Ensure students have calculators that are permitted for use on the Generic Fundamentals Examination. Operator Generic Fundamentals Reactor Theory – Reactor Operational Physics

Reactor Operational Physics This module introduces the student to actual PWR reactor operations. Among the topics covered are: Estimating critical conditions 1/M plots Identifying criticality Response to steam demand Use of control rods Use of boron Reactor trip response Decay heat INTRO

Terminal Learning Objectives At the completion of this training session, the trainee will demonstrate mastery of this topic by passing a written exam with a grade of ≥ 80 percent on the following TLOs: Explain the use of the estimated critical position calculation and nuclear instrumentation during reactor start-up. Describe the operation of a nuclear reactor during startup. Describe the operation of a nuclear reactor during power range operation. Describe the characteristics of and process used when performing a reactor shutdown. INTRO

Reactor Critical Conditions TLO 1 – Explain the use of the estimated critical position (ECP) calculation and nuclear instrumentation during reactor start-up. 1.1 Describe the reactivity variables involved in an estimated critical position calculation and how they are used to predict criticality. 1.2 Describe the nuclear instrumentation response during a reactor startup to criticality. It is important for the reactor operator to have a clear understanding of the theory and practices used for performing an estimate of critical conditions (ECC) and reactor startup. This chapter covers the following topics related to reactor startups: Determining target values for control rod height and boron concentration when the reactor will be critical on a start-up Nuclear instrumentation response TLO 1

Estimated Critical Conditions Calculations ELO 1.1 – Describe the reactivity variables involved in an estimated critical position calculation and how they are used to predict criticality. ECC compares current subcritical condition to previous known critical condition Plants do not have Keff meters Must determine how shutdown we are to: Determine critical boron concentration for desired rod height Recall rod height MUST meet RIL requirements Startup usually consists of: Withdraw shutdown banks to full out Achieve desired boron concentration Withdraw control banks to criticality in increments Based on procedural requirements Related KA - 192008 K1.07 Calculate ECP using procedures and given plant procedures. (3.5/3.6) Performing a safe reactor startup requires consideration of all core reactivities: Desired rod height and boron concentration Plant parameters to ensure criticality is possible Expectations of when the reactor should "go critical” (Anticipate criticality at any time during a S/U!) 1/M plot during actual startup validates the ECC Called an ECC, estimate of critical conditions, or ECP, estimated critical position Same concept, different names ELO 1.1

Estimated Critical Conditions Calculations Reactivities considered when calculating an ECP include: Critical control rod position Critical boron concentration Time in core life Power defect (if critical data was at power) Fission product poisons, xenon and samarium reactivity Samarium post-trip buildup offset by the buildup of Pu-239 PWRs have a minimum temperature for criticality Usually at NOT/NOP prior to startup Remind students that when a reactor startup is being commenced there is usually some decay heat (from previous plant operation) that is being maintained by some sort of “steam dump” system. This system is maintaining a desired steam pressure (and as such RCS temperature). This temperature is usually the temperature associated with Hot Zero Power (HZP) for that plant. Some plants have an ‘Effective Samarium’ value which includes adjustments for the buildup of Plutonium. Failure to include this “adjustment” could result in SHUTDOWN MARGIN being “non-conservative”. For example: The increase in Sm-149 post trip due to the Neodymium (Nd-149) and Promethium (Pm-149) already in the core may be around -400 to -500 pcm several days after a trip.  However, a good percentage of this is offset by the production of Pu-239 from the Neptunium (Np-239) in the core.  The “Effective Samarium” value might only be -300 pcm. This means you could be off by as much as 200 pcm! This example would show that the plant is shutdown by 200 pcm less than you thought (lower SDM). It also means that when calculating the ECC you will go critical 200 pcm earlier than expected! Check with your specific plant’s ECC procedure to see how much (if any) of the samarium post-trip value is included in the ECC calculation. ELO 1.1

Estimated Critical Conditions Calculations During ECC Calculation Plant curve book and/or computer data used to obtain values Adjust to desired RCS boron concentration “Estimated” critical rod height allows for: Minor errors in ECC Changing reactivities (like Xe-135) During Startup Reactor engineering tracks “estimated to actual” data 1/M predictions must track within established band After Criticality Achieved Validate startup calculations “estimated” to “actual” NOTE: The desired critical rod height usually allows for some sort of PLUS/MINUS reactivity tolerance (some plants use +/- 500 pcm). For instance, by choosing this height, if we go critical sooner we will be above the RIL (-500 pcm value) , if we go critical later we will be below full out position (+500 pcm value). Each plant has different “tolerances” and “procedural actions” if those tolerances are exceeded. ELO 1.1

Estimated Critical Conditions Calculations Knowledge Check Which one of the following is not required to determine the estimated critical boron concentration for a reactor startup to be performed 48 hours following an inadvertent reactor trip? Reactor power level just prior to the trip Steam generator levels just prior to the trip Xenon-135 reactivity in the core just prior to the trip Samarium-149 reactivity in the core just prior to the trip Correct answer is B. Correct answer is B. NRC Bank Question – P367 Analysis: A. WRONG. Reactor power level just prior to the reactor trip will affect the reactivity associated with both power defect and Xenon reactivity. This value is needed to calculate estimated critical Boron concentration. B. CORRECT. Steam generator level has no bearing on the amount of reactivity in the core (as long as it isn’t changing causing RCS temperature to change). C. WRONG. Xenon reactivity in the core just prior to the trip will impact the post-trip Xenon characteristics. This value is needed to calculate estimated critical Boron concentration. D. WRONG. Samarium in the core just prior to the trip is negative reactivity that must be accounted for when calculating estimated critical Boron concentration. ELO 1.1

Reactor Startup Nuclear Instrumentation ELO 1.2 – Describe the nuclear instrumentation response during a reactor startup to criticality. This section describes the response of the nuclear instrumentation during a reactor startup. Review of subcritical multiplication from a previous chapter Important to the reactor operator: Identifying proper versus improper response Correctly identifying criticality Related KAs - 192008 K1.03 Describe count rate and instrument response which should be observed for rod withdrawal during the approach to criticality. (3.9/4.0); 192008 K1.09, Define criticality as related to a reactor startup. (3.2/3.3) ELO 1.2

Reactor Startup Nuclear Instrumentation Source Range NI’s used during startup Addition of positive reactivity Slight positive startup rate (SUR) and count rate will increase When rod motion stops in a subcritical reactor: Count rate increases to new equilibrium level SUR will decay to zero As keff approaches 1.0 Longer time to reach equilibrium counts Greater change in counts for given reactivity addition Greater initial change in neutron population Prompt Jump ELO 1.2

Reactor Startup Nuclear Instrumentation Subcritical Multiplication Review Subcritical Multiplication Factor (M) = 1/1 - keff As reactor approaches criticality, “M” approaches infinity Hard to graph Use inverse for startups 1/M = CRI/CRF Called 1/M plot or ICRR (Inverse Count Rate Ratio) 1/M = 1-Keff (margin to criticality) Relationship between count rate and Keff On NRC Equation Sheet 𝐶𝑅1 1−𝐾𝑒𝑓𝑓1 =𝐶𝑅2 (1−𝐾𝑒𝑓𝑓2) ELO 1.2

Keff vs Count Rate Review Question Knowledge Check – NRC Bank With Keff at 0.95 during a reactor startup, source range indication is stable at 120 cps. After a period of control rod withdrawal, source range indication stabilizes at 600 cps. What is the current value of Keff? 0.96 0.97 0.98 0.99 Correct Answer is D. Correct Answer is D. NRC Bank Question – P2766 Analysis: When positive reactivity is added to a subcritical reactor, equilibrium condition occurs when fission neutron losses equal source neutron strength. Note that doubling the count rate by reactivity additions will reduce the margin to criticality by approximately one half. Therefore, with an original count rate of 120 cps, doubling counts to 240 cps requires a reactivity insertion of one-half way to criticality (approximately Keff = 0.975). Doubling counts again to 480 would result in a Keff of approximately 0.9875. Therefore to reach a count rate of 600 cps, Choice “D” is the only possible correct answer. Equation provided on SLIDE (next click) ELO 1.2

