IR 05000354/1988012
| ML20207C113 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 07/21/1988 |
| From: | Howe A, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20207C107 | List: |
| References | |
| 50-354-88-12OL, NUDOCS 8808050044 | |
| Download: ML20207C113 (140) | |
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i U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 88-12(0L)
FACILITY DOCKET NO. 50-354 FACILITY LICENSE NO. NPF-57 LICENSEE: Public Service Electric and Gas Company Post Office Box 236 Hancocks Bridge, New Jersey 08038 FACILITY: Hope Creek Generating Station EXAMINATION DATES: April 11 to April 15, 1988 CHIEF EXAMINER: [[/M /1 Allen G. Howe Senior Operations Engineer
/kv 7-4 o-48 Date APPROVED BY: pf ad 7- 2 / - 8'(
fv David J. Linge,Mhief, BWR Section, Date Operations Branch, Division of Reactor Safety SUMMARY: Written examinations and operating tests were administered to two (2) senior reactor operator (SRO) candidates and four (4) reactor operator (RO) candidates. Two (2) SRO candidates and three (3) R0 candidates passed these examinations. One R0 candidate failed the-written examination and the operating test.
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DETAILS TYPE OF EXAMINATIONS: Replacement EXAMINATION RESULTS:
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R0 SRO
~ Pass / Fail [ Pass / Fail [
[ Written [ 3/1 [ 2/0 [
[ Operating [ 3/1 [ 2/0 [
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1. CHIEF EXAMINER AT SITE: Allen G. Howe, Senior Operations Engineer l
2. OTHER EXAMINERS: D.Lange, Chief, BWR Section M. Evans, Operations Engineer l C. Gratton, LOLB, NRR i R. Eaton, LOLB, NRR l D. Moon, PNL !
C. Moore, PNL A. Thadanni, USNRC, NRR !
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3. The following is a summary of generic strengths and deficiencies noted on operating tests. This information is being provided to aid the licensee in upgrading license and requalification training programs. No licensee l'
response is required.
STRENGTHS a. Execution of Emergency Operating Procedures (EOP's)
b. Ability to locate plant equipment c. Knowledge of Technical Specifications d. General systems knowledge J
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DEFICIENCIES I a. Knowledge of where to escort visitors during.an emergency I b. No clear division of operator responsibilities during transients - at times, the operators would manipulate controls at the same panels, duplicating efforts ,
i c. The SR0 candidates sometimes tended to "micro-manage" a situation l rather than keeping plant overview in perspective' j 4. The following is a summary of generic strengths or deficiencies noted from the grading of the written examinations. This information is being provided to aid the licensee in upgrading license and requalification training programs. No licensee response is required.
SRO STRENGTHS a. Knowledge of core cooling mechanisms, xenon, cooldown rates, thermal limits, and subcritical multiplication. Ability to perform period calculations. (Questions 5.01, 5.03, 5.04, 5.06, 5.08, and 5.10)
b. Knowledge of control rod block signals, APRM/RPS logic, and backup scram design. (Questions 6.01, 6.02, and 6.08)
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c. Knowledge of E0P entry conditions, refueling procedures, and actions for emergency load reduction. Ability to use E0P curves and to classify events per the Emergency Plan. (Questions 7.01, 7.04, 7.06, 7.09, 7.10)
d. Knowledge of administrative requirements for the work control group. Ability to use Technical Specifications. (Questions 8.03, 4 8.09, 8.10)
SRO DEFICIENCIES I a. Ability to predict initial power response when a SRV opens and the initial and long term effects of HPCI injection on total steam flow and reactor vessel pressure. (Questions 5.05 and 5.11)
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b. Knowledge of FRVS initiation signals and the LOCA and LOP !
sequencers. (Questions 6.05 and 6.06) '
c. Ability to define a high radiation area. Knowledge of quarterly administrative exposure limits and the requirements to extent these limits. Knowledge of actions for a loss of feedwater heating.
(Questions 7.05 and 7.07)
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R0 STRENGTHS a. Ability to perform reactor period calculations. (Question 1.04)
b. Knowledge of RCIC response to reactor vessel level transients.
Understanding of feedwater control system, ADS logic, Rod Block Monitor, and RRCS. (Questions 3.01, 3.05, 3.06,.3.07 and 3.10)
c. Knowledge of actions when thermal power limits are exceeded.
(Question 4.02)
R0 DEFICIENCIES a. Knowledge of why shutdown margin decreases in the early part of the fuel cycle and why vessel heatup and cooldown rates are limited. Ability to predict the effects of operator action on available NPSH for RWCU pumps. (Questions 1.02, 1,07, and 1.09) i b. Knowledge of adverse effects of ECCS initiation when jockey pumps are not in service, LOP and LOCA load sequencers, signals which shift control area ventilation to outside isolate mode, design features protecting the reactor building from excessive delta pressure, reasons for HPCI minimum speed requirements,'and the ef fects of various APRM flow converter failures. (Questions 2.02, 2.04, 2.07, 2.08, 2.11, and 3.02) ,
c. Knowledge of the definition of a high radiation area, the requirement to place the mode switch to shutdown if a SRV sticks open, and methods of valve position verification. (Questions 4.05, 4.06, and 4.07) l l
5. Personnel Present at Exit Interview: <
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NRC Personnel '
J A. Howe, Chief Examiner C. Gratton, LOLB, NRR G. Meyer, Senior Resident Inspector l
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Facility Personnel
! J. Hagan, Maintenance Manager (Acting General Manager)
H. Hanson, Manager Nuclear Training G. Mecci, PTS Hope Creek Simulator L. Catalfomo, Assistant Manager Operations Training W. Gott, PTH Hope Creek Operations Training R. Hovey, Operating Engineer l
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6. Summary of NRC comments made at exit interview:
The chief examiner thanked the training and operations staffs for their cooperation during the examination.
The examiners felt site access was smooth and that housekeeping was adequate.
Two questions about the the station tagging procedures were discussed. One was about control room bezel covers that were not marked as required by the tagging procedures. The other referred to requirements about a paper caution tag that was hung on a control room panel since it was immediately removed from the panel when questioned by an examiner.
The written examination review was discussed. The facility staff was advised where to send the formal comments.
The generic strengths and weaknesses noted on the operating examinations were discussed.
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The simulator fidelity was reviewed. This is detailed in Attachment 5 '
of this report.
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Attachments:
1. Written Examination and Answer Key (RO)
2. Written Examination and Answer Key (SRO)
3. Facility Comments on Written Examinations after Facility Review 4. NRC Response to Facility Comments 5. Simulation Facility Fidelity Report j
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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: HOPE CREEK REACTOR TYPE: BWR-GE4 DATE ADMINISTERED: 88/04/12 EXAMINER: NRC REGION I CANDIDATE: ANSWER KEY Mf]
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 CATEGORY 25.00 25.00 ,
1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25.00 25.00 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25.00 25.00 3. INSTRUMENTS AND CONTROLS 25.00 25.00 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL l 100.00 % 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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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2 THERM 0 DYNAMICS, HEAT TRANSFER AND FLUID FLOW
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l QUESTION 1.01 (3.00)
l Concerning delayed neutrons:
a. WHAT is the source of delayed neutrons? (1.0)
b. STATE HOW reactor period changes (INCREASES / DECREASES / l REMAINS THE SAME) for a given reactivity insertion if I the EFFECTIVE DELAYED NEUTRON fraction is REDUCED. (1.0)
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l STATE HOW EFFECTIVE DELAYED NEUTRON fraction changes INCREASES / DECREASES / REMAINS THE SAME) as the core ages. (1.0)
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ANSWER 1.01 (3.00) j
a. (Delayed neutrons are) produced from the (beta) decay !
[+0.5] of fission products [+0.5].
J (Alternate Answer: produced from the decay of delayed neutron precursors [+0.5] created from the fission process
[+0.5])
b. decreases [+1.0]
c. decreases [+1.0]
REFERENCE 1. Hope Creek: LP RXPH23 LO #3, and #7.
KA 292001K102 292003K104 QUESTION 1.02 (3.00) l Concerning Shutdown Margin:
a. DEFINE shutdown margin. INCLUDE all applicable conditions. (2.0)
b. EXPLAIN WHY shutdown margin initially decreases during the early portion of the fuel cycle once fission product poisons have achieved equilibrium. (1.0)
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. l PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, l
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l ANSWER 1.02 (3.00)
(Shutdown margin is) the amount of reactivity by which the !
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reactor is suberitical or would be suberitical from its l present condition [+0.5] assuming all control rods fully l inserted except for the single control rod of highest worth I
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which is assumed to be fully withdrawn [+0.5] and the reactor is in the cold shutdown condition, 68 degrees F [+0.5], and xenon free [+0.5].
b. gadolinium (a burnable poison loaded into the core) depletes (due to burnup) [+1.0)
REFERENCE )
i 1. Hope Creek: LP RXPH19 LO #4, 6b, and #7. l KA 292002K110 292007K101 l l
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QUESTION 1.03 (3.00) l
The reactor is critical with the MINIMUM permissible stable period as allowed by procedure OP-IO.ZZ-003(Q), "Startup from Cold Standby to Rated Power." Reactor recirc loop suction temperature is 140 degrees F.
a. WHAT is doubling time if period remains canstant? (1.0)
b. If IRMs currently indicate 50 on range 2, HOW LONG will it take for power to reach the point of adding heat if period remains constant and heating power is given to be 40 on range 8 of the IRMs? (SHOW your work.) (1.5)
c. STATE HOW period will be affected (INCREASE / DECREASE /
REMAIN THE SAME) after the point of adding heat has been reached. (0.5)
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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWER 1.03 (3.00)
a. From OP-10.ZZ-003(Q), period equals 60 seconds [+0.5]. Thus doubling time equals 60/1.44 = 41.7 seconds [+0.5].
b. 50 on range 2 is equal to 0.05 on range 8 [+0.5]
P(0) = 0.05; P(t) = 40 P(t) w P(0)e**(t/ period) [+0.25]
40 = 0.05 e**(t/60sec)
t = 60 in (40/0.5) = 401 see or 6 min 41 sec [+0.75]
c. (Period will) increase [+0.5]
REFERENCE
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1. Hope Creek: LP RHPH24 LO #1a and fic.
KA 292003K108 292003K109 292008K113 l
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1. PRINCIPLES OF NUCLE /.R POWER PLANT OPERATION2 PAGE 5 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW l
l QUESTION 1.04 (2.00)
a. Concerning control rod worth during a reactor startup with 100% peak Xenon versus a startup with Xenon free conditions, WHICH statement below is correct? (1.0)
(1) Peripheral control rod worth will be lower during the 100% peak Xenon startup than during the Xenon free !
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startup.
(2) Central control rod worth will be higher during the 100% peak Xenon startup than during the Xenon free 1 startup.
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(3) Peripheral control rod worth will be higher during the 1 100% peak Xenon startup than during the Xenon free startup.
(4) Both central and peripheral control rod worth will be ;
the same regardless of core Xenon concentration.
b. ANSWER the following questions as TRUE or FALSE, given that the unit is at rated conditions and a Reactor Scram o
REGION I
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i CANDIDATE: _______ ________________
INSl@UCllgN@_IQ_C9Npig9IE1 Uze separate paper for the answers. Wri te answers on one side onl y.
Stcole Question sheet on too of the answer sheets. Points for each ouestion are indicated in parenthesec after the ouestion. The passing crede 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
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_ _ _ _ S_ C_ O_ R_ E_ ___V__ A_ L_ U_ E_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ C_ A_ T_ E_ G_ O_ R_ Y_ _ _ _ _ _ _ _ _ _ _ _ _ -
_?Qt99__ _29399 ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND THERMODYNAMICS -
_?Dt99__ _25t99 ___________ ________ 6. PLANT SYSTEMS DESIGN. CONTROL.
AND INSTRUMENTATION
.29199__ .39199 _ _ _ _ _ _ _ _ _ _ _ ___...._n_ 7. PROCEDURES - NORMAL. ABNORMAL.
EMERGENCY AND RADIOLOGICAL CONTROL M3 c 5 . 0_ 0_ _ _ _ 2_ 5 __. 0_ 0_ 8. ADMINISTRATIVE PROCEDURES.
g., ___________ ________
g 199 99__ ___________ ________% Totals l
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Final Grade All work done on thi s examination is mv own. I have neither given nor received aid.
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Candidate's Signature l
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS D,uring the admini strati on of this ex ami na t i on the following rules apolv:
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1. Cheating on the ex aminati on means an automatic denial of your appl i c a t i on ond cou.d result in more severe cenalties.
2. Roetroom trios are to be l i mi ted and only one candidate at a time may loave. You.must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
3. Une black ink or dark cencil gnly to facilitate legible reproductions.
4. Print vour name in the blank provided on the cover sheet of the ex cmi n a t i on .
5. Fill in the date on the cover sheet of the examination (if necessary).
6. Use only the paper orovided for answers.
7 Print vour name in the upper right-hand corner of the first Dage of each section of the answer sheet.
8. Consecutivelv number each answer sheet, write "End of Category __ " as cppropriate. start each category on a ngw page, write gnly gn gne side of the paper, and write "Last Pace" 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. l 11. Separate answer sheets from oad and place finished answer sheets face down on your desk or table.
12. Use abbr evi at i ons onlv if they are commonly used in facility literature.
13. The point value for each ouestion is indicated in carentheses after the ouastion and can be used as a guide for the depth of answer recuired.
14. Show all calculations. methods. or assumotions used to obtain an answer to mathematical problems whether indicated in the ouestion 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 ouestions of tho examiner only.
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 boon completed.
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1 8 *. When you complete your Ox omi nat i on , you Ghallt ,
, c. Assemble your ex ami nat i on as f oll ows:
(1) Exam ouestions on top.
(2) Exam aids - figures. tables. etc.
(3) Answer pages including figures which are part of the answer.
b. Turn in your copv of the examination and all pages used to answer ;
the examination cuest i on s .
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 l eavi ng , vou are found in this area while the examination is still . i in progress. your license may be denied or revoked.
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QUESTION 5.01 (2.00)
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LIST and describe TWO conditions which will assure adeouate core cool i no.
[2.03 QUESTION 5.02 (1.SO)
8. Briefly describe which centributor to NPSH ensures adequate NPSH will be ava'lable for the recirculation pumost 1. during low power saturated conditions. CO.75]
2. during full power operation. CO.753 OUESTION 5.03 (3.00) bk
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Annwer the f ollowing assuming that Xe is in eouilibrium before the t r ansi ent occurs.
c. Ex pl ai n whether Xe concentration will initially INCREASE or DECREASE and WHY during a power reduction from 100% to 60%. Cl.03 b. EXPLAIN what happens to Xenon concentration after a scram from high oower occurs. Be sure to discuss in your answer:
(1) WHAT happens to Xe production and removal.
(2) WHAT happens to Xe concentration over the first 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after the scram AND WHY I
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(3) WHAT will eventually happen to the Xe concentration over the next three (3) days if the plant remains shutdown. C2.03 ;
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QUESTION 5.04 (2.00)
A reactor oeriod of 85 seconds has been established. Reactor temperature is 185 F. With no operator action, what will moderator tcmperature be when the reactor returns to an infinite period. Assume tho fuel is near end of cycle. C2.03 Stcte all assumptions Show all work QUESTION 5.05 (3.00)
The reactor is operating at rated conditions when one (1) SRV inadvertent 1v opens and stays ooen. State HOW the following parameters would initially change and WHY. Assume no scram and no operator action, c. Total INDICATED steam flow CO.753 b. Turbine steam flow CO.753 c. Megawatt Electric Output CO.753 d. Reactor oower CO.753
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QUESTION 5.06 (2.00)
c. Calcu. ate the reactor cooldown rate for reactor oressure decreasing from 685 osig to 465 osig in 30 minutes. Use the attached set of reference material if reouired. Show all work. Cl.53 b. Have vor exceeded any reactor cooldown l i mi ts? Ex ol ain. C0.53
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QUESTION 5.07 (3.00)
c. You are currentiv operating near rated power early in core life CBOL3 when a partial loss of feedwater heating occurs.
Would control rods have to be INSERTED or WITHDRAWN to return reactor power to the level before the loss of feedwater heating cnd WHY? C2.03 b. If ~ the loss of feedwater heating were to occur at the end of the fuel cycle. would the resulting rod movement be CLARGER. SMALLER or NOT CHANGE 3? from the rod movement at the beginning of core life.
