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Home > Electronic Reading Room > Document Collections > Commission Documents > Commission Meetings, Slides, Transcripts, Meeting SRMs, and Full Written Explanation for Closed Meetings > 1997
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05/12/1997
1
UNITED STATES OF AMERICA
NUCLEAR REGULATORY COMMISSION
***
MEETING WITH BOILING WATER REACTOR VESSEL
AND INTERNALS PROJECT AND NRC STAFF
***
PUBLIC MEETING
***
Nuclear Regulatory Commission
Commission Hearing Room
11555 Rockville Pike
Rockville, Maryland
Monday, May 12, 1997
The Commission met in open session, pursuant to
notice, at 3:05 p.m., the Honorable SHIRLEY A. JACKSON,
Chairman of the Commission, presiding.
COMMISSIONERS PRESENT:
SHIRLEY A. JACKSON, Chairman of the Commission
KENNETH C. ROGERS, Member of the Commission
GRETA J. DICUS, Member of the Commission
NILS J. DIAZ, Member of the Commission
EDGAR McGAFFIGAN, JR., Member of the Commission
. 2
STAFF AND PRESENTERS SEATED AT THE COMMISSION TABLE:
JOHN C. HOYLE, Secretary
STEVEN BURNS, Deputy General Counsel
HUGH THOMPSON, Deputy Executive Director for
Regulatory Programs
THOMAS MARTIN, Acting Associate Director for
Technical Review, NRR
MICHAEL MAYFIELD, Chief, Electrical, Materials
& Mechanical Engineering Branch, RES
JACK STROSNIDER, Chief, Materials and Chemical
Engineering Branch, NRR
SAMUEL COLLINS, Director, NRR
CARL TERRY, VP, Niagara-Mohawk, Chairman,
BWRVIP Executive Committee
ROBIN DYLE, Project Engineer, Southern Nuclear,
Chairman, BWRVIP Assessment Committee
PETE RICCARDELLA, Structural Integrity Associates
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P R O C E E D I N G S
[3:05 p.m.]
CHAIRMAN JACKSON: Good afternoon, ladies and
gentlemen. The purpose of today's meeting between the
Commission, representatives of the Boiling Water Reactor
Vessel and Internals Project and the NRC Staff is to discuss
potential policy issues associated with the NRC staff
technical position regarding alternatives for augmented
inspection of the reactor vessel.
The NRC staff and representatives of the BWR
Vessel and Internals Project have interacted over the past
18 months with regard to a proposed alternative to augmented
inspection of the reactor pressure vessel.
In a recent Commission paper, SECY 97-088, the NRC
staff stated that no alternative to the expedited reactor
pressure vessel inspection requirements would be authorized
for boiling water reactor licensees until they have
completed at least one examination of essentially 100
percent of their reactor pressure vessel welds and have
shown that the examination performed provides an acceptable
level of quality and safety.
In its letter dated April 18th, 1997, the BWR
Vessel and Internals Project stated that an alternative to
the current augmented inspection requirements for all
domestic BWRs is warranted. The Boiling Water Reactor
. 4
Vessel and Internals Project stated that based on a
comprehensive study of the reactor pressure vessel design,
manufacturing process, in-service inspections to date,
operating experience, and extensive probabilistic analyses,
only longitudinal shell welds need to be inspected.
Many of these potential policy issues are linked
to the staff's determination concerning whether the BWR
Vessel and Internals Project's proposed alternative provides
an acceptable level of safety.
The Commission looks forward to the discussion
with both representatives from the project and the NRC
staff. I must say the Commission is interested in
understanding what, if any, technical issues relate to
policy issues that need to be resolved, and understanding to
what extent risk has been considered in the proposed
alternative and in the staff's proposed position, and the
implications of the staff's position on the industry's time
line for performing the augmented inspection.
I understand that copies of the presentations are
available at the entrances to the meeting. Unless my fellow
Commissioners have any opening comments, Mr. Terry, I guess,
you are leading this part of the discussion.
MR. TERRY: Thank you very much. We do appreciate
the opportunity to come here.
I'm Carl Terry. I'm vice president of Niagara-
. 5
Mohawk and also chairman of the BWRVIP executive committee.
There are a number of other people here from the VIP. In
fact, we represent a group of 21 utilities and 36 plants.
Eight of those utilities and 15 of those plants are
represented here today if we do get into more detailed
discussions.
The other thing, up here with me is Mr. Robin Dyle
from Southern Nuclear and Dr. Pete Riccardella, who are here
to support me in presenting. I do believe that our
presentation will come to point as far as the specific
questions you raised regarding what is our proposed
alternative, what are the risks associated with that, and
the associated benefits with going ahead with this
alternative approach.
CHAIRMAN JACKSON: You also should speak to why
you feel the Commission should be involved in resolving what
many might consider to be technical issues.
MR. TERRY: Okay. Thank you.
First off, we did provide some slides in advance
of the meeting. These slides are slightly different,
although technical content will not vary really
substantially.
MR. DYLE: They are right there in front of you.
MR. TERRY: And they are there in front of you.
As far as the presentation today, going to the
. 6
agenda slide, we're going to be -- after I make a few
remarks, Mr. Robin Dyle will provide additional detail
relating to the inspections that have been performed and the
details that we're proposing, along with information
relating to those in-service inspections.
Dr. Riccardella will go over the basis for our
safety assessment of this issue, and then Robin and I will
provide some summary remarks.
Going on to the introduction, what we are
proposing is an alternative for the BWR RPV shell weld
inspections. We believe that that's based upon a very sound
and thorough technical evaluation, as well as included in
that are deterministic and risk evaluations of these
inspections.
The proposed alternative we believe would result
in significant savings. These are both savings in exposure,
radiological exposure as well as cost, for the industry with
no measurable impact as far as safety.
It is important that this issue be resolved. The
reason that we asked to be here and talk to the Commission
is because we understand there is a disagreement between us
and the staff in terms of the recommendation. We believe it
has a sound basis and we felt this would be the most
expeditious way of addressing the issue. And w are here
today to request the Commission's approval of this proposed
. 7
alternative.
Just very briefly as far as a little bit of
backdrop against the rule, on the history slide, of course,
the rule was promulgated in September of 1992, and at that
time, there were opportunities to provide comments by the
industry. However, following the issuance of that rule, we
actually formed the BWR Vessel Internals Project. That came
out of a consolidated effort to address some issues related
to vessel internals specifically, but also, as part of that,
we did include inspections and evaluations relating to the
reactor pressure vessel.
As a result of that group effort, we did initially
meet with the NRC to discuss our technical approach in July
or August, rather, of 1995. The reason we did that is --
and we're going to get into this in a little more detail --
the primary issue really related to the fact that we
couldn't literally meet the rule without some relief,
anyway. So we got into looking at alternatives.
Following that initial meeting, we did submit a
detailed report with our proposal in September of 1995.
Around the middle of last year, we had a meeting with NRC
technical staff and senior management, and as far as we
know, while there were some requests for additional
information that came out, there really are no unresolved
technical issues that we know of as far as our submittal.
. 8
As far as the specifics on approach, we looked at
a number of options, whether it was an exemption to the rule
and other things, and we ultimately determined that an
authorization for a technical alternative would be an
acceptable legal approach to get this job done.
As far as our proposed alternative, in summary,
the current RPV shell weld inspection requirements call for
essentially 100 percent inspection of all circumferential
and longitudinal welds in the shell weld area.
What we are proposing as an alternative is to
inspect essentially 100 percent of the axial welds, i.e.,
the same as the rule with some minor access clarifications,
and zero percent of the circumferential welds. We believe
this can be handled as a technical alternative under the
current regulations.
With that background, unless there are questions,
I would like to move now to Mr. Robin Dyle. Robin is from
Southern Nuclear. He's technical chair of our Assessment
Subcommittee on the VIP and also he's active in a number of
ASME Code committees.
Robin.
MR. DYLE: Thank you, Carl.
On the Code and regulatory background slide, just
a few key points to make from there. The current Code
requirements are to do the 100 percent of the
. 9
circumferential and longitudinal seam welds, as Carl
mentioned; however, we believe that's inappropriate because
of a couple of things.
One, the Code treated the BWRs and PWRs as the
same when they promulgated the Code changes, and those of us
who were there understand that. There's no differences
accounted for at a Code level between the experience that
BWR would see from fluence, there's no differences in regard
to say a PTS event, which is not possible on a BWR but it is
on a P.
Secondly, there was no difference in the Code
between an axial and circumferential weld, and the stresses
are different. So there should be a technical basis for
treating those different as far as allowable flaw sizes in,
we think, the inspections that would be required.
Secondly, from the regulatory requirement
standpoint, when the staff put the rule together in 1992 and
invoked the 100 percent requirement from Section 11, they
did not consider the differences either in these two points.
They treated the Bs and the Ps the same. There was no
distinction made between, again, the issues such as PTS and
embrittlement related fluence, and there was also no
difference in the treatment of how you would evaluate a flaw
on the circumferential weld or a longitudinal seam weld.
So those two situations led us to think there were
. 10
reasons to look at this from a technical standpoint.
Again, as Carl pointed out, most BWRs physically
cannot meet the rule, and if you go back and look at the
construction history, the construction codes required a lot
of things, but one thing that was not in place at that time
for most of the BWRs was the Section 11 provisions for
inspection, and then as the later plants were being
constructed and designed, the rules that were there were not
very much in the way of what would be required during in-
service inspection. So they were designed and constructed
in such a way that there are physical limitations that would
prevent us from doing these examinations.
