ML17292B372
| ML17292B372 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 05/08/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML17292B371 | List: |
| References | |
| 50-397-98-01, 50-397-98-1, NUDOCS 9805140317 | |
| Download: ML17292B372 (22) | |
See also: IR 05000397/1998001
Text
ENCLOSU
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket No.:
License No.:
Report No.:
Licensee:
Facility:
Location:
Dates:
Inspector:
Accompanying
Personnel:
50-397
50-397/98-01
Washington Public Power Supply System
Washington Nuclear Project-2
3000 George Washington Way
Richland, Washington
February 9 through April22, 1998
John E. Whittemore, Senior Reactor Inspector
Maintenance Branch
Tai L. Huang, Nuclear Engineer, Reactor Systems Branch
Office of Nuclear Reactor Regulation
Jose March-Leuba, Nuclear Engineering Consultant
Oak Ridge National Laboratory
Approved By:
Dr. Dale A. Powers, Chief, Maintenance Branch
Division of Reactor Safety
ATTACHMENT:
Supplemental Information
980M40317 980508
ADOCK 05000397
6
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TV
A Y
Washington Nuclear Project-2
NRC Inspection Report 50-397/98-01
During the performance of an NRC inspection reported in NRC Inspection Report 50-397/97-11
to determine the adequacy of certain WNP-2 facility license power distribution and fuel thermal
limits, the inspectors identified three unresolved items. This current special inspection was
conducted to review the licensee's response to the unresolved items using inspection personnel
from Region IV, the Office of Nuclear Reactor Regulation, and Oak Ridge National Laboratory.
gggini~rin
~
The licensee operated the Siemens Power Corporation's fuel in Core Cycles 7-12 in
excess of a revised operating limit minimum critical power ratio based on revised and
conservative ANFB-1125 correlation constant uncertainty (Section E1.2).
On the basis of the November 25, 1997, licensee response,
the safety limit minimum
critical power ratio was not exceeded during the actual events and transients
experienced
by the plant during Core Cycles 8-12. The licensee's analysis to determine
ifthe limitcould have been exceeded during Core Cycles 8-12 did not use licensing
bases assumptions,
bounds, and parameters
(Section E1.2).
Administrative controls and operating limits in place during Core Cycles 7-12 would not
have ensured operation within the envelope of the licensing basis.
Therefore, had the
limiting transient occurred with design basis operating conditions, the revised safety limit
could have been reached or exceeded (Section E1.3).
~
The vendor development and implementation of the minimum critical power ratio
operating and safety limits for WNP-2 fuel were not adequate to assure that the limits
were accurate and conservative.
Licensee oversight of the fuel vendors'esign
processes
and controls for the nuclear fuel supplied to WNP-2 failed to detect that an
inadequate Technical Specification limitwas developed.
The failure to establish
measures
to assure that the design bases were correctly translated into technical
specifications was identified as an apparent violation of Criterion III,Appendix B to
10 CFR Part 50 (Section E1.4).
0
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The unit operated at or near full power during the onsite inspection period.
E1
Conduct of Engineering
E1.1
~B~r~nI
This inspection consisted of engineering followup review of the licensee's response to
the three unresolved items identified during NRC Inspection Report 50-397/97-011.
During that inspection, it was determined that the licensee had operated Core
Cycles 7-12 with nonconservative values for the operating limitminimum critical power
ratio (OLMCPR) for the Siemens Power Corporation's (SPC) fuel. These limits are
specified in the Core Operating LimitReport. These limits, according to 10 CFR Part 50,
Appendix A, Criterion 10, are specified acceptable fuel design limits (SAFDLs). The
licensee's reactor core, associated
coolant, control, and protection systems are to be
designed such that there is appropriate margins to assure that SAFDLs are not
exceeded during any condition of normal operation, including anticipated operational
occurrences.
The nonconservative and inappropriate limits occurred because the SPC
(previous fuel vendor) critical heat flux data base did not contain a sufficient number of
data points to statistically support the advanced nuclear fuel-boiling (ANFB-1125)
correlation that was used to provide the uncertainty measurement
for measuring the
performance of the SPC fuel for operation in Cycles 7-12. Performance measuring of the
other fuel in the core was not affected by the ANFB-1125 correlation. The net result was
that the OLMCPR for the SPC fuel was nonconservative because the uncertainty applied
to the calculation was not statistically supported by the available data.
