The monitoring of steel and concrete pressure vessels is essentially similar with due regard being paid to the fundamental design differences. In summary, the pressure-tight component of the concrete vessel (Fig 3.77) is a thin steel liner (of the order of 12.7 mm) which is surrounded by concrete. The concrete is con­strained by steel tendons which are tensioned to carry the stresses imposed on the structure when gas inside

the vessel is at pressure. In contrast, the stresses of a steel vessel (Fig 3.78) are carried by the pressure — tight membrane itself which, as a consequence, is typically 75-100 mm thick.

The boiler units in a concrete vessel are integral, whereas those for a steel vessel are external and con­nected to the vessel via large diameter stubs, welded to the vessel.

Large diameter penetrations are provided in con­crete vessels to carry the gas circulators, but these are more easily constructed and, as a result, are of simple design. Typically, the large penetrations in a concrete vessel number half those needed on a steel vessel.

Each type of vessel carries a large number of smaller penetrations, the majority of which accom­modate the access standpipes. Those on steel vessels are directly welded, which introduces a complication in design, but those on concrete vessels are more easily accommodated.

In each case, thermal insulation is associated with
the vessel. On concrete vessels, the insulation is in­ternal to reduce the transmisson of heat to the liner whilst on steel vessels it is on the outside of the vessel.

In general, it is the detection of a deviation from normal of a parameter that is used to monitor the vessels. For example, thermocouples are installed on or near the vessel material so that temperature changes can be monitored. On steel vessels, the metal tem­perature has to be maintained above a certain mini­mum when the reactor is shut down, to avoid metal­lurgical changes in the steel. For concrete vessels, the concrete temperature must not exceed the level at which it begins to break down due to the loss of ‘chemically-bound* water. Temperature monitoring is important during load changes and, in particular, during a start-up or shutdown procedure when tem­perature differentials of the vessel structure place con­straints on the operation. The detection of hot spots on the concrete vessel liners would indicate some fault

INTERFACIAL OXIDE TENDS TO SIND COMPONENTS

TO THE BOLT WHICH MAY ACT AS A DOWEL

IF THE BOLT FAILS AT OR OUTSIDE. PLANES A AND 0

Fig. 3.74 Effects of iruerfacia! oxide

Fig. 3.73 Effect of oxidation on a low silicon
thin-section material

or deterioration of the insulation, and may prompt remedial work at the next biennial shutdown.

The loss of coolant from either system may be indicative of some incipient failure. Thus routine checks of loss of coolant are carried out in the vessels to monitor this situation. The simplest procedure is that of loss of pressure over a given period of time. In this case, coolant-consuming operations such as refuelling are discontinued and the whole system is isolated as far as possible so as to eliminate losses from auxiliary systems. The loss of pressure is then determined over a number of hours and the result compared with that normally experienced.

Helium injection is also used to determine coolant loss. A known weight of helium is injected into the coolant circuit and its concentration is established. After a set time, the coolant pressure is returned to
its original level and the helium concentration is re­determined, thus permitting the losses to be calculated.

Determination of the CO2 concentration in a re­actor ventilation system is also used to monitor gas leakage, although in this case the loss will generally be from valves and seals associated with the reactor.

Strain gauges are embedded in the concrete of the pressure vessel and sometimes attached to the metal components. Routine checks of this instrumentation are carried out to monitor the vessel behaviour. The long term results from the gauges are received at in­tervals to ensure that the material is reacting in the manner predicted.

Whether constructed from concrete or steel, the total reactor structure weighs several thousands of tonnes and this weight is supported by the underlying rock strata. Inevitably, this results in some settlement of the structure together with the possibility of tilt. The main concern arising from uneven settlements is associated with the structural strength of the core and its supports, together with the uninhibited pass-

age ol control rods under the influence of gravity. Periodically a levelling survey is carried out in the reactor structure to determine settlement and tilt. The results are compared with previous years and predic­tions made to end of life (30 years). The operation ol the reactor beyond this life’ would be the subject ol a detailed assessment and new benchmarks estab­lished lor its continued operation.

Stressed steel cables are the restraining feature of concrete vessels and these are monitored throughout tbe life of the vessel. The residual load in a repre­
sentative number of cables is measured and the results plotted on a log/time base. A ‘best fit’ regression line is applied to the data, accumulated year by year, and a forecast is made of residual stress at end of life to ensure that the cables are performing as expected. A representative sample of the cables are withdrawn each year to check for possible physical and corrosion deterioration.

