ML20215C523
| ML20215C523 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 09/30/1986 |
| From: | Clark R, Higgins R, James Smith, Upton J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20215C516 | List: |
| References | |
| 50-255-OLS-86, 50-255-OLS-86-0, NUDOCS 8610100320 | |
| Download: ML20215C523 (139) | |
Text
{{#Wiki_filter:- U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-255/0LS-86-01 Docket No. 50-255 License No. DPR-20 Licensee: Consumers Power Company 212 West Michigan Avenue Jackson, MI 49201 Facility Name: Palisades Nuclear Plant Examination Administat ed At: Palisades Nuclear Plant, Covert, MI and Midland Training Center, Midland, MI Examination Conducted: June 24, July 22, 23 and 24, and August 27, 1986 Examiner (s): . Hi g ?[3d[f6 Date JV R. G. Clark 'l[30[f( Date-J. D. Smit h 3ddl Date J n Date [fd 7 Approved By: 4f&k hom)as M. Burdick, ChiefV o b Operating Licensing Section Da'te ' Examination Summary Examination administered on June 24, July 22, 23 and 24, and August 27, 1986 QeportNo. 50-255/0LS-86-01) Examinations were administered to two reactor operator candidates and eight senior reactor operator candidates. Results: Both reactor operator candidates and seven senior reactor operator candic'ates passed.
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8610100320 860930 PDR ADOCK 05000255 V PDR
REPORT DETAILS
- 1. Examiners R. L. Higgins, RIII R. G. Clark, PNL J. D. Smith, PNL J. W. Upton, PNL
- 2. Examination Review Meeting An examination review meeting is no longer conducted. A copy of the examinations and answer keys are left with the facility after the last candidate completes his exam. The facility has five working days in which to make comments to the NRC concerning the examinations and answer keys. The following paragraphs contain the facility comments concerning the reactor operator examination followed by the NRC response.
Question 1.11 Facility Comment: The examination answer key indicates a value of lambda equal to .08. Our student handout assumes a value of .1 for lambda, which would change the answer to the question of rod height equal to 92 inches (+ or - 2 inches). Recommendation: An alternate answer for rod height should be acceptable based on the value of lambda assumed.
Reference:
Palisades student handout SH-PQD0 Reactor Startup, Objective No. 4.24 (PQD00K5.20). NRC Response: Agree. The answer key was expanded to also award full credit if the candidate used a value of lambda equal to .1. 4 Question 2.01 Facility Comment: The examination answer key indicates a minimum of two gross radioactivity monitors for a liquid batch release. The reference cited on the key indicates only one monitor is required, which agrees with Palisades Technical Specifications. Recommendation: Change the answer key to reflect "B. One" as the correct answer for the minimum number of gross radioactivity monitors for a liquid batch release.
Reference:
Palisades Technical Specification Table No. 3.24-1, Student handout SH-PNPD Radwaste, Objective No. 4.7 (PNPD0G5.03). NRC Response: Agree. The answer key was change to "B." 2 4
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Question 2.02a Facility Comment: The examination answer key indicates that the diesel generator K-6A jacket water cooler is supplied by critical header A. The reference cited indicates critical header B is the cooling water supply for K-6A. Recommendation: Change the examination answer key to critical header B.
References:
Palisades P&ID M-208 SH.1A, Student handout SH-PNAA Service Water System, Objective No. 4.4 (PNAA0K1.01). NRC Response: Agree. The answer key was changed to " critical header B." Question 2.04c Facility Comment: The examination answer key indicates the motive force for filling a SIT is a HPSI pump. Depending upon the mode of operation assumed in answering the question, the containment spray pumps may also be used. Recommendation: Accept containment spray pumps as an alternate answer.
Reference:
S0P 4 Containment Spray and Iodine Removal System Procedure Section 7.1.1. NRC Response: Agree. The answer key was expanded to also grant full credit for the response " containment spray pumps." Question 2.07 Facility Comment: The examination answer key describes two flow paths by which decay heat is removed from the core to the " ultimate heat sink" immediately following a post LOCA Recirculation Actuation Signal. The answer key is not in accordance with the reference cited. Recommendation: Revise answer key to reflect the reference cited.
Reference:
Student Handouts: SH-PNMB Containment Spray System, Objective No. 4.14 (PNMB0G4.05), and SH-PNMC Containment Air Cooling System, Objective No. 4.4 (PNMC0K1.01). NRC Response: Agree. The following changes to the answer key were made: (1) Accepted "LPSI pumps" as a correct alternative to "minirecirc pumps"; (2) Replaced " cooling towers" with
" lake"; (3) Accepted "CCW" as a correct alternative to "CNTM air coolers."
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Question 2.10c Facility Comment: The examination answer key only lists one source of water to routinely fill the spent fuel pool. PMW or UW may also be used to fill the spent fuel pool on a routine bases. Recommendation: Accept PMW via P-90 or UW via P-91 as alternate answers.
Reference:
SFP0-1 Addition of Water to the Spent Fuel Pool. NRC Response: Agree. The answer key was modified to also grant full credit for any of the following responses: " utility water";
" primary makeup water"; or " gravity feed from the SIRWT."
Question 2.11b Facility Comment: The examination answer key lists "none" as the effect of a turbine trip signal on the feedwater regulating valves. This answer is based on the reference cited (PNGAOK3.01), which is a learning objective looking at a very specific interrelationships between systems that had been previously presented. There is an additional reference that was not cited that describes the interrelationship between a turbine trip and the feedwater regulating valves. As described in that reference the valves " remain in their as is position." Recommendation: Change the examination answer key from "None" to "as is."
Reference:
Student handouts: SH-PNGA HP and LP Turbines, Objective No. 4.15 (PNGAOK3.01), SH-PNFB Steam Generator Water Level Control, Objective No. 4.5 (PNFBOK5.03). NRC Response: Agree. The answer was changed to "as is." Question 2.12b,c,d,e,i Facility Comment: The answers listed on the examination answer key are based on the reference cited (EOP-1, Attachment 3, Revision 16) which provides Reset SIS repositioning criteria; however, the question deals with SIS Actuation response. Recommendation: Change the examination answer key to:
- b. trip
- c. all 3 start with standby power available
- d. closed
- e. closed
- i. no effect
References:
Palisades Logic Diagram Engineered Safeguards E-17 SH.3; E0P-1, Attachment 3, Revision 16; XE-17 SH.11. NRC Response: Agree. The answer key was changed to b. " trip"; c. "all 3 start with standby power available"; d. " closed"; e.
" closed"; and 1. "no effect."
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Question 3.01a Facility Comment: The examination answer key states "Both valves shut" which is not correct, only the main feedwater regulating valve on the affected steam generator will shut. Recommendation: Change the examination answer key to "the affected steam generator feedwater regulating valve would shut."
Reference:
Palisades P&ID M-207 SH. 1, Palisades System Information Manual Chapter 6b, Page 26, Section c.2. NRC Response: Agree. The answer key was changed to "the feedwater regulating valve for the affected steam generator would shut." Question 3.02 Facility Comment: The examination answer key lists the synchro-transmitters as providing the input signal to the nixie tubes; actually, the synchro-transmitters provide the input signal to the Position Indication Primary (PIP) which provides the input signal to the nixie tubes. The key lists the CRDM limit switches as the input devices to the matrix board display; however, the Secondary Position Indication (SPI) provides input to the matrix board display as well. Recommendation: Change the answer key to accept, for full credit, either the synchro-transmitters provide the input signal or the Position Indication Primary (PIP) provides the input signal to the nixie tubes. Change the answer key to also accept SPI as an input to the matrix board display. I
Reference:
Palisades System Information Manual Chapter 10b, Pages 10 and 11. NRC Response: Agree. The following changes were made to the answer key: (1) Accept either " PIP" or " synchro-transmitters" as providing the input signal to the nixie tubes; (2) Accept either " SPI" or " limit switches" as providing the input signal to the matrix board display. Question 3.03 Facility Comment: The examination answer key is not all inclusive of the automatic actions associated with loss of the instrument AC bus. Recommendation: Change the examination answer key to also accept che items listed in the references attached. 5
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References:
ONP-24.5 Loss of Instrument AC Bus Y01, Revision 11; ARP-24, Revision 36, annunciator number 30; and P&ID M-653 SH.3. NRC Response: Expanded the answer key to also grant credit for the following responses: (1) SPI is lost; (2) PIP is lost; (3) Rod matrix display is lost; (4) Startup, intermediate and power range recorders are lost; (5) Cooling tower pumps trip; (6) Pressurizer PORV temperature indication is lost; (7) All pressurizer temperature indications are lost; (8) Containment temperature and humidity indications are lost; (9) Letdown and charging temperature and flow indications are lost; (10) Boric acid storage tank levels are lost; (11) AFW discharge pressure indication is lost; and (12) CCW surge tank level indication is lost. Question 3.05d Facility Comment: The examination answer key lists the suction damper (P0-8001) closing as a result of an automatic action from the fuel handling building radiation monitor (RE-5712). The damper only closes automatically on the fan motor being de-energized. Recommendation: Change the examination answer key to eliminate the closing of the damper for full credit as an automatic action of the associated radiation monitor.
Reference:
Palisades P&ID M-658, Revision 12. NRC Response: Agree. The answer key was modified to no longer require
" closes suction damper (PO-8001)" for full credit.
Question 3.07 Facility Comment: The examination answer key does not include all of the Engineering Safety Features that could possibly be actuated on a feedline rupture inside the containment at 95% of full power. Recommendation: Add to the examination answer key as alternate answers to the following: Main feedwater regulating and bypass valves closing on the affected steam generator low pressure of 500 psi. 6 _ _ _ ___ - _ _ _ _ _ _ . _. __ , .._ _ _ . ~ --
Auxiliary feedwater initiation on two out of four low level in either steam generator of 28.7%. Control Room ventilation actuation on Containment High Pressure (CHP). Reactor trip on CHP or Low S/G Pressure.
Reference:
Palisades ARP-21 Annunciators: Page 4, Number 8, Page 5, Number 2., Page 9, Number 8.; Student Handout SH-PNPC Control Room Ventilation System, Objective No. 4.32 (PNPCG12.04). NRC Response: Agree. The answer key was expanded to also grant credit for the following: (1) " main feedwater regulating and bypass valves closing on the affected steam generator if pressure drops below 500 psi;" (2) " auxiliary feedwater initiation due to a level indication of 28.7% or less on two out of four steam generator level indicators;" (3) " control room ventilation actuation on containment high pressure;" and (4) " reactor trip on containment high pressure or low steam generator pressure." Question 3.08a Facility Comment: The examination answer key does not reflect the same answer as cited in the reference. Recommendation: Include in the examination answer key an alternate answer that reflects the cited reference.
