changed wording to make it clear that they are incorrect.
| |
| 15 The answer is too obvious. Almost anyone would know that a smallleak in containment would not result in a high range reading. Also, "last" should be capitalized in the stem. Resolution: Replaced question.
| |
| 16 As written, this question does not solicit the desired knowledge. The intent of the question is to determine if the examinee knows the cooldown limits for the conditions listed. However, it merely asks the examinee what he would do. Any answer which is under the limits would be satisfactory. With this knowledge, an experienced, test taker can eliminate some of the distractors.
| |
| Resolution: Clarified question.
| |
| 17 Distractor D is implausible. It is obviously wrong to watch the power
| |
| . increase from 50% to 100% under the given conditions. Resolution: Revised Distractor D.
| |
| 18 The f act that the answer is the only reference to the TS could be considered i a specific determiner. Resolution: Referenced TS in an additional distractor.
| |
| 19 The explanation for the answer is not fully supported by the reference in that there is no mention of Lo/Lo S/G level. Distractor A could be considered correct because it is the root cause of the trip. Distractor C is correct, but not complete. We doubt that these distractors cculd be defended as wrong answers. Resolution: Reworded stem and Distractor C.
| |
| 25 KA does not match the outline. Resolution: Changed outline.
| |
| j
| |
| | |
| 3 35 If true statements are to be used as distractors, the stem must more clearly describe what is being asked; i.e. the TS bases for rod insertion limits.
| |
| Resolution: Modified stem.
| |
| l 39 KA does not match outline. Resolution: Changed outline.
| |
| l 44 Justify using a negative stem. Resolution: Changed distractors to make stem l positive.
| |
| 45 The knowledge solicited is almost identical to that required to answer Question 19. Replace one of these questions. Resolution: Replaced this question.
| |
| 54 KA does not match outline. Resolution: Changed outline.
| |
| 58 Reword the stem. The operator can apparently close the spray additive tank isolation valve anytime. The intent of the question is to ask when he/she should close the valve. Resolution: No change is necessary. An interlock prevents closing valve.
| |
| 59 Distractor B is too similar to the answer. Rewrite. Resolution: Changed distractor.
| |
| 67 The explanation for distractor B indicates it is a correct answer. Also, the knowledge required is very similar to Question 7. Resolution: Replaced this question.
| |
| 95 Distractor B is implausible. Anyone in the nuclear industry knows that paperwork is never discarded. Resolution: Changed distractor and reworded stem.
| |
| | |
| Ura to-U.ts LOSS OF RHR COOLING
| |
| . Continuous Uma Psdi 146 of 148 FIGURE 5 TIME TO BOILING IN REDUCED INVENTORY TIME TO 80ll (MINUTES) 60 l
| |
| l 2
| |
| 50 /
| |
| VESSEL PRESSURIZED
| |
| /
| |
| /
| |
| /
| |
| 40 f f
| |
| /
| |
| /
| |
| /
| |
| 30 /
| |
| /
| |
| /
| |
| /
| |
| / ""
| |
| 20 l m - -
| |
| r 7 - s *
| |
| * ~
| |
| A _,s "~
| |
| m - - VESSEL NOT PRESSURIZED j f I #
| |
| 10 #
| |
| / l l
| |
| /
| |
| 0 0 5 10 15 20 25 30 DAYS AFTER SHUTDOWN
| |
| | |
| 7------_________- ]
| |
| OFN EJ-015 U LOSS OF RHR COOLING i ra Continuous Use Page 148 of 148 l "T FIGURE 6 TIME TO CORE UNCOVERY IN REDUCED INVENTORY
| |
| ; t3 l C1 TIME TO CORE UNC0VERY (MINUTES) l 500
| |
| -a i ! ! > i i i
| |
| vj '
| |
| I I i i i I l
| |
| 1 l
| |
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| |
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| |
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| |
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| |
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| |
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| |
| fe i 400 #' ' i I
| |
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| |
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| |
| /
| |
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| |
| 1 350 -
| |
| i i I '
| |
| I v i COMPLETE CORE UNC0VERY j '
| |
| i / I 300 ' '' ' I
| |
| / i i
| |
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| |
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| |
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| |
| , r 250 of i ,
| |
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| |
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| |
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| |
| 200 #
| |
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| |
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| |
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| |
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| |
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| |
| / J' 150 # #i ' ' '
| |
| / i , f . i i i i / '# ONSET OF CORE UNC0VERY '
| |
| : 1 /
| |
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| |
| 100 i
| |
| / i i ! i i i 8
| |
| / l ! l ! i 1 l i/ ! i I I i i t
| |
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| |
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| |
| i l I i l i e i e
| |
| ! i
| |
| ' ! I t i i I i i i i i I i I i r 0 '
| |
| 0 5 10 15 20 25 30 DAYS AFTER SHUTDOWN 4
| |
| i l
| |
| J
| |
| | |
| Revision: 9 SYS GS-120 POST LOCA CONTAINMENT HYDRCGEN Continuou2 Use RECOMBINER OPERATION Page 9 of 16
| |
| _s INIT/DATE 6.3 Hydrogen Recombiner A Startup MOTE WHEN time is available prior the startup, THEN the hydrogen recombiners should be placed in standby for warm up prior to startup.
| |
| I 6.3.1 At GS065A, turn hydrogen recombiner Power I Out Switch on.
| |
| o HIS-1B - ON EI l 6.3.2 Verify that red lamp on the Power out Switch plate is lit.
| |
| o Red Lamp - LIT (( l 6.3.3 Determine recombiner power setting by performing the following: [3
| |
| : 1. Record from Plant Operating Logs the Pre-LOCA containment temperature.
| |
| Containment Temperature /
| |
| l
| |
| : 2. Determine pressure factor (Cp) using Figure 1, RECOMBINER POWER CORRECTION FACTOR VERSUS CONTAINMENT PRESSURE CURVE.
| |
| Containment Pressure Cp / l
| |
| : 3. Multiply the reference power (OA 88-10-1, plaque on Recombiner Control Panels) by Cp to determine required recombiner power setting as ,
| |
| indicated on the power meter. l SGS01A Reference Power XC p =
| |
| l Power Setting / l I
| |
| e k
| |
| 1 J
| |
| | |
| I POST LCCA CONTAINMENT HYDROGEN l Continuous Use RECOMBINER OPERATION Page l' of ;E l FIGURE 1 DRY CONTAINMENT RECOMBINER CORRECTION FACTOR VS. CONTAINMENT PRESSURE
| |
| ---=,=,.im 46 li I
| |
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| Impf-Lata certaisutyt Pumuuset sPliel l
| |
| _ _ - - - - - - - - - - - - - - - - - 1
| |
| | |
| a Revision: 33 uLa vu-vus
| |
| + POWER OPERATION
| |
| < Continuous Use Page 34 of 34 I r
| |
| ~
| |
| l FIGURE 1 GENERATOR PERFORMANCE CURVE
| |
| * 1000 KILOVARS 1
| |
| I
| |
| , 1200 i
| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
| l ,' CURVE AB LIMITED BY FIELD HEATING ll ' '
| |
| , CURVE BC LIMITED BY ARMATURE HEATING CURVE CD LIMITED BY ARMATURE CORE END HEATING l ;
| |
| -1000 j j j j l
| |
| 0 200 400 600 800 1000 1200 1400 1600 1000 KILOWATTS
| |
| | |
| ES-401 Site-Specific Written Form ES-401-7 Examination Cover Sheet l
| |
| U.S. Nuclear Regulatory Commission l Site-Specific Written Examination ,
| |
| Applicant Information Name: Region: IV Date: 2-17-98 Facility / Unit WCGS License Level SRO Reactor Type W Start Time: Finish Time:
| |
| Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent. Examination papers will be collected four hours after the examination starts.
