[Federal Register Volume 63, Number 119 (Monday, June 22, 1998)]
[Notices]
[Pages 33968-33974]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-16539]
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NUCLEAR REGULATORY COMMISSION
[Docket Nos. 50-361 and 50-362]
Southern California Edison Company, et al.; San Onofre Nuclear
Generating Station, Units 2 and 3; Issuance of Director's Decision
Under 10 CFR 2.206
Notice is hereby given that the Director, Office of Nuclear Reactor
Regulation, has acted on a Petition for action under 10 CFR 2.206
received from Mr. Stephen Dwyer dated April 25, 1997, for the San
Onofre Nuclear Generating Station (SONGS), Units 2 and 3.
The Petition requests that the Commission shut down the San Onofre
Nuclear Generating Station pending a retrofitting of the steam
generators. As a basis for the request, the Petitioner asserts that the
ability of the steam generators to withstand a major seismic event is
seriously compromised by the degraded eggcrate supports discovered in
the SONGS Unit 3 steam generators.
The Director of the Office of Nuclear Reactor Regulation has
determined that the request should be denied for the reasons stated in
the ``Director's Decision Under 10 CFR 2.206'' (DD-98-06), the complete
text of which follows this notice and which is available for public
inspection at the Commission's Public Document Room, The Gelman
Building, 2120 L Street, N.W., Washington, D.C. 20555-0001, and at the
Local Public Document Room located at the Main Library, University of
California, P.O. Box 19557, Irvine, California 92713.
Dated at Rockville, Maryland, this 11th day of June 1998.
For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
Director's Decision Under 10 CFR 2.206
I. Introduction
By e-mail dated April 25, 1997, Stephen Dwyer (Petitioner)
requested that the Nuclear Regulatory Commission (NRC) take action with
regard to San Onofre Nuclear Generating Station (SONGS) regarding his
concerns about the ability of the SONGS steam generators to withstand a
major seismic event.\1\ Specifically, the Petitioner stated that the
ability of the SONGS steam generators to withstand a major seismic
event is seriously compromised by the degradation observed in the SONGS
Unit 3 steam generator internal tube supports (eggcrate supports)
during its 1997 refueling outage. The Petitioner requested an
investigation to determine if Unit 2 has experienced degradation
similar to that found in Unit 3 and also stated that further seismic
analysis should be performed for the SONGS steam generators and that a
retrofitting upgrade of the steam generator supports could be
accomplished at this time. On June 26, 1997, the NRC staff acknowledged
receipt of the Petition as a request pursuant to Section 2.206 of Title
10 of the Code of Federal Regulations (10 CFR 2.206) and informed the
Petitioner that there was insufficient evidence to conclude that
immediate action was warranted. Notice of the receipt of the Petition
indicating that a final decision with respect to the requested action
would be forthcoming within a reasonable time was published in the
Federal Register on July 3, 1997 (62 FR 36085).
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\1\ The Petitioner sought to add this concern to his Petition
dated September 22, 1996, wherein he requested the NRC to shut down
the SONGS facility ``as soon as possible'' pending a complete review
of the seismic design of the SONGS units based on information
gathered from the Landers and Northridge earthquakes. By letter
dated June 26, 1997, the NRC advised the Petitioner that his e-mail
request dated April 25, 1997, concerning the ability of the SONGS
steam generators to withstand a major seismic event, would be
treated as a separate 10 CFR 2.206 Petition. The Director's Decision
(DD-97-23) issued by the NRC on September 19, 1997, denied the
Petitioner's September 22, 1996, request to shut down the SONGS
units, providing a detailed discussion of the adequacy of the
seismic licensing basis for the SONGS facility.
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My Decision in this matter follows.
II. Discussion
A. Request for an Investigation to Determine if SONGS Unit 2 Has
Experienced Eggcrate Degradation Similar to Unit 3
1. Background
The SONGS units utilize Combustion Engineering Model 3410
recirculating steam generators. This model of steam generator contains
9,350 Inconel 600 (ASME Material Specification SB-163) U-tubes with a
nominal diameter and wall thickness of 0.75 and 0.048 inch,
respectively. Secondary side tube support structures consist of seven
horizontal full eggcrate supports, three horizontal partial eggcrate
supports, and upper bundle supports (i.e., two batwing diagonal
supports and seven vertical supports). The materials used for
fabrication of the steam generator vessels and internals (including
tube supports) are low-alloy and carbon steels, respectively. Figure 1
is a simplified cross-sectional diagram of the SONGS steam generators
that clearly displays the 10 eggcrate support levels, and Figure 2 is a
three-dimensional representation of the steam generators that gives
additional structural detail.
