Following agreements between Tony Blair and Jacque Chirac and confirmed by Gordon Brown and Nicholas Sarkozy, previously privatised British Energy was 85% re-nationalised by the French state with the intention of building a new fleet of nuclear power stations around the UK's shores. To meet a statutory duty to assess the designs of the new stations, a team of experts from the Health and Safety Executive's Nuclear Installations Inspectorate together with the Environment Agency was assembled.
 
The Generic Design Assessment (GDA) team when constituted was expected to take until 2012 to complete its work and found difficulty in recruiting its experts; most of those in the UK are either retired or approaching retirement. But the manning up of the team with French experts risks compromising the independence of the GDA, which it is supposed to maintain.
 
The GDA has issued a mellifluous report claiming that two of its fours steps in the assessment process have been successfully passed. It is hoped that the ageing of major reactor components so nearly bringing catastrophe will feature in the ensuing scrutiny.
 
The prospective generators are impatiently awaiting GDA’s decisions and political pressure is on it to speed up its processes. It is in all our interests that unwarranted haste will not dilute the detailed examination of the inherent disadvantages of the nuclear fuel cycle.

Ageing

Geriatrics is a branch of science dealing with old age, a science now applied to ageing nuclear reactors as the UK's joint HSE/NII * and Environment Agency's convened Generic Design Assessment team (GDA) considers the length of the operational life cycle of two candidate third generation power station designs. The GDA is beholden to "ensure that any … built in the UK meet the highest standard of safety, security, environmental protection and waste management." [1]

Two competing designs are under scrutiny, both of which are Pressure Water Reactors (PWR), the French EPR (Evolutionary Pressure-Water Reactor) by Areva and EdF and the US AP1000 by Toshiba-Westinghouse.

Once commissioned the entire internal components and surfaces of the reactors are subject to an ageing process influenced by irradiation, by thermal and by mechanical loads. There is also a phenomenon known as "stress corrosion cracking" (SCC) which also be-devils the chemical and food processing industries with its deleterious effect on alloy steels, exacerbated in a nuclear reactor by the irradiation of the neutron flux.

IAEA has published a report "Heavy Component Replacement in Nuclear Power Plants" [2]

The report considers as replacement objects the major PWR heavy components:

•    Steam generators

•    Reactor vessel heads

•    Reactor vessel internals

•    Pressurisers

•    Reactor coolant piping/recirculation piping

In France, Japan, Germany, Belgium, Switzerland, Korea, Spain, China, Slovenia and Sweden and in the US, SCC has led to the exchange of reactor vessel heads and entire steam generators. For example, at San Onofre power plant in California four replacement steam generators each weighing 640 tonnes were carried by barge and inserted through a 9 metre hole cut in the reactor concrete containment. [3]

Apart from the actual reactor vessel, the major components can be exchanged during a shutdown for refuelling, maintenance or inspection. In a complex operation, the control rods are dropped and the vessel head carrying the vessel internals is unbolted, lifted and transferred to the fuel pond, where the internals are submerged under water.

Once the fuel pack is under water and temporary shielding is installed, suitably clothed operators can repair or dismantle the defective components, but as they are moderately radioactive they are then placed in an on-site building for contaminated components colloquially known as a "mausoleum".

The IAEA report describes special provisions to facilitate the exchange of major components. In new designs it would be as well to incorporate suitable lifting appliances as well as consider the provision of exits for defective major components.

Steam generators

Cracking of the tubes in steam generators led to the contamination of the steam secondary circuits of the generator turbines. Initially, the solution was just the plugging of the defective tubes, but it was afterwards considered more economic to replace entire steam generators, because the reduction in the number of active tubes reduced the steam output. An investigation showed that the cause of the tube rupture was irradiation enhanced stress corrosion cracking (SCC) of the Inconel 600 alloy steel from which the tubes were made.