| |
| Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
| |
| Applicant's Signature -
| |
| Results Examination Value Points Applicant's Score Points Applicant's Grade Percent NUREG-1021 Interim Rev. 8. January 1997
| |
| | |
| Initial License Examination Answer Shcet Name: Date: 2-17-98 1 ABCD 26 ABCD 51 ABCD 101 ABCD l 2 ABCD 27 ABCD 52 ABCD 102 ABCD l
| |
| l 3 ABCD 28 ABCD 53 ABCD 103 ABCD l 4 ABCD 29 ABCD 54 ABCD 104 ABCD l 5 ABCD 30 ABCD 55 ABCD 105 ABCD 6 ABCD 31 ABCD 56 ABCD 106 ABCD 7 ABCD 32 ABCD 57 ABCD 107 ABCD 8 ABCD 33 ABCD 58 ABCD 108 ABCD 9 ABCD 34 ABCD 59 ABCD 109 ABCD l 10 ABCD 35 ABCD 60 ABCD 110 ABCD 11 ABCD 36 ABCD 61 ABCD 111 ABCD 12 ABCD 37 ABCD 62 ABCD 112 ABCD l l
| |
| 13 ABCD 38 ABCD 63 ABCD 113 ABCD l
| |
| ABCD ABCD ABCD ABCD 14 39 64 114 15 ABCD 40 ABCD 65 ABCD 115 ABCD 16 ABCD 41 ABCD 66 ABCD 116 ABCD 17 ABCD 42 ABCD 67 ABCD 117 ABCD 18 ABCD 43 ABCD 68 ABCD 118 ABCD 19 ABCD 44 .A B C D 69 ABCD 119 ABCD 20 ABCD 45 ABCD 70 ABCD 120 ABCD I
| |
| 21 ABCD 46 ABCD 71 ABCD 121 ABCD 22 ABCD 47 ABCD 72 ABCD 122 ABCD 23 ABCD 48 ABCD 73 ABCD 123 ABCD 24 ABCD 49 ABCD 74 ABCD 124 ABCD f 25 ABCD 50 ABCD 75 ABCD 125 ABCD
| |
| .______m_ - _ . _ _ _ _ . _ _ _ _ . . - . _ .
| |
| | |
| Question Number 001 l
| |
| The Control Room is performing STS SF-001 " Control and Shutdown Rod Operability Verification" on Control Bank "B".
| |
| l While returning Bank "B" to its fully wi^hawn position a Rod Control Urgent alarm (79A) is received and all rod movement stops.
| |
| What can be determined from this information?
| |
| A. There is a problem in the Logic Cabinet.
| |
| 1 B. There is a problem in the Power Cabinet. !
| |
| C. The Rod Control Urgent alarm is invalid.
| |
| D. InsufUcient information exists to determine cause of alarm.
| |
| 4 I
| |
| l l
| |
| l l
| |
| l l
| |
| 1 l
| |
| f
| |
| | |
| i Question 002 l
| |
| l l The following plant conditions exist:
| |
| l e Rx power 75%
| |
| * ANN 00-041 A, SEAL INJ TO RCP FLOW LO is alarming
| |
| * ANN 00-071C, RCP B THRM BAR CCW FLOW is alarming l
| |
| * RCP "B" No. I seal and bearing water temperature is 235*F Which of the following is the correct action to take?
| |
| A. Trip RCP "B" and perform Attachment D," Restoration of RCP Seal Cooling".
| |
| B. Reduce power to less than 48% and then secure RCP "B".
| |
| C. Immediately restore seal cooling to RCP "B" using Attachment D," Restoration of RCP Seal Cooling".
| |
| D. Trip the reactor and then trip" RCP "B".
| |
| i l
| |
| l 1
| |
| i
| |
| | |
| i l
| |
| l 1
| |
| Question 003 I
| |
| l i t 1 The plant has stabilized at approximately 75% power following a grid disturbance.
| |
| ''NB02 is de -energized and OFN NB 030 " Loss of AC Emergency bus NB01 (NB02)" is b ie ng l performed.
| |
| * Ann. 00-081C " Rod Bank LOLO Limit"is Lit.
| |
| l e Control bank 'D' is at 100 steps.
| |
| . CVCS is in service with the NCP running.
| |
| * All other equipment is functioning as designed. CVCS is in service with the NCP running.
| |
| i You Have lined up for ' Emergency Boration' and MCB indication BG FI-183A,"EMERG BORATE FLOW", reads zero; Your actions are:
| |
| l A. No action required because BG FI-183A has lost power but boration flow is present.
| |
| B. Align charging pump suction to the RWST using ' Red Train' RWST & VCT valves (BN HIS-ll2D & BG HIS-ll2B).
| |
| C. Emergency boration is not required due to the plant transient causing rod insertion.
| |
| l D. Commence boration using the reactor make up system.
| |
| i l
| |
| i i
| |
| i
| |
| | |
| l l
| |
| t Question 004 j i
| |
| l The plant has experienced a complete loss of wmponent cooling water.
| |
| Which is the preferred source of cooling water to 'A' CCP and 'A' SI pumps?
| |
| A. Circulating Water B. Closed Cooling Water C. Essential Service Water D. Domestic Potable Water l
| |
| l I
| |
| L _ _ __ __ ___ _ _- _ _
| |
| | |
| l Question 005 l
| |
| l The Pressurizer master pressure controller has failed to zero output in auto.
| |
| No operator actions have been performed.
| |
| RCS pressure will:
| |
| A. Decrease to PORV interlock setpoint.
| |
| l B. Increase to Reactor trip setpoint.
| |
| C. Stabilize at Program setpoint.
| |
| D. Cycle at the PORV setpoint.
| |
| l l
| |
| l t
| |
| -__-_____-___-a
| |
| | |
| j i
| |
| l l
| |
| k l Question 006 1
| |
| While performing diagnostic steps in E-0, the following conditions are observed:
| |
| * RCS T.., - 536 *F e RCS Press - 1800 psig
| |
| * PZR LVL - 4%
| |
| * MSIVs are Open e CTMT Temp. - 240*F e GE RE-92 pre-event was normal e SG A Press /WR Level- 915 psig/20%
| |
| * SG B Press /WR Level - 915 psig/55%
| |
| * SG C Press /WR Level - 915 psig/50%
| |
| * -SG D Press /WR Level - 915 psig/55%
| |
| * CTMT radiation Levels are normal e Aux. bldg. radiation levels are normal All plant equipment is functioning as designed.
| |
| Which one of the following is the most likely cause of this event?
| |
| A. LOCA inside containment.
| |
| B. LOCA outside containment.
| |
| C. Steam generator tube rupture. i D. Steam line break inside containment.
| |
| l l
| |
| l l
| |
| | |
| l Question 007 l
| |
| l A plant start up is in progress and turbine load is 28%.
| |
| Which of the following conditions would require the turbine to be tripped 7 A. ANN 00-116B 'COND A VAC LO' is alarming and turbine exhaust pressure is 6 inches HgA.
| |
| B. ALR 00-117C ' STANDBY EHC PMP START' is alarming and EHC fluid pressure is 1500 F5i8 C. ALR 00-017D 'PG XFMR UV' is alarming and PG-11 bus voltage is zero.
| |
| D. Turbine bearing oil pressure is 20 psig l
| |
| l
| |
| )
| |
| | |
| .e.
| |
| l Question 008 l
| |
| Initial conditions
| |
| * Train B outage in progress e 'B' CCP tagged out
| |
| * 'B' ESW pump tagged out
| |
| * 'B' Si pump tagged out A station blackout occurs. Power becomes available from the West Bus.
| |
| Which electrical lineups should be placed in service?
| |
| A. NB01 through the alternate feeder breaker.
| |
| B. NB02 through the normal feeder breaker.
| |
| C. NB01 through the alternate feeder breaker and NB02 though the normal feeder breaker.
| |
| D. Neither NB bus, energize PA02 from the West Bus.
| |
| l 1 \
| |
| l i
| |
| _ _ . _ - . _ - - - - - _ - _ - _ - - - - - - . _ - - - - - - - - - _ - - - - - - - - - - - - - - - N
| |
| | |
| l
| |
| ['
| |
| Question 000 '
| |
| l Step 3.b of EMG C-0, " Loss of All AC Power," checks RCS Letdown to Regenerative heat exchanger l valves BG HIS-459 and 460 closed.
| |
| Why is Letdown isolated early in EMG C-07 A. To ensure Containment Integrity is maintained.
| |
| B. To prevent flashing and two-phase flow in the letdown heat exchanger and upstream piping.
| |
| C. To prevent a dilution when hotter letdown flow enters the mixed bed demineralizers.
| |
| D. To ensure that RCS inventory loss is minimized.
| |
| l 1
| |
| l l
| |
| l 4
| |
| I
| |
| | |
| Question 010 l
| |
| l The plant is at 100% power.
| |
| l Due to an inverter problem NN02 was de-energized.
| |
| All required actions of OFN NN-021 " Loss of Vital 120 VAC Instrumentation Bus", have been completed and 'NN02' is energized from its 'SOLA' transformer.
| |
| Which of the following describes what must be done to allow continued operation?
| |
| A. Perform STS NB-005," Breaker Alignment Verification".
| |
| B. Restore NN02 to its inverter.
| |
| C. Ensure all 'A' train components are operable.
| |
| D. Ensure Turbine Driven Aux. Feed pump is operable.
| |
| 1 i
| |
| 1 1
| |
| ~
| |
| 4 i
| |
| j l
| |
| i
| |
| ...-_.________m __. _
| |
| | |
| Question 011 l
| |
| In the event both ESW pumps and Service Water pumps are unavailable OFN EF-033 " Loss of Essential Service Water", has an attachment to provide an alternate means to cool the CCPs.
| |
| Why is alternate cooling provided for these pumps?
| |
| A. To reduce the CCW heat load and prolong SFP cooling.
| |
| B. To reduce the heat load on their room coolers.
| |
| C. To maintain the pumps operable.
| |
| D. To maintain RCP sealintegrity, i
| |
| I i
| |
| l
| |
| | |
| l e,
| |
| l Question 012 i
| |
| Per OFN KC-016 " Fire Response", a continuous fire watch is required following a fire onsite if:
| |
| A. The fire lasts for longer than 30 minutes.
| |
| B. Offsite assistance was required. i 1 C. The fire has been out for 9 hours.
| |
| D. De fire area is hot to the touch.
| |
| {
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| Question 013 I 1
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| l l A The plant was stable in Mode 3 at 488 "F and 1900 psig.
