ML20151E147
| ML20151E147 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 07/07/1988 |
| From: | Brockman K, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20151E133 | List: |
| References | |
| 50-416-OL-88-01, 50-416-OL-88-1, NUDOCS 8807250388 | |
| Download: ML20151E147 (181) | |
Text
{{#Wiki_filter:- 7 ENCLOSURE 1 EXAMINATION REPORT 416/0L-88-01 Facility Licensee: System Energy Resources, Inc. Facility Name: Grand Gulf Nuclear Station Facility Docket No.: 50-416 Written examinations and operating tests were administered at Grand Gulf Nuclear Station near Vicksburg, MS. Chief Examiner: /sb/d ed ' so eva 38 p nneth E Q oc)nsn, Section Chief Date Signed Operating Licensing Section 2 Approved by: - I7 Jr F.~Munro, S(ction Chief 7!7!8 Oate Signed Operating Licensing Section 1 Summary: Examinations on April 26-27, 1988. Written examinations were administered to five candidates, four of whom passed. Operating tests were administered to three candidates, three of whom passed. Based on the results described above, one of one R0 and three of four SR0's passed. g72%hk 4 4 v
5 i i, REPORT DETAILS
- 1. Facility' Employees Contacted:
*H.: Shelly
*D. Bottemiller K. Beatty.
~* Attended Exit Meeting
- 2. Examiners:
*K E. Brockman, Region II G. H.' Hopper, Region II C. W. Rapp, Region II J. M. McGhee, EG&G
- Chief Examiner
- 3. Pre-Examination Review:
During the week of April 18, 1988, members of the examination team and the facility Training and Operations staff reviewed the written examination. This effort was part of a pilot initiative to improve examination validity and relevance. This_ pre-examination review provided an opportunity to ensure all test items were accurate and concise prior to the examination administration. Specific comments on this initiative will be made under seperate cover.
- 4. Post-Examination Review:
At the conclusion of the written examination, the examiners provided your staff with a copy of the written examination and answer key for review. Facility comments concerning the written examination are included in this report as Enclosure.3. The NRC resolutions to comments made by the facility reviewers are listed below. +
- a. SR0 Exam (1) Question 6.14:
Facility comment accepted. The answer key will be modified as suggested by the facility. The training material provided by the facility is incorrect. The facility is encouraged to ensure all training materials are correct and current. (2) Question 6.15: Facility connent accepted. The answer key will be modified as suggested by the facility. The additional information will be required for full credit.
p, 2
- b. R0 Exam (1) Question 2.05:
Facility coment accepted. The answer key will modified as suggested by the facility,
- c. Exam Changes Instituted During Administration SR0 Examination-(3) Question 7.17:-
Question was clarified to elicit response concerning the bases for the entire step, not just terminating and preventing injection. (4) Question 8.08: Part (b) was deleted because facility management could not provide adequate clarification as to the intent or this procedural allowance.
- d. Changes Instituted During Examination Grading-SR0 Examination (5) Question 5.17 (c):
Answer key was clarified to not require the MSIV closure setpoint for full credit. (6) Question 5.22 (a): Answer key was clarified to accept the alte nate response of "Born later than 10 E-14 sec after fission" for full credit. (7) Question 6.11 (a): Answer key was clarified to not require the phrase "with voltage restraint" for full credit. (8) Question 6.13 (d): Answer key was clarified to accept any response indicating more than one reed switch closed. (9) Question 6.19: Point value was changed to 1.00 to more accurately reflect the knowledge level required. (10) Question 6.21: Answer key was clarified to accept any response indicating that a +5 inch bias signal is applied to sensed water level.
+
in 3
.G (11) Question 7.05:
' Answer key was clarified to not require the phrase '.' Place flow controller _
in manual" for full credit. (12): Question 7.11: Answer key was clarified to accept responses directly associated with the recirculation system for full credit. The response "Perform complete loss of CCW" will also be accepted for full credit. (13) Question 7.17: Answer key was clarified to not require the phrase "by lowering RPV level" for full credit. (14) Question 8.17: Part (c) was. deleted because sufficient reference material was not available during examination administration to correctly answer this part of the qJestion. (15) Question 8.19: The response "TS 6.3.6.1.a" will not be required for full credit because sufficient' reference material-was not available during examination administration. G. Exit Meeting At the conclusion of the site visit, the examiners met with representatives of the plant staff to discuss the examination. No generic weaknesses were noted during examination administration. During grading of the written examination, responses to several questions indicated that SR0 candidates do not fully understand the safety significance, of certain system operations. Specifically, the candidates did not recognize the safety significance of recirculation FCV runback to prevent unnecessary scrams and setpoint setdown to prevent unnecessary ECCS actuations. Also, the candidates lacked complete understanding of important emergency procedure steps. Specifically, the difference in level control between EP-2 and EP-14. The cooperation given to the examiners and the effort to t.nsure an atmosphere in the Control Room conducive to the oral examinations was noted and appreciated. The licensee did not identify as pruprietary any of the material provided to or reviewed by the examiners.
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' ' S.; NUCLEAR REGULATORY COMMISSION
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SENIOR REACTOR OPERATOR LICENSE EXAMINATION
' FACILITY: _gB8ND_QULF_1____________
.)
-REACTOR TYPE: _BBB-gg6_________________
DATE ADMINISTERED: _QS/93/2D________________ EXAMINER: _ Bgg]QN_))_______________ CANDIDATE: _____dd_S[M___________ INSIBWCI19BS_ID_GBUDIDBIE1 Us o -, separate paper-for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after_the question. The passing grade requires at least 70% in each category and a final grade of at Examination papers will be picked up six (6) hours after
"<.least 80%.
the examination starts.
% OF l CATEGORY % OF CANDIDATE'S CATEGORY
__YGLUE_ _IQIBL ___EGOBE___ _YBLUE__ ______________G81EQQBY_____________ 2+ 9% _27,tZU__ 21461 ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND
#.I' 'JE THERMODYNAMICS NE %g . 3C It .2CAUx_b _-9Er@@ ___________. ________ 6. . PLANT' SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION Sr. Il 29x99__ _ESAGE -
___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 21.rd' * *. 6 4 _20sEU_'7 _2EAAE ___________ ________ 8. ADMINISTRATIVE PROCEDURES, s ( CONDITIONS, AND LIMITATIONS
. .u.
. rw - ew 1---w 112:Z0__ ___________ ________% Totals Final Grade All work done on this examination is my own. I have neither given nor received aid. Candidato's Signature
Gu__IlfEQ8Y_QE_NUQLEGB_EQWEB_ELGUI_QEE80Il0Ut_ELulRS t_QUQ PAGE 2 IGE 800DYUGul0S OUESTION 5.01 ( ,t . 00 ) Which one of the following is NOT a factor in decay heat gcneration rate 7 a) Time at power b) Power level c) Time since shutdown d) Xenon inventory ANSWER 5.01 (1.00) d) REFERENCE GGNS OP-NP-518 pg. 7 3.2/3.5 29200BK130 ...(KA'S) OUESTION 5.02 (1.00) Which one of the following is NOT a characteristic of subcritical multiplication? a) The subcritical neutron level is directly proportional to the neutron source strength, b) Doubling the indicated count rate by reactivity addition will reduce the shutdown margin by approximately one-half, c) For equal reactivity additions, it takes longer for the new equilibrium count rate to be reached as Keff approaches one, d) A single notch of rod withdrawal will produce an equivalent equilibrium count rate increase independent of the value of Keff.
.,= _
Gu__IMgQBY_QE_UyCLgnB_EQWEB_ELANI_QEEB611QNt_E(Q1QQt_6NQ PAGE 3
- IUEBdQQYU0dlGQ ANSWER 5.02 (1.00) l d)
REFERENCE GGNS OP-NP-515 Pg 4-7 2.9/3.0 292OO3K101- ...(KA'S) QUESTION 5.03 (1.00) The reactor trips f rom f ull power equilibrium Xenon conditions.
.Four (4) hours later the reactor is brought critical and power level is maintained on range 5 of the IRMs for three (3) hours.
Which one of the following statements is CORRECT concerning control rod motion during this three (3) hour period assuming eactor power remains constant? a) Rods will have to be withdrawn due to Xenon build-in. b) Rods will have to be inserted since the critical reactor will cause a high rate of Xenon burnout. c) Rods will have to be inserted since Xenon will closely follow its normal decay rate. d) Rods will approximately remain as is while Xenon establishes an equilibrium value for this power level. ANSWER 5.03 (1.00)
-a)
RE IRENCE BFNP Xenon & SAMARIUM LP pp 412 GGNS: OP-NP-514 Reactcr Theory Chap 6 LO# 7. 2.9/2.9 3.2/3.2 292OO6K105 292OO6K107 ...(MA'S) (***** CATEGORY 05 CONTINUED ON NEXT PAGE **$**) i l
Gu__IHEQBY_QE_NQQ(EGR_EQtjgR_ PLANT _QPEBBIlQN t _ELylO@t_8NQ PAGE 4 I6EBdQQYN0dlGS QUESTION 5.04 (1.00) Assuming 100*/. RTP conditions, which one of the following, acting INDEPENDENTLY, would cause the critical power to DECREASE 7 a) Inlet subcooling is INCREASED , 1 b) Reactor pressure is INCREASED c) Peak axial power is DECREASED d) Core flow rate is INCREASE 0 ANSWER 5.04 (1.00) b) REFERENCE GGNS Heat Transfer and Fluid Flow Chop 5 lou 5.1,5.7 3.3/3.7 2.9/3.3 2.8/3.2 2.7/3.2 2.6/3.1 293009K118 293009K122 293009K123 293009K124 293009K126
...(KA'S)
OUESTION 5,05 (1.00) Which one of the following is dependent on the radiation field intensity for the amount of hydrogen released 7 a) Zr + H O -> ZrO 4 H 2 2 b) 2H O -> 2H + 0 2 2 2 l c) 2A1 + 3H O -> Al O + 3H l 2 2 23 2
- d) Fe + H O -> FeO + H 2 2
;E Uc__IHEQBY_QE_NyQLEGB_EQUE8_ELGNI_QEE68IlQN t _ELylp@t_60Q PAGE 5 to 16E8MODIN@d]QS ANSWER 5.05 (1.00) b)
REFERENCE GGNS OP-PC-505 Pg 6-10 2.6/2.9 223OO2K509 ...(KA*S) QUESTION 5.06 (1.00) The reactor is operating at 95% power with recirculation flow control in Flux Manual when the operator rapidly reduces power to 60% by reducing recirculation flow. Approximately twenty five minutes later, the operator notices that power has been slowly increasing. Which one of the f ollowing BEST explains the cause of the power INCREASE? a) A fuel temperature decreace. b) A core void fraction decrease. c) A feedwater temperature decrease, d) A Xenon inventory decrease. ANSWER 5.06 (1.00) c) REFERENCE GGNS Reactor Theory pg 4-8 LO# 2.4 2.8/3.0 293OO8K129 ...(KA'S) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
Dt__IyEDBY_9E_NU96E0B_E99EB_ELGUI_9EEBBI1982_ELUIDS2_0ND PAGE 6 ISEBU9DYUB0193 QUESTION 5.07 (1.00) Which one of the following operating conditions would cause a centrifugal pump to draw the LEAST current?
-a) "runout" b) "cavi tati on" c) "shutoff head" d) "operating point" ANSWER 5.07 (1.00) c)
REFERENCE GGNS Heat Transfer and Fluid Flow pg. 6-9596 LOH 10.1210.14 2.5/2.6 2.4/2.5 2.6/2.7 293OO6K100 293OO6K112 293006K117 ...(KA'S) OUESTION 5.08 (2.00) State whether the FINAL value of each of the following will be HIGHER, LOWER, or THE SAME following an increase in recirculation flow at pcwer, a) Void fraction b) Reactor pressure c) Feedwater enthalpy d) Steam flow r
Ut__IHEQBy_QE_NQGLEAB_PQWEB_ PLANT _QPEBATIQNt _FLQIQSu_GNQ PAGE 7 IHEBdQQYN8dlGE ANSWER 5.08 (2.00)
- a. LOWER
- b. HIGHES
- c. HIGHER HIGHER REFERENCE GE BWR Academic Series Reactor Theory Chap. 4 pp. 43 -46 LOM 4.3, 6.3 GE DWR Academic Series GGNS: Heat Transfer and Fluid Flow Chap. 3 LOM 4 2.6/2.7 2.5/2.6 292004K111 293002K104 ...(KA*S)
DUESTION 5.09 (1.50) For each of the following, select the one condition which would result in the greatest reactivity change from a control rod INSERTION. a) An area of high relative flux OR low relative flux b) Edge of the core OR middle of the core c) A deep rod OR a shallow rod ANSWER 5.09 (1.50) a) High relative flux (0.50) b) Middle (0.50) c) Deep rod (0.50) REFERENCE North Anna ROP- pp . 6.11, 6.12, 6.19 Obj. B TPT CNTO Core Control 6-13 GGNS: RxTh sec. 5 OBJ 3.4 2.6/2.5 292005K112 ...(KA'S) (***** CATEGORY 05 CONT!NUED ON NEXT PAGE ****t)
.Dz__IyE98Y_DE_NUGLE68_E9BEB_ELONI_9EEBOIION2_ELUIDSi_BND PAGE 8 IUESb9DYNod10S QUESTION 5.10 (2.00)
State whether each of the following will INCREASC, DECREASE, or REMAIN THE SAME if a RWCU filter-demineralizer ruptures causing resin intrusion. a) Reactor water ph b) Reactor water conductivity c) Feed water activity d) Steam line N-16 activity ANSWER 5.10 (7.00) (0.50) eacht a) decrease b) increase c) remain the same d) increase REFERENCE GGNS 65-1~02-I-5 pg i 302/3.6 2.6/2.7 204000K301 204000K306 ...(KA'S) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
Ut__IHggRY_QE_NyGLEGR_EQWER_PL691_QEEG61[QUt_ELUIQQu_GNQ PAGE 9 I5EBdQQYNGdlGS QUESTION -5.11 (2.00) State whether each of the following statements concerning a reactor heat balance is TRUE or FALSE. a) If the feedwater temperature used in the heat balance calculation was HIGHER than the actual feedwater temperature then actual reactor power is HIGHER than calculated reactor power. b) If the reactor recirculation pump heat input used in the heat balance calculation was OMITTED then actual reactor power is LOWER than calculated reactor power. c) If the steam flow used in the heat balance calculation was LOWER than the actual steam flow then actual reactor power is LOWER than calculated reactor power. d) If the RWCU return temperature used in the heat balance calculation was HIGHER than the actual RWCU return temperature then actual reactor power is LOWER than cal cul ated reactor power. ANSWER 5.11 (2.00) a) TRUE b) TRUE c) FALSE d) FALSE REFERENCE 1st Law of Thermodynamics EIH: L-RO-667 (10) BSEP L/P 04-2/3-E p 66 GGNS: OP-AD-545 2.6/3.1 2.3/2.9 293OO7K111 293OO7K113 ...(KA*S) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
Ut__1HggRY_QE_UUQLEGB_PQWER_E(GNI_QPGBGIlgNt _ELU1GSt_GUQ PAGE 10 IhE80QQYUGdlGQ QUESTION 5.12 (1.50) State whether each of the following in TRUE or FALSE: a) As condenser vacuum is INCREASED, more energy can be extracted from the steam. b) The multipressure condenser arrangement provides a greater average vacuum than would a single prussure condenser. c) Air ejectors would not be needed if the main condenser was absolutely airtight. ANSWER 5.12 (1.50) (0.50) eacht a) TRUE b) TRUE c) FALSE REFERENCE GGNS Heat Transfer and Fluid Flow Chap. 4 LOM 10.4,10.6 2.5/2.6 2.6/2.7 293004K113 293004K114 ...(KA'S) (***** CATEGORY 05 CONTINUED ON NEXT PAGE 4****)
O t __IHEQQX _QE_UQQLE@B..,EQUEQ _CLQUI_QEESGIlgd u_ELQ1QS t _GUQ PAGE 11 ISEEdQQXUGulGS QUESTION 5.13 (1.00) Match each of the definitions in Column A with the appropriate parameter in Column B. Column A Column B c) Ensures plastic strain is 1) GEXL limited to less than 1%. 2) APLHGR
- 3) CPR
- 4) FLPD b) The ratio of bundle power 5) Axial PF required to produce OTB 6) PCIPMR in the bundle to actual 7) MCPR bundle power. 8) Total PF
- 9) LHGR c) The ratio of individual 10) Radial FF bundle power to core average bundle pcwor.
d) The ratio of nodal power to average nadal power. ANSWER 5.13 (1.00) a) 9 b) 3 c) 10 d) 5 (4 & O.25 pts each) REFERENCE GGNS: OP-HF-509/Rev. 1 L.O. #'s 1, 3, 4, 5 2.1/2.5 2.2/2.6 2.8/3.6 3.3/3.7 293009K101 293009K102 293009K106 293009K111 ...(KA'S) QUESTION S.14 ( .75) Select the three (3) correct responses within the following statement concer ning reactor water level instrumentation. Elevated reference leg temperature will (INCREASE / DECREASE) the density of water in the reference leg resulting in a (SMALLER / LARGER) sensed differential pressure and a (LOWER / HIGHER) indicated level. l l I L
5 t _ _IH, EQBy _QE _NyG(EQ8_EQWEB_CL6NI_QEGS$11QN t _ELylQQ t _QMQ PAGE 12 IUE800RYNQdlGS ANSWER 5.14 ( .75) (0.25) each: DECREASE SMALLER HIGHER REFERENCE GGNS Instrumentation and Controls pg. 2-31,32 LOM 4.1.3,4.2,4.3.1 3.2/3.2 2.8/2.9 291002K107 291002K10S ...(KA'S) < QUESTION 5.15 (1.50) State three (3) reasons why the threchold power for pellet-clad
-interaction DECREASES with fuel burnup.
ANSWER 5.15 (1.50) Any three (3) of the following at 0.50 each:
- 1) Neutron embrittlement (of the cladding)
- 2) Thermally induced pellet growth (pellet swell)
- 3) Inward motion of clad walls (creepdown)
- 4) Chemical embrittlement (Cd/I)
REFERENCE GGNS OP-HF-509 2.9/3.3 293OO9K132 ...(KA'S) OUESTION 5.16 (1.00) State the concern if the value of MAPRAT exceedeo 1.0.
Et__IVE98Y_DE_NU96E0B_E9 WEB _ELONI_9EEBOIl9N2_ELUIPS _i BUD PAGE 13
'IUEBd9DXUBUlgS ANSWER 5.16 (1.00)
F The clad temperature could exceed 2200 F (0.75) during a LOCA (core uncovery) (0.25). ' REFERENCE GGNS: Heat Transfer and Fluid Flow pg 9-24 LO# 4.3,4.4 200/3.6 3.1/3.6 2.6/3.1 273OO9K111 293OO9K113 293OO9K115 ...(KA*S) DUESTION 5.17 (1.50) The attached Figure 5-1 represents a Pressure Controller Failure
- OPEN to 130%. Referring to Figure 5-1, explain the cause(s) of each of the following recorder indications belcw.
a) Neutron Flux DECREASE (f rom 0 to 2.0 neconds) b) The INCREASE in reactor water level (f rom 0 to 7. 0 seconds) c) The INCREASE in reactor pressure (at approximately 9.0 seconds) ANSWER 5.17 (1.50) (0.50) each: a) Void formation (due to decreasing pressure). b) Increased voiding (swell) (as reactor pressure decreases) l c) MSIVclosure(at 849 psig turbine inlet pressure inRUt). REFERENCE GGNS: OP-LO-DT-LP-OO5 Rev. 2 LOM 1. 4.1/4.1 3.8/3.9 2.8/2.0 241000K301 241000K304 241000K321 ...(KA'S) (t**** CATEGORY 05 CONTINUED ON NEXT PAGE *****) 1 l l
Qu__IHEQBy_QE_UQQLEGB_EQWEB_CLGNT_QCE6GIlQNx_ELylDQu_GNQ PAGE 14 IEE8dQQYNQdigS QUESTION 5,10 (1.00) Explain why the moderator temperature coefficient for a 75% control rod density is MORE negative than for a 25*/. control rod dannity. ANSWER 5.10 (1.00) With a greater rod density a greater number of neutronn are "lost" to the control rods (increaned leakage) . Thus a change in rod denuity effects reactivity more by allowxng increased absorption by other fuel bundlen. (Can also explain why low rod density does not have a l arge reactivity of f ect since the absorption by other fuel bundles is already so large) REFERENCE EIH Reactor Physics L/P pp 1.7-9, 10, 13. BSEP: L/P 02-2/3-A pp 141 - 143 GGNS: OP-NP-513 Reactor Theory pg 4-11 LOM 2.3 3.5/3.5 292005K104 ...(KA'S) DUESTION 5.17 (1.00) Explain why a spr,rious HPCS initiation at 15% power produces a more significant power response than a spurious HPCS initiation at 85% power. ANSWER 5.19 (1.00) At low power the cold water (sprayed into the core exit region) has a more significant effect on moderator temperature and void content. C1.O] REFERENCE l GONS OP E22-1-501 l 3.8/3.S 2090v2A201 ...(KA'S) (V**** CATECORY 05 CONTINUED ON NEXT PAGE *****) l
n D1__IOE98Y_9E_UVGLE88_E9 WEB _EL80I_9EEEDIl9Bi_EL92DSi_SUD .
/E '15 ISEBU9DYUBd2CD OUESTION 5.20 (1.00)
Explain-how a one or two notch withdrawal of a challow control rod could cause a "reverno power" effect. J ANSWER 5.20 (1.00) L The power dec.' ease -caun'd by the increase in void f raction (0.50) (in the upper part of the bundle) may be greater than the power increase caused by withdrawal of a shallow control rod (0.50) (in the lower part of the bundle). (This would cause overall bundle power to decrease on withdrawal of a shallow control rod.) REFERENCE GGNS Reactor Theory Chap 5 pg 5-25 LO #1.6,3.3,3.4 , 2.6/2.9 3.0/3.1 201003K505 292OO5K112 ...(KA'S) QUESTION 5.21 ;i.50) The 6ttached Figure 5-2 represents a Trip of Doth Recirculation Pumps. Referring to Figure 5-2, explain the cause of each of thu following recorder indications. a) The INCREASE in reactor pressure (at approximately 5.0 seconds). b) The CONSTANT steam flow (after 13.0 seconds). c) The DECREASE in reactor power (f rom 0.0 to 3.0 seconds) . - ANSWER 5.21 (1.50) (0.50) nach a) Trip of main and RFP turbines (greatly reducing steam - demand). b; BPVs are controlling reactor pressure, c) Increased void formation (due to rapid flow coastdown). REFERENCE GGNS OP-LP-DT-LP-OO3 LOM 1. i
,e -,m- - - - - ,--.- ,-- - - - , - . - - - - - ,e- - - - , - , - , , . , , , . - - - - . , , - - - , ..,,---y.--------e, .
