Independent Review of NPP Modifications and Safety Upgrades

International Journal of Contemporary ENERGY, Vol. 1, No. 1 (2015) ISSN 2363-6440 ___________________________________________________________________________________________________________

___________________________________________________________________________________________________________ D. Grgić, V. Benčik, S. Šadek, I. Bašić: “Independent Review of NPP Modifications and Safety Upgrades”, pp. 41–51 41

DOI: 10.14621/ce.20150106

Independent Review of NPP Modifications and Safety Upgrades

Davor Grgić*, Vesna Benčik, Siniša Šadek, Ivica Bašić1

Faculty of Electrical Engineering and Computing, University of Zagreb

Unska 3, 10000 Zagreb, Croatia, [email protected] 1APOSS d.o.o., Repovec 23B, Zabok, Croatia

Abstract In order to fulfil operational and safety requirements nuclear power plants are often subjected to modifications and safety upgrades. That is especially true for old generation plants where power uprate and replacement of obsolete equipment play important role. Overview of experience acquired during recent reviews of RTD Bypass Elimination modification and PARs and Passive Containment Filtered Vent introduction in frame of nuclear power plant Krsko safety upgrade project was presented.

1. Introduction In order to fulfil operational and safety requirements nuclear power plants (NPPs) are often subjected to modifications and safety upgrades. That is especially true for old generation plants where power uprate and replacement of obsolete equipment play an important role. The changes affecting safety or licensing status of the plant are usually subjected to independent review as part of licensing process or as a support in regulatory body decision making process. After Fukushima accident many plants initiated safety upgrades to improve robustness in case of extreme weather and other site specific phenomena. Such upgrades are more challenging, are based on newly established design extended conditions, require information on equipment survivability, and include severe accident analyses. All this makes independent reviews both more important and more difficult.

For small countries and small organizations it is rather difficult to provide required multidisciplinary analyses and to maintain required knowledge for small number of unique requests. In such situation there are some benefits found in research institutes and universities where active research is present covering different problems related to NPP safety, and where intensive exchange of information with related foreign institutions is maintained. Another important aspect is active cooperation of different organizations sharing required knowledge in performing independent analyses and reviews. Overview of experience acquired during recent reviews of Resistance Temperature Detector Bypass Elimination (RTDBE) modification [1] and Passive Autocatalytic Recombiners (PARs) [2] and Passive Containment Filtered Vent (PCFV) [3] introduction in frame of NPP Krsko Safety Upgrade Program (SUP) was presented.

Keywords: NPP safety upgrade, Modification, independent reviews and analyses

Article history: Received: 02 February 2015 Revised: 17 February 2015 Accepted: 20 February 2015



2. Overview of modifications review process In order to show some of the problems related to review of plant modifications, as well as to highlight required type of knowledge and potential benefits of review process (beyond formal licensing support), overview of the review process for three modifications/safety upgrades implemented in NPP Krsko will be presented. First example is modification of existing system to improve plant operation and maintenance within existing licensing design base. Two additional examples are related to safety upgrade program used to improve plant safety for design extended conditions and beyond design accidents. In all three cases multidisciplinary knowledge was required ranging from reactor physics and radiation shielding, to heat transfer, safety analyses, severe accident calculation, mechanical and structural engineering and I&C. University of Zagreb Faculty of Electrical Engineering and Computing (FER) had overall responsibility for review and for integration of results of other institutions. University of Ljubljana Faculty of Civil and Geodetic Engineering (FGG) was responsible for seismic and structural analyses of PAR and PCFV modifications. Three small engineering and consulting companies (INKO Ljubljana and APOSS Zabok in case of PAR and PCFV modifications, and ENCONET Zagreb and APOSS Zabok in case of RTDBE modification) provided specific plant knowledge (knowledge of equipment, systems and plant procedures) and support during intensive walk down activity.

