Containment

ISSUE TITLE: Containment integrity (CS 1) ISSUE CLARIFICATION:

Description of issue

This issue is also applicable to NPPs with LWR.

In case of a severe accident resulting from a multiple failure of reactor components and safety systems, the pressure and temperature loading of the containment might eventually exceed the design limits and the containment could leak significantly. The challenge to containment integrity due to overpressurization can be mitigated by accident mitigation strategies such as:

• containment venting;

• hydrogen control measures;

• containment heat removal

The basic idea for containment venting is to open a controlled and filtered flow path to the external environment to relieve the pressure that is generated inside the containment due to various processes during the accident.

The main sources of hydrogen generation which can lead to combustible gas mixtures are:

• metal-water reaction involving the fuel element cladding;

• core melt-concrete-interaction;

• radiolytic decomposition of the water in the reactor core and the containment sump.

The effects of hydrogen burners and the possibility of containment failure due to over-pressurization were confirmed by the TMI-2 accident and by extensive research.

By implementing containment venting and/or hydrogen control measures, it may be possible to delay or prevent gross structural failure of the containment. This in turn would provide some additional time to mitigate the accident or to reduce the off-site-consequences of the accident compared to those produced by gross containment failure.

After a steam (D2O or H2O) leak in containment the pressure would rise to some value less than the design. The containment cooling system would take the enthalpy out and bring the pressure down to a level from which controlled depressurisation can start. Continued inleakage of instrument air could delay the depressurization or could cause a pressure increase. However current practice for most plants is to shutoff the instrument air after about 2 hours.

Safety significance

Severe accidents and DBAs like steam line break with loss of dousing can result in phenomena, such as the ones described above, which could result in containment failure. However, severe accidents in PHWRs are very slow due to the presence of the moderator and shield tank as heat sinks. This means the rise in containment pressure is also slow, giving time for severe accident mitigation and management.

Source of issue (check when appropriate)

• ____xx____ operational experience

• __________ deviation from current standards and practices

• ____xx____ potential weakness identified by deterministic or probabilistic (PSA) analyses

MEASURES TAKEN BY MEMBER STATES:

Argentina

The leaktightness of the Embalse containment is monitored within the surveillance containment program every five years. An important contributor to the containment leakage is the leak rate of room R001 (refuelling machine). Due to leakage detected during repetitive tests performed in Embalse NPP room R001, a specific procedure (TP-21080-02) to identify leakage points was created. In essence this procedure indicates to pressurize (reduced pressure) at both 0,210 Kg / cm² and 100 gr. / cm2 pressure values and to inspect inner room R001 surfaces by an adequate technique. Three tests were performed, the first two tests objective were to locate leakage points and the third one test, performed after corrective actions (fix) were taken, served to verify such corrective actions efficiency. The efficiency to detect leakage pressurizing to 100 gr. / cm² was demonstrated and this pressure value was incorporated to the procedure. It was important because of pressurize to 0,210 Kg / cm2 implicates evacuate water from fuel storage pools. The mentioned demonstrate that both procedure application and the corrective action taken are useful to improve containment performance. It was recommended to perform the repetitive test once in a year.

Canada

See SS 4, “Leakage from systems penetrating containment or confinement during an accident”, and See SS 5, “Hydrogen control measures during accidents”.

China

The design pressure of TQNPP containment is 124 KPa. The measuring range of containment pressure measuring instrumentation was originally designed as –20~150KPa(g), which acted as a parameter of PAM and was displayed in MCR and SCA. The basic design concept of CANDU containment only considers peak pressure generated from large LOCA not exceeding 124KPa(g). As a parameter of PAM, an accident resulting from the main steam tube rupture inside the containment shall be considered. AECL’s calculation indicated that peak pressures of the main steam tube rupture are 179KPa(g) (available for dousing system) and 400KPa(g) (unavailable for dousing system) respectively.

It is required by NNSA to change the measuring range of the containment pressure instrumentation to –20~400KPa(g).

There are 35 local air coolers inside the containment of TQNPP. 16 of them are safety-related. The 16 air coolers will be put into service when site design earthquake (SDE) occurs, 24 hours after LOCA.