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| A LOCA has occurred and the following conditions now exist:
| |
| * RCS pressure - 1800 psig
| |
| * CTMT pressure - 4 psig
| |
| * S/G pressure - 595 psig Which of the following signals should have actuated? I A. Main Steam LineIsolation.
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| B. Steam Line SafetyInjection.
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| C. CTMT pressure Safety injection.
| |
| D. RCS Low pressure Safety injection.
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| )
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| Question 014 A loss of coolant accident coincident with a loss of NB02 occurred several hours ago.
| |
| All " Red" train equipment is operating as required.
| |
| The crew has entered EMG ES-12," Transfer to Cold Leg Recirculation" and are performing step 10,
| |
| " Align CCP and Si pump suctions to RHR pump discharge."
| |
| EM HIS-8807A, "CVCS to SI Pump Suction Valve" will not open.
| |
| Recirculation flow from the CTMT sumps to SIP "A" is:
| |
| A. Not available unless EM HV-8807B is open locally.
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| B. Not available unless EJ HV-8804B is opened locally.
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| C Available from "B" train flowpath to suction of S1 Pump "B".
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| D. Available from "A" train flowpath to suction of SIP "A".
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| J
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| l Question 015 1
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| The following conditions exist:
| |
| e 100 % power e SJ RE-01,"CVCS Letdown Radiation Monitor," is inoperable e Containment Purge in progress e 1.3 gpm leakage from BG LCV-459 packing If fuel failure should occur, which of the following monitors would be the last to show an increased radiation reading?
| |
| A. GT RE-21B Unit Vent Monitor.
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| B. GT RE-33 CTMT Purge Monitor.
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| C. GT RE-31 CTMT Atmosphere Monitor.
| |
| D. GT RE-59 CTMT High Range Area Monitor.
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| _ _. - _ _ _ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ . _ _ _ _ _ _ _ _ _ . -_ ____ ________m
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| i Question 016.
| |
| The plant was operating at full power when offsite power was lost, all equipment operated as required and no SI occurred.
| |
| Power will be restored in 12 hours.
| |
| Plant staff determines a cooldown is required, the crew should conduct a natural circulation cooldown using EMG ES-04," Natural Circulation Cooldown" at:
| |
| A. < 100 *F per hour and maintain 125 *F subcooiing.
| |
| B.' < 50*F per hour and maintain 125 'F subcooling.
| |
| C. < 100*F per hour and maintain 75 'F subcooling.
| |
| D. < S0'F per hour and maintain 75 'F subcooling.
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| Question 017 Current plant conditions are:
| |
| * Control rods are moving outward e Power 50 % and stable
| |
| . T,,, 573 'F initially stable, but increasing slowly i
| |
| e Turbine load 625 MWe and stable e Average loop AT is 50 % and stable e Rods are in AUTO Under these conditions the RO actions are to: j A. Select Rod Control to MANUAL to stop the rod motion.
| |
| B. Select Rod Control to Group D to stop rod motion.
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| C. Monitor rod motion to ensure proper system response.
| |
| D. Monitor rod motion to verify that power remains less than 100%.
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| I Question 018 If two or more rods in the same group drop into the core OFN SF-01 I " Realignment of Dropped, '
| |
| Misaligned Rods (s) and Rod Control Malfunctions" requires the plant be shutdown.
| |
| The reason for this is:
| |
| A. Changes in Xenon concentration would make recovery hazardous.
| |
| B. Technical Specifications requires the plant be in HSB in 6 hours.
| |
| C. Inadequate shutdown margin remains in event of a reactor trip.
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| D. Two rods could not be recovered in the one hour allowed.
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| 1
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| f .-
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| Question 019 A Reactor trip from 100% power with no Safety Injection has occurred.
| |
| l The transient that caused the trip was initiated by the loss of the "C" Condensate Pump.
| |
| He Main Feed Water Pumps tripped due to:
| |
| A. The trip of the 'C' Condensate Pump.
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| B. De resultant AFAS(M) actuation signal.
| |
| C P-4, Reactor Trip Permissive.
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| D. The resultant Feed Water Isolation Signal.
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| Question 020 The following conditions exist:
| |
| * Pu. pressure - 1985 psig
| |
| * Pu temperature - 636 *F
| |
| * RCS temperature - 364 'F
| |
| * PORV BB PV-455A indicates open Which parameter set is correct?
| |
| PRT Press. PRT Temp.
| |
| A. 40 psig 267'F B. 25 psig 267*F C. 25 psig 636'F D. 145 psig 364 *F
| |
| | |
| Question 021 RCS leakage has increased to 0.9 gpm. The increase is coming from a weld where the BB FI-426 sensing line penetrates the loop.
| |
| Which of the following is required?
| |
| A. Plant operations can continue, identified leakage is less than 10 gpm.
| |
| B. Plant operations can continue, total leakage is less than I gpm.
| |
| C. Commence a plant S/D as this leakage could affect the leakage detection systems.
| |
| D. Commence a plant S/D as this is pressure boundary leakage.
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| Question 022 A plant event has occurred
| |
| . RCS Wide Range pressure - 1500 psig and decreasing
| |
| . RCS Wide Range Temperature - 557'F and stable e Main Steam Header Pressure - 1092 psig and stable
| |
| . S/G "A" WR level - 58% and increasing e S/G "B" WR level- 56% and increasing e S/G "C" WR level- 57% and increasing e S/G "D" WR level- 59% and increasing e Containment Pressure - 7 psig and decreasing Based on the above Main Control Board indications what has occurred?
| |
| A. A feedline break.
| |
| B. A steamline break.
| |
| C. A reactor coolant system break.
| |
| D. A steam generator tube rupture.
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| l
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| Question 023 Initial Conditions: ;
| |
| * Mode one,100% power, NOT/NOP
| |
| * NCP tagged out for maintenance
| |
| * 'A' CCP is in service The 'A' CCP trips and cannot be restarted. The only operator action is to start the 'B' CCP reestablishing seal injection flow at a minimum of 32 gpm.
| |
| With no further operator action which of the following adverse events would occur?
| |
| A. Reactor Trip on high pressurizer level.
| |
| B. Reactor trip on low pressurizer pressure.
| |
| C. Loss of Dow through the CVCS demineralizers resulting in increasing RCS activity.
| |
| D. Overheating of the letdown How resulting in bypassing the CVCS demineralizers on high temperature.
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| Question 024 l
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| t De following plant conditions exist:
| |
| * The plant has been shutdown for 10 days after 100 days of full power operation
| |
| * Plant is in Mode 5 e The RCS is drained to mid loop for RCP seal work I e The RCS is vented to atmosphere e Heat removal is being provided by RHR 1
| |
| Using the Attached pages, which of the following is the closest time it will take for the core to begin to uncover if RHR is lost.
| |
| A. 14 minutes.
| |
| B. 32 minutes.
| |
| C. 125 minutes.
| |
| D. 263 ininutes.
