ML18057A446
| ML18057A446 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 08/29/1990 |
| From: | Hasse R, Phillips M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML18057A445 | List: |
| References | |
| 50-255-90-17, NUDOCS 9009070208 | |
| Download: ML18057A446 (14) | |
See also: IR 05000255/1990017
Text
U.S. NUCLEAR REGULATORY COMMISSION
REGION I I I
Report No. 50-255/90017(DRS)
Docket No. 50-255
Licensee:
Consumers Power Company
1945 West Parnall Road
Jackson, MI
49201
Facility Name:
Palisades Site
Inspection At:
Covert, MI 49043
Inspection Conduc~~:
J~ly 30
Inspectors: ~a"S~I
Team Leader
8. Holian, NRR
R. Anand, NRR
G. Tan, NRR
E. Murphy, NRR
T. Rotella, NRR
August 3, 1990
Approved By~~hief
Operational Programs Section
Inspection Summary
License No. DPR-20
Date
I
Inspection on July 30 - August 3, 1990 (Report No. 50-255/90017(DRS)).
Areas Inspected:
Special, announced inspection of the adequacy of evaluations
performed pursuant to 10 CFR Part 50, Section 59 with primary focus on the
Steam Generatof Replacement Project (SGRP).
This inspection was conducted in
accordance with Inspection Module 37700.
Results:
No violations were identified.
The engineering supporting the SGRP
was ~uite thorough.
Some ~eaknesses were identified in the 10 CFR 50.59
evaluations with the primary concern being the failure to identify when
associated Technical Specification changes were required .
- ~: ~~',;:; ') (1 ~~ ;;~ ;~; g
~}:
I)
DETAILS
1.
Persons Contacted*
D. Joos, Vice President, ESSD
D. Vandewalle, Plant Technical Director
J. Kuemin, Plant Licensing
B. Gerling, Accident and Transient Analysis Supervisor
R. Rice, Operations Manager
R. Vincent, Plant Safety Engineering (PSE) Administrator
P. Bruce, PSE
J. Erickson, PSE
K. Toner, Projects Superintendent
T. Tai, Nuclea~ Support (Bechtel)
J. Lewis~ Steam Generator Replacement Project Licensing
US-NUCLEAR REGULATORY COMMISSION (NRC)
- J. Zwolinski, Assistant Director for Region III Reactors, DRSP
R. Pierson, Director, Project Directorate III-I, DRSP
- E. Swanson, Senior Resident Inspector
J. Heller, Resident Inspector
- With these exceptions, all personnel listed above attended the exit
interview held on August 3, 1990.
2.
.Introduction
The purpose of this inspection was to determine the adequacy of the
licensees evaluations performed pursuant to 10 CFR Part 50, Sectton 59.
This regulation allows the licensee to make c_hanges to the facility as
described in the FSAR, revise procedures described in the FSAR, or
perform tests and experiments not described in the FSAR without prior
NRC approval as long as no unreview~d safety question is generated or
technical spetification changes are required.
It also requires the
licensee to document the bases for concluding that these requirements
have been met when making changes pursuant to this regulation.
The inspection focused primarily on the licensee's steam generator
replacement project (SGRP).
A small effort was devoted to a review of
the l.O CFR 50.59 evaluations for changes submitted to the NRC pursuant
to 10 CFR 50.71 {annual FSAR updates).
The safety evaluations {which encompass the 10 CFR 50.59 evaluations)
for the SGRP reviewed by the inspectors had been prepared by the
licensee's contractor (Bechtel).
The licensee's review and approval of
these documents. had not been completed; therefore, it was necessary to
treat these as draft documents from a regulatory standpoint. This is
pertinent because they contained erroneous conclusions with respect to
technical specification changes required prior to plant startup (see
Paragraph 3.m).
Had these conclusions been final, they would represent
inadequacies in the 10 CFR 50.59 evaluation and a potential violation.
2
3.
- -
Steam Generator Replacement Project (SGRP)
The SGRP consisted of a series of facility change (FC) packages plus
several packages not falling within the definition of a facility change
(e.g., rigging for SG lifts and the interim storage facility for the old
SGs).