References:
Student Handout SH-PNEB Auxiliary Feedwater System, Objective No. 4.10 (PNEBG12.02). NRC Response: Modified the answer key to also accept " Feed Only Good Generator (FOGG) is unreliable." Question 3.09a Facility Comment: The reference cited in the examination answer key had nothing to do with the question that was asked. In order to answer the question, formulas and values for the constants and conversion factors that the I&C Department is given and uses while calibrating the instruments should have been provided to the operators during the test. Furthermore, our calculations do not support a calculated trip set point of 2300 psia which would result in a reactor trip per the answer key for Channel A. We assumed a normal Th reading of 585 degrees F for non affected channels, a Th reading of 600 degrees F for the affected channel, Tc readings 535.5 degrees F and a PCS pressure of 2010 psia. Using the attached I&C Department Monthly Technical Specification Test (MI-2A), we calculated the less than full power value (PY 1) 7
at 2.46 volts and the greater than full power ~value (PY 2) l at 3.012 volts. Using the standard I&C volt to psia ! conversion formula. {[(volt-1)/4]*1000}+1500 we'obtained a value of 1865 psia for a calculated trip set point for the less than full power condition and a value of 2003 psia calculated trip set point-for the greater than full power condition. These calculated trip set points result in a
-trip free condition on Channel A with a PSC pressure of 2010 psia.
Recommendations: Because the examinees were unable to calculate the absolute value for the set point, the examination answer key should be changed to give full credit to answers that state the assumption of a Channel A TM/LP trip or trip free condition and then demonstrate the ability to determine if a reactor trip occurred in accordance with the reference cited.
Reference:
Student Handout SH-PNNB Reactor Protective System, Objective No. 4.41 (PNNBG27.01); Palisades Technical Specification Test Reactor Protective Trip Units Procedure No. MI-2A, Revision 2. NRC_ Response: Agree. Since the exact trip setpoint for the TMLP trip can not be calculated with the information available in the facility literature, it is uncertain whether a trip would occur. The answer key was modified to award full credit if mention was made that the TMLP setpoint was raised. Question 3.09c Facility Comment: Based on the interpretation of the question "PCS flow indication to" the examinee could assume failure of the same signal that provides input to the flow indication as to the trip devices. Recommendation: Include in examination answer key an alternate answer that allows for an assumption of a trip condition on Channel C and therefore concludes with a reactor trip.
References:
Palisades P&ID M-204 SH. 1, Revision 33. NRC Response: Agree. If the candidate mentions that the flow indication to Channel C is a differential pressure signal, the candidate will be awarded full credit for stating that the reactor would trip. Question 3.10 Facility Comment: The reference cited states, "using all available indications." Recommendation: Accept alternate listings of "all available indications," especially with respect to Parts "c" and "d." 8
Reference:
Palisades E0P-1, Attachment 3, Revision 16, Page 1. NRC Response: Partially agree. The answer key will be modified as follows: (1) the response " core exit thermocouples" will also be accepted as a correct response for Part "a;" (2) the statement "on two channels" will not be required in part "d" for the " steam flow" response. Question 4.04 Facility Comment: The objectives in our licensed operator training program do not require trainees to memorize General Operating Procedures. However, we do include objectives similar to the following: Given a simulator scenario with only one start up detector available, using GOP-3, perform an approach to criticality. NRC Response: The candidates were not required to memorize the procedure, but they should be familiar enough with the procedure to know the major steps. Question 4.09a Facility Comment: The examination answer key indicates that the basis for establishing and maintaining PCS temperature at 525 degrees F to 530 degrees F is to maintain subcooling margin before initiating cooldown, which is not reflected in the reference cited. Recommendation: Revise the examination answer key to indicate that the basis for establishing and maintaining PCS temperature at 525 degrees F. to 530 degrees F. is to maintain shutdown margin before initiating cooldown.
Reference:
Palisades E0P8.2, Revision 15, Pages 1-7. NRC Response: Agree. The answer key was changed to " Maintain shutdown margin. . ." Question 4.10 Facility Comment: 4.10a. Based on the guidance given in our DNP-3 Loss of Feedwater, the value for either tripping the plant or reducing power is given as an approximation due to the necessary averaging of eight channels of indicators disagreeing within their expected allowed tolerances. 4.10c. and g. Neither of these states a specific value for the amount of PCS leakage, which would allow assumptions to be made by the examinees. 4.10f. The examination answer key does not reflect the reference cited. 9
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. Recommendation: 4.10a. Revise examination answer key to include as an alternate answer " Trip."
4.10c. and g. Revise examination answer key to include reasonable assumptions that could be made about the amount of PCS leakage, such as, " continue operations" due to PCS leakage within Technical Specification Limits. . 4.10f. Change the examination answer key to include as an alternate answer " reduce power" based on guidance given in the reference cited.
Reference:
4.10a. ONP-3 LOSS OF FEEDWATER, Revision 11. 4.10c. and g. Palisades Technical Specifications, Section 3.1.5. 4.10f. E0P-5 LOSS OF INSTRUMENT AIR, Revision 13. NRC Response: Partially agree. The following modifications were made to the answer key: (1) for part "a," the NRC disagrees with the facility, so the required answer remains " reduce power;" (2) for parts "c" and "g," the answer " continue operations if leakage is within Technical Specification Limits" will also be awarded full credit; and (3) for part "f," the answer " reduce power if the leak rate is small enough to allow time for component isolation and additional supply from the Feedwater Purity Building" will be awarded full credit. The following paragraphs contain the facility comments concerning the senior reactor operator examination followed by the NRC Responses. Question 5.02 Facility Comment: If the candidate uses only the Mollier diagram provided and follows constant enthalpy from interpolated points for 1715 psia /715 psia to 35 psia, he has to again interpolate between hundred degree lines which are one-half inch apart. Therefore, the allowable temperature band for an acceptable answer should be expanded. Recommendation: The acceptance band should be expanded to +/- 15 degrees F. instead of +/- 5 degrees F. The concept of isenthalpic expansion is what is important, not the exact final temperature.
Reference:
Mollier diagram from test, Attachment 5.02. NRC Response: Disagree. The answer key was not changed. 10
i Question 5.08 Facility Comment: The question infers exclusion of DNB as a possible answer. Recommendation: Specific reference to DNB should not be a criterion to award full credit. NRC Response: Disagree. The answer key was not changed. Question 5.13 Facility Comment: There are several factors which change over core life to influence flux which thereby changes control rod worth. The answer key concentrates on only one of those factors. Additional acceptable factors are suggested. Recommendation: Allow the following as an additional acceptable answer:
" Neutron flux increases over the life of the core due to several factors (reduction in boron concentration, depletion of burnable poisons, fuel depletion necessitating an increase in flux to maintain a constant fission rate, etc.).
This increase in flux results in a larger control rod worth since rod worth is proportional to neutron flux (thermal and epithermal)."
Reference:
Palisades student handout SH-PQBO, Attachment 5.13. NRC Response: Agree. The answer key was modified to also grant credit for the following responses: (1) " increase thermal neutron flux at EOL;" and (2) " depletion of burnable poisons and the reduction in boron reduces the competition for neutrons, causing the rods to be exposed to a greater neutron flux." Question 6.01.b Facility Comment: The question does not require a listing of the components supplied by CCW in containment. Recommendation: Allow the following as an additional acceptable answer:
" Prevent a loss of instrument air from causing a loss of CCW to the components in containment (cooled by CCW)."
NRC Response: Disagree. The candidate must state the specific CCW components in containment to receive full credit. Question 6.04 Facility Comment: The description of the trip devices listed as answers 2 and 3 on the answer key may be described differently by the applicants. The following descriptions are provided to the examiner for clarification. 11
Recommendation: Allow the following descriptions of the main feed pump trip devices as additional acceptable answers:
- 1. The manual trip button (electrical) at the main feed water pump (answer 2) is actually located on the local control console.
- 2. The manual trip ' lever (mechanical) at the main feed water pump (answer 3) is actually a button or plunger type trip device.
Reference:
Control Oil System Drawing for Utility Boiler Feed Pump Drive, Attachment 6.04. NRC Response: Agree. The answer key was modified as follows: (1) for answer 2, both phrases "at the main feedwater pump" and "on the local control console" were awarded full credit; and (2) for answer 3, full credit will be given for " button or plunger type trip device" as well as " manual trip lever." The original answers were taken verbatim from facility literature. The facility is admonished to revise its literature to reflect actual plant conditions. Question 6.06 Facility Comment: The backup sources of seal oil vary depending upon the operating mode of the main turbine (operating or on the turning gear). Therefore, additional acceptable sources are recommended. Recommendation: If normal seal oil pressure is decreasing, the seal oil pressure regulator automatically opens to supply backup oil from a variety of sources. These sources are:
- Air Side High Pressure Seal Oil Backup Pump
- Turning Gear Oil Pump
- Emergency Bearing Oil Pump
- Shaft Driven Main Lube Oil Pump Emergency Air Side Seal Oil Backup Pump
*These pumps will supply seal oil via the seal oil pressure regulator. The operating condition of each of these pumps is dependent upon bearing oil pressure and turbine conditions which will determine if each pump is available, not seal oil pressure. Therefore, the order of these four (4) pumps is irrelevant when considering decreasing seal oil pressure. It is recommended that the acceptable answers be evaleated accordingly.
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Reference:
Seal Oil Diagram, Attachment 6.08. NRC Response: Agree. The candidates will be awarded full credit regardless of the order in which their answers appear. Question 6.07 Facility Comment: One of the automatic trips of the diesel generator output breaker is the diesel engine trip. Therefore, any condition which causes an automatic trip of the diesel engine (thereby causing the breaker to trip) should be an additional acceptable answer. Recommendation: Allow the following additional acceptable answers:
-Low lube oil pressure
-Overcrank
-Overspeed
Reference:
Palisades Student Handout SH-PNHB, Attachment 6.07. NRC Response: Agree. The answer key was modified to accept " Low lube oil pressure," "overcrank," and "overspeed" as acceptable responses in lieu of the response " engine trip." Question 6.10.c Facility Comment: The allowable band for the approxima'te value of Primary Coolant Pump vapor seal pressure is too restrictive based upon operational experience. The PCP vapor seal pressure is manually controlled by throttling a manual valve (MV-2194) which controls PCP seal bleed off flow to the Volume Control Tank. Therefore, if VCT pressure is altered the vapor seal pressures will be effected also. The maximum allowable VCT pressure is 50 psig (High pressure alarm). At 50 psig in the VCT the vapor seal pressure of the PCPs will be approximately 100 psia (See control room traces from SH-PNJD, Objective No. 4.25 supporting information). Recommendation: Revise the allowable band for approximate values of PCP vapor seal pressure to 20 to 100 psi.
Reference:
Attachment 6.10.c, excerpts from SH-PHJD and ARP 4. Palisades P&ID M202, Sheet 1, Revision 32. NRC Response: Partially agree. The answer key was modified to accept answers between 20 and 50 psia. The facility is admonished to revise its literature to correctly reflect actual plant conditions. 13 1 l
l Question 6.10.d Facility Comment: The allowable band for the approximate value of primary coolant pump amperage is too restrictive based upon plant experience. Recommendation: Revise the allowable band for approximate values of Primary Coolant Pump amperage to 600 to 650 amps.
