IR 05000277/1988006
| ML20153F232 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 04/29/1988 |
| From: | Howe A, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20153F225 | List: |
| References | |
| 50-277-88-06OL, 50-277-88-6OL, 50-278-88-06OL, 50-278-88-6OL, NUDOCS 8805100275 | |
| Download: ML20153F232 (52) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT N (0L)
FACILITY DOCKET N and 278 FACILITY LICENSE N DPR-44 and DPR-56 LICENSEE: Philadelphia Electric Company FACILITY: Peach Bottom Units 2 and 3 EXAMINATION DATES: February 17 to February 19, 1988 CHIEF EXAMINER: d[44m , <//N///
Allen G. Howe, Senior Operations Engineer, DRS Date APPROVED BY: .
E 9-29 99 Davi'd J. Lange(/ Ch'ie 'R Section , Date Operations Branch, Division of Reactor Safety SUMMARY: Written examinations and operating tests were administered to four (4) reactor operator (RO) candidates. Two (?) candidates passed these examinations. One candidate failed the written examination and one candidate failed the operating test.
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l 8805100275 880429 PDR ADOCK 05000277 V DCD
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e DETAILS TYPE OF EXAMINATIONS: Replacement EXAMINATION RESULTS:
1 RO l l Pass / Fail i l
i I I l Written l 2/1 l l l l l 1 (Operating l 3/1 l l l l l 1 l0verall 1 2/2 l 1 I I i i CHIEF EXAMINER AT SITE: Allen G. Howe, Senior Operations Engineer OTHER EXAMINERS: M. Evans, Operations Engineer S. Pullani, Senior Operations Engineer W. Cliff, PNL J. Richardson, USNRC, NRR The following is a summary of generic strengths or deficiencies noted on operating tests. This information is being provided to aid the licensee in upgrading license and requalification training programs. No licensee response is require STRENGTHS a. Knowledge of off normal and operational transient procedures, b. Ability to locate and use piping and instrument diagram DEFICIENCIES a. Knowledge of radiation exposure limits for visitor b. Knowledge of the purpose and principals of instrument line restricting orifice .
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. -3-4. The following is a summary of generic strengths or deficiencies noted from the grading of written examinations. This information is being provided to aid the licensee in upgrading license and requalification training program No licensee response is require STRENGTHS a. Knowledge of reactivity effects during heatup, cooldown rates, steam table use, and subtritical multiplication, b. Knowledge of the core spray system, RCIC response to level and the SBGTS system, c. Knowledge of operator actions required for plant shutdown outside the control roo DEFICIENCIES Understanding of the reasons why xenon concentrations vary with power. Knowledge of power, pressure and level response to a lead reject transient. Understanding of the adverse effects of a loss of core bypass flo Knowledge of how an isolation of the RWCU system, while the du.ip valve is open, will cause piping stress. Understanding of why vessel level instrumentation accuracy varies from rated conditions to cold condition Knowledge that the action of the fast acting solenoids causes a turbine trip on a load reject. Understanding the effect of a condensate pump trip on the feedwater control syste Knowledge of general conditions when the operator should scram the reactor per A-7, (conduct of operations) and knowledge of the condi-tions when manual control of ECCS systems is allowe GENERAL Although two candidates passed the RO examination, the sectional and overall scores reflected individual weaknesses that were not common to the other candidates but should be reviewed by the training department for additional training during subsequent requalification trainin . Personnel Present at Exit Interview:
NRC Personnel A. Howe, Chief Examiner S. Pullant, Senior Operations Engineer l
M. Evans, Operations Engineer T. Johnson, Senior Resident Inspector
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R. Urban, Resident Inspector l
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4 Facility Personnel D. Smith, Vice President, Peach Bottom J. Franz, Manager, Peach Bottom J. Cotton, Superintendent of Operations-F. Polaski, Assistant Superintendent of Operations A. Donell, Quality Assurance ,
- Summary of NRC commente made at exit interview
The chief examiner thanked the training and operations staffs for their cooperation during the examinatio Tue examiners felt that housekeeping in Unit 2 was goo The written examination review resulted in a few comments requiring resolution. The reviewers stated that the examination was a good test.
> The generic strengths and weaknesses noted on the operating examinations-were discusse During the examination process it was determined that several precedures had been revised, added, and/or deleted and that some of these procedure changes were not submitted as part of the examination preparation packag The Chief Examiner requested that the facility determine if plant operators and operator candidates have been or will be trained on these change t Attachments: Written Examination and Answer Key (RO) Facility Comnents on Written Examinations after Facility Review i NRC Response to Facility Comments
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U. S. NVCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: PEACH BOTTOM 2&3 t
REACTOR TYPE: BWR-GE4 DATE ADMINISTERED: 88/02/17 EXAMINER: HOON/ MOORE ANSWER KEY CANDIDATE:
INSTRUCTIONS TO CANDIDATE:
Use separate pa)er for the answers. Write answers on one side onl Staple question sleet on top of the answer sheets. Points for each question are indicated in parentheses after the cuestion. The passing grade requires at least 70% in each category anc a final grade of at least 80%. Examination papers will be picked upsix(6) hours after the examination start % OF
CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY t
25.00 25.00 PRINCIPLES OF NUCLEAR POWER -
PLANT OPERATION, THERMODYNAMICS,
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HEAT TRANSFER AND FLUID FLOW
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i 25.00 25.00 PLANT DESIGN INCLUDING SAFETY t
AND EMERGENCY SYSTEMS
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25.00 25.00 INSTRUMENTS AND CONTROLS 25.00 25.00 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL ,
CONTROL 100.00 % Totals Final Grace
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All work done on this examination is my ow I have neither given nor received ai I Candicate's Signature
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: Cheating on the examination means an automatic denial of your application and could result in more severe penaltie . Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil only to facilitate legible reproduction . Print your name in the blank provided on the cover sheet of the examinatio . Fill in the date on the cover sheet of the examinWen (if necessary). Use only the paper provided for answer . Print your name in the upper right-hand corner of the first page of each section of the answer shee . Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer shee . Number each answer as to category and number, for example,1.4, . Skip at least three lines between each answe . Separate answer sheets from pad and place finished answer sheets face down on your desk or tabl . Use abbreviations only if they are commonly used in facility literatur . The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer require . Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or no . Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . If parts of the examination are not clear as to intent, ask questions of the examiner onl . You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been complete ,
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~18. When you complete your examination, you shall: Assemble your examination as follows:
(1) Exam questions on to (2) Exam aids - figures, tables, et :
(3) Answer pages including figures which are part of the answe Turn in your copy of the examination and all pages used to answer the examination question Turn in all scrap paper and the balance of the paper that you did ,
not use for answering the question ,
. Leave the examination area, as defined by the examine If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoke ,
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. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2 .
THERMODYNAMIC 5. HEAT TRANSFER.AND FLU 10 FLOW QUESTION 1.01 (2.50)
The reactor is brought critical at 6 n IRM range 2 with the MINIMUM permissible stable positive period & wed by procedure GP-2. Heating power is determined to be 405Fon range 8 of IRM' WHAT is doubling time if period remains constant? (1.0) HOW long will it take for power to reach the point of adding heat if period remains constant? (1.5)
QUESTION 1.02 (3.00)
Regarding the xenon transient following a significant decrease in reactor power from high power operation: If the decrease in reactor power was from 100% to 50%, WHY is the equilibrium xenon reactivity more than one-half the 100% equilibrium value? (1.0) WOULD the resultant peak from a 100% to 50% power maneuver occur in the same time as the result peak from a 100% to 0% power maneuver? EXPLAIN your answe (1.0)
NOTE: Consider all production and removal mechanisms in your answe WHAT would be the effect on the core axial power profile as a result of a down power maneuver from 100% to 50's?