E Ut__IVEQ8%_QE_UQGLE88_EQWE8_ELGNI_QEEBQllQU,_ELQ1QSt_GNQ PAGE 16 ISE8dQDYNGdlQS 3.9,'3.9 202001K303 ...(KA'S) l QUESTION 5.22 (1.50) Define. each of the following terms: c) Delayed neutrons b) Thermal neutrons c) Reactor- period i ANSWER 5.22 (1.50) (0.50) eacht a) Neutrons produced indirectly (delayed neutron precursors) from the fi asion process. (#&6r.as born > 10 s . E , .- f< ss ten e vc <1 $-) b) Neutrons in thermal equilibrium with their surrounding medium, c) The amount of time required for the neutron flux (reactor power)to :hange (increase or decrease) by a factor of "r-(2.718). REFERENCE GGNS Reactor Theory Chap 3 LOtt 3.1,5.8 3.0/3.1 2.7/2.7 3.7/3.7 3.3/3.3 292001K02 292001K103 292003K105 292003K107 ...(KA'S) (***** END OF CATEGORY 05 *****)
P i l6t__El6NI_SYSIEdQ_QEQLQN 4 _t CONI 6QLt_GUQ_lOQIBudEMIGIlQU PAGE 17
- QUESTION 6.01 (1.00) l-State whether a FCV inhibit WILL or WILL NOT occur for each of the following conditions.
a )' HPU tank level _ at "Tank Empty" setpoint b) . HPU oil' temperature at "Oil Warm" setpoint c) Drywell pressure at "Drywell Hign Pressure" setpoint
-d) Reactor water level at Level 3 setpoint ANSWER 6.01 (1.00)
'(0. 25 ) each
-a) WILL b) WILL NOT-c) 'WILL d) WILL NOT REFERENCE GGNS: OP-B33-2-501/Rev 2. LOH 3.c., 4., 5.a.1 3.7/3.9 3.3/3.4 3.4/3.4 3.3/3.3 2.8/2.8 202OO2A108 202OO2A208 202OO2A402 202OO2K112 202OO2K408
...(KA'S)
QUESTION 6.02 (1.00) Which one of the following BEST describes the response of the Standby Liquid Control (SLC) system if the TEST TANK OUTLET valve (F031) 'is OPEN and the SLC Keylock Control Switch for Pump 'A' is placed in the START position? a) SLC' Pump A does start when the Storage Tank Outlet valve (FOO1) reaches the full open position. b) The Storage Tank Outlet valve (FOO1) opens but SLC Pump A does NOT start. c) SLC Pump A does start but the Storage Tank Outlet valve (FOO1) does NOT open, d) The Storage Tank Outlet valve (FOOli does NOT open and SLC Pump A doeu NOT start.
PAGE le 6t__E4GUI_SYSIEdQ_QESIGNt_QQUIBQLt_QUQ_lM@IBQUEUISIlQN ANSWER 6.02 (1.00) c) REFERENCE GGNS: OP-C41-501 Print E1169-05 3.0/3.2 3.6/3.7 3.8/3.9 4.0/4.1 3.1/3.3 211000A104 211000A106 211000A109 211000A206 211000K402
...(KA'S)
QUESTION 6.03 (1.00) Which one of the following is the purpose of the recirculation FCV run-back7 a) Prevent cavitation of the recirculation pumps, b) Prevent runout of the feedpump. c) Prevent a low water level reactor scram, d) Prevent a rapid increase in recirculation flow. ANSWER 6.03 (1.00) C) REFERENCE GGNS: OP-B33-2-501/Rev. 2 LOM 4., 5.a.1, 3.1/3.1 3.8/3.9 3.5/3.5 3.1/3.2 3.3/3.4 3.1/3.1 3.3/3.3 3.6/3.6
-3.4/3 5 202002K109 202002K409 259001K116 259001K312 259001K411 ...-(KA'S)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
'bz__ELOUI_bYSIEUS DES 10Ni_G901896t_8dD_INSIBudEUIGl]QN PAGE 19 -QUESTION 6.04 (1.50)
Concerning the Reactor Recirculation System, state whether each of the following is TRUE or FALSE. a) With both recirculation pumps running or both recirculation pumps off, indicated core flow is equal to the addition of jet pump flows. b) When the Flux contro11e, is placed in MANUAL all Loop controllers;will trip to MANUAL. c) The Flux Controller CANNOT be placed in AUTOMATIC unless flow demand is GREATER than 40%. ANSWER 6.04 (1.50) a)- TRUE b) FALSE c) TRUE REFERENCE GGNS: OP-B21-501 OP-B33-2-501 Rev 2 LO4 3.c. 3.2/3.4 3.3/3.4 216000K110 216000K123 ...(KA'S) OUESTION 6.05 (1.50) For-each of the following abnormal conditions, state whether the LPCS system WOULD or WOULD NOT inject rated flow into the reactor vessel upon a valid initiation signal and Rx pressure decreasing below pump shutoff head. a) Pump suction valve from the Suppression Pool is closed. b) Flow transmitter has a failed 'HIGH' High Pressure Tap. c) Instrument air to the Testable Check valve is lost. l l J
b __ELONI_SygIgdg_DggigN _QQUIBQL2, 1 ^NQ,,10SIBudEUJ3IJQU PAGE 20 ANSWER 6.05 (1.50) (0.50) each: a) WOULD NOT b) WOULD c;' WOULD REFERENCE GGNS: OP-E21-501 OBJ 37B 3.4/3.4 2.6/2.6 3.2/3.2 209001A202 209001K102 209001K405 ...(KA'S) <- QUESTION 6.06 (1.00) State the two (2) affects on the RPS system of removing the RPG shorting links. ANSWER 6.06 (1.00) (0.50) each:
- 1. Allow SRM scram functions
- 2. Removes all neutron monitoring scram coincidence REFERENCE GGNS: OP-C71-501 Rev 1. 71 pg. 21 LOM 3.a. 3.c.,5.a.1,5.b.
3.7/3.9 3.5/3.7 3.3/3.5 212OOOK101 212OOOK402 212OOOK411 ...(KA'S) QUESTION 6.07 (1.50) State the three (3) requirementss) and applicable setpoints for the RHR minimum flow valve (F064) ta receive an AUTOMATIC open signal.
' 6 h__EL,'QUl _QYSIEdQ _DEQ1QU u _QQNISQL t _/30Q _ LNGIGydE UIGIlQU . PAGE 21 v ANSWER 6.07 (1.50)- 0 . (0.50) each: 1 Flow less than 1000 gpm ( +/- 100 gpm )
.2. Pump' running (breaker closed)
- 3. ' 8 second time delay REFERENCE GGNS: OP-E12-501 P. 12 3.2/3.3 3.'8/3.7 3.2/3.'3 3.8/3.7 3.8/3.7 4.1/4.0
-203OOOA103 203000A203 203000A301 203000A304 203000K403 ...(KA'S)
OUEST!ON 6.08 ( .75) a) State the ADS initiation requirement (s) that are NOT bypassed when ADS is manually initiated. b) State the number r~ SRV's that can be operated from the Remote Shutdown Fanel. ANSWER 6.08 ( .75)
- a. ECCS discharge pressure interlocks (0.50)
- b. Six (0.25)
REFERENCE GGNS: OP-E22-2-501 P. - 7 3.9/3.9.3.9/3.9 3.8/3.8 3.8/4.0 3.8/4.0 3.7/3.7 218000K105 218000K106 218000K402 218000K403 218000K501
...(KA'S)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
bi__ELONI_SYSIENg_ DES 10N2_GQUIB961_GUD_JUSIBUDENIGIlgN PAGE 22 QUESTION 6.09 (1.00> State whether each of the following values will fail CLOSED,
, OPEN, or AS 18 if Instrument Air System pressure is lost, a) CRD FCV b) RFP Minimum Flow Valve c) Feedwater Startup Flow Control Valve d) Division 2 Drywell Chillers Temperature Control Valves ANSWER 6.09 (1.00)
(0.25) each:
.a) AS IS b) OPEN c) CLOSED d) OPEN REFERENCE GGNS: ONEP 05-1-02-V-9 OP-C11-1A-501 p 13 SOI 04-1-01-P72-1 DCP 84-0064 3,O/2.9 3.0/3.0 3.2/3.2 3.6/3.8 201001K603 223OO1K601 259001K601 259002K601 ...(KA'S)
DUESTION 6.10 (1.00) Which one of the following would DEFEAT manual opening of the RHR Test Return Line valves? a) Reactor water level at Level 3 b) LPCI initiation c) Minimum flow valves not fully shut d) Containment Spray initiation l l
'st__P(6dI_QySIENS_DESLQNu_GQUIBQLt_6ND_LNSIBQUENI611QN PAGE 23 ANSWCR 6.10 (1.00) d ).
REFERENCE GGNS: OP-E12-501 pp 1450 OP-LP-SYS-LP-E12 Rev. 02 pg 21 of 46 LOH 3.d.,3.e.,4.c.,5.a.1.,5.a.2 3.4/3.6 3.5/3.7 3.2/3.4 3.2/3.2 3./3.4 3.0/3.0 2.8/2.9 3.5/3.5
'226001A105 226001K101 226001K102 226001K409 226001K613
...(KA'S)
QUESTION 6.11 (2.25) a) List five (5) trip functions which are BYPASSED when the Division III Emergency Diesel Generator is operating on a LOCA ini ti ati on signal. b) List three (3) UNIQUE signals which will 6!!TOMATICALLY start the Division III Emergency Diesel Generator and bring it up to speed and voltage. ANSWER 6.11 (2.25) a) Any five (5) of the following at 0.25 each:
- 1. Low lube oil pressure
- 2. High Jacket water temperature
'. Loss of excitation (generator lockout)
- 4. Reverse power (generator lockout)
- 5. Generatorovercurrent(withvoltagerestraint)
-6. High crankcase pressure b) Any three (3) of the following at 0.33 each:
- 1. Reactor vessel level low (-4t.6")
- 2. High drywell pressure (+1.79 psig)
- 3. Manual HPCS System initiation pushbutton
- 4. ESF Bus 17AC Undervoltar,e (timedelayed and instantaneous)
REFERENCE GGNS: OP-LO-SYS-LP-P81 LO #4a 4.0/4.2:3.8/3.7 264000K402 264000K408 ...(KA'S) (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
6s__EL8MI_SYSIEUS_ DESIGNt _QQUI6QLt_QUQ_IUSIByUEUIGIlgN PAGE 24
-QUESTION 6.12 f.oosf{f (W) i State three (3) of the four (4) conditions, including applicable cetpoints, under which the output of the flux estimator is the ACTUAL APRM signal.
- l. c o br69, ANSWER 6.12 (W)
(0.50) each: A> C;: ' - :- t " f r.y p _, _ m _ t;;, , ,, La7 ag Stefw
- 2) APRM flux signal greater than or equal to 110% of rated neutron flux U) Estimated neutron flux signal minus APRM flux rignal is greater than or equal to 5% of rated neutron flux.
^;
, L. J m u. a :'t':n m m
- '2'" "hits" .2thir "
..c ' ' ' '
- m ". c a 'r REFERENCE GGNS: SD-B33-2/Rev. 2 pg. 32 of 63 Flux Estimator Technical Manual 3.4/3.4 3.3/3.4 2.6/2.6 3.6/3.6 202KOO2 GOO 202OOO2K11 202OO2K408 202OO2K502 ...(KA'S)
QUESTION S.13 (2.00) Describe the condition (s) which will generate each of the following indications on the Operator Control Module, a) Channel Disagree b) Insert Required c) Rod Drift d) Data Fault l
1 I'_6c__EL,6NI_SYSIEd@_DE@l@N t _QQNISQLt_6ND_lNQISQUEUIGIlQN PAGE 25 y ANSWER 6.13 (2.00) (0.50) eacht a) (The RGDS finds) disagreement between the signals received from the 2 RACS. b) Any one control rod not fully inserted (at 00) with the mode switch in Refuel. c) Control rod pas =ing an odd reed switch position with no rod motion command. d) More than one reed switch closed per RPIS channel (except full in/ full out). ( Acc GPl* Ad5W e4 5 Ced CW M4 404 nMd 2. R C N Sw IrtM cusd D ) REFERENCE GGNS: OP-C11-2-501 3.2/3.2 3.5/3.5 2.5/2.7 3.2/3.2 3.8/3.2 3.7/3.7 3.0/3.3 3.3/3.5 201005A202 201005A206 201005K401 201005K406 201005K513
...(KA*S)
OUESTION 6.14 (2.00) Referring to Figure 6-1, answer each of the following concerning the MSIV Leakage Control System. a) Explain the initiation sequence f or i.'al ves COO 1 A, FOO2A, and FOO3A within the first 5 minutes. Assume M3IV leakage is minimal (as designed). Indicate approximate timing in your response. b) Following an initiation of the INBD MSIV-LCS there are two (2) system conditions which are indicative of excessive MSIV leakage. Describe these conditions and their resultant automatic action (s) on Subsystem A. Indicato component (s) affected, timing, and setpoints in your response.
bz__EL.ONI_SYSIgdS_pgglGN 2_CONIBg62_ONp_INSIBUMENIGIJgN PAGE 26 ANSWER 6.14 (2.00) { a) F001A, F002A, & F003A auto open on initiation F003A auto closes in -tr9 minutes (0.5 each) 2.p , , b) High Pressure - 5 psig between MSIV's after 1 minute ;9 Leakage High Flow - 22 SCFH after Hi Flow Leakage Timer -' times out (13 minutes) 71',a .I Closes F001A & rvi_u T d2A F003A Closes ONLY on High Pressure g ,i _ OlV - I REFERENCE GGNS: OP-E32/E38-501 pp 56 3c0/2.9 2.6/2.6 3.4/3.6 293001A106 293001G007 293003A105 ...(KA'S) OUESTION 6.15 (1.00) With regard to the Nuclear Instrumentation System: a) During a reactor shutdown, with the mode switch in STARTUP, the IRM's are reading 13 on range 5. State any automatic action (s) which could occur as a result of downscaling the IRM's to range 4. b) While operating at 100% power you bypass APRM Channal A. State the SPECIFIC effect, if any, that this has on the reactor recirculation system? ANSWER 6.15 (1.00) a) Rx scram (IRM's > 120/125 of scale and mode suitch not in RUN) C :- 3 :$$. 2 d
' Rob w rt)40 raw A t. Stocx ( tb4's i 18 8/itt 6 F s cels ) (p 15")
b) Rx recirc. flux controller automatically switches to APRM channel E. (0.5) REFERENCE GGNS OP-C51-4 L.O. 83 CP-B33-2 L.O. 84 OP-C51-2 L.O. #5 3.6/3.7 3.9/3.9 3.6/3.6 202002K607 215003K101 215005K109 ...(KA'S) (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
ht__EL8MI_SYQIENS_DESIGNt_GQNIBQLt_GNQ_INSIEUNENIOllgN PAGE 27 OUESTION 6.16 (2.00) Concerning the Division I Diesel Generator PARALLEL control handswitch: a) State the two (2) functions accomplished by placing the parallel control handswitch in the PARALLEL position. b) State what the GOVERNOR CONTROL is used f or:
- 1) BEFORE the output breaker is cloned.
- 2) AFTER the output breaker is closed.
ANSWER 6.16 (2.00) a) 1) Allows the output breaker to be shut (and parallel the diesel with normal source) (0.5)
- 2) Resets the diesel governor to the droop mode (0.5) b) 1. Diesel speed (f requency) (0.5)
- 2. Diesel load control (0.5)
REFERENCE GGNS: OP-P75-02 L.O. 85 SOI-P75-1 3.8/4.1 2.5/2.7 2.6/2.7 3.4/3.4 264000K101 264000K403 264000K406 264000K505 ...(KA'S) -QUESTION 6.17 (1.00) During a reactor start-up under Cold Conditions the operator adjusts the Control Rod Drive pressure control valve to maintain a +260 psid between CRD and reactor pressure. Explain how this pressure differential is AUTOMATICALLY maintained as reactor pressure increaues during the start-up.
st__Eh8NI_SYSIEdS_DESJGN2_99 NIB 9L2_8ND_JNSIBQMENISIJQN PAGE 28 ANSWER 6.17 (1.00) The FCV opens up as reactor pressure increases (maintaining a constant-flow and therefore a constant pressure to the PCV). REFERENCE GGNS: OP-LO-SYS-LP-C11-1A-03 LO #3e 3.1/3.0 3.1/2.9 201001A101 201001K408 ...(KA'S) OUESTION 6.18 (1.00) Explain the basis for the vacuum breakers installed on the SRV discharge lines. ANSWER 6.18 (1.00) Prevents drawing water from the Suppression Pool CO.50] when the line cools after discharge EO.503. REFERENCE GGNS: OP-E22-2-501 pp. 7 3.1/3.3 2.7/3.0 239002K403 239002K506 ...(KA'S)
- l. #6 QUESTION 6.19 (4 r59)
The reactor is at ~45% power as indicated by the APRM's. The Reactor Operator is continuocsly withdrawing control rod 28-53 from position 12 t , posi tion 48 as required by the rod pull shest. As the rod passes through position 18 the rod block annunciator alarms and the rod settles to position 20. A withdraw block is indicated on the Operator Control Module's Rod Motion Section. a) State the reason for the rod block. b) Explain why the system enforces a rod block under these conditions.
?t__EL6UI_EYF.IEdG_QEQ1GN t_CONIGOLt_@NQ_lNSIGQUENI@IlQU PAGE 29 ANSWER 6.19 ( )
- a. The operator attempted to withdraw the rod greater than four positions between the LPSP and HPSP. ( $.fy)
- b. To alert the operator of possible power peaking due to withdrawal of the same rod (or gang of rods) when above the LPSP and below the HPSP. ( p yg)
REFERENCE GGNS: OP C11-501 3.5/3.5 3.3/3.3 201005K403 201005K511 ...(KA'S) OUESTION 6.20 (1.00) Explain why it is necessary to simulate the fuel time constant as an input to the APRM High Thermal Power Scram signal. ANSWER 6.20 (1.00) (Thermal power increase is developed after the initial neutron flux increase due to the fuel time constant) Since APRM's sense flux, not thermal power (0.50), the simulated fuel time constant provides a trip signal that is analogous to thermal power (0.50). REFERENCE GGNS: OP-LP-SYS-LP-C51-4-02 pg. 9 of 28 LO #3a.,9b.,9c. 4.0/4.0 4.1/4.2 3.7/3.7 4.1/4.1 3.8/3.9 3.5/3.6 215000A104 215000A204 215000K101 215000K402 215000K407
...(KA'S)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
bt__EhBUI_SYSIEUS_pESIGN 1_COUIBQL.2_GUp_lUSIBydENIGIJQN PAGE 30 OUESTION 6.21 (1.00) During power operations at less than 45% steam flow, actual reactor water level i s maintained five (5) inches LOWER than
.sotpoint.
i a) Explain HOW actual reactor watur level is maintained five (5) inches LOWER than setpoint including any signal (s) that are utilized. b) State the BASES for maintaining actual reactor water level five (5) inches LOWER than setpoint at less than 45% steam flow. ANSWER 6.21 (1.00) (0.50) each: a) Sensed reactor water level (to the feedwater control system) is modified by total steam flow (so that actual water level inside the shroud is reduced). (/14 3o AccCPT 26st%mg igju a trds A 6 " Enf s 1s APPLCes to stasta voict Level.) b) f1inimize carryover REFERENCE GGNS: OP-C34-501/Rev 1 LO# 3.a.,4., 5.a.1 Feedwater Control System C34 SD/Rev 2 pg 9, 10 3.6/3.7 3.8/3.8 3.8/3.7 3.5/3.5 3.8/3.9 3.8/3.8 259001A304 259001 GOO 7 259001K108 259001K109 259002K103
...(KA'S)
QUESTION 6.22 (1.00) State the bases for the two (2) Setpoint Setdown level demand values of 54" and 18".
6t__ELGUI_SYSIEdQ_QESl@Mt_QQNIBQ6t_8NQ_lNQISydEUIGIlQN PAGE 31 ANSWER. 6.22 (1.00) (0.50) each 1). High'er level-demand (54") to prevent reactor water level from decreasing to Level 2. D Lower level demand (18") to prevent overfilling the reactor -(* vessel.
' REFERENCE
-GGNS: SD-C34/Rev 2 pg 16 of 27 OP-C34-501 Rev 1. pg 19 of 31 LO# 4.,5.a.2,5.b.,6.c.3.
t 3.8/3.9 2 9/2.9 3.5/3.4 3.1/3.1 3.8/3.8 3.6/3.6 3.4/3.4 3.0/3.0 3.5/3.6 3.6/3.7 259002A101 259002K103 259002K404 259002K412 259002K501
...(KA'S)
(***** END OF CATEGORY 06 *****)
Zt__ERQLgQUBES_ _UgBU662_6BNQBdBgz_EMEBGENgy_GNp PAGE 32 88DI9t9 GIG 86_G901896 QUESTION 7.01 (1.00) Which one of the following is NOT an indication of.an uncoupled rod in accordance with'ONEP 05-1-02-IV-1, Control Rod / Drive Malfunction 7 a) Failure of a CRD to operate on a normal insert or withdraw signal b) Rod overtravel alarm c) Loss of Position 48 indication d) Loss of the Full Out light ANSWER 7.01 (1.00) a) REFERENCE GGNS ONEP 05-1-02-IV-1 Rev 20 pg 1 3.1/3.1 2.8/2.9 3.2/3.3 3.7/3.8 3.6/3.4 201003A202 201003 GOO 8 201003KiO3 201003K402 201003K405
...(KA'S) l l
i f (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
" Z1__PQQGEQUBES - NORMAL 1_8BNgBMBL 2_EMgBQgNCY_8ND PAGE 33 a BBD196991GBL_G9 NIB 96 QUESTION 7.02: (1.00)
Referring to Figure 7-5, which one of the following is correct concerning step LP-31 of EP-14, Level / Power Control? a) Once RPV pressure decreases below the Minimum Alternate e Flooding Pressure (MAFP), adequate core cooling is assured.