2.1. RTD bypass elimination modification

Two concepts of Reactor Coolant System (RCS) coolant temperature measurements exist, RTD bypass concept (small auxiliary piping and manifold mounted RTDs), and thermowell mounted RTDs concept (RTDs immersed directly in hot and cold legs) [4]. Each approach has some benefits and drawbacks. Main benefits of RTD bypass elimination should be easier operation and maintenance due to elimination of small piping and related valves, and associated manipulation and adjustment during refuelling outage. From point of view of measurement accuracy and delay of measured temperature compared to real fluid temperature both concepts could achieve similar performance. In case of RTD bypass additional delay is mainly due to transport delay in small piping and to the smaller extent due to delay in RTD response. In case without bypass, transport delay is eliminated, but response of thermowell mounted RTD is slower (thermal inertia). In first case additional influence is present due to heat loss in small piping (larger surface to volume ratio) connecting hot leg and RTD manifold, making measured temperature different from RCS temperature.

RCS temperature measurement is additionally complicated due to hot leg streaming (HLS). Thermal stratification in hot leg pipe (often called hot leg streaming) of PWR plant is caused by different heating of core fluid streams in fuel assemblies having different power production. Calculated NPP Krsko reactor core exit coolant temperatures (C) are shown in Figure 1. Fluid streams having different temperature at core exit, after some mixing in outer plenum, enter hot leg pipe. Any perpendicular plane downstream hot leg nozzle will have some kind of 2D temperature distribution and temperature measured by immersed RTD will depend on circumferential position and insertion depth of RTD detector. In order to get representative average temperature of hot leg water typically 3 different measurement positions were used. Due to mixing influence of reactor coolant pumps the problem is usually not present in cold leg pipes.

Hot leg streaming presence is independent of type of measurement, but in RTD bypass case sampling streams are hydraulically mixed and then average temperature is measured using multiple RTDs, and in case without bypass averaging has to be done electronically using signals obtained from different positions on circumference of the pipe. That can make second approach more sensitive to temperature oscillations. In order to have two independent channels per loop, number of sampling points was increased to six, but two obtained measurements for hot leg temperature are not taken from the same place at the circumference of the pipe and are not the same (due to HLS). Measured temperature signal is used both in control and protection system and additionally it is used in calorimetric procedure during determination and correction of measured reactor mass flow rate. Related flow uncertainties are dependent on uncertainties present in prediction of average fluid temperature and they depend on circumferential profiles of coolant temperature in hot leg pipes.

NPP Krsko was changed way of RCS coolant temperature measurement from RTD bypass to thermowell mounted RTDs as part of NEK RTD Bypass Elimination project to improve operation and maintenance. The modification includes hardware changes (removal of existing RTD bypass piping and support, making new holes in hot and cold legs, modification of existing hot leg scoops, introduction of thermowell mounted RTDs, changes in process cabinets), changes in control and protection system (new narrow range RCS temperature measurement system processing, changes in rod control system constants, and redefinition of Over Temperature DeltaT (OTDT) and Over Power DeltaT (OPDT) protection set points). Modification covered additional things not directly related to RTDBE: changes in steam line pressure low set point (lead-lag controller time constants) and associated engineering safety features



Figure 1: Core-wide coolant exit temperatures calculated by RELAP5-PARCS coupled code response times, and changes in steam dump system and additional optimization. As direct or indirect consequences of described changes, Safety Analyses Report (SAR) and Technical Specifications (TS) were changed too.

As we can see the modification includes small change in mechanical components, introduction of rather simple changes in I&C system, and analytical part needed to verify response of control and protections systems. From point of view of analytical effort that was largest activity after power uprate of the plant and replacement of steam generators.

Main findings of RTDBE modification review evaluation can be summarized as follows:

• Proposed solution is rather common for 3 and 4-loops PWRs, additional insight is needed for 2-loopers,

• Conservative HLS profiles were used due to lack of experience with 2-loop plants,

• Used electronic equipment and RTDs are of proven design,

• Temperature uncertainties were properly recalculated,

• Fluid structure interaction for directly immersed RTDs is implicitly covered (based on NPP Vandellos experience),

• Electronic averaging and way how channels are organized can cause oscillations in Thot (proposed design solution: filtered signals and changed OTDT/OPDT protection),

• Initial design of the system (compensation of Tcold signal) is judged to be sensitive to overcooling transients,

• Additional possible complication present in signal conditioning is need for bias constants in two channels averaging process, proposed solution was found acceptable,

• Possibility for dual RTD failure is covered by analysis and additional reconfiguration of system if needed,

• RCS pressure boundary integrity is not negatively influenced by proposed RTDBE modification, and

• New OPDT and OTDT set points provide adequate protection of the plant.