Each package was supported by its own safety evaluation (SE). The
inspectors reviewed the SEs, FCs, and supporting analyses to determine
if any unreviewed safety questions were generated or technical
specification changes not previously identified would be required.
The
results of these reviews are discussed in the following paragraphs.
a.
FC-909, "Steam Generator Replacement"
This facility change encompassed the removal of the existing Steam
Generators (SGs) and the impact of the installation of the new
steam generators. Section 1.0 of the Safety Evaluation (SE)
supporting this change provided a description of the effects of
the replacement SGs on normal plant operation and the Palisades
Final Safety Analysis Report (FSAR) accident and transient
analysis. Systems interactions were also discussed. Sections 2.0
and 3.0 addressed the decontamination and coating of the existing
SGs and the installation of nozzle cover plates prior to SG
removal from containment.
Consumers Power Company (CPCo)
concluded that the overall SG performance changes associated with
the replacement SGs would not adversely affect current FSAR
conclusions or plant performance ~equirements .
Specifi~ areas reviewed by the inspector in~luded the impact on
net positive suction head (NPSH) for the reactor coolant pumps
(RCP) and the*affected transient .analyses.
The accidents
determined to require reanalysis were the Main Steam Line Break
(MSLB) and Stea~ Generator Tube Rupt~re (SGTR).
Certain specific
references, delineated for RCP NPSH considerations, and the MSLB
and SGTR accidents were examined in detail. A discussion of the
findings for FC~909 is provided below.
It should be noted with respect to FC-909 that the package did not
include a signed Design Input Checklist nor a completed Design
Review Sign Off.
However, the supporting documentation including
references were provided to the inspector with the proper .
signatures.
(1)
Reactor Coolant Pump NPSH and Core Uplift
The replacement SGs, which were manufactured by Combustion
Engineering (CE), are designed to physically match the
essential parameters of the existing SGs.
The replacement
_ SGs should ~losely duplicate the physical, thermal, and
hydraulic characteristics of the original SGs while *
ihcorporating a combination of design improvements.
The
overall heat transfer coefficient, U, times the tube area
available for heat transfer, A, is the same as the existing
3
SGs (i.e., (UA) original= (UA) *new).
Therefore, the design
limits of operationally interconnected plant systems should
be minimally affected.
The most significant effect of the
new SGs would be on primary coolant system (PCS) flow.
loop flow would be increased due to a decrease in
differential pressure across the new SGs.
Although the flow
increase would be primarily beneficial (e.g., fuel heat
transfer) other parameters such as RCP motor effects, core
upl.ift forces and RCP av~ilable NPSH must be evaluated.
The
inspector reviewed the licensee's calculations supporting
RCP NPSH and ~ore uplift considerations.
The calculations
adequately addressed both the RCP NPSH and core uplift
concern~. Starting of the fourth RCP should not occur until
RCS temperature reaches 450°F.
This is a change from
current operating practice but will not effect the Technical
Specifications.
(2)
Final Safety Analysis Report (FSAR) Accident Reanalyses
Section 1.3 of FC-909 provid~d an evaluation of the effects
of th~ replacement SGs on the FSAR accident and transient
analyses.
CPCo concluded that based on the results of a
disposition-of~events review and the detailed reanalysis of
main steam line break and SG tube rupture accidents that no
unreviewed safety questions would exist due to operation
with the replacement SGs .
(a)
Steam Generator Tube Rupture
Reference 17 of FC-909 provided a reanalysis of the
Palisades Steam Generator Tube Rupture Accident.
Although the outside diameter of the new SG tubes
would be unchanged (0.750 inches), the inside *diameter
has increased by 0.012 inches.
CPCo stated that the
SG tube rupture event with a loss of offsite power
upon reactor trip was reanalyzed in accordance with
the guidelines of SRP 15.6.3.
The results of the.
reanalysis indicated that the slightly larger inside
diameter of the U-tubes is not a significant factor in
the analysis. Rather, it was found that the
differences in methodology between the FSAR analysis-
of-record and CE's current analysis methodology had a
gr~ater impact on.the results.