The limits for fuel
supplied by current vendor, Asea Brown Boveri/Combustion Engineering (ABB/CE) were
adequate.
Pertaining to Cycle 12 only, the OLMCPR was not calculated in accordance with NRC-
approved topical reports referenced in Technical Specification 5.6.5.b, by the current
fuel vendor (ABB/CE). Specifically, a condition set forth in the NRC-approved generic
Topical Report CENPD-300-PA, "Reference Safety Report for Boiling Water Reload
Fuel," July 1996, had not been satisfied.
Restriction 7 of the report stated that an
independent evaluation of the current fuel vendor's US96A7 correlation should be
performed and compared against test data.
Had this condition been met, the
inadequacy of the ANFB-1125 correlation would have been identified. As a result, the
inspectors determined that the ABB/CE US96A7 correlation used by ABB/CE to describe
-4-
the performance of SPC fuel for Core Cycle 12, was deficient because absolute errors
were propagated by application of the SPC correlation. At the end of that inspection, the
inspectors could not determine if a revised operating limitdeveloped from a statistically
valid correlation, was exceeded,
during Core Cycles 7-12. This was an unresolved item.
The safety limitminimum critical power ratio (SLMCPR) was also affected because
it was
determined by analyzing to obtain delta (6,) MCPR, starting from the OLMCPR for limiting
Therefore, using the nonconservative OLMCPR as a starting point to
determine ifthe defined limiting condition for SLMCPR would be reached or exceeded,
also yielded an incorrect value for the safety limit. The inspectors were not sure ifa
revised SLMCPR was exceeded or could have been exceeded any time during Core
Cycles 7-12. This was an unresolved item.
As a result of the above unresolved items, the inspectors also questioned the adequacy
of the design control process to assure that the nuclear fuel design bases were
successfully translated into facility license limits. This was a third and final unresolved
item.
Because of the determination by the NRC inspectors that the OLMCPR and SLMCPR
were nonconservative, the licensee submitted a technical specification amendment
request for Cycle 13 (the current cycle of operation) to set the SLMCPR at 1.13 for the
SPC fuel in Cycle 13. This represented
a significant change from the previous limits of
1.07 for Core Cycles 7-12. The new limitwas deteimined by the application of a
conservative multiplier applied to the methodology and the Cycle 13 benchmark
calculations.
The OLMCPRs were changed accordingly and the amendment was
approved, allowing startup of the unit.
The licensee chose to respond to the unresolved items by performing investigation and
analysis to justify a SLMCPR limitof 1.10 and an appropriately adjusted OLMCPR. This
was thought to be an accurate and conservative estimate based on the conditions that
would exist when the SPC database
issue was finally statistically resolved.
Concurrent
with the effort by the licensee, the NRC was reviewing the SPC development of a
method to estimate the additional uncertainty in the ANFB-1125 correlation by
application of an additive constant.
Again, this was necessary because of the limited
experimental critical heat flux data available. At the time'of the inspection, the NRC
review of the SPC submittal was not complete, so the use of 1.10 as a valid limitcould
not be recognized.
4
'
The licensee's response letter to the unresolved items, dated November 25, 1997, stated
that the revised OLMCPR had been exceeded numerous times during Core Cycles 7-12,
but that the revised safety limithad never been exceeded,
nor could it have been
exceeded given the analyzed conditions. The response,
however, did not address the
adequacy of the design control process to assure the design and implementation of
appropriate facility license limits for operation of the nuclear fuel.
-5-
E1.2 Assessment for Potential to Exceed Technical Specification Safety and Operating Limits
a.
In
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o
2
3
The inspectors reviewed the licensee's November 25, 1997, response to the unresolved
items with respect to adherence to the: (1) OLMCPR and SLMCPR for the SPC fuel
during Core Cycles 7-12, and (2) graphs depicting MCPR for Core Cycles 8-12. The
inspectors also reviewed licensee initial and followup submittals requesting a license
amendment and the amendment
to change the cycle specific MCPR for Core Cycle 7.
b.
