The Nuclear Administrative Memoranda are a com­pendium of documents giving step by step instructions and administrative procedures for the preparation and assessment of safety submissions. The NAMs cover changes to the Operating Rules, identified operating instructions, plant operating instructions, modifica­tions to plant, maintenance schedules and instructions, experimental work to be carried out on operating reactors and any other submissions related to the safe operation of the nuclear power stations.

The intention of the memoranda is to ensure a consistency from station to station in the way they make submissions, and also in the method of assess­ment used by the various departments in the CEGB in considering these submissions.

Because of their importance to nuclear safety, sub­missions dealing with modifications to plant are the subject of a separate document which is approved by the HSE. The safety case for any station is based on the design of the plant and any changes to that plant could materially affect the case. Modifications to plant are divided into categories reflecting their nuclear safety significance. The highest category, Ca­tegory 1, contains those modifications which could significantly affect the risk of a radiological release or involve a change in the principles upon which the safety arguments were based. Category 1 submis­sions must be agreed both by the Nuclear Safety Committee and by the HSE before the modification is implemented.

Category 2 submissions, having some safety signi­ficance but not materially affecting the overall risk of a release of radioactive material and involving no change in the safety principles, need only the agree­ment of headquarters department and divisions before they are implemented. Finally, Category 3 modifica­tions, having no safety implication, may be imple­mented by the station without consultation.

In the latter case there is a requirement for a record of all such changes to be maintained on the station available for inspection by the HSE site inspector or by the HQ departments or divisions.

Within the context of the procedure, modifications to plant include the repair of any component (parti­cularly if it is pressure retaining), the replacement of one type of unit by another (unless it has previously been accepted as essentially identical) and any change to the safety arguments.

Oifferenrurgans of the body may receive different dose equivalent values due to the nature of the irradiation taking place. There will be varying risks associated with each organ, depending on the amount of dose received by the organ and the particular organ concerned. By considering the risk associated with each organ, it is possible to produce a total risk figure for the whole individual. This total risk is known as effective dose equivalent and is equal to:

T = WT HT

where H у is the average dose equivalent in the organ T
Wj is the weighting factor for the organ T.

The values of the weighting factors are:

The unit of effective dose equivalent is the same as dose equivalent, i. e., the sievert (Sv).

In the same way that dose equivalent may be committed, effec­tive dose equivalent may also be committed. If the time period chosen or integration is 50 years, then the quantity determined is committed effective dose equivalent, and has the sieved as the unit.

A6 Collective dose

A useful concept in radiological protection is collective dose. If dose equivalent or effective dose equivalent Is assessed for any number of individuals, which for some purposes may be as low as one or, for others, be equivalent to the world population, that assessed quantity is known as collective dose. The unit of collective dose equivalent and collective effective dose equivalent is the man sieved (man Sv).

A7 The old units

Each of the quantities described has been expressed in terms of the

St unit appropriate to each. These units were introduced as a requirement of the The Units of Measurement Regulations 1980, (Statutory Instrument 1070, 1980), which impfement a European Council Directive requiring the introduction of SI units in member states. The regulations require that SI units be used from 1 January 1986.

Previously, the radiological units used to describe activity, absorb­ed dose and dose equivalent were the curie (Ci), the rad and the rem respectively, A special unit, the roentgen (R), which is defined in terms of the amount of ionisation in the air that a particular quanti­ty of X or gamma radiation produces, has been dropped altogether.

Table 4.17 summarises the relationship between the old and SI units.

[1] Inelastic scattering Strictly, inelastic scattering is an absorption event in that, in accordance with the theory of the compound nucleus, the neutron is firstly absorbed by the target nucleus and a pos­sibly different neutron subsequently ejected in a random direction. After the collision the target nucleus is left in an excited state and reverts to the ground state by emission of a 7 ray. The energy of the ejected neutron is much lower than that of the initial neutron, being the difference between the kinetic energy of the bombarding neutron and the excitation value of the nucleus. It may be deduced that inelastic scattering can only occur if the kinetic energy of the impinging neutron is at least equal to the excitation energy of the target nucleus. For a given material there is, therefore, a minimum threshold energy below which inelastic scattering is not possible, typically 0.1 MeV for the heavy elements and greater than 1 MeV for the light elements.