It has already been mentioned that there are a large number of penetrations in a steel vessel. The pipework is attached to the stubs at these penetra-

KEY

EFFECT OF FAILURE OF INDIVIDUAL COMPONENTS ON WHOLE STRUCTURE INVOLVES PROBABILITY ANALYSIS AND REDUNDANCY CONSIDERATION

REPRESENTATION OF CHARGE PANS

Fig 3.T6 Pan 2 oxidation assessment for the overall structure

bellows have been inspected using ultrasonic and radiographic techniques to determine the presence of any incipient cracking. Where indications are found, these are generally round-out and re-welded. It must be remembered, that the latest examination techniques are further advanced than those used at the construc­tion stage, consequently indications not detected at the construction stage are tending to be located even though their significances may not be great. Indeed, bellows units base been removed from reactors and hydraulically over-pressurised to demonstrate that the units were sound despite indications found.

Water cooling systems are installed in concrete pres­sure vessels to maintain the concrete at a tempera­ture below about 60°C. The system consists of cool­ing water pipes attached to ribs welded to the steel liner. Water passing through these pipes extracts heat from the liner and the surrounding concrete. The heat results from thermal conduction of the coolant gas through the liner (a layer of insulation on the gas — side of the liner reduces this to a minimum) and by radiolytic heating. The heat in the water is removed by exterior coolers. The make-up to the cooling water system is monitored for excessive usage and the ex­terior of the reactor and cable ends are examined for the presence of water. These procedures are to check the absence of leaks in the embedded pipework. Rou­tine heat balances of the cooling system are performed to ensure that no gross deterioration of the internal insulation has occurred. The cooling system is usually divided into accessible sections so that individual areas can be monitored. Chemical control of the system is of paramount importance to ensure that no corrosion of the pipework can take place and that the gaseous products of radiolysis are removed. The performance of the external plant (pumps, etc.) is continuously monitored, since the loss of coolant to the cooling system cannot be tolerated.

The monitoring of pressure vessel integrity is a con­tinuous process and is carried out day-by-day and, indeed, hour-by-hour by the operating staff. Parti­cular attention is paid to differential temperatures dur­ing load changes such as start-up or shutdown.

The branch is responsible for ensuring that safety considerations are properly taken into account in the

siting, selection, design, construction and commission­ing of future nuclear stations. It is also responsible for the formulation of design safety criteria for use in the design of future systems.

The branch is led by the Head of Nuclear Safety Development. He is assisted by the three section man­agers responsible for PWR Projects, Analysis and Physics, and Future Systems.

2.10.1 Medical Branch

The branch is responsible for encouraging the pre­vention of ill health and the promotion of good health in all employees of the CEGB. The branch will also seek to prevent ill health arising in any other persons as a result of the CEGB’s activities. The branch provides a broadly based occupational health service, ranging from the routine health surveillance of certain staff to specialised advice on highly technical matters of relevance to the CEGB.

The activities of the medical branch are directed by the Chief Medical Officer who is responsible for the provision of all Occupational Health Services within the CEGB. The policies are implemented through its Area and District Medical Officers, HQ Medical Officers and nurses.

In this section the fundamentals of radiological pro­tection are discussed with reference specifically to CEGB practices, particularly on nuclear licensed sites. However, these fundamental principles obviously ap­ply to all radiation work, wherever it may be en­countered.

Appendix A to this chapter discusses the nature of radioactivity and the units employed in radiolo­gical protection. The purpose of this section is to describe the most common types of radiation encoun­tered, how the CEGB applies various protection meth­ods to ensure that personal radiation doses are kept as low as is reasonably practicable and that dose limits are not exceeded.

Radioactivity has existed since the formation of the planet and varies in its intensity around the globe. Natural radiation was first discovered in 1896 by Bec — querel who observed the blackening of photographic emulsions in the vicinity of uranium compounds. This phenomenon is still utilised today in the form of the film badge dosemeter to assess the radiation dose received by a worker. One year earlier than Bec — querel’s discovery, Roentgen observed a similar effect on photographic plates when these were placed near to high voltage equipment; these radiations were later christened X-rays. Since then, work by other famous scientists, notably Rutherford, the Curies and Chad­wick (who discovered the neutron in 1922), have all added to our knowledge of the physics of radioactive decay. As noted in the introduction to Section 4.1 of this chapter, it also became apparent in the early years that radiation could cause biological damage, which ultimately led to the deaths of many of the early workers. This fact led, in turn, to the develop­ment of radiation protection standards and the foun­dation of the International Commission on Radio­logical Protection.

