ML12216A018
| ML12216A018 | |
| Person / Time | |
|---|---|
| Issue date: | 12/01/1970 |
| From: | Office of Nuclear Regulatory Research, US Atomic Energy Commission (AEC) |
| To: | |
| References | |
| RG-1.002 | |
| Download: ML12216A018 (2) | |
Text
11/2/70 (Reprinted 12/1/70)
SAFETY GUIDE 2 THERMAL SHOCK TO REACTOR PRESSURE VESSELS A. Introduction gram. Since reactor vessel materials are ini-Proposed General Design Criterion 35 speci- tially ductile and their fracture toughness prop-fies design and operating conditions necessary erties are not significantly changed upon irra-to assure that the reactor coolant pressure diation during the initial 5 years of operation, boundary will behave in a nonbrittle manner. the potential for reactor pressure vessel fail-To provide protection against loss of coolant ure as a result of cold water injection is con-accidents, present designs provide for the in- sidered to be acceptably small during this jection of large quantities of cold emergency period. Sufficient data should be available from coolant into the reactor coolant system. The the HSST Program to permit a final judgment effect on the reactor pressure vessel of this cold within this 5-year period on the acceptability water injection is of concern because the reac- of the projected behavior of vessel material tor vessel is subjected to greater irradiation throughout its service lifetime.
than other components of the reactor coolant In the event that the results of the HSST pressure boundary and, thus, has a greater po- Program or other research indicate that the tential for becoming brittle. A suitable program potential for growth of defects in radiation em-which may be used to implement General Design brittled reactor pressure vessel n~aterial re-Criterion 35 to assure that the reactor pressure duces the available margin of safety against vessel will behave in a nonbrittle manner under brittle fracture to an unacceptable level, an loss of coolant accident conditions is described acceptable engineering solution to the problem in this guide. could be applied-for example, thermal anneal-ing of the reactor vessel material. Naval Re-B. Discussion search Laboratory data indicate that annealing of a PWR vessel at its design temperature The injection of cold water by the emergency (650'F) for a period of 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> should pro-core cooling system into a hot reactor pressure duce a recovery in fracture toughness proper-vessel after a loss of coolant accident raises the ties and reduce the transition temperature shift possibility that a vessel embrittled by irradia-due to irradiation by 30 to 50 percent (i.e., a tion and having a small internal defect could 100'F shift in transition temperature would be fail suddenly as a result of the large thermal reduced to 70-50' after annealing). Annealing gradient imposed and the resulting high BWR vessels at design temperatures and for stresses. Analyses by the reactor vendors indi- equivalent time periods would, if needed, pro-cate that cold water injected into a hot reactor vide an equivalent degree of recovery. Based on pressure vessel toward the end of the vessel's the calculation of potential irradiation effects in service life could cause incipient defects of the presently designed PWRs and BWRs, this de-maximum size expected to grow; however, the gree of recovery of material toughness proper-maximum crack depth is predicted to be no ties combined with the potential for repeating more than 30 to 60 percent of vessel wall thick-the annealing process, if required, appears to ness. The vessel is not expected to fail under be adequate to permit continued plant operation these conditions. The maximum crack depth with the same reactor pressure vessel through-expected cannot be firmly established since the out plant lifetime.
vessel material fracture toughness properties assumed in the analyses have not yet been com- C. Regulatory Position pletely confirmed. To assure that the reactor pressure vessel The additional data needed to resolve the will behave in a nonbrittle manner under loss uncertainties in the. fracture toughness prop- of coolant conditions, the following program erties of reactor vessel material are expected should be followed:
to be provided by the Heavy Section Steel Tech- 1. Data collection and research work on nology (HSST) research and development pro- the properties of reactor pressure yes-2.1
sel material should be continued in or- ently approved core or reactor pres-der to permit verification that expected sure vessel designs are proposed.
material properties assure nonbrittle 3. Should it be concluded that the margin behavior of the reactor vessel through- of safety against reactor pressure ves-out its lifetime under postulated acci- sel brittle failure due to emergency dent conditions. It is expected that this core cooling system operation at any determination can be made within 5 time during vessel life is unacceptable, years. an engineering solution, such as an-
- 2. During the 5-year period necessary to nealing, could be applied to assure ade-develop the needed data, the potential quate recovery of the fracture tough-reactor pressure vessel thermal shock ness properties of the vessel material.
problem which may result from emer- In the meantime, applicants should out-gency core cooling system operation line available engineering solutions and need not be reviewed in individual show that their designs do not preclude cases unless significant changes in pres- the use of such solutions.
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