ML20136C064
| ML20136C064 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 11/18/1985 |
| From: | Burdick T, Mcmillen J, Reidinger T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20136C053 | List: |
| References | |
| 50-454-OL-85-01, 50-454-OL-85-1, NUDOCS 8511210013 | |
| Download: ML20136C064 (11) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No.. 50-454/0L-85-01 Docket'No. 50-454 License No. NPF-23 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: Byron Nuclear Power Plant, Unit 1 Examination Administered At: Byron and CE Production Training Center Examination Conducted: SRO and R0 written and operating
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Examiners:
lhC&d T. Burdick A
l- -W Date f'h.h5 Date Approved By: J. I. McMillen, Chief //- /f-83' Operator Licensing Section Date Examination Summary Examination administered on September 16-20 and October 14-18, 1985 (Report No. 50-454/0L-85-01)
Ten Reactor Operator candidates and eight senior Reactor Operator candidates were examined. Six Reactor Operator candidates and five Senior Reactor
- candidates were issued licenses based upon satisfactory completion of the examinations. One Reactor Operator candidate was denied a license for failure of the walk-through examination. One Reactor Operator candidate was denied a
-license for failure of both the walk-through and simulator examinations. Two other Reactor Operator candidates were denied licenses for failure of the simulator examinations. Three Senior Reactor Operator candidates were denied
~1icenses for failure of the simulator examinations.
8511210013 851118 PDR ADOCK 05000454 G PDR
REPORT DETAILS
- 1. Examiners T. Burdick T. Reidinger F. Jaggar, INEL B. Picker, INEL P. Isaksen, INEL
- 2. Examination Review Meeting No meeting was held. The examiners left copies of the SR0 and R0 exams with the licensee for review and comment. Comments and their resolutions are enclosed as an attachment to this report.
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- 3. Exit Meeting On September 20, 1985, the Region III examiners met with members of the Byron plant staff and the NRC resident inspectors to discuss generic weaknesses identified during the operating examinations. One major concern the examiners had was the candidates' use of symptomatic procedures. It appeared that the candidates would not take corrective action unless directed to do so by the procedure which is not entered unless there is a reactor trip or safety injection. A number of situations arose in which the event did not cause an inmediate reactor trip or safety injection, such as steam generator tube ruptures, feed line breaks, or primary depressurization. In these cases the operators demonstrated a lower level of confidence and a reduced capability to terminate the event in a timely manner because they did not enter the symptomatic procedure for an extended period of time. When the examiners suggested that the operators should initiate the necessary reactor trip and/or safety injection to enter the procedures a licensee staff member stated such action causes an undesired shock to the plant systems.
On completion of the operating examinations during the week of October 14, 1985, the INEL contract examiners met with members of the plant staff and NRC residents to discuss generic weaknesses. They cited the following items:
- a. General unfamiliarity with Nuclear Instrumentation systems.
- b. Superficial usage and evaluation of Technical Specifications.
- c. E-0 does not verify total ECCS flow,
- d. Procedure usage and compliance.
- e. There is no clear acceptance criteria for channel check performance.
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Byron Reactor Operator Examination .
Facility Comment
~1.6 Candidates may have used 1x10 ' amps as point of adding heat. This information would have come from Byron Startup Test 52.33.
Examiner's Response Although 1% is approximately closer to 10 ' amps the answer key also allowed (1 to 10) x 10 ' amps as acceptable. Answer key was amended.
Facility Comment 1.11 a. Alternate answer: Condensate depression is the amount condensate is subcooled.
Examiner's Response Although the answer key is the technical definition, the alternate answer will be acceptable.
Facility Comment 1.12 Alternate answer: "a" section of curve more restrictive because in case of small break LOCA top of core could be uncovered longer. (uncovered first recovered last)
Examiner's Response The alternate answer by the facility is a translation of the answer key. It is part of the answer key. No amendments to answer key.
Facility Comment 1.13 b. Alternate Answer: Cold Shutdown surveillance requirement T.S. 4.1.1.2.b (Any 3 of 6)
(1) RCS boron concentration (2) Control rod position (3) RCS av'erage temperature (4) Fuel burnup (5) Xe concentration (6) Sm concentration Examiner's Response The facility alternative answer supplied three additional responses, answer key was amended.
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Facility Comment 1.15 Additional answers: Reference Lesson Plan LO15 pages 22 through 29 Source Range High Neutron Flux Intermediate Range High Neutron Flux Power Range High Neutron Flux (Low Range)
Power Range High Neutron Flux (High Range)
OPAT Steam Generator Low-Low Level Turbine Trip Containment High Pressure Manual SI, Low Steam Line Pressure Manual Reactor Trip Examiner's Response It is my opinion that the facility misread the question, such as they based their answer key on reactor trip's that are based on DNB. The question asked the reverse. No amendment to answer key.
Facility Comment 2.4 Answer Key: " Manual actuation" not required only whether the question was True or False.
Examiner's Response Examiner concurs to comment, it was intended only for clarification.
Facility Comment 2.10 Answer key is reversed, that is, b is the answer to a and a is the answer to b. Verified by Project 0AD.
Examiner's Response The error was corrected in the answer key.
Facility Comment 2.12 Training has indoctrinated the candidates to associate " protective trips" with reactor trips, turbine trips, pump trips. The lesson plan that is referenced in the answer key is associated with FW 009 interlock. For this reason we request this question be deleted.
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Examiner's Response The question will be deleted although the question was taken verbatim from the lesson plans provided for the training examinations. The lesson plan terminology doesn't support or reflect the facility indoctrination related to protective trips.
Facility Comment 2.13 SI minimum flow valves are manually closed by the operator when switching to Cold Leg recirc.; Example - BEP ES 1.3 step 5.
Examiner's Response The question solicited whct the basic reason for the flow valves for the safety injection system when RCS pressure is 1835 psig, to wit; provide flow path for pump recirculation at 1833 psig until pressure is below shut off head of pump. No mention in question for cold leg recirculation. The alternative first part answer will not be accepted. The second part of facility's comment is a translation of answer key. Their comment noted.
Facil.ity Comment 2.14 Additional answers: FSAR 5.4-23
-Limits Steam Generator internal stress
-Limits Main steamline piping stresses
-Limits RCS cooldown
-Limits containment pressure rise Examiner's Response The answer key will be amended only to accept "-Limits RCS cooldown" as the remaining listed answers are essentially indirectly derived from RCS cooldown.
Facility Comment 2.18 Letdown orifice isolation valve interlocks are:
- a. Pzr level > 17%
- b. Letdown isolation valves (LCV-459/460) are open
- c. Phase A isolation reset (This is a recent modification)
Charging pump running as stated in the answer key is not an interlock for these valves.
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Examiner's Response Since the recent modification was not supplied to the NRC examiners it will not be accepted. The answer key will reflect the first two answers listed in the facility's comment.
Facility Comment 2.20 d. Candidates could have included for the Main Generator 1,361,000 KVA 1,175 MW 0.9 Power Factor Examiner's Response The additional responses will be accepted.
Facility' Comment 3.7 Delta I exceeds + 10%
T.S. Table 2.2-1 (Recent T.S. change)
Examiner's Response The facility provided material that reflected a different target band initially. The answer key will reflect this data.
Facility Comment 3.19 Load rejection controller dead band has been changed to 4*.
Ref. PLS June 5, 1985.
Possible 5th answer is the error bistables have different setpoints.
Examiner's Response The examiner's notes that it is a recent modification and will be accepted.
Facility Comment 3.18 1. "A" Jockey pump auto starts at 145 psig "B" Jockey pump auto starts at 140 psig Ref: BOP-FP-3 Examiner's Response The jockey pumps setpoints are acceptable. Answer key will be amended although the lesson plans reflects slightly different setpoints.
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! Facility Comment 3.20 Additionally the motor operated discharge valve closes on an SI.
Examiner's Response Will accept the additional answer. The examiner's notes that the question and answer were taken verbatim from the facility question bank.
Facility Comment 4.6 Please consider the list of parameters listed in BEP ES 01 Step 14:
Subcooling Steam Pressure stable Hot leg temperature ,
10 highest TC Cold leg temperature There could have been some confusion as to whether the R0 was monitoring Natural Circ. cooldown or the establishing of Natural Circ.
. Examiner's Response The facility makes comment that there was confusion whether the R0 was monitoring Natural Circ cooldown or the establishing of Natural Circ. No R0 candidate during the examination period expressed any confusion about the question. The question stated Byron Natural Cooldown procedure and asked for specific parameters for cooldown. The examiner notes that the supplied answer is extracted from the Reactor Trip procedure vice the Natural Cooldown procedure. The additional answers will be accepted.
Facility Comment 4.8 a. During refueling when the more restrictive of the following reactivity conditions is not met.
Keff of .95 or less or Boron concentration of Greater than or equal to 2000 ppm. This means euergency boration is required if Keff is 0.95 or greater or Boron Concentration is 2000 ppm or less.
Examiner's Response The supplied answer is word for word for answer key. The examiner's notes the facility's comment about the exact interpretation of the answer key. No answer i
revision noted.
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Facility Comment 4.15 Per recent procedure revision BOA-ENV-1 the diesel generators are no longer started. They are checked and prepared for standby conditions.
Examiner's Response The examiner's notes the procedure revision and will amend the answer key.
Facility Comment 4.14 Possible optional answer: BRP 1000-Al 125 REM Thyroid for less urgent action.
Examiner's Response ,
The alternative answer is still optional. The examiner notes that the thyroid is not an extremity as called for in the question.
Facility Comment 4.18 Optional answer would be:
Place channel in test by tripping bistables and pulling fuses.
Examiner's Response The optional answer is acceptable.
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Byron Senior Reactor Operator Examination
. Facility Comment 6.1.c Lost steam dumps due to condenser vacuum interlock. There was still one,(1) CW pump running during this event.
Examiner's Response
' The examiner agrees and corrected the answer key.
Facility Comment 6.11.b Alternate Answer: Accumulator level 31% to 65%
Boron 1900 to 2100 ppm Pressure 602 to 674 psig Numbers were changed with issuance of T.S. to conform with MCB indications.
Ref: T.S. 3.5.1 Examiner's Response The examiner agrees and corrected the answer key.
Facility Comment 6.12 Group 6 lights Additional Answers: Containment Vent Isolation Phase B MSIV's
' Containment Spray Examiner's Response The examiner required all the additional answers offered by the facility as well as those on the answer key for full credit.
Facility Cornent
4 Examiner's Response l The answer key remains unchanged.
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! Facility Comment
.' '7.14 With the Byron Station organization the SCRE and STA titles are used
interchangeably with regard to scanning the BST's.
. Examiner's Response The examiner agrees and the answer key was corrected.
Facility Comment 8.14 BAP 300-18 designates Shift Supervisor as person who can perform this function. This is interpreted as either the Shift Engineer or Shift Foreman.
Examiner's Response
-The examiner agrees and corrected the answer key. ..
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4 MASTER COP 3 U.S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: Byron REACTOR TYPE: PWR - Westinghouse DATE ADMINISTERED: September 16, 1985 EXAMINER: T. D. Reidinger APPLICANT:
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 at least 70% in each category and a final grade of at least 80%.
% of Category % of Applicant's Category Value Total score Value 25 25 1. Principles of Nuclear Power Plant Operations, Thermodynamics, Heat Transfer and Fluid Flow 2f zy J5' I 2. Plant Design Including Safety and Emergency Systems 25 25 3. Instruments and Controls 255' 25.I 4. Procedures - Normal, Abnormal, Emergency and Radiological Control 3M TOTALS Final Grade %
All work done on this exam is my own, I have neither given nor received aid.
Applicant's Signature
Page 1 of 5 DATA SHEET
'Ev;10R THEORY FORMJ, AS:
P = Poet /T P = Po 10 'A(t)
S P= I f{th V 512x1010 fissions /sec p= 1*+ Teff T 1 + AT T = L.
p ..
T = Eeff - 0 ~~
Ap Ap -In K (final) 7 K (InTtE1 p = K-1 K
S'JR = 26.06 7
Keff = c Pr p Pth I4 p 3
-(B2 Lth?)
