IR 05000289/1988015
| ML20207C998 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 08/02/1988 |
| From: | Eselgroth P, Norris B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20207C974 | List: |
| References | |
| 50-289-88-15OL, NUDOCS 8808110021 | |
| Download: ML20207C998 (40) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 50-289/88-15 (OL) ; FACILITY DOCKET NO. 50-289 FACILITY LICENSE NO. DPR-50 LICENSEE: GPU Nuclear Corporation Post Office Box 480 Middletown, Pennsylvania 17057 FACILITY: Three Mile Island Unit 1 EXAMINATION DATE: June 22, 1988 CHIEF EXAMINER: A/Im 36<w d Ba' rry 3. Norris, Senior Operations Engineer Dite~ APPROVE 0 BY: Mt/ [ pf ts ' 85/88
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P(ter Wpselgroth, Cni4f, PWR Section Date
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SUMMARY: One RO candidate was re-administered Section 1 of the Written I examination; a waiver was granted for the other three (3) sections 1 of the Written examination and the entire Operating examination. The i examination was administered in the Region I office. The examination 1 was completed successfully and a license was granted. I l l l l l l 8808110o21 880803 PDR ADOCK 05000289 PDC
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REPORT DETAILS TYPE OF EXAMINATION: Replacement EXAMINATION RESULTS: l RO l l Pass / Fail l l l I l l l Written l 1 /0 l l l I I I i l0perating l Waived l l l l 1 I I l0verall l 1 /0 l 1 I l CHIEF EXAMINER: B. Norris, NRC Attachments: 1. SR0 Written Examination and Answer Key 2. Facility Comments on the Written Examination 3. NRC Response to the Facility Comments
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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION
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FACILITY: TMI-l REACTOR TYPE: PWR-B&W177 _ DATE ADMINSTERED: 88/06/22 EXAMINER: YACHIMIAK, E. __ CANDIDATE i INSTRUCTIONS TO CANDIDATE: 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%. Examination' papers will be picked up six (6) hours after - - the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGOR7 25.00 100.0 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25.00 _
% Totals Final Grade All work done on this examination is my own. I have neither given ..
nor received aid.
Candidate's Signature
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
3. Use black ink or dark pencil only to facilitate legible reproductions.
4. Print your name in the blank provided on the cover sheet of the examination.
5. Fill in the date on the cover sheet of the examination (if necessary).
6. Use only the paper provided for answers.
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7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.
9. Number each answer as to category and number, for example, 1.4, 6.3.
10. Skip at least three lines between each answer.
11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
12. Use abbreviations only if they are commonly used in facility literature.
13. The point value for each question is indicated in parentheses after the
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question and can be used as a guide for the depth of answer required. , i 14. Show all calculations, methods, or assumptions used to obtain an answer i to mathematical problems whether indicated in the question or not.
15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
I 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
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i- . j 180 When you complete your examination, you shall: i
. 1 a. Assemble your examination as follows: (1) Exam questions on top. l (2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer. ) b. Turn in your copy of the examination and all pages used to answer < the examination questions.
c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your 11 cense may be denied or revoked. 9 l
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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Page 4
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THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW l l
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l l QUESTION 1.01 (2.50) List FIVE (5) ways an operator insures core peaking values are not exceeded during plant startup or power operation.
QUESTION 1.02 (2.00) HOW would a 10 degree DECREASE in feedwater temperature affect (Increase, Decrease, Not Change) each of the following parameters? Assume the plant is at 15% power and is NOT in Track. Consider each separately.
a) RCS Flow
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b) Reactor Power [4 X 0.50] c) OTSG Level d) Pressurizer Level QUESTION 1.03 (1.00) Describe the behavior of RCS pressure both before AND after the OTSGs boil dry it a small Break LOCA event were to occur without the availability of any feedwater. Assume automatic ESF actuation does not occur and no operator actions are taken.
QUESTION 1.04 (1.75) . Utilizing steam Tables or the Mollier Chart, calculate the approximate value for each of the parameters specified below. Consider each separately.
a) Temperature that steam would have exiting the OTSGs at 1000 psig with 45 degrees F. superheat. (0.50) b) Amount of subcooling available (in degrees F.) with the RCS at 1500 psig and 560 F. (0.50) c) Temperature of steam exiting a slightly open (throttled) pressurizer PORV with RCS pressure at 1400 psig and PRT pressure at 5 psig. (0.75)
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' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Pago 5 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
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QUESTION 1.05 (2.00) Briefly state the production and removal mechanisms for Xe-135 and Sm-149.
QUESTION 1.06 (1.75) A reactor is subcritical by a calculated amount of 5% delta k/k. By HOW much (in ppm) should the boron concentration be diluted to cause the count-rate to double? Justify your answer by calculation (see Attachment 1).
State any necessary assumptions and show your work.