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| | |
| Question 025 A reactor trip has occurred from full power.
| |
| The crew is performing the immediate actions of EMG E-0 " Reactor Trip Or Safety Injectica".
| |
| The following items are observed:
| |
| * 'A' trip breaker is open e 'B' trip breaker is closed .
| |
| e 3 rods are stuck out = mid core
| |
| * SE NI-35D SUR -1/3 DPM e SE NI-36D SUR -1/3 DPM e SE NI-35B 4 x 10-5 amps
| |
| * SE NI-36B 5 x 10-5 amps Which of the following actions must be performed?
| |
| A. Manually trip the reactor.
| |
| B. Attempt to manually drive the stuck rods in.
| |
| C. Continue with step 2 in EMG E-0 " Reactor Trip Or Safety Injection".
| |
| D. Transition to EMG FR-Si " Response to Nuclear Power Generation /ATWT".
| |
| | |
| Question 026 The crew is responding to a Safety Injection caused by a SGTR on the S/G "B".
| |
| Assume all signals actuated as required and none have been reset.
| |
| Which of the following radiation monitors will continue to give radiation readings consistent with the accident?
| |
| A. SJ RE-2, S/G sample monitor.
| |
| B. GE RE-92, Condenser air discharge monitor.
| |
| C. BM RE-25, S/G blowdown radiation monitor.
| |
| D. FC RE-381, TDAFW pump turbine exhaust monitor.
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| I Question 027 i
| |
| EMG E-3 " Steam Generator Tube Rupture", is being performed in response to a tube rupture on S/G "A". I The cooldown hasjust been completed but the target temperature value selected by the crew was higher j than that stipulated in the procedure.
| |
| This error could result in which one of the following conditions?
| |
| A. Filling the Pressurizer solid during the subsequent depressurization.
| |
| B. Decrease the time for termination of the primary to secondary leakage.
| |
| C. Decrease in pressure of the ruptured S/G with increases leakage from the RCS.
| |
| D. Loss of RCS subcooling before RCS and ruptured S/G pressures are equalized.
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| 1
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| I g
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| Question 028 l
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| l A plant stan up was in progress per GEN 00-003," Hot Standby to Minimum Load," when an inadvertent Safety Injection occurred.
| |
| 1 All equipment functioned as designed except the "B" train trip breaker did not open.
| |
| The initiating signal has been cleared and the Si reset switches have been depressed.
| |
| If a valid auto Si signal is received, it will actuate:
| |
| A. Both trains.
| |
| B. "A" train only.
| |
| C. "B" train only.
| |
| D. Neither train.
| |
| 9
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| | |
| Question 029
| |
| . A small break LOCA occurred at full power. The crew is performing EMG ES-Il " Post LOCA Cooldown and Depressurization", and are at step 21 " check if one CCP should be stopped".
| |
| Plant conditions were:
| |
| * SI pumps running- 2
| |
| * RCPs running - 1
| |
| * Subcooling - 90 *F e PZR level- 25%
| |
| The crew secured 'A' CCP.
| |
| The following plant parameters were reported after the CCP was secured.
| |
| * RCS temperature - 426 *F e RCS pressure - 415 psig
| |
| * PZR level- 10% and decreasing Based on this information what can be stated about the LOCA?
| |
| A. The LOCA has not changed.
| |
| B. The LOCA has decreased in size.
| |
| C. The LOCA has increased in size.
| |
| D. There is insufficient information to make any statement.
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| Ouestion 030 l
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| The plant experienced a LOCA in the Aux Building and the crew is performing EMG C-12, "LOCA Outside Containment". The RO has closed EJ HV-8809A "RHR to accumulator l injection loops 1 & 2". RCS pressure is reported to be stable.
| |
| 1 What does this tell you about the break?
| |
| A. The break was on the accumulator injection line and has been isolated.
| |
| B. The break was on the RHR injection line and has been isolated.
| |
| C. The break location is not known but has been isolated.
| |
| D. The break location is not known and is not isolated. I i
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| Question 031 initial Conditions:
| |
| The crew is performing " Bleed & Feed" per EMG FR-H1 " Response To Loss Of Secondary Heat Sink".
| |
| . S/G "A" level - 9% WR e S/G "B" level - 8% WR
| |
| * S/G "C" level - 10% WR e S/G "D" level - 6% WR
| |
| * CTMT pressure - 3.5 psig
| |
| * RCS Thot - 553*F and decreasing "A" MDAFW pump is restored to service.
| |
| Which of the following is the expected course'of action?
| |
| A. Continue bleed & feed until Thot < 550 F.
| |
| B. Commence feeding all S/Gs at 40,000 lbm/hr.
| |
| C. Commence feeding" S/Gs "B" & "C at 40,000 lbm/hr.
| |
| D. Feed S/G "C" as necessary to restore narrow range level.
| |
| l
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| | |
| Question 032 The purpose of' Attachment C " Establishing Minimum ECCS Flow to Remove Decay Heat", of EMG C-11 " Loss of Emergency Recirculation", is to provide minimum ECCS flow required to:
| |
| l A. Restore emergency coolant recirculation capability.
| |
| B. Prevent heat up and minimize RWST depletion.
| |
| C. Cool down and minimize RWST depletion.
| |
| D. Commence a cool down.
| |
| )
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| l t
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| Question 033 At EOL a loss of power has occurred. EMG C-0," Loss of All AC Power," is in progress and the intact steam generators are being depressurized to 260 psig.
| |
| The Reactor Operator reports a positive startup rate on the source and intermediate range channels.
| |
| What action is required?
| |
| A. Control the S/G ARVs or TD AFWP to stop the depressurization and allow the RCS to heat up.
| |
| B. Control the S/G ARVs or TD AFWP to stabilize temperature until Xenon builds in.
| |
| C. Slow down the cooldown rate to 50 'F/hr to allow Xenon to maintain reactor subcritical.
| |
| D. Increase the cooldown rate so ECCS accumulators can inject before point of adding heat is reached.
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| I u____.____._____ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - . _ _ _ __
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| | |
| 7 l Question 034 Initial Conditions:
| |
| * The unit is at 100% reactor power.
| |
| * KA HV-29 " Instrument Air to Containment," fails closed.
| |
| With no operator action 24 hours after the event pressuriar level will:
| |
| A. Stabilia at program level.
| |
| B. Decrease to a lower level then stabilia.
| |
| C. Increase to a new higher level then stabilia.
| |
| D. Continue to increase until the reactor trip setpoint is reached, l
| |
| d' e
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| l l
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| Question 035 l
| |
| l
| |
| [
| |
| Which of the following correctly describes the reasons for Rod Insertion Limits?
| |
| A. Ensures acceptable power distribution limits and minimum shutdown margin are maintained.
| |
| B. Ensures the reactivity transients associated with postulated accident conditions are controllable within acceptable limits.
| |
| l C. Ensures that negative reactivity control is available during all modes of operation.
| |
| D. Ensures the radial flux di&rence is limited to maintain the minimum DNBR in the core at or above the safety analysis DNBR limits.
| |
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| l 1
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| Question 036 During full power operation a spurious Phase A Isolation occurs. No operator action is taken.
| |
| Which of the following describes the effect upon Loop D RCP seal operation?
| |
| A. No. I seal leakoff flow is stopped.
| |
| B. No. 2 seal leakoff flow rate increases.
| |
| C. No. 3 seal standpipe makeup rate increases.
| |
| D. Seal injection flow is stopped.
| |
| 1 >
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| Question 037 CCW flow to the letdown heat exchanger is reduced by half.
| |
| The resultant plant response is?
| |
| (assume no protective actuation's occur and that temperature limits are not exceeded or alarm)
| |
| A. RCS Tavg will decrease.
| |
| B. RCS Tavg will increase.
| |
| C. Reactor power will decrease.
| |
| D. Turbine MWe output will increase.
| |
| 1 E___________________ .j
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| | |
| 4 Question 038 l The Cation Demineralized saturated at MOL is placed in service at EOL during coastdown.
| |
| Assume rod control is automatic and control rods are parked at 232 steps.
| |
| Which one of the following will occur? ,
| |
| A. The control rods will step inward.
| |
| B. The control rods will step outward.
| |
| C. Primary coolant pH will decrease.
| |
| D. No control rod motion will occur 1
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| ._-____________-______O
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| | |
| Question 039 The plant is operating at full power with 'A' train CCW supplying the service loop. An event has occurred and the following is observed:
| |
| e 'A' & 'C' CCW pumps are running e BB HIS-13,14,15, & 16 are open
| |
| * EG HIS-62 is closed What has occurred?
| |
| A. RCP thermal barrier leak.
| |
| B. Loss of battery charger NK21.
| |
| C. Loss of NG03.
| |
| D. Loss of NNO1.
| |
| l L
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| m Question 040A During a normal controlled cooldown we can block the Low Steamline Pressure Si signal when RCS pressure is Excess cooldown protection is provided by A. < 1920 psig ; lii-I containment pressure.
| |
| B. < 1920 psig ; Negative Steamline Pressure rate.
| |
| C. < 1830 psig ; lii-I containment pressure.
| |
| D. < 1830 psig ; Negative Steamline Pressure rate.
| |
| I l
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| i L___
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| | |
| Question 074 While touring the Radwaste Bldg. you notice water dripping from a valve body which has its bonnet removed.
| |
| The work area is separated from the surrounding area by HP boundary tape and is posted as a Contaminated Area..
| |
| Which of the following actions should NOT be performed?
| |
| A. Immediately notify HP and the Control Room.
| |
| B. Confine and cover the spill.
| |
| C. Leave the area.
| |
| D. Monitor yourself for contamination.
| |
| w___-______-__ _.