EXPLAIN YOUR ANSWER. INCLUDE THE CONTRIBUTIONS OF THE REACTIVITY COEFFICIENTS CAssume the rods used in each transient are eaual worth.]
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QUESTION 5.08 (2.50)
The f oll owing is a partial output of program P-1 from the process computer. The segment shown is in AREA 4 for MAPRAT:
FOR MAPRAT MAPRAT LOC MAPLHGR LIMLHGR
1.051 13-16-4 11.96 11.38 1.011 31-16-4 11.39 11.38
.788 13-30-4 8.97 11.38
.779 13-26-4 8.89 11.41
.779 13-20-4 8.89 11.41
.753 13-28-4 9.04 12.00 Anower the following concerning MAPRAT:
a. What is the definition of MAPRAT Cword or couation3? CO.53 b. Is the core coerating within li mi ts? If so WHY? If not. WHY NOT? [1.03 c. Answer the following TRUE OR FALSE concerning MAPRAT:
1. Maintaining MAPRAT limits ensures that the APLHGR limits are met. CO.53
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2. Maintaining MAPRAT within limi ts ensures that transition boiling will not occur in >99% of the fuel bundles. CO.53
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QUESTION 5.09 (2.00)
For each of the cairs of conditions listed below. state which condi' tion would have the GREATER differential rod worth and EXPLAIN WHY,
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a. Reactor moderator temoerature of 200 F or 500 F C1.03 b. For an INSERTED ROD next to a FULLY WITHDRAWN rod or next to a FULLY INSERTED rod. CAssume the average core flux is constant] C1.03 (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
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QUESTION 5.10 (1.00)
As o subcritical reactor nears criticality, the length of time to reach an ocuilibrium count rate after an insertion of a given fixed amount of oositive reactivitv... (SELECT THE CORRECT ANSWER)
a. ... increases primarily because of the increased population of del aved neutrons in the core.
b. ... increases because of a larger number of neutron life cycles required to reach equilibrium.
c. . . . decreases pri maril y because of the increased population of del ayed neutrons in the core. I d. ... decreases because the source neutrons are becoming less important in relation to the total neutron population.
C1.03
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QUESTION 5.11 (3.00) '
The reactor is operating at 50% oower. HPCI initiates and injects. Compared to their steadv state values at 50% power. state whether the following pcrameter will Cinitially and after 10 min] INCREASE. DECREASE OR REMAIN l THE SAME. (Assume no operator action) l l
a. INITIALLY (when injection occurs):
1. Reactor oower CO.3753 2. Reactor water level CO.3753 3. Total steam flow . CO.3753 i
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4. Reactor oressure CO.3753 l
b. AFTER 10 MIN. OF HPCI OPERATION )
1. Reactor oower CO.3753 2. Reactor water l evel CO.3753 3. Total steam flow CO.3753 4 Reactor pressure CO.3753
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QUESTION 6.01 (3.00)
For each of the f ollowing parameters STATE if it DIRECTLY initiates a ceram, a rod block. both or neither CDo not include setooints) :
Ascume power operations. [3.O]
1. Drywell oressure = 2.3 psig 2. APRM - Channel A = 95% Channel B = 90% Total Recirc Flow = 70%
3. Drvwell temoerature = 150 F 4. Main Steam Line Radiation monitor = 7000 mrem /hr (NFPB= 1000 mrem /hr)
5. Main Steam Line tunnel temperature = 215 F 6. Reactor Vessel Level = +10" QUESTION 6.02 (1.00)
Using the attached drawing (SPVAH):
Will a failure of the "b" valve prevent the scram back-up function from working? Explain. C1.03
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4t__EbeSI_SYS]EDS_ PES]QN3,CgN]@g63,@Ng_!$5}@UDEN]@gN PAGE 9 . e . QUESilON 6.03 (3.00) c. Describe FIVE AUTOMATIC actions that would occur in the RHR system if the RHR system was aligned for suppression pool cooling and a valid LPCI initiation signal was received. Use the suppli ed diagram and assume the f ol l owing No operator action HVF O24 is throttled open RHR Pump B i s running HVF 048 i s shut HVF 003 is open H.VF 047 is open
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The reactor pressure is initially 600 psig and it decreases to 30C psig. A7 33 9,, ,a j , C2.03 b. What THREE system i nterl oc ks must be satisfied to restore the RHR system to the suporession cool cooling line up (given in part a). Include any setooints or operator actions required to satisfy these interlocks. [1.03
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Podb\t QUESTION 6.04 (2.50) ga3 ken Assume the reactor has just st r a nimed for an undetermined reason with a core age near the beginning of the fuel cycle. With the mode switch in RUN and pressure decreasing, answer the following about the Nuclear Steam Supolv Shutoff CNSSS] system isolations a. WHY is the reactor i sol at ed at 756 psig with the mode switch in RUN? . C1.03 b. To reset the i sol a t i on. LIST the two actions and/or checks tnat are reouired. Assume that the mode switch i s no longer in 'RUN'.
[1.03 c. What combinations of manual ce.-t oushbuttons A-B-C-D must be depressed to cause a manual isolation? CO.53
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List the FIVE signals that will isolate the Reactor Building ventilation system and start the Filtration. Rec i rc ul at i on j cnd Ventilation System CFRVS). Give setpoints where applicable.
C2.03 i l l QUESTION 6.06 (2.00) 1 With the reactor at rated power. Hope Creek experiences a Loss of Offsite Power event and a reactor scram. Approximately 13 see into the
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evsnt. drywell pressure is 1.9 psig. Answer the following: I c. Assume that the Diesel Generators are functioning properly. STATE l which load secuence program is in control CLOP or LOCA3 to restore loads to the 4160 bus and WHY. C1.03 j l b. What is the difference between the two load secuence programs CLOP j and LOCA3 in the way the bus is subsecuently reloaded. C1.03 ,
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QUESTION 6.07 (2.50) c. What is the purpose of the excess flow check valve in ! cn RPV instrument line' CO.53 b. How would Indicated RPV level be affected (INCREASE. DECREASE or NO CHANGE) if a leak cevel oped in the reference leg? WHY? CO.53 l c. For the following l evel instruments and ranges. state all j automatic action (s) which will occur, if any, from signals provided I by these instrumentst
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1. WIDE RANGE. LEVEL 2 UPSET RANGE. LEVEL 8
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3. NARROW RANGE. LEVEL 3 Cl.53
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QUESTION 6.08 (2.00) Tho olant is coerating at 100% power and flow. APRM ch'annel A reads 118% and Channel C is bvoassed. Instrument technicians are investicating the problem with Channel C while you research Tech Specs.
A plant auxiliary operator in training shifts RPS B power supp1v to its clternate power supply without informing the control room.
What automatic actions will occur? Exclain. C2.03 poEbb t QUESTION 6.09 (3.00) qvesduwd c. What causes a Rod Drift Alarm? L1.03 b. During a reactor heatup with pressure at 800 psig. if both CRD pumos become inoperable and 2 accumulator trouble lights energize WHY must the reactor be manually scrammed? [2.03 QUESTION 6.10 (2.00) For EACH of the f oll owing. answer TRUE OR FALSE: The Main Steam Line Radiation Monitoring System... l c. ... monitors the cross gamma radiation from the main steam lines at a l oc ati on just upstream of the inboard MSIV's b. ...will initiata a PCIS Grouc I isolation, but not a direct SCRAM when the tric level of 3X normal full power background is reached c. ...will activate a "Main Steam Line Downscale" annunciator to a alert the operator of an equipment malf unction while at power
d. ...will trip the mechanical vacuum pump, if running, and close its j suction valves when the trip level of 1.5x normal full power background is reached. [2.03 (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) __
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'QUESTlON 6.11 (2.00) 90Cshm Tho reactor i s at 100*/. Dower when a complete loss of instument air occurs. What is the first cause of a reactor scram and WHY? CAssume no cosrator actions 3. [2.O]
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~ PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 13 hhpiOLOhl(h(_~(hhIhh(~~~~~~~~~~~~~~~~~~~~~~~~ .
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QUESTION 7.01 (3.00) For the following plant carameters: 1. STATE the Emergency Operating Procedure (s) that would be utilized.
2. STATE ALL of the entry condition (n) met for each EOP utilized.
[ Assume all parameters are occurring simultaneously.] RPV Water Levels -65" Devwell Pressure 9 psig ' Suppression Pool Temperature 145 F Drywell Temperatures 245 F Racctor Pressure: 1052 psi g Rocctor Power 6%
[3.03 QUESTION 7.02 (3.00)
Stco 4 of the SCRAM PROCEDURES directs that RPV level be rGBtored and haintained between +12.5" and +54" using available - wcter supplies.
. a. LIST FIVE (5) of the seven water supplies and include the maximum discharge cressure for each source. [1.53 b. What is the bases for UPPER and LOWER level l i mi t stated in this step. C1.53 I i QUESTION 7.03 (3.00) l You have entered RPV/ PRESSURE VESSEL CONTROL PROCEDURE.
You are instructed by the procedure to inject Boron. Prior to starting SLC.
you are also instructed to prevent automatic initiation of ADS. Explain why? (Include any adverse ef f ects in your answer) C3.03 l l l
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7t__P60CEDU8ES_ _NQ85@L 1,@BNQ@d@(t_g0g5GENCY_@ND PAGE 14 SADIOLQQlC@L_CQNI@Q( QUESTION 7.04 (2.00) c. OP-OO9 "Refueling". contains the following steo:
"Within 8 hours pri or to the start of CORE ALTERATION requiring control rod withdrawl. verify that the RPS shorting links are removed" l
Note 5.1.4 of OP-OO9 states that "Removal of the RPS shorting links is i NOT required if adequate shutdown margin has been demonstrated 1AW Tech ) Soet 3.1.1".
, EXPLAIN why the shorting links can stav in place if the shutdown margin is adequate. Include in your answer the f unction of the shorting links. C1.03 i l i
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b. Which of the followino are core alterations as defined by Tech Soec Section 1.7: CAnswer ves or no] l 1. Removal of the startuo source 2. Insertion of an IRM detector 3. Relocation of a fuel bundle to the fuel cool 4 LPRM removal C1.03 l l l l
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QUESTION 7.05 (3.00) l l I a. STATE the exoosure limit associated with each of the followino radiation categories: 1. Radiation Area 2. High Radiation Area l 3. Locked High Radiation Area C1.53
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6. In accordance with SA-AP.ZZ-024(Q) Radiation Protection Program.
the initial cuarterly whole body dose limit may be extended provided certain reouirements are satisfied. STATE the . initial cuarterly whole body dose ADMINISTRATIVE limit and LIST the , requirements [there are 23 for extending the limit. ! C1.53 l l
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QUESTION 7.06 (2.50) a. You have entered the Containment Control EOP on High Drvwell temperature. The procedure directs you to spray the drvwell. The f ollowing plant parameters are relayed to yous Drywell temperature - 250 F Suppression Pool Temperature - 217 F dra vnbcr
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Reactor pressure - 1015 psig Drywell Pressure - 1G psig a. Can crywell sprays be initiated? Justify your answer using the attached reference material.
C1.03 . b. If drywell sprays are initiated when containment parameters are outside those allowed in the EOP's, what adverse effect would this have on the containment and WHY? C1.51 CUESTION 7.07 (2.00) Whi l e operati ng at rated power, the following annunciator alarm comes in
'Feedwater Heater Trio'. You confirm a loss of feedwater heating.
Assuming there is no scram, what immediate action (s) are reouired oer OP-AB.22-118 LOSS OF FEEDWATER HEATING? CTWO ANSWERS REQUIRED 3 C2.03
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QUESTION 7.08 (1.50) Which of the following is (are) symptom (s) that the #1 orifice of a rocirculation pump seal is plugged? [1.53 a. #1 seal pressure decreases b. #1 seal pressure increases c. #2 seal pressure decreases d. #2 seal pressure increases ; i e. controlled leakage throuch #2 orifice increases f. controlled leakage through #2 orifice decreases ;
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o. seal temperature increases 4 seal temperature decreases l h. 1 QUESTION 7.09 (2.00) ; i l Comolete the following statement using the selections from below: :
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When an Emeroency Load Reduction is recuired [1.A.W. RE-CI.ZZ-OOI. Core Oparations] and rods are to be ' stuffed' to reduce oower .
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___ (a) ___ first and insert it ___ (b) ___ into the core. C2.O] l l l a ) sel ec t a shallow rod use the rod presently sel ec t ed sel ec t a deep rod sel ec t a peripheral rod I l b)a minimum of 2 notches fully
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as directed by the Reactor Engineer l
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7 PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND
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QUESTION 7.10 (3.00) l For each of the f ollowing situations given below, classif y the l evsnt using the HOPE CREEK EMERGENCY EVENT CLASSIFICATION GulDE.
1. The reactor has scrammed and the MS!V's have closed due l to high radiation. Primary leakage is 80 gom. CO.753 l l 2. HPCI system f ails during a surveillance test. [0.753 ) l 3. A man is injured while working on contaminated equipment ] and he needs evacuation to the hospital. CO.75] 1 4 Offsite power lost at 1:20 AM with failure of D/G's A & C to start. It i s now 1: 40 AM. CO.75] l l
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' QUESTION 9.01 (2.00) O. Prior to authorizing restart of the reactor after a SCRAM. the Operations Manager and the SNSS review the Post Reactor Scram /ECCS Accuation. Review Report. Li st THREE items checked during this revi ew. C1.53 b. Who (bv title 3 must ultimately give approval to restart after receiving a startuo recommendation from the SORC? CO.53 QUESTION B.02 (3.00) c. You are the refueling SRO. Li st any three of the the minimum number of personnel (bv title) who are also recuired on the refuel floor to load fuel into the core at Hope Creek. [1.53 b. Fall in the blank from the cotions belows Members of the refuel crew should work no more than .,___, continuous hours.
(1.03 ANSWER OPTIONSt 6. 8. 10. 12. 14. 16
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c. TRUE OR FALSE: Communications are reouired to be established betveen the refuel floor and the control room at all times during fuel l oatii n g . CO.53 QUESTION 8.07- (2.50) a. Who (by title 3 i s included in a Work Coordination Control Group CWCCG3? CTwo answers reauired3 C1.03' b. What i s the f unction of the WCCG? [1.03 c. The WCCG does not review all work cermi ts. What is the criteria for submitting work permits to the WCCG? CO.53 (***** CATEGORY 08 CONTINUED ON NEXT PAGC *****) _ _ - . _ . _ - _ __ . _ _ . _ .
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QUEST]DN B.04 (1.00) c. The maximum STEADY STATE core flow authorized by RE-DI.22-Ol. Core Operation Guidelines shall not exceed _______________. (0.33 b. Answer TRUE or FALSE : With the RWM found to be inoperable below 20'/.. the only rod movement allowable is initiating a MANUAL SCRAM or selecting SHUTDOWN with the reactor manual sel ecter swi tch. CO.53 I0 ' QUESTION 8.05 M * Sb'43I' @ # . As the Emergencv Coordinatc' you have declared a Site Area Emergency CSAEJ. You now wish to termlante the event.
Antwer the following TRUE OR FALSE concerning the terminati on of an SAE: Mik , =- m -- __ . - % i. e i. h i s event. C ors 3- ); b. To terminate, both of the f ollowing must be true-None of the applicable a-tion levels defined in the ECG are applicable AND...
-The plant is in a recovery status (0.53 .
c. NRC concurrence IS NOT reoutred tc UPGRADE an events but NPC concurrence IS reouired to DOWNGRADE an event. [0.53 OUESTION 8.06 (3.00) o. What are the maximum surveillance time extension periods?
(2.03 b. 'On the spot changes' to procedures, may be accroved by members of the *
unit manacement staff provided what TWO criteria are satisfied. [1.03 (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
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. 'OUESTION 8.07 (3.00)
The plant is at 7 5'/. p o w e r and in procedino to fu;) oower in accordance I with 01-006.