During the night of 11 February 2006, a major leak between the primary and secondary systems reached a flow rate of 500 litres per hour and led to the shutdown of the French Cruas 4 reactor. The investigations carried out by EDF determined that the leak came from a point of contact of a tube with the upper tube support plate on one of the steam generators. The tube failure was attributed to vibration fatigue due to a build-up of deposits on the support plate. The tube plate in the replacement steam generators has been modified to avoid the “clogging” as ASN describes it. See the ASN Annual Report 2007 pages 335-336. See ASN website for annual reports

In the replacement steam generators, the tube bundles were made from thermally-treated Inconel 690 alloy tube material, which it is hoped will prove more resistant to SCC and chemical attack, but only the passing of time will tell.

Up until 2005 IAEA reports that 83 steam generators have been replaced.

Reactor vessel heads

The failure of the reactor vessel heads, which is the flanged upper closure of the vessel bolted on the lower main part of the flanged vessel, was initiated by the stress corrosion cracking on one of the "penetrations" into the head, which are branches through which the control rods are inserted. The penetration material was Inconel 600 alloy with weld metal 52/152. A leaking head was first noticed because of a precipitation of boron leaving a white trail.

The most significant failure was noticed during refuelling of the Davis-Besse reactor in the US in 2002, when a large cavity, 17 cm long and 11 cm wide in the 15 cm thick carbon steel reactor head material was discovered. The hole extended down to the stainless steel internal cladding, which fortunately held the internal pressure, though it "ballooned" out somewhat.

When it was discovered that this degradation was generic, it was decided to exchange 54 vessel heads in France.

Up until 2005 IAEA reports that 93 reactor vessel heads have been replaced.

Reactor vessel internals

The reactor internals hold the fuel elements are therefore subject to intense radiation. Not surprisingly the internal components suffer irradiation embrittlement, fatigue, corrosion and radiation induced creep, relaxation and swelling, plus mechanical wear.

Although in some cases just cracked baffle bolts were replaced, most chose to replace the entire internals.

Pressurisers

Pressurisers are separate vessels connected to the cooling water pipework. Most of the SCC occurred in the penetrations for instrumentation, which were repaired by welding in new branches. However, some utilities decided to replace the entire pressuriser.

Reactor coolant piping/recirculation piping

A leak in a reactor heat removal circuit was discovered in France in 1998 and the failure was attributed to thermal fatigue.

Cause of major component failure

In particular the failure of components was attributed to SCC in the proprietary nickel steel Inconel 600 used for steam tubing and penetrations and in the weld metal employed. The replacement items make use of Inconel 690 or Inconel 800 as an alternative for tubing, fuel module parts and steam generator tubing.  The 82/182 weld metal has been replaced by alloy 52/153.

The first steam generator replacement was in the US in 1979, while the first reactor vessel head replacement was in France in 1993. No reports of a failure in a replacement major component have emerged, but as 30 and 16 years time has elapsed respectively, the next round of inspections are crucial to know whether the alternative metals have performed better.

Stress corrosion cracking

Stress corrosion cracking is a well-known phenomenon in the chemical and processing industries and occurs in alloy steels under stress, from say hoop stress in a vessel wall due to an internal pressure, together with contact with corrosive chemicals at a high temperature. In the case of a PWR nuclear reactor, all the internal surfaces of the various components in contact with the cooling water are subject to irradiation, heat, stress and to a corrosive boric acid solution acting as a neutron absorber.

The stress corrosion cracking in PWRs seems to be exacerbated by the addition of boric acid to the cooling water to enhance its moderating effect. The bombardment of a boron atom by a neutron releases an atom of tritium, a radioactive isotope of hydrogen. Hydrogen is added to the primary coolant to suppress the build-up of oxygen from radiolysis. Lithium hydroxide is added to control the pH (acidity-basicity) of the reactor cooling system. Lithium also produces tritium when irradiated.

Both hydrogen and tritium are known to penetrate alloy steels, so in this case the strain on the steel due to the applied stress from the pressurisation assists the ingress of hydrogen and tritium between the alloy grains. Once a tiny crack is initiated it continues to grow until it breeches the wall of a tube or the stainless steel cladding of the vessel.

