ML20203D703
| ML20203D703 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 04/04/1986 |
| From: | Brockman K, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20203D695 | List: |
| References | |
| 50-321-0L-86-01, 50-321-L-86-1, NUDOCS 8604220300 | |
| Download: ML20203D703 (175) | |
Text
{{#Wiki_filter:_ . . . . ..__ _ _ _ _ _ _ _ _ _ _ _ ENCLOSURE 1 ; EXAMINATION REPORT 321/0L-86-01 l Facility Licensee: Georgia Power Company P. O. Box 4545 l Atlanta, Georgia 30302 Facility Name: Edwin I. Hatch Facility Docket Nos.: 50-321 and 50-366 Written and oral examinations were administered at Plant Hatch near Baxley, Georgia. Simulator examinations were administered at Plant Hatch near Baxley,
- Georgia.
Chief Examiner: L [dMur- 4'/4/76 Date Signed K./E. Brockawin /
. Approved by: Ie N # [/+ '
Jokh F. Mb @o, #cting Section Chief YN/74 Date Signed l l Summary: t Examinations on February 17-20, 1986 Written, oral, and simulator examinations were administered to seven candidates, I all of whom passed. Three operators were administered written and simulator i examinations, as appropriate; all three passed and are considered to meet the l requirements of 10 CFR 55.31(e) for resumption of licensed duties. l i i l l l ! i l I 8604220300 860415 PDR ADOCK 05000321 V PDR
REPORT DETAILS
- 1. Facility Employees Contacted:
*H. Nix, General Manager
*P. Fornell, QA Supervisor -
*C. T. Moore, Training Manager
*R. S.-Grantham, Operations Training Supervisor '
K. Elliott, Simulator Instructor : B. Smith, Simulator Instructor
- Attended Exit Meeting
- 2. Examiner
*K. E. Brockman, Region II, NRC J. F. Munro, Region II, NRC G. S1y, PNL i J. Hanek, EG&G M. Spencer, EG&G
- Chief Examiner
- 3. Examination Review Meeting l
At the conclusion of the written examinations, the examiners provided
- R. S._ Grantham, with a copy of the written examination and answer key for review. The coninents made by the facility reviewers are included as l
Enclosure 4 to this report, and the NRC Resolutions to these coninents are
- listed below. '
(1) Question 1.01 (Repl) and 1.01 (RQ) NRC Resolution: All answers listed are contributors to the lowering of the PCI threshold. The utility response of Fission i l Product (chemical) embrittlement from I or Cd is acknowledged as an additional reason (NEDE 21493). Answer key modified to allow for additional response. ' ( i I (2) Question 1.17(Repl)and 1.10/5.08 (RQ) i NRC Resolution: It is acknowledged the question may not be specific enough to elicit only the answer on the initial answer The key key. (The key noted "other responses as appropriate"). has been expanded to include the responses suggested by the utility (NEDE248108); other accurate responses will also be accepted.
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Enclosure 1~ 2 April 15, 1986 i g (3) Question 2.04 (Repl) NRC Resolution: The graph is detailed enough to distinguish between 320 and 330 KVAR. However the error with respect to
- a power of ten in the distractors is considered significant enough to elicit multiple responses. Question deleted from exam.
(4) Question 4.){b,c(Repl) NRC Resolution: The operators responsibilities in performing this - procedure are limited such that he should be familiar with his direct actions and limitations as opposed to specific nuances. Some knowledges are agreed to be more oriented to the STA; part b is, therefore, deleted from the exam. Part c is reweighted to 0.5 and the answer key modified to require only that an 00-3 must be run to verify reactor power. l (5) Question 8.04b(RQ) NRC Resolution: Disagree. It is essential that the significance of dnily information be known to the senior operator. If he knows that only Unit 2 allows for negative grace periods on surveillances, the question is simplistic. This knowledge is appropriate. No change to the exam or answer key.
- 4. Exit Meeting At the conclusion of the site visit the examiners met with representatives of the plant staff to discuss the re:sults of the examination. Those individuals who clearly passed the operating examinations were identified.
One item requiring inspector followup was identified. It was verified that training on the Safety Parameter Display System (SPOS) had been given to neither the replacement candidates nor these particular (requalification) operators. The utility acknowledged both the impact of and commitment to provide this training. It was agreed that this training would be provided to all candidates from this exam prior to declaring SPDS as being fully operational. This will be carried as Inspector Follow-Up Item No. 50-321/0L-61-01. The cooperation given to the examiners and the effort to ensure an atmosphere in the control room conducive to oral examinations was also noted and appreciated, i The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners. l
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-MA5TEK-l ENCLOSURE 3 U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION l FACILITY: HATCH 1R2 l
REACTOR TYPE: BWR-GE4 DATE ADMINISTERED: 06/02/17 EYAMINER* BROCKMAN, K. APPLICANT: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . INSTRUCTIONS TO APPLICANT! Use separate paper for the answers. Write answers on one side univ. I 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 sin (6) hours after the examination starts.
% OF CATEGOPY % OF APPLICANT *9 CATEGORY VALUE TOTAL SCOPE VALUE CATEGORY l -_-_____ -----_ ------_____ ______ - __ -----___.._-_-_-__--______--____
ng -e n4 go _I_I $_ _I_I_I --__-______ ._____._
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, 93 g g HEAT TRANSFER AND FLUID FLOW
-e em
,5 0
_II II__ _I I_6 _ ___-_-____- ________
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
,15" 15
_'_"_- _I'5 31_I_-
- 3. INSTRUMENTS AND CONTROLS 8 PROCEDURES - NORMAL, ADNORMAL, i _IIII"I__ _'4I_I_2' ___________ ________ 4.
! 29 lb EMERGENCY AND RADIOLOGICAL CONTROL 97.7 5 1 r: 1. 100.00 TOTALS ____'5-__- __--_. __-_-______ ________ l l FINAL GRADE _________________% l All work done on this examination is my own. I have neither gaven 7or received aid. i 5PPL5C5UI~5~555U5IUR5~~~~~~~~~~~~~~ l l l
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2
~~~~is55566Y5555C5-~555T~TR5U5f5R 556~FLUi6 ft60 QUESTION 1.01 (1.00)
The THRESHOLD power below which PCI failures do not occur is known to DECREASE with fuel burnup. STATE two (2) reasons for this decrease in the PCI threshold. l QUESTION 1.02 (1.00) ! The reactor trips from full power, equilibrium xenon conditions. Four (4) hours later the reactor is brought critical and power level is main- ! tained on range 5 of the IRMs for several hours. Which of the following statements is CORRECT concerning control rod motion during this period?
- a. Reds will have to be withdrawn due to '<enon build-in.
- b. Rods will have to be rapidly inserted sirice the critical reactor will cause a high rate of nenon bornout.
- c. Rods will have to be inserted since xer>on will closely follow its normal decay rate.
l d. Rodu will a; proximately remain as is as the xenon establishes ! Its equilibriun value for this pouer level. QUESTION 1 03 (2.00) l A significant amount of e::c e s s reactivity must be loaded into a core'at BOL l so that 1007. Power can be attained at the end of a fuel cycle. For each of the following, LIST the approximate value of K-excess which must be loaded to overcome that negative reactivity component at rated-equilibrium tanditions.
- a. Moderatur temp increase l b. Void fraction increase l
l c. Samarium buildup l d. Xenon buildup l l l l . i l I (***** CATEGORY 01 CONTINUED ON NEXT PAGE **ses) l l t 1 l l l \ _ _ _ _ _ _ _ - _ _ _ _ _ __ _ _ - _ . _ - _ - _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ . - _ _ _ - _ . _ _ - _ _ - - - _ _ - - - _ - _ - - _ - _ _ _ _ w
. s 10 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION > PAGE 3 ;
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~~~~5U5R566 U555C57~5E5T~TR505fER dU6~FL6i6 Ft60 i
QUESTION 1.04 (1.00) STATE for which cor.dition the reactivity coefficient contribution l would be MORE NEGATIVE. EXPLAIN your choice. Mcderator temperature coefficient for a 75% CON 1ROL ROD DENSITY,
-OR-
! Moderator temperature coefficient for a 25% CONTROL l ROD DENSITY. GUESTION 1.05 (2.00)
- a. STATE the design feature in the reactor vessel which ensures proper flow distribution through the core fuel bundles. (0.5) l 6. EXPLAIN how the rectreviation flow distribution would react during l a power increase by rod pull if this design feature were NOT PRESENT.
I INCLUDE IN YOUR RESPONGE THE REASON (S) FOR THIS REACTION. (1.5) l OUcSTION 1.06 (2.00)
- a. DEFINE ' Critical Power'. (1.0)
- b. Which one of the following conditions would tend to INCREASE the Critical Power level assuming all other variables remain unchanged? (1.0)
NOTE: ASSUME NORMAL FULL-POWER OPERATING CONDITIONS
- 1. Inlet subcooling is DECREASED
- 2. Reactor pressure is DECREASED
- 3. The a:n a l power peak is RAISED
- 4. Coolant flow rate is DECPEASED
(***** CATECORf 01 CONTINUED ON NEXT PAGE *****) l l l l i i i
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l l 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PACE 4
~~~~iAEER667sARICi? AEEi iEE sFEE As5 FE0i5 FE6E GUESTION 1.07 (1.00) l Which of the following radioactive isotopes found in the reactor coolant WOULD NOT indicate a leak through the fuel l cladding.
- a. Co - 60 j b. Xe - 133
! c. I - 131
- d. Mr - 87 OUESTION 1.08 (2.00)
! Gaturated steam with 100% quality enters the main condenser at 4.5 psia and with a flow rate of 6E+6 lbm/hr. Condensate exits as a saturated liquid. Circulating water enters the condenser at 62 des F and exits at 77 des F.
- a. CALCULATE the cireviating water flow rate. (1.5)
- b. STATE whether condenser vacuum would INCREASE, DECREASE, or REMAIN THE SAME, if the circulating water flow rate were OECREASED. (0.5) l OUESTION 1.09 (1.00) l STATE for which condition the reactivity coeffletent contribution uov1d be MORE NEGATIVE. EXPLAIN your choice.
Moderator voi<f coefficient for a 1% INCREASE in void fraction at 10% void frection in the core,
-OR-Moderator void coefficient for a 1% INCREASE in void
' r : .. t i o n at 70% void fraction in the core.
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5
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TU55566 5555C5I~U55T ~T Edh5f5R~506'FL0i5 Ft50 00ESTION 1.10 (2.00) As part of the scram procedure, the operator is directed to insert the SRM's and IRM's.
- 2. Following a wevere LOCA, EXPLAIN how these systems could be used to detect gross core damage. (1.0)
- 6. EXPLAIN how these systems could be used to provide a crude indication of water level if level could not be confirmed by normal instrumentation. (1.0)
GUESTION 1.11 (1.25) ARRANGE the following in order of INCREASING heat transfer coef-ficient (Lowest Coefficient to Highest Coefficient).
- a. Free Convection. Air
- b. Boiling Water (Free Convection)
- c. Dailing Water (corced Convection)
- d. Forced Convection. Water
- c. Forced Convection, Air
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 6
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T5ER566Y5555C5I~5EhT"iRd55F5R b5b~iLU5b~FLb5 CUESTION 1.12 (1.00) l l A reactor heat balance was performed (by hand) during the 00-08 l shift due to the Process Computer being 00C. The GAF's were l computed, but the APRM GAIN ADJUSTMENTS HAVE NOT BEEN MADE. ! Which ONE of the following statements is TRUE concerning reactor power? I a. If the feedwater flow rate used in the heat balance calev-( lation was LOWER than the actual feedwater flow rate, then the actual power is HIGHER than the currently calculated power.
- b. If the reactor recirculation pump heat input used in the heat balance calculation was OMITTED, then the actual power is LOWER than the currently calculated power.
- c. If the steam flow used in the heat balance calculation was l LOWER than the actual steam flow, then the actual power is LOWER l than the currently calculated power.
l d. If the RWCU return temperature used in the heat balance cal- ! culation was LOWER than the actual RWCU return temperatore, then the actual power is HIGHER 'han the currently calevlated power. QUESTION 1.13 (3.00) Atttched Figure #595 represents a transient that could occur at a BWR. Given (1) Master Recire Flow Controller Fails High
'C) No operator actions occur (3) Recorder Speed = 1 division = 1 minute EXPLAIN the cauce(s) of the following recorder indications! I i l
- a. Reactor Water Level DECREASE (Point A) (0.5)
- b. Reactor Power INCREASE (Point B) (0.5)
- c. Core Flow DECREASE (Point C) (0.5)
- d. Reactor Pressure INCREASE (Potnt D) (0.5)
- e. Total Feedwater Flow INCREAGE (Point E) (0.G) l i f. Total Steam Flow INCREASE (Point F) (0.5) l I (***** CATEGORY 01 CONTINUCD ON NEXT PAGE resse) l I
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 7
~ ~~~~iEER566iUE55C5I~555I~TR55555R dU6~5EU56~fL U QUESTION 1 14 (2.00)
During your Shift, an SRV inadvertantly opens from 100% power and 1000 psia. Use a Mollier Diagram or the Steam Tables to answer the following'
- a. STATE the tailpipe temperature. assuming atmospheric pressure in the Suppression Pool and No Reactor Depressuri:ation. (0.5)
- b. If the Suppression Pool Pressore were to INCREASE, STATE whether the Tailpipe Temperature would INCREASE, DECREASE, or REMAIN THE SAME. (0.5)
- c. If the reactor is depressurized when the SRV is opened. STATE whether the Tailpipe Temperature will INITIALLY INCREASE, DECREASE, or REMAIN THE SAME. (0.5) !
- d. STATE the Reactor Frassure at which the Tailpipe Temperature would be at its MAXIMUM value (during the depressurization). (0.5)
(ASSUME A SATURATED ?YSTEM AND IPSTANTANEOUS HEAT TRANSFER) OUESTION 1.15 (1.00) Attached Figure i 214 illustrates four potential temperature profiles (relationships) for a COUNTER FLOW heat e::c ha nge r . Which of these graphs most accurately shows the temperature profile (relationship)? NOTE' THE TEMPECATURE PROFILES FOR EACH RESPONSE ARE PLOTTED ON THE SAME TEMPERATUPE SCALE (i.e.. The TOP line is the HOT flow and the 00TTOM line is the COOLING uATER flow). QUESTION 1 16 (1.00) Reactivity is defined as which of the following?
- 2. The ratio of the number of neutrons at some point in this '
generation to the number of neutrons at the same point in ; the previous generation.
- b. The fracttonal change in neutron population per generation.
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- c. The factor by which neutron population changes per genera-tion.
- d. The rate of change of reactor power in noutrons per socond.
(as*** CATEGORY 01 CONTINUCD ON NEXT PAGE ****s)
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- 1. PC:INCIPLES OF NUCLEAR POWER PLANT OPERATION, PACE 8
~~~~iSEE566Y5555C5I~5E55~YE 55EER~d56~ELb56~ELbh QUESTION 1.17 (1.00)
LIST two (2) Design Characteristics which allow for the use o f' Hybrid Control Rods? (***** END OF CATEGORY 01 asses)
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 9 GUESTION 2.01 (1.00) l l Which one of the f ollowing accur ately describes Low- Low Set (LLS) l logic, as applied at Plant Hatch?
I
- a. Lowers BOTH the opening and closing setpoints of the LLS valves,
- b. Controls the operation of all relief valves, e:teepting E and H. ,
- c. Is activated by a position switch which confirms any SRV opening.
- d. Is applicable to Unit 2 ONLY.
QUESTION 2.02 (1.00) The Unit 2 Vital AC Power 120/240 v Distribution Cabinet 2A is ! normally supplied from 600 v Bus 2D through a Dattery Charjer and a Static Inverter. If the Static Inverter fails ...(CHOOSE ONE)
- a. ... the 125 vde battery will maintain power to the Vital AC Cabinet for up to 6 hours,
- b. ...the power supply can be manually transfered to the alternate 600 v Evt 2C / Vital AC Transformer 2A by depressing a transfer PB.
- c. ...the power supply will automatically transfer to the alternate 600 v Bus 2C / Vital AC Transformer 2A,
- d. ...the power sv., ply can be manually transfered to the alternate 600 v Dus 2C / alternate Static Inverter by depreusing a transfer PB.
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 10 l
QUESTION 2.03 (2.00) With reaard to the Unit 2 RCIC System;
- a. Which of the following is the unty normally CLOSED valve in the RCIC steam supply flow path in the STAND-BY lineup? (1.0)
(1) Steam Supply Valve (F045) (2) Outboard Steam Isolation Valve (F000) (3) Trip Throttle Valve (4) Governor Valve
- b. For each of the situations listed below, STATE whether final RCIC injection into the reactor would continue or reinitiate AUTOMATICALLY, would require CONTROL ROOM Operator Action, or would require LOCAL Operator action. Assume that RCIC had auto initiated prior to any of the following situations.
(1) The Test Bypass Valve to the CST (F022) FAILS OPEN. (0.25) (2) A 125% Overspeed Trip is received due to low control oil pressure. Control oil pressure is then returned to normal. (0.25) (3) After decreasing to 50 psis, RCIC Steam Line Pressure increases to 150 pois. (0.25) (4) After increasing to +60 inches, Reactor Vessel Water (0.25) Level decreases to -60 inches. MUESTTn" 2;0M - '4-;00) - The main generator is on line at 010 megawatts when a hydrogoti' beak in the generator reduces hydrogen pressure to 15 psig 'Osing attached Figure i 259 (Estimated Capability Curve), w jhiv6he of the following is the maximum leading REACTIVE load all,queT on the generator, if a power factor of 0.00 is to be maint ened.
- a. 320 KVAR
- b. 330 KVAP
- c. s MVAR
- a,n v o .o .
(***** CATEGOPY 02 COHlINUED ON NEXT PAGE *****)
1 O I l L 2. PLANT DESIGN INCLUDING SAFETY AND EMERCENCY SYSTEMS PAGE 11 i l GUESTION 2.05 (1.00) Which of the following will NOT occur when one Bus of 125 VDC power is lost to ATTSS
- a. Either HPCI or RCIC in INOP.
- b. Core Spray actuation limited to the remaining division.
- c. Either the inboard or outboard Group II valves isolate
- d. Half of the LLS SRV's will not cycle in LLS Mode.
QUESTION 2.06 (2.00) Concerning the CRD System'
- a. LIST the two (2) causes of a CRD Accumulator Trouble alarm (Setpoints NOT required) and EXPLAIN the action which must be taken to determine the cause. (1.0)
- 6. Shortly after resetting a reactor SCRAM, it is reported that Cooling Water flow is LOWi however, the CRD flow indicator is reading FULL SCALE. EXPLAIN this apparent discrepancy. (1.0)
GUESTION 2,07 (2.00) Concerning the Unit 2 RHR System, while operating at 100% RTP. For each of the following situations, STATE whether the statement is TRUE or FALSE.
- a. The Inboard Injection Valve (F015) may be opened ONLY if the Outboard Injection Valve (F017) is shut. (0.5)
- b. The Inboard Containment Spray Valve (F021) CANNOT be opened unless the Outboard Valve (r016) is shut. (0 5)
- c. The Shutdown Cooling Isolation Valves (F000 & F009) are interlocked shut with the plant in thewe conditions. (0.5)
- d. The Torus Isolation Valve (F024), the Toros Spray Valve (F027), and the focus Cooling Valve (F020) may all be in the OPEN posttion at *he same time.
(0,5) (arres CATCGORY 02 CONTINUED ON NEXT PACE *****)
- 2. PLANT DESIGN INCLUDING SAFETY AND EHERGENCY SYSTEMS PAGE 12 OUESTION 2.08 (1.00)
, STATE how the integrity of the Core Spray piping between the l RpV and the Core Shroud is confirmed, and how the indication (s) would change if the intagrity was LOST. l GUESTION 2.09 (1.00) , l EXPLAIN HOW/uMY ati SRV discharge pipe (tailpipe) could be damaged due to its vacuum breaker SLICKING SHUT during repeated actuctions (lifting / resecting) of the SRV. QUESTION 2.10 (1.00) Which of the following lists of responses (a - d) CORRECTLY describes the main turbine's response to an OVERSPEED condition? RESPONSES SPEEDS (a) * (b) (c) (d) 100% ' Master
- All ICV's All ICV's ' Master
- ICV's begin begin to b es t ri to ICV's begin to t h r- o t t l e throttle throttle to throttle
' Slave ' Slave' ICV's begin ICV's begin to t h r o t, t l e to throttle ;
All ICU's All ICV's All ICV's All ICV's Fiell Clowd Full Closed Full Closed Full Closed 110% nocit an t e 11 Electrical Mechanical Electricai O'Spd Trip O'Spd Trip O'Spd Trip O'Spd Trip 111.5% Electrical Mechanical Electric-21 Mechanic 11 0'Spd Trip O'Spd Trip O'Gpd Trip O'Spd Trip
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 13 i
I ( QUESTION 2.11 (1.00) The High Pressure Coolant Injection (HPCI) System utilizes a steam turbine to drive a pair of centrifugal pumps to inject water into the reactor pressure vessel under emergency conditions. The main HPCI pump ...(CHOOSE ONE)
- a. ... takes its svetion directly from the CST and discharges to the HPCI booster pump, which then ' boosts
- the pressure high enough fa reactor injection.
- b. ... takes its suction from the discharge of the HPCI booster pump which shares a common shaft with the main pump via a speed reduction coupler.
- c. ... discharges, at leastr 5000 gem at rated pressure into the
'A' feedwater line, opstream of the feed header flow detectors.
- d. ...is driven by a two-stage, Terry turbine, which receives its steam supply from the *D* main steam line, downstream of the flow '
rest.ictors. GUESTION 2.12 (1.n01 Regarding the Standby Gv; Treatment System (SGTS)* LIST the tue (7) autometic actions whico will occur in the oper-ating train of SCTS, given that the deluge sprinklers for that train receive a valid initiation signal. QUESTION 2.13 (1.00) The Full-Core Display on the center panel has a OLUE scram light for each control rod. CISCRIOE what is DIRECTLY indicated when this light is illuminated. (***r* CATCCORY 02 CONTINUED UN NEXT PAGE *****) a - . .
-c.--..-_ , , . , . _ . , . . , . - _ . - , _ . . , _ , _
, __ -...-_,._-,,_m , , , _ _ , . _ _ ,m_ . . _ . , _ , . . _ _ , _ - .
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERCENCY SYSTEMS PAGE 14 QUESTION 2.14 (1.00)
The plant is operating normally at power when Pump A Controlled L eakage (FS 'A') alarms LO (0.1 gpm) and you note an INCREASE in No.2 Recire Pump seal pressure with NO CHANGE in No. 1 seal pressure. Which of the following failures would cause these indications?
- a. Failure of No. 1 seal
- b. Failure of No. 2 seal
- c. Plugging of the No. 1 internal restricting / breakdown orifice
- d. Plugging of the No. 2 internal r es tr icting/br eak down orifice NOTE: NO OTHER ALARMS ARE PRESENT GUESTION 2.15 (1.50)
Concerning Primary and Secondary Containment Systems for Unit 2:
- a. STATE the Bases for inerting the Drywell and the Torus (0.5)
- b. EXPLAIN the Bases for having a MIPIMUM and MAXIMUM Torus Level. (1.0)
GUESTION 2.16 (1.00) Concerning the SRM Detection System:
- a. STATE how an increase in the gas fill pressure would effect (IN-CREASE, DECREASE, or NO CHANGE) the sensitivity of the detector? (0.5) i b. STATE how elimination of the discriminator would effect (IN-I CREASE, DECRCASE, or NO CHANGE) the control room SRM count rate? (0.5)
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 15 GUESTION 2.17 (2.00)
The displays of the indicator lights for ATTS can indicate that the system is in its normal operating made with NO ALARM CONDITIONS present? For each of the following STATE the status (ON/0FF) and the indicator color (If ON) when in this normal mode. NOTE
- FIGURES 9.9(3) AND 9.9(6) PROVIDED FOR REFERENCE
- a. STATUS /MTU Indicator
- b. GROSS FAIL /hTU Indicator
- c. POWER /P925 Indi,cato,r QUESTION 2.18 (1.50)
With regard to the diesel fire pumps:
! a. STATE the source of feed to the Fire Protection Storage Water Tanks, given that the Fire Water System has initiated. (0.5)
- b. Which one of the following is the initiation signal for the diesel driven fire pumps? (1.0)
- a. Low Fire Mater Header Pressure
- b. Low Air Pressure (Dry Pipe Systems)
- c. High Fire Water Header Flow Rate
- d. Smoke Detector Actuation
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****) I t r 1 1 [ 1 l l [ . .