In order to perform the evaluation the knowledge of the present status of the plant and related analyses performed earlier, as well as information on similar attempts in other countries were needed. Some additional independent calculations were performed to support evaluation process. We used CFD (FLUENT [5]) calculation to determine approximate HLS profiles for different plant fuel cycles and operational conditions. Initial information, coolant temperatures at the exit of



each fuel channel, is coming from whole core PARCS [6] calculation, Figure 1. The maximum core exit coolant temperature difference is influenced by selected fuel loading pattern (different fuel cycles were analysed), cycle burnup (usually beginning of the cycle was limiting) and assumed core coolant flow rate (best estimate and thermal design flow rate). Hot leg temperature distribution calculated for beginning of one of the plant fuel cycles is shown in Figure 2. O1 to O3 are used to label existing (old) RTD positions and N1 to N3 to label new RTD position (second channel per loop). It is clear that measured temperature depends on RTD position and insertion depth (black lines were used to show RTD positions, and black circles to show influence of insertion depth). The calculation was used to compensate for not enough information available on RTD bypass elimination in two-loop plants, and to clarify some measurement results related to temperature gradients for 18-months cycles and low leakage fuel patterns. As a consequence of this activity additional conservatism was introduced in temperature gradients used in modification design. The measurements performed during system commissioning showed that HLS gradients obtained with CFD calculation were conservatively high.

In order to check applicability of new RCS temperature measurements and protection set points, independent calculations were performed as part of the review for Rod Withdrawal at Power (RWAP), rod drop, Main Steam Line Break (MSLB, spectrum of breaks), and load rejection transients. RELAP5 [7] and RELAP5 coupled to PARCS were used for calculation.

In Figure 3 calculated measured and compensated temperature differences for loops 1 and 2 (DT1 and DT2) were shown together with OPDT protection set points, for small break SLB. Labels rtdb and rtdbe were used to mark situation before and after modification, and lead-lag, and lag 2 and 7 s were used to mark different Tcold compensations used after RTDBE. Small break SLB has

Figure 2: Hot leg temperature distribution and RTD positions

Figure 3: Setpoint response during MSLB for different OPDT implementations

0 5 10 15 20 25 30 35 40 45 50T I M E (sec)

DT

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DT1 rtdb SL DT2 rtdb SL DT1 rtdb DT2 rtdb DT1 rtdbe lead-lag DT2 rtdbe lead-lag DT1 rtdbe lag 2 s DT2 rtdbe lag 2 s OPDT1 DT1 rtdbe lag 7 s OPDT2 DT2 rtdbe lag 7 s

N E K M SL B A R -02 0.0304 m2



characteristics of slow reactivity initiated accident and intervention in OPDT protection can delay plant trip (as shown in Figure 3). As we said, the runs were performed using nodalization of the plant before and after modification. That way modification impact was checked without performing licencing type of the calculation (that was done by vendor). Within review just quantitative change influenced by modification was checked.