Reactor Startup Nuclear Instrumentation Once again the key thing about this graph is: Counts continue to increase after rods movement stopped until a new higher equilibrium count rate is reached. For equal amounts of reactivity added the a) prompt jump is greater, b) the change in counts is greater, and, c) the time to equilibrium counts is greater, as the reactor gets closer and closer to a Keff of 1.0. The SUR response for these types of reactivity additions is the same. It increases initially, then decreases back to 0.0 when new higher equilibrium count rate reached. Consider the term JUMP-SMILE-JUMP: JUMP – the increase in SUR (prompt jump) that occurs when rods first withdrawn SMILE – the increase in SUR as rods continue to be withdrawn (constant rate of withdrawal versus a “step” withdrawal) Jump – the slight decrease in SUR when rod movement stopped (SUR will continue to decrease in a subcritical reactor back to 0.0). If reactor slightly supercritical, SUR still drops slightly but remains “positive”. Also keep in mind that indications of SUR are hard to see during a reactor startup until you get close to a Keff of 1.0. That is because the “net reactivity” in the core is still “NEGATIVE” Figure: Reactor Startup – Time (Rod Pulls) vs. Count Rate Increase ELO 1.2

Reactor Startup Nuclear Instrumentation Difficult to determine between critical and subcritical SUR decreases to zero in both cases with steady counts More apparent when reactor is supercritical Reactor is called “critical” (actually supercritical) with a: Constant positive startup rate Increasing counts Linear increase in the log of reactor power No control rod motion (or positive reactivity addition) ELO 1.2

Reactor Startup Nuclear Instrumentation As neutron level increases, intermediate range nuclear instruments indicate increasing power level Power is raised to a power level in the intermediate range Below the POAH No reactivity effects by temperature changes Above Source Range Minimal reactivity effects of source neutrons Critical data recorded (ECC verified) De-energizing SR covered in next TLO. ELO 1.2

Reactor Startup Nuclear Instrumentation Data typically recorded: Time and date of criticality Power level Rod position RCS temperature Boron concentration As will be discussed later, the rod height (when critical below the POAH) will be at the same rod position regardless of the power level critical data is recorded at! ELO 1.2

Reactor Startup Nuclear Instrumentation Knowledge Check – NRC Bank A nuclear reactor startup is in progress and the reactor is slightly subcritical. Assuming the reactor remains subcritical, a short control rod withdrawal will cause the reactor startup rate indication to increase rapidly in the positive direction, and then... stabilize until the point of adding heat (POAH) is reached; then decrease to zero. rapidly decrease and stabilize at a negative 1/3 dpm. gradually decrease and stabilize at zero. continue increasing until the POAH is reached; then decrease to zero. Correct answer is C. Correct answer is C. NRC Question P265 (order of answers modified) Analysis: WRONG. This will only occur if the reactor was “supercritical”, then power would increase until the POAH (where fuel and moderator temperature increases would add negative reactivity, then SUR would decrease and stabilize at 0.0) WRONG. This is what happens to SUR on a reactor trip (explained in a later slide). CORRECT. As counts continue to increase (at a decreasing rate), SUR will continue to decrease (at a decreasing rate) until counts are stable. Then SUR is 0.0. WRONG. This is impossible unless rod motion is NOT stopped (besides, you will have other problems)! ELO 1.2

Reactor Startup Nuclear Instrumentation Knowledge Check – NRC Bank A reactor startup is in progress with a stable source range count rate and the reactor is near criticality. Which one of the following statements describes count rate characteristics during and after a 5- second control rod withdrawal? (Assume reactor remains subcritical.) There will be no change in count rate until criticality is achieved. The count rate will rapidly increase (prompt jump) to a stable higher value. The count rate will rapidly increase (prompt jump), then gradually increase and stabilize at a higher value. The count rate will rapidly increase (prompt jump), then gradually decrease and stabilize at the original value. Correct answer is C. Correct answer is C. NRC Question P365 ANALYSIS: When positive reactivity is added to a subcritical reactor, initially a prompt jump will occur due to the positive reactivity addition rate (prompt neutrons take very little time to be born). Delayed neutrons will start to build in after approximately 0.1 seconds. After the rod motion is stopped (and the reactor remains subcritical), counts will continue to increase. An equilibrium condition occurs when fission neutron losses (Keff < 1.0) equal source neutron strength, causing startup rate (SUR) to stabilize at 0.0 once the new steady state neutron count rate is reached (the stem indicates the reactor remains subcritical). ELO 1.2

Nuclear Reactor Startup Operations TLO 2 – Describe the operation of a nuclear reactor during startup. 2.1 Explain the use of inverse multiplication (1/M) plots during the approach to criticality and how they predict when criticality will occur. 2.2 Describe why shutdown control rod assemblies are withdrawn prior to starting up the reactor. 2.3 Describe how reactor response changes when criticality is reached and the parameters that must be monitored and controlled as a nuclear reactor approaches criticality. 2.4 Describe reactor response and operator responsibilities when operating a reactor in the intermediate range, as well as above and below the POAH. 2.5 Describe the basic startup sequence for the secondary (steam) plant, including how reactor power is affected as steam flow is increased. TLO 2

Predicting Criticality Using a 1/M Plot ELO 2.1 – Explain the use of inverse multiplication (1/M) plots during the approach to criticality and how they predict when criticality will occur. 1/M plot used to predict criticality Rod withdrawal to criticality Normal method Dilution to criticality Sometimes used on initial startup after refueling Related KAs - 192008 K1.05, Explain characteristics to be observed when the reactor is very close to criticality. (3.8; K1.06 Calculate ECP using a 1/M plot. 2.9 3.1 Calculate ECP using 1/m plot. (2.9/3.1)/3.9) In case of loading fuel, 1/M plot not used to predict criticality, used to ensure that reactor will not go critical as each fuel element loaded. You might want to provide your students with a copy of your plant specific 1/M (ICRR) plot sheet. ELO 2.1

Predicting Criticality Using a 1/M Plot Recall 1/M = CRinitial/CRfinal CRinitial value ALWAYS initial count rate prior to Control Bank withdrawal CRfinal value stable count rate on most current rod withdrawal As keff approaches 1.0, 1/M value approaches zero Summary of 1/M steps on next slide. ELO 2.1

Predicting Criticality Using a 1/M Plot 1. Start with 1/M =1. 0 with controlling bank rods at 0. (1/M is also referred to as ICRR (inverse count rate ratio). Plot the 1.0 data point. 2. Follow plant procedures for incremental control rod withdrawals, stopping at each interval, allowing time to reach count rate equilibrium, then calculating the 1/M value. 3. On the 1/M form, plot the 1/M data point at each incremental rod position. 4. Using the last two data points on the 1/M plot, predict the critical (1/M=0) rod position by drawing a straight line between the last two data points and extending that straight line until it crosses the horizontal axis (noting critical rod height or boron concentration). 5. Repeat steps 2 through 4 to extrapolate successive and more accurate predictions of critical control rod position. After several points have been plotted on a 1/M plot, a conservative estimate of critical rod position can be obtained by extrapolating slope between last two data points (or baseline and last data point) on plot and reading rod position off horizontal axis. ELO 2.1

Predicting Criticality Using a 1/M Plot Last two data points are extrapolated down to 1/M = 0 to determine estimated critical rod height. Based on things Like SR detector location, rod shadowing, IRW of last withdrawal, initial extrapolated points might be below RIL or above FOP (full out position). Reactor Engineering will validate if response is correct and allow subsequent rod withdrawal. Figure: 1/M Plot for a Reactor Startup ELO 2.1

Predicting Criticality Using a 1/M Plot 1/M examples provided for following plant types: Westinghouse Babcock & Wilcox (B&W) Combustion Engineering (CE) Instructor Note: You can skip over the slides associated with plant types not specific to your type, if desired. None of the startup specific 1/M plots are tested by the NRC (except the fuel loading “conservatism” one on Slide 34. ELO 2.1

Predicting Criticality Using a 1/M Plot Baseline 1/M value (1.0) established after fully withdrawing Shutdown banks (Westinghouse) Safety groups (B&W) First two of six CEA groups (CE) Baseline is plotted at intersection of rod position fully inserted on x- axis and 1.0 line on y-axis IMPORTANT CONCEPT: always allow indicated source range count rate to stabilize for data points ELO 2.1

Figure: Typical ICRR Plot for Westinghouse Plant Westinghouse 1/M Plot Figure: Typical ICRR Plot for Westinghouse Plant ELO 2.1

Westinghouse 1/M Plot In this example, control rod bank A initially pulled to 50 steps After allowing count rate to stabilize, count rate is recorded and 1/M is calculated 1/M = 9.0/9.4 = 0.958 1/M then plotted against current rod position (A bank at 50 steps) Extrapolating a line from baseline data point to next data point yields approximate critical rod height Compared to estimated critical conditions (ECC) calculations In this case, small change in keff makes it too early for an accurate prediction Process repeated ELO 2.1