Tho NSS is then notified that the .not or operator for the HFCI Outboard ! Stoam Supply Valve CHV-FOO33 has failed so that it cannot be moved from th3 control room. l Ccn the power increase continue to full power? Why or why not? State ALL applicable Tech Soet LCO's and action statements which apply. Cuss- the i I attached Technical Speci ficati ons t devel op your answer. Be sure to roference each apolicable Tech Soe- 'CO and action statement by number.3
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C3.03 QUESTION 8.08 (2.00) For each 'USE' described b el ow. MATCH it with the appropri at e tag: [2.03 l USE TAG _____ _____ c. Di sti nct l y marks electrical RED equipment that is safe for work b. Alerts plant oersonnel of YELLOW malfunctionino eculoment which has been identified to the work order control system c. Identify el ec t r i c al and mechanical WHil.E WITH blocking ooints between any circuit GREEN DORDER or eaut oment that is energized (or could q be) and the deenergized eaui9 ment upon i which work is to be performed ' d. Indicates a position other than the LIPH' RLUE normal operating oosition of electrical eculpment WHITE (*s*** CATEGORY 08 CONTINUED c,N NEXT PAGE *****) l
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QUESTION B.09 (3.00)
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l Tho reactor is operating at 100% oower. You are informed that the RCIC Stcam Line delta Pressure (Flow)-High instrument i s inoperable. According to Tech Specs. what actions if any, shnuld you take? C3.003 (Une the attached Tech Specs to form your answert be sure to document coplicable sections used as cart of your answer] QUESTION 8.10 (3.00)
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o. While operating at p o' .wr (condi t i on 1). the Station Service Water pumo
"A" has been out of service for repairs for two days. Maintenance continues on the Dumo and it is expected to be returned to service in four days.
A Dioe break large enough to render SSW oumos "B" and "D" inoperab'le occurs in the SSW pumo room. Both pumps are taken out of service.
Using the attached Tech Specs. determine whet action (s) you will take and re+erence all Tech Specs you use to develop your answer. C1.03 o. How would your answer differ if the above scenario were to happen while tne plant was performing refueling operations? [ Assume olent heat losses are sufficient to maintain Condition 5. 3 (2.03
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QUESTION 8.11 (1.00) l The reactor has been at rated power for several weeks. The RO reports that a check of the reactor's thermal output for the last hour was 3320 MWt. I
Recctor power was reduced immediately to 3290 MWt and it was found that the rocctor was at 3320 MWt for only one hour. Have any limits been exceeded? l If so WHY7 If not WHY NOT? CTech Specs are attached 3 [1.03 i
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ANSWERS -- HOPE CREEK -88/04/12 b , w y es b ANSWER 5.01 (2.00) c. Maint ai n reactor vessel l evel . Maintainino l evel C atove two-thirds core active height) assures adequate cooling and adeouate driving head for natural circulation. . (1.03 on C.cs Ssee% wee . Spry Co<.At'% 4 5 % b b 1 forI0) b. Minimum of one LP ECCS oumo injecting water into the core C1.03 REFERENCE OP-ED.ZZ-102CO3 Containment Control Section 3.0 302HCOOO.OO-100-01. Objective 1.1.4 K/A's... 205000 G010 3.2/3.3 205000G010 ...(KA*S) ANSWER 5.02 (1.50) 1. The (57 foot] el evat i on difference between normal reactor vessel water and the recirc oumos Corovides adeouate NPSH during low oower saturated conditions, with minimum Dumo speed 3. CO.753
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2. cat, full power with recirc at 100'/. f l ow ) . feedwater flow provides I subcooling Cto the recire oumo suction to ensure adequate l NPSH). CO.753 REFERENCE 302/304HC-145.03-HTFF-42-01. E.O. 3.1 K/A's... 291004 K1.06 3.3/3.3 293006 K1.10 2.7/2.8 293006 K1.08 2.5/2.6 291004K106 293OO6K108 293OO6K110 ...(KA'S) ,
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4t_ IHEORY,O[, NUCLE @R,PQWCR,PL@yl_QEER@llgNt,[LylGS3 ANp PAGE 23 IHERMOQyNAMjCS . ANSWERS -- HOPE CREEK -88,/04/12 msy Qw.M ANSWER 5.03 (3.00) o. Xenon c onc en tr at i er, wi l l increase as reactor oower decreases from 100% to 60 %CO.53. Thi s increase is attributable to the reduction of Xe ' burnout' caused by decreased flux levels at the new lower power.
CO.53 b. Cl3 After the scram, neutron flux amoroaches zero and therefore Xe concentration can be described by the following Xe Conc. = h d; + a vc I = rd [,p 0 0 CVerbal discussion is acceotable] C2] Because Iodine decavs to Xe at a rate faster than Xe decavs to Cesium. Chalf-life's of 6.6 hrs versus 9.1 hrs] Xe concentration builds as the pre-scram lodine concentr ati on decavs.
CO.53 lodine concentration will decrease as it decays to Xe resulting in a reduction of Xe produced. Xe concentration will peak when the rate of production of Xe f rom Iodine decay ecuals the rate of removal of Xe from the decay of Xe.
- CO.53 This peak will be 6 to 10 hours after the reactor is shutdown.
deoending on power level., C33 Xe concentration will eventually reach zero when oroducti on from lodine is exhausted. and the remaining Xe concentration decavs to Cesium. CO.53 REFERENCE RXPH-33 op.4-91 L.O. 1.2.3 K/A... 292006 Kl.07 3.2/3.2 292006 K1.03 2.9/2.9 292006 Kl.04 2.9/2.9 292OO6KlO7 292OO6K104 292OO6KlO3 ...(KA*S) l I I
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ANSWER 5.04 (2.00) Psriod = B - dk/ dkt.1 dk(Period +1)= B dk = B/ (1+ Periods.1) dk = .0054 / (1 + 8.5) = .000b7 CO.53 T mod = dk / aloha T
= .00057/.0001 = 5.7 F C1.03 Thorefore moderator temoerature would increawel = 185 F + 5.7 F = 190.7 F CO.53 REFERENCE RXPH 24. L.O. 3.4 K/A... 292000 K1.11 3.7/3.8 292000 K1.12 3.6/3.7 292008 K1.15 3.7/3.7 292OOOK111 292OOOK112 292OO8K115 ...(KA'S)
I ANSWER 5.05 (3.00) c. DECREASE CO.253: reactor pressure decreases from the SRV openino and because the d/o cell used to measure steam flow is located downstream of the SRV. indicated steam flow will decrease [0.53. . b. DECREASE [0.253: as reactor cressure decreases. the EHC system will shut down on the TCV's and turbine steam flow will decrease CO.53.
c. DECREASE CO.253: with less steam flow to the turbine, electrical output will decrease [0.53,
d. DECREASE CO.253t as the safety relief valve lifts, reactor power will decrease as reactor oressure decreases [0.53 REFERENCE HC LP 106 Transient Analysis. op 12-13 Learning Objective 1.1 I K/A'S... 293008 K1.21 2.9/3.0 293008 K1.22 3.5/3.6 293OOOK121 293OOOK122 . . . ( K A '. S ) _ , _ - - - . ,. .-- . . .
St..U.KOGY,0[,NyCLggg,EgWg5,E6@y],9Eg@@JJgy3,E(y, lgs3,AND PAGE 25 THERMODYNAM1CS ANSWERS -- HOPE CREEK -88/04/12 . ANSWER S.06 (2.00) c. CONVERT PRESSURE TO PSIA: 805 +14.7 = 900 osia 495 + 14.7 = 500 psia CO.53 OBTAIN CORRESPONDING TEMPERATURE FROM STEAM TABLES: 900 osia -> 532 F 500 osia -> 467 F CO.53 DETERMINE THE TEMPERATURE CHANGE : , 532 - 467 = 65 F in 30 minutes CO.53 b. NO: the TS cooldown rate l i mi t is 100 F/hr and this has not been exceeded.
CO.53 REFERENCE , l Saturated steam tables. HTFF 10-01. L.O. 2 HC T.S. 3.4.6.1.b K/A's... 293003 K1.23 2.8/3.1 1 293OO3K123 ...(KA'S)
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ANSWER 5.07 (3.00) l e. Rods should be i nser t ed C 1. 03. The transient will cause a reactivity increase in the core from the decrease in void coefficient CO.53 and reduction i n the RCS temoeratureCO.53. Thi s reactivity increase will be offset by an insertion of control rods.
b. The resultant change weald be a smaller rod movement (0.53.
This is because alpha (mod 3 and aloha Cvoid] become less negative CO.253 and alpha Cdocoler] CO.253becomes more negative recuiring less rod motion. I REFERENCE l 305HC-OOO.OO-RXPH27-01 L.D. 2.3 305HC-OOO.OO-RXPH28-Ol L.O. 2 305HC-OOO.OO-RXPH29-01 L.O. 3d K/A's... 292004 K1.04 3.3/3.3 292004 Kl.02 2.5/2.6 ; 292OO4K114 292OO4K102 ...(KA'S) I l
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JHERMQDyN@MjQS : ANSWERS -- HOPE CREEK -88/04/12- ,
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l ANSWER 5.98 (2.50) c. MAPRAT = MAPLHGR (actual ) / MAPLHGR LCO : OR the ratio of the avg. ! alcnar linear heat generation rate (APLHGR) to the APLHGR coerating limit. CO.53 b. NO for lines 1 and 2 CO.53. MAPRAT's in excess of 1.0 enean that ) MAPLHGR (limitino] has been exceeded. CO.53 ; I c. 1. TRUE CO.53 2. FALSE CO.53 l l REFERENCE 302HC-OOO.OO-109-OO L.O.4 op. 11-25 ) HC Heat Transfer and Thermal Limits Section 16 j K/A's... 293009 K1.13 3.1/3.6 293009 Kl.12 2.9/3.5 294001 A1.15 3.2/3.4 294001 A1.08 3.1/3.6 ) 294001A115 294001A108 293OO9K113 293OO9K112 ...(KA'S) !
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ANSWER 5.09 (2.00) I
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a. 500 F Co.253: As moderator temocrature increases. thermal diffusion length increases. thus the control rod is excosed to higher neutron flux and its worth increases.CO.753 b. Next to a withdrawn rod CO.253: Local neutron flux will be j higher here making the rods' worth greater CO.753 I I REFERENCE 305 HC COO.OO-RXPH31-01 Objective 5 ;
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ANSWER 5.10 (1.00) b.
I REFERENCE 305HC-OOO.OO-RXPH-21-01 op. 4-7 K/A's... 292003 kl.01 292OO3K101 ...(KA'S) -
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FLUIDS. AN
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ANSWER 5.11 (3.00) c.1. decrease 2. increase 3. decrease 4. decrease b.1. Increase gQ gbGimk (cal 2. t r. ; r ; _ . - @ men et kk SWE 3. decrease 4. decrease . CO O .375 EA3 REFERENCE , 302HC-OOO.OO-026-04 primary objectives 10. 12 FSAR Section 15.5-1 K/A's... 206000 A2.19 3.9/4.3 206000 K1.02 4.0/4.1 206000 K1.03 3.8/3.0 206000N103 206000K102 206000A217 ...(KA'S)
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ANSWERS - . HOPE CREEK -88/04/12
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- ANSWER 6.01 0 (3.'0)
1. scram 2. rod block 3. neither 4 scram ' 5. neither 6. scram C6@ 0.5 ea.] REFERENCE 302HC-OOO.OO-022-03 Enabling Obj ecti ve 5 K/A's... 212005 K1.01 3.7/3.9 212005 K1.02 3.7/3.9 212005 K1.05 3.3/3.6 212005 K1.13 3.5/3.6 212OO5K113 212OO5K105 212OO5K102 212OO5K101 ...(KA'S) ANSWER 6.02 (1.00) No. The valve 'b' is the downstream backup scram vent valve. It is , orovided with a bvoass check valve that will allow venting of the hoeder via the upstream valve ("a"31n the event that the downstream vc1ve fails to operate.
(1.03 REFERENCE 302HC-OOO.OO-OO6-02 Enabling Objective 3f
K/A's... 201001 K4.04 3.6/3.6 201001K404 ...(KA*S) ' l I
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4t..fL993.!YS}EUS_QESlGyg,CgyJRgLg,@yp,10SJyUyEy}@Jjgy PAGE 29 ANSWERS -- HOPE CREEK -03/04/12
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ANSWER 6.03 (3.00) c. All RHR pumps start All HVF 048 HX Bypass valves open HVF O24A(B) will shut At 450 osig in the reactor, all HVF 017's will open
[At 1400 gom system flew.) all HVF 007' will close [5p O.4 ea3 b. HX bypass valve cannot be repositioned for 3 minutes after a valid LPCI initiation signal HVF O24 cannot be opened without first depressing the ' AUTO CL OVRD'
PB HVF 017 cannot be shut without first deDressing the ' AUTO OP OVRD'PB
[30 0.33 ea3 REFERENCE OP-SO.BC-OO1(D)
302HC-OOO.OO-020-04 Enablino Objective 10.
K/A's... 203000 K4.01 4.2/4.2 203000 V4.10 3.9/4.1 203000 A3.01 3.8/3.7 203000 A4.06 3.9/3.9 203000K410 203OOOK401 203OOOA406 203000A301 ...(KA'S) ANSWER 6.04 (2.50)
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\.O c. The i sol at i on stops the reactor depressurization and c ool down . t? . 7. 3-
'Mr Ortcci-t act chenrr becawee-essu=ct:cn; m;de for MCIC ere not calid 'c- 211 critical power calculat;cne [-et crassurra br!ca 705-csic cr cc r '!:x; 1r:; : n e ,- 10 '/. eted fic-.: [0,52 a
l c. -Reset the isol ation logic by depressing the ' RESET * PB j-Place all MSIV control switches in the ' CLOSED' position <
[20 0.5 ea3 I e B '
l c. Any combination of A ORdt with[and] 4 OR D [0.53 (er b l. 6 MSiv reset fb isetea,-ggNe+ced] REFERENCE I 302HC-OOO.OO-045-02 Enabling Objective 9 and 10 l HC Tech Spec Bases Section 2.1.1 l K/A's... 223002 K1.01 3.0/3.9 223002 K4.06 3.4/3.5 j 223002 A1.02 3.7/3.7 223OO2K406 223OO2K101 223OO2A102 ...(KA'S)
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ANSWERS -- HOPE CREEK -88/04/12-
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l l ANSWER 6.05 (2.00) c. -30" RPV Level 1.68 osig Drywell pressure Hi Rad at the Refuel Floor Ventilation Exhaust 2x10E-3 uC/cc.
Refuel Accident Signal Hi Rad Reactor Buildino Ventilation Exhaust 1x10E-3 uC/cc. Main Steam Line Break Outside the Drywell Signal Manual PCIS initiation C5 & O.4 ea) REFERENCE 302HC-OOO.OO-042-02 Enablino Objectives 13 and 14 K/A's... 288000 K4.01 3.7/3.9 288000 K5.01 3.1/3.2 288000 GEN 05 2.6/3.4 288000 GEN 11 2.8/3.6 288000K501 288000K401 288000dO11 288000 GOOS ...(KA'S) i ANSWER 6.06 (2.00) i c. LOCA CO.53: a LOCA program will supercede a LOP program when attempted to run concurrent 1v.CO.53 b. The LOCA load secuencer reloads only those units recuired to perform a safe plant shutdown under emergency conditions C1.03 REFERENCE 302HC-OOO.OO-068-02 Enablino Objectives 8.9 K/A's... 262001 K3.02 3.8/4.2 262001 K4.05 3.2/3.5 262001 A3.05 3.4/3.5 262OO1K405 262OO1K302 262OO1A305 ...(KA*S) l i e i
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ANSWER 6.07 (2.50) l e. The excess flow check val ve i s installed in an RPV instrument " line to isolate the line if a break occurs downstream of the check valve. CO.53 b. The leak would cause a decrease in reference leg Cdensity. l evel or.oressure J. Indicated level would increase. CO.53 c. 1. N3SSS Isolations. HPCI Initiation. PCIS/MSIV boundary leakage.
RCIC Initiation.*Recire Pump Trio / ARI 2. None ' 3. RPS Trio. NSSSS Isol ation S/D Cooling, ADS Permissive. Recirc Run Back C3 0 0.5 ea)
' hi it %.ept 8Ac t d.ik ss cicar Mr/ m i s +q g ,, ; g,, g q ,,g y,3 REFERENCE 302HC-OOO.OO-02-01. Table 1 Encbling Ob.iective 2a.5 K/A's... 261000 K5.07 3.6/3.8 216000 K5.12 3.5/3.6 216000 K1.02 3.8/4.0 216000 K1.14 3.8/4.1 216000 K1.07 3.9/4.1 261000k507 216000K512 216000K114 216000K102 21600K107 ...(KA'S) .e, ANSWER 6.08 (2.00)
j When transferrino RPS B bus power supplies. RPS B is momentarily
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desnergized causino a half scram in Channel B Co.53 coupled with a half ceram from hioh power on Channel A CO.53 a full scram will occur (1.03.