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QUESTION 1.07 (1.50) WHEN during non-accident plant operations does the reactor vessel experience the greatest stresses and WHAT are TWO (2) parameters that can be controlled to limit these stresses? QUESTION 1.08 (1.00) Explain HOW the Excore Nuclear Detectors can be used to determine core water level during a LOCA.
. QUESTION 1.09 (1.00) WHAT is the approximate steady-state value of SUR after a reactor trip from 100% power and WHAT is the source for these neutrons? e (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) I _. __ _
l 1, '. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Pago 6 l THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
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QUESTION 1.10 (2.00) ! A variable speed centrifugal pump is operating at 1/4 rated speed in a CLOSED system with the following parameters: Pump Flow = 75 gpm Pump Power = 50 KW Pump Head = 300 psid a. Calculate the new values for these parameters when the pump speed is increased to 1/2 rated speed. (1.50) b. HOW (INCREASES, DECREASES, NO CHANGE) does the available Net Positive , Suction Head (NPSH) vary by an increase in system flow rate? (0.50) 9 QUESTION 1.11 (2.00) HOW (Increase or Decrease) and WHY does differential rod worth change when: a) moved close to another rod? b) moderator temperature increases? QUESTION 1.12 (1.50) d) WHAT is the FIRST control room Indication that the Point of Adding Heat (POAH) has been reached on EACH of the following types of instrumenta-tion and HOW does it change (Increase, Decrease). (1.05) 1) Nuclear 2) Primary 3) Secondary b) As indicated by the Intermediate Range (IR) neutron flux, the POAH for a startup performed after a reactor trip from 100% power will occur at (a Higher, a Lower, the same) level as a startup performed after a refueling outage? (0.45) l
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I i QUESTION 1.13 (1.75) I
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Using Attachment 2: ) l a) LABEL each of the Regions (I-IV) by name. ' b) IDENTIFY the point (A,B,C, or D) at which DNB occurs.
QUESTION 1.14 (1.75) Using Attachment 3, Identify the power level (0%, 50%, 100%) for EACH of
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the intervals listed (%-E) which would create the given xenon transient Curve.
QUESTION 1.15 (1.50) a) WHAT are TWO (2) reasons for the Doppler Coefficient becoming More Negative over core life? (1.00) b) HOW does Clad Creep affect the Power Doppler Coefficient (More Negative or Less Negative) as core life ages? (0.50)
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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Pago 8 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
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ANSWER 1.01 (2.50)
- no asymetric rods (5 X 0.50]
- control rod group overlap
- control rods are within the proper rod index limits
- Imbalance is within its Technical Specification limits
- Tilt is within its Technical Specification limits REFERENCE B. Reactor Theory 10.9,10.15 TMI OPM Volume 8, Section N-7, page 108 K/A 001000 K5.01 3.3 *
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K/A 001000 K5.04 4.3
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K/A 001000 K5.07 3.3 K/A 001000 K5.52 3.0 K/A 001000 K5.58 2.7 001000K558 001000K552 001000K507 001000K504 001000K501
..(KA's)
ANSWER 1.02 (2.00) a) No Change b) inereese- A4 (_/uq e_ w STdon O A 3 c) Decrease 7 d) Derramen ff/g 4 Y C' J C }
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REFERENCE J , TMI OPM Volume 8, Section N-8, page 7 K/A 059000 K1.04 3.4 K/A 059000 K1.05 3.1 059000K105 059000K104 ..(KA's) ANSWER 1.03 (1.00) pressure initially decreases (0.50] until the OTSGs boil dry then rapidly increases (0.50]