When the rule was put forth, I, representing the
owners' group, and some others met with the staff about what
was the appropriate way to approach this, because we knew
plants couldn't do these examinations, we knew the staff
would not want to see 30 or 36 individual relief requests.
So we tried to come up with a generic approach, and the next
slide really is where we are.
When we went off to do this, we said let's not see
how much we ought to reduce the inspections; let's start
from ground zero and say what would be the right thing to do
for a BWR vessel? What should we inspect? Where should we
focus our inspections?
Quite frankly, we were surprised at what we came
. 11
up with and the recommendation that we have, because we
would have thought there would have been more required. But
when you go through the technical evaluation again, we think
what we're proposing is legitimate.
What we focused on first was safety. We have to
operate the plant safely. We have to deal with risk, we
have to deal with exposure of personnel.
The second impact or the third impact there would
be the cost. There is a cost associated with this, and we
looked at that.
Then the last thing, we recognize that the staff
has to have defense in depth. That's just a given. They
have to know that what we're doing as an industry provides
enough assurance that we're operating the plant safely.
So those were the criteria we used and that was
the approach that we took going in to try to figure out
where we ought to go.
If you would skip two slides over to slide number
9, labelled BWR Fabrication. Just a couple of real quick
points I would like to make from a background standpoint.
When we went back and looked at this from a fabrication
standpoint, here are some of the items that described the
vessels. It's shells with rolled plates, you have vertical
seam welds and you have circumferential welds.
There are three different welding processes that
. 12
could be used and there are different cladding steps. One
machine clad for most of the shell courses, and then a
manual back clad on the field welds.
The bottom line is the seam welds and the cladding
all receive a post-weld heat treat, so that's a good thing
to have happen.
Also, repair welds, when they were necessary, they
were documented and tracked to the same degree that all the
vessel welds were done so that you have a high quality
repair there, and we know where those repairs are. They're
located and we can go find them.
On the next slide, when you get into the
fabrication inspections, and I won't go through all the
details of those, but there are multiple inspections
required. You can see radiography down to penetrant. And
then what we've listed there on the presentation are the
acceptable flaw sizes that we're concerned with, and
construction code.
Typically, if you look at the way these things
were put together, the vessel would get an RT and an MT;
then you have a PT after cladding to make sure the cladding
was put on; and both of these steps ensure that you don't
have surface breaking flaws. Then there was a hydrostatic
test performed, and then there was another magnetic particle
inspection done. All of this to assure that we don't put
. 13
the vessels in service with large defects, and that's where
we are. We think the inspection summary will show that,
also.
The next slide on the operations just again is a
background to try to point out a little bit of the
difference between the BWRs and PWRs.
The BWR, as you're, I'm sure, aware, operates so
that the steam region moderates the vessel responses. You
have the normal heatups and cooldowns along the steam
saturation curve.
One of the key things is the vessel temperatures
are normally 100 to 300 degrees above the P-T curve. So
we're always in the ductile region, you're always on the
upper shelf.
The pressure test after each outage is limiting
integrity challenge, and that's done normal operating
pressure but at a lower temperature, so it stresses the
vessel a little more. But the plant's in cold shutdown and
the pressure is carefully controlled and you have the rods
in. So if you were going to challenge the vessel, that
would be the right place.
The bottom line is, is that if you verify
integrity when you've done the pressure test, you're good to
go for a cycle. The worst that you could ever postulate
happening would be a leak during operation. There would not
. 14
be any brittle failure. And that's important to know.
CHAIRMAN JACKSON: How does the leak test tell you
that?
MR. DYLE: Because if you look at the evaluations
we've done, and Dr. Riccardella will get into it in more
depth, if you don't fail during the leak test from the
structural evaluation, you'll go up and you'll be in the
ductile region, so that if you did have anything, the only
thing you would have would be a leakage. You wouldn't have
a brittle failure of the vessel at operation because your
temperature --
DR. RICCARDELLA: The ductility of the material is
temperature dependent, and so it tends to be more brittle
when it's cold than when it's hot, and we conduct this
pressure test or leak test when the vessel is cold.
CHAIRMAN JACKSON: Okay.
MR. DYLE: If you would, turn to Slide Number 13.
It's labelled 1997 ISI Summary. And just to give you an
update, this is an update from what was originally provided
in our report, BWRVIP-05.
We now have responses from 37 domestic units and
three international units, and we've got all six designs
represented in the results.
Of interest here is back in 1995 when we looked at
this, we had over 440 cumulative years of operation.
. 15
Obviously, we have more in that range. There's some plants
that have now operated seven to ten years, and some of them
out in the 25- to 30-year range. So we've got a broad
perspective.
There are over 16,000 total feet of category B-A
weld that could be inspected, and the category B-A comes
from Section 11. Of that, over 8,000 feet has undergone a
full code examination, and an additional 700 feet has
received a partial code examination where you may have had
limitations, could only do one side of exam, or limitations
due to transducers.
On the next slide --
CHAIRMAN JACKSON: Let me ask you a question. I'm
told that in 1990, that inspections of BWR reactor vessel
heads at Quad Cities and Fitzpatrick identified surface
cracking and sub-surface flaws. Now, can you discuss the
implications of those within the context of the conclusions
that you reach?
MR. DYLE: Pete?
DR. RICCARDELLA: Yes. That cracking mechanism
was specifically addressed in the evaluation, and you'll see
how we did address it when we get into the probabilistic
fracture mechanics.
MR. DYLE: It was -- surface cracking wasn't
associated necessarily with the actual shell welds; it was
. 16
in the head region.
Back to slide 14, just going through the summary
briefly, as I said, there's over 8,000 feet that's been
examined, full and partial code examinations. Of that, over
7,000 feet has been examined using techniques which satisfy
Regulatory Guide 1.150.
We asked the EPRI NDE Center to evaluate those
techniques and their conclusion was, along with ours, that
if a procedure was used that satisfied the Regulatory Guide,
there was a high degree of probability that we would find
the flaws of concern when we did our inservice examinations.
So we're confident that those exams are valid and give us
good information about the status of the reactor vessel.
To date, out of all the examinations we've done,
there's been 17 indications that required evaluation.
There's been others that were acceptable to code evaluation
criteria. These 17 were all sub-surface. When you do the
fracture mechanics that's required by WB 3600, they are
found to be acceptable. And of that, the cumulative length
of these indications were 31 inches or .03 percent of the
weld length that we've examined.
COMMISSIONER ROGERS: How many of those were in
circumferential welds?
MR. DYLE: I would have to go back and look for
the exact number. The majority of them are in
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circumferential welds, which -- but again, they were sub-
surface, they were manufacturing type defects, and they
weren't anything that occurred inservice. So they would
have been there all along historically. I could go back in
the report and pull the data out and try to get that number
for you.
The last item on the page just simply gets to the
cost in man-REMs. The average cost when we did the survey
is about $3.3 million per interval, which is a ten-year time
frame. The interval comes from Section 11. Some units
would be less; some would be significantly higher.
Also, the average exposure associated with this
was 12.2 man-REM, and that's just to do the inspection.
Those numbers would go up for plants that do examinations
from the outside diameter. Also, as the plants age, that
number could get worse also.
The conclusion of the survey, shown on Slide
Number 15, is that the inspections done to date demonstrate
the shell seam welds are free from unacceptable fabrication
defects which you would expect from the manufacturing
processes that were used. We also found no flaws developing
during operation.
This evidence supports the conclusion there's on
degradation mechanism that's affecting the seam welds and
all of these things combined together supports the reduction
. 18
in inservice inspections that we're proposing.
The next slide is what we propose to do in the
future, and that would be that we'll use a demonstrated
technique and procedure. We're going to do the right kind
of NDE, we'll make sure it can accurately size and detect
the flaws of concern, and it will enhance our ability to do
that.
Also, as we do these vertical weld examinations,
the way they'll be done is in such a way that when you run
across a circumferential weld at the intersection, that weld
will also be interrogated at the intersection. What this
allows us to do is to continue to collect data on the most
risk-significant welds and not do the inspections on those
that are not risk significant.
CHAIRMAN JACKSON: Let me ask you a question about
terminology.
MR. DYLE: Yes, ma'am.
CHAIRMAN JACKSON: What do you mean when you say a
risk-significant weld? Aren't all reactor pressure vessel
welds essentially risk significant?
MR. DYLE: I think when Dr. Riccardella gets
through, you'll see that there are orders of magnitude
difference between the vertical seam welds and the
circumferential seam welds.
CHAIRMAN JACKSON: That may be the case, but are
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you telling us that we should believe that circumferential
welds are not risk significant? That's your basic position?
DR. RICCARDELLA: I think, first off, understand
that certainly a failure of either vertical or
circumferential welds is significant, and that's not our
point here at all.
What we really want to get to is the risk
contribution that's made by doing or not doing inspections
of these welds which is coupled to the probability of
circumferential welds actually failing. We're certainly not
here to tell you that it's unimportant that circumferential
welds fail. That would be significant.
MR. DYLE: It's a relative contribution, yes.
That concludes --
CHAIRMAN JACKSON: So you don't mean the risk
significance of the weld; you mean the probability of
failure of the weld?
MR. TERRY: Right. And we're talking about the
risk significance of the decision to inspect or not inspect.
That's really the key point here.
DR. RICCARDELLA: The probability of failure is so
small as to make the risk insignificant.
MR. TERRY: I think Dr. Riccardella, when we get
to his presentation, you'll understand more precisely where
we're coming from.
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MR. DYLE: That concludes my remarks, unless
you've got any questions about that. Dr. Riccardella, who
was one of the primary authors and did the fracture
mechanics evaluation, is next.