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 16 l 1
l GUESTION 2.19 (2.00) LIST the four (4) DC automatic shutdown signals that remain OPERADLE If the diesel is started by using the local Emer3ency Start PB. NOTE! CONSIDER DIESEL GENERATOR 2A QUESTION 2.20 ( .50) TRUE or FALSE: When synchrontzing the Main Generator to the grid, the synchroscope . circuitry provides an INTERLOCK to preclude the Output Breaker l from closing, unless the frequency is a p p r o::i m a t e l y in phase, as l shown by the synchroscope being between 11 and 1 0' Clock. i l l l l l l l l (messa END OF CATEGORY 02 *****) l
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- 3. INSTRUMENTS AND CONTROLS PAGE 17 QUESTION 3.01 (1.00)
Given: Unit 2 in control of D/G *B' D/G *B" Mode Switch in TEST (Surveillance being performed) Electrical distribution NORMAL (rull Power Lineup) D/G *B' is at rated speed and voltage, but not synchronized, when power is lost to 4160 volt Bus 2F. Which of the following accurately describes the system operation?
- a. Bus 2F can be powered by D/G 'B' when the operator takes the Output Breaker Switch to CLOSE and has the SYNC SCOPE activated,
- b. Bus 2F will be powered by D/G 'B' automatically, after 12 seconds; appropriate loads will be picked up sequentially.
- c. Bus 2F can not be powered by D/G 'B' while it is in the TEST moder given these conditions.
- d. Bus 2F can be powered by D/G *B' when the operator resets the Lockout Relay, activates the SYNC SCOPE, and takes the Output Breaker to CLOSE.
QUESTION 3.02 (1.00) The Reactor Manual Con'.rol Systnm includes several manual control switches en the 603 Paneli one of these is the ' Rod Out Notch Overri,de" (RONOR) switch. This fbree position switch ...(CHOOSE ONE)
- c. ...must be held in the ' notch override
- position while the 'CRD Control' witch is held in the ' God In' position if continuous rod intertien is desired.
S. ..., if held in the ' Emergency Rod In" position, will bypass all red insert blocks e >:c e p t those imposed by the RSCS. j
- c. ..., if held in the " Emergency Rod In* Position, will illuminate an amber light above the switch. j
- d. ..., when used for emergency rod insertion, bypasses the automatic sequence timer and acts directly on the insert bus.
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****) 9
- 3. INSTRUMENTS AND CONTROLS PAGE 18 GUESTION 3.03 (2.00)
Regarding the SRV's and the associated Low Low Set (LLS) logic: There are three lights associated with each SRV - RED, GREEN, and AMBER. EXPLAIN what each of the different colored lights indicates, and STATE whether each would be energtred or de-energized during the time its SRV was open as a result of reactor pressure reaching the SRV's relief setpoint.
-ASSUME THAT THE VALVES IN GUESTION ARE LLS VALVES-AND WERE ACTUATED BY THE LLS FUNCTION GUESTION 3.04 (2.00)
Hith regard to the Off gas Radiation Monitoring System
- a. LIST the three (3) combinations of radiation instrument trip signals that will cause an Off gas System auto-isolation. (1.5)
- b. LIST the Off gas System valve (s) which CLOSE on an auto-isolation. (0.5)
QUESTION 3.05 (1.00) The Unit 1 Primary Containment Atmospheric Cor. trol System can be used to vent the primary containment under normal operating con-ditions. To do this, the operator must manipulate the key-locked bypass switches. The direct effect of doing this, and the method by which this is done, is to ...(CHOOSE ONE)
- a. ... override the HI MSL PRESSURE (? 850 psig) isolation signal by HOLDING the bypass switches in the BYPASS position.
- b. ... override the MI MSL PRESSURE (: 850 psig) and the LOCA isolation signals by HOLDING the bypass switches in the BYPASS position.
- c. ... override the HI MSL PRESSURE ( : 850 psig) isolation signals by PLACING the bypass switches in the DYPASS position.
- d. ... override the HI MSL PRESSURE ( -
850 psig) and the uuCA isolation signals by PLACING the bypass switches in the BYPASS position. (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****) 1
30 INSTRUMENTS AND CONTROLS PAGE 19 QUESTION 3.06 (1.00) A normal cold plant startup is in progress with the RSCS Sequence Mode Selector (SMS) Switch in ' Withdraw." Which of the following operator actions will cause the dim back-lights for the 'A-12" sequence rods to extinguish -AND- the dim backlishts for the 'A-34' sequence rods to illuminate? NOTE: PICK THE MOST RESTRICTIVE CONDITION (i.e., the least amount of operator action required)
- a. FULLY WITHDRAW all 'A-12' sequence rods; SELECT 'A-34' on the Rod Sequence Selector (RSS) Switch.
- b. The position of the 'A-12" sequence rods is immaterial; SELECT 'A-34' on the Rod Sequenca Selector (RSS) Switch.
- c. The position of the 'A-12' sequence rods is immateriali SELECT 'A-34' on the Rod Sequence Selector (RSS) Switch. and SELECT any *A-34' rod.
- d. FULLY WITHORAW 311 'A-12' sequence rods; SELECT 'A-34' on the Rod Sequance Selector (RSS) Switch, and SELECT any "A-34' rod.
QUESTION 3.07 (1.00) STATE the TREND (INCREASE, DECREASE, NO CHANGE) that you would expect to see as a result of a " Jet Pump Riser Failure" for each of the following,
- a. Failed Jet Pump Flow,
- b. Core Differential Pressure.
- c. Reactor (APRM) Power.
- d. Indicated Core Flow.
QUESTION 3.00 (1.00) Given a Low CST Level on Unit 1 AND a High Torus Level on Unit 2. DESCRIBE the responses of the HPCI and RCIC Systems for BOTH units. (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
30 INSTRUMENTS AND CONTROLS PAGE 20 QUESTION 3.09 (1.00) The main turbine is at 1800 rpm in preparation for synchronizing the main generator to the grid (i.e., the main generator output breakers are still open). DESCRIBE what will happen if the "All Valves Closed' pushbutton is depressed? INCLUDE IN YOUR DESCRIPTION ALL (If Any) TURBINE SUPERVISORY VALVES (STEAM VALVES) WHICH WILL RESPOND AND THE RESPONSE OF EACH. QUESTION 3.10 (1.50) The Main Turbine first stage pressure switches provide permissives and/or control signals for several plant functions. LIST three (3) of these permissives/ control functions. QUESTION 3.11 (2.00) Unit 2 is operating at 100% RTP when APRM 'A' fails upscale and results in a reactor half-scram. Utilizir3 the attached RPS trip logic diagrams (Figures 1224 A thru F) DESCRIBE in a STEP-DY-STEP fashion (with regard to the opening /elosin3, ener-giring/deenergining of ALL applicable contacts and relays) how the
~
APRM vpscale trip results in an actuation of the scram solenoid valves. NOTE' IF THE ATTACHED DIAGRAMS CAN NOT DE EASILY READ, ASSIGN THE CONTACTS / RELAYS, ETC NUMBERS AND REFER TO THEM IN YOUR ANSWER. (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
- 3. INSTRUMENTS AND CONTROLS PAGE 21 QUESTION 3.12 (2.00)
For each of the following situations (1 and ii) select the correct Feed-water Control System / plant response from the list (a through e) which follows. An answer may be used more than once, and NO operator actions are taken.
- a. Reactor water level decreases and stabilizes at a lower level.
- b. Reactor water level decreases and initiates a reactor scram,
- c. Reactor water level increases and stabilizes at a higher level.
- d. Reactor water level increases and initiates a turbine trip.
- e. None of the above.
- i. The plant is operating at 70% power, in 3-element control, when One (1) MSIV Fails Shut.
ii. The plant is operating at 100% power, in dP control. when one Feed-water Flow Detector fails to indicate Zero Flow (i.e., Fails Down). GUESTION 3.13 (1.00) Concerning the Rod Block Monitor (RBM) on Unit 1: Assume a Control Rod surrounded by four (4) LPRM strings is sel-ected. LIST the detector assignments to RBM Channel A (i.e., the number of LPRM's per level). (1.0) GUESTION 3.14 (1.75) Concerning the Rod Worth Minimizer (RWM) on Unit 2:
- a. When a select error occurs on the RWM, STATE whether the operator ccn still move the rod. (YES or NO) ASSUME THE RWM IS NOT BYPASSED NO RCD SLOCKS EXIST PRIOR TO SELECTING THE ROD. (0.5)
- b. EXPLAIN the bases for your decision in part (a). Consider in your e::planat ion both an attempted insert and withdraw action. (1.25)
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
- 3. INSTRUMENTS AND CONTROLS PAGE 22
. ,,.crvrnu , 4e re nns ELL 'TLb The S DS F'rimary Display provides information concerni 'fne ADS, amon3st other parameters.
- a. DESCRIBE how the ADS Valves are differo lated from the other SRV's. (0.5)
- b. STATE the significance of + following COLORS. as they are used for indication of AD alve status: (1.0)
(1) Green (2) 0 .se
- c. _ T the two (2) time periods over which TREND Displays
':re ci er (0.5)
QUESTION 3.16 (1.50) LIST ALL requirements which could/must be met in order to actuate ADS. (Automatic Actuation) QUESTION 3.17 (1.00) A Control Rod is declared INOPERABLE and valved out of service on Unit 2. DESCRIBE what must be done to the RSCS logic to allow continued rod withdrawal. A BLACK-AND-4HITE ROD PATTERN HAS NOT YET SEEN ACHIEVED IN THE STARTUP SEQUENCE. (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
- 3. INSTRUMENTS AND CONTROLS PAGE 23 GUESTION 3.18 (2.00)
STATE for each of the following conditions, which Unit (if any) will recieve the 1B EDG.
- a. Select Switch Position -
Unit 1 Unit 1 Condition - LOCA/LOSP Unit 2 Condition - LOSP
- b. Select Switch Position -
Unit 1 Unit 1 Condition - LOSP Unit 2 Condition - LOCA/LOSP
- c. Select Switch Position -
Unit 2 Unit 1 Condition - LOCA/LO3P Unit 2 Condition - LOCA/LOSP
- d. Select Switch Position -
Unit 2 Unit 1 Condition - LOCA Unit 2 Condition - LOSP l I l (***** END OF CATEGORY 03 *****)
40 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 24
~ ~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~R5656L6556dL C6sTR6L GUESTION 4.01 (2.00) 34AB-OPS-009-2, ' Inability to Shutdown With Control Rods",
states that'
... if at any time, either condition b(1) or b(2) exists, and either _____ (a.1) ____ or ____
(a.2) ____ , and if it is obvious that the reactor cannot be shutdown and, in the judgment of the Shift Supervisor, or in his absence, a licensed operator, a hazard exists to the environs, personnel, or the plant, utilize the standby liquid control system per 34S0-C41-001-2.*
- a. LIST conditions (a.1) and (a.2). (1.0)
- b. LIST conditions b(1) and b(2). (1.0)
GUESTION 4.02 (1.00) 34AB-OPS-002-2, ' Pipe Creak Inside Primary Containment', lists numerous conditions indicative of a break. Which one of the following supports the suspicion or a SMALL break inside con-tainment,
- a. Decrease in reactor water level; pressure and/or temperature in-crease in Drywelli airborne activity increase in Drywelli increased DWFDS operating frequency.
- b. Decrease in reactor pressure; pressure and/or temperature increase in Drywelli generator load decreasei DWFDS high level.
- c. Reactor Scram from low water level; Pressure and/or temperature increase in Drywelli generator' load decrease; increased DWFDS operating frequency.
- d. Reactor Scram from high Drywell pressure; generator load decrease; airborne activity increase in Drywelli DWFDS high level.
QUESTION 4.03 (1.00) 34S0-B31-002-2, ' Reactor Recirculation Pump Speed Changes", cautions the cperator not to start an idle pump unless the other pump is operating at < 50% loop speed. EXPLAIN the reason for this procedural CAUTION. (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
- 4. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 25
~ -~---------------------- ~~~~REU5bl65555L C6UTR5L QUESTION 4.04 (1.00)
You enter an area posted with the following sign' l_________ CAUTION i i_________l 1 I Your purpose is to conduct a normal i I Surveillance inspection. I I I I IContamin.1 l ___ Area __l LIST the minimum protective clothing requirements which an RWP for this area could require. GUESTION 4.05 (1.00) 34AB-DPS-006-2, ' Post Accident Venting', is very specific as to what conditions require entry into this procedure. LIST the two (2) conditions. (Setpoints NOT required) QUESTION 4.06 (1.00) 34S0-C11-003-2, ' Control Rod Drive System", cautions the operator to check the (position of the) Reactor Recirculation Pomps' discharge and suction valves prior to CRD startup. STATE the basis for this caution. ( * *.v *
- CATEGORY 04 CONTINUED ON NEXT PAGE *****)
P l
l l l 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 26
~~~~ -
i R5656L6656dL 66UTR6L'~~~~~~~~~~~~~~~~~~~~~~~ l ____________________ 4 l GUESTION 4.07 (2.00) . 42FH-ENG-010-2, ' Control Rod Movement', provides numerous l STANDARD PRACTICES which apply when rods are being moved for the purpose of chan3i n3 Power level. l
- m. With LHGR > 8 Kw/ft, STATE the period of time you must wait I between successive notch withdrawals of the same rod. (0.5) l l b. With HIGH POWER and HIGH FLOW conditions, STATE which type of rod - SHALLOW, DEEP. CENTRAL, or EDGE - should NOT be with-l drawn nor inserted. (1.0)
- c. With power < 30%, STATE the verification which is l utilized when latching the first rod in any group. (0.5) l QUESTION 4.08 (2.00) l A pipe break occurs and results in a decreasing reactor water l level and decreasing reactor pressure. LIST the four (4) criteria which would require the MANUAL initiation of ADS, given that auto-initiation has not occured.
00ESTION 4.09 (1.00) Concerning a " Cable Spreading Room Fire', Procedure 34AB-FPX-040-0:
- a. STATE who is responsible for directing the control room response to cable spreading room fire. (0.5)
- 6. TRUE or FALSE (0.5)
For an exposure fire (in the cable spreading room) involving com-bustables WITHOUT electrical insulation involvement, the Contr.ol Room Operator in the affected unit will initiate, as necessary, a rapid load reduction, will trip the turbine generator, will manually SCRAM the reactor, and will place the Mode Switch in SHUTDOWN. l l l (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) l i l l I I l l \ \
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 27
~~~~RE65UL655Chl 56NTRUL'~~~~~~~~~~~~~~~~~~~~~~~
GUESTION 4.10 (1.00) Per Procedure 10AC-MGR-003-OS, " Preparation and Control of Pro-cedures', for which of the following periods are SRO PROCEDURE CHANGES valid?
- a. One Time Use Only
- b. Not to Exceed 7 days
- c. Not to Exceed 30 days
- d. Until the next bienniel review QUESTION 4.11 (1.00)
A plant startup is in progress and condenser vacuum is being established in accordance with 34GO-OPS-001-2, " Plant Startup'. Which one of the following is the proper sequence for component / subsystem startups.
- a. Steam Packing Exhauster, Steam Seal Header, Mechanical Vacuum Pump, Steam det Air Ejector,
- b. Steam Seal Header, Steam Packing Exhauster, Mechanical Vacuum Pump, Steam Jet Air Ejector.
- c. Mechanical Vacuum Pump, Steam Packing Exhauster, Stesm Seal Header, Steam Jet Air Ejector,
- d. Steam Pading Exhaus ter , Mechanical Vacuum Pump, Steam Seal :le ader , Steam Jet Air E jector -
GUESTION 4.12 (1.00) 34GO-SUV-001-OS, ' Control and Surveillance of Locked Valves", provides svideance concerning the proper way to confirm and verify a locked valve position. DESCRIBE the proper way to CONFIRM Locked Valve positioning and Locking Device integrity. (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
l
- 4. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 28
~~~~R h5 UL65iBAt CBNTE5t GUESTION 4.13 ( .50) 30AC-OPS-001-OS, ' Equipment Clearance and Tagging', states:
"Under no circumstances will equip-ment be operated or manipulated when tagged with a HOLD tag.
LIST any exceptions to this statement. (If no exceptions exist, state so!) GUESTION 4.14 (1.00) A CAUTION in 34AB-OPS-002-2, ' Pipe Break Inside Primary Containment *, states that'
'A Pipe Break in the Drywell could result in Reactor Level Reading inaccurecies due to increased temperature.
Refer to Section 'E' ... for erratic correction method before proceeding. Erratic level indication could be due to water in the sensing line flashing to steam.' Using Section "E*, enclosed, and given the following readings 2T47-R620 192 deg F 2r47-R621 228 deg F 2B21-604A +26 inches l 2B21-604B &23 inches CALCULATE the best estimate of Actual Water Level. SHOW ALL INTERHEDIATE STEPS USED TO DETERMINE YOUR ANSWER (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
1
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 29
~~~~RA5i5t55fEAL EBATR5L QUESTION 4.15 (2.00)
STATE which Emergency Classification is appropriate for the following definitions.
- a. Events are in progress or have occurred which involve actual or potential substantial degradation of the level of safety of the plant.
- b. Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant.
- c. Events are in progress or have occurred which involve
. actual or imminent. substantial core degradation or melting with the potential for loss of centainment integrity,
- d. Events are in pro 3ress or have occurred which involve an actual or likely major failure of plant functions needed for protection of the public.
QUESTION 4.16 (1.00) Concerning 34AB-OPS-002-2, " Pipe Break Inside Primary Containment *:
- a. STATE when Drywell Cooling Fans are REQUIRED to be placed in service. (0.5)
- b. STATE when Drywe'l l Spray is REGUIRED to be initiated. (0.5)
- s. e c 0UESTION 4.17 ( 2 rM)
Concerning Procedure 42CC-ERP-007-2:
- a. LIST the_ maximum Steady-State power allowed. (0.5) b FATE 5 r ~ c'te: ,r;uer ::1:victicr,; ;; :; s t b e ,, s d = i d 2cre see.f.-1iance uit' tS: 1 1 i t ., t o t : ; . {G.31
- c. DESCRIBE what actions the operators must take after a power adjustment with Recirculation Pumps. (Include actions required, c,5 qvs-tity equi-ed- pe*iedicity- and ultisete limits.) (-Frt)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
r- ,
- 4. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 30
~ ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~R5656L6656AL C6UTRUL QUESTION 4.18 (2.00)
LIST four (4) pieces of information which must be entered onto a RED TAG in order for it to be properly hung. (Information which an operator must write on the tas!) GUESTION 4.19 (1.75) The Procedure Numbering System consists of a twelve character alpha-numeric designator, with three dashes (e.g., 31S0-E11-034-1S). This catagorizes each procedure in seven ways. MATCH the Column A characters (using the er: ample above) with the Column B information, in order to most accurately describe the numbering system. COLUMMN A COLUMN B
- a. 1st (3) 1. Safety Status b, 2nd (1) 2. Sequence Number
- c. 3rd/4th (S0) 3. Unit Applicability
- d. 5th/6th/7th (Ell) 4. Procedure Type
- e. 8th/9th/10th (034) 5. Responsible Department
- f. 11th (1) 6. System Identification 3 12th (S) 7. Responsible Section
(***** END OF CATEGORY 04 *****) (********s**** END OF EXAMINATION *:s*************) l l l l i
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l E. EEMI9E LEEL_BE601ts_G9BBEGngti_E93_yg3ey_Ib11t!wtJgeIn
- 1. O,b t a i : Orywell Temperature Re,disegn (u. ti,e 185' Elevat n' l e' ape ratur e Indicator ET4T PCOO gaves Lisa s reading for t e
' fide end ET47-RC.C1 g i .i e s this . esdirig f or the south
- 2. Refer to riguae 1 and ott in the correc{ ion factor for eact temperature reading. .
3, 4 the correction f.ctor f ound using tise (Jorth Drywel Temperature Reading ttie icvel it.Ji ca ti ore visown on Leve Indiretor 2021-REO4A. .
- 4. 5- the correction facter found Lating tlie South Drywel Tec, si sture Pe-a ding . th= I=v=1 ...w. . tion sisown on Leve
'Indireier 2001 RC04C.
~
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?. .'5 . ' .T."
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 31
~~~~iA5RE66YU355C5-~U5dT5R3Y5FER~5U6~FLUi6 FLUU ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN: K.
ANSWER 1.01 (1.00)
- 1. Neutron embrittlement of the cladding.
l
- 2. Thermally induced pellet growth. l
- 3. Inward motion of the cladding walls (creepdown). l
- 4. Firsion Product (Chemical) embrittlement of the cladding (from Cd/I). (20 0.5 each)
REFEPENCE GGNS* MCD, PCIOMR, p EIH* L-RG-673 ANSWER 1.02 (1.00) a REFERENCE BFNP: XENON & SAMARIUM LP, pp 4, 12 GGNS: OP-NP-514 EIH: L-RO-606 l ANSWER 1.03 (2.00)
- a. 4.77% (4- .48%)
- b. 3.8% (+- .33%)
- c. 1.0% (&- .10%)
- d. 3.0% (+- .30%) *0.5 each)
REFERENCE GGNS: OP-NP-513; OP-NP-514 EIH: L-RO-604; L-RG-605 t _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ }
e a
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 32
--- isEEs567sAsiC5- REAi iEAssFEE Es5 FEUi5"FE5E i ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, M.
l ANSWER 1.04 (1.00) 75% CONTROL ROD DENSITY. (0.5) (The increased Control Rod Density causes greater competition 'or the l l thermal neutrons; this necessitates greater pin power for the same net H l Power output). Higher pin power results in a greater Void Fraction l which causes a more negative coefficient. (0.5)
-- OR --
With a greater rod density, a greater number af neutrons are ' lost' to the control rods (increased leakage). Thus, a change th rod density effects reactivity more. by allowing increased abscrbtion by other fuel bundles. (Can also explain why low rod density does not have a large reactivity effect, since the leakage tc other fuel bundles is already so large) (0.5)
-- OR --
As rod density decreases, the moderator fills the rods' cavities as they are withdrawn, f. h u s increasing the Moderator-to-Fuel ratio. This shifts the operating point towards the overmoderated region. Thi- results in the Moderator Temperature Coefficient becoming l e e ., negative as the Control Rods are withdrawn (less C/R Density). (0.5)' PEFERENCE EIH: Reactor Physics L/P, pp 1.7-9, 10, & 13; L-RO-604 90EP: L/P 02-2/3-A, pp 141 - 143 GCNS: GP-NP-513 ANSVER 1.05 (2.001 Crificed Fuel Support Pieces (Core Grificing) (0.5) S, Ac power increases, the flow to the central (higher powered) bundles would decrease (0.25); flow to the peripheral (lower powered) bundles would increase (0.25). This is due to the increased two phase flow resistance that is developed in the higher powered bundles where there is greater boiling (0.5) and the resultant ' restriction
- that this would pose to flow, thus sending it preferentially through the lower povered bundles (0,5). (1.5)
REFERENCE General Electric NEDE 24810, September 1983 GGNS: Thermal Limits LP. p 28 EIH: GPNT, Vol V, Chapter 2.2.J 3 , -
s ,_e %AJ n. 4
. -2 . , PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 33
's ;c- =---------------------------------------------
~
' JHERM0DYN AMICS , HEAT TRANSFER AND FLUID FLOW ANSWERS -- HATCH 182
-86/02/17-BROCKMAN, K.
ANSWER 1.06 (2.00)
- a. The assembly power which would cause the onset of transition boiling at some point in the assembly. (1.0)
N b. 2 , REFERENCE BFNP TP.ANSITION BOILING & ATLAS TESTING LP,P.5-6 GEXL CORRELATION & CRITICAL POWER LP,P.3 GGNS: MCO, THERMAL LIMITS, P.26,32-33 EIH: L-RO-672 ANSWER 1.07 (1.00) a REFERENCE BFNP: BFN Mitigating Rx Core Damage, pp 17 - 18; RO 85/02/01 GGNS: MCD, SECTION 4.4
<EIH: L-RG-540 (MCD)
ANSWER 1.08 (2.00) NOTE: 0.9 - EQUATION; 0.5 - VALUE SUBSTITUTIONSI 9.1 - MATH h(out) = 1129.4 Btv/lbm; h(in) = 125.9 Btv/lbmi c(p) = 1.0 Btv/lbm-des F STEAM 0=m { h(out - in) } = 6E+6 (1129.4 - 125.9)
= 6.02E+9 B',u/hr HEAT REJECTED BY THE STEAM WATER 0=m { c(p)
- T(out - in) }
m=0/{ c(p)
- T(out - in) } = 6.03E+9/(1.0
- 15) l = 4.01E+8 lbm/hr
- b. Decrease (Absolute pressure increases)
REFERENCE 1st, Law of Thermodynamics; Bernoulli's Equation GGNS: OD-FH-504 (Obj 4) l, EIli .L-RG-666 3 f
' e s l
10 PRINCIPLES OF NUCLEAR POWER PLANT OPERATIONr, PAGE 34
~~~~ ~
TE5RE66 UdbEC57~55dT~TR5U5F5R 5U6~FL0i6 FL60 ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K. ANSWER 1.09 (1.00) 70% void fraction in the core (0.5) There is a larger % change in water volume for the same increase (3.45% vs 1 1%) (0.5) -OR- The voids produced at 70% VF have a larger effect on core reactivity since they are in an area of higher neutron flux. (0,5) REFERENCE EIH* Reactor Physics L/Pr pp 1.7-9, 10, 13; L-RO-604 BFNP Reactivity Coefficient LP, PP 2, 3; RO 85/03/01 ANSWER 1.10 (2.00)
~
- a. By observing the Full-in and Full-out travel lights (the operator could determine if geometric disto?cion had occured.