-b) Once RPV pressured decreases below the Minimum Alternate Flooding Pressure (MAFP), steam flow through the core does not provide adequate core cooling.
c) If there are no SRVs open and pressure remains above the Minimum Alternate Flooding Pressure (MAFP), sufficient natural circulation flow through the core exists to provide adequate core cooling. d) If at least 2 SRVs are open and pressure remains above the Minimum Alternate Flooding Pressure (MAFP), insufficient natural circulation to provide adequate
, cor e coo, ling exists and injection must be reestablished
< to increase RPV water level.
ANSWER 7.02 (1.00) b) REFERENCE GGNS.OP-EP/SPDS-510 pg. 13 LO# 3
, 4.4/4.5 4.:O/4.2 4.3/4.5 4.1/4.5 295037K204 295037K209 295037K302 295037K303 ...(KA'S)
(t**** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
ZL__PRQCEQUGEQ_ _NQBd86t_8BBQ800Lt_EdE8@ENGY_8NQ PAGE 34 B001RLQ51GOL GQNIBQL QUESTION 7.03 (1.00) Which one of the following is correct if ALL rod position indication is lost-in accordance with ONEP 05-1-02-IV-6, Loss of RPIS7 a) Immediately move the control rods to posi tion with an loperable position indication, b) Obtain an OD-7 printout before and after moving any control rod. c) Withdrawal of control rods is allowed using only singl e notch movement. d) Control rod movement is allowed only by scram. ANSWER 7.03 (1.00) d) REFERENCE GGNS ONEP 05-1-02-IV-6 Rev 12 pg 1 3.4/3.4 3.3/3.3 3.5/3.7 3.8/3.8 2.9/3.8 3.1/3.2 3.7/4.0 214000A303 214000A402 214000 GOO 5 214000K106 214000K304
...(KA'S)
DUESTION 7.04 (1.00) According to ONEP 05-1-02-III-3, Decrease in Recirculation System Flow Rate, which one of the following reactor scrams will occur as a result of a recirculation pump shaft seizure? a) APRM Flow-biased Thermal Power b) RPV High oressure c) RPV High Level d) AFRM High Flux
~
y_ l 3 - E69G60V866 _NQBdQLt_6BNOBdG(t_EMEBQENQY_GNQ PAGE 35
.00D1060GIC66_CQNIBQL i
ANSW'ER 7.04 (1.00) c) l I REFERENCE GGNS ONEP 05-1-02-III-3 Rev 16 pg 3 3.6/3.7 4.1/4.1 3.4/3.4 3.5/3.6 3.4/3.5 3.6/3.6 3.9/3.9 3.7/3.7 202001K101 202001K102 202001K105 202001K122 202001K123
...(KA'S) .
l l QUESTION 7.05 (2.00) In accordance with ONEP 05-1-02-IV-1, Cor; trol Rod / Drive Malfunctions, list the immediate operator actions required to place the standby CRD pump in service following a trip of the operating CRD pump. ANSWER 7,05 (2.00) (0.50) eacht i) (Pl ace flow controller in manual and)decreaseoutputto zero.
- 2) Verify FCV is shut.
- 3) Start the standby CRD pump.
- 4) When charging header pressure has returned to norn.al ,
manually adjust { low controller cutput (to 54-66 gpm and place i n automati r.:) . REFERENCE GGNS: ONEP 05-1-02-IV-1 Rev. 20 pg 2 3.1/3.1 3.1/3.2 3.7/3.5 i 295022A101 295022G010 295022K202 ...(KA*S) (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
'31__8@QGEQQBES_2_NQBdObt_00808066t_EUE8GENQY_8NQ- PAGE 36 8001060 GIG 66..GQUIBQL QUESTION 7.06 (2.00)
State ~the' basis for each of the following procedural precautions, a) .SOI-04-1-01-G33-1, Reactor Water Cleanup, cautions to operate the RWCU system at the maxi mum permi'ssible temperature end f]ciw rate to both feedwater lines during low
?. feedwater flow conditions.
b) SOI-04-1-01-N64-1, Off-Gas Sy-+ tem, cautions that the charcoal-adsorber beds should be bypassed during initial Off-Gas System startup. a. c) SOI-04 21 -O!-E13-1, RHR System, cautions not to operate RHR in the LPCI injection modo unless required during an emergency condition. d) SOI-04-1-01-P75-1, Standby Diesel Generator, cautions not to operate the diesel generator without' air pressure. ANSWER 7.06 ' (2.00) (0.50) each: a) Limit feedwater nozzle temperature transients b) Prevent wetting of the charcoal adsorbers c) 'Impingment damage to in-core instrumentation could occur . d) DG auto shutdown features are inhibited. REFERENCE GGNS SDI-04-1-01-G33-1 SOI-04-1-01-N64-1 SOI-04-1-01-E13-1 SOI-04-1-01-P75-1 2.9/3.8 2.9/3.8 3.5/4.4 3.4/4.1 203OOOGOO5 204000 GOO 5 264000 GOO 5 271000 GOO 5 ...(KA'S) f OUESTION 7.07 (1.50) t List the three (3) reasons for controlling RPV level differently in EP-14, Power /Lovel Control, than in EP-2, RPV Control.
Zu__ESQGEQUEEQ_ _UQBMG(u_GENQBdGLt_EUEBQESQY_GNQ PAGE 37 c BQQ106901GGL_GQUIBOL ANSWER 7.07 (1.50) Each of/the following at (0.50) each: 1
- 1) Minimize borcn dilution l 2b, Promote mixing of baron 3)' Minimize suppression pool temperature rise
<< 1 REFERENCE GGNS: OP-LP-EP/SPDS-LP-010 pg. 4 of 19 LO# 4 4.4/4.5 4.0/4.2 4.3/4.5 4.1/4.5 295037K204 295037K209 295037K302 295037K303 ...(KA'S) ,
1 1 1
-QUESTION 7.08- (1.50)
List the plant conditionn that must be maintained during the
. performance of training startups per IDI-1, Cold Shutdown to Generator Carrying Minimum Load.
ANSWER 7.08 (1.50) Each'of the following at (0.50) each:
- 1) Reactor unpressurized
- 2) < 1% rated thermal power
- 3) < 200 F reactor coolant temperature REFERENCE GGNS 101-1 Rev 3' pg. 29 TS 3.10.3 4.2/4.2 294001A102 ...(IGY S)
OUESTION 7.09 (1.00) State the status of each of the following when generator load is at 15%. a) Reactor Recirculation pumps b) Reactor Recirculation Flow Control Valves c) Reactor Recirculation Loop Flow controllern d). Reactor Feedwater pumps v-
3t_ ESQGEQQBE@_ _NQBdQLu_GENQBd6Lt_EdE8QENQY_ANQ PAGE 38
.68910690;GGL_GONISOL ANSWER 7.09 (1.00)
EacP. of the following at (0.25) each: l a) Both pumps on LFMG b) maximum open position c) Manual l d) One RFPT operating l l REFERENCE GGNS 10I-2 4.2/4.2 l 294001A102 ...(KA'S) QUESTION 7.10 ( .50) In accordance with ONEP 05-1-02-V-5, Loss of Feedwater Heating, state the immediate operator action to a loss of feedwater heating due to a generator load reject or turbine trip within bypass valve capacity. ANSWER 7.10 ( .50) Trip the recirculation pumps to the LFMG (0.50). REFERENCE GGNS ONEP 05-1-02-V-5 Rev. 19 pg 2 3.8/3.9 3.9/3.9 3.6/3.8 3.8/3.9 3.1/3.3 259001K312 295001A202 295014A102 295014K106 295014K207
...(KA'S)
QUESTION 7.11 (1.50) In accordance with ONEP 05-1-02-V-1, Loss of Ccmponent Cooling Water, state the immediate operator actions relating to the Reactor Recirculation System if recirculation pump motor temperatures continually INCREASE during a partial loss of Component Cooling Water. l f
Zs__EBOCEDQBEQ_n_UQEd66,_OEUQBd86t_EdEEGENGy_GNQ PAGE 39 880106001G86_GQNIE06 ANSWER 7.11 (1.50) Each of the following at (0.50) each:
- 1) Close FCV to minimum
- 2) Shift to LFMG (at high temperature alarm)
- 3) (Perf orm actions f or complete loss of CCW)
AL .'_,c'.c -=1 n. x,7 trip pumps (within one (1) minute) ~' -
= ,- c-, ning C C 's' p u o.p - . . ' L m ,; 1m c' 1"' a mr_
t c a t ar _. d e REFERENCE GGNS ONEP 05-1-02-V-1 Rev 11 pg 2 3.5/3.6 3.3/3.4 3.4/3.6 3.3/3.4 3.3/3.4 3.1/3.9 3.4/3.3 3.1/3.2 2.9/3.1 3.1/3.2 3.7/3.7 3.4/4.2 295018A201 295018K101 295018K201 295018K202 295018K302
...(KA'S)
QUESTION 7.12 (2.00) State the limits that require the posting of each of the following areas in accordance with 01-S-08-2, Exposure and Contamination Control. a) Contamination Area b) High Contamination Area c) Potential Airborne Radioactivity Area d) Very High Radiation Area ANSWER 7.12 (2.00) Each of the following at (0.50) each: a) Alpha - 20 dpm/100 cm-cm; Beta / Gamma - 1000 dpm/100 cm-cm b) Alpha - 200 dpm/100 cm-en.; Beta / Gamma - 10000 dpm/100 cm-cm c) Alpha - 1000 dpm/100 cm-cm; Beta / Gamma - 50000 dpm/100 cm-cm d) In excess of 1000 mrem in any one hour REFERENCE GGNS AP 01-S-08-2 Re/ 15 pg 4,7,0 3.3/3.8 3.3/3.6 294001K103 294001K104 ...(KA'S) (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
Zi__EBQGEQUBEQ_ _NQBUOLt_0EURBUGLt_EME8GENQY_8NQ PAGE 40 86010LQQ1GOL_GQUISQL QUESTION 7.13 (2.00) HList four (4) of the five (5) bases f or the performance of EP-12, Emergency,Depressurization. ANSWER 7.13 (2.00) Any four (4) of the following at (0.50)- each: - J
- 1) Minimize energy. addition to the primary or: secondary d"(( ^^"" "
containment.
- 2) Minimize discharge of reactor coolant from the RPV.
- 3) Minimize radioactive release to the secondary containment.
- 4) Reduce energy contained in the RPV.
- 5) Maximize injection flow into the vessel from motor driven systems.
REFERENCE GGNO OP-LP-EP/SPDS-LP-OOB Rev 01 pg 5 of 13 LOM 2. 6.0/4.1 4.0/4.1 218000K101 218000K102 ...(KA'S) QUESTION 7.14 ( .1. 50 ) Excepting TS/ Log requirements, list the three (3) additional conditi'ons that must be met prior to placing the reactor mode switch in RUN per IOI-1, Cold Shutdown to Generator Carrying Minimum Load. ANSWER 7.14 (1.50) (0-50) each
- 1) All operable APRMs indicate > 4%
- 2) Main Steam Pressure > 850 psig (and LO PRESS alarm clear)
- 3) Recorders transferred to APRMs l REFERENCE l GGNS 101-1 Rev 37 pg 48
! 3.3/4.2 215005 GOO 5 ...(KA'S) (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
. 7. . PBOCEDURES - NORMAL1_ADNgRMBL2_EMEBOENgy_8ND PAGE 41 BSDIOLOG1986_GOUIB96 QUESTION 7.15 (1.00)
Referring to Figure 7-1, state the basis for requiring SLC be initiated prior to suppression pool temperature exceeding 110 F at step RC/Q-9 of EP-2, RPV Control. ANSWER 7.15 (1.00) Insures that the reacter will be shutdown (0.50) before Emergency RPV Depressurization is required (0.50) (by the Heat Capacity Temperature Limit). REFERENCE GGNS: OP-LP-EP/SPDS-LP-OO3-01 LO# 3. 3.8/3.8 3.9/4.1 4.0/3.9 4.1/4.3 4.2/4.4 4.0/4.2 4.3/4.5 3.8/4.1 295015G010 295015K102 295015K104 295037K101 295037K103
...(KA'S)
QUESTION 7.16 (1.00) State the basis for the Maximum Core Uncovery Time Limit. ANSWER 7.16 (1.00) (Maximum time the core may remain uncovered with no heat transfer to water or steam) without exceeding the 2200 F limit at any point (1.00). REFERENCE GGNS: OP-LP-EP/SPDS-LP-OO9 pg 17 or 19 3.6/3.8 4.0/4.2 216000K122 295037K209 ...(KA'S) QUESTION 7.17 (1.00) Referring to Figure 7-2, state the basi s f or + n~ .,r'i,g -- d-p; . "tig all : - j e - & 4 e, n in+n &n r:e.o m.m_ q q r_ ggg - .. g s n r n r.
*Hy step LP-13 of EP-14, Level / Power Control.
m
- 7. PBgggDURES - NORUBL2_6DNQRdAlz_EdgS9ENGY_AND PAGE 42 88DIDL991086_GQNISQL l
ANSWER 7.17 (1.00) Cause Natural Circulation core flow to decrease (by lowering RPV l ovel )(0. 25) and decrease reactor power by increasing voi d fraction (0.75). REFERENCE GGNS: OP-LP-EP/SPDS-LP-014 pg. 9 of 19 3.8/3.9 216000K119 ...(KA'S) OUESTION 7.18 (1.00) Referring to Figure 7-3, state the reason for terminating all injection flow at step RF-30 of EP-13, RPV Flooding. ANSWER 7.18 (1.00) Water level must be reduced (0.50) to bring the instrumentation back on scale and verify proper operation (0.50). REFERENCE OP-LP-EP/SPDS-LP-013 4.4/4.4 3.9/4.5 295031G012 295031K201 ...(KA'S) QUESTION 7.19 (1.00) Referring to Figure 7-4, state the basis for prohibiting containment spray actuation while in the shaded area of CN-T-1. ANSWER 7.19 (1.00) (Combination of eveparative and convective cooling) results in depressurization rater, which exceeds the negative design pressure of the containment. ! REFERENCE GGNS OP-LP-EP/SPDS-LP-OO4-01 LO# 11 l 3.6/4.1 3.7/4.1 4.0/4.1 l 275024A209 295024K303 295024K308 ...(KA'S) l l (***** CATEGDPY 07 CONTINUED ON NEXT PAGE *****)
3t__ESQGEQUBES_ _dQBdQLt_8ENQBd86t_EMESQENGY_8NQ PAGE 43 6801960 GIG 8L_GQNISOL QUESTION 7.20 (1.00) According to 04-1-01-N19-1, Condensate System, state the alternate method of determining if adequate rejection flow exists for CRD pump suction if it cannot be determined by flow measurement. ANSWER 7.20 (1.00) If CRD Oxygen concentration is greater than hotwell Oxygen concentration then adequate reject flow does not exist. REFERENCE GGNS 04-1-01-N19-1 3.1/3.1 2.8/2.8 2010001A20 201001K101 ...(KA'S) OUESTION 7.21 (1.50) Referring to Figure 7-6, state the basis for terminating and preventing all injection flow to the RPV, except for baron and CRD inj ecti on, at step ED-4 of EP-12, Emergency Depressurization. ANSWER 7.21 (1.50) Injection of large volumes of cold, unborated water (0.50) could add sufficient net positive reactivity (0.50) to induce reactor power excursion which could damage the core (0.50). REFERENCE GGN5 OP-LP-EP/SPDS-LP-OO8 Rev 01 pg 6 of 13 LO4 1. 3.2/4.3 3.8/4.5 295026 GOO 3 295026G012 ...(KA'S) l (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
= .M , z Zs__E80GEQUEES_ _UQ8dGLt GEUQBUGLt_EdEBGENGY_8MQ PAGE 44 BQQ1QLOGIGGL_G90IS96 OUESTION 7.22 (1.00) 101-3, Plant' Shutdown, cautions that shutdown cooling flow should not be permitted to decrease below 1000 gpm. linswer each of the following concerning this caution.
a) State the basis for this caution, b) State the adverse consequences of allowing shutdown cooling flow to decrease below 1000 gpm. ANSWER 7.22 (1.00) Each of the following at (0.50) each:
- 3) Results in RHR minimum flow valve opening d ', Draining of the RPV to the suppression pool.
REFERENCE GGN9: 10I-3 Rev 29 pg 17 3.6/7.6 3.2/3.3 2.7/2.8 2.8/2.9 3.6/3.6 3.3/3.2 3.4/3.4 2.9/2.9 3.2/J.1 3.2/3.2 3.1/3.9 205000K102 205000K302 205000K407 205000K502 205000K604
...(VA'S)
(***** END OF CATEGORY 07 *****)
e 19t__8DDINISIBBIlYS_EB9CEDUBESi_G9NDlIIONSi_8BD_LIU1IBIIONS PAGE 45 s OUESTION 8.01 (1.00) Which one of the following individuala DOES NOT approve SNM Movement Plans for core alterations. a) Refueling SRO
.b) Technical Superintendent c) Reactor Engineering Superintendent d) GGNS General Manager or Manager, Plant Operations ANSWER 8.01 (1.00) a)
REFERENCE GGNS: 09-S-02-300 pg 4 OP-LO-AD-LP-OO1-02 LOM 2,4 2.3/2.9 230000 GOO 1 ...(KA'S) QUESTION 8.02 (1.00) Which one of the following is NOT a purpose of the Protect $ve Tagging System according to 01-S-04-1, Protective Tagging System. a) Provi des administrative controla necessary to prevent the 'elease of radioactive materials to the environment. b) Establishes the division of tagging authority between the Operations and Maintenance Sections. c) Provides detailed instructions for the use of Red Equipment Clearances and Information Tags. d) Provides administrative controls nece,sary to control equipment used in order to prevent personnel injury or equipment damage.
ac__0RdlNIE18811YE_E8QQEQU8EGu_QQNQlllQNSu_@NQ_LidlI@IlQNQ' PAGE 46 ANSWER 8.02 (1.00) a)- RtFCr.ENCE GGNS: OP-AD-501 LO4 1 AP-01-S-06-2 2.7/3.7 3.6/4.2 3.3/4.3 3.5/3.8 294001A103 294001A110 294001A111 294001K116 ...(KA'S) QUESTION 8.03 (1.00) The unit is operating at 75% RTP when the Shift Supervisor is given a MNCR stating that two (2) SRVs may not operate in the Relief, Safety, or ADS modes. Referring to the attached Technical Specifications, which one of the following BEST describes the allowances / limitations imposed by Technical Specifications f or the above situation. a) Be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours. b) De in at least HOT SHUTDOWN within 12 hours and reduce reactor pressure to less than or equal to 135 psig within the next 24 hours. c) Be in at least STARTUP within the next 6 hours and in at least HOT SHUTDOWN within the following 6 hours. d) Be in at least STARTUP within next 6 hours and in at least HOT SHUTDOWN within the following 6 hours and in COLD SHUTDOWN within the next 24 hours. ANSWER O.03 (1.00) a) REFERENCE GGNS TS 3.5.1 OP-PB-601 Rev 1 LOM 4 3.5/4.4 3.6/4.3 203OOOGOOS 218000 GOO 5 ...(KA'S) (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) W
'.Qu__GDd1hlSIBOI1YE_EBQGEQUBEGu_G9N01IlRNet_GNQ_(ldlIGIlQNS PAGE 47 OUESTION 8.04 (1.00)
Unit 1 is operating at 60% RTP with LPCS inoperative for the past 24 hours and no estimated repair time. DG 12 han failed to start twice during performance of scheduled surveillance and has been declared inoperabic. The cause of DG 12 failure to start has been determined tn be a seized shaft. Referring to the attached Technical Specifications, which one of the followino BEST describes the allowances / limitations imposed by Technical Specifications for the above situation. a) Se in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours, b) Power Operations may .ontinue for 12 hours then be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours, c) Power Operations may continue for 72 hours then be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. d) Power Operations may continue for 6 days then be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. ANSWER 8.04 (1.00) a) REFERENCE GGNS: OP-PB-601 LO#4 TS 3.8.1.1 2.9/3.9 3.4/4.1 262OO1 GOO 5 264000 GOO 5 ..(KA'S) (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
f-St__0DdLNLSIBGIIVE_EBQGEQUBESt_GONQlI10NSt_8NQ_LldlIGIlONS PAGE 48 QUESTION 8.05 (1.00) The reactor is defueled with tNo (2) control rods withdrawn from the core and fuel loading is scheduled to begin. Referring to the attached Technical Specifications, which one o? the following BEST describes tha Technical Specification allowances / limitations for the above condition, a) Fuel loading may NOT begin until all control rods are inserted.
- 6) Fuel loading may begin and continue as long as Shutdown Margin requirements are met.
c) Fuel loading may begin; however, the fuel assemblies surrounding the removed control rods may not be loaded. d) Fuel loading may begin AFTER one of the control rods is inserted and the fuel assemblies surrounding the remaining removed control rod may not be loaded. ANSWER 8.05 (1.00) a) REFERENCE GGNS: TG 3.9.10.2 OP-PB-601 Rev 1 LO4 4 3.0/4.1 234000 GOO 5 ...(KA'S) DUESTION 8.06 (1.50) State whether each of the following concerning on-duty operations shift composition is TRUE or FALSE. a) A SRO is required in the control room in Operational Conditions 1, 2, 3, and 4. b) Two (2) licensed operators are required in the control room during a reactor start-up. c) The shift composition, including the shift superintendent, may be one less than the minimum required for a maximum of two (2) hours.)