Reference:
Control Room Log Sheet No. 1, May 10, 1986, Attachment 6.10.d. NRC Response: Partially agree. The answer key was expanded to "600-630 amps." The facility is admonished to revise its literature to reflect actual plant conditions. Question 6.11 Facility Comment: The noun name of the valve downstream of the blender (CV-2155) is commonly described as the " blender outlet valve" vs. the " boric acid makeup valve." Recommendation: Allow an additional acceptable answer of " blender outlet valve" as a description of CV-2155.
Reference:
Attachment 6.11, Palisades P&ID M-202, Sheet 1A, Revision 3. NRC Response: Agree. The answer key was modified to also grant credit for the response " blender outlet valve" in lieu of " boric acid makeup stop valve." Question 6.12.a Facility Comment: An alternate acceptable answer is to describe only the configuration of the pressurizer pressure " controller" (due to the wording of the question). ("60 to 70 psi lower than the normal pressure setpoint of 2010 psia") is too restric-tive based upon operational experience. Ideally, the pressure controller should be set 75 gsi lower than the desired PZR pressure. This setpoint will allow the PZR spray valves to start to open thereby maintaining the desired setpoint (2010 psia); however, this is an ideal setpoint which assumes the calibration of the pressure controller and spray valves are exact. In reality, the operators adjust the setpoint of the controller to whatever pressure is necessary to maintain the desired pressure. Therefore, the acceptable band for the controller pressure should be expanded to "approximately 50 to 75 psi lower than the normal pressure setpoint of 2010 psia." Recommendations: 1. Allow an alternate acceptable answer to be restricted only to the configuration of the pressurizer pressure
" controller."
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- 2. Expand the allowable range of the pressurizer pressure controller setpoint to "approximately 50 to 75 psi lower than the normal pressure setpoint of 2010 psia."
References:
Attachment 6.12.a, excerpts from: Palisades Student Handout SH-PNKD (Objectives No. 4.14 and 4.13), Palisades SOP 1, Revision 16. NRC Response: Disagree. The answer key was not changed. Question 6.12.b Facility Comment: Due to the configuration of the pressurizer pressure control system, the spray valves will be partially open when pressure is being maintained at the desired pressure (as described in the comment to part "a" of the question). Upon an increasing pressure transient the spray valves will continue to open thereby limiting a pressure increase. According to the original design of the pressure control system these valves would not open until 2085 psia. Therefore, in accordance with our present configuration, during a transient which results in increasing pressure, the spray valves would respond more quickly than if the control system was configured according to design. During normal operation the pressurizer pressure will fluctuate if operated according to design. The controllers for the proportional heaters will not provide the fine control necessary to maintain the desired pressure. The result is a "saw toothed" trace on the pressure recorder. With our present configuration of the control system the spray valves provide a more smooth pressure control during normal operation. Recommendation: Allow an additional acceptable answer of "provides for smoother pressure control."
Reference:
Attachment 6.12.b, excerpts from Palisades Student Handout Sh-PNKD. NRC Response: Agree. The answer key was expanded to also grant credit for the response " smoother pressure control." Question 6.14 Facility Comment: The answer to this question (and the plant training material) is incomplete. While it is true that the water is less dense during the recirculation phase this does not fully explain why the flowrate through the containment spray pumps increase post-RIAS. 15
Prior to RIAS the suction pressure of the spray pumps is j determined by the level, pressure and temperature of the water in the SIRW Tank. Post-RIAS the suction pressure of I the pumps is determined by the level, pressure and l temperature of the water in the containment sump. The net i effect of the change in these parameters is to cause the head of the containment spray pumps to decrease upon shifting of the suction to the containment sump, primarily because the suction pressure has increased dramatically. The net effect is an increase in total spray flow. Recommendation: Allow "because the suction pressure increases post-RIAS" as an acceptable answer.
Reference:
Attachment 6.14, FSAR, Table 6-7. NRC Response: Agree. The answer key was modified to also grant full crecit for the following response: "The pump must add greater head during the injection phase than during the recircu'ition phase. During the injection phase the pump's suction is from the SIRWT and it is pumping against high containment pressure. During the recirculation phase the pump's suction is from the containment sump, so the high containment pressure adds to the pump's suction pressure." Question 6.18 Facility Comment: The answer key only lists valve numbers as acceptable answers. Tht. noun names of these valves are provided for clarification. Recommendation: Allow the following valve names in addition to valve numbers: CV-0738, S/G Surface Blowdown Valve CV-0739, S/G Surface Blowdown Valve CV-0770, S/G Bottom Blowdown Valve l CV-0771, S/G Bottom Blowdown Valve ) CV-0704, Blowdown Tank to Mixing Basin Valve
Reference:
Attachment 6.19, SOP-7, Revision 9. NRC Response: Full credit was awarded if the candidate used the correct valve name instead of the valve number. Question 7.03.b Facility Comment: The answer to this question is in E0P 1, Step No. 4.10, which is a subsequent action. Our objectives in the lesson plan for E0P 1 state that students must be able to state - symptoms, automatic actions, and immediate actions from memory, without the aid of references. In accordance with NOREG 1021, ES-402, students were given the clear expectation 16
I . that they did not have to memorize specific information in subsequent actions unless there was a specific objective over a particular section of the subsequent actions, but only had to " describe generally the objectives and methods." We reviewed each of the subsequent actions of each E0P and ONP and identified several such objectives, but did not determine that memorizing the steps in Section 4.10 of E0P 1 was necessary for successful task performance, since the operator would be required to use the procedure in executing the task. Recommendation: Allow tne following as an acceptable answer:
" Isolate the steam generators and emergency borate."
NRC Response: Disagree. Operators must know what to do during circumstances which require prompt action, whether that prompt action is called "immediate action" or not. The candidate must state how to isolate the steam generator and how much to emergency borate in order to get full credit. Question 7.04 Facility Comment: In accordance with ES-402 the candidate is expected to
" describe generally the objectives and methods" of subsec,uent actions. Therefore, exact setpoints and datails should not be required for full credit.
Recommendation: Accept the following answer as the minimum information required for full credit:
- 1. Unable to maintain adequate PCS subcooling.
- 2. Unable to maintain adequate PZR level.
- 3. No steam generators are available for removing heat.
NRC Response: Disagree. Operators must know what to do during circuinstances which required prompt operator action, whether that action is called an "immediate action" or not. The answer key was not changed. Question 7.05 Facility Comment: In accordance with ES-402 the candidate is expected to
" describe generally the objectives and methods" of subsequent actions. The basis for this question is Step No. 4.9 of E0P 2.1, therefore, exact setpoints and details should not be required for full credit. )
17 l L
Recommendation: Accept the following answer as the minimum information required for full credit: , "The two conditions are: battery discharge amps must be minimized quickly and maintained at the reduced discharge rate." NRC Response: Partially agree. The answer key was modified to grant full credit for the response " battery current is reduced to less than 150 amps within 30 minutes of the onset of the loss of AC Power." Question 7.06 Facility Comment: We do not have an objective addressing memorizing Section 4.1 of E0P-4, but have decided to add an objective requiring students to know these general conditions from memory. Since the basis for the question is a subsequent action (allowing time for reference to procedures) the following answer is suggested. Recommendation: Accept the following answer as the minimum information required for full credit:
- 1. Primary Coolant Pump seal bleed off temperature is alarming (setpoint of alarm is 180*F.).
- 2. Primary Coolant Pump bearing temperature is alarming (setpoint of alarm is 175 F.).
- 3. All or most CRDM seal Bleed-off temperatures are alarming (setpoint of alarm is 200 F.).
- 4. CCW flow is interrupted and cannot be restored (greater than 10 minutes).
NRC Response: Disagree. The operator must know the conditions and actions to take when a prompt operator action is required, whether or not that action is called "immediate action." The answer key was not changed. Question 7.07 Facility Comment: The wording of this question does not exclude the candidate from providing conditions listed in the immediate actions (see Step No. 3.4) which necessitate tripping the Reactor. The following answer is therefore suggested. 18
Recommendation: Allow the following as the minimum required information for awarding full credit:
"Any two (2) of the following three (3):
-air pressure less than 50 psig
-instrument air pressure dropping rapidly
-indication of erratic equipment behavior" NRC Response: Disagree. The operator must know the conditions and actions to take when prompt operator action is required, whether or not that action is called "immediate action."
The answer key was not changed. Question 7.10 Facility Comment: In accordance with ES-402 the candidate is expected to
" describe generally the objectives and methods" of subsequent actions. The basis for this question is subsequent action Step No. 4.24.3.6 of E0P 8.1, therefore, details should not be required for full credit. Our analysis indicated that students do not need to state this information from memory. On-the-job performance would be done with the use of the procedure as an aid.
Recommendation: Allow the following as the minimum information required for full credit:
" Establish HPSI flow to the PCS, through the PORV's (feed and Bleed)"
Reference:
Attachment 7.10, E0P 8.1, Revision 18 NRC Response: Disagree. The operator must know the conditions and actions to take when prompt operator action is required, whether or not that action is called "immediate action." The answer key was not changed. Question 7.11 Facility Comment: The answer key omitted one of the four immediate actions of E0P 9 (Activate Site Emergency Plan) for the " fuel handling accident on the reactor refueling side" event. Recommendation: Allow any three (3) of the following four (4) actions for full credit:
- 1. Dispatch personnel immediately to secure the equipment access door and the personnel air lock doors.
- 2. Notify the Control Room and Evacuate ,
Containment. l i 19
- 3. Evaluate status of containment penetrations and initiate actions necessary to prevent air leakage from containment.
- 4. Activate the Site Emergency Plan.
Reference:
Attachment 7.11, E0P-9, Revision *2. . NRC Response: Agree. The answer key was expanded to also grant credit for the response " activate the Site Emergency Plan." Question 8.03 Facility Co:nment: This question does not match the condition statement in our objective which reads: "Given the Plant Administrative Procedures and an employee's work history for the last 168 hours, determine if the employee would exceed restrictions if assigned overtime." Our objective does not require the student to state this information from memory. Recommendation: For comment only. No recommended action for the purposes of this examination. NRC Response: Disagree. A senior reactor operator must know how much time operators are allowed by administrative procedure to work during a week. Question 8.05 Facility Comment: The answer key does not state that the Shift Engineer must activate the Site Emergency Implementing Procedure during an accident. Recommendation: Allow the following as an additional acceptable answer in lieu of " Function as site emergency director until relieved."
" Implement the Site Emergency Implementing Procedures" (thereby establishing the Shift Engineer as the SED)
Reference:
Attachment 8.05, EI-1. Revision 11 NRC Response: Agree. The answer key was expanded to also grant full credit for the response " implement the site emergency implementing procedures." Question 8.06 Facility Comment: This question does not match the condition statement in our objective. We do require students to know from memory that a refamiliarization requirement exists, but we do not require memorization of the details of that section of the procedures. We expect the supervisor to look up the procedure when making the determination that a refamiliarization shift is required for somebody. 20
Recommendation: For comment only. No recommended action for the purposes of this examination. NRC Response: Disagree. A senior reactor operator must know whether or not the operators under his supervision require refamiliari-zation training in order to satisfy the currency requirements of the plant administrative procedures. Question 8.16.a Facility Comment: Plant Administrative Procedure 4.02 allows a valve / breaker found out of position to be repositioned with the authoriza-tion of the Shift Supervisor. In such a situation, it would be unnecessary for the person performing the system checklist to note the condition on the " Record of Exceptions" sheet. Recommendation: Allow the following additional answer for part "a" of the question:
" Notify the Shift Supervisor and reposition the valve / breaker as authorized."
