Atomic Safety and Licensing Board (ASLB) hearing for Clinch River Breeder Reactor, 1982
Same for the sentence in the middle of the second paragraph, which states that additional sensitivity analyses have shown that containment temperatures and pressure margins would accommodate a wide range of material releases to containment in excess of those predicted from realistic assessment.
Part of that information is contained in the documents that you referred to.
And are you relying upon the analyses in those documents for that conclusion?
Yes.
In the final sentence of page 54,
you state that the applicants have identified criteria and requirements
for features to mitigate the mechanical and thermal challenges resulting from HCDAs.
Can you describe those criteria to me, Mr. Dietrich?
Could I answer that, please?
Of course.
The details, the requirements that have been specified are of two general types, one which
we call structural margin beyond the design base which are discussed in volume one of
CRBRP3 and the other classification is thermal margin beyond the design base which are discussed
in volume two of CRBRP3.
The general character of the requirements in the case of the structural margins is to
provide dynamic loadings and leakage requirements on the reactor coolant boundary so as to limit
the leakage of materials from the reactor coolant boundary in the event of a hypothetical
core disruptive accident.
In the case of the thermal margin beyond the design base requirements, the nature of those
requirements are to specify requirements for certain features that are added to the plant
specifically to mitigate the consequences of core melt accidents and to provide certain
additional requirements on some other features that are already in the plant but these are
supplementary requirements on those features also to mitigate the consequences of the core
or melt aspects of a hypothetical core disruptive accident.
Do you have these criteria listed at any point in CRBRP 3?
Yes, we do.
The testimony identifies where they are listed.
The top of page 55, we refer to the sections of both volume 1 and volume 2 of CRBRP 3,
list the specific requirements.
Do you have the page numbers to those sections?
No, I don't have the documents in front of me,
but the section numbers are defined in the testimony.
volume 1 volume 2
page numbers for the page numbers for the criteria
by criteria i'm assuming you mean requirements
well you you uh use the term criteria on page 54
we distinguish between criteria and requirements we use both terms at the bottom of page 54
where are the criteria documented
for
for CRBRP 3 volume 1
appendix
should be appendix B
but this particular copy
seems not to have an appendix B
does anybody have an appendix B?
uh just a clarification uh the testimony cites to the requirements specifically
the question was the page number of the requirements i understand
requirements start on page 5-13 of volume 1 of crv or p3
the structural criteria which is the other aspect of the requirements and criteria referred to
are provided in Section 5.3 starting on page 5-25 of CRBRP 3, Volume 1.
And you mentioned Appendix B as containing other criteria?
I believe that Appendix B may provide additional background information
on the criteria that are referred to in Section 5.3.
Yes, that is correct.
A copy of Appendix B.
Have you introduced Appendix B into evidence?
Mr. Dietrich, are the criteria mentioned on page 54 a complete set of criteria, or are
they still under development?
MR.
Could I answer that, since the criteria are in our area?
We consider them to be a complete set of criteria.
MS.
Were they developed based upon detailed analysis of the Clinch River Breeder Reactor Plant?
Yes, as indicated in these documents that we're referring to.
That level of analysis was used to specify the requirements and the criteria.
The criteria themselves are more general so that they do not really depend on a specific
analysis.
So what stress or strain limit can you go to in analyzing components to these beyond the design base structure margin loadings?
So they do not really depend upon specific analyses, but the requirements are dependent on the analyses that we have in the Volume I and Volume II documents.
Am I correct that the structural margin criteria are on page 523?
They start on page 5-25.
And the thermal margin criteria are in appendix B?
No.
though uh volume two of crbrp3 is where the thermal margin of requirements and criteria are
on page 56 of your testimony second full paragraph you state that an energetic hcda
and CRBRP would be predicted only when
coolant cladding and fuel relocation occur in such a way that
the reactor undergoes a sustained super-prompt critical power excursion
which produces significant fuel vaporization.
Are you relying upon the detailed analysis in CRBRP 3
and the underlying documents for that conclusion?
