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14 декабря, 2021
3.2.1 AEC Criteria
In July 1967, the AEC published a set of 70 design criteria for use by nuclear power plant designers. The AEC had the following objectives in doing this: (a) the provision of a record of the criteria which had already been used by designers to that date; (b) the provision of a standard set of criteria so that all designers could follow the same rules. This would make licensing somewhat easier; and (c) the provision of a safety check list to ensure that all safety considerations would be covered before a license was issued.
The criteria were published as “General Design Criteria for Nuclear Power Plant Construction Permits” (8). These criteria were reduced to 58, rewritten and republished in February 1971 (9). They are split into six general categories.
Criteria
I. Overall plant requirements 1-6
II. Protection by multiple fission product barriers 10-19
III. Protection and reactivity control systems 20-29
IV. Fluid systems 30-46
V. Reactor containment 50-57
VI. Fuel and radioactivity control 60-64
The criteria certainly fulfilled the objectives above but they had certain disadvantages for fast reactor designers.
(a) The criteria were too vague, partly because of the state-of-the-art and partly because of a desire not to hamper the designer in his design choices.
(b) They apply to the nuclear reactor of the day: the light water reactor (LWR) system. They do not necessarily apply to fast systems, although many of the criteria are general enough to include all reactor systems.
(c) They were almost law and therefore difficult to change. Fast reactor designers have undoubtedly obtained relevant criteria of their own, but no AEC criteria specifically for fast reactors have yet been prepared. However the American Nuclear Society is moving to prepare such criteria through one of its standards subcommittees.
Figure 4.13 shows the state of the system in its initial state of compression at point A. It has an internal energy of a Btu/lb. If it now undergoes an isentropic expansion all the way to 1 atm, then it finally arrives at a state with an internal energy of у Btu/lb.
Thus, in expanding, work has been done by the core on the surrounding material. From the first law of thermodynamics
work done dW = —dU = (a — y) (4.26)
Internal energy Fig. 4.13. Calculation of available work energy: expansion of fuel from high pressure condition during the core disruptive accident; Kr—reduced volume, ST—reduced entropy. |
This work is available to do damage to the structure around the core. The rest of the energy (y Btu/lb) remains in the fuel material to be released on a somewhat slower basis to the surrounding sodium through heat conduction. This residual heat forms a large proportion of the total energy release and following transfer to the sodium, which may vaporize, further damage can be done.
The calculation of the exact isentropic expansion process can be performed if generalized tables are available. However, this is not so for uranium oxide.
Hicks and Menzies derived an approximate analytical method (9):
work = U,— U,-U, In (UJUj) (4.27)
The available work is the change in internal energy minus the unavailable work during the irreversible change from state 1 to state 2.
work = U, — 17,— unavailable work during
the energy generation phase (4.28)
= 17,-17,- — S,) (4.29)
= U,-U,-T, ldQ! T (4.30)
Assuming that dQ = cv dT for a constant volume process, but that for a liquid cv is approximately cp and then using an average value of cp, Eq. (4.34) becomes:
work = U, — U, — cpT, ln(7’2/7’1) (4.31)
This expression is similar (17) to the Hicks and Menzies version in [Eq. (4.27)] but it does not assume that cpT, = U,. Actually U, = cp(T,— Tm) where Tm is the base temperature chosen as the fuel-melting temperature.
This melt-down occurred when doing reactivity addition runs for period tests at low flow. The order to scram was misunderstood and a slow shutdown initiated rather than the fast scram. The whole core melted down in a few seconds. The accident was of course not directly due to the above instability considerations (43).
The detailed sequence of events for this accident was given in Section 4.4.4.1.
There are many possible alternative types of containment building. Some of the suggested varieties (14b) include:
(a) Underground containment, in which excavation alone may cost millions of dollars.
(b) Hemispherical containment with a wall and footing below grade.
(c) A prestressed concrete containment designed with compressive stresses of over 2000 psig at pouring, which can give a leakage rate below
(d) Conventional wall and roof panels for which leakage rates of less than 1% can be achieved with differential pressures of up to 0.5 psig for metal panels and 5 psig for a concrete building.