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| | |
| l L
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| Question 075 A steam break due to a failed open safety valve on S/G "A" has occurred.
| |
| l EMG E 2," Faulted S/G isolation" has been completed on S/G "A".
| |
| S/G "A" has not blown dry and RCS and S/G are still decreasing.
| |
| The crew has returned to EMG E-1," Loss of Reactor or Secondary Coolant" and have reached the step that has them check RCS and S/G pressures. ,
| |
| i What action is required?
| |
| A. Stay in EMG E-1 and return to step 1. l l
| |
| l B. Stay in EMG E-1 and continue with the procedure.
| |
| 1 C. Return to EMG E 2 until S/G "A"is depressurized.
| |
| D. Return to EMG E-2 and check for additional faulted S/Gs.
| |
| l 1
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| a.
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| l Question 101 1
| |
| A small break LOCA concurrent with a loss of power has ecurred.
| |
| NE01 is inoperable, all train "B" ECCS equipment performed as required.
| |
| 'Ihe RCS is at saturation.
| |
| Which of the following would be an indication that an inadequate decay heat removal situation has developed? l RCS Temp RCS Pressure A. Increasing Stable B. Increasing increasing C. Stable Stable D. Stable Increasing i
| |
| i
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| | |
| l Question 102 i-l The MINIMUM level of classification that the shiR chemist and shiR HP report to the control room is:
| |
| . Notification of Unusual Event.
| |
| B. Alert.
| |
| C. Site Area Emergency.
| |
| ! D. General Emergency.
| |
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| . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - a
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| | |
| Question 103 A small break LOCA is in progress.
| |
| While performing EMG E-1," Loss of Reactor or Secondary Coolant," a steam line break occurs.
| |
| The blowdown cools the RCS by 150 'F in a 15 minute period to 220 'F.
| |
| Thermal Stresses on the reactor vessel wall will:
| |
| A. Be removed immediately when the cooldown is stopped.
| |
| B. Remain until the thermal gradient is reduced by a soak.
| |
| C. Be removed when the RCS is depressurized.
| |
| D. Remain until a heat up is commenced.
| |
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| t n._ _ _ _ _ _ _ _ . _ _ - _ _ - _ _ _ . _ - _ _ _ _ _ _ _ . _ . - . . _ _ _ _ . _ _ _ _ - _ _ _ _ _ . - _ _ - . _ _ . _ _ _ . _
| |
| | |
| Question 104 t
| |
| ! Assume rods are withdrawn 50 steps each increment during a reactor start-up.
| |
| What will be the effect on the magnitude of the resulting neutron level change after each rod pull and the time required for the source range count rate to stabilize?
| |
| A. Magnitude increases, time required decreases.
| |
| B. Magnitude decreases, time required increases.
| |
| C. Magnitude decreases, time required decreases.
| |
| D. Magnitude increases, time required increases.
| |
| | |
| Question 105 1
| |
| Given the following conditions: i l
| |
| * Reactor startup in progress
| |
| * Source Range channel N31 indicates 8 x 10+4 cps-
| |
| * Source Range channel N32 indicates 7 x 10+4 cps
| |
| ! *- Intermediate Range channel N35 indicates 2 x 10-10 amps e Intermediate Range channel N36 indicates 4 x 10-1 I amps
| |
| * Permissive P-6 light is NOT LIT and NOT blocked Assume the Source Range channels are indicating properly.
| |
| Which of the following describes the conditions of the Intermediate Range instruments?
| |
| A. N35 is overcompensated.
| |
| B. N35 is undercompensated.
| |
| t
| |
| ! C. N36 is overcompensated.
| |
| D. N36 is undercompensated.
| |
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| 1 1
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| Question 106 l <
| |
| ! The plant has experienced a LOCA from full power. The crew is transitioning to EMG E-1 when the following parameters are reported:
| |
| * Containment pressure 8psig e RCS pressure 900 psig l
| |
| * RCS temperature 400*F '
| |
| l e Steam header pressure 1092 psig i e S/G "A' NR level 23 %
| |
| l
| |
| * S/G "B' NR level 20%
| |
| e S/G "C' NR level 21 %
| |
| * . S/G "D' NR level 19%
| |
| Aux. Feedwater flow is = 200,000 lbm/hr and cannot be increased.
| |
| l l What actions are required?
| |
| A. Transition to EMG FR-HS and restore S/G levels.
| |
| B. Continue in EMG E-1 because Aux. Feed cannot be increased.
| |
| C. Transition to EMG FR-H1 and restore feed using feed and condensate sys;em.
| |
| D. Transition to EMG FR-H1 but return to EMG E-1 as the S/G's are not required for heat sink.
| |
| l l I
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| 1 l-l 1
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| f
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| ,. Question 107 I
| |
| A LOCA is in progress. CTMT conditions are currently 2.8 psig aRer spiking to 12 psig while CTMT radiation indicates 3E+3 R/hr aRer spiking to 9E+6 R/hr.
| |
| Adverse containment conditions values:
| |
| A. Must still be used until the engineering staff performs an evaluation on the radiation effects to the instruments. -
| |
| B. Can be returned to their normal values based on the radiation conditions returning to the non-l adverse CTMT value.
| |
| C. Must still be used until the engineering stafTperforms an evaluation on the pressure and temperature i effects to the instruments.
| |
| l l D. Can be returned to their normal values based on pressure conditions returning to the non-adverse l CTMT value.
| |
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| i Question 108 j
| |
| The crew is operating with pressurizer level instrument BB LI-459 out of service. All of the required actions for this instrument failure have been completed. The surveillance for BB LI-460 is about to l- become late.
| |
| l t
| |
| j The Supervising operators should:
| |
| A. Trip the reactor and commence cooldown to mode 4.
| |
| l l B. Bypass BB LI-459 and declare BB LI-460 inoperable. l l
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| C. Bypass BB LI-459 and perform surveillance on BB LI-460.
| |
| D. Continue plant operation and submit LER on missed surveillance.
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| i 4
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| l Question 109 f
| |
| You are the refueling SRO. During fuel movement all containment area radiation monitors alarm.
| |
| * An assembly is being moved by the refueling machine
| |
| . The fuel transfer cart is in the Fuel Building Where are you directed to store the fuel assembly?
| |
| l j A. In the fuel building
| |
| ! B. In the RCCA change fixture i
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| C. In the reactor vessel l D. It can be left suspended within the mast.
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| i f
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| {
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| l l .
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| l Question 110 The unit is at 10% power with a unit shutdown in progress.
| |
| In preparation for the upcoming outage, mechanical maintenance is preparing to enter containment and store some equipment.
| |
| The equipment size requires both containment personnel airlock doors to be open for about 10 minutes.
| |
| The ShiA Supervisor can authorize this action:
| |
| A. Any time in Modes 1 or 2, provided an operator is standing by to close the door if a CISA signal occurs. 3 i
| |
| B. AAer the unit enters Mode 3, and temperature is reduced below 500 'F.
| |
| C. AAer the unit enters mode 4.
| |
| D. After the unit enters Mode 5.
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| 4 l
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| l j
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| | |
| Question 111 Wolf Creek maintains 11 personnel on crew in addition to the Shift Chemist, Shift HP and Shift Engineer required by Technical Specifications.
| |
| Assume minimum crew manning.
| |
| If one of the Nuclear Station Operators assigned to the Fire Brigade becomes ill and has to leave, what actiors is required?
| |
| A. No action is required other than continue trying to call in a replacement.
| |
| B. No action is required as long as long as the NSO who left is not the Fire Brigade Leader.
| |
| C. A replacement must be on site within 2 hours unless another operator is assigned to the Fire Brigade.
| |
| D. A replacement must be on site within 2 hours or a 30 day LER must be initiated.
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| l l
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| 1 1
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| l i
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| Question i12 i
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| l The plant is at 7 % power when an RCP is inadvertently tripped. The reactor does not trip.
| |
| What action must be taken prior to restarting the RCP?
| |
| l l A. Close the MSIV on the affected loop.
| |
| l B. Shutdown the reactor.
| |
| C. Allow the motor to come to a stop.
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| D. Decrease power below 5%.
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| l l
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| l l !