COMMISSIONER ROGERS: Well, I have a question. I
don't know where the best place is, but what about the
possibility that the weld materials of the circumferential
and the vertical welds are not the same? What could be the
implications of that possibility?
DR. RICCARDELLA: In our analysis, we've taken
into account statistically the possible variability in the
properties of both types of welds. We've analyzed the
probability of failure considering the variability in the
material properties, and as you see, the results come out -
- the results that come out are very striking.
COMMISSIONER ROGERS: All right. Why don't you go
ahead.
DR. RICCARDELLA: What I will present is an
overview of the methodology that we used in conducting this
probabilistic fracture mechanics evaluation, some key
features of the analysis and conservatisms in the analysis
as well as just a quick overview of the results and
conclusions.
As has been mentioned, the details of this
analysis were presented in this BWRVIP report which was
. 21
submitted to the staff in September of '95. That was
followed by a two sets of requests for additional
information which we responded to. I think that the overall
volume of paper submitted on this topic was probably about
four inches thick worth of response to the RAIs, and our
understanding is that all of the technical questions on our
analysis methods and conclusions have been answered and that
there are no technical issues remaining unresolved on this
analysis.
On the next slide, I'll talk a little bit further
about the inherent flaw tolerance of BWR and specifically
the differences between a PWR and a BWR in this area.
One of the major points is that the BWR vessel is
about twice the diameter of a PWR vessel. This creates a
much larger annulus of water between the core and the
vessel, and the result is lower irradiation fluence in the
vessel and, therefore, lower irradiation embrittlement.
The reference temperature, that is the brittle to
ductile reference temperature for a BWR varies from -- at
end of life varies from 60 to 150 degrees F versus almost
twice the value, 300 degrees F, for a PWR. As a result, the
material remains ductile. This is for both longitudinal and
circumferential welds. The material remains ductile during
all normal and transient operating conditions.
This results in an inherent flaw tolerance for
. 22
longitudinal seam welds for the limiting pressure test
condition and the ASME code quarter-inch reference flaw of a
safety factor of four against brittle fracture, which is
more than twice -- which is twice the code required safety
factor of two.
It also leads to the fact that a through-wall
crack that's ten times as long as it is deep does not exceed
the fracture toughness of the vessel even in the worst
irradiated beltline region.
These first two points are made for longitudinal
seam welds. Circumferential cracks exhibit even higher
safety factors. This is because fundamentally, the pressure
stress in a circumferential weld is half the stress in a
longitudinal weld.
You've asked about potential service degradation
mechanisms. Two that immediately come to mind are fatigue
and stress corrosion cracking.
Fatigue is relatively inconsequential in the
beltline and in the shell region of a BWR. The vessel
system cycling events are very slow and the fatigue usage
resulting from these events is very low. There is no rapid
cycling or severe thermal fatigue cycling mechanisms that
are applicable to the BWR vessel shell region.
Stress corrosion cracking you mentioned the Quad
Cities had -- it's definitely a concern in BWRs, both for
. 23
stress corrosion crack initiation in the cladding as well as
the potential for stress corrosion crack growth in the low
alloy steel vessel material. The SCC in the cladding has
been observed in the field. The SCC growth in the low alloy
steel material has been observed only in the laboratory; it
hasn't been observed in the field. But both of these
mechanisms were specifically addressed in the probabilistic
fracture mechanics analysis.
On the next slide, I show an overall schematic of
the analytical approach. I think you can read this.
Basically it's a Monte Carlo probabilistic fracture
mechanics evaluation technique where we select samples from
a weld, either from a longitudinal seam weld or from a circ
weld. I show here we're sampling an axial or longitudinal
weld. A crack is assumed to exist in that sample, and the
probability of that crack comes from two sources as shown in
the arrows leading to the upper box on the right-hand side,
probability of crack size.
We have included both the probability of a
manufacturing defect existing in the vessel in accordance
with the standard Marshall distribution. This is the
distribution that is -- the well known distribution that's
been known in PTS evaluations and has been verified with
respect to destructive examination of the Midland vessel.
In addition to that, we take into account the
. 24
potential for cracking to initiate in the cladding, and so
we have two potential sources of cracks -- of cracks being
distributed in this sample that we selected.
Then, with operating time, we consider the
potential for crack propagation, again in a probabilistic
manner considering IGSCC crack growth data and the stress
distribution both due to normal operating stresses plus
potential clad stresses, and then we have the ability to
superimpose upon this the inspection or non-inspection.
So we can have certain -- depending on what
percentage inspection we assume, we can have certain of
these samples that come through the Monte Carlo analysis
subjected to inspection and others not inspected, in which
case, if we consider inspection, then we superimpose a
probability of detection on that inspection and so then we
have a remaining probability that this crack will exist, and
then we make a comparison of the resulting crack size to the
critical crack size, and in doing that, we look at the
initial material properties, RTNDT, the possible variation
of copper and nickel content in the weld, and the fluence
versus time in the weld. So we make a time comparison of K
versus KIC.
This is the basic analytical technique that we use
to address this problem.
The next two slides, I talk about the key features
. 25
of the analysis, and I will point out that the starting
point for this analytical methodology was the method
developed by the NRC to address PWR pressured thermal shock,
namely the VISA code which was developed at Battle Northwest
-- at Northwest Laboratories.
This includes a probabilistic treatment of the
vessel fracture toughness and the radiation embrittlement
concerns; the assumed fabrication defects in the vessel,
specifically the Marshall distribution with all of the --
all of the defects in the Marshall distribution were
artificially moved to the vessel ID surface, which is
conservative from the standpoint of a radiation
embrittlement, but we did this to be -- and also
conservative with respect to stress corrosion crack growth,
because that's where the corrosive environment is. We did
this to be consistent with the NRC methodology for PTS.
As in the VISA code, it's a multiple random
variable, Monte Carlo analytical approach that we used.
We did have to add -- on the next slide -- some
features to the methodology to make it specific to analyze
BWR vessel ISI, and those include some items I've already
mentioned: the treatment of stress corrosion crack
initiation in the cladding; the treatment of stress
corrosion crack growth in the low alloy steel; the effects
of periodic inservice inspection. And because the resulting
. 26
probabilities are so low, we couldn't just use a brute force
Monte Carlo technique. I mean, you'll see in some cases we
would have had to take 10 to the 40th iterations. So what
we did is we implemented an importance sampling technique
out of the literature to speed up and basically to make the
calculations feasible.
These are the new features that we added in the
analysis. I should mention that we did, for the features
that are consistent with the current VISA code methodology,
we did benchmark our methodology against the VISA code, show
that we got essentially equivalent results, and that
benchmarking is documented in the submittals that we made.
On the next slide -- I'm sorry. Previous slide,
please.
Another key feature of the analysis is, you know,
as you go through these Monte Carlo iterations, a sample
either progresses to failure or it doesn't, and the
probability of failure is the number of samples out of the
total which have progressed to failure.
But what we found was that there were two types of
failures that were falling out of the analysis. One is the
crack would just grow to the point where we can't analyze it
anymore. It got to be 80 or 90 percent through-wall. But
we still haven't reached a point where K exceeds K1C. We
still haven't predicted a fracture. This is what we would
. 27
call a leak scenario.
The second type is that somewhere during that
crack propagation, due to the combination of a large flaw
and a low fracture toughness condition, you would predict K
exceeds K1C, and therefore we would predict a brittle
fracture.
What we found was the overwhelming majority of
cases, even where we did predict failure, were leakage type
failures. Something like, you know, 99 out of every 100
failures that we predicted in the analysis were leaks, and
only occasionally did we predict a brittle fracture type
failure, and when we did, that occurred during the system
leak test.
As Robin mentioned earlier, the critical condition
from the standpoint of a low pressure stressing of this
vessel is the leak test, which is conducted in a cold
condition when the reactor is in cold shutdown.
CHAIRMAN JACKSON: So you're arguing that leak
before break for the reactor vessel is acceptable?
DR. RICCARDELLA: Absolutely. And it --
CHAIRMAN JACKSON: Why is that acceptable?
MR. TERRY: That's not our argument. I think our
argument --
DR. RICCARDELLA: We're doing inspections. We're
saying that the analysis demonstrates that if -- in the very
. 28
unlikely event that we're going to have a problem with this
vessel, that that problem would be a leak, not a break. And
you will see a little bit further when I present the results
exactly how that manifests itself.
Let me just identify some of the conservatisms in
the analysis. They are listed here. I have already
mentioned the flaws in the Marshall distribution, even
though they're generally expected to be distributed through-
wall, we've pushed them all to the ID surface.
We have included a conservative treatment of
stress corrosion cracking in the cladding. Basically what
we said is if our analysis predicts stress corrosion
cracking in the cladding, we instantaneously assume that
that cladding is through-wall. We take no credit for time
for the crack to propagate through the cladding.
We also arbitrarily assume that it lines up with
one of these Marshall type manufacturing defects; that is,
we haven't assumed that -- as soon as we predict that the
cladding is violated, we assume that it's violated over the
entire inside surface of the vessel and, therefore, the
Marshall defects will be exposed to the BWR environment and
will propagate by stress corrosion.
The rates of stress corrosion cracking in the low
alloy steel are based on earlier test data which are shown
to be very conservative. More recent test data really shows
. 29
no stress corrosion crack growth in the low alloy steel, but
still we based the analysis on the more conservative data.
As I already mentioned, we have used conservative
vessel fracture toughness and radiation embrittlement
correlations.
On the next slide, I have a plot, a typical plot
of the results of a probabilistic fracture mechanics
analysis. There are three curves on this plot. The upper
horizontal dash line represents the PTS screening limit;
that is, the vessel failure probability that is inherent in
the NRC's PTS screening limit.