REFERENCE 302HC-OOO.CO-022-03, pp. 25. 41 Enabling Objective Sa. 6a. 7 K/A's... 212000 K3.05 3.7/3.8 212000 K4.01 3.4/3.6 212000 K4.02 3.5/3.7 212OOOK402 212OOOK401 212OOOK305 ...(KA'S) - f
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J , et .PL9NI,SyS] EMS,9CSIGN3_CQN160Lg,9Np,1NS16ydENI91}ON PAGE 32 ANSWERS -- HOPE CREEK -88/04/12-
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ANSWER 6.09 (3.00) c. The motion of a control rod passino an odd reed switch that is not selected by the RMCS. C1.03 b. With reactor pressure less than 900 osig, it cannot be assured that a complete scram will occur if the scram accumulators are not adeouately e chargedC1.03. Therefore with no CRD oumps available and evidence that more than one accumulator is below the alarm setooint with the reactor below 900 psio. the reactor is scrammed [1.03 REFERENCE 302HC-OOO.OO-114-OO Enabli ng Obj ecti ve 3 302HC-OOO.OO-OO7-02 Enablino Objective 4 302HC-OOO.OO-OO6-02 Enabling Objective 17 K/A's... 202000 K4.03 3.6/3.6 295022 AK3.01 3.7/3.9 295022AK30 201002K403 ...(KA'S) ANSWER 6.10 (2.00) c. false b. false c. true d. false (40 0.5 eal REFERENCE 302HC-OOO.OO-022-03 Enabling Objective Sa K/A's... 272000 K1.01 3.6/3.8 272000 Kl.OB 3.6/3.9 272000 K3.04 3.7/3.8 272000 K4.02 3.7/4.1 272OOOK402 272OOOK304 272OOOK108 272OOOK101 ...(KA'S) ! l l ANSWER 6.11 (2.00)
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The reactor will scram on low level C1.03 because the Feedwater Pumo ' rocire valves f ail open uoon a loss of instrument air.C1.03 l out9. (bm hte re s petes (w (il C e d M *. - 4 M % rck M *L*b REFERENCE h t e a, e.N MVpHWC Sc h 302HC-OOO.OO-75-OO Instrument Air oc on e454 V q p. @ ,c. Sc e 302HC-OOO.OO-58-OO Feedwater ,_gg,g, g , ggf 302HC-OOO.OO-114-OO AB Procedures Objective 3a.b.
K/A's... 295019 AK2.03 3.2/3.3 259001 K6.01 3.0/3.0 , 6 ,* ,, ggg 295019AK2O 259001K601 ...(KA'S) k d.rIC F op SN a i l
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?*.. PROCEDURES , NORMAL.,@BNORMALt_EMERGgNCY,@Np PAGE 33 *
RADIOLOGICAL.QQN~ROL ANSWERS -- HOPE CREEK -80/04/12 i ANSWER 7.01 (3.00) RPV/ REACTOR POWER CONTROL: EO 101 Co.53 RPV LEVEL < -30 CO.253 D/W PRESSURE > 1.60 PSIG Co.253 REACTOR POWER > 4% CO.253 REACTOR PRESSURE > 1037 PSIG CO.253 CONTAINMENT CONTROL: EO 102 CO.53 SUPPRESSION POOLTEMP. > 95 F CO.333 D/W TEMPERATURE > 135 F CO.333 D/W PRESSURE > 1.60 PSIG CO.333 REFERENCE EO 101 RPV / POWER CONTROL , ED 102 CONTAINMENT CONTROL k/A's... 295004 G011 4.3/4.5 295025 G011 4.2/4.2 295026 G011 4.4/4.6 295031 G011 4.2/4.7 295031G011 295026G011 295025K201 295025G011 295025A202 295004G011 ...tkA'S) ANSWER 7.02 (3.00)
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c. CONDENSATE . 720 osio FEEDWATER 1200 osio CRD 1500 osio RCIC 1250 osio ' HPCI 1250 osig CS 300 osio LPCI 340 osi g C5 sets at O.3 per set]
[ Note to oraders cressures are to be accepted to a 'olus or minus' 5% tolerance]
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' D. +54" - Corresponds to the turbine trip setpoints (main and fosdoumo) and component i sol at i on s CHPCI and RCIC3 that prevent coutpment damage. CO.753
+12.5" - Corresponds to the low level scram setpoint so that tho scram can be reset Cbarring other signals to RPS3.CO.753 REFERENCE OP EO.ZZ-100CZ3 o. 4 K/A's... 295006 AK2.02 3.0/3.0 295006 AK3.01 3.0/3.9 295006 AA2.03 4.0/4.2 ._ --. - . ___-
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7.,, PROCEDURES, _ NORMAL,,ABNORMALt. EMERGENCY _AND PAGE 34 R AD,10L O91G@b _ CONT R06
. ANSWERS -- HOPE CREEK -08/04/12-295006AA30 295006AK2O 295006AA20 ...(KA'S)
ANSWER 7.03 (3.00) If ADS were to initiate durino or after reactor shutdown with Boron, the following may occur: laroe volumes of low pressure water would dilute the Boron adding posi ti ve reactivi ty: C1.03 cold water addition would add posi tive reactivity: C1.03 a resulting power excursion could be large enough to cause substantial core damaae C1.03 REFERENCE OP EO.ZZ-101CZ3 o. 20 K/A's... 295037 Gen 7 3.7/3.9 ANSWER 7.04 (2.00) c. Shortino links, when removed. Will activate the SRM Hi Hi scram - C2x10E5 cos3 and cause all NMS scrams to be non-coincident. If adequate sh u t down margin is demonstrated, then there i s no need for this orotection during core alterations. C1.03 b. 1. YES 2. NO 3. YES 4. YES C4 & O.25 EA3 REFERENCE 302HC-OOO.OO-013-02. o. 2B HC Tech Soec 3.1.1 and section 1.7 K/A's... 215004 GOO 7 3.4/3.5 * 215004 GOO 7 ...(KA'S) ,
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ANSWERS -- HOPE CREEK -88/04/12-
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ANSWER 7.05 (3.00) o. 1. 5 mR/hr or 100 mR in 5 consecutive days 2. > 100 mR/hr 3. > 1000 mR/hr (3 & O.53 b. 1000 mrom/QTR-Review dose status with individuals supervisor-Process TLD badge Cno RCA access until processing comolete] [3 @ 0.53 REFERENCE HC SA-AP.ZZ-046(Q) Radiation Access Program Control HC SA-AP.ZZ-024 Radiation Protection Program K/A's... 294001 K1.03 3.3/3.0 294001K103 ...(KA'S) ANSWER 7.06 (2.50) I c. No. drywell sprays should not be initi ated CO.53. Per DW/T-7. a check of the Drywell Sorav Initiation Pressure Limit curve i s reaut red.
Because the given parameters are in the shaded area of the curve.
spraw initiation is not allowed CO.53.
i b. With plant parameters above th'e initiation curve (SP-L-43. initiation of drywell soravs could cause steam condensaticn excessive enough to cause a fat' lure of the primary containment due to the negative l oressure C 1. 5 3.
REFERENCE HC Containment Control EOP Bases DW/T-7 K/A's... 226001 GOO 5 3.2/4.0 226001 G010 3.4/3.5 226001 A1.01 3.6/3.8 226001G010 226001 GOO 5 226001A101 ...(KA*S) l
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ANSWER 7.07 (2.00) 1. Reduce recircul ation flow so that thermal power reduces (by approximately 20%) (1.03 2. Insert conrol rod CI.A.W. the rod pattern 3 to restore power to the 100% load line. (1.03 REFERENCE OP AB.ZZ-110 LOSS OF FEEDWATER HEATING K/A's... 295001 A2.02 3.1/3.3 295001 GOO 4 3.4/3.4 295001 GOO 4 295001A202 ...(KA*S) ' ANSWER 7.08 (1.50) c. f, g (0.5 ea. all rea'd for full credit.)
REFERENCE 302HCOOO.OO-19-04 REACTOR RECIRCULATION SYSTEM. Objective 10.
K/A's... 202001 A2.10 3.5/3.9 - 202OO1A210 ...(KA'S) - ANSWER 7.09 (2.00) j a. Select deeo rods first (1.03 , l b. Insert fully (1.03 I REFERENCE
RE 01.2Z-001. Core Operations Guidelines
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P ANSWER 7.10 (3.00) CO.753 each) 1. Alert 2. None 3. Unusual Event 4. Site Area Emergency REFERENCE HOPE CREEK EMERGENCY EVENT CLASSIFICATION GUIDE KA 294001 A1.16 2.9/4.7 , 294001A116 . . . (KA'S)
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. ANSWER B.01 (2.00) c. -The report a s complete and properly prepared-The root cause has been properly identified-All eculement functioned satisfactorily-Corrective actions required to be complete prior to restart have been identified and initiated Cany 3 D O.5 ea) 6. General Manaoer CO.53 (orOpedes, Ma> r A +W.Gtu absw h REFERENCE OP-AP.ZZ-101(Q) Post Scram / ECCS Actuation Review 302HC-OOO.OO-113-OO Learning Objective 3 OP-AP.ZZ-OO2 Learning Objective 4 293OO9K121 293OO9K107 293OO9K109 293OO9K110 293OO9K112
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ANSWER 8.02 (3.00) c. Refuel Floor Supervisor Refuelino Platform Operator
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Reactor Engineer Qualified Rad Protection Personnel CAnv three D0.5 ea) b. 12 [1.03 c. TRUE CO.53 REFERENCE SA-AP.22-049(Q) Refueling Operations p. 17 - 302HC-OOO.OO-116-OO Instructional Objective 9.O. 4.0 K/A's.. 294001 A1.03 2.7/3.7.
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D_ U_ R_ E. S. . . A.N. D. .L. .I M. I. T. A. T. .I O. N. S. PAGE 39 ANSWERS -- HOPL CREET -89/04/12-e t F -
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ANSWER 8.03 (2.50) < e. A licensed SRO CO.53 , Taggino desk operator CO.53 , b. Continuous review and interface between the WORK IMPLEMENTING 1 ORGANIZATION and the NSQRg CWCCG SRO resoonsible for knowing oresent plant conditions / interface with SNSS and WCCG TDO responsibl e f or ; preparing safety taos] (1.03 l c. Any work permit that requires SNSS signature. [0.53 l REFERENCE ' SA-AP.ZZ-053(O) Work Control Coordination !
K/A's... 294001 K1.02 3.9/4.5 294001K.102 ...(KA*S) ,
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ANSWER 8.04 (1.00) a. 100 M1bm/hr (0.53
l b. False CO.53 ' REFERENCE HC Tech Soecs 3.1. 4.1 RE 01. 2 2 -001 (O )- Core Ooeration Guidelines K/A's... 201001 GOO 5 3.3/3.9 201006 GOO 1 3.5/3.8 201006 GOO 1 201001 GOO 5 ...(KA*S) i, o ANSWER 8.05 M)-
-c. 444-svr- C O.3 D - d e\t 4< d b. f al se CO.53 l c. false CO.53 !
REFERFNCE Hope Creek Emergency Classification Guide. Att. 3 K/A's... 294001 A1.16 2.9/4.7 294001A116 ...(KA'S) I
Fe. ADMINISTHATIVE, PROCEDURES t ,CQNQlllQNS._AND,LIMllA]IQNS
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ANSWERS -- HOPE CRI.EA -88/04/12-
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ANSWER 8.06 (3.00) c. Per (section 4. 0. 2 of 3 Tech Specs the l i mi t may be extended 25'/. C1.03 as lono as a three consecutive interval periods do not exceed 3.25 times the interval period C1.03 b. -The chance does not alter the intent of the procedure-One of the two staff reviewers holds an SRO license C2 & O.5 ca)
REFERENCE Hope Creek Tech Specs 4.0.2 and 6.8.3
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K/A's... 294001 A1.01 2.9/3.4 , 294001A101 ...(KA'S) ANSWER 8.07 (3.00) Yoo. power increase to 100*/. is allowable. CTS 3.0.4 does not aopiv3 HPC1 to now inoperable, oower coerations may continue for fourteen C143 days provided that RCIC. ADS. CS and LPCI prove operable per TS 3.5.1. [1.03 Lookino at the Primary Containment Isolation Valve. TS 3.6.3 states that all i solation valves listed in TS Table 3.6.3-1 be coerableC1.03. except os otven in TS 3.6.3.a.4 which is met by closino at least one i sol at i on velve in the line within four hours. C1.03 REFERENCE : HC Tech Spec 3.5.1 and 3.6.3 j l K/A's... 206000 GOO 6 3.0/3.9 206000 G011 3.7/4.4 206000G011 206000 GOO 6 ...(KA'S) < ANSWER 8.08 (2.00) ! o. YELLOW b. LIGHT BLUE c. RED d. WHITE C4 & O.5 ea3 l
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REFERENCE SA-AP.ZZ-015(Q) Tagging Procedures l K/A's... 294001 K1.02 3.9/4.5 l 294001K102 ...(KA*S) l l l
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ANSWERS -- HOPE CREEK -80/04/12-
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ANSWER O.09 omM i Wh (3.00) According to T.S. Table 3.3.2-1 the RCIC isolation valves must be shut (1.03 and RCIC declared inoperable C1.03. With RCIC inoo. the ol ant can continue to operate for 14 days (1.03.
[er C: 0. fat'c # erd sc he 5,take w k * b M **^dN REFERENCE HC Tech Soecs Section 3.3 K/A's... 217000 G011 3.4/4.3 217000 GOO 5 3.3/4.3 217000G011 217000 GOO 5 ...(KA*S) ANSWER 8.10 (3.00) b%ik bb0 (pol c. None of the action statements listed in TS 3.7.1.2.a apoly. TS 3.0.3 must therefore be aantied.CO.53 The unit must be claced in at least STARTUP within the next 6 hours.
' HOT SHUTDCWN within the f ollowino 6 hours. and COLD SHUTDOWN within the f ollowino 24 hours [A 53 li.
b. Accordino to TS 3.7.1.2.0. the associated SACS system must be declared inoperable CO.53 and tSo actions recuired by TS 3.7.1.1 must be taken.
CO.53 REFERENCE Hope Creek Tech Specs 3.0.3. 3.7.1.1, 3.7.1.2 K/A's... 205000 GOO 5 3.1/3.9 205000 GOO 6 2.5/3.7 l 205000 GOO 6 205000 GOO 5 ...(KA*S)
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ANSWER B.11 (1.00) I t% No (C.53. the reactor may exceed rated thermal power (3293 MWt] by 44 for I u; tc .: 5:r . 322^ "9t in 1 :: ther ; */. e t, s e . . i. - U u,m ,, , . i w- eu e ' cret! - d::: mrt ex.:t.'^.52 A3 g g We age Mc M po,se.,r- w cm l REFERENCE ShiC t- m .et eccec( mkd Wer %\ [#. 'O l OP-!O.ZZ 103, Cold Shutdown to Rated Power K/A's... 292OO2G.5 3.3/4.1 292OO2GO5 ...(KA'S) ,
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Pvtac Sevce E ectnc and Gas Company P O. Boi 236 Hancocks Bhdge. New Jersey 06038 l l Nuclear Depadment April 19, 1988 l l l Mr. David Lange i Operations Branch I Division of Reactor Safety U.S. Nuclear Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19406 I
Dear Mr. Lange:
EXAMINATION REVIEW COMMENTS - HOPE CREEK LICENSE EXAMINATION l Attached please find comments concerning the written examinations i administered on April 13, 1988. These comments have been discussed with the lead examiner and other members of the NRC examination team during a two-hour examination review session conducted immediately after the written examination had been completed.
i The following format has been used to document specific comments: l a, listing of NRC question, answer and reference; i b. facility comment including a recommendation for resolution; i and c. support documentation.
These comments are presented in the same order as originally numbered on the RO and SRO examinations.
Tnc En+ re; P eopic M . i 6 a t u 5 0 ;-
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Mr. David Lange -2- 4/19/88 .