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| I j
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| Question 113 t
| |
| At 20% power, a reactor coolant pump trip occurs during a load transient causing Tavgt o droop to 549 F, pressurizer level drops to 18%, and pressurizer pressure drops to 2210 psig.
| |
| i Identify which parameter below has the most restrictive technical specification limit action statement time requirement.
| |
| A. Tavg.
| |
| B. Pressurizer Level.
| |
| I C. Pressurizer Pressure.
| |
| D. RCS loops in operation.
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| I I
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| i i
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| l l
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| i l
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| | |
| Question i14 Technical Specifications 3.4.8, " Reactor Coolant System Specific Activity", requires Tavg be reduced to less than 500 'F if RCS activity limits are exceeded.
| |
| l \
| |
| j Which of the following is the basis for reducing Tavg to less than 500 *F?
| |
| A. Limits the site boundary dose rate to a small fraction of the 10 CFR part 100 limit with an assumed
| |
| : one gpm steam generator tube leak.
| |
| B. Minimizes the effects ofiodine spiking following a power change in the event of an assumed one gpm steam generator tube leak.
| |
| C. Ensures the 2 hour dose at the site boundary does not exceed the 10 CFR part 100 dose guideline limits in the event of a steam generator tube rupture.
| |
| D. Prevents the accidental release of activity in the event of a steam generator tube rupture.
| |
| t i
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| l l
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| l l-Question 115 i
| |
| A planned modification requires a door in the Auxiliary Building be open for over an hour.
| |
| . A breach permit requires Shift Supervisor or Work Control Center authorization:
| |
| A. Only if the breached barrier is for a pressure boundary or train operability.
| |
| B. Only if the breached barrier is listed in attachment A," Door List" of the procedure.
| |
| C. Is not required if the door is a security door only.
| |
| ]
| |
| D. Is not required if the door is a fire door only.
| |
| l j
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| I l
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| I
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| , Question 116 l
| |
| l Which of the following is NOT a responsibility of the Shift Supervisor per AP 16C-002 " Work Controls"?
| |
| A. Incorporating and coordinating work activities into the plant schedule.
| |
| B. Ensure plant is in the proper condition for work activities.
| |
| C. Determine if equipment can be returned to service.
| |
| D. Review and evaluate action requests.
| |
| | |
| Question i17 A transient occurred while in Mode 3 that resulted in the pressurizer going solid and pressure increasing to 2835 psig.
| |
| Which of the following actions must be taken?
| |
| A. Within one hour reduce pressure to within its limit and be in hot shutdown within 6 hours and cold shutdown within the next 30 hours. .
| |
| B. No action is required, the pressure reached did not exceed the safety limit.
| |
| C. Be in hot shutdown and reduce pressure to within its limit in one hour. I D. Reduce RCS pressure to within its limits within 5 minutes. j 1
| |
| )
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| l I
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| i l
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| l l
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| Question 118 l Initial conditions:
| |
| i . "A" train RHR is out of service
| |
| = "B" train RHR is operable and in service Refueling Pool level is greater than 23 feet above the flange e Refueling operations are in progress During the last 24 hour period the operating RHR train was secured to ease the loading of fuel around the hot legs at:
| |
| 0050 to 0120 RHR pump B secured 1030 to 1130 RHR pump B secured At 1700 the refueling SRO requests the B RHR pump be secured to load fuel in the area of the hot leg.
| |
| You are the supervising operator and order the RHR Pump to be:
| |
| A. Left running because 8 hours has not passed since the last time the pump is secured.
| |
| B. Secured but note that the pump must be restarted in no more than one hour.
| |
| C. Secured but note that the pump must be restaned in no more than 4 hours.
| |
| D. Secured but note that the pump must be restarted in no more than 30 minutes.
| |
| | |
| Question 119 You are the Shift Supervisor. Maintenance informs you that they are ready to detension the first reactor vessel head bolt.
| |
| IAW WCGS procedures, whose authorization is required to change modes?
| |
| A. Chief Operating Officer.
| |
| B. Outage Manager.
| |
| C. He Plant Manager.
| |
| I D. The Manager Operations.
| |
| i l
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| l I I l
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| 6
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| _ _ _ _ _ _ ___m ___..___ _ __ _
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| l i
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| ~
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| l
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| ! Question 120 l
| |
| l l An operator is required to complete a valve lineup in an area where the radiation level is 50 mrem / hour.
| |
| The operator's current TEDE is 1750 mrem.
| |
| How long can he work in this area and not exceed WCGS administrative limits?
| |
| A. One hour.
| |
| B. Five hours.
| |
| l l C. Twenty-five hours.
| |
| D. Sixty-five hours.
| |
| l t
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| l l
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| l l
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| l l
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| I l
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| I.
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| | |
| Question 121 An unplanned containment entry is required to look for RCS leakage.
| |
| The Shift Supervisor must notify which of the following plant staff to get the containment entry authorized?
| |
| A. Security Shift Lt.
| |
| B .' Call Superintendent.
| |
| C. HP Supervisor ALARA or H? Supervisor Operations.
| |
| D. Manager Operations, i
| |
| l I
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| l
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| | |
| l l
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| l 1
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| Question 122 l
| |
| l Liquid radiological releases are approved by the: .
| |
| l A. Operations Radwaste Supervisor or designee.
| |
| B. Manager Chemistry.
| |
| C. Shift Supervisor.
| |
| I i
| |
| l D. System engineer responsible for tank being released. l l
| |
| l l
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| l 1
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| 4 l
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| l l
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| l l
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| l
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| | |
| l Question 123 While performing EMG C-21," Uncontrolled Depressurization of all Steam Generators", a red path was diagnosed on suberiticality. ]'
| |
| After commencing EMG FR-SI," Response to Nuclear Power Generation /ATWT", an orange path is diagnosed on containment. I If suberiticality becomes green during the performance of EMG FR-Si the operators should-1 A. Complete EMG FR-Si then transition to EMG FR-Zl. !
| |
| B. Complete EMG FR-Si then return to EMG C-21.
| |
| C. Perform EMG FR-Z1 then return to EMG FR-Sl.
| |
| D. Perform EMG FR-Z1 then return to EMG C-21.
| |
| l l
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| 1 l
| |
| Question 124 A General Emergency has been declared following a LOCA, the following conditions exist:
| |
| * RCS Pressure is 750 psig
| |
| * 350 'F superheat GT RE-59 and GT RE-60 read 2000 - 2200 R/hr !
| |
| * CTMT pressure is 8 psig i
| |
| * No red or orange paths exist at this time
| |
| * HP has not completed dose projections What Protective Action Recommendation (PARS) should be made? i A. Evacuate JRR, WCL, CTR and Subzones 2 10 miles downwind.
| |
| B. Evacuate JRR, WCL, CTR and Subzones 2-5 miles downwind.
| |
| C. Evacuate JRR, WCL, and CTR.
| |
| D. Evacuate JRR and WCL.
| |
| | |
| Question 125 During performance of EMG FR-S I, " Response to Nuclear Power Generation /ATWT', the Shift Engineer reports a Red Path indicated on the Core Cooling Status Tree.
| |
| l The SO should:
| |
| A. Immediately transition to EMG FR-Cl," Response to inadequate Core Cooling".
| |
| l l B. Transition to EMG FR-Cl only if the entry into EMG FR-SI was due to an orange path, i C. Remain in EMG FR-Si and perform all actions of EMG FR-Cl in parallel.
| |
| l D. Remain in EMG FR-Si until the status goes green then transition to EMG FR-CI.
| |
| t l
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| l l
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| l
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| I CHIEF EXAMINER SRO f XAM DIFFICULTY ANALYIS - WOLF CREEK 2/17/98 Legend:
| |
| L - Level of knowlege required A - Knowledge at the comprehension / analysis level K - Fundamental knowledge only C "X" indicates a chief examiner comment exists D - Overall difficulty of question on a scale of I to 5, with 5 being the most difficult Question L/D C? Question L/D C? Question L/D C?