Then I show two curves. The upper curve
designated by triangles is the probability of leakage, and
then the lower curve is the probability of actual failure.
This is the point that I was alluding to earlier. All of
the BWR vessel probabilities are lower than the PTS
screening limit, but the probability of a break is much,
much lower, it's several orders of magnitude lower versus
the PTS -- versus the probability of a leak.
Also, I would address that all of the
probabilities shown on this chart are for longitudinal seam
welds. We can't even plot the probability of failure or
leakage associated with a circumferential weld because it's
so many orders of magnitude lower than these.
CHAIRMAN JACKSON: Where is the uncertainty? I
. 30
mean, these show these as point curves, but whenever you do
a probabilistic analysis, you know, there's a certain
uncertainty in that analysis, and where would that show up
in these curves?
DR. RICCARDELLA: You know, in terms of analytical
uncertainties, we have repeated these analyses over and over
and we show that they're accurate to within plus or minus a
factor of two. I'm not sure if that's what you're asking
about, or if you're asking about, you know, potential
uncertainties for things that we haven't considered, you
know, that we haven't considered in the analysis.
CHAIRMAN JACKSON: I'm asking you about both.
DR. RICCARDELLA: Okay.
CHAIRMAN JACKSON: I mean, there's a certain
uncertainty that gets propagated through a probabilistic
analysis, and any time you have a probability distribution,
--
DR. RICCARDELLA: Yes.
CHAIRMAN JACKSON: -- okay, you're really not
talking just simple multiplication or carrying through of
point values; you have to recalculate what the distribution
looks like.
DR. RICCARDELLA: That's true.
CHAIRMAN JACKSON: And so --
DR. RICCARDELLA: Yes. Those uncertainties are
. 31
within a factor of plus or minus two on the probability of
failure. But, you know, the main point that I would like to
make is that these curves are for longitudinal welds, and
we're not talking about changing anything for longitudinal
seam welds. I would like to make that point with the next
slide, which is a table.
In this case, what we've looked at, in this table,
the effect on probability -- both probability of failure and
probability of leakage of the current requirements, that is
the essentially 100 percent of all welds, versus the
proposed program, which is essentially 100 percent of seam
welds, of longitudinal welds. We have broken this down by
the contribution of irradiated longitudinal welds,
unirradiated longitudinal welds, and circ welds. And the
plot that I showed earlier is what gave the number, for
example, irradiated longitudinal seam welds, a probability
of failure of 5.68 times 10 to the minus 8. That --
CHAIRMAN JACKSON: With what confidence?
DR. RICCARDELLA: Let's see. I would say within
an accuracy of plus or minus a factor of two, but --
CHAIRMAN JACKSON: But with what confidence?
DR. RICCARDELLA: I haven't got a confidence
number, confidence interval right at my fingertips.
CHAIRMAN JACKSON: Okay.
DR. RICCARDELLA: But the point is, whatever the
. 32
confidence, it's exactly the same under the proposed program
because we haven't changed anything on longitudinal seam
welds when we go from the current requirements to the
proposed program. We're talking about the exact same
inspection. And likewise, for the unirradiated portion of
longitudinal seam welds. We're not proposing any change.
Where we're talking about a change is in welds for
which, to the best that we can calculate it -- and here I'm
not going to state much confidence in this value other than
to state that it's extremely low. We calculated a number of
10 to the minus 40th for the contribution to probability of
failure from circumferential welds; many, many orders of
magnitude less than that from longitudinal welds. We
basically had trouble in any of our Monte Carlo iterations
showing a failure, predicting a failure due to a
circumferential crack in a circumferential weld.
So what we're saying is that the probability of
failure, both failure or leakage, are both already lower
than the PTS screening limit and they don't change at all
with our proposed program.
So the conclusion slide basically just restates
this point. The calculated vessel failure probability is
already orders of magnitude lower than the PTS screening
limit. This is based on conservative analyses; they could
actually be lower if we took some of the conservatisms out
. 33
of the analysis. The proposed change in inspection scope
has an insignificant impact on the already small failure
probabilities.
MR. DYLE: Thank you, Pete.
Just a couple of slides and I'll turn it back over
to Carl for his closing remarks.
If you look at the slide for impact of
implementing the shell weld recommendations, and again, from
looking at the probabilistic fracture mechanics, as Pete
pointed out, we're not changing anything on the longitudinal
seam welds. So comparing apples to apples, there's no
change in risk with the program regarding those. But we are
talking about removing the circumferential welds, but we
don't believe there's any realistic change in the plant
safety or risk by not examining those circumferential welds.
Also, we can save at least 200 man-REM in exposure
by reducing the number of inspections we do, and that number
can go higher for the plants that do OD examinations. As
the plants get older and become more contaminated, that
number will be greater, also. But that's just from the
survey that we've done of what it takes to do the
inspections.
There is no consideration in this number for craft
support like insulators, scaffold builders and things of
that nature. This is just associated with performing the
. 34
inspections.
CHAIRMAN JACKSON: Do you use similar techniques
for doing these inspections as the Japanese use in their
reactor pressure vessels?
MR. DYLE: To the best of our knowledge, yes. I
know they are working on developing some new tools that
we're watching. I believe you may have seen one of them
demonstrated at the EPRI NDE Center on one of your visits,
and we're eager to see how well that works out. As yet,
that has not been done in the field and we're not sure what
limitations there will be. But yes, we are eagerly looking
for that.
Also, one other thing is we tried to do this in a
generic sense in a hope that we could reduce the number of
requests for exemptions and relief requests that the staff
would have to deal with, because there are so many plants
that will not be able to fully meet the rule. They're going
to have to deal with exemptions, and this would reduce a
number of those.
Finally, there is a significant cost savings to
the industry to implement this which would save in excess of
$50 million.
CHAIRMAN JACKSON: Commissioner Dicus?
COMMISSIONER DICUS: The 200 man-REM, is that
total for all plants?
. 35
MR. DYLE: That's total for all plants for one
ten-year interval, yes.
The next slide on the current status, where we
think we are today with this, we have submitted our
technical documentation in the form of the VIP report.
We've responded to the staff's RAIs, we provided additional
calculations and information on the NDE techniques.
We submitted a request for a technical alternative
that's currently pending, and we think we've resolved the
technical issues and are awaiting a response to that
technical alternative, and that's where we believe we are
today.
With that, I'll turn it over to Carl.
MR. TERRY: Thank you, Robin.
In closing, again going back over what we've told
you, the BWR vessels were fabricated free of large defects.
Robin went over the degree of inspections that were done
during the course of that fabrication.
We also talked about the survey results of the
ISIs that have been done to date, and they indicate no
significant flaws.
In summary, we've looked at about a mile and a
half of weld. We found less than three feet of indications,
and those were sub-surface indications and are not service-
related type flaws.
. 36
As far as BWRs, the cold pressure test that we do
generally at the end of the outages is the limiting BWR
condition. Certainly a failure at any time is not good, but
certainly that's -- that's certainly the least risk
significant time if a failure were to occur.
ISI of the circumferential welds is really of
little value. We see no impact on safety by not doing these
inspections, and that's really what's shown by the
probabilistic fracture mechanics work that we've done.
As far as the cost savings for reduced
inspections, they are substantial with no measurable
increase in risk. The inspection recommendations are
consistent with what we believe is the right focus, which is
to focus the industry and regulatory resources on those
issues that really add value from a safety standpoint.
Our alternative specifically is, again, to inspect
essentially 100 percent of the axial welds, longitudinal
welds, and zero percent of the circumferential welds.
Finally, in closing, by adopting the proposed
alternative, the BWR utilities will continue to perform a
substantial amount of inspections on the RPV shell welds.
We see no predicted leakage or failure for
circumferential welds, and I would point out here that this
is something that is unique to the BWRs as far as this
condition. The continued inspections of circumferential
. 37
welds does not add any measurable safety benefit, while it
offers substantial savings on the order of 200 man-REM and
$50 million for the utilities.
Rapid adoption of this is really critical. We are
coming for most plants or a number of plants to the end of
this current ten-year interval. This proposal, by the way,
is applied for the interval inspections; however, we are
coming to the end of the current ten-year interval, making
the current review and request for exemption particularly
critical and, therefore, we request the Commissioners'
approval of this proposed alternative.
CHAIRMAN JACKSON: Commissioner Rogers?
Commissioner Dicus?
COMMISSIONER DICUS: One quick question. You're
meeting with ASME, I understand, or you have met with them?
Could you just very quickly characterize what has come out
of those meetings?
MR. DYLE: In our discussions, the item has been
discussed at task group and working group and sub-groups
responsible for this issue, and the code case, which is
based on the report of doing 50 percent of the longitudinal
seam welds and zero of the circumferential, has passed all
the way to that point. It is at subcommittee and it is
waiting a letter ballot. I'm responsible for writing a
white paper to go with that for the members of subcommittee
. 38
to vote on that.
I have reason to believe there will be a large
majority of positive votes there because most of the members
also had a chance to vote on this and review the story as it
came up through the various committees. And we've deferred
writing the white paper so we could roll in any information
that might come forward from this meeting so that the code
committee is fully aware of everything that's been done.
CHAIRMAN JACKSON: Commissioner Diaz?
COMMISSIONER DIAZ: Just a couple of comments.
Obviously, this is a highly technical issue. We certainly
appreciate you bringing it up to the attention of the
Commission. But I kind of feel inadequate at judging the
technical merits of it.
I do believe there is some substantial benefit
from addressing the issue again and trying to have the
staff, you know, make an additional analysis on your
proposal. I certainly don't feel that I can, at this point,
address the technical issues on it.