[2] Uranium 238 capture cross-section, Fig 1.6 (b). oc exhibits the /v dependency in the low neutron

[3] U-235 is a fissile material; it can undergo fission with neutrons of any energy but is much more likely to do so the less energetic, or slower, the neutron, Fig 1.6 (c).

• The bulk of the energy appears as kinetic energy of the fission products. This energy is given up in the form of heat within the fuel.

• Not all the energy release is instant. About 13 MeV comes from fission product decay and is delayed.

• Some of the energy produced is associated with the neutrons and gamma radiation. As neutrons and gamma rays can travel large distances in matter some of the energy is released as heat some distance from the place of fission.

[5] Some of the energy is associated with antineutrinos which have an extremely low probability of inter­acting with matter; this energy is lost from the system.

Using appropriate conversion factors and a value of 200 MeV release per fission it can be shown that 3.1 x 1010 fissions per second will produce 1 watt of power; hence a 1000 MW (thermal) power station requires 3.1 x 1019 nuclei to fission every second. However, this very large number is small compared with the number of nuclei in a cubic metre of solid matter, of the order of 1028.

Fig. 1.30 Reactor layout and containment system of the CANDU/PHWR

pressurised in the acronym above refers to the pres­surised DiO coolant which flows in opposite direc­tions in adjacent tubes and passes its heat to the secondary coolant via the steam generators. System pressure is maintained by a pressuriser on one of the legs of a steam generator.

Adoption of pressure tubes in preference to a pressure vessel has a number of advantages:

• The failure of a pressure tube has not the same significance as the failure of a pressure vessel.

• The moderator can be kept at low temperature, giving a lower thermal neutron energy spectrum. Also, in the analysis of possible major accidents, the potential thermal energy represented by the mod­erator temperature is minimised.

[7] The reactor can, in principle, be made indefinitely larger.

• Access to individual pressure tubes makes on-load refuelling possible.

The fuel used in Candu is natural UO2 clad in zirca — loy. A bundle of fuel rods make a fuel assembly,

[8] A thermal neutron reaction with 10B which is gen­erally present as a grain boundary impurity at an overall concentration of 2 atom ppm.

[9] I he rate ol increase ol power, on w hich depends [he чгаіп rale (hence stress) in the clad and. pos­

[10] is the structural factor

[11] Contamination from added chemicals, make-up water and ion exchange resins.

[12] Some metal cations can be removed by solubilisation,

and although the concentrations will be very

[13] Achievement of a low oxygen concentration can be assisted initially by the use of hydrazine. During power operation, with the specified hydrogen con­

[14] Plant for collecting and treating active liquid waste and cooling pond water treatment.

[15] A store near the cooling pond for shield cooling air and ventilation filters (sealed), sludge, sand and resins from pond and active effluent treatment (in sealed drums) and Ctb filters (contained).

[16] The tensile and compressive stresses in the concrete, under all operating, test and shutdown conditions, are always specified below limits. (These limits are specified in BS4975, the British Standard for con­crete pressure vessels.)

[17] CB — background count derived from fission

products emitted from fuel surface con­tamination.

[18] Fail sate properties, i. e., any failures that do occur should have a high probability of failure in the direc­tion of safety.

[19] Station waste vault for incombustible materials and

[20] Close control of the product is maintained by checks on chemical analysis, neutron absorption cross-

action, material cleanliness, hardness and erain

size.

[21] Mechanical and electrical plant.

[22] The design and layout of other systems significant to safety (including cabling) incorporates segrega­tion and other means of hazard protection comen — surate with their safety role.

[23] To allow access to the fuel for its replacement and to allow access from outside for instrumentation

and control rod drives.

[24] The 3500 kW rated motor is a single-speed (1485 г/ min), drip proof, air cooled, three-phase, squirrel cage induction motor with a Class F thermalastic — epo. xy insulation system. The rotor and stator are of conventional design. The motor houses the RCP assembly thrust bearing, which is a double­acting Kingsbury type (accommodating either up­ward or downward thrust) comprising pivoted seg­mental shoes and a shaft-mounted runner. A high pressure oil lift system provides the initial oil film during start-up and the thrust bearing is self-lubri­cating at speed. Also mounted within the motor is an anti-reverse rotation device, which prevents

[25] Control the feedwater flow, in order to maintain steam generator level within satisfactory limits.

