Recollections of a UK Boiling Water Reactor

INTRODUCTION

This document is a summary of some thoughts and

recollections from operating experience with the

Steam Generating Heavy Water Reactor (SGHWR)

at Winfrith. Whilst the reactor systems were

complex and numerous, this document

concentrates on aspects that may be common to

issues that affect more conventional boiling water

reactors. SGHWR had many unique features such

as a rapid acting liquid shutdown system, direct

emergency cooling water injection and vertical

primary circuit pressure tubes, however, it was

basically a boiling water reactor (~330 MW(th))

with direct production of steam fed to a

conventional 100MW turbine alternator.

I hope the readers of this report and those hearing

the associated presentation will find the

information in it of interest and value.

My thanks go to the staff at Winfrith, both past and

present, who were colleagues at the time and have

helped me find some of the old photographs and

drawings that I have used.

BACKGROUND AND CREATION OF THE PROTOTYPE

In the early 1960s, the UK was looking for options for moving on from the first generation of Magnox

power stations which were just getting into their stride. The prototype advanced gas cooled reactor

started operation at Windscale in 1962, but other options were being considered. With an eye to

developing the export market, a design that could be “modularized” was looked for.

Out of this work the United Kingdom Atomic Energy Authority (UKAEA) developed the Steam

Generation Heavy Water Reactor design. The concept was radically different from previous power

reactors built in the UK in that a single pressure vessel was replaced by pressure tubes, and it was

light water cooled, direct cycle and heavy water moderated; however, it shared many characteristics

with the already proven CANDU design.

2

The Authority gained approval from government to

construct a medium size, electricity generating

prototype at its reactor development facility

recently opened at Winfrith Heath in Dorset. The

project involved a number of private sector power

industry partners including International

Combustion and AEI. Construction started in 1962

and commissioning in early 1967. The reactor first

went critical in September 1967, was first

connected to the local electrical grid system on the

24th December. It reached nominal full power for

the first time in late January 1968.

There was an extensive development programme

over the following years, but once the commercial

One of the steam drums development of the SGHWR concept was set aside

in the late 1970s, the plant continued to operate on

a semi-commercial basis, reliably generating electricity to feed both the rest of the Winfrith site and

the local electricity grid.

Rising costs, ageing materials and enhanced safety standards led to the decision to cease generation

in 1990 after 23 years of operation. The plant then entered a post operational clean out stage.

Since then, the plant has gone through periods of decommissioning interspersed with times of care

and maintenance when funding was low. Final decommissioning is now in hand with the objective

of the land reverting to heathland by 2023.

SGHWR BASICS

3

PRIMARY CIRCUIT The SGHWR primary light water cooling circuit was in two symmetrical halves, each having 52

vertical pressure tubes (with associated feeder and riser pipes), two primary circulating pumps and

one steam drum. The fuel assemblies (consisting initially of 36 pins filled with low enriched uranium

dioxide pellets and latterly with 57 pins) were suspended on hanger bars within the pressure tubes.

Force-circulated water was allowed to boil as it passed over the fuel with about 10% being converted

to steam. The steam and water were divided in the water separator/driers located within the steam

drums and then passed forward to the turbine alternator where the two circuits’ outputs were

combined.

Provision was made for a limited degree of superheating in eight separate pressure tubes in the

periphery of the reactor core, but development of fuel for these channels was never completed.

Feed water from the conventional feed plant was delivered to the ends of the steam drums.

MODERATOR AND REACTIVITY CONTROL Whilst some neutron moderation occurred within the light water coolant surrounding the fuel, the

majority of this function was provided by heavy water in a segregated circuit. The heavy water was

kept at near atmospheric pressure and at low temperature and was contained in a calandria through

which the pressure tubes passed.

Short term reactivity control was

achieved by raising and lowering the

level of heavy water in the calandria

and medium-term control was by

changing the concentration of boric

acid in solution. Long term control

was by differential fuel loading.

Reactor shutdown was ensured by

rapid dumping of the heavy water

into storage tanks, but reactor trip

systems triggered valves that

flooded 12 segregated tubes inside

the calandria with lithium borate

solution which shut the nuclear

reaction down within 2 seconds.