Westinghouse 1/M Plot Figure: Typical ICRR Plot for Westinghouse Plant with Criticality Prediction ELO 2.1

Westinghouse 1/M Plot Useful prediction of critical rod position does not occur until control rods are at B bank at 172 steps and not highly accurate until C bank is at 119 steps Initial “estimates” might be HIGH Bank C at 119 steps predicts critical control rod position early, which is ideal for early anticipation of criticality If 1/M plot does not predict rod position within limits of ECP: Startup must be terminated, cause must be identified Errors more commonly associated with reactivity balance calculation of ECP Reactor engineering staff notified Westinghouse reactors with the current fuel loading strategy tend to project a critical rod height way high from the first two points on the 1/M plot. The current fuel loading strategy places twice burned fuel on the outside of the reactor, which means that the fuel assembly that is in front of the source range detectors is one of these heavily burned assemblies. Furthermore, a D bank rod is one assembly further in, in front of the source range detectors. Until the D bank rods begin to move , the subcritical multiplication is partially suppressed by this heavily burned fuel and the inserted D bank rod. Consequently, the first 1/M projection is way high. ELO 2.1

B&W 1/M Plot Figure: 1/M Plot for B&W Plant Startup ELO 2.1

B&W 1/M Plot Baseline ICRR typically established after first four control rod assembly (CRA) groups withdrawn Referred to as safety groups 1/M plots at B&W plant may be used: While shutdown and diluting boron concentration Plant heatup to 525F Approach to criticality Plant cooldown Plot 1/M against safety rod positions During refueling operations ELO 2.1

Combustion Engineering 1/M Plot Figure: ICRR Plot at a CE Plant ELO 2.1

Combustion Engineering 1/M Plot Baseline typically established after first two of six CEA groups withdrawn 1/M plot normally established for CEA groups 3, 4, 5, and 6 withdrawal Other uses of 1/M plots at CE plants include: While shutdown and diluting boron concentration During refueling operations ELO 2.1

Conservatism During 1/M Plot Most conservative approach to criticality represented by line A Least conservative approach to criticality represented by line C Equal reactivity additions represented by line B Conservatism related to “initial” data points Line A Initial extrapolation shows early criticality More conservative Line C Initial extrapolation shows late criticality Less conservative The “conservatism” tested by the NRC relates to anticipating criticality. Obviously Operator’s are ALWAYS diligent in anticipating criticality, it is merely a justification for NRC bank questions (See P1770, P3665, P6034). For example, if your 1/M for Line A shows an early criticality, you might withdraw “less” rods on the next rod pull (being more conservative). If Line C, you might desire to withdraw “more” rods on the next rod pull (being less conservative). Figure: Conservative and Non-conservative 1/M Plots ELO 2.1

Predicting Criticality Using a 1/M Plot Knowledge Check As criticality is approached during a reactor startup, equal insertions of positive reactivity result in a __________ numerical change in the stable source range count rate and a __________ time to reach each new stable count rate. larger; longer larger; shorter smaller; longer smaller; shorter Correct answer is A. Correct answer is A. NRC Bank Question – P267 Analysis: Part 1: When positive reactivity is added to a subcritical reactor, an equilibrium condition occurs when fission neutron losses equal source neutron strength. Consider this simple analogy (assumes change in reactor power is a product of flux times reactivity). Neutron population initial = 2; reactivity added = 5. 5 x 2 = 10. Increase in neutron population of 8 (10 – 2). Neutron population now = 10; reactivity added = 5. 5 x 10 = 50. Increase in neutron population of 40 (50 – 10). Considering that a power increase is an “exponential” function, not a “product” function, it is even easier to grasp this concept. Also recall the thumbrules previously presented. “If “X” reactivity is added to a subcritical reactor and counts double, it takes “X/2” for counts to double again. Part 2: The time to an equilibrium count rate is longer due to more generations before a stable Keff can be reached. ELO 2.1

Reactivity Control Mechanisms for Reactor Startup ELO 2.2 – Describe why shutdown control rod assemblies are withdrawn prior to starting up the reactor. Normal order of reactor startup process: Withdraw shutdown banks/groups first Dilute to critical boron concentration Announce commencing reactor startup Withdraw control banks/groups in prescribed increments When slightly supercritical, announce reactor critical If problem during startup, or error in ECC Procedures might require trip or insert shutdown banks/groups Related KA - 192008 K1.02, List reactivity control mechanisms which exist for plant conditions during the approach to criticality. (2.8/3.1) ELO 2.2

Reactivity Control Mechanisms for Reactor Startup Many procedural requirements to abort reactor startup Outside of allowable ECC band (+/- pcm allowance) Reactor critical below RIL/PDIL Loss of operator control of plant/reactor SUR > 1.0 dpm, etc. Regardless of reason Negative reactivity available from shutdown banks/groups Ensure reactor can be shutdown ≈ -3000 to -4000 pcm from shutdown banks/groups ECC – Estimated Critical Conditions RIL – Rod Insertion Limits PDIL – Power Dependent Insertion Limits (same concept as RIL) Mention that Technical Specifications are part the plant’s operating license requirements. Should an event occur that could cause an unanticipated criticality accident or the probability of one is increased during a reactor startup, then: Reactivity mechanisms exist to ensure the reactor can be immediately and safely shutdown Shutdown rod banks (other names for CE and B&W) are pulled prior to actually starting 1/M plot Provides a reserve of negative reactivity that can be quickly inserted ELO 2.2

Reactivity Control Mechanisms for Reactor Startup ELO 2.3 – Describe how reactor response changes when criticality is reached and the parameters that must be monitored and controlled as a nuclear reactor approaches criticality. Indications of subcritical and exactly critical similar Counts constant SUR = 0 Related KAs - 192008 K1.01, List parameters which should be monitored and controlled during the approach to criticality. (3.4/3.5); K1.10, Describe reactor power response once criticality is reached. (3.4/3.4) K1.11, Describe how to determine if a reactor is critical. (3.8/3.8) ELO 2.3

When the reactor is supercritical! Criticality So how is a critical reactor clearly identified? ANSWER: When the reactor is supercritical! Normally when the reactor operator identifies the reactor as critical, it will actually be slightly supercritical. ELO 2.3

Reactivity Control Mechanisms for Reactor Startup A supercritical reactor is identified by: Constant positive startup rate Increasing counts Linear increase in the log of reactor power No control rod motion Normally, criticality achieved in source range But could be in the intermediate range Equilibrium neutron counts is not reached as in subcritical multiplication ELO 2.3

Reactivity Control Mechanisms for Reactor Startup Difference between criticality and subcritical multiplication: Note that neutron level is plotted on a log scale, therefore a straight line on this plot results in an exponential rise in neutron flux. Understanding differential rod worth based on time in core life and how it relates to SUR provides the operator better control of the reactor. For example, using estimated values for typical Westinghouse plant (you may replace with your plant specific data, if desired): Assume criticality estimated with Bank “D” at 100 steps Currently, D at 98 steps/inches with SUR = 0 (operator assumes still subcritical). Two steps away from “estimated criticality”. DRW estimated at between +5 to+10 pcm/step(inch) (for this time in core life per plant curve books). We will assume highest DRW and use +10 pcm/step Based on SUR calculations, it takes about 150 to 200 pcm added to a critical reactor to get a 1 DPM SUR (perform two calculations on board using reactivity equation, using .007 and .0054 for BOL and EOL Beta-Bar-Effective) Conservatively then, 0.5 DPM based on above info would allow for rods to be withdrawn about 10 steps (10 steps x +10 pcm/step = +100 pcm), which is about 0.5 SUR. Your calcs may show about 8 steps or so. What this means, is understanding rod worth you could withdraw rods from Bank D at 98 to Bank D at 108 and still be well within SUR limitations However, during this withdrawal, if SUR reaches 1.0 dpm, release rods and see what you have relating to a critical reactor. The whole point of this is not to go crazy during a startup and “push the limits” but to also realize you don’t need to “sneak up on criticality” and pull rods in 2 – 3 step increments. Obviously, follow any procedurally provided directions! This is just a training example of understanding previous topics and how they relate to this concept. Figure: Neutron Count Rate Increases during a Reactor Startup (Log Scale) ELO 2.3

Monitoring the Approach to Criticality During rod withdrawal, SR NI’s increase Exponential on Analog scale Linear on Log scale When rod motion is stopped in a subcritical reactor Counts increase (at decreasing rate) until new higher equilibrium level reached As keff approaches 1.0 longer time to reach equilibrium SUR indication more prevalent NOTE: The “Smile” part of Jump-Smile-Jump really has “teeth” since rod movement is done in steps. The graphs have been “smoothed” for simplicity sake. SUR = 0 SUR Indication During Startup ELO 2.3