The air coolers have two functions, i.e. agitating air to prevent excessively high H2 concentration in local areas and functioning as a heat sink to take out the accumulated heat inside the containment. It is required in safety review that the air coolers are needed to be in service after an earthquake. So power supply for the coolers shall also meet seismic requirements. However, the original design is not seismically qualified. The design change has been made to meet above requirement. The electricity distribution system from emergency power electricity distribution room to the local air coolers is designed to meet seismic requirements. This requirement also applies to the control system from the second control area to the local air coolers.

Many vessels of CANDU design are lined with epoxy resin, such as containment, spent resin bay and waste resin storage bay. It is required to use stainless steel liner instead of in TQNPP. Considering that the containment is a dry vessel and its inner surface is accessible to repair the liner if broken.

However, the spent fuel bay is a radioactive water bay and stores waste with high radioactivity. Once the epoxy liner breaks, it is very difficult to repair the bay. Consequently, the design changes have been completed that the spent fuel bay and the spent resin storage bay epoxy liners are instead of by stainless steel liner in TQNPP.

India

The reactor building of Indian PHWR consists of a primary containment (PC), designed for peak pressure following design basis accident, and a secondary containment (SC) which completely envelopes the primary containment. Both the containments are usually painted from both sides to reduce the leakage rate in case pressurization of PC following an accident. The Technical Specifications postulate the depressurisation of containment in Indian NPP depending upon design.

For older units depressurisation is envisaged from 1.2 to 6 hours after LOCA Annulus part formed in between PC and SC as kept at negative pressure during normal operation as well as during accident condition to prevent the ground level release of activity. To minimise repressurisation of RB, provision exists to reduce instrument air supply to RB.

The proof testing of primary containment is done at design pressure before the operation of reactor.

Main steam line break (MSLB) accident is governing event for primary containment design. Unless there is steam generator tube leak MSLBA does not have radiological consequences, the ILRT of primary containment is carried out at LOCA peak pressure which is below the design pressure of PC.

Indian PHWR Containment is qualified even for severe accidents like LOCA with ECC failure. In such accidents, moderator acts as a backup heat sink. Thermo-mechanical behaviour of coolant channel under such accident enhances the heat transfer path to moderator. Thereby, the fuel temperature is maintained at lower value leading to lesser hydrogen generation. It is also planned to qualify the containment under the severe accident like moderator cooling failure during LOCA + ECCS failure. Under such cases, fuel melts, coolant channel and calandria tubes fail and debris can be held in calandria by providing cooling through calandria vault water. Under such accident, hydrogen generation is of concern. This is addressed through the hydrogen management measures.

Ultimate load capacity of the Inner containment structure of a few Indian NPPs was calculated which about twice the design capacity.

Korea, Republic of

The inside surfaces of containment buildings and spent fuel storage pool structures in PHWR plants are lined with non-metallic, not steel, liner, so the maintenance of the non-metallic liner is very important to ensure leak tightness.

In 1998, Wolsung 1 NPP discovered some degradations of the non-metallic liner, such as cracks and swelling, in the spent fuel storage pool structures.

The spent fuel storage bay (SFSB) is surrounded by water-proofing membrane which prevents the release of leaked cooling water from SFSB to the ground. The inflow of water to the sump located between SFSB and the membrane indicates the leakage of cooling water from SFSB or the loss of function of the membrane.

As the leakage of cooling water from SFSB causes direct release of radioactive material to environment, the collaboration of CANDU-6 owners is required to resolve this matter with consideration of detailed inspection, necessity of repair and future design change.

The leaktightness of the CANDU-6 plant containment is ensured by prestressed concrete structure and rigid type non-metallic liner applied on inside concrete surface. Since the inherent nature of concrete has permeability and no practical capability for preventing the occurrence of cracking, leaktight integrity of containment is compromised by the cracking of concrete surface which directly causes the tearing of liner and leads to bringing down the leaktightness function.

The lessons learned from the failure of ILRT requirement of CANDU-6 plants which had been appropriately designed and constructed according to related codes and specification require the

improvement of liner material and related technical standards including the strengthening of qualification test provisions and periodic inspection requirements for containment.