Reference:
Attachment 8.16.a, Administration Procedure No. 4.02 NRC Response: Agree. The answer key was expanded to also grant full credit for the response " Notify the shift supervisor and reposition the valve / breaker as directed." Question 8.19.b Facility Comment: The plant terminology used for part "b" of the answer may not be familiar to the examiner, therefore, clarification is provided. The floor plugs are identified by the room under the floor plug (i.e., the West Safeguards room floor plug is actually located in the CCW room floor and the Safeguards room ceiling). NRC Response: Noted. Question 8.28.a Facility Comment: The question does not sufficiently limit the scope of possible correct answers. For example, one possible correct answer could be "1250 mrem" in accordance with 10 CFR 20 criteria. Recommendation: Evaluate the candidates answers in accordance with the assumptions stated. NRC Response: Disagree. The question refers directly to a specific limitation stated in the facility's administrative procedure. The answer key was not changed. l 21 l l I
- 3. Exit Meeting On July 24, 1986, the PNL examiners met with representatives of the plant staff to discuss the results of the simulator and oral examinations. The following personnel attended the Exit Meeting:
PNL: R. G. Clark J. D. Smith J. W. Upton Palisades: W. G. Merwin R. B. Heimsath S. Ghidotti R. J. Frigo NRC Senior Resident Inspector: E. R. Swanson The examiners made the following observations concerning the training program:
- a. The candidates had been well trained in the knowledge and use of plant P& ids. They indicated competency in the use of electrical drawings as well as piping diagrams.
- b. The candidates were responsive to the questions of the examiners and demonstrated their intent to convince the examiners that they could operate the power plant in a safe manner.
- c. The overall rating of the simulator teams of candidates for communication would be barely satisfactory. Numerous examples were provided in the meeting.
- d. The candidates did not appear to be well trained in the implementation of the Emergency Plan. This deficiency was noted in both the simulator portion of the examination and in the operating portion. In the simulator examinations the candidates (as the Shift Supervisor) were reluctant to make a classification because that would be the Shift Engineer's responsibility. Upon followup at the plant, the candidates seemed to have difficulty in classifying an event in which the severity of the conditions would determine the classification.
22
P ,
. fs U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION
( FACILITY: _P6 LIQ 8QES _ __ _ _______ REACTOR TYPE: _PWB-GE________________ DATE ADMINISTERED: _@6/9@Z12____,,____ _ _ __ EXAMINER: DUDLEY 2 _N.__ _ _ _ _ _ _ CANDIDATE: _ _ _ _ _ _ ___ INSIByCIlgNS_IQ_CBND198IEi paper for the answers. Write answers on one side only. Use separate sheets. Points for each Staple question sheet on top of the answer The passing question are indicated in parentheses after theand question. a final grade of at Crade requires at least 70% in each category up six (6) hours after least 80%. Examination papers will be picked the examination starts.
% OF
% OF CANDIDATE'S CATEGORY CATEGORY
_y@(UE__ _ _ _ __ _ _ _ COIEGQBy___ _______ __YOLUE_ _IQIOL ___SGQBE___
-_ 1. PRINCIPLES OF NUCLEAR POWER
_2Ez99__ _2Ez99 ___________ _ _ PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ________ 2. PLANT DESIGN INCLUDING SAFETY _2Ez99__ _2Ez99 =_ --____ AND EMERGENCY SYSTEMS
-- - 3. INSTRUMENTS AND CONTROLS
_25z99__ _25z99 ___________ _-
- 4. PROCEDURES - NORMAL, ABNORMAL,
_2Ez99__ _2Ez99 EMERGENCY AND RADIOLOGICAL r CONTROL q ,- Totals 199299__ Final Grade I have neither given All work done on this examination is my own. nor received aid. _ _ _ Candidate's Signature L
f NRC RULES AND GUIDELINES FDR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil gely to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary) .
- 6. Use only the paper provided for answers.
~/ . Print your name in the upper right-hand corner of the first page of gach section of the answer sheet. ~
- 8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a Ogw page, write gnly gn gng sidg of the paper, and write "Last Page" on the last answer sheet.
1.4, 6.3.
- 9. Number each answer as to category and number, for example,
- 10. Skip at least thrge lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in f acility litgtatutg.
- 13. 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 required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE DUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the gxamingt only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in I completing the examination. This must be done after the examination has been completed.
l l l
- 10. When you complete your examint; ion, you shall:
- a. Assemble your examination as follows:
(1) Exam questions on top. (2) Exam aidc - figures, tables, etc. (3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
D k , . _ , __ -.. - - -
12_ _ EBI NQ1 EL E S_ DE_ NgGL E 88_ EQWEB_ EL 9NI_ DEE8GIlg N 2 PAGE : JUEBDQDXNBulGE4_MEBI_IB8NEEEB_eNp_ELylp_ELQW QUESTION 1.01 ( .50) Which one of the f ollowing descriptions best supports the reason why X non reactivity increases sharply after a trip from 1000 hours at 100% power? A. Xenon decays less rapidly due to a reduction in the neutron flu >:. B. Iodine half-lif e is meseh shorter than Xenon half-lif e. C. Iodine production is greatly reduced and Xenon production is greatly increased due to the reduction in neutron flux. D. Due to reduced neutron absorption, Iodine concentration increases, and Xenon decays directly from Iodine, thus Xenon increases. QUESTION 1.02 ( .60) Select the most correct statement from the following.
~
A. I'f two centrifical pumps are in parallel then the combined pump head will be approximately the sum of the individual pump heads. B. If two centrifical pumps are in series then the combined power requirements will be approximately equal to the cube of the individual pump power. C. If two centrifical pumps are in parallel then the combined flow will be approximately equal to the sum of the individual pump iIows. , I D. If two centrifical pumps are in series the flow of each pump will be approximately equal to the square of the individual pump speed. QUESTION 1.03 (2.50) A nearly instantaneous load reduction fron 100% power to 50% power is made. The plant is now at a steady state power level of 50% with Tave at its proper value and no rod motion. Ecw much makeup water must be added during the next six hours to maintain Tave at its proper value with no rod motion? Assume the initial PCS boron concentration at 50% power is 500 pp: cnd the plant burnup is 2.6 GWD/MTU. Portions of the Technical Data Book have been provided. Show all calculations, reference any figures used, and state all assumptions. l l l (***** CATEGDRY 01 CONTINUED ON NEXT PAGE *****) l L ___ _ _ _ _ _
Iz__EBINQlELE5_DE_NUgkEBB_EQWEB_EL9NI_DEEB911QB2 PAGE 3
- IBEBDDDYBedIGE,_HE91_IB9NEEEB_9Np_ELylp_ELQW QUESTION 1.04 (1.50) ..
- c. For an operator taking data for a 1/M plot, how will the value of Keff affect the time elapsed before a stable count _ rate can be obtained after withdrawing rods ?
- b. How will the initial count rate affect the count rate at criticality?
DUESTIDN 1.05 ( 1.00) With all systems in manual and no operator action, what effect (increase, decrease, or no change) will decreasing the circulation water temperature have on the following?
- a. Condenser vacuum l
- b. Condensate Pump NPSH l 1
QUESTION 1.06 (2.00) l 1
- c. List two plant evolutions which could result in a waterhammer. (0.8)
- b. Give two examples of how waterhammer can be minimized. (1.2) l QUESTION 1.07 (1.50 )
Indicate whether each of the f ollowing will INCREASE, DECREASE, or REMAIN l UNCHANGED as the discharge valve of a running, motor operated, centrifugal pump is throttled (valve moved in the shut direction): I
- a. Pump motor amps i f
- b. Pump discharge flow
.c. Pump discharge pressure )
l l l i (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****) ! t -
Iz__EBINCIELE5_QE_MQQLE68_EQWEB_EL9NI_QEEB9IlpN 2 PAGE 4
- 'IVEB599YN951QS,_UE91_IBBNSEEB_6NQ_ELQ1p_ELQW QUESTION 1.08 (2.40)
- o. Why does nucleate boiling heat transfer remove more heat than )
non-boiling heat transfer? ;
- b. Why does film boiling heat transf er remove less heat than non-boiling heat transfer?
QUESTION 1.09 (2.50)
- c. In Figure 5.2 in the Technical Data Book, explain why rod worth at 80 inches increases over core life. (1.0)
- b. Explain qualitatively what effect a dropped rod would have on reactor power and Tave if the plant was initially at 100% power with all rods withdrawn. Assume all systems in manual and no reactor
~
trip occurs. (1.5) DUESTION 1.10 (2.50)
- a. The overall power coefficient is a combination of three coefficients.
List those coef ficients in order from the largest to the smallest contributor to the overall power coefficient. (1.5) b.-What are two reasons for the Total Power Defect increasing with power? (1.0) DUESTION 1.11 (2.00) If the reactor is critical at 10 E -4 % power and rod group 4 is at 80 inches at EOC, how far should group 4 be withdrawn to establish a O.3 DPM startup rate? Show all calculations, reference any figures, and state any assumptions. (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****) w . _ _ . ._. . _.
PAGE 5 lt__EBINclegg5_DE_N99LE98_EDNEB_EL9NI_DEEBBIIQN, INEB5DDYN951R52_NESI_IB9NSEEB_9ND_ELUID_ELDN DUESTION 1.12 (3.00) .. What effect, if any, will each of the following plant upsets have on subcooling margin? Assume all systems are in manual. Justify your answers and consider each upset separately.
- a. A pressurizer PORV begins leaking while at 80% power.
- b. One PCP is secured f or physics testing while operating at 10% power.
- c. Steam Generator level is being controlled 4% above setpoint while l at 30% power.
QUESTION 1.13 (3.00) Determine and explain what effect, if any, each of the following changes in plant parameters will have on shutdown margin. Consider each parameter separately.
- a. During a reactor startup the levels in both steam generators are rapidly increased 5% .
- b. After raising power from 50% to 100% conditions are stabilized and no further operator actions are taken for six hours.
- c. Rods are borated to all out position from some intermediate position.
I i i (***** END OF CATEGORY 01 *****) t
'2i__EL6NI_p5 Sign _INCLUplNg_g8EEIy_8ND_EDEBQgNCy_gySIEdg PAGE 6 QUESTION 2.01 ( .60)
What is the minimum number of gross radioactivity monitors required for a liquid batch release through a three inch discharge line? Choose the most correct answer. A. None B. One C. Two D. Three OUESTION 2.02 (2.00) Indicate whether the f ollowing service water loads are supplied by the non-critical header, critical header A, critical header B, or critical header A and B. a." Diesel generator K-6A Jacket water cooler
~
- b. Component cooling water heat exchanger E-54B
- c. Engineered safeguards pumps seal cooling
- d. Auxiliary building air conditioning condensers
- e. Containment air coolers QUESTION 2.03 (1.50)
- a. What design feature of the reactor vessel prevents complete draining of the vessel?