[26] To impose an absolute liability (even if due to unavoidable accident) on licensees for injury to per­sons and damage to property of any person, attri­butable to the radioactive, toxic, explosive or other hazardous properties of nuclear matter. Licensees must provide cover by insurance or otherwise, up to £5 million for any one cover period.’

[27] To control exposure to personnel, the site is zoned into areas defining the radiation and contamina­tion levels. The areas are designated by a Duly Authorised Person appointed for this purpose. The method by which areas are zoned must be included within the Safety Rules. There is a commitment

[28] Fault conditions which may be due directly to break­down of plant or the malfunction of some piece of equipment or system. The fault condition may or may not lead to a loss of generation or shut down

[29] Major overhaul has to be done on some items of plant which form a major component of a system (e. g.. boiler feed pumps, CV pumps and intakes).

[30] A group of plant operators who carry out the acti­vities of running the plant and refuelling the reac­tors, headed by a foreman.

[31] Inspection of reactor internal structures for sound­

[32] Automatic protective devices will shut down the reactor if power diverges at a doubling time of less than about 20 s, equivalent to a net reactivity of about + 200 mN.

[33] Turbine-generator and dump condenser controls to utilise the steam generated in the boilers.

Reactor power output is gien by the formula:

Power = gas flow x specific heat x temperature rise

[34] Control rod position which is related to the re­activity which is available by movement of the control rods. Magnetic synchros or resistance po­

[35] Data processor alarm from the *B’ thermocou­ples for any channel above target temperature. This margin is common to all channels, but since each channel has itv own target the alarm level is channel-specific.

[36] It may identify a leaking boiler so it can be iso­lated.

[37] The fuel dement identification number.

[38] The flask load is selected to ensure that the final heat burden of the flask is not greater than that required by the safety case. The heat burden varies

[39] Direct charge — this is the cost of actually carrying out the work specified and is dependent on the

[40] To provide specialist advice to all levels of manage­ment on nuclear health and safety aspects of siting, and to inspect and approve services in relation to

[41] Thermoluminescence dosemeters (TLD) have been located around each power station at a distance of between 1 km and 10 km and disposed at intervals of 15°. In addition TLDs have been placed at centres of population up to 30 km from magnox power stations and 15 km from AGR sites.

• Integrating electronic dosemeters have been placed at some of the TLD sites close to the power stations.

• Gamma radiation monitors have been installed at 60° intervals round the perimeter fence of each

[42] Liaise with the senior police officer and officers in charge of other emergency services.

• Arrange controlled access to the flask and trans­porter.

Following the satisfactory completion of its checks and inspections in the fuel inspection room, each fuel element is individually transported to the NFC where pressure testing and bottom end eddy current testing takes place. If these tests are passed the No 1 element in the fuel stack is carefully seated onto the bottom stringer component which has previously been located on the assembly tube platform within the NFC glove — box. (This component is a safety device known as an ‘anti-gapping unit’ (AGU) which, by the action of a spring, prevents the formation of inter-element gaps during refuelling.) A top-end eddy current test is then carried out and if successful, the platform lowered in the fuel assembly tube. The No 2 element is placed in the glovebox room for its final pressure and bottom-end eddy current tests. When it has been seated on the top of the No 1 element, the enrichment monitor is used to check the latter and the No 2 de­ment top-end eddy current test is performed. This sequence continues until the eight-element stack is complete, whereupon a top component, either a ‘top reflector’ (a graphite sleeve used to reduce axial neu­tron leakage from the active core) or a ‘central iner­tial collector’ (a device for collecting particulate matter from the reactor gas) is added, together with a ‘sta­bilising brush’ assembly (which provides lateral stabi­lity of the stringer when in situ) in the reactor core. As stated previously, a completed fuel stack involves the use of two different types of fuel element at any given enrichment. Positions 1 *6 contain fuel pins with 14 ASGs whilst the top two positions will contain 22 ASG pins.

When the stack is complete, the charge machine (positioned over the NFC access hole) lowers a plug unit with tie-bar attached into the cell. The tie-bar is carefully threaded through the central guide tubes in the elements and components and, when special end-fittings have been attached to the tie-bar at the underside of Lhe AGL, stringer assembly is completed. The latter can then be lifted out of the NFC and placed in a storage tube until required for refuelling.