Ot__BDUINISIBAII$g_BBQGEDUBgS2_GQUp1IIQUS1_OND_LIUlIGIIONS PAGE 49 ANSWER 8.06 (1.50) (0.50) each: a) FALSE b) TRUE c) FALSE REFERENCE GGNS: 01-S-06-4 OP-LO-AD-LP-OO1-02 LO4 4 2.7/3.7 294001A104 ...(KA'S) OUESTION 8.07 (1.00) Answer TRUE or FALSE to each of the following concerning 01-S-06-12, Surveillance Program Procedure. a) A system that is made inoperable by a Technical Opecification surveillance procedure is NOT subject to the applicable Technical Specification ACTION statement during the performance of the surveillance precedure. b) An "Unacceptable" surveillance determination may be changed to "Acce table" f ollowing a satisf act ory retest. ANSWER 8.07 (1.00) a) FALSE b) FALSE REFERENCE GGNS: 01-S-06-12 Rev 14 pg 31 OP-LO-AD-LP-OO1-02 LO4 4 2.9/3.4 3.7/3.7 294001A101 294001K101 ...(KA'S) (***** CATEGORY 00 CONTINUED ON NEXT PAGE * * * * *)
T
-at_ BDd1NIEIBBIIME_EBQQEQUBESt_QQUQlI1QNSt_88Q 11dlIGIlgNQ PAGE 50
- 1. o o QUESTION 8.08 (+-Se)
Answer TRUE or FALSE to each of the following concerning 01-S-06-5, Incident Reports / Reportable Events. a) An IR is NOT ruquired when performing SRV actuations during tents. the plant is voluntarily shut down to investioat poten tu . on .qt 'p nt fif u r- , R is NOT required if the inve , .
,< ee -< t omponent failure
- was a tribu =_- - shut down c) Voluntary entry into an action statement for reasons such as preventivo maintenance does NOT require an IR to be initiated.
- l. 0 0
' ANSWER 8.08 (4 r4M)
(0.30) each: a) FALSE
-u; iF uu D 6 L&Tti-D c) TRUE REFERENCE GGNS 01-S-06-5 Rev 16 pg 8,11 OP-LO-AD-LP-001-02 LO4 2,4 3.4/4.5 239002G003 ...(KA'S)
QUESTION B.09 (1.00) Explain why the Reactor Coolant System Pressure Safety Limit, as measured in the steam dome, in 1325 psig.
Qi__8DdblSIBOIIME_EGQQEQUBEQu_QQUQlIlONS_QUQ_LINIIGIlgNS t PAGE 51 ANSWER 8.09 (1.00) (Steam dome pressure of 1325 psig is) equivalent to the pressure limit (1375 psig) at the lowest elevation in the reactor coolant system. REFERENCE GGNS: Technical Specification Bases 2.1.3 OP-PB-601 Rev 1 LO4 7 3.3/4.1 2.8/3.8 2900002 GOO 290002 GOO 6 ...(KA*S) QUESTION 8.10 (1.50) State whether each of the following statements concerning 02-S-01-17, Control Of Limiting Conditions for Operation, in TRUE or FALSE. al The Shift Supervisor will fill out an LCO Report any time the plant enters the action statement of a Technical Specification except for equipment made inoperable due to MWO/MWP maintenance activities, b) Tracking LCOs may be written at any time to track conditions the must be corrected before a system can be decle. d operable, c) When equipment is placed into a necessary condition to meet the requirements of an LCO, place a Red Tag on the control switch for that piece of equipment. ANSWER 8.10 (1.50) a) FALSE b) TRUE c) FALSE REFERENCE GGNS O2-S-01-17 Rev 7 pg 2,4,6 OP-LO-AD-LP-OO1-02 LOM 4 4.2/4.2
~294001A102 ...(KA'S)
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) t
'Dz__0DdiUIDIB6IIME_E8gCgggggs,_CgypIIlggs,_00p_ lid 1IGIlgus PAGE 52 QUESTION 8.11 (1.50)
Procedure 01-S-02-1, Description and Use of the GGNS Operations Manual, requires that all directives shall be strictly adhered to cxcept under three (3) specific circumstances. State these three (3) specific circumstances.
' ANSWER 8.11 (1.50)
(0.50) each:
' '. ) Imminent danger to plant workers !!) Imminent danger to important plant equipment
- 3) Imminent danger to the general public REFERENCE
",GNS 01-S-02-1 Rev 14 pg 3 OP-LO-AD-LP-OO1-02 LO4 2,4 4.2/4.2 294001A102 ...(KA'S)
DUESTION 8.12 (1.00) While operating at 100% RTP, HPCS suction automatically transferred to the Suppression Pool. After verifying that suppression pool level and CST level are within allowable limits, the HPCS suction was manually overridden to the CST. Referring to the attached Technical Specifications, state the required action for the above condition. ANSWER B.12 (1.00) 3.5.1.c.1 (Restore the inoperable di vision to OPERABLE within 14 days) REFERENCE GGNS TS 3.5.1 TSPS 078 OP-PB-601 Rev 1 LO4 4 3.2/4.0 206000 GOO 5 ...(KA'S) (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
Sz__0Dd1NISIBBIIME_EBQCEQQBgD1_CQNQlIIQUS1_8NQ_LIMIISIJQUE PAGE 53 QUESTION 8.13 (1.50) In accordance with 01-S-06-2, Conduct of Operations, state the three (3) actions that the Shift Superintendent shall take if it in determined that an emergency condition exists. ANSWER G.13 (1.50) (0.50) each:
- 1) Assume the Emergency Director position
- 2) Implement the GGNS E-Plan
- 3) Notify the On-Call Duty Manager REFEPENCE GGNS O2-S-06-2 Rev 20 pg 21 6.3.2 OP- LO-AD LP-OO1-02 LOM 2,4 2.9/4.7 294001A116 ...(KA'S)
QUESTION 8.14 (2.00) Answer each of the following concerning 01-S-06-29, Independent Vcrification Program, a) State three (3) situations where independent verification does NOT have to be performed. b) State how an independent verification should be perform. l
Sz__GDdlNISIB8IlyE_EBgggpUBES2_GQUDlIIONSt_8ND_LIdlIBIJQNS PAGE 54 ANSWER 8.14 (2.00) a) (0.50) each:
- 1) If the component is located in a high or sery high radiation area.
- 2) Components that are inaccessible.
- 3) (Throttle and instrument) valves that have seals attached.
b) In series or after performance of the activity (not in parallel) (0.50) REFERENCE GGNS: 01-S-06-29 Rev 1 pg 3,6 OP-LO-AD-LP-OO1-02 LO4 2,4 3.3/4.3 3.5/4.2 294001A111 294001A112 ...(KA'S) OUESTION 8.15 ( .50) Technical Specification 3.9.1 requires that the Mode Switch be locked in Shutdown or Refuel prior to performance of refueling operations. State what is meant by the term ' locked". ANSWER 8.15 ( .50) Mode Switch Key removed REFERENCE GGNS: 02-S-01-9 Rev 13 pg 3 OP-LO-AD-LP-OO1-02 LO4 2,4 3.4/3.8 3.0/4.1 234000 GOO 1 234000 GOO 5 ...(KA'S) (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
Ot__0DdlulSIBGIIME_PBQCEQUEESt_GONQlIlQUSu_QNQ_(ldlIGIlQUQ PAGE 55 QUESTION B.16 (1.00) State the meaning for each of the f ollowing colors as used by the SPDS. a) Red b) Green c) Yellow d) White ANSWER 8.16 (1.00) (0.25) each a) Alarm or action level b) Normal, unchallenged EP status c) Suspect data quality d) Static information (titles, labels, scales, etc.) REFERENCE GGNS: LPTS-03/3 LO# 12,13 3.2/3.4 2.9/4.7 294001A115 294001A116 ...(KA'S) t . SW QUESTION 8.17 (W) Referring to Attachment B-1, state the emergency classification for each of the fellowing events. a) Following a normal power increase, routine reactor coolant sampling indicates dose equivalent I-131 is 4.2 microCi/ml. l b) After placing the mode switch in RUN, drywell total leakage l was calculated to be 56 gpm. c; Upp -- p c;-1 t :-p ~ r ^ ' " te >inteined l e a r- th=n im C DdLCT6D d) RCIC steam supply line ruptures and fails to isolate.
St__0RULUISIBOIIME_EBQGEDUBESt_GQUDlIlQUSu_QND_LidlI@IlgNS PAGE 56 i . ro ANSWER 8.17 (W) (0.50) each: a) Unusual Event b) Alert
'J,u , u. I m.-_-r+ d 6L6rEh d) Site Area Emergency REFERENCE GGNS: 10-S-01-1 Attachment I EPTS-6 Rev 3. LO# A.1 3.4/4.5 239002 GOO 3 ...(MA'S)
OUESTION 8.18 (1.00) State the Technical Specification bases for the restriction on recirculation loop flow rate and flow control during single recirculation loop operation. ANSWER 8.18 (1.00) Flow rate - ensure vessel internal vibration remains within limits (0.50) Flow control - reduce valve wear and ensure valve does not enter into the cavitation region (0.50) REFERENCE GGNS TS Bases 3.4.1.1 OP-PD-601 Rev 1 LOM 4 3.9/3.9 3.4/4.2 202001G001 2020010005 ...(KA'S) (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
r. Di__0Bd1NISIBGI1YE_EBQGEQUBESt_GQNQlIlQNQt_QNQ_ Lid 1IGIlQNQ PAGE 57 OUESTION 8.19 (2.00) The unit is in Operating Condition 2 with reactor pressure at 100 poig. LPCI 'C' has been declared inoperable due to a failed curveillance and investigation has begun. The RCIC steam cupply has been manually isolated to allow f or maintenance on the RCIC turbine. Di vi si on 1 of the ADS initiation logic has been bypassed due to on-going troubleshooting activities. Suppression Pool water level has been determined to be 18 feet. Referring to the attached Technical Specifications, state the applicable LCOs for the above situation. ANSWER 8.19 (2.00) TS 3.5.3.a (1.00) TS 3.5.1.b.1 (1.00) REFERENCE GGNS: TS 3.7.3,3.5.1,3.5.3 OP-PB-601 Rev 1 LOM 4 3.3/4.3 3.5/4.4 3.6/4.3 203OOOGOO5 217000 GOO 5 218000 GOO 5 ...(KA'S) QUESTION 8.20 (1.50) 101-2, Power Operations, rev ires all turbine bypass valves to be fully closed when withdrawi ., control rods with reactor power above the Low Power Setpoint State the Technical Specification basis for this requirement. ANSWER 8.20 (1.50) The reactor power input for rod control is determined from first stage (turbine) pressure (0.50). With the BPVs open, the RC&TS senses reactor power to be less than actual reactor power and the potential exists for a non-conservative rod withdrawal (1.00). REFERENCE GGNS: 03-1-01-2 TS 3.4.1.4 OP-LO-AD-LP-OO1-02 LOM 4 2.0/4.0 i I t 1
St__GRd181 SIB 8I1YE_EBQCEQUBE@u_CQNQlIlONgt_GNQ_(ldlIQIlQU@ PAGE 50 20100SGOO5 ...(KA'S) OUESTION 8.21 (1.00) Regarding Technical Specification 3.4.4, Chemistry Limits, state tho basis for each of the following, a; lhe more restrictive chloride limit when in Operational Condition 2. b) The surveillance requirement for conductivity monitoring to ensure chemistry limits are acceptable. ANSWER 8.21 (1.00) a) The oxygen concentration and temperature (in Mode 2) promote chlorido stress corrosion (so chloride must be limited to a smaller amount). b) When the conductivity is within limitn, the ph, chlorides, (and other impurition affecting conductivity) must also be within their acceptable limits. REFERENCE GGNS TS Bases 3.4.4 pg D3/4 4-3 OP-PD-601 Rev i LOM 4 2.9/3.6 2.7/2.9 256000A108 256000 GOOS ...(KA'S) DUESTION 8.22 (2.00) State the Technical Specification definition for each of the following termn. a) Fraction of Limiting Power Density b) Limiting Control Rod Pattern c) Ch an r.el Check d) Rod Density
Ot__GR51MLSIBGIlYE_EBQQEQUBESt_QQNQlIlOUgx_@UD_LIMlIGI1QUS PAGE 59 ANSWER 0.22 (2.00) (0.50) eacht a) The LHGR existing at a given Ir; cation divided by the limiting LHGR for that bundle type (LHOR/ Lim LHGR). b) A (control rod) pattern which results in the core being on a thermal hydraulic limit. c) Qualitative assessment of channel behavior dur'ing operation by observation. d) The number of control rod r.otches inserted as a fraction of the total number of control rod notches. REFERENCE GGNS: Technical Specification Definitions 1 . 1 4 , 1 . 2 0 ,, 1 . 5 , 1 . 3 7 3.3/4.3 3.5/4,2 294001A111 294001A112 ...(KA'S)
.,7 TEST CROSS REFERCNCE PAGE 1
~ QUESTION VALUE REFERENCE 05.01 1.00 RWCOOO1749 05.02 1.00 RWCOOO1750 05.~O3 1.00 RWCOOO1737 05.04 1.00 RWCOOO1746 05.05 1.00 RWCOOO1747 05.06 1.00 RWCOOO1755 05.07~ 1.00 RWCOOO1825 OS.08 2.00 RWCOOO1741 03.09 1.50 RWCOOO1744 03.10' 2.00 RWCOOO1751 05.11 2.00 RWCOOO1738 05.12 1.50 RWCOOO1758 05.13 1.00 RWCOOO1740 05.14 .75 RWCOOO1753 05.15 1.50 RWCOOO1748 05.16 1.00 RWCOOO1752 05.17 1.50 RWCOOO1739 05.18 1.00 RWCOOO1742 05.19 1.00 RWCOOO1743 05.20 1.00 RWCOOO1754 05.21 1.50 RWCOOO1756 05.22 1.50 RWCOOO1757 27.75 06.01 1.00 RWCOOO1778 06.02 1.00 RWCOOO1765 06.03 1.00 RWCOOO1777 06.04 1.50 RWCOOO1771 06.05 1.50 RWCOOO1772 06.06 1.00 RWCOOO1759 06.07 1.50 RWCOOO1760 l 06.08 . 7 5, RWCOOO1761 l 06.09 1.00 RWCOOO1763 l 06.10 1.00 RWCOOO1764 06.11 2.25 RWCOOO1769 06.12 1.50 RWCOOO1779 06.13 2.00 RWCOOO1762 I 06.14 2.00 RWCOOO1766 06.15 1.00 RWCOOO1767 06.16 2.00 RWCOOO1768 06.17 1.00 RWCOOO1770 06.18 1.00 RWCOOO1773 06.19 1.50 RWCOOO1774 06.20 1.00 RWCOOO1775 06.21 1.00 RWCOOO1776 06.22 1.00 RWCOOO178 ) 28.50 07.01 1.00 RWCOOO1795 l
-- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ J
le
. .- /~' 4 TEST CROSS REFERENCE PAGE 2-
. QUESTION VALUE REFERENCE 07.02 1.00 RWCOOO1792 07.03 1.00 RWCOOO1794 07.04 1.00 RWCOOO1797 07.05 2.00 RWCOOO1781 07.06 2.00 RWCOOO1783 07.07 1.50 RWCOOO1786 07.08 1.50 RWCOOO1787 07.09 1.00 RWCOOO1788 07.10 .50 RWCOOO1790 07.11 1.50 RWCOOO1796 07.12 2.00 RWCOOO1798-07.13 2.00 RWCOOO1799 07.14 1.50 RWCOOO1801 07.15 1.00 RWCOOO1782 07.16 1.00 RWCOOO1784 07.17 1.00 RWCOOO1785 07.18 1.00 RWCOOO1789
'07.19- 1.00 RWCOOO1791 07.20 1.00 -RWCOOO1793 07.21 1.50 RWCOOO1800 v7.22 1.00 RWCOOO1802 28.00 08.01 1.00 RWCOOO1807 08.02 1.00 RWCOOO1BO9 08.03 1.00 RWCOOO1805 08.04 1.00 RWCOOO1810-08.05 1.00 RWCOOO1814 08.06 1.50 RWCOOO1804 08.07 1.00 RWCOOO1812 08.08 1.50 RWCOOO1815 08.09 1.00 RWCOOO1822 08.10 1.50 RWCOOO1824 08.11 1.50 RWCOOO1803 08.12 1.00 RWCOOO1806 08.13 1.50 RWCOOO1813 08.14 2.00 RWCOOO1816 08.15 .50 RWCOOO1817 08.16 1.00 RWCOOO1818 08.17 2.00 RWCOOO1819 08.18 1.00 RWCOOO1820 08.19 2.00 RWCOOO1823 08.20 1.50 RWCOO01808 08.21 1.00 RWCOOO1811 08.22 2.00 RWCOOO1821 28.50
----m-112.75 DOCKET NO. 416 J
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{ , e e, , READY ... l d I l
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sr. = w p (t es ? U. S. N(JCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION ' ' ' FACILITY: Grand Gulf Unit I REACTOR TYPE: DWR-GE6 DATE ADMINSTERED: OO / 04 /,WWM 26 EXAMINER: REGION II CANDIDATE INSTRUCTIONS TO CANDIDATE: Use separate paper for the answers. Write answers on one sice only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 29 44 25.50 2L.17 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
- 25. G 8 25.75 25.37 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS SYdY 24.50 20.it 3. INSTRUMENTS AND CONTROl.S
'dME RGENCY AND RADIOLOGICAL CONTh0L
/00. 2 (
-i O I . SP- % Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.
Candicate's Signature MUERCOPY
o~ NRC RULES AND GUIDELINES FOR LICENSE EXAh!INATIONS During the administration of this examination, the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave . You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil on_Iyl to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the time you START and STOP on the cover sheet of the examination.
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of each page of the exam.
- 8. The exam has one question per page. Write the answer beneath the question (start just below *"** CATEGORY . . . ) . Write only on one side of the exam and any extra answer sheets.
- 9. Number each answer continued on additional paper as to category and number, for example,1.4, 6.3.
- 10. Attach continued answers to back of question to which it applies.
- 11. Place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used aa a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
.)
q
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
- 18. When you complete your examination, you shall:
- a. Assemble your examination as follows:
(1) Exam questions with answers on top. (2) Exam aids - figures, tables, etc. (3) Scratch paper used during the examination.
- b. Turn in your copy of the examination and all pages used to answer the examination questions,
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked, s
.I'.,, PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Page 2
' THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW f
i OUESTION 1.01 (1.00) Which.ONE of the following defines a DELAYED neutron ? 1
- a. A fast neutron that is born at lower energies than a prompt neutron.
w?
- b. A neutron born at the samo kinetir. energy l evel as the energy of
-the moderator atoms.
- c. A neutron with a higher probability of being absorbed by a boron atom than a thermal neutron.
- d. A neutron born within 10E-14 seconds of a fission event.
l (***** CATEGORY 1 CONTINUED ON HEXT PAGE *****)
I?
?
-o OUESTION 1 ~. 02 -(1.00)
LThe total amount of reactivity that must be.added to bring a reactor to a critical condition is equal to thes
.a. Excess Reactivity
-b. Reactivity Defect
- c. Shutdown' Margin
- d. Subtritical Factor i
l U 4 (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
~s i
(DUESTION- 1.03 (1.00) Which ONE of the f ollowing describes the behavior of neutron power following a prompt insertion of a large amount of negative. reactivity?
- a. Neutron power decreases linearly with time after the initial j pre.at drop.
b.- Af ter .the ini tial prompt drop, neutron power decreases on a constant negative period of 12.5 seconds unless a reactor scram occur 5.
~
- c. After the initial prompt drop, neutron power decreases on an 80-second period regardless of the size of the negative reactivity insertion.
- d. ' Neutron power drops immediately to "Beta" (delayed neutron fraction) times the neutron power prior to the prompt insertion of negative reactivity. !
\
9 L f b t b r (***** CATEGORY 1 CONTINUED CN NEXT PAGE *****) , r., r--w n-.e- w- w
t V
. QUESTION 1.04 (1.00)
Which ONE of the following identifies the purpose of using control rod sequencing and flux shaping?
- a. Shaping or power rods are sequerited to extend control rod usable life.
- b. '1ermal flux distribution is shaped by rod sequencing to compensate for flow instabilities caused by excessive two phase flow within the high power channels.
- c. Rod sequencing and flux shaping are used to compensate for reactor power anomal i es.
- d. Axial and radial thermal flux distributions are controlled to minimize power peaking and control rod worth and to optimize fuel burnup.
1 (***** CATEGORY l CONTINUED ON NEXT PAGE *****)
. m. .
QUESTION 1.05 (1.00) Which ONE of the f ollowing thermal limits protects the fuel from
! clad rupture due to PLASTIC STRAIN (deformation)?
- a. LHGR
- b. APLHCR
- c. MCPR
- d. MAPRAT i
j s (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
Jh . QUESTION 1.06. (1.00) Which ONE of the f ollowing conditions would tend to INCREASE , the Critical Power l evel assuming all other variables remain
. unchanged?
NOTE: ASSUME NORMAL FULL-POWER OPERATING CONDITIONS
- a. Inlet subcooling is DECREASED b.. Reactor pressure is DECREASED
- c. The axial power peak is RAISED
, 'd. Coolant flow rate is DECREASED k I (****r CATEGCRY 1 CONTINUED ON NEXT PAGE *****) f i 1 5
F. t - QUESTION 1.07 (2.00) The reactivity worth of ar single control rod will ____________, (For EACH statement below indicate INCREASE, REMAIN THE SAME, nr DECREASE) i
- a. -i f the void content around the rod increases.
.b. if the moderator. temperature decreases.
- c. if'an adjacent cor' vi rod is withdrawn.
- d. if Xe-135 concentration around the rod decreases.
i
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L (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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QUESTION 1.08 (0.50)
If.the power level .is increased by control rod withdrawal at the beginning core life, how does the stesdy state to steady state void fraction change? (ANSWER INCREASE, REMAIN THE SAME, OR DECREASE) L 4 J 4 l l l (***** CATEGORY 1 CONTINUED ON NEXT PAGE * * ** * ) l l
4 _ QUESTION 1.09 (1.00) Which ONE of the _f ollowing describes the effect of an INCREASE in the amount of non-condensit.le gases in a steam turbine condents tr?
- a. Condenser pressure decreases (vacuum increases)
-i
- b. Circulating water outlet temperature decreases s
- c. Steam cycle ef ficiency decreases
- d. Condensate depression increases
(.***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) L
6) x EQUESTION 1.10 (1.00) Given a full open FCV, use the simplified pump laws to CALCULATE the f ollowing: a.. Reactor Recirculation pump power at slow speed (450 rpm) if power at fast speed (1000 rpm) is 416 KW.
- b. Reactor Recirculation loop flow at slow speed (450 rpm) if-loop flow-at fast speed.(1800 rpm) is 40,000 gpm.
I L i l' (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
; QUESTION 1.11 (1.00)
The' effective delayed neutron fraction (beta-bar effective) changes over-core life because:
- a. as lambda-bar effective decreases, beta-bar effective must also decrease,
- b. as the core ages, a increased percentage of fission is from Pu-239, which has a smaller delayed neutron fraction than U-238 and U-235.