In the November 25, 1997, response,
the licensee evaluated the impact of a revised
additive uncertainty constant and a chosen SLMCPR of 1.10, and concluded that:
~
The OLMCPR associated with an SLMCPR of 1.10 was exceeded numerous
times for each of the previous Core Cycles 7-12.
~
The SLMCPR was not exceeded for any of the actual operating events or
transients experienced during Cycles 7-12.
~
The SLMCPR could not have been exceeded during Cycles 7-12 for any
postulated anticipated operatio'nal occuriences (AOOs) ifthe actual plant
operating conditions during those cycles were used instead of the design basis
assumptions
used for reload transient analyses.
The inspectors agreed with the licensee's first two above conclusions, but disagreed
with the derivation of the third conclusion.
The licensee's position was that the limiting
transient should be analyzed assuming actual conditions that existed during Core
Cycles 7-12. Therefore, because the turbine bypass system was not declared
inoperable during high power operation for any of the previous cycles, the limitingAOO,
turbine trip without bypass, need not be analyzed.
Consequently,
in the response,
the
analysis was performed assuming that automatic operation of the turbine bypass was
available.
The inspectors'osition was that the design basis assumptions prescribed for the reload
transient analyses define the limitingAOO during high power conditions in the Final
Safety Analysis Report (FSAR). Therefore, analysis without bypass capability should be
performed to validate the analysis.
However, the licensee determined that it was
acceptable to analyze a turbine trip with the turbine bypass available and a'ssume it
would operate during the postulated event.
This approach was different from,
and'xplicitly
contrary to, the design basis assumptions for the analysis in Chapter 15 of the
FSAR.
0
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Limi Viol 'o
The inspectors reviewed WNP-2 operational data graphs for Cycles 8-12 (July 1992-
March 1997) to identify the operating conditions with lowest margin to critical power
during these five operating cycles.
Table
1 below shows the most limiting conditions for
each of these cycles using the old additive constant uncertainty values, which resulted in
an SLMCPR value of 1.07.
Table 1:
Most Limiting Operating Conditions Using Original Additive Constants
Uncertainty
Date
7/29/92
" 1/12/94
4/15/95
1 2/1/95
11/21/96
Cycle
10
12
Power %
99.7
99.9
99.9
100.0
100.1
Flow%
96.0
93.2
92.8
96.4
95.8
- FLCPR
0.996
0.987
0.980
0.979
0.955
1.26
1.31
1.30
1.27
"Fraction of Limiting Critical Power Ratio
In the November 25, 1997, response, the licensee evaluated the impact of a revised
correlation uncertainty on the SLMCPR and concluded that the SLMCPR value should
be at least 0.03 higher than assumed.
This conclusion was based on a revised
ANFB-1125 correlation uncertainty value of 0.0195, which was the least conservative
value under review by the NRC at the time of the inspection.
The application of the
0.0195 correlation uncertainty constant
resulted in the derivation of the 1.10 SLMCPR.
The response
also indicated that other uncertainties associated with transient hMCPR
and less certainty of the database
could result in a difference between corrected and
uncorrected SLMCPR of 0.06.
From discussion with NRC review personnel, the
inspectors determined that the correlation uncertainty value could be as high as 0.026,
which would result in larger MCPR penalty. Therefore the final approved safety and
operating limits were likelyto be more restrictive (higher), but the least restrictive (lowest)
expected value for SLMCPR was 1.10.
The licensee stated that their study to develop the response concluded that the
OLMCPR would have been exceeded
on numerous occasions during each of the
evaluated cycles (7-12) ifthe revised correlation uncertainty constant and resultant
additive MCPR constant uncertainty had been used to develop the limitfor the SPC
fuel. Also according to the licensee, the most significant departure from the revised
OLMCPR occurred during Core Cycle 8 where it was exceeded
by about 5 percent.
