IR 05000423/1985044
| ML20136D971 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/13/1985 |
| From: | Dante Johnson, Keller R, Kister H, Kuscitto D, Ruscitto D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20136D930 | List: |
| References | |
| 50-423-85-44, NUDOCS 8511210400 | |
| Download: ML20136D971 (106) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 85-44(OL)
FACILITY DOCKET NO. 50-423 FACILITY LICENSE NO. CPPR-113 LICENSEE: Northeast Nuclear Energy Company P.O. Box 270 Hartford, Connecticut 06141-0270 FACILITY: Millstone 3 EXAMINATION DATES: September 3 - 6, 1985, and September 9 - 12, 1985 CHIEF EXAMINER: b ./ . #
0. Ruscitto, Examiner
//!4f/6 Date REVIEWED BY: #hJe 0. J6hnson,'eadL Exami r Nh Date M
s ilb 6 R. Keller, Chief, Proj ct Section 1C Date Approved by: /47 N if. Kist'e(, Chief, Projects Branch N /Dat6 Division of Reactor Projects Summary: Oral, written and simulator examinations were given to 14 Senior Reactor Operator (SRO) and four Reactor Operator (RO) candidates. One R0 candidate was given a written (first) ratake examination. Eleven SR0 and three R0 candidates passed all portions of the examination and will be issued licenses. Three SR0 and two R0 candidates failed the written examination, and their licenses were denie f* DR ADOCK 0500 i
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REPORT DETAILS TYPE OF EXAMS: Initial X EXAM RESULTS:
l R0 l SR0 I l Pass / Fail l Pass / Fail l l l l l I I I IWritten Exam l 3/2 l 11/3 l l l l l l l l l l Oral Exam l 5/0 1 14/0 l l l l l l l l l l Simulator Examl 5/0 1 14/0 l l 1 I I I I I i 10verall l 3/2 1 11/3 I I I I I l l l l CHIEF EXAMINER AT SITE: D. Ruscitto OTHER EXAMINERS: B. Keller, NRC 0. Coe, NRC R. Schreiber, PNL L. Defferding, PNL Summary of generic deficiencies noted on oral and simulator exams:
Minor deficiencies with no safety significances were noted in the following areas: Knowledge of how and where to trip protection and instrumentation bistables when I and C assistance is unavailabl Knowledge of the approximate contact dose rate levels present on recently expended fuel assemblie . Summary of simulator deficiencies:
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5. Summary of generic deficiencies noted from grading of written exams: '
R0 Exam (Five Candidates)
l Core characteristic changes over core lifetime ECCS mini-flow recirc valve automatic features and interlocks Location and purpose of Radiation Monitoring System components SR0 Exam (14 Candidates) Bases for S/G code safety setpoints and capacit OTAT and OPAT setpoint changes with plant parameter ( Reason for loading emergency diesels greater than 50%.
- -. Difference between Containment Instrument Air and Instrument Ai Basis for pressure and temperature limits on RHR syste :
6. Interface with Plant Staff During Examination Period l Liaison with plant staff was good and the simulator instructors were helpful in scenario review and modification during the course of several scenarios, which facilitated the examination proces . Personnel Present at Exit Interview: !
NRC Personnel
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D. Ruscitto, Chief Examiner D. Coe, Examiner B. Norris, Examiner +
R. Starostecki, Director, Division of Reactor Projects i
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,' Facility Personnel J. Opeka, Vice President, Nuclear Engineering and Operations R. Test, Director, Nuclear Training W. Romberg, Station Superintendent J. Crockett, Unit Superintendent
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K. Burton, Operations Supervisor M. Moehlman, Nuclear Training J. Black, Simulator Project Manager ,
H. Haynes, Simulator Technical Systems Manager !
R. Lueneburg, Simulator Program Supervisor l R. Stotts, Nuclear Training i l Summary of NRC Comments made at exit interview:
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Generic deficiencies during oral / simulator examinations and interface with plant staff were summarized. No preliminary results were give . Examination Review At the conclusion of the written examinations, the examiners met with licensee personnel to review the examination and answer keys to identify ;
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any inappropriate questions relative to plant specific design and to ensure that the questions will elicit the answers in the key and that they reflect the most current plant conditions. The following licensee personnel presented the comments to the examiners:
! J. Crockett R. Stotts
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l Attachments: I Written Examination and Answer Key (RO)
l Written Examination and Answer Key (SRO)
l Facility Comments on Written Examinations made after Exam Review I
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U. 5. NUCLEAR RECULATORY COMMISEIGN REACTOR OPERATOR LICENSE EXAMINATION FACILITY: HILLSTONE 3
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REACTOR TYPE: PWR-WEC4
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DATE ADMINISTERED: 85/09/04
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EXAhINER: COE, _________________________
APFLICANT: _________________________
INSTRUCTIONS TO APPLICANT:
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Use separate paper for the answer Write answers on one side onl Staple question sheet on top of the a.n s w e r sheet Foints for each question are indicatec in parentheses after the question. The passing grade requires at least 70% in each category and a final stade of at lcast 80%. Enamination papers will be picked up sin (6) hours after the exanianstion start M'
% OF CATEGORY % OF APPLICANT'S CATEGORY .
VALUE TOTAL SCORE VALUE CATEGORY
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25.00 25 00 FRINCIPLES OF NUCLEAR POWER
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PLANT OPERATION,.THERH0 DYNAMICS, HCAT TRANSFER AND FLUIC FLOW
_ 1 __ _ 1 ___________ ________ PLANT DESIGN INCLUDING SAFETY AND EMERGENCY S1 STEMS
_ Ibb___bI ___________ ________ INSTRUMENTS AND CONTROLS
"3.00 PROCEDURES - NORMAL, ADNORMAL,
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_ ___________ ________ ENERGENCY AND RADIOLOGICAL
- CONTROL 100.00 100.00 TOTALS
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FINAL GRADE _________________%
All work done on this enan.ination is n.y own. I have neither givcn nor received ai ~~~~~~~~~~~~~~
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QUESTION 1.01 (2.50)
Prior to Millstone 3's initial criticality, two neutron sources will be intentionally placed in the core, a. What are each of these sources called and what materials are responsible for neutron production in each? C0.83 b. What is their purpose? CO.83 c. What, if anything, must be done in order to cause each of these sources to start emitting neutrons? Why? CO.93 GUESTION 1.02 (2.50)
During a reactor startup the reactor is e::actly critical when the Intermediate Range indication is observed to increase fron: 3E-8 amps to 4E-7 amps in 90 second What is the start up rate? CO.53 If the effective delayed neutron fraction is 0.005 and assuming on average neutron precursor decay constant of 0.08 see-1, how much reactivity was added in part (a) above? [1.03 c. If the above stable SUE had been caused by a step change in reactivsty power would initially increase abruptly before establishing the stable SU What is this phenomenon called and what causes it to occur? [1.03 (zuuma CATEGORY 01 CONTINUED ON NEXT PAGE musum)
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. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3
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OUESTION 1.03 (3.00)
In each of the following cases, which of the two hillstone 3 control rods will have the higher observable rod worth or will they be the same?
Briefly explain why. Assume all factors other than those specifically ocntioned are identical in both case I. A new RCCA which hes never been used at powe II. An identical RCCA which has been usec through 10 core cycle I. A single ' black' control rod containing a given amount of strongly absor bing n.ater ial .
II. Ten smaller diameter control rods having the same length and using the same total amount of material as in I. above. Each tod maintains its ' black' characteristi c. I. An RCCA in a reactor with reactor coolant temperature at 557 II. An identical RCc3 in a reactor with reactor coolant temperature st 587F. This RCCA is at the same height as in I. abov An RCCA stuck fully withdrawn with all other RCCA's fully inserted (reactor shutdown)
II. An identical RCCA which is dropped while at 100L powe [0.75 each]
QUESTION 1 04 (2.50)
The hillstone 3 reactor has been operating at 75% power for 3 day Boron concentration is 1200 ppm (BOL). Due to an instrument malfunction a reactor trip occurs. No operator action is taken and the plant temperature stabilizes at 557F through stean, dump actuatio Use the attached graphs Fig. 1-1 through Fig.1-7 to answer the following question Clearly show your work, including which graph each number is taken fro Assume Shutdown bank tod worth is 4700 pcm and that Bank D was at 150 steps prior to the reactor tri Determine actual Shutdown Margin (SDM) at 557F. Do not account for a single stuck ro Assume cooldown from 75% power conditions to 557F takes only 60 second [0.53 i Determine actual SDM at 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after the tri C0.83 Determine how much boron dilution would be required to achieve criticality 75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> after the trip if all Shutdown rods were withdraw E1 23 l
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QUESTION 1.05 (2.00)
The Millstone 3 reactor is operating at 50% power, BOL, when a steam dump fails open. Assuming rods are in manual, no operator action is tsken, and no reactor trip occurs, give a step-by-step explaination of what happens to reactor power and Tave including why each change occur QUESTION 1.06 (3.00)
Indicate at which tin.e ir. core life (BOL or EOL) the following occidents are more severe (result in a longer time spent at higher power).
Briefly explain wh hain steam lir.e br eak [1.0] Total loss of coolant flow [1.03 Rod withdrawal accident ftom low in the source range prior to any significant reactor coclant temperature increas [1.02 00ESTION 1.07 (2.50)
Answer the following questions regarding Reactor Safety Limits: Adher ence to Reactor Safety Limits ensures two specific occurances are prevented. What are these occurances and how do Reactor Safety Limits prevent t h e n. ?, E. e sure to identify the reactor parameters which are limite [2.03 b. How are Limiting Safety System Settings (LSSS) related to Reactor Safety Limits? CO.53 OUESTION 1.08 (1.50)
A cooling water pump is operating at 1500 rpm. Its capacity is 250 gal / min at a discharge pressure of 15 psis which requires 40 MW of powe Determine the pump capacity, speed and power requirement if the pump discharge pressur e drops to 10.0 psis due to reduced spee e (***** CATEGORY 01 CONTINUED ON NEXT PAGE maamm)
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QUESTION 1 09 (2.50)
The Millstone 3 reactor is operating at steady state with the following parameterst Pstm = 1065 psis Feedwater temperature to S/G's = 440F All S/G steam flows = 3.7 million Ibm /hr a. Determine the thermal output of the primary system in terms of percent rated core thermal power. Show your work and state your assumption C2.03 b. Other than the reactor core and S/C, what other heat sources and sinks tend to add or ren.ove heat from the primary tystem durin3 normal power operation [0.53
' QUESTION 1.10 (3.00)
Answer the following questions concering natural circulatio List 5 major plant parameters available to help provide indication of natural circulation and how they would respond if natural circulation were taking place? A single parameter may be defined as a certain combination of individual paranieters which give i m p o r t a r.t information about the actual core condition C1.03 How would the following occurances affect natural circulation (help, hinder, or no effect)* E:rtefly enylain wh [0.5 each]
1) S/G 1evels fall to low in the narrow rang ) Pressuri:er pressure increases fron saturation for That to 100 psia greater than saturatio ) S/G 1evels rapidly increased from below the narrow range level indication into the normal operating ban c. What reactor plant design features tend to enhance natural circulation flow? CO.53 (samma END OF CATEGORY 01 *****)
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GUESTION 2.01 (1.50)
Answer the following questions concerning the Pressuri:er Relief Tank (PRT)! Give 2 reasons for using nitrogen overpressure in the PR [0.53 b. Briefly describe the automatic feature of any valve or valves other l than the nitrogen regulator which connect directly to the PRT. E1.03 l
00ESTION 2.02 (2.50)
On Figure 2-1 draw the reactor vessel head vent piping and valves. Attach Figure 2-1 to your snswer sheets. Include:
1) All major flow paths, valves, their method of positiening, their normal position when the reactor is at power, where each valve may be operated from, and whether or not a valve can be modulated open from its remote control poin [2.03 2) Any interconnections to another system clearly showing the precise point the systems interconnec QUESTION 2.03 (1.50)
Answer the following questions concerning the Reactor Coolant Pumps!
a. List all the design features which protect the RPCCW system from a f ailure/ruptur e in the thermal barrier cooling cotl? C1.03 l
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b. Explain why VCT pressure is important to the operation of the RCP seal CO.53 OUESTION 2.04 (4.00)
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Clearly explain any automatic features or interlocks associated with the operation of the mini-flow rectre valves on Emergency Core Cooling System pumps including setpoints where applicable and the reason for having occh interlock or automatic feature. Clearly indicate the specific pump associated with each recire valve.
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QUESTION 2.05 (2.75)
Use Figure 2-2 (ECCS Standby Mode) to answer the following questions:
a. List the interlock conditions which must entst in order to line up the loop i hot les to the RHR A pump suctio E1.03 Trace the norsially expected flowpath of water from the sump to the hot less during hot les recirculation following a LOCA. Use Fig. 2-2
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as an answer page, use a felt tip markere and show open valves with an 'O' and closed valves with an 'X'l or else describe the flow path in words using a blank answer sheet clearly indicating open and closed ,
valve [1.753 !
OUESTION 2.06 (3.00)
Answer the following questions regarding the Containment Depressurt:stion System (CDS)!
a. List the sequence of all automatic actions related to the CC5 that you would expect to occur following a valid High-3 Containr.ent pressure signal during a Loss of Coolant Accident. Do not include alarm Assume no operator action, include all automatic actions associated with the RWST levci, and continue until the RWST is empt C2.53
, b. How would the above sequence be different if a Loss of power (LOF)
signal existed when containment pressure reached tne high-J setpoint?
CO.53 f
GUCSTION 2.07 (2.25)
The reactor is operating at a steady state 25% power, all control systems are in auton.ati Turbine load as increa' sed to 100% and the stest pr essure cetector for el S/G sticks at the 25% value. Euplaine in detail, how and ehy this will affect 41 steam gener ator leve Assume no operator actio QUESTION 2.00 (3.00)
Sketch the flow path of electrical power from DC A to bus VIAC-3. Label all busses, components, and breakers. Ind1eate ;
voltages, and include alternate flow paths. Creaker numbers are not '
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QUESTION 2.09 (3.00)
Downstream of isolation valves LCV112B/C there are five separate oajor piping connections into the combined CCP suction headere not including either the RWST or RHR discharg List each of these and briefly describe its intended us QUESTION 2 10 (1.50)
List the source of cooling water, if any, for the following loads during o Loss of Power (LOP) condition:
a. The three containment air recirc cooling coil b. Containment instrument air compressor c. Reactor coolant punip motor air cooler c. Neutron shield water tan e. MCC and rod control air conditioning unit [0.3 each]
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QUESTION 3 01 (3.00)
- Answer the following questions concerning the Power Range NI's!
a. What happens when N41 is selected on the power mismatch bypass switch?
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List 3 systems which are affecte [1 03 b. What two functions do the rod stop bypass switches perform? [1.03 c. List two reasons why the detector current comparator alarm function is automatically defeated when power is below 50%. E1.03 GUESTION 3 02 (2.00)
l Answer the following questions about Figure 3-1, the Gas-filled detector l curve!
l a. Label the horizontal and vertical axe b. Indicate in which region (I-VI) each of the following instruments j operates l 1) source range NI
! 2) intermediate range NI l 3) power range NI l 4) portable frister (RM14)
QUESTION 3.03 (2.50)
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Answer the followir.g questions regarding the main steam s/ stem s. While operating at 80% powere 42 S/G safety valve fails wide open with centrol rods in auto at 215 steps. Will the S/G 1evel control system detect the addec staan flow? E:riefly explain your answe [0.53 b. How will nuclear and RCS temperature instrumentation alert the operator that a problem has occured in (a) above? [0 53 c. What three si3nals cause automatic closure of the MSIV's? Indicate
! coincidence, and what conditione if any, will block each signal.