CHAIRMAN JACKSON: Commissioner McGaffigan?
Well, thank you very much.
We will hear from the NRC staff.
MR. TERRY: Thank you.
CHAIRMAN JACKSON: We know who you are.
MR. THOMPSON: I was afraid of that. You know
. 39
where we live.
CHAIRMAN JACKSON: Mr. Thompson, please.
MR. THOMPSON: Thank you, Chairman Jackson. Good
afternoon, Chairman Jackson and Commissioners. Thank you
for the opportunity to discuss the staff's position on
augmenting examination requirements for boiling water
reactor pressure vessels, as we spelled out in our
commission paper, SECY 97.88.
At the table with me from NRR is Sam collins,
director of NRR; Tim Martin, the acting associate director
for technical review; Jack Strosnider, chief of the
materials and chemical engineering branch and, from the
office of research, Michael Mayfield, chief of the
electrical, materials and mechanical engineering branch.
First I would like to thank Mr. Terry, Mr. Dyle
and Dr. Riccardella as well as the other members of the BWR
vessel and internal projects for their extensive discussion
and evaluation that went into the development of their
report on BWR reactor pressure vessels shield weld
inspection recommendations. Although our judgments differ
on how to use the results of their effort, this is an
excellent example of their proactive effort in working with
the Staff to develop appropriate requirements for inspection
and repair of BWR internals, including the BWR core shrouds,
jet pump assemblies, core spray piping as well as a number
. 40
of other BWR internal components and systems. We believe
that these cooperative efforts will resolve safety issues
and they benefit everyone.
The staff has carefully reviewed the industry's
report and agree that it contains substantial technical
arguments for deducing the scope of BWR pressure vessel weld
examinations. However, we believe that this reduction
should be for inspections following the initial base line
inspection that is required by both our regulations and the
ESM code.
Our focus today is on the integrity of the reactor
vessels, the one component for which there is no redundant
safety system. It is vital that its integrity be
maintained.
Historically, our ability to predict component
degradation has not been perfect. Also, the ASME consensus
has evolved over time and currently requires 100 percent
examination of the reactor pressure vessel belt line welds
every ten years. Today, the staff's presentation by
Mr. Strosnider will focus on the need to maintain the
defense in depth and to validate the assumptions of the
industry's probabilistic model.
I would like to turn the rest of the briefing over
to Mr. Strosnider.
MR. STROSNIDER: Thank you. Good afternoon.
. 41
First, I would like to indicate that, as
Mr. Thompson said, in fact I would like to reemphisize that
the industry analysis has provided some substantive
arguments for reducing the scope of inspections. So you are
not going to hear a general condemnation of their analysis.
All right.
But I am going to go through some issues that the
Staff considered that led us to conclude that it is
appropriate to perform a base line examination before we
consider this sort of reduction. Those are the things that
I want to focus on.
Specific areas for discussion are listed in the
first viewgraph. I want to talk a little bit about the
safety significance of the vessel, the rule which you have
probably heard enough about now to understand what its
intention was, the need for inspections, some discussion
about the NRC and ASME inspection philosophies, visions that
do exist for relief or alternatives and then our
conclusions.
On the next viewgraph talking about safety
significance, stated quite simply the assumption is that the
reactor pressure vessel failure is an incredible event.
Quite frankly, when I got ready to present this particular
slide, it was a little difficult for me because we just take
that as a given that pressure vessel failure is not
. 42
something that is credible. The engineered safety features
of the plant are not designed to cope with reactor pressure
vessel failure. They are not specifically designed for
that, either catastrophic failure or leakage. So the
consequences of such an event have not really been fully
evaluated.
Pressure vessel integrity must be maintained at
the highest level of quality and nobody is questioning that
statement. An important part of that, Staff's position is
that an important part of that is maintaining defense in
depth and that is accomplished through inspections and
evaluation of inspection results to understand the current
condition of the reactor vessel and any potential future
degradation modes.
Moving on to the next viewgraph, just a little bit
more about the augmented inspection rule. Going back in
history to the early to mid-'80s, relief had been granted to
the boiling water reactors for performing some of the code
required examinations. These were granted under 5055(a) of
the regulations. The main reason was the inability to
access these locations. The tooling just wasn't available.
However, and the Staff recognized the small amount
of inspection that was being performed and, also, at the
same time, advances in inspection capability that had
occurred, and some of this in particular was overseas where
. 43
we found that people were doing more examinations, and also
recognition of some viable degradation mechanisms that I
will talk about later, the decision was to promulgate this
rule.
Did require expedited implementation of
inspections. This is basically what was required by the
ASME, except on a faster schedule because of the concern
that time had gone by without any significant inspections.
It revoked all the prior reliefs that had been granted and,
as I indicated, these were granted largely on the basis that
they were just physically unable to do the examinations and
it was related to tooling.
Some of the units at that time had inspected less
than 10 percent of the shell welds and that is still true
today. Even though, as you heard in the earlier
presentation, there has been a fairly substantial sample of
welds inspected, there are plants out there that have not
looked at 10 percent of the shell welds in their plants.
I'm sorry, have looked at 10 percent or less.
So the rule was promulgated in '92. The one major
comment, public comment that was received on the rule was to
provide some flexibility in schedule, specifically for those
plants that were near the end of the 10-year inspection
interval, that they wanted some flexibility in being able to
implement this, do some planning and develop the appropriate
. 44
tooling. So, in fact, the rule was modified such that
plants that were within 40 months of the end of the 10-year
interval could go to the next interval, next first period of
the next interval. A little bit complicated, but we gave
them some extra time to implement the inspections.
Also, it was recognized that even with
improvements in some of the tooling and inspection
capabilities, that there still may be some areas which are
inaccessible and we are talking about where there are lugs
or attachments physically inside the vessel such that you
just can't get to the weld that you want to examine.
Moving on to the next viewgraph, I want to talk
about the need for inspections. First, I would point out
the purpose of the reason we perform inspections, just in
general. We want to identify problems that we didn't
anticipate and, as was noted earlier, prediction of
degradation in other components has not always been real
reliable. Although in hindsight, some of these degradation
modes can be explained, it was really inspections and
inspection activities that identified them and examples
include stress corrosion cracking in BWR piping.
When this issue first came up, it showed up in
some small diameter piping and the thought at the time was
it wouldn't happen in large diameter piping. Inspections
confirmed eventually that it did.
. 45
BWR internals, there have been a number of areas
where cracking has been found through inspections and that
includes, for example, the access cover holes in the inside
of the vessel, the core shroud, which has been getting a lot
of attention lately.
So one of the things we want to do is identify
things we haven't anticipated. The other thing is that the
evaluation of the inspection findings is really a proactive
way of looking at the condition of the vessel and, as I said
earlier, looking at what potential degradation could
possibly occur in the future.
So when indications are found, and it was
mentioned in some of the recent examinations indications
have been found, they were evaluated, they were found
acceptable by the code which is what we would expect, that's
what we want. But we also look at those and say, well, what
kind of degradation is it? Yes, it is subsurface, it is not
exposed to the environment. So, you know, we don't have to
be as concerned about that as if it were open to the
environment and might therefore see some more aggressive
growth.
So those are some of the reasons we do the
research.
CHAIRMAN JACKSON: Let me ask you a question. Is
the code meant to be predictive? I mean, is there an
. 46
established relationship between code-identified
degradations and failures?
MR. STROSNIDER: I would say the answer to that is
no. There is -- there is work going on now in the risk
informed arena which I think is taking into account more
looking at what areas as susceptible and what the
consequences are. But I think when some of the early code
inspection requirements were developed, it was largely just
go out and do a sample across the system. For example, look
at 25 percent of the reactor cooling system welds, class one
welds, pick those and that should be an adequate sample to
tell us if there are any problems.
CHAIRMAN JACKSON: Commissioner McGaffigan?
COMMISSIONER McGAFFIGAN: Could I ask, why
wouldn't sampling work in this instance, when their
probabilities are ten to the minus fortieth, I haven't seen
those since I was studying neutrino cross-sections some time
ago.
CHAIRMAN JACKSON: Yeah, we know about those.
COMMISSIONER McGAFFIGAN: Which are small.
But why would -- they are proposing no testing of
or inspection of the circumferential welds but why -- why
wouldn't a sampling technique be adequate?
MR. STROSNIDER: It is a good question. It is one
that we have considered. I will get to that, but I will
. 47
give you a little preview, which is basically that reactor
pressure vessels and the reactor pressure vessel welds are
not all the same. Okay? You have to realize that there was
a discussion about the sort of inspection that was done
during fabrication of the vessels. However, that inspection
was different, whether it was radiography or surface, in
some cases ultrasonic. It changed as the code changed in
time. So not all vessels saw the same fabrication
inspections.
The welds made in the vessels because of the
fabrication process are different. For example, there was a
question earlier about are the circumferential welds
different than the axial welds. When you look at the
process for fabricating these vessels, the ring sections are
made up of plates and there is an automatic process once the
ring section is laying down the cladding, welding process.
Then the rings are welded together and, in most cases, the
back cladding as it is called, the cladding over the welds
that join the ring sections together, were done manually.
So there is a difference.
In the manual welds, what we have seen is that
they are not controlled as well, the heat input may be more
difficult to control and those may be areas that are more
susceptible to degradation. Also, some of the issue that
comes up is repair. There have been and it was indicated
. 48
repairs were made during fabrication.