In 1938, the first nuclear fission reaction took place, this being the splitting of the atoms of heavy ele­ments, e. g., uranium and plutonium, giving a release of energy. In 1942 the first fission reactor was op­erated in a converted squash court at an establishment in Chicago, developed by the scientist Fermi. This of course led on to the development of the atomic bomb, but more constructively, the civil nuclear power programmes throughout the world.

In order to ensure that a nuclear power station is operated within the limits and constraints assumed in the safety case, it is necessary to provide the op­erator with clear concise definitions of these limits and constraints. It is essential to impress upon the operator the overriding importance of compliance with any instructions which embody these limits and constraints.

Notwithstanding any instructions the licensee may give to his employees as a responsible operator of a nuclear power plant, the HSE requires as a condi­tion of the site licence that such instructions are drawn up. The licence condition also indicates the statu­tory requirements for making any changes to these instructions.

There are three separate but closely related docu­ments or sets of documents specified in the licences for the latest stations, whilst for the earlier stations there were only two.

The first document contains the Operating Rules. These specify the actual limits in terms of tempera­ture, gas pressure and power to which the reactor may be operated and the principles covering the degree of essential plant availability required. The Operating

Rules can only be changed with the explicit approval of the HSE after the licensee has followed his own procedures for agreeing such changes internally. It is emphasised that the licensee has no authority for making any change to the Operating Rules, which would lead to a relaxation of the limits or constraints, under any circumstances. Clearly if as a result of some investigation or operating experience a limit was found to be insufficiently rigorous, the licensee would be expected to act accordingly.

In order to change or suspend an operating rule, the station manager is required to prepare a submis­sion for the Nuclear Safety Committee. This submis­sion explains the need for the change, provides the necessary safety arguments supporting the change and, if appropriate, details additional precautions which may be necessary to ensure compliance with the safety principles. If agreed by the Nuclear Safety Committee, the submission is sent to the HSE by the Director of Health and Safety for their agreement with a request for the necessary formal consent.

If the requirement for change is urgent, agreement may be given by the Director of Health and Safety on behalf of the Nuclear Safety Committee. It is still necessary however to obtain the consent of the Health and Safety Executive.

The second document contains the Identified Op­erating Instructions (IOIs) and details the interpre­tation of the limits contained in the Operating Rules in terms of the actual measurement of parameters from the plant. It also gives details of the constraints on plant availability amplifying the principles em­bodied in the Operating Rules. The IOIs are a recent innovation brought about by the increase in complexi­ty of the AGR plant and consequently of the safety case, compared with the magnox reactors. Because the alteration of the Operating Rules is a lengthy process involving finally the consent of the HSE, it was agreed that the detailed interpretation of the Operating Rules could be relegated to a document requiring a less formal method of alteration. Indeed it is these details which, because of operating experience or of minor plant faults, are most likely to require amendment. Such amendment, of course, may only be carried out if the limits and principles in the Operating Rules are not violated. The amendment of the IOIs requires only the agreement of the head­quarters departments and divisions of the CEGB. The HSE is informed of the change within 14 days of its implementation; their explicit consent is not required. It is likely on the latest stations that in future the HSE will require to be informed of changes to the IOIs immediately before implementation.

The Plant Operating Instructions (POIs) form the final level of documentation directly related to the Operating Rules. They are the detailed instructions on the operation of the station as a whole. They give the step by step instructions on, for example, starting up and shutting down the unit, refuelling, operation

under fault conditions, etc. It is required that the POIs should be consistent with the Operating Rules and identified operating instructions and for this rea­son all changes are furnished to. the HSE within 14 days of their implementation. The changes themselves are the subject of a procedure involving only the station staff and the agreement of other organisations is not required.

The format of Operating Rules has remained very similar for all stations. The detail varies to take ac­count of a particular plant design and its mode of operation. There are sections covering:

• Fuel elements — clad temperature limits.

• Pressure vessel cooling system — setting of safety relief valves.

• Circuit activity — burst can detection.

• Fuel handling.

• Essential supplies — availability of plant and of consumable supplies such as CO2 or diesel fuel.

• Control and safety circuits — trip settings.