- b. What is the design purpose of having the reactor vessel supported by steel pads resting on sliding plates?
l DUESTION 2.04 (2.10)
- a. What would be the probable cause of the pressure instrument (PT 0338) downstream of the 12 inch check valve to the Safety Injection Tank (SIT) reading 2000 psig? See attached figure.
- b. How would the pressure be reduced?
- c. What is the source and motive force used to fill an SIT?
(***** CATEGORY O2 CONTINUED ON NEXT PAGE *****) L __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .__
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2t__P(@NI_DE@lGN_lNCLUQ1NG_@@EEIy_6ND_EdEBGENCY_SYSIE55 PAGE 7 QUESTION 2.05 ( 2. 510 ) For each of the following valves indicate whether the v41ve will fail open, fail shut, f ail as is, or will not be affected by a loss of instrument air.
- a. Pressurizer spray valve
- b. Condensate sump recirculation valve (CV-0730A)
- c. Main feedwater bypass valve (CV-0734)
- d. Main steam isolation valve
- e. Atmospheric steam dump QUESTION 2.06 ( 2.00 )
A double ended load center is a 480 volt bus which can be supplied by two different power sources.
- a. How is paralleling of double ended load centers prevented? ( 0.8 )
- b. What are three reasons for not paralleling double ended load centers? (1.2)
QUESTION 2.07 (2.00) Describe the two paths by which decay heat would be removed immediately following a post LOCA Recirculation Actuation Signal. Start at the fuel cladding and continue to the ultimate heat sink. Include major components or systems. QUESTION 2.08 (2.50)
- c. What two design features would prevent control rod damage during a rod drop? ( 1. :2)
- b. Would any control rods drop if power was lost to MCC1? Assume the selector switch in the spreading room was selected to MCC1. Justify your answer. (1.3)
(***** CATEGORY O2 CONTINUED ON NEXT PAGE *****) u . - - .
PAGE E
,2t__ELON1_QE@l@N_lNQ(UDINQ_S@ EELY _@ND_EMEBGENQY_EYSIEd@
QUESTION 2.09 (2.40) I 1 What is the reason for the limitation allowing operation with one Primary Coolant Puinp (PCP) seal f ailed but requiring immediate shutdown if more than one PCP seal fails? See item (i) on the precautions and limitations provided. QUESTION 2.10 (2.40) State the components which are routinely used to:
- a. Maintain the clarity and chemistry of the spent fuel pool.
- b. Reduce the temperature of the spent fuel pool.
- c. Raise level in the spent fuel pool.
QUESTION 2.11 (2.00) What effect, if any, does a turbine trip signal (45 psig auto oil pressure) have on:
- a. Reactor Protection System
- b. Feedwater regulating valves
- c. Cooling tower fans
- d. Emergency diesel generator QUESTION 2.12 (3.00)
What response, if any, should be observed for each of the following components af ter a Saf ety Injection Actuation Signal?
- a. Service water outlet valve (VHX-4)
- b. Containment air coolers (V-1B,2B,3B, and 43)
- c. Component Cooling Water (CCW) pumps (P-524, B, ind C)
- d. CCW to spent fuel pool and evaporators (CV-0944H)
- e. VCT outlet valve (MD-2OB7)
- f. Gravity feed valves (MO-2169 and 2170)
- g. CCW to containment (CV-0910)
- h. 1E bus feeder breaker (152-303)
- 1. Emergency diesel generators .
- j. Instrument air compressors
(***** END OF CATEGORY O2 *****)
PALISADES NUCLEAR PLANT b e d on 4.,g Ne SOP 1 SYSTEM OPERATING PROCEDURE Rsvisien 16 Page 3 of 44 TITLE: PRIMARY COOLANT SYSTEM
- g. During heat-up or cooldown of the primary coolant system, at least one primary coolant pump must be operation if shutdown cooling is not operating.
- h. Primary coolant pumps shall not be started if rolling in the reverse direction.
- i. Failure of more than one of the three main pressure seals on a Primary Coolant Pump requires immediate shutdown and cooldown in accordance with GOP 8 and GOP 9. Operation with one main seal and low pressure seal failed is possible if additional leakage is within limitation of the radwaste system. Maximum expected leakage with all main seals failed should not exceed approximately 30 GPM.
(See ARP 5, " Primary Coolant Pump Seal Pressure Off Normal," for failed seal criteria.)
- j. Primary coolant pump seals shall be vented prior to starting of the pump following any operation which could introduce air into the l system including " Steam Generator Air Sweeps."
- k. Primary coolant pump seals shall be provided with a source of cool, clean, borated water (flush flow) during any filling of the primary coolant system f rom a level below the upper seal or af ter reactor head replacement, during venting of the seals, during primary coolant pump run for " Steam Generator Air Sweeps" and during seal run-in operation of the primary coolant pump (s).
- 1. Primary coolant pump seals shall be run in for a period of two hours subsequent to any repair or replacement of the seals, or if the primary coolant level has been below the elevation of the upper seal for a week or more.
- m. If pressurizer sprays are operated when differential temperature between spray water and pressurizer water is greater than 200*F, -
log time, differential temperature and pressurizer pressure in Reactor Logbook. A cumulative total of spray cycles with greater than 200*F differential temperature will be kept inside the cover of each Reactor Logbook. The abnormal operating limit is 350'F, J but can be exceeded in an emergency situation.
- n. A reactor coolant pump shall not be started with one or more of the l PCS cold leg temperatures less than or equal to 250*F unless; 1) i the pressurizer water volume is less than 700 cubic feet; or 2) the secondary water temperature of each steam generator is less than i 70*F above each of the PCS cold leg temperatures. (Reference c) l ;
I
- o. No more than two (2) primary coolant pumps shall be operated when the primary coolant temperature is less than 250*F.
- p. No more than three (3) primary coolant pumps shall be operated when primary coolant temperature is less than 400 F.
l l pa0679-0284a-89 e ;
PAGE 9
, Its_JNSIBUDENIS_AND_C9 NIB 9bE QUESTION 3.01 (2.40)
What effect, if any, would a failure high of steam generator level instrument, LIA-0702A, have on: (Figure PNFB-6 is provided)
- a. Main feedwater regulating valves to steam generators E-50A and B.
- b. Bypass valves to steam generators E-50A and B.
- c. Main feedwater pumps to steam generators E-50A and B.
- d. Overall plant operations if no operator action is taken.
GUESTION 3.02 (2.40) List.four control room indications, not including alarms, which provide control rod position indication, and explain HOW the indication signals are produced. QUESTION 3.03 (2.50) What five automatic actions would occur if the instrument AC bus was deenergized during power operations? QUESTION 3.04 (1.40) List FOUR automatic diesel generator trips and indicate whether each trip is overridden by a Safety Injection Actuation Signal. DUESTION 3.05 (2.40) 1 Explain what automatic actuations, if any, are associated with each of the following radiation monitors. Do not include alarm indications.
- a. Stean generator blowdown monitor (RE-0707) j
- b. Stack monitors (RIA-2318)
- c. Engineered safeguards room west (RE-1811) )
i
- d. Fuel handling building (RE-5712)
I (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****) f
I S E C SYS M OVERRIDES i Low Pressure Deerride s. - - . . ==9 , I
- Tr p 1 L I I l l Hi signai select -
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- 9 Lp 0735 E-6A g 3
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PAGE 10 21__JNSIBydENIg_AND_CDMIBQL5 QUESTION 3.06 (2.40) What signals, including power levels, type of channels, and number of . channels, produce each of the following?
- a. Bypass of the reactor trip on "High Startup Rate" (1.2)
- b. Removal of voltage from the startup channels ,( 0.6)
- c. Bypass of the reactor trip on " Loss of Load" (0.6)
QUESTION 3.07 (2.00) If during reactor plant operations at 95% power a feedline rupture were to occur inside the containment, what are the FIVE Engineering Safety Features (ESFs) that could possibly be actuated and what signals will cause these actuations? Include setpoints and 1,ogic. QUESTION 3.08 (3.00)
- a. Why are the eight motor operated valves associated with the auxiliary f eedwater system normally positioned to the open position rather than the auto position?
- b. What actions, if any, should an operator take if one minute after the receipt of an Auxiliary Feedwater Actuation Signal the auxiliary feedwater pumps P-BA and P-8B have not started and P-8C is running?
Feed flow to SG E-50A is 100 gpm and to SG E-50B is 50 gpm. Pressure in SG E-50A is 930 psia and in SG E-50B is 900 psia. Justify your answer and assume all switches are in automatic. (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
PAGE 11 Ic__IN51BydEN15_@ND_CQNIBQLS DUESTION 3.09 (3.00) For each of the following pairs of simultaneous instrumentation failures indicate whether the reactor should or should not have tripped. JUSTIFY your answer and consider each pair of failures separately. Figure PNNB-7 is supplied.
- a. Th indication to channel A fails to 600 F and RCS pressure indication to channel B fails to 1700 psia.
- b. SG A level indication to channel A fails to 20% and SG B level indication to channel A fails to 20%.
- c. PCS flow indication to channel C fails to 0% and loop 2A differenbial pressure indication to channel D f ails low.
- d. Power supply to channel A drawer fails and the upper detector on nuclear instrument power range safety channel B fails high.
DUESTION 3.10 (3.50) What indications must be checked to verify compliance with each of the following Safety Injection Actuation Signal reset criteria. Include the number of channels that must be checked.
- a. PCS is greater or equal to 25 F subcooled with Primary Coolant Pumps (PCP) running. (0.9) 1
- b. PCS is greater than or equal to 25 F subcooled with PCP's not .
running. (0.8)
- c. Pressurizer l evel is greater than 20% and constant or increasing.
i (0.6)
- d. At least one steam generator is available f or removing heat. (1.2) l I
i l l l (***** END OF CATEGORY 03 *****)
FIGURE PNNB-7 I t { _) l,I ,e.:
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4 t__PBQCEgyBES_;_NgBb963_9pNQBMBL3_gBgBQENGy_9ND PAGE 12
. BSD196991996_G9NIBQL QUESTION 4.01 (2.00) .
- a. What is a licensed operator 's responsiblility concerning the RPS and RPCIC panels when there is fuel in the reactor?
- b. Under what circumstances, if any, can an operator depart from an operating procedure?
- c. What actions should an operator take if , while perf orming a checklist, he finds a valve in the Safety Injection System shut which should be open?
- d. Who is responsible f or maintaining the fuel status board?
QUESTION 4.02 ( 1.50 )
- Answer the f ollowing questions concerning nonmal reactor startups.
- a. What is the MINIMUM temperature for criticality?
e
- b. What is the MAXIMUM startup rate permitted under normal conditions?