Containment response for relevant accidents was analysed using GOTHIC [8] calculations. General conclusion was that influence of the changes due to RTDBE was small. Rest of the potential problems were assessed during start-up measurements. Thot oscillations were less than expected, as well as expected streaming gradients. One of the things not properly taken into account in initial review was sensitivity of the cards in protection cabinets to ElectroMagnetic (EM) environment that exist in control room during relay manipulation in some auxiliary plant systems. The review was based on the assumption that the electronic equipment was replaced with the same components and with the same card design. Old cards were not tested from point of view of Electromagnetic Compatibility and that was not required for new cards too. The combination of small changes in card design and sensitivity of newly introduced lead-lag Tcold compensation resulted in inadvertent OPDT trip of the plant at the power close to the nominal power. Improved design using lag compensation, verified by calculation, instead of lead-lag Tcold compensation solved the problem. This was example where careful review was able to recognise weakness of the proposed initial solution, but not all aspects of the problem (design was sensitive both to overcooling events and EM noise). The final result of the modification is reliable operation of the plant with easier maintenance and decreased doses to maintenance personnel. Most of the safety analyses performed as part of the review were done by FER. The definition of relevant scenarios, review of the documentation and changes in plant procedures, and overview of hardware changes implementation were done by small engineering companies (APOSS and ENCONET) having people with required knowledge and experience. Taking into account different type of work needed to cover all aspects of this modification review, the benefit of this kind of cooperation is obvious.

2.2. Safety Upgrade Program PAR installation

NPP Krsko was capable for safe operation within current licensing limits for all internal and external events. Plant had implemented hydrogen control for all Design Bases Accidents (DBAs) using redundant electrical hydrogen recombiners. They rely on availability of electrical power and are not able to provide satisfactory protection for

severe accident conditions. Beyond design bases accidents were addressed only through plant specific Emergency Operating Procedures (EOPs) and Severe Accident Management Guidelines (SAMGs). As part of post Fukushima actions, and due to additional requirements introduced by Slovenian Nuclear Safety Administration (SNSA), the plant decided to introduce PARs. The installation of Passive Autocatalytic Recombiners should prevent dangerous concentration of combustible gases and mitigate the consequences of hydrogen and carbon monoxide presence inside the containment even in a case of core melt and core concrete interaction. As part of the modification NPP Krsko eliminated existing safety related electrical recombiners, and replaced them with two safety related PARs (DBA). Additional 20 non-safety related PARs were installed to ensure containment integrity for Design Extension Conditions (DEC) and for Beyond Design Basis Accident (BDBA) conditions. Independent review was performed as part of licensing process. The review was specific due to passive nature of proposed safety upgrade and due to need to address DEC conditions (the conditions that are more challenging/severe than original conditions used in plant design) for equipment design and installation. The findings that follow illustrate nature of the performed review.

PARs are passive devices and their impact to existing systems and structures should be minimal. They are not introducing materials that can affect hydrogen generation or debris forming. Their impact on containment free volume and metal mass is approximately neutral taking into account removal of existing electrical recombiners. They are seismically designed and tested (as well as their supports), and probability of their drop or collapse is negligible up to and including DEC conditions. The impact of PARs to structures supporting them is evaluated for DEC conditions that include 0.6g Peak Ground Acceleration (PGA) and corresponding seismic response spectra. The impact is small due to their small mass compared to mass of structures supporting them. The robustness and adequacy of PARs were tested using shaking table with appropriate excitation. The supports were evaluated using analytical tools (FEM model), and that performed calculation showed required robustness of the design. Impact of hot gases exiting PARs is limited due to selection of their location, shape of PAR outlet, and limited exhaust temperature (250 to 350 C depending on H2 concentration at time of actuation) and flow rate. Still, screening analysis was performed for cable trays above PAR at selected locations and additional protection was provided. PARs are designed for DBA and DEC environmental conditions in containment. Due to used material they are not sensitive to temperature, radiation or pressure. Catalyst surface is protected by hydrophobic polymer. The polymer is able to sustain



environmental conditions in containment up to the first PAR actuation and after that it is probably destroyed, but its presence is not needed anymore. There is no DBA accident being able to develop ambient temperature to damage the polymer. Taking into account PARs position they are exposed to limited radiation doses to polymer coating during normal operation and effect of selfshielding is present for all situations (streaming) except for airborne radioactivity (limited influence during normal operation and DBA accidents).

The locations selected for PARs guarantee that they are not exposed to direct mechanical, high temperature or jet impact from other systems. They can’t be immersed in water and due to its high elevation only gases and aerosols are candidates to enter PAR enclosure. Top of the PAR is protected from direct spray influence that can bring boron and other chemical to active catalyst surface. The locations of two safety related PARs are shown in Figure 4 (small green boxes are PARs, PCFV

Figure 4: PAR locations in NPP Krsko containment

Iodine filters are light blue boxes close to fan cooler ducts).