Inability to conduct full detector movement would indicate that internal misconfiguration had occured). (1.0)
- b. By observing the neutron level while moving the nuclear instrumentation. A significantly HIGHER (approximately 300 times) count rate would be seen for the UNVOIDED areas of the core as opposed to the VOIDED. (1.0)
REFERENCE EIH: L-RG-540 (M.9) (MCD) ANSWER 1.11 (1.25)
- 1. a
- 2. e:
- 3. d 4 b
- 5. c (0.25 each) t
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 35
~ ~~~~iEERE66YA551C5I~5555~iREU5E5R 506'_f[U16'f[6U ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
REFERENCE GGNS: MCD, 2.0-1, p 14 EIH: L-RO-650; L-RG-666 ANSWER 1.12 (1.00) b REFERENCE EIH' L-RO-667, p 10 BFNP: Rx Heat Balance LPi RO 85/03/05 ANSWER 1.13 (3.00)
- a. Increased Recire Flow from the Annulus. (0.5)
- b. Increase in Recire Flow (0.5)
- c. Recire Pump Runback ( Feedwater Flow < 20% ) (0.5)
- d. Increasing Reactor Power (0.5)
- e. FWCS response to Decreasing Water Level (0.5)
- f. EHC controlling Reactor Pressure (with TCV's) (0.5)
REFERENCE EIH: L-RG-732 (45) ANSWER 1.14 (2.00)
- a. 295 des F (+- 15 des F) (0.5)
- b. Increase (0.5)
- c. Increase (0.5)
- d. 450 psia (+- 50 psia) (0.5) i REFERENCE Steam Tables /Mollier Diagram
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 36
~ ~~~~isERs557sAsics! sEAi TRd55F5R dU6~ FLU 56~FL5E ANSWERS -- HATCH 182 -86/02/17-BROCKMAN, K.
ANSWER 1.15 (1.00) d REFERENCE ist and 2nd Laws of Thermodynamics BSEP* L/P 04-2/3-E, pp 7-15 EIH: L-RG-666, pp 15 - 18, Fig 6 ANSWER 1.16 (1.00) b REFERENCE DPC. Fundamentals of Nuclear Reactor Engineering, p. 96 001/000-K5.56 (2.8/3.1) GGNS: OP-NP-511 EIH: L-RO-602 ANSWER 1.17 (1.00)
- 1. Hf pins are located in the high flux regions of the control rods.
- 2. New roller and pin materials are used.
- 3. Thinner sheath materials are used to maintain the weight of the rods constant with the previous rods.
- 4. The last three pins in each blade of the cruciform are Hf, thus maintaining the control rod's worth effectively constant to previous.
- 5. Control Rod tube crack ing over design lifetime is reduced (structural lifetime is extended somewhat).
- 6. Nuclear lifetime is increased (due to Hf daughters having reasonably large thermal neutron absorbtion cross-sections).
- 7. Decrease in He (Alpha) build-up. (2 9 0.5 each)
REFERENCE EIH: L-RQ-605, pp 2, 3 l l i
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 37 ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
ANSWER 2.01 (1.00) a REFERENCE EIH: GPNT, Vol V, Chapter 5.1.II.B; L-RO-703 ANSWER 2.02 (1.00) b REFERENCE EIH: L-RG-733, pil, Fig 4 ANSWER 2.03 (2.00)
- a. (1) (1.0)
- b. (1) (Continue) AUTOMATICALLY (2) LOCAL Operator Action (3) CONTROL ROOM Operator Action (4) (Reinitiate) AUTOMATICALLY (0.25 each)
REFERENCE EIH GPNT, Vol VI, Chapter 8.1; L-RO-737 m e . . ,_ _
- r, ,, ., n .
tt tt TCD a REFEPENCE - o ._rn ~70 - - ANSWER 2.05 (1.00) e REFERENCE EIH: L-RQ-704
1 1
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 38 ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
ANSWER 2.06 (2.00)
- a. Low Nitrogen Pressure (of 965 psis) (0.25)
High Water Level in the instrument block (of 60 ml) (0.25) At the local control panel, the back lit button must be depressed. If the light goes out, the cause is wateri if the light stays lit, the cause is gas pressure. (0.5)
- 6. The CRD FCV is downstream of the flow element. (0.25) All of the indicated flow is gcing through the Charging line to recharge the accumulators. (0,5) The sensed high flow is sending a signal to close the FCV, (and thus Cooling Water flow is low). (0.25)
REFERENCE EIH: L-RG-719 ANSWER 2.07 (2.00)
- a. TRUE (0.5)
- b. TALSE (0.5)
- c. TRUE (0.5)
- d. TRUE (0.5)
REFERENCE EIHi L-RG-741 ANSWER 2.08 (1.00) The D/P between the bottom of the core (above the core plate) and the spray sparger pipe just outsid - of the RPV (downstream of the injection valve) is monitored. (0.5) If the integrity of the core spray header is lost, the D/P will INCREASE (0.5) REFERENCE EIHi GPNT, Vol VI, Chapter 8.21 L-RG-739
20 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 39 ANSWERS -- HATCH 182 -86/02/17-BROCKMAN, K. ANSWER 2.09 (1.00) Following the SRV's first actuation, the steam in the discharge line would condense, causing a vacuum in the line (0.33). This would result in Suppression Pool water being drawn up into the line (0.33) which could cause an overpressuri:stion of the line on the next actuation (0.33) - OR - causing water hammer or exces-sive jet forces (thrust) on the quencher (0.33). REFERENCE EIH NUREG/BR-005/Vol 5, No 4, Power Reactor Events, January, 1984, p5 l Event Summary NO 1.2 (Hatch Unit 2 - 08/25/02) ANSWER 2.10 (1.00) l B REFERENCE EIH GPNT, Vol. VI, Chapter 5.5-2; Vol. VII, Chapter 9.4-19, 20; L-RO-705 ANSWER 2 11 (1.00) l b i REFERENCE EIH* L-RG-725 (Fig. 5.3.1); L-RO-740 (7,24); GPNT, Vol.VI, Chapter 8.1 l ANSWER 2.12 (1.00) H<aters Trip F in Trips I2olation Valves (Dampers) Close (2 req'd 0 0.5 each ) l REFERENCE E 71! ' GPDT, Vol. V, Chapter 3.3-7, 9, 103 L-RO-748 (4,6) l _
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 40 ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
ANSWER 2.13 (1.00) That both the inlet (0.33) and outlet scram valves (0.33) for that rod are open (0.33). REFERENCE i BFNP: LPt29,P.6 EIH: GPNT, Vol VII, Chapter o.2.1, Chapter 9.3 l ANSWER 2.14 (1.00) d REFERENCE BFNP: LPt7,P. 28 EIH: L-RG-714, Fi 3ure 714-6; L-RO-714 GGNS SD B33-1, pp 5, 6i OP-833-1-501, p 5; ARI B33-FAL-L603A ANSWER 2.15 (1.50) I a. Reduce the concentration of oxygen for fire prevention considerations (0.5).
- b. Provide sufficient water for steam condensation (+0.5) l and enough air space for the primary containment non-condensibles (+0.5)
REFERENCE OFNP: Lesson Plan 16, Objectives C, D, H, I, pp 1, 14, 26, & 32 EIH: L-RO-707 ANSWER 2.16 (1.00)
- a. increase (+0.5)
- b. increase (&O.5)
REFERENCE l Gas Detector Theory BFNP: Lesson Plan 19, p. 19. EIH: L-RG-716 l l
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 41 ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
ANSWER 2.17 (2.00) l a. On / Green
- b. Off i
- c. On / Clear (Ind 2 0.53 Color 2 0.25)
REFERENCE EIH L-RG-704 (6) ANSWER 2.18 (1.50)
- a. Deep Well Pumps (0.5)
- 6. a (1.0)
REFERENCE l GGNS: OP-P64-501, p7 EIH: L-RG-745 ANSWER 2.19 (2.00)
- 1) Engine Overspeed
- 2) Low Lube Oil Pressure
- 3) Engine Start Failure
- 4) Differential Lockout (0.5 each)
REFERENCE l GGNS: OP-P75-501, p 18 EIH: L-RO-742 ANSWER 2.20 ( .50)
- FALSE i
L
l i . . i: l
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 42 ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
REFERENCE EIH: EIH Simulatori i l i l _ - . _ . . - - , , , . . - _ , _ , , , . , _ , , . , _,__,,,.,-.r---- -- ,=4 w- - - e-- v - - - - - - - rr --1--e- - - -v -, -- - - - , - - - - - - - - - - - - -
- 3. INSTRUMENTS AND CONTROLS PAGE 43 ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
ANSWER 3.01 (1.00) c REFERENCE EIH: GPNT, Vol VI, Chapter 7.2; EIH Simulatori L-RO-742 ANSWER 3.02 (1.00) d REFERENCE EIH: GPNT, Vol. VII, Chapter 9.2.1-4, 5, 11, 13; L-RO-723 ANSWER 3.03 (2.00) RCD - Solenoid control valve has energized (0.33); Energized (0.33) GREEEN - Power available to the solenoid control valve (0.33) Energized (0.33) AMCER c ressure in tailpipe (>85 psis) (0.33)
**nergized (0.33)
REFERENCE CIH GPNT, Vol. VI, Chapter 5.1.II.B.2i L-RO-703 AMSWER 3.04 (2.00)
- a. 2 upscale Hi Hi Hi radiation trips 1 opscale Hi Hi Hi radiation trip and 1 downscale/inop trip 2 downscale/inop trips - 1 from each channel (0.5 each)
- b. Off gas System Outlet and Drain valves, OR Discharge valve to the stack, cooler condenser and moisture separator drain valves. and holdup line drain valve. (0.5)
REFERENCE EIH; GPNT, Vol. VI, Chapter 6.8-21; Vol. VII, Chapter 9.7-1-7i L-RO-708; L-RG-746
i . . i
- 3. INSTRUMENTS AND CONTROLS PAGE 44 ANSWERS -- HATCH 1&2 -86/02/17-BROCKMANr K.
l I r ANSWER 3.05 (1.00) l
- e REFERENCE l EIH
- L-RG-7443 L-RO-752 (1984), DCR 79-442 ANSWER 3.06 (1.00) a REFERENCE EIH: GPNT, Vol. VII, Chapter 9.2.3-1,21 L-RO-722 ANSWER 3.07 (1.00)
- a. Decrease
- b. Decrease
- c. Decrease
- d. Decrease (0 25 each) l REFERENCE BFNP: BF-01-60, pp 28, 29 EIH: EIH Simulator, Malfunction 1363 L-RG-714 GGNS: SIM MAL 11 ANSWER 3.08 (1.00)
Both Systems' will Auto Swap to Torus Suction for Both Units. (1.0) REFERENCE EIH: L-RQ-7378 L-RO-740 l l l \
30 INSTRUMENTS AND CONTROLS PAGE 45 ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K. ANSWER 3.09 (1.00) All Control Valves (TCV's and ICV's) will Close (0.5) The Stop Valves (TSV's and IV's) will Close (0,5) REFERENCE EIH: L-RG-705 ANSWER 3.10 (1.50) Recireviation Pump RPT Breaker Permissive RSCS Bypass TSV, 10% Closure Scram Bypasss TCV, Fast Closure Scram Bypass (3 0 0.5 each) REFERENCE EIH: GPNT, Vol V, Chapter 4.1-9; GPNT, Vol VI, Chapter 5.5-10; GPNT, Vol VII, Chapters 9.2.2-11 and 9.2.3-5 ANSWER 3.11 (2.00)
- 1) APRM 'A' fails upscale -> relay K12A deenergizes (0.4)
- 2) -> NMS contacts K12A in RPS Trip Logic A1 open (0.4)
- 3) -> Relays K14A & E deenergize (0.4)
- 4) -> Contacts K14A & E open (0.4)
- 5) -3 Scram pilot solenoid valves for RPS A deenergize and vent (0.4)
REFERENCE BFNP L/P 128 EIH: L-RO-720, Fig 720-ta, -1b, -2a, -2b, -1a, -3b GPNT. Vol. VI, Chapter 9.3.1-2, 3, 4 BSEP: SSM 28-2-A
l \
- 3. INSTRUMENTS AND CONTROLS PAGE 46 ANSWERS -- HATCH 182 -86/02/17-BROCKMAN, K.
ANSWER 3.12 (2.00)
- i. e it. e REFERENCE BFNP: LPt12, p 24; TRANSIENT 820; 01-57, p 53; RO 85/01/02 EIH: L-RO-726 BSEP: RTN 026; H0 17-2/3-B, Section 3.2 GGNS* OP-C34-501 ANSWER 3.13 (1.00)
CHANNEL A Level A 0 Level B 2 Level C 4 Level 0 2 (0.25 each) REFERENCE EIH: GP t4 T Vol. VII, Chapter 9.1.3; L-RO-719 ANSWEP 3.14 (1.75)
- a. Yes CApply in accordance with part (b)] (0.5)
- b. It can be moved out one notch (0.25) before a withdraw error will block further movement. (0.25) If the rod was inserted, it will move as far as the operator wants (0.25) unless it is the third insert error. (0.25) If it is the third insert error, then i t u- >1d insert only one notch (0.25) (1.25)
REFERENCE EIH: GPNT, Volume VII, Chapter 9.2; L-RO-721 l l
- 3. INSTRUMENTS AND CONTROLS PAGE 47 ANSWERS -- HATCll 1&2 -86/02/17-BROCKMAN, K.
1 I'd
"' C " E R 3.15 (2.00) -
- a. ADS Valves are in Double Boxes (0,5) , . . -
- b. (1) No Demand (0.25) or Demand Mot.CO '25)
(2) ADS Armed (0.5) ~~~'
- c. 60 Minutes (0.25r'and 6 minutes (J.25)
REFERE#er _ . .-g . nn -~
- nn,s- n c n __ ; g .
ANSWER 3.16 (1.50) 11 min 51 see (Or 13 Min) (0.15) -OR- 1.85 i D/W Pressure (0.15)
- 121.5
- RWL 100 psis RHR/CS
+
- 8.5 RWL ,
2 min T.D. . (0.3 each) REFERENCE EIll: L-RO-767; 34AR-OPS-010-2 ANSWER 3.17 (1.00) The INOPERACLE Control Rod must be Bypassed (0.7) using the Remote Gypass Switch. (0.3) REFERENCE EIH: GPNT, Volume VII, Chapter 9.2.3; L-RO-722 l l l l 1 1
I l
- 3. INSTRUMENTS AND CONTROLS PAGE 48 ANSWERS - HATCH 1&2 -86/02/17-BROCKMAN, K.
ANSWER 3.18 (2.00)
- a. Unit 1
- b. Unit 2
- c. Neithet
- d. Unit 2 REFERENCE EIH L-RG-742
. - - _ _ . - - . ~ . . - - - . . - - _ - - . .-
j
- 4. PROCEDURES - NORMAL, ADNORMAL, EiiERGENCY AND PAGE 49 L ~~~~Rd656[U5565E~EUUTRUE~~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K. ANSWER 4.01 (2.00)
- a) 1. RPV level cannot be maintained (0.5) l 2. Suppression Pool water temperature cannot be maintained l
below 110 deg F (0.5) b) 1. Five (5), or more adjacent control rods not inserted below the 06 position (0.5)
- 2. Thirty (30), or more total control rods not inserted below the 06 position. (0.5)
REFERENCE EIH: 34AB-OPS-009-2 ANSWEP 4.02 (1.00) d REFERENCE EIH: 34AB-OPS-002-2 ANSWER 4.03 (1.00) To preclude establishing excessive stresses in the jet pumps due to flow reversals - OR - to prevent excessive radial bearing loads (on the recire pumps) REFERENCE EIH 34S0-B31-002-2; GPNT, Vol V, Chapter 4.1 AN3kER 4.04 (1.00) Cloth and Rubber Gloves (0.5) ; Cloth and Rubber Boots (0,5) RErEPENCE ' EIH: GET Handbook: 40AC-HPX-004-0 l l l l
l l l l
- 4. PROCEOURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 50
-~~~~~~~~~~~~~~~~~~~~~~~ ~~~~R5616E65fE5E~E6ATR L ANSWERS -- HATCH 1&2 -86/02/17-CROCKMAN, K.
ANSWER 4.05 (1.00)
- 1. Prevent Excessive Primary Containment Pressure (57.5 psig)
- 2. Control Combustible Gas Mixtures (HYDROGEN < 4 v/o and OXYGEN < 5 v/c) (0.5 each)
REFERENCE EIH 34AB-OPS-006-2 ANSWER 4.06 (1.00) To prevent hydrains the Recirculation Pump (to CRD System pressure) upon startup of the CRD pump. - OR - To preclude section and/or discharge valve actuator inoperability due to excessive dP. REFERENCE EIH: 34SO-C11-003-2; 34SO-031-002-2; L-RO-714 ANSWER 4.07 (2.00)
- s. 2 Minutes (0.5)
- b. Shallow (1.0)
- c. ' Print Notch Error
- function (of RWM) (0.5)
REFERENCE EIHi 42FH-ENG-010-1/2 ANSWEP 4.08 (2.00)
- 1. Inability of High Pressure Systems (HPCI/RCIC/CRD) to restore level.
- 2. One or more Low Pressure Systems running (LPCI, CS, or 1 CP & CDP)
- 3. Reactor Pressure > 350 psig.
- 4. Reactor Level < 121.5' (0.5 each)
r i L . . l I
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 51
~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~R565UL65EU5E~UU5TRUL ANSMERS -- HATCH 182 -86/02/17-BROCKMAN, K.
REFERENCE EIH: 34SO-821-001-25 34AB-OPS-002-2 ANSWER 4.09 (1.00)
- a. Unit 1 (0.4) Shift Supervisor (0.1)
- b. FALSE REFERENCE EIH: HNP 34AB-FPX-040-0, pp 1, 2 ANSWER 4.10 (1.00) c REFERENCE EIH: 20AC-MGR-003-OS, p 44 ANSWER 4.11 (1.00) a REFERENCE BFNP: OF-0I-66, pp 5-7 EIH: 34GO-0PS-001-2 BSEP: GP-02, pp 16, 17, 21 ANSWER 4.12 (1.00)
Turn the handwheel in the DESIRED POSITION direction (1/4 turn mart)(0.5); Verify the locking device integrity and proper installation by attempting to misposition the valve (0.5) REFERENCE EIH: 34GO-SUV-001-OS
l 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 52
- -----------------~~--~~~ ~~~~R565UL6G CdL E5 sir 5L ANSWERS -- HATCH 182 -86/02/17-BROCMMAN, K.
ANSWER 4.13 ( .50) None (0.5) REFERENCE
- EIH
- 30AC-OPS-001-OS AMSWER 4.14 (1.00)
- 1) R620 -> 604 A -:) -7' R621 -> 6040 -> -12' (0.25)
- 2) R604A = 26 - 7 = 19' (0.25)
- 3) R6048 = 23 -12 = 11' (0.25) l
- 4) Level = [(2) + (3)] / 2 = 15' (0.25)
REFERENCE EIH: 34AB-0PS-002-2; L-RO-712 ANSWER 4.15 (2.00)
- a. Alert
- b. (Notification of) Unusual Event .
- c. General Emergency
- d. Site Area Emer 3ency (0.5 each)
REFERENCE EIH GET Handbook, pp 57, 50, 60, 61; 10AC-MGR-006-OS BFNF: BFN-IPD, IP-1, p 1; RO 85/04/01
.._..--_._s.___ .- . . . _ . ~ , . _ . . . . - , . _ , , , . . . , _ , _ - . . _ _ _ , . , . . , . . . . _ _ - . . . _ _ _ __ ._
40 PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 53
~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~R565d[U5555E~d65TR6L ANSWERS -- HATCH 182 -86/02/17-DROCKMAN, K.
ANSWER 4.lo (1.00) Drywell Cooling Fans when CH-2] >= 2% (0.5) Drywell Spray if Pressure (0/W) >= 17 psig (0.25) and RVL is Normal (0.25) REFERENCE EIH: 34AB-OPS-002-2 n .i ANSWER 4.17 (M))
- a. 4 2436 CmWt (0.5)
- b. E err- 5 _- . s ---- - -
--(0.5)
- c. After a power adjustment with ra pumps, the operator eust run(5 consecutive)00-3's (0.5)(ctreulation at an interval >= 30 sec '=.25), cg
(-t,+) to ensure CTP ( m a :< i m u m ) is between 2420 & 2 4 3 6 C S W t . ' ^ 2 '_- ) ) . REFERENCE EIH: 42CC-ERP-009-2 <R1); 34GO-OPS-005-2 ANSWER 4.18 (2 00) Clearance Number Gequential Tag Number Component Identificatton Required Position of Component (0.5 each) REFERENCE EIH: 10AC-OPS-001-OS
i i i l \
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 54
~~~~Ed656E6656dE~66ETE6E~~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- HATCH 182 -06/02/17-BROCMMAN, K. ANSWER 4.19 (1.75)
- a. 5
- b. 7
- c. 4
- d. 6
- e. 2
- f. 3
- c. 1 (0.25 each)
REFERENCE EIH: L-RG-7158 10AC-MGR-003-OS
TEST CROSS REFERENCE PAGE 1 QUESTION VALUE REFERENCE 02.01 1.00 KEB0000582 01.02 1.00 MEB0000583 01.03 2.00 KEB0000504 01.04 1.00 KEB0000585 01 05 2.00 KEB0000586 01.06 2.00 KEB0000587 01.07 1.00 KEB0000508 01.08 2.00 KEB0000590 01.09 1.00 KED0000591 01 10 2.00 KEB0000592 01.11 1.25 KEB0000503 01.12 1.00 ME00000594 01.13 3.00 KE00000595 01.14 2.00 ME00000596 01.15 1.00 KEB0000597 01 16 1.00 - JNM627 0t.17 1.00 KE000004%8 25.25 02 01 1.00 F L'J 0 0 0 00 3 6 02.02 1.00 'Eim000t43 02.03 2.00 ME00000233
- n. e . c e- coo;5g 02.05 1 00 KEB000T;286 02.06 2.00 KEB0000288 i 02.07 2.00 KEB0000324 02.08 1.00 KEB0000325 02.09 1.00 KE00000326 02.10 1.00 KEB0000598 02.11 1.00 KE00000599 02 12 1.00 KEB0000600 02.13 1.00 KEB0000601 02.14 1.00 KE00000603 02 15 1.50 KED0000608 02.16 1.00 KE00000609 02 17 2.00 KEB0000614 02 18 1.50 KE300006L5 02.19 2.00 KED0000616 02 20 .50 KED0000629 25.50 03.01 1.00 KED0000032 03.02 1.00 KE30000003 03.03 2.00 MEB0000004 03.04 2.00 KE00000085 03.05 1.00 KEB0000172 03.06 1.00 KEB0000244 03.07 1 00 KEB0000610
F- s'
.a I
TEST CROSS REFERENCE PAGE 2 QUESTION VALUE REFERENCE 03.00 1.00 KEB0000611 03.09 1.00 KE80000612 03.10 1.50 KEB0000613 03.11 2.00 KEB0000617 03.12 2.00 KEB00006L8 03.13 1.00 KEB0000620 03.14 1.75 KE00000621 D .15 2,;o -EL GOdell - 03.16 1.50 KEB0000632 03.17 1.00 KED0000637 03.19 2,00 U t:D 00004:01 21.75 04.01 2.00 bE30000016 04.02 1.00 ME00n00066 04.03 1 00 KEB0000193 04.04 1.00 KEB0000239 04.05 1.00 MEB0000246 04.06 1.00 KEB0000247 04.07 2.00 KEB0000254 04.08 2.00 KEB0000266 04.09 1.00 KEB0000260 04.10 1.00 KEB0000272 04.11 1.00 KEB0000622 04.12 1.00 ME00000623 04.13 .50 NE00000624 04.14 1.00 KE80000625 04.15 2.00 KE80000626 04.16 1.00 KE00000633 04.17 1 %.00 UEU0000634 04.18 2.00 KE30000635 04 19 1.75 KED0000636 25.25 101.'5 __w
~ 6D _ 6
- M A .$ T E V -
. ENCLOSURE 3 U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: H A T Cll id2 9EACTOR TYPE: E: W R - G E 4 DATE ADMINISTERED: 06/02/17 EVAMINER: EROCKMAN, K.