- c. What is the MINIMUM number of PCPs required to be running prior to startup utilizing control rods?
QUESTION 4.03 ( 1.00 ) From the Emergency Operating Procedure (EOP) Index provided, list the EDP which should be used to:
- a. Determine power supplies and room locations f or motor operated valves.
- b. Determine the reset criteria for SIAS and SIS equipment.
t (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
h.__P8QQEQUBE5_,_MQBd6(a_6SNQ856(,_[UEB@ENQY_6NQ PAGE 17
'6001DLQQ1G96_GQNIBQL QUESTION 4.04 (2.50) .
Explain how an approach to criticality should be conducted with only one operable startup detector. Limit your explaination to the order cnd levels of positive reactivity additions made by dilution or i rod withdrawal. . State your initial assumptions concerning boron concentration and rod position. Actual values are not required. QUESTION 4.05 (2.50)
- a. What actions, if any, are required if there is indication that ,
FOUR full length control rods have not inserted following a reactor trip? (1.0) I
- b. Why are two Primary Coolant Pumps (PCP) tripped when PCS pressure l drops below 1300 psia and all PCP's are tripped at a lower PCS
- i pressure following initiation of Safety Injection after a reactor trip? (1.5)
QUESTION 4.06 ( 3.0-0 )
- c. Explain why each of the f ollowing procedural steps is an indication of natural circulation.
- 1. Delta T is less than 50 F
- 2. SG 1evel > -84% wide range
- 3. Loop Tc constant or decreasing (1.5)
- b. During depressurization while on Natural Circulation PCS pressure is 650 psia, core exit thermo-couples are at 500 F, Th is 500 F and Tc is 480 F. Has a void formed in the PCS? A page of ONP-21 is provided. ( 0.5)
- c. What are the maximum and minimum PCS pressures allowed in the PCS if PCS temperature is 290 F and f orced circulation has just been reestablished by starting a PCP? SDP 1 figure 1-2a is provided. ( 1.0 )
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
PALISA1st'S NUCLEAR PLAfff EEEEhCY OPERATING PNOCEDURES INDEX - . . 81ENhlAL o REVISI0ff REVIEW PERPORDED DOCl2 eff SPONSOk f PROCEDURE NUPSER --- PROCEDURE TITLE IRDBER n ___ 16 86/03/17 RJFriso EOP 1 REACIOR TRIP as 14 84/08/23 ILIFriso LOSS OF AC POWER 5 EOP 2 1 86/01/24 RJFriso EOP 3 IDSS OF SERVIG WATER 2 83/12/05 RJFrigo EOP 4 LOSS OF CDl900 err CDOLINC l 13 83/12/05 RJFriso EOP 5 IDSS OF INSTRIDeft AIR 16 86/03/17 ILIFriso EOP 6 MAIN STEAM LI E BREAK / MAIN FEEDWATER LINE BREAK INSIDE CDerIAllteft 14 84/03/17 RJFrigo E0P 7 MAIN STEAM LINE BREAK /NAIN FEEDWA1ER LINE hREAK OUTSIDE 000fIAlleeft 18 86/03/17 RJFriso EOP 8.1 LOSS OF CODIANI ACCIDEffT 15 86/03/17 RJFriso EOP 8.2 S1TAH UENERATOR TU8E RUF2URE 12 86/01/24 RJFriso EuP 9 RJEL HANDl.ING ACCIDENT Canceled E0P 10 (DNTRCL ROOM EVACUATION 1 85/04/25 RJFriso ' EOP 10.1 FIRE WilOf THitEATEMS SAFETY RELATED EQUIPteff 15 85/05/28 RJFrigo EOP 10.2 ALTERNATE SAFE SHUTDOWI PROCEDURE l 13 83/12/05 RJFrigo EOP 11 LOSS OF CDef!AllGENT INTEGRITY 12 86/01/24 RJFriso EOP 12 ABNORMAL RELEASE OF RADIOACTIVITY p*Se 1' 1NM:X-19101 5/21/86 ,
PAGE 14 4 2__PBQCgpuBES_ _NDBd662_OpdQBd@L2_EMEBGENCy_9ND
'B9D196DD1996_GDNIBg6 " QUESTION 4.07 (2.50)
During a serious emergency, operators may be called upon to assist in search and rescue or recovery operations in the plant.
- a. In such cases, what dose could an operator receive To bring an injured worker to safety? (0.5) 1)
- 2) To eliminate the further escape of radioactive effluents ? (0.5)
- b. What are the possible effects of receiving radiation exposures of the levels of 50 rem ? Include short and long term effects. (1.0)
- c. Who must authorize this voluntary radiation exposure up to the emergency limits ? (0.5)
QUESTION 4.08 (3.00) What immediate actions should be taken if the Shift Supervisor determines that the plant cannot be maintained in hot standby due to a fire in the cable spreading room? QUESTION 4.09 (3.00) Explain the basis or reasons for the following subsequent actions in the Steam Generator Tube Rupture procedure, EOP-8.2.
- a. Establish and maintain PCS temperature at 525 F to 530 F (Listed in order of preferrence).
- b. Reduce and maintain PCS pressure as low as possible in the operating region.
- c. Maintain PCS pressure approximately 0 - 100 psi higher than the isolated Steam Generator.
- d. If affected Steam Generator level approaches full scale, drain to the Miscellaneous Waste System.
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
PA!,ISADES WUCt.EAP PLA'IT Proc No ONP 21 I OFF-Pf0RMAL PROCEDURE Rsvisicn 12 P:33 7 cf 9 l TITLE: NATURAL CIRCULATION Question 4.06 1 4.8.9- During PCS depressuration, monitor for void formation. If substantial void formations are indicated by one or more of the following, go to Section 4.9 and 4.10. l a. During steady state conditions, charging a known voluene of l water / boric acid into the PCS does not result in a corresponding increase in pressurizer level.
- b. Pressurizer level increases significantly greater than expected while operating auxiliary spray.
- c. If pressurizer level control system is in automatic, an unanticipated letdown flow greater then charging flow.
NOTE: During periods of abnormal pressurizer level behavior, charging and letdown _centrols should be placed in manual since automatic actions may be opposite to those which should be teken.
- d. Qualified core exit thermocouples read higher than saturation
. . conditions or are erratic. l
- e. Increased core AT abcve full power (50*F).
- f. Startupneutrondeteck:orsshowingerraticindication.
- g. Loop Th s increasing or erratic.
- h. Loop Tes erratic.
1
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U 38 "g PALISADES PLANT PkYSSURE - TEMPERATURE LIMITS of r FOR C00LDOWN TO 1.3 X 10" nyt
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7
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,3 o ' .3 i c i,ii i c, i iio sio ,i i , i. i c i .i . ,i,i si i ic. c iiji.icjr.sijii.si.co j,ii, l 0 ,,o oo ,,,, ,o, ,o, ,,u ,,o oo oo ,o, 50 70 90 110 130 150 170 190 210 230 250 270 290 310 330
' Temperature degrees F i __ _ - _ _ _ . _ _ - - _ _ _ _ - . - __ _ _ __ _ _ _ _
dz__EBOGEggggS_ _NQBb8L,_9pNQBbg63_gbEggENgy_8ND PAGE 15 BBD196991Q86_GQNIBQL QUESTION 4.10 (4.00) For each of the f ollowing situations indicate whether the. operator should continue operations, reduce power, shutdown the reactor, or trip the reactor. Consider each situation seperately,
- a. Reactor power is 70% and one of the two condensate pumps' trip.
- b. Reactor power is 20% and condenser vacuum has dropped to 17" of Hg.
- c. Reactor power is 50% and a steam generator tube leak is identified.
- d. Reactor power is 80% and component cooling water has been lost for two minutes. There are no temperature alarms.
- e. Reactor power is 80% and non-critical service water has been lost for two minutes. There are no temperature alarms.
- f. Reactor power is 100% and instrument air header pressure has decreased from 90 to 80 psig over the last 10 minutes and is continuing to decrease.
- g. Reactor power is 75%, total leakage is 3 gpm and a CRDM seal has been confirmed to be leaking.
h.' Reactor power is 100% and a fuel element failure has been identi fi ed by coolant activity as being above Technical Specification limits.
- i. Reactor power is 90% and a single rod drops into the core.
J. Reactor power is 100% and a major fire is reported in the HPSI pump room. (***** END OF CATEGORY 04 *****) (************* END OF EXAMINATION ***************)
~
)
EQl! ATit)NS REACTOR THEORY RADIATION FI.lllDS/TilERHG/ HEAT TRANSFER s.
,. l .
- 8 *I ~
P=Pe = P,10 N=Ne k = AtonVt = A202V2 A = IN Q = A1V1 = A2V2 1 = pb+ 0 10 D or T =
~#
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~
in " out +0 stored
~ ~
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~
- l N*I 8
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--1-k H
Iid -1d2 2 R/hr x time - R
- line source
"(
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" net " O(Ddoppler + Dmod + Dvoid T1' 3 g, + TisRad ** ***
- g c head loss a AP
+DXe + 8Sm
- OPu + P=h + pambient =k Y
E F = pA c
, Boron + prod + , fuel , AP2 pg. = APa hase p xK o pg, ) k = f(quality I Pressure)
Pump laws speed a flow k, = kg + !. k (speed)2a Pressure 6k=k-1 HATil (8 Peed)3a power Q = kAaT = hAaf = UAa7 3g, , 26.06 a y =b 9 ~ *
- PAT
"# Q = mah .