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QUESTION 3.04 (1.50)
Answer the following questions regarding the Steam Dump electro pneumatic control system with either (s) always true, (b) sometime true and sometime falser or (c) always fals Assume no malfunctions e:<1st unless specifie If answer (b) is chosen, briefly explain wh ) The supply air which acts on the steam dump valve diaphragm passes through four solenoid valve CO.53 2) The lo3tc si3nal which tells the steam dump valve when to open and how much to open passes throvsh the I/P converte [0.53 3) When in the steam pressure mode of operation above the low-low Tave setpoint, a siniultaneous failure of both turbine impulse pressure channels (Icw) will cause steam dump actuatio [0.5:
QUESTION 3.05 (3.00)
Answer the folicwing List all inputs er coreditior.s which wovid cause actuation cf the following slynals. Give setpoints and coincidence where applicablei 1) SIS 2) CDA 3) LOP What compcnent recieses these three signals ande dependin3 or, their c o mbina t t ori e alters the response of plant Engineered Safety Features? -
GUESTION 3 04 (2.50)
As reactor pcwcr and turbine load are reduced from 100% to 5% cescribe when each RCP Shaft Speed reactor trip becomes effective. Where permissives are 2nvolved, include all applicable coincidences and oetpoint (mmmmm CATEGGRY 03 CONTINUE 0 ON NEXT PAGE us***)
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QUESTION 3.07 (2.50)
Answer the following questions: What limits are the following RCS trips designed to protect against?
1) OTDelta T 2) OpDelta T 3) Low pressuri:er pressure 4) Hi pressuricer pressure [0.4 each]
b. For a given RCS temperature, how does the low pressurizer pressure trip serve.to limit the range of OT0 elta T setpoint? CO.9]
OUESTION 3.08 (3.00)
Indicate what happens to the Rod Control System (rods in, rods out, no change) ano DRICFLY e:: plain why the change will or will r.ot occur for the following ecnditicns. Rods are in auto unless otherwise specifie Reactor power is 17% where the controlling turbine first stage impulse pressure transmitter (PT505) fails hi3 b. Reactor power is 100% and loop 1 That fails hig c. Reactor power is 50% with rco control in manual. Due tc instrument testing the following conditions existi
" ~ Mid durehy mt..o... .d'..f ;.; 2.a ':' cr; ; ;;t'. . :: t r' . . mm g)<..
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Auctioneered high nuclear power is 51% and turbine pewer (input to rod control only) is 75%
All indications have been stable for the last hour The Bank Selectcr switch is then placed in AUT (xxxxx CATECORY 03 CONTINUED ON NEXT PAGE ***x*)
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QUESTION 3.09 (3.00)
Answer the following questions regarding the Radiation Monitoring System (RMS): Describe the location, purpose, and automatic control features, if any, associated with the two Categor y 1E AREA radiation monitor pair Channel designations are not require E1.03 b. Following a high concentration alarm what automatic function (s) does the turbine building drains monitor (30A5-RE50) initiat Ci.03 c. The 3HVRmRE10A and B ventilation vent monitors read out one display channel during normal operation and two display channels during accident cond2tion What are thE two 20cident parametets being measured? [1.02
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QUESTION 3.10 (2.00)
The reacter is operating at 40% power steady state with four loop Describe the initis1 cutomstic actior.s f or the fc11owing cecorances (slar not required): . /4 low-low level in 1/4 S/G' CO.53 /4 low-low level in 2/4 S/G' [0.53 /4 hi-hi level in 1/4 5/G' CO.53 Steam pr essure input to the Hein feed pump speed control system fails hig CO.53 i,
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DUESTION 4.01 (3.00)
Answer the following questicns concerning the Frecautions and Limitations during OP 3201 Reactor Plant Heatu a. A plant heat-up must be stopped immediately under what conditions of source rense channel indication? Assume no instrument failure or as1 functio CO.753 b. What precaution should be taken when Auxillery Spray must be used and why? CO.753 c. Why is FCP B the prefered pump for single pump operation? CO.753 d. In Mode 4 at least one RCP should be in operatio In Mode 3 at least two RCF's shecid be in operatio In either of these erses, all RCP's may be secured for up to one hour as long as what two conditions are met? CO.753 DUESTION 4.02 (2.00)
Onswer the following questions concerning a reactor startup:
a. What is the minimun. temperatura at which the hillstone 3 resetor may be taken critical? Co.53 b. Within what tin,e per iod pr ior to criticality must this minimum temperature in part a. be verificd7 CO.53 c. After criticality is achiesede loop 1 Tsve is noticed to be less than the minimum required temperature for critical operations. The other channels ate above this limit. Assuming no instrument malfunction, what action, if any, is required? C1 03 (musan CATEGORY 04 CONTINUED ON NEXT PAGE mansu)
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QUESTION 4.03 (3.00)
Answer the following questions regarding Reactor Coolant Pump operation:
a. An unexplained increase in RCP A seal water return causes the RCP HI RANGE LNG FLOW HI alarm. Return flow indication is noted to be spm-and stead The RCP A lower bearing water temperature is trending steadily upward. All other RCP's are operating normall Assuming these conditions continue, at what bearing temperature must action be taken to isolate 41 seal and how much time do you have to take such action in accordance with AOP 3554 FAILURE OF RCP SEAL?C1.03 If the high seal return flow rate was indicated on all RCP's and in addition the VCT temperature was noted to be excessively high (195 F),
how would your required actions be different when one RCP lower bearing water temperature reached the limit in (a) above? [1.03 c. Answer question (b) assuming two RCP lower bearina water temperatures reached the limi E1.03 OUESTION 4.04 (3.00) Ir. accordance with OP3346A: if the Emergency Diesel Generator (EDG) is operated at less than 50% espacitr for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the diesel should be operated at greater than 50% for one hour. Why? [1.03 List four diesel shutdown parameters which require close observation during emergency mode operation.to prevent dama3e to the diesel. E1.03 What are two conditiens thet will automatically stop the diesel under emergency mode of operation? [1.01 QUESTION 4.05 (3.50)
Answer the following questions utili=ing information found in FR-S.1, RESPONSE TO NUCLEAR POWER GENERATION /ATW What four conditions must you observe to verify a reactor trip? E1.03 b. What action must be taken if the reactor will not trip on a manual reactor trip initiation? E0.53 c. What actions are required to initiate immediate boration of the RCS?
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QUESTION 4.06 (3.00)
Answer the following questions regarding EOP 35 FR-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK List the two conditions which warrant entry into this procedure. Be specific. C2.03 b. What action is required if, during this procedure, the RWST level decreases to less than 520,410 gallons? [1.03 GUESTION 4.07 (2.00)
List two SI actuation criteria which may not necessarily cause an automatic SI signal. Gsvs setpoints (and adverse containnier.t setpoints)
if applicabl QUESTION 4.05 (1.00)
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Provide the Technical Specification definition for CHANNEL CHECK (as
,. performed by a control room watchstander).
GUESTION 4.09 (1.50)
In accordance with Technical Specifications, temporary changes to operating procedures may be made provided what three conditions are met?
- GUESTION 4.10 (1.00)
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If no Control Point henitor is established, what responsibility does each individual have when entering and leaving ar. RWF work area and what areas of the RWP must he/she fill in? [1.03 (xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE xxxxx)
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. PROCEDURES - NORhAL, ABNORMAL, EMERGENCY AND PAGE 16
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____________________
55siE5E------------------------
GUESTION 4.11 (2.00)
A Health Physics survey in an area of the plant shows the following results:
Beta-samma dose rate 4.0 mrem /hr Mixed (fast and thermal) neutron dose rate 6.0 mrem /hr loose surface contamination 200 dpm/100 sq cm beta-samma 50 dem/100 sq cm alpha How should this area be classified radiologies11y? [1.03 Assuming the loose. surface contamination was removed, how long can a radiation worker who has a con.plete NF:C Form 4 (exposure record)
work in this area without exceeding the Millstone initial administrative exposure limit' EO.53 Who's approval is required to exceed the limit in (b) above? E0.53
.
(xxxxx END OF CATEGORY 04 xxxxx)
(xxxxxxxxxxxxx END OF EXAMINATION *******xxxxxxxx)
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. PRINCIPLES OF NUCLEAF. POWER PLANT OPERATION, PAGE 17
--- isEER557sEEiCi- sEIT iEEs5 FEE Es5 FEUi5 FE55
____________________________________________
ANSWERS -- MILLSTONE 3 -85/09/04-COE, ANSWER 1.01- (2.50)
- o. primary source - californium; secondary source - antimony /berylium l CO.4 each]
i b. To raise the flun level by suberitical multiplication to a j
'
level which allows verification of proper source range NI operation. [0.83 c. Primary - nothing. It spontaneously emits neutrons from decay of C CO.453 Secondary - Power oper ation is r equired to activate the source which then e. wits neutrons (with a decaying 60 dcy half life after powet operations cease). CO.453 REFERENCE Topic 1 Lesson 5 pp 28-29 Reactor Theory Fig. 5-6 to 5-11 ANSWER 1.02 (2.50) sur(t)
F = Po(101 CO.23-7 -5 1.5 sur 4X10 = 3X10 (10) CO.23 solve for sur, sur = 0.75 DPH CO.13 b. T = 26.06/sur = 26.06*/0.75 = 34.74 see CO.33 rho = beta *eff/(1 + lambambar X T) = 0.005/(1 + 0.05(34.74)) CO.53
= 0.001323 = 132 pem CO.23 c. Prompt jump CO.23 caused by the immediate multiplication of prompt neutrons which are limited only by their relatively short generation ti CO.63 REFERENCE-Reactor Theory pp. RT-11.2 and 1 .
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.
. PRINCIPLES OF NUCLEAR POWER PLAtT OPERATION, PACE 16
-~~ isEEs55isEsi5s- REEi iEsssFEE Es5 FEUi5 FE5E
____________________________________________
ANSWERS -- MILLSTONE 3 -85/09/04-COE, ANSWER 1.03 (3.00) Same CO.25] Hafnium is used in Millstone 3 RCCA's and does not noticeably burn ou [0.53 II CO.25] Total surface area is greater for I [0.5] II [0.25) Due to increased thermal diffusion and slowing down length more neutrons can reach the rod to be absorbe [0.53 I CO.253 Rod worth is proportional to (local flu::/ core ave. flux)
square In case I. this value is much greater than in case I [0.53 REFERENCE o. Topic 1 Lesson 5 pg 26 and Reactor theory Fig. 14-2 and pp. 1 b. Reactor theory pp. 14.3-1 c. pg. 1 d. pg. 1 ANSWER 1.04 (2.50) +1310 pem f r o ni Power Defect curve EO.12-4100 from IRW curve CO.12-4700 given SD rods worth E0.13
________
-7490 pe [0.23 pcm from part (graders note * no double jeopard Full credit given for correct process) [0.23-(4500-2600)=-1900 from Xe curve C0.22
-$ 10-589)= -22 from Sn, curve EO.23-Ni2 I5 5?? i;s- to.23 pcm from part CO.23
+4700 SD rods out E0.2]
+(2600-300)=+2300 from Xe curve [0.2]
453-589)= -/6 from Sm corve EO.23
_____________
-l^? pcm E.2] 43A4 pcm/10.2 pcm/ ppm = -44e ppm dilution CO.2]
~ 655 655 @2 REFERENCE Millstone 3 curve book
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. PRINCIPLES OF NUCLEAR POWEF, PLANT OPERATION, PAGE 19
--- isEER55isEsiCi- REEi isEssFEE Es5 FEUi5 FE5E
____________________________________________
ANSWERS -- HILLSTONE 3 -85/09/04-COE, ANSWER 1 05 (2.00)
1. Tcold initially decreases due to greater heat removal fron. the primary causing Tave to decreas CO.43 2. MTC causes a positive reactivity insertion due to the lowered Tave E0.4]
3. Reactor power increases due to the reactivity insertio [0.43 4. As reactor power increases, Tave stops decreasing due to increased heat input and negative reactivity is inserted by FTC due to higher fuel temperature [0.4]
5. Final T4ve then stabilizes at s lower value at which positive reactivity from MTC excetly offsets the ne;,etive reactivity from FT Reactor power-is increased by the amount cf the steam dump capacit CO.4]
REFERENCE Roactor theory pp. RT-18.4 and RT-1 ANSWER 1.06 (5.00)
c. EOL - MTC is more nes.3tive and thus imparts greater positive reactivity from the drop in coolant temperatur M D,c3 (Doppler only Power coefficient (DOPC) is less ne and thus imparts less reactivity from the power increase)gative IJk"If BOL - MTC is less negative and thus imparts less negative reactivity from the coolant heat-u Wr"TT O,03 (00PC is more negative and thus imparts more positive reactiFity as power drops which tends to hold power u p ) .LE . C r EOL - Beta-eff is less, making SUR greate [0.5]
DOPC is less negative and thus imparts less negative reactivity after the fuel temperature begins to increase. [0.5]
REFERENCE a. Topic 8 Lesson 5 pg. 28 b. pg. 45 c. pg. 48 a. through Reactor Theory Chapt. 17 l
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. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 20 l
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--- isEEE557DEsiCE- REEi iEEsiFEE ED5 FEUi5 FE5E
____________________________________________
ANSWERS -- MILLSTONE 3 -85/09/04-COE, ANSWER 1.07 (2.50)
a. DNB [0.23 - The plant is operated within the acceptable region set by Tave, Thermal power, primary pressure, and flow CO.2 each3 which provides a reasonable assurance that a minimum '
DNBR (of 1.3) will be maintaine [0.23 RCS integrity [0.23 - Protected by appropriately set code safeties [0.33 and reactor trip settings. CO.33 LSSS's are automatic protective device setpoints which are chcsen so that protective action will prevent e::ceeding a Saf ety Limit. [0.53 REFERENCE Thermodynamics text pp. 260-263 Topic 7 Lesson 1,2,3 pg. 44 a n d T c l .,f p c, Q g 7
' ANSWER 1.06 (1.50)
N1=1500rp Vi=250 spa Hp1=15psis Pl= 40 kW N2 = 1500 sqr roct(10.0/15) = 1224.7 rpm CO.53 V2 = (1224.7/1500)(250 spn.) = 204.1 spm E0.53
P2 = (1224.7/1500) (40 KW) = 2 KW [0.53 REFERENCE Thermodynamics text pp. 322-324
- . . ._-__- -__ - .. .
,
.
. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, FAGE 21
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~~~~ TUERU66 Ud55C5I~5E5T TRhU5E5R kU6~ FLU 5C FLOU
____________________________________________
ANSWERS -- MILLSTONE 3 -85/09/04-COE, ANSWER 1.09 (2.50)
a. Assumptions: No heat loss occurs from primary system to surroundings Feedwater is at saturated conditions to.2 each3 Pstm = 1065 psis = $' h psia; enthalpy = 1191.0 Btv/lbm CO.43 Feedwater enthalpy = 419.0 Btv/lba CO.23
________
change in enthalpy = 772 Btu /lbm EO.23 Gdot = mdat X delta h = (3.7 M lbm/hr)(772 Stv/lbm)(1 MW/3.41 MBtv/hr)
! = 837.65 MW CO.43 l Gdot for all 4 S/G's = 4(837.65 MW) = 3350 MW CO.23 3350/3411 = 98.2% of rated thermal power EO.23 RCP's and pressuricer heaters add heat C0.253 ambient losses and CVCS system removes heat CO.253 REFERENCE Thermo te::t pp. 341-343
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. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 22
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~~~~iEER566YUIE5C5~~UEdT"YRdU5E5R DUD 5LUYU~i LUU
____________________________________________
ANSWERS -- MILLSTONE 3 -85/09/04-COE, ,
ANSWER 1.10 (3.00) . RCS differential temperature approximately 25% to 80% of" full power i on wide range RTD's (less than or equal to 100% power delta T will also be an acceptable answer)
2. Thot RTD's should indicate stable or decteasing temperature . Core exit thermocouples steady or decreasin . S/G pressure should be decreasing, following RCS average temperature 5. RCS subcooling based on core exit TC's 6. Teold at saturation temperature for S/G pressure
[0.2 each, 5 required] . no effect CO.2] - unaffected heat transfer area of 5/G U-tubes [0.33 2. help [0.23 greater succooling CO.33 3. hinder [0.23 - rapid cocidown of reactor coclant in S/G U-tubes forming a cold water slu [0.33 Maximum elevation of S/G (heat sink) from core (heat source) [0.25]
Minimize flow resistance - Pipe bends, valves, restrictions in ,
RCS flow path CO.25]
REFERENCE Thermo text pp. 355-358 Mitigating Core Damage Chapt. 1 pp 1.59-1.62
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I PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 23
_______________________________________________________
ANSWERS -- MILLSTONE 3 -35/09/04-COE,0.
l ANSWER 2.01 (1.50)
, Displacesairsothat[1) corrosionisminimi:ed)
l-ev hydrogen is maintained in solution preventing r^
formation of an explosive aiuture "'w.cnidLI) ) Drainage valve (AV8031) to Reactor plant gaseous drains system -
can be controlled in auto by PRT level CO.53 2) Vent valve (PCV469) To Reactor plant gaseous vents systems - if open l it closes if PRT pressure exceeds 6 psi [0.53 REFEF:ENCE N;/l glso etccc S c/eYr uf 5e'c,1 e Topic 1 Lesson 4 pp. 12-13 . .
g gg ANSWER 2.02 (2.50) *C ibe- ISI* b'C)
l (valve numbers are not required) See attached drawin valve operator normal position controlled f t on
_________ _________ _______________ _________________
V-8070 hanual open local only CO.33 SV-8095A/B Solenoid closed Control Rm Aux SD panel (ASP) CO.43 SV-8096A/B Solenoid open Control Rm ASP CO.43 MV-809S hotor closed Contrul Rm [0.33 SV-8097A/E Solenoid , closed Control Roi ASP C0.43 can be modulated CO.23 Interconnection to CVCS Just upstrean. of the Excess Letdown Heat Exchanger Jter':7 and to the PRT M REFERENCE
{c.2.& [.b o 2S)
Topic 1 Lesson 5 pp 9,10 Topic 2 Lesson 1 Fig CV-1 l
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I PLANT DESIGN INCLUDING SAFETY AND EhERGENCY SYSTEMS PACC 20 l ....................-- .............................--.
l ANSWERS -- MILLSTONE 3 -55/07/04-CCE,C.
I l
l ANSWER 2.03 (1.50)
o. A downstream valve automatically closes if hi-flow is sense CO.253 An upstream check valve prevents backflow CO.253 The piping between the auto closed valve and upstream check valve is designed for full RCS pressure CO.253 A relief valve protects the high pressure piping from overpressure due to heatup CO.253 b. VCT pressure is kept greater than 15 psis to ensure 42 seal is not robbed of a trickle cooling flo [0.53 REFERENCE Topic 1 Lesson 2 a. ps. 24 and Fig. RP-21 b. P3 28 c. pg. 27 ANSWER 2.04 (4.00)
A,0e and C CCP's Each pumpt individus1 recirc valve plus a comnor. rocire line valve shuts on SI CO.53 to provide 3reater charstn3 flow during the injection Co.5 A and B SIP's Each pumps t rid i v'i d u s t rectre valve plus a common iectre line valve are intericcked with RHR discharge header-to SI/CCP suction valves ( C 804 A / E. ) such that the rectre valves and the 8804A or B valve may nct be cpen at the same time CO.53. Thas prevents a path for c orit e m ina t e d contatnnent sump water to reach the RWST CO.5 T.$~f A and B RHR Each pumps rectre valve will automatically open if pump dischar3e flow drops below '. spm Co.5 This prevents overheatins due to running against a shutoff head CO.4 A and D CRP onlyCO.23Rectre valve shuts when flow is greater than 2000 spa and opens when flow less than 1000 spm with the pump running longer than 30 seconds CO.5 This prevents overheatins due to runnin3 against a shutoff head CO.4 REFERENCE Topic 3 Lesson 4 pp. 39,50,46,78 LSK 27-76 f'Ls (3 7+
l
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. PLANT DESIGN INCLUDING SAFETY AND ENERGENCY SY3TEMS PAGE 25
____.....________...._.....________.. _________________
ANSWERS -- MILLSTONE 3 -85/09/04-COE ANSWER 2.05 (2.73i o. Valves 8837A, 8838A, 8804A, and 8812A must be shut CO.2 each3 and RCS pressure must be less than ,&&B' psi 3 [0.23 see attached drawing C1.753 SI76~
REFERENCE o. Topic 1 Lesson 6 pg. 7 and PLS pg . 3~.2 b. Topic 3 Lesson 4 pg. 81 c. Topic 3 Lesson 4 pg. 44 d. Topic 3 Lesson 4 Fig. EC-49 ANSWER 2 06 (3.00) . Quench Spray System actuates the following at the same time DCp discharge valves open (2)
CAT valves open (2)
QSF 's star t (2 > g200 eM2/o tecc.d5 CO.2 each3 2. CRP's start after-5 -;r-t;. (to allow sun.p level to fill) all four
. pumps supply spray flow M Co.6]
3. RWCT low-low setpoint g RHR pumps trip off (to allow sufficient reserve quench spray h6 '
flow capacity)
""r" 1: _ 1:1 1:_ ::t;; E . 0 [h. e]
Ocr'_ fi__': ; .- 1 ; : ;;rt.:1!, :12 . 't :;;;-: .- '"C" 0 . RWST empty OSp's trip off M b. OSP's would be sequenced on by the Er.gineered Safeguar d Sequencing Fanel instead of starting immediately A .,5 ]o, g REFERENCE o. Topic 3 Lesson 5 pg. 25 b. pg. 18 l
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. FLANT DESIGN I.iCLUDING I'iFET- A rJ: EMERGENC7 5YSTEn3 F .'.C Z Ic
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ANEWERI -- NILL 3 TONE . -55/C7/04-COE, ANSWE .07 (2.25)
As power increases, steam ficw detector celta-p increases CO.353 As power ineteeses, steen pr essure deeresses CO.33
.
a-sta = K(P-stm)(sqroot(delta-P))
==>since the stean. pressure component stays constant (it shovic 93 down) while the delta-P incteases, indic ated stean, flow will be higher than actual stean. flow E0.43 The suminiins rie t wor k for flow will send a signr1 tc the total centroller to open the feed res>1 sting valv CO.43 As level star ts te increase. t r. e level error will s:gn ] f or the F R '. - t cicsc C0.43 Ev e r.tu a l l y , the flow artcr will be cancelled out by the Io.el error, and the TR'> will oe positioned such that steam flow equal: foe flow at s o n.s h2 sher leve CO.42 REFERENCE HP 3, System Description. Tcpic s' Lesson *r p3s 1E-;C, ;;
Topic B. Lesson 4- pgs o7-73 ANSWER 2.06 (3 00 see attached drawing REFERENCE N111 stone 3 Requirec Drawings #12 and 323 ANSWER 2.07 (3.0G, 1. Chemical addition - to add LiGH and hydracine 2~. hanval Coretion - to borate if auto n.aLeup system is unavallat.le 3. Emergency Corstion - to borate rapidly curing reactivity transients or casualties 4 Gravity Beration - to borate if SAT punps are inoperable 5. san.e as above- tut separate lines exist (- CO . 6 ea ch 3 - !? ref ssire .
REFERENCE Topic 2 Lesson 3 Rt Temn 1_6..' ks ss Sc<t / er.$ , . /c.Y .c / : . g e'r Fig. PG-2 arid p '
Top,'t 2 Quwl 0 -$ t3S t tdt \
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I
. FLANT DESIGN I N C L U D Ir.G SAFEIf AND ErtERGEr4CY 5tSTEns FAGE 27
_______________________________________________________
ANSWERS -- MILLST0r4E 3 -65/09/04-COE, ANSWER 2.10 (1.50) /3 none 2/3 RFCCW none c. none RPCCW o. Service water C0.3 each]
REFERENCE Reactor Flant Ch111ec Water oc 11-12, 38-39
. _ . . . _ _ _ _ _ . . . _ . _ _ _ _ - - . .
.
. INSTRunENTS AND CONTROLS FAGE li
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ANSWERS -- HILLST0r1E 3 -65/09/04-COE, ANSWE F: 3.05- (2.50)
c. Yes-CO.253 flow cetectors are upstream of safeties CO.253 Nuclest power will ir. crease above turbine power [0.253 and Tsve will decrease from Tref (Tave-Tref deviation alarm) CO.25]
c. 1) 2/3 high 2 contsinn.ent pressure E0.453 2) 2/3 cr.. 1/ 4 S/G Icw steam pressure - blockeo by s t e a n. l i n e SI cicek swi tc h ot- by loop isoistion valves shut f or a particular l o c.p L'.c3 3) 2/3 S/G hi negative rate steam pressureA CO.453 ,
REFERENCE a. Main Steam system Fiq. 1
[,/u [,, d w k e .., (f u. eu li- <. ST s CAF /4 /2 C 7' A/r'N"'k (_'-l') l *
c. hain Steam systeni p Topic 7 Lessons 1,2,3 pg. 63 ANSWER 3.04 61 50)
j 1) answer (c) (diaparsom air comes from the volume becster relay 1 [0.53 2) answer (b) trip open losit bypasses the I/P converte:. [0.53 L
3) answer (c) EC.53
'
REFERENCE ,
Topic 5 Lessen 2 F19. SD-4,SD-7,anc SD-7
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' IN51RUnEtiTE AliL CD'4TROLE FAGE 30 AN5WER5 -- n1L;510r,E 5 - 5 5 / 0 9 / 0 3 - C O E . L> .
ANSWER 3.05 (3.00)
e. 1) hanual 1/2 LO.33 F r e s s u r t ::e r Pressure low 1860 osia /4 C0.453 H1-1 containn.ent ptessuie : ' A r ; /rp5/n 2/3 CO.453 Low' steam line pressure 585 psic 2/3 on 1/4 CO.45]
2) hanual 2/2 pushbuttons CO.33
.Hi-3 containn.ent pressur e 10 psio 2/4 00.453
'
.
et- 2 C PSist 3) Under voltage on 4160 emet geracy bus tfor a set perloc of time's C 0. :.3 EDG 0/F breaker trics w.;11e d'upplying load CO.23 Engineered E4fe3us c.; Sequencing Fant; C 0 . :'.i REFERENCE Topic 3_ Lessorb 4 P9 E' 4 s
, Topic 3 Lesson'5 99 17, if, :' O Topic 7 Le s s or, 1,~,5 Table 4 pg L7 '
_ s ANSWER 3 . 0 (2.50's - ,
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o. 1) DN ) excessive fuel power ano temperatures - or- -Et<C f 15Ivt /(d M 3) DN ) protects RCS integrity from overpressure CO.4 each] Since RCS pressure is an input to OTDelta T and lower pressure lowers the setpoint, the low pressuri er pressdre reactor tric places a lower limit on OTDelta T range at a given RCS temperature. CO.93 i
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. INSTRuhErJTS ArdD CONTROL 3 FAGE 31
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____________________________
l ANSWERS -- rtILLS10t<E 3 -65/09/04-COE.L.
!
REFERENCE.
! Topic 7 Lesson 1,2,3 pp. 44, 47, 52, 53 ANSWER 3.08 (3.00)
a. rods out CO.253 Tref will be man so Tave/ Tref mismatch and NI/Turcine power misa.atch will both si.e a rods out signal CO.753 l rods in CO.253 Loco 1 Tave increases and auctioneered hioh Tave also incresse T a v e. / T r e f n. i s a.s t c h gives a rods in signal CO.753 r e A.r. ro T-c . .a e.-eaenee- C O . 2 5 ] the pcwer mismatch etreult of the reactor control unit
,
resporeds only to r ett o, c h e r. e cf devisticn betweter. turoint arid nucit a r
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'
power G . ' C .r b .i t r+ i m t h E u!.'78 t h *' dA. fc y-[v 7"i, ,a RE F E F,ENCE [b. 7'O Topic 6 Lesson 2 Fig. RS-5 and pp 55, 19, 20 ANSWER S.09 s5.00) ) The containment structure nach rance internal monster (SRnsmRE04/05)
[0.253 locatet ir.r.loe the centainmerit above tne oper sti flost (0.".53 supply cata on ractation levels inside containment followino an sec1 dent. [0.253 2) The fuel drop accioent monitors (3RnS*RE41/42) C0 25] Iccated inside the coritainn.er t up st r ean. of the put se duct a r.t s k e C O . 2 5', .
isolates the containment purce ssstem and stopt purse ar;c ennsust accident C0.253 isns M in tne event of a fuc1 dr op b. Diverts the effluent cf the turbine buildino some from vard orains to the tur bane F1snt C on pone nt coolitis dr a an sun.ps CO.53 c. Chan A - iodine (particulate) C0.5J Chan E. - rioble 3as CO.53 REFERENCE c. RMS System Description oo. 14, A4 b. ps. 66 c. pg. 33 l
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- Ir45TRUMENTS AND CONTROL 5 F A r.E .L'
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ANSWERS -- nILLSTONE 3 -65/09/04-COE, ANSWER 3.10 (2.00)
o. Reactor' trip and start both HD AFW pump b. Reactor trip and start all AFW pump c. ' Turbine trip and FW svstem isolation (no reactor trip because 40%<F-9). (Control system see's large decr ease in delta F and acts to raise it to s conform'with programmeo delta Fi> thus main feeo pump speeo speeds u .5 each3
,
REFERENCE Topic 4 Lesson'2 p pp. 19, 23 Feeowater chapter po. 22 Topic 6 Lesson 9 Fig. SL-3 and pp. 10-12 (NOTE: A traininq material discrecancv exists'between Feedwater chant o : 22' aric T opic 4 Lesson 2 pg 19 regar ding the n0AFW star t on low low 5/G 1evel) CAF l
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. PROCEDURE 5 - NORnAL, ADnCRnAm, EhERGENCY AND PAGE 33 C5sisoE------------------------
' -
--- REciBE55iCEL
ANSWERS -- MILLSTONE 5 -55/09/04-COE, '
ANSWER 4.01 (3.00)
a. Anv une::pected increase bv a factor of two or creater (will also accept any unexpected increase) CO.753 b. Use letdown flow through the Regenerative heat exchancer to oreheat the water to avoid tnermal stres E0.753 c. Loop 2 has the surae line and one sorav line associated with it. CO.753 d.-1) reo oper ations are per r.itted thct would cause boron dilution and 2's core outlet temperature is maintained > 10 cegrees F subcoole C0.753 REFERENCE o. OP 3201 p b. ps. 11 c. pq. 24 pp. 26,27
' ANSWER 4.02 (2.00)
s. 551 des'rees F E0.53 minutes C0.53- restore Tave to within its lin.It within 15 min. CO.53 or be in HOT STANDDY thode 31 within the nent 15 minutes E0.53 REFERENCE Tech.-Spec. 3.1.1.4 and 4.1. (note: a discrepancv exists cetween T/S 4.1.1.4 and OP 1207 orecaution 4 . 5 's l
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. PROCEDURES - t40Rr.AL. AE:143 FN L s EhEF.LENCY A N ? ;, L :. 54
= RADIOLOGICAL C0hTR3L-ANSWEF.5 -- h1LL5'Or4E 3 - E.5 / 0 9 / 04-C O E , L .