The inspectors determined through the review of graphs depicting MCPR for Core
Cycles 8-12, and applying licensing bases analysis values,
that the limitof 1.10
may have been exceeded
by as much as 10 percent, ifthe higher correlation
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uncertainty of 0.026 had been applied.
The inspectors noted that operating in excess
of the OLMCPR in the core operating limit report would have violated Technical Specification 3.2.2 ifthe appropriate OLMCPR value had been included in the core
operating limit report.
f
Limi Durin
c al0 era in
Ev n
In the November 25, 1997, response,
the licensee determined that, from 1991 to 1997,
WNP-2 was subject to five transients that could have potentially challenged the
SLMCPR. However, upon comparing relative severity, none of these transients
resembled the limiting transient described in Chapter 15 of the FSAR. The conditions for
these transients are summarized in Table 2, and, based on review of the data, the
inspectors determined that the SLMCPR was not exceeded
in any of these transient
events.
Table 2:.
Operational Events and Their Impact on SLMCPR
Date
11/19/91
8/15/92
8/3/93
Event
FW Controller
Failure
Instability
Event
MSIV Closure
Initial
- FLCPR
0.864
0.860
Initial
1.435
1.940
1.442
hCPR
0.05
0.37
0.04
1.385
1.570
1.402
2/18/95
4/4/95
Turbine Trip with
Bypass Available
Turbine Trip with
Bypass Available
0.842
0.952
1.473
1.303
-0
0.01
1.473
1.293
- Fraction of Limiting Critical Power Ratio
- Unknown, Unstable
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Limi Violation
The most limitingAOO for WNP-2 was identified as a turbine trip without bypass (TTNB).
According to the Cycle 11 transient analysis report, the TTNB AOO resulted in a 8 CPR
of 0.21 for SPC Atrium-9 fuel, when the uncorrected ANFB-1125 correlation was used.
As previously stated, the licensee had evaluated the effect of the revised correlation
constant for Atrium-9 fuel and concluded that a b,CPR correction of 0.03 was required for
the new limit (1.10). Thus, the TTNB AOO was estimated to result in a BCPR of 0.24
when the revised constant was used.
-8-
Table 3 shows the most limiting CPR for SPC fuel that could have been reached or
exceeded during a TTNB AOO using the above assumptions
at the worst point in each of
the previous five cycles.
This table indicates that, had a limiting TTNBAOO occurred,
the safety limitwould have been reached or exceeded regardless of whether the safety
limit MCPR value was 1.10 or 1.13.
Table 3:
Projected CPR During Design Basis Turbine Trip Without Bypass Starting
From Actual Conditions on Indicated Date
Date
7/29/92
1/1 2/94
4/1 5/95
12/1/95
11/21/96
Cycle
10
12
1.02
1.07
1.06
1.03
1.10
Licensee representatives
did not agree that the revised SLMCPR of 1.10 could have
been reached or exceeded under the conditions existing on the dates above, and stated
that additional information would be provided to substantiate their position.
In addition,
the inspectors and licensee representatives
disagreed on other issues related to the
assumptions
used in the analysis.
These disagreements
are listed below.
The licensee's analysis used a control rod scram insertion time that was based
on historical surveillance results and not the maximum insertion time for which
operation is allowed by the technical specifications.
The licensee's analysis assumed
a predicted actual axial power shape that was
based on previous end-of-cycle conditions, instead of the axial offset assumed
in
the FSAR licensing analysis.
The licensee's analysis assumed
an all-control-rods-out condition instead of the
limiting rod patterns used in the FSAR licensing analysis.
During the inspection, the licensee was unable to supply the same data for Core Cycle 7,
as that supplied for Cycles 8-12, without significant added effort. The inspectors
determined that they had sufficient data.
However, as shown above, during each of the
evaluated cycles, the potential existed to reach or exceed the revised SLMCPR during a
turbine trip event with the bypass system unavailable.
This transient described in the
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licensing basis as the limiting event that is analyzed
from full power operation in
Chapter 15 of the FSAR. The inspectors believed that the potential existed to reach or
exceed the revised SLMCPR during all five of these cycles.
However, the licensee's
representatives
indicated they would be able to provide additional information to refute
the inspectors'osition.