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If you have any questions, comments, or need additional infor-mation, please call L. Catalfomo, (609) 339-3810 or W. Gott, (609) 339-3769. They will provide the requested information or will see that you are contacted by the appropriate person, h'
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0"Y Stanley LaBruna General Manager - Hope Creek Operations Attachment C Mr. C. Grotton USNRC Washington, D.C. 20555 , 1 - White Flint North Mail Stop 10-D-18 Mr. D. Moon Battelle Pacific Northwest Laboratories 3160 George Washington Way Sigma 3 Building 3000 Area P.O. Box 999 Richland, Washington 99352 I l
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ATTACHMENT
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A. QUESTION 1.03 (3.00) The reactor is critical with the MINIMUM permissible "Startup stable from
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period as allowed by procedure OP-10.ZZ-003(Q) loop suction Cold Standby to Rated Power." Reactor recirc temperature is 140 degrees F.
WHAT is doubling time if period remains constant? (1.0) a.
b. If IRMs currently indicate 50 on range 2, HOW LONG will it take for power to reach the point of adding heat if
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period remains constant and heating power is given to (1.5) be 40 on range 8 of the IRMs? (SHOWyourwork.)
c. STATE HOW period will be affected (INCREASE /0ECREASE/ REMAIN THE SAME) after the point of adding heat has (0.5) been reached.
ANSWER 1.03 (3.00) From OP-IO.ZZ-003(Q), period equals 60 seconds [+0.5]. Thus a.
doubling time equals 60/1.44 = 41.7 seconds [+0.5]. , b. 50 on range 2 is equal to 0.05 on range 8 [+0.5] P(0) = 0.05; P(t) = 40 P(t) = P(0)e**(t/ period [+0.25] 40 = 0.05 e"(t/60sec) ) t = 60 in'(40/0.5) = 401 see or 6 min 41 soc [+0.75] c. (Periodwill) increase [+0.5] l
; REFERENCE , 1. Hope Creek: LP RHPH24 LO #1a and fic, i KA 292003K108 292003K109 292008K113 a
B. If the candidate lists an incorrect value for period, credit should be received for calculations if they are completed correctly using an incorrect value for period.
C. No reference required. , d .I i l Page 1 of 41
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A. QUESTION 1.07 (2.50)
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Concerning Net Positive Suction Head (NPSH) with regard to the Reactor Water Cleanup (RWCU) pumps: a. PROVIDE a brief definition of AVAILABLE NPSH. (0.5) b. Given: (1) both RWCU pumps are initially in service (2) both RWCU filter demineralizers are in service with flow control in automatic, (3) the RWCU filter demineralizer bypass valve (BG-HV-F044) is fully closed, (4) the RWCU heat exchanger bypass valve (BG-HV-F104) is fully closed.
STATE how the AVAILABLE NPSH is affected (INCREASES, DECREASES, or REMAINS THE SAME) by each of the following separate events: 1. Reactor water level decreases from normal level to just above the low level scram setpoint (no change in feedwater flow). (0.5) 2. Reactor wat.er temperature decreases. (0,5) 3. One of two RWCU pumps trip (no operator action). (0.5) 4. The operator opens the RWCU blowdown flow control valve (BG-HV-F033) to increase blowdown flow (no other operator action taken; reactor water level remains constant). (0.5)
.
ANSWER 1.07 (2.50) a. the difference between the pressure at the eye of the pump and saturation pressure [+0.5) b. 1. decreases l+0.5; 2. increases + 3. increases 0.5 4. l+0.5l [+0.5] remains the same REFERENCE 1. Hope Creek: LP HTFF-42-01 LO #3.1.
2. Hope Creek: LP HTFF-43-01 LO #3.
KA 293006K108 293006K110 ) Page 2 of 41
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B. An alternato answer should be "the difference between the actual i and saturation pressures at the suction of the pump" sinco this - is where we measure pressure and temperature.
C. See next page for supporting documentation.
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Page 3 of 41
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LESSON NAME: PUMPS 302-145.03-HTFF-48-01
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INSTRUCTIONAL CONTENT:
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KEY / AIDS C. Centrifugal Pump Terminology 1. Net; positive suction head
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a. The., absolute. pressure,_above.the vaporspresot'reeofathe. fluid pumped, lvaiV rs , %t'sthespiamphuetion klar'.. 7 o ve.and accelerate _the .
;
Obj. 3.1 flui e ,ciaa the impeller. , b. NPSI :, it tsat c. Factsrs that /tet net positive iI suction Mcad 1) Vapor pressure of the fluid. ! 2) Temperature of the fluid.
3) Speed of the pump impeller.
d. Available NPSF. must be > required NFSH.
2. Cavitati:n - the fe:matien and obj. 3.2 su= sequent c:llapse of vapor bubbles al:ng :ne pump impelle: vanes due to inadequate SPSH.
u. .
. .r .e . . 2. ,. 9.-6 e- 3 a. Pressure fr p :hrough the impeller eye causes :he f;uid :: vaporize.
, b. ?: essure increase a: the impeller
:::e: edges causes them to ]
c:11 apse.
3. Gas 3tnfing
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2. Des::ibes a c:ndi:ica vnen a gas ! Ob), 3.3 su:n as at gets into ene pump ana l re:arfs fluid flew. l
!I b. Pumps are equipped with constant ;
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ven:s and.o: manaal ven:s to
;:even: :nis f:;m cccur :ng.
i l 302:'0 . OAT r. : 10'10e96
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Page 4 of 41 ggy, ; i l
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A. QUESTION 1.09 (1.50) a. STATE why reactor vessel heatup and cooldown rates are - administratively limited. (1.0)
, ! b. STATE the Technical Specifications heatup and cooldown rate limits for Hope Creek Generating Station. (0.5)
ANSWER 1.09 (1.50) a. (Heatup and cooldown rates) limit the magnitude of the cy;lic stresses placed on the vessel and compenents [+0.5) to maintain reactor coolant system integrity (or prevent failure ' of the reactor pressure boundary) [+0.5]. .
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b. 100 deg F/hr (heatep and cooldown) [+0.5]
REFERENCE
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1. Hope Creek Technical Specifications, Basis 3/4.4.6.
- KA 293010K104
i B. The answer fvr part "a" is divided into two parts: e limit the magnitude of cyclic stresses.
! l e maintain reactor coolant system integrity.
As stated, the question does not necesr rily evoke both responses i, The purpose of limiting the cyclic otresses is to maintain RTV
; integrity. Limiting the cyclic stresses should be a sufficient ; response for full credit.
C. No reference required.
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Page 5 of 41 i
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A. QUESTION 1.11 (1.50) Concerning core thermal limits: a. STATE the core thermal limit that is established to ensure that peak fuel clad temperature at,any location in the core will not exceed 2200 degrees F during a (0.5) ; loss of coolant a'ccident.
b. STATE two (2) adverse affects to the fuel that could occur if clad temperature exceeded 2200 degrees F in a loss of coolant accident. (1.0) i i ANSWER 1.11 (1.50) a. average planer linear heat generation rate (APLHGR) limit l
[+0.5]
b. 1. gross cladding failure [+0.5] 2. hydrogen gas generation (Alternate Answer: zirconium-water reaction [+0.5]) REFERENCE l Hope Creek: Heat Transfer and Thermal Limits, Section 16. ; 1. < KA 293009K110 293009K111 223001K509 l , B. Consider giving crc.dit t'or part "a" if the student responds with MAPLHGR (Maximum Average Planer Linear Heat Generation Rate) since this is the primary item that is checked on the l Pl edit relative to APLHGR.
C. See next two pages for supporting documentation.
Page 6 of 41
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C. MAPLHGR Limit Determination A saximus value of APLHGR that results;in.a.22000 F. clad. temperature during a LOCA is calculated for each fuel-type.in'the core. The peak cladding temperature is a function of such parameters as: Pellet-clad conductance Fuel rod internal pressures Pellet dimansional changes . Clad ductility The above parameters vary as a function of exposure. In some cases the
;
MAPLHGR limit is based on peak cladding temperature, but 8 x 8, 8 x 8R, and P 8 x 8R fuel clad would not reach 2?00 0 F even if operating at 13.4 Kw/ft. In these cases the limit is determined by the LHGR Limit and the
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local peaking factor applie nle to that node. The HCGS Technical Specification MAPLHGR is the LHCR of the highest powered rod in a node, dividet ty its local peaking factor. A typical MAPLHGR curve is shown in Figure 16.2 for two different fuel types.
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12.5-12 - nieuca (k=/ft) e
- 12.5 - -
it 10.5 - 4 , , , , , 5 10 is to: 25 33 AvtRAct P(ANAR (IP01VRE (t%J/t) vage 8 of 41 _, _ _ _ ._. _ ._ -- . . ___
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A. QUESTION 2.05 (2.00) I a. The reactor is at 100% core thermal power and both reactor auxiliaries cooling system (RACS) pumps trip on low i expansion tank level. If expansion tank level cannot be ' restored, LIST four (4) REACTOR BUILDING loads which will ; require operater attention to prevent automatic protective trips or possible component damage. (1.0) b. In the event a reactor auxiliaries cooling water heat exchanger develops a tube leak, DESCRIBE how the RACS expansion tank level should respond, considering system design. INCLUDE in your description the reason WHY the level would respond this way and the REASON for this design feature. (1.0)
:
ANSWER 2.05 (2.00) a. 1. RWCU pumps [+0.25] 2. RWCU filter demineralizers 3. reactor recire pumps [+0.25] [+0.25] j 4. CRD pumps [+0.25] b. RACS expansion tank level should increase [+0.3]. (The reason i ' why is that) Service Water system pressure is maintained greater thar. PACS pressure and so leakage through a heat exchanger tube should be from the Service Water system to RACS
[+0.4]. The design is intended to encure that any heat exchanger tube leakage is towards the potentially contaminated system, RACS [+0.3].
REFERENCE 1. Hope Creek: LP 302HC-000.00-081-03 LO #4 and #7. l KA 201001K106 202001K107 j l
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B. The question asks for "Reactor Building loads...." and the key includes RWCU filter demineralizers. The filter demineralizers are not loads, but the non-regenerative heat exchangers (NRHX) are. NRHX outlet temperatures are maintained less than 140 to protect the filter demins. The key should allow the NRHXs to be listed in lieu of the filter demineralizers for part "a.2".
C. No reference required.
Page 9 of 41
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. A. QUESTION 2.06 (3.00) Some Safety Relief Valves (SRVs) utilize two air pilot solenoid valves. Concerning a SRV with two pilot solenoid valves: a. STATE how many solenoids must be energized to cause the SRV to open electrically. (0.5) b. ARE any of these SRVs one of the SRVs controlled by Low-Low Set logic? (0.5) c. LIST the locations OUTSIDE the control room where a provision exists for an operator to manually open at least one SRV. (1.0) ; DESCRIBE how the capacity of an SRV accumulator to open an
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d.
SRV is affected -(INCREASES / DECREASES / REMAINS THE SAME) if primary containment instrument gas is isolated and drywell pressure INCREASES. PROVIDE a brief explanation for your answer. (1.0)
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ANSWER 2.06 (3.00) a. one [+0.5] b. no [+0.5] c. 1. remote shutdown panel (10C399) [+0.5] 2. lower relay room (10C631) [+0.5] d. The capacity decreases [+0.5]. As drywell pressure increases, accumulator pressure relative to the drywell decreases, and it is this difference in pressure that is the driving force for the SRV diaphram operator [+0.5]. REFERENCE 1. Hope Creek: 302HC-000.00-046-03 LO #5 and #8.
rA 239002K109 239002K405 239002K408 B. The wording of the question is confusing, and it can be answered two different ways. Part c. pertains to "at least one SRV" and the answer given in the Key is correct. However, the lead-in to the question says, "Concerning a SRV with two pilot solenoid valves" and this pertains only to the ADS SRVs (A, B, C, D& E).
In this case, the answer to part c.is the lower relay room only (SRVs A&E). Because of the question wording, full credit should be given if only the lower relay room is listed.
C. See next two pages for suppcrting documentation.
Page 10 of 41 _ _ _ - . . _ __
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3 - LESSON NAME: MAIN STEAM SYSTEM 302HC-000.00-046-03
' '- INSTRUCTIONAL CONTENT:
KEY / AIDS 5) Capacity - 8.18x10 5 t 9.5x105 lbm/hr saturated steam (depending en set pressure) c. Instrumentation and Control M-41-1 Sht 2 Table 1 1) . ADS;SRV'ss(A-E) Obj 5.b a) fach2 ADS.SRV,-hasitwof(2)isolenoids i associated;withidt. Energi:ing any - I one (1) of the two (2) solenoids will cause that respective SRV to , cpen (solenoid A or B).
Fig. 15 o See the ADS Lesson Plan for
" .ig. 3 specific system actuation signals and interlocks.
b) Each ADS SRV has two (2) sets of controls on panel 10C650C, one set of controls for each solenoid. , c) The control switch is an AUTO-OPEN maintained centact control switch ( p.b.): o AUTO - the respective scienoid will de-energi:e and will respond to the ADS icgic commands.
o OPEN - the respective sciencid will energi:e and the SRV will open.
Cbj.5d d) ,In_a'dditi'on,f t.ie :A and .E SRV's have
; AUTO-OPEN.keylockcswitches in the ilower relay room on panel Hil-P631.
The key is remcvaole in the ACTO position only: o AUTO - de-energite the respective "B" solenoid and allcws it to rescond to the ADS logic ccmmands.
302EC:10 DATE: 08/12.'97 Page 11 of 41 EEV : 3 _ __ _
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_ -_ _ LESSON NAME: MAIN STEAM SYSTEM 302HC-000.00-046-03 __
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~ INSTRUCTIONAL CONTENT: ' 1-
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KEY / AIDS l l
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o OPEN - energizes the respective "B" solenoid and l the SRV will open.
All "A" ADS solenoids are powered Obj Sc e) from 125VDC Bus B (lBD417) while i the "B" ADS solenoids are powered j l from 125VDC Bus D (lDD417).
, l j 2) Non2 ADSMSRVisMF,GiJkM7MiR)
!
i ! a) Each non-ADS SRV has one (1) ' ' solenoid asscciated with it.
Energizing this solenoid will cause * that respective SRV to open (solenoid A).
b) Each non-ADS SRV has an OPEN-CLOSE il
, 1 Fig 15 maintained contact control' switch !
Fig 3 (p.b.) on panel 10C650C: Fig 7 I o OPEN - respective "A" solenoid I energizes and the SRV will open. i o CLOSE - respective "A" solenoid de-energizes and the SRV will close. t
!
c) :The.following75RV's<can be operated ' Obj 5d .from the-remote shutdown panel (C399): f e ,F013Fi(non-ADS;SRV), . e (F013El(non-ADS 7 Low-Low Set
,$RV ) -
0 D13&Td3hyAD'S35RV) NOTE: If the RSP-C399 Cnannel B transfer switch is placed in l EMERGENC't, the F,3 5 M SRV's bec0me inoperable fr m the main centrol room.
i
- DATE: 08/12/97 302HC:10 REV.: 3 Page 12 of 41
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M vn i l l l A. QUESTION 3.01 (2.00) l I a. WHY shouldn't the RCIC turbine be allowed to run at ' speeds below 2500 rpm? (1.0) b. If RCIC had automatically initiated and then tripped on mechanical overspeed, WOULO it automatically inject if level fell again to "-45", WOULD it automatically begin injecting? If not, WHAT action must be taken to achieve RCIC injection? (1.0) l
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ANSWER 3.01 (2.00) ! a. Because the RCIC lube oil pump is shaft driven [+0.5], low I turbine speeds could cause bearing damage [+0.5]. ( b. No [+0.25], before the turbine trip throttle valve can be l reset to allow RCIC injection, the mechanical overspeed I trip linkage must first be reset at the RCIC turbine l
[+0.75].
REFERENCE 1. Hope Creek: LP 302HC-000.00-030-01 LO #6.
KA 217000K404 217000A506 . l l
l B. Since transmittal of original reference material, Lesson Plan 302HC-000.00-030 has been revised to include a discussion of NRC lE Notice 82-26. This document discusses damage to the RCIC Exhaust line check valve and included guidance to minimize operation below the recommended turbine rated speed. For this reason, a response that includes discussion of RCIC exhaust line check valve cycling or chattering or exhaust line water hammer should not be counted as incorrect and should be considered for credit.
C. See next page for supporting documentation.
Page 13 of 41
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LESSON NAME: REACTOR CORE ISOLATION COOLING SYSTEM
INSTRUCTIONAL CONTENT:
l I KEY / AIDS or down. The amount of pilot' valve plunger l movement is proportional to the EG-M control box ' l signal. This plunger movement varies the hydraulic pressure to the remote servo.