1 + (82 Lth2 )
2 l Pr = e-(g2 Lf) p=eN1 Neff5
! II s l C1 (1- Kerrl) = C2 (1 - Keff2) it = 1 TW =t{l5ItialC (final) 7 Af + Ap 8 2 ( N_ f2 + A'-th? )
1 -
OT = 1 f At p At At 'At
l' l
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- Page 2 of 5 ,
D4TA SEEET THERIODYMat41CS Ato FLUID'ECHANICS FORT 4JLAS: .-
6=$Ah ,
Q = 2 s LAT t
Q = UA(AT) 1 + in R /R1 2 + in R /R2 3
\
K K2 K3 Q = K (pAT)~
~
n=61n-60ut 6==6A*R4 61n n = (h in - hout) REAL (hin - hout) IDEAL .
np = W ACTUAL W SUPPulED PV11 =PV22 T1 T2 p1AV I 1 = P2AV 2 2 m = p4V
- d. = KA / A Px p 3
dyc = KAg ]~ = KA AT UT' = KA Ap / AP ATm = AT (IN) - AT (OUT)
In [ AT (IN) )
AT (OUI) .
G= I (th 8.8 x 109 TCL - Tes = p2, 4k 6=kAAT AX i
6 = A AT total (Jg + Lxg ... + Lxy KA KB KN I
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Page 3 of 5 i DATA SHEET CENTRIFUGAL PLW LAWS: _
Speed 1 Flowl (Speed1 )2 Headi (Speedi )3 Poweri
= = _-
Speed 2 F10W2 (Speed2 )2 Head 2 { Speed2 )3 Power 2 RADIATION AND CHEMISTRY FORMJLAS:
R/hr = 6CE Ix = 10 e-**
d2 CVI1=CV22 t1/2 = 0.693 A C = Co e-Gt A = AN G = Dilution Rate A = Aoe -At I = Io(ilu 10 CONVERSIONS:
1 gm/cm3 = 62.4 lbm/ft3 Density of water (20*C) = 62.4 Itmy'ft3 = Ig/c~m3 1 gal = 8.345 lbm 1 ft3 = 7.48 gal Avogrado's NJober = 6.023 x 1023 1 gal = 3.78 liters Heat of Vapor (H 2O) = 970 Btu /lbm 1 lbm = 454 grams Heat of Fusion (ICE) = 144 Btu /lbm e = 2.72 1 AW = 1.66 x 10-24 grams s = 3.1415 Mass of NeJtron - 1.008665 AMJ 1 KW = 736 ft-lbf/sec Mass of Proton = 1.007277 AMU 1 KW = 3413 Btu /hr Mass of Electron = .000549 AW 1 HP = 550 ft-lbf/sec One atmosphere = 14.7 psia = 29.92 in. Hg. abs.
1 HP = .746 KW 'F = 9/5*C + 32 1 HP = 2545 Btu /hr *R = *F + 460 1 BTU = 778 ft-lbf
- C = 5/9 ( *F-32) 1 EV = 1.54 x 10-16 BTU h = 4.13 x 10-21 M-sec *K = *C + 273 IW = 3.12 x 1010 fissions /sec g = 32.2 lbm - ft Ibf - sec2 1 CAL /GPAM = 1.8 Btu /lbm C2 = 931 W V/ AMU 1 in. = 2.54 CM C = 3 x 108 m/sec o = 0.1714 x 10-8 Btu l
Tir ftZ *R4 C
Page 4 of 5 DATA SHEET AVERACE THERMAL CONDUCTIVITY (K)
Material BTU /hr-ft *F Cork 0.025 Fiber Insulating Board 0.028 Maple or Oak Wood 0.096 Building Brick 0.4 Window Glass 0.45 Concrete 0.79 1 Percent Carbon Steel 25.00 1 Percent Chrome Steel 35.00 Aluminum 118.00 Copper 223.00 Silver 235.00 W, iter (20 psia, 200*F) 0.392 Steam (1000 psia, 550*F) 0.046 Uranium Dioxide 1.15 Halium 0.135 Zircaloy 10.0 MISCELL NE005 INCORMATION:
E = mc2 KE = 1/2 mv2 PE = mgh Vf = Vo + at
. AE = 931 Am GEOMiTRIC OBJECT AREA VOLUPE TRIANGLE A = 1/2 bh //////////////////
SQ'JARE A=S2 /////////////////j RECTANGLE A=LxW //////////////////
CIRCLE A = wr2 /////////////////j' RECTANGULAR SOLID A = 2(LxW + LxH + WxH) V=Lxwxh RIGHT CIRCULAR CYLINDER A = (2 wr2)h + 2(wr2) V = wr2h SoHERE A = 4 wr2 V = 4/3 wr3 CusE //////////////////////// v=S3
Page 5 of 5 DATA SHEET MISCELLANEOUS INFORMATION CONTINLED:
10 CFR 20 Appendix 8 Gamma Table 1 Table II -
Energy Col I Col 11 Col 1 Col 11 EV per Air Water Air Water uc/ml uc/ml uc/ml Material Half-Life Disint.:_ratior uc/ml Ar-41 1.83 h 1.3 Sub 2x10-6 ,---- 4x10-8 -----
~
2.5 S 3x10-7 1x10-3 1x10-8 5x10-5 Co-60 5.27 y 8.04 d 0.36 S 9x10-9 6x10-5 1x10-10 3x10-7 1-131 Kr-85 10.72 y 0 . G'. Sub 1x10-5 ----- 3x10-7 -----
2.52 h 0.59 S 9x10-7 4x10-3 3x10-8 1x10-4 N1-65 2.41x104 y 0.005 S 2x10-12 1x10-4 6x10-14 5x10-6 Pu-239 Sr-90 29 y - S 1x10-9 1x10-5 3x10-Il 3x10-7 9.09 h 0.25 Sub 4x10-6 ----- 1x10-7 -_--_
Xe-135
- Any single radionuclide with T1 /2> 2 hr 1x10-10 3x10-6 which does not decay by a or spa.taneous 3x10-9 9x10-5
. fission Neutron Energy (EV) Ne*trons per cm2 Average flux to deliver a ,ivalent to I rem 100 mrem in 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> Thermal 970 x 106 6'07 0.02 t.00 x 106 280 (EUTRONS) 0.5 43 x 106 30 cm2 see 10 24 x 106 17 LI E AR ABSORPTION COEFFICIENTS y (cm-1)
Energy ( E V) Water Con- ete Iron Lead 0.5 0.090 0.21 0.63 1.7 1.0 0.067 0.15 0.44 0.77 1.5 0.057 0.13 0.40 0.57 2.0 0.048 0.11 0.33 0.51 2.5 0.042 0.097 0.31 0.49 3.0 0.038 0.088 0.30 0.47
SECTION 1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATIONS, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 1.1 Byron Unit 1 is operating at 100% power with all rods out and the core at end of life. The SR0 instructs you to decrease reactor power to 50% by boration leaving the rods out in order to maintain Delta I on target. He says that by not moving rods he will not affect power distribution and therefore AFD will remain constant. Do you agree or disagree with his assessment?
Justify your answer. (1.0) 1.2 Why is nitrogen-16 of major concern during plant operations but not a concern when the plant is shutdown? (1.0) 1.3 Answer the following True or False:
- b. During an increase in power from equilibrium xenon
._ conditions, xenon concentration initially decreases. (.33)
- c. Slowing the rate of a power decrease lowers the height of the resultant xenon peak. (.33) 1.4 a. Are neutrons best shielded by light or heavy material? (.25)
- b. Are gammas best shielded by light or heavy material? (.25) 1.5 a. Explain why the effective delayed neutron fraction changes from BOL to E0L, and how this affects reactor control. (1.0)
- b. What two factors cause the delayed neutron fraction to be different than the effective delayed neutron fraction? (1.0) 1.6 With a constant startup rate of +.25 DPM and an initial power level of 10 10 amps, how long will it take for reactor power to reach 1%? State all assumptions and show all work. (1.5) 1.7 If the effective delayed neutron fraction is .0055 and there is 50 pcm of positive reactivity in the core, how much more positive reactivity, in pcm, can be added to the core before the core becomes prompt critical? (1.0)
- 1. 8 Why is the most negative reading on the source range meter
.5 DPM? (1.0) 1.9 Why does the moderator temperature coefficient have its least negative value at the beginning of the fuel cycle? (1.0) e
4 1.10 In preparation for reactor startup, you are directed to dilute the RCS with 400 gallons of water in 100 gallon increments. After completing the first 100 gallon addition, you observe the source range instruments have doubled their count rate. Would you proceed as directed to dilute the plant with a second 100 gallon addition? Explain.
(1.0) 1.11 a. Define condensate depression. (0.5)
- b. What is the advantage and disadvantage of condensate depression? (0.5) 4 1.12 Fuel damage from a loss of coolant accident is strongly core height dependent as such that the F(Q) (total peaking factor) have different limits for each core elevation per j Figure 3.2-2. Explain why the K(Z) curve is more limiting in Sec, tion (a). (1.0)
, 1.13 a. How does the reactor operator ensure adequate shutdown margin when the reactor is operating? ,
(0.5)
- b. What three factors must be considered to ensure adequate shutdown margin is maintained during cold shutdown?' (1.5) 1.14 How will a rod-drop affect reactor power in the following ,
situations? Assume rods are in manual and a reactor trip does not occur. Carry out the event until stable conditions are 4
achieved.
- a. Reactor is just critical at 10 8 amps. (1.0)
- b. Reactor is at 50% power. (1.5) 1.15 Byron Reactor Protection Chapter states that all reactor trips except two are based on DNB. Name these two reactor trips. (1.0) 1.16 Why will a reactor trip from 100% power without a turbine trip result in severe pressurized thermal shock? (1.5) 1.17 a. ,
? (1.0)
- b. Why does RT NDT increase with reactor operation? (1.0) 1.18 a. How does the required NPSH of a centrifugal pump change with an increase in flow? (0.5)
- b. Name two indications of pump cavitation. (0.5) j
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1.19 How does an increase in each of the following parameters affect
- DNBR? Consider each parameter separately,
- a. RCS temperature (0.5)
- b. RCS pressure (0.5)
- c. RCS flow (0.5)
- d. Reactor power (0.5)
END OF SECTION 1 O
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F/& 3.d 4 3/' 2-3 BYRON - UNIT 5 ! & 2
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l SECTION 2 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 2.1 The reactor is in hot shutdown and. primary pressure is 1845, you are the R0 and Mr. Goodwrench (I&C) wishes to conduct maintenance on the turbine first stage pressure instrument.
Would you authorize him to do the maintenance? Defend your answer. (1.5) 2.2 According to Byron Technical Specifications the Power Range Nuclear Instrumentation is only required in power operations.
In spite of this fact, no more than one channel can ever be removed from service during shutocvn modes. Explain why. (1.0) 2.3 Which of the following will actuate tr,e Unit 1 Rod Control Urgent Failure annunciator light? (1.0)
- a. ' Internal failure of the power cabinet because of failure of the slave cycler.
- b. Internal failure of the logic cabinet due to the removal of a printed card.
- c. Internal failure of the logic cabinet from phase failure detector,
- d. Normal alarm occurring during the dropped rod recovery procedure.