The main sources of radioactive wastes arising at nu­clear power stations stem from the nuclear fuel dis­charge route, the reactor coolant and from the repair and maintenance of the reactor structure and asso­ciated control machinery.

2.6.3 Gaseous wastes

Discharges to atmosphere from the older magnox reactors are contained largely in air which is used to cool the steel reactor pressure vessel exterior and the concrete radiation shielding which surrounds the reac­tor. The air is subject to some neutron bombardment giving rise to neutron activation products. The pre­dominant radionuclide is argon-41 which has a half-life of 109 minutes. Treatment of the cooling air consists of filtration using glass fibre filter media or similar material before discharge. The argon-41 discharged is almost directly proportional to reactor power.

The reactor heat transfer medium (coolant), which is carbon dioxide at pressure, is also periodically discharged to atmosphere. This operation is necessary in order to maintain chemical purity of the coolant and also to depressurise the reactor for maintenance work. When discharges are made promptly after re­actor shutdown, the principal radionuclide is argon-41 arising from neutron actuation of argon in the car­bon dioxide, but smaller quantities of other neutron activated species are present including carbon-14 and tritium. Treatment of the effluent consists of filtra­tion using sintered metal or ceramic filters designed for the temperatures and pressures of discharge. Later magnox and AGR power stations use water for cool­ing the concrete pressure vessel, consequently the argon-41 discharges are much reduced and arise main­ly from the occasional discharge of reactor coolant. Typical argon-41 discharges for magnox and AGR power stations are given by Heap and Short fS]; and are summarised in Table 4.4.

Table 4.4

Typical Argon-41 discharges for AGR and magnox
power stations

Early magnox power station

(Air cooled pressure vessel) 400 to 5200 TBq

Later magnox and AGR power station

(Water cooled pressure vessel) 40 to 150 TBq

In the PWR, the reactor coolant is high grade water. Radioactivity arises in the water through neu­tron activation of dissolved air and of other elements present in the coolant. Many of these nuclides have short half-lives. Some of the fission products formed in the fuel diffuse into the coolant and some of these are gases. These gases are removed from the coolant through a gas-stripping process and are discharged from the reactor through discharge stacks above a radioactive waste treatment building. The principles of dilution and dispersion after filtration apply to these discharges in a similar manner to the gaseous discharges from gas-cooled reactors. Typical annual discharges of gaseous waste from a PWR reactor are given by Passant [9] and are shown in Table 4.5.

The main discharges of gaseous wastes are via stacks above the reactors. Minor discharges from ventilation systems are from lower level stacks. In all cases the height of the stacks is such as to ensure adequate dilution and dispersion with the atmosphere to safe levels.

The concern in the short term transient is that the magnox temperature of one or more fuel elements may reach its melting point. In this event the uranium would be exposed to the coolant and rapid oxidation would take place, releasing uranium oxide and fission products via the breach into the environment. Even worse, the magnox might ignite, releasing sufficient heat to melt the uranium. In this case, the fission product release would be considerably higher.

The studies therefore aim to demonstrate that start­ing from a steady state, full or part load condition
the probability, in the event of the breach occurring, of the magnox temperature reaching 640°C is less than one in one hundred. These steady state condi­tions, described in terms of a maximum clad tempera­ture and channel power, indicate the limits to which the operator may run his reactor and remain pro­tected against this particular fault.

The probability is evaluated by statistically com­bining the uncertainties in the various properties and parameters used in the fault studies and taking due account of errors in the measurement of reactor con­dition available to the operator. In addition, it is im­portant that the settings of the protection equipment are consistent with those assumed in the fault studies. Figure 4.3 shows a typical short term transient.

Prior to the introduction of the Ionising Radiations Regulations 1985, there were a number of different sets of legislation dealing with radiological aspects on the licensed site. Notable among this legislation were the health physics conditions laid down 4n each site licence, the Ionising Radiations (Sealed Sources) Regu­lations 1969 and certain aspects of the Radioactive Substances Act, including transport requirements. The 1985 regulations brought most of the requirements under one legal umbrella without radically changing, in practical terms, the legalistic nature of the Depart­ment’s tasks.