WHY7 (1.0)
QUESTION 1.03 (3.00)
An EHC load reject occurs at 100% core thermal power with the EHC system aligned for normal 100% power generatio DESCRIBE and DISCUSS how the following parameters respond during the first five minutes subsequent to the opening of the generator output breakerg
$ W Reactor Power Reactor Pressure Reactor Water Level (***** CATEGORY 01CONTINUEDONNEXTPAGE*****)
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, '
, PAGt THERMODYNAMICS, HEAT TRANSFER.AND FLUID FLOW QUESTION 1.04 (2.00)
l PBAPS Unit 3 is taken critical during startup and a steady-state ;
period is established. Afterthepointofaddingheat(P0AH),
the reactor seriod lengthens to infinity, and the reactor operator notes that tie moderator temperature has changed from 240 degrees F to 260 degrees WHAT reactivity coefficients turned reactor power? LIST them in order from the largest effect to the least effec (1.0) HOW much positive reactivity was added to establish a stable positive period after criticality was obtained? (1.0)
QUESTION 1.05 (2.50)
SELECT the appropriate response for each of the following
, statements concerning Control Rod Worth: (MORE/LESS) control rods would need to be pulled to make the reacter critical at 545 deg F, as opposed to 140 deg (0.5) An INCREASE in the Void craction will result in a/an (INCREASE / DECREASE) in individual control rod wort (0.5) Control Rod Worth will (INCREASE / DECREASE) with an INCREASE in moderator temperatur (0.5) Control Rod Worth at End of Cycle would be (GREATER /LESS)
than at the Beginning of Cycl (0.5) Control Rod Worth will (INCREASE / DECREASE) as the adjacent control rods are withdraw (0.5)
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" PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4 .
THERMODYNAMICS, HEAT TRANSFE.1. AiiD FLUID FLOW QUESTION 1.06 (2.75)
ANSWER the following questions concerning "CRITICAL POWER." DEFINE "Critical Power." (1.0) WHICH one of the following conditions would tend to INCREASE the Critical Power level assuming all other variables remain unchanged? (0.75 )
NOTE: ASSUME NORMAL FULL-POWER OPERATING CONDITIONS (1) Reactor pressure is INCREASED (2) The axial power peak is RAISED (i.e., power profile peaks higher in the core)
(3) Coolant flow rate is INCREASED WHAT fuel failure mechanism is associated with exceeding
"Critical Power?" (0.5) WHAT thermal limit has been established to ensure critical power is not exceeded? (0.5) ,
QUESTION 1.07 (2.50)
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As a reactor operator coming on shift, you are told that the previous shift performed a reactor shutdown and commenced a cooldown from 1000 psig at 0600. It is now 0730 and you note that wide range reactor pressure is 200 psi Your shift is to place the reactor in shutdown cooling, HAS the previous shift exceeded the Technical Specifica-tion maximum allowable cooldown rate? (INCLUDE in your answer the PBAPS TECHNICAL SPECIFICATION COOLDOWN LIMIT and the assumptions and calculations used.) (1.5) HOW many more degrees of cooldown are necessary before RHR can be unisolated for shutdown cooling? (INCLUDE your assumptions and calculations.) (1.0)
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, PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, P"AGE 5 ,
THERMDDYNAMICS, HEAT TRANSFER AND FLUID FLOW i
QUESTION 1.08 (2.00) !
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Concerning the Bypass Flow in the reactor core: DEFINE core bypass flo (1.0) STATE the two most significant consecuences that would occur if bypass flow were significantly recuced at full powe (1.0)
QUESTION 1.09 (3.00)
Concerning Net Positive Suction Head (NPSH) with regard to the Reactor Recirculation Pumps:
, PROVIDE a brief definition of NPS (0.5) STATE how the AVAILABLE NPSH changes for a reactor in Cold Shutdown (INCREASES, DECREASES, or REMAINS THE '
SAME)ineachofthefollowing: Reactor water level decreases from normal level to just
, above the low level scram setpoint (no change in feed-waterflow). (0.5) Reactor water temperature decrease (0,5)
- Recire, pump speed is changed from 20% speed to 30%
spee (0,5) The reactor vessel pressure increases from 0 psig to 200 psi (0.5) STATE what effect can occur if NPSH requirements are not
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, PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
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QUESTION 1.10 (1.75) ,
During a reactor startup the source range count rate response for a single notch of control rod withdrawal will change as criticality is approached. Assuming a constant moderator temperature and a uniform control rod worth: WILL the AMOUNT OF CHANGE in count rate INCREASE, DECREASE, '
or REMAIN THE SAME for a single notch of control withdrawal as criticality is approached? (0.5) WILL the AMOUNT OF TIME necessary for count rate to stabilize between the control rod notch withdrawals INCREASE, DECREASE,-
or REMAIN THE SAME as criticality is approached? (0. 5,\ WHAT criteria are used by the operator to determine when criticality has been achieved? LIST three (3) criteri (0.75)
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, PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PhGE 7 .
QUESTION 2.01 (2.75)
Regarding the Reactor Recire. Flow Control System and the Reactor Recirc System: STATE for EACH of the recirc speed limiters: the purpose of the limiter (1.0) the initiating logic signals (INCLUDE setpoints). (1.0) DESCRIBE HOW the Recire Pumps respond to a valid ATWS signa INCLUDE specific electrical breakers affected (breaker I.D. numbers not required). (0.75 )
QUESTION 2.02 (2.50)
The reactor water cleanup system is in operation with one pump and one filter demineralizer in service. A reactor startup and heatup is in progress with wide range reactor pressure indicating 400 psi The RWCU dump valve is open, rejecting water to the main condenser to control reactor water level. Suddenly, the operator receives a RWCU low pump flow alarm and notes that system flow is 0 gpm and the previously running pump has stoppe Given that the containment inlet and outlet isolation valves did not close, STATE four (4) possible causes
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of the pump tri (1.0) STATE whether the RWCU dump valve (CV-55) will isolate CONCURRENTLY with any of the pump trip (0.5) la the above example, if the operator also notices that an RWCU isolation has also occurred, STATE HOW the RWCU dump valve position at the time of the isolation can cause significant stress upon the RWCU system piping and component (1.0)
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, PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS P' AGE 8 .
QUESTION 2.03 (2.50)
The main turbine is on line with the reactor operating at 92% of rated core thermal power. One main steam relief valve fully opens though reactor pressure is less than its setpoint, causing a
"Safety Relief Valve Open" alarm, In addition to the relief valve position indications and torus temperature and level indications, LIST four (4)
OTHER means or indications by which the operator could use to quickly verify the valve has opene (1.0) Suppose in the example above, the relief valve reseated but the vacuum breaker in the relief valve discharge piping to the suppression pool failed to open. EXPLAIN HOW plant systems and components could be adversely affected on a subsequent lift of the relief valv (1.5)
QUESTION 2.04 (3.00)
You are making a tour of an emergency diesel generator room and you notice that the engine control transfer switch at the Emergency Diesel Panel (EOP) is selected to the "NORMAL" positio EXPLAIN WHY this emergency diesel generator could not be considered operable if this switch was selected to "AT ENGINE." (0.5) When aligned for standb (automatic) operation, suppose the voltage drop (droop selector switch at the local voltagecontrolcabinetandthegovernordrop(droop)
selector switch at the local engine generator cabinet were both in the "PARALLEL" position. EXPLAIN HOW the diesel and generator would respond if the diesel generator were to subsequently auto start on a LOCA signal concurrent with a complete loss of offsite powe (INCLUDE in the explanation WHY this would be undesirable.) (2.0) The diesel has a "Start Failure Relay" that locks out the diesel if it does not achieve 250 RPM within 10 seconds of receiving a start signal. STATE why such a relay is needed in the event a diesel fails to star (0.5)
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. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PhGE 9 .