SUR Indication During Startup Initial changes to SUR hard to see Still “net” negative reactivity in core Recall SUR phrase Jump – Rod motion started Smile – Continued rod motion Jump – Rod motion stopped When subcritical, phrase is: Jump-Smile-Jump-Decrease to 0 Smile Jump Decrease to 0 Jump Startup rate animated in on mouse clicks NOTE: The “Smile” part of Jump-Smile-Jump really has “teeth” since rod movement is done in steps. The graphs have been “smoothed” for simplicity sake. The initial “jump” is the prompt jump due to the prompt neutrons produced from the fissions The “smile” is a function of the constant rate of rods being withdrawn. The final “jump” (or drop) is a function of removal of the reactivity addition rate (rho-dot) when rod motion stops. SUR = 0 SUR Indication During Startup ELO 2.3

Reactivity Control Mechanisms for Reactor Startup Reactor operator is monitoring: Neutron level and startup rate Source range and intermediate range until SR blocked Control rod position Rod step counters verified by individual rod position indicators Ensure any stuck/dropped rods are identified Boron concentration, chemistry provides samples as needed Moderator temperature minimum temperatures are met steam dumps are maintaining temperature ELO 2.3

Estimated Critical Conditions Calculations Knowledge Check Near the end of core life, critical rod position has been calculated for a nuclear reactor startup 4 hours after a trip from 100 percent power equilibrium conditions. The actual critical rod position will be lower than the estimated critical rod position if... the RCS temperature is maintaining 3 degrees higher than its normal percent power temperature. actual boron concentration is 10 ppm lower than the assumed boron concentration. one control rod remains fully inserted during the approach to criticality. the startup is delayed until 8 hours after the trip. Correct answer is B. Correct answer is B. Not a direct NRC bank question, but several in the bank just like this one. Analysis: For “actual” critical rod position to be lower than “estimated” rod position (negative reactivity), something had to have added positive reactivity. HINT: In some bank questions this simple test-taking technique will help answer difficult questions. Disregard the STEM for a minute and look at each choice. Since the concept being asked is “reactivity”, A adds “negative”, B adds positive, C adds negative, D adds negative. From deduction “C” must be right. WRONG. If RCS temperature was maintained higher, that would have added negative reactivity. CORRECT. If boron concentration is 10 ppm less than estimated, that adds positive reactivity. WRONG. If a rod remained fully inserted (adding less positive reactivity), that would add negative reactivity. Or, it would make the remaining rods have to be farther out (positive reactivity) WRONG. If the time after the trip was delayed from 4 to 8 hours, Xe-135 concentration would be higher (more negative reactivity in the core), requiring rods to be further out. ELO 2.3

Critical Reactor Above and Below POAH ELO 2.4 – Describe reactor response and operator responsibilities when operating a reactor in the intermediate range, as well as above and below the POAH. PWR inherent safety feature is moderator and fuel temperature coefficients If temperature(s) increase Negative reactivity added to the reactor Power decreased to stop temperature increase How do these coefficients respond above and below POAH? Related KAs - 192008 K1.08, List parameters which should be monitored and controlled upon reaching criticality (3.5/3.7); 192008 K1.12, List parameters which should be monitored and controlled during the intermediate phase of startup (from criticality to POAH) (3.5/3.6) 192008 K1.13, Discuss the concept of the point of adding heat (POAH) and its impact on reactor power. (3.4/3.6) 192008 K1.14, Describe the reactor power response prior to reaching the POAH. (3.1/3.1) 192008 K1.15, Explain characteristics to look for when the POAH is reached. (3.4/3.4) 192008 K1.17, Describe reactor power response after reaching the point of adding heat. (3.3/3.4) ELO 2.4

Reactor Operation – Source Range Criticality normally achieved in SR ECC must be validated Power raised to IR to validate Above SR where source neutrons add to neutron population Below POAH where temperature affects neutron population Withdraw rods to establish positive SUR Verify overlap between SR and IR When IR operation verified: De-energize SR, block any possible low power trip NOTE: When the reactor is operating critical between the SR and the POAH, the operator is in charge of SUR (rods). Source range blocked and de-energized to prevent burnout of its detectors from high neutron flux (not all plants may do this) Source range no longer needed as it will be “pegged” upscale Actual plant specific steps may be discussed, if desired. ELO 2.4

Reactor Operation – Intermediate Range Power increase is stopped and stabilized around 10-8 amps, (around 10-2 to 10-3 % power) Rod height slightly below where “critical” called NOTE: Rod height required to stabilize power is the same regardless of what power you stabilize at! Take critical rod height data (validate ECC) NOTE: (provided that reactor power is below the Point of Adding Heat ) - When the reactor was “called” critical, it was actually slightly supercritical. At that point rods are withdrawn to achieve a +0.5 SUR to the IR. When the desired level is reached, rods are inserted to stabilize power. The rod height you will stabilize at will be slightly below where you were at when you called the reactor critical (because you were actually supercritical). Regardless of the power level in the IR you stabilize at, rods are always at the same height. All you did was insert some positive reactivity to increase neutron population for a “longer or shorter” period of time. There are a few NRC bank questions that test this concept. ELO 2.4

Reactor Operation – POAH After critical data, withdrawn rods to increase power to POAH Usually 0.5 DPM SUR Might reduce SUR lower prior to POAH Verify overlap between IR and PR detectors Ensure reactor stabilizes at POAH 10-8 (10-3%) data is typically reviewed by plant’s nuclear engineering group and is used to: Update ECC/ECP correction factors Update reactivity curves Validate core design data Until POAH is reached, reactor fission process is not contributing sufficiently to heat RCS Negates any effects from negative moderator and fuel temperature coefficients When POAH reached, moderator and fuel temperature will increase, causing power to be turned from negative moderator and fuel temperature (Doppler) coefficients The POAH varies between plants some and depends on ambient heat losses but normally is in the range of approximately 5 x 10-6 to 10-5 amps in the intermediate range or approximately 1 percent power in power range. Where power stabilizes (POAH) is a function of the amount of positive reactivity in the core (SUR) and the time in core life (power coefficient) ELO 2.4

Point of Adding Heat (POAH) RCPs are major heat producers to RCS Up to ≈ 1 % reactor power At about 1 % reactor power Heat production from fission increases RCS temperatures Defined as point of adding heat (POAH) In reality you have to stabilize “slightly above” the POAH. That is, as power is increasing towards the POAH, fissions are increasing, but temperature increases are offset by heat losses. When power is increased high enough (1-2% power) fission rate will cause fuel and moderator temperatures to increase. They must increase high enough to add the required amount of negative reactivity to turn power and stabilize it. Therefore, you really stabilize “slightly above the point where heat is added). We just call that the POAH. NRC Bank questions test the concept of “below”, “at”, or “above” the POAH. If the power increase to the POAH was a normal positive SUR from rod withdrawal, the correct answer is EITHER “AT” or “ABOVE” the POAH (since they both really mean the same thing). The only time they require you to differentiate between AT or ABOVE is if the power increase was caused by positive reactivity added from some sort of steam demand (steam leak, for example). Then the correct answer will be ABOVE the POAH. One important concept when raising power to the POAH. Care should be exercised, especially at BOL. FTC and MTC are there smallest at BOL. Therefore power must raise higher to add the negative reactivity to offset the positive that was added by withdrawing rods. Be careful that you don’t increase power at too high of a rate because you could pass into Mode 1 – (>5% pwr). Even though procedures might limit you to 1 DPM, most plants come into the POAH at a SUR much smaller (like 0.25). ELO 2.4

Critical Reactor Above and Below POAH Knowledge Check – NRC Bank For a slightly supercritical reactor operating below the point of adding heat (POAH), what reactivity effects are associated with reaching the POAH? There are no reactivity effects. The increase in fuel temperature will begin to create a positive reactivity effect. A decrease in fuel temperature will begin to create a negative reactivity effect. An increase in fuel temperature will begin to create a negative reactivity effect. Correct answer is D. The correct answer is D: NRC Bank Question – P70 Analysis: A. WRONG. The POAH is defined as the power level where the reactor produces enough heat to cause a measurable temperature rise in the fuel and coolant. Therefore, a rise in both fuel (due to Doppler broadening) and coolant (due to a negative MTC) temperatures inserts negative reactivity, which will reduce SUR. B. WRONG. The POAH is defined as the power level where the reactor produces enough heat to cause a C. WRONG. The POAH is defined as the power level where the reactor produces enough heat to cause a measurable temperature rise in fuel and coolant temperatures. D. CORRECT. Fissions are occurring at a high enough rate to cause fuel temperature (and moderator temperature) to increase. The subsequent increase in temperature adds the negative reactivity to turn power and stabilize it at around 1 - 2 % power. For a given SUR, power will stabilize slightly higher at BOL because FTC and MTC aren’t as negative. ELO 2.4