• Carriage of nuclear matter on site.

• Moderator.

• Impurities in the gas circuit.

• Structural temperature.

Each rule or set of rules within a section usually has associated with it a section in the identified operating instructions. It is a general principle that only those limits and constraints which are under the direct control of the operator are included in the Operating Rules. If operation outside a particular limit or with a prohibited plant state is prevented, either by the inherent design of the station or by the use of safety interlocks, an operating rule is unnecessary.

A1 Radioactivity

The nature of radioactivity is such that the atoms of a radioactive substance change from one chemical element to another. During this change process, or decay, radiation of some form is emitted, and the original parent atom is transformed into a new daughter atom. This daughter atom may itself be radioactive and the disintegration process continues until a stable non-radioactive atom is formed.

The rate of change of parent atoms to daughter atoms is known as the activity of the parent substance. The unit of activity is the becquerei (Bq) and is defined as that amount of radioactive material giving one disintegration per second.

Because a radioactive substance is decaying from one element to another, it follows that the activity of the parent will decrease with time as the number of atoms available for decay decreases. The time taken for the activity to fall by one half of some original value is known as the half-life, and is a constant for any particular isotope of an element. Thus, knowing the activity of a nuclide at a particular time zero, the activity of the substance at a time r may be calculated from the equation:

A = Ao exp (-f)

where A is the activity at time t

Ao is the activity at time r = o, is the decay constant of
the nuclide and is equivalent to loge(2)/hatf life.

A2 Absorbed dose

The radiation being emitted by a radioactive source will ultimately interact with the material it meets. The type of interaction depends on the type of radiation and its energy, but whatever processes are involved this leads to an energy loss of the radiation and thus an energy deposition in the material. The amount of energy deposited per unit mass of material is known as the absorbed dose. The unit of absorbed dose is the gray (Gy) and is defined as the absorption of one joule per kilogram of material.

A3 Dose equivalent

In radiation protection, the primary interest is in the effect of radia­tion in humans and it is not necessarily the case that the biological effect is uniquely dependent on the absorbed dose.

Different types of radiation differ in their biological effectiveness for the same amount of absorbed dose. A measure of this effec­tiveness is the queiity factor (Q). The product of the quality factor, a dimensionless quantity, and the absorbed dose is known as dose equhraient. The unit of dose equivalent is the sievert (Sv), and has the same dimensions as absorbed dose.

The latest values of Q for the different radiations are:

X, gamma, electrons 1

Neutrons 10

Singly-charged particles (e. g., protons) 10

Multiple-charged particles (e. g., alphas) 20

A4 Committed dose equivalent

If a radioactive substance is taken into the body it will be treated by the biological systems in the same way as the stable form of the substance. The material will thus be transported to various locations in the body, for example, strontium-90 will go to the bone; iodine — 131 to the thyroid gland; tritiated water will be distributed throughout all of the body water.

The substance will also be excreted by the body, as with the stable form, but it will also be subject to radioactive decay. Therefore, a combined reduction process occurs which will have a radioactive half-life of the nuclide present and a biological half-life of the stable chemical substance. The combination of these is known as the effective half-life.

At the time of intake, not ell the radioactive disintegrations wilt have taken place and, depending on the effective half-life value, it may be some time before the material may be regarded as not being present. Prom the time of intake to this later time, that part of the body being irradiated is thus committed to receiving a dose.

The committed dose equivalent is defined as the dose received over 50 years from the time of intake and has the seme units as dose equivalent, i. e., sieverts (Sv).

It is essential that fuel being used in the reactor is able to withstand the onerous conditions imposed upon it both during the on-load charge/discharge process as well as in normal service. Demonstration of fuel element integrity is therefore a vital part of safety cases which relate to refuelling. For this rea­son, several important checks need to be made on new — AGR fuel elements before they can be declared fit for reactor use. These include careful visual in­spections, eddy current tests, a pressure test and en­richment check, and each is conducted at different stages following initial box-opening. The relevant parts of the fuel route which relate to the handling of new fuel are the fuel store, box opening room, fuel inspection room and the new fuel ceil (NFC) itself, in which stringer assembly takes place. The NFC is a shielded facility situated next to the fuel inspection room from which the fuel required for stringer as­sembly is transported. It is connected to the central block of the charge hall above it by an access hole to which the charge machine can be coupled. In the cell, a fuel assembly tube containing an adjustable platform is positioned below a glovebox, which is used for the introduction of fuel and other stringer com­ponents. The assembly tube is long enough to accom­modate a completed fuel stack. Figure 3.54 shows a schematic layout of the relevant parts of the fuel

Fig. 3.54 Schematic arrangement ot new [uel
handling facilities

Each of the various fuel element checks, tests and
inspections are carried oui at different stages before
stringer assembly can take place.

route to clarify the sequence of events which takes place prior to stringer assembly. The layout relates only to the fuel route at Hinkley Point B, although similar facilities exist at the other AGR stations.