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i QUESTION 2.05 (2.50)
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The Core Spray System is in a standby lineup. Normal power is available. Rx pressure is 100-psi STATE ALL logic signals required to cause automatic initiatio (1.0) Upon receiving a valid auto initiation signal, if the "A" core spray pump motor supply breaker were to fail to close, STATE whether core spray loop I WILL or WILL NOT pump water to the vesse (0.5) If both the "A" and "C" core spray pumps were operating at rated loop flow in the full flow test mode (torus to torus)
for surveillance and one pump were to trip, STATE how the remaining operating pump would be adversely effecte (0.5) Should the loop A testable check AIR OPERATOR MECHANISM (AO-13A) bind and become immovable in the CLOSED condition, STATE whether core spray loop I WILL or WILL NOT pump water to the vesse (0.5)
QUESTION 2.06 (2.00)
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- TheAreactor operator on shift suddenly notes that "A" and "B" Emergency Service Water (ESW) Pumps auto started, as well as the
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Emergency Cooling Water (ECW) pum). The ECW pump subsequently trips with no operator action. T1e Service Water system is still in service and operating with normal flows and pressures. All systems had been previously aligned for normal operatio STATE ALL the conditions that cause an automatic start of the ESW pump (0,5) STATE the reason why the ECW pump subsequently trippe (Assume no abnormal conditions caused the ECW pump to trip.) (0.5) Given the conditions described above, WHAT loads are being supplied cooling water by the running ESW pumps if no operator action is taken subsequent to the ESW pumps /ECW pumps auto star (1.0)
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, PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 10 .
QUESTION 2.07 (3.00) LIST four (4) major shysical barriers against fission product release to t1e environment BESIDES the primary containmen (1.0) DESCRIBE the conditions under which significant concentra-tions of hydro en could develop inside containment (do not discuss radiol tic decomposition of water or corrosion of containment pi ing and structures). (1.5) EXPLAIN HOW the Containment Atmospheric Control (CAC)
system limits the danger of explosion in the event large amounts of h STATE the specific limit (s) ydrogen are produce for the containment gas concentra-tion (s) as required by Technical Specification (0.5)
QUESTION 2,08 (3.00)
The reactor is o)erating at 90% of rated core thermal power,100%
rod pattern, wit 1 the main generator on line. The "A" stator water cooling pump is tagged out (blocked), DESCRIBE the respense of the EHC system, reactor recirc pumps, and reactor power to a trip of the "B" stator water cooling pump, given the transient does not induce conditions requiring a turbine trip or RPS scram.
- INCLUDE in your discussion the time delay setpoints for major component trips and also final reactor powe ASSUME that no operator action is take (2.0) DESCRIBE the two (2) main turbine trips that are enabled
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on a total loss or stator water cooling. Include in your description all time dolay setpoint (1.0)
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, PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 11 QUESTION 2.09 (2.50)
The reactor is at normal operating pressure and temperature, operating at 70% of rated core themal power. The reactor operator manually scrams the reactor. Prior to the scram, control rod 18-31 was fully withdrawn, LIST ALL pressure sources other than the accumulator nomally available to drive control rod 18-31 inward during the scra (1.0) For the reactor conditions described above, if accumulator 18-31 was not precharged with nitrogen and the gage at the accumulator read 0 psig prior to the scram, STATE whether control rod 18-31 WOULD or WOULD NOT Tully inser (0.5) STATE the final position of control rod 18-31 if the scram pilot valves for HCU 18-31 failed to re)osition when the scram was initiated, assuming all otler scram actions properly functioned. EXPLAIN WHY for full credi (1.0)
QUESTION 2.10 (1.25)
The RCIC system is in operation on your shift to demonstrate operability for Technical Specifications, DESCRIBE the RCIC system response if reactor water level exceeds +45 inche (0.75) STATE whether operator action (WOULD/WOULD NOT) be required to permit the RCIC system te inject to the reactor if a reactor low-low water level condition occurs subsequent to the high level condition described in part "a" abov (0.5)
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, INSTRUMENTS AND CONTROLS
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QUESTION 3.01 (3.00)
For EACH of the following reactivity control systems, STATE: the purpose for the syste . the conditions at which the system is ENABLE . what provisions, if any, exist to manually bypass the ENTIRE system during operation, INCLUDING the administrative requirements that must be met to bypass a system where such a provision has been mad rodworthminimizer(RWM) (1.5) rodsequencecontrolsystem(RSCS) (1.5)
QUESTION 3.02 (3.00)
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You are theAReactor Operator on shift during a refueling outag Refueling operations are complete and you have been told during your turnover that your shift will be draining the reactor vessel into the normal shutdown band. Yarway and narrow range water level indication are presently upscale on all channels. Bottom head drain temperature is 110 deg STATE which indication -(YARWAY RANGE or NARROW RANGE) should
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come on scale first as actual water level is lowere (0.5) EXPLAIN WHY Yarway and narrow range indications will disagree by as much as 30 inches during cold shutdow NOTE: Be sure your explanation includes principles of dete tor operatio (1.5) During a small break LOCA, STATE whether Yarway indication could be HIGHER or LOWER than actual leve ALSO STATE how large a deviation could be expecte (1.0)
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i I-QUESTION 3.03 (2.00) i- If the IRMs are indicating 20 on Range 4 and an operator down ranged to Range 3. WHAT trips, if any, would occur 7 WHY7 (0.75) With the mode switch in STARTUP and IRM "C" reading 11 on Range 7, WHAT trip (s) if any, would occur if IRM "C" was down ranged to Range 67 WHY7 (0.75) An associated APRM downscale concurrent with an IRM upscale HI-HI will cause an RPS scram signal. Wii!CH APRM channel is associated with IRH channel "H"? (0.5)
QUESTION 3.04 (3.00)
The reactor is in cold shutdown with the "B" loop of RHR in shutdown cooling at a flow of 10,000 gpm using RHR pump "B".
All other RHR pumps are secured and aligned for standby opera-tion. The "D" high pressure service water pump is cooling water to the "B" loop RHR heat exchanger RHR providing RPV head spray is not in service, DESCRIBE WHAT automatic valve and pump actions should occur in the RHR system if reactor water level decreases to -10 inches with no operator actions take LIMIT the description to only those components in the shutdown cooling flow pat (1.0) DESCRIBE W!!AT operator actions, if any, are required to inject into the RPV with the "A" loop of RHR the "B" loop of RHR if water level continues to fall to -140 inche ASSUME all automatic actions properly occu (2.0)
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, INSTRUMENTS AND CONTROLS P' AGE 14 .
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QUESTION 3.05 (3.00)
ConcerningtheReactorProtectionSystem(RPS): EXPLAIN WHY a fast transfer of the E32 4.16KV bus will not cause an RPS half scram, but transferring RPS bus "A" to its alternate source of power will cause a half scra (1.0) Following a reactor scram from 100% power, several scram set)oints that were in effect at power are bypassed by eitler automatic or normal post scram operator action LIST three (3) of these trips and STATE the conditions that must be in effect for the bypass to occu (1.5) One of the RPS inputs is main turbine stop valve (TSV)
position. STATE whether a SCRAM, HALF SCRAM, or NO HALF SCRAM would occur if TSV #1 and TSV #2 both drifted to 85% of full open position with the reactor at 85% of rated core themal powe (0,5)
QUESTION 3.06 (3.00)
ConcerningtheAutomaticDepressurizationSystem(ADS): Once ADS has commenced blowdown, STATE ALL the operator actions that e. auld be taken in Unit 2 to reclose the relief valves prior to reactor prescure decreasing Selow 50 psi (1.0) STATE which signal in)ut to ADS logic must in all cases be manually reset when tie signal clear (0.5) The ADS logic received a modification from its initial design that added a 9.5 minute timer and a keylock hand-switch to each logic train. STATE the purpose (s) of this additional timer and handswitch. DESCRIBE in your statement how the timers and handswitenes affect the logic. INCLUDE setpoints associated with these device (1.5)
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, INSTRUMENTS AND CONTROLS ,
P1GE 15 ,
QUESTION 3.07 (3.00)
Unit 2 is operating at 100% core thermal power with the main generator producing 97% rated electrical power. The EHC system is aligne r nomal 100% power generation. An electrical fault 3-=. causes generator output breaketyto open. You may use the attac d EHC system figure to answer the following question DESCRIBE the sequence of events (i.e., signals generated andcomponentsactuated)thatwillleadtothefirst scram signal to the reac;or protection system (RPS).