TURBINE AND BLED STEAM SYSTEMS The turbine, alternator and feed water systems were basically commercial systems adapted slightly

for the circumstances. The systems varied from those used at the time in coal fired plants in two key

respects: the turbine was designed for saturated steam and much of the feed water plant was

enclosed in concrete cells. These latter cells were required to allow for the likely low-level

contamination of the systems with radioactive material and to shield the normal working areas from

the associated radiation dose rates.

Due to their potentially radioactive content, non-condensable gasses arising in the condenser had to

be removed by electrically driven air pumps before being filtered and discharged to the atmosphere.

4

A feed water polishing plant was provided that treated 100% of the feed flow to limit the level of

impurities. This used powdered ion exchange resin on pre-coated leaves that filtered and treated

the feed water flow.

Two 100% condenser extract pumps and two 50% electrically driven boiler feed pumps were

provided. The feed system included conventional low-pressure and high-pressure feed heaters and

a deaerator tank, but all of these were housed in concrete cells.

REFUELLING The design included the capability for on load refuelling using a carousel arrangement in a large

pressure vessel that could be connected to the primary circuit at the reactor charge face. However,

this system was problematic and was never commissioned. Instead, an off-load system was used

that operated at atmospheric pressure during reactor maintenance outages.

OPERATING EXPERIENCE

The following notes are personal recollections of operation issues and problems associated with the

working of the reactor and its systems. It concentrates on those issues that are associated with it

being a boiling water reactor and does not

explore aspects of the heavy water, pressure

tube or reactor control systems. The

operational problems were common to those

experienced with contemporary boiling water

reactors.

By nature, many of the problems related to

the operation of the systems as a whole, and

cannot be treated in isolation; however, they

are summarized below by plant system, from

steam, through feed water to primary circuit.

STEAM SYSTEMS

Live steam, direct from the reactor primary circuit, came with a significant “load” of nitrogen-16.

This short-lived radionuclide is generated by neutron absorption in oxygen in the primary circuit

water and is carried forward to the turbine in the steam. This was recognized as an issue at the time

of design and a large shield wall was provided between the HP cylinder of the turbine and the

instrumentation area where routine access was needed. The following diagram shows the radiation

dose rates that were generated around the turbine. Whilst this proved successful, there remained

the problem of on-load maintenance of the turbine. Problems with the turbine governor systems or

control valves would either require shutdown or maintenance staff to work in high dose rate areas

to resolve the issues.

Turbine floor during reactor operations

5

Radiation dose rates on the turbine floor

Being a direct cycle plant, with limited steam drying capability (but note superheating proposals

referred to above), steam at the turbine stop valves was wet/saturated. This caused erosion of the

turbine blades which required significant work during reactor outages. If the turbine was not

scheduled for partial dismantling, then the work was in difficult and unpleasant conditions with the

added problems associated with low level radioactive contamination.

Problems in the early operational days with fuel can failures (see below) and “tramp” uranium on

new build fuel assemblies led to the presence of low levels of uranium in the primary circuit. As a

consequence, fission products could be found in the steam circuits. Whilst this caused problems

with maintenance (see below) it could also cause short term airborne issues when there was a plant

steam leak. Whilst, on a conventional plant, a steam leak could be either lived with in the short term

or easily repaired, at SGHWR, the

associated airborne xenon and krypton

would cause operational and radiation

protection issues.

Both activated corrosion products and

longer-lived fission products were carried

forward in the steam systems. This

resulted in contamination of the turbine

and bled-steam systems. Whilst the levels

of radiation and contamination were never

high enough to prevent hands-on

maintenance, they did make the operations

Turbine floor during overhaul

6

more complex. The operations had to be carried out in radiologically controlled areas with

contamination control requirements, exposure control and waste management issues.

Of course, corrosion and erosion products were also carried forward through the feed system and

into the primary circuit where they were exposed to neutron irradiation to become radioactive and

cause further issues as discussed below.