I think that's a generic statement that would be true for any oxide-fueled
so we can cool
reader-reactor
the specific answer is no
turning to page 526
Volume 1 of CRBRP3, you state that the methods of analysis, evaluation guidelines,
and other information of Appendix B must be used in the application of these criteria.
How must these methods and guidelines and other information be used when the criteria are applied?
As discussed in Appendix B, that's the purpose of Appendix B.
On page 58 of your testimony, you state that depending on the initial conditions assumed,
the best estimate evaluation of the top HCDA for CRBRP predicts a stable, partially damaged core
having the power level which varies between well below nominal to slightly above nominal
or permanently shut down core considering a reasonable range of fuel ejection and fuel
sweep out assumptions in both cases the hcda is predicted to be non-energetic and is not
predicted to proceed beyond the initiation phase that's not a conservative estimate is it mr teacher
Which estimate are you referring to?
I'm referring to the best estimate evaluation mentioned in that paragraph.
I believe that's the best estimate as it states, yes.
That's not a conservative estimate, is it?
i don't believe we have intentionally biased that in a conservative way no
and your prediction that the hcda is not energetic is not based on a conservative estimate is it
the the cases that are referred to in that paragraph are not we're not intentionally
biased to be conservative however we have run more pessimistic cases which gave us the
same result did you rely upon the the analyses in crbrp3 for that conclusion
those conclusions are based on the analyses in crbrp3 that's correct
are the conclusions in the final paragraph of that page based upon the detailed analysis in crbrp3
yes
they are based on analysis
that's in CRBRP3
and do you rely upon the detailed analysis
of CRBRP3 for your
conclusions in that final paragraph
that paragraph reflects the analysis
that's in CRBRP3
that's correct
and you rely upon those analyses for your conclusion
yes
on page 59
on page 59
of your testimony
you refer to another best estimate evaluation
which predicts a non-energetic entry
into the meltdown phase
that's not a conservative estimate either
is it Mr. Dietrich?
I believe that particular calculation does have some bias in the conservative direction, yes.
But not in every parameter of that evaluation, does it?
I don't know that it does in every parameter.
At the end of that paragraph, you state that as a result of that best estimate evaluation,
you predicted non-energetic termination of the initiation phase.
Ned, did you rely upon the detailed analyses of CRBRP3 for that conclusion?
Objection.
This entire line of questioning can be just completely foreshadowed.
The documents exist.
We've amended our proffer.
It is a complete waste of time for the board, the parties, the witnesses.
I mean we're sorting pages this is clerical work
On the top of
page 16
you
predict
a non-energetic initiation phase termination.
Again, was that based on conservative assumptions?
The sensitivity studies were based on pessimistic assumptions
which were introduced to evaluate the effect of the more pessimistic phenomenology on the accident sequence.
In that sense, they were conservative, yes.
Can you tell me where the analysis on the second paragraph of page 60 is documented?
Those cases are documented in CRBRP3.
Where are the sensitivity studies on the first line of page 60 documented?
The same place.
Do you rely on the detailed analyses of CRBRP 3 for your conclusions at the end of the second paragraph on page 60
that these results increase confidence that the LOF initiation phase in the CRBRP is inherently non-energetic.
That conclusion is the result of my expert opinion,
having evaluated the work that was done and reported in CRBRP 3.
Do you rely upon the analyses in CRBRP 3 for that judgment?
yes along with my expert knowledge of the subject
on page 61 of your testimony first full paragraph
is the best estimate you referred to there a conservative one
I do not believe there's any intentional bias in that estimate, no.
The number of paragraph two on that page, when you state there is sufficient volume,
is that based on a conservative assumption?
Not intentionally conservative, no.
Trevor, read or react?
Objection.
No vote.
They use the parameters of the Clinch River breeder reactor plant and the components that
are designed to go into the Clinch River breeder reactor plant.
From that standpoint, they are specific.
such analyses would apply to a general reactor size and type.