ANSWER 4.03 (3.00s a. 230'destees F ,'5 niinutes E0.5 each";
b. . lower reactor power (to < P-8), and stop the affected RCP CO.5 each3 2 trip the reactor (estry out E-0) and stop the affected RCP E0.5 es: M ,)
"'T'"%54p,. 3 os: Tr.c. %s,- A< + uc r +e,q.<o,$ le b. OF33010 pq. 32
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c. OP33010 p3 3; g, ./ 3 g. p ,
ANSWER 4.04 (3.00)
a. Fire'ha:ard E0.53 due to e:< c e s s i v e oil buildup in eenaust lines E0.53 erankesse pressure iacket coolant pressure Lube oil temperature
.iacket coolant temperature CO.25 each3 low ~ lobe oil pressure ( 2 / 3 ) U .23 --
( h e& D (I'Y * <m y b'b
'
engine over speed [c ,2] : ~, *
r5: '6 f 6 REFERE14CE OP3346A para. 6.2 and'6.5 p system descriFt2on. EDG pp. 41-45
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' PROCEDURES - NORnAL, ABNORnAL, EnERCENCY AND FAGE 35
- ________________________________________________
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RADIOLOGICAL C0rJTRGL
____________________
ANSWERS -- MILLSTON /09/04-COE, ANSWER 4.05 (3.50: ) rod bottom lights lit 2) reactor trip creaker and bypasses open 3).RPI at zero 4) Neutron f lu:< decr easino CO.25 each] trip bus 325 ar.d 3;N load center supply br eaker s E0.53 )~ start a CCF, or check running 2) start both BAT oumps 3) open emergency oot at;or. valve si open charg2 res flow c or.t r e l valve 5) ene:L pressuriner pressure s 2350 psia [0.4 each]
REFERENCE EOP 35 FR-5.1 pp. 3,4 ANSWER 4.06 (5.00) ) From EOF 35 E-0 REACTOR TRIF OR ST when minimum AFW flow is not verifie [1.03 2) S/G NR level - 35 s with total FW flow to S/G's < 275 opm (red oath condition). E1.03 ECCS shoulo be alioned for cold leg recir C1.03 REFERENCE EOP 35 FR-H.1 p ,9
' ANSWER 4.07 (2.00)
1) RCS subcoolino based on core TC's is less than 35 F CO.753 or-60 F or adverse c o nt ai nm e rit [0.253 2),pressuricer level cannot be maintained creater than 7% E0.753 or 50% for adverse containn.en [0.253 REFERENCE Foldout for EOP E-0 series procedures l
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.
. PROCEDUREE - t4DRh AL , ALNORnAL. EnERGENCY AhD PAGE Sc-RADIOLOGICAL CONTROL AN'SWER5'-- h!LLST0r4E 3 -85/09/04-COE, ! ANSWER 4.08 (1.00)
Tne qualitative assessment of a enannel durinc operation bv observatio This shall include, where possible, comparison with other i nd e p e r.d e channels and/or indications measuring the same paramete E1.03 REFERENCE Tech Spec definition 1.6 94 1-1 ANSWER 4.09 s1.b0)
1) the intent of the ortatnal orocedure is not altere [0.53 2) the chan3e is app;oved by two member s of plant r snager ent staff, at least 1 of whom holos an SRO license on the unit affecte [0.5]
3) change is docun.ented and revieweo by PORC/SORC as appropriste and
. approved bv Unit / Station Superintencent within 14 cavs of imolementation E0.53 REFERENCE Tech Spec 6. ANSWER 4 10 (1.00) Individuals read and 1,nitial the RWP indicatinc an onderstandino nf work l conditions,: radiations and contamination control [0.5]
l Loss time in and out ano individual cosimeter readinas and total time l and enposure after thcir Isst entry. LO.53 l
REFERENCE l SHP 4912 Rev 5 i p9 9
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4 PROCEDURES rJ0Rn AL r A B r406 n A L , EnERGEr4Cf Arde PAGE 37
~~~~Ed555L55i5dE
____________________
C5s?E5E------------------------
ANSWERS -- rtILL5 Tor 4E 3 -85/09/04-COE, ' ANSWER 4.11 (2.00)
a. radiation area (> 0.5 mrem /hr) [0.5]
neutron radiation area (> 2 . 5 n.r e n./ h r ) CO.53 3, ...nn _ _ _ _ , , . . - ____z._ .mr < _ . - -
_
[0.5]
c. HP supervisor iders snee E0.53 REFERENCE n. SHF 4/06 pp. 2,3 anc.c. 5HP 4902 c, c . -
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_________________________
REACTOR T(FE; FWR-WEC4
_________________________
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_________________________
EXAhINER: NORRIS, .
_________________________
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. _________________________
INSTRUCTIONE TO APFLICai;T:
__________________________
Use sepsrat,e paper for +he ansaer Write snswers on ont side onl Ettple question sneet on top of the sosser s r.e c t s . P c i r. t s for etch question are indicatec i. n parentheses after the ques *,1cn. The passing 3rade r equires et least ' O *. In each cetegory end a final gr&de of st least 80 Examinstian pape-I will o'c picked up si: i c. hours after the easminst 1: n start OF
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L n. h. i-DFERATION. FLUIDS. A b' D T H E R -i 3 C 'r N A n IC E ne -
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AND INSTRUMENTATIOr,
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. PEOCEDUREE - NORMAL, ADNCRMAL, EhERGENCY I.ND RADIDLOGICAL CONTROL
"
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_ _' _" _ _ ___'_"_'_'C'_ __ ___________ ________ ADMINI2TRATIVE FFOCEDURES, CONDITIONS, AND LIh1TATIONS 100.00 100.00 TOTALE
________ ______ ___________ ________
rINAL GRADE ________________ *
All wcrk done on this e:: a n i n s t i o n ts ms cun. I hsse neither 31ven no: received ai ___________________________________
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THERnGO nan:C3
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Q UES TI0 ri 5.C1 (~.0 In acecrdance with 10CFR50.46(b,(1), clacains temperature is not pern. tte; to e:,ceed & certsin setpoin What is that setpctnt - :0.53 Why is this limit impcsed? [1.': Wha'. hsppens to tne :ltcalcy as temperature increases pa st the n. e l ', l o g rcint? CD.53 OUESTICN J . . : "> >
The reacter is c p e r s t i r. g s ', e : c . - s t a t e s t. 50% ,:curr. steam gen:ratcr pressure is 102: ps13 How n. u c ." w:..lc Tsag have to incr ease 10 causc t h :-
fir st stesm a c r. s t, c r tsfety ssice +c lift w i t.- power constant .s t 5 0 '. '
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Stste all as s ump ti cr;s ans s. n o u sll u;r GUESTION 5.03 <3.CC U p l a r. why dels.-ec neutrons .ts s e such a large effcct en r 4 a c t,c r operationsi incivce in voor 6:scursacn the produc ti c of both prempt and d e l a ,. e c neutron : ....
t; . Wts cres E+ t s- c 4 1eetive chsnac o v ;: : cor e lifei '
E 0. 0 3la kf Calc.;1st.e anc e .. p l e : .- . .w 'ssts ter ti,e start-up Ra c- 1.ti c. e c i a t : 1-A following a r es etor tri .C3
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u, u --t.:, . A C n .., . v 4 a .u..t, 1
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Assume that the p; ant has ,ost coc.pletec a n.aintenance perlad cf 70 days, early i r. cere life; E::p l a in why the i r> t r i n s i c neutron sources alent may not be sufficient to n.e e t all requirements for s reactor startu [2.03 What are the installed sources and how do they react to previce neuttoris f c' s t e actt: s tar tup 0 [1.03 ( A .* * * * C A T E C O R ': 25 CONTINUED ON NEXT FAGE .*****,
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GUESTION 5.05 (3.50) The re actor ecclant s y s t e Is maintained withir. sp ecit'ica t ton s by use of the f oll ow1 rig che::.i str y contrels, enpla.in each r e acti cri '
E5.03 (1) Lithium H y c.r c': i c e - pH control (2) Hydtanine - c u y g e r, c or.tt ol (3) H y d r o g e r. - cxygen contrcl m Why is withium-7 used vice r.s t u r a l occuring Lithiup> [0.53 GUE3 TION 5.06 ( .,C'
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-6 You ar e i r. the pr oces s cf start:n3 ut tne p l t r. t witn t' e 2. t o r power at IC smps. The reactor cparator i n a d v e r t e r. ',1 y moves rods in for ~_0 step A S s unie roc Werth is o p e::. < 1 ri a n d s ri aveisge neutron p r e t i.i r s c r .d e c a. y c o r. s t a n t, cf 0.1 see-1; wit:. c. c forther operstor s c +. i n c : Wnst 25 the retelting S U i r.inic c i s t e l y a'tc; red r.c t i o n s t e p ~~ [1.25: What is the p o t. c r stter cu m i n u t ,: ~ CO. 52 QUEETION 5 . 0 (2.0D Definc cendensate c c p r c : 2 : c c. . C1.0: Define : s . i t. s t i o n e e :. p l z i r, how it :s c f f :- : t c t' by cterectin3 condensate deptessie [0.T5: How is p l a r.t efficienc, affecte-; c ;, eccessise ennoenstre c e ;.r e : s l o n c- ,
Co3 4 e. sd 4 (**zt* CATEGORY 05 CCNTINUED ON NEXT PAGE *****)
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r . THEOR'r GF r.UCLEAR F 0WER F ,_ A ra 7 0FERATION. FLUID 5, A;K : AGE ,
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GUESTION 5.08 (5.00)
The plant is operating at 5 3 !. Power when the il 5/G nain Stean. Isclation
alve f ails shut due to loss of electrical powe .
Usin3 tne below inat;al conditionst calculate the new itead state values for the listed parameter 2 Assume no op.erstor at t i ore. ;;' -- : , : :: . 2
_ . . - .. : , and rio r eactor tri State a'l assumptions and show all ucr Initial conditions: Tass = Sc7 F / g, /,,,, c, , / u /,r , ,, a u /c m ,/ec
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(,, .m . . . . . - $.cr e,, ,- .
"'W r<sec h ' e en /nl.1 < h .m nue. f 6 Tur bine power Tav3 (for the affected 1ccp Tavg (f or the ritn-af f ec ted loops /G pressure 'for the affected 1 capi E/G pr essurc (for tne rian-af f ec ted loops-GUES TIO ri 5.07 ,~.0 Indicate at which time in c cr e- life (COL cr EGL) the following a c c i d e r.1, s ar e r.o r e se .e r e . r e sul t :n a longer t 2 n.e spent 5t h;grer peuet ).
E.r t e f l y enplain wh Main stesm line br ed C1.07 Tetal loss cf coela-t 1;;e :1.D3 Rod wit.7drawal sccide-t frcm Icu in the scurce rsnge pr:ct to -s c. .
signifitst.t r e s: tor coolar.t tcmt er atur c increas .03 GUESTIGN " 10 t1.50)
Explain how incicated reacter scwcr would cnan3e (incresce, 2ecrease, cr ren.ain the same) for the following condition A s s ur.e actusi powtr r e n t i r,s ccnstant at 75%.
CONSIDER EACH CASE SEPARATEL Cooldown of the reactor ecclan Core agin Decrease in boren concentratio ( z s z. x x EN- Or CATEGORi C5 ***.sa
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. FLANT SYSTEhi DI51GNr CONTFCL, AN: INS TRUr.EN T A TIOr> CAGE 5
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GUESTION 6 01 (2.23; What are the cases for *ne first and fifth coce s a f .: t
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retpcints? C .5" How niany coce safcty .alvts per s t e e n- generator e c r, be c e c i a.: e c; ino during power o p .E r a t i c r. ' What 1 :::i i t a t t a n s tre placed on
- gp. poperatians?
eraole (setpcInts E t i r.i e f r a n.e s r,o t requi ed) E0.~53 c . C ,.,
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The reactcr ts ope st:ng at a steady st6te 25', power, all contrci systems ar e in attonati Turoine Icac i t. 1r. reaseo tc 100. and ;. n a c ar,tr ol l i n g steac, prestorc cetector f .; r 41 E ,' O s t i c i. s s i. tne Z :. . a 'i v .: . 2 ;.p l s i n , in detail, now ana wh. t h a a. L1: 1 e f f e c ". 41 stern gene r stor level - c a s.u ni e no cperator actio State sli 22:cmpricns ano shoi. ail w o r (A naoner,ea l ans wer is Mc 7 r ef a 'r
DUESTION 6.0: ( 1. 5 0 ',
C e s e r t b e: the location anc purpcse of the labsttnth seal: cc ,b : R e s c t.a r Cool ant Punip s .
QUE3 TION 6.04 (5.C0, Wnat 13 t h -; eignt: icance of the "FZF F E'L TH TEhF rC ' s l a r ai What sctioni shoult y o .i t El c - [1.03 Why ts r. : t r c y r, a c c e d to t,h e FFTT C1.Gi Give the s e r,ve n c e of e v e r.t - es pressure incret;es t r. e PET - include setacints as spptopr:st C1.03 GUESTION o.C5 43.00e The reactor is op(rat.n3 at SC'; p cwe r when the locp f;cw circuitry for loop 1 1tils lo Will the r esetor i t 1 p
Include setpo:nt,s a r.d .'o r logic as apprcFritat C1.OJ Same question cs above except that power is at 3 E1.0]
- . S a nie questicn as abcve encept that pcuer is at 87, and loop 2 circuttry Elsc fs11s lo [1.0:
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. FLANT CiETEh5 DEE!5N. CONTiOL, A N L: I N 5 T F. LP,E N T AT I O N FAEE o
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QUESTION 6.04 11.50)
What three ccnditions are nece:ssary bcfore the s t c s a. cump control s i s t e n. 12 sble to admit s t e t.c to thc- n.a l n concense' for cooldown cel . 553 F?
-GUE5 TION 4 07 (4.00) Indicate whethcr the OT Delta-T anc the OF Gelta-T sctpoints will i n c r e s s t- , cecrease, or not change if the f ollowing changes cc cur .
CONSIDEF EACH CHANGE INDEF ENDENTL'r . Pressursrer pr es aut e de^teasas 130 ps:'. The N-41 lower c e t .s e t o r fatis ic . Dserdilutien of the RCE. whicn causes rods to : r. s e r t s ' oi r
to maintain constant lcac anc Tav E ,0] Justify you? answers f or p s: t 1. accv E2.03 GUE5 TION 4.03 ( .GC) Howe i s s r. cperatore c3n you tell if power is lost tc the i n t e. r m e d i a t e ran3e detector prior to a r e r.c t o r stcrtup? E0.7EJ What kould be the significance ci cne IF: channel ceinj overccmpcnsatec duri reactor star tupi 0.5J What Lculd ce the significance of one IR channel bring undcreempenss+cd dur i r,3 a r e ac tor shut oown? :D.5J If one IR channel fail- . 413 n th eir,3 a r e.s c h e r shut 0-n, aca ccn the i operstor continue to 5.hutdovn?