FEED HEATERS These were conventional items of equipment,

but were located in the feed heater cell to

allow for the control of the radiological issues

anticipated. Whilst this was generally

successful, maintenance again was difficult.

Being within a cell, major work (like removing

a heater shell) meant lifting the cell roof on

the turbine floor. Not only did this “break-

through” the normal contamination barrier,

but it could only be done when no

maintenance was being undertaken on the

turbine, as the cell roof was the lay down

area for this latter work. Laying out feed heater shells

The initial material for some of the feed heater tube bundles was found to be excessively prone to

erosion and corrosion resulting in difficulties with primary circuit impurity control. These were

replaced by more appropriate materials in mid-life.

FEED SYSTEM As above, much of the feed system suffered from a degree of erosion and corrosion; all of these

causing problems with primary circuit impurity control. However, as with other boiling water

reactors of the time, extensive use of Stellite as a valve seat material caused particular problems.

Stellite is high in cobalt and any material released into the primary circuit became activated in the

neutron flux as cobalt-60. This became the dominant aspect of radiological protection during the life

of the reactor. Most of the staff exposure was due to this material. Whilst much of the Stellite was

replaced early in the reactor life with other materials, the damage was done and the material is a

significant component of the waste materials even now after 28 years (over 5 half-lives).

Condenser tube failure was a significant threat to the primary circuit conditions and great emphasis

was placed on the response to such events. Major leaks and the associated rise in feed water

conductivity would require a manual reactor trip.

The feed system was a significant source of oxygen ingress. This was a significant contributor to

corrosion and its associated problems. As such, significant efforts had to be made to minimize

ingress and to reduce the dissolved oxygen levels by extensive feed water dosing.

In the plant design, a decision was made to adopt 100% condensate polishing using the Powdex

system (combined filtration and ion exchange). The system allowed the clean-up plant to be placed

7

upstream of the deaerator, rather than at the condenser outlet more typical of conventional ion-

exchange approaches.

The Powdex process is a powdered ion-

exchange resin that is pre-coated onto filter

leaves. Whilst this was generally very

successful, it did generate a significant

volume of radioactive waste. This material

has since been encapsulated in cement in

stainless steel drums and is due to be

disposed of to LLWR this year (2018).

Additionally, the Powdex plant had to be

operated with a significant degree of care.

Any breakthrough of the resin into the feed

system would cause phosphate

contamination which then led to primary

circuit chemistry problems and higher levels of radiation on the turbine systems.

PRIMARY CIRCUIT As noted above, contaminants in the feed water caused, not only materials issues in the

components, but radiological difficulties when it came to maintenance work. Blowdown of the

primary circuit was used to limit the concentrations, but, because of its radioactive content, this

blowdown had to be routed into the feed system rather than to waste. The design discharged the

water into the low-pressure feed system pipework immediately upstream of the polishing plant.

Unfortunately, the design was imperfect, as the flashing of the high temperature blowdown water

caused vibration issues, particularly with the polishing plant itself. Latterly, a blowdown cooler was

installed that resolved this issue.

Impurities in the feed water concentrated in the primary circuit. This plated out in all areas, but

when this occurred on the inside of the pressure tubes and the surface of the fuel, the material

would then become activated in the neutron flux. Subsequently, it became remobilized and resulted

in radioactive contamination of other parts of the primary circuit, steam systems and feed water

pipework. This impurity deposition reached levels in the early stages such that fuel pins were

affected and there were some failures allowing escape of fission products into the primary coolant.

This was addressed both by improved feed water control and by the adoption of a 57-pin fuel

assembly design.

Particular problems arose with the presence of “dead legs” in the primary circuit where radioactive

material would concentrate and form a radiation hot spot.

In common with other light water reactors across the world at the time, trials were undertaken with

primary circuit decontamination techniques in an effort to reduce radiation exposure for the

maintenance workforce. Initial operations used the TURCO process, but this proved too corrosive to

the key primary circuit components and would limit the number of times the process could be

conducted. It was, however, successful in the object of reducing primary circuit radiation dose rates.