A spring loaded relief valve within this oil !
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system maintains the operating actuator hydraulic oil pressure at approximately 325 psig.
g. Remote Servo - The remote servo contains a piston connected to the turbine governor valve by linkage. As the hydraulic pressure from the EG-M actuator varies, the piston is positioned within the remote servo, and the linkage positions the i turbine governor valve to vary turbine speed.
' C. Flowcaths M-49-1 1. RC:C Steam Supply , M-50-1 Ooj. 13a a. Steam is supplied to the RC:C turbine frcm the
"A" Main Steam line ,(before the MS:V's) via the folicwing RCIC components:
1) FE 4155 2) HV F007 3) HV F008 4) HV F045 5) HV 4282 6) TV 4283 M-49-1 Obj. 13b 2. RCIC Steam Supp'y Line Drains
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a. C ndensate f:: ed in -he RC: stea. supply '_ine wnile the RC:C Sys en is :n standby f;;ws : :ne main condenser v;a :he fol'.:w;ng :: penen s: 1) Drain PO-2) 'V F025 2) HV FO:6 4) ST 0003 j
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5) LV F054 Cbj. 13c
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' Discuss 3. RC:C turbine exhaus: fl0ws to ta.e :rus {
'IE 6IN suppressi:n pocl via :he RC:C urcine 1 l
No. 82-26 exhaus checx valve (FC 7003) * hen va'.ve
. MV F059. ! ;
M-50-1
' 4. Turbine and Va've *eakoff
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Obj. 13d
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i Page 14 of 41 Date: 18 /1 - '8'
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JN 82-26
' July"22, 198?.
Page 3 of 4
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impac on the turbine exhaust back pressure thereby causine the turoine to trip. All camaged components of the RCIC exhaust check' valves will be replaced and tne valves will be re:uilt to assure prcperly conditionec anc wo-king valves are installed.
Discussien: All of the above failures deal only witn the RCIC5Urbine exhaustjcheck 29alve. ! However, as noted by Georgia Power C:moany, the HPCI exhaust system nas :ne j same type valve in a similiar system configuration, inus it is reasonacle to
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expec: similiar problems with tne HPCI tur:ine exnaust eneck valve also. In f act, both services have been icentified in :ne generic correspondence by l General Electric pertaining to this :ccic.
I The first of :ne 9eneric corre.s.Po.nde.n.ce is Servic.es.,Information Letter (SIL)
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No. 30, "HPCI/RCIC Turbin_e Exhaust L~ine Vacu..um Breake.rs," catec October
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31, 19U. In :nis SIL, General Electric icentifiec :ne preolen of ;cssiele i can: age to :ne exnaus: line : neck valve ar.c rec:mtrenced the installa:icn of , vacuum breakers baseo en tests c:ndu:*ec at Br:wns Ferry enc Peach Sott m. ) The se::nc or :ne generic corres:oncence is Acclication Information Occument !
( AD) No. 56, "Hign Pressure Core Injecten anc Reac;or Core Isola icn C: ling l Turbine Exhaus: Check Valve Cycling," ca:e: Decem:er 13, '.;S1. In :nis AI:, i General E;ec r c icen fiec : .e :cssi:li :auses of f ailure as imorc:er system j
:: erat;en, im re:er enec< valve s1:1cg, inacecuate enec< valve cesign, er l teacecwa:e exnaus line cesign. To-minimize:tne possibility.cf future proolems, I
'
:ney rec:meno that:
'.. .janual starts and mentnly sys:en s;rveillance :esting sncule ce :er- l ccmec in ac: rcance wi:n :ne C:ern:rg snc Maintenarce Instrue:icns l
,sce:1'ically, gracually increas ng :ne tur:ine s;eec un:i' the ra:ec j
:e: iscrar;e fica * s ecnievec s not rec:r rencec). 1 1. e exnaus :necx /aive. :ne etra s; 1:ne vacuum Oreake , an: :ne exraas:
're s;a ;er 3. aIC e esigne: n 3:: r ance wi*.n ne reca; emen*,s/
'e::.-'*ecc:*.ier.5 3 : v e a. *n *,ne .;E
*
js'.a?. :eti;n 3:e jf1:a;;;c.
3. System coerat un Delow ttheirec;mencecc_ uroi._ne_ :r.ated:sceec,
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snoui.c...te
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tm101m12eo. l
. Se exnaus*. cre < /aI/O i h u e '. :e I;;a".ec as Ol:50 as :csi: I e *. *ne
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c;n.31rmen*,
- 'he ur tae es*aus* ;r.tc< valve ' ~eals 3* UI Oe /1suaI'y ins ec*e:
On a c w :"e icne:s.: swer as !*. ise 'j if kel'n; Ou* age, Page 15 of 41 _ _ _ __ - . _ - ,_ _ _ , _
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. A. QUESTION 3.05 (2.50) l The reactor is operatinv at 95% of rated core thermal power.
i'aed water control is lii three-element control with RPV water j level controlling at the 35" setpoint of the controller, a. LIST three (3) process variable signal inputs (i.e., 1 the "three" elements) for the water level controller. ) INDICATE WHICH of the three is the predominant input relative to the cuput of the controller. (1.25) b. DESCRIBE hoa interlocks in the feed water control system would respond if the "A" primary c-ondensate pump tripped.
INCLUDE setpoints. (0.75) l l c. If a low reactor water level scram occurs, given that the feed water system is :apable of feeding the RPV > af ter the scram, S~i;TE the RPV water level at which the feed water control system will attempt to maintain. (0,5) f L l i l ANSWER 3.05 (2.50) a. Reactor water level [+0.25), total feed flow [+0.25), total steam flow [+0.25] (are the three inputs). Reactor water level is predominant [+0.5]. b. feed water flow will be limited [+0.5) to 70% [+0.25] l c. +18" (RWLC setdown setpoint) [+0. 5] j REFERENCE 1. Hope Creek: LP 302HC-000.00-059-02 L0 #3, #4, and #7. l 2. Hope Creek: LP 302HC-000.00-020-05 L0 #9.
l 3. Hope Creek: LP 302HC-000.00-058-02 LO fil.
KA 256000K304 256000K102 259002K102 259000K103 259002K104 259000K404 B. The Hope Creek level control system does not use total feed flow, it uses individual feed pump flows. The system is also not level dominant, it is individual feed flow dominant.
The Key should read "Reactor water level, individual feed pump flow and total steam flow with the individual feed pump being dominant".
C. See next two pages for supporting docunentation.
Paco 16 of 41
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REACTOR WATER LEVEL CONTROL 302HC-000.00-059-02 LESSON NAME: _ , . h INSTRUCTIONAL CONTENT: _
I KEYS / AIDS ' 2. Controls either: 1
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M-31-1 SET 1 a. RFPT governor i M-06-1 l b. Start-up valve position l
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$ 1) S/U valve (LV-1754) opens f CF first (3" valve) f i
S/U valve LV-1785 opens after
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Od#
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2) the LV-1754 S/U valve 3) LV-1785 is the 12" S/U level control valve DCP 5030 3. Automatic control will normally be in single element on the startup valve during low power operation and three element control during high power so peration
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4. Single element means that vessel water level is monitored as the controlli'ng
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OBJ.3 - parameter a. Used as control parameter normally during ic power operation b. Vessel level changes are slower i l during low power operation c.
Control signal comes from one of M-42-1 SET 1 two Harrow Range Rosemount Level
, Detectors (PDT N004 A or B via selector !
switch) . ikdardul
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5.
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'l)- g :57
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OBJ 3 (Three element means thatL(flow 4_(2),totalwaa tfeedwater:
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iflow,tand?(3)Kreactorisesselaw ater.ilevel signals are used as controlling parameters a. The same reactor water level signal is used as in single element DATE: 03/17/86 300EC-34:10 Page 17 of 41 REV. 2
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'0- REACTOR WATER LEVEL CONTROL 302HC-000.00-059-02 I LESSON NAME: ^
l --
.c m'
INSTRUCTIONAL CONTENT: A\ / v v
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. l %" * 6. -roe snomount, trea Rotu roR reca MFtr utanp,riar m
,,4, KEYS / AIDS lv , gu7p , , ,,, , y ,a n , , ,, m , ,,,, , ,
* 'j ' ,i Flota s/cHAs. Auo rge ersurnd EMM9 3rsnm cenrnus W 70t21 I;25"E " 2 " ~ " " * " r*
A w cett1 M-06-1 ! M inil:isu:1 f;;d 7""
$;y E dell ;. :ed ? ad C1 c. Total main steam flow is a M-41-1 SBT 1 summation of individual. main steam line flow rates as measuredat the . main steam line flow restrictions.
- (PDT N003A, B, C and D) w n w sou x, _ and feed , d. rue rarm. sn f5 THC ff70h k a.
. , , . .,. .~ ~w- . . . _.. wwwam u <- Mcy 3 p s. _ r
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:::::: ;; ;; :;;;: 1. s_ ng t<MICil Af/463C N ch-ey: 5 '
Ornnso, nni * A v r A n / o n s'*o f M1anent qui;h;; aGting-3 Dss f>N C Y Gf IU VA burs Creewan ora rur nu,we d o goresArius /d e. Three element is only (master used atlevel higher DU A MO, . power levelsJig?# W . control)
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1) At low power levels both
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feed, and steam flo,w sign.als are less accurate and~more unstable
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R Aeove AttpoxlM ATriY 201o5::::: s to weg 2) nnc= t.eee :_;n:1 Ot:51e (a? determine du Meg e**rt-up te: ting; 3 element feedwater control will be used r -s f.
Steam flow, feed flow and vessel level are used in 3 element control
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i OBJ 7 ') N' A Aw.e.
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a A R. m 2)
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On a loss of.1peatet feed flow
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p g;.4M signal feedwater control will w pou; initially increase and level M * Y' W :b W wL11 increase TM4 - Vassel /*'d a) Should receive a high level alarm (Lev 7 at 39") DATE: 03/17/86 300HC-34:10 Page 18 of 41 REV.t 2 _ . . . _ _ _ _ - . . _ _ _ _
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A. QUESTION 4.04 (1.00) The reactor is operating with the reactor mode switch in "RUN."
The "RPV PRESSURE HIGH" annunciator is received. Reactor power and total core flow are still at rated. You then check reactor pressure indication, both by the recorder and by the control room instrument display system (CRIDS). BOTH indica?.e reactor pressure is 1080 psig. SPECIFY ALL emergency operating procedures (EOPs) for which an entry condition exists. '
(1.0)
ANSWER 4.04 (1.00)
/,O OP-E0.ZZ-101(Q) (RPV control) [+Dr5]g,vup,/:0 REFERENCE i
1. Nope Creek: OP-E0.ZZ-101(Q).
KA 295025SG10 295025SG11 l I
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B. Since a scram condition does exist (>1037 psig) an entry condition for OP-EO.ZZ-100, Rx scram, does exist, although you are led i back to OP-EO ZZ-101. Credit should not be taken off if the answer includes OP-EO.ZZ-100 along with OP-EO.ZZ-101.
C. No reference required.
I I Page 19 of al
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A. QUESTION 4.05 (2.25) , You are touring the reactor building and notice that the health physics department has posted a "High Radiation Area" sign outside of the HPCI pump room.
a. DEFINE a "High Radiation Area" in tems of the exposure you could receive in such an area. (0.75) b. ASSUME that you DO NOT have a completed NRC Fom 4 on file. STATE the quarterly WHOLE BODY exposure limits that have been established for you by: 1. the NRC (0.5) ) 2. Hope Creek Generating Station (0.5) ) c. STATE which one of the following types of radiation would be responsible for the greatest portion of your exposure i if you were in the HPCI pump room during HPCI turbine I operation. (0.5) i
alpha j beta ' gama i neutron ANSWER 4.05 (2.25) a. (an area accessible to personnel where) a major portion of the body could receive in excess of 100 mrem in one hour [+0.75]
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b. 1. 1250 mrem (NRC) 0.5) l 1000 mrem (HCGS) [+[4.5)
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2.
c. (3) gama [+0.5) REFERENCE 1. Hope Creek: SA-AP.ZZ-024(Q).
KA 294001K103 B. The Key for part b. reflects the HCGS limit if a current quarterly estimate has been completed when an NRC form .I has not. However, if the carrent quarterly estimate has not been completed, then the HCGS limit is 300 mr/qtr. This second limit, 300 mr/qtr, should also be acceptabic for full credit.
C. See next two pages for supporting documentation.
, Page 20 of 41
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' SA- . .z3-024( O)
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FlulHE 2 * ' I AIMINISIRATIVE IEEE I.IMI15 AND IT112aSIO4 REOllIREMENIS EXCElTION APPROVAL TO APPLICAl'I( N ACrlON " IX1iE 11NEL EXCEED BY l i --
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_ _ _ _ ~ _ . _ _ Prows 3 TLD lakje. Limit cannot te 1000 mran/UrH InitiebOuarterly dose Ikny RCA access rosovat until NHC-4
l Ilmidfbr: individuals. alariro T1.D promss- is coipletel.
Caspl@-a2 current isvj arwt unti1 in-quartec estimate !
( 10ClR20.102a) creased <kr.se leve1 dpprOVed.
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_ . _ _ . . . . . . _ _ . . . _ _ _ . Ds tQ (D Cl) N > '
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i O O i i F i giar'e 2 ROV- .I
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S A ... 22-024( O) FIGIRE 2 i AININISI1tATIVE 110GE LIM 115 AND FXnNSI(M HFE)tilHtMENIS APPLICATHW ACTI(N EXCEL'rICN APimVAI. 'IU DO6E LEVEL ' EXCEFD llY _
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Mi'nidWnistrative Process TI.D ba<kje. I.imit is twrwed Qualif le<l Stat i<m Ita<liatisn 300 mren/tTTR ogum receipt of Protect ion gersmael.
limit for all personnel Deny access to HCA working in the protectal until increasal current quarter area. 1imit is appr(ned. estimate ( 10CFH20.102a) or c< stylet al NRC-4 fonn.
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m .__ cn
$ 1000 mrun/UrR Ouarterl_ dose limit for No access to HCA RCA entry
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* irxtividuale less tlian 19 allowed. restricted untit y
y tut at least ,18 years old next <1aarter. - with a corplete NRC-4 ca o fonn.
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NOTE: An irtlividuhl approaching (within 100 mrun) a dose limit shall le rest rictol f r<sn ent er inu) liiqh Padiati<m Areas and may te restricted f r<an enterisw) tim IN'A.
Visitors shall have a dose limit of 300 mron/UrR. Extension ag.ptovals via the Fitysre 2 aat twiri7at ions shall lie obtained to exceed that limit.
I l Figune 2
* l * V - I S A- A P. 2 2-02 4( O)
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_ _ , _ - ,- . l A. QUESTION 4.08 (2.00) The Technical Specifications specify the maximum allowable mismatch between reactor recirculation pump speeds. STATE the BASIS for administratively controlling this mismatch in speeds.
(2.0) ANSWER 4.08 (2.00) The limits ensure an adequate core flow coastdown [+0.5] from eitherrecirculationloopfollowingaLOCA[+0.5]. 1 l
(Alternate Answer: to comply with ECCS LOCA design criteria
[+1.0])
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REFERENCE 1 1. Hope Creek: Technical Specifications, Basis, 3/4.4.1. I j KA 211000SGS i B. The point value for this question is 2.0 points. In the Key, the answer is given in two parts, each worth 0.5 points for a total of 1.0 points. The alternate answer is given a value of 1.0 points. Since the alternate answer should be accepted l in lieu of the primary answer, either the total point value of this question must be changed or the value of the acceptable responses must be changed.
The reference KA item, 211000SG6, addresses Standby Liquid Control System and, therefore, does not pertain to the question or response.
RO training does not include universal instruction on all Technical Specification Basis. The references cited does not identify material where this item is covered in the RO training program.
Since there is no explicit documented link between KA and any learning objectives of the RO training program, it is recommended that this question be deleted.
