ML19259D211
| ML19259D211 | |
| Person / Time | |
|---|---|
| Issue date: | 07/23/1979 |
| From: | Wu T Advisory Committee on Reactor Safeguards |
| To: | Plesset M Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML19259D206 | List: |
| References | |
| ACRS-CT-1157, NUDOCS 7910170375 | |
| Download: ML19259D211 (3) | |
Text
f.
23 July 1979 To:
D r.
M. S. Ples set, ACRS, From:
T. Y. Wu, Consultant Subjec t:
ACRS ECCS Subcommittee Meeting, June 19-20, 1579, Washington, D. C.
Review and comments on meehng subject.
During this meeting various important aspects concerning the two main subjects: (1) B&W Small Break Analysis, (2) NRC Water Reactor LOCA/
ECCS Research were reviewed, audited, commented and discussed by the NRC Staff with the Subcommittee. The scope of coverage was quite exten-sive as it included an overview of the capability of the existing methods for small break analysis, an assessment of the B&W guidelines for small breaks, a close post-event examination of the guideline-operator interface as well as a survey of separate effects undertaken in research development.
Emerging from this meeting a clearer view seems to have develop.ed on the extent of the present knowledge of the phenomena, exceptions to guide-lines, desirable modifications and adaptation to the procedures in ue, and the directions of new work. Instead of giving fut dar analyses of the meeting proceedings, this b'rief report is intended to offer a few comments in connection with the main subjects.
1.
STABILITY OF THE INTENDED THERMO-HYDRAULIC PROCESS DURING LOCA.
In the event of a small break LOCA (~ 7 in or less) in a B&W 205-FA PWR, the decay heat removal from the reactor vessel depends on continual (design-basis)* natural circulation and pool boiling as the two principal modes. Alternating removal of decay heat can be effected through safety valves and by condensing inside steam generator tubes. Further, decay heat can be removed from the reactor coolant system by blowdown through the break and by a release of steam on the steam generator secondary side.
To ease up the post-LOCA operation or shutdown, it is extremely essential to avoid possible impediments and obstructions to these various modes of decay heat removal. Assurance of this goal amply exemplifies the impor-tant role of a sound knowledge of thermo-hydraulic processes that can play in safeguarding nuclear reactor safety.
i j
9 m
I s
/
79101703 7
In view of the small break analyses carried out by NRC Staff, several consultants including C.liichelson and others, it seems quite plausible s
that diHerent combinations of impediments to decay heat removal indeed appear to exist and need to be;better un[erstood. 'A likely case is when steam bubbles generated by pool boiling and other high temperature loca-tiens seek their way to form a large steam bubble in the U-bend at the top of a steam generator, natural circulation is then interrupted and water boiling in the reactor core becomes more vigorous due to loss of circula-ting flow. Since the local flow around the U-bend, whether blocked or not, is at least two-dimensional in nature, it is thought that an accurate evalua-tion of the flow situation would be beyond the reach of any 1-D model, such as the Code RELAP-4. To understand the eHects due to a steam bubble formation in the U-bend on obstructing natural circulation, it should be much more effective to carry out investigations especially for the local 3-D ilow with emphasis on the physics that underlie the mass and heat transport. It would be fruitful to pursue in parallel engineerhg solutions by developing effective remedies for overcoming such hazards. For example, one may explore possible effectiveness in employing new U-bend head sprays and/or safety relief valves which could be automatically controlled by a set of differential pressure tranducers in order to detect if the U-bend is blocked by a steam bubble and to facilitate its removal.
2.
ADEQUACY OF DESIGN AND INSTRUMENTATION Judging fror the present piping network arrangement in the TMI-2 type Blew 205-FA PWR especially involving the pressurizer surge line (which is square-bent in subsidence), the pressurizer watee would not drain but remain almost stable in volume aside from smail variations due to steam condensation in the pressurizer and due to entry of hot-leg steam into the pressurizer. These two and other possible modes all indicate that adequacy of the design should be reexamined with a thorough analysis of the physical processes involved.
The current round of review analyses has substantiated the belief that the water level in the pressurizer is not a reliable indicator of natural circulation and core cooling conditions. This conclusion raised an impor-tant question if it is necessary to develop a separate device as a direct y
I smer
- ~
- m.
... :.. - : 1--- - v u. :.:., r.y. -,. - n :- - ;. c... :,:... : ;...a..,... -
s., s.;. - - -
indicator of water level over the reactor core in addition to the existing pressure ind temperature gages-indhe+eactor vessel. Among the usabip.. -...
types known to the pr aiession, one may consider selection between the resis tance-type, conductance-type and the type using ultrasonic waves for an optimum water level gages adapted to reactor vessel applications. It should be clear that with the capability of indicating the water level over the reactor core much of the undertainities and frustrating guess work one time faced can be eliminated.
3.
AN OVERALL POST-EVENT EVALUATION AND SOME FOPsESIGHT The current round of post-event review of operation end guidelines for small breaks, analysis of design related to safety, case studies of numerical code applications, as well as audit of operator training has been conducted with a conscientious 'and commending effort and spirit.
Almost in every aspect there have emerged views and beliefs on uncer-tainties of the physical processes, limitations to the present knowledge, inadequacies of modeling with too crude or overly simplified assumptions (such as neglecting rapidly transient phases), and insufficient mnigation of potentially hazardous situations. While much of these concerns will require confirmation by further studies, I think the time is opportune to take a full benefit cf the situations for making a firm commitment to a full scale overall investigation. Ideally, this wide range study should stress on the basis of physical understanding and its interface with human engineering. It should be conducted with imagination and insight -
imagination for thinking out all possible yet hidden sources of ~ hazards, insight for extracting from the heap those important problems which need to be better understood and integrated into the system engineering. It can be of great value co have such a special task force established and organ-ized in order to assist the NRC Staff. There is no lack of good ey imples of ' thinking-tank' organizations that have successfully served some gov,ernmental agencies.
j*
c N
~. - -
..,... -. _ - - -