- c. av U-235 is fissioned and-effectively used up, there are fewer to.c 'issions which result in fewer delayed neutrons,
- d. The microscopic cross sections for the isotopes producing delayed neutrons progressively become smaller with core age.
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
200ESTION 1.12 (1.00)
~ Step SP/L-20 of EP-2 asks the operator if Suppression Pool temperature is above 212 degrees F and directs RPV injection to be taken from
.outside containment if the answer is yes. Which ONE of the following explains why this action is taken?
- a. If Suppression Pool temperature is at or above 212 degrees F, the water is boiling and Suppression Pool level may be falling faster than available makeup will allow.
- b. If' Suppression Pool temperature is at or above 212 degrees F, ECCS pumps taking a suction on the Suppression Pool will lose net positive suction head,
- c. If Suppression Pool temperature is at or above 212 degrees F, cooler water from outside sources is used to depressurize containment.
- d. If Suppression Pool temperature is at or above 212 degrees F, cooler water f rom outside sources is added to minimize the thermal stresses on the Suppression Pool / Containment wall.
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
QUESTION 1.13 (3.00) Determine whether bach of the following statements is TRUE or FALSE.
- a. Void Coefficient magnitude is directly proportional to core size.
- b. The Moderator Coefficient is less negative at EOL.
- c. The two isotopes that make the largest contribution to the Doppler effect are U-235 and P-240.
Ng4d>c
- d. An increase in flow through the retctor core will addin? t t i s :'
reactivity by decreasing the void fraction and thus increasing reactor power.
- e. As the burnable poison within a fuel bundle burns out, the Void Coefficient becomes more negat!ve.
- f. Late in core life, the large reduction in fuel molecules and the decrease in moderator densi+/ during a plant HEATUP can result in posi ti ve reacti vi ty addi ti on and a power increase.
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
QUESTION 1.14 (3.00) O.. List the sources of reactivity addition to the reactor during the approach to criticality if reactor coolant temperature is not changing. [1NCLUDE FOSITVE AND NEGATIVE. SOURCES] (1.5)
- b. lWhich ONE of the following statements describes reactor. power and period when criticality-is ANNOUNCED during a reactor startup? CASSUME MODERATOR. TEMPERATURE IS NOT CHANGING 3 (1.0)
- 1. Prompt jump in neutron counts with decrecsing . reactor
. period.
- 2. Increasing neutron counts and a stable positive period with no rod motion.
- 3. Single notch withdrawal causes prompt increase in counts and an infinite period.
- 4. Neutron population levels off at an equilibrium level with a constant positive period.
- c. From control room neutron monitoring instrumentation, HOW can the operator tell when the heating range has been reached?
(Rod position and recirculation flow are held constant.) (0.5)
/-
(***** CATEGORY l CONTINUED ON NEXT PAGE *****) t -
4 QUESTION 1.15 (1.00) Which ONE of the following defines the Onset of Transition Boiling?
- a. the point of. heat addition in which Lubbles formed on.a heated surface submerged in water become so dense they combine and form a vapor film over the heated surface.
- b. .the point'of heat addition in which the energy from a heated surface submerged in water forms small steam bubbles at the matarial surface which then collapse .
- c. the most efficient method of heat transfer of a heated surface to its environment.
.d. the heat transfer method in which energy from a heated surface is directly transferred to the fluid it is submerged in.
4 (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
iOOESTION 1.16 (1.00)
. Which DNE of theEfollowing defines Critical Heat-Flux?
'a. the. amount of heat flux which corresponds to departure from nucleate boi' ling.
- b. the-amount of heat flux (D/A) required to.be added to a heated surface submerged in' water in order to transfer all the energy by radiation.
,c. ~the amount et heat flux required to be added to a heated surface in order.for the transfer method to become the most
~ efficient.
- d. the heat flux generated by the core which compensates for ambient losses and maintains the reactor critical with no
.further rod movement.
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
-4 QUESTION. 1.17 (2.00)
- a. List FOUR conditions necessary for natural circulation to occur in~any system. (1.0)
'b. Which ONE of the following describes why forced core recirculation is required in BWRs? (1.0)
- 1. Allow for less feedwater' reheating and thus increase overall plant efficiency.
.2. Maximize cooling to incore instrumentation to assure accurate reactor power indication.
- 3. Maximum reactor. vessel heatup rate will'not be exceeded if the minimum required forced flow is maintained.
- 4. Forced flow is used to prevent thermal stratification of reactor trater and. fuel rod overheating due to inadequate cooling.
l (***** CATEGORY l CONTINUED ON NEXT PAGE *****) 1 <- 1
OUESTION 1.18 (2.00) Attached Figure 1 illustrates a transient that could occur at a BWR. GIVEN: (1) A fast opening of BOTH recire. FCVs at 11% per second. (2) No operator actions are taken. (3) Valve c l ocu :#Uegins at time = 0 seconds, movem ent EXPLAIN the cause(s) of the following recorder indications:
- a. The DECREASE in core inlet flow after '10 seconds (Graph D). (0.5)
- 6. The PEAK in vessel pressure in circle on the graph (Graph F). (0,5)
- c. The DECREASE in feedwater flow between '23-30 seconds (Graph D). (0.5)
- d. The reactor SCRAM at ~1.5 seconds (Graph A). (0.5) e
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
QUESTION 1.19 (1.00) Without. core orificing,'the coolant flow through a high power bundle will be-less than the flow through a low power bundle because: a.- the= channel. quality increases.
.b. the two phase flow friction multiplier decreases.
- c. the fuel rods expand due to thermal effects,
- d. the channel bypass flow - increases.
(***** END OF CATEGORY 1 *****)
02e PLANT DESIGN !NCLUDING SAFETY AND EMERGENCY Page 21
.OUESTION~ 2.01 (2.00)-
The HPCS Diesel Generator Mode Select Switch was inadventently lef t in the "Maint." position (all other control s normal ) . A valid LOCA signal is received. HPCS Bus 17AC becomes deenergized and the operators take the f ollowing mi tigating actions in the indicated order:
- 1. Placed the HPCS Pump Control Switch on 1H13-P601 to TRIP.
- 2. Places the HPCS D/G Mode Select Switch to Automatic.
EXPLAIN the responses of BOTH the HPCS D/G AND the HPCS Pump to the operator actions. Limit your discussion to diesel and pump responses
-to the electrical signals / control manipulations AND the reasons for those responses.
l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
QUESTION 2.02 (1.00) i Which ONE of the following is the only normally CLOSED valve in the RCIC steam supply flow path in the 'at power' standby lineup?
- a. Steam Supply Vtive (F045)
.b. Outboard Steam Isolation Valve (F064) l
- c. Turbine Trip Throttle Valve
- d. Turbine Governor Valve
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
OUESTION 2.03 (3.00) List SIX of the seven automatic isolation signals which will activate a Group 1 isolation signal, in addition to the manual initiation? CINCLUDE SETPOINTS FOR FULL CREDIT] I s (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
-QUESTION 2.04 (2.00)
~
' List SIX of the;eight.RHR system valves which_ receive CLOSE signals .,.
when Reactor water level decreases to +11.4". CSpecific valve numbers are not required for full credit, but if (numbers are not included, description of valve should be specific enough to identify the valve. A valve duplicated in A and B loops
-shculd onl y be li sted ONE time. ]
/
l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
-p QUESTION 2.05 (1.50)
Concerning the MSIV Leakage Con *.rol System,
- a. How will the components of the Inboard MSIV Leakage Control I system respond if the system is initiated with a MSIV still OPEN? ( 1. 0)
- 16. Explain the function. the pipe heators serve? ( 0, 5 ) ~
f
=4 k u
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
QUESTION 2.06 (2.00) Determine if EACH of the following statements conce ning the Low Pressure Core Spray (LPCS) system is TRUE or FALSE.
'a.- On a Loss of 125 VDC/24.VDC, the LPCS actuation signal on low RPV water level is delayed for 3 seconds to allow the level transmitters to regain loop power and stabilize when power is restored, b.. An Automantic Depressurization System (ADS) initiation permissive signal is generated when the LPCS pump breaker is closed.
- c. The "LPCS PUMP DISCH PRESS ABNORMAL" annunciator indicates the minimum flow valve'has failed to open on a pump start.
- d. The LPCS Injection Valve, 1E21-F005, opens automatically when a LCCA initiation signal is present and the pump discharge pressure reaches 575 psig.
4 l i I l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
d: _ ,y
' QUESTION 2.07 (2.00) t.
'c . . Explain the purpose of the 30 second time delay associated with automatic-initiation of the Drywell Purge Compressors following a LOCA event.
- b. List TWO control processes used by the Combustible Gas Control System, other than purging, to control the drywell and containment hydrogen concentration following a LOCA.
1 0 (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) 1
7 -.
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, QUESTION 2.08 (1.00)
~
Choose the'~ correct Offgas~ system component, sequence from4 the fol1owing list. a._ Offgas condenser, gas cooler, prefilters, charcoal adsorbers
- b. Catalytic recombiner, holdup line, cooler condenser, gas cooler
- c. Cooler condenser, catalytic recombiner, water separator, after; f11ter-
- d. Preheater, dessicant dryer, holdup line, cooler condenser o
i 3, , i l i J i (***** CATEGORY 2 CONTINUCD ON NEXT PAGE *****) l
J Id.! : .2. QUESTION 2.09 '(1.00)~ What prevents ~ f1 coding of the drywellif the suppression pool makeup system is inadvertently initiated when the reactor vessel.is
- pressurized? ,
1
,N'. .T .
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
m _ . - . h.
' QUESTION ~ 2.10 (2.50)
- Answer the'following questions cancerning the Reactor Feed Pump
. Turbine lube oil system. .
. a.- Explain 1theidesign feature'of the RFPT lube oil system.that allows testing the automatic start feature of the standby
-lube,pil pump. (1.0)
~
b'.- What condition (s) wi11 automatically start the emergency
' lube oil pump? CINCLUDE SETPOINTS3 (0.5)
- c. . Explain why.the RFP turbine will trip if 480 VAC is l' o s t to BOTH of the AC oil pumpa and the emergency oil pump functions ~
as designed. (1.0) e a f l
.(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
y , 9 QUESTION- 2.11 (2.50) i' Answer ; the f ol lowing questions concerning the Standby Liquid Control
- (SLC) . syst em.
- a. Explain HOW the injection (squib) valves provide for zero
,. leakage from the' system into the reactor and HOW the valves
, t operate to allow flow when an actuation signal is generated.
- b. List THREE indications available to the operator to verify that SLC is INJECTING into the vessel f ollowing system ini tiation?
c / 6 (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
D QUESTION 2.12 (1.00) The LPCS system piping inside the reactor vessel is monitored for integrity by the measurement of the pressure differential between the
- LPCS-injection piping inside the drywell and :
- a. the HPCS injection piping inside the drywell.
; .y .
- b. the above core plate pressure tap inside the dr ywell .
- c. the A RHR system injection piping inside the drywell.
- d. the SLC injection line below core plate pressure tap inside the drywell.
l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
.OUESTION' 2.13 (2.25) LIST and EXPLAIN HOW the Control Rod Hydraulic system design features, components, and/or interlocks provide the following functions;
- a. Constant control rod speed,* system flow during normal rod movement. (0.5)
- b. Prevent pump runout while a scram signal is present. (1.0)
- c. Prevent an excessive pressure difference across the drive piston during normal rod movement follcWing a scram. (0,5)
(t**4* CATEGORY 2 CONTINUED ON NEXT PAGE *****)
rGUESTION 2.14 (1.50) List SIX (6) actions which occur within the' Control Room Ventilation System when an. isolation signal i s generated. 4 (*wt** CATEGORY 2 CONTINUED ON NEXT PAGE *****) l 1
QUESTION 2.15 (0.50)
.With regard to the diesel fire pumps:
- a. STATE the. supply header pressure at which the "A" and "B" diesel fire pumps will AUTOMATICALLY start. (0.25)
- 6. STATE the number of start attempts'which EACH diesel will make. (0.25)
(***** END OF CATEGORY 2 *****)
-J
- 3. - INSTRUMENTS AND' CONTROLS Page 36 f
-QUESTION. 3.01 I (1.d,A The Off-Gas System i5 equipped with a Post-Treatment Radiatici Monitoring System. Liat THREE signals which will result in e channel trip. (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
At 1r
%;s ,
QUESTION 3.02 '(2.00)
- a. List the control rod notch withdrawal restrictions imposed by the Rod Withdrawal Limiter (RWL) in each of the following plant conditions: CASSUME SINGLE NOTCH WITHDRAWAL MODE]
- 1. _ Reactor' power.between the Low Power Set Point and the High Power Set Point (0.25)
.2. Reactor power above the High Power Set Point (0.25)
~
- b. Explain why the Rod Pattern Controller notch restraints are not required or enforced above 20% power. (1.0) c.- If the operator receives a control rod block which can not be readily. identified by the alarms or indications at the P680 panel, where can he find a status display of ALL Rod Blocks which will help identify the source of the block? (0.5)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
i 4
. QUEST IOf; - 3.03 -(1.00)
"Answer the following questions concerning the Safety Parameter Di spl ay System-(SPDS);
- a. Explain the difference between a white plot (graph) and a yellow plot on the SPDS screen. (0.5) b.- TRUE or FALSE :
If the following statement is displayed at the bottom of the SPDS screen, the information displayed is not being updated. (0.5) DKP: STALLED CPUA: 6UNNING (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
l.
' OUESTION.~ 3.04 - ( 1. 50)
- a. List tha automatic initiation signals and setpoints for the ARI/RPT system. (1.0)
- b. Explain why a 32 second time delay is include in the ARI/RPT reset circuitry. 'O.5)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
L: QOESTION 3.05 -(1.50)
. Answer the;following questions concerning the APRMs:
- . ' . 'How many Local Power Range Monitoring (LPRM) detectors must be available to an APRM channel.for the APRM to remain operable? (0.5)
Ib. Describe how the APRM circuitry would respond (alarms and/or trips) if-the-flow reference signal is lost because of component failure with the reactor at rated power. (1,0) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
GUESTION 3.06 (1.00) List FOUR conditions which allow the retraction of the SRM detector without actuating a rod withdrawal block. CINCLUDE ANY APPLICABLE SETPOINTS3 (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
-QUESTION' 3.07: -(1.00)
List. TWO locations available- (indications) to the operator to determine the indicated power output of an INDIVIDUAL Local Power Range Monitor (LPRM) and DESCRIDE any manipulations or operations , necessary to obtain that output. (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) +. -
' QUESTION 3.08 ( 1. '50 )
EXPLAIN what condition will generate EACH of the following indications on the Operator Control Module (P680):
-a. Data Fault
- b. Scram Valves
- c. Insert Required f
4 (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
^1 OUESTION- 3.09 (0.75)
L List THREE interlocks which must be satisfied in order to start a circulating water pump from the Control Room. CASSUME ELECTRICAL POWER IS AVAILABLE AND ALL MOTOR PROTECTION RELAYS ARE RESET) i 1 L i (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
- --- ~,---, , , - - - , . mm, - - _ _ . . _ _ -.. - - . . . . . - - - , _ - . - - - . -
1
-QUESTION- 3.10 (2.50)
Answer the following with regard to the Recirculation Pump Speed Control
- Se qu er.c e s t
- a. While MANUALLY transferring from fast to slow speed breaker SA o p e n s', but SB remains closed. Explain the response of EACH pump -
<< to this occurrence. (0.5)
- b. . STATE the-TWO methods by which the "Incomplete Sequence Relay" can be reset after its operation. (0.5)
-c' ' List four signals ~which will automatically transfer the Reclec Pumps from fast to slow speed. CINCLUDE SETPOINTS] (1.5)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
r :p '. QUESTION 3.11 (1.50) List THREE trip / actuation signals generated as condenser vacuum DECREASES _from 27" Hg to 10 " Hg. CINCLUDE SETPOINTS3 (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
- _a
~- a QUESTION 3.12 (1.50) Answer the following concerning the Automatic Depressurization System (ADS) Safety Relief Valves (SRVs).
- a. List the modes of operation which are still functional if DOTH Control Room handswitches for an ADS SRV are in the OFF position,
- b. List the modes of operation which are still functional if BOTH Shutdown panel handswitches for an ADS SRV are in the OFF position.
- c. Explain WHAT the indicating lights above the handswitches on the apron section of H13-P601 mean when they are illuminated and HOW the signal is' generated.
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
I OUESTION 3.13 (1.50)
~ List TWO control systems / components (outside of the feedwater system) which receive inputs from.the.Feedwater Control System (C34) Reactor I Water' Level detectors. For EACH signal identify BOTH the_ INPUTS and what
' FUNCTION that input is used for (IDENTIFY ANY INTERLOCKS, CALCULATIONS, OR TRIPS).
I (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
i v. QUES'T ION 3.14 (1.50) List-THHEE controls available to the control room operator which would stop Rec,irculation Flow Control Valve movement: if the valve starts to
. ramp open while operating in Loop Manual Control.
a l
's l
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
C'
-QUESTION -3.15 (2.75)
- a. List the Standby Gas Treatment system automatic (no operator action) initiation signals and setpoints. CDO NOT INCLUDE INOP CONDITIONSh (2.0)
- b. List the conditions which will automatically restart a STANDDY ,
SBGT train. (0.75) l J i '* h :
;s 1
l I l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
. QUESTION 3.16 (1.50)- List the automatic (no operator action required) Control Room Vantilation isolation signals and setpoints. (***** END OF CATEGORY 3 *****)
e- <;
'4.c IPRO'CEDURES - NORMAL,- ABNORMAL, EMERGENCY Page 52
~ ----~~------~~~-
~~~~3bb~Rbb5bLb55 bbl CUUTR5t s,4 QUESTION 4.01 (1.50)
The. Power -Opsrations procedure, 101 03-1-01-2, requires all turbine bypass valves to be fsily closed when withdrawing control rods with
-reactor power above thu Low Power Setpoi.it. What is the basis for
.this; requirement?
I l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) l l
f
, t 4
il 3
-QUESTION '
4.02 (1.50) With the exception of breaker position, what THREE items should
- , an operator check on a RACKED IN breaker, if applicable-during the performance of a system lineup checksheet per Control and Use of Operations Section. Directives, 02-S-01-27 Consider Local checks only, and a 4.16 KV I.T.E. Circuit Breaker _as an example, e
i t k (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) ,
. _ _ - - . . , - . . =
w.
,b QUESTION 4.03 (2.00)
The Conduct of Operations procedure, 01-S-06-2, has specified four conditions any one of which meet the requirements of "Adequate-Core Cooling". -State these FOUR conditions.
;4 4
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) 4
l. 1 OUESTION 4.04 (1.50) Determine if the following statements _are TRUE or FALSE.
- a. The Control Room. operator is recuired to log all centrol rods
'that remain bypassed at the end of his dhift.
b.. Chart recorders'are required to be checked within one hour of shift change and marked with date,. time, and operator i ni ti al s.
- c. Only the Control Room Operator or the Assistant Control. Room Operator may make entries in the Control Room Operators Logbook, i
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) s
CE
, ( 1 I
l QUESTION 4.05 (1.25) l E' ' ' While executing the RC/0 steps of EP 2, the operator is directed to j trip the recirculation pumps if flow reduction and rod insertion has J not. reduced power to l ess than 4%. l I
- a. Explain how tripping the recirculation pumps while power is )j above 4% will cause a power reduction. (0.25)
, 1
- b. Explain Why the procedure does not direct the recirculation pumps to be trippped if the reactor is not shutdown but power is below 4%, (1.0) 1 1
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) e
}
(*<
-QUESTION 4.06 (3.00)
' List the entry conditions (INCLUDING SETPOINTS) for EP 3, Containment Control.
Q", I i F P l i k P (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
?
b
u OUESTION 4.07' (1.00) Explain the basis for the Suppression Pool Load Limit Curve (Attached Figure 2).
- l t
1 i l l l l (***** CATEGOFtY 4 CONTINUED ON NEXT PAGE *****) i l L .._ _ _ _. _ _
o 5
- 7 QUESTION 4.08 (2.00)
Answer EACH of 'the _~f ollowing questions concerning radiological practicas at Grano Gulf TRUE or FALSE;
.a.- Partial PCs may be approved by Health Physics i nstead of the minimum full pcs for such activities as surveys or
.nspections.
- b. In order to enter a high radiation area, an operator munt carry an alarming dosimeter and a dose rate meter.
'c. The ALARA program has been i mplemented to provide more ef ficient scheduling of Health Physics personnel for better plant support.
- d. Health Physics surveillance may be used i n place of an RWP whenever approved by the HP Supervisor.
P 6 Y (***** CATEGORY 4 CONTINUED ON NEXT PAGE * * * * *)
-MY-T- -ct3- - - - -r' ) e y y w- w-ve-- ?7-T'- Y &** =-- a --'e WT'T #2hR
.s
~
. QUESTION 4.09, (1.00)
. List TWO of the three criteria used to determine if a system is in a~"standby" condition on the Pre-Sta tup Checksheet per 03-1-01-1, Cold Shutdown . to Generator Carrying Minimum Load.
l 1 (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) L.... . _ _ _ _ _ _ _ _ _ _ . _ _ _ _
QUESTION 4.10 (2.25)
- a. List THREE automatic actions expected to occur during a partial or complete loss of component cooling water per 05-1-02-V-1. (0.75)
'b. In addition to verifying the automatic actions occur, whet are the IMMEDIATE operator actions for'a complete lonn of Component Cooling Water oer 05-1-02-V-17 (1.5) f I
i 1 (***** CATEGORY 4 CONTINUED CN NEXT PAGE *****)
L' b QUESTION 4.11 (1.50) i r Describe ti;a ' reason f or ' each of the f ollowing precautions concerning the-Standby Diesel Generator System from SOI 04-1-01-P75-11
- a. Do not operate the diesel generator without air pressaire.
- b. Do not parallel two diezel generators to the same power supply,
- c. Before maintenance is perfccmed on a diesel-that has been shut down because of high crankcase pressure, ensure that the diesel l has been thoroughly cooled to ambient conditions f or at least 15 minutes.