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REFERENCE E. Abnormal Transient Procedures 4.6 TMI OPM Volume 7, Section N-2, Pages 142-146 ' K/A 000074 EA1.12 4.1 000074A112 ..(KA's) ANSWER 1.04 (1.75) / a) T(sat) @ 1000 psig = 545 F. (0.25] T(steam) = 500 45 - 543 F. (0.25] () // $ e // 3 5 bW CONc] b) T(sat) @ 1500 psig = 596 F. (0.25] subcooling = 596 - 560 = 36 F. (0.25] c) using the Mollier Chart: * plot the point for saturated steam at 1400 psig [0.25] draw a line of constant enthalpy to the 20 psia pressure curve (0.25] at the point of intersection, T(exit) = 260 F. +/- 10 F. (0.25] REFERENCE C. Heat Transfer and Fluid Flow 2.5,2.6,2.12 TMI OPM Volume 7, Section N-2, pages 84-96 K/A 193003 K1.25 3.3 K/A 193004 Kl.15 2.8 193004K115 193003K125 ..(KA's) ANSWER 1.05 (2.00)
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Xe production: directly from fission (0.35] decay from Iodine (0.35] Xe removal: decay to Cesium (0.35] neutron absorption (burnout) (0.35] Sm production: decay from Promethium (0.30] Sm removal: neutron absorption (burnout) (0.30] REFERENCE B. Reactor Theory 11.4,11.13 TMI OPM Volume 8, Section N-7, pages 79,80,94 K/A 192006 K1.03 2.7 K/A 192006 K1.04 2.8 K/A 192006 K1.16 1.8
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19 00 K 92b06K116 192006K104 192006K103 ..(KA's)
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ANSWER 1.06 (1.75) assume -10 pcm/ ppm [0.25] CR1 1 - K1
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1 - K1 = 5000 pcm [0.50] CR2 1 - K2 CR1
--- = 5000/1 - K2 1 - K2 = 2500 pcm [0.50]
CR2 (2500 pcm)/(10 pcm/ ppm,)_ = 250 ppm boron to be diluted [0.50] REFERENCE B. Reactor Theory 5.5 TMI OPM Volume 8, Section N-6, page 64 K/A 192008 Kl.03 3.9 192008K103 ..(KA's) ANSWER 1.07 (1.50) during plant cooldown !0.50h d'VC 00- /!d# "[ CO' 5 ] temperature OR cooldown rate [0.50}= pressure (0.50) 0.C. // e g v [y Ache C'e, g g' REFERENCE
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C. Heat Transfer and Fluid Flow 7.6 ' TMI op 1102-11 K/A 193010 K1.07 3.8 193010K107 ..(KA's) ANSWER 1.08 (1.00) decreasing water level will uncover fuel (voiding) causing a decrease in . neutron moderation (0.50) a11'owing more neutrons to reach the upper ! detectors (0.50] l
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_. _ 11 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Pags 11
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REFERENCE B. Reactor Theory 8.14 K/A 015020 A2.02 3.3 015020A202 ..(KA's) ANSWER 1.09 (1.00)
-1/3 DPM [0.50)
the longest lived neutron precursor (bromine-87) [0.50) REFERENCE
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B. Reactor Theory 3.2,'6.8 TMI OPM Volume 8, Section N-7, pages 12,13 K/A 192003 K1.03 2.3 K/A 192003 K1.06 3.2 192003K106 192003K103 ..(KA's) ANSWER 1.10 (2.00)
(N1/N2)^1 = Q1/Q2 (0.25) (0.25/0.50)^1 = 75.0/Q2 Q2 = 150 gpm (0.25) (N1/N2)^3 = P1/P2 [0.25) (0.25/0.50)^3 = 50/P2 P2 = 400 KW [0.25) (N1/N2)'2 = H1/H2 [0.25] (0.25/0.50)^2 = 300/H2 H2 = 1200 psid [0.25)
b. DECREASE [0.50) REFERENCE , D. Component Fundamentals 8.3 TMI OPM Volume 7, Section N-3, pages 89,100,101 K/A 191004 K1.05 2.3 K/A 191004 K1.15 2.6 191004K115 191004K105 ..(KA's) l
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ANSWER 1.11 (2.00) a) Decreases (0.50) due to the competition from the other rod (0.50] b) Increases (0.50] because the rod sees more flux (diffusion length increase) (0.50) .
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_1 . . PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,
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REFERENCE B. Reactor Theory 10.4 TMI OPM Volume 8, Section N-7, page 43 K/A 192005 Kl.07 2.5 192005K107 ..(KA's) ANSWER 1.12 (1.50) a) 1) SUR [0.20] decreases [0.15] 2) PZR [0.20] level increases (0.15] 3) OTSG pressure (0.20] increases [0.15] b) Lewes.[0.45] . H(GH6P-. ' REFERENCE B. Reactor Theory 8.12 TMI OPM Volume 8, Section N-7, page 102-104 K/A 192008 K1.13 3.4 192008K113 ..(KA's) ANSWER 1.13 (1.75) a) I - Single-Phase Convection II - Nucleate Boiling III - Partial Film Boiling [4 X 0.35] IV - Film Boiling b) B [0.35] . REFERENCE C. Heat Transfer and Fluid Flow 5.10 TMI OPM Volume 7, Section N-2, Page 17S K/A 192006 Kl.11 3.1 K/A 192006 K1.12 3.1 192006K112 192006K111 ..(KA's) ANSWER 1.14 (1.75) A. 50% B. 100% C. 0% [5 X 0.35] D. 100% E. 50%
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1, PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,
* PagD 13 . THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW .
REFERENCE B. Reactor Theory 11.7 TMI OPM Volume 8, Section N-7, pages 83-90 K/A 193008 K1.03 2.8 K/A 193008 K1.06 2.8 K/A 193008 K1.07 2.6 K/A 193008 K1.08 2.6 193008K108 193008K107 193008K106 193008K103 ..(KA's) t ANSWER 1.15 (1.50) a) Pu-240 [0.50] and fission product inventory increases [0.501 b) Less Negative (0.50] * REFERENCE B. Reactor Theory 8.5,8.7,8.11 TMI OPM Volume 8, Section N-7, pages 73,120 K/A 192004 K1.02 3.0 K/A 192004 K1.07 2.9 192004K107 192004K102 ..(KA's)
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TEST CROSS REFERENCE Page 1 QUESTION VALUE REFERENCE
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1.01 2.50 ZZZ0000001 1.02 2.00 ZZZ0000002 1.03 1.00 ZZZ0000003 - 1.04 1.75 ZZZ0000004 1.05 2.00 ZZZ0000005 1.06 1.75 ZZZ0000006 1.07 1.50 ZZZ0000007 1.08 1.00 ZZZ0000008 1.09 1.00 ZZZ0000009 1.10 2.00 ZZZ0000010 1.11 2.00 ZZZ0000011 1.12 1.50 ZZZ0000012 1.13 1.75 ZZZ0000013 1.14 1.75 ZZZ0000014 1.15 1.50 ZZZ0000015 ______
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.- . ATTACHMENT 1
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Cycle' efficien'cy = (Net work ." f = ma v = s/t - out)/(Energy in)
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w = og s = V,t + 1/2 at
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-At
a = (Vf - V,)/t A = in A=Aeg KE = 1/2 m PE = agn V f = V, + a t w = e/t i=tn2/h/2=0.693/t1/2 ' 2 t W W = v AP-A= t04 1/2*" * @((t 1/2) * l*bIl ti = 931 am - m.= Y,yAo t,I,e
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h = [nCpat
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6=UAaT I = [ ,e # Pwr = W 7ah I = I,10' O TVt. = 1.3/u ; sur(t) HVI. = -0.693/u P = P,10 p=pe/T o t SUR = 26.06/T SCR = S/(1 - K,ff) CR x = S/(1 - K ,ffx) SUR = 26p/t= + (s - p)T CR j (1 - K,ffj) = G 2(I ~ beff2} l T = ( t*/o ) + (( a - o y Io ] N = 1/(1 - K,ff) = CR j/CR, , T = v(o - a) x = (1 - K ,ff,)/(1 - K ,ffj) T = (8 - o)/(Io) SDM = ( - K,ff)/K,ff a = (K ,ff-1)/K ,ff = AK,ff/K,ff 1* = 10 seconds I = 0.1 seconds-I
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o = ((1*/(T Keff)3 + (I,ff/ (I + IT)] l
=1d I jd) 2 ,2 gd
P = (tev)/(3 x 1010) I jd) 22 2 I = cN R/hr = (0.5 CE)/d (meters) R/hr = 6 CE/d2 (f,,g) Water Parameters s Miscellaneous Conversions 1 gal. = 8.345 lem. 1 curie = 3.7 x 1010eps 1 gal. = 3.78 liters 1kg=2.21lbm 1 ft3 7.48 gal. I hp = 2.54 x 10 Btu /hr * Oensity = 62.4 lbm/f t3 1 m = 3.41 x 10 6tu/hr Oensity = 1 ga/c..r 3 lin = 2.54 cm HCat of va00r1 ration = 970 Stu/ lcm *F = 9/5'C + 32
.9e st of fusion = 14 3:u/1tra *C = 5/9 (*F-32)
1 A t:a = 14. 7 p s i = 29.9 i n . ric .
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1 BTU = 778 ft-lbf I ft. H O = 0.4335 Itf/in.
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.- .. i%Iwed 2 } l e GPU Nuclear Corporation l g g7 Post Office Box 480 Route 441 South Middletown, Pennsylvania 17057 0191 717 944 7621 TELEX 84 2386 Writer's Direct Dial Number: June 24, 1988 l
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C311 -88-2083 Mr. Robert Gallo Operations Branch Chief Division of Reactor Safety US Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406
Dear Mr. Gallo:
Three Mile Island Nuclear Station, Unit 1 (TMI-1) Operating License No. DPR-50 Docket No. 50-289 Comments on NRC Reactor Operator Examination Administered on June 22, 1988 - Randy E. Kennedy In accordance with NUREG-1021, ES-201, paragraph H.1, enclosed are THI-1 comments for the NRC Reactor Operator (RO) Examination administered by the NRC at Region 1, King of Prussia, for Randy E. Kennedy on June 22, 1988.
Sincerely, i
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. D. kill Vice President & Director, THI-1 HDH/DVH/spb:1197A g cc: V.--Pttssell h)
R. Conte Document Cor, trol Oesk Enclosure i l i l l GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation
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QUESTION 1.01
Add the following as possible answers: Control Rods Axial Power Shaping Rods (APSR's) Burnable Poisons
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REFERENCE OPM Volume 8 Section N-7 pages 16, 19, 59
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SECTION N-7 REVISIVH 0 d. Flux Distribution Control of the flux distribution is an important aspect to power reactor operations. Control rods can be used to assist in this spatial canagement of reactor power.
While soluble boren provides uniform reactivity control throughout the core, control rods can be used to control the reactivity in selected regions of the reactor.