r$V ,, ., , E y Q = cot '~*', p , 3.1 x 1018~ log x = c lug x all = m e. aT P j E = No lE ~lE*~ LEY U" #' aT v log xy = log x + lon y H = U + pV Defect = Coeff x a Parameter as = ^3 T pV = nRT P1VI , E1V1 Tl T2 t CV + C2V2 - C (Vi + V:)
~
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. Tems It per sat $st 1st Est. f.alrapy
- Sat fahr Sat 1eme
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800E 11873 1 1873 32 3 *
.. 3s B C IC?95 '
1974 4 16764- 8 8041 2.17&2 21st?- 7t - O 0160?C 2139 0 2839 0 4 008 19732 30772 29 01:249 0 016039 25M I 26342 0 008) 2 IESI 1 8732 53 6 018 80?2 8 1878 8 ECl22 1 1543 Elu3
., ass
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- 1585 2 ISM 2 20 C' 7 10p2 leu 2 0 0s;lt 2 079E 2 1197
$4 8 0 ?ot?S 00406 let' d 148 4 22 0$3 10633 IOES I 8 08 N 2 0E M FitM 52 8 h mM --
96 8 0 2?lt) 0 016028 13136 18&3 6 24 059 10619 1086 0 0 0876 20%3 21070 ts e ,sa:.4 24*,v' 68 3 0236s3 0 016631 36 8 12S22 '12522 26 060 19603 8066 3 0 0516 2 0491 13008 sta ta l 0 ?M!! 0 016033 17076 17076 82 0 0 274u 004:36 28 % 0 10597 10877 0 05S5 2 0391 f or,sg ga g 3129 2 3129 2 3C 0$9 3058 5 30856
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. 03:6'6 004083 32 0St 10514 10t91 0 0637 2 0:9: " 2 0674
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*f 66 9 989 0 Sti l 38 0 % 10543 10 % 4 0 0670 2 004s 2 0'64 *Rs '
68 8 0336! > C 04046 926 5 926 5 1609 30$$2 10912 0 0708. I MM 2 270s as 70 0 S M'9? 0 C16%0 068 3 068 4 38 052 1t154 e 10921 12 4 O Mua 0040ss $14 3 314 3 0 0745 1993C 2 05s5 at 14 0 04:Mr 00405 40 0a 9 10529 RM30 O C 783 198X 2 0557 32 3 764 1 764 1 42 0s6 50518 10938 MG 0a4470 004363 7174 717 4 0 0E 71 1973E 2 0t?9 ts t 44 083 10 % 7 1044 7 0 0255 19 Ele 2 0 s 72 78 8 0 474U 004007 6738 6739 46 043 30495 10956 35 8 0 0th ISRC 2 041$ Ma se s 05m C C!f t'2 6333 633 3 4:C37 loss a log s si t C SAM 0040'7 $*i b 595 6 0032 1942t 2 0*H Os t ta s 057. C olm SC C33 30873 IM73 00H9 1 9338 2 C3:3 82 t S60 3 %C) S20?9 30461 10582 0100E 1 9742 20044 86 8 0 6 '.t a C Cit 3f 7 227 S S?? S Os t BI I Sa C26 10aS C 30% C 0 1043 1 9151 2 C:53 38 lits!S ' 004093 49E E 4968 M C22 10829 10H 9 C IC75 1906C 2 0135 38 9t t t 691 M C Olt" 4Ll1 of f 1 3: t C'. St C:I 104:7 Il0C # 0 1115 I E97C 20CEi te t C ;:t Mi AC ) as: 1 6C C.4 la e CM~ tCat..: 8:t ? dit 3 ICA E Ilti t 0 11S? I sle: 2003 tt et e C t-: . CCJ;.7 6: C:C IGs: S 1802 S 011t f 1119: I tse: ed I 39: P 25? S 44 Ki 103* 3 1103 3 to t 0 8ila C C ali) 3706 37C 9 0 12.8 117x 199?t SE E M OC3 IC36 2 11042 0 326C 13617 1 9576 se t 1989 09s974 0 016130 SSC 4 350 4 67 999 1037 1 1105 1 187 C 3 0Pf S 0 016137 333 1 331 1 69 995 0 1795 135'30 19t?S tes s tes t 3 0t E S C C 4144 313 1 10359 1105 9 01331 Isaas 13775 88:3 Sill 71 992 103al 1106 3 tat t 11M' C C;64) 796 tf 796 4 73 99 03364 lant 1 3 '25 Its e tes t l E.i* OC448 28;28 1033 6 3107 6 0 1402 13273 1 96'S tes t 2803C 75 98 3C32 5 1101 5 08437 IJ!l8 136M test Ues 12?S! O tti!ES 76537 75539 112 0 l it"". 77 90 10314 11093 01477 13105 I E77 taa g 0 016:73 2$137 251 38 7% 96 103: 2 1110 7 till l a?W 0 016180 2382' 23822 01SC7 52021 33S?8 1128 till I S;13 C Citllt 225 6. ??S tS 81 97 10?t i lill e 01582 1 7938 33480 Ines 63 97 10279 1111 9 ti8 e 1 60:5 0 C4196 2142C 21421 01577 1 78 % S k33 1st s 05 97 IQ68 1312 7 01611 17De 13384 tiSJ ; /2.s M*.*J 1730 I HT1 O C16?08 20325 2C326 0797 103 4 1113 6 122 8 1 791. 0 C16213 92 94 197 95 016d6 1 7693 33339 tNI spa a I M:. 404221 M 96 BC'e n line s 01600 3 7413 .33293 1I23 8323 18324 91 96 1023 3 1115 3 i 175 0 19*M 0 0lt ??9 74 08 174 09 03 96 0 1715 1 7S33 13747 1N I ' 173 0 f lC#e 004238 10722 3116 5 01789 1 7453 IS?C2 IM S l 6546 16547 9596 1021 0 1817 0 0 1733 32374 13157 ta t '
- ute 2 223 O C1C247 15737 157J3 132 8 2 3'4t 97 9E 3019 8 1137 3 018t? 1 7295 13117 C oll?% 149 H I49 66 MM 1914 7 till6 13B 9 1pt 2 s'i7 C C16?tt 34280 34241 101 M OllM 3 7217 13068 3328 tMt 26X' 004pa 135 % 1017 5 1819 5 01064 1 7140 I tc a tas t IM S7 ID3 M 3044 1120 3 uit 2 7834 004264 I?t 09 129 11 0 1913 5 7063 33960 tais 10595 10:5 2 1121 3 0 1951 14104 13137 13La 189 0 2 0P5' 004'93 122 98 I?3 00 10795 1014 0 !!?? 0 I47 0 3001 0 016303 18721 01185 1 6910 18f95 teII *
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cc Teble 1. Setureted Steam: Temperature Table-Conthiued Ats bess 50 etat tolome Intnaipp (ntropy fear 10 Se $al S ai Sat Sal 5at 5st fem: Iam Se an bou e feet basr liewe Iva0 vape- tiers tea 0 vape. Fav I s e, og De hg hg Se sea Sg I g ove tes t 75110 0 014510 SC11 50 22 548 00 900 2 11N 2 01631 ISs40 laill Set e It7 s P as: 0 016S?? 40 172 18 199 160 03 009 0 1139 0 01M7 ISal) 1307L 9:70 e -90s t 0 7C? S C16S34 46 2!2 46 749 IS2 01 M78 1139 8 ' 01644 13M6 130a0 sea s 906 0 0Mt 001697 da 333 m a00 SW 07 90t S l140 5 &2??S 15279 130Ds les B let a $ H 7, 8tl6SH 42621 42 638 IM 03 MS3 listJ , 027% 11213 3.2969 _lesL_ 998 9 *
- 9 MC 0 016572 40 94'l . 40 957 158 04 Gla i 1842 3 02787 13148 ! ?934 194 9
~ ' ' ' *
- W8 9M7 001085 M3M 39 3Sa MC 05 982 8 1842 9 01818 1508? 1 7900 182 4 isa s 30 168 0 016M8 37 t44 37 874 142 05 98; 6 1143 7 02848 15017 17MS Ita l 196 1 10 6CS 0016 Ell 36 k8 3639 IH 06 des 1844 a 02979 14952 1 7431 196 9 990 I 13 CSI C OIM24 as 9 a 34 970 IM OS 9791 1145 2 02910 14Mt i 77s8 198 e 0 016t37 33 622 33639 A68 09 9M9 1146 0 01940 1 4874 I ??64 3et t 3:08 11 S?6 sea s 12 S;2 0 0:6e6a 3: 135 31151 172 11 944 18476 0 30C1 14697 1 7696 Sea t Pee 8 33 Mt 0 0Vll 28 842 288'8 17614 9728 11890 0 3063 14571 1 7632 308 8 3i78 la 6k 0 016?ll 25 18? 24 799 10017 9703 IISC $ 0 3121 1 4447 17Mt 212 8 -
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- 218 9 IS 901 0 016'47 24 474 24 39a 16410 967 8 1152 0 CJ181 J 4323 175CS file -
376 8 17 Itt 0 016775 23131 23148 IH F3 965 2 1153 4 03741 Is2Cl 1 7447 379 8 fu s 18 5 % 0 016805 21 S?9 21 S45 19227 962 6 11W 9 0J300 14381 11383 27a g 273 8 2C C;S 9 0166M 20 0 % 20 073 196 31 960 0 1156 3 03M9 TJM1 1 7370 221 8 237 t 21M7 0 01686a IS FC: 18 713 200 35 4574 IIS7 8 C M17 3J4a2 I ??t0 217 8 . . . . , ,
.- tu s 23 216 0 0.48 % 1749 17473 20440 954 8 1159 2 03476 IJ72$ I;7201 236 9 . ';*;/.y
~ MI3 24 Mt 0 0169?6 16 30s 16 321 20845 952 1 ,160 6 0 3533 13609 1 7'142 348 8% -- Mae 26 t?t 4 0169M 1$ 743 15 260 21250 Set t l167 0 03H1 IJaH !?Ot$ 244 4 - M89 28 7H 0 0169 % 14 2M le ft: 214 66 946 4 lilla 036al IJ379 1 7076 . ..aaa s 257 8 3C t!3 S C17tJ2 13 358 13 376 22067 944 1 :llW 7 CJ706 132M 1 6972 M74 fu e 33 05: 0 0170 % 12 S20 12 538 "22469 Mle 11M I CJ763 IJISa 14917 2tes
-
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- 3et t 67 0t5 3es e 71119 0 017a9 6 09 % 6 1130 2738 907 0 llE 9 0 44?6 118'? 143:3 3e4 8 ass a 75a}} S C1753 5 76SS S 7130 24C 904 C 110 C 044?9 11776 167M 3Bi t 317 e 79 953 0 01757 $4%6 $ 4742 242 t 9C10 1183 1 0 4533 18676 1 6209 312 8 04 641 0 01761 5 1673 $1449 206 3 097 9 IIM I 0 4S86 18576 14162 Sit t 316 0 37t t 09 64? 0 C1766 48M1 4 9133 290 4 Sw p 11 52 0 a640 1 1477 Illil 379 0 874 8 W 876 0 01770 4 Wit 4 45 % 294 6 Stil llM 2 0 4697 113?t 1 607) 37a 8 378 9 10C 2a! 0 01De 4 4030 4 4208 298 7 SIA S til? ? 04745 1828: I 6025 370 0 337 3 IC*. 9" 0 0:779 4 1788 419'E 30? 9 80S 3 1188 2 0 4798 1 1183 I Sil: 327 0 334 8 111 8?; 0 01783 3 9681 3 9859 307 1 842 1 1185 1 048SC !!0tt I S936 SN I 3et s 317 957 0 01787 3 7699 3 7878 311 3 078 8 litC l 049C2 1 9990 15892 Sea t het 174 43: 0 01792 3 5838 3 63:3 315 5 4MS til:0 04954 l esM R S689 asa t F X. .-M 34?St 315 7 0722 lit: 1 8 5006 3 0?99 IS6X 3a8 8 p 3488 317 8 33f l42 13913t C Cl?9 ?