| |
| 1 N3 35 K/2 X 69 N3 2 N3 36 A/3 70 N2 3 N3 37 A/3 71 N3 4 K/2 38 N3 72 A/2 5 A/3 39 N4 X 73 A/3 6 A/2 40 K/2 74 K/2 7 A/3 41 N3 75 N3 8 N3 42 A/2 76 K/1 X 1 9 K/2 43 K/3 77 K/2 10 A/3 X 44 K/2 X 78 K/3 l 11 A/3 X 45 K/2 X 79 K/3 12 K/2 X 46 N3 80 N3 13 N4 X 47 A/3 81 A/3 14 N4 X 48 K/2 82 A/3 15 K/1 X 49 K/3 83 N3 X 16 N4 X 50 A/3 84 K/3 17 A/2 X 51 N3 85 K/3 18 K/2 X 52 N2 86 K/3 19 A/2 X 53 K/3 87 N3 X 20 A/3 54 A/3 X 88 K/3 21 K/2 55 N3 89 K/3 22 A/2 56 A/3 90 K/3 23 N3 57 A/3 91 K/2 X 24 N2 58 K/2 X 92 K/3 25 N2 X 59 A/3 X 93 N3 26 K/2 60 K/4 94 K/2 27 K/3 61 A/3 95 K/1 X 28 A/4 62 K/2 96 K/2 X 29 A/4 63 N3 97 K/2 30 A/3 64 N3 98 A/3 31 A/2 65 K/4 99 N4 l 32 K/2 66 N2 100 K/3 33 K/3 67 K/2 X f
| |
| l 34 A/4 68 N3 A = 59 D,,, = 2.71
| |
| | |
| CHIEF EXAMINER SRO WRITTEN EXAM COMMENTS - WOLF CREEK 2/17/98 Question No. Comment 10 The answer has a specific determiner in that it is the only choice referring to the inverter and NN-02, which were discussed in the stem. Resolution:
| |
| Eliminated reference to NN-02 in answer.
| |
| 11 Because of the structure of the distractors, the difficulty of this question is minimal. The examinee only has to know that the seal water temperature increases to eliminate the distractors. Resolution: Changed one of the distractors.
| |
| I 12 Why is distractor A not correct? Resolution: Replaced distractor.
| |
| 13 KA does not match outiine, i.e. EK1.2 vice AK1.01. Resolution: Changed outline.
| |
| la KA importance rating in the outline is incorrect, i.e. 4.1/4.7 vice 3.9/4.1.
| |
| Also, from the references provided, it is not clear that Distractors C and D are plausible. Resolution: Changed reference sheet to match outline.
| |
| Distractor C and D are plausible. Changed wording to make it clear that they are incorrect.
| |
| 15 The answer is too obvious. Almost anyone would know that a smallleak in j containment would not result in a high range reading. Also, "last" should be l capitalized in the stem. Resolution: Replaced question.
| |
| 16 ,As written, this question does not solicit the desired knowledge. The intent of the question is to determine if the examineo knows the cooldown limits for the conditions listed. However, it merely asks the examinee what he would l do. Any answer which is under the limits would be satisfactory. With this knowledge, an experienced test taker can eliminate some of the distractors.
| |
| Resolution: Clarified quescion.
| |
| 17 Distractor D is implausible, it is obviously wrong to watch the power increase from 50% to 100% under the given conditions. Resolution: Revised Distractor D.
| |
| 18 The f act that the answer is the only reference to the TS could be considered a specific determiner. Resolution: Referenced TS in an additional distractor.
| |
| 19 The explanation for the answer is not fully supported by the reference in that there is no mention of Lo/Lo S/G level. Distractor A could be considered
| |
| ! correct because it is the root cause of the trip. Distractor C is correct, but not complete. We doubt that these distractors could be defended as wrong answers. Resolution: Reworded stem and Distractor C.
| |
| 25 KA does not match the outline. Resolution: Changed outline.
| |
| 1
| |
| | |
| i i
| |
| 35 If true statements are to be used as distractors, the stem must more clearly f describe what is being asked; i.e. the TS bases for rod insertion limits.
| |
| Resolution: Modified stem.
| |
| l 39 KA does not match outline. Resolution: Changed outline.
| |
| i l 44 Justify using a negative stem. Resolution: Changed distractors to make stem I positive.
| |
| 45 The knowledge solicited is almost identical to that required to answer Question 19. Replace one of these questions. Resolution: Replaced this question.
| |
| l 54 KA does not match outline. Resolution: Changed outline, i
| |
| 58 Reword the stem. The operator can apparently close the spray additive tank isolation valve anytime. The intent of the question is to ask when he/she should close the valve. Resolution: No change is necessary. An interlock prevents closing valve.
| |
| 59 Distractor B is too similar to the answer. Rewrite. Resolution: Changed distractor.
| |
| ; 67 The explanation for distractor B indicates it is a correct answer. Also, the l knowledge required is very similar to Question 7. Resolution: Replaced this question.
| |
| 76 This question is too easy. It is not written at the SRO level. Resolution:
| |
| Replaced question.
| |
| 83 KA does not match outline. Distractor A is implausible. Tripping the reactor is almost never a requirement for this type of a problem. Resolution: Revised l
| |
| distractor.
| |
| I 87 KA does not match outline. Resolution: Changed outline.
| |
| l 91 Justify negative question. Resolution: Changed stem and distractors.
| |
| ; 95 This question is too easy for an SRO applicant. Replace. Resolution: l j Replaced question.
| |
| l.
| |
| ! 96 Distractor A is implausible. The security lieutenant never makes operating decisions. Resolution: Revised Distractors A and D.
| |
| (
| |
| | |
| "cvsasuu; a f FN EJ-015
| |
| ~
| |
| ,OSS OF RER COOLING Continuoua U2a Pa#d 146 of 148
| |
| (' FIGURE 5 TIME TO BOILING IN REDUCED INVENTORY TIME TO BOIL (HINUTES) 60 I
| |
| 2 50 /
| |
| VESSEL PRESSURIZED l 1
| |
| /
| |
| -/
| |
| /
| |
| 40 7 }
| |
| /
| |
| /
| |
| /
| |
| j f 30 /
| |
| [
| |
| }
| |
| [
| |
| / s 20 l l
| |
| - s # "
| |
| }{ s # # ~
| |
| f AW~
| |
| j s ' ' VESSEL NOT PRESSURIZED ll ,
| |
| ,, r 10 I /'
| |
| Y 4
| |
| 0 5 10 15 20 f 30 OAYS AFTER SHUT 00WN
| |
| | |
| 7-i ; Revision: 5 ) !
| |
| ~f OFN EJ-015
| |
| " LOSS OF RHR COOLING {
| |
| . ril Continuous Use Page 148 of 148 {
| |
| )
| |
| lN ,
| |
| ' I FIGURE 6 i
| |
| )
| |
| TIME TO CORE UNCOVERY IN REDUCED INVENTORY
| |
| {
| |
| l \
| |
| t t)
| |
| ? TIME TO CORE UNC0VERY (MINUTES) 1
| |
| ! ) 3 1 1 1 1 I t i i
| |
| ,3 i i
| |
| ! i i t t I 1 l
| |
| {
| |
| i i i i i
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| ' i i 450 ' '
| |
| l , ,
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| |
| l I i l f' i -, , -
| |
| 400 I
| |
| / r i s i
| |
| [
| |
| s 350 i
| |
| i i / i COMPLETE CORE UNC0VERY f
| |
| imr i 300 ' #
| |
| 7 i
| |
| /
| |
| / I
| |
| , f i 250 i / i i i/
| |
| /
| |
| /i 200 # '
| |
| i i / i i i y 1
| |
| / i t l 2 / " ,
| |
| / i J7 l
| |
| / /
| |
| 150 # '
| |
| #' I ' ' '
| |
| i f i . f . i i i e i
| |
| !/ i# ONSET OF CORE UNC0VERY i if s 7 i i
| |
| i f t A i 100 I ' ' '
| |
| i i ! /c i i i i i
| |
| / !
| |
| ! I I
| |
| / i i i i i jr i ! I i t i i 50 6
| |
| i t ) I I i i
| |
| ! i i I i i i i i i I i j i i i i I l l i l t i i i i l 1 0 ' '
| |
| 0 5 10 15 20 25 30 l
| |
| DAYS AFTER SHUTDOWN 1
| |
| u______________________--_------ - --
| |
| | |
| Rsvicion: 9 SYS GS-100 l POST LOCA CONTAINMENT HYDROGEN Continuous Uco RECOMBINER OPERATION Page 9 of 16 INIT/DATE 6.3 Hydrogen Recombiner A Startup l
| |
| NOTE WHEN time is available prior the startup, THEN the hydrogen recombiners should be placed in standby for warm up prior to startup.
| |
| 6.3.1 At GS065A, turn hydrogen recombiner Power Out Switch on.
| |
| o HIS-1B - ON ET l 4
| |
| 6.3.2 Verify that red lamp on the Power out Switch plate is lit. f o Red Lamp - LIT E[ l 6.3.3 Determine recombiner power setting by performing the following: []
| |
| : 1. Record from Plant operating Logs the Pre-LOCA containment temperature.
| |
| Containment Temperature /
| |
| l
| |
| : 2. Determine pressure factor (Cp) using Figure 1, RECOMBINER POWER CORRECTION FACTOR VERSUS CONTAINMENT PRESSURE l CURVE. 1 Containment Pressure Cp / l
| |
| : 3. Multiply the reference power (GA 88-10-1, plaque on Recombiner Control Panels) by Cp to determine required recombiner power setting as indicated on the power meter.
| |
| SGS01A Reference Power XC p =
| |
| Power Setting /
| |
| b .
| |
| L
| |
| | |
| navision: 9 SYS GS-120 POST LOCA CONTAINMENT HYDROGEN Continuouc Uso RECOMBINER OPERATION Page l' of IE
| |
| =
| |
| FIGURE 1 DRY CONTAINMENT RECOMBINER CORRECTION FACTOR VS. CONTAINMENT PRESSURE
| |
| - rectus scri aa lI I' j
| |
| - 1i .
| |
| // I! i
| |
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| |
| /// \\
| |
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| l l s / ; I ) ll !!