2.4 Indicate whether the following statement is true or false. On Unit 1, the automatic reactor trip signal will cause both the undervoltage coil to deenergize and the shunt coil to operate. (1.0) 2.5 a. List the radiation monitoring systems that respond to a high radiation signal and initiate automatic actions to prevent the release or spread of radiation at Byron Unit 1. Be specific, list five (5) out of the eight (8) systems. (.75)
- b. List all the automatic actions which are initiated in the event a high radiation signal is detected for five radiation monitoring systems. (.75) 2.6 Mr. Goodwrench has the one (1) diesel in test mode and loaded 3000 KW and a spurious SI signal is received from the results of a faulty test procedure. How does the diesel generator and its electrical load respond in this condition? (1.25) 2.7 While performing a cooldown with RCS temperature at 450* and l pressure at 1100 psig, Mr. Goodwrench (I&C) requests that he be allowed to calibrate two (2) 0-30 psig lower containment pressure
! transmitters simultaneously. He states that this is acceptable since the containment spray logic will not be made up. Would you allow this work to occur? Defend your answer. (List two reasons.) (1.25) i 4 l
2.8 Would the failure of a turbine first stage pressure channel l result in an inadvertent actuation of the steam dump?
! Explain. (1.0) 2.9 Byron Unit 1 is being cooled down for maintenance and the decay heat load is constant. Will the RHR flow rate and component cooling flow rate require any adjustment to maintain cooldown rate? Explain. (1.25) 2.10 On the Output of the Main Generator two potential transformers exist per phase. Explain the affect on Main Generator power operation if:
- a. One Regulating PT is inadvertently racked out. ( 75)
- b. One Metering PT is inadvertently racked out. (.75) 2.11 What'is Byron's Unit 1 four pump zero power rod insertion limit? (.75) 2.12 What protective trips are active only if either the F MO
, regulating values or FW gy ass valves are open? M gc/o(1.0) 2.13 Explain the differences if any of the purpose or operation of the minimum flow valves between the safety injection system and the Residual Heat Removal system when a SI signal is actuated and system RCS pressure is 1835 psig. (1.0) 2.14 What are two purposes of the flow restrictor on the main steam line? (1.0) 2.15 What are two reasons why the drain valves on extraction steam system automatically modulate to maintain a water level in the drip legs during power operations? (1.0) 2.16 a. What are two reasons for having a 30 second delay between the turbine tr;p and the generator trip? (1.0)
- b. After 30 seconds delay, what then causes the generator trip to occur? List two. (1.0) 2.17 a. On a phase A signal (T-signal); the CC isolation valves close. Name component /s which are not supplied when these valves close. (1.0)
- b. On a phase B signal (P-signal), additional CC isolation valves close. Name component /s which are not supplied when these CC isolation valves close. (1.0) 5
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! 2.18 List what interlock /s is/are associated to open the letdown l orifice isolation valves. (1.0) 2.19 What design feature is installed in the charging line to '
prevent overpressurization damage to the portion of the charging line which contains the regenerative heat i
exchanger? (1.0) 2.20 Your turbine and diesel generators are distinctly different in many ways. Explain the difference between them in regards to the following:
- a. Field excitation and control (0.5)
- b. Generator cooling method (0.5)
- c. Synchronous speed, control, (droop) (0.5)
- d. Load ratings (0.5)
END OF SECTION 2
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SECTION 3 INSTRUMENTS AND CONTROLS 3.1 The Byron Turbine Unit 1 impulse is used as an indication of electrical load. For what protective purpose /s is/are the L.P turbine inlet pressure measured? (1.25) 3.2 2/2 coincidence is required from both level channels LT-112 and LT-185 on the VCT to transfer charging pump suctions to the RWST. True/ False (1.0) 3.3 The power level signal used by the FW bypass valve is the average nuclear power signal /s from the power range nuclear instruments. True/ False (1.0) 3.4 The centrifugal charging and SI pumps can be started automatically by the ESF sequencer when their control switch is in pull to lock position. True/ False (1.0) 3.5 What signals are required in order for the RHR pump suctions to automatica12y align to take a suction from the containment
- recirculation sumps? (1.0) 3.6 What automatically start signal is blocked on the component cooling pumps when they are being sequenced on following a loss of electrical power? (1.0) 3.7 What two inputs determine the OTAT setpoint and how would these inputs have to change to cause a decrease in trip setpoints? (1.5) 3.8 Which four rod stop, manual and/or automatic, can be bypassed to allow continued manual control rod operation? Include setpoints. (1.5) 3.9 How will an undercompensated intermediate range detector respond after a reactor trip? (1.25) 3.10 Explain how an inadvertent safety injection could occur when conducting a normal plant cooldown. (List 3 ways.) (1.5) 3.11 List the automatic actions which occur in the Flux-Doubling /
Boron Dilution Protection System as source range instruments count rate doubles. (1.25) 3.12 Explain the basic principle of the Reactor Vessel Level Indicating System (RVLIS). (1.25) 3.13 List the automatic actions which occur in the fire protection system as ring header pressure decreases from 160 psig to l 120 psig. Include setpoints. (1.25) l 7
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3.14 The local diesel panel has a " Speed Voltage Mode Selector" switch to select Droop or Isochronous modes. Explain the diesel operation for each mode. (1.0) 3.15 List five trips for the motor driven feedwater pump.
-(Setpoints not required) (1.25) 3.16 You as a R0, inadvertently reset the Rod Control Startup Reset pushbutton while conducting a startup. What instruments have to be manually reset to conduct the startup? (1.0) 3.17 List the failed positions of the following values,
- a. Feedwat x isolation valve (09) (0.5)
- b. Letdown back pressure regulating valve CV131 (0.5)
- c. Steam generator PORV's (loss of air) (0.5)
- d. Main Steam Isolation Valve Bypass Valves (0.5) 3.18 The motor driven auxiliary feedwater pump isolation valves can be operated from the Hot Shutdown panel. True/ False (1.0) 3.19 The steam dump system has three controllers. How do the two (2) temperature controllers differ? Provide four (4) examples. (2.0) 3.20 Describe the difference on the Main Feedwater System on a safety injection and a reactor trip with low Tave. (1.0)
END OF SECTION 3 8
r SECTION 4 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL s
4.1 What reactor trip and safeguards actuation circuits can be administratively bypassed for maintenance on a single channel? (1.20) 4.2 Byron Nuclear Station utilizes a unique " Green Board" concept.
- a. Explain the " Green Board" concept. (1.0)
- b. At what reactor power level does the " Green Board" concept begin to alert the Reactor Operator to systems and components being in an abnormal condition? [ff) 4.3 Per BAP 300-18, Removing and Returning Equipment Out of Service, where may the master out of service cards be placed? (1.0) 4.4 What two criteria constitute adverse containment conditions? (1.0) 4.5 List four criteria that must be met before a spurious safety injection can be terminated. (1.25) 4.6 List four primary plant parameters per Byron Natural Cooldown procedure you would use as a R0 to verify natural circulation in progress. (1.25) 4.7 Based upon the priority established through the Byron' Administrative Procedure BAP 300-11, the status trees for critical safety functions have color coding of thi' end points.
- a. Name and define each end point. (2.0)
- b. List the critical safety function status trees in order of priority. (1.5) 4.8 a. List one criterion which would require emergency boration during refueling. (0.5)
- b. List four criteria which would require emergency boration during other plant modes. (2.0) 4.9 During the performance of ATWAS procedure why is a manual safety injection not preferred? (1.25) 9,
4.10 In the event of Generating Stations Emergency Plan (GSEP) actuation on backshift:
- 1. Who should normally assume the Station Director position? (0.5)
- 2. Who would assume the Station Director position if the preferred person is not available? (Indicate order of preference.) (3 required) (1.5) 4.11 Match the following casualties, abnormal conditions and malfunctions to the applicable procedure or guideline.
(Answers can be used multiple times) (1.0)
- 1. Loss of Instrument Bus a. Byron Emergency Procedure
- 2. Loss of all AC power b. Byron Contingency Actions
- 3. , Loss of Emergency Coolant c. Byron Functional Restoration Recirculation
- 4. Mode 5 to 4 checklist d. Byron Operating Abnormal Procedure
- 5. Stuck or misaligned rod e. Byron General Procedure
- 6. Response to void in Reactor f. Byron Operating Procedure Vessel
- 7. Loss of reactor coolant g. Byron Emergency Procedure Specific Events
- 8. Natural circulation cooldown h. Byron (BZP)
(no accident)
- 9. Safety Injection procedure
- 10. Power descension 4.12 Whose permission /s is/are required to restart the reactor if I the root cause of a reactor trip has not been determined?
(List four) (1.0) 4.13 List 2 criteria for terminating venting'from the reactor vessel vent. (Setpoints required) (1.0) i
'4.14 a. What is the maximum planned radiation doses in an
! emergency situation that require personnel to search for and remove injured persons in high radiation area? (Include extremities) (1.0)
- b. What is the maximum radiation doses in which to enter a hazardous-area to protect valuable equipment?
(Include extremities) (1.0) 4.15 During the past few months there have been tornados sighted in different areas of Illinois and Wisconsin. What are the RO's general responsibilities if a tornado warning is in effect? (List 2) (1.0) 4.16 What emergency action levels automatically require activation of the Technical Support Center? (1.0) 10'
4.17 How many Fire Brigade personnel must be maintained onsite at all times? (1.0) 4.18 What actions must be taken in general if a power range channel fails during reactor startup? (List 2) (1.0)
END OF SECTION 4 ne 5
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U.S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: Byron REACTOR TYPE: PWR - Westinghouse DATE ADMINISTERED: September 16, 1985 EXAMINER: T. D. Reidinger APPLICANT:
INSTRUCTIONS TO APPLICANT:
Use separdte paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%.
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% of Category % of Applicant's Category Value Total Score Value 25 25 1. Principles of Nuclear Power Plant Operations, Thermodynamics, Heat Transfer and Fluid Flow if 2 JF 2. Plant Design Including Safety and Emergency Systems 25 25 3. Instruments and Controls 25rI 25. I 4. Procedures - Normal, Abnormal, Emergency and
- s Radiological Control TOTALS Final Grade %
All work done on this exam is my own, I have neither given nor received aid.
1 Applicant's Signature
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SECTION 1 - ANSWER KEY 1.1 Disagree. A decrease in reactor power will decrease AT. The negative MTC will add positive reactivity to the top half of the core and negative reactivity at the bottom of the core causing the AFD to swing positive.
Ref.: Reactor Theory, Chapter I-5 1.2 Nitrogen-16 is the most abundant activation product and it has an extremely high gamma decay energy (0.5). It
-has a very short half life so it is not a concern when the plant is shutdown (0.5) OR Nitrogen-16 is not produced in shutdown modes.
Ref.: Reactor Theory 1.3 a. True
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- b. True
- c. True Ref.: Reactor Theory, Chapter I-5.64 1.4 a. Light (.25)
- b. Heavy (.25)
Ref. : Reactor Theory 1.5 a. Effective delayed neutron fraction decreases over core life due to the depletion of uranium 235 and the buildup of plutonium 239. This decrease makes the reactor more responsive for a given reactivity insertion. (1.0)
- b. 1. Ceduced leakage of delayed neutrons because they are born at lower energies than the i
average prompt neutron. (0.5)
- 2. No fast fission of uranium 238 by delayed neutrons because they are born at energies below the threshold energy for fast fission l or uranium 238. (0,5) l Ref.: Reactor Theory, Chapter I-3 l
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. . - _,._ .__~_._.._,__ ___ _ - , _ , _ _ - - , _ , _ . _ _ _ _ . _ . _ _ , - _ . . .. _.._ _ _ ... .,_ -_..._ __ _ _...___..... _
. 7b /0 1.6 1% is approximately 10 5' amps to 10 x 10 _6 amps is acceptable). (0.5)
P = Po 10 SUR x time (0.3)
SUR x time P/Po = 10 log P/Po = SUR x time time = 1 log P/Po (0.3) 3UR time = 1 log 10 5/10 10 (0.3) 25 time = 4 log 105 time'='4 x 5 = 20 min. (0.1)
Ref. : Reactor Theory 1.7 .0055 = 550 pcm (0.5) 550 - 50 = 500 pcm (0.5)
Ref. : Reactor Theory 1.8 The stable negative startup rate after shutdown is due to the longest lived delayed neutron precursor, which decays at a negative startup rate of .33 DPM.
Ref. : Reactor Theory 1.9 Boron concentration is the highest.
Ref. : Reactor Theory 1.10 No. Based upon the inverse linear relationship between SD margin and counts, it appears that the 100 gallon dilution reduced the SD by 50% since counts have doubled.
To' proceed would be contrary to good judgement since the second addition would take the reactor supercritical.
1.11 a. Temperature difference between the saturation temperature for the existing condenser vacuum and the temperature of the condensate. M 4 4 88 # sens p A 4 S&& C09dedP.