APPLICANf: _ _ _ _ _ _ _ _ _ _ _ _ _ , _ _ _ _ _ _ _ _ . , INSTRUCTIONS TO APPLICANT! Use 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 et least ' 0 *. In each category and a final grade of at least 80%. E n.a m i n a t i o n papers will be picked uP ,:- hours after the examinatton start 3. '*W)
". O F CATEGORY % OF APPLICANT'S CATEGORY
/ALUE TOTAL SCORE UALUE CATEGORY
_b21 __ .. . _________ ________
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, 1:C A T TRANGFER AND FLUID FLOW
__ i __ _ _I ___________ ________ 2 PLANT DESLGN INCLUDING SAFETY
. AND EMERGENCY SYSTEMS
.3.73
.. i_ __ _ i ___________ ________
- 3. INCIPUMENIS Alic CONTROLS
,- 9 ne
__-_5 _-__ _'_"__'So__ __--_______ _____---
- 4. PPOCEDUREC - NURMALe AUNORMAL, CONTROL Lt SC 62,5^ 100.00 TOTAli FINAL GRADE _________________I All work done on this enaminatton ts my own. I have neither given nor received aid.
5PPL5C55iI5~5 5U5iUR5~~~~~~~~~~~~~~
. l
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2
~~~~IU5R5665Ud55C5 IUEdi~YE b5kER~h56"FLUEb~fLUU
'20ES T IO N 1.01 (1.00)
The THRESHOLD power below which PCI failures do not occur is krown to e DECREASE with fuel burnup. STATE two (2) reasons for this decrease in the PCI threshold. QUESTION 1.02 (1.00) The reactor trips from full power, equilibrium ':enon condi tions . Four (4) hours later the reactor is brought critical and power level is main-tained on range 5 of the IRMs for several hours. Which of the following
- t it ement + is CORRECT concerning control rod motion during this period?
- 3. Pods will have to be withdrawn due tu :enon build-in.
- b. Rods '4111 have to be rJpidly inserted sinco the critical reactor will cause a high r ite of : .3 n o n burnout.
- c. tods uill h2"o to be inserted since tenon will closely follou iM nurmal d e e r/ rate.
- d. ?,v h util a; pco tmately rsmain a- t- r the tenon establishes its a.zuil t h r io n reloe for t5: 3 p)oor level.
QUF5"!CH 1.63 /:.no)
- 3. STATE the design '<tature in the reactor vessel which ensures proper flow distribo', ion through the core fuel bundles. (0.5) l
- b. EMM.AIN how the rectrculation flow distribution would react during a ,c o v e r inctw :e by rod pull .f this design feature were NOT PRESENT.
INCO'SC Ill i ' P U re 9tgpqNSc THE REASON (S) FOR THIS REACTION. (1.5) t l l l I
'***** CAT:GOGY 01 CONTINUCD ON NEXT PAGE *****)
i l
- 1. Pe!NCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3
~~~~iUEE5607Ud55C37"UEdi~ Teds 5 FEE ~dsU~iLU56"EL E QUESTION 1.04 (2.00)
- a. DEFINE 'Crtticil Power *. (1.0) a,
!J h i c h one of the followin3 conditions would tend to INCREASE the Cett'c31 Poocr level insuminj all other variables remain unchanged? (1.0)
NOTE: AG3UME NORMAL FULL -POWER OPER ATING CONDITIONS
- 1. Inlet subcooling 13 DECREASED
- 2. Reactor pressure is DECREASED
- 1. The a: t 31 ,
, o w .t r pe n is RAISED
- 4. Coolant flow '
<te is DECREASED QUESTION 1,OS (1,noi Which of the following radioactive isotopes found in the reactor caelant WOULD NOT indicate a leak through the fuel c ] .ni d i n 3
- 2. Co - 60
- b. Xe - 133
,_ ! - 131
- d. Vr - 37 OUESTION 1.Os (2.00)
As part of the scram procedure, the operator is directed to innert the GRM'; 3nd IRM's.
- a. Followinj a severe LOCA, EXPLAIN how these systems could be
.> sed to detect gross core damaje. (1.0)
- b. EXPLAIN how these systen.s could be used to provide a crude indication of water level if level could not be confirmod by norm;:1 instrumentation. (1 0)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4
-- inisa66 sAsics- sEEi isissFEE As6 EcGi6 FC60 GUESTION 1.07 (1.00)
A reactor heat balance was performed (by hand) during the 00-08 shift due to the Process Computer being 00C. The GAF's were computed, but the APRM GAIN ADJUSTMENTS HAVE NOT BEEN MADE. Which ONE of the following statements is TRUE concerning reactor power?
- a. If the feedwater flow rate used in the heat balance calcu-lation was LOWER than the actual feedwater flow rate, then the actual power is HIGHER than the currently calculated power,
- b. If the reactor recirculation pump heat input used in the heat balance calculation was OMITTED. then the actual power is LOWER than the currently caleviated power.
- c. If the ste ui flow used in the heat i.mlance calculation was LOWER than the actual , team flow. then the actu31 power is LOWER than the currently esleulated power.
{ d. If the TACU taturn temperature v re d in the heat balance cal-ov13 tion w a s '_g u C 9 th an the , e t o :51 RPCU return tenperature, then the acto;1 ,ower
) la !I I C '!E:- than the cutrently calculated power.
QUESTION 1. 0 '3 (3.00) I Attached Fig'.re 65?5 represents a transien' that could occur at a. BWR. I Civen: (1) Master Recirc Flow Controller Fails High
." 2 i 90 operator a c t i o n .s occur (2) Recorder Speed = 1 division = 1 minute i
EXPLAIN the cause(s) cf the fallowing recorder indications;
- a. Teactoc u2ter ' owl DECREASE (Point A) (0.5)
L. , Reactor Power INCOCAgr (Point B) (0.5) ! c. Core Flow DECREASE (Point C) (0.S) di 79 x tor c' r e u s e r e INCor.CC (Point D) (0,5) e, Total "eed a s ter F'ou IK 9 EASE (Point E) (0.5)
- r. Tot,1 "; p 31 12 ., " 1 r " C A "- (Point F) (0,5) i m s* CATEGORY 01 CONTINUED ON NEXT PAGE xxxxx)
. l V
t
- 1. OPINDI'PLES OF NtJC'_E AR POWER PL ANT OPERATION, PACC 5
~~ ~~II5555C675555C 5I~H55T~TR555555~ 56~ELU U~EL65 l
{ QUESTION 1.09 (2.00) f Ouring your Shift. an SRV inadvertantly opens from 100% power an 1000 psia. Use a Mollier Diagram or the Steam Tables to answer the following*
- a. STATE the tailpipe temperature, assuming atmospheric pressure in the Suppression Pool and No Reactor Depressurization. (0.5)
- b. If the Suppression Pool Pressure were to INCREASE, STATE whether the j Tailpipe Temperature would INCREASE, DECREASE, or 3EMAIN THE SAME. (0.5)
- c. If the reactor is depressurized when the SRV is opened. STATE whether the Tailpipe Temperature will INITIALLY INCREASE, DECREASE, or PEMAIN THE SAME. (0.5)
- d. STATE the Reactor Pressure at which the Tailpipe Temperature would be at its MAXIMUM value (during the depressurization). (0.5)
(ASSUME A SATURATED SYSTEM AND INSTANTANEOUS HEAT TRANSFER) l l QUESTION 1.10 (1.00) LIST two (2) Design Char-acteristics which allow for the use of Hybrid Control Pods? l l ( *r*** END OF CATEGORY 01 *****)
O
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 6 00ESTION 2.01 (1.00)
The Unit 2 Vital AC Power 120/240 v Distribution Cabinet 2A is normally supplied from 600 v Bus 2D through a Battery Charger and a Static Inverter. If the Static Inverter fails ...(CHOOSE ONE)
- a. ... the 125 Yde battery will maintain power to the t> ital AC Cabinet for up to 6 hours.
- b. ...the power supply can be manually transfered to the alternate 400 v Bus 2C / Vital AC Transformer 2A by depressing a transfer PB.
c- ...the pouer supply will automatically transfer to the alternate 600 v Bus 2C < Vital AC Transformer 2A. ts ...the pooer si'ppl, can be manually transfered to the alternate A00 v 90: 20 alternate Slette Inverter by depressing a transfer PB. tr*r** CATECDPY 02 C0ilTIHUED ON NEXT PAGE ****s) l
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7 GUESTION 2,02 (2,00) uith regard to the Unit 2 RCIC Gystem
- 3. Which of the following is the only normally CLOSED valve in the RCIC cteam supply flow path in the STAND-0Y lineup? (1.0)
(1) Steam Supply U21ve (r045) (2) Outboard Steam Isolation Valve (F008) (3) Trip Throttle Valve (4) Governor u a lve
- b. For each of the situations listed below, STATE whether final RCIC injection into the reactor would continue or reinitiate AUTOMATICALLY, would require CONTROL ROOM Operator Action, or would require LOCAL Operator action. Assume that RCIC had auto initia+.ed prior to any of the following situations.
(1) The Test Bypass Valve to the CST (F022) FAILS OPEN. (0.25)
'2) A 125% Overapeed Trip is received due to low control oil pressure. Corit, el oil pressure is then returned to normal. (0.25)
(3) After decre asing to 30 psig. RCIC Steam Line Pressure incre:ces to 150 psig. (0.25) (2) After increcstog to +60 inches, Reactor vessel 'A a t e r (0.25)
-;uel decree m to -60 inches.
(***r* C ATCOOG:Y 02 CONTINUCD ON NEXT PACE *****)
2 o' . A N ' DESIGN INCLUDING '3 AFCT Y AND EMERGENCY S Y'3 T E M S PAGE O QUESTION 2.03 (2.00) Concerning the Unit 2 RHR System, while operatin3 at 100% RTP. For each of the following situations, STATE whether the statement i- TRUE or FALSE.
- 2. The Inboard Inj-ction Valve (F015) may be opened ONLY if the Outboard Injectioi Valve (F017) is shut. (0.5)
S. The In'eard Canta:nsent Spray 'J a i v e (F021) CANNOT be opened l oii l e ., s tha Gotbcard "alve (F016) is shut. (0.5) {
. The 9 h v i '.w r. C.olin3 Isolation 'J a l v e s (F008 i F009) are t titeoloM od .ho' oith the plant in these conditions. (0.5)
T;. r.o .i- I .1. ti..a -21ve (F02d), the Toros S p r a y ') 31 v e en ~ , ,:,.; t '.- r . :. . . . - c _."r . i t n .3 -/(F020) e i v emay all be in the n qN :.o; i ; o.. U.. . ;i. e ttme. (0,5) oOEST:an 2 na n n '. 3
- A rT h.m 4 :.e .o t e ;, .t, , , ' ' the Core Sprey pipitig between the i 7 ' cnd the Core Shroud is confirmed. and how the indication (s) l
- u l. d :htnge if the integrity was LOST. l l
l 70:P.rInN :,05 <1,00s j cr :.A 9 400/44Y an S P ') discharge pipe Itailptpe) could be
.io ed due to its racuus breder STICKING SHUT during repeated i
.E' o n s, el:rting'setestingi of the S R ') .
l (***** CATICORY 02 CONTINUED ON NEXT PAGE *****)
2, PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 9 GUESTION 2.06 (1 004 uhich of the following lists of responses (a - d) CURRECTLY rf e t e r i b e s the main turbine's response to an OVERSPEED condition? RESPONSES SPEEDS (a) (b) (c) (d) 100% " Master
- All ICV's All ICV's ' Master' ICV's begin bejin to begin to ICV's begin to throttle throttle throttle to throttle
- Slave ' Slave' ICV's begin ICV's begin to throttle to throttle All ICV's All ICV's All ICV's All ICV's tvil Clo ed roll Closed Full Closed Full Closed 110% Mechanic il Electrical Mechanical Electrical O'Spd Trip O'Spd Trip O'Spd Trip O'Spd Trip 111.S% Ilectric+1 Machenical Electr: cal Mechanical O'Spd Tr ip O'Spd Trip O'Spd Trip O'Spd Trip 00ESTION 2 07 (1.00)
The F u l l o r e Otaplay on the center panel has a BLUE scram light for each control rdd. OESCPI3E uhet is DIRECTLY indicated When this light is i l l u m i n a '. o d . (**2** CATE60PY 02 CONTINUED 04 NEXT PAGE *2:**) i i 1
- 2. PLANT DESIGN INGLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 10 GUESTION 2.03 (1.00)
The plant is operating normally at power when Pump A Controlled Leakage (FS ' A ' elarms LO (0 1 ;pui and you note an INCREASE in No.2 Recire Pump se31 preswre with NO CHANGE in No. 1 seal pressure. Which of the followir.3 feilures would e euw these indications? 3 F;ilur+ of No, *
, ;9al t- r: 11.re of Un. tal c "1.ojginj of +he 'c.
- internal restricting / breakdown orifice
- gg.iig of 'he o 2 internal restricting / breakdown orifice NOTE' NO OTHER ALArd0 &: "_ MESENT O'J E S ' I O i; 2.n9 (^ nn The disp 1 f - ' " the indicator lights for ATTS can indicate that the system is in its normal operating mode with NO ALARM CONDITIONS present? For each of the following STATE the status (ON/0FF) and the indicator color (If ON) when in this normal mode.
NOTE: FIGURES 9,9(3) AND o.0(6) PROVIDED FOR REFERENCE a> STATUS /hTU Indicator
-- SPOSS FAIL /MTU Indicator
- c. P0uER/P925 Indic tor
(***** CATECOPY 02 CONTINUED ON NEXT PAGE r****)
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 11 00ESTION 2.10 (1.50)
With regard to the diesel fire pumps
- a. STATE the source of feed to the Fire Protection Storage Water ranks, given that the Fire Mater System has initiated. (0.5)
- b. which one of the following is the initiation signal for the diesel driven fire pumps? (1.0)
- a. Low Fire Water Header Pressure
- b. Low Air Pressure (Dry Pipe Systems)
- c. High Fire Water Header Flow Rate
- d. S:noke Detector Actuation 00ESTION 2.11 (2.00)
LIST the four (4) DG iutomatic shutdown signals that remain OPERABLE if the diesel te started by using the local Emergency Star t PB. NOTE: CONSIDER DIESEL GENERATOR 2A (***** END OF CATEGORY 02 *****)
- 3. INSTRUMENTS AND; CONTROLS PAGE 12 GUESTION 3.01 (1.00)
Given: Unit 2 in control of D/G *B' D/G *B' Mode Switch in TEST (Surveillance being performed) Electrical distribution NORMAL (Full Power Lineup) D/G *B' is at rated speed and voltage, but not synchronized, when power is lost to 4160 volt Bus 2F. Which of the following accurately describes the system operation?
- a. Bus 2F can be powered by D/G 'B' when the operator takes the Output Br eaker Switch to CLOSE and has the SYNC SCOPE activated.
- b. Bus 2F will be powered by D/G *B' automatically, after 12 seconds; appropriate loads will be picked up sequentially.
- c. Bus 2F can not be powered by D/G 'B' while it is in the TEST mode, given these conditions.
d' ?.,3 ?F can be powered by D/G 'O' when the operator resets the Lockout 9elay, ecti"ates the SYNC GCOPE, and takes the Output Breaker t.o CLOSE. QUESTION ?. 3 2 (2.001 Regarfing the SRV's and the associ:sted Low Low Set (LLS) logic: There are three lights associated with each GRV - RED. GREEN, and AMBER. D'. P '. A I .*1 what each of the different colored lights indicates,- and SMTE uhe',her each would be energized or de-energized during the t t. e :t= 7:') wa. open as a result of reactor pressure reaching the 5 7 '.' ' = 'elief :etpoint.
- A'3 m t a C T AT THE ') ALVES IN Q'J ES TIO M A 't E L ' S '! A L 'J E S -
AMD '!E'tE AC TU ATED BY THE LLS FUNCTION
/r7* :x C AT C00 r:Y O '] CONTINUED ON NEXT PAGE *****)
O
- 3. INSTRUi1ENTS AND: CONTROLS PAGE 13 GUESTION 3.03 (1.00)
STATE the TPENO (INCREASE, DECREASE. NO CHANGE) that you would expect to see as a result of a " Jet Pump Riser Failure" for each of the following.
- a. Failed Jet Pump Flow.
- b. Core Differential Pre;svre.
- c. Reactor (APRM) Power.
- d. Indicated Core Flow.
QUESTION 3.04 (1.50) The Main Turbine first stage pressure switches provide permissives and/or control signals for several plant functions. LIST three (3) of these permissives/ control functions. QUESTION 3.05 (2,00) Unit 2 is operating at 100% RTP when APRM 'A' falls upscale and results in a reactor half-scram. Utilizing the attached RPS trip logic diagrams (Figures 8224 A thru F) DESCRIDE in a STEP-BY-STEP fashion (with regard to the opening / closing, ener-gizing/deenerginin3 of ALL applicable contacts and relays) how the APRM vpscale trip results in an actuation of the scram solenoid valves. NOTE: IF THE ATTACHED DIAGRAMS CAN NOT DE EASILY READ, ASSIGN THE CONTACTS / RELAYS. ETC NUMDERS AND REFCR TO THEM IN YOUR ANSWER. QUESTION 3.06 (1.75) Concerning the Rod 'A o r t h M i n i m i z a r ( U! M ) on Unit 2:
- 3. When a select error occurs on the R !J M . STATE whether the operator can still move the rod, tYES or NO) ASSUME THE R'Jti IS NOT SYPASSED HO ROD RLOCUS EXIST PRIOP TO SELECTING THC ROD. (0.5)
S. E!"". A I N the bases for your decision in part (a). Consider in jour m:pi nna tion bo t;, m 2 *,t. eiiip t e d insert and withdraw action. (1.25) (**rra CATEGORY 03 CoilTINUCD ON HEXT PAGE *****)
~
3 INSTRUMENTS AND-CONTRC. PAGE 14 __________________________'_'3 _ D c LL 1 E b y s The So0S Primary Display provides inf ormation concernin- . e ADS, amongst other parameters. I DESCt: ICE how (he A03 '>alves are differa . ated from the other S'V's+ (0.5)
- 6. STAT the sijnific+nce of +' e following COLORSr as they are used for indication of AD" alve status: (1.0)
(1) Green (2) 0 _nge
- c. ST the two (2) time periods over which TREND Displays 2: c _. >:' (0.5)
GUESTION 3.08 (1.50) LIST ALL requirements which could/must be met in order to actuate ADS. (Automatic Actuation) GUESTION 3.00 (1.00) A Control Rod is declared INOPERABLE and valved out of service on Unit 2. DESCRIBE what must be done to the RSCS logic to allow continued rod withdrawal. A BLACK-AND-WHITE ROD PATTERN HAS NOT YET SEEN ACHIEVED IN T'IE ST ARTUP SEQUENCE. (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
- 3. INSTRUMENTS AND -CONTROLS PAGC is QUESTION 3.10 (2.00)
STATE for each of the following conditions, which Unit (if any) will recieve the 19 EDG.
- a. Select Switch Position -
Unit 1 Unit 1 Condition - LOCA/LOSP Unit 2 Condition - LOSP
- b. Select Switch Position -
Unit i Untt 1 Condition - LOSP Unit 2 Condition - LOCA/LOSP
- c. Select Switch Position -
Unit 2 Unit 1 Condition - LOCA/LOSP Unit 2 Condition - LOCA/LOSP
- d. Select Switch Position -
Unit 2 Unit 1 Condition - LOCA Unit 2 Condition - LOSP (***** END OF CATEGORY 03 *****) l l
t k
- 4. PROCEDURES - NORMAL, AGNORMAL. EMERGENCY AND PAGE 16
~~~~ ~ ~~~~~~~~~~~~~~~~~~~~~~~~
R565UL6G555L b NTR L QUESTION 4.01 (2.00) 34AB-OPS-009-2, ' Inability to Shutdown With Control Rods', states that'
... If at any time, either condition b(1) or b(2) exists, and either _____(a.1) ____
or ____ ( e.2} ____, and if it is obvious that the reactor cannot be shutdown and, in the judgment of the Shift Supervisor, or in his absence, a licensed operator, a hazard exists to the environs, personnel, or the plant, utiltre the standby liquid control system per 34SO-C41-001-2.'
- a. LIST conditions (3.1) and (a.2). (1.0)
- b. LIST conditions b(1) and b(2). (1.0)
GUESTION 4.02 (1.00) 34AB-OPS-002-2, " Pipe Dreak Inside Primary Containment', lists numerous conditions indicative of a break. Which one of the following supports the suspicion of a SMALL break inside con-tainment.
- 3. Decrease in reactor water leve1 ; pressure and/or temperature in-crease in Drywelli airborne activ2ty increase in Drywelli increased D4FDS operating frequency.
- b. Decrease in reactor pressure; pressure and/or temperature increase in Drywall: ganerator load decreasei D4FDS high level.
- c. 9esciu' Cer am from low water level; pressure and/or temperature increase ir Orywelli generator load decrease; increased DWFDS u p e r s ', i n g f r aquer.cy ,
d Reactor Scram fcem high Drywell pressure; generator load decrease; airborne activity increase in Drywelli DWFDS high level. ( v x
- rx CATESOPY 04 CONTINUED ON NEXT PAGE *****)
1 I l l I
- 4. PROCEjuRES - N06 MAL, ADNGRMAL, EMCRGENCY AND PAGE 17
~~~~E5050L6656IL C6UTRUL' " "~ "'" ~
QUESTION 4.03 (1.001 You enter an area posted with the following sign: _________ i CAUTION I l_________t i I Your purpose is to conduct a normal l I S u r v e i ll a r.c e inspection. I 1 I I ICantamin.1 l___Arec. I LIST the minimum protective clothing requirements which an RWP for this area could require. QUESTION 4.04 (2.00) 42FH-ENG-010-2, ' Control Rod Movement', provides numerous TIANDARD PRACTICES which apply when rods are bein<3 moved for the purpose of changin3 power level. 7 With LHGR , 8 Kw/ft, STATE the parlod of time you must wait betseen successive notch withdrawals of the same rod. (0.5)
- b. With HIGH POWER and HIGil FLOW conditions, STATE which type of rod - SHALLOW, DEEP, CENTRAL, or EDGE - should NOT be with-drawn nor inserted. (1.0)
- c. With power < 30%, STATE the verification which is utilized when latching the first rod in any group. (0.5)
(***** CriEGORY 04 CONTINUED ON NEXT PAGE *****) l l 1
i,
- 4. FROCEDURES - N0kMAL, ADNORMAL, EMERGENCY AND PAGE 18
~~~~Rh656LU5iCAL E6NTR L'~~~~~~~~~~~~~~~~~~~~~~~
GUESTION 4.05 (2.00) A pipe break occurs and results in a decreasing reactor water level and decreasing cactor pressure. LIST the four (4) criteria which would require the MANUAL initiation of ADS, given that auto-initiation has not occured. QUESTION 4.06 (1.00) 34GO-SUV-001-05, " Control and Surveillance of Locked Valves', provides guideance concerning the proper way to confirm and verify a locked valve position. DESCRIBE the proper way to CONFIRM Locked Valve p'sitioning and Locking Device integrity. QUESTION 4.07 ( .50) 30AC-OPS-001-OS, ' Equipment Clearance and Tagging', states *
'Under no circumstances will equip-ment be operated or manipulated when tagged with a HOLD tag.
LIST any exceptions to this statement. (If no exceptions exist, state so!) QUESTION 4.08 (2.00) STATE which Emergency Classification is appropriate for the following definitions.
- a. Events are in progress or have occurred which involve actual or potential substantial degradation of the level of aafety of the plant.
- b. Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant.
- c. Events are in progress or have occurred which involve actual or imminent substantiel core degradation or melting with the potential for loss of containment integrity.
- d. Events are in progress or have occurred which involve an actual or likely major failure of plant functions needed for protection of the public.