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| |
| / / / 11 I I ! / fl./l ! ll M&38ENT
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| |
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| b .
| |
| Furt.,Ka Garfalguerf reunitet 195341 l
| |
| | |
| 3 Rovision: 14 PROTECTIVE ACTION RECOMMENDATIONS EPP 01-10.1 5
| |
| n Reference Us2 Page 8 of 27 e
| |
| ATTACHMENT A (Page 1 of 1)
| |
| { PROTECTIVE ACTION RECOMMENDATION CHART A
| |
| 9\ [ Commitment Step 3.2.1) sesmucmous J ,. These preemenve oceano . ; are for paenneg purposes orey. Preammad enammme nest take sameng conemone ne---- -.. Cenemms to consent are scame vuest to es authe bened on pasnt aanmanns. mesmer, evacumman raiens. :
| |
| gJ ames sec. > ms suuno insee candmone euo samount, och me appreenman assen Courey. NRC and FEMA emonte may yum - aparent men mese speahed by me how chart
| |
| : 2. Proescawe ecmon .-. .- lw oueude as sumannes masenger entuepsemd a most the proposed ease hmen PetECAUTsont Shanonne of the putec shouhl be conmeered se en seernesse to evacummon r she sees secaned arms & . nught be greater than the enes recorved romenne visaore er me moscend seesman m some a onest by easmy name enesved m carryme out me preencase accort I
| |
| M A. John Reenond Reserne (JRR) and Wbr Cresa Lehe (WCQ me recommenced for suncumann en a preaniennery ensaeure igen escearemon of a See Arne l Emergency
| |
| : e. Propaceed severe core esmase mecated by care cachng orange pom, or core anchng red pem. or heat est sed pem. Amuel m6 mend by GTREse or so ruedag a 2.8E + 03 Rhtr or e300 nuoroC#pm del.
| |
| Consen opereennel-START : reenemessi _ :
| |
| (See sneen A)
| |
| \ / No t
| |
| No ir DoyEMIIWee b General AssueMmmener11 Pmpeamons ressy wei Ernergency N roteese of Y when you con mese a Dedered ? . ,7 tenney Proencave Acean Recommuneston?
| |
| Yes you have Yes Propeenns rency weh
| |
| -N """"P'""'"""""*
| |
| emmy PraescDe Y-- -
| |
| '1r
| |
| ,, -N TEDE g1mem OR Theroid
| |
| * 5 Rom ?
| |
| Actumi or or propeasd unmment bream '***'
| |
| af mreefoamn N '"''**""***'e'ty. No-control or fac (See None B) y ,,
| |
| ,, Pnn.eme Accan Recaneneneumon Proences Acaon Ye,s Yes R----
| |
| Y Evacumen JRR wcL cTR.
| |
| Y 1r and sa asununnd Sunsonne Proomarm Asam Pruewsrve Acmon P.nemarn Acamn M_.._. vacume JRR. pre l R_._ Recommeneston g g g ,saceedgsome Evacuses JRR. Evocuses JRR. Eveauses JRR.
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| We' cm. -
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| Sutaones 210
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| = cm -
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| Sutmones 2-5 Weu - cm (See Proceuman) rGE _-
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| WCL & CTR
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| ,RR.
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| f ends' rr e ,,r. ,o, ,o r,,,e me6 mess dommwed endes downennd m .(See (See Precouman) (See Proceumon) Precaukon)
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| (See Procauten)
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| - END -
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| WCGS INITIAL LICENSE WRITTEN EXAMINATION KEY RO l
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| l Question answer Question answer Question enswer Question answer I
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| 1 A 26 D 51 D 76 C 2 D 27 D 52 D 77 B t
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| l 3 B 28 C 53 D 78 A 4 C 29 C 54 D 79 C 5 D 30 D 55 A 80 C 6 D 31 D 56 B 81 B 7 A 32 B 57 A 82 B 8 A 33 A 58 A 83 A 9 D 34 A 59 C 84 B 10 B 35 A 60 A 85 B 11 D 36 B 61 B 86 A 12 D 37 A 62 A 87 A 13 C 38 D 63 C 88 D 14 A 39 D 64 B 89 C 15 D 40 B 65 A 90 D 16 B 41 B 66 B 91 C 17 A 42 A 67 D 92 B 18 B 43 C 68 B 93 C 19 D 44 D 69 B 94 D 20 B 45 D 70 B 95 A 21 D 46 D 71 C 96 A 22 C 47 B 72 D 97 A 23 A 48 D 73 C 98 A 24 C 49 D 74 B 99 A 25 A 50 A 75 A 100 D l
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| WCGS INITIAL LICENSE WRITTEN EXAMINATION KEY SRO Question answer Question answer Question answer Question answer i A 26 D 51 D 101 B 2 D 27 D 52 D 102 B 3 B 28 C 53 D 103 B 4 C 29 C 54 D 104 D 5 D 30 D 55 A 105 C 6 D 31 D 56 B 106 D 7 A 32 B 57 A 107 A 8 A 33 A 58 A 108 C 9 D 34 A 59 C 109 C 10 B 35 A 60 A 110 D l
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| 11 D 36 B 61 B 111 C 12 D 37 A 62 A 112 D 13 C 38 D 63 C 113 A 14 A 39 D 64 8 114 D 15 D 40 B 65 A 115 B 16 B 41 B 66 B 116 A l 17 A 42 A 67 D 117 D 18 8 43 C 68 B 118 A 19 D 44 D 69 B 119 D 20 B 45 D 70 B 120 B 21 D 46 D 71 C 121 C 22 C 47 B 72 D 122 C 23 A 48 D 73 C 123 A 24 C 49 D 74 B 124 C i 1
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| 25 A 50 A 75 A 125 B
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| P l0 Examination Changes The following changes were made based on the following reasons.
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| . NRC comments made 1-30-98.
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| * Final operations validation of the examination and examination changes.
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| Final review of all examination questions based on the understanding achieved from working with the Chief Examiner comments.
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| Question 5: Restructured stem to improve question readability.
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| Question 7: Removed alarm titles from distractors. Determined they did not add value to the question.
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| 97@ g y.*,
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| Question 8: Changed stem tense. "Lineupp" for " lineupy. j Question 23: Improved grammar of the stem. Rewrote distractor C.
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| ? hw Question : Removed the term "using the attached pages" from the stem.
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| Question 25: Restructured the stem statement based on operations validation comments.
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| Question 36: Rewrote the stem statement to improve readability.
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| Question 37: Rewrote the stem statement to improve readability.
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| Question 52: Restructured scope of the question to replace implausible distractor.
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| Question 54: Changed the statement "no actions have been" to "With no operator action. .."
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| Question 55: Changed distractor D. Distractor D was partially correct.
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| Question 62: Improved stem and distractors to remove a possible specific determiner.
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| Question 63: Restructured the stem statement to improve readability.
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| Question 65 Improved sentence grammar.
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| Question 70: Removed the stem statement "using attached pages" Question 72: Rewrote stem and a distractor to remove fluff.
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| l l Question 75: Rewrote the stem statement to improve readability.
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| j Question 77: Restructured the initial conditions to match plant conditions.
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| Question 79: Rewrote stem statement to remove ambiguity.
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| l l
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| Question 85: Change tense of distractor D to match the question.
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| Question 89: Restructured the distractors to improve discrimination between them.
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| Question 90: Rewrote distractor A to more grammatically correct.
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| Question 94: Removed "using the attached . . . " from the question stem.
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| Question 97: Improved the grammar in the question.
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| Question i15: Restructured distractors B and C to improved discrimination. Interchanged distractors B and C to have more C answers.
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| )
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| .~
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| i l
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| l
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| ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _}}
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