C. None Page 23 of 41
l A. QUESTION 4.10 (2.75) l ConcerningprocedureOP-AB.ZZ-145(Q),"StatorCoolingMalfunction": l
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a. LIST ALL automatic actions associated with a complete loss of stator water cooling. INCLUDE setpoints. (2.0) b. As a subsequent operator action, the operator is directed to "reduce generator vars to zero." STATE the purposo for this action. (0.75) i ANSWER 4.10 (2.75) a. 1. turbine generator runback to 23.3% of rated [+0.4] 2. turbine trip j
, a. (in 2.0 minutes) if generator amps are still >70% of '
rated [+0.4] l b. (in 3.5 minutes) if generator amps are still >23.3% l of rated [+0.4] l 3. full recirc runback to 30% (#1 limiter) [+0.4] l l 4. recire pumps trip if generator amps are still >23.3% of rated (in 3.5 minutes) [+0.4] i l b. (the purpose is) to minimize heating of the generator stator
[+0.751 REFERENCE 1. Hope Creek: OP-AB.ZZ-145(Q).
2. Hope Creek: LP 302HC-000.00-114-00 L0 il and #3.
3. Hope Creek: LP 302HC-000.00-062-02 LO #4.
KA 241000K616 B. The answer is complete except that two of the numbers have changed.
The 23.3% is now 23.49% and the 70% is now 79.11%. The Answer Key should be changed to reflect the correct numbers.
C. See next two pagos for supporting documentation.
Page 24 of 41
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HOPE CREEK GENERATING STATION OP-SO.BB-002(Q) -<Rev. 9 REACTOR RECIRCULATION SYSTp.OPERAT mM _ __ . C0TTR0'_ . n. n.a y a ~ ha ! Remarks / Reason for Revision: Rev. 4 Incorporates changes made as a result of modifications.
Rev. 5 Incorporates Scorp Tube Positioner Lockup operation and.
Caution on critical speed.
Rev. 6 Incorporates an OSC addressing the Output A(B) select pushbuttons on SIC R620, and adds Recirc Pump Cr!tical Speed Ranges.
Rev. 7 Add procedures for operating with a single recirculation loop in service and adds procedure for isolating a recirculation pump due to multiple seal failures.
. Rev. 8 This revision was a disapproved on-the-spot change.
Rev. 9 This revision incorporates the method of disengaging and engaging the scoop tube positioner disc brakes installed under DCP 4-HM-0006 and adds isolation for recirculation MG vent dampers.
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!/,
Approved: '
~ [//
Op(rat 1ons Manager - HCO k[
'Date
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l OP-SO.BB-002(Q) Page 25 of 41 Rev. 9 _. .._. -_. ._ - _ - _ . ._ --_ _ __ - - . _.
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3.2.9 The following, conditions'fy11139auselarfull Recirc Pump rrunback%o464% peed:
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3.2.9.1 Reactor water level < level 3 (+12.5") 3.2.9.2 Total feedwater flow < 20% for > 15 seconds 3.2.9.3 Loss of 1 Circ Water Pump with 3 or less Cire Water Pumps running and Condenser pressure > 5.8" HgA 3.2.9.4 Loss of 1 Primary Condensate Pump with total feedwater flow > 75% 3.2.9.5* Loss ~of/StatorfWater' Cooling: a. Generator Stator Cooling Inlet Pressure low-low (13 psig) p_r_ b. Generator Stator Cooling Outlet Temperature high (82*C) p_n - c. Main Generator Stator Current to
*
Stator flow 15% mismatch: STATOR REQUIRED ALARM RUNBACK AMPS FLOW (FLOWi (FLOW) 30,022 (100%) 661 gpm 595 gpm 562 gpm 27,020 (90%) 535 gpm 482 gpm 455 gpm 24,018 (80%) 423 gpm 381 gpm 360 gpm 21,015 (60%) 324 gpm 292 gpm 275 gpm 18,013 (50%) 238 gpm 214 gpm 202 gpm 15,011 (50%) 165 gpm 149 gpm 140 gpm 12,009 (40%) 106 gpm 95 gpm 90 gpm 3.2.9.6 HVF031A(B) PUHF A(B) DISCH VLV < 90% open.
- Main Turbine will: trip'11 Main Generator. Stator Current is not iless than 79.11% (23,752 amps) in 2 minutes' n .23.49% (7,055
, amps) in 3 to 5 minutos.
Page 26 of 41 pey, 9 OP-SO.BB-002(Q)
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l A. QUESTION 5.01 (2.00) LIST and describe TWO conditions which will assure adeouate co're cooling. l C2.03 ANSWER 5.01 (2.00) l a. Maintain reactor vessel level. Maintainino level Cabove two-thirds l core active height) assures adecuate cooling and adecuate drivino l head for natural circulation. C1.03 ' b. Minimum of one LP ECCS pumo injecting water into the core C1.03 REFERENCE OP-EO.ZZ-102CO3 Containment Control Section 3.0 , 302HCOOO.OO-100-01 Objective 1.1.4 l K/A's... 205000 G010 3.2/3.3 205000G010 ...(KA'S) i B. OP-EO.ZZ-102, Containment Control Section 3.0, states that: Three viable mechanisms of adequate core cooling exist; they are - 1. Core Submergence 2. Spray Cooling l l 3. Steam Cooling Any two of these three should be co'nsidered as correct, as well f as the original response on the Answer Key. , l C. See next two pages for supporting documentation.
l Page 27 of 41 l l _ . _ _ - ,-_ l
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CONTAINMENT CONTROL l 1.0 PU RPOS E The purpose of this procedure is to provide cloe.r and concise instructions for the operators responding to an emergency or conditions that may degrade into an emergency. l The Primary Containment Control Procedure provides the means to maintain primary containment integrity and protects equipment in the primary containment.
This procedure is entered whenever suppression chamber water temperature, drywell temperature, drywel2 pressure, or suppression chamber water level is above its hi.gh operating' I limit or below its low operating limit.
2.0 REFERENCES 2.1 BWROG EMERGENCY PROCEDURE GUIDELINES Rev. 3
2.2 HOPE CREEK GENERATING STATION FSAR l l 2.3 S A-AP. Z 2-0 0 2 ( O) , "Station Organization and Operating i Practices," Rev. 0
.
2.4 Closing Documents incorporated j C D- 0 96B (DW/T train and DW/P train) i C D- 6 21 D (step SC/L-2 9) ' CD-733A (SC/T train) CD-138X (DW/P train) CD-088X (SC/L train)
.
l CD-087X (SC/L train) CD-086X (SC/L train) CD-085X (SC/T train) CD-084X (SC/L train) C D-26 8 X (SC/T-3 ) CD-ISB X, Cou2/r. g 3.0 DEFINITIONS l-1 Interpretations, definitions, and discussions regarding the usage of key EOP words and phrases are provided below. This information is provided to promote a uniform understanding of the actions intended by the Cautions and steps in the EOPs.
3.1 Adequate Core' Cooling - defined to be heat removal from the reactor suf ficient to restore and maintain peak fuel cladding temperature (PCT) at or below 2200 *F.
OP-EO.22-102(Q) Page 28 of 41 Rev. O _ _ --.
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. .threehviable:mechanismssofi adequate core cooling exist; fin : orde r o f p re f e re n'ce ~,'. they' a re : ) .I1. . . Core Submergence 2..iSpray. cooling 3 ~.T58 team. Cooling Core submergence is the mechanism of core cooling l whereby each fuel element is completely covered with water. Indicated RPV water level at or above the top of the active fuel thus provides direct confirmatior. l that adequate core cooling exists. Assurance of l adequate core cooling is provided when, in the I judgement of the operator, RPV water level can be 1 maintained above the top of the active fuel. I Spray cooling is the mechanism of core cooling whereby each fuel element is sprayed with suf ficient flow to r.emove all heat generated in the fuel bundle.
If RPV water level cannot be maintained or confirmed : above tne top of the. active fuel, one core spray system l operating at or above design conditions (system flow, ; spray sparger dif ferential pressure, etc.) will provide l adequate core cooling by the spray cooling mechanism. l Assurance of adequate core cooling by sprap cooling is provided when, in the judgement of the operator, core I spray system operation at or above design conditions can be maintained.
. Steam cooling is the mechanism of core cooling whereby steam updraf t up through the uncovered portion of each fuel bundle is sufficient to remove all heat generated in the fuel bundle. A high fuel-to-steam differential temperature is required for steam cooling. Water in the lower core region and lower plenum is the source of the steam, and RPV pressure control via SRV operation is used to establish and maintain the required steam updraft through the core. The steam cooling mechanism of core cooling is employed only when there is no source of makeup to the RPV availablo.
3.2 Assure - make certain that a specified state or condition is established and will be maintained; encompasses an implied action to operate appropriate systems, as available, to accomplish the stated objective. Both direct and indirect indications may be used to determine that the specified state or condition has been achieved and will be maintained.
O P-SO . 2 2- 10 2 ( O ) Page 29 of 41 gey, o
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g, QUESTION 5.11 (3.00) The reactor is operating at 50% cower. HPCI initiates and injects. Compared to their steady state values at 50% power. state whether the following parameter will finitially and after 10 min] INCREASE. DECREASE OR REMAIN THE SAME. (Assume no operator action) a. INITIALLY (when injection occurs): ! 1. Reactor oower CO.3753 l 2. Reactor water level CO.3753 3. Total steam flow CO.37b3 1 4. Reactor oressure CO.3753 b. AFTER 10 MIN. OF HPCI OPERATION l
,
1. Reactor oower CO.3753 ) 2. Reactor water level CO.3753 3. Total steam flow CO.3753 4. Reactor oressure CO.3753
I ANSWER 5.11 (3.00) 4.1. decrease 2. increase l 3. decrease l
4 decrease
l b.1. increase 2. increase 3. decrease l l 4. decrease CB & .375 EA) REFERENCE l 302HC-OOO.OO-026-04 primary objectives 10. 12 l FSAR Section 15.5-1 j K/A's... 206000 A2.19 3.9/4.3 206000 K1.02 4.0/4.1 1 206000 K1.03 3.8/3.8 206000K103 206000K102 206000A217 ...(KA's) , l l l l Page 30 of 41 i
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However,
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B. The Answer Key has level increasing and remaining high.
the HCGS level control system is set up such that level will initially increase but will then return to the initial value * withi'n 10 minutes. The Answer Key should be changed to reflect this (i.e., change 5.ll.b.b.2 to remain the same or' return to initial value).
C. See Attachment to RO - Question 3.05.
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l l i l A. QUESTION 6.04 (2.50) Assume the reactor has just scrammed for an undetermined reason with a l core age near the beginning of the fuel cycle. With the mode swi'tch in 1 RUN and pressure decreasing, answer the following about the Nuclear l
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Steam Supp1v Shutoff CNSSS3 system isolationst a. WHY is the reactor isolated at 756 psig with the mode switch in RUN? l C1.OJ l l b. To reset the i sol at i on , LIST the two actions and/or checks that are ; required. Assume that the mode switch is no longer in 'RUN'. l C1.03 I c. What combinations of manual reset pushbuttons A-B-C-D must be j dooressed to cause a manual isolation? CO.53 l ANSWER 16.04 (2.50)
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a. The isolation stoos the reactor deoressurization and cooldown.CO.53 The setooint was chosen because assumotions made f or MCPR are not valid for all critical power calculations Cat oressures below 785 ) osio or core flows less than 1 0 */. r a t e d flow.3 CO.53 l b. -Reset the isolation logic by depressing the ' RESET * PB-Place all MSIV control switches in the ' CLOSED' cosition 1 C2 & O.5 ea3 i C 0 I c. Any combination of A OR 'h wi th C and 3 )(, OR D CO.53 REFERENCE 302HC-OOO.OO-045-02 Enabling Objective 9 and 10 I HC Tech Spec Bases Section 2.1.1 J K/A's... 223002 K1.01 3.8/3.9 223002 K4.06 3.4/3.5 223002 A1.02 3.7/3.7 223OO2K406 223OO2K101 223OO2A102 ...(KA's) ! l B. The Answer Key reflects a generic response, to item a, that is not addressed directly by HCGS Technical Specifications.
The reference cited, HC Tech Specs Basis 2.1.1, pertains to Thermal Power and would not apply to the conditions established by the question; i.e., the reactor has just scrammed. A response of either stops the reactor depressurization or stops the reactor ; cooldown should be accepted for (1.0).
Item c is improperly worded. The switches intended to be identified are apparently the MSIV Manual Initiation Pushbuttons A, B, C and D l and NOT the Reset Pushbuttons. The answer of None of the reset pushbuttons will cause a manual isolation should be accepted l for full credit, as well as accepting the original Answer Key response for full credit.
C. No additional support documentation can be provided for item a, as none exists.
Page 32 of 41 I
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i l A.GUESTION 6.07 (2.50) a. What is the puroose of the excess flow check valve in an RPV instrument line? C O,5 3 b. How would indicated RPV level be affected (INCREASE. DECREASE or NO CHANGE) if a leak d evel op ed in the ref erence leg? WHY7 (0.53 c. For the following level instruments and ranges. state all automatic action (s) which will occur, if any, from signals provided ) by these instrumentst l l 1. WIDE RANGE. LEVEL 2 1
2. UPSET RANGE. LEVEL 8 j 3. NARROW RANGE. LEVEL 3 C1.53 ANSWER 6.07 (2.50) a. The excess flow check valve is installed in an CPV i n s t r u r eci t line to isolate the line if a break occurs downttream of the check valve. CO.53 b. The leak would cause a decrease in reference leg Cdensity, l evel or oressure J. Indicated level would increase. CO.53 c. 1. NSSSS I sol a t i or. s . HPCI Initiation. PCIS/MSIV boundary leakage. 1
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RCIC Initiation. Reci re Pumo * '/ ARI 2. None 3. RPS Trio. NSSSS isolation S/D Cooling, ADS Permi swi ve. Recirc l Run Back C3 D O.5 ea) l l REFERENCE l 302HC-OOO.OO-02-01. Table 1 l Enabling Objective 2a.5 l K/A's... 261000 K5.07 3.6/3.8 216000 K3.12 3.5/3.6 1 216000 K1.02 3.8/4.0 216000 K1.14 3.8/4.1 l 216000 K1.07 3.9/4.1 l 26100C U17 216000K512 216000K114 216000K102 21600K107 ,
...(KA _, l B. The acceptable response to item c.1 should include RRCS initiation which includes Recirc Pump Trip and ARI.
C. Np additional reference provided; see 302HC-000.00-024-02 previouuly p;ovided.
Page 33 of 41
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_ . . _. .. 6.11 <2.00) A. QUESTION The reactor is at 100% cower when a complete loss of instament air cc ctar s. What is the first cause of a reactor scram and WHY7 CAssume no ' o .rator actions). C2.OJ ANSWER 6.11 (2.00) The reactor will scram on low level C1.03 because the Feedwater Pumo recire valves fail open upon a loss od instrument air.C1.03
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CEFERENCE 302HC-OOO.OO-75-OO Instrument Air 302HC-OOO.OO-58-OO Feedwater 302HC-OOO.OO-114-OO AB Procedures Objective 3a.b.
K/A's... 295019 AK2.03 3.2/3.3 259001 K6.01 3.0/3.0 295019AK2O 259001K601 ...(KA'S) I B. References cited do not support the response in the Answer Key.
There are a large number of variables implicit with this question and, therefore, multiple pos;-ible answers. Albeit, the HC Simulator cause and Effect Document does identify the ensuing reactor scram as being the result of RPV level 3 caused by Primary Condensate Pumps recirculation valves going full open on loss of instrument air, it must be understood that the simulator is not an engineering model. Nor are students trained to simulator specific effects for a given malfunction at a specific plant condition. To date, the plant has not experiencea a complete loss of instrument air; therefore no actual plant data has been used to model this malfunction. Air line supply sizes and valve activation pressures were used in the air system modeling. The results of the malfunction are what fell out of the model ar? represent 4 our best-guess at what may happen at the plant given a break at this specific location and power level. It should be noted that effects differ at icwer power levels.
It is recommended that either the question be removed from the exam or allow any reasonable response full credit.
C. No additional reference material required.
Page 34 of 41
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M 1 l O A. QUESTION 8.01 (2.00) a. Prior to authorizing restart of the reactor after a SCRAM. the Operations Manager and the SNSS r evi ew the Post Reactor Scram /ECCS Accuation Review Report. List THREE items checked durina this review.
C1.52 b. Who (by title) must ultimately give approval to r3 start after receiving CO.5J a startuo recommendation from the SORC? ANSWER 8.01 (2.00) a. -The recort is complete and Oroperiv prepared-The root cause has been crecerly identified-All ecui p- '- St functioned satisfactorily-Corrective a tions recuired to be ccmclete Orior to restart have been identified sd initiated Cany 3 C O.5 ea] b. General Manaoer CO.53 REFERENCE OP-AP.ZZ-101(Q) Post Scram / ECCS Actuation Review 302HC-OOO.OO-113-OO Learning Objectiva 3 OP-AP.ZZ-OO2 Learn:ng Objective 4 293OO9K121 293OO9K107 293OO9K100 293OO9K110 293OO9K112
...(KA*S)
B. Accept answer as stated in the Answer Key for full credit.
Do not remove credit for including: Operations Manager may grant approval for taking the reactor critical in the absence of the General Manager.