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
40 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 19
~------- --- EE5i5E55iEEE E5 sis 5E-----~~--------
QUESTION 4.09 (1.00) Concerning 34AB-OPS-002-2, " Pipe Break Inside Primary Containment *:
- a. STATE when Drywell Cooling Fans are REQUIRED to be placed in service. (0.5)
- b. STATE when Drywell Spray is REQUIRED to be initiated. (0.5)
QUESTION 4 10 (2.00) LIST four (4) pieces of information which mus' be entered onto a RED TAG in order for it to be properly hung. (Information which an operator must write on the tas!i GUESTION 4.11 (t.75) The Procedure Numbering System consists of a twelve enaracter alpha-numeric designator, with three dashes (e.g., 3100-E11-034-1S). This catagorizes each procedure in seven ways. MATCH the Column A characters (using the example above) with the Column B information, in order to most securately describe the numbering system. COLUMMN A COLUMN B
- a. 1st (3) 1. Safety Status
- b. 2nd (1) 2. Sequence Number
- c. Ord/4th (50) 3. Unit Applicability
- d. 5th/6th/7th (Ell) 4. Procedure Type e, oth/9th/10th (034) 5. Responsible Department
- f. 11th (1) 6. System Identification 3 12th (S) 7. Responsible Section
(***** END OF CATEGORY 04 *****) (********'<**** END OF EXAMINATION ***************)
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. I E. EE6 CIDE _LEWL _BE601tN _GQBBEGI19S_EP__!eB9eX_iti3Ie wtJteIn
- 1. O,b t a i : Orys. ell T wperature Re,dirige (we tiie 185' Clevatspn.
Temperature Indicator ET4' RTO givv. this ie. ding for the l north side end ET47-RC21 J i'r e s this resdirig for the south I gide. C( 2 R*fer to riguae 1 at:d obtain the cor redion f actor for each temperature reading. l 3, 4 the correction f.ctor f ound usirig tise North Orywell Temperature Reading the !avel it.Jicataun shown on Level Indirator 2021-RE.O'A. . 4, 5- the correction facter f our.d us ..ig time South Drywell , T ee,.:.c atur e Pe e ding 16. level ... . . tion sliown ori Level
*Irifice4or 2001 FC04D.
C. The aver age of thes.e twe Laie=ssew level re dangs will give
*.ier t.evea.
e 9 t . 3 . s t ir..a t e o f the A.: t .:31 G =.c iu. I
, C t L _ c.o - ,,e_ d % t ,.A,:. c AW\Ttoe o, som eA c.Ts oc 4 Th,_ kncw v. muw \ m c.
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Q\uw,,5 b sk. FIGURE FOR 7.07
y, -
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c. GeorgiaPower 4 7 or 7
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nli ... ,.. l l l 8 i!l l l ' .I_ . 0 . i iie I ! 1 I l . I I i. i. 140- ISO 220 - 260 300 - 0 DRnfEf.L TEMPE AATURE AT 10!,' EI EVATIOE ("r) [4 A [ ! V / '/ teitrente enll .i FIGURE FOR 7.07 0 _ . . T. 7J . ~..iT." l
10 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 20
--- iAEER56isERI5s-~sEEi iEEsiEEE Es5 FE0i5 FE60 ANSWERS -- HATCH 182 -86/02/17-BROCKMAN, K.
ANSWER 1.01 (1.00)
- 1. Neutron embrittlement of the cladding.
- 2. Thermally induced pellet growth.
- 3. Inward motion of the cladding walls (creepdown).
- 4. Fission Product (Chemical) embrittlement of the cladding (from Cd/I). (2 0 0.5 each)
REFEPENCE GGNS* MCD, PCIOMRr p7 EIH; L-RO-673 ANSWER 1.02 II.00) e REFERC4CE 9FNR; XENON & SAMARIUM LP, pp 4, 12 GGNS* OP-NP-514 EIH L-RG-606 ANSWER 1.03 (2.00)
- a. Grificed Fuel Support Pieces (Core Grificing) (0.5)
- b. As power increases, the flow to the central (higher powered) bundles vould decrease (0.25); flow to the peripheral (lower powered) bundles oculd increase (0.25). This is due to the increased two-phase flow recistance that is developed in the higher powered bundles where there is greater boiling (0.5) and the resultant ' restriction
- that this would pose to flow, thus sending it preferentially through the louer powered bundles (0.5). (1.5)
REFEPENCC General Electric NEDE 24810, September 1903 GGit0* 'hermal Limits LP, p 20 EIlli GPHT. Vol u, Chapter 2.2.J l J
t
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 21
--- iREER557sERIC5- REEi isissFEs Es5 FEGi5 FE5s ANSWERS -- HATCH 1a2 -86/02/17-BROCKMAN, K.
ANSWER 1.04 (2.00)
- 3. The assembly power which would cause the onset of transition boiling at some point in the assembly. (1.0)
- b. 2 REFERENCE RFMP* TRANSITION DOILING & ATLAS TESTING LP,P.5-6 GEYL CORRELATION !. CRITICAL POWER LP,P.3 GGNS: MCD- Tl:ERMAL LIMITS, P.26,32-33 EI:t * '
_ -RG-67 2 AMP,NER 1.05 (1.00) 3 REFERENCE BFNP DFN Mitigating Ry Core Damage, pp 17 - 10; RO 85/02/01 GGNS: MCD, SECTION 4.4 E Ot: L-PO-530 (MCD) At8G9ER 1.06 (2.00)
.e . By observing the Full-in and Full-out travel lights (the operator could determine if geometric distortion had occured.
Inability to conduct full detector movement would indicate th=t internal misconfiguration had occured). (1.0)
- b. By observing the neutron level chile moving the nuclear instrumentation. A significantly HIGHER (approximately 300 times) count rate would be seen for the UNVOIDED areas of the core as opposed to the VOIDED. (1.0)
REFERENCE EIH; L-RG-540 (M.0) (MCD) l i l
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATIONr PAGE 22
--- iREEF557 SERIES- REEi iEAssFEE As5 FEUi5 FE5E ANSWERS -- HATCH 1&2 -86/02/17-BROCKMANr K.
ANSWER 1.07 (1.00) b REFERENCE EIH* L-RG-667, p 10 BFNP' R:< He a t Balance LP: RO 05/03/05 ANSWER 1.00 (3.00)
- a. Increased Recire Flow from the Annulus. (0.5)
- 6. Increase in Rectre Flow (0.5)
- c. Recirc Pump Runback ( Feedwater Flow < 20% ) (0.5)
- d. Increasing Reactor Power (0.5)
- e. FWCS response to Decreasing Water Level (0.5)
- f. EHC controlling Re4ctor Pressure (with TCV's) (0.5)
REFERENCE EIH; L-RG-732 (45) ANSWER 1.09 (2.00)
- 3. 295 deg F (4- 15 deg F) (0.5)
- b. Increase (0.5)
- c. Increase (0,5)
- d. 450 psia (+- 50 psia) (0.5)
REFERENCE Steam Tables /Mollier Diagram
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 23
~ ~~~~IUERU66YU555C5,~UE5T~IR5U5EER 5U6~ELU56~iL6U ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
ANSWER 1.10 (1.00)
- 1. Hf pins are located in the high f l u :. regions of the control rods.
- 2. New roller and pin materials are used.
- 3. Thinner sheath materials are used to maintain the weight of the rods constant with the previous rods.
- 4. The last three pins in each blade of the cruciform are Hf, thus maintaining the control rod's worth effectively constant to previous.
- 5. Control Rod tube cracking over design lifetime is reduced (structural lifetime is evtended somewhat).
- 6. Nuclear lifetime is increased (due to Hf daughters having reasonable large thermal neutron absorbtion cross-sections).
- 7. Decrease in He (A1:her buildup. (2 0 0.5 each)
REFERENCE Elll
- L-RQ-605, pp 2, 3 1
I l l l l I ! l l
1 I l ! 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 24 ANSWERS -- HATCH 182 -86/02/17-BROCKMAN, K. ANSWER 2.01 (1.00) b REFERENCE l L Ill
- L-PG-733, pli, Fig 4 ANSWER 2.02 (2.00)
- a. (1) (1.0)
- b. (1) (Continve) AUTOMATICALLY (2) LOCAL Operator Action (3) CONTROL ROOM Operator Action (4) (Reinitiate) AUTOMATICALLY (0.25 each)
REFERENCE EIli: GPNT, Vol VI, Chapter 8.1; L-RG-737 ANSWER 2.03 (2,00) l
- a. TRUC (0.S)
- b. FALSE (0.5)
- c. TRUE (0.5)
- d. TRUE (0.5)
REFERENCE EIll: L-RQ-741 ANSWER 2.04 (1.00) The D/P between the bottom of the core (above the core plate) and the spray sparser pipe just outside of the RPV (downstream of the injection valve) is monitored. (0.5) If the integrity of the core spray header is lost, the D/P will INCREASE (0.5) ! REFERENCC EIH GPNT, Vol VI, Chapter 0.23 L-RG-739 i
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25 ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
ANSWER 2.05 (1.00) Followins the SRV's first actuation, the steam in the discharge line would condense, causing a vacuum in the line (0.33). This would result in Suppressior. Pool water being drawn up into the line (0.33) which could cause an overpressurization of the line on the next actuation (0.33) - OR - causing water hammer or exces-sive jet forces (thrust) on the quencher (0.33). REFERENCE EIH: NUREG/BR-005/Vol 5 No 4, Power Reactor Events, January, 1984, p5 Event Summary NO 1.2 (Hetch Unit 2 - 08/25/82) ANSWER 2.06 (1.00) a REFERENCE EIH: GPNT. Vol. VI. Chapter 5 5-21 Vol. VII, Chapter 9.4-19, 20; L-RQ-705 ANSWER 2.07 (1.00) That both the inlet <0.33) and outlet scram valves (0.33) for that rod are open (0.33). REFERENCE
'3 F N P : LPt29.D.6 EI'1: GPNT, 'l o l VII. Chapter 9.2.1, Chapter 9.3 ANSWER 2.08 (1.00) d REFERENCE OFNP: LPt7,P. 28 EIH' L-RG-714, Figure 714-6; L-RO-714 GGNS SD B33-1, pp 5, 6; OP-833-1-501, p 5; ARI B33-FAL-L603A
- 2. PLANT DESIGN INGLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 26 ANSWERS -- HATCH 1&2 -8 6 / 02 /17 -E:R O C K M A N , K.
ANSWER 2.09 (2.00)
- a. On / Green
- b. Off
- c. On / Clear (Ind 0 0.5; Color 0 0.25)
REFERENCE EIH: L-RG-704 (6) ANSWER 2.10 (1.50)
- a. Deep Well Pumps (0.5) b+ a (1.0)
REFERENCE l GGNS: O P -P 6 4 -5 01, P 7 E Ill: L-RO-745 l AnsuER 2.11 (2.00) l l l' Engine Overspeed j :) Low Lube Oil Pressure
- 3) Engine Start Failure
- 4) Differential Lockout (0.5 each)
REFERENCE GGNS: OP-P75-501, p 18 EIf t : L-RQ-742 l I l l l l I
O
- 3. INSTRUMENTS AND; CONTROLS PAGE 27 ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
ANSWER 3.01 (1.00) C REFERENCE EIH GPNT, Vol VI, Chapter 7.23 EIH Simulatori L-RG-742 ANSWER 3.02 (2.00) RED - Solenoid control valve has energized (0.33); Energized (0.33) GREEEN - Power available to the solenoid control valve (0.33) Energized (0.33) AMorP - Pressure in tailpipe (>G5 psig) (0.33) Energized (0.33) PEFERENCE EIH: GPNT, Vol. VI. Chapter 5.1.II.D.21 L-RQ-703 AMSWER 3.03 (1.00)
- a. Secrease
- b. Decrease
- c. Decrease
- d. Decrease (0.25 each)
REFERENCE SFNP OF-0I-68, pp 28, 29 EIN' EIH Simulator. Halfuncti3n #36; L-RG-714 GCNS: SIM MAL 11
-b 1
f
\
- 3. INSTRUMENTS AND. CONTROLS PAGE 28 ANSWERS -- HATCH 182 -86/02/17-BROCKMAN, K.
ANSuER 3.04 (1.50) Recirculation Pump RPT Breaker Permissive RSCS Bypass TSV, 10% Closure Scram Bypasss TCV, Fast Closure Scram Bypass (3 0 0.5 each) REFERENCE EIH: GPNT, Vol V, Chapter 4.1-9; GPNT, Vol VI, Chapter 5.5-10; GPNT, Vol VII, Chapters 9.2.2-11 and 9.2.3-5 ANSWER 3.05 (2.00)
- 1) APRM ,A'
- fails upscale - ~:- relay K12A deenergi es (0.4)
- 2) -> NHS contacts K12A in RPS Trip Logic Al open (0.4)
- 3) -> Relays K14A & E deenergize (0.4)
- 4) -> Contacts K14A & E open (0.4)
- 5) -> Scram pilot solenoid valves for RPS A deenergine and vent (0.4)
REFERENCE BFNP: L/P D28 EIH: L-RO-720, Fig 720-1a, -ib, -2a, -2b, -3a, -3b GPNT, Vol. VI, Chapter 9.3.1-2, 3, 4 BSEP: SSM 28-2-A ANSWER 3.06 (1.75)
- a. Yes CApply in accordance with part (b)] (0.5),
- b. It can be moved out one notch (0.25) before a withdraw error will block further movement. (0.25) If the rod was inserted, it will move as far as the operator wants (0.25) unless it is the third insert error. (0.25) If it is the third insert error, then it would insert only one notch (0.25) ( 1.25)
REFERENCE EIH: GPNT, Volume VII, Chapter 9.25 L-RO-721
- 3. INSTRUMENTS AND' CONTROLS PAGE 29 ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
sucor p 3,7- ;,00 p
- a. ADS Valves are in Double Boxes (0.5)
- b. (1) No Demand (0.25) or Demand . .25)
(2) ADS Armed (0.5)
- c. 60 Minutes ov) and 6 minutes (0.2S)
REF
;P' L PO '52 .1' U 6 ) ; DCR-ia7 ANSWER 3.08 (1.50) 11 min 51 see (0.25) -OR- 1.85 t 0/W Pressure (0.25)
- 121.5 RWL 100 psij RHR/C3 s 150 psig Rn 2 min T.O. (0.5 each)
REFERENCE ElH* L-w0-767) 34AR-OPS-010-2 AN1WER 3.09 (1.00)
't h e INOVERAULE Control Rod must be Bypassed (0.7) using the Remote 0/pr : 'iuitch. (0.3)
REr ERENCE EIH: GPNT, Volume VII, Chapter 9.2.33 L-RG-722 i
30 INSTRUMENTS AND CONTROLS PAGE 30 ANSWERS -- HATCH 182 -86/02/17-BROCKMAN, K. ANSWER 3.10 (2.00)
- a. Unit 1
- 6. Unit 2
- c. Neither
- d. Unit 2 REFERENCE EIH: L-RO-742
- a. PROCEDURES - N0dMAL, ADNORMAL, EMERGENCY AND PAGE 31 R
~~~~ d656EU6563E"_CUNTRUE~~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K. ! ANSWER 4.01 (2.00) l a) 1. RPV level cannot be maintained (0.5)
- 2. Suppression Pool water temperature cannot be maintained below 110 des F (0.5) b) 1. Five (5), or more adjacent control rods not inserted below the 06 position (0.5)
- 2. Thirty (30), or more total control rods not inserted below the 06 position. (0.5)
REFERENCE EIH: 34AD-OPS-00?-2 l ANSWE7 4.02 (1.00) d PEFERENCE EIP: 34AB-OPS-002-2 1
- ANS JC0 4.03 (1.00) 1 Cleth and Rubber Gloves (0 5)
Cloth and Rubber Boots (0.5) l PEFERENCE SIH: GET Handbook; 40AC-HPX-004-0 l A NS!JER 4.04 (2.00)
- a. 2 Minutes (0.5)
- 6. Shallow (1.0)
- c. ' Print Notch Error' function (of RWM) (0.5) l l
l l 1 l
6 l l
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 32
~ ~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~U565ULUU CdL 6UUTRUL l --------------------
ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K. REFERENCE EIH: 42FH-ENC-010-1/2 ANSWER 4.05 (2.00)
- 1. Inability of High Pressure Systems (HPCI/RCIC/CRD) to restore 1esel.
- 2. One or more Low Pressure Systems running (LPCI. CS, or 1 CP & CDP)
- 3. Resctor Pressure ' 350 psig.
- 4. Reactor Level : 121.5' (0.5 each)
REFERENCE EIH: 34SO-021-001-21 34AD-0PS-002-2 ANSWER 4.06 (1.00) Turn the handwheel in the OEGIRED POSITION direction (1/4 turn man)(0.5); Verify the locl<in3 device integrity and proper installation by attempting to misposition the valve (0.5) REFERENCE CIM: 34GO-300-001-os ANGW:P 4.07 ( .50) None (0.5) REFERENCE EIH: 30AC-OPS-001-OS l
40 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 33
~~~~ ~~~~~~~~~~~~~~~~~~~~~~~~
R56 ULUb5 EEL'UU5TRUL ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K. ANSWER 4.08 (2.00)
- a. Alert
- b. (Notification of) Unusual Event
- c. General Emergency
- d. Site Area Emergency '0.5 each)
REFERENCE EIH: GET Handbook, pp 57, 58, 60, 61; 10AC-MGR-006-0S BFNP: BFN-IPD, IP-1, p li RO 85/04/01 ANSWER 4.09 (1.00) Drywell Cooling Fans when CH-23 r= 2% (0.5) Drywell Spray if Pressure (0/W) .= + 17 psis (0.25) and R9L is Normal (0.25) REFERENCE EIH: 34AB-0PS-002-2 ANSWER 4.10 (2.00) Clearance Number Sequential Tag Number Component Identification Required Position of Component (0.5 each) REFERENCE EIH: 10AC-OPS-001-OS
- 4. PROCEDURES - NORMAL, AB 1RMAL, EMERGENCY AND PAGE 34 RADIOLOGICAL CONTROL ANSWERS -- HATCH 1&2 -86/02/17-BROCl(MAN, K.
ANSWER 4.11 (1.75)
- a. 5
- b. 7
- c. 4
- d. 6
- e. 2
- f. 3 3 1 (0.25 each)
REFERENCE EIH: L-RG-715; 10AC-MGR-003-OS l i I 1 l
\
l
o TEST CROSS REFERENCE PAGE 1 QUESTION VALUE REFERENCE 01.01 1.00 KEB0000647 01.02 1.00 KEB0000648 01.03 2.00 KEB0000649 01.04 2.00 KEB0000650 01 05 1.00 KE80000651 01.06 2.00 HEB0000652 01.07 1.00 ME80000653 01.08 3.00 KE00000654 01.09 2.00 MEB0000655 01 10 1.00 ME00000656 16.00 02.01 1.00 ME30000657 o:' . 02 2.00 KED0000650 02.03 2.00 MEB0000659 02.04 1.00 t*EB0000660 02.05 1.00 UED0000661 02.06 14 CO KEB0000662 02.07 1.00 KEB0000643 02.08 1.00 KEB0000664 02.09 2.00 KEB0000665 02.10 1.50 KEB0000666 02.11 2.00 KE00000667 15.50 03.01 1.00 KE00000668 03.02 2.00 ME00000669 03.03 1.00 MED0000671 03.04 1.50 KE80000672 09 05 2.00 HED0000673 03.06 1.75 HE00000674 4'" 2 . '; 0 FLE0000673 03.08 1.50 KE80000676 03.09 1 00 KEB0000677 03 10 2.00 MED0000678 15.75 04.01 2.00 KEB0000679 04.02 1.00 KEB0000600 04.03 1.00 KEB0000681 04.04 2.00 KE00000602 04.05 2.00 KEB0000683 04.06 1.00 KE80000684 04.07 .50 KEB0000605 04.08 2.00 KEB0000606 04.09 1.00 MEB0000687 04.10 2.00 KE00000688
. _ _ _ _ i
O
~~
TEST CROSS REFERENCE PAGE 2 QUESTION VALUE REFERENCE 04.11 1.75 KEB0000689 16.25 de-6aus se d 63.50 1 1 i 1
o %
-MRETER-ENCLOSURE 3 U. S. NUCLEAR REGULATORY COMMISSION SENIOR ":E ACTOR OPERATOR LICENSE EXAMINATION FACILITY: HATCH 18.2 REACit' TYPE: BWR-GE4 DATE ADMINISTERED: 06/02/17 EXAMINER: EROCKhAN, K.
APPLICANT: ,________,___,,_______ _ _ _ , INSTRUCTIONS TO APPLICANT: Use separate paper for the answers. Write answers on one side oniv. Staple question sheet on top of the answer theets. Points for each question are indicated lei parentheses after the question. The passing grade requires et least 70% in each category and a final grado of at le'st 90%. E::a m i n a t i o n papers will be picked up , hours after the examination starts. 5' '%)
% OF CATEGORY *: OF AP*LICANT'S CATEGOPY
/ALUE TOTAL 1CORi '1 A L U E CATEGORY 1 < ~e in 0
_151_0_-- _IIlli _______.---- -__._--__ 5. TilEORY OF NUCLEAR PONER PLANT OPERATION. FLUIDS, AND
,q THERMODYNAMICS
_Ine Ijo 5 _ 1__l-- _ --_________ ________
- t. PLANT SYSTEMS DESIGN, C001ROL, AND I"STPUMENTATION le '5 a'
__Ili___ _'I_I'__l _____-_____ ________
- 7. PROCLDURES - NORMAL, ABNORMAL, ENERGENCY AND RADIOLOGICAL CONTROL
- _I__.- - I __-________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMIIATIONS si . '3-
- 12. 100 00 TOTALS
_ _ _ _ _' S___ __-_-- ___________ _____--_ FINAL GRADE _________________% All work done on this e::a m i n a t i o n is my own. I have neither gaven nor received aid. 5PP~5C5UII5~555U5iURE~~~~~~~~~~~~~~
e 1
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS, AND PAGE 2 QUESTION 5 01 (2.50)
Attached Figure 4 070 represents a transient that could occur at a BWR. G I'1E N ' (1) An SRV is manuclly lifted at time T = 2 min
'2) The SRV is closed at time T = 4.5 min
'3? No other operator a c t ioris occur (4) Recorder speed = 1 dtvtston = 1 minute EXPLAIN the cause(s) of the following recorder indtcations:
- 6. Presture DECREASE (Region A)
- b. Total Steam Flow REDUCTION (Recton 9)
- c. Turbine Sterm ficw REDUCTION (Region C)
- d. Level INCPEASE (Point D)
- e. NEW STAOLE (loueri Leve' (Region E) 00E3fION 5.02 (1.50)
MArr9 the 2ppropriate Thermal Limit (ce),
.: . L i ie 2 r Heat Guneration Rate (LHGP)
L. A er tje Planar Linear Heat Generation Pate (APLHGR) _ h t o i m u ai Ce tttczl Pouer Ratto (MCPR) to FAILUPE MECI' ANISM AND to etch LIMITING CONDITION given below: FAILURE M E C H.;N I S M LIMITING CONDITION F'. Clad relting caused by Lt. Coolant transition c'ecay n-at / stored heat botiln3 f r 11.n'i o j a LOCA F2. Cl-d crack.ing from the surface L2. Clad pla'atic strain ecoming vapor ' blanketed' l '. F '3 . C12d crack ing caused by L3. Manimum elad temp-t ity, stre'ss from pellet er'ture of 2200 dog F
;2n ion
(***** CATCGUPY 05 CONTINULD ON NEXT PAGC *****)
o 1 1
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3 QUESTION 5.03 (1.00) j The reactor trips from full power, equilibrium :< e n o n conditions. Four (4) hours later the reactor is brought critical and power level is main-tained on range 5 of the IRMs for several hours. Which of the following statements is COPRECT concerning control rod motion during this period?
- 3. Rods will have to be withdrawn due to ' tenon build-in.
- b. Rods will have to be rapidiv inserted since the critical reactor will cause a high rate of <enon burnout.
- c. Rods will have to be inserted since ::enon will closely follow its normcl decs/ rate.
- d. Rods will appro: imately remairi au ts as the :enon establishes its equilibri"m value for this power level.
QUESTION 5.04 (1.00) STATE for which condition the reactivity coefficient contribution uov1d be MOPE FIEGATIVE. EVPLAIN your choice. Moderator void coe rricient faa 1 1% IFICREASE in void fr3ction at 10% "old f r ac tiori in the core.
--OR-Moderator votd coeff1ctent for a 1% U4 CREASE in void I. .* _en it ;* n.
- void fraction to the core.
(taver c n T .: r; 0 ? Y O'3 CON TINUCD ON NEXT PAGE *****)
e 1
- 5. THEORY OF NUCLEdR POWEP P: ANT OPERATION. FLUIDS, AND PAGE 4 QUESTION 5.05 (1.00)
A reactor heat balance was performed (bv hand) during the 00-00 i shift due to the Process Computer being 00C. The GAF's were l computed, but the APRM GAIN ADJUSTMENTS HAVE NOT BEEN MADE. l Which ONE of the following statements is TRUE concerning reactor power?