PARs of similar design with the same catalyst material are installed in other NPPs (in Germany and USA) and are supported by extensive experimental program. Representative experimental conditions are used (to the extent possible – there is no test being able to produce exact atmosphere composition as the one present during real severe accident, but most important catalyst poisons were present). The vendor of the PARs is present in that business since the beginning and has big experience in that area.

Two safety related PARs are able to replace existing electrical recombiners for DBAs (LOCA). We have performed independent calculation and the requirement to keep hydrogen concentration is fulfilled with margin. Obtained recombination rate (GOTHIC code was used for calculation) is lower than the one obtained by vendor. They used experimentally based correlation dependent solely on conditions in the volume where PAR is located. In GOTHIC calculation PAR was located on flow path and recombination depends on developed flow due to natural convection.

In Figure 5 H2 concentration is shown for post-LOCA hydrogen build up used in SAR Chapter 6 DBA calculation. The cases are shown for concentration with (er) and without electrical recombiner actuation (nor), and for different number of PARs (par, 2par, 22par), for actuation after one day (1d), for actuation when H2 concentration reached 3.5% (vf35) and for reduced recombination efficiency. In our calculation, contrary to vendor calculation, PARs were not so effective in H2 removal as electrical recombiners, but in all cases they were more than able to keep peek H2 concentration below required 4%.

Sizing calculation performed by vendor estimated required number of PARs in conservative way. The protection was based on oxygen “starvation” approach. When we did reverse calculation (GOTHIC calculation with MAAP [9] combustible gas production or standalone MELCOR [10] calculation), decreasing number of PARs (less than 20) was showing only limited decrease in reacted combustible gas (or depleted oxygen) mass. In Figure 6 H2 recombination rate for safety related PAR Station Black Out (SBO) sequence is shown, for different number of PARs (only two safety related or all 22 PARs), with and without reactor containment fan coolers (with condensation (fan coolers in operation) higher H2 concentration was experienced). The effective H2 concentration reduction was demonstrated for analysed accident sequences when all installed PARs are available.



Figure 5: DBA post-LOCA H2 mass

Figure 6: SBO sequence H2 recombination rate for PAR01 Beside independent calculations review resulted in some suggestions for changes in PAR locations, and additional protection of PARs during regular refuelling/maintenance in containment. One of the reviewer’s suggestions was to keep and extend functionality of H2 monitors to DEC conditions even after

PARs introduction, due to possible side effects of used oxygen starvation approach. The suggestion was not completely supported by the plant.

Most of the safety analyses performed as part of the review were again done by FER. FER was responsible for



plant walk downs too. MAAP severe accident calculations were done by APOSS as well as evaluation of modification influence to SAMG procedures. FGG did evaluation of all seismic responses and stress calculations. INKO evaluated selection of DEC conditions and influence to licensing documents and did review of project documentation. Again, different knowledge available to different members of the team was valuable advance of this kind of cooperation.

2.3. Safety Upgrade Program PCFV installation

The installation of Passive Containment Filtered Vent System was another measure initiated as part of NPP Krsko safety upgrade program. It should ensure containment integrity (preclude pressures that can challenge containment integrity) during Design Extension Conditions and Beyond Design Basis Accidents. To provide depressurization of containment while minimizing releases of radioactive particles to the environment NEK installed five aerosol filters inside containment, one iodine filter in auxiliary building and associated piping equipped with rupture disc, valves, expansion orifices and instrumentation. The venting gas first passes (from right to left in Figure 7) the aerosol filter modules and leaves the containment via piping through a containment penetration, passes to the iodine

filter and discharges to the environment through the stack (top of the figure).

When evaluating modifications influencing plant behaviour in DEC and BDBA conditions, clear regulatory requirements present in the case of plant licensing for DBA, are usually not available. The modification should be evaluated from point of view of interaction with existing plant licensing base and is it able to fulfil stated protection goals for plant and environment. In case of PCFV modification our findings were as follows.