Essentially, control rods affect core reactivity by means of influencing two factors in the six factor formula. The major effect is their influence on the thermal utilization factor, f. Secondly, control rods have the ability to alter the thermal neutron flux distribution in the core. This results in a change in the thermal non-leakage probability, Ltha 2. Poison and Construction The poison material selected for use in all the control rods is an alloy of silver-indium-cadmium, in a 80 percent - 15 percent - 5 percent mixture, respectively. This alloy has several features that make it an excellent selection as a control poison, a. The melting point of the alloy is substantially higher than its constituent elements.
b. It is a good absorber of both thermal and epithermal neutrons. Figure 2 and Table 1 give values of the absorption cross section for the three elements as compared to some other commonly used control poisons, c. Ag-In-Cd alloy does not yield a gaseous product under irradiation. Consequently, internal pressure and swelling of the absorber material will be limited and will not cause excessive stretch or stress on the control rod clad.
For a complete discussion of the control rod and control rod assembly construction, refer to Section B-3, "Rx Vessel and Internals," within Volume I of this nanual.
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SECTION N-7 REVISION 0 3. Types of Control Rods There are 69 control rod assemblies contained in the cycle 5 core. These assemblies are organized into 8 groups. Groups 1
- 4 are safety rod assemblies, groups 5, 6, and 7 are regulating rod assemblies, and group 8 is an axial power shaping rod group. The function of each type of assembly f ollows:
a. Regulating or Control Rods. Regulating rods provide fine control to reach a desired power level ur to compensate for small changes in reactivity due to temperature, pressure, poisons, etc. These rods are usually designed to move rapidly over a limited range and they have a smaller reactivity worth. These rods should not be able to produce large changes in reactivity but they should have trip capability.
b. Safety Rods. The function of the safety or shutdown rods is to shutdown the reactor rapidly (trip), if needed.
c. Axial Power Shaping Rods (APSR). The function of the axial power shaping rods is to offset any axial flux imbalance caused by control rod position, Xenon distribution, or other factors that cause the flux to be greater in one half of the reactor core than in the other. The poison section of the rod is contained in the lower 25 percent (3 feet) of the total rod. By properly placing these rods, the operator can cause the flux to be redistributed to control power imbalance, axial flux profile, and to insure even fuel burnup over core life.
Figure 3 provides a core map and shows the location, function, and group designation of each of the 69 control rod assemolies.
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SECTION N-7 REVISION O D. Burnable Poisons 1. Functions While cycle 5 will not utilize this type of reactivity control, burnable poison may be used in later core loads for three basic purposes: a. Extend core life Burnable poisons themselves do not extend core life but because they insert negative reactivity, the reactor designer can load more fuel into the core resulting in higher allowable core excess reactivities. By careful design, the nuclear engineer can arrange and size the amount of burnable poison and fuel such that the increase in core reactivity due to poison depletion nearly matches the decrease in core reactivity due to fuel depletion.
b. Flatten core flux distribution By specifying the proper location of the burnable poison within the core, the nuclear engineer can also flatten the flux for a more even fuel burnout and power distribution.
c. Maintain the moderator temperatur.e coefficient within prescribed safe limits The presence of burnable poison will also allow a lower soluble boron concentration at the beginning of cycle life. This will result in a more negative (or less positive) moderator temperature coefficient, as discussed in a later section.
2. Type and Construction When a burnable poison is utilized at TN!-1, it will be in the form of B4 C rods contained within full length poison rod assemblies which are inserted into selected fuel assemblies.
For a complete discussion of poison rod asser21y construction, see OPM Volume 1, "Reactor Vessel and Construction."
3. Use of Burnable Poisons - Disadvantages l Because some residual poison will exist at the end of the core cycle, the fuel loading for criticality will be increased for reactors utilizing lumped burnable poisons. This means that more fuel will remain in the core at the end of the fuel cycle, even though use of burnable poisons has provided a longer cycle length,
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QUESTION 1.02 Answers should be as follows: a) no change b) no change c) decrease d) no change The ICS would be in fully automatic operation in order to not be in track. The feedwater demand will be modified (decreased) by the drop in feedwater temperature and no other changes should take place.
REFERENCE OPM Volume 3 Section F-3 page 93 Lesson Plan 11.2.01.055 Integrated Control System pages 41-43 attached. Entire lesson plan also included in package.
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SECTION F-3 REVISION O Section 0 FEEDWATER SLESYSTEM Int ztcuction In the two previous sections we have seen how the Unit load demand subsystem develops a demanded lead signal and how the Integrated Master uses that signal to coordinate overall plant control and develop turbine and turbine bypass valve control signals. This section will describe what the feeowater subsystem does with the mocified signal coming frcrn the Integrated Master stcsys tem.
Subsystem Description (refer to GPU prints) Af ter the Integrated Master subsystem modifies the oemanced load signal it is sent to the input of the Steam generator /Rx Master control station IC 9.15.
If the control staticn is in automatic the signal is then passed on to IC 8.16 and IC 9.16. IC 8.16 is a limiter which will prevent a signal being passed to the Reactor Control subsystem if the demanded load is less than 15%. IC 9.16 is a function generator which ccnverts the megawatt demand sional to a feedwater cerand signal. It is this signal and its manipula: that we are concerned with and will discuss in this section.