There are a number of different vendors or shops
that were involved in fabricating these vessels. At least
four. Some of the vessels actually went through one, two or
in one case three of those shops during fabrication. The
vessel was partially fabricated, shipped to another vendor
for additional fabrication and shipped to another one to be
finaled.
So there is a question about whether the welds we
are looking at really represent a homogeneous statistical
population, to which you could apply sampling. And one of
our concerns is that where repair welds may have been made,
that those are particular areas that ought to be looked at.
And we think the best way to catch that is by doing a one
time base line examination.
You know, we have to keep that in perspective. We
do not expect that there are significant, huge flaws in
these reactor vessels or I would be here taking an even more
aggressive decision on this. But we do recognize from the
evaluations that have been done that there is the potential
that the wrong -- the wrong elements could wind up in the
same place. It is a low probability. But we believe that
it is appropriate to go confirm the assumptions that are in
the analysis and the evaluations to make sure it really is
as low as we think it is.
. 49
The situation we are talking about, and even in
the industry's assessment, they talk about the potential for
stress corrosion cracking in the cladding, lining up with
some pre -- some fabrication defect that is in the
underlying base metal. And perhaps if you go on beyond that
and say, well, this was the area of a large repair, was the
stress relief, post-repair stress relief effective, what
kind of environment are you in in a particular plant? If
you add all those up in the wrong place, you might have the
potential for a viable degradation mechanism. And a large
part of our conclusion is we ought to verify that that
doesn't exist out there.
CHAIRMAN JACKSON: Commissioner Diaz?
COMMISSIONER DIAZ: Yes, just in the same vein,
wouldn't a 100 percent examination of the longitudinal welds
provide you with a very reasonable sample of how the
pressure vessel is standing up?
MR. STROSNIDER: What I am suggesting is that the
circumferential welds and the axial welds are not
necessarily the same population of welds because of
differences in fabrication.
COMMISSIONER DIAZ: I know, but that is not the
question. The question is, wouldn't a 100 percent
examination of longitudinal welds give you a very good
program to verify at least, you know, a portion of the
. 50
industry's analysis?
MR. STROSNIDER: I am sure you could make some
statistical inferences from that if you understood how many
repair welds were in that sample versus how many repair
welds are in the circumferential welds, things of that
nature.
CHAIRMAN JACKSON: Are you saying that is not
known?
MR. STROSNIDER: I would say, number one, it
hasn't been analyzed. It would take a tremendous amount of
effort to pull out all those records. We also -- one of the
bullets on the next viewgraph talks about the concern for
undocumented repairs.
I would point out that what we have also concluded
is following an initial base line to verify the condition of
the vessels that a sampling program may in fact be
appropriate depending upon the results of that base line
example.
COMMISSIONER DIAZ: Define a base line.
MR. THOMPSON: Our definition was essentially a
100 percent of accessible. Essentially 100 percent.
MR. STROSNIDER: Let's move on to viewgraph number
six and some of this I think I may have already covered in
response to questions.
I want to point out that inspections have
. 51
identified degradation in reactor pressure vessels and these
are some of the instances that, in fact, were called out in
the backfit analysis that supported promulgation of the
rule.
At Hatch One, there was some pre-service
ultrasonic testing done. This was actually in the industry
report, which identified defects in the recirculation to
shell weld nozzles that required repair so they had to be
ground out and repaired.
COMMISSIONER DIAZ: I'm sorry, I couldn't hear
you.
CHAIRMAN JACKSON: Hatch One.
MR. STROSNIDER: Yes, at Hatch One during
fabrication inspections, ultrasonic testing did identify
defects in the recirculation nozzle to shell weld that
exceeded -- from what I can read it exceeded the code
acceptance criteria and required repair. So there were
defects in some of these vessels during fabrication. There
were repairs made. And there were varying degrees of
inspection.
COMMISSIONER DIAZ: But a nozzle is always a high
stress point so it is not the same as the rest of the
vessel.
MR. STROSNIDER: True, but this was not service
induced. This was fabrication. And it may be a more
. 52
difficult spot to weld, that's true.
The state of the art inspection methods have
identified indications requiring code evaluation. I have
heard mention of Brown's Ferry did inspections in 1993.
They were using state of the art inspection methods.
Fifteen indications required evaluation by code. They would
not have been evaluated under the old inspection procedures
but they were under the new, detected and evaluated under
the new procedures. They were found acceptable; they were
subsurface.
In 1995, Pilgrim also performed a state of the art
inspection. They found no indications requiring flaw
evaluation and this is the information we have available. I
wanted to point that one out because in terms of the reactor
vessels being similar and there are differences, these were
in fact made by different vendors, different results from
the inspections.
With regard to viable degradation mechanisms
existing, first, it is a given that the BWR environment is
an aggressive environment. It can support crack growth.
Certainly in stainless steel, we have seen this in piping
and internals. Ferritic, as was indicated, some of the
early data show that stress corrosion could be supported in
some of the ferritic base metal. Some of the more recent
data says no, there is some mixed results on that.
. 53
With regard to actual experience, there was a
mention of the Quad Cities Unit Two, indications that were
found in 1990. These were not in a shell weld, they were in
the flange, the head weld. There were 34 surface flaws
found during that inspection. The longest one was 30 inches
long. It penetrated, at its deepest point, through the
cladding and about two-tenths of an inch into the heat
effective zone in the base metal. So about seven-tenths of
an inch deep.
CHAIRMAN JACKSON: Is there a difference between
the, you know, are there sufficient differences between the
construction of the reactor vessel head and the reactor
pressure vessel itself to make the head more susceptible to
these degradation mechanisms?
MR. STROSNIDER: Using the same welding processes,
there may be some difference, perhaps, in how easy the fit-
up is and I can't say there is anything particularly or -- I
don't know, staff is shaking their head no difference.
I can't really add anything beyond that.
COMMISSIONER DIAZ: The environment is not the
same.
MR. STROSNIDER: No.
COMMISSIONER DIAZ: There is a different
environment in the head.
MR. STROSNIDER: There is a different environment.
. 54
That is certainly true, in that you are in a steam
environment in the head.
I just comment, we got into looking at differences
in environments on the core shroud where we thought all the
cracking was going to be up high because of the more
aggressive environment and it didn't turn out that way.
What you have to remember is you have a lot of
competing parameters in developing and sustaining cracking
and it includes the environment, it includes the stresses,
it includes the material properties and it -- you have to be
careful in trying to assume you know how all those are going
to come together.
So that was the experience at Quad Cities. It was
evaluated that that flaw was found that it was acceptable as
it was found. There was some grinding done on it to smooth
it out and then it was found acceptable for continued
service. But the grinding, of course, reduces the stresses
there and makes it less susceptible to any continued growth.
The backfit package that went along with the rule
in 1992 referenced some experience with stress corrosion
cracking in feedwater nozzles siphons where again cracking
was initiated in stainless steel but grew into the ferritic
material. It occurred at Brunswick and also at a Chinese
plant.
Finally, this one was interesting, Fitzpatrick,
. 55
this was also I believe in 1990. They found a surface crack
in the reactor vessel head. This was higher in the head
than at the flange weld. Interesting. This was an unclad
head. There was no stainless steel cladding on this vessel.
When they went back and took a close look at this,
it turned out that the surface indication that was there was
some sort of fabrication scratch or defect. It happened to
be in the area of some subsurface slag inclusions that were
about 12 inches in length. The maximum depth at that
location was about two inches.
Those appear to have been fabrication, not service
induced defects but one of the things that we heard and that
we have been considering is what's the likelihood that the
wrong situations could add up at the same time. This is in
a location where, in all likelihood, had it been clad it
would have been done manually and those are areas where we
know there is a greater susceptibility to stress corrosion
cracking of the clad and if that sort of crack joined up
with this sort of preexisting defect, it might be a concern.
As you heard, the analysis does make an
assumption, okay, that in fact if you grow through the clad,
you sample from a distribution and have that match up with
some fabrication defects. One thing I point out here to
recognize is a lot of the Monte Carlo analysis is often
assumes independence of all these different parameters. In
. 56
this case, they have tried to address that but I think the
point is there may not be independence because some areas
are just more susceptible to having these adverse
conditions.
COMMISSIONER DIAZ: May I make a comment?
CHAIRMAN JACKSON: Please.
COMMISSIONER DIAZ: You know, this is not my area.
I am here, you know, apples and oranges. You are mixing
flanges and heads that are carbon steel that are not, you
know, stainless steel with defects from manufacture and
putting all that together in the context of the reactor
pressure vessel. And I don't think they are the same thing,
you know, from the little of what I know. I think they are
completely different issues.
I mean, we know that there is a stress corrosion
cracking issue with boiling water reactors and we have
always known that. They have taken care of that.
Now, the question is, have we ever found a
deficiency or degradation in a reactor pressure vessel, in a
boiling water sufficient to say, hey, this is not acceptable
and you have to do something about it? Have we ever found
one?
MR. STROSNIDER: I am describing what has been
found and the inspections that were performed.
COMMISSIONER DIAZ: No, you have not said that
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there is one that has actually been significant to the point
that it is not acceptable to the staff or, at least, that is
what I heard.
MR. STROSNIDER: That's correct.
COMMISSIONER DIAZ: So all of them have been
acceptable to the staff so the staff concluded that they did
not really degrade to the point that it posed a safety
question; is that correct?
MR. STROSNIDER: That is absolutely true and as I
indicated earlier, that is our expectation. I hope that we
never find and I don't think we will find flaws in a reactor
vessel that compromise its integrity.
COMMISSIONER DIAZ: The if is not the issue. The
question is, have you found one and I guess your answer is
no.
MR. STROSNIDER: No, we have not found one.
COMMISSIONER DIAZ: Thank you.