Third paragraph on page 65, you state that requirements have been established to ensure that containment integrity can be maintained without venting for a sufficient time to implement evacuation procedures and to provide long-term mitigation of HCVA consequences.
Do you rely on CRBRP 3 detailed analyses for that conclusion?
Yes, Volume 2.
Do you rely on those same detailed analyses for the other conclusions in the same paragraph?
Yes.
On page 67, first paragraph, Mr. Dietrich, you mention in the second-to-last sentence
that the full range of experimentally observed penetrations can be accommodated without venting
the containment before a day.
Are you referring to detailed analysis
and experiment results containing CRBRP3
for that conclusion?
We're depending upon analyses, including sensitivity studies,
and the results of experiments that
are discussed here, which are also discussed or referenced
in CRBRP3, volume 2.
and you rely upon those analyses for this conclusion?
Yes.
And you also rely upon those analyses for your conclusion
that it has been found that the radiological consequences are
not sensitive to the vent time over a water?
On table 5-2 on page 72,
The SEFOR, the words that were used to characterize design in SEFOR, FFTF, and Clinch River, and I
did say that in the groupings of accidents into the three levels, if you look at the
words that are used in FFTF, ACDAs are included within the level three. However, I didn't
mean to imply by that that the functional application of those rules is any different
the three designs in it in SEFOR and in FFTF as in Clinch River the treatment of HCDAs is
is different than what the application criteria for the design basis accidents
It's not specifically drawn out that way in SEFOR, nor is it, what I've seen on FFTF, done that way.
In Clinch River we have specifically made the beyond the design basis evaluations separate into another category.
uh i don't believe they were treated significantly different in three plants
and you're referring
but you want to add on that i could just add from the standpoint of accidents the
purpose of the construction and operation of the SEFOR reactor was to do experiments to characterize
the reaction to characterize the response of the plant to deliberate transients induced in that
reactor core and it was on that basis that the individuals involved with the review of that
reactor took whatever position they did with regard to the evaluation of hypothetical core
disruptive accidents certainly both the fastbox test facility and crvp are designed to operate
at steady state power aside from descent and descent from power and will not be intentionally
subjected to transients the approach to the hypothetical core disruptive accident on fftf
is identical to that for crvip thank you mr swanson did you get a yes or no answer to your
question i don't think i asked a question that very well lent itself to yesterday no i was trying
to characterize the testimony and i realized that we got a certainly a more specific answer today
than we did yesterday and that's satisfactory mr strobridge yesterday you were talking about the
you were asked about the fermi accident and there was some questioning regarding the blockage of
two sub assemblies i want to ask you given the general size and type facilities such as
the proposed clintura facility and also given the state-of-the-art design features such as most
multiplicity of flow paths debris filters and screens what in your opinion is the
the likely result of a similar initiating event as occurred at Fermi 1?
In other words, a plate coming loose.
What would be the likely result of that type of initiating event occurring at a general
size type facility as I just described?
I think I can best answer that by referring you to figure 3-15 in the applicant's testimony,
which will help me discuss what the differences are.
This is page 44 of the testimony.
We have, extending down into and through the core support structure,
what we call core-inlet flow modules.
That portion is identified over in the right-hand picture,
and then the details of what those core flow modules look like
on the two left-hand pictures,
inserted into the core modules are the fuel assemblies.
The difference that we have between Clinch River and Fermi
is basically in the design of those inlet core modules
and the fact that we have a multiplicity of holes
around the inlet core modules
below what is called the debris barrier flange.
we have holes up through that debris barrier flange and then we have another sequence of
holes called auxiliary ports above the elevation of that debris barrier flange
and any of those sets of holes are adequate even including some complete blockages of some
part of the total number of holes that are there to provide adequate cooling to the fuel assemblies
that are inserted into these modules.