PCFV system is and should be passive and that to largest amount limits its interaction with other plant systems. The influence due to additional mass of heavy components and piping should be properly addressed up to and including DEC seismic event. Containment isolation function should be kept. New isolation configuration for line penetrating containment (two valves in series outside containment) is introduced in the plant and all isolation valves should be covered with leakage testing. PCFV containment penetration and redesigned PCFV plant stack are not affected by DEC seismic event due to relative AB and RB displacement. Additional source of radioactivity is introduced in AB after PCFV actuation, but its influence is small compared to potential benefits.

Figure 7: PCFV system layout



The adequacy of MAAP model used in analyses by vendor (model was not described in detail) was found acceptable when supported by reviewer’s independent calculations (MAAP and MELCOR). The analyses were based on SBO sequences and that accident was found to be acceptable as limiting sequence. If active PCFV usage is planned, it was our position that, other sequences should be covered too. A detailed radiological analysis of the offsite doses was missing, but verification of the main requirement (release to the environment shall be <0.1% of reactor core fission products inventory) was implicitly fulfilled by design of the system (cumulative decontamination factor of filters). In order to use PCFV in active mode (below rupture disc opening pressure), our position was that, explicit calculation of radiological consequences is needed. As consequence of that, the system was approved only in passive mode till fulfilment of the requirement. That is because PCFV actuation at lower pressure can influence current licensing base (currently there is no radiological effluents, except design containment leakage, if the containment is intact, and it should be below rupture disc opening pressure). It was recommended to provide radiological impact analysis for passive mode of PCFV operation, too (noble gases are always discharged), and that was done by plant after modification implementation.

It was our position that the design and operation of aerosol filters (metal fibre) was supported by number of experiments. That includes retention capability and heat transfer used to cool the filters. For Iodine filter, experimental support is mainly limited to retention capabilities of active material (Zeolite). We have found that most important factors, like decontamination factors and retention capabilities were covered by performed experiments and Factory Acceptance Testing (FAT). Our position was that aerosol filters are robust enough and that Iodine filter with Zeolite is potential weak point due to requirements for specific inert conditions for active material after installation. That can question (maintenance of N2 pressure within iodine filter and related surveillance) passive nature of the system. Important part of the review was shielding review for the Iodine filter located in AB. The filter is acting as a heat source after PCFV actuation, but that is of less importance taking into account rather large volume and open connections of the room where it is located.

Used aerosol metal filters and Iodine filters (and related shielding) are rather heavy. Important part of the review was need to check allowable loads of civil structures and seismic response during DEC conditions (DBA and DEC spectra with PGA of 0.6g). From point of view of PCFV integrity, most attention was paid to the locations where PCFV piping connects two buildings (RB and AB), due to their relative motion during seismic event. The seismic and environmental impact on PCFV stack

outside containment was addressed properly both in design and in review process. During review process additional requirements were stated for locations where PCFV piping penetrates roof/floor concrete slabs (sleeve and seals).

As part of the review we asked for strict testing (leakage testing using appropriate pressure difference, at least up to DBA conditions) of all valves and rupture disk which are acting as containment isolation boundary. That should be taken into account both in design of the system and in the plant testing procedures.

Radiological monitor and flow measurement in exhaust PCFV stack were requested as part of PCFV design. It was not possible to do that in passive way and it was not implemented initially. It was our suggestion to provide the measurement with appropriate active system. Finally it should be mentioned that implementation of such modification/system in existing plant is difficult task which need proper planning and enough time for implementation. In case of NPP Krsko rather big effort was needed to implement the modification on time and some changes in project were needed during realization due to limited available time for preparation of the modification.

The review was supported by independent calculations of containment and PCFV response using MAAP, MELCOR and RELAP5 (PCFV piping). In addition some CFD (FLUENT) calculations were performed for distribution of the PCFV discharge outside ventilation duct. In Figure 8 velocity streamlines of the discharged fluid for specific wind conditions and simplified plant building arrangement were shown.

Tasks distribution between organizations participating in the review was similar to the one described in the case of PAR modification review.