MR. COLLINS: Commissioner, I guess it is
important to know that I think part of what Jack is trying
to stress is because we have not done the 100 percent
examinations we have not established a base line which would
indicate what the potential is for that to occur other than
an in-process issue, which would be a leak. And, of course,
that has been avoided.
COMMISSIONER DIAZ: I understand the difference.
. 58
COMMISSIONER ROGERS: Just before you leave that,
though, it does seem to me that you have -- you do have a
total disagreement with the industry on this question of
whether there is a viable degradation mechanism for welds.
I mean, you have cited a number of examples of degradations
that you have found but I didn't hear you mention any in a
weld.
Their statement, their concluding statement was,
an absence of degradation mechanisms substantiates vessel
integrity, dot, dot, dot. And you are saying there is a
viable degradation mechanism and so it seems to me there is
a total conflict on that issue.
MR. STROSNIDER: Yes, and the real issue here,
first of all, there is a degradation mechanism which
everyone acknowledges in the stainless steel cladding.
There are cracks that have been found, service induced in
the cladding. The question is, will it grow into the
ferritic base metal, all right? And as I indicated, and I
think as was indicated in their presentation, some of the
early data indicate that you could grow cracks if you have a
high enough driving force. Some of the more recent data
says, no, you wouldn't expect that.
All right.
We have not seen an example where it has really
been given a chance. Probably the closest was quad cities.
. 59
That was found early in the inspection and the defect was
corrected. The analysis that the industry did did suggest
that if you had cladding flaws growing into significant
fabrication defects where you get a high enough driving
force, something like 30 KSI root inch applied stress
intensity factor that there could be a mechanism.
So, as I indicated, the data are not all that
clear, all right? And given that uncertainty, our
conclusion is that we should go take a look.
The last thing on this viewgraph I wanted to talk
about was the potential for undocumented repairs. I am not
sure how much difference it makes whether they are
documented or not in terms of the potential for degradation
although, as was said, there was a lot of work done, a lot
of procedures in place to document this sort of thing.
However, the research office says the reactor
vessel down at Oak Ridge National Laboratory which we have
been doing examinations on, looking at welds, looking at
density of defects and that sort of thing. And one of the
things they found in that reactor vessel was a significant
repair to one of the shell welds which was not documented.
It was not in the documentation that we acquired with the
vessel. I don't know if Mike wants to expand on that at all
but --
MR. MAYFIELD: Just that it turned out to be a
. 60
quite large defect or repair, in some cases according to the
laboratory running as much as three-quarters of the way
through the wall thickness. It spanned several feet. The
only indication in any of the documentation is that there
were -- there was a repair based on high-low mismatch that
you get when you line up the two rings but there was
certainly no suggestion of the extent of this repair in any
of the documentation that we acquired.
COMMISSIONER ROGERS: Would that have been done at
the time of fabrication?
MR. MAYFIELD: Yes, sir.
COMMISSIONER DIAZ: Of course, repairs are part of
the industrial process.
MR. MAYFIELD: Yes. And, in and of itself, we
weren't bothered by it. It is just that it is one more bit
of information that feeds into this puzzle.
CHAIRMAN JACKSON: Okay. Let's move on.
MR. STROSNIDER: Moving on to viewgraph number
seven, again, the need for inspections, the conclusion that
we reached here is that we think again a base line
inspection, which I will define as essentially all the welds
they can get access to and take a look at is appropriate in
order to verify the low probabilities that we are seeing.
As I said earlier, you are not going to hear a
condemnation, general condemnation of the analysis that was
. 61
done by the industry. We think it had a lot of insights and
that there is a lot of merit to it but we do think there are
enough questions, looking back at the history, that it is
appropriate to go do that sort of base line examination.
What we are looking for is what we consider a very
low probability event. But we are talking about the reactor
pressure vessel and we feel that the safety significance of
the vessel warrants doing that sort of inspection.
Having done that, we do think that the analysis
that has been presented, after we look at the results of
that base line, provide perhaps good basis for going through
a sampling inspection and that could mean significant impact
on the resources expended in subsequent intervals.
Going on to slide number six, just a discussion on
the NRC and the ASME code inspection philosophy. You heard
some of this. Basically, the code has evolved over time.
It currently does require 100 percent inspection,
essentially 100 percent inspection, which means 90 percent
recognizing some of the limitations. Anything less than 90
percent requires actually some granting of relief or
alternative by the NRC under 5055(a).
I should point out that some of the NRC certainly
was a proponent in some of these code changes that went to
larger examination percentages. But our position has been
consistent with the ASME code for some time which, actually,
. 62
since 1975 has required at least 100 percent base line
examination. Essentially 100 percent.
You heard that the industry is pursuing with the
ASME codes some changes in these requirements. In fact, we
encourage that in one of our letters, particularly with
regard to those inspections that might be performed
subsequent to a base line.
CHAIRMAN JACKSON: Is that to say then that if the
code is changed, the staff will change its position?
MR. STROSNIDER: No.
But we will certainly assess the changes in the
code and see through our rulemaking process if that is the
appropriate answer. And, as I said, we have encouraged
after a base line examination the notion that the
evaluations performed support a sampling sort of inspection.
CHAIRMAN JACKSON: Where in the process is the BWR
owner's group in its request to change the code? I mean,
how far along?
MR. STROSNIDER: As Mr. Dyle indicated, it has
been through several committees. I am not sure I can give
you all the way up through the subcommittees.
CHAIRMAN JACKSON: I mean, how much longer do you
think this is going to take? Is it hard to predict?
MR. STROSNIDER: I don't know. Is there someone
who was at the code meetings from the staff that can address
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that?
Gil Millman?
MR. MILLMAN: Pardon my laryngitis; I have been at
code meetings for the last week.
This particular code case did come up to the
Subcommittee on Nuclear In-Service Inspection last December.
At that time, Mr. Dyle withdrew it and on the basis that it
would go forward only when there was a technical basis
document supporting it and so it waits at the subcommittee
for that action.
CHAIRMAN JACKSON: I see.
Commissioner Diaz?
COMMISSIONER DIAZ: I don't know whether the
question is valid any more but you said no to whether this
type of change in the position, you know, regarding the
ASME. Does that mean the staff's position is independent of
the ASME?
MR. STROSNIDER: Well, in general.
COMMISSIONER DIAZ: In total?
MR. STROSNIDER: In general, the process that we
go by is the Code of Federal Regulations endorse industry
codes and standards. Sometimes we endorse those with some
exceptions or with some additions and my comment is
basically that we will not only observe but we have people
who will participate in the code activities and make sure
. 64
that our concerns are identified early.
When the code reaches conclusion, either in a code
case or in a change to the code, we will assess that as part
of the rulemaking process and see how it would be endorsed
in the regulations.
But we don't -- it is not a given that we just
take it the way it's --
CHAIRMAN JACKSON: Have there been cases where the
staff -- the staff's position has not been consistent with
the code and the staff has come out with a more conservative
position?
MR. STROSNIDER: Yes.
MR. COLLINS: Yes.
CHAIRMAN JACKSON: Okay.
MR. STROSNIDER: One other comment is we did -- we
went out last week basically a poll looking to see what the
positions are internationally with regard to this type of
inspection. We have three responses so far, one from MITI,
the Ministry of Industry and Trade in Japan. They require
100 percent each 10 years, every 10-year interval --
CHAIRMAN JACKSON: Of vertical --
MR. STROSNIDER: Of the shell welds.
CHAIRMAN JACKSON: Of all of them?
MR. STROSNIDER: Yes, longitudinal and
circumferential.
. 65
COMMISSIONER ROGERS: BWRs as well as PWRs?
MR. STROSNIDER: Yes.
We do understand also that there is some
discussion with their industry about possibly changing that
at some point.
The Spanish do 100 percent of axial and
circumferential each 10 years and also in Sweden they do 100
percent.
I would point out that a lot of this is driven by
what is in the ASME code and that is an international code
so there are other countries who follow that and in fact do
follow pretty much what the NRC is doing.
I would also point out, though, that Sweden has
been leading, perhaps, in the area of risk-informed in-
service inspection and they still do this sort of
inspection.
Viewgraph nine, talking about granting relief and
I think the main point I wanted to make here is that we
recognize that certainly with the current tooling there are
some limitations as to what can be inspected.
In the industry submittal, they talk about,
however, some of the improvements that have been made and
they talk about an inspection in 1983 at a BWR 3 facility
where they were able to get 41 percent of the
circumferential welds and 52 percent of the longitudinal.
. 66
In a more recent 1993 examination, this was at a BWR 4 so
there might be some slight differences, but they achieved 78
percent of the circumferential welds and 91 percent of
longitudinal. So there has been progress in terms of the
tooling and the technology.
You also heard mention the device that has been
demonstrated at the EPRI NDE center that was developed by
the Japanese. You understand there is at least one U.S.
company looking at commercializing that in the U.S. and it
is basically a submersible device which is, as I understand,
self-propelled and can move around. It is very thin. The
word we got is it could get probably 90 percent of the welds
in most of the vessels out there. I don't know how far that
is from actual implementation. We know there have been
demonstrations at the NDE center and they are ongoing in
Japan.
I think the point here is that progress can be
made in terms of improving the inspection technology. And
some of this, again, we haven't seen all the details but it
sounds like it would have reduced setup time and even
personnel exposure as opposed to putting big manipulators on
top of the vessel, being able to put in some submersible
which you can operate from some distance.
COMMISSIONER DIAZ: That is not commercially
available in this country. Will it be in the next five
. 67
years?