The health physicist, under the requirements of the 1985 regulations, acts as Radiation Protection Ad­viser to the management and its representatives, the senior authorised persons (NR). In addition to this role, the health physicist is responsible for monitoring radioactive discharges from the site and measuring the radiation dose to personnel in accordance with the legal requirements. A number of other legal functions are also carried out which stem from these tasks including the calibration and maintenance of instruments, the operation of the district survey, the running of the dosimetry service and the control of radioactive sources amongst others. It should however be recognised that ultimately the Station Manager is the responsible per­son for safety in legal terms, although much of the actual work is devolved to the health physicist.

The CEGB’s interpretation of the various legal re­quirements dealing with the protection of personnel from the radiological hazard, is reflected in the CEGB Safety Rules (Radiological), which are more fully dis­cussed in Section 4.3 of this chapter. The Safety Rules specify the operational conditions under which man­agement must seek radiological advice from the health physicist. For maintenance work, this advice takes the form of a Health Physics Certificate issued by an ac­credited health physicist for a specific task. The senior authorised person (NR), who will issue the associated Safety Document, is under no obligation to accept the certificate and may specify alternative radiological precautions to be taken. However, in the event of an incident, the management will be under obligation to justify not taking the health physicist’s advice.

6.1 Station emergency arrangements

6.1.1 Legislative requirements

Amongst the comprehensive conditions of each site licence is the requirement to: ‘make arrangements, to be approved by the Health and Safety Executive, for dealing with any accidents or other emergency on the site’. These arrangements must include provision for:

• The measures to be taken when an accident or emergency occurs.

• The reporting of the accident or emergency to spe­cified persons.

• The supply and maintenance of suitable and suf­ficient:

Protective clothing and equipment for the use of persons dealing with the emergency.

Radiation and contamination monitoring instruments.

Facilities for changing clothes, washing and de­contamination.

Means of communication with persons, local au­thorities or other bodies with whom it is necessary to co-operate.

Proper instruction and training of all persons who have duties in connection with the emergency ar­rangements, together with the maintenance of a register of the instruction.

A demonstration of the approved emergency arrange­ments, to members of the Nuclear Installations In­spectorate of the Health and Safety Executive, is required each year.

An obvious method to improve the economics of the fuel cycle is to increase the discharge irradiation limit. Considering a 500 MW(E) station, increasing the discharge limit from 5500 MWd/t to 6000 MWd/t effects a saving of about £350 000 per annum at 1984 prices. The discharge limit may be determined by re­activity needs and coupled with the limiting metal­lurgical behaviour of the canning material and the uranium bar, it is unlikely that the channel discharge irradiation can be raised beyond 6500 MWd/t.

Axial flux flattening Marginal fuel cycle improve­ments can be achieved by actual flux flattening. This requires the loading of absorbers at the level of the mid-element positions with no absorbers above or be­low ihem. These absorbers would require to be sup­ported by non-absorbing support struts made of, say, "magnox.

Photography is one of the obvious and most easily employed techniques available for inspection. Early devices were commercial cameras deployed into the above-core area via a standpipe. Usually, these were
limited to three movements of freedom (rise and fall, axial rotation, angular deflection). In addition load­ing procedures were slow and the rate of production low. The need to increase the coverage of inspection led to the use of motorised film cameras. These are incorporated, together with a flash system, in a pro­tective pod which is deployed into the reactor via a guide tube. Photographic inspection of specific items requires more complex equipment. To this end, ca­meras steered by closed circuit TV (CCTV) through the viewfinder are employed for close-up photography. Additional movements of freedom are obtained by pod rotation and by viewing through a rotating mir­ror. The use of TV necessitates the provision of a steerable lighting system together with cooling faci­lities. Typical photographs produced by these systems are illustrated in Fig 3.58. Photographic inspection of the below-core area involves the hazard of the passage of photographic film through an intense radia­tion field (if access is via the core). This has been overcome by deploying the camera mechanism into the subdiagrid area and then posting the film to the

camera utilising a pneumatically propelled ‘bullet’ car­rier. Core transit time is extremely short and film fogging avoided.