The outpJt of IC 9.16 is sent to the jnouts of FW 1.12 and FW 3.10. The e fmetion @nerator FW l.12 is used to cevelop and expected feedwater temperature for the demanded loaa. This signal is compared to actual feecwater temperature via FW 3.12 and any difference is sent to FW 3.10 via FW 3.11 (wnich is a function generator which calculates amomt of signal correction needed)to correct the feedwater demand input to FW 3.10 for temperature. The purpose for this mocification has been discussed in the Basic Overview section.
The output of FW 3.10 is sent to surmer anplifier FW 4.10. Here the cemanced feedwater signal can be modified by Tave and/or neutron error. The purpose o.f these modifications have been described in the Basic Overview section. The , signal developcent for Tave and neutron error will be described in tre Reactor l control subsystem. Our discussion of Tave error in this section will be limited to the fact that it is sent to FW 4.10 for modification of the j feedwater demand signal, however this section will discuss how the neutron ! error signal modifies feedwater demand (neutron cross-limits). l l The neutron error signal developed in the Reactor control subsystem is sent to l fmetion generator FW 4.15 where it is converted to a useful signal for feedwater modification. This sipal is then inverted by FW 4.14 and sent on to FW 4.13 and Fw 5.13 where a 5% nestive signal is added to a positive ) output of FW 4.14 by FW 4.13 and a 5% positive signal is added to a negative output by FW 5.13. The output signals of FW 4.13 and FW S.13 are then sent to
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Revision 3 10/20/87 11.2.01.055 INSTRUCTOR NOTES a. Correct ratio of flow between the two steam generators b. BTU availability of the steam generators c. Level conditions in the steam generator 2. FW Demand Calculator
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a. Purpose: To correct FW demand for a given load such that there is balance of BTU exchange between the primary and secondary side of the steam generator.
(1) To prevent disturbing the ratio when feedwater flow demand is corrected by a function of the
"ftccwater temperature error" 0-1 b. Feedwater temperature error - is generated by comparing the expected average feedwater temperature in both loops, as a function of load with the
, actual feedwater temperature.
NOTE: A functions generator calculates the expected average f eedwater temperature for various loads.
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Revision 3 LO/20/81 11.2.01.055 INSTRUCTOR N:'ES c. Example:
(1) With a fixed primary flow and a fixed aT (T - Tc) across the primary side of the SG, a fixed BTU is available.
D-2 (2) Now - if colder fluid is ( introduced into the secondary \ side of the steam generator at a fixed flow rate, more heat would I be required to raise this fluid to the proper SG outlet
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l l conditions or if if the flow ' rate on the secondary side were reduced, the total number of BUT's required to raise this 1 colder fluid to the desired outlet conditions would also be reduced.
THEREFORE: By properly reducing the secondary side flow rate, for a given FW temperature reduction, the proper steam generator outlet
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INSTRUCTOR '40TES conditions can be maintained, aTc 0-3 d.2,3 Feedwater Oemand and %Tc Control (a) Purpose: To divide the
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unit f eedwater demand as necessary into separate demands (one for each SG) based upon the plant conditions.
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(b) Now remember:
i. The reactor receives coolant from each SG.
11. Each SG receives the
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same reactor coolant temperature (Th) 111. The amount of heat i
transferred in the ' steam generator will , determine the ;
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reactor. Since we l
, want to balance the i
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l QUESTION 1.04 a.
Answer should be 590 F. (545 F + 45 F = 590 F) Degrees of superheat should be added to saturation temperature.
REFERENCE OPM Volume 7 Section N-2 page 90 l
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SECTION N-2 REVISICN O
.Fsat = 1097.55 psi sf = 0.7575 BTU /lb m *F seg = 0.6222 BTU /lb m *F sg = 1.3797 BTV/lbm- F Using the expression for quality based on entropy.
x= ~ 'f 5fg s = xsfg + sf
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s = (0.97)(0.6222 BTU /lbm- F) - 0.7575 BTU /lbm *F
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s = 0.604 BTU /lbm- F + 0.7575 BTU /lbm - F s = 1,36 BTU /lbm.'F Thus, the specific entropy of 97 percent quality steam at 556 F is 1.36 BTV/lbm- F.
. Use of Superheated Steam Tables The superheated steam tables give the energy transfer properties et superheated steam as a function of the pressure P and tne temperature T. Tne next table is a portion of a typical superheated stean table. For each value of pressure, the saturation temperature is given as well as the values of specific volume, specific enthalpy and specific entropy at the saturation temperature for both saturated water and saturated steam and at tenperatures above saturation for superheated steam. The number of degrees of superheat, represented by Sh, is given at each temperature and pressure for superheated steam. Not all tables give values for Sh. To calculate this value, the Saturation temperature is subtracted from the temperature of the steam.