(Assume no operator action.) (1.5) WOULD any automatic changes in the setpoints for either LOAD LIMIT or LOAD SET occur in this transient? If so, DESCRIBEthechange(s),INCLUDINGthefinalsetpoint value(s). {1.5)
QUESTION 3.08 (2.00)
Reactor power is 85% of core rated thermal power and total core flow is 70%. Neither rod block monitor (RBM) channel is bypasse Control rod 18-31 is selected, STATE the effect upon the RBM system when the reactor operator bypasses APRM channel "E" with its respective APRM joystick for an I&C surveillanc (0.5)
, If control rod 18-31 is selected and APRM channel "B" indicates 885r, STATE whether the RBM system WOULD or WOULD NOT enforce a rod block. INCLUDE a calculation AND an explanation to support your answe (1.0)
l (M c(C red ved rodynAv If the operator were to deselectYcontrol rod 18-317and did not select another control rod, WOULD the RBM system respond with a rod block? (0.5)
(*"" CATEGORY 03 CONTINUED ON NEXT PAGE ""*)
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PEACH BOTTOM QUESTION 3.07 FIGURE PAGE 15a e
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, INSTRUMENTS AND CONTROLS ,
PhGE 16
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QUESTION 3.09 (1.50)
The reactor is in cold shutdown with bottom head drain temperature of 120 deg F. The standby gas treatment auto initiates with the "C" fan. Both filter train isolation dampers are ope STATE the conditions that could cause standby gas treatment (SBGT) to auto initiat (1.0) STATE whether the initiating signal was from Unit 2 or Unit (0.5)
QUESTION 3.10 (1.50)
The reactor is operating at 100% rated feedwater flow and 99%
rated core themal power. The "A" condensate pump trip DESCRIBE HO'l interlocks in tne following systems would respond: feedwater control system ( EHC system (0.75)
(***** END OF CATEGORY 03 *****)
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- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PkGE 17
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RADIOLOGICAL CONTROL QUESTION 4.01 (3.00)
Regarding Administrative Procedure A-7, "Shift Operations:" STATE which individual by title is required to authorize a startup subsequent to a shutdown or scra (0.5) Appendix 5 of A-7 lists the specific duties of the control room operato STATE the three (3) conditions under which the control room operator is re;ponsible for and has the authority to shutdown the reacto (1.5) Section 7.1 of A-7, "Shift Operations" contains guidance for
"On-Duty" senior licansed operators and licensed operators concerning their PERSONAL CONDUCT while on shift. STATE two (2) of these guidelines that help ensure the units are operated as safely and as reliably as possibl (1.0)
QUESTION 4.02 (2.75) gmb /4WI
~ The control room becomes uninhabitable because of aAfDe and the decision has been made to immediately evacuate the control roo LIST the seven (7) immediate actions thc operator is to take PRIOR to exiting the control room as delineated by procedure SE-1, "Plant Shutdown from the Emergency Shutdown Panel - Procedure." (2.0) Once at the emergency shutdown panel, procedure SE-1 instructs the operator to place all the pistol grip hand switches on the emergency shutdown panel in the "pulled-out" position. STATE the purpose of this actio (0.75 )
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
O
. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PkGE 18 .
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RADIOLOGICAL CONTROL QUESTION 4.03 (2.00)
A reactor startup is in progress. Reactor pressure is 920 psi Reactor power is 2% by APRM. The "A" reactor recirc pump trips and a few minutes later, the "B" recirc pump trips, LIST the immediate operator actions of 0T-112, "Recirculation Pump Trip." (INCLUDE actions for one recirc pump trip.) (1.0) STATE whether an automatic reactor scram from the reactor protection system WOULD or WOULD NOT occur in the above situation and SUPPORT your answer with an explanatio (1.0)
QUESTION 4.04 (2.00)
ON-105, "Control Rod Uncoupled-Procedure," provides instructions to follow in the event of an uncoupled control rod, LIST three (3) indications of an uncoupled control ro (1.5) HOW many recoupling attempts are allowed by ON-1057 (0.5) ,
QUESTION 4.05 (2.50)
The reactor was in hot standby with a bottom head drain terr.perature of 480 deg F. The high pressure coolant injec-tion (HPCI) system auto initiated on a valid initiation signal while the reactor core isolation cooling (RCIC)
system remained in standby. Based on the responses of these two systems alone, STATE whether TRIP procedure entry condi-tion (s) existed and if so, SPECIFY which entry condition (s)
AND also which procedure (s) should have been entered. Assume the HPCI and RCIC systems are initially properly aligned in standby for automatic initiation and are fully operabl (2.5)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
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. .. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 19
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RADIOLOGICAL CONTROL ,
QUESTION 4.06 .(1.50)
A hallway is surveyed and is routinely found to have a radiation field present whereby a person would receive 8 mre.i/hr. No loose surface contamination or airborne contamination is presen STATE what entry requirements an individual would have to satisfy to physically enter this hallwa (1.0) At PBAPS, STATE the station's administrative quarterly dose limit for an individual who does NOT have a completed NRC Form 4 on fil (0.5)
QUESTION 4,07 (2.75)
During reactor heatup and pressurization, procedure GP-2, "Normal Plant Startup" is the controlling procedure, STATE what administrative restriction exists if the reactor is brought critical and the EHC system is not availabl (0.75) GP-2 administratively limits heatup rat STATE the maximum allowable heatup rate allowed by GP- (0.751 DESCRIBE HOW the EHC pressure control function is first demonstrated properly functional in GP-2 during reactor startup/heatu (1.25)
QUESTION 4.08 (1.50)
t l The reactor producing 100% rated core thermal powe The
'
"Drywell Hi-lo Press" alarm is received. The reactor operator then notes that drywell pressure is 0.8 psig, LIST all immediate action (1.0)
l STATE what condition must be veri.ied prior to the l initiation of the standby gas treatment system to l vent the drywell of excess pressur (0.5)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
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' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 20 RADIOLOGICAL CONTROL .
QUESTION 4.09 (2.00)
A surveillance test is to be performed on the "A" loop of core spray for unit 2. According to procedure A-41, "Procedure for Control of Safety Related Equipment," the individual who performs this test must receive prior approva After shift supervision approval is received to perform this test, STATE which other )erson must also approve the performance of this test and )e advised of the anticipated effects of this tes (0.5) LIST what information must be logged by the reactor operator or control operator concerning this. tes (1.0) STATE the TYPE of verification that must be performed upon completion of this test to ensure that equipment and components affected by this test are properly returned to servic (0.5)
QUESTION 4.10 (2.00)
Concerning Trip Procedure General Notes: STATE the conditions under which manual control of ECCS systems is allowe (1.E)
.,
b' . STATE the conditions under which an automatic initiation
- of a safety function can be assumed NOT to be due to a true initiating even (0.5)
QUESTION .50)
- Radiation workIpermits (PWPs) control work performed in the l radiologicall mcontrolled area (RCA). Operations personnel I have two 'bc.et.eiGRWP's" in effect at all times, one at each l unit, i.llowing operators to perform certain functions. STATE three (3) operations functions these two RWPs together l
allo (1.5)
(***** CATEGORY 04CONTINUEDONNEXTPAGE*****)
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~ PROCEDURES-NORMAL, ABNORMAL,EMERGENCYAND PAGE 21 RADIOLOGICAL CONTROL QUESTION 4.12 (1.50)
Procedure ON-103, "Off Gas Stack High Radiation - Procedure" consists of a series of checks to determine the source of the '
activity followed by actions to be taken based upon the results of those check g y ,L.madj STATE which SYSTEM besides off g ischarges to the (* * W /
off gas stac (0.5) STATE two (2) actions that must be taken if the source of activity is in fact from unit 2 off ga (Assume the reactor is at rated power with the main turbine generator on line with a normal full power lineup.) (1,0)
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(************ END OF EXAMINATION ************)
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' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 22 .