5-e i 1 l c k (***** CATEGDRY 4 CONTINUED ON NEXT PAGE *****)
.'OURSTION 4.12 12. 50)(/. 78)
Answer the following with regard to Integrated Operating Instruction 03-1-01-3, Plant Shutdowns
- a. With the Rwcirculation Pumps secured in cold shutdown, reactor coolant temperature stratification will probably occur. What are (HREE indications monitored to ensure reactor cool ant temperature is less than 200 degrees F? (0.75)
- b. Expl ain why the operator is cautioned not to allow shutdown cooling flow to decrease below 1000 gpm when Shutdown Cooling is in operation. (1.0)
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
2,$d QUESTION 4.13 (N
- a. List FOUR symptoms of a loss of feedwater heating. (1.0)
- b. List the IMMEDIATE operator actions for a reduction or loss of feedwater heating which does not result in a scram, per ONEP 05-1-02-V-5. (1.0) l
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
a sk 5: a
^
.. !QUESTION 4.14' (1.00) h L.When is Mode 5 entered when rnovi ng from Cold Shutdown to refueling for
-core alterations according to IOI 03-1-01-5, Refueling?
L (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
-j;-
b _ [ I- , , . - - .- .-. ,. , - , .
4 , QUESTION 4.15 (1.25)
- a. List the 10CFR20 QUARTERLY legal exposure limits for items 1, 2 AND 3 below for a radiation worker with no Occupational Radiation Exposure Hi story (NRC Form 4 equivalent) on file. (0.75)
- 1. Whole body
- 2. Extremeties
- 3. Skin
- b. List the maximum 10CFR23 OUARTERLY legal exposure limits for a radiation worker if the Form 4 is on fil e and verified. (0.5) l
(***** END OF CATEGORY 4 *****) (********** END OF EXAMINATION **********) [ l
OP-LO-DT-LP-011 l i 1 256.- - I i 2co. 3 I
; m.
3 //, NEUTRON FLUX y 100. t 10.
- 0. . <.
~
100.. 58 - f. CORE INLET FLOW r;
\w '
- 0. _,.
50. f
/ C. LEVEL (INCH-REF-SEP-5KIRT) 50.
I 150.
/
I
- 50. , L. FEEDWATEK FLOW s
-50. t so.
- . -- 6. VE55E 5TEAMFLOW *
-50. t 0
E '
;, [.
VESSEL PRES RISE (?;I)
-a. ..
O. 10. 20. 30. 40. I T!Pt (5tc) l i I l FIGURE /. FAST OPENING OF BOTH RECIRCULATION FCVs AT 1.1", PER SECONO i
1 l l l CURVE SP-L-2 l SUPPRESSION POOL LOAD LIMIT l 28
' 26 /'/'/l//l/
fi '
' l /l /l/l l '
/ / / '
$ g5'5 U 2A
/ /> ', ', 's 's 's', f fl o
l 0 22 l ! z
- 9 20 m
m W , l $. . .. a. a_ o ooo o oo ooooooo ooo oooo o o o o a - NM V O W N CO CD Q=N w . RPV PRESSURE (PSIG) l . FIGURE 2 - CURVE SP-L-2 SUPPRESSION POOL LOAD LIMIT
EQUATION SHEET
. ~.
f = ma v = s/t v = mg s = v,e + 1:at 2 Cycle efficiency = Net Work (out)- E = mC
- a = (v, - y )/t Kg = imv 2 ~E g v g=v +a A = AN A = A,e PE = agh w - e/t A = in 2/c3g = 0.693/t q W = vaP-e (eff) = (t,,)(ts ) .
AE = 9314:2 - (Gq+e)b Q = $CpaT , I=I ,4X Q " UAAT I = I e-ux , ,
-Pur = Wg $"
I=I o 10"* P=P 10 5UR(t). T7L = 1.3/u P=P o et /T HVL i= 0.693/u
- SUR = 26.06/T -
T = 1.44 DT SCR " S/(1 - K,gg) SUR = 26 /A[o g 8 CR g = S/(1 - K,gg ) eff T2
~
T = '(t*/o ) + [(i-' n)/A,gfo ] 1 eff 1
- 2 0 ~~
T,= t*/ (, ;; M = 1/(1 - K,gg) = CR g/CR g T = (3 - p)/ Aef.,o M = (1 - K,gg)0/ (1 ~ Keff)1 8 " ( eff-l)IKeff " AKeff/Keff m = g - gg p= [t*/TKygg .] + [I/(1 + A,ggT )] t* = 1 x 10" seconds
-I P = I4V/(3 x 1010) A,ggA = 0.1 seconds I = No Idty=Id22 UATER PARAMETERS Id =Id2 g
1 gal. = 8.345 lbm 2 R/hr = (0.5 CE)/d (c:ecers) I gal. = 3.78 liters R/hr = 6 CE/d2 (g,,t) , 1 fc3 = 7.48 gal. MISCEI.L\NEoUS CONVERSIONS , Density = 62.4 lbm/ft 1 Curia = 3.7 x 10 dps 10 ' Density = 1 gn/cm i kg = 2.21 lbs Heat of vsg ori:stion = 970 Etu/lbm I hp = 2.54 x 10 3BTU /hr Heat of fusica = 144 Bcu/lbs 0 1 Mw = 3.41 x 10 Btu /hr 1 Atm = 14,7 psi = 29.9 in. I'g. 1 Btu = 778 f t-lbf 1 ft. H 2O = 0.433' lbf/in 1 ' inch = 2.54 cm F = 9/5 C + 32 "C = 5/9 ( F - 32) sW=- - -
-:1s - P'RINCIPdES OF NUCLEAR POWER PLANT OPERATION, . Page'67
-~~~iREER55vnERi5s- seEi isss5 FEE Es5 FEUi5 FE50 3
- ANSWER ~ ~ 1.01. (1.00)
.a 'C1.03 i ; REFERENCE GGNG, OP-NP-10A/REV 0,.LO 3.6, P. 19a (3.0)
H
'292001K102 ..(KA's)
ANSWER 1.02 ( 1 '. 00 ) i ,
- c. C1.03 ,
REFERENCE
-i
-GGNS, OP-RT-501/REV 1, LO 5.1, PP. 35a & 36a '
i(3.2)-
~ 292002K110 ..(KA'c) e ANSWER 1L03 (1.00) c C1.01 i i --REFERENCE !
GGNS, OP-RT-503/REV 1, LO S.6, PP. 45a & 46a (3.7) 292003K106 ..(KA'n) ; 1 i l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) l l
t
- 1. . PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Page 68
~
~~IUEUUUUEUUU555,~UUUY~5UUU5 FUR ~UUU~FLUIU~FL.UU
-ANSWER 1.04 (1.00) d [1.03 REFERENCE EGNS, OP-RT-505/REV 1, LO 3.2, P. 21a (2.0) 292005K110 .. WA's)
ANSWER 1.05 (1.00) a [1.03 REFERENCE GGNS, OP--HF-509 / REV 1, LO 3.3, PP. 1Sa & 16a (2.0) 293009K107 ..(KA'n) ANSWER 1.06 (1.00) b {\.f>] (***** CATEGORY 1 CONTINUED ON NEXT PAGE *tt**)
g . C l} .( 71'. PRINCIPLES OF. NUCLEAR POWER PLANT OPERATION, Page 69
- ~
T
~~~ UERR559EERIC57 SEAT iRER5FER En5 FLUiB~FL6U '\*';__..___..--__--_____..__.___.. . REFERENCE.
GONS,,OP-HF-509/REV 1, LO 5.7, PP. 26a-28a
- BFNP- TRANSITION BOILING & ATLAS TESTING. LP,P. 5-6 GEXL CORRELATION & CRITICAL POWER LP,P 3-(3.3/2.9/2,8/2.7/.2.6) 293009K126 293009K124 293009K123 293009K122 293009K117
. . '(K A 's) l ANSWER 1.07 (2.00) a.,Docreane 'CO.503
- b. Decrease CO.503
- c. Increase CO.503
- d. Increase CO.503 REFERENCE GGNS, LOP-RT-505/REV'1, LO 2.5, PP. 12a, 13a,& 16a (2.5) b 292005K109 ..(KA's)
- ANSWER 1.09 (0.50)
Increane CO.53
' REFERENCE- I GGNS, OP -RT-5F 'REV 1, LO 4.3, P. 4-23a (3.1) l 292000K119 .(KA'a)
(*****. CATEGORY 1. CONTINUED ON NEXT PAGE *****) i
l l
' ' l '. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Page 70 i t
~ ~ ~
- ~~~~35ERbUDYU35IC5[HEET TREU5FER UUU~ FLU 5D FL5W l'
ANSWER 1.09 (1.00) c C 1. 0 3 -
' REFERENCE GGNS, OP-LO-GYS-LP-N19, P. 17 OP-HF-305, LO 5.5, P. 42.b
, EIH,. Heat Transf er !< Fluid Flow, LO 5.5.2, P. 5-53 (2.7) 293007K107 ..(KA'n) 1 ANSWER l'. 1 0 (1.00) j a. (Power in proportional to the speed change cubed.) I . speed change u 450 rpm /1000 rpm =- 1/4 3 i !- (1/4) n 1/64 of fast speed power b.$ 416 KW/64 = M KW CO.53
- b. .( Flow is proportional to the speed chango.)
upeed change = 450 rpm /1800 rpm = 1/4 Flow.in 1/4 of fact speed flow. 40,000 gpm/4 m 10,0E,0 gpm CO.53 REFERENCE I I I GGNS,-OP-HF-306/REV 1, LO 10,14, P. 6--96o (2.8) : 1 , 291004K105 ..(KA'n) l l l (***** CATEGORY 1 CCNTINUED ON NEXT PAGE *****) L
..e-.-_..._._.---.._--------._-.,_....--...-_.--_._.
e?ll L T t, , v
. ' L 1. . PRINCIPLES OF' NUCLEAR POWER PLANT OPERATION, Page-71
~~
TUER5UUEUEU565[~HEdi~TE555 FEE ~5UU~ FLU 5U~FLUW
.___.p______________________.________.__
q: r- .\ _
~
ANSWER: 1.11 (l'00)
- ~
- b. C1.03 : s i
iREFERENCE' GGNS, OP-RT-503/REV 1, LO 4.6, P.3-30a .;. : (2. 3) . 292003K104 , .- (KA' n)
- o. .
~ ANSWER- 1.12 (1.00)
.b El.01
. REFERENCE GGNS, OP-LO-EP/SPDS-LP-004, 2 LO 3, P. 17 ,
(3.3/2.5) '{ 291004K*.14 291004K106 ..(KA'n) ANSWER 1.13 (3.00)
- a. False E O. 53 - _,
- b. True EO.53 ,
i.
.. False
- c. E0.53 l t
1 I l .d. .True -EO.53 l 1
- e. True to.5] [
t
. a I
l
-f, Trun E0.5]
f I f i l_ . (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
1,- PRINCIPLES OF.-NUCLEAR POWER.PLANTLOPERATION 3 Page 72
---~isEER559hERIEs~~REEi iERAsFER EA5 FEUiB FE69 e
REFERENCE' GGNS, OP-RT- 504 Rev. ' 1. , LO'2.3,3.6,4.3,& 7.3,.PP. 4-11A,.4-344, 4 TliRU '4-49 (2.5/2.9/2.5/2.1) 292004K113 292004K111 292004K105 292004K102- ..(KA's) ANGWER 1.14. (3.00)
- p. -a.. Control rod withdrawal CO.5]
Fuel temperature increase EO.53 Fiscion ' product poison concentration change CO.5]
- b. 2 ,[1.01
- c. Either of the following answere at 0.5 each; The Operator can notice that period han become longer, or that power change on IRMn/SRMs.is leveling of f (turning around duo to power overshoot).
REFERENCE
-GONS, OP-RT-507 Rev. 1, LO 1.4 & 2.3, P. 7-7A thru 7-9A
- ( 3. 8/4.1 )~
L :292008K107 292008K102 ..(KA's)
,t
- ANSWER 1.-15 (1.00) a C1.03 T a J
(***** CATEGORY 1 COr4T INUED ON NEX T' PAGE * * * * * )
. _ . _ . . _ . , _ . . _ . . . _ - _ _ _ . . _ - _ - . _ . . . - . _ ~ _ _ _ . . . . _ ..-_.___._________.m.--
#p_i n {.
' l l' . .. PRINCIPLES:OF NUCLEAR POWER PLANT OPERATION, .Page 73- i
~~'~YUER5UbE555555[~5EhY~iRU55FER E55 FLUIU~FI56W t
!REFERENCEE
. GGNS, OP-HF-508/REV~1, LO:2.7, P. 12a
.(3. 0) ,
( 29300GK109 ..(KA*n)
' ANSWER 1.16 (1.00) a E1.03 a
-REFERENCE.
OGNS, OP-HF-508/REV 1, .LO 2.7, P. 12a
'(2.9) ,
v, 293008K110 ...(KA'n) , 4ANSW5R< 1.17 (2.00)
- a. IAny 4 of the following at 0.25 each;
- 1. .Continuoun li quid flow path.
2 .- Density difference between two fluids.
- 3. Heat e,aurce.
I. p 4. Flow renintance lenu than natural circulation head. l S. Heat sink. [- L in 4 C1.03 i: REFERENCE , t i GGNS, OP-HF-500/REV 1, LO 10.1 & 10.2, PP. 53a-55a ( 2. 9/ 3'. 2) s 293000K125 293000K134 ..(KA'n) p. I b s (***** CATEGORY 1. CONTINt!ED ON NEXT PAGE * * * * * )
- o. \
l l l 2 j l '. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Page 74 l l THERt10DYNAM ICS, HEAT TRANSFER AND FLUID FLOW [ l l t l
- s. ANSWER 1.18 (2.00) <
s.Q l
- a. Recire, pump trip on low venuel level 2.
( (
'b. The spike in reactor poser,
- c. The vessel level recovery -OR-The feed pump trip on high vensel level 8.
- d. APRM high f l u:< .
[4 9 0.5 each] REFERENCE GGNS, FSAR, Chapter 15.4 OP-LO-DT-LP-011, LO 1 3 Table 2 (3.0/3.6) 295014K211 202001A205 ..(KA'n) ANSWER 1.19 (1.00) a [ 1. O 'l i REFERENCE GGNG, OP-HF-508/REV 1, LO 9.3, P. 46a l EIH, GE Heat Transfer & Fluid Flow, LO D.9.2, P. 0-46 DSEP , HEAT TRANSFER, CH. 9 Page 9-51 L.e s r, can Objective = Third from top of page 9-1A (no number assigned) (2.7) 29300D}f130 .(KA'c) t
. - . - --~,-n,..~,.,.,, - ., ,n ,. . . . --
ll . - l 1 l2( : PLANT: DESIGN. INCLUDING SAFETY'AND EMERGENCY Pago 75 .
.:________.- a__________._________________________ l
- SYSTEMS 1
ANSWER 2.01~ (2.00) The HPCS D/G Would start CO.53 due to LOCA C O . 25 ]' !and undervoltage :I uignalu^ CO.253. 'HPCS would not auto start-[0.51 ,.taking the control s witch to the. trip posi' tion (braakn the auto start signal) puts the
- pump'in manual. override.
E0.53 IREFERENCEl L G G N S ,- O P - L O - S Y S - L P - E 2 2 - 1 ,- LO 3.d - SOI.04-1-01-PO1-1, pp. 34 OP-LO-SYS-LP-P81, - LO 4. a !< 8. a, Table 2 , SD-E22/REV.2, P. 14 , (3.0/2.5/3.1)
.209002K109 209002K107 209002K104 ..(KA's)
ANSWER. 2.02 (1.00) a E1.03 , REFERENCE; L GENSr OP-LO-SYS-LP-E51, LO 4.a, FIGURE 1 l ~(3.0/3.8) t i F 217000G008 217000A212 ..(KA'n) I l:; I I r b i (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) ; l::.: .~ - . i
7,_-___.-_____._-_. O ' t . 4 2 y, . ' 2.5 iPLANT.-DESIGN INCLUDING. SAFETY AND EMERGENCY Page 7'6
-___g ___ ________________-___________________
f.<
.' } '
C 4 s
; ANSWER 2.03- ( 6 , 00 ') ~ E
. i et e cAny 6 of the fi.11owing.0 0.25 for. signal and 0.25 for setpoint; : J1. RPV Level 1- -150.3" , i
- 2. MSL Radiation High or Inop 3 x FPD
- 3. MSL Flow High 169 PSID (or 140 */. FLOW)
- 4. MSL Low Pressure 849 PSIG -i
- 5. Main Condenser Low Vacuum 9 " Hg Vac
!6. MSL Ambient Temp High 185 deg. F
-7. MSL Vent Diff-Tamp High' 101 deg. F j l
l. fREFERENCE. l'
-GbNS, OP-LC-9YS LP-M71, LO4, PP. 16 - 1(7 '
j:l { (3.8/3.1) e , 1223002K110 223002K101- ..(KA's)
- f. ~
L p; ANSWER- 2.04 - (2.00) , An'y:6 of.the following 9 U.'33 each; L4 .1. lGhutdown Cooling' Upper Pool _ Isolation Valves (1E12-F037A/D) I- 2. Shutdown' Cooling ~ Injection Throttle Valve (1E12-F053A/B) ;
- 3. . Shutdown Cooling Outboard Isol. Suction Valve (1E12-F008) ;
- 4. . Shutdown Cooling Inboard Isol. Suction Valve (1E12-F009) 5 .' RHR' Discharge to Radwante Inboard Inolation Valvo (1E12-F049) f
-6. .RHR Discharge to Radwaste Outboard Isolation Valve ( 1 E 12--F040 ) !
7.- RHR Head Spray Injection Valve (1E12-F023) 5
- 8. RHR Sample Valve (1E12-F060A/D)
- 9. HRHR Sample Valve (1E12-F075A/D)
{ I a i
-(*****
CATEGORY 2 CONTINUED ON NEXT PAGE *****) i
.7 ., .-
, ; .7 -
,a . .
'y .
=l'
-j2.N P'LANT' DESIGN' INCLUDING SAFETY AND EMERGENCY Page 77
>SYG1 EMS
^
b 1 REFERENCE $
' GGNS, OPi LO-SYS-LP-E12-02, LO 3.D'& 5,C, PP.. 36 - 37
-(3.7/3.4) c
-203000K413 203000K111 ..(KA's)
ANSWER 2.05 U 50) . MAe /ulap waw/ /rh assoe.kled usK -fb opea stry xemaint h isolaid :[o.k]
.3,.
Csaic} ^ At ;,demirut:13
.ga. ,
5o rf gee.' /ther tivehifg 3,ioes
.. a _1 -_ , , s 4 h e co n 4/ thi!il/io'd
# r - ~_ .- 4 ~ , q7 ,
seguease.(p.S q % 77; pcr _ f h u. s a. uf" ' - " '
< , -J 11 c. u - - :'. -. "
O U2. "
- b. (prior to discharge to the SBGT systom)
Evap) {,0.S orate;anycondennate REFERENCE y
~GGNG,.OP--E32-38-301/REV 1, LO 3.A & 3.B, PP. /,8, C 10, Y /M -
GG-E32/E30/REV 2, P. 13 OF 53 (3.2/2.4) 239003K40S 239005G004 ..(KA's)
-ANSWER 2.0'6 (2.00)
- a. True. E0.5]
- . . b. False CO.51
- c. False C0.53 1
^e, Falco CO.53 i-I' ry-,m a w -
-et-,-gm-3rre , e t w -. w-rc--, ,W-e.w-- ee , --r,--,e., ,we-.<= ,c..--we-,.c -~e v,+-- e.---w-ge-+r*=
j[ s -1 12 ; - PLANT' DESIGN INCLUDING: SAFETY AND EMERGENCY Pago 70' SYSTEMS- A ' l _______' ' c REFERENCE LGGNS, DCP 84/b016, Ammendment' to OP-LO-SYS-LP--E21-03
,OP-LO-SYG-LP7 E21-03,'LO 4,6,& 7,. Tables'1-& 4,.P. 11
-(3.7/2,5/3.0/3.2) 209001'K401 709001K302 '2c9001K107 209001K105 ..(KA's) '
ANSWER '2.07 (2.00) ' P
+_. .a. : Allows'LSS: actuation to complete sequencing E0.53 and completion r .
of reactor blowdown and isolations CO.5]
- b. Hydrogen recombination -LO.53 Hydrogen burnin'g- (or -ignition) CO.53
REFERENCE:
GGNS, OP-LO-SYS- E6) ; LO 2, P. 7& 11
'(3.5/3.1/7.7) <
'223001K403 22D001K40o 223001K404 ..(KA's)
ANSWER. 2.08 (1.00) i b E1.03 l
- REFERENCE j CGNS, OP-N64/65-501/REV 2, LO 3, FIGURE 1 i (3.5) 271000G009 .-(KA's)
(***** CATEGORY 2. CONTINUED ON NEXT PAGE *****)
I l
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Page 79
------- l
' ANSWER 2.09 (1.00)
The top of'the drywell' weir wall han sufficient free board height from tuppression pool HWL (to account -f or the increased suppresulan pool-level). [1. 0) REFERENCE GGNS, OP-E30-501/REV 2, LO 3.b, P. 6 Perry - MPL G43; Rev. 1, page 13 (3.1) 295029K207 ..(KA's) ANSWER 2.10 (2.50)
- a. A calenoid operated dump valve isolates the prennure switch and bleeds off the prennure (to nimulate a lons of pump discharge header prensure) C1.03
- b. Low bearing oil header pressure C0.253 below 50 psig [0.253
- c. A check valve is inntalled CO.5] which prevents the emergency oil pump from nupplying the control oil E0.53 l REFERENCE GONG, OP-NC1-501/REV 1, LO 3hib L 3h26, PP. 15 L 24 (2.7/2.5/2.9/2.8) l 259001K111 . ( K A ' s .)
l 259001K609 259001K603 259001K40! )
-(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
s pl 6 t,
- 02. -iPLANT DESIGN INCLUDING SAFETY AND EMERGENCY- Page 80 y '
4 . T 1 .. . .. , . i.' ANSWER. 2.11- (2.50) 4 E a. The- valve. inlet fitting has an end cap- (or pl.ug) T O. 5 3 ' which is sheared off to allow any' flow past.the valve CO.51, b.- 1. SLC pump discharge presnure greatur than reactor pressuru [0.5]
- 2. SLC utarage tank level decreasing CO.53 s
+
- 3. Nucinar' instrumentation indicaten reatter power is
+
-. d ec reas i n g 00.53 s
s REFERENCE
- l. -
j GGNS, OP.-LO-SYS-C41,. LG 3.h and C.e, PF. 10. ; 9, & 20
~(3.1/3.1/3.9)
N 211000G008 -211000K504 211000K401 ..(KA's) 4 4 i i , iANSWF.R' 2.12 (1.00) i ' , 1, p c
- C 1. 0 3
! REFERENCE i I' GGNS,.OP-LO--SYS-LP-E21, LO 3.'b, P. B 4' (3.0)
; 209001K404 ..(MA'G) 4 e.
u (***** CATEGORY 2 CUNTINUED ON NEX' PAGE *****)
):
p ',
.2r . PLANT. DESIGN ~ INCLUDING-EAFFhY AND EMERGENCY Page 81
___r__.__.._-.___.__._..._.____--_..,._._ SYSTEf1G f . . _ . - 1 1-' xANSWER 2.13' f2,25).
- a. Stabilizing'vhlven E0.253 maintain constant flow through the pr essure control - val ve . CO. 2*i i , thus maintaining RPV/dri ve water differential pressure constant E0.253. i
- b. 1) E0.253 in the charging water heador Restrict.ing(crifico limits flow to leon the 165 gpm) E0.253
- 2) Flow element ~for Flow Control Valves is located between the pumps and the charging water header E0.253 no a high c.harging flow closes the FCVn E0.253
- c. Equali: ing valves (0.253 reprennurize the exhaunt water header after a reactar ncram E0.25]
REFERENCE GGNG, OP- LO~SYS-L P-C11 -- 1 A , LO 3B, PP. O,9,15,&16
. CPS Student Handbook Vol 6, LP 74005, LO 1,2 LP 74005, P. 7, D, ?< 13 (2.S/2.6) 201001K402 201001K401 ..(KA'n)
L l l : t I t ( I l l l r J (***** CATECORY 2 CONTINUED ON NEXT PAGE *****)
s._....____..._______.._...__.___._.. _ _ _ _ _ _ . - c c
,1 ,
l- $ t W. . LPLANTc.UEGIGN IMCLUDING SAFETY AND EMEROENCY Paga 82-l * : _ '. a . SYSTEMD' .- .p.; g: i s; . j'pi .. j i e)., Y. < t. ANSWER: 2.14 (1.50)
- .. .Anvc6 of the following-G 0.25 cach;
- 1. Utility' exhaust fan.icolation valveu close.
c L2. Utility exhaust' fan' trips. t
- 3. Control Room purge' fan exhaust line isolation valves close.