CCTV has many attractions as an inspection tech­nique and certainly is appropriate where the device handling time is such that film fogging precludes the use of a photographic technique. However, high radia­tion levels lead to lens browning and break-up of the TV picture. Again, steerable light sources and rotating shewing mirrors together with multi-linked deployment equipment provide versatile viewing systems. Tracked vehicles (tanks) carrying a TV camera and controlled by an umbilical cable have been employed at the charge pan level; the advantage of this type of system is that a large area can be covered from a single entry point. This method has been used successfully in the inspec­tion of a large number of fuel channels following remedial work (600 channels in 8 hours). Hard copies of the TV screen are obtained using a printer or by photographic means. However, this method of inspec­tion does require technically experienced operators to interpret the views as opposed to photographic prints which can be produced by staff not trained in inspec­tion, and assessment carried out at a later date. Video recording has been used for technical assessment at a later date but this may result in additional visits to the area to resolve queries.

Photogrammetry, originating from map making techniques, has been used by the CEGB to investigate movement of components brought about by relatively small temperature changes. This has successfully de­monstrated the satisfactory operation of temperature compensated devices. Considerable progress has been made in the area of holography, particularly for the examination of AGR fuel. Holography allows a rea­sonable volume of the reactor to be recorded and photographed. Currently the volume is limited to about a i-2 m cube, although detailed examination can be made of components. Television and photo­graphic systems require the provision of light sources. A novel type of laser-powered camera providing TV — like records but with high resolution and great depth of field has been developed and this reduces the pro­blem of illumination. A low power laser produces a light intensity exposure in the beam greater than that obtainable with a high power light source. The angle of view and zoom facilities are controlled elec­trically. The reflected light is amplified by a photo­multiplier and the signal display on a TV monitor, the raster being synchronised with that of the laser.

9.1.4 Inspection — sampling and measurement

There are occasions when a detailed material analysis is required to assess component condition. The drilling of components with the collection of swarf is used where the material composition is unknown. Extensive use of laser techniques has been employed where trace dements, for example, silicon and sulphur, need to be determined with greater accuracy than that provided by the material specification or the component size and access precludes drilling techniques.

The measurement of oxide thickness is carried out by a number of techniques. Again, a laser has been employed to ‘drill’ a hole through the oxide layer with through-penetration being detected by the increase re­flectivity of the parent metal surface. The energy used for penetration is a function of the oxide thickness. An eddy current device utilising the difference in per­meability of the steel/oxide layers has been extensively employed to determine oxide thickness.

Ultrasonic techniques are used to determine inter­facial oxide thickness and to measure the strain con­dition of components, particularly bolted assemblies. Whilst the principle of application is relatively simple, a great deal of research and development has to be employed to obtain satisfactory results. For example, in order to test a socket-headed set screw in a charge pan component, it is necessary to build a unit incor­porating the following features:

• Precise location facilities.

• TV viewing.

• Grinding attachment for bolt preparation.

• Ultrasonic probe.

• Indexing facility.

• Umbilical cable for services and signals.

A novel ‘air-abrasive’ technique has been used to se­cure metallic samples without destroying the surface finish of the sample. Here, four air-directed streams of abrasive particles impinge on the surface of the ma­terial to be sampled. The streams are arranged to cut out a small four-sided pyramid of the material which is recovered for sampling purposes. The surface retains the undamaged oxide layer that can be examined and no changes are introduced into the material as a result of the cutting action, as in the case of drill samples.

Direct measurement of components has been achi­eved by photographic stereogrammetry using the prin­ciples of aerial mapping. The system has been sensitive enough to detect and quantify very small changes in the clearances of components brought about by a relatively small change in temperature. Thermocou­ples and transducers have been installed into reactors to measure deflections of components brought about by temperature changes. These measurements have been possible on a reactor operating at power. This type of measurement has been required to demonstrate the continued satisfactory operation of a component with the reactor at power.

The International Atomic Energy Agency (IAEA) is the United Nations body which is concerned with the peaceful uses of atomic energy, and to this end it develops and publishes safety standards. During the late 1950s, the Agency undertook to produce such standards for the transport of radioactive materials, and the first IAEA Transport Regulations were pub­lished in 1961. The Regulations are subject to periodic revision, and further editions were published in 1964, 1967, 1973 and 1985 [12].

The Regulations cover all modes of transport, road, rail, sea, air and inland waterways, transport being deemed to include all the operations associated with the movement of radioactive material, including load­ing and unloading of packages and their storage dur­ing transit. They apply to most forms of radioactive material, i. e., to material which has a specific activity greater than 70 Bq/kg, and to contaminated objects.