MRM0 DYNAMICS, HEAT TRANSFER AND TLUID FLOW ANSWERS -- PEACH B0TTOM 2&3 -
88/02/17-M00N/H00RE ANSWER 1.01 (2.50) From GP-2, period-equals 50 seconds [+0.5]. Thus doubling time e is 50/1.44 = 34.7 seconds [+0.5]. dnge2iseualto nge 8 [+0.5]
P(0) = 0.06 P( ) = 40 Period = 50 seconds P(t) = P(0) e ^(t/ period)
40 = 0.06 e ^(t/50 sec)
t = 50 in 40/0.06 = 325. sec or 5 min 25 sec [+1.0]
REFERENCE Peach Bottom: LOT-1430; 1530 L0 # K108 ...(KA'S)
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,. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PhGE 23 .
THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N/M00RE ANSWER 1.02 (3.00) {Xe(ec .) --> Production = Removal . The Production term is a flux cependent term; the Removal term is dependent on burnup anddecaywithonlyburnupbeingfluxdependent.}
While all the production term would reduce by one-half [+0.5],
only the burnup portion of the removal term would reduce by one-half; thereby leaving the xenon concentration at a higher equilibrium level [+0.5] --0R-- Since the major production and removal terms are based on two independent variables
[+0.5], flux and decay, Xe versus power is not linear [+0.5].
+1.0 maximu , No [+0.5], because xenon peak following a scram is approximately the square root of the power change (10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />)
[+0.25] due to the loss of the burnup tenn [+0.25].
+1.0 maximu Power would peak higher in the core [+0.5] because the amount ,
of xenon that would build into the core would be greatest where p density had been the greatest arior to the down power maneuver. This in turn would make tie bottom of the core less reactive and force power production higher in the core until equilibrium xenon concentration was again
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established [+0.5] .
REFERENCE Peach Bottom: LOT 1510, LO #3, 4, and K105 292006K106 292006K107 ...(KA'S)
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.- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PhGE 24
THERMODYNAMICS, HEAT TRANSFER.AND FLUID FLOW ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N/M00RE j ANSWER 1.03 (3.00) Reactor power will rapidly increase due to a pressure increase [+0.5]. Power will then decrease due to the TCV fast closure scram [+0.5]. Reactor pressure will rapidly increase due to the rapid closure of the TCV's [+0.5] . Pressure will then decrease due to the scram and the opening of the bypass valves which will then attempt to maintain reactor pressure at 920 psig [+0.5]. Reactor water level will initially drop as steam f Mw is abruptly interrupted [+0.5]. The feed control system will respond to increase level and level should then rise to the level controller setpoint2 [+0.5].
' ( M may o-x e m s % p ps. e te.g nr + vs- % 8' d l )s REFERENCE Peach Bottom: LOT 1600, LO #1 and # K101 241000K102 241000K103 ...(KA'S)
ANSWER 1.04 (2.00) . mod temp coeff [+0.25] fuel temp coeff [+0.25] void coeff [+0.25]
[+0.25] for correct order Assume (1.) contribution from void and fuel temperature l coefficient insignificant and (2.) moderator temperature
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coefficient = 1 x 10**-4 k/k/deg [+0. 5]
l [1 x 10**-4 (k/k)/deg F] x [(260 - 240) deg F] = 20 x 10**-4 k/k added [+0.5]
l REFERENCE Peach Bottom: Reactor Theory, Student Handout, Sections 26 through 3 . Peach Bottom: LOT 1440, L0 #3 and K114 ...(KA'S)
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- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION PkGE 25 .
THERMODYNAMICS, HEAT TRANSFER:AND FLUID FL6W ANSWERS -- PEACH BOTTOM 2&3 -81/02/17-M00N/M00RE ANSWER 1.05 (2.50) more decrease increase less increase
[+0.5] each REFERENCE Peach Bottom: LOT 1490, LO 6, K109 ...(KA'S)
ANSWER 1.06 (2.75) The assembly power which would cause the onset of transition boiling at some point in the assembly. [+1.0] (3) [+0.75]
< fuel clad cracking due to a lack of cooling [+0.5]
- MCPR limit [+0.5]
REFERENCE Peach Bottom: LOT 1380, LO # . Peach Bottom: LOT 1360, L0 #3, # K119 293009K120 293009Kl?2 293009K123 ...(KA'S)
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.- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, P'GE A 26
- THERMODYNAMICS, HEAT TRANSFER.AND FLUID FLOW ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N/M00RE ANSWER 1.07 (2.50) The previous shift DID EXCEED the cooldown limit [+0.5] of -
100 deg F/ hour [+0.5].
(Tsat for 1000 psig = 546 deg F; Tsat .for 200 psig = 388 deg F; cooldown rate = (546-388) deg F/1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />
= 105 deg F/hr) [+0.5] +- 2 deg F (of cooldown required).
(Tsat for 200 psig = 388 deg F; Tsat for 75 psig = 320 deg F; (388-320) = 68 deg F) [+1.0]
REFERENCE Peach Bottom: LOT 1150, L0 # . Peach Bottom: LOT 1160, L0 # . Peach Bottom: Technical Specifications, 2.2.2 and 3.6. K402 293003K123 ...(KA'S)
ANSWER 1.08 (2.00)
- (Core bypass flow is) that )ortion of total core flow that does not flow inside tie fuel channels [+1.0]. . Excessive vriding in bypass region resulting in unreliable ifrei readings. [+0.5] Inadequate cooling of LPRM detectors resulting in premature LPRM detector failures. [+0.5]
REFERENCE Peach Bottom: LOT 0010, L0 # K132 293008K133 ...(KA'S)
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.. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PhGE 27
THERMODYNAMICS, HEAT TRANSFER AND TLul0 FLOW ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N/M00RE ANSWER 1.09 (3.00) The differencepressur and saturation between static [+0.5 p]ressure at the eye of the pump . decreases '+0. 5' increases '+0. 5' decreases '+0. 5' increases l+0.5l Cavitation of the reactor recirculation pumps [+0.5]
REFERENCE Peach Bottom: LOT 1290, LO # K109 293006K110 ...(KA'S)
ANSWER 1.10 (1.75) (The amount of change will) INCREAS [+0.5] , (The amount of time necessary for count rate to stabilize between control rod notch withdrawals will) INCREAS [+0.5] . positive stable period [+0.25]
- increasing count rate no rod motion [+0.25] [+0.25]
REFERENCE Peach Bottom: LOT 1530, L0 # K103 292008K108 ...(KA'S)
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.- PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS P' AGE 28
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ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N/M00RE ANSWER 2.01 (2.75) . (Purposes)
The 60% speed limiter reduces reactor power to reduce the rate of reactor vessel inventory loss in a loss of feedwater situation, either actual or anticipated [+0.5].
The 30% speed limiter ensures the system is properly aligned for operation [+0.25] and also ensures adequate NPSH in a low feed water flow situation [+0.25]. (InitiatingConditions)
60% speed limiter EITHER a. Rx water level < 17 inches AND an individual feed pump flow < 20% [+0.25]
OR b. total feed water flow > 90% AND an individual condensate pump not running [+0.25]
30% speed limiter I T*M e &~-
EITHER total feed water flow < 20% [+0.25]
OR recirc pump discharge valve <90% full open k[+"0.25] Both the reactor recire pumps trip [+0.25]. Both recire MG set drive mctor breakers open [+0.5].