'4 . Control Room purge. fan trips.
,- 5. Both Standby Frosh Air Units ( f i l ter trains) start.
~
i ~
- 6. Frenh air in19t valven nhut.
i L 7. Filter train recirculation valven open. b :8. 'ACU outside air makeup valves close, j- : I
' REFERENCE F .GONS, OP-251-501/REV 1, LO 4b, PP. 12-13 C3.1 )
290003K401 ..(KA's)
' ANSWER 2.15 '(.S0)
- a. '125 psig CO.253
- b. six (6) CO.253
- fREFERENCE
. GGNS,-CP-LO-SYS-LP-P64, LO 3b & 3c, PP. 12-14 & Table 4 (3.3/3.3/3.0)
. 286000Kb05 206000K407 206000K403 ..(KA'n)
(***** END OF CATEGORY 2 *****)
- 3. - INSTRUMENTS AND CONTROLS '> age 83 l
l {.
- -ANSWER. 3.01 (1.60) 1
- 1) Downstale
- 2) Upacal e Hi gh-High-Hi gh
- 3) INOP (3:D0.5each]
REFERENCE GGNG, OP-LO-SYS-LP-D17, LO 5, PP. 47 (3.1/3.1) 271000K400 271000v102 ..(KA's)
. ANSWER 3.02 (2.00)
- a. 1. 4 notchec 00.253 2, 2 notchou CO.253
-b. (As power in increaued) above 20%, increated voiding reduces the individual rod Worth 00.53 and thuc reduces the severity of (or
, limits the eenergy addi tion during) a rod drop accident [0.53. l-
- c. ' Status panel) in RACG paneln ( H 13-P65.1 and H13-P652) CO.53 REFERENCE GGNS, OP LO -3YD-LP-C11--2 -03, LO 3a,3b,&S:
GD-C11-2 REV 2, PP. 20 & 3G GOI 04 01 -C 11 --2 REV 16, P. 21 (3.3/2.0) 201005K502 201005K510 ..(KA's) (***** CATEGCRY 3 CCNTINUED ON NEXT PAGE *****)
ll~ }. P l= .- . . f.3. - -INSTRUMENTS AND CONTROLS Page 84 j (
._9 ___, _. ~_ _ _-
l J ANSWER' '3. 03 - -(1.00) av -A white-plot indicates good or-valid data C 0 .'2 5 3 while a yellow plot indicates sunpeei- data (or data which failed the vali.dation
-procenn) 'CO.253
- b. False CO.53
'= REFERENCE
' GGNS,- Emergency Procedure /SPDS Trai ni ng OP-LO-EP/SPDS-LP--002-01,
( ' (3. 2) - LO 3 & 5, PP. 6, 7,; . & 9.-
-294001A115- ..(KA'n) !
l ANSWER 3.04- (1.50) , i t
'a. Low RPV water Ievol CO.253 -41.6" (level 2) CO.2S] J High.RPV dome pressure CO.253 1095 psig CO.25 ?
l
.b. Prevents renetting'the ATW3 trip until the scram air header is !
completely bled down CO.253 and all rods have inserted CO.253. -l
, REFERENCE- .
GGNSi DCP G7/3001 REV O (Ammendment to OP-LO-SYS-LP-C11-1A-04) 7 (3.6) , I i i 2010010007. ..(KA'c) ! t l l f I t L (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) '
~ ._._ _. -
- 3. INSTRUMENTS AND CONTROLS Page 65 ANSWER 3.05 (1.30)
- a. 14 total C0.25] and at least 2 per (axial core) level E0.25]
Co. Af) ' ,
- b. APRM upsccio alarm
, .s L-o . J '
en <,-
+ '
'- -' " H ' ^"'
4 - _ _ -
-1 Akm f (0.25)
Ss bd 4t> .To>rPs.*QklIkli) kd % 99-Aott 8ls% (0.2 C.} sds . REFERENCc. 4Lekerl 4tif (flow blesed sserm .ss3 rkt] bo ll hS GGNS, OP-L O-SYS- LP-C51-4-02, LO 4a, 4b, & 9c GD-C51-4/REV 2. P. . 18 & 20 (3.6/3.3) 215005K116 215005K104 ..(KA'c) ANSWER 3.06 (1.00)
- 1. The SRM in indicating greater than 100 count s.
- 2. Both acnociated IRMu are on rango 3 or above.
- 3. The reactor mode switch is ir run.
- 4. The SRM in manually bypanned.
(4 @ 0.25 each)
. REFERENCE CCNS, C"' L O-- E W G-- L P -- C 51 - 0 3 . LO 4a, PP. 15-1" SD -C51-1/REV 2. P. 11 (2.0) 215004M404 ..(KA'c)
(**t** CATECORY 3 CONTIN 11ED ON NEXT I' AGE *****)
~ . - . . _ _ - - - . . = . _ . . _ _ . _ . . . . . -.-7.~.._-.--.--..--
i, . ; 1 & iL3; x1NSTRtJMENTS 'AND CONTROLS Page 86 _L__________________________
- m a
i e J- l - t 4 1-ANSWER 3.07 (1.00) f- Any.- 2of .the f al l owi ng; . 1.- ,P680.fu11-core-display =[0.253 -if adj acent ~ rod is selected. LO.25] : a-i
- - 2. .On the,back panel
- APRM meter. CO.253 select individual
, detector output. C0.253 3._.Procons computer CRT CO.253 have STA'cbtain readout. (select { computer data. point for.the detector). 00.251 ? ; ! i I' REFERENCE l 1: i~ i j .GGNS, OP-LO-SYS-LP-C51-3-02, LO 4, PP. 13-18 1: -SD-C51/REV 2 3 P. 7 j l' _( 3.6/2.6/3.2) .; j . j 215005K112 215005K107 215005K104 ..(KA's) j i i !~ !
- ANSWER 3.08 (1.50) . l h . i
- n. More'than one rend switch closed per RPIS channel (except full '
in/ full out) CO.51. !
, b. All scram valves are not in the name position 00.53. -j
- c. .The-selected rod munt be fully inserted !
before'any-other control rod can be moved CO.51.
- REFERENCE i f
l GGNG, OP-LC-5 VS- t P-C 1 1 -2 "33, LO 6.a and 6.b, PP. 17 3 18, f< 25 I
-(3. 31 - l 201005K107 . . ( K A ' ta i
i I i (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) ! 1 l
. . . . . . . . - . - . .- . .. - . -. .~ ..~.. .- - ..~.. . - . ... - _. . - . . . .-. .. .-. -.
- 13. INSTRUMEh!S AND, CONTROLS Page 87-
------ u--~~----------~~~---
ANSWER 3.09.; (0.75) i
\
- 1. Condenser-outlet v&lve (F011). munt be fully open. CO.25]
2.-.Adequat'e Lube water fl ow .(not low or >13 gpm) E0.25] i
, 3. Cire Water-Pump discharge valve (F002) must be f ull y shut EO.25] i 3 REFERENCE.
.GGNS, OP-N71-501/REV 2, LO 3c, P.5 (3.5) 245000G009 ..(KA's)
. ANSWER - 3.10 (2.50)
-t
- a. A' pump will not transfer to the LFMG, it will coast down to zero i
- rpm. CO.25] B pump will romain in fast speed E0.25]
I .
- b. Depress the "STOP" EO.253 or . "STOF LOCK" pushbuttons on the CD-5 -
flandswi tch. CO.253 : c.' LL3 11.4" , FW Flow <227. for 15 sec. l Steam dome temp / pump suction delta T <7.4 deg F l EOC RPT 40 psi.TSV Closure Fluid Pretsurc i 44.3 (44) psi TCV Fant Closure Fluid Pressure i l- t C1.5 = .2 parameter, .1 netpoint]
> REFERENCE ;
GGNS, OP-LO-SYS-LP-D33-1 LO 3h & 5, PP. 27-29 f
.(2. 9/3. 3) i i
202001K416 202001K226 ..(KA's) , k t t i (***** CATEGORY 3 CONTINUED ON NEXT "AGE *****)
.s f
+ s f3.-EIW3YRUMENTS AND CONTROLG Page 88 ANSWER 3.11 - ( 1. 50 ) '
l1, .
,Any 3 of the following;. [0.25 for. signal and 0.25 for setpoint]
1 '. Condenser High. Pressure elarm 23" Hg (or 23.7" or 24.1")
- 2. Turbine trip 21" Hg
- 3. Reactor Feed Pump TurLine trip 16" Hg l ,.
'4. Main Steam Dypass Valve Closure 12" Hg t-
- REFERENCE l
t - GGNS, OP-LO-SYS-LP-N19, LO 3f & 6, P. 22 and Table 3 l - (2.5/3.1) 4 \ \ !. 259001K120 2590DiK113 ..(KA'n) l l' ANSWER 3.12 (1.50)
- a. . Gafety [0.25] and ADS CO.253
- b. Safety (only)CO.53 l
- c. SRV.open or clored an indicated by tailpipe pr'esnure switch E0.53 l REFERFNCE GONS, OP-LO-SYS-LP-E22-2, LO 3e & 3f, PP. 8, 16s & 19 (3.9/4.2)
-- 218000K301 218000K105 ..(KA's)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
Page-89.
;i3. 1 INSTRUMENTS AND-CONTROLS-s
~
i t ANSWER' 3.13' . ( 1. 50) ~ , l '. Reactor _ Recirculation System CO.5] f or' FCV' runback (level
- i) 32" with < 2 RFPs running) E0.2S]
- 2. Main'.Tarbine EO.53 Jovel-for high water (53"). trip. CO.25]
REFERENCE GGNS, OP-C34-501/REV 1, LO 3b 8< 6b, PP. 9-11 (2.6/3.6/3.2/3.1) 2d?002K116 259002K114 259002K105 202002KtD9 ..(KA's . ANSWER- 3.14 (1.50) ,
- 1. Shutdown HPU.at H13-P634 using'the SHUTDOWN pushbutton. [0.5)
- 2. Shutdown HPU using manual pushbuttonatH13-P680[05]
- 3. Shu Jown HPU at H13-P6M by placing both nubloops in Maintenanco. {o,s]
\
REFERFNCE
~GGNS, CP-B33-2-501/REV 2, LO 3b, PP. 12, 26-27 SD-D33-2/REV 2, PP. 20, 27-30 (3.0) 202002G209 ..(KA's)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE 4****) y E I - - _ _ - _ . - - _ _ _ . - _ _ _ _ - _ . _ _ _ _ _ - . _ _ _ _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____m.__
- 3. ...p. :
y - - -->; g - 4' E - *,s .
- ,; - , t
- d85 IINSTRUMENTS AND CONTROLS ~ Page 9B ,
e.
;)(A'NSWER 3.15. (2.75)
.i i
- a. ,- 1. Fuel Handling Area Ventilation-Exhaust Hi Hi Rad LO.25]
i 3,6~.mr/hr EO.253 -- ; [ 2. Fuel Pool Sweep Ventilation Exhaunt Hi'Hi. Rad [0.253 - 30 mr/hr C0.25]
- 3. Low reactor water' level LO.253 -41.6" E0.253
- 4. High drywell prennure. 00.253 1.23 psig- C0.25] P
-b.~ .1. -Low-filter train' flow CO.253 -
'2. :Recircul'ation fan flow low CO.253 3.- Enclosure Bldg..prescure high- 00.253 "
REFERENCE i .. E GGNS, OP-LO-SYS-LP-T48, L O NG !< 5 A , PP. 9,13,14 , !< 16 , (3.7/2.5) I 261000K403 261000K401 ..(KA's)
- t
' ANSWER 3.16 -
(1.50) j .' ' i !' 1 1' . Hi Hi Radiation CO.253 5 mr/hr (0.253
- 2. Law. Reactor water level 00.25'l -41.6" CO 25]
i.
- 3. Hi Drywell Prosnure E0.253 1.23 puig CO.253 !
- - i I
- f. REFERENCE r
~
GGN9, 'JP-Z51-501/REV 1, LO 5, PP. 12 (3.4/2.9/3.2) i !L 29BD3K104. 29003K102 29003K101 ..(KA'c) I t (***** END OF CATEGORY 3 *****) t l r , - . _ _ , , _ , - - . _ _._ ,. _ _ . _ . _. _ _ _ _
a -- , i
~4..-PROCEDURE 9 -NOFbAL, ABNORiiAL, EMERGENCY Page_91 at;n7Aa5sE575c6515EE66aiNBE-~~---------~~--
k
-ANSWER- 4'. 0 1
-; ( 1. 50 )
The:Rx power input.for rod; control is determined from first stage pressure CO.53. -With.the bypass vals.!s open,- the RCS sensen Rx power as lean then actualfand.the potentia 17 exist for a non-conservati ve rod wi thdrawal . t i . 03 ' SREFERENCE
;3 GGNG, 03-1-01-2, P. 3 (T.S.'3.1.4.1) 13'.0/3.7)
, 241000K405 241000K101 ..(KA's) f i' lJANSWER 4.02 -(1.50)
-3'of the following checks should be made:
Dreaker charging _springn charged. CO.53 L - Charging motor switeh on. [0.53 1
- Control fason inntal;ed CO.53 I
~
Doors closed and-bolted CO.53 i REFERENCE 3-GSNG,-Procedure 02-8-01-2 ATT III/REV 17, P. 2 (3.7) 294001K101 ..(KA's) i
- - . . . , . , - . . - . - _ _ ~ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ , _ . _ _ _ _ _ _ _ _ . . . , . _ _,,,,~__,,m,,. ,
f: p 4 r (*****' CATEGORY 4 CONTINUED ON NEXT PAGEJ*****)i-p, c i. 4n PROCEDUR'ES NORMAL,-ADNCRMAL,-EMERGENCY 'Page 92 {-{4. [~-~~Es5555i6EE5iEEEE6Eis5E-~~~~~~~~~~~~~~~ i; 1 l [. ANSWER 4.03 (2. 00)- 1 t
- 1. RWL maintained above~TAF.
{ 2. CGio is being sprayed by either HPCS of LPCS
- 3. Sufficient'-ntoam flow through the core
(- 4.-Reflooding flow of' a LPCI pump in injecting'into the core with l RWL'high enough to produce two phase' flow up through the core, i l i [4 0 0. 5 each] [ I i f l' REFERENCE s l
- GGNSL 01-G-06-2, Definition 16, P. 5 i . (4.6/4.4/4.1/4.0) i 295031K304 295031K303 295031K3L*2 295031K101 ..(KA'n) '
i i
' ANSWER. 4.04 (1.50) ;
i 1 i a. True [0.53 r [ b.- Falne CO.53 j j c. Fal n e' [0.53 ' 1 i 1 !
) REFERENCE t
3 : i
- 2 GG. ;S , 02-S-01-5, P. 5, 17 & 19 (3.2) ,
i 294001A106 ..(MA's) f i r : t f i l f (stist. CATEGORY 4 CONTINUED ON NEXT PAGE *****) l
.. ..v. . . . .
. . ~. -- - . . . . . . . -. _ . . - . . . _ . . -
@ - } g. .
, p. ;;
,c
+
y s c4L LPROCEDURES -fNORM.^*,' ABNORMAL,'. EMERGENCY-Pagn 93: W___________________..___ AND' RADIOLOGICAL CONTROL t 'k
)
R JANSWER- ' 4. 05 ; (1.25)
. a.: -Rapid ~ flow reduction causes core voiding CO.253 '.;
b.-
~
(Delow 4% power:,-tripping.the pumpu yields very little power chango -E0.53 and recirculation flow wil1 aid in boron mixing-if injection is required CO.53. P L
~ REFERENCE t
f j; :GONS, OP-LO-EP/SPDS-LP-003, LO 3,.P.18
=-( 3. 8 )
L 295015K103 ..(KA's) i. ANSWER 4.06 (3. 00 ). : t i 1.- Suppronsion Pool Temperature above 95 deg. F E0.SJ
- 2. Suppression Pool Level below 18.34 Ft. (accept 10.3 - 18.5) CO.53 l
! 3. Supprennic.5 Pool Level above 10.G1 Ft. (accept 10.8 - 10.9) r 3. '5 3 o I 4. Drywell Temperature abovo 135 dog F E0.53
~
! 5. Drywell Pressure above 1.f?3 psig CO.53 l t
- 6. Containment Temperature asave 90 dog F [0.33
..-REFERENCE ; i GGNS, EP 3. OP-LO-EP/SPDS-tJ'-004, LO 2, P. 4 (4.3/4.0/4.2/4.2/4.3) , i i 295030G011 295029G01', 295028G011 2950260011 295024G011 - ! ..(KA'n) f 4 l 1 t' ! .(***tt CATEGORY 4 CONTINUED ON NEXT PAGE *****)
. . ~ . . . _ _ . . . . _ . _ _ _ . _ _ _ _ . . . _ . . _ _ _ _ _
. - . . . - . - _ . . - - . . . . . - . . - . . . _ . . ~ . . . . . . - . . . - -
- . . - = _ . . ~ ,
N L ,
4-. ,l 4 ..
c4.,1 PROCEDURES --NORMAL, Ah'ORMAi,- EMERGENCY. -Page'94_ +
~~~~En5~nE5i6t557EEE E5ATR5t' -~~~~~~-~~~
i l LANSWER' 4.'07 (1.00) i I i The ;curvo gives the maximum .Guppression Pool Lovel at which olie { or moro SRVs may bo.apened with no loads in the pool exceeding .: 2dksign li mi ts. E1.03 l i 1 i i REFERENCE. - i l
= GGNS,LO-EP/SDPS-LP-004, LO 3, P. 19 ,
j~ (3.4) a t ); 295029K101 ..(KA's) l t )' ANSWER
- -4.00 .( 2.00) l
, t
- a. True CO.53. ;
1 a
- b. Falso E0.53
} c. False CO.53 l ! d. False CO.53 , i 1 i i- -
' REFERENCE .
,' ) r I GGNS, OG-S-01-21/REV 3, ATT 1 P. 2, ATT 2 P. 1 l 28-S-01-24/REV 16, PP. 3& 16 ! j 08-S--01-80 /REV 7, P. 1 ! I (3.3)
- L a
l 294001K103 . . (KA" n) ; i j t 4 e k I I l 1 h' h 1 I i-d I t I 1' , a 1 4 a (***** CATEGORY 4 CONTINUED ON NEXT FADE *****) l
-.w-nn~.-..
- n
('
*
- r
- t. ,
f 14.<_ PROCEDURES _ NORMAL, , ABNORMAL,'EMERGENC, Page'95 l
^
.--- a;L-----..-------------------------------
' AND RADIOLOGICAL CONTROL:
P L
-ANSWER- 4.09- .( 1. 00 )
.Any?2 of the following at 0.5 each;
- 1. 'All.checksheets in the 301 are complete.
2.; The system will respond properly upon automatic o- manual initiation.
- 3. Equipment .may -tur t ' agged provided the aystem is not prevented -
from meeting its intended fu ction, i
. REFERENCE.
GGNS,.03-1-01-1,-P. 6 ,
-(3.7) 294001K101 ..~(KA's) .
1 ANSWER 4.10 (2.25) h
- a. Any 3 of the following 3 0.25 each; ;
- 1. Standby CCW pump starts e. Aaw pressure, f
,- 2. Makeup valve to surge tank opens on low level. ! t
- 3. Fuel Pool Heat Exchangeru isolate on low flow.
4.. RWCU system isolaten on Domin Inlet high temperature. I
- b. 1. Scram the reactor. [o.5]
a
- 2. Manually trip the recirc pumps within 1 minute or when pump or motor temperatures increase. [0. f] ,
- 3. If uurge tank level i n l ow and cannot be restored, secure any i running CCW pumps. [0 s]
i I (***** CATEGORY 4 CONTINUED ON NEXT PACE *****) <
. _ , . . . . . .. m._. ~ . _ . . . - . . - .. _ . ._ _ _ .. _ _ _. -. .. _ .._ . .... _ . - _ _ _
N;
.;T3
._._\
.f
'34, LPROCEDURES --NORMAL,1 ABNORMAL, UMERGENCY. Page 96- , t
.AND' RADIOLOGICAL CONTROL 3
. REFERENCE 1
'GGNS,_05 i l V-1, P. 11 (344) .