If a plate broke loose from any place
and somehow got into the lower part of the system
and came up against the lower portion of the fuel assemblies,
the lower portion of the core modules in the Clinch River case,
that would not block off probably any of the flow holes,
but depending on how it's oriented,
it could get up in and block off, say, two holes on one side of the inlet module,
leaving holes on the other sides of the inlet modules, leaving paths up through the debris barrier flange,
and leaving open holes into the modules above that debris barrier flange.
So therefore my conclusion would be that the consequences of any kind of a plate coming loose
and getting into the lower plenum of the reactor vessel and coming up against the bottom of those flow modules
would not be significant. An event such as the Fermi event would not
lead to any significant consequences in the
clinch or pubertum reactor plan. I see. Thank you. You were also asked today
about calculated low population zone doses
and whether or not you had performed calculations
beyond those that might be obtained at the end of 30 days,
and you indicated that you had.
You were not asked what the result would be
relative to the numbers that you indicated in your testimony
at the end of 30 days.
So I was wondering if you could indicate what those results were.
Yes. I believe the word that I used was I made an assessment of that.
The assessment indicated that the aerosol effects,
effects, which I discussed somewhat earlier, over the 30-day period will reduce the concentrations
of the materials within the reactor containment building, of everything except the double
gases, by three to four orders of magnitude from time zero to the 30 days.
That reduction due to the aerosol effect is sufficient to result in only a very small
amount of material still remaining in the reactor containment building at the end of
the 30-day period. Furthermore, we noted that approximately 90 percent of the 30-day dose
occurred as a result of the releases in the first day and approximately 98 percent of
the 30-day dose resulted from releases in the first week. Therefore, only 2 percent
of the total 30-day dose resulted from releases over the last 23 days out of the 30-day period,
which indicates to us that radiological consequences of releases for time periods
longer than 30 days would not be significant. Furthermore, there's another aspect, and that is
there's no good reason to believe that any accident could maintain the pressure of containment at the
design basis pressure for the whole 30-day period, let alone periods beyond 30 days.
And so the reduction in pressure would be another basis for believing the leak rates would in fact
be lower over a longer time period than they would be over the first 30 days. So based on
the assessment and those results I quoted, we have no reason to believe that the beyond 30-day
doses would add significantly to the doses we were quoting for the 30-day period.
I see. In that same line, one of the
assumptions that you indicated in your site suitability source term analysis
was that you assume aerosols
for the full 30-day period. Is that correct?
Yes. And what was your basis for that
assumption? Simply
trying to take into account the physics of the situation,
there is nothing to my knowledge that would indicate that aerosol
would be effective only over some portion of the time period,
such as one day. So the physics would indicate that the aerosols
should continue to be effective.
The aerosol calculation itself takes into account the fact that
there is less aerosol reduction
is a physically bounding sort of event
and it doesn't make any difference
of what the specific sequence is
that might lead up to such an event.
One could postulate a sequence where pump A fails before pump B, but it's not important
because we've identified the bounding events in section 3.2 and considered those as our
design basis accidents.
Two minor points.
There was some questioning, historical question on the parallel design.
First point is when was the PSAR docketed by NRC for CRBRP?
We submitted the PSAR in the spring of 1975. It was docketed by NRC on June 13, 1975.
Secondly, when was the present containment-confinement design concept submitted to NRC?
That plan design was submitted in Amendment 18 to our PSAR on April 30, 1976.
And when was the parallel design withdrawn?
The parallel design was withdrawn in Amendment 24 to the PSAR on July 22, 1976.
Mr. Clare, you were asked questions about the failed fuel monitoring design features in your testimony,
in particular the last paragraph on page 45.
You indicated that no analysis was relied upon there.
What is the basis for your conclusions in regard to failed fuel monitoring?
The basis for our conclusion that the failed fuel monitoring system
described in PSAR Section 754 will in fact perform its function is the experience base
which has been generated in the various sodium-cooled nuclear power plants around the world, where
these two features, which consist specifically as fission gas detectors, merely gamma detectors
detecting various isotopes from gamma spectroscopy
and by neutron detectors monitoring a portion of the sodium loop,
which have been typical features provided in a number of nuclear power plants.