E i,. _,r.,,.
QUEETION 6.07 (I. 50) Using Figure c.1 (sttached) describe the preferred and alterr.ste floupaths for eras r 3ency borctio :1.5J bo List the conditions that require emergency boratic [1.03 l
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6o F L ANT S Y3 7Ed5 DES IGr. . : G ;4 T r,0 ; , AN; IN ;TRUdENT ATION PAGE 7
GUESTI0tl 6.10 (3.00>
Sketch the flow F.a t h cf clectrical power fron OG A to bus VIAC- Label s11 busses, c o ::.p o n e n t s , ar.d breaker Indi ate vc1: ages and show alternate flow path Breaker numbers are not require xn=2r END OF CATEGORY Of a n * z * ',
. PROCED:JEEE - NCRnAL, ABv0RMALr E i E F C E N C't AND .: AGE 8
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QUESTION 7.}1 (2.50) What operator actions are rcquitec if criticalitv is achieved prior to t h e n.i n i n.u m rod position on the Estinated Critics 1 Position caleviatio (ECP)? ;-) . 5 2 Wnat if criticality is not achieved oy the m a::i mu m rod position) EO.52 The Reactor Ccolant System lowest cperating temperature s72.3) is not allowed to sc below 551 F during a reactor startu Whzt is the bases for tbis limit C1.53 OUE5 TION 7.02 (3.001 In acccedar.ce with CP3346A ' E.u r g e r.c y Siese' Generator *: if the EDC ts oper ated at l e s s t h ar. 50;. c e p a c i t y f or 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 2, the d:etel should be operateo at greater than 50% for one hou Wh,? [ When-opetttir.9 the EDG in the e n.e r genc y mode, the sa c ke t Coola Temperature and tae CranLcasc Pressare shatcouns sre cutout cf the trip circuit. 'Wnat a c ti or.s e si any, c e r. tne ope r a tor tste t ni a i n t a i r, tnese p a r a n.e t e r s withir, lietts a r.d therefore minimize dan.33e to the EDG?
A s s u r..e t h t t t h e Iced cannot be o c- c r e a s e d . [1.03 List the tua autcastic protective f ea tur c s that will alwa.n stop the ED E1.C]
GUE57 ION 7.03 ::.;5, You are the on-duty '5 C O a r.c a Radiaticn Werk Fore.it (RWF) 11 trought to you f or approva In acconcar.cc with SHF 4712, your appr os al z.: 3na tur e signifies what three conciticns with respect to the plant?
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. PROCEDURES - riC R M A L , AENGRnAL, EriEF GEN C Y AND CAGE ?
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RAOIOLOGIC:m CGHTFOL
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GUESTION 7.04 (2.50)
In accordance with EOF 3503 "Shutdcun Gutside Centrcl Rocmr' answar the followin3: How do you verify the reactor trip? :1.02 Why does the FTCCedure caution e,:41nst placin3 the e r.e t ee n: , switcngect busses in local' C0.52 How would a cor tr ol voor. oper ator know if contic) hsd been shifted tc the AST C1.03 GUESTICr, 7.05 (1 00 In accorcance with EOP 35ECA-0.0, what are the uimediste actions on a 1 css of sll A/C power 0 GUESTION ~.06 ( 3 . " ',
In a ccot t'ance wi t h GF 15 : 'Auxillcry Feetuster Eystem'i What acticn must be t a k 'e n to feed all steam :enerate:s trcm cne motor driver, Ava111er y Feeowawr P un+ i (~/, 0 ] E'.:: You are feeding the s t s. a m gene.catcrs during a plant cccidown using thc turcinc cravon tunil2xry 1ee5Leter o v n. ;. f011cwing e iost of til A/C power w h e r. .ov rsce the OW2T LEVEL LC-LC alar What ections woult you eapect to be teken? [ /,o,3 M What disadvantases 3e si.sociated with the alternate watcr s.. py lie s ?
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[p, c u GUESTIOr4 7.07 (1.00i Answer the fallcuing with respect to GP 3330 ' Reactor Plant Cccponcot Cooling Water' E ADP 3561 ' Loss of Reactor Plant Component Cocling Water':
, Why should the shift fi co A cr E train RPCCW to C Swing RPCCW pumF and
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heat enchan3er be done swiftly? [1.53 i
l What are three symptoms cf a loss of RPCCW? C1.03 Why must you limit RPCCW ficw to 8100 3pm? [0.53 (***** CATECORY C7 CGNTINUCC ON NEXT FAGE *****)
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I PROCEDUREB - NORMAL, A E;N D R n A L . ENERGEt4C1 AND FACE .0
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GUESTI0t* 7.05 s1.50)
In accordance wtth OP 3319A ' Condensate System *: When init2 ally starting up the s y s t e ni . why should you ensure proper motor amperage en the pump before letting go of the switch? CO.752 Prior to placing tne systen; in oper ationi if s t e a ni generator pressure is less than 700 psis, ensure the n.ain feedwater isolaticn valves are close Kny? CO.753 GUESTION 7.09 (3.53; Answcr the f 11cwing i r, accord 6nce wito AGF 0562 ' Loss of Instrument Air'; The plant 2s opercting st 53% power when you rece2ve the CThi li4 S T AIR FRE55URE LO elarm. ycur pressure indicaticr. confirms that pressure is lo Ali S ut o n.s t ; c ertions ital to nappe What operstor a:tions wavid you enFect to c4 taken to restore air ,ressure' : Wherc can tacse
&ctions be per f or meo ASSUME that each sction t E l- e n fails to r. top the
- 2 r e s s u r e dro C1.52 E
- rlefly e;plair. hou the following s y s ten.s will be affected by a loss cf C o r. t a i n m e r, t , A i r if tne p i a r. t is operating at 5 0 ?.
Ir,s /reme&
Feed Water CVCS RCP Seal Water C1.5] How would RHR be effected by e loss of Contai nn.e nt, Ai r in hode 4) :0.53 i t,, , / rso aner, / 1 (***** CATEGORY 07 CONTINUED ON NEXT FACC *****)
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RADIOLOGICAL C ; r; F. J L
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GUESTION 7.1C (2.75i In crder to n.atntain t, h e plant at 100% pcwer, work must be performed inside the c o n t s i nn.e n t i r, a radattion field of 850 MREh/HR 3 sn,m s a,r.d 3 0 hR A C .'H thermal and fast neutro The c.aintenance man selectec is 25 years old ar d has a l i f e t i r.ie exposure through last quarter of 43 REM on h i s N F: C F o r n, 43 additionally, he has accumulated 1.0 REM so far this quarte How long n.sy the n.an work in this ar ea without enceeding his 10 CFR l i n. i t Show all wor ti .20 During a decisted en.ergency. this indi,1dval . ol ur.te c t s t :. a r.t c r
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a high raci stian ar e a cr.d p er f or m work necessar. to ,t'e.ent further etfluent ieieas In ac c or ds.nc c witn the E t a 11 :ar. F r c et d u r c . what is his cia n i m u r, allcwed whole bot ext.osure3 CO. G Whose suthc.r i: s t ion I r: necdcm in pit t [0.70 l
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QUESTION 6.01 (2.50, The plant is operating at IC'. Icad anc ycv are the on-shift 5.uft Superviso Enplein what actions. If any, would nced to be caten for the following condition CONSIDER EACH CA!E $EFAEATEL Your on-shift E:0 P op e r r t orests his les ir. the plant arid you se nt him to the hctpital for t r e a t rie n t . C1.052 The on-con.t res ST A calls and u ps be won't t.c i C1.!5 j l
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QUESTI0ri 6.00 1.75)
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The Technical Specificatiant, stata that ' E .., c h Evaveillance Requireme !
sh&11 be perforn.e6 with1r. tne sFOC;.fied tibc :nterial with; ! A m a n i m u cwable entenston nct to er uco 25*. of the servet11:nce interval, but The combined time interval (cr anv t rir e consecutivet sor vc t ilence:
intervtit snt:1 r. o t e ,:c e e f ?. 25 tin.cs ths : p o : s f i e r.' sur.,9211ance intervel.'
What ave the ba, set for the abe'. 9 L
GUESTI0tl 5.00 c;.25 <
In accordance with CF1 01, 'The reactor c o o l a n t p r e s si. r 4 and t emp * * a tur et i shovid not euceed ______ptia and ..,___r when the Fetidus) Hest E ( n. o '.1 1 syste:m i f, t ri se r vtcc. '
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Fill in the blar.ts enc provide the betts for ea: OUEST~CN G.0c 2.00i Answer the fc110 wing as they apply to ACF-GA-0.004 'Statten T433tn3'! i Undet what conditions uovic you authat 1::e ' Or et a t or - 2 n- At tenosi.co '
l tagging? C1.03 What settons niv e t be taten if the Operator-in-Attendance n.ust leave the :
area? -
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I r. accordance with EFIF 40;(A & 4510ft:
- 4 As the Entft Supervtscr. ween n.ost you a s s u n.c the ren cnsististics c'
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00CCTICH G . i: P /: .00;
l Ant.use the fallowiny -tth re n ect to Acr-0A-;..cC 'uark cro.rs':
1 Whteh of the below at rat a rccpc.s.o41st, of the 05,!001 P. 0 3 1. F1 ace app:cpriato 6 4 f . r. .- t 4 -) w 2. Ids. ret i f y t re. t I ;' * e g u t r p r.. t r, t :
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fer Technical Grectricettoe, wnen coes ,:ontatnn.rt A n ta p i t , e .i t , t *
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The safety I t n. i t s for the re4ctwr core are deters.tned fron, a e c na t r.a t u n of resttot toutre pttasuf1:et pretsutte a n t) loop t e n.p e t s t u t 6 .
l If a safet,y Inntt in violatede what t pis a d i a t e a c t i o n n u f. t be tek sh i l
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CO.53 l Wnet are the tasett f or $ttyshj withtr. the reactor safety i n na t tervet"
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The reactor plant is at 05*. powe rour permts ton i t, r queetted to coniseenee wor 6 on orie of the RCS a:cun.ulatorsi this wori will involve ;
the partial draininj of the a t:comu la to r ! j Would you authorin the wori? Eaplair,. ti.0) f t What is the dest 3n 1: 411s for the accumulator volutt.oS C1.03 ; What 4 tions would ytv tais for an LI ACCunULATOR LEU"L LC (1sta.? L . 00 l l
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Equations p= 1 Q = Mah +1 tK,ff I +1r e e Q = mcpAT SUR = 26 e t' + (5 p) t a = UAat
,
.
hg = Ky* C3 (1-Kg ) = C2 (I'E2 )
P = Po 10 sur (t)
3= 3.14 p = pa ,t/r i
e = 2.72 >
'
SUR = g CR = S t
1-K,ff 7 = j ' + S '~ @
1 = CR
' '
M CR
1 = 0.693
,, , ,
th 1 = 10 Conversions I curie = 3.7 x 102*dps '
1 kg = 2.21bs 1 gal = 3.78 liters 1 gn/cm 8
= 62.4 lbs/ft*
1.in = 2.54 ca 1 ft' = 1.48 gal 1 yr = 2.15 x 10' se .
I gal. = 8.3453 lb MW = 3.41 X 10' BTU /HR -
-
.
, THEORY OF r!bCLE Af; FCWER FLANT OFEEATION, Fuu!OS. AND FACE 15
________________________________________________________
ThEnt
_____'0DntAMICS
_________
ANSWERS -- MILLSTOrlE 3 -a5/07/04-NORRIS, E: . ANSWER 5.01 (2.00) The limit imposed by 13CFR50 is 2200 CC.52 The 11 nii t is i n.p o s e d t o n. i n i r.il :: e the e f f e:: t t of the srceloy resettori with Water'stest CO.51 ,which 3 i ses off heat and hycrogeny CO.53 (Zr + 2H2O --; ZrG2 + 2H2 + Hest) F "w A.e 6 c ours c lar/</.oj e.n 4n#/r e"e 4 " T ', e reaction will becenc self-sustain ng (sbout 4320 F- [0.5:
REFERENCE 10CFR50 46(b)(1)
MFS System D e s c r i p t i c r. , ~cp;c 7. Lessons 1 - 0, page 4 NNE Mitigating Core Der. age. Lessen 1 pages 1.50 & . 51 '
AtJ5WER 5.02 (2.30>
Steam ganera,cr press.ure of 1020 psig = 1035 psis whicn re16t:3 ta 543.7 F (200 psis = 54 F1 [0.53 (1035 psia = 0??07 F)
(1050 pria 2 55 F)
F t r a t, S/G ssfetv valve lifts at 1155 Ests = 1CGC p:u ?.; F CO.52 Steam ten.persture must aner esse 15.5 F l0.53 (567.2 - 546.7 = id,5)
Therefore, Tav3 must inctsas( by 13.5 F C).52 REFERENCE nF3 System Cescriptior., Topic 1, Lesscr. 3, pages 39 & 41 H P 3 S y r., t e Des er ip t 2 nri, Topic 1, L e s s o r. le page 37 Steam Tables i
r b _ - - _ - - _ _ _ _ - _ _ _ _ _ _ _ _ _ _
.
,.
i l
r f
' THEORY OF NUCLEAF F0WER PLANT JFERATIOti. F i.J 1 C S . ANC PACE 10.
,
.... ..........._............. .........................
! THEhriOD f rurt!Ci
..............
l ANSWERS -- HILLST0HC J -05/09/04-NORFIS. E. 3.
!
, ANSWER - 5.03 (2.001 Prompt neutrons are produced almost immediately after fission 00.253 with a very short lifet.mc (2 x 10E-5 see)(3.253 Delayco neutrons are produced from the cecav cf daughters of fitsten prodvetsto.253with a much Acnger lifetime (12.7 sec) [0.253 Average neutron 3eneration time is approutmatelr equal to th) delayed neutron tern, CO.;53 wM eh it why deleyee neutrons (21&nt.c.11y c :,rit t o f the re sponse F. t e c of the reactoa [0. !] Eeta-eff = E: e t a - t. r t n In pc.r tanc e isttot Beta-b.ar is a weightN a. era 3e of the Octa's of the different fuel types (U-235, U-256, Fu-;3i (0.2538 as the ccre eges.. the pcreeritt3e of Fu-239 incacases which avses the value ot' Eota-o.sr to decrease ever etre lifet0.053 The SUR immediately following a reactor trip is based Fi the lon3 cst 11ved del eye d neutr or, pr eevr s ot with a r.siflift of "55 sec CO.53 i e 55 s4c/0.673 = 79.3o a *00 See C3.;53 BUL = -26.04/T = -26 04/;4 - 0 . 3 : 3 c. p [0.053 REFERCHCC
tit 3 System Description, Top ic S . ;_c s son , pg ;0 Resetos- Theory E.c o L . Topar p 3 n. : - 4 1. figurt 4 '?
AH5WER 5.04 <3.C0i N ' "'/" "J h *' re '8 4 * W M"*1^!N ,..
-. : :: ,- ; , .: , , Mst /1 due & * (:--theMw
_ _ . . __ . .. G*fl-
' -
. :*~
,
..-t;--:' '-!.