During a presentation by CPCo on* June 6, 1990 and
during this inspection, the inspector questione.d the.
use of minimum primary coolant system pressure as an
assumption in the reanalysis of the worst case SGTR.
On August 3, 1990 a conference call.between CPCo, CE,
and the NRC was conducted to ascertain the basis for
this assumption.
CE representatives stated that the
assumption of minimum primary coolant system pressure
4
(b)
for the limiting SGTR accident was in error and was
nonconservative for Palisades.
CE realized this
nonconservative assumption subsequent to the June 6,
1990 presentation to the staff but prior to this
inspection and had corrected their error. The SGTR
reanalysis documentation provided during the
.
inspection was claimed by CE to have delineated this
change.
However, the inspector found the reanalysis
to be unclear in this regard. CPCo agreed with the NRC
on this point.
The results of the reanalysis
demonstrated that all *calculated doses remained within
General Design Criterion 19 guidelines and within 10
CFR Part 100 acceptance criteria. Therefore, the
inspector found the 10 CFR 50.59 supporting reanalysis
to be acceptable because the Palisades licensing basis
stated that .the acceptance criterion for the SGTR
accident is 10 CFR Part 100.
Main Stem Line Break (MSLB)
Reference 18 of FC-909 provided a reanalysis of the
worst case MSLB.
During the inspection, several
discussions were held with CPCo to ascertain the basis
of certain assumptions used in the reanalysis
including the reliance on conclusions made by CPCo in
a 1985 submittal. to the NRC addressing single failure
issues .. The staff S~R referenced in the MSLB
-
reanalysis addressing CPCo's submittal is entitled,
"Single Failure Issues for Main Steam Isolation Valves
and Main Feedwater Isolation Valves," dated February
28, 1986.
As a result of the MSLB reanalysis, iri March 1990, CE
identified that the worst case MSLB for Palisades was
-not a double-ended 100 percent break,. as was
previously believe_d, but a small break (approximately.
30%).
Corrective action by tPCo was to install a
.containment high pressure (CHP) initiated feedwater
regulating valve closure actuation circuitry to
provide a quicker response to the transient.
The 30
percent break size results in a higher peak .
containment pressure due to the delay in receiving a
low SG pressure initiated feedwater valve closure.
The new SGs have a nozzle on the steam outlet which
limits a full double-ended MSLB to a 30 percent
blowdown.
The CHP-initiated feedwater valve closure
modification provides feedwater isolation in time to
limit the peak containment pressure to 52.45 psig
which is below.the Palisades containment design limit
of 55 psig and below the previous MSLB analysis-of-
record peak containment pressure of 54.1 psig.
5
The inspector concluded that the FC-909 supporting *
reanalysis for MSLB does not present an unreviewed
safety question. Also, during the inspection, CPCo
committed to amend the Palisades Technical
Specifications to include the feedwater regulating
valve CHP actuation logic and the Low SG Pressure.
Surveillance Requirements (SR).
The SR frequencies
will be consistent with other safety related
equipment.
(3)
Impact on Technical Specificitions
Section 4.14 of the Palisades Technical Specification
specifies the plugging limits (acceptance criteria) to
be applied to the results of steam generator tube*
-
inspections. These limits were developed spetifically
for the 0.75 inch diameter x 0.48 thick tubes in the
existing SGs.
These limits are not appropriate for
the 0.75 inch diameter X 0.042 thick tubes in the new
SGs.
The licensee stated on page 43 of the attachment
to the safetj analysis for Facility Change 909) that
it would propose a new 40% wall thinning plugging
. limit, consistent with the ASME Code,Section XI, for
inclusion in the Technical Specifications.
Plugging limits w~re not an issue duri~g the
preservice inspection of the new steam generators
(discussed on page 38 of the attachment to the safety
analysis) since no tubing flaws were found.
However,
experience indicates that tubing flaws can develop
durinij the initial operating cycle.
New plugging
limits should be in place prior to the first inservice
inspection (ISI) to ensure that any flaws found are
assessed ~gainst criteria appropriate to the new SGs.
The licensee planned to submit a proposed Technical
Specification change to incorporate new plugging
limits prior to.the first ISi of SG tubes.