(e) Designs which include internal or external expansion volumes, such as in the case of the CP-5 at ANL which had a hold-up volume with a floating neoprene diaphragm to allow a hold-up of any fission gas release for up to 20-30 days.
As the LMFBR system contains plutonium, any gross fuel aerosol release to the containment system may necessitate leakage dilution factor of 10~3 or 10-4. Such values cannot be achieved with a single barrier because such values cannot be tested. However it can be done with two enclosure buildings
Fig. 5.6. Cross section of the EBR-II containment. [Courtesy of the Argonne National Laboratory (14b). Identification key: |
1. 5-ton crane |
11. |
Reactor vessel cover |
2. 75-ton crane |
12. |
Neutron shield |
3. Crane bridge |
13. |
Basement |
4. Concrete missile shield |
14. |
Sodium purification cell |
5. Gripper-hold-down mech. |
15. |
Na-to-Na heat exchanger |
6. Control rod drives |
16. |
Reactor |
7. Storage rack drive |
17. |
Subassembly storage rack |
8. Rotating plugs |
18. |
Concrete biological shielding |
9. Blast shield |
19. |
Subbasement |
10. Primary coolant auxiliary pump |
20. |
Primary tank |
with a combined leakage of the right value, say 2 and 0.5 vol%/day or 10 and 0.1 vol%/day. Thus, for plutonium containment, two barriers are usually required although one containment volume may be merely a rather leaky aerosol settling volume.
There are many other containment design requirements. The building must house the plant with convenience and layout is of importance. The design must account for lateral stability, withstand windload, snow and roof loads, and lightning loads, and it must allow these loads to be transmitted to the foundations. It must accommodate internal mechanical and heating loads, cope with penetrations, and be designed to withstand seismic loads.
The windload, for example, is calculated according to the equation
F = PAS = Cdou2AJ2g (5.9)
where q is the air density; и is the wind velocity; g is the acceleration due to gravity; As is the exposed surface area; P is the wind pressure; F is the wind force; and Q is a shape factor.
The internal loads must also include those which arise from reactor transients, and particular attention must be paid to the design of air locks and various penetrations to achieve the leakage rates required.
Figure 5.6 shows the EBR-II containment building and its two barriers; the inner containment barrier comprising the reactor vessel and vaults that withstand an internal pressure of 75 psig, and the outer containment shell of steel in the form of a cylinder that would withstand at least 24 psig with an overall leakage rate of 0.25 wt%/day. Similar designs modified to include a refueling cell have been suggested for the large 1000 MWe LMFBR plants.
Certain other laboratories also deeply concerned with fast reactor technology development and safety are the Pacific Northwest Laboratory (PNL), run by the Battelle Memorial Institute; the Liquid Metal Engineering Center (LMEC), run by Atomics International Division of North American Rockwell, Inc.; and the Hanford Engineering Development Laboratory (HEDL), run by Westinghouse Electric Corporation.
These laboratories are also subject to AEC contracts and under commitment to the AEC for certain projects. HEDL’s Westinghouse management is now engaged in bringing the FFTF facility into being, while LMEC provides sodium technology information and acts as a clearing house for failure data.
If we assume the separability of time and space dependence in the flux, for a step change of dk at t = 0, the flux and the precursor concentrations will eventually attain a steady period
Ф(Е, t) = ф0 exp(/a>) фі(Е, t) = фі0 exp (no)
After substitution in the single-point Eqs. (1.10) and cancelation of фі01ф0 and exp(/co), the in-hour equation is obtained:
q = dk/keS = (l*(o/keff) + £ «&/((о + A*) (1.12)
i-l
This equation relates all the N + 1 time constants of transient ф and фі behavior for the reactivity level q. Thus
A+l A+l
Ф=Фо^ап exp(/ft>„), фі = фі0 £ bin exp(ton)
n—1 П“1
For Q > 1 there is at least one dominant positive root m0 and the neutron flux increases with time. Figure 1.3 shows co0 as a function of q.
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