The review of the additional information is discussed later in
Section E1.3.
In a letter from the licensee to the NRC dated February 28, 1991,
and supplemented
by
letters dated March 21, 1991, and April26, 1991, the licensee requested to amend the
facility license.
The amendment sought was to modify the facilitySLMCPR and the
associated
bases to reflect cycle specific analysis resulting from use of a new reload
methodology.
The new methodology was necessary to account for the loading of fuel
from a new vendor (SPC).
The inspectors noted that the amendment was granted to change the safety limitfor the
SPC fuel to 1.07.
The 1.07 limitwas in effect for Core Cycles 7-12. As noted above, it
was necessary for the licensee to change the limit in a more conservative direction once
the NRC identiflied the undesirable conditions at the vendor facility.
~nisi
The inspectors concluded that the licensee had operated the SPC fuel in Core
Cycles 7-12 in excess of a revised OLMCPR based on revised correlation constants and
additive uncertainties applied to the SLMCPR of 1.07.
On the basis of the November 25,
1997, response, the SLMCPR was not exceeded during the actual events and transients
experienced by the plant during Core Cycles 8-12.
Data was not provided by the
licensee to assess
whether the limitwas exceeded during Core Cycle 7. The licensee's
analysis to determine ifthe limitcould have been exceeded during Core Cycles 8-12 did
not use licensing bases assumptions,
bounds, and parameters.
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The licensee provided additional information on March 9, 1998, to refute the NRC
position that the potential existed to reach or exceed the revised SLMCPR during Core
Cycles 8-12. The inspectors reviewed this information in order to make the
determination ifthe potential still existed.
Findin
and Obs
rva io s
At the end of the on-site portion of the inspection, licensee representatives
and the
inspectors agreed on the selection of a specific time in Core Cycle 8 for analysis to
provide the additional information. The March 9, 1998, additional information submittal
documented two licensee calculations using the NRC-approved models WNP-2
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RETRAN and VIPRE calculational codes.
The conditions selected for analysis by the
licensee were the actual conditions that occurred near the end of Cycle 8. The simulated
transient used for the analysis was the AOO of TTNB because
it was'he most limiting
transient for the analyzed baseline conditions according to Chapter 15 of the FSAR. The
inspectors determined that the conditions and transient selected for analyses were
appropriate.
One calculation simulated a load rejection without bypass using the actual plant
conditions (as opposed to licensing bases initial conditions).
For this calculation, the
licensee used a condition near the end of cycle with a minimum fraction of limiting CPR
(MFLCPR) value of 0.993, which was considered limitingfor Cycle 8. The actual core
power, coolant flow, vessel pressure,
and rod pattern were used instead of the licensing
bases initial conditions. The results of the licensee's calculation indicated that, by using
the actual initial.operating conditions for this event, the loss of CPR margin (b,CPR) was
as much as 0.125. This reduction was accomplished through a combination of the lower
initial power, flow, and pressure and because some rods were partially inserted in the
core, which provided for a faster negative reactivity insertion during the scram.
The
licensee concluded that this additional margin compensated
amply for the erroneous
safety limitvalue assumed during operation; thus, a load rejection without bypass would
not have caused a safety limitviolation should it have occurred during the Cycle 8 most
limiting conditions.
The other calculation reproduced the licensing-bases
event of load rejection without
bypass event for Cycle 8. This calculation indicated that the hCPR during a licensing-
bases load rejection without bypass was 0.259. This reduction in margin was
significantly higher than the reduction assumed for actual conditions (0.125).
During
actual operation, the licensee was not aware of the potential safety limitviolation and did
not take action to minimize its impact. Therefore, administrative controls and operating
limits would not have precluded facilityoperation within the licensing basis envelope.
L
As observed
in Table
1 of this report, WNP-2 was operated with MCPR values as low
as 1.26 during Cycle 8.
Therefore, margin was insufficient to preclude reaching or
exceeding a revised safety limitof 1.10 for the limiting transient initiated just within the
envelope of licensing basis.