- a. If the feedwater flow rate used in the heat balance calcu-lation was LOWER than the actual feed 9ater flow rate, then the actual power is HIGHER than the currently calculated power.
- b. If the reactor recirculation pump heat input used in the heat balance c.siculation was OMITTED. then tho actual power in
'0WER than the corrently calculated power.
- c. If the te am fluo vaed in the hest bilance calculation was f
i L O W E ;' t:un t:i. 2ctual steam flou. then the actual power is LOWER l than the ;uretot1 / calcola+.ed power. l l d. If the "" 3 C U iaturn ten;wrature used in '. h e heat balance cal- _v1-tto at .nuf~
- i.a n ti o re t i'a l Pm C U r e t .ir o temperaturer then
+.9 >et- 1 , ;. -
HIC:lCP thmi f 'ie cor,m.y calcol3 tad pow +r. J OUESTION 5.61 ' l ' j Att .:h e d r, y,c. PWS s ,3 r s entc a tran_1"nc that avid occur at a BWR. l
: no t (14 d: ;ter :' e c i r c Flow Contreller F,21s High i
< ) +3 me, ito, actions occur
's *
.,.hr
{
"p ad - 1 d i v i ', i o n = 1 minute c'rp' y 1 , ,,s.
of f. h e 'ollowing recorder indications
- 2 , % u '. u r Y-te _c :1 DECPCA'3E (Point A) (0.5)
') , N ;, t o c "' aver I N C P C V;' 5 (Point D) (0.5) c, C- ~ '. o o DECPEASE (Point C) (0,5)
"' a c c t o r Pre: 9vre INC0FACC (Point D) (0.5)
- e. Total Feedwater Flow INCPEASE (Point E) (0.5)
- f. Total Steam Flow IN Cr:E ASE (Point F) (0.5)
( * * * * *' CATEGORY 05 CONTINUED ON NEXT PAGE *****)
e t
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5 QUESTION 5.07 (2.00)
During your Shift, an SRV inadvertantly opens from 100% power and 1000 psia. Use a Mollier Diagram or toe Steam Tables to answer the following'
- a. STATE the tatlpipe temperature, assuming atmospheric pressure in the Suppression Pool and No Reactor Depressurl ation. (0.5)
- b. If the Suppresston Pool Pressure were to INCREASE, STATE whether the Tailpipe Temperature would INCREASE, DECREASE, or REMAIN THE SAME. (0.5)
- c. If the reactor is deprescurl:ed when the SRV is opened, STATE whether the failpipe Temperature will INITIALLY INCREASE, DECREASE, or REMAIN THE SAME. (0.5)
- d. STATE the Reactor Pressure at which the Tailpipe Temperature would be at its MAXIMUM value (during the depressurization). (0.5)
(ASSUME A SATURATED SYSTEM AND INSTANTANEOUS HEAT TRANSFER) GUESTION 5.00 (1.00) LIST two (2) Design Characteristics which allow for the use of Hybrid Control Rods? QUESTION 5.09 (1.00) Attached Figure 1404 illustrates the ' Combined Head / Pressure Curves for Two Pumps.' Select from the figure the appropriate system operating point l (numbered i through 6) for each of the following conditions. j
- a. Pumps A and B running in SERIES with the pump discharge valve (0.5) throttled shut from the initial condition. l
- b. Pumps A and B running in PARALLEL with the pump discharge valve (0.5) ;
fully open. l (***** CATEGORY 05 CONTINUED ON NE/T PAGE *****)
. t
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND PAGE 6 QUESTION 5.10 (2.00)
Concerning General Electric's Preconditioning Interim Operating Management Recommendations (PCIOMR):
- a. Starting with the fuel at a threshold of 11.0 kw/ft, a ma:timum ramp increase is begun at time 0000 and the final desired power of 13.0 kw/ft is achieved at 2000. At this time, the required soak is performed FOR 10 MINUTES, at which time the load dispatcher directs a power reduction that takes nodal power down to 11.0 kw/ft. SELECT the valid preconditioned valve for this node.
ASSUME THE MAXIMUM RAMP PATC IS .10 Kw/ft/hr
- 1) 11 0 k w ,' f t
- 2) 11.8 kurft
- 3) 12.5 ku/ft 46 tl.0Lo'f'
- b. SELECT the minimum time which would be required to raise power back to 13.0 ku/ft. gioen the above maximum ramp rate.
- 1) Itmedtale (Patte to 13.0 1.w/ft. w/o restrictions)
- 2) 3 houre 3: 12 ,,v r A 2D *: t I<**v* END Or CATECORY 05 *****)
. i
- 6. P. ANT SYSTEMS DESIGN. CONTROL, AND INSTRUMENTATION PAGE 7 00ESTION 6.01 (2.00)
Regarding the S R') ' s and the associated Low Low Set (LLS) logic: There are three lights associated with each SPV - RED. GREEN, and AMCER. EXPLAIN what each of the different colored lights indicates, and STATE whether each wouid be ener:11:ed or de-energized dering the time its SRV w>s open as a result of reactor pressure reaching t'ie S P ' ' -= relief etpoint.
-ASSUME T'AT TI'E UALUES IN GUES T I"'J ARE LLS VALVES-AFD 'C E AC T U A TED OY THE LLS FUNCTION QUESTION 6.02 (2.00' Concerning the Unit 2 Pitt' System, while operatinj at 100% RTP.
Cor each of the folloulog situations. STATE whether the statement is TRUE or FALSE.
- a. The Inbased Injection Valve (f015) may be opened ONLY if the Outboard In j+etion Value (F017) is shut. (0.5)
- b. The Inbo ar d Cor t a toment Spray Valve ( F 021) CANNOT be opened uni-;3 the 'utbt '~d 4 ; 1 "! (F016) is shut. (0.5)
- c. *he 9hutdoon Cooling Isolation Valvec 'F008 & F009) are int er locb d shut ut: .i the p1 ant in these conditions. (0.5) 4- The Toro; 012t on Valve (F024). the Torus Spray Valve (F^.Z?). a: d th. reros Coel t ric Valve 'F020) may all be in the nrtN : o 2 ., ' .on ,+ me u ,,, e time. (o,5)
Ir,3 rat:,q9y n6 CONTINUED ON NEXT PAGE *****)
. t
~
- 6. P_ ANT SYSTEMS DCSIGN. CONTROL. AND INSTRUMENTATION PAGE O OUESTION 6.03 (1.00)
Whtch of the following lists of responses (a - d) CORRECILY describes the main turbine's response to an OVERSPEED condition? l RESPONSES SPEEOS 'a)
. (b) (c) (d) l 1027. ' Master' All ICV's All ICV's ' Master' ICV's bejin begin to begin to ICV's begin to throttle throttle throttle to throttle
' Slave ' Slave' TCU's begin ICV's begin to throttiv 'o throttle All Ir"'s All ICV's All ICV's AL1 ICV's
.t Cl o .ed Full Closed Full Closed Full Cloned
- 10 7', .i e c h a n t e i l r'ectrical Mechanical Electrical l d'Spd Tcip O'Spd Tctp O'Spd Trip O'Spd Trip 1
l L i '. 57; Elec tr ic il Michanical Electrical Mechantcal l t ' +; ; , *.;p r.'9pd Tr1p O'Spd 1. ip O'Gpd Trip QUIS r !ON 6,n1 '1,50) l Concerning Pr:m.:r/ <id Secondery C ant s innier.t Systems for Unit 2! 04TE t'.. ? 3, , e r i n e r '. i n j tiie Drvuel! and the T o r o ', (0.5)
- b. "Xr' A tH ' :io 'C a s e ; for b r/ ing 4 MINIMUM and Mr.<1 MUM Torus Level.
(1.0) G U E'3 T IO N 6.05 (1.50) The M it i n Turtfine f tr 3t stage prossure switches provide permissives
.nd/c. control signals for m "> c a l plant functions. LIST three (3) of + he m permissives/contr 1 Nnct ions.
t 1
'***** C A T E G o t' Y 06 CONTINUED ON NEXT PAGE *****)
l
o 1
- 6. PLANT SYSTEMS DE' SIGN. CONTROL, AND INSTRUMENTATION PAGE 9 1
OltESTION 6.06 (2.00) The displays of the indicator lights for ATTS can indicate I that the system is in its normal operating mode with NO ALARM CONDITIONS peesent? For each of the following STATE the status (ON/0FF$ and the indicator color (If ON) when in this normal mode. ' NOTE: FIGURES 9.9(1) AND 9.9(6) PROVIDED FOR REFERENCE
- a. STATUS /MTU Indicator
- b. GROSS FAIL /MTU Irid i c a t o r
- c. POWE.7/P925 Indicator OVESTION 6.07 ( .50)
TRUE or FALSE: When synchronizing the Main Generitor to the gride the synchroscope etcevitry provides an INTERLOCK to preclude the Output Dreaker from closing, unless the frequency is aPPronimately in phase, as shown by the synchroscope being between 11 arid 1 0' Clock. l l (***** CATECORY 06 CONTINUED ON NEXT PAGE *****) i
. i
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 10 nn ,nns
' ' 'O'D
_ m ,r n 7 g g The SPDS Pr1 mary Display provides information concer ig the ADS, amongst ;ather parameters,
- a. DESCRIBE how the ADS Valves are diff ntiated from the other SRV's. (0.5)
- 6. STATE the sijntficance o .,e following COLOKS, as they are used for indication of f Valve status (1.0)
(1) Green (2) je
- c. ST the two (2) time periods over which TREND Disp ldY5
;re jlve^ (0.5)
GUESTION 6.09 (1.50) LIST ALL requirements which could/must be met in order to actuate ADS. (Automatic Actuation) l OUESTION 6.10 (2.00) STATE for each of the following conditions, which Unit (if any) will I recieve the 18 EDG. l
- a. Select Switch Position -
Unit 1 Unit 1 Condition - LOCA/LOSP Unit 2 Condttton - LOSP
- b. '? ? l e c t hitch Position -
Unit 1 Unit 1 Condition - LUSP tintt 2 Condttton - LOCA/LOSP
- c. Select Switch Position -
Unit 2 Untt 1 Condition - LOCA/LOSP Untt 2 Condition - LOCA/LOSP
- d. '3 e l e c t Switch Position -
Unit 2 Unit 1 Condition - LOCA Unit 2 Condition - LOSP . (***** END OF CATEGORY 06 *****)
. i
- 7. PROCEDURES - NORhAL, ADNORMAL, EMERGENCY AND PAGE 11
~~~~ - ~ ~ ~~~~
RE656L55iCAL C5NTRdL~ ~ ~~~~~~~~~ QUESTION 7.01 (2.00) 34AB-OPS-009-2, ' Inability to Shutdown With Control Rods', states that:
"... If at any time. either condition b(1) or b(2) e::i s t s ,
and either _____ that the r (a.1) e ra c____ t o r orcannot____ (a.2) ____ , and if it is o b i t o u .5 be shutdown and, in the judgment of the Shift Supervisor. or in his absence. a
- 11. : s ed ope r.a t or . a h a z a r d e s: 15 t s to the environs, personnel, I or th3 plante utilize the -s t a n d b y liquid control system l
; .. e 7 4 '3 0 - C 4 '. - 0 01 -' . *
- 1. i_IST conditiam 3 1) and (- . 21 (1.0)
- b. LIST cond;tions b ' 1 '- and b(2). (1.0)
GUESTION 7.02 t'.00) 14AD-OPS-002 "P1;m Si ed Inside Primary Contitnment', lists numerous canditions :ndicatire of a break. Which one of the followinj supports the cucpicion of a SMALL break inside con-t r i rem ent . ' a, Decreew in rewtor water level; pressure and/or temperature in-cr3 rue in O'visellt etrborne activity increase in Drywell; increased ntfrrC oper 3 tog r e q u *e n e ; .
- b. "w.reate in resc'oc pressure pressure and/or temperature increase ti, ~. oof 1' - g e n e r a t o ,- load decrease
- DWFDS hijh level.
' : .e : 'or 'I vie f , om low water level: pressure and/or temperature rcr> 2 :n <
,aall: generator load decrease: increased DWFDS
.,-.,t!.g f' r s , i c ,> .
j- C'e 3 c t o r '3 r _ 0 from *itgh OfyWell pressure" jener3 tor lo 3d decreasei
. ir bur ie activit; tro: i23se in Dry 9 ell; DWFDS high level, t7 *r: rmr30py 07 r0NTINUr0 GM MEXT P Ar;E t '< * * * )
e i
'. FPOCEDURCS - N O R41 AL , ACNORMAL. EMERCENCY AND PAGE 12
~
~~~~id656E66563L 665TRUE~~~~~~~~~~~~~~~~~~~~~~~~
C'IESTION 7.03 (1.00) Yeo enter e r. 3r es potted with the follouing sign: _________ l CAUTION I g---_--_--; j l I Your purpose 12 to c or.d ve t a normal I i > Su -ve t 11arice i n s p e c + t o ri . I I i ! I 1 l \ l l IContamin.1 ) l l ___ Area __l l s
- IST t'ie .i i n i m . m pra'e:tt>e clothing requirements which an RWP for 'hi-; aree could ie;vice.
Oi!ES' ION 7,04 (1 00)
'JoAC-Or0-006-2 *^ost Accident Venting *, 10 very specific as to what conditions require entry into this procedure. LIST the two (2) conditions. (Setpoints NOT required)
QUESTION 7.05 (1.00) l l 3dSO-C11-003-2 ' C orit r o l Rod Drive System", cautions the operator ! to check the (position of the) Reactor Recirculation Pumps' discharge and suction valves prior to CRD st:rtup. STATE the basis for this caution. (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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- 7. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE
~
13
~~~~53656L665CEL C05TR6L'~~~~~~~~~~~~~~~~~~~~~~~
GUESTION 7.06 (2.00) ! 42FH-ENG-010-2, ' Control Rod Movement', provides numerous l STANDARD PRACTICES which apply when rods are being moved for the purpose of changing power level.
- c. With LHGR - 8 Kw/ft. STATE the period of time you must wait between successive notch withdrawals of the same rod. (0.5)
- b. With HIGH POWER and HIGH FLOW conditions, STATE which type 1
of rod - SHALLOW. DEEP, CENTRAL. or EDGE - should NOT be with-drawn nor inserted. (l.0)
- c. With power 30%, STATE the verification which is utilized when latch.ng the first rod in any group. (0.5)
GUESTION 7.07 '.2.00) A pip break cccurs ano r?sults in a decreasing reactor water
- 1. 3. . e l and decreeuing r e .a c t o r p r e s s e_i r e . LIST tho four (4) criteria ub : , 5 soield +;uire the i5ANUAL initiation of A D '3 , given that auto-
- nitiction hr; not occured.
l l GUESTICN 7,00 (1.00) Concerning a " Cable Spreading Room Fire', Procedure 34AB-FPX-040-0: a, STATE who is resporisible for directing the control room response to eable spreading room fire. (0.5)
- b. T9UE or FALSE (0.5)
For en exposuce Ftre (in the cable spreading room) involving com-bustables WITHOUT electrical insulation involvement, the Control Room Operator in the affected unit will initiate, as necessary. a 52pid loed reduction, will trip the turbine generator, will menoally SCRAM the reactor, and will place the Mode S' aitch in S"UTDCUN, l (ra"** CATECORY 07 CONTINUED ON NEXT PAGE *****)
o i 7 PROCEDURES - NORMAL, ACNORMAL, EMERGENCY AND PAGE 14
~~~~
R5656[665CdE~6UNTR6E~~~~~~~~~~~~~~~~~~~~~~~~ QUESTION 7.09 (2.00) STATE which Emergency Classification is appropriate for the fo11owin3 definitions.
- 1. Events are in progress or have occurred which involve actual or potential substantlal degradation of the level of safety of the plant.
- b. Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant.
- c. Events are in progress or have occurred which involve actual or imminent substantial core degradation or melting with the potential for loss of containment integrity.
- d. Events are in progress or have occurred which involve an actual or likely mejor failure of p l a n t, functions needed for protection o' the public.
GUESTION 7.10 (1.00) Concerning 34AB-OPS-002-2, ' Pipe Break Inside Primary Containment **
- a. STATE when Drywell Cooling Fans are REQUIRED to be placed in service. (0.5)
- 5. GTATE when Dryuell Spray is REQUIRED to be initiated. (0.5)
(***** CATECORY 07 CONTINUED ON NEXT PAGE *****) 1
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15
- ~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~EE656L55 FEEL E5NYR6L QUESTION 7.11 (1.75)
The Procedure Numbering System consists of a twelve character alpha-numeric designator. with three dashes (e.g., 31SO-E11-034-1S). This catagorizes each procedure in seven ways. MATCll the Column A characters (using the exemple above) with the Column B information, in order to most accurately describe the numbering system. COLUMMN A COLUMN D
- a. 1st (3) 1. Safety Status
- b. 2nd (14 2. Sequence Number
- c. 3cd/4th (30) 3, Unit Applicability
- d. 5th/6th/7th (E11) 4. Procedure Type e, Gth/9th'10th ' W14 ' 5. Responsible Department C. lith (1) 4. System Identification
( "> i 7 Respons ble Section 3- 12th . ( * *'* *
- END OF CATEGORY 07 *****'s
+
O I 8 ADMINISTRATIVE P'ROCEDURES, CONDITIQUS, AND LIMITATIONS PAGE 16 QUESTION 0.01 (1.50) Unit 2 Technical Specific:tions define SHUTDOWN MARCCN as...
"Shutdeun Mar jiri shall be the amount of reactivity by obich the reactor is soberitical, or would be soberiti:al from it3 present condition, assuming...'
LIST the *Vm cunditions which complete the definition of SHUT-DOWH A M:cTv OUESTION 9 . n ~_ (2.00i Given' h- _!an Pcur 60% Cc w Fle' 40% i / T :o ) .o
~
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i ur - 4. 33 5.6 Ec. closed a G L P t' ^ ^ i; ,, <C"003C ONE) (1.0) 0,;
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If the MCPR Printout from the Process Computer for Rod 29-17 read 1,9'3, which of the following would be required per Unit 2 Tech Specs? (1) No Action Required (2) Initiate corrective action within 15 minutes and continue the corrective action so that MCPR is within the applicable limit within 2 hours, or reduce thermal power to - 25% within the next 4 hours. (3) Initiate corrective action within 15 minutes and continue the corrective action so that hCPR is within the applicable limit within 4 hours, or reduce thermal power to 25% within the next 8 hours. (4) Place the reactor in HOT SHUTDOWN within 2 hours and reduce ' steam dome pressure to < 785 psig and core flow to 10% within the next 4 hours. (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 17 GUESTION 0.03 (2.50)
Unit 1 Technical Specifications provide guideance with respect to refueling operations. C0hPLETE the following, as per U1 TS's.
- a. When conducting refueling operations, at least
___ (1) ___ SRM channels must be operable and ___(2i___. with a minimum acceptable count rate of ___ (3) ___. Refueling operations may commence __ (4) ___ 3fter reactor uhutdoun. (2.0)
- b. TRUE or FALSE You are conductinj a spiral reload on Unit 1. Up to four (4) fuel assemblies will be loaded into their previous core positions nent to the four (4) 3RM's. Untii these assemblies have been loaded.
the (Minimum Requ i r ed ) SPM count rate requirement is not ne: esser; (0.5) GUESTION 3 04 '2 03)
- , Unit 2 it in Oper.ationel 1 Condition with NO outstanding deficien-cies. The Surveillance Assignment printout sheet identifies tomorrow as the ' Latest Octe" for the RCIC quarterly flow testi RCIC becomes INOP before completing the surveillance scheduled for today. SELECT the
.: t a t e m e r t ohich accurately describes the surveillance requirements as per 40AC-rec-001-OS- " Tech Spec Surveillance Program." (1.0)
(1) Tho <vcvoillance must be performed immediately AFTER naturning 11.3 Tystem (or subsystem) to an OPC't ACLE condi tion .
<2) A "ICSED SUOVEILLANCE SHECT must be inttiated since the surveillance was not completed today. as scheduled.
- 3) The surveillance shall NOT be documented as officially missed until the END of the ' Latest Date.'
/3) Since PCIC is already INOPERABLE. a MISSED SURVEILLANCE Sl!EET need no' be issued to track the missed surveillance.
- b. Th-- ' Car 1Lest Gat"* bloc!' of the Surveillance Assignment printout sheet reads '00000* for
osed by the Technical Specifications in this it
- e,n c e ? -
- e. No action required; function not required to be OPERABLE.
- b. Place one Trip System in the tripped condition, within one tiour' no additional action is required.
e 3e in Cold Shutdown within the next 30 hours. 1 m e.:e ono T . i;. Systeti in the tripped condition, within one
'i our AND be in Cold Shutdown uithin the next 30 hours.
e- Lock the Reactor Mo h Suitch in the Shutdown Position, u i '. h i n one hour
/***** EHO 0" CATEGORY 00 ***r*)
(******7r***** END OF EXAMINATION ***************)
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3/4 LIMITING CONDITIONS FOR 0PERAllON AND SURV61LLAh d stVU ntMcN b 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION ( 3.0.1 Limiting Conditions for Operation and ACTION requirements shall be applicable during the OPERATIONAL CONDITIONS or other states specified for each specificatiog 3.0.2 Adherence to the requirements of the Limiting Condition for Operation and associated ACTION within the specified time interval shall constitute compliance with the specification. In the event the Limiting Condition for Operation is restored prior to expiration of the specified time interval, completion of the ACTION statement is not required. l 3.0.3 In the event a Limiting Condition for Operation and/or associated ACTION requirements cannot be satisfied because of circumstances in excess of those addressed in the specification, the facility shall be placed in at least l HOT SHUTDOWN within 6 hours and in COLD SHUTDOWN within the following 30 hours unless corrective measures are completed that permit operation under the permissible ACTION statements for the specified time interval as measured from initial discovdry or until the reactor is placed in an OPERATIONAL CONDITION , in which the specification is not applicable. Exceptions to thqse i j requirements shall be stated in the individual specifications. , 3.0.4 Entry into an OPERATIONAL CONDITION or other specified applicability state shall not be made unless the conditions of the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION statements unless otherwise excepted. This provision shall not prevent passage thru OPERATIONAL CONDITIONS required to comply with ACTION requirements. 3.0.5 When a system, subsystem, train, component, or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable . l Limiting Condition for Operation, provided: (1) its corresponding normal or. i emergency power source is OPERABLE; AND (2) all of its redundant system (s), l subsystem (s), train (s), component (s), and device (s) are OPERABLE, or likewise satisfy the requirements of this specification. Unless both conditions (1) and (2) are satisfied, the unit shall be placed in at least HOT SHUTDOWN within 6 hours, and in at least COLD SHUTDOWN within the following 30 hours. _ This specification is not applicable in Conditions 4 or 5.
'" SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be applicable during the OPERATIONAL CONDITIONS or other states specified for individual Limiting Conditions for Operation unless otherwise stated in the individual Surveillance Requirements. ,
4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with: .
- a. A maximum allowable extension not to exceed 25% of the surveillance l 1
interval, HATCH - UNIT 2 3/4 0-1 Amendment No. E, 49
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%0C) d5 %9C WOd W3dr 10 HATCH - UNIT 2 3/4 2-7d Amendment No. 39 l
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o . 3/4.3 INSTRUMENTATION 3/4.3.1 ' REACTOR Pit 0TECTION SYSTEM INSTRUMENTATION LIMITINGf CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor. protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2. Set points and interlocks are l 1 given in Table 2.2.1-1.
, APPLICABILITY: As shown in Table 3.3.1-1. \
ACTION:
- a. With the requirer.ents for the minimum number of OPERABLE channels not .
satisfied for one trip system, place at least one inoperable channel .. in the tripped condition within one hour. ~r W
- b. With the requirements for the minimum number of OPERABLE channels not.
satisfied for both trip systems, place at least one inoperable channel
.M .
in at least one trip system
- in the tripped condition within one hour and take the ACTION required by Table 3.3.1-1. g-J <-
- c. The provisions of Specification 3.0.3 are not applicable in OPERA-(- TIONAL CONDITION 5.
a
~
SURVEILLANCE REQUIREMENTS .? .-
. a 4.3.1.1 Each reactor protection system instrumentation channel shall be
demonstrated OPERA 8LE by the performance of the CHANNEL CHECK, CHANNEL FUNCTION TEST and CHANNEL CALIBRATION operations during the OPERATIONAL M;. CONDITIONS and at the frequencies shown in Table 4.3.1-1. .
. . g.
4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of - M
, all channels shall be performed at least once per 18 months and shall . .
.._M include calibration of time delay relays and timers necessary for proper .