The eggcrate supports consist of 1- and 2-inch carbon steel strips
interlocked perpendicular to each other as shown in Figure 3. The
eggcrate supports limit lateral motion of the tubes and, at the same
time, allow free flow of fluid around the tubes.
During the 1997 refueling outage for SONGS Unit 3, the licensee
discovered that portions of the eggcrate supports had experienced
degradation, ranging from minor wastage of the eggcrate material to
severe thinning in localized areas. The significant degradation
observed during this refueling outage was confined mainly to the
periphery locations of the eggcrate supports. The secondary sides of
the steam generators in both units were inspected during their 1997
refueling outages and during their 1998 mid-cycle outages and, as
discussed below, significant degradation was limited to the periphery
locations of the SONGS Unit 3 eggcrate supports.
The licensee has extensively researched the cause of the eggcrate
degradation and has concluded that the degradation was caused by a form
of flow accelerated corrosion (FAC), a general term describing
processes that use assistance from fluid flow to remove the protective
oxide layer from base material. Removal of the protective oxide layer
exposes the base material to the fluid environment, allowing further
material removal through corrision and/or erosion processes. The carbon
steel
[[Page 33969]]
eggcrate material utilized in the SONGS steam generators can be
susceptible to FAC in the presence of sufficiently high fluid
velocities.
The licensee concluded that the FAC occurred during recent
operation of Unit 3 primarily as a result of steam generator secondary
side increased fluid velocities caused by the buildup of deposits on
the steam generator tubes. This buildup of deposits on the tubes
significantly reduced the available flow area within the tube bundle
causing flow diversion to the periphery of the tube bundle. The flow
diversion to the periphery was also affected by the increased steam
quality of the fluid within the tube bundle. The buildup of deposits on
the tubes changed the heat transfer characteristics of the tubes
causing the steam quality to increase in the central region of the
steam generators. This resulted in an increase of the flow resistance
in the central portions of the steam generator, forcing more flow to
the peripheral regions, with resulting higher velocities. The resulting
large velocity gradients at the periphery initiated vortices which
further elevated local velocities that were capable of dislodging the
protective oxide layer of the eggcrate material and initiating erosive
FAC.
The chemical cleaning of the SONGS Units 2 and 3 steam generators
during the 1997 refueling outages removed the deposit buildup and
restored fluid flow to their original design values (i.e., nominal
conditions). The licensee stated in its October 17, 1997, letter that
with the flow area restored to nominal conditions, the high fluid
velocities that lead to FAC would no longer exist, thus stabilizing
eggcrate support degradation. The licensee has also made changes to the
chemistry control program for the secondary system at SONGS Units 2 and
3 to reduce the feedwater iron transport. This is expected to prevent
the level of deposit buildup observed in the steam generators before
chemical cleaning was done in 1997. The staff concurs with the
licensee's evaluation that FAC was caused by deposit buildup on the
steam generator tubes and that removal of the deposits should restore
the steam generator secondary fluid flow to within nominal design
values, thereby eliminating continued significant eggcrate degradation.
To confirm that FAC has been stopped by the chemical cleaning of the
steam generators, and to assure that no significant degradation of the
eggcrate support structure goes undetected, the licensee has committed
to conduct periodic inspections of the secondary side of the steam
generators in both units during future outages. The licensee will
conduct periodic inspections of the secondary side of the steam
generators to check the level of deposit buildup on the tubes and to
verify that future degradation of the eggcrate, if any, remains within
the assumptions used in the analysis to demonstrate continued
operability of the steam generators (discussed later in this Decision).
2. Description of the Eggcrate Inspections
The SONGS licensee inspected the steam generator secondary side
support structures, which include the eggcrate supports, in both SONGS
units during their 1997 refueling outages and during their 1998 mid-
cycle outages. The results of these inspections are contained in the
licensee's letters dated May 16, 1997, and June 5, 1997 (SONGS Unit 2
and Unit 3 refueling outage inspections results, respectively), and
letters dated March 10, 1998, and April 15, 1998 (SONGS Unit 2 and Unit
3 1998 mid-cycle outages, respectively).
The objective of the inspections for both units was to provide
video documentation of all areas in which indications of support bar
degradation was suspected and to verify that other areas did not
exhibit these same characteristics. The extent and results of these
video inspections are summarized below.
The inspection of the secondary side of each steam generator was
divided into six areas: (1) general inspection, (2) inner tube bundle,
(3) batwings and vertical straps, (4) eggcrate periphery, (5) eggcrate
interior (blowdown lane),and (6) stay cylinder. Each of these areas was
inspected to the extent necessary to understand, with a high degree of
confidence, the amount of degradation present. The majority of these
areas did not exhibit any significant degradation and therefore the
design function of the support structures was not adversely impacted.