'
. I t .- . z , n ; . ':- _; -- - ' -
~~
. ;- *i:- '... 6cesuse of that, there may be an insufficient neutrer level te provide the t egoir ec' Sout ce Ren2e indication cr, s t a r t .ip :1. 0 be Primary source: Californium-252 CO.053 Reactiont spontaneous fissior, t0.253 (Cf-25; --? FF-1 + FF-; + :n)
Secondary source: A n t int o ny-Ec r y l l i u m to.253 Resettent photo neut r o [0 053 (Sb-124 - . Te-124 + G a ni n.a ih13h energy >>
( B e - 9 + G a n.n a - 6 2He-4 + ri)
i i
i
I*
,
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__________________....... ...____.......__ .......__.___
T U C R n L C ': N A r. lC G
...... .. . AriSWERS -- r1ILL3 TONE 3 -G5/09/04-NORT 0, RE F E R E tJC E Reactor Theorv Ecol., T Pi: 2. pgs *4 E, figs 5-11 ANSWER 5.05 (0.5C)
i 4 (li L 1 - O H d i s e t. s u c i s t s.- s i re witer tc form Hycronyl t o r. s C O .M , which shifts the c or e c ritr a t i ate o f Hy d t o r.1 v n. lores to coritt ol the pH level of the RCS C O .{a (pH r 1/CH00+ )
(;. H y rJ r + :: . r. e t ( a c '. s with c :: r g e r. t c i e t rt. r.2ttc. p r. a r a. wtter L O .X *
(.h t . 12 restricte. t.o u s, a o r, s (, a r t y p s t'rca. cold shyf.co.cI ~
(3) Hydtopere ies:t.S with i j.c os.y gers to fetr ueter t { ., wher .r tac pres.rn;4. at a high gamru f l .n , CO.73 '
G Do (J a t o t a l 11 t h a v rd Bisis L:-o which I o f R. 6 t r a t 1 u n, th-L' whe r< 100J0Cttd to 4 h a y r.e u t . o f lv N . '* ]
R2 r E F Eh'CC rii .3 C y S t e m C E S. r i p t. t a r. , I4pic !, Lesson 1, pg) 5-6, d 13-14
.
k THEDRY GF NUCLEAf POWEF P_ ANT OFERATION. FLUILE. AND PACE 1E
_______ _ ___ _ ______________________________________
______________
ANSWERS -- MILLSTONE 2 -85/09/04-NORRI5, E. ANSWER SsO6 (2.00) one step = 5/6' ==2 6 FCm/in = 3.75 pcm/ step CO.12 Beta-eff = 0.00o9 CO.15J 1ambda = C0-151 tho = (3.75 p e ni / s t e p ) ( -2 0 steps) = -75 pea or -0.00075 [0,1 T = (Eeta-eff - the)/(rho)(lacbda)
= (0.0065 - :-0.00075})/(-0.00075)(0.1)
= (0.00765),'(-0.000075)
= -10 CO.5]
EUR = 26.3c/T = 26.36/(-1024 = -0.255 dpm [0.25 ( S U R 'i ( t 's P = Fo10 CO.25:
-8 {-0. 2 5 5 } -!1 >
= (1C )(10 )
-6
= (1C )(0.556) s
.-?
= 5.56 - 10 a ra p s CO.51
,
REFERENCC MP3 - Sun, mar y c f Core Nuclear Characteristics Resctor Tneory. Topic 10, pages 5-5 MP3 System Cescription, Topic cr Lesson 27 pg 17 MP3 - E1mulator Cur ves f or Intest al Rod Worth ANSWER 5.07 00) , The d i f f e r e r.c e in t mperature for the e::: sting condenser .acuun and the teraper atur e o.1 the condensat E .03 As the condensate enters the eye of the condensate pump, thE fluid flashes to stean, due to the lower Pressure; as the stean; bubbles travel the length of the impeller bladest they collapse (causing excessive wear). [ Decr easing conder.s a t e depression causes increased cavitatio [0.25] Reduced pl ar t ef ficienc E0.25]
REFERENCE Millstone HTF~. pages 152-163
. THE0R( OF iJUCLEAR FCWER FL A 4 T GF EF ATIOt;, FLUI:3, A tlD i..;E I?
TH E Rh0D r N Ar1IC 3
_----
AtJ5WERS -- hlLL5T0r!E 3 -35/0~/04-NORRIS, E. . RIhrnhst u,'Wr *." ,s l a ,- t,,. . ', aw L < .f -
At1SWER 5.08 (3.S0) Turbine power - stays ccns f ' tant $/:f4 re /// "Le 'B r4 4 ce</
at 33% power EC.63 Tav3 (affetted l o c. p ) - incteases [0.23 final value - 577 F Ccqual to Th) CO.4] Tavg (non-affected loops) - decreases 00.23
- ' i n a l .elve - total r e a c t c .- Power :u.s cat cnangec; howcVer, the ,t o w. r that ecco of the nor.-ef f e c ted 1 c o p s .1eu s t ti a r. . f e r t c- t h e 5/G has i n c r e s s.i d by a tactor of 1/3 tc c caip e r.s s t c fcr 11 100 . .
G r :, =m Cp (Th - Tc1 for cac.. Icop 1, 3 _-_; -; y initial Dcits T ws: 2C F, a r.d ;r u s t t r. c r ._ a s e L .- fcctor 1/3
- - -
final Delte i cauals Co.67 7 final Tavs = Tn - Del'.a T/2 = 577 - ._c.c7/2
= Sol.67 F [C.43 /G ressure (affectec locp, - increases CO.22 (Ir2tial value - 970 psi 3 = s a tur a tior; pressur e for 555 F)
final salve - saturation pressure for Th = 5~7 F
= 13:5 pr23 (ssfeties will lift; CO.43 ,' C Fressure .ncr.-aficcted iceps; - decreasc CO.2]
final vaive - as with part c anovs, the ap pr op r i t.ts delta-T 'isv3-Istn)
must aneresse by 1/3; iritial De:ta T = isvg - Tstr.: = So7 - 555 = ;c F final Tstm = Ta.>3 - Delta T = 563.67 - 14(4,'3)
= 565.67 - 15.67 = 545 F [0.43
~
REFEREt1CE MF' 3 Systen Description, Tcpic 6, Lesson 2 Transient E Accident Analysis, Chapter 4, page 4.11
.
. THEORY OF NUCLEAR F0 WEE FLAUT OFERATION, FLUID 5, ANC FACE 20
____ _ _ _______________________________________
___________ .
ANSWERS -- nILLSTONE 3 -35/C7/04-NORRIS, E. ANSWER 5.09 (3.00)
a. EOL - MTC is more negative and thus imparts greater positive reactivity from the drcp in coolant temperatur C/, c) E+,+1 r., 4. ., _ - . . ._ ce;_._..i __ .. ._ . ,_ .. .
_ . -
. . _ , _ _
_
_ _
impstts less reactivity ficm the power incr e as s'.C b. BOL - MTC'is less negative anc thus inparts less negative reactivity from the coolant heat-u .. - _ i i ft. 0) M
_...-_.._.__
._ __
m .
. ... _. _ ._- _
, _.- .
.__ .... .
-
- - :- 1*: , __ .-- *: 5: 2d : ' - -
,
- 53 c. EGL - Eeta-eff is less. making SUR g.~ eat.e CO.53 DOF C is less neest ive and thus t ra p a r t s less negative ieactivity after i, h c fuel t ei:.p e r a t u r e begins tc increas CO.5:
REFERENCE a. Topic 3 Lesson 5 pg. CE b. Ps. 45 c. pg. 45 a. throu3h Reactor Thect/ Chapt. 17 ANSWER 5.10 '1.50) Decrease CO.13 Less neutron learage due to decreased Tav3 [0.43 Incresse CO.13 Neutr or, f l .u- . shif ts towar d the core edges as the cor e rge CO.43 Increase CO.13 Decreased neutr on absctFtion by boro CO.43 REFERENCE MP3 System Description, Topic 6, Lesson 4, pgs 19-24 Wd N M }7/se a <c ef de e hanje. O G yy*, e,[f,a A/ perge/ca 07 [c W
!
l
_ _ _ _ ,
-
.
. PLANT SYSTEh5 DIEIC9, CONTRDL. AtJD INSTRUnEt T ATIDF FA.LE 2:
______________________________________________________
ANSWERE -- MILL 5 TONE 3 - 0 5,- v i / 0 4 - t4 C.P i,I E , B. ANSWER 6.01 (2.25, First valve (set st 1185 psi 3)
ASME requires the first valve to lift at or below the stesm generator design pressure (1185). [0.753 Fifth valve (set at 1225 psig)
Ensures that .s i l ,sives are cren ar.c passing 110*. cf design steam flow
when steam pressure nas esched 110*. of desig [0.753 Up to three salves per s t c: a n generato CO 25]
The p l a r,t 25 l i m 2 t e t' to ;< reduced power level due to the .p '
_
'
/6}k/ ,4lNvin) O yacj/y . [
- * ' ' '
',^ . [2.53 REFERENCE N cc</r .se h u J MF3 System Description. Tcpic 1 Losicn 3 ogs 1E-17 OF 3204. 99 5 TS 3.7. pg 3. 2 -:
ANSWER 6 . ;- : (2.25'
As power incresse > t.'lta-F .c le :It .-) increases CC.353 As power increasesi steam pressure decreases CO.3J
, .-
. s. .
m - s. t m = K(P-stm)(delta-F)
= = > s : r.c e the s t e s ,1, pressure con.conent stays c o n s t a t it sbculd 30 down) while the delts-F increases, indicated stesm fi;L w:11 de higher than actus1 steam flow CO.43 The summing networt f or flou ail' send a s 13 r.c l to tha total c o r.t t :11 e r to open the feed regulating vsl- CO.42 As level starts te intresse, the level error will signal for the FRU to close CO.42 Eventually, the flow error will be tsneelled out by the level error, and the FRV will be positiened such that steam flow equals feed ficw at some higher leve [0.43 REFERENCE NP 3 System Description, Topic 6, Lesson 9, pas 18-20, 23 Topic S. Lesson 4, pgs 67-75 i
. PLANT SYSTEh0 DESIGN, CONTROL, AND INSTRunEniATION FAGE 22
______________________________________________________
ANSWERS -- eILL5 TONE 3 -6 5 / 0 7 / 0 4 -N O F, RI E . E. , ANSWER 6.03 (1.501 Against the vertical surface of the :mpeller wear ing r in CO.253 To minimize flow to the inlet of the therrasl bstrier nest e:tenanger 1 the event normal seal injection is los CJ.53 Adjacent to the p u nip shsit, below the therms 1 bstrier hest enchen3e [0.25]
When the pump n4ust pr ovide i t s curi seal c o ol i r:9 the flcw will be diverted past the heat sachanger tubes instead of cirectly along the shaf EC.53 REFERENCE hP3 Systen, Description, L e s s c , Topic 2, pg 5 ANSWER 6.04 (3.00) Easts is keeping tiie t a r< k temperature belcu 120 F (to th? tano is capeble of handlin-a the desi d i sgh a r g e) EG.53 fml. c,lhe j Tany h a lore.gnD'4TJ Loua ha l caoyu.rdad gs i n g p r i m a r etade y water and crain tc n e Resctor,j t,
Plant Gaseous Drs. in [0. J
~~ ' " *
'
o, . ; c -. _ . . - _ .: _ _ - . _ .
-
i-.: .:._ '._ : . ._
tv ts-i 4 5-.-- . . _ , i .- i m - ::..:__,- rn n- _:mj f, prevent tne hydroger, from'ccn. ng out Of solution which could provide a potentially e r:p l o si v e atmosp her e !' ": psig CO.13 - P C 'J 469 (vent) signalled to closeCO.33
- ' FIR REL TK PRESEURE HI' alarmEO.23 115 pais CO.12 - F o p t .> r *. disks blow outEO.33 REFERENCE hP3 System Description, Lesson 1, Topic 4, p3s 11-13 i
.
. PLANT SYSTEh5 DEEIC.N. CONTFDL. AND IN5TF UMENT ATION PAGE 25
______________________________________________________
ANSWERS -- MILLETOr;E 3 - C S ,' o r/04-NORRIS. E: . ANSWER 6.05 (3.0G; Reacter trip - YES CO.5:
power ' '/ F 10 CO.253 1/4 Icgic CO.253 Reactor tr ip - NO CO.53 power ' SPC but
' , 10 :0.252 2/4 logie 4870 CO.253 Reacter trip - NO [0.52 power < 1 07; CO.253 2/4 Icgic CO.253 F:EF E RE N C E HP3 System Description. Tcpic 1, Lesson 1, Pgs 22-23 ANSWER 6.06 (1.50-Steam Dump h0DE SWITCH tc 'STEAn FRE55URE' CC.52 E:o t h S t e a n. D u n:p I r.t e r l o ;; 6 Euitches to 'E:iPAES INTERLOCM' E0.53 3/3 Condenser Vacuum (>25' Hg) CO.5]
REFERENCE MP3 System Descripticr.. Topic 5, Lescon 2, pgs 16-17 & 22-23
,. PLANT SYSTEhE DESIGN, CONTROL, AND INETF:UriENT ATIO*4 PAGE 24
______________________________________________________
ANSWEF:E -- h1LLE. TONE 3 -85/09/04-NORRIE, E: . ANSWER 6.07 (4.00)
a OT Delta-T OF Delta-T decrease oc change
':: :: : n. c /a g e
' decrease _ : : _: f,,f,g i : ::::f;gc),g , E0.33 each] . OT Delta T - Indicated pr essur izer pressure will be celow the nominal pressure which will insert a negative tern, into tne s.etpoint GP Delta T - nct affected by changes in pressure OT De:ts T - Delta f l u. - n o r n.a ll y nesst2ve terno is subtreeted fron, the setpcint (minus :, minus = positive); if lower detector f s i l s. . Delta flu:, becomes positise, which esuses the.setpcint to de:rease DP Delti T - . . - .- :: :- :- ". . g ffa,y c' he a // 4 7 [Surd w, r+ llg
'- GT Delta T - <- i n _- i- !!! - - - - 0 -I ' ; '_- ; - -
-
- - - '
-
__ ' :: -
_:
4 '
$ .:' 21: :.:: 2-:
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'
OF Delts ~ - '-_: - - :_:- :- - :: :i: :
6 F/wp f3 /3 a , / g ,,Je / w/
a E0.33 for each discuss:cn3 REFERENOE N 7" '" /"*7"///
HP3 System Description, Topic 6, Lesson 5, pas -14 L Topse 7, Lessor s 1-3, pgs 44-52 & RA-1 fS. Yabk J.ht,pp g - 7 M,a gic 61 a(ways reduta -f& Jr frNl
. . F war 4T 5'r 5TEh5 CESIGN. CONTROL, Ar4 D INSTRunENTAT2Cr, RAGE 25
ANSWERS -- MILLETONE 3 -E5/09/04-NORFI E' ANSWEP 6.08 (2.00) 'IR 1(2s LOSS OF COMPEr45 ATING VOLT AGE' E0.252
'IP LOSS OF DET UOLTAGE' 00.253 There is an electrics 1 signal p?cvided tc the n.eter equisalent to the low scale r e e d i r. g il v 10E-11snps) which prevents the instrument fro-pegging downscale dur ir 3 nee rr sl operaticn [0.251 Detector o u t : t.t as loat' t r. c n sett21 rie utr on ;eve [ C . 2 5 '.
Indicated level st:1 inc r c a u at s higher tnan actual rate, rising a felse high sicrtuF r z.t c i n d i t z.t i o n . [0,~53 Detector output is higher than scttil neutron level CO.25]
The source isn3t dctcetors will not suton.ctically ener gine .1 the affectcc char.nel reeds higher than 5
.