*~
functioning.,of the trip system. , 4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip ' M function of Table 3.3.1-2 shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one logic train Ml jl such that both logic trains are tested at least once per 36 months and one I channel per function such that all channels are tested at least once every -! v N times 18 months were N is the total number of redundant channels in a ' specific reactor trip function. . t l
*If both channels are inoperable in one trip system, select at least one !
inoperable channel in that trip system to place in the tripped conditions, < L except when this could cause the Trip Function to occur. ' HATCH - UNIT 2 3/4 3-1 Amendment No. [ f
TABLE 3.3.1-1 g REActom PA0TECT10N SYSTFM INSTRUMENTATION MINIMUM NUMBER N Z APPLICABLE OPERATIOMAL OPERA 8LE CHANNELS C0ct0ITIONS PER TRIP SYSTEM (a) ACT1001
, FUNCTIONAL UNIT
- 1. Intermodlate Range Monitors:
E
- (2C51-K601, A, 8, C, D, E, F, G, H) * -4
- m a. Neutron Film - High 2,5**
8 8 3 1
. 3, 4 2 2 . .
- b. I nope ra t ive 2,5'** 3 1 3, 4 2 2
- 2. Average Power Range Moni tor:
(2C51-K605 A, B, C, D, E, F) .
- a. Notetton Fitm - Upscale, 155 2, 5 2 1
- b. Flow Referenced Simulated 1he rma l Powe r - Upsca le 1 2 3
- c. Fixed Neutron Flux -
Upscale, 118% 1 2 3
- d. Inope ra t ive 1, 2, 5 2 4
- e. Downscale 1 2 3 F. LPRM 1, 2, 5 (d) MA
%" 3. Reactor Vessel Steam Dome Pressure -
A High (2821-N678 A, 8. C, D) 1, 2 g(J, 2821-NO45 5 A, 8, C, D) w e N
- 4. Reactor Vessel Water Level -
Low ( Level 3) (2821-N680 A, 8, C, D) 1, 2 5 2(J. 2s21-N681 A, e, C, o)
- 5. Main Steam Line isolation Valve -
4 3 Closure (NA) I
- 6. Main Steam Line Radiation - High I, i' * ' 2 6 (2011-K603 A, B, C, 0)
- 7. Drywe l l Pressure - High 1, 2 2 5 (2C71-N650 A, 8, C, D)
E,
=
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m (' i ( i . . TABLE 3.3.1-1 (Continued) g REACTOR PROTECTION SYSTEM INSTRUME4TATION c> APPLICA8LE MINIMUM IIUISER Z OPERATIONAL OPERA 8LE CHA8000ELS
, FU80CT1088AL U8tI T CONDITIONS PER TRIP SYSTEM (s) ACT1000 E 8. Scram Discha rge Volume Water' ,
w Level - High ( 2C11-N013A,8,C,0) 1, 2, 5'"' 2 4 H Level - High (2C11- % 8,C.D) 1, 2, 5 2 4 - ro l
- 9. Turbine stop Valve - Closure (NA) I t 'l 4lh) 7
- 10. Turbine Control Valve F.est Closure, Trip Oil Pressure - Low 1"8 288 7 (2C71-fe005A.8,C D)
- 11. Reactor Mode Switch in Shutdown Position (NA) 1, 2, 3, 4, $ 1 8
- 12. Manual Scram (NA) 1,2,3,4,5 1 9 N.
Y w l R
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TABLE 3.3.1-1 (Continued) REACTOR PROTECTION SYSTEM INSTRUMENTATION . _ ACTION ACTION 1 - InOPEitATIONALCONDITION2,beinatleastHOTSHUTOOWN within 6 hours. - l, In OPERATIONAL CONDITION 5, suspend all operations involving CORE ALTERATIONS or positive reactivity changes and fully, insert all insertable control rods within one hour. Locli the reactor mode switch in the Shutdown position within ^ ACTION 2 - one hour. - - ACTION 3 - Be in at least STARTUP within 2 hours. I '- ., ACTION 4 - In OPERATIONAL CONDITION 1 or 2, be in at least HOT SHUT 00WN I within 6 hours. I E' q 47 In OPERATIONAL CONDITION 5, suspend all opet .tions involving t 4 CORE ALTERATIONS or positive reactivity char.ges and fully $ ,. insert all insertable control rods within one hour, g ACTION 5 - Be in at least HOT SHUT 00WN within 6 hours. Si ACTION 6 - Be in STARTUP with the main steam line isolation valves closed witihin 2 hours or in at least HOT SHUT 00WN within 6 hours. f.L'j ACTION 7 Initiate a reduction in THERMAL POWER within 15 minutes and il::
-c.
be at less than 30% of RATED THERMAL POWER within 2 hours. , ACTION 8 - In OPERATIONAL CONDITION 1 or 2, be in at least HOT SHUT 00WN ' within 6 hours. - f, .> r In OPERATIONAL CONDITION 3 or 4, immediately and at least
~
b '- once per 12 hours verify that all control rods are fully : inserted. ,m
'.. 4-In OPERATIONAL CONDITION 5, suspend all operations involving #:
' CORE ALTERATIONS or positive reactivity changes and fully .
insert all insertable control rods within one hour. ] E i sw 4 HATCH - UNIT 2 3/4 3-4 Amendment No. 8
TABLE 3.3.1-1 (Continued) (,, REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION 9 - In OPERATIONAL CONDITION 1 or 2, be in at least NOT SHUTOOWN within 6 hours. In OPERATIONAL CONDITION 3 or 4, lock the reactor mode switch in the Shutdown position within one hour. In OPERATIONAL CONDITION 5, suspend all operations involving CORE ALTERATIONS or positive reactivity changes and fully insert all insertable co'ntrol rods within one hour. TABLE NOTATIONS
- a. A channel may be placed in an inoperable status for up to 2 hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
S
- b. The " shorting links" shall be removed from the RPS circuitry during CORET ALTERATIONS and shutdown margin demonstrations performed in accordance 4 with Specification 3.10.3.
( ~- c. The IRM scrans are automatically bypassed when the reactor vessel mode switch is in the Run position and all APRM channels are OPERABLE and on scale.
- d. An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than eleven LP8M inputs to an APRM channel.
- e. These functions are not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed.
- f. This function is automatically bypassed when the reactor mode switch is in other than the Run position.
- g. This function is not required to be OPERABLE when PRIMARY CONTAINMENT
- INTEGRITY is not required.
,, l <
- h. With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.11.1 or 3.9.11.2.
~
- 1. These functions are bypassed when turbine first stage pressure is $250*
psig, equivalent to THERMAL POWER less than 30% of RATED THERMAL POWER. J. Also trips reactor coolant system recirculation pump MG sets.
- k. Also trips reactor coolant system recirculation pump motors.
'# Final setpoint to be determined during startup testing.
" Initial setpoint.
HATCH - UNIT 2 3/4 3-5 Amendment No. 8,'29, 53
l o . INSTRUMENTATION 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION ( LIMITING' CONDITION FOR OPERATION l l 3.3.7 The iso 1'ation actuation instrumentation channels shown in Table 3.3.2-1 shall be OPERABLE with their trip setpoints set consistent with the ~ 1
, values shown in the Trip Setpoint column of Table 3.3.2-2 and with '
ISOLATION SYSTEM RESPONSE TIME as shown in Table 3.3.2-3. , APPLICA8ILITY: As shown in Table 3.3.2-1. l l ACTION: '
- a. With an isolation actuation instrumentation channel trip setpoint i ',
' less conservative than the value shown in the Allowable Values column of Table 3.3.2-2, declare the channel inoperable and place l the inoperable channel in the tripped condition
- until the l
cha'nnel is restored to OPERABLE status with its trip setpoint - adjusted consistent with the Trip Setpoint value. l ' 4 ,
- b. With the requirements for the minimum number of OPERABLE channel $
not satisfied for one trip system, place at least one inoperable,
. channel in the tripped condition
- within one hour.
- r
. c. With the requirements for the minimum number of OPERABLE channels I
( '. not satisfied for both trip systems, place at least one inoper-able channel in at least one trip system" in the tripped condition within one hour and take the ACTION required by Table 3.3.2-1.
- d. The provis' ions of Specification 3.0.3 are not applicable in l OPERATIONAL CONDITION 5.
2 SURVEILLANCE REQUIREMENTS 4.3.2.1~ Each isolation actuation instrumentation channel shall be
'demonstr,ated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST AND CHANNEL CALIBRATION operations during the OPERATIONAL 4 CONDITIONS.and at the frequencies shown in Table 4.3.2-1. 3
$ i
- With a design providing only one channel per trip system, an inoperable channel need not be placed in the tripped condition where this would l cause the Trip Function to occur. In these cases, the inoperable '
channel shall be restored to OPERABLE status within 2 hours or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken. ( **If both channels are inoperable in one trip system, select at least Cs one inoperable channel in that trip system to place in the tripped condition, except when that would cause the Trip Function to occur. HATCH - UNIT 2 3/4 3-9 Amendment No. 8
R . I ) ( . . j TA8LE 3.3.2-1 h
--e ISOLAT1001 ACTUATIOII INSTRUMElliAT I000 n
Z . VALVE CROUPS MINIMUM IIUISER APPLICASLE
' OPERATED BY OPERABLE CHANIIELS OPERATIONAL' g TRIP FUIICTl006 . S IGNAL( a ) PER TRIP SYSTEM (b)(c) CONDITION.
ACTim M 1. PRIMARY CONTA111 MENT ISOLAT1006
- ' i. 8 N a. Reactor Vessel ihter Level
- 1. Low (Level 3) 2, 6, 10, 2 1,2,3 29 (2821-N680 A, 8, C, D) 11, 12
- 2. Low-Low (Level 2) 5, E,
- 2 1, 2, 3 d 20 (2821-N682 A, 8, C, 0)
- 3. 2 1,2,3 20, Low-Low-Low (Level 1) 1 .
(2821-N681 A, 8, C, 0)
- b. Drywell Pressure - High 2, 6, 7, 10, 2 1,2,3 20 l (2C71-N650 A,'8, C, 0) 12, g, *
- c. Mein Steam Line
- 1. Rediation - High 1, 12 g,(d) 2 1, 2, 3 21 (2D11=K603 A, 8, C, 0)
- 2. Pressure - Low I 2 1 22 (2821-N015 A, 8, C, 0)
- 3. Flow - High 1, # 2/line 1, 2, 3 21 (2821-II686 A, 8, C, D) ,
w (2821-N687 A, 8. C, D)
) (2821-N688 A, 8, (2821-N689 A, 8, C,
C, D) D) w
/
w
- d. Main Steam Line Tunnel Temperature - Hign 1 2/line 1, 2, 3 21 (2821-N623 A, 8, C D)
(2821-N624 A, D, C, D) (2821-N625 A, 8, C, D) (2821-N626 A, 8. C, D)
~
- e. Condenser Vacuum - Low I 2 1, 2,,3 23 (2821-N056 A, 8, C, D)
- f. Tur81ne Bul161pg Area Tempera t.are - High 1 ,
28 1, 2, 3 21 (2061-ROO1, 2061-R002, 2U61-R003, 2061-R004) 3 W
=
tt
,mwans -
'n w
* . :. : . . , :. ..c -
. . .'., ?.4gf,fgg *M .. . . . , . .:: .#.\u ., . ' ' ..k::b..R
. TABLE 3.3.2 1 (Continued) z
$ 1SOLAT1000 ACTUAT1018 IItsTRupeENTAT1981 9 MINIMUM leUISER APPLICABLE s VALVE GROUPS OPERAT[0 BY OPERABLE CHANIIELS OPERAT 4010AL SIG9tALiel PER TRIP SYSTEM (blic) COIIDITION ACTien E TRIP FuseCT1088
- 2. SEC0f00ARY CONTAlaspeENT I SOLAT 1001 m /
- e. Reactor Sullding Exhaust 1,2,3,5 and** ' 26 Radletion - High - 6, 10, 12,
- 2 (2011-K609 A, S, C, D)
- b. Drywell Pressure - High 2, 6, 7, 10, 2 1, 2, 3'- M (2C71-N650 A, 8, C, D) 12, #, *
- c. Reactor Vessel Water 24 Level - Low Low (Level 2) 5. #,
- 2 1, 2, 3 (2821-18642 A, B, C, D)
- d. Refueling Floor Exhaust Radletion - High 6, 10, 12, g,
- 2 1,2,3,5 and** 24 (2011-K611 A, 8, C 0)
- 3. REACTOR WATER CLEA80*JP SYSTEM RSOLAT1018
- e. A Flow - High (2G31-le603 A, s) S 1 1, 2, 3 25 w b. Area Temperature - High 5 1 1,2,3 25 8
(2G31-80662 A, D, E, H J M) l N
- c. Aree Ventilation A Temp. - High 5 1 1, 2, 3 25 (2G31-N663 A, 0, E H, J, M; 2G31-N661 A, D, E, H, J, M; 2G31-le662 A, D, E. H, J, M)
- d. SLCS Initiation (NA) 58*8 NA 1,2,3 25
- e. Reactor vessel Water Level - Low Low 5. #,
- 2 .
1, 2, 3 25 le (Level 2) (2821-N682 A, C. C D) 1 = 3 rv r......
, x o
) ,) _l
m -
~ . . . ,
e~ TAsLI 3.3.2 1 (Continued) g
--e ISOLAfl018 ACTUATICII INSTRUMENTAT ICII c4 2 VALVE Oft 00PS MINIMUN IOUMBER APPLICASLE e
OPERATED BY' OPERASLE CHANIDELS OPEftAT I0 stat. TRIP FussCTeost $10stAL( s ) PER TRIP SYSTEN(b)(c) COIIDIT ICII ACTION E
- 4. ' HICH PRES $URE COOLANT litJECTl001 SYSTEM ISOLATI0el .
m a. HPCI Steam Line Flow 4 High (2E41-N657 A,5) 3 1 1,2,3 bf - l
- b. HPCI Stese Supply Pressure - l Low (2E41-N658 A,8,C,D) 3, a 2 1, 2, 3 M I
- c. HPCI Turbine Exhaust Diaphrege a Pressure - High (2E41-N655 A,8,C,0) 3 2 1. 2, 3 N s
- d. HPCI Pipe Penetration Room Toeperature - High (2E41-II671 A, 8) 3 1 1,2,3 26 l
- e. Suppression Pool Area Ambient Tempera ture-High (2E51-II666 C, D) 3 1 1,2,3 26
- f. Suppression Pool Aree
& Toep.-High (2E51-se665 C, D: 3 1 1, 2, 3 26
- w 2E51-se663 C, D:
) 2ESI-14664 C, D) w g. Suppression Pool Area Temperature .'., Tleer Relays (2E41-M603 A, e) 8 3 '8 1 1,2,3 26 w
- h. Emergency Area Cooler Temperature-High (2E41-h670 A, B) 3 1 1,2,3 26 3 Drywell P re s su re-Hi gh a (2E11-N694 C, D) 6 1 . 1,2,3 26 l J. Logic Power Monitor (2E41-n1) asA' a ' 1 1, 2, 3 27 l l
s . et 3 O , M.* *s.& w , e i ,
- h .w :.
ISOLA AT1001
.c
-e VALVE CAOUPS MINipalpl IguemER APPLICASLE O OPEllATED SY OPERASLE CHAIIIgELS OPERAT ICIIAL
- TR1P FusICT1011 8ISISAL(e 1 PER TRIP SYSTEft(b)(o) C W ITIGN ACTION e ,
c- 5. REACTOR CORE ISOLAT1001- 2 ExxPLI E 5Y3 En 150LATl04
- a. RCIC Steam Liene Flow-High 4 1 ~1, 2, 3 ,I' . N
~ (2E5i-N65T A,s) .
- e. '9:
- b. RCIC Steen Supply Pressure -
Low (2E51-II658 A, 8, C, D) as , 9 2 1,2,3 N l
- c. RCIC Turbine Exhaust .
Diaphrege Pressure - High 4 2 1,2,3 . N (2E51-N685 A, 8, C, D) l
- d. Emergency Area Cooler Temperature -
High (2E51-N661 A, 3) 4 1 1, 2, 3 N l
- e. Suppression Pool Ares Ambient Tempe ra tu re-Nigh 4 1 1, 2, 3 26 (2E51-N666 A, 8) l l %" F. Suppression Pool Aree & T-High 4 1 1,2,3 24
* (2E51-N665 A, 8; 2E51-N663 A,8; l w 2[51-N664 A,B) 3
% g. Suppression Pool Area Temperature Timor Relays (2[51-M602 A, B) 4g,, 1 1,2,3 M
- h. Drywell Pressure - High (2E11-II694 A, 8) 9 1 1, 2, 3 26 l
- 1. Logic Power Monitor (2E51-K1) NA'"' 1 1,2,3 27
- 6. SMUTDOWIl COOLIIIC SYSTEM I SOLAT ICII
- a. Reactor vessel Idster Level-Low (Level 6, 10, 11, 2 2 3, 4, 5 26 3)(2821-N680 A, 8, C, D) 12 .
- b. Reactor Steas Does Pressure-High 11 1 1,2,3 29 l (2831-N679 A, 0)
\
r+ l h l l
%e * . . .
w 'e
*4
. .f *fS .
t 43 4 , j
. .. g I- -
l 1
. , . , . . :, p a' p .3
- . .y ,
;iM ~klt$,ey6L;.. %d'. kU2.
f .,-
. ( .
. ,.c. ..nb % e.i . )
,...-)
TA8LE 1.3.2-1 (Carticued) 150LAT!0N ACTUAT!ON INSTRUNENTAT!0N
,L
( .. ACT!0N ACT!0N 20 - Se in at least HOT $ HUT 00WN within 6 hours and in COLD SHUT 00WN within the next 30 hours. ACTION 21 - Se in at least STARTUp with the main steam line isolation valves closed within 2 hours or be in at least NOT SHUTOOWN within 6~ hours and in COLO SHUTOOWN within the next 30 hours. ACTION 22 - Se in at least STAATUP within 2 hcurs.
' ACTION 23 -
Be in at least STARTUP with the Group 1 tsolation ' valves closed within 2 hours or in at least HOT SHUTOOWN within 6 hours. ACT!fM 24 - Establish SECONDARY CONTADetENT INTEGA!'TY with the standby gas treatment systee operating within one hour. l ACTION 25 - Isolate the reactor water cleanup systes. . \ . l ACTION 26 - Close the affected system isolation valves and declare the i affected systee 'noperabic. . ACTION 27 - Vertfy power availability to the bus at least once per 12 urs ' or close the affected system isolation valves and declare
- affected systee inoperable.
ACTION 28 - Close the shutdown cooling supply and reactor vessel head pray isolation valves unless reactor steam does pressure t 145 psig:
.)
NOTES -'
- # Actuates operation of the main control room environmental control systes in the pressurization mode of operatte,n. l Actuates the standby gss treatment systee. l When hand 1tng irradtated fuel in +.he secondary containment. -
- a. See Specification 3.6.3, Table 3.6.3-1 for valves in each valve group.
- b. A channel may be placed in an inoperable status for up to 2 hours for required surveillance without placing the trip system in the tripped condition provided at least one other OPERA 8LE channel in the same trip-system is monitoring that parameter. '
. c. With a design providing only one channel per trip system, an inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, the inoperable channel shall be restored to OPERA 8LE status within 2 hours or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken.
- d. Trips the mechanical vacuum pumps.
e. A channel is OPERABLE if 2 of 4 instruments in th'at channel are OPERA 8LE.
- f. May O* sypassed with all turbine stop valves closed.
l
- g. ? a u s only RWCU outlet isolation valve 2G31-F004.
I
- h. As.ia oly.
s
- 1. Adjustable up to 60 minutes.
HATCH - UNIT 2 3/4 3-15 Amendment No. j, 39
CONTAINMENT. SYSTEMS 3/4.6.5 SECONDARY CONTAINMENT (' SECONDARY CONTAINMENT INTEGRITY
~.. .
LIMITING CONDITION FOR OPERATION
-s 3.6.5.1 Hatch-Unit 2 SECONDARY CONTAINMENT INTEGRITY and Hatch-Unit'l secondary containment integrity shall be maintained. .
-s APPLICABILITY: CONDITIONS 1, 2 and 3. .
ACTION: ( Without Hatch-Unit 2 SECONDARY CONTAINMENT INTEGRITY and/or without Hatch-Unit 1 secondary containment integrity, restore Hatch-Unit 2 SECONDARY CONTAINMENT INTEGRITY and Hatch-Unit 1 secondary containment integrity within 4 hours or be in at least HOT SHUTDOWN within the next + 12 hours and in COLD SHUTDOWN within the following 24 hours. - SURVEILLANCE REQUIREMENTS 4.6.5.1.1 Hatch-Unit 2 SECONDARY CONTAINMENT INTEGRITY shall be demon-strated by:
- a. Verifying at least once per 31 days:
- 1. All equipment hatches are closed and sealed, and
- 2. At least one door in each access to the secondary l containment is closed. '
- b. Verifying at least once per 92 days that each secondary contain-ment ventilation system automatic isolation damper is CPERABLE
' or-secured in the closed position per Specification 3.6.5.2.
- c. At'least once per 18 months:
- 1. Verifying that one standby gas treatment subsystem will draw down the secondary containment to 2 1/4 inch of vacuum water gauge in s 120 seconds, and .
- 2. Operating one standby gas treatment subsystem for one hour and maintaining 21/4 inch of vacuum water gauge in the secondary containment at a flow rate not exceeding 4000 CFM.
4.6.5.1.2 Hatch-Unit 1 secondary containment integrity shall be demon-strated per Hatch-Unit 1 Technical Specifications. HATCH - UNIT 2 3/4 6-36
( 3/4.'6 CONTAIMENT SYSTEMS 3/4.6.1 PR$DYChNTAINMENT , PRIMARY CONTAIMENT'!NTEGRITY
~
l l LIMITING CONDITION FOR OPERATION - l ' I 3.6.1.1 PRIMAP.Y CONTAINMENT INTEGRITY shall be maintained. APPLICABILITY: CONDITIONS 1, 2* and 3. ACTION: 1' Without PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT , INTEGRITY within 1 hour or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. . 9. 9 a4 ,
~ T$Q 4 l
SURVEILLANCE REQUIREMENTS ' 1 ,,
.h s- .
4.'6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be demonstrated: .' .' - '_
- a. At least once per 31 days by verifying that; ;j*
. a.
- 1. All penetrations
- not capable of being closed by OPERABLE .
containment automatic isolation valves and required to be . .$ closed during accident conditions are closed by valves, - blind flanges, or deactivated automatic valves secured in .x. position, and
].}
- 2. All equipment hatches are closed and sealed. ...
.+
~
~. s y s} %.;
y -
, , M '
~@,
. 4!
, n. *See Special Test Exception 3.10.1 :
i j *Except valves, blind flanges, and deactivated automatic valves which L
! are located inside the containment, and are locked, sealed or otherwise . ! ; secured in the closed position. These penetrations shall be verified
- ( closed during each COLD SHUTDOWN except such verification need not be j 'g, '
performed more often than once per 92 days. i i HATCH - UNIT 2 3/4 6-1
SURVIILLANCE RfoulREMENTS LIMITING CONDITIONS FOR OPERATION
.~ *
- 2. primary Containment Inteerity 4.7.A.2. Leak Testine te Verify .
Primary Containment Inteerity Primary containment integrity is required: Primary containment integrity shall be demonstrated by the
- a. Prior to withdrawing following test procedures:
control rods for the Tvne A Tenth - Inteerated leak purpose of going critical. a. tate Test t' LRT1*
- b. Whenever the reactor is critical.
. Primary containment integrity is confirmed if the leak rate does -
- c. Whenever the reactor water not exceed the maximum allow- -
temperature is above 212*F of 1.2 able weightleak rate La,he percent of t contained -- and fuel is in the reactor air per 24 hours at the peak vessel. f.
- test pressure. i.
An exception is made while per- . .( forming low power physics tests (1) Type A tests shall be perfonned ' at atmospheric pressure at power under the program established a levels not to exceed 5 ftft, during in Appendix J of 10 CFR which time primary containment Part 50. (Reference 1). integrity is, not required. i I
- La - Maximum allowable peak pressure test leak rate - 1.2 weight percent per day
- Maximum allowable reduced pressure test leak rate I Lt l Lam - Measured peak pressure test leak rate - values are subject to change with each ILRT performed i
Lt m - Measured reduced pressure test leak rate - values are subject to change with each ILRT performed Lao - Allowable operational leak rate for peak pressure tests - values are i subject to change with each ILRT performed Lt o - Allowable operational leak rate for reduced pressure tests - values are subject to change with each ILRT perf'ormed (All leakage rates measured in weight percent of contained air per 24 hours) P a
- Peak test pressure - 59 psig Pt
- Reduced test pressure - 29.5 psig HATCH - UNIT 1 3.7-2 i
, - - - . - - , - - . - - , - , . - - , . _ , . , - ~ . .-,..-...~_.m.