Critical Reactor Above and Below POAH Knowledge Check – NRC Bank A reactor is critical below the point of adding heat (POAH). The operator adds enough reactivity to attain a startup rate of 0.5 decades per minute. Which one of the following will decrease first when the reactor reaches the POAH? Pressurizer level Reactor coolant temperature Reactor power Startup rate Correct answer is D. Correct answer is D. NRC Question P569 Analysis: A. WRONG. When the POAH is reached, the reactor produces enough heat to cause a measurable temperature rise in the fuel and coolant. Therefore, a rise in coolant temperature results in a corresponding rise in pressurizer level. B. WRONG. When the POAH is reached, the reactor produces enough heat to cause a measurable temperature rise in the reactor coolant temperature. C. WRONG. The increase in temperature adds the negative reactivity to turn power and stabilize it at around 1 – 2 % power. Power will rise slightly until the negative reactivity from fuel and moderator stops the power rise. D. CORRECT. There was “net” positive reactivity in the core when SUR was 0.5 DPM. As fuel and moderator temperatures increase adding negative reactivity, there is less “net” positive reactivity in the core causing SUR to decrease (eventually to 0.0 DPM). ELO 2.4

Steam Plant Startup ELO 2.5 – Describe the basic startup sequence for the secondary (steam) plant, including how reactor power is affected as steam flow is increased. Basic overview provided Sequence as follows: Main steam lines warmed Steam pressures equalized Main steam valves opened Condenser vacuum established Power raised to accommodate turbine Power transferred from steam dumps to turbine Related KA - 192008 K1.21, Explain the relationship between steam flow and reactor power given specific conditions. (3.6/3.8) Between Westinghouse, Combustion Engineering (CE) and Babcox and Wilcox (B &W) this process is fundamentally the same but somewhat different due to plant design differences. Explain the control of TAvg during this startup phase. ELO 2.5

MODE 1 Operation (> 5% Pwr) Once MODE 1 operation approved: Steam dump loading increased by withdrawing rods Causes temperature to increase Raises steam pressure Causes steam dumps to open When turbine generator ready to sync Close generator output breaker Go to “Raise” on turbine speed control transfers load from steam dumps to turbine Continue power ascension using rods/boron Steam dumps may be used in manual to raise reactor power by increasing steam demand Steam dumps open RCS temperature drops Reactor operator withdraws control rods to restore TAvg Reactor power at new higher power level to match increased steam demand NOTE: when synched to the grid the actual turbine speed doesn’t increase steady-state to steady-state. Once the breaker is closed and the Operator goes to RAISE on turbine speed control, the turbine speed initially increases, control valves open, more load is placed on the turbine, then speed decreases back to the speed required to maintain grid frequency. An INCREASE IN SPEED is really an INCREASE IN LOAD. More information will be provided is 192005 – Motors and Generators. ELO 2.5

Steam Plant Startup Knowledge Check – NRC Bank A nuclear power plant has been operating at 80 percent power for several weeks when a partial steam line break occurs that releases 2 percent of rated steam flow. Main turbine load and control rod position remain the same. Assuming no operator or protective actions occur, when the plant stabilizes reactor power will be __________; and average reactor coolant temperature will be __________. higher; higher unchanged; higher higher; lower unchanged; lower Correct answer is C. Correct Answer C. NRC Bank Question – P1370 Analysis: With the reactor stable (critical) at 80%, a partial steamline break results in more steam being withdrawn from the S/G’s, thus cooling off the RCS. A reduction in moderator temperature inserts positive reactivity, initially resulting in a positive SUR. This causes reactor power to rise (approximately 2% power) and stabilize. Therefore, power will increase and Tavg will decrease. Even though when power rises, RCS temperature also rises, temperature will still be less than when this transient started. ELO 2.5

Nuclear Reactor Power Range Operation TLO 3 – Describe the operation of a nuclear reactor during power range operation. 3.1 Describe the monitoring and control of reactor coolant system temperature and power during power range operations. 3.2 Describe the process of raising reactor power to rated core power. 3.3 Describe the effects of control rod motion and boration/dilution on reactor operation in the power range. 3.4 Explain how boron concentration affects core life. Some duplication of previously presented information. TLO 3

Reactor Power Range Operation ELO 3.1 – Describe the monitoring and control of reactor coolant system temperature and power during power range operations. Requires precise balancing steam flow and RCS average temperature (TAvg) Steam flow changes affect TAvg and require continuous attention by operator Flux distribution is core also maintained within limits Related KAs - 192008 K1.18, Describe the monitoring and control of TAvg, TRef, and power during power operation. (3.6/3.5) K1.16, Describe monitoring and control of reactor power and primary temperature during 0 percent to 15 percent (B&W). ELO 3.1

Ramped Tavg Program Increase steam demand Tcold decreases Lower Tcold, lower steam pressure Lower steam pressure, less efficient the secondary cycle Explained in Thermodynamics Ramped Tavg program means Higher Pwr, higher Tavg Higher Thot, relatively constant Tcold Even though Tavg programs vary greatly by reactor types, NRC GFE Bank questions are asked based on the standard Westinghouse design so a little more time is spent explaining that version. As steam flow increased, reactor must produce more heat, therefore fuel temperature must increase Before fuel temperature can increase, TAvg starts to decrease from reactor producing less power than demanded by steam flow Moderator temperature decrease adds positive reactivity, causing reactor power to increase (no operator action) Reactor power increase heats fuel, adding negative reactivity to core because of fuel temp coefficient Without operator action, these two reactivities balance each other, resulting in: New higher power level at a reduced RCS temperature ELO 3.1

Figure: Typical Westinghouse TAvg versus Ramped Tavg Program CE and Westinghouse plants have a programmed TAvg function Based on referenced TAvg (TRef for Westinghouse plants) TRef - a function of power level Turbine 1st Stage Pressure (increases as power increases) Operator’s responsibility to match TAvg to TRef Note that TAvg will rise (or slide) as a function of reactor power (TRef) Reason for sliding TAvg upward (or holding TCold constant) is to enable steam pressure to remain higher at 100 percent turbine load More efficient secondary cycle, also, With a constant TAvg, steam pressure would be too low for existing turbine design and too expensive to build a turbine that would work Figure: Typical Westinghouse TAvg versus Power Program Graph ELO 3.1

Figure: Typical B&W TAvg versus Power Program Graph Ramped Tavg Program B&W uses different method Ramped initially Up to about 20-30% power Held constant after that This slide can be hidden if not applicable to your plant. Between 0 and about 20-30 percent reactor power, value of TAvg ramps up rapidly, then levels out between about 30 and 100 percent reactor power B&W plants have an ICS that automatically maintains TAvg B&W plants utilize a vertical once through steam generator Design allows for approximately 35-50F of superheated steam B&W plants have advantage of higher quality steam entering high-pressure turbine Eliminates need for moisture separators Figure: Typical B&W TAvg versus Power Program Graph ELO 3.1

Axial Flux Distribution Tech Spec – Mode 1 > 50% pwr Westinghouse plants Ratio of upper detectors to lower detectors Top – Bottom AFD is slightly negative CE plants Ratio of lower detectors to upper detectors Lower – Upper DI is slightly positive ΔΦ or ΔI - difference in power production between the upper and lower half of the core as indicated by the delta between the power range upper and lower detectors Power Range detectors are Ion Chambers and produce a current (I) output proportional to neutron Φ AFD must be maintained in specified band during reactor operation to ensure more uniform axial flux distribution across core ⇒ preventing high peak power in either top or bottom of core (Technical Specifications) High peak power results in high fission product concentration in that location Decay heat generated by fission products could overheat fuel during loss of coolant accident Control rod position used to maintain AFD within allowed operating range during reactor operations Under most operating conditions, AFD limitation more restrictive than rod insertion limits Figure: Upper and Lower Power Range Neutron Detector Locations ELO 3.1