This calculation confirmed the inspection finding that, by
operating with the erroneous. operating and safety limitvalues, the facilitywas operated
in a mode that potentially could have resulted in exceeding a revised safety limit based
on a conservative application of the ANFB-1125 correlation.
~nisi
Administrative controls and operating limits in place during Cycles 7-12 would not have
ensured operation within the envelope of the licensing basis.
Therefore, had the limiting
transient occurred with design basis operating conditions, the revised safety limitcould
have been exceeded.
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'o
The inspectors reviewed procedures and documentation that licensee personnel
depicted as representing the integrated process for the determination of technical
specification core power distribution limits. Discussions were also held with fuel design
engineering and fuel management
personnel involved in the fuel and core thermal limits
determination process.
The inspectors conducted a review of the licensee's process for core reload design
control to identify and evaluate features and elements that would challenge the adequacy
of the vendor supplied core operating and safety limits. This effort necessitated
additional review of fuel management
and procurement processes.
b.,
fv
I
According to licensee personnel, the process of core reload was processed
as a design
modification. The inspectors reviewed Procedure PPM 1.4.1, "Plant Modifications,"
Revision 23. This upper-tier site document provided the administrative guidance for
planning, implementing, and tracking modifications, but did not reach to the level of
providing requirements for detailed review of vendor-provided limits. The inspectors also
reviewed Procedure El 2.8, "Generating Facility Design Change Process," Revision 14.
This lower-tier engineering document was the engineering instruction for processing
design modifications. The inspectors were unable to identify any attribute in this
procedure that would result in an evaluation of the adequacy of vendor-provided
cycle-specific operating and safety limits to meet the licensing basis.
Licensee representatives
stated that the supplier's quality surveillance program was
instrumental in the process for core reload design control. The inspectors reviewed
Procedure SQI 7.4, "Source Surveillance Activities," Revision 11. This procedure
describe'd the methods for the quality assurance
organization to perform surveillance
activities of vendors who were a source of services, material, and hardware.
This
procedure provided the administrative guidance needed to schedule, plan, perform, and
report source surveillance activities. The inspectors determined that this procedure did
not cause the WNP-2 staff to perform activities associated with validating the adequacy
of vendor-supplied license limits.
Licensee representatives
also stated that design review of fuel vendor analysis was
performed in accordance with Fuel Management Instruction FMI 4-3, "Reload and Fuel
Design Reviews," Revision 5, for every new core, prior to startup.
This instruction
described requirements and responsibilities for conducting design reviews of reload
nuclear fuel, including review of the methods and procedures used by vendors in making
fuel design decisions.
The procedure required the development of a review checklist and
suggested
topics to be included. According to the procedure, the checklist was expected
-12-
to consist of a series of questions to be answered during the core design review. The
inspectors reviewed the design review packages,
including the completed design review
checklists, for Core Cycles 10 and 11. None of the questions in the checklists required
evaluating the validity of the vendor-proposed
fuel operating and safety MCPR limits.
In summary, licensee oversight of the previous and current fuel vendors did not assure
that the core power distribution and fuel thermal limits submitted by letters dated
August 2, 1990, February 28, 1991, and May 20, 1991, were accurate and conservative.
10 CFR Part 50, Appendix B, Criterion III requires that established measures
assure that
applicable regulations and the design basis are correctly translated into specifications,
drawings, procedures,
and instructions.
The inspectors determined that established
measures
did not assure the translation of regulatory requirements and the design basis
for the reactor core into the facility license Technical Specifications.
Consequently, the
licensee failed to ensure that sufficient margin was available in its SLMCPR SAFDL.
This was an apparent violation of Criterion III of Appendix B to 10 CFR Part 50
(50-397/9801-01).
The inspectors concluded that vendor development and implementation of MCPR
operating and safety limits for WNP-2 fuel were not adequate to assure that the limits
were accurate and conservative.
Licensee oversight of the fuel vendors'esign
processes
and controls for the nuclear fuel supplie'd to WNP-2, failed to detect tliat
an'nadequate
technical specification limitwas developed.