In addition to the integrity on LOCA pressure which can be confirmed by ILRT, when the integrity on ultimate pressure (dual failure condition) is concerned, the design change of liner material to steel which has ductile nature to large deformation and leak resistance has to be considered.

The reactor containment buildings of the Wolsong nuclear power plants are made of prestressed concrete of the bonded type. As time goes by the prestress level is reduced due to concrete creep and shrinkage, along with tendon relaxation. Therefore, in-service inspection of the prestressing system is required to verify whether the effective prestress provided in the structure has reduced the safety margins used in the design.

In-service inspections of CANDU nuclear power plants, such as the Wolsong units, aren't performed in the containment building directly. The loss of effective prestress of the prestressing system in containment is estimated by comparing theory predictions applied to the design and the test results from experimentation of the test beam. The test beams are constructed of the same materials and methods as for the containment structure and stored at a site subjected to similar environmental conditions.

The test beam is very different from the size, shape and bonding state of the prestressing system in containment, so it is impossible to evaluate quantitively the variation of effective prestress in the containment building by the test beam approach.

For Wolsong stations, in order to complement the indirect characteristics of the above method, additional strain gauges method, in which strain gauges are embeded in the containment structure, is used. But, it is difficult to assess the structural integrity because this measured data is small and the strain gage is very sensitive to change of temperature after which the analysis result isn't reliable.

To ensure the leak tightness function of the spent fuel storage pool, the following activities were performed.

− inspection for degradations (every 3 month)

− measurement of the inflow water in the sump below the spent fuel storage pool (once a month)

− analysis of the radionuclide of water in the sump (once a month)

− establishment of a research program for the integrity of non-metallic liner Pakistan

The containment of KANUPP was tested for leak rate at the design pressure of 27 psig at the time of commissioning in 1971. Since then, the reactor building was tested after every two year at two (2) psig and the leak rate at the design pressure of 27 psig was predicted by extrapolation. According to the revised KANUPP Final Safety Report (KFSAR), the containment building pressure will not rise above 13.4 psig in case of a critical break (where maximum probability of activity release exist) in PHT system headers. Therefore, testing of Containment Building at 13.4 psig will ensure the intended level safety. This has revealed the increased safety margin of the containment barrier.

KANUPP containment building was tested at 5 psig in 1993. The leakage rate was found to be within limits. The feasibility of testing at higher-pressure i.e. 10 psig and 15 psig has been carried out. For execution of these tests at KANUPP, installation of a remote monitoring system is required in containment building. The test at 10 psig will be carried out in 2006 after the installation of these measuring instruments.

Romania

Containment integrity requirements are defined as part of the Licensing Basis and they are reflected in the safety reports. In addition to the safety envelope the operational envelope is defined for containment systems, as part of the Operation Policy and Principles. These envelopes are defined for the DBA postulated. In order to check the compliance of the containment integrity with the requirements commissioning tests were performed and there are tests included in the in service inspection procedures. The containment integrity is monitored also based on the results of the support documentation on the containment behaviour for a long term after a LOCA. The leak rate tests, which are going on in this program have some points of interest, as for instance the definition of some of the refuelling machine area as part of the containment for these tests and the evaluation of the results These requirements and results are currently under evaluation as part of the strategic policy for relicensing unit 1.

On the other hand there are requirements for the SAM procedures from which evaluations for containment monitoring procedures and necessity for hardware improvements are expected as part of the Periodical Safety Review for Unit 1. The review of the passive parts of containment systems (as for instance the pre stressing cables) is also part of these programs.

It is expected that in 2001 more details and updates are to be included in this document.

The Licensee will manage the performance of this project in a framework of a long term Research and Development Strategic Safety Program.

ADDITIONAL SOURCES:

• INTERNATIONAL ATOMIC ENERGY AGENCY, Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety: Concrete Containment Buildings, IAEA-TECDOC-1025, IAEA, Vienna (1998).

• Strategic Policy for Cernavoda NPP Unit 2 licensing process, CNCAN 1997.

• FSAR Cernavoda Unit 1, 1995.

• Strategic Policy for Cernavoda NPP Unit 1 relicensing in May 2001, CNCAN March 2000.