- 10 E - 11 a c4 p s CO.25] Tne oper stor tsn n.snus11. ener gire the 5F detectors u2tn tne EF E:l o: L-Reset switc CO.25]
REFEEENEE nF3 System Description, 7.3 p . c 4, Lesscn pgs 15 - 1 <3 OP 3:07 pg 5 ANSWEF 6.09 ( 2 . 5.' See atteched sheets Freterrec path 1.03 Alternate patb [0.53 C . Rod below the Icw-Icw insertion limi [0.2Y each:
2. All rods fsil to inser- cn a reactor tra . An uncontrclied cocidcwn c- -'::t_ i+ i-- ::
. ,__ ,, .: 2: ,. :n . .; ; . . me :.- ;; ; .--tu :; _, . .
An uncen fu//ed Modedb iic ren REFERENCE MPo System Descriptione icpic 2 f 4 9 ,7 ,f, g, gb Les2cn 2, pgs 32-33 & PG9-PG11 ADF 3566 pg /A ANSWER 6.10 fe.00)
see attached drswing
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. . . . .
.
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! PLANT 5 5TEMS DESIGNr C o r1 T R O L , AND It4STRUMENTATI0ti FACI 22
- . ______________________________________________________
ANSWERS -- n:LLST0reE 3 -85/09/C4-NGRFIG, E. 3 REFERENCE Millstone 3 Required Drawings 1322 and 323
,
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.
. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND FAGE 27
-
~~~~IIID5dE5555hL
.____________________
55sTEoE-----------------~~~----
. ANSWERS'-- MILLSTONE 3 -83/09/04-NORRIS, E. ANSWER- 7.01 (2.50) . . Terminate the startup by inserting all control bank CO.5] . Terminate the startup by fully inserting all control rod CO.25] heasure boren concentration and recalculate the EC CO.253 . The-moderator temperature coefficient is within its analy ed temperature range The trip instrumentation is within its normsl operating range _The P-12 interlock is above its setpoint 4 The pressuricer.is capable of being in an operable ststus with a steam buoble The reactor vessel is above its minimum RT-NDT t e a.p c r a t u r e CO.3 each]
REFERENCE OP 3202r Fages 4, 9& 10 I TS 3.1.1.4. pgs 3/4 1-6 L B3/4 1-2 Arl5 WER 7.02 (3.00; Fire hazard CO.53 due to excessive oil buildup in euhaust lines CO.52 . Js eke t - Cool aret Tempersture - incresse the servicc wcter flow to the
. o fh e r4'e Scea lk jacket uater heat enchanger CO.332 4 r7J M rf er A ~ - ensure the 3-way valve is operatinc ~
c c c ejr, /r //( properly CO.33]
Cr ankca se Pressure - nothins can b e c o n c o_C e n /7,/ go e gg,, f, CO.23] Low lube oil pressure (2/3) f #ar 72sc, gg ,,, ,/ CO.53 naine overspeed %
4 te ren 4 r ~ h (pri m h /
l REFEREr1CE-OP3346A, para 6.2 & 6.5, pg 4 System Description, Diesel Generator, pgs,42-43 L S X a?t{- f 9 h {' ?'l- U T ")
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. . __ . . _ . _ _ . . _ _ - . _ _ . _ _ _ _ . . . . .
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t I PROCEDURES - NORMAL, A E r40 RM AL , EMERGENCY AN1 PAGE 28 l - ________________________________________________
l
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RADIOLGCICAL CariTRO ______._____________
ANSWERS --' MILLSTONE 3 -65/09/04-NORRIS, En ,
ANSWER 7.03 (2.25)
' No evoluticns'are planned which could change the radiological conditions as listed on the RW [0.752-4 Any evolution that would change the radiological conditions will be reported to H [0.75] The work on the RWF does not Jeopardine the plan [0.75]
REFERENCE
'SHF 4912r para 4.2, pgs 3-4
,
ANSWER 7.04 -(2.50) .. Rod bottom lights lit .CC.25:
2.' Reactor - toip anti bypass brealer:- open EO.25] Digit,al rod position indicators at nerc ;0.25]
4.. Neutron flux decrecsing - CO.25] " Loss of. automatic actions of c
'oth safety injection and diesel generator-s e qu e n e i E0.53 L ' AUX SHUTDOWN FANEL COCR OFEN' ,'(affw.yhsk Ma y, /rc/
CO.33 cach]
'RHS VALVES LOCAL CONTROL' m f, f .vt ec
'SIL VALVES LOCAL CONTROL' ,,.n ,y REFERENCE ECP 3503, pgs 3 1 5 h
EOP'FR-S.1, pg 3
.MP3, System Descriptien, Topic 3r Lesson 4, Pgs 13-14 ANSWER 7.05 (1.00)
' verify rn trip .25 each] verify turoine trip ,-[ a f h u y d' * # p M p #
' check that RCS is isolated - - verify AFW flow e M [#' '"'J REFERENCE l rr / Sc, & llfA lJ EDP 35ECA-0.0, pas 3&4 fujenrJe $c in h4 I
.
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! PRCCEDURE5 - NORMAL, AE40Rh AL , EMERGENCY AND FAGE 27
- _________________________..______________________
<
,
FADI0 LOGICAL C0t. TROL
____________________
ANSWERS -- HILLSTONE 3 -85/09/04-NORRIS, B. ANSWER 7.06 (3.00) (Manually)cpenthe AFW cischarge header cross-connect valve : 52 Otsy the r- .= if :- 10: y ~:t--
- is ~ - :';' '::f" :-:.E:
C o .. m ; . m m .1 1. . . , ; ! !:- -_ t'-
'
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.
n__ ,
-, ,_
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,, . ,,,g, T# fhift suction to the Concensate Storage Tank g j, pf E" 52 T4 ~ ~ ~ ~ ~ ' , , , _
-
-
.-._--- _ __ __. _. _ _ .._4-. .
.,m _
,,
e,
-- . _ _ _
, The CST is aerated .
fe,5) ::.:52 The Service Water System causes. corrosion (see weter) (c.y] ::.:EJ REFERENCE OF 3322 p a E 2:
ANSWER 7.07 (3.00) There is a momentar/ loss of RPCCW flow C0.52 The service water p u n.p s will be momentarily opereting below minimum required fic CO.53 Service wcter pr es s.or e will be momenterily grerter than RPCCW pressure (possi'le c in-leakage of sea water). CO.5] RFCCW Sur se Tank A or B Level (<112') tany three at 0.33 each]
RPCCW Train A er 5 Non-Safety Header Auto Isolation RPCCW Tr ain A or B Suction Pressure (<15 psis)
RFCCW Header Supplf Flow High RPCCW Pun 4p Auto Trip / Deer current Loss of service water cooling to any on-service RFCCW heat enchanger Any high tempereture/ low flow s l a t o> s on con.ponents served by RPC.CW ce To avoid tube vibration effects in the RFCCW heat e:: c h a n g e r s . CO.5]
REFERENCE AOP 3561, pg 2 OP 3330A, pgs 6 & 20-21 O
i
_ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ ._ .
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.
7 PROCEDUREE - NORMAL. AENORMAL, EMERGENCY AND FAGE 30
--- RE5iBE55fc;L C5sTE5L
- ----~~~~--------------~~
____________________
ANSWERS -- MILLSTONE 3 -E5/09/04-NORRI5, E. ANSWER 7.08 (1.50) When starting the pump, if the breaker has not closed (proper smperage)
the other pumps will Auto-Start if their switch is not in Pull-to-Loc [0.752 To prevent seat leat.ege Fest the SGWLC valve [0.75]
REFERENCE OF 3319A. p3s 5 ' & 15 ANSWER 7.09 (3.50) (105 psla)-start stancby CIA compressor [0.32-locs) or c onti c 1 rocm E0.23 (100 psis)-manually cFen the cutsiac c o n t a i n m e r, t isolation valve to admit i n s tr umer.t s2r EC.33-Icc61 only CC. ]
( 85 ps; a )-msrmslly c1cse Service Air Header valvc E0.15]
-msneally open Sersice Air to Instrument Air X-Connect [C.153-local only EO.2] Feedwater '.1 -
_ -,; ' -
. .1: ::--i nc ef[rc /, mg/,e / f f T /1 CO.51 RCF Ses) Lster Outlet Valves Fail Open EO.5]
CUCS Grifice Isc1stion Valves Fail Closed CO.57
" " '- '- RHR --
~_ -
- -
(1 css of ten.per stor e controli E0.53 ne ~a , hNcl, S u/7 Nl6 7 7fl REFERENCE 5 & W Owg Nos. 12179-En-138C-2 & 12179-EM-103A-2 AOP 3562,pg 2 MP3 System Description, Topic 2, Lesson 1, pg CV-1
!
l.
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! PRDCEDUREE - r4 3 R H A L , A E.N O R M A L , E M E R G E rJ C. ? AND FACE 3:
--- EA5i5t5ETEEE
E5siE5E----------~~~-----------
ANSWERS -- MILLSTONE 3 -55/09/04-NORRIS, .
ANSWER 7.10 (2.75) (N-18) = 50 REM [0.25]
Total lif etinte to date = 48 + 1 = 49 REh Total lifetime available = 50 - 49 = 1 REM CO.253 Tots 1 this quarter evcilable = 3- 1 = 2 REM E0.253 Lifetime is mere restrictive than quarterly limit 0.85 REM /HR gLams + f.03 RAD /HR)(10 QF) neutr on =1.15 REn/HF. dote Tste 1.0 REh/1.15 REM.'HR = 0.67 HR$ = 52 MIN CG.5]
CO.2 for using tne conservative q u e l i t ,, 'artor 3 REM whole body one time exposure to.753
. Director, Site Energency Oper ations :0.753 REFERENCE 10 CFR 20 3HP 4700, pas 6-11 & 14-15
_ _ _ __
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.
t ADhINISTRATIJE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 32
._______________'__________________________________________
l ANSWERS -- HILL 5 TONE 3 -55/09/04-NOGFI5 B. ANSWER 8.01 (2.50) You may operate for up to two hours with one less than minimum complement E0.753 pr ovided that immediate action is t al, e n to bring the complement up to minimum EO.5 The on-shift STA will have to wait for a relief to come in EO.7531 ct if the on-coming 55 or SRO meets the STA requirements, the STA siticr may be vacant E0.53.(C/q,',(je ,<,pej,f,y ,,, ,,;j f he le y r W %p< * * " *
REFERENCE ACP 6.01, para 6.1.5, ps B T.S., Section 6. ANSWEF: 8.02 (1.75) To allow for crerstional f l e a l t t l i t ,> . (c.E5 J :
- - : . . . : i 11 u _ : ' . r, .. : : :. c:; :i ;- ::. _ , g- ;:- ;f :a ;.
-'
- --
- ^ '. l o c. c c _ ;- c. ;* - '. T' ~5:
2F4. The tolerances for ir.tervals is restrictive to ensur e the reliability associated with toe surveillance is not degraded.fv ?) E0.62 REFERENCE T.S. 3/4.0, pgs 3/4 0-2 d,B3/4 0-3 ANSWER 8.03 (2.25) Fressure < 440 psia (fro pren) E0.375]
Temperature ,
350 F (f w<i) E0.5753 P r e s s u r'e : based on-the static pressure of the RCE added to the discharge head of the F:HR pump not to exceed t*t design pressur e of the RHR syste [0.753 Temperature * based on the design capacity of the heat exchangers to reduce temperature of the RCS to 120F within 24 hrs after S E0.753 REFERENCE OP3201, para 4.17, pg 12
,
. ADMIN!5TRATIVE FROCECURES, CONDITIONS, AND LIMITATI0tJC FACE 33
ANSWERS -- MILLSTONE 3 -55/07/04-bCREIS, .
ANSWER B.04 (2.00) Only used when the 0-I-A can assure the safety of the persennel for which the tagging was requeste E1.03 <1 hr - work stopped a work order in custody of OIA CO.53
>1 hr - valves and/or br ess er s TAGGED as listed on Tss-Log sheet CO.53 REFERENCE ACP-GA-2.06A, pgs 4, 6, S 25-26 ANSWER S.05 (2.00) Channel Calibraticn - the adjustment of the channel such that it responds within the required range Enc securacy of kr.own values of input E0.5 The coannel eslibration includes the senscrs anc circuitry EO.5 Channel Operational Test - the injection of s simulated signal intc the channel ss close to the sensor as pt3ctical to verify cperability of the circuitry E0.53 (Does not include a check of the senscr, deduct 0.25 if the answer indicstes senscr is checked.) Adjustments msy be made as necessary CO.5 REFERENCE Technical Specifications, pg 1-1 ANOWER G.06 (3.00) y, /76;v e If the emergency concition is classified as anAlert,(Site ,
Area Emetgency, or General Emetgency) E1.03 . W ill assembly help or hinder the situation? [0.53 Per sonnel dose rates exceeding limits [0.53 (limits: 210 mrem /hr or I-131 levels : 10 n MPC) Personnel exposure exceeding limits CO.53 (limits: >500 mrem whcle body or 500 I-131 MPC hours) Actual or potential personnel hazar ds e::is [0.53 . ADMINISTR A TIVE F riOCEDURES , CGNDITIGh5, ANE LIMITATIONE FACE 34
__________________________________________________________
ANSWERO -- NILL 5 TONE 3 -85/09/04-NORFIS, E. REFERENCE EPIP 4010A. pg 5 EPIP 4 010 E:, pg 3 ANSWER 8.07 of Nont
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(2.00)
ao 13 sthis is the responsibility of the Lead Department Head) [0.5] . If performing PORC/50RC appr oved procedur es [0.253, th-t require Control Fccm approval [0.253r and work. is documented [0.25 . I t h a s. c e e r- d e t e r n. i n e d that the vori does not affect plant operaticn
[0."53, and does not require DC [0.25] or equipment tagging [0.2523 PEFERENCE ACP-GA-2.02C, pas ~
'
T !
11-1l3 g)J6( ,.c c ; ja a nd desah AN5WER 8 . 0 (5.00) All penettstion3 required to be closed during accicent conditions are either:
CaFable of being closed by an operable containment aito-icciction valve system, or [0.53 Closed by manosi valves, blind flanges, or deactivated autenatic valves secured in their closed position [G.53 All equipment hatches' closed and sccle [0.53 Each sir Icek opercbl [0.5] Containment lestage rates within limit [0.5] Ses11ns nechEnisn associated with each penetratior, is cpor abi [0.53 REFE6ENCE Technical Specification, pg 1-2
. _ _ . _ . _ _ _ . - - . - - _ . - . _
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. ADMINISTRATIVE FROCEDURES, CONDITIONS, AND LIMITATIONS FACE 35
ANSWERS -- hILLSTONE 3 -85/09/04-NORRIS, .
ANSWER 8.09 (2.50) Go to HOT STANDBY (within one hour). CO.53
/. o . Prevent cladding peri oration ;0[. 5:]which would r esult in the
,
release of fission products to the RCS : . 5: .C/,qJ
" .