--.----- , . , - . - - - . - . - - - - . , - - - - - . . - - - - . . - - - - - - - - , - - - , . - - . . - - - - - . , . - - . - - - - - , - . ~ - - .
l 0 i
-* SURVEILUUICE REcu1REMENTE LIMITING CONDITIONS FOR OPERATION 7 l
.- )
C. Secondarv Cantainment* l C. Secondary Centainment ' j 1 i 1. gecondary Containment Inteerity 1. Survei' lance m ile Inteerity I Rainta' eed l a, Integr.ity of the secondary con- ' tainment shal,1 he maintained Secondary containment surveillance - during all modes of Unit 1 plant shall be performed as indicated ' operation except when all of the below: following conditions are met: (1) The reactor is subcritical and Specification 3.3.A. is est. a. A preopertional secondary contain- l eent capability test shall be _. l (2) The reactor water temperature is conducted after isolating the j below 212'F and the reactor secondary containment and placing coolant system is vented. the standby gas treatment system ; filter trains in operation. Such i (3) No activity is being performed tests shall demonstrate the capa-which can reduce the shutdown bility to maintain a minimum - i margin below that stated in 1/4-inch of water vacuum under Specification 3.3.A. calm wind (<$ aph) conditions with each filter train flow rate not l
- (4) The fuel cask or irradiated fuel more than 4000 cfs. . ,
is not being moved in the re-
. actor building. b. Secondary containment capability to maintain a sinteum 1/4-inch (5) All hatches between Unit 1 of water vacuum under cela wind secondary containment and (<$ aph) conditions with each Unit 2 secondary containnsnt filter train flew rate not more {C; are closed and sealed. than 4000 cfm saall be demohstrate6 '
at each refueltag outage, prior te (6) At least one door in each refueling. access path between Unit 1 - secondary containment and Unit . 2 secondary containment is closed.
- b. Integrity of the Unit 1 secondary containment shall be maintained during all modes of Unit 2 plant operations except Operational Condition'4 as defined in the Unit 2 Technical Specifications.
*For secondary containment during 1982 refueling outage, see page 3.7-12a.
HATCH - UNIT 1 3.7-12 Amendment No. 22. pp. M. 91
SutVEILLANCE RE0uitENENT3 (. LIMITING CONDITIONS FOR OPERATION 4.7. C .1. Survei' 1ance tettle Inteeritw Maintacnea (Cent'd)
. . b. Secondary containment capability to 1 maintain a minimum 1/4-inch.of water i * *
. iiscuum under calm wind ( 5 mph) I
. conditions with each filter train '
f1sw rate not more than 4000 cfm ! shall be demonstrated at each ; refueling outage, prior to refueling. j 3.7.C.2. Violation of Secondarv 2. Surveillance After Jnteurity Violated containment inteerity , i
- s. Without Match-Unit 1 second- After a secondary containment viola-ary containment integrity, tien is determined the standby gas restore Match - Unit 1 sec- treatment system will be operated .
j ondary containment integ- tamediately after the affected zones j rity within 4 hours, or pere are isolated from the remainder of i fore the following (as appli- the secondary containment. The' l t cable): ability te maintain the remainder
- j (1) Suspend irradiated fuel of the secondary containment at * )
and/or fuel cask handling 1/4-inch of water vacuum pressuPe in the Match-Unit 1 sec- under cals ( 5 mph) wind conditions ondary containment. shall be confirmed. j
-(2) Se in at least Not Shutdown - !
within the next 12 hours and . l aset the Conditions of 3.7.C.1.a. within the next 24 ) hours. . j
- b. Without Match-Unit 1 secondary containment, refer to the follow-ing Match-Unit 2 Technical Specification, for LCO's to be followed~for Match-Unit 2:
(1) Section 3.6.5.1.
. (2) Section 3.g.5.1.
Primary Containment Isolation Valves D. Primary Containment Isolation Val'ves D.
.1. Valves Reauired to be Doerable .1 . Surveillance of Goerable Valves Surveillance of the primary ton-During reactor power operation, all primary containment isolation tainment isolation valves shall be valves listed in Table 3.7-1 and perforined at follows:
all reactor coolant system instru-ment line excess flow check valves a. At least once per operating shall be' operable except as stated cycle the operable isolation in Specification 3.7.0.2. valves that are power operated I and automatically initiated l
' shall be tested for simulated automatic initiation and the closure times specif.ied in 1
. Table 3.7-1.
3.7-13 Amendment No. ##, 56, 97, 100 HATCH - UNIT 1
i
- 5. T OF NUCLEnR POWEP rLANT OPERATION- FLUIDS, AND PAGE 20
_____'ICOPY___________________________________________________ TuEPMCDYNAM!CO ANS'4 E? S -- MATCM 13.2 -86/02/17-BROCKMAN, K. A N C 22 7' 5.01 (2,50)
- a. P-essure decrease is from the reduced head losses associated with the steam flow (flow through the SRV requires less driving force). (0.5) b, Indicated total steem flow is reduced since less steam is
'seen" by the flou-restrictors. The SRV is upstream of the flow sensors. - OR - EllC reduces flow in response to pressure dactease. Les5 flow is "seen* by the flow restrictors. (0.5)
. Tvrbine steem flow is reduced by the EliC sys tem proportionate
- 6. o the SRV release (interactive with pressure control). (0.5) a 'eoel increste is e qerienced in the annulus as the pressure r e.L.'c t ion c a v ie s "oid l'o r m a t ions in the core. (0.5) 3 'evel reduction is controlled by the FWCS). Level is stabili_ed at point where the level error mat Nes the steam l
'1:w/"w.d flow aismetch tignal. (0.5) l RCCEPEMCE C:lt
- GPNT, Vol VII. Chapter 10.4 AUSEP 5.02 (1.50)
1 b P2 c r; ,
_! c
', i i3. b (.25 each)
REFERENCE
- 5. ! M : CP9T, Vol VII, Chapter 10.2-23 ANSUIR 5.03 (1.00) 2
-REFEREMCE 01 7 : /ENON & SAMARIUM LP, pp 4, 12 l GGNG: OP-NP-514 l
l l
l
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 21 ANC .1S -- HATCH 1&2 -86/02/17-BROCKMAN, K.
EIH: L-RG-606 ANSWER 5.04 (1.00)
'0% void fraction in the core (0.5)
There is a larger % change in water volume for the same increase (3.45% vs 1.1%) (0.5) -OR- The volds produced at 70% VF have a larger effect on core reactivity since they are in an area of higher neutron flux. (0.5) REFERENCE Elll: Peactor Physics t./P. pp 1.7-0, 10, 13; L-RG-604 SFNo* ?e ctivity Coef ficier.t LP, pp 2, 3; RO 85/03/01 ANSWER 5.05 (1.00) b REFERENCE EIH: L-RO-667r p 10 CFNP: Rx Heat Balance LP; RO 85/03/05 ANSWER 5.06 (3.00)
- a. Increased Recire Flow from the Annulus. (0.5)
- b. Increcse in Recire Flow (0.5)
- c. Recire Pump Ronback ( Feedwater Flow ' 20% ) (0.5)
- d. Increasing Reactor Power (0.5)
- e. FtJCS response to Decreasing Water Level (0.5)
- f. EMC controllins Reactor Pressure (with TCV's) (0.5)
REFERENCE EIll: L-RG-732 (15) I l I
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 22 ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
ANSWER 5.07 (2.00)
- a. 295 deg F (+- 15 des F) (0.5)
- b. Increase (0.5)
- c. Increase (0.5)
- d. 450 psia (+- 50 psia) (0.5)
REFERENCE Steam Tables /Mollier Diagram ANSWER 5.08 (1.00)
- 1. Hf pins are located in the high flux regions of the control rods.
- 2. New roller and pin materials are used.
- 3. Thinner sheath materials are used to maintain the weight of the rods constant with the previous rods.
- 4. The last three pins in each blade of the cruciform are Hf, thus maintaining the control rod's worth effectively constant to previous.
- 5. Control Rod tube cr ack.ing over design lifetime is reduced (structural lifetime is extended somewhat).
- 6. Nuclear lifetime is increased (due to Hf daughters having reasonably large thermal neutron absorbtion cross-sections).
- 7. Decrease in He (Alpha) build-up. (2 0 0.5 each)
REFERENCE EIH' L-90-605, pp 2, 3
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 23
--- inEER667sIsiCS
' ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
ANSWER 5 09 (1.00)
- a. 1 ,
- b. 4 (0.5 each)
REFERENCE BFNP' PUMP CH',RACTERISTICS, PUMP HEAD, PUMP LAWS LP,P.4 GGNS: OP-NP-504; OP-NP-514 Elit : L-RG--6o5 ANSHER 5.10 (2.00)
- a. 2
- b. 3 (Or, as appropriate for answer given in (a))
REFSRENCE General Electric NEDE 21493, February 1982 EIH: GPNT, STA Training Manual, Section 9 GGNS! MCD, PCIOMR:
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 24 ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
ANSWER 6.01 (2.00) RED - Solenoid control valve has energized (0.33); Energized (0.33) GREEEN - Power available to the solenoid control valve (0.33) Inergi:ed (0.33) AMBER - Pressi.ere in ta11 pipe (>85 psig) (0.33) Energtzed (0.33) REFERENCE
- EIH' CpHT. Vol. VI, Chapter 5.1.II.B.2
- L-RQ-703 i ANSWER 4.02 (2.00)
- a. TRUE (0,5)
- b. FALSE (0,5)
- c. TRUE (0.5)
- d. TRUE (0.5)
PEFEREtlCE EIli" L-PO-741 ANg aro 6.03 (1 00) a REFERENCE EIH' GPNT. Vol. VI. Chapter 5.5-2; Vol. VII, Chapter 9.4-19, 203 L-RO-705
6o PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 7 PAGE 25 ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K. ANSWER 6.04 (1.50)
- a. Reduce the concentration of ouygen for fire prevention considerations (0.5).
- b. Provide sufficient water for steam condensation (+0.5) and enough air space for the primary containment non-condensibles (+0.5)
REFERENCE BFNP: Lesson Plan 16, Objectives C, 0, H, I, pp 1, 14, 26, & 32 EIH: L-RG-707 ANSWER 6.05 (1.50) Recirculation Pump RPT D r e a k.e r Permissive RSCS Cypass T3V. 10% Closure Scram Sypasss TCV, Fast Closure Scram Bypass (3 0 0.5 each) REFERENCE EIH: GPNT, Vol V, Chapter 4.1-9; GPNT, Vol VI, Chapter 5.5-10; GPNT, Vol VII, Chapters 9.2.2-11 and 9.2.3-5 ANSWER 6.06 (2.00)
- a. On / Green
- b. Off
- c. On / Clear (Ind 9 0.5; Color 0 0.25)
REFERENCE EIH: L-RG-704 (6) l l I l 1 i l l
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 26 ANSWERS -- HATCH 18.2 -86/02/17-BROCMMAN, K.
ANSWER 6.07 ( .50) FALSE REFERENCE EIH: EIH Simulator; b '- S I ED s u. r ,.,c n,
.amn , . mas
- a. ADS Valves are in Double Boxes (0.5) /
- b. (1) No Demand (0.25) or Demand Me,t t M (2) ADS Armed (0.5)
- c. 60 Minutes (AA f 5nd 6 minutes (0.25)
REFCP" __
- u. t_c.n_ 57 <1091 - gcn ;57 -- -
ANSWER 6.09 (1.50) 11 min 51 see (Or il Min) (0.15) -OR- 1.85 t D/4 Pressure (0.15)
- 121.5 RWL 100 psig RHP/CS + 95 '
RWL 2 min T.D. (0.3 each) PEFE<ENCE EIH: L-RG-767; 34AR-OPS-010-2 i
a e
- 6. PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE 27 ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
ANSWER 6.10 (2.00)
- a. Unit 1
- b. Unit 2
- c. Neither
- d. Unit 2 REFERENCE EIH: L-RO-742
i
- 7. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 20
------~~--------------~~ ~~~~R565UL55iEEE E5sTR6t ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
i ANSWER 7.01 (2.00) a) 1. RPV level cannot be maintained (0,5)
- 2. Suppression Pool uater temperature cannot be maintained below 110 deg F (0,5) b) 1. Five (5), or more adjacent control rods not inserted below the 06 position (0.5)
- 2. Thirty (30), or more total control rods not inserted below the 06 position. (0.5)
REFERENCE EIHi 34AD-0PE-009-2 i ANSWEP 7.02 (1.00) d REFERENCE E DI* 34AD-OPS-002-2 i ANENEP 7.03 (1.00) Cloth and Rubber Gloves (0.5) Cloth and Rubber Ooots (0.S) REFERENCE EIH: GET Handbook; 60AC-HPX-004-0 . ANSWER 7.04 (1.00)
- 1. Prevent E:< c e s s i v e Primary Containment Pressure (57.5 psig)
- 2. Control Combustible Gas M i:: t ur e s (HYDROGEN < 4 v/o and OXYGEN 5 v/o) (0.5 each) l REFERENCE ETH: 3 4 A D - O P '3 - 0 0 6 - 2 i
1 l
t o . I i
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 20 l ------------------------------------------------
RADIOLOGICAL CONTROL l _-_-__------_-----_- ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K. l ANSWER 7.05 (1.00) To prevent hydroing the Recirculation Pump (to CRD System pressure) upon startup of the CR0 pump. - OR - To preclude cuction and/or discharge valve actuator inoperability due tc excessive dP. FEFERENCE . EIH: 34SO-C11-003-2; 34S0-831-002-2; L-RO-714 l l ANSWER 7.06 (2.00)
- a. 2 Minutes (0.5)
- 6. Shallow (1.0)
- c. ' Print Notch Error
- function (of RWM) (0.5) l l REFERENCE i EIH; 42FH-ENG -010-1/2 l
l AriSWEP 7.07 (2.00)
- 1. Inability of High Pressure Systems (HPCI/RCIC/CRD) to restore level.
2, noe or n. ore Low Pressure Systems running (LPCI, CS, or 1 CP & CBP)
- 1. Reactor Prensure 350 psis.
J. Praetor Level 121.S* (0.S each) rcccocVCE e E T:t: 1430-47.' 001 ?4AB-OPS-002-2 ANS4EP 7.00 (1,00)
- a. Unit 1 (0.4) Shift Supervisor (0,1)
- b. FALSE REFERENCE EIH; HNP 34AB-FPX-040-0, pp 1, 2 l
{
. .. s
- 7. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 30
~~~'RE5 5t55iEEL E5sTR5t ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
ANSWER 7.09 (2.00)
- a. Alert
- b. (Notification of) Unusual Event
- c. General Emergency
- d. Site Area Emergency (0.5 each)
REFERENCE EIll: GET Handbook, pp 57, 50, 60, 61; 10AC-MGR-006-OS BFNP' DFN-IPD, IP-1, p 1; RO 85/04/01 ANSHER 7.10 (1.00) Orywell Cooling Fans when Cli-2] >= 2% (0.5) Drynell Spray if Pressure (D/W) >= 17 psig (0.25) and RVL is Normal (0.25) RFFERENCE EIH: 34AB-OPS-002-2 ANSWER 7.11 (1.75)
- a. 5
- b. 7
- c. 4
- d. 6
- e. 2
- f. 3
- c. 1 (0.25 each)
REFERENCE Elll: L-RO-715; 10AC-MGR-003-OS l 1 t i i
80 ADMINISTRATIVE PROCEDURES. CONDITIONS, AND LIMITATIONS PAGE 31 ANSWERS -- HATCH 182 -86/02/17-BROCKMAN, K. ANSWER 8.01 (1.50) (1) Highest worth rod (0.25) fully withdrawn (0.25) (2) Xenon free (3) Cold (60 deg F) (0.5 each) REFERENCE EIH: U2 TS, 1.0 "SDM' ANSWER 8.02 (2.00) l a. (3)
- b. (1)
REFERENCE EIH* L-RO-672 (15); U2 TS 3.2.3 ANSWER 8.03 (2.50)
- a. (1) 2 (2) rully inserted - OR - located in the quadrant where core alterations are taking place and in an adjacent quadrant.
(3) 3 cps (J) 24 hours b, TPUE REFERENCE E!H: U1 TS, 3.10.C.1, 3.10.C.2, 3.10.H AN5WER L.3a 2.00
- s. 3) (1.0)
- o. '00000' indicatas tnat tnere is no Negstive G r :- c e fertos for tnis surve111snee.ic.51 It aculd be spplicable for Unit 2. 0.5) s' 51 n e s- Unit i silows :. : n A re D nei n e s 257. Intervs1 .arlances REFEREr4CE
, E H: 40AC-REG-04-0 p 12 l l l t i
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 32 ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
ANSWER 8.05 (1.00) c REFERENCE EIll: 10AC-MGR-003-OS, p 44 ANSWER 8.06 (1.00) Turn the handwheel in the DESIRED POSITION direction (1/4 turn ma:t )(0.5 ) ; Verify the lockin3 device integrity and proper installation by attempting to mi'sposition the valve (0.5) REFERENCE E I!!! 14GO-SUV-001-03 A N'3 J ER B.07 ( .50) None (0.5) l R'IFERENCE l LI!!! 30AC-OPS-001-OS ANSWER 0.00 (2.00) Clearance Number Sequential Tag Number Component Identification Required Pocilion of Component (0.5 each) REFERENCE TIll
- 10AC-OPS-001-OS l
i i 1 L }
l t
- 8. ADMINISTRATIVE P:ROCEDURES, CONDITIONS, AND LIMITATIONS - PAGE 33 ANSWERS -- HATCH 1&2 -86/02/17-BROCKMAN, K.
ANSWER 8.09 (1.00) Operations SUPERINTENDENT Superintendent-Plant Engineering Services Maintenance Supervisor-Electrical (2 0 0.5 each) l REFERENCE EIH; HNP-0-ADM-0508 ANSWER 0.10 (1.50) The activity should NOT be initiated. (0.5) Unit 1 TS's 3.7.A.2, 3.7.C.1, and 3.7.C.2 establish the need for Secondary Containment Integrity for Unit i during this condition on Unit 2. Unit 2 TS 3.6.5.1 is referenced to and provides the l requirement to maintain Secondary Containment Integrity. (0.5) TS 3.0.4 precludes entry into an Action Statement. (0.5) (1.0) REFERENCE EIH: Ut TS 3.7.A.2, 3.7.C.1, & 3.7.C.2; U2 TS 3.0.4, 3.6.1 1, a 3.6.5.1 ANSWER 0.11 (1.00) d REFEPENCE ETH: U2 TS 3.3.1 & 3.3.2 l l
f . o. 7 i i TEST CROSS REFERENCE PAGE 1 QUESTION VALUE REFERENCE 05.01 2.50 KED0000070 05.02 1.50 KEB0000153 05.03 1.00 KEB0000583 05.04 1.00 KEB0000591 05.05 1.00 KEB0000594 i OS.06 3.00 KEB0000595 03.07 2.00 KEB0000596 0'5.08 1.00 ME80000628 05.09 1.00 KEB0000642 05 10 2.00 UEB0000643 16.00 06.01 2.00 UE90000004 06.02 2.00 MEB0000324 06.03 1.00 KE00000998 06.04 1 50 eE00000600 06 05 '
,50 -
J 0 ^ ' N, ! 3 06.Oo 7.00 l'LD0000614 0,< .0/ ,So i: i - in- >9 0 6 q <> ;0 K: 3 0 0 0 0 4 1.'
' m
- i. o 'ij i i. , ' -
cf yt . .o.- .) q o.) a t a
- (i n ,
u
.,/ a : 0 0 3 0,' p >
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a ra.
+ , l a g. ,
,n ., j, l ' ' O li i s [ "*, d 4/ # , () N U'f '/ .'i l s n , ..
, ,,c.t.,
nf . ,
,, ,s
,e '; v _r .
11 00.01 1.53 1:: :))00005 03.C2
.0' KEB0000054 0(* 0.3 2.50 KEB0000110 08.04 2.00 KEB0000134 09.05 1.00 KEB0000272 08.06 1.00 KEB0000623 00.07 .50 KEB0000624 08.00 2.00 KEB0000635 08.09 1.00 KEB0000644 00.10 1.50 MEB0000645
F....
~
TEST CROSS REFERENCE PAGE 2 GUESTION VALUE REFERENCE 08.11 1.00 KEB0000646 16.00
--O-9 6 63.75
}
g. Enc 1ccura 4 \ l Plant E. 2. Hatch l, Exon Review-Comments Critten responses are requested regarding NRC resolution of the following quOctions contained in the Plant Hatch NRC examination of February 17, 1986: s - ) w R3 Initial License Exam 1.01, RO Requel Exam 1.01 D l 3>
- 25 Pellet Clad Interaction .
N ha C:ncern: The answer key does not reflect that there is one major reason i.e. fission product (chemical) for the lowering'of the threshold, cabrittlement of the cladding. Neither creepdown, nor pellet growth is a primary reason. As auch, only fission product embrittlement is taught sa tho reesc!n for the threshold lowering, although the phenomena of creepdown
< cnd pellet "yrowth are discussed with reference to PCIONR.
Recommendetions: Modify the key to allow full credit for fission product As a minimum, this reason [ chemical) embrittlement of the cladding. Chruld receive more than half-credit. R forences: L-RQ-673, VI.D. and VI.E.; GPC Hatch Nuclear Training Manual, Ch. 10.2, page 10.2-16 RO Initisi License Exam 1.17, RO Requal Exam 1.10, SRO Requal Exam 5.08 C:ntrol Roda Cencern: There is ambiguity in the wording of the question and the key as to whether advantages or dealgn characteristica are the desired , responses. For this reason, the candidate may list other advantages or dcaign characteristics than are contained in the key.
~
Officia' Copy
g - J R: commendation: Modify the key to also include as acceptable alternate cnewers:
- 1. Hafnium pins are located in the high flux regions of the control rods.
- 2. New roller balls and pin materials are used.
- 3. Thinner aheath materials are used to maintain the weight of the hybrid rod constant with that of the non-hybrid rods.
- 4. The last three rods in each blade of the cruciform are Hafnium, thus ocintaining reactivity of the rods constant.
5 ._ Eliminate / reduce control rod tybe cracking over design life (increased ctructural lifetime).
- 6. An increase in the nuclear lifetime of the roda.
References:
Gpc Hatch Nuclear Training Manual, Ch. 2.4, pp. 2.4-16, 2j4-17; LIMQ-605, IV.A. 3 and 4 RO Initial License Exam 2.04 Main Generator Hydrogen Cooling Cancern: Due to all of the available choices being a factor of ten lower than the intercept point, "d" is technically correct. However, it is believed that this factor of ten problem was unintentional, and as much was subtle enough that most exsminees would not notice this problem, ccpecially since they are familiar with the Hatch specific graph which has MVAR's in units of 100's. Also, the resolution of this graph due to the problems inherent with all copiers is such that the discernment of less than one gradation difference between the desired response a. 320 KVAR, cnd b. 330 KVAR is difficult to make. An examinee could easily have read the intercept point as 325 KVAR and as such chosen a. 320 KVAR, or as 330 KVAR and chosen answer b. 330 KVAR. Recommendation: Modify the key to allow responsen (a), (b), or (d) for full credit. The only acceptable alternative to this modification would b3 to delete this question. 345U-N40-OO1-2, Figure 1 (Attached) I
References:
1
l RO Initial License Exam 4.17 b. and c. _ Cancern: The referenced procedure, 42CC-ERP-OO9-2S, " Maintaining Rated Thermal Power", states "The STA will perform the requirements of this procedure with the aid from an on-shift licensed operator if neccessary." The only time emaistance would be required would be if the maximum oteady-state power level were exceeded and a reduction in recirc flow were required or if power level was less than 2420 Mwt and an increase in rceirc flow was required. In either case this easistarce would be directed via the shift supervisor as the STA is not allowed to directly crder power changes. As such, questions (b) and (c) are not RO tasks, but STA tasks. The plant specific Job Task Analysis also makes this _ distinction.
~
R0commendaElon: Delete questions (b) and (c) from the RO Initial License Examination.
References:
42CC-ERP-OO9-2S, 34GO-OPS-OO5-2S CRO Requalification Examination 8.04 b. C:ncern: The significance of this item is not auch as to warrant its ocmorization by an SRO. In the event that the shift supervisor were to rcceive a surveillance sheet with 00000 in the earliest date block, and was unsure of the meaning, he would certainly not use memory as the source of his information. It is for this reason that procedures are written. l It is unreasonable to expect an SRO to memorize the entire spectrum of dotails contained in the administrative procedures. 1 i R commendations: Delete 8.04 b. I l l mas
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