The general inspections were performed in the steam generators from
the top of the moisture separator can deck and included the general
area, U-bend, and annulus regions. The areas inspected included I-
beams, I-beam to shroud attachments, drains, vertical supports,
batwings and the batwing hoop, and baffle anti-rotational keys. These
inspections identified no significant degradation in either unit in
these areas.
The inner tube bundle consists of that area between the outer or
peripheral tubes to the inner tubes of the stay cylinder. The inner
bundle inspections were performed in both steam generators from the can
deck. A small camera was dropped down in between the tubes in a number
of different locations to assess the general material condition of the
eggcrates away from the periphery area. For the steam generators in
both units, the inspections indicated that the inner bundle did not
exhibit the degraded characteristics of the periphery eggcrates found
in the Unit 3 steam generators during the 1997 refueling outage.
No indications of thinning were detected during the inspections of
the interior batwing and vertical strips on either unit.
Comprehensive peripheral eggcrate inspections were performed in
both steam generators in the two units from the can deck. This included
the lattice bars and tube to lattice bar interfaces at each eggcrate.
The area near the periphery of the eggcrate supports in the Unit 3
steam generators experienced the maximum thinning, as shown in Figure 3
and discussed above. As stated earlier, minor isolated instances of
thinning were observed in the peripheral eggcrate locations in the
SONGS Unit 2 steam generators, but overall the thinning was
considerably less than that observed on SONGS Unit 3.
Inspections of the blowdown lane eggcrates were performed in the
steam generators through the 6-inch handhole at the secondary face of
the tubesheet from the handhole to the stay cylinder. This included the
lattice bars and the eggcrate rings. The inspection scope was to sample
the eggcrate area nearest the tubes on both the hot- and cold-leg sides
of the blowdown lane. Minor amounts of eggcrate degradation were found
in the steam generators of both units, with the Unit 3 steam generators
exhibiting the larger amount of degradation in this area.
For the inspection of the overall condition of the eggcrates and
ring in the stay cylinder, a support plate inspection device was used.
Little or no degradation was found in this area in either unit.
3. Summary of SONGS Unit 2 Eggcrate Inspection
The licensee's initial assessment of the Unit 2 stream generator
eggcrate supports, conducted after the degradation issue was identified
in the SONGS Unit 3 steam generators, was reported in its letter dated
May 16, 1997. The licensee concluded that the Unit 2 eggcrate supports
were in very good to excellent overall condition, based on the limited
video examinations of the eggcrates performed in support of the
chemical cleaning process. Although the licensee considered operation
for the normal period of operation between refueling intervals to be
acceptable on the basis of this limited examination,
[[Page 33970]]
the licensee conservatively performed a more extensive video
examination of the eggcrates during a mid-cycle outage that began on
January 24, 1998. As reported in its March 10, 1998, letter, the
licensee observed minor isolated instances of thinning in the periphery
areas of the eggcrate supports, but overall the thinning was
considerably less than that observed on SONGS Unit 3.
The NRC reviewed the program established by the licensee to conduct
the video examinations of the eggcrate supports during the SONGS Unit 2
mid-cycle outage and reported its findings in Inspection Report 50-361/
98-10; 50-362/98-01, dated May 29, 1998. This program was similar to
the licensee's program for inspecting the Unit 3 eggcrate supports
during its mid-cycle outage. The primary difference between the
inspection programs for the two units was that a larger portion of the
Unit 3 eggcrate structures was inspected. The staff concluded in its
inspection report that the scope of the SONGS Unit 2 secondary side
visual inspections was satisfactory and the results supportive of the
licensee's conclusion that no steam generator tubes needed to be
removed from service due to insufficient support from any secondary
side support structures, which includes the eggcrate support
structures.
4. Actions Taken as a Result of Observed Eggcrate Degradation
Following the secondary side inspection activities conducted during
the SONGS Unit 3 1997 refueling outage and 1998 mid-cycle outage, the
licensee plugged and stabilized (by insertion of a steel cable inside
the subject tube) some Unit 3 steam generator tubes as a precautionary
measure due to the degradation observed in certain eggcrate supports.
No tubes in the Unit 2 steam generators were removed from service. Once
the tube is removed from service in the above described manner, support
from the eggcrate structures is no longer needed. The criterion
established by the licensee for removing tubes from service is
described in detail below.
B. Concern About the Seismic Adequacy of the SONGS Steam Generators
The Petitioner asserts that the degradation of the steam
generators, eggcrate supports could seriously weaken the supports and
make the steam generators vulnerable to seismic events.