-295010G010 1
.(KA's) i ANSWER 4.11- (1.50)
- a. ' Th ts auto shutdowies -f unctions are inhibited with a losn of concrol air to the pneumatic cantrol!nr. CO.53
- b. The diesel generator voltage regulatorn will nn+ 4 _in c ti on '
properly with'two dieuels in parallel. CC. J a ! c. High tran!c cane precsure indicates the possible ignition of.an ' explosive gas mixture. CO.53. j l REFERENCC l EGNS, SOI 04-1-01-P75-1 REV 32, PP. 3& 4 v (3.5) 264000G013 ..(KA'n) - l E b .-
/> 95 j ANSWER 4.12 (9M)
- a. RPV pr ensure [0.253 [
Recire loop temperature CO.253 , Dottom Head Drain temp C0.253 ,
- 5. RHR pump minimum flow valve will open CO.53 and durnp vessel inventory to the Supprension Pool, E0.53 ,
r i i i ' '***** CATEGOF,Y 4 CONTINUED ON NEXT PAGE *****) ! t
,._. . .. _ . . _ . _ . . _ . .. . . . . . . . __m _. . _ . _ . _ , _ . .
;4. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY Page 97
- ~~~~~~~~~~~~~~~~
~~~~EUB~RE5i5L551 EEL E5UTR5t a -- -- -- --- . .--
t REFERENCE l GGNS, 101 03-1-01-3 REV 29, PP14 & 17 (2.7/3.8/3.2) i F 205000K302 2C5000K403 2
'05000K467 ., (KA's) ;
i ,
.7,00 ,
1 . ANSWER 4.13 G:. 4) i
- a. Anyfof the following at 0.25 each;
.1. Increano in reactor power.
- 2. APRM/LPRM upscale alarmc I
- 3. Feedwater heater Hi-Hi level alarmn/extractico uteam isolationc.
'4. 'Dacreaning Feedwater temperature to vessel.
- 3. Feedwtater heater string condennaio isolation /bypaus.
- 6. Decrease in core flow. .[
! b. 1. Reduce recirculation flow (or trip to LFMG) [0.25) to reduce thermal power to at least 30*/. below the power level prior to the event CO.25] or to thr minimum f1OW position EO.253.
- l '
- 2. Insert control rods an required by the Reactor Eng2neer 00.253.
l REFERENCE ; ) I CGNS, ONEP 05-1-02-V-S RtV 19, Pf *, 2-3 r (4.0/4.0) i 295014G010 295014'i107 ..(UA's) a l l i l 1 , l : t C 4 (***** CATEGORY 4 CJNTINUED ON NEXT PAGE ****J) >
7 s u; s ._ k
/
- 4. .PROCEt,URES .
. NORMAL3 'ADNORMAL, EMERGENCY Page 98L
'r
~ - - - - - ~ ~
~~.. ER5~nA515E55iEAt E6niR5t :
t }\ - . - i; ANSWER' 4 . 1.4 _ (1.00) When the first head bolt in detennioned (less than f ully -l [- terqued) . [ 1. 2 3 ' [ t REFERc.NCE j.
~
I GGNS, 101 03-1-01-5,,P. 8 , (3.3)~ ! 295023G301 ...(KA'c) *
?
i . . ' ANSWER .4.15 (1.25) !
- a. 1. 1.25 rem /qtr [0.25]
I
- 2. 18.75 rem /qtr E0.251
, 3. 7.5 rom /qtr E0.253 j ! b. 3 rem /qtr CO.253 not to exceo'J 5 ( N-'18 ) CO.25] [ !' REFERENCE ; d. i' : j GGNS, 01-G-00--2 REV 15, p.15 [ (3.3) . 294001K103 ..(KA*n) f f q-
?
i (***** END OF CATEGORY 4 *****) f' (*Ut******* END OF EXAMINATION *********t)
, i
~ c m>
. GRAND GULF UNIT 1 REACTOR OPERATOR EXAM FACILITY / EXAMINER COMMENT RESOLUTIONS EXAM OATE: APRIL 26, 1985
< Ouestion Number:- '2.05 Facility Comment: The question asks how'the inboard MSIV Leakage Control System would-respond if the system is initiated with.a MSIV still open. Clarification' was provided .during the-examination that the open MSIV was an inboard valve.
The answer key addresses system response for the system being initiated before the MSIV opens. The answer given is correct for that situation. For the situation given in the question, however, the answer should be: The leakage control line associated with the OPEN MSIV remains isolated and the remaining three-lines will initiate as per the normal initiation sequence. We recommend the answer key be changed to reflect _this answer as correct. NRC Resolution: Answer key has been modified to accept the facility answer as the correct response based on material contained on Page 14 of OP-E32/38-501/REV 1. Portions of this lesson plan contain ir. correct and misleading information and should be corrected as soon as practical. EXAMINER'S COMMENTS Question Number: 1.10.a
~
Examiner Comm3nt: Answer to part a is not correct (decimal is missing). NRC Resolution: Answer key has been changed to reflect 6.5 KW as the correct response.
. Question Number: _ 3.05.b Examiner Comment: Review of the examination and reference material revealed additional correct responses to the question.
NRC Resolution: Answer key is modified to accept the following as the correct answer. l 1 APRM upscale alarm [0.25] APRM Inop signal / trip [0.25] Signal to RCIS for Rod Block [0.25] Upscale thermal trip (flow l biased scram signal to RPS) [0.25]
e GYMTIEM BVEmiiY MSOLRCEE, WC. JC* G CESV4. J7 May 3, 1988 g U. S. Nuclear Regulatory Comission Region II 101 Marietta St., N.W., Suite 2900 Atlanta, Georgia 30323 Attention: Dr. J. Nelson Grace, Regional Administrator
Dear Dr. Grace:
SUBJECT:
Grand Gulf Nuclear Station Unit 1 Docket No. 50-416 License No. NPF-29 Coments Regarding NRC License Examinations AECM-88/0097 On April 18, 1988 representatives of Region II Staff presented proposed license examinations to System Energy Resources, Inc. (SERI) for review and coment as a part of a pilot examination review process. SERI is grateful for the opportunity to participate in the program and believes the process was advantageous for both the NRC and SERI. On April 26, 1988 examiners from the NRC gave written examinations to one (1) Reactor Operator and four (4 Senior Reactor Operators as a part of the Grand Gulf Nuclear Station (GGNS Operator Licensing Process. Upon completion of the examinations, a copy of the examinations was supplied to SERI with the request that SERI review the examinations and furnish any appropriate coments to the NRC. The attached comments are supplied pursuant to that request. Should you require additional information, please contact V'. K. E. Beatty of the GGNS Plant Staff at (601) 437-6301. Yours truly, JGC:rg Attachments J12AECM88050301 - 1a "ve* 2" um I cm
" " "" I # #2' Oh5ljh? { ^
f
AECM-88/0097
-Page 2
=.
} cc: Mr. O. D. Kingsley, Jr. (w/a) j Mr. T. H. Cloninger (w/o)
- Mr. R. B. McGehee-(w/a)
Mr. N. S. Raynolds (w/o) Mr. H. L. Thomas (w/o) Mr. R. C. Butcher (w/a) Mr. L. L. Kintner, Project Manager (w/a) Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 14B20 Washington, D.C. 20555 Mr. K. E. Brockman, Operating Licensing Section (w/a) i U. S. Nuclear Regulatory Commission Region II 101 Marietta St., N.W. , Suite 2900 Atlanta, Georgia 30323 Mr.' James McGhee '(w/a) EC&G Idaho
'1520 Sawtelle Dr.
Idaho Fails, Idaho 83415 i 4 J12AECM88050301 - 2
Attachment I to AECM-88/0097 t COMMENTS REGARDING NRC OPERATOR EXAMINATIONS I. REACTOR OPERATOR EXAMINATION Question 2.05: The question asks how the inboard MSIV Leakage Control system would respond if the system is initiated with a MSIV still open. Clarification was provided during the examination that the open MSIV was an inboard valve. The answer key addresses system response for the system being initiated before the MSIV opens. The answer given is correct for that situation. For the situation given in the question, however, the arswer should be: The leakage control line associated with the OPEN MSIV remains isolated and the remaining three lines will initiate as per the normal initiation sequence. We recommend the answer key be changed to reflect this answer as correct. II. SENIOR REACTOR OPERATOR EXAMINATION Question 6.14: The answer key for part a. reflects a 1.5 minute timer. This timer is now a 2.0 minute timer (see Attachment II). We recommend that the 1.5 minute timer be changed to 2.0 minutes. The answer key in part b. refers to a F004 valve. This reference should be F002A (reference OP-E33-38-501 Figure 6-1). We recommend F004 be changed to F002A, Question 6.15: The answer key to part a. should be revised to include a rod withdrawal block (IRM's > 108/125 ofscale)(reference: OP-LO-SYS-C51-2 lesson plan). We recommend the above be added to the answer key. J12AECM88050301 - 4
ATTACHMENT II TO AECM-88/0097 GRAND GUIJ NUCLEAR STATION ADMINISTRATIVE PROCEDURE l 01-S-02-3 Revision: 20 l s i Attachment VII Page 1 of 1 l
... l l k&l.I TEMPORARY CHANGE NOTICE COVER SHEET l DIRECTIVE # OY-/-O/ -dS/., - l TCN#,,fb DATE///AP/
i
. <- i l- OTHER CURRENT TCNs ISSUED l l l l TITI.E /dk 172rw Jte/e r?sa t/a+ vt. L dc e e Co- r 2 s) l l SAFETY RELATED [ ] NO LATEST REVISION NO. /$ l l l _l t
REASON FOR CHANGE: (Describe why the TCH is being issued, and if appropriate,
. summarize what procedural changes are involved.)
.a .
C-09 " & c 23 E 4Go/A2,jW f.m, _feypo , ,,,, .,. g M N c. K O j; f _ c9 f
&l INSTRUCTIONS FOR ENTRY: Write the TCN Numb er in the lower r tTon of
, directive's e r sheet, initial, and date . r. i
(- l '. Replace the same directive pages wit T. N
?h pa es a hdt. YO' vi CHANGES TO BE MADE: (Attach marked-up cop i 0 TER opdirectivepa 72addtMb 3rd} nm'l I pages required and list here.) arr A- 4;4e,u IW m
. i bf Y '
PSRC APPROVAL REQUIRED $( .'YES, i
'9 3 '/ < !
' L(( = .
u o e.,.n; y 7 ~d 9 'n )- Gn me . . , . . . qq Gd I l a $g REviEWANR. APPROVAL'{bthsDWM gate b[?hbs l . Prep , Sect. Supvg3upt. 'Date SRO Licen'se HoTder Date l ' 4 l f/ l REVIEW AND APPROVAL l l l [A [ /'[ M l l Techn Q/1 RYvi5 te QUalit M ogrims Date Tech Support Supt. Date l l - l l l FINAL APPROVALj l A/ l l PSRC Chatraan Date Date i 5, I I N 5-01-S-02-3 ATT VII
ATTACHMENT II TO AECM-88/0097 WiAND GULF NUCLEAR STATION SYSTEM OPERNflNG INSTRUCTION l 04-1-01-E32-1 Revision 18 l l Attachment VI Page 1 of 1 l '.b :j l (- System Main Steam Isolation Valve Leakage Control MPL No. 1E32 W ' Checksheet Name Automatic Timers Checksheet - Instruction Step 4.1.1 e SAFETY RELATE _D l TIMER I PANEL lSETPOINT l i INITIALS l l DEV. 1st 2ndi l TIMER DESCRIPTION NO. NUMBER (Min.) l l l Steamline Depressurization N600A IH13-P655 1.0 l I I l Steamline Depressurization N600E '1H13-P655 1.0 l l l N600J '1H13-P655 1.0 l l Steamline Depressurization l
', I l Steamline Depressurization N600N . 1H13-P655 1.0 l J..o , l l
i Bleed Valve Closure N601A 1H13-P655 W l l l 1 z.o l I l Bleed Valve Closure 'N601E 11H13-P655 e 1 6 I 1 z.o . I
)
IN601J 1H13-P655 4,6-- l l J Bleed Valve Closure I k l I t. . o l l N601N 1H13-P6551 4.~ l Bleed Valve Closure I I I Steamline Flow 'N602A 1H13-P655 13.0 1H13-P655 13.0 l l Steamline Flow lN602E l l l 1H13-P655 13.0 l l Steamline Flow N602J l l _l IN602N 1H13-P655 13.0 l l Steamline Flow l l l N604 1H13-P654 6.0 l l Bleed Valve Closure l l ' l Dilution Air Flow Alarm Bypass K46 11H13-P654 0.5 i I l l l l Dilution Air Flow Alarm Bypass K49A 1H13-P655 0.5 , l l l l Dilution Air Flow Alarm Bypass 'K49E l1H13-P6551 0.5 I I I l 0.5 l l Dilution Air Flow Alarm Bypass K49J 1H13-P655 l l . _l l Dilution Air Flow Alarm Bypass IK49N 11H13-P655l 0.5 l l I Note Exceptions: Date Performed by: Date
.i Reviewed by:
(Shift Supervisor) 04-1-01-E32-1 A*aT VI
~
- ENCLOSURE 3 t
s m euewnov co ;
~.. 3 c.\\ : \ 6 nesamtes.nc. '
DN G CE%7 A ore m May 3, 1988 w arLcenvo U. S. Nuclear Regulatory Comission Region II 101 Marietta St., N.W., Suite 2900 Atlanta, Georgia 30323 Attention: Dr. J. Nelson Grace, Regional Administrator
Dear Dr. Grace:
SUBJEr.T: Grand Gulf Nuclear Station Unit 1 Decket No. 50-416 License No. NPF-29 Coments Regarding NRC License Examinations AECM-88/0097 On April 18, 1988 representatives of Region II Staff presented proposed license examinations to System Energy Resources, Inc. (SERI) for review and ( coment as a part of a pilot examination review process. SERI is grateful for the opportunity to participate in the program and believes the process was advantageous for both the NRC and SERI. On April 25, 1988 examiners from the NRC gave written examinations to one (1) Reactor Operator and four (4) Seniar Reactor Operators as a part of the Grand Gulf Nuclear Station (GGNS) Operator Licensing Process. Upon completion of the examinations, a copy of the examinations was supplied to SERI with the request that SERI review the examinations and furnish any appropriate coment: to the NRC. The attached coments are supplied pursuant to that request. Should you require additional information, please contact Mr. K. E. Seatty of the CGNS Plant Staff at (601) 437-6301. Yours truly,
. . \. d JGC:rg Attachments
(- gi l G3 ft 3f f J12AECM88050301 - 1 ww.wsce,mn
5 AECM-86/0097 Page 2
! cc: Mr. O. O. Kingsley, Jr. (w/a) !- Mr.T.h.Cloninger(w/c) 2 Mr.N.R.
Mr. S. B. McGehee Reynolds (w/o (w/a)) Mr. Mr R.H. C.~L.Sutcher Thomas(w/a (w/o)) Mb L. L. Kintner, Project Manager (w/a) Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comission Mail Stop 14B20 Washington,~D.C. 20555 Mr. K. E. Brockman, Operating Licensing Section (w/a) ,s U. S. Nuclear Regulatory Comission Region II 101 Marietta St. , N.W. , Suite 2900 Atlanta, Georgia 30323 iir.JamesMcGhee(w/a) ES&G Ide.ho 1570 Sawtelle Dr. t Idai:o Falls, Idaho 63415 4 ( I 5 i 4 I a124tCs8050301 - 2 1
y _ L- . . Attachment I to AECM-88/0097 hh COM ENTS REGARDING NRC OPERATOR EXAMINATIONS I. R,EACTOR OPERATOR EXAMINATION Question 2.05: The question asks how the inboard MSIV Lei.kage Control system would respond if the system is initiated with a MSIV still open. Clarification was provided during the examination that the open MSIV was an inboard valve. The answer key addresses system response for the system being initiated before the MSIV opens. The answer given is correct for that situation. For the situation given in the question, however, the answer should be: The leakage control line associated with the OPEN MSIV remains isolated
! and the remaining three lines will initiate as per the normal initiation sequence.
We recommend the answer key be changed to reflect tnis answer as correct. II. SENIOR REACTOR OPERATOR EXAMINATION Question 6.14: The answer key for part a. reflects a 1.5 minute timer. This timer is now a 2.0 minute timer (see Attachment II). We recommend that the 1.5 minute timer be changed to 2.0 mir.utes. g The answer key in part b. refers to a F004 valve. This reference should
\ be F002A (reference OP-E33-38-501 Figure 6-1). We recomend F004 be changed to F002A, Question 6.15: The answer key to part a. should be revised to ir.clude a rodwithdrawalblock(IRM's> 108/125 of scale) (reference:
OP-LO-SYS-C51-2 lesson plan). We recomend the above be added to the answer key. s J12AECM88050301 - 4
e O 6
. ATTACHMENT II TO AECM-88/0097 GRAND GULF NUCLEAR STATION ADMINISTRATIVE PROCEDURE l 01-S-02-3 i Revision: 20 l
'[$!y.
. l Attachment VII l Page 1 of 1 l u M- l l
,- TEMPORARY CHANGE NOTICE COVER SHEET
^
l DIRECTIVE L O Y-/ - O / -d3/, - l TCH f b DATE ///AP/ l l l- OTHER CURRENT TCHs ISSUED l 1 , I l TITLE Mn .rir+>-~ JrA w mn L dc , , Co-rzs) l l SAFETY RELATED ( ) NO LATEST REVISION NO. /$ l l l t REASON FOR CHANGE: (Describe why the TCN is being issued, and if appropriate, summarize what procedural changes are involved.) a . CAS H G r E.3E & G o / A, 2, ,},j H ps.7 _fmo ,,,., ., MNC $ O ll f - c3 f 0 0 'l f
.- INSTRUCTIONS FOR ENTRY: Write the TCN Numb er in *.he lower r on of directive's e e sheet, initial, and date .
"r. i
( t' viti.t the n nige w g)gap es a ydt.tro. vI Replace the same directive pages ( l; CHANGES TO BE MADE: (Attach marked-up co o d t e rtl ddl 1 pages required and list here.) c- aOAe. a r 2b C' /
,n - wr i . Tr - Y -
' ' 3 ' ' '
~hES, I 'O IT '<"
-L((ib PSRC APPROVAL REQUIRED '[
q n n ., , ..- 6L.!gL;y;,*.='qgfg' r l , REVIEWMIL APPROVAL { , $qq l Ah Prep 4 oli/77
/,oite Y
Sect. Supvf($upt. 'Date
~
o SRO license H6Tder Date l l i l f/ l l ), ~ QUalit M ogrims g ,. /h fglREVIEWANDAPPROVALl Date
&h Tech Support Supt. Date l l TechnW1 Re' vies te 1
l l l FINAL APPROVALj l A/ l PSRC Chair'=an ' Date Date l ,
..k 01-S-02-3 ATT VII t
ATTACHMENT II TO AECM-88/0097 SYSTEM OPERATING INSTRUCTION ULAND GULF NUCII.AR STATION l 04-1-01-E32-1 Revision 18 l l Attachment VI Page 1 of 1 l
'. l
.. I 41 Jg.f) Ef.g(* System Main Steam Isolation Valve Leakage Control MPL No. II32
'!f L Checksheet Name Autoastic Timers Checksheet Instruction Step 4.1.1 e SAEITY RELATED l l TIMER l PANEL lSETPOINT l l INITIALS l TIMER DESCRIPTION NO. '
NUMBER (Min.) DE '.' . 1st 2ndl l l l I . l Steamline Depressurization 'N600A . 1H13-P655 1.0 l l l i l 1 l Steamline Depressurization N600E 1H13-P655 1.0 l ! l l l l Steamline Depressurization .N600J 1H13-P655 1.0 l
; I
. I ,
I Steamline Depressurization N600N 1H13-P655 1.0 l 40 l l l Bleed Valv_e Closure N601A 1H13-P655 4--G-- ! z.o l
'1H13-P655 , e N601E l e l Bleed Valve Closure 1 z.o . l l Bleed Valve Closure l
N631J 1H13-P655 @ l I
)
- z. . o - A I
l Bleed Valve Closure N601N l1H13-P6551 '4.~ ll l l 1 Steamline Flow 'N602A 1H13-P655 13.0 ! f
-{
lN602E 1H13-P655 13.0 l l Steamline Flov ' l l N602J 1H13-t655 13.0 l l Steamline Flow l l
'N602N 1H13-P655 13.0 _l i Steamline Flow l 1
'N604 1H13-P654 6.0 l l Bleed Valve Closure I i
l Dilution Air Flow Alarm Bypass K46 1H13-P654 0.5 l l I i 1 l Dilution Air Flow Alarm Bypass K49A '1H13-P655 0.5 l l [ ;'4', l l l Dilution Air Flow Alarm Bypass X49E 11H13-P655 0.5 l I l
} l .I I l l .
l_ Dilution Air Flow Alarm Bypass 'K49J 1K13-P6551 0.5 _l I I I . lK49N l1H13-P6551 0.5 I I l i I Dilution Air Flow Alar:n Bveass Note Exceptions: Performed by: Date
.t n
h Date S .i. Reviewed by: (Shif t Supervisor) 04-1-01-E32-1 ATT VI L /
g y (, ENCLOSURE 4 SIMULATION FACILITY FIDEllTY REPORT Facility Licensee: System Energy Resources, Inc. Facility Licensee Docket No.: 50-416 and 50-417 Facility Licensee No.: NPR-29.and CPPR-119 Operating Tests administered at: Grand Gulf Nuclear Station Operating Tests Given On: April 27, 1988 During conduct of the simulator portion of the operating tests identified above, the following apparent performance and/or human factors discrepancies were observed. I (a) Discrepancies impacting examination administration: The available initial conditions do not have sufficient decay heat. The CRD pumps failed to isolate on a valid NSSSS signal. A method of monitoring comunications is not available. (b) Items that would enhance examination administration and operator training: Reactor Water Level transmitters cannot be failed to a HIGH or LOW condition. Individual control room alarms cannot be failed to an inoperative condition. The nonnal plant paging system is not available. The simulated telephone system is "ring-down" vs "dial-up" used in the plant. 1/0 Override does not have the capability to override controls to an ACTUATE condition. Loss of Feedwater malfunction is not variable. N >}}