APROS model validation in case of fast transients in a boiling water reactor

This Master´s thesis investigates the performance of the Olkiluoto 1 and 2 APROS model in case of fast transients. The thesis includes a general description of the Olkiluoto 1 and 2 nuclear power plants and of the most important safety systems. The theoretical background of the APROS code as well as the scope and the content of the Olkiluoto 1 and 2 APROS model are also described. The event sequences of the anticipated operation transients considered in the thesis are presented in detail as they will form the basis for the analysis of the APROS calculation results. The calculated fast operational transient situations comprise loss-of-load cases and two cases related to a inadvertent closure of one main steam isolation valve. As part of the thesis work, the inaccurate initial data values found in the original 1-D reactor core model were corrected. The input data needed for the creation of a more accurate 3-D core model were defined. The analysis of the APROS calculation results showed that while the main results were in good accordance with the measured plant data, also differences were detected. These differences were found to be caused by deficiencies and uncertainties related to the calculation model. According to the results the reactor core and the feedwater systems cause most of the differences between the calculated and measured values. Based on these findings, it will be possible to develop the APROS model further to make it a reliable and accurate tool for the analysis of the operational transients and possible plant modifications.

Boiling water reactor technology status of the art report / [edited by C. Victor Pearson].

"Reactor technology."

APROS code benchmarking inboiling water reactor accident analyses

Diplomityö käsittelee kiehutusvesilaitosten transienttien ja onnettomuuksien analysointia APROS-ohjelmiston avulla. Työ on tehty Teollisuuden Voima Oy:n (TVO) Olkiluoto 1 ja 2 laitosyksiköiden mallin pohjalta. Raportissa esitetään ohjelmiston käyttämiä yhtälöitäja laskentamalleja yleisellä tasolla. Työssä esitellään laitoksen yleispiirteet turvallisuustoimintoineen ja kuvataan ohjelmaan suureksi osaksi aiemmin luotua laskentamallia. Työssä on luetteloitu voimassa olevatlisensiointianalyysit, joiden joukosta on valittu laskentatapauksia ohjelmiston suorituskyvyn arviointia varten. Lisäksi työhön on valittu laskentatapauksia muilla kuin lisensointiin käytetyillä ohjelmilla lasketuista analyyseistä. Lisäksi on suoritettu vertailulaskuja konservatiivisen ja realistisen mallin erojen esille saamiseksi. Laskentatapauksia ovat mm. ylipainetransientti, jäähdytteen menetysonnettomuus ja oletettavissa oleva käyttöhäiriö, jossa pikasulku ei toimi (ATWS). Diplomityön edetessä laitosmallia on kehitetty edelleen lisäämällä joitakin järjestelmiä ja tarkentamalla joidenkin komponenttien kuvausta. Työssä ilmeni, että APROS soveltuu jäähdytteenmenetysonnettomuuden ja suojarakennuksen yhtäaikaiseen analyysiin. APROS.n vaste nopeisiin transientteihin jäi kuitenkin vertailutasosta. Tämän työn perusteella APROS-mallia kehitys jatkuu edelleen siten, että se soveltuisi entistä paremmin myös nopeiden transienttien ja ATWS-tilanteiden kuvaamiseen. Työssä olevaa lisensointianalyysien kuvausta tullaan käyttämään hyväksi selvitettäessä laitoksen turvallisuuden väliarviossa tarvittavien analyysien määrää ja laatua. Nyt saatuja kokemuksia voidaan hyödyntää myös mahdollisen kolmiulotteisen sydänmallin hankinnassa APROS-ohjelmistoon. Tässä diplomityössä esitettyjä parannuksia voidaan käyttää hyväksi SAFIRtutkimusohjelman hankkeiden suunnittelussa.

Assessment of CTF boiling transition and critical heat flux modeling capabilities using the OECD/NRC BFBT and PSBT benchmark databases

The need to refine models for best-estimate calculations, based on good-quality experimental data, has been expressed in many recent meetings in the field of nuclear applications. The modeling needs arising in this respect should not be limited to the currently available macroscopic methods but should be extended to next-generation analysis techniques that focus on more microscopic processes. One of the most valuable databases identified for the thermalhydraulics modeling was developed by the Nuclear Power Engineering Corporation (NUPEC), Japan. From 1987 to 1995, NUPEC performed steady-state and transient critical power and departure from nucleate boiling (DNB) test series based on the equivalent full-size mock-ups. Considering the reliability not only of the measured data, but also other relevant parameters such as the system pressure, inlet sub-cooling and rod surface temperature, these test series supplied the first substantial database for the development of truly mechanistic and consistent models for boiling transition and critical heat flux. Over the last few years the Pennsylvania State University (PSU) under the sponsorship of the U.S. Nuclear Regulatory Commission (NRC) has prepared, organized, conducted and summarized the OECD/NRC Full-size Fine-mesh Bundle Tests (BFBT) Benchmark. The international benchmark activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD) and Japan Nuclear Energy Safety (JNES) organization, Japan. Consequently, the JNES has made available the Boiling Water Reactor (BWR) NUPEC database for the purposes of the benchmark. Based on the success of the OECD/NRC BFBT benchmark the JNES has decided to release also the data based on the NUPEC Pressurized Water Reactor (PWR) subchannel and bundle tests for another follow-up international benchmark entitled OECD/NRC PWR Subchannel and Bundle Tests (PSBT) benchmark. This paper presents an application of the joint Penn State University/Technical University of Madrid (UPM) version of the well-known subchannel code COBRA-TF, namely CTF, to the critical power and departure from nucleate boiling (DNB) exercises of the OECD/NRC BFBT and PSBT benchmarks

Potential issues related to emergency core cooling system strainers performance at boiling water reactors: application to Cofrentes NPP (Spain)

From the 60s to the 90s, a great number of events related to the Emergency Core Cooling Systems Strainers have been happened in all kind of reactors all over the world. Thus, the Nuclear Regulatory Commission of the USA emitted some Bulletins to address the concerns about the adequacy of Emergency Core Cooling Systems (ECCS) strainer performance at boiling water reactors (BWR). In Spain the regulatory body (Consejo de Seguridad Nuclear, CSN) adopted the USA regulation and Cofrentes NPP installed new strainers with a considerable bigger size than the old strainers. The nuclear industry conducted significant and extensive research, guidance development, testing, reviews, and hardware and procedure changes during the 90s to resolve the issues related to debris blockage of BWR strainers. In 2001 the NRC and CSN closed the Bulletins. Thereafter, the strainers issues were moved to the PWR reactors. In 2004 the NRC issued a Generic Letter (GL). It requested the resolution of several effects which were not noted in the past. The GL regarded to be resolved by the PWR reactors but the NRC in USA and the CSN in Spain have requested that the BWR reactors investigate differences between the methodologies used by the BWRs and PWRs. The developments and improvements done for Cofrentes NPP are detailed. Studies for this plant show that the head loss due to the considered debris is at most half of the limited head loss for the ECCS strainer and the NPSH (Net Positive Suction Head) required for the ECCS pumps is at least three times lower than the NPSH available.