The SG replacement program did not require a change to
the inspection frequency and sampling requirements in
Section 4.14 of the Technical Specifications.
Nevertheless, the licensee planned to propose-updated
requirements in this area, consistent with Standard
Technical Specification requirements.
b.
FC-912, "Auxiliary Feedwater System Piping Modification"
This facility change involved a portion of the existing auxiliary
feedwater system piping to be rerouted to accommodate the location
of the new steam generator auxiliary feedwater nozzles.
The 3"
and 4" diameter auxiliary feedwater piping from the steam
generator nozzles would be removed back approximately 5 feet at
6
. I
steam generator E50A and approximately 11 feet at steam generator
E50B.
These lines would be shortened approximately 20 inches and
reoriented approximately 7 1/2 degrees from the old steam
generator nozzle locations to fit the new steam generator nozzles.
The modification was being performed under ASME B&PV Code, Section
XI.
The new piping was designed to meet seismic category I
requirements and was to be in accordance with the codes and
standards required by FSAR Table 5.2-3.
The piping was procured
in accordance with ASMt Section III, Class 2 requirements.
The licensee had analyzed for potential new pipe break locations.
No new jet impingement or pipe whip targets were identified.
The
auxiliary feedwater line would be rerouted to minimize horizontal*
runs to prevent water hammer in the auxiliary feedwater system.
The new position of the piping being installed had been evaluated
to assure that the stress levels would not exceed code allowable.
The licensee's analyses indicated .that the removal and subsequent
replacement of the auxiliary feedwater lines would not constitute
an unreviewed safety questions.
c.
FC-893, "Blowdowh Piping Modification
This facility change involved the replacement of the 2 inch
nominal
pip~ size (NPS) steam generator bottom blowdown system
- piping with 4 inch NPS piping from the steam generatqr nozzle up
to but not including the containment isolatitin valve outside
containment.
The replacement would consist of removing the
existing 2" piping and associated supports and installing new 4"
piping, pipe supports and thermal insulation.
High point vents,
low point drains, and sample connections were alsd to be added.
Although the existing piping had been uninsulated, the replacement
piping would be insulated with a flexible insulation to prevent
any increase in heat load due to the ihcreas~d heat transfer area.
Modifications to the blowdown piping outside containment beyond
-
the isolation valve would be made later.
The piping and piping supports were designed as seismic Category I
in accordance with the requirements of ASME Section XI.
The
existing containment penetration was to be ~odified to accept a 4"
pipe.
The licensee had analyzed the penetration sleeve to
demonstrate that it w_as capable of carrying the design loads
associated with the 4" pipe. Also, the licensee performed
calculations to ensure that the*concrete temperature around the
penetration remained within acceptable limits.
The licensee's
analysis postulated pipe breaks in accordance with the high energy
line break (HELB) criteria given in FSAR Section 5.6. The
analysis concluded that a HELB in the replacement 4 inche pi~e
would not increase the consequence of an accident as it relates to
the containment temperature or pressure responses.
Pipe whi~ and
7
jet impingement targets had been identified and whip restraints
were to be added where the consequences of unrestrained whip were
unacceptable.
The existing structures have been evaluated for the
additional loads.
The licensee had committed to perform
nondestructive examination (NDE) of all welds performed during
this modification. All butt welds in piping were to be 100%
radiographed and ex_amined by liquid penetrant or magnetic particle
techniques.
Based on the above, the inspectors concluded that the replacement
and/or modification of the blowdown piping did not involve an
unreview~d safety question.
The licensee committed to amend the
plant Technical Specifications to revise the containment
penetration description.
d.
FC-894, "Replace Existing Surface Slowdown Pipe With 4" SG
Recirculation Pipe"
This facility change involved the replacem~nt of the 2 inch
nominal pipe size (NPS) steam generator surface blowdown system
piping with 4" NPS piping from the steam generator nozzle up to.
but not including the containment isolation valve outside the
containment.
A1so, high point vents and low point dr~ins would be
provided in the new piping. Since the system .would only be used
during cold shutdown, only the line from the steam generator to
the containment shield wall was to be insulated.