Axial Flux Distribution Initial power ascension (using rods) Rods withdrawn to offset power defect Flux shifts upward (less depression from rods) DT across core increases Tends to shift flux downwards (colder water at bottom) Rods have bigger influence on flux distribution Flux follows rods until fully withdrawn Continued power ascension (using boron) Flux starts shifting down in core Core DT increases Peak flux is below core midplane at 100% power ELO 3.1

Raising Reactor Power to Rated Core Power ELO 3.2 – Describe the process of raising reactor power to rated core power. Numerous considerations must be observed during power increase to full rated load Technical specifications place certain operating limits on core Other reactivity inputs besides those coming directly from power increase must also be observed and compensated for by operators Related KA - 192008 K1.19, Describe means by which reactor power will be increased to rated power. (3.5/3.6) ELO 3.2

Raising Reactor Power to Rated Core Power Power raised to 100% with combination of rods/boron Offsets Power Defect due to fuel/moderator temperature rise Rods withdrawn initially to offset power defect Minimizes Peak/Average power (due to flux depression) Increase steam demand by opening turbine throttle valve(s) Lowers Tcold back to reactor Moderates neutrons better increasing fission rate Increased fission rate raises temperatures slightly But not back to where it was W/D Rods or dilute boron to keep Tavg on program Steam flow increases from various steam loads at startup RCS temperatures controlled by operator with rods, boration, or dilution RCS temperature will drop from greater steam flow Operator responds by withdrawal of rods or dilution to restore RCS temperature, in affect raising power Feedback process to control RCS temperature and reactor power: Reactor operator using control rods or adjusting boron concentration Automatic operation of steam dumps controlling steam pressure/TAvg ELO 3.2

Raising Reactor Power to Rated Core Power As load (steam flow) increased: Match TAvg to TRef Explained in previous slide(s) Maintain axial flux distribution within power distribution limits Maintain control rods above Rod Insertion Limits Adequate Shutdown Margin Compensate for xenon build up (negative reactivity) Compensate for power coefficient Reactor power increased to 100 percent load by process of increasing turbine load either manually or automatically using the plant’s installed turbine load control system. ELO 3.2

Raising Reactor Power to Rated Core Power Rod Insertion Limits/Shutdown Margin As power increased, rod height requirement increases Ensures rods overcome cooldown from power defect On a reactor trip Rods insert negative reactivity Power Defect adds positive reactivity Difference is Shutdown Margin (SDM) Review from Chapter 2 SDM decreases from BOL to EOL Power Defect adds more positive reactivity at EOL More than the small increase in IRW at EOL Maintain control rod position above minimum level for ensuring required shutdown margin Shutdown margin required for all reactor operating modes: full power to cold shutdown/refueling When reactor is critical, called available shutdown margin If reactor trips, available shutdown margin from control rods compensates for positive reactivity added from power defect Only happens if control rods sufficiently withdrawn to ensure necessary negative reactivity Rod insertion limits higher as power increased because power (Doppler) defect is larger ELO 3.2

Reactivity Control in the Power Range Compensation for Xe-135 build up (negative reactivity) When power increased, xenon must be compensated for If Xenon-free “Building in” adding negative reactivity If Xenon previously in core “Burning out” initially adding positive reactivity, then, Rods/boron adjusted to compensate for Xe-135 ELO 3.2

Reactivity Control in the Power Range Compensation for Sm-149 Brand new (clean) core “Building in” adding negative reactivity Previous operating history “Burning out” adding positive reactivity Rods/boron adjusted to compensate for Sm-149 ELO 3.2

Compensation for the Power Defect Reactivity from Fuel Temperature Defect, Moderator Temperature Defect, and Void Defect BOL ≈ -1800 pcm FTD: -1300 pcm, MTD: -450 pcm, VD: -50 pcm EOL ≈ -2300 pcm FTD: -1450 pcm, MTD: -800 pcm, VD: -50 pcm Recall from previous topic: Power ascension from HZP – HFP Fuel Temperature Defect largest part of Power Defect Values for Power Defect vary greatly between plants due to variances in design/size. The above values are merely sample values and may not be representative of your plant. Reference the “Various GFE values.pdf” file to review some of these differences between BOL and EOL. ELO 3.2

Other Reactivity Issues Insertion of reactivity simultaneously via two methods normally not allowed i.e. control rods and boron dilution at the same time Reactor response to boron dilution has a lag time for it to mix in large volume of RCS Operators must plan ahead when using boron concentration changes for reactivity control Reactivity effects from control rods is instantaneous and useful if reactivity adjustment needed quickly Reactivity changes must be made deliberately During a normal power increase from off line to 100 percent, significant dilution required ELO 3.2

Raising Reactor Power to Rated Core Power Knowledge Check – NRC Bank How do the following parameters change during a normal ramp of reactor power from 15 percent to 75 percent? Main Turbine First Reactor Coolant System Stage Pressure Boron Concentration Increases Decreases  Decreases Decreases Increases Increases Decreases Increases Correct answer is A. Correct answer is A. NRC Bank P570 Analysis: CORRECT. As power is increased from 15% to 75%, more steam is flowing through the main steam lines (and thus, main turbines). While this does cause a drop in main steam header pressure, main turbine ahead first stage pressure will increase due to higher mass flow rate of steam through the turbine first stage. Due to the increased negative reactivity from the rise in Doppler broadening and the higher moderator temperature at elevated power levels, reactor coolant system Boron concentration must be lowered from 15% to 75% power to insert positive reactivity to raise temperature back on program for the higher power level. The initial rise in power is usually offset by withdrawing control rods first, then by lowering boron concentration. B. WRONG. While the steam generator and main steam header pressure will lower as power rises, main turbine ahead first stage pressure will increase due to higher mass flow rate of steam through the turbine first stage. C. WRONG. Due to the increased negative reactivity from the rise in Doppler broadening and the higher moderator temperature at elevated power levels, reactor coolant system Boron concentration must be lowered from 15% to 75% power to insert positive reactivity. D. WRONG. While the steam generator and main steam header pressure will lower as power rises, main turbine Due to the increased negative reactivity from the rise in Doppler broadening and the higher moderator temperature at elevated power levels, reactor coolant system Boron concentration must be lowered from 15% to 75% power to insert positive reactivity. ELO 3.2

Reactivity Control in the Power Range ELO 3.3 – Describe the effects of control rod motion and boration/dilution on reactor operation in the power range. Control rod movement at power results in changes to core axial flux distribution Minimal radial flux impact because operated in “groups/banks” Axial flux distribution limits used to ensure: Even power production (kW/ft) top to bottom Uniform depletion of fuel Acceptable axial xenon distributions Operation within core peaking factors Assurance of assumptions made in plant safety analysis Related KA - 192008 K1.20 , Explain the effects of control rod motion or boration/dilution reactor power. (3.8/3.9) T.S. requires that these limits be maintained within limits, or proper corrective actions instituted. ELO 3.3

Reactivity Control in the Power Range Axial flux differences (AFD) sensitive to many core-related parameters: Control bank position Core power level Axial burn up Axial xenon distribution and, to a lesser extent, Reactor coolant temperature and boron concentration Quadrant power tilt ratio (QPTR) Peak radial flux/Average radial flux issues caused by: mis-positioned rod, or, dropped control rod Normally AFD is more restrictive ELO 3.3

Axial Flux/Xenon Oscillations Recall [Xe-135] dependent on flux Perturbations to flux affect xenon concentration In high flux portion of core, initially [Xe-135] decreases Eventually iodine decay causes increase of xenon to decrease flux Opposite effect occurs in other portion of core with lower flux Xenon oscillations normally self-dampening Unless improper control rod movement Oscillations more prevalent at EOL Review the concept of Xenon oscillations from 192006 – Fission Product Poisons, if necessary. ELO 3.3

Boron Concentration Changes Most power changes offset by changes to [B] Minimize effect of rod movements on flux Recall [B] changes (reactivity addition) faster at BOL About 10 times as much water at EOL for given dilution If [B] too low at EOL Coastdown method used to keep temperature on program Base loaded means that nuclear units are not used by the system dispatcher to raise load during system peaks or decrease load with low system loads. Operation in this manner minimizes potential of operation outside allowable axial flux band that could result in unacceptable xenon oscillations. Control rod movement for reactivity control, when base loaded, undesirable because it distorts natural axial neutron flux distribution Effort made to maintain controlling rod bank fully withdrawn In general, minimizing control rod movement over core life and handling reactivity changes with boron/dilution serves to minimize axial flux shifts ELO 3.3

Control Rod Movement Control Rod Movement while on turbine control Withdraw rods Fission rate increases Temperature increases Fission rate decreases Temperature high in operating band Insert rods Opposite effect Based on this, rod movement while on turbine control No change in power (no change in steam demand) No change in shutdown margin (SDM) ELO 3.3