Sh = T - Tsat where Sh = number of degrees of superheat (*F) T = temperature of superheated steam being considered (*F) T t = temperature of saturated steam at the given pressure l l l 90.0 0370K
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l QUESTION 1.07 Question seems to ask for more than one operation due to the word operations being plural in the question. The answer should also allow Plant Heatup in the key as an additional operation that causes the Reactor Vessel to experience greater stress than normal.
REFERENCE OPM Volume 1 Section B-3 pages 29-30
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SECTION 3-3 9EvlS:">. C
5.0 TECHNICAL SPECI ICAT10NS 5.1 Pressurization beatup anc Coolco n Limitatiens acclicacility Apolies to pressuriration, neatup anc ceciccan of the reactor ecclant system.
Objective To assure that tencerature anc pressure changes in the reacter ecclant system co not cause cyclic Icacs in excess of cesign for reactor coolant systen conoonents.
Scecification 5.1.1 Fnr ccerations until five effective full pcwer years, t*e reacter ecolant pressure and the system heatup anc ecolccwn rates (with the excepticn of tne cressurirer) shall te limited in accercance ith Figure 24 ano Figure 25 anc are as follcws: Hea tuo /Cecldcwn d A11:wable concinations cf pressure and temperature shall te te One right of and belcw the limit line in Figure 24. Heatup anc coolcown rates shall not exceed those shown en Figure 24 Irservice Leak and Hycrestatic Testino Allewacle conninations of pressure and ten;erature shall be to tre ! j
rignt of and belcw the limit line in Figure 25. Heatup and coolcewn rates shall not exceed those snown on Figure 3.1-2.
5.1.2 Prior to exceecing five. effective full power years of operation, Figures 24 and 25 shall be upcated. The highest precicted, adjusteo reference temperature of all the beltline materials shall be used to cetermine the ' adjusted reference temperature at the end of the service period.
Bases All reactor coolant system conconents are cesigned to withstand tne l effects 9f cyclic loads due to system tercerature and pressure changes.\l) These cyclic loacs are introduced by unit loao transients, reacter trips and unit heatup and coolcown coeratiens. i The nuncer of thermal and leading cycles used fer cesign purocses are sh0=n in Table 4-8 of tne FSAR. The maximum unit heatup and , , - 29 - 05414
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SECTICN S-3 FCvl517. O coolcown rate of lO0cF in any gne hour (1.6cF/ min) satisfies stress limits for cyclic operation.(2) The 200 psig pressure lirit fcr the seconcary sice of the steam generator at a tencerature less than 100*F satisfies stress levels for tenceratures telow the DTT.(3) The heatuo anc cocleown rate limits in this specification are nct intenced to limit instantaneous rates of tercerature enange, but are intenced to limit tencerature changes sucn that there exists oc one hour interval in which a temperature change greater than t e linit takes place.
As a result of fast neutron irradiation in tne beltline regicn cf the core, there will be an increase in the RT tot with acc A latec nuclear cceratiens. The acjustec reference tecceratures nave ceen calculated by accing the precictec raciation-incucec RT FOT anc tne unirraciatec RT TOT for each of the rea,: tor coolant celtlina naterials.
The crecietgJ LT NDT was calculatec using the respective neutron fluence after Five effective full pc.er years of operation. The analysis cf tre reactor vessel material containec in the first surveillance capsule removec from the Three Mile Islano Nuclear Station Un*,t 1 confirmed tnat the current techniques usec for crecictinc t;.4 change in inpact properties due to irraciation are conservattve.
Analysis of the activation cetectors containec in the first surveillance capsule bncicates that the average f ast flux curin'; Cycle 1 was 1.45 x 10"O n/cm?-sec maximum at the pressure vassel wall. Extrapolation of the Cycle 1 flux basec on precicte: fuel reloac anc burn-up concitions incicates that the maximun average fast neutron (E 1 MeV) flux curing six full power years of operation will be 1.68 x 1010 n/m2.sec at the reacter vessel wall anc 9.33 x 109 n/cn2-sec at tM 1/4 T location. The fast neutron exposure curing five ef fecti'e full power years of d there is 1.5 x 10 coeration) 3.7 x 101 n/cm2 at the 3/4 T location.18 n/cm2 at tre 1/ Based on the precictec RT POT af ter five effective full ocwer years l of operation, the pressure-temperature limits of Figure 24 anc 25 i have been established in accorcance with the requirenents of 10 CFR 50, Appencix G. The protection against noncuctile failure is j assuned by maintaining the coolant pressure celow the upper linit
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of these pressure tercerature limit curves. I Tre pressure linit lines on Figures 24 anc 25 have been establisme: consicering the following: l
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QUESTION 1.12 b.
Answer should be HIGHER.
The decay heat level is higher after a trip than aftnr a refueling outage, so the POAH is higher after the reactor , trip.