REFERENCE Peach Bottom: LOT 0040, L0 #1, #2, #3, 14, and # X414 202002X402 202002K406 ...(KA'S)
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, PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS P GE 29 .
ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-H00N/H00RE
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ANSWER 2.02 (2.50) . Iow pump flow high pump vibration high pump flow pump motor supply breaker trip pump bearing cooling water outlet high temperature pump motor thermal overload trip Any four (4) [+0.25] each, +1.0 maximu The RWCU dump valve (CV-55) WILL NOT concurrently isolat [+0.5] If the dump valve is open at the time of the RWCU system isolation, the system will ra) idly depressurize, the water in the piping will flasi to steam [+0.5] in the high temperature hammering) portions the system pipingof the and system, shocking](water components [+0.5 .
REFERENCE Peach Bottom: LOT 0110, LO #5 and # K106 204000K401 204000K402 204000K407 ...(KA'S)
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, PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 30 .
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ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N/M00RE ANSWER 2.03 (2.50) . relief valve tail pipe thennocouple/high temp alarm relief valve tail pipe acoustic monitor total steam flow / total feed flow mismatch generator MWE reduction
[+0.25] each Following the relief valve's first actuation, the steam in its discharge line would condense causing a vacuum in that line
[+0.5]. This would cause suppression pool water to be drawn into the line [+0.5] to an elevation significantly above the elevation of the suppression pool surface. On a subsequent lift of the relief valve, the inertia of the additional water in the discharge line would cause over pressurization of the line while the additional water was being cleared [+0.5].
(Alsoacceptable: On a subsequent lift of the relief valve, the additional water would cause excessive loading on the diffuserandtorusstructures.)
REFERENCE Peach Bottom: LOT 1210, L0 # K403 239002K406 239002K501 239002K503 ...(KA'S)
ANSWER 2.04 (3.00) The emergency diesel generator would not auto start on a LOCA signal [+0.5] . The diesel would start and align to the bus [+0.5]. As loads cycled onto the bus, bus frequency would drop as load increased [+0.5], and bus voltage would drop as load increased
[+0.5]. This would be highly undesirable because ESF equipment could be operating at speeds less than designed and drawing currents higher than designed as bus loading increased
[+0.5]. This relay prevents air bleed-down of the starting air receivers in the event the diesel failed to start. [+0.5]
REFERENCE Peach Bottom: LOT 0670, L0 #3 and # .. . . _ . . . - __-._ _
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.. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PhGE 31 .
ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N/M00RE Peach Bottom: System Procedure, S.8. K106 26400K403 26400K406 26400K407 ...(KA'S)
ANSWER -2.05 (2.50) . low reactor water level (-130 inches) [+0. 5] high drpell pressure (2 psig) (AND Rx pressure < 450 psig) +
,0.5] will pump water to the vessel [+0.5] The remaining pump would be operating beyond rated flow conditions (pump runout). [+0.5] will pump water to the vessel [+0.5]
REFERENCE Peach Bottom: LOT 0350, LO # K408 209001K410 209001K503 209001SG4 ...(KA'S)
ANSWER 2.06 (2.00) Either a Diesel Generator Start (or 22 seconds after the DG low speed relay (LSR) energized) [+0.25] OR a valid LOCA signal (or 22 seconds after Maximum Credible Accident (MCA)
relay energized) [+0.25]. The ECW aump tripped because normal ESW pressure was establisled by the ESW pumps. [+0.5]
. the Emergency Diesel Generators [+1.0]
REFERENCE Peach Bottom: LOT 0680, L0 #2, #3, and # . Peach Bottom: LOT 0410, L0 # A306 264000K104 264000SG7 ...(KA'S)
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.. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
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PhGE 32 .
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ANSWERS -- PEACH BOTTOM 2&3 - 3/02/17-M00N/M00RE ANSWER 2.07 (3.00) . fuel pellet [+0.25] fuel cladding [+0.25] primary system piping ;+0.25; secondary containment ,+0.25 b.- In a LOCA with inadequate core cooling [+0.5] the zirconium in the fuel rod cladding [+0.5] could react with the water [+0.5]
in a zirconium-water reaction to produce explosive concentrations of hydrogen in containmen The containment fs inerted with nitrogen [+0.25] to reduce oxygencontenttolessthan4%[+0.25].
REFERENCE Peach Bottom: LOT 0130, L0 # . Peach Bottom: LOT 0160, L0 il and # K103 223001K404 223001K509 ...(KA'S)
ANSWER 2.08 (3.00) . The EHC load set will run back [+0.2] to 25% 1:+0.1]
causing all bypass valves to fully open [+0.2ll.
l The "A" reactor recirc pump will trip [+0.3] in seconds [+0.2]. The "B" reactor recire pump will trip [+0.3] in 1 seconds [+0.2] .
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l Reactor power will stabilize at 50-55% of core rated thermalpower[+0.5].
l The turbine will trip if generator amps are not less than 26,530 [+0.1] in 2 minutes [+0.4] or less than 7726 amps
[+0.1] in 3.5 minutes [+0.4].
REFERENCE Peach Bottom: LOT 0630, LO #5 and # K407 241000K123 241000K125 241000K405 ...(KA'S)
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. . PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS ,
PhGE 33 .
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ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N/M00RE ANSWER 2.09 (2.50) . Rx pressure [+0.5] CRD pressure pump0.5] p[+ressure --0R-- charging water header would fully insert [+0. 5] Rod 18-31 should fully insert (position "00") [+0.4]
scram valves would vent off the
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because the backup [+0.3] causing the HCU 18-31 scram scram air header valves to fail open [+0.3] to effect the scram of control rod 18-3 REFERENCE Peach Bottom: LOT 0070, L0 #7, 201003K404 201003K601 201003K602 ...(KA'S)
ANSWER 2.10 (1.25) , The RCIC turbine steam supply valve (M0-131) closes (this is NOT a turbine trip). [+0.75] (operator action) WOULD NOT (be required) [+0. 5]
REFERENCE Peach Bottom: LOT 0380, L0 #2 and # K102 217000K402 217000SG7 ...(KA'S)
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.. INSTRUMENTS AND CONTROLS -
PhGE 34
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ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N/M00RE ANSWER 3.01 (3.00) .- The purpose is to limit peak fuel enthalpy in a postulated rod drop accident to 280 cal /gm [+0.3] Enabled at 25% (decreasing) Rx feed water flow [+0.4] OR
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25% (decreasing) Rx steam flow [+0.4]. A bypass switch has been provided to manucily bypass the RWM and requires a second licensed operator to verify that the operator at the reactor console is following the control rod program [+0.4]. . The purpose is to limit peak fuel enthalpy in a postulated rod drop accident to 280 cal /gm [+0.3]. Enabled at 21% (decreasing) power as measured by main turbine first stage shell pressure [+0.8]. No provision to bypas [+0.4]
REFERENCE Peach Bottom: LOT 0090, L0 #1, #2, and # . Peach Bottom: LOT 0100, L0 #1 and # ' Peach Bottom: LOT 0280, LO #1 and # K104 201002K105 201002K106 201006K501 ...(KA'S)
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. INSTRUMENTS AND CONTROLS *
PAGE 35 l l
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ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N/M00RE l
ANSWER 3.02 (3.00) narrow range [+0.5] At cold conditions, variable and reference leg temperatures are at conditions other than the calibrated conditions expected for the reactor vessel and drywell when operating .