The tirculation system is a non-safety system-except the portions
from the steam generators to the containment isolation valves (CV-
739 for steam generator 50A and CV-738 for steam generator 508),
which is designed as seismic Category I. Modification to the
recirculation lines outside the containment beyond the isolation
valve was to be made later.
The licensee had postulated pipe breaks in accordance with the
high energy line break (HELB) criteria. Pipe whip and jet
. impingement targets had been identified and whip restraints were
- to be added.
The existing containment penetration was to be
,
modified to accept a 4" pipe.
The penetration sleeve had been
analyzed to.demonstrate that it was capable of carrying the design
loads associated with the larger pipe.
The existing structures
had been evaluated for the additional loads imposed by the new
pipe supports.
The* licensee committed to perform non-destructive
examinatio~ (NDE) of all welds performed during this modification.
All butt welds in the piping were to be 100% radiographed and
examined by liquid penetrant or magnetic particle techniques.
Based on the above, the inspectors concluded that the replacement
and/or modification of the recirculation piping did not involve an
unreviewed safety question.
The licensee committed to amend the
plant Technical Specification to revise the containment
penetration description.
8
e.
FC-915, "Component Cooling Water (CCWl Surge Tank Room
Modifications"
This FC provided for an opening of approximately 48" by 21" in the
ceiling of the CCW surge tank room and drilling a 2" hole in the
floor of the room.
The room would be returned to its original
configuration after the SGRP was completed.
The openings were
required to provide ~quipment access to the containment post-
tensioning system tendons located on the west side of containment
buttress B.
The inspector reviewed the FC package, the SE, and toured the
affected area. Several concerns were identified with the SE:
o
The SE did not address the potential for *the release of*
airborne radioactivity through these openings ..
o
Some potential .for damage to the CCW surge tank and other .
components in this area exists while thg. de-tensioning
process is in progress.
The heat load on the CCW system
during this period were not addressed in the SE.
The SE *
noted that the surge tank was required only for "long term"
CCW operatio~. The duration of "long term" or heat loads
during that period wee not addressed;
Further, no
contingency actions were identified if damage to the CCW
system should occur.
The incorporation of these issues into the SE will be tracked as
an open item (255/90017-01).
f.
. FC-910, PCS Piping Replacement
The facility change involved modification of Primary Coolant
System (PCS) piping to facilitate steam generator (SG)
replacement.
T~is included replacing the first cold leg elbow at
each SG outlet nozile and cutting each hot leg pipe at the SG
nozzle weld and machining the pipe ends to accept the new SG~.
One or both hot leg elbows. were to be replaced if necessary to
facilitate SG replacement.
The pipe-to-elbow and elbow-to-SG
nozzle circumferential butt welds would employ the GTAW process
and were to be accomplished by automatic welding machines using a
"narrow gap" welding procedure.
The inspectors reviewed both the facility change package and the
accompanying safety evaluation.
Inspector review activities at
the site were supplemented by a review by NRR headquarters
personnel (from the Materials and Chemical Engineering Branch) of
the narrow gap welding technique and the ongoing program to
qualify this technique in accordance with ASME Code,Section IX
requirements.
9
The inspector's review uncovered no issues involving either a
potential unreviewed safety question or a need for changes to the
Technical Specifications.
g.
FC-911, "Main Steam Piping"
This facility change involved removal and subsequent replacement
of the main steam line to accommodate the replacement of the SGs:
Removal was to be accomplished by cutting the steam line at the SG
main steam nozzle and in the vertical riser section. During
reinst~llation, the vertical riser portion would be extended
approximately 32 inches to compensate for the added height*of the
replacement SGs due to the incorporation of a flow restrictor at
the main steam nozzle.
The inspectors reviewed the facility change package, including the
Design Input Checklist and Design Documents Checklist, and the
associated 10 CFR 50.59 safety evaluation.
No areas of concern
were identified.
The inspectors nqted that internal reviews of
the facility change package had not yet been completed by Bechtel
and CPCo.
In addition, certain pipe support drawings and
mechanical and civil calculations, as identified on sheets 7, 8,
and 9 of the Design Document Checklist, had yet to be incorporated
into the package.