REFERENCE , OPM Volume 8 Section N-7 pages 100-103
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SECTION N-7 REVISION O
!. Transient Reactivity Effects 1. Power Maneuvers Below the Point of Adding Heat If the reactor is critical below the point of adding heat, a positive change in reactivity will establish a constant start up rate after a prompt jump as described by the Inhour Equation. If no further changes in reactivity are made, this start up rate will continue until the point of adding heat is reached (e.g., the Power Range) likewise a negative reactivity additi'>n to a critical reactor will establish a constant negative start up rate until we decrease power low enough to level out on subcritical multiplication.
2. The Point of Adding Heat a. Defini tion: The point in fission power above which a tission rate increase begins to raise total core sensible heat production, fuel temperature, and moderator temperature.
The point of adding heat will be dependent on decay heat l evel . That is for a given set of conditions the point of initial sensible heat will occur at a unique position j on the inte In any case this will occur between10gmediategange.
and 10- amps on the intermedate range.
, Figure 43a shows a typical plant startup after critical data is taken at 10-0 amps the rods are pulled enough to create a stable start up rate (e.g., 0.5 OPM). Power will continue to increase until it reaches the point of adding heat where the coefficients for fuel temperature (20) and Moderator Temp. (sMT) slow down the startup rate and cause power to level off.
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TpA g --o- the StMt1 c) C4 5 a 2, 2C Oruy Ht A r Figure 43 Point of Adding Heat 101.0 0323K
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SECTION N-7 REVISION O b. Contributing Factors
"Background" heat in the core can over-shadow the fission heat being generated at lower power levels. This does not mean that we do not produce heat from fission while at the lower power levels - only that at the lower power levels the heat production is not significant enough to cause a fuel temperature increase that can be measured or fel t. Remer2er - reactor power level is measured over many decades. The exact point of beginning sensible heat production is variable since fission heat is added to the background heat production - decay heat.
First let us consider the effect of RC pumps running in the RCS with no core decay heat. Each running pump adds 4 MW of heat into the RCS fluid. So if four pumps are running, this is 16 MW thermal or 16/2535 = 0.63 percent power. Figure 438 Case 1 shows this power as a horizontal line. Notice that the left axis is logari thmic . If fission heat production is increased at a constant start up rate, then total heat production would deviate from the RCP heat line as shown. The point where this line increases is the point of initial sensible fission heat production or the point of ad#ing heat. In this example, this corresnonds to 3 x 10 o , amps on the intermediate range.
' Now let's consider case 2 where a certain amount of reactor core decay heat is present. The amount of decay heat the core produces is determined by power history (e.g., time since last power operation). For a graph of percent decay heat vs. time see Figure 44.
In case 2 (Figure 43.c) a decay heat production of 2 percent or approx. 50 MW thermal is assumed. The total heat production befo;*e the addition of fission heat is then 50 + 16 or 66 MW thermal or 2.6 percent of total power.
In this case, it can be seen tha sensible heat is approx. 5 x 10 pamps the point or 0.5 of initial percent . fission power. Fission heat production must be raised i high enough (higher than case 1) to become a significant ' addition to the existing heat sources to raise fuel temperature (become sensible). The major variable in this phenomenum is core decay heat production. Hi gher rates of decay heat production will raise the reactor power where sensiblg heat production begins. At TM1 i t varies from 5 x 10-' Amps to 10 0 Amps on the Intermeoiate Range Instruments. This does not mean that 102.0 0323K
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SECTION N.7 REY!SION O 5 x 10 7 Amps is ever 1 percent full power. A misconception held by some people is that when the point of sensible heat is reached the reactor is at one percent power. It only means that fission heat production - when added to decay heat production - is significant enough to change fuel temperature, which broadens resenance capture i peaks and produces a negative reactivity feedback. When th? fuel temperature is increased, the heat is transferred to the coolant - causing a temperature increase and density decrease. This moderator temperature increase also produces a reactivity change - thus, the moderator temperature coefficient follows the Doppler coefficient, helping to reduce X effective back to unity.
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Tir.e After Thin (Seconds) Figure 44 Decay Heat Generation Following Reactor Trip From 100 Percent Power 103.0 0323K
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l v i ATTACHMENT 3 NRC RESPONSE TO FACILITY COMMENTS
Question No.
, 1.01 The answers recommended by the facility for inclusion into'the answer key do not, by themselves, provide indications'for an '
operator which would allow him to insure that core peaking values were not being exceeded. No changes were made to the [ answer key, i 1.02 The information provided by the facility's comment identified an alternate set of acceptable answers to the question ir one ; assumed that automatic feedwater. control was in place.
~ The answer key was modified to include these alternate answers if this assumptiun was stated.
1.04a The facility's comment was incorporated into the answer key.
1.07 The facility's comment was reviewed and it was determined that - the question does appear to address more than one operational condition. The answer key was modified to include both plant , heatup and cooldown. In the future, the question, when used, will be reworded such that only the one (1) condition which creates the greatest reactor vessel stresses will be asked for.
1.12 The facility's comment is correct and the answer key was l ' changed.
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