Subsequently, UKAEA worked with the Central Electricity Generating Board to develop the Low

Oxidation State Metal Ion (LOMI) process. This less aggressive approach resulted in very good

decontamination performance with acceptable levels of circuit corrosion. This process was later

Powdex unit

8

refined with the use of a pre-treatment phase using nitric acid and potassium permanganate to

enhance the dissolution of chromium. The disadvantages of such systems are the increased volume

of waste generated and the presence of heavy metals in plant effluents. The LOMI process

continues to be used in the nuclear industry.

OPERATIONAL CONTROL OF RADIATION EXPOSURE AND RADIOACTIVE CONTAMINATION The use of a direct cycle approach (combined with a lack of experience in the associated issues) led

to a significantly higher level of radiation exposure to the workforce when compared to a gas cooled

reactor or PWR.

In the early years of operation, radiation protection was not as demanding as it became following

the introduction of the Ionising Radiations Regulations (IRRs). Members of maintenance and

operations staff were exposed to levels of radiation exposure that would today seem

unacceptable. The exposures of some key members of the maintenance team with specific skills

and experience were particularly difficult to manage.

After the introduction of the IRRs, additional efforts were made to reduce exposures and this,

together with growing experience and improved decontamination meant that both individual

and total exposures were

significantly reduced.

Low level contamination of the

steam and water circuits meant that

most maintenance had to be

undertaken in radiologically

controlled areas – over barriers and

with separate active area clothing.

This meant not only low-level

exposure to the workforce, but loss

of time and effort in crossing

barriers etc, the requirement for

monitoring effort and the

generation of amounts of secondary

wastes. In particular, working on feed system, air

pumps, feed heaters etc in enclosed spaces results in a significant loss of efficiency

and makes working areas more cramped and unpleasant.

OTHER OPERATIONAL ISSUES The SGHWR was a very rewarding plant to operate, however, most of the stories about reactor

kinetics, control systems, start-up and shut-down, and steam system dynamics don’t relate directly

to it being a boiling water reactor, but to its use of heavy water as a moderator and the associated

systems. However, the dynamic links between reactor power, steam pressure in the drums and

mechanical governor in the turbine did lead to some interesting excursions when the grid frequency

changed rapidly. Significant changes were made to the automatic control systems to improve

system stablility.

Air pumps cell during maintenance

9

Also notable was the steam void reactivity co-efficient. When the reactor was designed, it was not

known whether or not this co-efficient would be positive or negative and additional mechanisms

were included to allow this to be adjusted. It worked out in practice to be significantly negative – i.e.

the reactor power would fall if the steam generation rate rose, giving negative feedback on most

transients. This was a significant factor in favour of this design over the RBMK pressure tube reactor

of the USSR which has a significantly positive co-efficient – a significant factor in the Chernobyl

accident.

SUMMARY AND CONCLUSION

SGHWR had a number of design and operational shortcomings associated with being a direct cycle

plant; however, it was remarkable that the facility generated successfully for 23 years bearing in

mind the speed and economy of its design and construction.

Around the world, much experience has been gained from operating boiling water reactors of all

types. International collaboration and the embedding of the lessons learnt in the design,

construction and operation of an advanced design has every prospect of providing safe and cost-

effective power for current and future generations.

ABOUT THE AUTHOR

John Lindsay joined the UKAEA at Winfrith in 1983 after gaining a degree in Electrical Engineering at

Nottingham University and spending five years with British Rail. After undergoing three years

training and gaining sufficient plant experience, he was authorized as a Shift Manager. When on

duty, he was responsible for the plant operations and management of the operations team of

operators, supervisors and health physics staff. He remained in this post for the latter years of the

plant’s operations and the first two years post final shutdown.

Once the major hazards (spent fuel and heavy water) were removed from the plant, he left the

UKAEA to join the Nuclear Installations Inspectorate, (now the Office for Nuclear Regulation). After

23 years with the NII/ONR he returned to Winfrith working for Magnox as safety manager for the

decommissioning of the adjacent Dragon high temperature reactor plant.