Visual inspections Before detailed examinations are made of the genera! condition of elements following transportation, individual identification checks are car­ried out. Once the fuel transit box lid has been re­moved in the box opening room and prior to cutting open the polythene bag containing the element, pre­liminary checks are made that the fuel element iden­tification code is in accordance with that specified on delivery paperwork in order to ensure that the element has been correctly numbered and packed. This particular check also facilitates verification that the appropriate type of fuel element (i. e., enrichment and number of ASGs) has been chosen for fuel stack as­sembly. Following a top-end eddy current test to check for damage sustained during transport, more detailed inspections are made in the fuel inspection room, in which a mirror-bench and an illuminated rack are provided. Here a cross-check is made that each pin top-end cap enrichment lettering matches that con­tained in the fuel element code on its outer sleeve, in order to verify that the pins have been properly assembled into the correct element during manufac­ture. Following this a careful examination of the element’s physical integrity begins. Firstly the upper region of the outer sleeve is checked for scratches and chips, and the top brace is examined to ensure that no distortion, looseness or excessive pin-brace clearances exist. The top end-caps are surveyed once more for any obvious signs of damage or defects. With the aid of a bench light the inside surfaces of the element are then examined in order to check pin alignment and to establish that no coolant flow ob­struction exists, following which the lower surfaces of the element are viewed by suspending it above the mirror-bench. As before, inspection is aimed at lo­cating any graphite cracks or chips which may be present. The grid can then be checked for tightness and distortion and the lower ends of the fuel pins examined to ensure that they are satisfactorily secured to it.

Eddy current testing It is possible that minor cracks or imperfections can be present within the graphite sleeves which are not detectable by visual inspection. Thus both ends of the sleeves are tested at different stages before stack assembly, using specially provided eddy current scanners. These devices detect element induced signal anomalies created by structural weak­nesses in the graphite.

Proof pressure testing Provided an element has passed all the visual examinations outlined, a pneumatic pres­sure test rig is used to create a large pressure dif­ferential across the sleeves. This encourages the pro­pagation to the point of failure of any inherent de­fects which have so far remained undetected. The rig provides a test pressure drop of about 6.5 bar and is calibrated to clearly indicate either ‘pass* or ‘fail’.

Enrichment monitoring A fuel enrichment monitor, consisting of a detector and counting unit, gives a final check on element enrichment as it is assembled in its fuel stack within the NFC. The monitor is fre­quently calibrated using an element of known enrich­ment, and in subsequent use the equipment compares individual readings with a ‘band of acceptance’ for the appropriate enrichment being monitored.

Rejects

From time to time elements are ‘rejected’ as a con­sequence of failing to pass the various tests and in­spections outlined. Under these circumstances the fuel is put to one side in clearly labelled boxes and is sub­jected to a separate ‘joint inspection’ involving BNFL as well as CEGB inspectors. This can often result in some elements being returned to BNFL Springfields for refurbishing or repair.

During the fission process some of the neutrons pro­duced are captured by uranium, forming a group of radionuclides known as the actinides which have higher atomic numbers than actinium whose atomic number is 89. The significant actinides and their half-lives are given in Table 4.3.

Actinides are significantly different from fission and activation products in that they emit alpha particles during radioactive decay and have long half-lives. A notable exception is plutonium 241 which emits beta particles. In quantity, the actinides produced are si­milar to the fission products.

Since heat is removed from the reactor core almost exclusively by the ССЬ coolant, any reduction of the mass of gas passing up the fuel channels will result in an immediate increase in core temperatures.