3. Conclusion

Independent review of NPP modifications and safety upgrades can be, depending on type of modification, very challenging process. It requires multidisciplinary knowledge and at least familiarity with plant systems and procedures. There are important differences between upgrades of existing systems within current licensing basis and introduction of safety upgrades to provide protection from BDBA events. It is rather difficult to maintain expertize required for safety reviews in small countries and for small number of nuclear objects. One approach can be to join capabilities of universities, research institutes and small consulting companies having required diversity of knowledge. The safety review, if properly prepared, can be not just part of formal licensing, but can improve and optimize modification itself.



Figure 8: PCFV duct discharge, velocity streamlines and pressure contours Acknowledgements The authors gratefully acknowledge the work done by other members of review team, as well as the support provided by NPP Krsko and Slovenian Nuclear Safety Administration, and active cooperation of Westinghouse Electric Company, vendor of the equipment, systems and services, during modifications review process.

Abbreviations AB – Auxiliary Building BDBA – Beyond Design Basis Accident DBA – Design Basis Accident DEC – Design Extension Condition EM – Electro-Magnetic EMC – Electro-Magnetic Compatibility EOP – Emergency Operating Procedure

FAT – Factory Acceptance Testing FER – Fakultet Elektrotehnike i Računarstva

Zagreb FGG – Fakulteta za Gradbeništvo in Geodezijo

Ljubljana HLS – Hot Leg Streaming FEM – Finite Element Method LOCA – Loss of Coolant Accident MSLB – Main Steam Line Break NEK – Nuclear power plant Krško NPP – Nuclear Power Plant OPDT – Over Power DeltaT OTDT – Over Temperature DeltaT PAR – Passive Autocatalytic RecombinerPCFV – Passive Containment Filtered VentPGA – Peak Ground Acceleration PWR – Pressurized Water Reactor RB – Reactor Building RCS – Reactor Coolant System RTD – Resistance Temperature Detector



RTDBE – Resistance Temperature Detector Bypass Elimination

RWAP – Rod Withdrawal at Power SAMG – Severe Accident Management GuidelinesSAR – Safety Analysis Report SBO – Station BlackOut SNSA – Slovenian Nuclear Safety AdministrationSUP – Safety Upgrade Program TS – Technical Specifications References [1] Bajs, T., Bašić, I., Benčik, V., Grgić, D., Independent

Evaluation Report of "Krško RCS Temperature Measurement Optimization", FER-ZVNE-FR01/13-0, Zagreb, 2013.

[2] Bilić-Zabric, T., Bašić, I., Grgić, D., Independent Evaluation Report of "Severe Accident Hydrogne Control System (PAR)", FER-ZVNE-FR02/13-0, Zagreb, 2013.

[3] Bilić-Zabric, T., Bašić, I., Grgić, D., Final Independent Evaluation Report of "Passive Containment Filtered Vent (PCFV) System", FER-ZVNE-FR03/13-0, Zagreb, 2013.

[4] Hashemian, H.M., Maintenance of Process Instrumentation in Nuclear Power Plants, Springer-Verlag Berlin Heidelberg, 2006.

[5] ANSYS FLUENT User's Guide, ANSYS, Inc., Release 14.5, October 2012.

[6] Downar, T., Lee, D., Xu, Y., Kozlowski, T., PARCS v2.6 U.S. NRC Core Neutronics Simulator User Manual, School of Nuclear Engineering, Purdue University, 2004.

[7] RELAP5/mod3.3 User’s Manual, NUREG/CE-5535/Rev 1, Information Systems Laboratories, Inc., Rockville-Maryland, Idaho Falls-Idaho, January 2002.

[8] GOTHIC Containment Analysis Package, Technical Manual, Version 7.2b(QA), NAI 8907-06 Rev 17, March 2009.

[9] MAAP4 User’s Manual, Electric Power Research Institute, 1994.

[10] MELCOR Computer Code Manuals, Vol. 2: Reference Manual, NUREG/CR-6119, December 2000.

Independent Review of NPP Modifications and Safety Upgrades

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