The intergranular migration of the tritium allows it to enter the secondary steam circuit through the steam generator tubing and into the building containment through the wall of the vessel. This means that the turbine is contaminated and the atmosphere in the containment contains radioactive tritium. Once the cooling water has passed the alloy steels of the tubes, weld metal and internal cladding through a developed crack it will reach the ferritic outer wall of the vessel and of the header. When this occurs it results in the main pressure withstanding part of the vessel being subject to rapid corrosion from the boric acid in solution, as happened at Davis-Besse in the US, where it resulted in the large hole in the head. This was described as a "near-miss"!

The effect of the irradiation on the alloy steels is to induce an internal swelling leading to micro-cracks from an embrittlement and a loss of ductility. The production of the micro-cracks initiates SCC, so that it is thus enhanced by irradiation.

The mechanism of irradiation enhanced stress corrosion cracking is notwithstanding the subject of research. [4] López and Ferguson established that Inconel 690 is subject to hydrogen-induced SCC, but did not study the effect of tritium in the water. It is unlikely that the high temperature conditions in a PWR with boric acid in solution and with irradiation leading to the presence of tritium could be repeated in a laboratory, but test samples of Inconel 690 under stress and subject to corrosion cracked.

It may be that the use of Inconel 690 instead of Inconel 600 will extend the life of the PWR major components, but the eventual appearance of stress corrosion cracking in it must be assumed.

Ageing management

The response of the industry to what could be a catastrophic sequence of events is to introduce the science of ageing management. ("aging" in the US). IAEA updated its document "Assessment and Management of Ageing of Major Nuclear Power Plant Components" in 2007" (IAEA-TECHDOC-1557) [5]

Because the capital costs and initial construction carbon emissions are so great, the economic and climate alleviation status of nuclear power depends on extensions to the life-cycle, so that the capital and emissions can be amortised over the operational life, which has typically averaged around 25 years. In France, for example, some 34 of the 58 working reactors are nearing their 30 year inspection at which licenses for a further 10 years will be issued. It is hoped that many will gain further reprieves up until 60 years, the life now claimed for the Areva EPR.

Inconel 690 has been adopted for many internal components of the EPR and a 90 tonnes stainless steel neutron heavy reflector surrounds the fuel assembly in order to reduce the irradiation of the vessel wall. However, the irradiation of the vessel wall is but one factor and the boric acid, lithium hydroxide, with the released hydrogen and helium circulate the entire primary water cooling circuit, including the steam generator tubing. Accessible metal test specimens are located on the vessel inside in the cooling water, but unless these are contrived to be put under stress, they will not assess their resistance to SCC or irradiation.

In France the ASN will subject the Flamanville EPR to the same rigorous 10 yearly inspection and licensing routine, so the 60 years claimed operational life is subject to successive steps of ten years. [6]

An international consortium of nuclear organisations and generators are building a test reactor at Caradache in France which is expected to be operational in 2014. One of its tasks is to study the ageing of plant materials. [7] The experimental reactor will endeavour to simulate the progress of time by increasing the irradiation, stress and chemical loading on test specimens. It is debatable whether such tests are legitimate, but in any case the results will not be known until say 2020, some five years after the test reactor is commissioned in 2014.

In existing nuclear power plants, various instrumentation devices, employing techniques such as ultrasonics, have been applied externally to determine internal degradation and from the data time-material status plots have been constructed. The financial benefits of life extension are considered to be worthwhile, but the costs of major component exchange and the inconvenience may mean that they are not. The length of the operational life cycle is however basic to the financial analysis and to the claims for low carbon emissions so that the large emissions at the front-end can be levelled over the cycle, even though the emissions resulting from the replacement activity, the decommissioning and the final waste management feature during and at the end of the cycle.

The claims of reactor lives to be double those currently achieved must therefore be viewed circumspectly.

Spent fuel ponds

There is however one ageing problem unable to be remedied by exchange of components, viz. the spent fuel ponds. The internal spent fuel ponds are lined with stainless steel, because corrosive boric acid is added to the pond water to absorb neutrons from the interaction of the fuel elements. Spent fuel has to remain under water until it has sufficiently "cooled" to be removed and placed in dry casks.