- b. Advantage: decreases cavitation in the condensate pumps. (.25)
Disadvantage: reduces plant efficiency. (.25)
Ref.: Heat Transfer, Thermodynamics, Fluid Flow i
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1.12 In Section (a) the highly restrictive values of K(Z) is to so preclude high power astJ0 rate.
in the core. In thislevels area in the }acccu..".-12" (3"-4")' or This restriction could occur if the break size were large enough to cause a blow down of the core area but not large enough to allow the displacement of steam out the break when the ECCS s, tarts reflooding the core.
Ref.: T/S, 10 CFR 1.13 a. Rod insertion limits are observed. (0.5)
Ref.: TS 3.10-15
- b. 1. Control rods fully inserted /M% (0.5)
M dff>%#
2.
3.
Core burnup Boron concentration [$
(0.5)
(0.5)
Ref.: Reactor Theory 1.14 a. Reactor becomes subcritical and power decreases into the source range.
- b. Reactor becomes subcritical. (0.5)
Moderator and fuel temperature drop due to the steam load, thus inserting positive reactivity to counter the dropped rod. (0.5)
Reactor stabilizes at 50% power, with fuel and moderator temperature less than their original value. (0.5)
Ref.: Reactor Theory 1.15 High RCS pressure (0.5)
High pressurizer level (0.5)
Ref.: Rx Protection - L015 1.16 A reactor trip will virtually stop heat input into the RCS while the turbine is still drawing off steam, causing the RCS temperature and pressure to fall dramatically. (0.5)
A low pressure SI will be activated, causing the pressurizer to fill and the RCS to be repressurized. (0.5)
The combination of rapid temperature drop and repressurization will result in severe shock to the reactor vessel. (0.5)
Ref.: Byron Lesson Plans 3
1.17 a. The temperature below which a material will undergo brittle rather than ductile failure.
- b. Neutron embrittlement.
Ref.: TS 3.1-5 1.18 a. Increases
- b. Any two of the following:
- 1. Fluctuating motor amps (.25)
- 2. Fluctuating discharge pressure (.25)
- 3. Fluctuating flow (.25)
' 4'. Excessive pump noise (.25)
- 5. Excessive pump vibration (.25)
.. Ref. : Heat Transfer Fluid Flow 1.19 a. Decrease M c./M M % g4g
- b. Increase ##t. FWe*/I$t, & gg/g4
- c. Increases #4 #dVMM [#M
- d. Dec* ease 80 k#56 % M-N Ref.: Heat Transfer, Thermodynamics and Fluid Flow, p. 228 4
SECTION 2 - ANSWER KEY 2.1 No. Low pressurizer pressure trip is 1885 psig. If the turbine first stage pressure instrument is placed in test for maintenance purposes, it will de-energize interlock circuit P-7, this will unlock the low pressurizer pressure trip functions and will result in a reactor protection actuation.
Ref. : L016 013-000-K1.16 (3.4)
- p. 77/88 012-000-K4.01 (4.0) 2.2 If more than one (1) PR NI channel is de energized, P-10 will block source range high voltage causing a loss of source range. Source range is required in shutdown modes.
Ref. : LOO 6 012-000-K6.04 (3.6) 015-020-K3.06 (3.9) 2.3 (b)
Ref. : LOO 4, p. 27/47 001-050-K4-01 (3.8) 2.4 False,hnual actuationf/ / OIM u/NM Ref. : LO15, p. 19/88 2.5 A. Component Cooling Water Monitor -IRE (PR009)
B. High radiation signal initiates closure of the component cooling surge tank vant valve to prevent outgassing of containment cooling water t0 the aux. bldg.
A. S/G Blowdown Detector - (OR 008)
B. High radiation doses blowdown sample valves.
A. Blowdown Filter Radiation Monitor - (PR016)
B. Blowdown flow from mixed bed demineralizers is stopped to prevent contamination of other condensate.
A. Control Room Outside Air Intake - (PR031)
B. Auto on high radiation, the outside air intake dampers close, the makeup fan starts and main control room turbine building air intake dampers open.
A. SJAE/GS and Steam Exhaust Monitor - (PR027)
B. Auto on high radiation, bypass valves and the off gas vent filter system are energized.
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A. Gas Decay Tank Effiaent - (PR002)
B. Auto on high radiation, the gas decay tank discharge isolation valve closes.
A. Volume Reduction Areas / Cubicle Vent Exhaust Monitors -
(PR040A)
B. Auto on high radiation, the bypass dampers close and the fans start to route the exhaust through the filter unit.
A. Liquid Radwaste Effluent Monitor - (PR001)
B. Auto on high radiation, the release tank discharge isolation valves close.
Ref.: LO32 072-000-K1.01 (3.5)
T/S, p. 3/4 - 3/39 072/000-K2.02 (3.9) 2.6 a. 'The diesel operating in test mode with a SI signal will return the diesel to standby operation, W F P (.625)
- b. The emergency loads will automatically energize by offsite power. (.625)
Ref. : Technical Specifications 3/4, 8-6, para. 10 064-050-K36 (4.1)
L-024 D/G 2.7 No! Containment sprays require logic 2/4 for containment Phase B high pressure signal to actuate (.625), also required in Modes 1, 2, and 3 per Technical Specifications. Mr. Goodwrench should be directed not to take two (2) channels to test per Technical Specification which require three (3) channels operable. Steamline isolating (MSIVs close) would occur thus isolating stm to stm dumps, possible SI actuation (.625).
Ref.: L014, L015, p. 40/48 T/S 3.6 T/S 3/4, p. 3-15 026-020-K36 (4.1) 2.8 No. A failure of a turbine first stage pressure channel should not result in an inadvertent actuation of the dumps unless simultaneous failure of both channels occur since one channel feeds the loss of load arming circuit and the other channel feeds the load rejection error signal circuitry.
Ref.: L010 041-020-K35 (3.6) 2.9 The RHR flow must be increased to continue the cooldown at a constant rate since the temperature is dropping. The reactor coolant flow through RHR is increased and flow through the bypass line is reduced. Component cooling doesn't change flow rate.
Ref.: LO22, p. 19, p. 32 6
2.10 / M ain Generator will trip due to a phase imbalance on the machine. l l
[.4Thevoltageregulatorwillshifttomanualandthe unit should remain in service.
Ref.: Byron Question Bank 2.11 47 steps, Bank C Ref.: T/S 3/4 1-22 2.12 Low S/G pressure Low S/G 1evel tow FW flow trips NM j Ref.: ,
L011, p. 37/72 2.13 The SI minimum flow bypass valves provide a flow path to the RWST until reactor coolant pressure is low enough (1500 psig) for the SI pumps to discharge to the RCS.
- The minimum flow valves for the RHR remain open until '
flowinJHRtrainincrease above 400 a o s per l minute.L d 6 /
NMf!fW )
Ref.: L013 2.14 1. It limits steam flow during a steamline break.
- 2. A D/P is produced as a steam flow signal input.
Ref.: L010, p. 15 2.15 1. Maintenance of a water level prevents the water from backing up into the turbine
- 2. Prevents steam from being dumped to the next stage Ref.: LO25, p. 13 2.16 a. Maintain power to the RCPs for 30 seconds after a Rx trip / turbine trip (.25). To prevent compounding an overpower accident with a simultaneous loss of flow accidents, g RCP overspeed on LOCA (.25),
- b. Prevent turbine overspeed (.50)
Ref. : LO35, p. 21
- c. Generator motoring; reverse power Ref.: LO35, p. 21 7
2.17 a. Excess letdown HXs
- b. RCPs Ref.: L015, p. 40, L022, p. 14 2.18 Letdown orifice isolation vplves automatically shut and cannotbereopenedunless{ ;t.)
- a. Letdown isolation valves (LCV-459/460 are open b.
jbd A. /d8NM MS gump ia i umi,ug ad C
._._......3..3
- c. Pzr level > 17%
Ref.: L01, p. 24, p. 94 2.19 Overpressure protection is provided by spring loaded check valves that are installed in the bypass lines.
Ref.: LOO 1, p. 23 2.f0 a. The main generator is excited by a permanent magnet generator whereas the diesel generator is initially flashed by the battery and then provides its own excitation.
- b. Main generator cooling is by way of hydrogen and stator cooling water whereas the diesel generator is cooled by air.
- c. Thesynchronousspeedoftheturbine/ 800 rp( W) whereas that of the diesel is 600 rpnt iesel speed droop is manually added by proper switch alignment duringsurveillanceruntestin.b d.
llW //20 The turbine load rating isAK /MW whereas the diesel is rated for 5500 KW continuous and 6050 KW for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. 78'#/8/NC ,o /F Ref.: LO35, p. 6 L024, p. 27 8
SECTION 3 - ANSWER KEY 3.1 This pressure is used to develop a mismatch signal between LP pressure and generator megawatts for overspeed protection. If LP turbine inlet pressure is greater than generator megawatts by 30% then OPC
-is energized cyclically.
Ref. : LO12, p. 27/59 3.2 True.
Ref.: L001, p. 32/102 3.3 False.
Ref.: 'L015, p. 46/72 3.4 False.
., Ref.: L014, p. 17/43 3.5 SI signal RWST - Lo-2 alarm Ref.: L014, p. 13/43 3.6 Low CC pump discharge pressure Ref.: L22, p. 11 3.7 Tavg - increase Pzr pressure - decrease OI exceeds a (optional)
Ref.: LO:3, p. 25 3.8 1. (C-2) power range high flux -103%
- 2. , (C-1) intermediate range high flux -20%
- 3. (C-11) control bank D above 223 steps
- 4. (C-5) turbine impulse pressure below 15% load
(.1875 each) (Total 8 answers)
Ref.: LOO 6, p. 21, 27 LOO 4, p. 25 3.9 Intermediate range counts will not decay nearly as rapidly as they normally would OR Causes an abnormally high indication since the gamma flux is not subtracted.
Ref.: LOO 6, p. 39 9
3.10 1. Pressurizer pressure drops to 1829 psig and the low pressure safety injection was not manually blocked when pressurizer pressure below P-11.
- 2. Steamline pressure is reduced below 640 psig prior to manually blocking the low steamline pressure safety injection.
- 3. Steamline pressure is reduced below 640 psig when pressurizer pressure is greater than P-11.[/#/#
Ref. : LO15, p. 38 3.11 As SR count range average counts increase by a factor of 2, an output is sent to the CVCS system to open parallel valves LCV 112 D&E (0.5) and shut series valves LCV 112 B&C (0,5). This action aligns the efueling water storage tank to the CVCS pump suctions (.25).
Ref.: LOO 6, p. 15 3.12 The basic principle of RVLIS operation is the detection of a delta T between adjacent heated and unheated thermocouples (T/C) (.3125). The HJTC sensor consists of a chromel-alumel T/C near a heater and another chromel-alumel T/C positioned away from the heater. In a fluid with relatively good heat transfer properties, the delta T between adjacent T/C is very small (.3125). In a fluid with relatively poor heat transfer properties the delta T between the T/C is large (.3125) So if a T/C is uncovered the delta T is large and the RVLIS indicates the level change (.3125).
Ref.: LOO 2, p. 76 3.13 1. Jockey pump startspsig at g/% /yf lVD - b d _
- 2. Motor driven fire pump starts at 135 psig YP'BP
- 3. Diesel driven fire pump starts at 125 psig
(.208 each) (Total 6 answers)
Re f. : LO27, p. 8 3.14 Isochronous mode maintains one diesel speed regardless of load.