C. See next page for supporting documentation.
Page 35 of 41
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j '. OP-AP.22-002(Q) l l 5.4 Taking the Reactor Critical 5.4.1 Taking the Reactor Critical - Approval for takingithe; reactor.criticalashall;beTigranted by
,, 'the General * Manager only after recommendations have been made by the SORC. Thejoperations CD-905D , Manage rgy pan tM.ttne tnooess a ry approvale f o r itakingt:the reactor 1criticaltiin:1.the 7 bsence of Aher Genera 19 Manager but again only af ter recommendations have been made by SORC. ,
C D-14 2 D 5.5 Post Trip Review 5.5.1 A post trip review shall be conducted after every reactor trip. The post trip review will be conducted IAW OP-AP.ZZ-101(O), Post Reactor Scram /ECCS Actuation Review and Approval ', Requirements.
5.6 shifts and Shift Manning *
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5.6.1 The work week for operations personnel shall start at 0000 (midnight) on Monday.
5.6.2 There shall be five shifts that rotate on assigned schedules. The shifts are designated
"A" through "E".
. 5.6.3 The normal working day consists of three operating shifts: 2300 to 0700 (Mids), 0700 to 1500 (Days), and 1500 to 2300 (Swings). . 5.6.4 A licensed senior reactor operator shall be on site at all times when fuel is in the reactor. l l 5.6.5 In OPERATIONAL CONDITION 1, 2 or 3, two licensed ) SRO shall be on site - one of these SRO's shall l be in the control room at all times. In l OPERATIONAL CONDITION 4 or 5, one SRO shall be i on site - this SRO need not be in the control room at all times. The control room SRO may from time to time act as a relief operator f0r a licensed reactor operator assigned to the control room.
5.6.6 A licensed reactor c;erator (RO) shall be in the control room at all times when fuel is in the reactor. Whenever the reactor is in OPERATIONAL CONDITION 1, 2, or 3, an additional licensed RO shall be assigned to the control room.
OP-AP.22-002(Q) Page 36 of 41 Rev. 5 _- _ _ ._ - _ ___
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u l l l A. QUESTION 8.05 (1.50) ha the Emergency Coordinator you have declared a Site Area Emergency I CSAEJ. You now wish to terminate the event. l l Answer the following TRUE OR FALSE concerning the termination of an SAE 1 I a. Ontv the Emergenev Coordinator may terminate this event. CO.53.
b. To terminate. both of the f ollowing must be trues-None of the apolicable action levels defined in the ECG are soplicable AND...
-The plant is in a recovery status CO.53 c. NRC concurrence IS NOT reautred to UPGRADE an events but NRC concurrence IS recuired to DOWNGRADE an event. [0.53 ANSWER 8.05 (1.50)
a. false CO.53 b. false CO.53 c. false C O . 5.1 REFERENCE HoDe Creek Emergency Cl assi f i cati on Guide. Att. 3 K/A's... 294001 A1.16 2.9/4.7 294001A116 ...(KA'S) B. The Emergency Coordinator (EC) may be the SNSS, Emergency Duty Officer (EDO) or the Emergency Response Manager (ERM). During a Site Area Emergency, the EC position will be filled by either the EDO or the ERM. Only the EDO/ERM may terminate or de-escalate this event, but when they do, they are acting as EC. The Answer Key should be changed to TRUE.
C. See next two pages for supporting documentat..on. j l l l
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*Pg. 2 of .10 I. EMERGENCY COORDINATOR
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< Loc Sheet (Pa. 1 of 3) _ Instructions - Name 1. This is a permanent record 2. Each' step shall te initialed by the responsible individual. Date Time 3. Emergencyfoordinatora(EC)-respons-21bil'it'y'Aay Abe af ulf ilEdMy : Loc a t ion Se nior :Nu'cles'r' Shif t Supe r-
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visor (SNSS) Emergency Duty Of ficer- (EDO)
; Emergency Response Managar-( E RM )
A. Declare a SITE AREA EMERGENCY EC Initiating ECG Section - Condition Declared on at nrs.
cate time S. Not:fication Notifications must be completed within the time limits specified in the Communications Log.
1. Ensurw completion of !nittal Contact Message form EC (ICMF) (page 8). Approve. Provice to CM1 for required notifications.
. 2. Initiate Notification cy directing ccaraunicators: EC a CM1 - Implement Communicators Log ( At?.achment 27) l CM2 - Complete status forms (Attacnment 29).
, as designated by EC and those attachments.
HCGS 3 e 't . 4 Page 38 of 41
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*ATT 3 Pg. 5 of 1.0 II. TERMINATION (Emergency Coordinator)
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Instructiens Name 1. This is a permanent document.
2. Each step shall be initialed Da te Time Lo ca tion t Note . Only the EDO/ERM may terminate or de-escalate ; this event. , I
< 1. Terminate the Event when either of the following Ec conditions are met a. None of the action levels defined in the ECG are applicable.
OR
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b. The plant is in a recovery status
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2. Redute the event class when both of the following
. iu; conditions are met: l l
a. None of the action levels for a Site Area l Emergency as defined in the ECG are applicaole, ) and:
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b. One or more of the action levels for an Alert 0: Unusual Event as defined in the ECG is still l applicaole. l l l EC 3. Upon completion of step 1 or Step 2 (acove): l a. Ensure completion of the Reduction in Event Status Message Form. (See page 9 of tnis j procedure.) ! I b '. Di rec t CM1 and CM2 to make the designated j notifications on the Communications Leg. Time ' limits do not apply.
. 4. Ensure collection of all documentation and forward in EC accordance with reporting requirements of section III of this Attacnment.
HCGs Page 30 of 41 gev. 4
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A, QUESTION 8.09 (3.00) The reactor is operating at 100*/. o ower . You are informed that the RCIC Steam Line delta Pressure (Flow)-High instrument is inoperable. According to Tech Specs, what actions if any, should you take? C3.003 CUse the attached Tech Specs to form your answert be sure to document Coplicable sections used as cart of your answerJ ANSWER G.09 (3.00) According to T.S. Table 3.3.2-1 the RCIC isolation valves must be shut C1.0) and RCIC declared inoperable C1.03. With RCIC inoo. the ol ant can continue to coerate for 14 days (1.03.
REFERENCE HC Tech Spect Section 3.3 K/A's... 217000 G011 3.4/4.3 217000 GOO 5 3.3/4.3 217000G011 217000 GOO 5 ...(KA'S) B. The Tech Specs handout used during the test did not include page 3/4 3-16. This is where action 23 is located. Without knowing what action 23 was, the rest of the answer could not be completed as per the Answer Key. Full credit should be given for the table number (3.3.2-1) and the action number (action 23).
C. See next page for documentation.
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This POSE mis 5*3 - TABLE 3.3.2-1 (Continued) ISOLATION ACTUATION INSTRUMENTATION ACTION ACTION 20 - Be in at least HOT SHUT 00W within 12 hours and in COLD SHUTOCWN within tne next 24 hours.
t ACTION 21 - I Be in at least STARTUP with the associated isolation valves closed I within 6 hours or be in at least HOT SHUT 00WN within 12 hours and !
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in COLD SHUTOOWN within the next 24 hours.
ACTION 22 Be in at least STARTUP within 6 hours.
ACT10N"23 -
;Closetthe 2f f ecteosystee d seletion wa l ves .within wnshouri and declare thecaffacted4systes Anoperable.
ACTICN 24 - Restore the manual initiation function to OPERABLE status within 48 hours or be in at least HOT SHUTOOWN within the next 12 hours ! and in COLD SHUTDOWN within the following 24 hours.
ACTION 25 - Restore the manual initiation function to OPERABLE status within 8 hours or close the affected system isolation valves within the next hour and declare the affected system inocerable.
ACTICN 25 - Establisn SECONDARY CONTAINMENT INTEGRITY with the Filtration, Recirculation and Ventilation System (FRVS) operating within one hour.
ACTION 27 - Lock the affected system isolation valves closed within one hour and declare the affacted system inoperable.
ACTION 28 - Place the inoperaDie channel in the tripped condition or close the affected system isolation valves within one hour and declare the affected system inoperaDie.
ACTICH 29 - Plkce the inoperable channel in the tripped condition or establish SECONDARY CCNTAINMENT INTEGRITY with the Filtration, Recirculation, and Ventilation System (FRVS) operating within one hour.
NOTES I
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hhen handling irradiated fuel in the secondary containment and curing CORE
ALTERATICNS anc operations with a potential for draining the reactor vessel.
Vhen any turoine sten valve is greater than 9C% open anc/or when the key-locxed Dyoass switen is in the Nors position.
e
** Refer to Soecification 3.1.5 for applicability.
Within 24 hours prior to the planned start of the hydrogen injec* ion test, witn reacter power at greater tnan 22% of RATED THERMAL 80VER, the normal full power raciation background level and associated tric set:oints may be changed the test. cased on a calculated value of the raciation level expectec curing ' The background radiation level and associated trio set:oints may te acjustec curing the test program based on either calculations or seasure-ments of actual radiation levels resulting from hydrogen injection. The I background radiation level shall be determined and the associated trio setcoints shall be set within 24 hours of re establishing normal radiation levels af ter coteletion of the hycrogen injection test or witnin 12 nours of establishing reactor ;cwer levels colow 22% of RATED THERMAL PCWER, while these functions are required to be OPERABLE.
(a) A channel say be placea in an incoerable status fer up to 2 hours for re-cuitec surveillance without placing the trip system in the tri:csa concition proviced at least one other OPERABLE cnannel in the same trip system is monitoring that parameter.
(b) Also trips and isolates the mechanical vacuum pumos.
(c) Also starts the Filtration, Recirculation and Vantilation System (FRVS).
h0PE CREEr Page 41 of 41 Amendment. tao. a
ATTACHMENT 4 NRC Resolutions of Facility Comments on Written Examinations.
The following represents the NRC resolution to the facility comments made as a result of the current exam review policy.
Only those comments resulting in significant changes to the master answer key, or that were "not accepted" by the NRC, are listed and explained below. Comments made that were insignificant in nature and resolved to the satisfaction of both the examiner and the licensee during the post exam review are not listed (i.e.: typo errors, relative acceptable terms, minor set point changes).
SR0 EXAMINATION QUESTION 5.01 NRC RESOLUTION: NRC will accept two of the three mswers for full credit in addition to the answer key. The facility should review and update their training material and/or procedure to avoid confusion between the Learning Objective in the lesson plans and the material presented in the procedure.
QUESTION 5.11 NRC RESOLUTION: Comment accepted. Answer key will be changed to accept "return to initial level"; or "remains the same" with an understanding that the level is at the pre-transient le/el.
QUESTION 6.04 NRC RESOLUTION: Comment accepted. Full credit of (1,0) point will be given for either "stops the reactor cooldown" or
"stops the reactor depressurization". For item c, NRC will accept both responses for full credit (i.e.
Answer Key response or NONE of the RESET pushbuttons) QUESTION 6.07: NRC RESOLUTION: RRCS initiation will be accepted as long as it is clear by the candidate that all functions of RRCS are not initiated (e.g. SLC).
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l l , Attachment 4 -2' ! QUESTION 6.11 ! NRC RESOLUTION: Upon reviewing the possible responses,-the following i will be accepted for full credit: Loss of instrument air leading to a loss of scram air header pressura; . loss of air to the outboard MSIV's causing a slow ' closure resulting in a reactor trip from valve position or reactor pressure; or opening of the i condensate recirc valves resulting in a complete-loss of feed and a reactor trip on low wat'er level.- Any . reasonable response in this area will be given full l - credit. , QUESTION 8.01 i J NRC RESOLUTION: Comment accepted. It must be clear in the candidates- i response that the OPS Manager is acting for the ' General Manager. ! > QUESTION 8.05
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NRC RESOLUTION: Question 8.05.a will be deleted vice changing the , answer to TRUE. Confusion over who is acting as the EC has invalidated this question. ' l ; QUESTION 8.09
i NRC RESOLUTION: Comment accepted. Full credit will be given for- l - listing the proper table and action number as the response, r j ADDITIONAL ANSWER PEY CHANGES THAT ARE NOT ADDRESSED IN THE COMMENT -
: INCORPORATION CHANGES-
' QUESTION 8.10 I Answer key point value change to match question point value.
Each part of section a. will be worth (1.0) point. Part b.
j point value will remain unchanged.
QUESTION 8.11 l In accordance with the reference material, the correct response i to question shall read: l l "NO (0.5) the reactor may exceed rated thermal power (3293 MWt) ] by up to 2'4 as long as the average thermal power over any eight hour shif t does not exceed rated th9rmal pnwer. (0,5)"
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Attachment 4 3 i RO EXAMINATION i QUESTION 1.03b NRC RESOLUTION: Comment accepted. It is general grading policy to award partical credit and to not "double jeopardize" incorrect answers.
QUESTION 1.07b , NRC RESOLUTION: Comment accepted. It is general grading policy to a award credit for answers that are correct though they vary from the wording of the answer key.
j QUESTION 1.09b
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NRC RESOLUTION: Comment not accepted. Both parts of the answer together constitute a complete answer. ' i QUESTION 1.11b NRC RESOLUTION: Comment not accepted. MAPLHGR is the maximum APLHGR in the core, not the APLPGR limit. The question clearly states: "STATE the core thermal limit."
(Though this comment is "not accepted," partial credit will be awarded if the candidate responded with
"MAPLHGR," consistent with general grading policy.)
' l QUESTION 2.05b j NRC RESOLUTION: Comment accepted.
' QUESTION 2.06b NRC RESOLUTION: Comment accepted.
QUESTION 3.01a NRC RESOLUTION: Comment accepted, however, the answer key was amended
- to require discussion of RCIC exhaust line check valve cycling in addition to bearing lube oil requirements for full credit.
" QUESTION 3.06b , ' NRC RESOLUTION: Comment accepted, based upon supplemental corrected reference material provided.
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QUESTION 4.04b NRC RESOLUTION: Comment partially accepted. The intent of the answer ' key answer was not to-penalize the candidate if he .
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failed to list OP-EO ZZ-100 since it would immediately > direct usage of OP-E0.ZZ-101. It the candidate , additionally lists OP-E0-ZZ-100, he will not be penalized.
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QUESTION 4.05b i NRC RESOLUTION: Comment partially accepted. 'If the candidate
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qualifies 300 mR/qtr with a statement to the effect d that it applies in the event a current quarterly estimate has not been completed, then full credit will-be awarded. Otherwise only half credit will be awarded.
QUESTION 4.08b NRC RESOLUTION: Comment partially accepted. Point value for the question is 2.0 points. The answer key was corrected to be consistent with this value. The referenced KA i was in error and has been changed to 202002SG6 vice 211000SG6. However, the question will not be deleted.- Step 5.4.20.1 of procedure OP-10-ZZ-003(Q) contains this technical specification as a limitation, and
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learning objective #1 of Hope Creek lesson plan 302HC-00-112-01 states that the trainee, from memory, will be able to correctly state the basis / purpose of a limitation. The corrected KA does apply. , QUESTION 4.10b r NRC RESOLUTION: Comment accepted, based upon supplemental corrected
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reference material provided.
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ATTACHMENT 5 SIMULATION FACILITY FIDELITY REPORT Facility Licensee: Public Service Electric and Gas Company Post Office Oox 236 Hancocks Bridge, New Jersey 08038 Facility Licensee Docket No.: 50-354 Facility License No.: NPF-57 Operating Tests administered at: Hope Creek Simulator Operating Tests Given On: April 12, 13, 14 1988 During the conduct of the simulator portion of the operating tests identified above, the following apparent performance and/or human factors discrepancies were observed : 1. In high power initial conditions all of the LPRM downscale lights are lit. They should only be lit when an actual downscale condition exists.
2. During an ATWS transient, some of the power, pressure and level instrumentation started switching from upscale to downscale and back at a frequency of about twice per second. The switching lasted for about one minute then the indicators responded as expected.
Overall, the fidelity and response of the simulator was good. }}