NE 405/505 Exam 2 Spring 2015 (80%) 1) A PWR with UTSGs is operating at 100% power, BOC, with control rods all out in automatic control when a failure in the speed pump controller results in all feed pumps running under speed. The Feed Control Valves are nearly full open at the initiation of the fault, and do not have sufficient capacity in their full open position to meet demand at 100% power. a) What broad accident type is this, i.e. power distribution anomaly, heat up, cool down, inadequate core cooling, etc? ANSWER: Heat Up Accident b) What safety concerns are associated with this accident? ANSWER: DNB, Over pressurization of the primary c) Given the attached trip logic diagram, what potential trips protect against this accident, and what is the primary trip protection function, (e.g. prevent over pressurization of the containment)? ANSWER: High Pressurizer Pressure => Over Pressurization High Pressurizer Level => Over Pressurization Low-Low Steam Generator Water Level => Loss of heat sink Low Feedwater Flow => Loss of Heat Sink DNB Trip => CHF Assuming all other control systems operate in their normal automatic mode, indicate all control system actions using logic flow diagrams (up/down arrows, etc.) and sketch the response of the following variables up to the point of reactor trip: i) Control Rod Position ii) Reactor Power iii) Rx Fuel Temperature iv) Average Coolant (Moderator Temperature) v) Feed Control Valve Position vi) Turbine Control Valve Position vii) Steam Pressure viii) Steam generator level Annotate your graphs to explain their shape. You may assume the upset is insufficient to result in an immediate reactor trip. ANSWER: Control Functions i) Feed Control Valves m FP FD : mfd mst LevelSG : FCV @ Full Open Position
ii) Turbine Control Valves The reduction in feed flow results in an increase in the bundle inlet enthalpy, and an increase in the vapor generation rate m m P P W FD g SG IMP T P ( P ) TCV P W IMP IMP REF IMP T iii) Control Rods PSG TSG TAVG since 0Rx Pwr MOD T ( T ) CR and Rx Pwr AVG AVG REF E1 T ( T ) ( ) : E2 (Rx Pwr) ( W ) ( ) AVG AVG REF REL T REL The control rods stop moving in when these two signals balance and the reactor settles at a new lower power level where the negative reactivity from the control rods is balanced by the positive reactivity from the reduction in fuel temperature iv) Pressurizer T Prz Level Prz Pressure AVG Assuming constant letdown Prz Level (Prz Level) REF m CHG Prz Pressure > (Prz Pressure) REF Q HTR m sp
d) As a reactor safety engineer, you are to analyze the potential severity of this upset. Of the following parameters, which values (if any) would you change from their nominal condition? Give the change and the reason for this change. i) Moderator and Doppler temperature coefficients ANSWER: If not already, Moderator coefficient at its least negative and Doppler at its most negative ii) Rods in auto or manual ANSWER: Since the rods want to move in to arrest the rise in T ave, rods should be set in manual iii) Turbine Bypass Valves ANSWER: It is possible, though unlikely, that the TBVs would open it T ave >>(T ave ) REF, so they should be set in manual
iv) Pressurizer Heaters ANSWER: Depends on whether you are analyzing for the potential for DNB, or Over pressurization Off for DNB, on for Over pressurization, though the heaters will likely go to their minimum value since system pressure is increasing v) Pressurizer Sprays ANSWER: Depending on whether you analyzing for the potential for DNB, Over pressurization On for DNB, off for Over pressurization vi) Pressurizer PORV ANSWER: Depending on whether you analyzing for the potential for DNB, Over pressurization On for DNB, off for Over pressurization (20%) 2) Short Answer a) Why is core inlet temperature an input to the DNB trip? ANSWER: For a given G and P, Tin q crit b) Why does the reactor trip logic include both low RCS flow, and a Reactor Coolant Pump Breaker Trip? ANSWER: The Breaker Trip responds faster than a low flow trip in the event of pump failure c) What is the major difference in the system behavior for a RCP shaft seizure versus a RCP shaft break? ANSWER: A shaft seizure terminates flow in the affected loop immediately, while a shaft break decouples the impeller from the pump and fly wheel allowing for reverse flow in the affected loop. d) What is the role of the Accumulators in the Safety Injection System? ANSWER: Provides for a low pressure, passive injection of primary system cooling water in the event of a LOCA e) What two accident categories are represented by a rod ejection accident? ANSWER: Reactivity anomaly and Small Break LOCA f) How is hydrodynamic slow stability addressed in PWRs? BWRs?
ANSWER: PWRs => Limit the maximum core voiding BWRs => Canned assemblies and assembly inlet orificing g) What concerns are addressed by maximum assembly burnup limits? ANSWER: Material limits h) Why do BWR fuel assemblies have higher enriched rods in the center of the assembly, and lower enrichment rods on the periphery of the bundle? ANSWER: Local peaking factors due to higher moderation on the bundle periphery and lower moderation in the high void region at the center of the bundle i) Why is the inter-assembly gap (spacing) so different between PWRs and BWRs? ANSWER: The control elements (control blades) for BWRs traverse the inter-assembly gap. j) In Fuel Assembly Design, why does the Nuclear Designer focus on the fuel rod pitch, and not the rod diameter? ANSWER: Rod diameter is set by T&H considerations. For a given rod diameter, the rod pitch is directly related to the W/U ratio and must be set to insure the lattice is under moderated.