Reactivity Control in the Power Range As reactor operated, fuel atoms deplete Adds negative reactivity (Keff < 1.0) RCS temperature cools down Adds positive reactivity (Keff = 1.0) RCS temperature now “low in program band” Control rods normally remain full out Boron concentration lowered to bring temperature band to program Adds positive reactivity (Keff > 1.0) Raises power Raises temperature adding negative reactivity (Keff back to 1.0) Recall DBW increases over core life, but IBW decreases ELO 5.4 Explain how boron concentration affects core life. ELO 3.3

Reactivity Control in the Power Range Knowledge Check A reactor is operating at steady-state 50 percent power near the end of a fuel cycle when the operator withdraws a group of control rods for 5 seconds. (Assume main turbine load remains constant and the reactor does not trip.) In response to the control rod withdrawal, actual reactor power will stabilize __________ the initial power level and reactor coolant temperature will stabilize __________ the initial temperature. at; at at; above above; at above; above Correct answer is B. This is a review question from 192005 – Control Rods Correct Answer is B. NRC Bank Question – P1054 Analysis: In the power range, withdrawing control rods will initially raise reactor power due to more thermal neutrons being absorbed by the fuel compared to control rods (less leakage). Because main turbine load has not been changed, the reactor is now producing more power than the steam plant is removing, resulting in a positive power mismatch. A positive power mismatch results in average coolant temperature rising because more heat is being produced than removed from the RCS. The combination of the initial rise in power (inserting negative reactivity due to the Doppler coefficient) and the negative reactivity inserted due to the rise in average coolant temperature (MTC is negative at EOL) will result in reactor power turning and stabilizing at the initial power level with a higher average coolant Temperature. ELO 3.3

Boron Effects Over Core Life ELO 3.4 – Explain how boron concentration affects core life. Fuel depleted (burned up) over core life Negative reactivity addition Positive reactivity must be added to compensate Control rods normally maintained full out Boron concentration reduced Related KA - 192008 K1.22, Explain how boron affects core life. (2.6?/3.8?) ELO 3.4

Boron Effects Over Core Life At BOL, boron concentration high Offset Kexcess in core Boron concentration changed to keep temperature on program Boron concentration initially reduced (making MTC negative, or less positive) Compensate for buildup of Xe-135 Boron concentration might be increased (for short period early in core life) If burnable poisons deplete faster than fuel Boron concentration then continually decreased to EOL Compensate for fuel depletion and other FPP buildup Related KA - 192008 K1.22, Explain how boron affects core life. (2.6/3.8) Some reactor designs incorporate fixed burnable poisons installed in fuel assemblies to compensate for reactivity associated with excess fuel in new core Fuel designers have gotten more clever over the years, and they have put a small amount of Sm-149 in the new fuel as well If boron concentration high, positive moderator temperature coefficient can result - not good! Plant technical specification limits for positive MTCs are restrictive; core designs consider these limitations Early on in core life as Xe-135 builds in, boron concentration quickly reduces to point where MTC becomes negative (less positive) Recall BOL - IBW the highest, DBW the lowest, rate of reactivity addition from boron dilution the highest, water required to dilute the lowest EOL - IBW the lowest, DBW the highest, rate of reactivity addition from boron dilution the lowest, water required to dilute the highest ELO 3.4

Boron Effects Over Core Life Knowledge Check – NRC Bank A high boron concentration is necessary at the beginning of a fuel cycle to... compensate for excess reactivity in the fuel. produce a negative moderator temperature coefficient. flatten the axial and radial neutron flux distributions. maximize control rod worth until fission product poisons accumulate.  Correct answer is A. Correct answer is A. NRC Question - P1072 ELO 3.4

Reactor Shutdown and Decay Heat TLO 4 – Describe the characteristics of and process used when performing a reactor shutdown. 4.1 Explain how reactor power decays following a reactor trip. 4.2 Explain the effect that control rods have during a normal reactor shutdown. 4.3 Describe the relationship between decay heat generation, power history and time after reactor shutdown. TLO 4

Reactor Shutdown and Decay Heat ELO 4.1 – Explain how reactor power decays following a reactor trip. This section discusses reactor power response on a reactor trip Also mentioned: Prompt drop Decay of delayed neutron precursors Related KA - 192008 K1.23, Explain the shape of a curve of reactor power versus time after a scram. (2.9/3.1) This section provides knowledge for operator to recognize correct response of a reactor trip as indicated on available instrumentation ELO 4.1

Reactor Shutdown and Decay Heat A reactor trip is a rapid, usually unplanned, automatic or manual shutdown of a reactor by a rapid insertion of control rods Usually caused by equipment failure, uncontrolled transient, or reactor protective function Emergency procedures used to mitigate effects Cause of reactor trip may be immediately determined or could be difficult to determine ELO 4.1

Reactor Shutdown Power Decay Prompt neutrons are gone immediately Neutrons remaining are from delayed neutron precursors Shortest lived decay off first Usually within ≈ 3 minutes Longest lived remain longer Longest lived account for negative 80 second period following the trip (- 1/3 DPM) When longest lived are gone, subcritical multiplication maintains neutron count rate Counts/keff vary over next 3 days with xenon-135 Half-life of longest-lived delayed neutron precursors (bromine-87, 56 sec.) results in reactor period of -80 seconds or -1/3 DPM SUR after a reactor trip Subcritical Multiplication and SR NI’s energization should be about 15-20 minutes after an uncomplicated reactor trip. NOTE: Keep in mind that as xenon-135 builds in over the next several hours, counts (and keff) will decrease, then steadily increase until xenon-135 has decayed away 3 days after a trip. Figure: Reactor Trip Power Decay Response ELO 4.1

Reactor Shutdown Procedure ELO 4.2 – Explain the effect that control rods have during a normal reactor shutdown. Similar to a reactor startup, the plant shutdown process will be a carefully scripted procedure and followed precisely This section gives an overview of how rods effect a reactor shutdown Related KA - 192008 K1.25, Explain the necessity for inserting control rods in a predetermined sequence during normal shutdown. (2.9/3.1) ELO 4.2

Reactor Shutdown Procedure Reactor shutdown consists of: Turbine power reduction Maintain TAvg equal to Tref Boration initially, then rod insertion Ensures: Axial flux maintained in its acceptable band Adequate shutdown margin (RIL) The use of boration and/or rod insertion during a reactor shutdown will be a function of maintaining axial flux distribution within limits. Since AFD is applicable above 50% power, normally boration is initially used to keep temperature on program as power is reduced. If AFD shifted too high in the core, rods could be inserted to shift flux downward. ELO 4.2

Reactor Shutdown Procedure When acceptable low power reached Trip turbine Ensure proper steam dump operation Reactor shutdown continued by either: Tripping reactor, or, Manual insertion of control rods In reverse order of their bank overlapping sequence Maintains a relatively constant DRW Minimizes flux distribution problems Turbine/reactor steps vary widely by plant and are only provided as an overview. Only testable concept on this KA is the effect control rods have on flux distribution during a shutdown. NOTE: “MANUAL rod insertion” usually means rods are inserted in GROUPS/BANKS and not individually. MANUAL INDIVIDUAL rod insertion would have impacts on radial flux distribution. ELO 4.2

Reactor Shutdown Procedure Although a reactor is shut down, it must be continuously monitored [Xe-135] initially increases to a peak SR counts decrease, SDM increases [Xe-135] decreases until xenon free (70-80 hours) SR counts increase, SDM decreases Any RCS cooldown will require an increase in [B] Ensure adequate SDM maintained Be sure to review SDM questions in 192002 – Neutron Life Cycle, now that all of the concepts relating to SDM have been presented. ELO 4.2

Reactor Shutdown Procedure Knowledge Check Which one of the following describes the process for inserting control rods during a normal reactor shutdown? Control rods are inserted in reverse order, one bank at a time, to maintain acceptable power distribution. Control rods are inserted in reverse order, one bank at a time, to maintain a rapid shutdown capability from the remainder of the control rods. Control rods are inserted in reverse order, in a bank overlapping sequence, to maintain a relatively constant differential control rod worth. Control rods are inserted in reverse order, in a bank overlapping sequence, to limit the amount of positive reactivity added during a rod ejection accident. Correct answer is C. Correct answer is C. NRC Bank Question – P2971 Control rods are inserted in reverse order in a bank overlapping sequence to maintain a relatively constant differential control rod worth. ELO 4.2