E -et r :t : r ::: .- t'u i- t ; 37
-
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_
. ; _ ; - :,; 7 : ,5,,
REFERENCE TS p3s 2-1 & B2-1 ANSWER -0.10 (4.00)
t No CO.253; the TC require 5 all RCS accumulators to be operable when in Mode 1 E0.75 (An tr ri s we r of Yes will be acceptable with sn sppropriate exp % nation.) To ensure that an adequate volume of borated water [0.53 is provid to the cor e f ollowing a lorse [0.25] RCS pipe rupture CO.25 c., Restore level to the neraal band (with1r. ane hour ) C1.03 or commence placins the plant in at leest hot staridby [1.0 REFERENCE TS 3/4.5.1, pas 3/4 5-1 & B3/4 5-1 MP3 Systen. Description, Topie 3, Lesson 4, ps 61 t
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September 12, 1985 MP-S-T-822-85 TO: Doug Coe USNRG,Refon1
. Y/b
. b y - //k / ~
FROM: on Stotts Millstone unit 3 Training SUBJECT: Comments'on Written Examinations of September 4, 1985 Attached is my compilation of comments on your written examinations administered to our operators and senior operators on September 4, 198 All comments have been resolved in discussions with Barry Notris, David Ruscitto and yo This list reflects those agreement Thank you for your cooperation as well as that of Mr. Norris and Mr. Ruscitt RGS/rm Attachment cc: Training File OT- M. Moehlmann osto nty 3 a3
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ATTACHMENT RESOLUTIONS TO NRC EXAM COMMENTS REACTOR OPERATOR EXAM 7, 1. l' 10 cycles may produce a slight difference; answer key will l,0 Neallow t ex-<.afor pi *A"1 is~*("~
worth more" s is d'+*if explanation G <l'+r % **isk'provided.dg+ps Lle c""~
1.04 a,b,c Use of the proper term " Shutdown Margin (SDM)" will key *
candidates to include +1400 pcm for stuck rod, as the Tech Spec definition dictates; answer key will allow inclusion of this value. p,& m b) ~ p 4 w s p .' k lly McivEs e s> L )rw~
s k is s+ .06 af (@ same as SRO #5.09)
1.07 a Will also allow Tech Spec Bases for answer. - ?
I 2.01 a (same as SRO #6.04 b) - ? I 2.04 A & B Rl!R min flows open if less than 555 gpm (references o fc ,
provided LSK 27-7E, pLS page 74) /
2.06 a Answer key item #2: time delay is 200 and 210 seconds (reference provided). V o,p *
1 tem #4: feature is completely removed freference provided)( 0,l Key will reflect changes D q # '2 9 gu- 4M (~cv- +,2J 2 - J(,yf gA4 J ttf -1.~ M 3.01 a Axial Flux Difference Monitoring is not included in the correct answer '"*
rwt(reference ut qM ed . provided).
y pt s(nrau G.J- l blxk (Ap to;A '
%, M (1,.4 ~
3.01 b Agreed that this switch performs only one function: " defeats input to overpower Rod Stop"; answer key will allow for this 0. F ,
fac .02 Will also accept " output" as label for vertical axis, o te .
3.05 a 1) CAF = 15 psia 2) 111-3 we use 25 psia OA '
3.06 The 97.75% trip has been removed; now only one trip called
" low-shaft speed" at 95.75%; we have no P-17; P-7 (10%) is the Q only permissive (references provided).
3.07 a 2) Ansycr key will also accept " excessive kw/ft" u !
3.08 c Correct answer will read " rods out from Tave/ Tref deviatio O,k ,
3.10 Answer key, as written, is correc .03 b,c In both cases, answer key will allow for operator tripping reactor on high VCT temperature. /vD 'T LI)
Ae? 35T/ pg 3-1-i
.
ppyr --
- - .
. . :
1 4.04 InboBQcases answerkeywil
-
reactor'onhig(h'VCTtemperatur.-allow]oroperltortrlpping
-
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g7 4.04 b Will accept any 4 from a provided list of shutdown parameters (reference provided). OI -
4.04 c
" Generator provided). Differential" is added to answer key (reference 7 ,
.
SENIOR REACTOR OPERATOR EXAM 5.01 b Answer key too specific ior required answers. A broader range ;
should be allowed. Concern was resolved. Some latitude will be allowed during gradin {
li 5.03 a :
Answer long...".
key states: "With very short lifetime," "with very kl Utility feels these may be implied rather than '
state This concern was resolved. Examiner is looking for those concept c Concern that students may use an alternate method for proving '
the 1/3 dpm SUR. This concern resolve Alternate methods are acceptabl .04 a Concern that the first 1.0 points in the answer key were not ;
directly asked for. This concern resolved by changing the answer ke l
'
b Concern that an acceptable alternative for "photoneutron reaction" and " spontaneous fission" should be the equation Concern resolved. Equations are sufficient for full credi .05 a Concern that full credit should be given for writing the equations. This concern was resolved as follows for parts 1, 2, ) must include some discussion as to how OH~ affects p ) equation adequate for full credit and second half of answer key, "this is restricted to use on startups..."
was eliminate ) equation adequate for full credi .06 a Concern that reasonable values for 5 gg should be allowe This concern resolved. Any acceptabfe value will be allowe .07 b Answer key should not include the words: " causing excessive wear." This concern resolved by eliminating the words from the answer kc !
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!
-2- !
f
--
_
l i
5.08 Concern that many variables may occur in the answers given by examinees which are not stated in the question. For example, turbine load may decrease with stage pressure feedback not used by MP3. This concern resolved. Examiner will use great latitude in gradin j
!
5.09 a Concern that MTC discussion should be adequate for full credi ;
This concern resolved by changing answer ke .
b Concern that MTC discussion should be adequate for full credi This concern resolved by changing answer ke t c
Concern that discussion of Doppler Only Power Coefficient not I necessary to fully answer the question. This concern not resolved. Examiner wishes to reserve j .
research is completed. Not wpm udgemen fWtwu% untilg5therj7Q l-fo cessco'S4 - wly r>oPC 04 +k pw 0-3'$
/
!
6.01 b Concern that the words "due to the high $ setpoints lowered" should not be required for full credit. This concern resolved by removing those words from the anever ke .04 a Concern that the significance in this question is that an abnormal condition exists for temperature, not that_the design discharge capability is affected. This concern resolved as follows: the words "so the tank is capable of handling the design discharge" were put in parentheses in the ke Also concerned that the words " cool the tank" should be .i acceptable for full credit for second part of this questio Thic concern resolved by enclosing remainder of the answer key for this part in parenthese i j
b Concern that the answer key reflects concerns for corrosio MP3 uses a stainless steel PRT and therefore this is not necessary. This was resolved by changing the answer ke Hydrogen explosion hazard discussion is adequate for full M credi .05 Concern that answer key is incorrect for P-8 permissive setpoint at MP3. This concern resolved by changing answer key to reflect correct value of 48%. pg ,g engM 6.07 Concern that answers under OPAT should all be "No Change" since l Al input to OPAT is set to 0 at MP3. This concern resolved by '
changing answer ke (References provided) g( ) Concern that OTAT will decreas (if anything) because of the way Ai input is used by OTAT circuitry. This concern resolved by changing answer ke (References provided) 4
,
i-3-
- _ ,_-. - ,- , . , . - - -.
HWEEW -
. .
j eScrff 6.08 a Concern that proper names for an nciator tiles should not be f'
required and that other possible indications should be allowed. p " Q This concern resolved. Examiner will use latitude in gradin p?
g 6.09 b Concern that listing the 5 entry conditions from A0P 3566 should be acceptable for full credit. This concern resolved by changing the answer ke (References provided) py .
7.02 b Concern that other acceptable answers will be allowed. This o
'
concern resolved. Examiner will use latitude in grading, rd E i c Concern that " Generator Differential" should be included in the i answer ke This concern resolved by changing the answer key (References provided) cy !
7.04 c Concern that some examinees may approach question from the d point that they are at the ASP and list lights and indications seen there. This concern resolve Examiner will accept [@LX either the answers in the key or an explanation of lights and '
indicators at the AS !
7.06 a Concern that word " manually" should be eliminated from answer key. This concern resolved by enclosing the word in parentheses in the answer ke b Concern that " shift to CST" is the only part of the answer .
necessary for full credit. This concern is not resolve .F -
Further research necessary for full resolutio .02 Concern that part 2 of the answer should not be included in the ke This concern resolved by changing the answer ke .03 Concern that certain amount of latitude should be allowed on numbers to fill in the blanks. This concern resolve ~
Examiner will use latitude in gradin Also concerned that answer key is too specific for basis on the pressure and temperature. This concern resolved. Examiner will ;
'
use latitude in gradin .07 Concern that acceptable answers should be("#3" or "none." This concern resolved by addition to answer key.- .
ggg%
~
M A A Act 8.09 b Concern that part 2 of the answer key is not part of the stem question and should, therefore, be eliminated. This concern resolved by eliminating part 2 of the answer key for this question and adding y to each of the answers in part ' ATTACHMENT 3 R0 EXAMINATION COMMENTS AND RESOLUTIONS Question / Answer Comment ,
1.03 a Comment that 10 cycles may produce a slight difference is not accepted since the referenced training material clear-ly states "...no noticeable change over design life."
1.04 a, b, c Comment that use of the proper term Shutdown Margin will key candidates to include a stuck rod is not accepted since the question specifically ex-cludes consideration of a stuck rod.
1.06 a, b A discussion of MTC effects during these accidents is sufficient for full credit since the Doppler effect is relatively very small.
1.06 c Comment that discussion of DOPC is not necessary to fully answer the question is not accepted since this is the only power turning reactivity effect given the conditions set by the question.
1.07 a Answers based on the Technical Specification for Reactor Safety Limits will be accepted as they are equivalent in content to the referenced training material.
2.01 a Corrosion is not required for full credit since the PRT is stainless '
steel. Referenced training material is inconsistent on this point.
2.04 555 gpm is the correct setpoint for RH3 pumps as per LSK 27-7E, PLS page 74. Training material is not current.
2.06 a Time delay of 200 and 210 seconds is correct instead of five minutes, as per Emergency Generator Loading Se-quence Fig. Item 4 is deleted from l
,
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2.06 a (cont'd) the answer per Dwg. #25212-2691 Training material is not curren .01 a AFD Monitoring deleted from answer as per NI Technical Manual Process Control Block Diagram 10 Series. Training material is not curren .01 b Will accept " defeats input to over-power Rod stop" for full credit since this is a correct answer the way the question is worde .02 Will accept " output" as a label for the vertical axi i l
l 3.05 a CAF = 15 psia and Hi-3 is also 25 psia. -
3.06 The 97.75% trip has been removed. An-swer changed as per revised Reactor Trips table of Topic 1, Lesson 2 which was not current when provided to NRC examiner .07 a Will accept " excessive Kw/ft" as an alternate answe .08 c Question was changed during the exam-ination to delete "Auctioneered high Tave... matched." Correct answer is now " rods nove out due to Tave/ Tref deviation."
4.03 b, c Answer key will allow for tripping the reactor on high VCT temperature in accordance with A0P 3561 page .04 b Due to confusion caused by question wording, will accept shutdown param-eters taken from LSK 24-9.3 .04 c " Generator Differential" added to answer as per LSK 24-9.3 In addition to the above, the following errors were discovered in the answer key after grading commenced:
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4 Question / Answer Comments
- 1.04 b, c Mathematics' error correcte ~ 1.09 1050 psia should have been 1080 psia (typo)
= 2.01 Due to question wording, safeties and i reliefs are also correct respons '05
. Correct value is 375 psig as per PLS pg. 52 2.09 Additional correct answer is " Seal water heat exchanger / excess letdown -
normal flow path," as per Topic 2 Lesson 1 Figure CV- .03 Add " blocked when steamline SI is not blocked" as per Main Steam System pg. 2 .11'- A 250 mrem /qtr neutron exposure limit exists as per SHP 4902 pp. 10-11 which becomes more limiting than the beta-gamma limit in the answer ke ,
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. 4 SENIOR REACTOR OPERATOR EXAM COMMENTS AND RESOLUTIONS
. Question / Answer Comment 5.01 b Answer key too specific for required
answers. A broader range should be allowed. Will be considered during gradin .03 a Answer key states: "With very short lifetime," "with very long...". Util-ity feels these may be implied rather than stated. Will be considered during gradin .03 c- Concern that students may use an al-ternate method for proving the 1/3 dpm SUR. Alternate methods will be con-
sidered during gradin .04 a Concern that the first 1.0 points in the answer key were not directly asked
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for. Answer key changed to that re-quired by the questio .04 b Concern that an acceptable alternative for "photoneutron reaction" and "spon-taneous fission" should be the equa-tions. Equations are sufficient for full credi .05 a Concern that full credit should be given for writing the equations. This concern was resolved as follows for parts 1, 2, 3.
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1) must include some discussion as to how OH affects p ) equation adequate for full credit and second half of answer key, "this is restricted to use on startups..."
was eliminate ) equation adequate for full credi .06 a Concern that reasonable value for Beff should be allowed. Other acceptable values will be considered during grading.
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5.07 b Answer key should not include the words: " causing excessive wear," since it was not asked for in the questio This concern resolved by eliminating the words from the answer ke .08 Concern that many variables may occur in the answers given by examinees which are not stated in the questio For example, turbine load may decrease with stage pressure feedback not used by MP3. Will consider other assumptions during gradin .09 a, b, c Same comment and resolution as R0 1.0 .01 b Concern that the words "due to the high $ setpoints lowered" should not be required for full credit. This concern resolved by removing those words from the answer key and substi-tuting "due to the limited relieving capacity of the code safeties," in accordance with the basis for Tech Spec 3.7. .04 a Concern that the significance in this question is that an abnormal condition exists for temperature, not that the design discharge capability is affect-ed. This concern resolved as follows:
!- the words, "so the tank is capable of l handling the design discharge" were
! removed from the required answe Also concerned that the words, " cool l the tank" should be acceptable for full credit for second part of this ques-l tio Remainder of answer was not
! specifically called for in the question
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and was deleted.
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6.04 b Same comment and resolution as R0 201. .05 Concern that answer key is incorrect for P-8 permissive setpoint at MP This concern resolved by changing answer key to reflect correct value of 48%. Supplied reference material not consisten ..
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6.07 Concern that answers under OPAT should all be "No change" since AI input to OPAT is set to 0 at MP3. This concern resolved by changing answer key as per Tech Spec Table 2.2.- Concern that OTAT will decrease. (if anything) because of the way AI input is used by OTAT circuitry. This con-cern resolved by changing answer key
, as per Tech Spec Table 2.2- .08 a- Concern that proper names for annunci-ator tiles should not be required and that other possible indications should ,
be allowed. Will be considered during gradin .09 b Listing the five entry conditions from AOP 3566 is acceptable for full credi .02 b Concern that other acceptable answers will be allowed. Will be considered during gradin .02 c Same comment and resolution as 7.02 .04 c Concern that some examinees may approach question from the point that they are at the ASP and list lights and indications seen there. Will accept either the answers in the key l or an explanation of lights and indi-cators at the ASP.
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7.06 a Concern that word, " manually" should be eliminated from answer key. This concern resolved by deleting the word from the answer key since it was not i asked for in the questio .06 b Concern that, " shift to CST" is the only part of the answer necessary for full credit. The rest of the answer was not olicited by the question and was delete .02 Concern that Part 2 of the answer should not be included in the ke Part 2 was not elicited by the l
question and was delete .
7 8.03 Concern that certain amount of lati-tude should be allowed on numbers to fill in the blanks. Also concerned that answer key is too specific for basis on the pressure and temperatur Will be considered during gradin .07 Concern that acceptable answers should be "#3" or "none." Due to poor wording within ACP-QA-2.02C either "#3" or
"none" will be accepted as a correct answe .09 b Concern that Part b.2 of the answer key is not asked for by the stem question and should, therefore, be eliminate This concern resolved by eliminating part 2 of the answer key for this
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>'s question and adding .5 to each of the answers in Part 1.
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