If a breach occurs in the pressure circuit, the coolant gas will be lost at a rate depending on the size of the hole. In the case of a reactor with a steel pressure vessel, it is assumed that any of the main ducts be­tween the vessel and the boilers may fail. This leads to a double breach of the pressure circuit and the possibility of major damage both to the structure of the core and to essential plant situated elsewhere on the site. Furthermore, if the breach occurs in the duct on the outlet side of a gas circulator, total stagnation of flow up the fuel channels must be assumed during part of the period of the depressurisation. Finally, when the circuit pressure falls to atmospheric, unless the breach can be closed, air will enter the pressure vessel. For these large breaches of the pressure cir­cuit, the studies are done in two parts; firstly, the short term transient is investigated as the cooling ca­pability is lost and the reactor is tripped by the pro­tective equipment; secondly, the long term behaviour is examined when air is assumed to replace the CO2. In the second case, the options open to the operator to limit the effects of the accident are investigated. Although pressure circuit failures leading to smaller hole sizes are less dramatic, all need to be studied to ensure that the protective equipment is adequate. Since the potential maximum hole size in a concrete pressure vessel is very much smaller than for a steel vessel, the investigation of the long term transients are much simpler.

In view of the specialist nature of the department’s work, it follows that a high degree of expertise is required. Thus the Health Physicist is usually a grad­uate in a science subject and will have spent many years working within the discipline. Working under the health physicist will be the assistant health phy­sicist and number of health physics assistants together with about thirty-five industrial grade staff. In ad­dition, the work is very closely linked to that of the station’s Medical Department.

The assistant health physicist and health physics assistants will typically be of graduate status, particu­larly at the higher grades. In recent years, more per­sonnel are being recruited with specialist health physics qualifications which, in earlier years, were not avail­able. The industrial staff comprising foremen, moni­tors, change room attendants and possibly labourers, will have been specially trained in most aspects of operational health physics with some qualifying in City and Guilds in the subject.

The supervisory staff do not work on a shift sys­tem, but there is such a pattern for the industrial staff. However, the health physicist, the assistant health physicist and at least two other health physicist assistants, are available on a standby rota so that shift staff may call upon them should the need arise. In addition, the standby rota fulfils the requirement for a professional health physicist to be always available for emergency purposes. In order to be on the standby rota, the health physicist must be ‘Accredited’ in ac­cordance with the requirements of the Safety Rules (Radiological). This means that the person in ques­tion must have sufficient knowledge of radiological protection and the plant to be able to proffer his or her advice when the need arises. The accreditation is thus site specific and in order to obtain the qua­lification candidates must convince the Health and Safety Department of their suitability, which involves an interview.

Depending on the local management structure, the health physicist reports directly to the Station Manager or indirectly through a Technical Services Manager, who will also have charge of the Chemistry and Re­actor Physics Departments. In the latter case however, the health physicist is guaranteed direct access to the station manager on matters of radiological safety.

The problems of providing the necessary ‘on job’ training for the operator on a base-load station make simulation facilities essential, and these have been provided in the CEGB from the beginning of the nu­clear programme. Basic generic simulation was pro­vided for the magnox reactors up to 1987 by means of an analogue computer-driven simulator (see Fig

4.11) representing the major operating parameters at each of the five planes in a single fuel channel, thus demonstrating the operational effects on detailed tem­perature and flux patterns.

Nuclear training simulators range from a basic prin­ciples training simulator, a large scope magnox simu­lator, to the more complex AGR simulators (see Fig

4.12) with complete replication of the station control room. Although the use of simulation techniques is important, they only form part of the overall training required by an operations engineer. For a student to obtain the necessary benefit from the simulator he must have a good knowledge of the fundamentals of plant kinetics. It is also important that simulator training is given in varying degrees of complexity, to avoid a mechanistic approach to dealing with plant problems. Throughout the training programme, there­fore, the emphasis is on the development of an under­standing of the principles, the systems, and the system — to-system interactions.

5.9.1 Nuclear training audits

Following the issue of the nuclear training specifi­cation, and the production of the station training document by each nuclear station, a biennial national training audit was instituted by the Director of Per­sonnel. The audit team, comprising managers from nuclear power stations, headquarters departments and NPTC; visits each station and Berkeley Nuclear Lab­oratories to audit the training function against the location training document, reporting on its findings to the Director of Personnel.

5.9.2 Nuclear power training advisory committee

This is a group of managers from various parts of the CEGB, and including a representative from SSEB. The committee maintains a watching brief on the nuclear training function and advises the NPTC Man­ager and the Education and Training Branch Manager on the content of courses, future training needs and equipment, the current nuclear operational require­ments, and the methods of assessing the competence of operating staff on the completion of training.