0 C1K1 3 8077 32473 326C3 323 9 064 9 lit? ? 0 5058 1970$ 13763 357 8 3H g 0 01806 30M3 3 1044 328 1 065 5 til3 6 0311C 1 9611 15721 3MI 145 474 3sa s 153 0*! 0 01811 2 9392 2 9573 3323 062 1 Ilk e 0 $161 19517 SS6'8 368 8 3365 858 6 1195 2 0 5212 1 0424 1 9437 ana t 364 0 16c 9c3 0 018:6 2 8307 2 8184 asal 3st a 189 113 00101 2 6691 2 6873 340 0 BMI 3195 9 0 5763 18332 ISMS 3770 177 649 00 06 2 SaSI 2S633 MSO 051 6 1896 7 0 $314 802a0 IS%a 3778 376 e IOL SIF G 01131 2 4279 2 4462 349 3 048 I 3197 4 0336S 18149 15513 37E 8 spel 300 0 195 77% 00llM 2 3170 2 33S3 353 6 tas t It98 0 O S416 1 8057 1973 tes t 205 2w t otta? 2 2120 2 2334 3579 es08 Il9s ? O len 9 99o4 I W3? Esa t 362 2 E372 1899 3 6 5516 0 9676 I S3t? asal ase t 21$ 22C 0 01547 2 1126 2 1311 357 8 27% Sit 0 018S3 2 0188 2OM9 366 5 833 4 1199 9 0SW7 0 ti.e I S3S? 3B7 8 J - .. 0 01854 1 9?91 1M77 370 0 829 7 1200 4 S M17 GM96 I S313 3e6 8 3es t 236 193
""** eas t 2472% 0 0196s 1 9444 l et 3C 3758 075 9 1201 0 G W6? 096C7 I S77a des t ..__~
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E Table 1. Satursted Steam: Temperature Table-Continued Aes P,ess spec.t.c mvme Int 7ia'pr . taltop?
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. ~ 477 474 8 5 574 6?
WSil S<l2 832958 88a950 459 9 Mel 12943 8 6S99 8 79M 84515 ella mee ME IS 8 02000 8 79716 9 Bl?!? 464 S TH 6 .320s I _ e 8648 8 7871 i e518 _mas eti _ . esa t Sc2039 0 7M13 07868 e691 7W 7 1203 8 O utt 87735 l esat ege t S47 81 610 10 802C17 87M47 87%St 473 8 779 7 1203 S 867at 0 7Er lesas est es2 e 0 C?C?6 0 7079a e 72020 478 $ 778 6 lin)! 4 6713 0 761a 1es:7 es2 9 633 Cl G H42 8 7128 1437G 406 8 , , see s &W il S C20M 9 64365 87010C 463J filS 82C2 7 06t90 0 7a43 1 4333 ges t let t 6K Bf SC2043 06w48 0 67897 487 1 714 3 I?c22 tes e 70578 0 c?0S3 06?t38 0 64931 492 7 709 0 1701 7 0 69M 0 7357 142W tes t 497 5 703 7 1701 i O 6*87 0 7271 142St act t les 8 ??l C 0 07M7 0 60630 06Mt? 9 7C 4 0 7185 lant St2a 51:1 75772 0 C?C72 0 68738 06 ?t9 SC?) 493 1 37tT S 0 C2081 0 S$997 0 58079 SC7 I 662 7 11958 8 7045 6 7099 I 4143 5163 lits 7H 76 9 7??? 9 70f ? f atas ave s 81753 9 02091 O SM68 CS$9% S12 0 8470 1199 0 an.e w- o f,. 479 8 gla s Sal os C C21C7 0 51314 0 H916 516 9 681 3 1198 2 til8? Sil?6 3 410s 524 B S21 8 67t h I1973 8 7231 06639 l eC70 $78 8 128 5 87031
- 0 SC?lt3 C2112 047947 C assa)0 05795$ 526 8 Wl6 litt e SU40 . 067S2 l aC32 Sils 537 0 90C M 50C 7C 8 7329 06MS IJ953 184 0
$36 8 931 17 0 C2134 0 a6123 0 a8:57 131 7 E34 11934 w . *-
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,g' tal e IC2t at S CIltt C albat 041:17 638 5 11905 0 7S2S 1 6311 I M37 967 8 106: 55 0 C7182 039379 0416M 152 0 S$2 8 1097 5$ O C21W 0379M 04016C SS7J 632 0 liti2 8 7673 e 6222 1 3757 ~ 3e6 8 SH S 00??O' 03tS07 0 38714 562 4 625 3 1187 7 8 76 ?S S&I37 13757 888 4 teal llH IS M76 615 5 1184 1 0 7674 0 6081 IJ716 INI Mat : 17C lf S CO221 03SCM 037370 0 7775 0 5950 1 367S tu t O C2:!S 0 33781 O n976 5U9 611 5 1168 5 Mit 70' 77
$78 3 60a S 118: 7 0 7775 0 St% I 36M l'21 872 t 1741 ?t e C:Jat 0 324?S 0346M $972 116; l 0 742S 0 5766 IJS12 5768 l281 74 G C#W 0 31162 CJM26 543 7 5748 St99 1875 0 0 7876 0 5673 1.35$! lac e 988 8 !!?6 !? O CM79 0 79937 0 32214 St9 I taa s 0 c:29$ 0 28'M 9 31046 594 6 b82 4 1176 5 0 75D C SSIS 1 350' tes t 13'.77 6C: 1 574 7 1174 0 0 7973 0 Sets 13464 M1i 600 t lef t C 0 t:311 027Me C 295:9 MEI 31DE 08 3: 0$n: 13c: M:s lat? ! C 73:6 0 76a 99 C28M7 601 7 St. t Sit t Il7C 2 C K12 C S2H 1337: te. e Hit 14 v' 6 C C?ut C2Sd:S 01777L 611 4 550 4 1157 7 4 813a 0 5196 1333c met Set t 15a32 00'3M 02a38: 015747' 517 1 0 8187 0 5097 IJ7&s taa s taa g 1575 7 0 02382 CJ3W 0 25757 622 9 S4?J 3165 1 1637 3 0 C24% 07239a 0 24796 628 8 133 6 116? 4 0 8:a3 C ate? IJ233 att Het 574 7 1195 0 8'W C at96 1319 812 0 Gif t 16f( ! 0C7472 0 7144? 023P!S D88 1735 9 0 02444 020$16 CJ2960 HQ8 $156 11 % 4 8 048 0 4794 13141 614 3 018 6 178E 9 G C2468 0 1961S 0170t1 6469 506 3 11537 8 8833 046ft .!JO9? 879 8 579 1 19390 B C:sfi 0 38737 0 21226 653 1 4966 IIal 8 O NSI 04S83 IJoal s24 9 67a t 0 8S14 S a474 1 2984 873 8 it9? a 0 0?$18 0 17880 0203M ES95 48t 7 Ilot t 8?s t 0 8578 04M4 125 # 422 8 632 8 IW7C 0 0?S19 O D064 0 19513 Ett 4764 1842 7 2002 3 0C %6 0 16726 ILII792 6724 e65 7 1134 1 0 8628 04M1 IJE75 634 I 834 3 70'9 9 C C?t95 C ISs27 0 180?) $73 ) 4546 1133 7 e tFRf Calm 32t?! taa l 68( t 0 5'46 Dac!S IJ76: pa8 6aa t 2118 3 8 C?625 C latJa 017M1 tait ea3 3 1179 0 2!M I 0 0?tS 7 O D876 01694 692 9 4311 3124 0 8 88X 03893 12699 eas t Sal t 0 3858 03767 IJE M 032 8 00?t91 0 13174 0 15416 700 0 418 7 1118 7
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PALISADES TECHN2 CAL DATA BOOK 4/2/85 LIST OF EFFECTIVE PAGES
- 1 Figure Title g 6
I, l 1 l I
. 1 i
2.0, Xenon Worth 2.1 Equalibrium Zenon vs. Reactor Power 1 2.2 Xenon Worth vs. Time. Trips fron'Various Powers 1 .. 2.3 Xer.on Worth vs. Time, Step Decreases from 100% Power 1 2.4 Xenon Worth vs. Time, Step Increases to 100% 1 from various Power Levels ~
. .41 0, Power Defect 3.1 Power Defect vs. Power 2
~
3.2 100% Power Defect vs. Burnup 2 3.3 Programmed PCS Temperature vs. Reactor Power 1
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1 PALISADES TECHNICAL DATA BOOK 10/26/84 3 LIST OF EFFECTIVE PAGES Fipure Title Rev . 4.0, Reciprocal Boron Worth 4.1 HZP Reciprocal Boron Worth vs. Burnup 1 5.0, control Rod Worth 5.1 Integral Rod Worth 1 5.2 Group 4 Worth at 80 Inches vs. Burnup 1 6.0, Boron Rundown
. 6.1 Predicted Boron Concentration vs. Burnup 1 7.0, Boration I
7.1 Boric Acid Batching Graph 0 7.2 Boraticn Rate Addition 0 7.3 Boration Volume Addition 0 8.0, Dilution . 8.1 Gallons PMW per ppm Boron 0 i 8.2 Formula Sheet (Pages 1 & 2) 0 8.3 Boron Dilution Rate (120*F) 0 8.4 Baron Dilution Rate (532*F) 0 8.5 Dilution Make-up Volume (120*F) 0 8.6 Dilution Make-up Volume (532*F) 0 l l EP0584-0189A-TC03-DM01 ( l
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1 Figure 8.2, Pcge 1 of 2, Rav 0 F DBMUI. A SHEET
!. Boron Addition A. HOT V Gal. B.A. = 5 77 X 10 i n' (B.A.T.k. PPM-PC(Initiel)
(B.A.T.k. PPM-PC(Final) t B. COLD in (B.A.T.k. PPM-PC (.In i t i al ) V Gal. B.A. = 8.k8 X 10 (B.A.T.k. PPM-PC(Final) C. V Gal. B.A. for Gal. of Water Desired PPM desired PPM .
= to Borate X increase nerease B.A.T.k. PPM II .' Dilution A. HOT in (PC initial)
V Gal. PMW = 5 77 X 10 (PC Final)
- t. .,
B. COG V Gal. PMW = 3.h8 X 10 in (PC Initial)
- (PC Final)
III. Blend Ratio B.A.T.k. PPM - 1 = #of Gal. PMW PC PPM 1 Gal. B.A. Pai!EWED BY
.ows s, e'4 ReSctor Engnser I
- (IY -- ^/dL.
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Figura 8.2, Pgga 2 cf 2, Rav 0 IV. Mixing Water and Concentrated B.A. I A. r-Final Concentration = ((Initial Initial Gal. Gal. + Gal. H O2 X Initial Cone.) Added) e de F = - Initial Gal. F C Desired Cone. V. Mixing 2 Tanks of Different Concentrations. A. Final Cone. = (Cone. Of Tank "A" X"A" Gal. Tk. "A")+(Cone. 0f Tk."B" X Gal. Tk. "I (Gal. Tk. + Gal. Tk. "B")
;u:, c:;c;,,1,,,. Ga. Of Water to Borate X Desired PW Increase Pm Of Conc. B. A. - Pm nnd PPM Increase VI. To Predict Or Determine Evap. Bottoms Concentration Gal. Of Feed Increase in Bottom Cone. Evap. Concentrate
- 7 To Evap. Cone. Of Feed To Evap. Volume VII. Pressure, Tenperature Volume, Of A Gas PV 1y P 22
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