[+0. 5] . The density increase of the colder water introduces error into both reference indications and variable legs p[+roportional 0. 5] . For ato the heights given actual of the reactor water level, the variable leg height for the Yarway is significantly greater than the variable leg height for the narrowLgera . instrument [+0.5] . (A%e , m ,% muu d ~,
&~M w < , m 4 r m e. m W .L u ,s ) Indicated level could be HIGHER [+0.5] by as much as 30 inches
[+0.5].
REFERENCE Peach Bottom: LOT 0050, L0 # . Peach Bottom: LOT 3070, L0 #5 and # K122 216000K501 216000K507 216000K510 ...(KA'S) ,
ANSWER 3.03 (2.00)
- None[+0.25[ . The IRMs would indicate 20 on Range 3, 0-40 scale [+0. An IRM upscale trip (rod block) would occur [+0.25] because the IRM would read '10 on the 0-125 scale [+0.5]. APRM channel "B". [+0.5]
l REFERENCE Peach Bottom: LOT 0250, LO #5 and # K101 215003K102 215005K102 215005K306 ...(KA'S)
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.. INSTRUMENTS AND CONTROLS PkGE 36 .
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ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N/M00RE ANSWER 3.04 (3.00) . Shutdown cooling suction inboard and outboard containment isolation valves (M0-18 and MO-17) would auto clos [+0.5] RHR pump "B" will tri [+0.25] Loop B LPCI injection valve (M0-258) would auto clos [+0.25] The operator must depress the shutdown cooling valve reset pushbutton for the "A" loop of RHR. [+1.0]
The operator must depress the shutdown cooling valve reset pushbutton for the "B" loop of RHR. [+1.0]
REFERENCE Peach Bottom: LOT 0180, LO #2 and # . Peach Bottom: LOT 0370, LO #3, #4, f5, and # .203000K401 205000K404 216000K105 223002K108 ...(KA'S)
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. INSTRUMENTS AND CONTROLS PkGE 37 .
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ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N/M00RE ANSWER 3.05 (3.00) A fast transfer of the E32 4.16 KV bus is fast enough that the inertial energy stored in the flywheel [+0.25] of the RPS motor generator (MG) set will sustain adequate RPS bus "A" voltage and frequency until power is restored to the MG motor by completion of the fast transfer [+0.25]. A transfer of RPS bus "A" deenergizesto its thealternate RPS bus source of power (and causes briefly)but a half scram totally
[+0.5]. . Main steam line isolation scram [+0.25]. Auto bypassed whenthemodeswitchisNOTinRUN[+0.25]. Main condenser low vacuum scram [+0.25]. Auto bypassed when the mode switch is NOT in RUN [+0.25]. Turbine control valve fast closure scram [+0.25]. Auto bypassed when reactor power is <30% as measured by main
,
turbine first stage pressure [+0.25]. Turbine stop valve closure scram [+0.25]. Auto bypassed when reactor power is <30% as measured by main turbine first stage pressure [+0.25]. , Mode switch in SHUTDOWN scram [+0.25]. Auto bypassed after a 10 sec. TD from when the mode switch was manually placedinshutdown[+0.25]. Scram discharge volume (SDV) high level scram [+0.25].
Auto bypassed when mode switch is in SHUTDOWN or REFUEL AND manual keylock bypass switch in BYPASS [+0.25].
pc.g] %pbbp mJ 9, gr q- m _ n_ _ rm , m "M N r w- Wk Any three (3), +1.5 maximu %
l whew h w a.s~l+L a half scram [+0.5] g g p g p o ,g-]
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REFERENCE Peach Bottom: LOT 0300, L0 #4, #5, #6, #7, and # K110 212000K111 212000K114 212000K201 ...(KA'S)
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- 3 .__ INSTRUMENTS AND CONTROLS ,
PhGE 38 .
ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N/M00RE ANSWER' 3.06 (3.00) . Depress both the "A" and "B" timer reset pushbuttons to break the seal i [+0.33] Shutdown the RHR and core spray pump [+0.33] Place the "A" and "B" keylock switches in "Inhibit."
[+0.34] high drywell pressure [+0.5] The additional timer in each logic train automatically inserts a high drywell pressure permissive signal [+0.25] if reactor water level is not restored to a level greater than -130 inches within 9.5 minutes of level falling below -130 inches [+0.25]
The purpose of this feature is to make the logic responsive to a LOCA with the break physically outside the containment [+0.25].
s The keylock switches for each logic train of ADS disable their respective logic trains to prevent ADS relief valve actuation
[+0.5]. The surpose of this feature is to provide a positive means to disaale ADS when under certain accident conditions ADS actuation would be highly undesirable [+0.25].
(ALTERNATEANSWER: The purpose of this feature is to provide a positive means to disable ADS when directed by procedure. [+0.25])
REFERENCE Peach Bottom: LOT 0330, LO #2, 13, and # K107 218000K403 218000K501 ...(KA'S)
ANSWER 3.07 (3.00) An EHC load reject signal [+0.5] will cause an actuation of the EHC system fast acting solenoids [+0.5] which in turn will cause a turbine control valve (TCV) fast closure scram [+0.5]. Yes [+0.5] . LOAD SET will runback [+0.5] to minimum (zero load) [+0.5]. (LOAD LIMIT will not change).
REFERENCE Peach Boctom: LOT 0590, LO # K101 241000K102 241000K103 241000K104 ...(KA'S)
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., INSTRUMENTS AND CONTROLS
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PkGE 39 .
ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N'H00RE ANSWER 3.08 (2.00) to APRM "C" for RBM its APRM"A" referenc will automatically]
[+ shift from APRM "E" Would enforce a rod block [+0.34]. Even upranged to the high setpoint, with total core flow at 70%, rod block setpoint would be [(0.66)(0.7) + 0.41] 100 = 87.2% [+0.33] . APRM channel "B" (at 88%) is the reference channel for RBM channel
"B" and so regardless of the LPRM average of channel "B" about control rod 18-31, the RBM LPRM averaging and gain change circuit will ensure at least 88% power is used fer rod block detennination [+0.33] . Yes [+0.5]
REFERENCE Peach Bottom: LOT 0280, LO #3 and # K101 215002K103 215002K401 215005K103 ...(KA'S)
ANSWER 3.09 (1.50) Low Rx water high drywell level pressure (2 psig (0 inches))
refuel floor exhaust high radiation (16 mr/hr)
Rx building exhaust high radiation (16 mr/br)
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[+0.25] each (Alsoallowlossofpowerto"A"RPSinUnit3) unit 3 [+0.5]
REFERENCE
, Peach Bottom: LOT 0210, LO #2 and # K401 ...(KA'S)
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- INSTRUMENTS AND CONTROLS
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PkGE 40 .
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ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N/M00RE ANSWER 3.10 (1.50) The feedwater control system would limit feed pump speed control signal to 90%. [+0.75) The EHC system would runback at 1%/sec until feed water flow is <95%. [+0.75] ,
REFERENCE Peach Bottom: LOT 0550, LO #13 and #1 K102 256000K301 259002K113 259002K116 ...(KA'S)
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- . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PhGE 41 ' .
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- RADIOLOGICAL CONTROL ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-H00N/M00RE ANSWER 4.01 (3.00) . . _ _ . _ .
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I.__ "..... [+0.5]
- . The safety of the reactor is in jeopard [+0.5] Operating parameters exceed reactor protection system setpoints and automatic shutdown does not occur. [+0.5] When there is doubt as to whether safe conditions exis [+0.5] . (On duty SLO's and LO's) must be alert and attentiv . (On duty SLO's and LO's) must be aware of and responsible for the plant status at all time . (On duty SLO's and LO's) must prohibit distracting activities in the control room, e
Any two (2) [+0.5] each, +1.0 maximu REFERENCE Peach Bottom: LOT 1570, LO #2a and #3 A103 ...(KA'S)
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.. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 42 .