The failure to establish
measures to assure that the design bases were correctly translated into technical
specifications was identified as an apparent violation of Criterion III,Appendix B to
Miscellaneous Engineering Issues
E8.1
Iv
I
-
7 -: failure to establish measures to assure that
design bases are translated into technical specification limits in accordance with
Criterion III of Appendix B to 10 CFR Part 50, for Core Cycles 7-12. This item was
evaluated in Section E1.2.
E8.2
Iv
I
- 9
7 1-0: failure to operate the reactor core in
accordance with a revised OLMCPR for Technical Specification 3.2.2, for Core
Cycles 7-12. This item is evaluated in Section E1.3.
E8.3
I
d
nr
Iv d I
5 -
7/9711-: failure to assure that a revised SLMCPR
could not be exceeded during AOOs for Core Cycles 7-12. This item is evaluated in
Section E1.3.
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X1
Exit Meeting Summary
V
The inspectors met with licensee personnel on February 12, 1998, for a briefing of
preliminary findings. Licensee representatives
disagreed with the inspectors conclusion
that the potential existed to exceed the revised safety limitfor Core Cycles 8-12.
Consequently, the licensee's representatives
indicated their intent to provide additional
information for NRC review. The licensee's additional information was submitted in its
March 9, 1998 letter. Subsequently,
a supplemental telephonic exit meeting was held on
April22, 1998. At that time, the licensee was informed of the apparent violation
discussed
in this inspection report and the NRC's intent to hold a predecisional
enforcement conference.
Qs
tl
SUPPLEMENTAL INFORMATION
PARTIALLIST OF PERSONS CONTACTED
MLi K5lK
D. Atkinson, Manager, Quality Assurance
P. Bemis, Vice President, Nuclear Operations
S. Bian, Supervisor, Reactor/Fuels Engineering
R. Casarant, Engineer, Supplier Quality
D. Coleman, Acting Manager, Regulatory and Industry Affairs
T. Cong Hoang, Engineer, Fuel Design
J. Fisher, Engineer, Fuels Design
P. Inserra, Manager, Licensing
W. Kiel, Regulatory Programs Specialist
W. Moore, Consulting, Engineer, Fuel Design
D. Richey, Engineer, Fuels Design
G. Smith, Plant General Manager
J. Teachman, Principal Engineer, Fuel Design
R. Torres, Manager, Reactor Fuels Engineering
R. Webring, Vice President, Operations Support
QX~~ZS
W. Harris, Asea Brown Boveri/Combustion Engineering
J. Ingham, Siemens Power Corporation
S. Boyinton, Senior Resident Inspector
H. Wong, Chief, Projects Branch E
C. Poslusny, Jr., Project Manager,.Office of Nuclear Reactor Regulation
INSPECTION PROCEDURES USED
Followup - Engineering
~Oned
50-397/98-01
ITEMS OPENED, CLOSED, AND DISCUSSED
Failure to Control Design of Facility Technical Specification
Limit(Section E1.4)
t
e
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GLzzd
50-397/9711-01
50-397/9711-02
50-397/9711-03
Adequacy of Design Control of Facility License Limits
(Section E8.3)
Potentially Exceeding OLMCPR During Core Cycles 7-12
(Section E8.3)
Determination of Potential to Exceed SLMCPR During
Anticipated Operational Transients (Section E8.4)
El 2.8
FMI 4-3
PPM 1.4.1
SQI 7.4
LIST OF PROCEDURES REVIEWED
Generating Facility Design Process,
Revision 14
Reload and Fuel Design Reviews, Revision 5
Plant Modifications, Revision 23
Supplier Quality Instruction, Revision 11
LIST OF DOCUMENTS REVIEWED
Nuclear UtilityProcurement Issues Committee Audit of Siemens Power Corporation,
July 24, 1997
Procurement Quality Assurance Vendor File, Siemens Power Corporation
Updated Final Safety Analysis Report, Chapter 15
Critical Power Ratio Graphs, WNP-2 Core Cycles 8-12
Core Reload Design Review Reports, WNP-2 Core Cycles 9-11
License Response to Unresolved Items, November 25, 1997
Cycle 13 Core Operating Limit Report
WNP-2 Technical Specifications
License Response to Unresolved Item 50-397/9711-03, March 9, 1998