Droop mode gives a lower speed at higher load to permit paralleling.
l Ref. : LO24, p. 20 l
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3.15 1. ground
- 2. transformer 142-2 86 relay trip
- 3. low L0 pressure
- 4. high LO temperature
- 5. under voltage on 6.9 KV bus 156
- 6. loss of all CD/CB pumps
- 7. safetyinjectionsignal
- 8. low suction pressure
- 9. high-high S/G level Re f. : L0ll, p. 20 3.16 group step counters (.33)
P/A converter (.33)
^ Bank overlap unit (.33)
Re f. : LOO 4, p. 24 3.17 as is open shut shut Ref.: L0ll, p. 37 LOO 1 LO10, p. 21 LO10, p. 26 3.18 True Ref.: C033, p. 6 3.19 1.* The load rejection controller is armed by C9 and C7 whereas the trip controller is armed by C9 and C8. (0.5)
- 2. The load rejection controller compares T-average to T-reference whereas the trip controller compares T-average to T-no load. (0.5)
- 3. The load rejection controller has a deg. F dead-band to allow control rod action. (0,5)
- 4. The controllers have different gain and setpoints for valve positioning. (0.5)
![ C r % "4'. /4SOf!V6 064tt x?uf?/H<.q peQuq Ref.: LO10, p. 31 11
_ _ _ _ _ _ _ _ _ _ - ..a
3.20 The difference on the main FW system between a SI and a Rx trip with lo-Tavg is that on an SI signal the FW pumps trip while on a Rx trip with lo-Tavg signal th 9 g FW pump go on rec nggp a
Ref. : 8 n Question Bank i
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SECTION 4 - ANSWER KEY 4.1 a. Source range high flux trip
- b. Intermediate range high flux trip
- c. Containment Hi-2 spray actuation Re f. : Precautions, Limitations 4.2 a. When that particular panel is in its normal configuration, the only indicating lamps (green or blue lens) should be illuminated.
- b. Reactor power equal to or greater than 30% power.
Ref.:' 'BAP 399-4 4.3 May be placed on an out of service peg board or maintained on file with the Master Equipment outage form.
' Ref.: BAP 300-18 4.4 1. Containment pressure greater than 5 psig
- 2. Containment radiation greater than 104 R/hr Ref.: BEP-0, p. 2 4.5 Containment conditions normal RCS pressure > 2000 psig RCS subcooling > 26'F Pzr level > 25%
Any S/G 1evel > 4% NR AF flow > 485 gpm
(.3125 each) (Total any 4 answers)
Ref.: BEP-1 l
l 4.6 1. Average of ten highest core exit T/C - Trending i down
- 2. RCS hot leg temperature - Trending down
- 3. RCS subcooling - increasing
- 4. RCS temp / pressure within limits of Tech. Specs.
Ref.: BEP ES.2A, p. 5 Q S/>t fMSM 1MN s) m so ay ive s+r ny nn u~pmswe 13
4.7 End Points are:
Green - The critical safety function is satisfied. No operator is called for.
Yellow - The critical safety function is not fully satisfied - operator action may eventually be needed.
Orange - The critical safety function is under severe challenge prompt operator action is necessary.
Red - The critical safety function is in jeopardy - i immediate operator action is required.
b) 1) Subcriticality 2) Core cooling 3) Heat Sink
- 4) RCS integrity 5) Containment 6) Inventory Ref.$'BAP300-11,p.6 wNiv $. nere rerWK1he of h Wowspp GW/s4Pr cow)M 15 W W 4.8 a. 1. A Keff of .95 or less or
. 2. A boron concentration of greater than or equal to 2000 ppm.
- b. 1. uncontrolled cooldown
- 2. insufficient shutdown margin
- 3. failure of reactor makeup control system
- 4. unexplained or uncontrolled reactivity increase
- 5. failure of more than 1 RCCA to fully insert following reactor trip or shutdown
- 6. control rod bank height below the rod insertion low-low limit when the reactor is critical Ref.: BOA PRI-2 4.9 Would result in possible loss of heat sink due to tripping of main feedwater pumps while the reactor is at power.
! Ref.: BCA-1, p. 2 l
l 4.10 1. Shift Engineer
- 2. a. Shift Foreman
- b. Station Control Room Engineer
- c. Nuclear Station Operator (Senior Experienced Personnel) 14
4.11 1. d.
- 2. b.
- 3. b.
- 4. e.
- 5. d.
- 6. c.
- 7. a.
- 8. g.
- 9. a.
- 10. e.
Ref.: 80A, BCA, etc.
4.12 Vice President of Nuclear Operations or Div. VP and General Manager of Nuclear Stations or Operations, Manager-Nuclear Station plus Station Superintendent Ref.: BAP 100-A13, p. 24 4.13 Containment hydrogen concentrator > 3%
_ RCS subcooling (corrected for subcooling error)
Pzr level < 20%
RCS pressure decreases by 200 psi Venting period (greater than calculated)
Ref. : 1 BFR - I.3, p. 8 4.14 a. 75 rem - W.B.; 200 rem - extremities
- 2. es n n1 f 6444'f/McP M/3 Sgrt Gengra g n
- 3. Stop any surveillances that would make any ES equipment inoperable
- 4. Stop fuel handling and/ processing of radioactive n.aterial Ref.: BOA ENV-1 4.16 1. '
' 'e Emergency 2.
- 3. General Emergency Ref.: BZP 400-1 15
4.17 5 (1.0)
T/S, p. 6-1 Ref.Notify 4.18 a.
- (AA* Y TSV*)
the Shift Engineer to initiate 1 805 LC0AR 3.1.1-la (0.5(64c4)
- b. Select bypass power mismatch rod stop upper / lower current comparator comparator channel defeat C.
Ref.:
7Wf 8/f 759/t '>
BOA - INST - I p, ff M Q h? $60*'V & t HWY g, Wg 40ptyfbrW fW*'Y l* >
O 16
CO2 Pri'nt Dirgn:stics for: BYRON OL EXAM T tal Formatting Exceptions = 20 Total Listed Below = 20 The Following Two Formats Will Be Used:
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MASTE CC'PY U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: BYRON 1 REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 85/09/16 EXAMINER: T BURDICK APPLICANT: _________________________
INSTRUCTIONS TO APPLICANT Uco separate paper for the answers. Write answers on one side only.
Stcple question sheet on top of the answer sheets. Points for each qv;stion are indicated in parentheses after the question. The passing grcde requires at least 70% in each category and a final grade of at least 80%. E:: amination papers will be picked- up six (6) hours after tha e:: amination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 00
_'5 00_'_I____ I I__
_'5 ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS
_ I __ _ 1 ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 25.00
________ _'____[5.00 ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 25.00
________ _'5.00 _____ ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIM 1 RATIONS 100.00 100.00 TOTALS FINAL GRADC ___,_____________%
All work done on this e:: amination is my own. I have neither given nor received aid.
EPPL5C5UiI5~55GU55UR5~~~~~~~~~~~~~~
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND P' AGE 2 QUESTION 5.01 (2.50)
Ascume one RCi' trips at 30% power without a reactor protection cyctem actuation or a change in turbine load. Indicate whether l the following parameters will Increase, Decrease, OR Remain the Sece.- ,.
- c. Flow in the operatin3 Reactor Coolant Systems loops. (0.5)
! b. The ratio of core flow compared to the total loop flow. (0.5)
(Core flow / Total loop flow)
- c. Reactor vessel delta-P. (0.5) i d. Actual Core delta-T. (0 5)
- o. An (RCS) operating loop steam generator temperature. (0.5) l OUESTION 5.02 (1.50)
- c. Provide TWO reasons for Xenon contributing more negative reactivity at full power than Samarium. (1.0) j b. What unuld happen to the magnitude of the equilibrium reactivity l due to Sarmarium, if reactor power was changed from 50% to 100%. (0.5)
QUESTION 5.03 (2.00)
For each of the following conditions, which of the two choices would the INDIVIDUAL (differential or integral as indicated) rod worth be 3reater?
Rod Worth Condition Choice 1 Choice 2 A. Integral Tavs 150-F 500-F (0 5)
D. Integral Core life BOL EOL. (0.5)
C. Differential Rod position 180 steps 215 steps (0.5)
D. Differential Rod in Bank C an inserted the rod (0.5) which is next rod withdrawn to a module with l (***** CATEGORY 05 CONTINUED ON NEXT PAGE xxxxx) l l
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3 QUESTION 5.04 (1.50)
State the core location that would require the highest heat flux to reach DNB and explain the basis of your choice.
(Start your answer with TOP OF, MIDDLE OF, or BOTTOM OF the coro.) -
(1.5)
GUESTION 5.05 (2.00)
Consider two separate startups. Assume the only difference between the two is that during startup ti rod speed is twice as fast as startup
- 42. Qualitatively COMPARE and BRIEFLY EXPLAIN the differences between the two startups in re3ard to the fo11owin3 (Tha time of rod pulls and time between rod pulls are the same.)
- o. Critical rod height (0.5)
- b. Time to criticality (0.5)
- c. Power level at criticality (0.5)
- d. Startup Rate at criticality (0.5)
QUESTION 5.06 (1.30)
Ycu have just completed a reactor startup and power level is at the point of edding heat. For the following situations, indicate whether the final, otable power level will be bisher, lower, or the same. Explain your answers assuming the core is at mid-life. Treat each situation separately.
- a. Steam' dump pressure setting is raised by 20 PSIG.
- b. A large steam leak develops outside of containment.
- )
(***** CATEGORY 05 CONTINUCD ON NEXT PACE l
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4 OUESTIOli 5.07 (1.00)
Assume RCP's are tripped followins a LOCA. After the break has been isolated which of the followir.s situations would be MOST desirable?
P P7R T PZR T HOT T COLD
- c. 600 " 520 500 480
- b. 800 530 530 520
- c. 1000 545 540 530
- d. 1200 567 575 565 QUESTION 5.08 (1.50)
- a. How does the enthalpy change across the steam generator secondary as power increases? (0.5)
- b. What factor affects this change in enthalpy? (1.0)
GUESTION 5.09 (1.00)
Radionuclidet. such as Xenon, Krypton, Cesium and Iodine would be present in tho reactor coolant following a fuel clad failure. How would each of these bo removed f r o n. the RCS?
00ESTION 5.10 (2.00)
- c. Wher, would REFLUX E:0ILING be used for core coolin3? (0.5)
- b. How does it work? (1,5) i
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5 i
QUESTION 5 11 (2.00)
Shawn on figure 5-1 are typical axial flux profiles at full power for BOL cnd EOL.
- o. Why does the flux peak shift to below the core centerline as power is increased'at BOL7 (1 0)
- b. State the reason for each flux peak CA and 83 on the EOL curve. (1.0) i QUESTION. 5.12 (1.00)
Name two changes that take place in nuclear fuel which tend to cause an increased fuel temperature as the core aSes.
QUESTION 5.13 (2.50)
- o. Name and define F 9(Z) (1.0)
- b. Why must FqtZ) be modified by K(Z)? Two reasons are required. (1.5) i OUESTION 5.14 (1 00) l l Enpress 100 pcm in units of %DK/Kr and DK/K.
l l QUESTION 5.15 (2.00)
A typical Westinghouse four loop plant trips from 20% power during startup '
l following a 40 day refueling outage. The operators find it impossible to cointain T eve in specification due to a cooldown but steam dump and all ,
other systems are normal. They decide to close the main steam isolation i
volves. State why yno feel thin is or is not normal plant behavior.
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i (msens END OF CATEGORY 05 mamas)
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- 6. PLANT SYSTEMS DESIGH, CONTROL, AND INSTRUMENTATION PAGE 6 OVESTION 6.01 (2.00) .
In June tnit one enperienced an automatic OTDT reactor trip. The trip took picce as operators were unloading the turbine because the circulating COter pumps had tripped.
- o. Why did operators try to unload the turbine?
- b. What would have happened if the operators had not unloaded the turbine?
c.Whydidn'tsteamdumphelpprevggj}histrip?
- d. The OTDT tr ip took place AFTER /PORV lif ted. Why?
DUESTION 6.02 (2.00)
Ecclier this year unit one was subjected to a loss of offsite power test.
During that test a safety injection occured as well as unenpected RCS pressure increases.
- c. The SI occured due to low SG pressure but SG pressure was greater than the setpoint for SI. Why did SI initiate above the setpoint?
- b. Following the loading of the emer3ency diesel generator pressuri:er pressure began increasing. Why?
- c. Eventually RCP's were restarted. Which one's were started first? Why?
Why?
00CSTION 6.03 (2.00)
During a loss of offsite power test this year an SI occured. The operator Cbterved that the A train RHR pump had started but the B train pump had not and manually . tarted it per the procedure. The D RHR pump was initially d:clared inoperable but later determined to be opecable. Why did the pump roturn to oppenble status when the operator had to manually start it during tho SI?
l (manns CATEGORY 04 CONTINUED ON NEXT PAGE *****)
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 7 QUESTION 6.04 (2.00) .