In its letter of May 16, 1997, the licensee committed to perform an
evaluation of the effect of the degraded eggcrates on steam generator
tube integrity in the SONGS Unit 3 steam generators before return to
power from the Unit 3 1997 refueling outage. This initial evaluation
was provided by the licensee in its letter of June 5, 1997, and
included the effects of a postulated design-basis earthquake. The
licensee submitted the final version of the degraded eggcrate support
evaluation for SONGS Unit 3 on October 17, 1997. As stated in the
previous section, the amount of eggcrate support degradation observed
in SONGS Unit 2 was considerably less than that observed in Unit 3.
Therefore, the staff concludes that demonstrating the ability of the
SONGS Unit 3 steam generators to withstand a design basis seismic event
will demonstrate the adequacy of the Unit 2 steam generators as well.
The staff's review of the seismic adequacy of the SONGS Unit 3
generators is detailed below.
1. Methodology and Acceptance Criteria
The Petitioner did not specifically request the staff to evaluate
the eggcrate supports assuming other design loads concurrent with
earthquake loads. However, to provide additional conservatism, and to
conform with General Design Criterion (GDC) 2 of 10 CFR Part 50,
Appendix A, the licensee, in its October 17, 1997, letter, evaluated
the ability of the eggcrate supports to perform their intended safety
function assuming the most limiting combination of load conditions.
GDC 2 requires, in part, that the design bases for structures,
systems, and components important to safety reflect appropriate
combinations of the effects of normal and accident conditions with the
effects of natural phenomena such as earthquakes. The earthquake for
which these plant features are designed is defined as the safe-shutdown
earthquake (SSE).\2\ The Petitioner's concerns on the adequacy of the
seismic design of the SONGS units, based on information gathered from
the Landers and Northridge earthquakes, were addressed previously by
the staff in DD-97-23 (see footnote 1).
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\2\ The SSE is defined, in part, as ``that earthquake which is
based upon an evaluation of the maximum earthquake potential
considering the regional and local geology and seismology and
specific characteristics of local subsurface material. It is that
earthquake which produces the maximum vibratory ground motion for
which certain structures, systems, and components are designed to
remain functional.'' See 10 CFR Part 100, Appendix A, Section
III.(c),
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Appendix A of Standard Review Plan,\3\ (SRP) Section 3.9.3,
``[American Society of Mechanical Engineers] ASME Code Class 1, 2, and
3 Components, Component Supports, and Core Support Structures,''
delineates acceptable design limits and appropriate combinations of
loadings associated with normal operation, postulated accidents, and
specified seismic events for the design of Seismic Category I fluid
system components (i.e., water- and steam-containing components). This
appendix also provides that necessary plant features important to
safety meet the appropriate design limits specified in Section III of
the ASME Boiler and Pressure Vessel Code (ASME Code) when the component
is subjected to concurrent loadings associated with the normal plant
condition, the vibratory motion of the SSE, and the dynamic system
loadings associated with the faulted plant condition. Faulted plant
conditions are those operating conditions associated with postulated
events of extremely low probability, such as loss-of-coolant accidents
(LOCAs) or main streamline break (MSLB) accidents. The design limits
and loading combinations utilized by the licensee in the October 17,
1997, evaluation of individual steam generator tubes are the same
design limits and loading combinations that were reviewed and approved
by the staff at the time of plant licensing. This evaluation is
contained in Chapter 3 of NUREG-0712.\4\ Therefore, the staff finds
acceptable the licensee's use of these design limits and loading
combinations in evaluating the impact of the degraded eggcrate supports
on individual steam generator tubes.
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\3\ The Standard Review Plan (SRP) is published as NUREG-0800,
and is used as guidance for the Office of Nuclear Reactor Regulation
staff responsible for the review of applications to construct and
operate nuclear power plants.
\4\ NUREG-0712, ``Safety Evaluation Report related to the
Operation of San Onofre Nuclear Generating Station, Units 2 and 3,''
Chapter 3, February 1981.
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The evaluation of the potential for lateral movement of the entire
steam generator tube bundle (whole bundle evaluation) was not
explicitly addressed during the staff's review performed at the time of
plant licensing. Also, the ASME Code does not provide specific design
limits for the whole bundle evaluation. The whole bundle evaluation
contained in the October 17, 1997, letter performed by the licensee to
verify that the structural integrity of the eggcrate is maintained to
ensure that it does not shift in a way that could damage the tubes.