Although the stress on the liner will be just that resulting from an hydraulic head, it will be irradiated and attacked by hydrogen and tritium from the boron in solution, so that cracks may appear. Once cracks appear through the liners, the concrete and reinforcing bars will be subject to corrosion leading to leakage and potential groundwater pollution. Also the level in the pond may lower to uncover the elements resulting in a meltdown and possible fire.

For any repairs to be carried out, the pond would have to be emptied and decontaminated. The ponds will have to retain their contents at least ten years after the operational life of the reactor and to avoid lethal radiation the elements must remain under a depth of water, water which has to be constantly filtered, cooled and circulated. The only solution is to have a standby pond ready to take the contents of the main pond in order for it to be emptied.

A pond's life continues well after the last fuel element is removed, so in the case that an EPR endures for its 60 years, the associated spent fuel pond cannot be accessed for repair for at least 70 years from the commissioning of the reactor.

Design rejection or modification

It remains to be seen what influence the GDA can actually apply. Its independence has been questioned by its close relations with Areva and EdF. It is politically unacceptable to turn down the EPR completely, but its design is supposed to be standard and modifications, like duplicating the spent fuel pond, would void its standardisation. The first EPR is yet to be commissioned and if there are no "teething" problems it will be a first in such a development. It is named as "evolutionary" because it differs little from its predecessors, the N4 and KONVOI, but it does employ different alloy steels and an internal neutron reflector, both variants needing time testing.

It is unlikely that should the GDA require unacceptable design changes for EdF's ventures, that the alternative Toshiba-Westinghouse design would be acceptable to it, though E.On and RWE appear to be open-minded. The AP1000 is as yet untried.

It was claimed that Sizewell B was more expensive than it need have been and its construction start delayed due to the previous NII design changes. One way out of the difficulty might be to insist that EdF builds a Sizewell C and D to the designs of Sizewell B. This would have the advantage of interchangeability of parts on the same site and the staff would be already suitably trained. However, the same ageing process required the replacement of the Sizewell B reactor vessel head in 2006 and to the plugging of a tube in a steam generator. HSE/NII assume that thermally treated Inconel 690 in the replacement head will offer a longer term resistance to SCC, but laboratory experiments show it is subject to it.

Geriatric design assessment

From the moment the first fuel charge in a reactor goes critical the ageing process begins. Over 200 major components in PWRs have been exchanged at huge expense and with loss of generation during the changeover. So ageing of nuclear reactors is generic.

Reactor life cycles are equivalent to the human life cycle in that ageing is progressive and sometimes needs vital organ surgery to maintain life. Some systemic and organ failures are catastrophic and cause premature death as nearly happened at Davis-Besse.

It is incumbent on the Generic Design Assessment to consider the validity of the ageing management processes applied by the generators and to consider whether the life-cycle claims thus reliant are valid. Also whether the exchange of alloy steels for those previously subject to near catastrophic failure are like to endure much longer.

Should the HSE/NII adopt an inspection regime similar to that conducted by the French ASN on a ten-year basis be applied in the UK?  The GDA may recommend such.

The GDA when constituted was expected to take until 2012 to complete its work and found difficulty in recruiting its experts; most of those in the UK are either retired or approaching retirement. The manning up of the team with French experts risks compromising the independence of the GDA, which it is supposed to maintain.

The prospective generators are impatiently awaiting GDA’s decisions and political pressure is on it to speed up its processes. It has already issued a favourable report on Steps 1 and 2 of its assessment -  it remains to be seen what its reaction to this article will be, if any.    

[1]  http://www.hse.gov.uk/nuclear/reactors/

[2]  http://www-pub.iaea.org/MTCD/publications/PDF/Pub1337_web.pdf

[3]  Steam generators arrive for San Onofre

[4]  http://espace.library.uq.edu.au/eserv/UQ:10054/Lopez_sif04.pdf

[5]  http://www-pub.iaea.org/MTCD/publications/PDF/TE_1557_web.pdf

[6]  http://annual-report.asn.fr/

[7]  http://new-jmtr.jaea.go.jp/contenpage/symp/materials/A-1.pdf

 * Health and Safety Executive/Nuclear Installations Inspectorate