The inspectors review of Facility Change 911 uncovered no issues
involvin~ either a potentlal unreviewed safety question or a
potential need for changes to the Technical Specifications.
h.
FC-914, "Containment Construction Opening"
This FC provides for the installation of a 28' by 26' opening in
the containment structure to permit the removal of the existing
SGs from the containment and transfer of the new SGs into the
containment.
The containment will be returned to its original
condition following the completion of the SG transfer.
The inspectors focused their review of this FC package and SE on
the finite element analyses for various loads and load
combinations during the existence of the containment opening and
after its closure.
The inspectors concluded that the licensee had
made a comprehensive and thorough analysis of the containment
during the existence of the opening and after its closure;
however, some concerns were identified that still needed to be
addressed:
o
There was a large discrepancy between the stress levels
calculated at the containment foundation by two different
analyses. This discrepancy should be reconciled or the
larger valves should be used to determine if the stress
levels are acceptable.
10
o
The finite element stress analysis of the containment
averaged stresses in adjacent elements (i.e., an overstress
in one element was averaged with understresses in adjacent
elements).
The basis for the acceptability of this
technique should be provided.
o
The results of the tests performed by the licensee on the
concrete mix to be used to close the containment opening
should be provided to sup.port the creep and shrinkage va 1 ues
used in the analysis of the reconstructed portion of the
containment.
o
The containment liner will. be cut as a part of the process
for providing the containment opening.
The section removed
will be welded back in place.
An analysis should be
.
performed to demonstrate the structural integrity of the
repaired *1ining.
o
The reinforcing steel used in the reconstructed area was to
be spliced to existing reinforcing steel via Cadweld
splicing. Assurance should be provided that the relative
location of these splices in adjacent bars does not
jeopardize the structural integrity of the containment.
o
It would be necessary to remove the containment prestressing
tendons which would cross the containment opening.
These
tendons will be inspected and the undamaged ones re~sed. An
analysis should be performed to show that the creep and
relaxation that has occurred in thes~ tendons will not
result in an increase in prestress loss after reinstallation
and tensioning.
-
o
The prestressing tendons are contained in a sheathing which
i~ filled with grease for corrosion protection of the
tendons. The sheathing crossing the containment openirig will
be removed and discarded: New sheathing will be installed
in this ar_ea when the open*ing is closed.
The details of
connecting the new sheathing to the existing sheathing
should be provided.
o
It will be necessary to withdraw some tendons from their
sheathing and then reinsert it. The potential for damaging
this sheathing should be addressed.
o
The integrity of the Cadwell splices of the reinforcing bars
is very important~ Mockup training on the installatiori of
these splices should be provided.
o
The licensee plans to perform a structural integrity test
(SIT) at 115% of containment design pressure after repairing
the containment.
The effect of this test on containment
integrity considering the loss of prestress in the
11
undisturbed tendons (which will riot be retensioned) should
be evaluated.
The resolution of th~se concerns will be tracked as an open item
(255/90017-02).
i.
FC-904, "Auxiliary Building Modification for Containment Access"
This FC provided for a 6'-6" by 7'-2" personnel and equipment
.
access opening in the auxiliary building east wall. The wall would
be returned to its original configuration upon completion of the
SGRP.
The inspectors reviewed the FC package and SE.
One concern
was identified. The SE failed to address the potential release bf
airborne radioactivity via the access opening.
The resolution of this concern will be tracked as an open item
(255/90017-03).
j.
FC-895, "Wide Range Instrumentation* Tubing"
This FC revised the configuration of the SG wide range
instrumentation tubing. This modification was necessary because
of the different configuration-of the instrumentation taps dn th~
new SGs.
The inspectors reviewed the FC package and associated SE.
The
only operationally ~ignificant change was increasing the wide
range level readings from -138% - +100% to -140% - +150%.
No unreviewed safety question or Technical Specification changes
were identified.
k;
FC-913, "Secondary Sample System Piping"
This FC added sampling points to the new SG blpwdown p1p1ng and at
the sampling connections on each SG.
Rerouting of sample lines
was also required beca~~e of configurational differences between
the old and new SGs.