MR. STROSNIDER: Not right now, no. And I don't
know. Like I said, the industry is following that. As
Mr. Dyle indicated, they are aware of it.
COMMISSIONER DIAZ: In other words, it is a long
term thing. It is not something that is going to be
available next year?
MR. STROSNIDER: I don't know what the schedule
is. As I said, it has been demonstrated and is -- there are
some in-vessel demonstrations going on in Japan.
CHAIRMAN JACKSON: I've seen it. EPRI is working
on it.
MR. STROSNIDER: So with regard to granting
reliefs and, as I pointed out, the rule does -- and it
specifically included, and I am looking at slide number 10
now --
CHAIRMAN JACKSON: Let me go back to slide nine.
You say the industry proposal is for NRC to grant a large
number of reliefs from requirements based largely on
probabilistic assessments and I note that in your paper, the
Staff stated that it had concluded that rejection of the
project's probabilistic arguments to support authorization
of inspection alternatives, et cetera, is consistent with
the Commission policy on the use of probabilistic risk
assessment.
. 68
Can you explain, you know, the basis of that
statement and is the staff's current position risk informed
and can you relate that to ongoing efforts with respect to a
risk-informed ISI and IST, okay?
MR. STROSNIDER: A statement that was in the
Commission policy, let me see if I can actually get the --
well, this I can just read. This was a quote from the
Commission policy statement that use of probabilistic risk
assessment methods, the staff used the safety goals in
making regulatory decisions regarding backfitting new
generic requirements but not to make specific licensing
decisions including granting relief from unnecessary
requirements.
That is a quote from the policy statement.
MR. COLLINS: It is on page 4.
MR. STROSNIDER: August 19, 1995. I was looking
for the policy statement but it is in the paper.
But I guess I would also point out that, to try to
keep this in context, the evaluation that was submitted by
the industry is really not full-blown risk assessment. It
doesn't go out to the consequence stage administration that
sort of thing. It doesn't assess what happens if you have a
leak, for example, and it does include some deterministic
arguments with regard to fabrication and that sort of thing.
So it is sort of a mix.
. 69
But we thought that was an issue that we at least
questioned when we looked at it and said, well, is this an
appropriate basis for granting release and it would be
release for essentially all the BWR plants. Does it
maintain defense in depth as we think is appropriate?
CHAIRMAN JACKSON: Commissioner McGaffigan?
COMMISSIONER McGAFFIGAN: The difference is 30
orders of magnitude between longitudinal welds and
circumferential welds and in their analysis. You have gone
through a long explanation as to why there might be
something there that no one has foreseen and therefore you
want to inspect them all but 30 orders of magnitude, have
you looked at that difference and that analysis and found a
flaws in it?
MR. STROSNIDER: There are no specific problems
that we have identified in the way the analysis -- in the
modeling itself. It has to do with looking at assumptions,
input parameters and, quite frankly, our experience in
trying to predict what may or may not happen. I refer back
to some probabilistic assessments on piping and that sort of
thing where people failed to take into account loadings and
they found degradation. They weren't in the model.
So one of the reasons you do inspections is to
find out what you are not smart enough to put in your model.
As I said, you are not going to hear a
. 70
condemnation of the analysis that they have done and it does
show a significant difference.
MR. THOMPSON: Commissioner, to get to your point,
as Jack explained, we are dealing primarily with the up-
front assumptions that you predicate that risk questions on
and the uncertainties that are involved as articulated by
the staff here with the fabrications and the records and the
history and the repairs and the lack of a base line. Lacing
that base line, the staff really is missing a key piece of
information to predicate the change under 5055(a) which is
allowable if you are able to meet the statement of an
acceptable level of quality and equivalent acceptable level
of quality and safety. That is essentially where we are.
MR. STROSNIDER: On viewgraph number 10, I just
briefly indicate that, as I said earlier, that the rule
acknowledged when it was promulgated that there could be
some areas that are difficult to access and in fact the
wording in the augmented inspection rules where people are
unable to do inspections, they may propose alternatives.
Quite frankly, it takes a little bit of thinking
but it is our assessment of the industry's proposal, we
think, that proposal can be used to justify some of these
areas where you just can't access them. But we also think
that in terms of defense in depth that you should do the
scope of the inspection that you can do.
. 71
So, the conclusions are that we -- again, we think
the industry's analysis has merit. It has added a lot of
insights to pressure vessel integrity issues. We have
concluded for the reasons we just discussed that a base line
examination of those welds that can be accessed should be
performed. That the report and the work they've done can be
used to support relief where, in fact, they just can't
access some of these welds and that future modifications to
the inspection requirements may be appropriate after
completion of the base line.
It would be our plans to complete that in a safety
evaluation that we could issue in probably about six weeks
or so.
CHAIRMAN JACKSON: Let me ask you three questions
quickly. If uncertainty isn't in part influencing the
staff's position, are there alternatives such as pilots or
targeted implementation or some other strategy to provide
some additional information to support the staff's position?
MR. STROSNIDER: Well, I think the question came
up. One obvious thought that comes up there is could you
deal with this on a sampling basis and draw inferences from
the sampling basis. And --
CHAIRMAN JACKSON: That is one example. But one
could take a -- and I guess this is a different -- it
depends on what you mean by sample. You could take all the
. 72
plants and have a sample of areas. You can take a subset of
plants and do 100 percent. That's a sample. Et cetera, et
cetera.
Have you given some thought to these kinds of
alternatives?
MR. STROSNIDER: Well, we thought about that and,
again, the conclusion we reached was do as much as you can
at this point and then look at a sampling basis because
after you have gone through and looked at all the welds and
confirmed the -- you know, really given confirmation of the
quality that was there when they were originally fabricated
and, as we pointed out, there have been improvements in
inspection techniques, we can see things today we couldn't
see then, you have confirmed that in fact you don't have all
the wrong conditions at the same location, you have
confirmed that there is something you didn't anticipate,
then you basically we think can go to a sampling method
where you are monitoring for any sort of degradation that
might show up.
CHAIRMAN JACKSON: So basically you want a
database which you believe you don't have at this stage of
the game, is that the point?
MR. STROSNIDER: Yes.
CHAIRMAN JACKSON: Have you discussed this at all
with the ACRS?
. 73
MR. STROSNIDER: We have not had any recent
discussions. The ACRS was involved in the original
promulgation of the rule back in '92. They looked at that
and supported it, as I understand it.
CHAIRMAN JACKSON: Do you intend to document the
technical basis for your rejection of the industry group's
proposal then in a safety evaluation report?
MR. STROSNIDER: Right. We would document the
discussion basically that I just gave you and a safety
evaluation which I would expect to complete in about six
weeks.
CHAIRMAN JACKSON: And what kind of time line are
we operating under?
MR. STROSNIDER: Well for, as I say, issuing the
safety evaluation, I would put a target of about six weeks.
It is important, and I think the industry pointed
out, when you look at the rule and where the plants are in
their inspection intervals, that many of these examinations
would need to be performed in the next year or two and the
planning has to be done, equipment has to be available. So
we recognize that a decision of position needs to be made
sooner rather than later.
CHAIRMAN JACKSON: Okay, is that it?
MR. THOMPSON: That concludes our presentation.
We would be prepared to answer any questions.
. 74
CHAIRMAN JACKSON: Commissioner Dicus, questions,
Commissioner Diaz?
COMMISSIONER DIAZ: Yes, I just have a quick
comment. Knowing the difference between these reactors and
the difference between circumferential and longitudinal
welds, I actually don't see, although you might have it in
six weeks, a basis for denial of the industry request. It
seems to me like 100 percent longitudinal inspection program
with some beef behind it, I mean, to get it done in a very,
you know, reasonable period of time will provide a good base
line. And from there, during that period of time, we might
be able to develop a program that will provide some basis
for the circumferential welds.
I actually see no technical information that has
been presented that says this is, you know, unreasonable or
is not adequate protection of health and safety. Because
most of the things that have been presented are peripheral
to the main issue of how the pressure vessel is attacked and
how are the -- you know, the differences in stresses between
circumferential and longitudinal welds.
So unless I see something different, I don't see
why a program that actually addresses 100 percent
longitudinal wells as soon as possible, will not be a good
base line to consider, you know, than the circumferential
welds.
. 75
CHAIRMAN JACKSON: I would like to thank the
representatives of the BWR Vessel and Internals Project and
the NRC staff for briefing the Commission regarding the
issues associated with the staff's technical position
regarding alternatives for augmenting inspection of the
reactor vessel. As I mentioned in my opening remarks, you
know, the Commission is not a commission of technical
experts and so, I don't believe in an hour and a half we can
sit here and sort through all of that. It is important for
the Commission to understand aspects of the technical basis
for the staff's position so that if there are any policy
issues involved, the Commission can make informed decisions.
It is also important for the public and the
industry and as well, as the discussion today has revealed,
the international regulatory community to understand the
staff's positions. So given the recognition of the
important role that the reactor pressure vessel does play in
implementing the Commission's defense in depth philosophy
but given that you have even said yourself that the project
has proposed some technically sound discussions for
implementing a reduced scope augmented inspection, the staff
should complete, on an expedited basis, the development of
the safety evaluation report on the Boiling Water Reactor
Vessel and Internal Project proposed alternative and to
consider whether there is a tiered approach to getting at
. 76
the issue. And if it is not technically possible, you
should tell us that.
This safety evaluation report would then serve as
the staff's documented and defensible basis for resolution
of the issues and any -- document any open issues and would
facilitate any commission decisions if they are appropriate
on any of the related policy issues.
So unless there are any further comments, we are
adjourned.
[Whereupon, at 4:50 p.m., the meeting was
concluded.]
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