Chocistry informs you, the Shift En3ineer, that the Li-7 concentration is too high and the cation bed must be placed in service to reduce it.
- o. How much letdown flow can be diverted throV3h the cation bed?
- b. Is letdown flow diverted through the cation bed alone or in addition to the mined ' bed domin?
- c. Besides Li-7, name two other elements removed by the cation bed.
- d. Is the cation bed valved into service in the control room or locally?
QUESTION 6.05 (1.00)
Why is the high temperature alarm setpoint for the letdown line set lower ;
thcn the setpoint for the excess letdown line? i 00ESTION 6.06 (1.00)
- c. When in low temperature overpressure protection armed? C.253
- b. What are all the inputs si3nals for low temperature over pressure protection? C.753 1
00ESTION 6.07 (1.00)
- o. What antal regions of t:.e reactor vessel is monitored by the RVLIS?
- b. When are the level indications considered valid?
- c. What type of detectors are used?
- d. Where is the output from RVLIS displayed? l l
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(***** CATEGORY 06 CONTINUED ON NEXT PAGE mamma)
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 8 QUESTION 6.08 (1.00) .
- c. Why do the containment cooling fans shift to slow during accidents?
- b. Containment charcoal filter systems have deluse fire protection. What is the source of heat?
l QUESTION 6.07 - (2.00) gucivsf '
Earlier this year the Zion station unit one was at 20% power with feedwater control and rod control in manual. The Shift Engineer noticed that the steam dump valves were open 20%. The reactor operator was using a ceanuter trend S; v' 5 TCm3 T**Sor der to moEitoE Tave and Tref at the time. 955b # process M te "Ul * *' '- ^' t '
dT44'
- c. 'What caused 20% steam dump at this point ' assuming the system was not malfunctioning?
- b. Why would the operator use the process computer trend recorder at this point?
OUESTION 6.10 (1.50)
Nsce sin of the seven reactor trips that are subject to the P7 permissive.
No setpoints or coincidence are necessary.
QUESTION 6.11 (3.00)
- b. State the design liquid volume, pressure and boron concentration of the accumulators.
QUESTION 6.12 (1.50)
ECCS component status lights are arranged into six groups. Each stoup has a significant meaning. State the meaning of each group of lights.
(***xx CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx)
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 9 QUESTION 6.13 (1.00) -
When is the CCW pump automatic start feature for low CCW header pressure blocked?
QUESTION 6.14 (2.00)
- e. Can the em'ersency diesel generators be started without service water cooling available? E.53
- b. How would a loss of service water affect diesel senerator operation in both normal and emergency circumstances. El.5]
OUESTION 6.15 (1.00)
Indicate whether the followins statements are true or false concerning the response of the Rod Control System. (No explanation is required)
- s. An utsent failure in a power cabinet sends a signal to the logic cabinet inhibiting all rod motion.
- b. At 3% below the DP delta T/0T delta T trip setpoints automatic rod cotion is inhibited.
- c. If turbine power falls below 15% automatic rod withdrawl is blocked. '
- d. At 103% reactor power automatic rod insertion is inhibited.
QUESTION 6.16 (1.00)
What are two (2) conditions which would cause an automatic bumpless trans-for of the main turbine from ai'tamatic control to manual control?
~(***** END OF CATEGORY 06 xxxxx)
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND DAGE 10
~ ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~R5656LU5565L 6UUTR6L QUESTION 7.01 (1.00)
Name five alarms that are listed as symptoms of a reactor coolant pump seal failure by the procedure.
QUESTION 7.02 - (2.00)
'Nsme five of the eiSht reasons for workers to evacuate a controlled area as quickly as possible according to Radiation Protection Procedure BRP 1000-A1.
QUESTION 7.03 (1.00)
Whsn you observe the word ' DANGER' in association with radiation protection signs what does it mean?
OUESTION 7.04 (2.00)
How does a type 1 RWP differ from a type 2 RWP in regard to its application and duration?
QUESTION 7.05 (1.00)
The procedure for power ascension, BGP 200-3, cautions the operator not to exceed 10 % reactor power while acceleratins the main turbine to synchro-nous speed.-Why?
QUESTION 7.06 (2.00)
A caution appears in the power ascension procedurer.BGP 100-3, just after the step involving trip-testing.the turbine. It states that if the VPL can't be increased to 120% IMMEDIATLY after relatching the turbine below 1700 rpm then set the reference demand to 50 rpm below actual turbine speed PRIOR to increasing the VPL to 120%.
- a. WHAT problem may occur if this is not done? E.53
- b. WHY does this problem occur? [1.03 BE SPECIFIC
- c. HOW does this precaution prevent the problem? C.53
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
- 7. PROCEDURES - NORMAL, ABNORNAL, EMERGENCY AND PAGE 11
~~~~ R5D56EUU5CdE~66OTR6E~~~~~~~~~~~~~~~~~~~~~~~~
QUESTION 7.07 (1.00)
BOP AP-41, 'Synchronizins a SAT to a bus fed by a DG', states as a precau-tion that when a diesel generator is in the emergency mode and paralleled to the grid DO NOT return the emergency mode speed / voltage C/S to auto if it is currently in the manual emergency position. Why?
00ESTION 7.08 (2.00)
Should the RHR system be totally inoperable during refuelins what are the four op_tions available for core cooling according to the procedure, 1 BOA Refuel-4?
OUESTION 7.09- (2.00)
The procedure for a stuck or misaligned rod calls for verification of DRPI operability using FOUR possible means. Name them.
OUCSTION 7.10 (2.00)
According to the procedure for condensate /feedwater malfunctions, what are
.thw maximum power levels for' Lassume all other equipment is available]
- e. one main feed pump available
- b. no heater drain pumps available
- c. two CC/CB pumps available
- d. one CC/CD pump available -
QUESTION 7.11 (1.00) .
The site emergency plan implementing procedure requires that state and local vjenries be notified after an emersency is classified. What is the time limit ,
e offsite notification?
(***** CATEGORY 07 CONTINUED ON NEXT PAGE xxxxx)
7.- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 12
~~~~R5656E6556 E~U6UTR5E------------------------
QUESTION 7.12 (1.00)
Name the two responsibilities of the Statirn Director which may not be dale 3ated.
QUESTION 7.13 ~ (2.00)
- a. State the four symptoms for a reactor trip as given in the procedure.
- b. State the four symptoms for safety injection as given in the procedure.
OljECTIOri' 7.14 (1.00)
- c. Who is designated to scan the BST's during an emergency?
, b. What.is the minimum frequency of BST scanning during an emergency?
QUESTION 7.15 (2.50)
Name all the criteria for tripping RCP's during an emergency as specified in 1 BEP-0 GUESTION 7.16 (1.50)
The Byron operating procedures contain low level sters that are preceeded by either letters, open bullets or closed bullets. What is the purpose of each?
4 (xxxxx END OF CATEGORY 07 xxxxx)
- 8. ' ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 'PAGE 13
{, f - 87 / 6[ Cd6 c)u CJMcc7 C6%A QUESTION- 8.01 ( .50) .
Functional checks
- a. should not be utili=ed as a means of independen'. verification if the check-could interrupt plant operation.
- b. used as a method of independent verification, should be examined to ensure they' test the entire portion of the system which was affected by previous actions.
i c..are acceptable if the indication utilized is either null or zero and L if the readios is directly observed and recorded by the observer.
- d. may~blso be satisfied by a visual verification, apart in time and documentation, of equipment alignment by a second qualified person.
g GUESTION 0.02 ( .50)
Daily Orders:
I a. .may be used in place of temporary procedures where no safety concern i
is~ involved.
- b. should not be in effect for greater than one month.
- c. .sfrecting other departments will be brought to their attention by the Shift Engineer.
- d. required to last greater-than one week shall be stated as such.
OUESTION 8.03 ( .30)
Operatins Lo3*
- a. errnts will be voided by drawins a single line th~ rough the entry and initialed and dated by the person discoverin3 th'e error.
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- b. entries shall be made usins black ink and each shift should begin its log entries on a new line.
- e. late entries should be preceeded by the word " note" and the time at which the entry was made.
- d. out-or-tolerance data taken shall be written in red' ink.
(varvu CATEGORY 08 CONTINUED ON NEXT PAGE *****)
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V 8.. ADMINISTRATIVE PROCEDURES, CONDITIONS, LIMITATIONS PAGE 14
___________________________________________'AND _______________
GUESTION 8.04 ( .50) .
. The Temporary Alterations procedure:
- a. is intended to document alterations which exist on operable systems and corkponents.
i
- b. .is applicable to situations involving Out-Of-Service equipment.
.c. excludes cases involving an approved procedure which does not require Shift ens i neer notification or independent verification.
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- d. is not intended to apply to situations affectin3 safety systems.
~
QUESTION 8.05 ( .50)
Binders containing DOA's are colored
.a. oranse
- b. red
- c. blue
- d. yellow OUESTION 8.06 ( .50)
Tha'most important critical safety function is:
, a. containment
- b. RCS integrity
- c. suberiticality
- d. heat' sink (xxxxx.CATECORY 08 CONTINUED ON NEXT PAGE *****)
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 15 GUESTION 8.07 ( .50)
Tha hi 3h est priority in the status trees is designated by the color
- a. green
- b. red
- c. yellow , -
- d. oranse GUESTION 8.08 ( .50)
Status'~ tree scanning during emergencies should be continuous if any condition code higher than ________--____ is found to exist.
- a. yellow
- b. red
- c. 3reen
- d. orange GUESTION 8.09 ( .50)
BGP flow charts utilize the symbol of a box to represent a step
- a. to be. completed and initialed by the NSO.
- b. that is to be approved prior to startin3 by the initials of the SRO.
- c. with an option to either do the step or bypass around the step.
- d. that'is to be approved prior to starting and initialed after completion by the SRO.
4
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 'PAGE 16 QUESTION 8.10 ( .50) -
Out-Of-Service cards are colored;
- s. blue on white
- b. red on white
- c. black on yellow
- d. red on yellow OUESTION 8.11 ( .EO)
The equ'ipment outase forms for unit one are colored:
- a. white
- b. yellow
- c. steen
- d. blue GUESTION 8.12 ( .50)
The proper order for tassins a pump out of service would bet
- a. suction valve, discharse valver control switch, breaker.
- b. discharge valve, suction valve, breaker, control switch.
- c. breaker, control switch, discharge valve, suction valve.
- d. control switch, breaker, discharge valve, suction valve.
(xxxxx CATECORY 08 CONTINUED ON NEXT PAGE **xxx) r i
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- 8. PAGE 17 i .____ ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS l
QUESTION 8.13 ( .50) l The ______________ will place an asterisk next to those isolation points i requiring an input to ESD on Equipment outage forms. l
- a. Shift En3 neer i
- b. Shift Foreman
- c. Center Des'k'NSO -
- d. Unit'NSO QUESTION, 8.14 ( .50)-
Verbal concurrance made by telephone for temporary 00S card lifting may be obtained by the*
- a. person requesting the action.
- b. Shift Engineer.
- c. Plant Superintendent.
- d. Shift Foreman.
QUESTION 8 15 (2.00)
In March 1985 an NSO inadvertently. actuated ESF when he meant to close an MSIV. What new policy was-established to help prevent this kind of error in the future?
QUESTION 8.16 (2.00)
A policy was established at Byron prohibiting the us.e of radio communica-
.tions in the containment building. Why?
(mmmmu CATEGORY 08 CONTINUED ON NEXT PAGE *****)
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 18 QUESTION 8.17 (2.00) .
The maintenance department notifies your the Shift Engineer, that they have completed adjustments to a limit switch on 1 PRO 66, a containment isola-tion valve for process radiation monitorins. The Unit NSO also states that the valve cycles freely and that the position indication is functioning properly. Are you satisfied that this valve is operable? Why?
QUESTION 8.18 (2.00)
When a pressurizer PORV is declared inoperable due to other than seat leakage what.must be done within one hour according to Tech Specs?