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. RADIOLOGICAL CONTROL ,
ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N/M00RE ANSWER C.02 (2.75) runback recirc flow to minimum [+0.3] transfer house loads [+0.3] manually scram and execute T-100 [+0.3] place the drywell instrument air in service [+0.3] (4.C deydLW close MSIV's [+0.3] kanC+4-4 gg,h4 establish torus cooling [+0.3] obtain master keys [+0.2]
(allow [+0.3] if all six preceding actions are not produced) Placing the switches in the "pulled-out" position transfers control of the' associated components from the control room to the emergency shutdown pane [+0.75]
REFERENCE ; Peach Bottom: SE-1, "Plant Shutdown from the Emergency :
Shutdown Panel - Procedure."
295016AK20 2950165G10 ...(KA'S)
ANSWER 4.03 (2.00)
- If one recire pump is operating, then drive in deep rods fuD y as required to prevent a scram. [+0.33]
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If neither recirc pump is operating, then scram and enter T-10 [+0.34]
If a scram condition occurs, enter procedure T-10 [+0.33] f An automatic scram WOULD NOT occur [+0.5] because power level ;
prior to the loss of both reactor recire pumps was not '
sufficiently high to produce 1 control rod pattern that would produce a power under natural circulation conditions that could exceed a f E bic: d sctam setpoint [+0.5].
WD k. k /13en REFERENCE Peach Bottom: LOT 1540. LO / . . . . - . . - - - - - - - -- . - - - - _ - . - . - -
, _ _ _ _ _ _ _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - . - -
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- . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PhGE 43 .
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RADIOLOGICAL CONTROL i
ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N/M00RE R
295001AK20 295001SG10 ...(KA'S)
ANSWER 4.04 (2.00) r . rod overtravel alann when fully withdrawn [+0. 5] control rod withdrawal with no apparent nuclear response
[+0.5] no control rod drive wn sr "stall flow" observed when perfonning an uncouplir check at position 48 [+0.5]
4 Rexs poi.% ;~ue.A.k y M [+cA] three [+0. 5] 5 REFERENCE Am fsree.l3) fIEvnadm Peach Bottom: LOT 1550, LO #1 and # SG15 ...(KA'S)
ANSWER 4.05 (2.60)
A trip procedre entry condition DID exist [+0.5]. The entry condition was .igh drywell pressure > 2 psig [+0.5], and the TRIP procedures which should have been entered were T-102, "Containment Control" [+0.5]; T-101, "RPV Control" [+0.5]; and T-100, "Scram"
[+0.5].
REFERENCE Peach Bottom: LOT 1560, LO # G11 ...(KA'S) ,
i ANSWER 4.06 (1.50) The individual would be required to be wearing his personal dosimetry consisting of his TLD [+0.5] and a self reading dosimeter [+0.5]. > mrem /qtr [+0.5]
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REFERENCE Peach Bottom: LOT 1730, L0 #1 and #2.
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294001K103 ...(KA'S) ,
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4 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 44 .
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RADIOLOGICAL CONTROL .
ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N/M00RE ANSWER 4.07 (2.75)
' The reactor operator, upon achieving criticality, must insert control rods to make the reactor slightly suberitical. [+0.75] deg F/hr [+0.75] At a reactor res ure o psi setpoint set a psig [+0.25]g with the EHC pressure, bypass valves are verified to open as necessary [+0.5) to maintain reactor pressure at psig as control rods are withdrawn [+0.5].
REFEERCE Peach Bottom: LOT 1530, L0 #2, f4, and # SG10 216000"' 241000SG10 241000SG1 ...(KA'S)
ANSWER 4.08 (1.50) . maximize drywell cooling tenninate drywell inerting [+0.34]]
[+0.33 if a scram conditions occurs, enter T-100 [+0.33] Drywell temperature c:st be less than 212 deg [+0.5]
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REFERENCE Peach Bottom: LOT 1540, LO f1 and # EA20 2950245G11 ...(KA'S)
AKSWER 4.09 (2.00) reactor operator [+0.5] (ako 4 - hedy\w ,.u . A,c.o.} surveillance test number [+0.33]
equipment affected [+0.34) (au, > poc,An. 4aH<4 start date/ time [+0.33](au Ca ora t te 4 AA h p.ht. %u mL double (verification) [+0.5]
(also allow independent verification for full credit)
REFERENCE Peach Bottom: LOT 1570, L0 # _ _ _ _ _ _ _ _ _ _ _ _ _
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- . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PhGE~45 .
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RADIOLOGICAL CONTROL ,
ANSWERS -- PEACH BOTTOM 2&3 -88/02/17-M00N/H00RE 294001A103 294001A106 ...(KA'S)
ANSWER 4.10 (2.00) B at least two indications [+0.5] that either:
1 misoperation in the automatic mode is confirmed [+0.5] or 2 adequate core cooling is assured [+0.5]. Automatic initiation can be assumed not to be due to a true initiating event when supported by at least two independent verifications. [+0.5]
REFERENCE Peach Bottom: LOT 1560, L0 #3 and # SG7 ...(KA'S)
ANSWER 4.11 (1.50) [+0.5] (operator)[+0.5]
blocking rounds inspection [+0.5]
REFERENCE Peach Bottoir: LOT 1760, LO # K103 ...(KA'S)
ANSWER 4.12 (1.50) standby gas treatment (SBGT) [+0.5] . place the unit 2 main turbine sealing steam supply on auxiliary steam [+0.5] drop load on unit 2 to maintain off gas stack radiation below the alert limit [+0.5]
REFERENCE Peach Bottom: LOT 1550, LO # K107 271000K111 E71000K302 272000K103 ...(KA'S)
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,' IQUA7'ca :.nEir
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..- = .- .
f o ma v a s/t Cycle efficiency.= (Net' sort
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out)/(Energyin)
s = Va t + 1/2 at2
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w = mg ,
E = me-KE = 1/2 mv a = (Vf - Vg )/t A = tri A=Ae'"
g PE = mgn '
Vf = V, + at' w = e/t x = tn2/t 1/2 = 0.593/t1/2
- t y , y 3p 1/28ff * E( *U)( **) 3
[(t1/2) * (*b)3 '
f.E = 931 un
,
I=Ie" g ,
Q = mCpat Q = UA I = I n
e~"*
Pwr = Wf I=I n 10- N T'/L = 1.3/u P = P 10 sur(t) HVL = -0.593/u t
P = Po e /T .
SUR = 25.05/T SCR = S/(1 - K,ff)
CR x
=
'/(1 - K,ffx)
SUR = 25a/t* + (s - o)T CR)(1 - K ,ff)) = CR 2 (I - eff2)
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T = (1"/c ) + [(S - o )/ 0 3 M * I/II - Neff) * CS /CR l g T = 1/(o - 8) M * (I - Keffo)/II - Keffl)
T = (s - o)/(4o) SDM = (1 - Keff)/Keff
,
a = (Keff-N/Keff = deff/K eff S***
[* A= seconds $
a = ((t=/(T K,ff)] + (T,ff /(1 + IT)]
=Id I)d) 2 ,2 2 '#
P = ( nV)/(3 x 1010) I d) 7 22 2 I = oN R/hr = (0.5 CE)/d (meters) .
R/hr = 6 CE/d2 (feet)
Water Parameters Miscellaneous Conversions I gal. = 3.345 le curie = 3.7 x 1010eps 1 ga;. = 3.78 liters 1 kg = 2.21 lom 1 ft* = 7.48 ga I np = 2.54 x 10 3 Stu/nr -
Density = 62.4'10m/ft3 1 w = 3.41 x 100 Stu/nr Density = 1 gm/ lin = 2:54 cm Heat of vapori:ation = 970 Stu/lem '
'F 9/5'C + 32 Heat of fusion = 144 Stu/lem 'C = 5/9 (*F-32)
1 Atm = 14.7 Psi = 29.9 in. H9- 1 STU = 778 ft-lbf i ft. H O
= 0.4335 Bf/i .
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