This is not an ASME Code evaluation; however, ASME Code techniques were
used by the licensee to generate and assess the results. The staff has
reviewed the specific ASME Code techniques utilized by the licensee,
and concludes that they provide conservative results, and are,
therefore, acceptable for the whole bundle evaluation.
[[Page 33971]]
Furthermore, the loading combinations used in the licensee's whole
bundle evaluation are the same loading combinations used in the
individual tube evaluations, and are the same loading combinations that
were reviewed and approved at the time of plant licensing.
2. Degraded Eggcrate Support Assumptions
The staff reviewed the assumptions used in the licensee's October
17, 1997, evaluation regarding the amount of eggcrate support judged to
be available, and verified that these assumptions were supported by the
results of the licensee's inspections.
For the individual steam generator tube analysis, the licensee
calculated the maximum loads that could occur assuming that adequate
support was not available at two consecutive eggcrate levels (see
Figure 1). The staff finds this assumption conservative and acceptable
because the licensee has removed from service all tubes where two
consecutive eggcrate levels were found degraded to the point where
adequate support could not be assured.
For the whole bundle analysis, the licensee used the inspection
results to sort the eggcrates into categories based on a conservative
estimate of the remaining thickness of the eggcrate lattice bars. The
staff reviewed the sorting criteria used by the licensee, and concludes
that the material strength assumptions established by the licensee for
the degraded eggcrate supports are conservative, and appropriate for
evaluating the ability of the eggcrate structures to perform their
intended function.
The visual inspections performed by the licensee during the 1998
mid-cycle outages for both units confirmed the appropriateness of these
assumptions pertaining to the amount of eggcrate support degradation
used in the licensee's evaluation.
3. Evaluation Results
Using the above described methodology and assumptions, the licensee
determines that the peak calculated loads on the individual steam
generator tubes would remain below the allowable design limits approved
by NUREG-0712 during and following a postulated design basis
earthquake.
The results of the licensee's whole bundle evaluation confirmed
that the eggcrate structure will provide sufficient support to ensure
that the tube bundle will not impact the eggcrate support ring during
and following a postulated design basis earthquake.
The staff finds these results acceptable, and as detailed above,
also finds acceptable the methodology and assumptions used by the
licensee in the generation of these results. The staff concludes,
therefore, that the amount of degradation observed in the eggcrate
supports will not prevent the SONGS Units 2 and 3 steam generators from
performing their intended safety functions.\5\
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\5\ Since the amount of support degradation in SONGS Unit 2 was
observed to be considerably less than that observed in Unit 3, the
NRC staff concludes that the licensee's October 17, 1997, evaluation
of SONGS Unit 3 steam generator structural integrity and the staff's
review of that evaluation support the adequacy of SONGS Unit 2 steam
generators to withstand a design basis event and perform their
intended safety function.
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4. Confirmatory Actions
The licensee's 1998 mid-cycle inspection of the SONGS Unit 3 steam
generators confirmed that the condition of the Unit 3 eggcrate internal
supports remained within the analytical assumptions used in the
licensee's evaluation contained in its October 17, 1997, letter and
also supported the licensee's contention that the phenomenon (FAC) that
led to the degradation of the eggcrates had been arrested by the
chemical cleaning of the steam generators.
Furthermore, the licensee has committed in its letters to the NRC
(April 15, 1998, for Unit 2 and October 17, 1997, for Unit 3) to
inspect the eggcrate supports during future outages to assure that
their condition remains within the analytical assumptions used in the
licensee's evaluation. These inspections will continue to be conducted
until it is established that further inspections are not required.
In summary, on the basis of the video inspection results for the
steam generators in both units, and the staff's review of the detailed
evaluations performed by the licensee, the staff concludes that the
SONGS steam generators are fully capable of performing their intended
safety function during and following a postulated SSE, and no
retrofitting upgrade of the steam generators is required.
III. Conclusion
As explained above, there is no evidence of significant degradation
of the SONGS Unit 2 steam generator eggcrate supports, and the
extensive analyses demonstrate the ability of the steam generators in
both SONGS units to perform their intended safety function.
Accordingly, the Petitioner's requested action, pursuant to Section
2.206, is denied.
A copy of this Decision will be filed with the Secretary of the
Commission for the Commission to review in accordance with 10 CFR
2.206(c) of the Commission's regulations. As provided by this
regulation, the Decision will constitute the final action of the
Commission 25 days after issuance, unless the Commission, on its own
motion, institutes a review of the Decision within that time.
Dated at Rockville, Maryland, this 11th day of June 1998.
For the Nuclear Regulatory Commission.
Original signed by
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
Attachments: Figures (3)
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[FR Doc. 98-16539 Filed 6-19-98; 8:45 am]
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