GUESTION 3.19 (2 00)
Accordin3 to CAP 300-22 MOV's that perform a safety function should be evaluated for operability whenever they are manually closed or placed on their backseat. Why?
DUESTION 8.20 (2.00)
- a. In the event of an emersency who will direct recovery and procedural flow in the control room? [1.0]
- b. How soon should the SCRE be relieved of line duties after an emergency occurs? [1.03 fu' ( ; 8/90 3CO ~ Z L QUESTION 0.21 (2.00)
State f.ive of the eight abnormal conditions listed in BAP 300-28 that require immediate suspension of core alterations.
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(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
8,'_ ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 19 QUESTION -8.22 (2.00) .
Personnel entering containment are required to verify the position of incore detectors beforehand.
- a. What is the desired status of the detectors? [1.53
- b. What is the significance of the incore detector position?
E.5]
GUESTION 8.23 (2.00)
- a. How are LCOAR's containing actions that must be completed in one hour or less. identified?
- b. Where are document forms for LCOAR's in progress kept while not directly being used?
(***** END OF CATEGORY 08 *****)
(************* END OF EXAMINATION ***************)
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o Table 1. S0turated Sloam: Temperatuee Table-Continued aa ben Spec.i.c toivme taina'p, ini,ep, lemt t e s+- Sa* 1st Sat $st $at Sat tems lam Se er Lit #C luar Wapo- Lited less Wapo- Litee (esp Vapo' f ane g o. e e, ey he h e, hg s, g ,, ,g, t 460 0 4649' O t1961 0 97467 0 99s74 mit 7631 1208 0 0 6a05 03711 1470s ame s
, ana l 4 B: $4 00196 093Sl! 0 95557 M61 het I?os ? O pva 03213 1AM7 asa s 464 0 904 8:- 00:64 0 8tst5 0 9146? 460 7 het 1704 6 0 6507 08177 l a676 ass e
- *' 4770
- 574 6* 00:1 64 040 5 e sg3M e55J 743J 120s3 u$$1 0 0042 l ets? 4778 ellt $4a li 641997 037tM 03495C 4393 744 3 1204 3 0 6S99 0 79M 1Athh 476 8
.. aas e SM !* . 0 67000 0 79716 0 81717 464 5 739 6 .1208 1 _ 0 5648 8 7871 1 4519 StL ,,,*
aes t St*t 00702 0 PM13 0 7M77 469 1 734 7 I?c3 8 SMW 0 77t5 lut; ess a esa e lit 1: O C7C:* 0 7E4! 0 7MM 4730 77t ? 1703 t 0 674! 0 770c l aa44 esa t it:t 8 il C.: 0 t7c7e a 70798 8 7712t- 4785 774 6 12G31 0 6753 0 76)a 344;$ ag7 g ele s Gh t; 807C3s 06061 0 70100 483J 7193 12G2 7 S taas e.734s s esso ege t .. ,
est t Lt; 8; 8 0704) O tMat 0 67897 4879 714J IPC21 0 6890 Sies) 1 4333 set t tes t 73 70 0 0M*) 0 6?t3r 0 6a 95. 4977 709 0 120i f CHM 0 7357 l a7E taa s tes t 731 4' O C?u? 0 60$K 0 6255: 4976 703 7 120 1 06108 0 7271 3 4753 asa s lit t h? 7; O C?tF2 O tt?lt 0 6C?tt 502 3 69t1 1700 5 0 7030 0 71th 34224 SIP s tie s FW 74 - 0 C205. C D197 038079 5073 652 7 !!9ht 0 7045 0 7095 14111 5163 l?t t 5:? 11 0 OMt. OM8M t thtSi .517 C 647 C litt t 0 7133 J 7013 1 414( 17t 8 N7 Het 84 ! (. S C:1* 0 5:614 0 13116 $16 % 641 3 litt ? 0 718? 0 617t 1 4106 174 4 ~ w e 4 '-.;*
lis t 47; s. O C'::: 0 4this C 6'it! S2I t 671 $ lit?J 0 723; 0 62.35 147; 174 8 S3: t tX M t C?:.*3 04754* C 53C 7; 126 e eni llte a 0 77s; . 0675? I aC3: 637 2 636 8 li; ;* C C;.34 44t;23 0 4&;T 131 7 M3L 18924 0 732b 96Mt IJth) 534 8 WI '
St* 75 0 C:;41 0 a43P 0 ott!! SM i 6573 11H ? O 7371 0($?7 L395a
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i 8*18 176 6 t t'49 C IM:: CJ70! &af t SN ? Ili!! 8840 0 4649 3 J01: sit t ,
s se it . t ;743 C it*3* CJ:!h 651 . 8Ct 11464 C la ti 0858) 13Gs. 87s t t?l t Ib a E t?).a C l ?tt' C10N Ett i ofi
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- 1131 : 0 5624 04h; IJt'i 63. I 6a8 t ??tt t t'* t- C 854?- 0117.?; $79 : 4546 1133
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND 'PAGE 20 ANSWERS -- BYRON 1 ,
-85/09/16-T BURDICK ANSWER 5.01 (2.50)
- c. Increase
- b. Decrease
- c. Decrease ' ^ ^
- d. Increase
- e. Decrease [0.5 each3 (2.5)
REFERENCE l CER PTD.R0 FUN HT/ THERM 0/FF ANSWER 5.02 (1.50) l c. 1. Hi 3h er fission yield.
- 2. Larger (thermal) absorbtion cross section (1.0) l
- b. Remains the same (0.5)
REFERENCE CE PTD RO FUN COREOPS ANSWER 5.03 (2.00)
A. 2 (0.5)
B. 2 (0.5)
C 1 (0.5)
D. 2 (0.5)
REFERENCE CE PTD RO FUN PWR COREOPS 6 ANSWER 5.04 (1.50)
Dottom of CO.51, because the temperature is at the lowest value [0.53 and the pressure is at the highest value CO.53. (1.5) a
c.
4 .
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 21 ANSWERS -- BYRON 1 -85/09/16-T BURDICK REFERENCE CE PTD RO FUN COREOPS ANSWER 5.05 (2.00)
- s. Critica1' rod height: same for both E0.23 Rod height is dependent upon the amount of reactivity needed for criticality E0.33 (0.5)
- b. T i nie to criticality shorter for startup 41 E0.23 The reactivity addition rate.is greater, thus a shorter time CO.33 (0 5)
- c. Powbr level at criticality lower for startup 41 E0.23 The higher rod speeds do not allow for neutron population buildup as great as in startup 82 E0.33 (0.5)
- d. SUR at criticality: Higher for startup 41 E0.23 This is because of the higher reactivity addition rates CO.33 (0.5)
REFERENCE CE PTD RO FUN PWR COREOPS ANSWER 5.06 (1.50)
- a. Loweri the steam dump pressure setting increase causes a RCS temper-ature increase. MTC and FTC(doppler) both add negative reactivity to lower reactor power.
- b. Higher; the increased steam flow will result in a lower RCS temper-ature. MTC will add positive reactivity and power will rise. Also lower; if low SG pressure SI setpoint is reached Rx will trip.
c Lower; the negative reactivity inserted by boro'n will cause power to decrease.
REFERENCE PTD PDFUN COREOP3 ANSWER 5.07 (1.00) c.
I I
4 *
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND 'PAGE 22 ANSWERS -- BYRON 1 -85/09/16-T BURDICK REFERENCE CE PTD RO FUM HT/ THERM 0/FF ANSWER 5.08, (1.50)
,. a. The erithalpy change across the SG secondary decreases as power increases
- b. fs'edwater temperature increases REFERENCE CE PTD-RO FUN HT 6-46 ANSWER 5.09 (1.00)
- c. Xenon-desas
- b. Krypton-desas
- c. Cesium-ion exchanSe
- d. Iodine-ion exchanse (.25 each)
REFERENCE B/B FSAR AND CVCS SYS DESC. CHAP 46 ANSWER 5.10 (2.00)
- a. Under accident conditions using natural circulation with a voided core.
- b. Steam exits the core E.53 is condensed in the SG tubest.53 and the condensate returns to the core via the hot les to repeat the cyclet.53 REFERENCE l CE PTD ROFUN HT/ THERM 0/FF
s
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 23 ANSWERS -- CYRON 1 ,
-85/09/16-T BURDICK ANSWER 5.11 (2.00)
- a. Due to the moderator temperature decrease in the lower part of the core causing better thermalization and more power than in the upper half.
- b. The upper 3eak is due to a higher fuel density as a result of lower core depletion whereas the lower peak is due to the moderator affect as stated in part a.
REFERENCE CE PTD ROFUN PWR COREOPS 8-20,21 ANSNER 5.12 (1.00)
- a. fuel densification
- b. reduced gas thermal conductivity
- c. clad corrosion
- d. crud buildup E2/4 0 .5 each]
REFERENCE CE PTD ROFUN PWR COREOPS PAGE 2-45 AND HT PAGE 3-63 ANSWER 5.13 (2.50)
- a. Heat Flu:: Hot Channel Factort.25] is the maximum local heat flux at core elevation Z divided by the average heat fluxC.75]
- b. The coolant temperature increases as it rises through the core and DNBR is lower as a resultE.5]. Also, during a LOCA the upper part of the core is uncovered first and reflooded last leavins it uncovered longer and reaching higher temperatures [1.0]
REFERCl!CC CC PTD'ROFUN HT PAGE 13-34
i 5.- THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 24 ANSWERS -- DYROH 1 -85/09/16-T BURDICK ANSWER 5.14 (1.00) 100 pcm = .1% DK/KE.53 and .001 DK/KE.53 REFERENCE-CE PTD ROFUN PWR COREOPS CHAP 2 l
l ANSWER 5.15 (2.00)
Yes it is normal.E.53 Since the plant just came out of refueling the core does not have enouah decay heat.to sustain Tave under the post trip steam load condition .E1.53 Edue to AFW cooldown and MFP3 l REFERENCE l
CC PTD ROFUN PWR COREOPSi 1BEP ES 0.11 BGP100-2,page 3 t
i l
f i
T
D *
- 6. PLANT SYSTEMS DESIGN. CONTROL, AND INSTRUMENTATION PAGE 25 ANSWERS -- BYRON 1 -85/09/16-T BURDICK ANSWER 6.01 (2.00) -
- a. loss of vacuum
- b. Iow vacuum turbine trip
- c. SD interlocked off by ri c eter N N t O
- d. trip setpoint was lowered by PORV E.5 each]
REFERE!8CE LER 4G4/05-062 ANSWER 6.02 (2.00)
- a. rate compensation E.53
- b. due to pressuriner heater restoration CL - ___C;egU ) " YA E.53
- c. D and C E.253 for spray flow E.253~//
- d. e::pansion of the cold water pocket formec in idle pump from seal injection flow after the pocket is heated E.53 REFERENCE LER 454/05-035 ANGWER 6.03 (2.00)
The D RHR pump was manually started because it hadn't sequenced on the acergency diesel generator yet. The loss of offsite power test caused a loss of B traire power but A train was unaffected so A RHR pump started imaediatly.'
REFERENCE LER 454/85-027
- 6. PLANT. SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 26 ANSWERS -- BYRON 1 -85/09/16-T BURDICK ANSWER 6.04 (2.00)
- c. 75 spm be with the mi::ed bed
The letdown path'is through demins and the excess letdown path is not.
Tho demins can't withstand high temperature.
REFERENCE CVCS LP, pages 28, 39 ANSWER 6.06 (1.00)
- a. Less than or equal to 380 degrees F.
- b. 1h pressure
- 2. WR Teold
- a. upper head and plenum above the core
- b. Upper head any timei plenum area with RCP's off
- c. thermocouplesfl46'M'N#'**
- d. SPDS and MCB a
- 6. PLANT SYSTEhS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 27 ANSWERS -- BYRON 1 -85/09/16-T BURDICK
~
REFERENCE RCS LP, pages 75-78 ANSWER 6.08 (1.00)
- c. To protect the motor from overload due to moisture in the air.
- b. Fission products.