Mr. Strawbridge, you were asked whether you had done new meteorology calculations
and secondly, whether you had done calculations which treat organs other than whole body, thyroid, bone, and lung.
And you answered yes.
Could you describe the results of those calculations?
Yes.
I have the results in the form of two tables to supplement Table 4-2 on page 51 of the testimony,
where in the first of those two tables, which I have identified as table 4-A, 4-2A,
I have done the calculation now for the additional organs, including liver, bone surface, and red bone marrow,
and also did the additional calculation specified recently by the NRC on what's called the mortality risk equivalent whole body dose
dose and have included those in this particular table using the same meteorology as all the
other calculations within our testimony.
I have furthermore performed similar calculations but updated the meteorology to the latest
meteorology referred to in section 2.3 of the PSAR and have similar calculations for
all those different organs and the mortality risk equivalent whole body dose.
The end result of those calculations is that there is a minimum factor of 15 on all of
the whole body and organ doses and a factor of at least 12 on the mortality risk equivalent
whole body dose compared to the guidelines that have been identified by NRC as appropriate
guidelines for the CP stage.
Mr. Chairman, I'm going to hand out two tables.
I would like to have those marked for identification
and I may ask applicants to submit those.
26 and 27 respectively.
They may be marked by identification.
And if I might be a little more specific on that identification.
There are two documents.
One is entitled or is captioned as a comment, Table 4-2-A.
And Mr. Trowbridge explained that.
And like that mark, for identification as applicants, it's a 26.
I'm sorry.
What's the life you've taken?
Twenty-six.
And twenty-six.
And a second document, which is captioned Table 42B, requests that that be marked for
identification as applicants exhibit 27.
I'm sorry.
one final question
to the entire witness panel
and that is
are the conclusions
drawn in the testimony
that you have sponsored
while it relies
on specific CRBR information
do you believe that it's
fairly representative
of a reactor of the general size
and type
and in your opinion
does it show that it is feasible
to design to meet
the conclusions you have drawn
Yes.
I have no further questions.
Mr. Edgar, in the first place, would it be appropriate to
to characterize applicants exhibits 26 and 27
as addenda or amendments to applicants exhibit one?
Yeah, they're supplemental too.
Supplemental.
Judge Linenberger, the question was asked in redirect.
The witness had done the calculations
and this is the answer to the questions
about what the results are.
All right, sir.
Secondly, is it appropriate for the board to assume that this panel of witnesses has no substantive corrections to make to
Appendix Exhibit 1, their pre-file testimony?
I'll ask that to this panel right now.
I'm not particularly interested in taking up time with punctuation mistakes or spelling mistakes, but any matters of substance.
Does the panel have any corrections as to matters of substance?
No.
All right, sir. Then a few questions, if you gentlemen can hold up a little longer.
through a circuit where there's no temperature control,
the over 1600 degrees Fahrenheit thank you the overall sodium primary sodium
loop has been characterized as operating near atmosphere and I understand that
characterization is primarily in contrast with the light water systems
where the primary coolant is considerably higher.
However, somebody did indicate yesterday
that when the circulating pumps are running,
there are places in the sodium primary loop
where the pressure is on the order of 10 atmospheres.
Now, approximately how much does the 10-atmosphere pressure
uh elevate the boiling temperature if you have any idea don't guess if you have a number fine if not
if the boiling point at 10 atmospheres would be somewhere between 1700 and 1800 degrees
so the order of 1 to 200 degrees elevation
yes and I presume we're talking Fahrenheit here
does that mean then that
if the sodium in the loop
under full power normal operating conditions
were suddenly relieved of the approximately 10 atmospheres on it.
What does this say with respect to whether it would immediately start boiling or not?
The normal hot leg temperature is on the order of 1,000 degrees,
which is several hundred degrees below even the boiling point at atmospheric temperature,
at extreme atmospheric pressure,
and therefore any relief of pressure during normal operation
would not result in boiling at any point in the loop.
I gather in terms of accident analyses
that considerable...