Reactor Decay Heat ELO 4.3 – Describe the relationship between decay heat generation, power history, and time after reactor shutdown. Decay heat is heat the reactor continues to release for a very long time following shutdown Decay of fission fragments in the core This section addresses impacts on amount of decay heat Related KAs - 192008 K1.26, Define decay heat (3.1/3.2) K1.27, Explain the relationship between decay heat generation and: a) power level history, b) power production and c) time since a reactor shutdown. (3.1/3.4) ELO 4.3

Reactor Decay Heat Of the reactor heat produced at power About 93% is instantaneous energy kinetic energy of fission fragments and fission neutrons About 7% is delayed energy Betas and Gammas from decay of fission products Referred to as decay heat If trip occurs at 100% power Prompt drop to 7% of rated thermal power If trip occurs at 50% power Prompt drop to 3.5% of rated thermal power Half as many fission fragments exist in the core Half as much delayed energy ELO 4.3

Reactor Decay Heat Decay heat first hour following reactor shutdown After an hour, still greater than 1 percent of full power Decay Heat Thumb rule (estimated for an average four-loop Westinghouse reactor): Basically after 1 sec, for each subsequent time constant, decay heat drops to about one-half. Time Decay Heat Immediate 7% 1 sec 6% 1 min 3% 1 hr 1.5% 1 day 0.75% 1 week 0.375% Etc. Figure: Decay Heat Production for the First Hour Following a Reactor Trip ELO 4.3

Reactor Decay Heat Knowledge Check – NRC Bank A nuclear power plant has been operating for one hour at 50 percent power following six months operation at steady-state 100 percent power. What percentage of rated thermal power is currently being generated by fission product decay? 1 percent to 2 percent 3 percent to 5 percent 6 percent to 8 percent 9 percent to 11 percent Correct answer is B. Correct answer is B. NRC Question P2972 Analysis: Between 6-7% rated thermal power is contributed from delayed energy at 100% power. Reducing power to 50% results in half as many fissions occurring in the core, thus half the delayed energy production (approximately 3% to 3.5% rated thermal power). ELO 4.3

Reactor Decay Heat Knowledge Check A nuclear power plant has been operating at 100 percent power for six months when a reactor trip occurs. Which one of the following describes the source(s) of core heat generation 30 minutes after the reactor trip? Fission product decay is the only significant source of core heat generation. Delayed neutron-induced fission is the only significant source of core heat generation. Fission product decay and delayed neutron-induced fission are both significant sources and produce approximately equal rates of core heat generation. Fission product decay and delayed neutron-induced fission are both insignificant sources and generate core heat at rates that are less than the rate of ambient heat loss from the core. Correct answer is A. Correct answer is A. NRC Bank Question – P4336 Analysis: A. CORRECT. Thirty minutes after a trip, power is relatively stable in the Source Range. Both short and long lived delayed neutron precursors have decayed to their daughter products and most delayed neutrons have been produced. Therefore, decay heat (fission product decay) is the only mechanism adding heat into the RCS (besides RCP heat). B. WRONG. The decay (not fission) of the delayed neutron precursors adds heat to the RCS. Post-reactor trip, Keff is less than one and power begins to lower. Thirty minutes after the reactor trip, power will be in the source range. In the source range (significantly lower than the POAH), the fission strength is not strong enough to add heat to the RCS. C. WRONG. The decay (not fission) of the delayed neutron precursors adds heat to the RCS. Post-reactor trip, D. WRONG. Thirty minutes after the trip, decay heat will be approximately 2% of rated thermal power (a major source of heat addition to the RCS), significantly higher than the ambient heat loss rate of any commercial PWR. ELO 4.3

SDM Review Question Knowledge Check Reactors A and B are identical except that reactor A is operating near the beginning of a fuel cycle (BOC) and reactor B is operating near the end of a fuel cycle (EOC). Both reactors are operating at 100 percent power with all control rods fully withdrawn. If the total reactivity worth of the control rods is the same for both reactors, which reactor will have the smaller Keff five minutes after a reactor trip, and why? Reactor A, because the power coefficient is less negative near the BOC. Reactor A, because the concentration of U-235 in the fuel rods is higher near the BOC. Reactor B, because the power coefficient is more negative near the EOC. Reactor B, because the concentration of U-235 in the fuel rods is lower near the EOC. Correct answer is A. This is a review of SDM from 192002 – Neutron Life Cycle Correct answer is A NRC Bank Question – P4924 Analysis: Reactor A (BOL) has a smaller power coefficient. Therefore, following a reactor trip, less positive reactivity will be inserted five minutes after the trip due to the power coefficient. If less positive reactivity is added from the Power Defect, and the same amount of negative is added from rods, then the reactor will be shutdown by more (Greater SDM). The greater the plant is shutdown, the lower the Keff. ELO 4.3

SDM Review Question Knowledge Check A reactor is shutdown with the reactor vessel head removed for refueling. The core is covered by 23 feet of refueling water at 105°F with a boron concentration of 2,000 ppm. Which one of the following will decrease Keff? Refueling water temperature decreases by 5°F. A depleted neutron source is removed from the core. A spent fuel assembly is replaced with a new fuel assembly. Refueling water boron concentration decreases by 5 ppm. Correct answer is A. This is a review of SDM from 192002 – Neutron Life Cycle Correct answer is A NRC Bank Question – P5324 Analysis: HINT: In a shutdown reactor, a “decrease in Keff” means negative reactivity was added (reactor MORE shutdown). A. CORRECT. At low temperatures and high boron concentrations (typically refueling conditions) the moderator temperature coefficient (MTC) is positive. This is because the RCS is over-moderated at this point; for an increasing temperature, the thermal utilization factor increases more than the resonance escape probability lowers, therefore, positive reactivity is inserted as temperature increases. Therefore, decreasing reactor water temperature inserts negative reactivity, thus lowering core Keff. B. WRONG. Removing a depleted neutron source in the core may slightly lower the shutdown count rate, but will not reduce the effective multiplication factor within the core. Source neutrons affect the stable count rate in a shutdown reactor, but do NOT change Keff. C. WRONG. Inserting a new fuel assembly into the core with fresh, unburned U-235 in the core will raise Keff; more fuel is available for fission. D. WRONG. Reducing boron concentration removes neutron poison from the core, thereby inserting positive reactivity, increasing core Keff. ELO 4.3

KA to ELO Tie KA # KA Statement RO SRO ELO K1.01 List parameters which should be monitored and controlled during the approach to criticality. 3.4 3.5 2.3 K1.02 List reactivity control mechanisms which exist for plant conditions during the approach to criticality. 2.8 3.1 2.2 K1.03 Describe count rate and instrument response which should be observed for rod withdrawal during the approach to criticality. 3.9 4.0 1.2 K1.04 Relate the concept of subcritical multiplication to predicted count rate response for control rod withdrawal during the approach to critical. 3.8 2.1 K1.05 Explain characteristics to be observed when the reactor is very close to criticality. K1.06 Calculate ECP using a 1/M plot. 2.9 K1.07 Calculate ECP using procedures and given plant procedures. 3.6 1.1 K1.08 List parameters which should be monitored and controlled upon reaching criticality. 3.7 2.4 K1.09 Define criticality as related to a reactor startup. 3.2 3.3 K1.10 Describe reactor power response once criticality is reached. 2.3, 2.4 K1.11 Describe how to determine if a reactor is critical. K1.12 List parameters which should be monitored and controlled during the intermediate phase of startup (from criticality to POAH). K1.13 Discuss the concept of the point of adding heat (POAH) and its impact on reactor power. K1.14 Describe reactor power response prior to reaching the POAH. K1.15 Explain characteristics to look for when the POAH is reached. K1.16 Describe monitoring and control of reactor power and primary temperature during 0% to 15% (B & W). K1.17 Describe reactor power response after reaching the point of adding heat. K1.18 Describe the monitoring and control of T-ave, T-ref, and power during power operation. K1.19 Describe means by which reactor power will be increased to rated power. K1.20 Explain the effects of control rod motion or boration/dilution on reactor power. K1.21 Explain the relationship between steam flow and reactor power given specific conditions. 2.5 K1.22 Explain how boron concentration affects core life. 2.6 K1.23 Explain the shape of a curve of reactor power versus time after a scram. 4.1 K1.24 Explain reactor power response to a control rod insertion. K1.25 Explain the necessity for inserting control rods in a predetermined sequence during normal shutdown. 4.2 K1.26 Define decay heat. 4.3 K1.27 Explain the relationship between decay heat generation and: a) power level history, b) power production, and c) time since reactor shutdown.