REFERENCE FSAR 6.2 ANSWER 6.09 (2.00)
- c. The reactor power is greater than turbine power and excess power is soins.out the steam dump system. [1.03
- b. Thu trend're' corder is capable of a much narrower ranse for fine control than isithe field instrumentation. [1.03 REFERENCE
'LER 295/85-05 ANSWER 6.10 (1.50)
Icw flow RCP UV RCP UF Icw PZR pressure high PZR level turbine trip RCP breaker trip C6 0.25 each3 REFERENCE LOO 6 page 28
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 'PAGE 28 ANSWERS -- BYRON 1 -85/09/16-T BURDICK ANSWER 6.11 (3.00)
- o. SOH RATED HEAD RATED FLOW RHR 200 psig 160 psis 3000 spm SI 1520.psig 1100 psis 400 spm CCP 2600 psis 2500 psis 150 spm
- b. 6995-7217 gallons, 1900-2100 ppm, 65^ ps+3 a [12 0 .25 each3 RIFERENCE
- 0 01 ~ "/ #
LO13, p'bges 10-12 ANSWER 6.12 (1.50)
CROUP 1-readiness status 2-injection status 3-containment isolation [2b,byj 4-cold les recirculation 5-hot les recirculation 6 phase B and containment spray M W AS/#d CL/T T E.25 each]
REFERENCE LO13 pages 23-24 ANSWER 6.13 (1.00)
Following a loss of power when pumps are sequenced on the diesel.
REFERENCE ,
<LO22 page 11 0
S
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 29 ANSWERS -- BYRON 1 -85/09/16-T BURDICK ANSWER 6.14 (2.00)
- a. yes C.53 -
- b. Normally the diesel would overheat and trip due to high jacket water tewmperature. C.753 In an emergency the diesel would continue to overheat with no trip feature for protection and failure would be imminent without operator action. [.753 REFERENCE LO24 g ANSWER 6.15 (1.00)
- a. false
- b. true
- c. true
- d. false [.25 each]
REFERENCE LOO 4, page 25 ANSWER 6.16 (1.00)
- 1. load control
- a. loss of vital power supply b. error detected by computer that wonid hinder auto control
- 2. spaed control
- a. loss of vital power supply b. error detected by comput,er etc.
- c. los of 2 speed channels.E% G.5 each3 3 ma , Lny Sa t-t-. fe W REFERENCE d LO12, page 26 l
_ _ _ _ _ _ _ _ . _ _ . _ . _ _ _ l
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 30
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~R5650L66565L"C6NTR6L ANSWERS -- DYRON 1 -85/09/16-T BURDICK ANSWER .7.01 (1.00)
- c. RCP seal leakoff flow low
- b. * * * '
high
- c. DP-low
- d. *
water injection filter DP hi S h
- e. outlet temperature high
- f. * '
water injection flow low 3 standpipe level high
- h.
- J
- low ES 9 .2 each3 REFERENCE BOA RCP-1 ANSWER 7.02 (2.00)
- c. When told to do so by the RAD-CHEM stoup.
- b. Failure of personnel protection equipment.
- c. Unexpected deterioration of radiation conditions.
- d. Uncertainty about dose status or dose equivalent limit is reached.
- e. ' ASSEMBLY' siren sounds.
f.' Work ansignment is completed.
g.-Injury.
- h. Ilnenpoeted .3rea radiation alarm and the area dose rate is unknown.
[5 9 .4 each] ,
REFERENCE CRP 1000-A1, page 9 ANSWER I.03 (1.00)
- 0 ANGER' is designated exclusively for high radiation areas only.
REFERENCE DRP 1000-A1, pese 10
. ., .. . _ _ . . - . _ _ --.m -
/
4 e
I
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE- 31
~~~~RA5i5E5EiEAE E5tlTR6E~~~~~~~~~~~~~~~~~~~~~~~~
1 ____________________
ANSWERS -- BYRON 1 -85/09/16-T BURDICK 1.
ANSWER -7.04 (2.00)
Type:1 RWP's are for ro.utine access or work in areas where a whole body
" dose equivalent'to 50 mrem / day will not be exceeded E.53 whereas the type 2 RWP'is for greater expected doses or jobs' involving contaminated / airborne problems. C.53 3.
t'
. Type l'is good for one calender" year E.53 whereas type 2 is good for the duration of the~ job.C.53 e REFERENCE BRP.1000-A1, pages 12 and 15 i
- - AflSWER 7.05 (1.00)
F If the turbine trips while reactor power is above 10% then the reactor 4
trips.-
m ' REFERENCE BGP 100-3 ANSWER 7.06 .( 2.00)
! 'A. . Failure to do so could result in a reactor trip due to excess steam pres sure in the turbine.t.53 B. Excess pressure is a result of pilot valves "
- opening as speed drops before the stop valves are open.C.53 With the throt-l tle valves: closed below1700 rpm the RPS sets a turbine trip signal when 1
above.10%' power- as indicated by the high turbine' steam pressure.E.53 C. .Roducins reference demand to 50' rpm below actual speed will prevent the
, pilot valves from opening too far and causing the pr~oblem as described.C.53 REFCREFCE LER-454/05-051
- ANSWER - "' 0 7
. (1.00)
To prevent.inochronous operation while in parallel.
- REFEpEHCE CDP AP-41 4
= , _ . - . . .
f..
~
f i
'7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND 'PAGE 32
~~~~~R 5656E655E5E~E6UTEUE~~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- BYRON 1 ,
-85/09/16-T BURDICK A N S',lE R 7.08 (2.00)
- e. e:: cess letdown and charging.
- b. Refueling purification pump suction on the cavity and discharge to the SFP with the refuelin3 Sate and sluice gate open.
- c. SI pump injection.
- d. SI accumulator injection. E.5 each]
REFERENCE 1DOA defuel-4. page 3 ANSWER 7.09 (2.00)
- a. incore f lu:: map
- b. OPTR
- c. thermocouple symmetry check
- d. bank demand and DRPI comparison E.5 each3 REFERENCE 180A R00-3 ANSWER 7.10 (2.00)
- a. 560_Mw
- b. 100 Mw
- c. 700 Mw
- d. 350 Mw Epercent power also acceptable] E.5 each]
! REFERENCE 1 00A SEC-1
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 33
~~~~E A656E6656EE~66U T RUE ~~ ~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- BYRON 1 ,
-85/09/16-T BURDICK ANSWER 7.11 (1.00)
Within 15 minutes after event classification.
REFERENCE ,_.
BZP 310-5, page 2-3 ANSWER 7.12 (1.00)
- a. declaration of an emergency condition
- b. decision to notify and recommend protective actions to offsite authorities E.5 each3 REFERENCE BZP 100-T1, page 4 ANSWER 7.10 (2.00)
! o.1. any reactor trip first out annunciator l 2. rapid decrease in neutron level indication
! 3. all shutdown and control-rods are fully inserted; rod bottom lites lit
- 4. rapid decrease in unit load to =ero power bei. any SI first out annunciator lit
- 2. group 2 injection monitor lites lit
- 3. ESF equipment actuated
- 4. SI actuated bypass permissive lite lit E.25 each]
^
REFERENCE 1 BEP-0, page i ANSWER 7.14 (1.00)
- a. STA &^ SC806
- b. 10-20 mtnotes E.5 each]
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND 'PAGE 34
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~R5656L66I65[~C6UTRUL ANSWERS -- BYRON 1 -85/09/16-T BURDICK REFERENCE BAP 300-11, page 8; BAP 300-22, page 7 ANSWER 7.15 (2.50)
- b. Phase B isolation
- c. RCS pressure low AND
- o. any SI pump flow E.5 each3 REFERENCE 1 BEP-0 l ANSWER 7.16 (1.50)
- c. Letters preceeding steps mean the sequence of steps must be adhered to.
- b. A closed bullet means all steps must be completed in any order.
- c. Open bullets mean only steps that apply can be completed in any order.
E.5 each]
REFERENCE BAP 300 11, page 4
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 35 ANSWERS -- BYRON 1 -85/09/16-T BURDICK ANSWER 8.01 ( .50) b.-
. REFERENCE BAP 100-13, p.,se 2 ANSWER 8.02 ( .50) c.
REFERENCE BAP 300-3, page 1 and 2 ANSWER 8.03 ( .50) 3 REFERENCE BAP ?,00-4, page 1-4 ANSWER 8.04 ( .50) a.
REFERENCE BAP 300-5, pages 1-6
( ANSWER 8.05 ( .50) d.
REFERENCE CAP 300-11, page 5 ANSWER 0.06 ( .50) c.
r l
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 36 ANSWERS -- BYRON 1 -85/09/16-T BURDICK
^
REFERENCE BAP 300-11, page 6 l
ANSWER 8.07 ( .50) b.
REFERENCE BAP 300-11 ANSWER 8.08 ( .50) i n.
REFERENCE I -DAP-300-11, page 8 ANSWER 8.09 ( .50) d.
REFERENCE DAP 700-11, page 11 1
ANSWER 8.10 ( .50) b.
l l REFERENCE CAP 300-18, page 3 ANSWER 8.i1 ( .50) c.,
REFERENCE BAP 300-18, page 3
l
~
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 37 ANSWERS -- BYRON 1 -85/09/16-T BURDICK ANSWER 8.12 ( .50) d.
REFERENCE BAP 300-18, pase 4 ,
ANSWER 8.13 ( .50) c.
REFERENCE BAP 300-18, pa3e 9 ANSWER 8.14 ( .50)
- b. 04 d.16 REFERENCE.
BAP 300-18, page 13 ANSWER 8.15' (2.00)
A second independent verification in non-emergency cases related to the canipulation of ESF and/or RPS related controls. Also the operator must reverify his hand is on the proper switch if he releases it momentarily prior to actuating it.
REFERENCE LER 454/85-034 ANSWER 0.16 (2.00)
The radio transmissions are believed to cause undesired response on plant instrumentation which could result in unwarted plant or system behavior.
REFERENCE LER 454/05-020
U
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 39 ANSWERS -- BYRON 1 -85/09/16-T BURDICK ANSWER 8.21 (2.00)
- a. Unenpected increase in count rate of any responding nuclear monitoring channel by a factor of 5 following the addition of a fuel assembly after the initial eight fuel assemblies have been loaded.
- b. Unenpected increase in'the count rate of all responding nuclear monitor ing channels by a factor of 2 following addition of a fuel assembly after the initial eight assemblies have been loaded.
- c. Unenpected change in RCS boron concentration of greater than 20 ppm as determined from two successive samples or RCS temperature changes by 10 dheroes from the baseline values.
- d. Containment evacuatioon alarm.
- c. Less than two nuclear monitoring channels have counting rates greater than or equal to 2 ces with both primary source bearing assemblies loaded in the core.
- f. Data fron. at least one responding detector is not being recorded.
- 3. The entrapolated ICRP based on count rate data indicates that critical-ity may occur within the next few assembly additions.
- h. Less than two nuclear monitoring channels are responding.
REFERENCE BAP 300-28, pages 5 and 6 ANSWER 8.22 (2.00) 4 a. stored 0.53 OR bottom of the vessel C.53 AND tagged 00S E.53
- b. radiation hazard C.53 REFERENCE CAP 1450-1, page 1 ANGWER U.23 (2.00)
- a. Circle on page one in the title section.
- b. Ori the 'LCDAR IN PROGRESS
- board hanging from o clipboard.
l L
So ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 'PAGE 40 ANSWERS -- BYRON 1 -85/09/16-T BURDICK REFERENCE CAP 1400-6, page 2
+ 4D -
0 L.. .a
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 38 ANSWERS -- BYRON 1 -85/09/16-T BURDICK ANSWER 8.17 (2.00)
No, the valve must also meet stroke timins limits per TS E4.6.3.13 to be
.daclared operable.
ANSWER 8.10- - (2.00)
Either restore it to operable [13 or close its MOV [.53 and remove power frem the MOV [.53 REFERENCE TS 1.4.4 ANSWER 8.19 (2.00)
If the valve is designed to auto stroke on an emergency signal from the position it was placed then thermal overload devices may trip rendering the volve inoperable.
REFERENCE DAP 300-22e page 4 ANSWER 8.20 (2.00)
- e. The first available SRO
- b. 10 minutes REFEFENCE CAP 300-22, page 6 I
I l
L. ,