The inspector reviewed the SE for this FC and identified no
unreviewed safety question or Technical Specification changes.
1.
Other SGRP Support Activities
The inspectors reviewed the FCs and associated SEs for two
activities not directly affecting the plant:
(1) the installation
of a feinforced earth retaining wall outside the containment
structure (FC-903); and (2) the construction of cast-in-place
reinforced spread footings to support the lift system outside
containment (FC-908).
These packages were reviewed to assess any
potential impact on plant facilities in the vicinity.
12
m.
These facilities appeared to be well engineered and no concerns
were identified,
Summary of Technical Specification Changes Required as a Result of .
the SGRP.
There were six Technical Specifications (TS) that need to be
addrssed as a result of the SGRP.
The licensee addressed these
changes in the SE for FC-909, "Steam Generator Replacement."
The
SE concluded that none of these changes were required before
startup with the new SGs.
This conclusion was in error,and
contrary to the requirements of 10 CFR 50.59, which states that
prior NRC approval is required if an applicable TS change is
involved.
The TS chahges and when they are applicable are
discussed below:
o
TS 2.3.l, Item 5
This TS specifies the low SG level trip setpoint to be "not
lower than the centerline of the feedwater ring which is
6'0" below the normal water level." The feedwater ring in
the *new SGs is 5'8" below the normal water level. The. *
location of the feedwater ring must be revised.
The basis
for this TS is unchanged and there is no apparent impact on
any margin.
Since this TS is applicable during operation, it must be*
revised prior to startup.
o
This TS specifies that the primary-to-secondary leakage in a
steam generator must not exceed 0.3 gpm.
The basis for this
TS is a maximum throu~h wall track in a SG tube of 1/4"
superimposed on a 64% wasted area.
This assures structural
integrity of the tube.
The licensee plans to change the
basis to be consistent with IWB-3521 of ASME Section XI.
This specifies that the maximum through wall crack assure
structural integrity if superimposed on a 40% wasted area of
a tube.
(This basis is tied to TS 4.14 which specifies the
maximum permissible wall thinning.
As such, it would be
updated coincident with the update of TS 4.14.)
o
TS Table 3.5.l
This table lists the containment penetrations and valves;
including the line sizes .. The line sites for the SG
blowdown and recirculation systems must be revised. This TS
is applicable during operations and must be changed prior to
startup.
_ 13
o
TS 4.4;a
This TS specifies that the4P across the SG tubes must not
exceed 1380 psi.
The basis for this TS is to prevent a tube
rupture assuming a 64% wasting of a tube wall (originally
0.048" thick).
The licensee plans to change this basis to a
40% wasting of a tube wall to be consistent with ASME
Section XI.
This TS would be conservative as it currently reads on
startup. It would continue- to be conservative after TS 4.14
is changed.
o
Among other issues, this TS specifies the maximum tube
wastage (thinning) allowable for the SGs.
Since the new SGs
have tubes with thinner walls (0.042" vs 0.048"), this TS
must be changed before the first ISI of the SG tubes.
o
This TS states the codes and editions for the p1p1ng systems
and components.
This must be revised to reflect the later
editions used for the new SGs and replacement piping. This
TS applies during operation and must be changed prior to
startup.
-
The licensee agreed to submit the TS change requests commensurate
with the required dates.
The revision of the SE for FC-909 to reflect the required schedule
for TS changes will be track~d as an open item {50-255/90017-04).
4.
FSAR Updates
A sampling of modification packages and drawing changes from the most
recent FSAR update (Rev. 10) were reviewed to determine if all aspects
of the changes were addressed.
No discrepancies were noted.
5.
Open Items
Open items are matters which have been discussed with the* licensee which
will be reviewed further by the inspectors, and which involve some
action on the part of the NRC or licensee or both.
Open items disclosed
during the inspection are discussed in Paragraphs 3.e, 3.h, 3.i, an 3.m.
6.
Exit Interview
The inspectors met with licensee representatives (denoted in Paragraph
1) at the conclusion of the inspection on August 3, 1990, and summarized
the purpose, scope and findings of the inspection.
The licensee stated
that the inspectors had no access to proprietary information.
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