ML18153C762
| ML18153C762 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 09/16/1991 |
| From: | Burnett P, Crlenjak R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18153C760 | List: |
| References | |
| 50-280-91-20, 50-281-91-20, NUDOCS 9110080100 | |
| Download: ML18153C762 (11) | |
See also: IR 05000280/1991020
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W.
ATLANTA, GEORGIA 30323
Report Nos.:
50-280/91-20
and 50-281/91-20
Licensee:
Virginia Electric and Power Company
5000 Dominion Boulevard
Glen Allen, Virginia 23060
Docket Nos.:* 50-280 and 50-281
License Nos.:
Facility Name:
Surry 1 and 2
1991
Approved by:
--
. V. Crlenjak, Ch'ef
Operational Programs Section
Operations Branch
Division of Reactor Safety
SUMMARY
Scope:
Date Signed
Dak:e signed
This routine, announced inspection addressed the areas of Unit 2,
cycle 11, startup tests, thermal power analysis, power
distribution monitoring, nuclear instrument calibrations, and
followup of an unresolved item.
Results:
Following completion of the low power tests for Unit 2, cycle 11,
it was determined that control rod F-6 (in control bank D) had
been unlatched from its drive and fully inserted throughout the
tests.
Discussions were held with licensee personnel to
determine and evaluate their conclusions that reperformance of
the tests was not necessary following relatching of F-6.
The
licensee's conclusions were found to be acceptable (paragraph
2. c) *
New feedwater flow venturis were installed in Unit 2 during the
recent outage.
Tests to compare performance of the new venturis
with the steam flow venturis used for routine thermal power
measurements were ongoing during this inspection.
Review of the
test results to date indicated that no single procedure was
9110080100 910920
- DR
ADOCK 05000280
2
capturing contemporaneously all of the data necessary to perform
the analysis and obtain a cross calibration of the steam flow
ventures.
This issue was identified as an inspector followup
item (paragraph 5).
Review of monthly core performance characteristics for Unit 1,
cycle 11, confirmed that it has been operating with acceptable
hot spot factors, reactivity balance, and power distribution
throughout the cycle (paragraph 6).
The interval for surveillance of Unit 2 hot channel factors was
found to be inviolation of Technical Specifications in one
instance (paragraph 8) .
1.
Persons Contacted
Licensee Employees
REPORT DETAILS
K. L. Basehore, Supervisor, Nuclear Engineering
w. R. Benthall, Supervisor, Licensing
- R. M. Berryman, Manager of Nuclear Analysis and Fuel
- R. E. Bilyeu, Licensing Engineer
D. Dziadosz, Supervisor of Core Design
- D. s. Hart, Supervisor, Quality Assurance
- J. W. Henderson, Lead Reactor Engineer
- M. R. Kansler, Station Manager
- D. c. Lawrence, Reactor Engineer
- M.A. Paul, Reactor Engineer
- J. A. Price, Assistant Station Manager
- E. R. Smith, Jr., Site Quality Assurance Manager
- T. B. Sowers, Superintendent of Engineering
Other licensee employees contacted included engineers,
technicians, security force members, and office personnel.
NRC Resident Inspectors
M. Branch, Senior Resident Inspector
s. G. Tingen, Resident Inspector
- J. W. York, Resident Inspector
- Attended exit interview on July 12, 1991.
Acronyms and initialisms used throughout this report are
listed in the last paragraph.
2.
Startup Tests for Unit 2, Cycle 11 (72700, 61708, 61710)
a.
Precritical Tests
NPT-RX-007 (Revision 1), Hot Rod Drops, was performed at
hot, full-flow conditions on June 2, 1991.
All measured
drop times satisfied the TS 3.12.C.1 requirement of being
less that 2.4 seconds from interruption of power to dash pot
entry.
The timing measurements were made from individual
chart records of rod drop speed versus time.
The trace for
rod F-6 was not as smooth, prior to deceleration into the
dashpot, as that of the other rods.
However, the test
engineer's conclusion that the trace was the result of
otherwise undetected electrical noise was not unreasonable;
the drop time for F-6 was within the span of the other
results.
Evaluation of the incore power distribution map obtained at
b.
2
30 percent RTP (see paragraph 2.c) confirmed that control
rod F-6 was fully inserted in the core (unlatched).
The
reactor was cooled down, the head removed, and rod F-6
latched to the rod extension.
Following a return of the system to hot, full-flow
conditions, the procedure was performed again with
acceptable results and traces for all control rods.
The inspector independently evaluated the rod drop time
traces for several of the control rods for both sets of
measurements.
The inspector's results were consistently
shorter than those reported by the licensee.
Discussions
with the test engineer produced agreement on the
identifications of the point of power interruption and the
point of dashpot entry on the chart records.
However, the
licensee used a ruler and the selected chart speed, 6
ins/sec, to determine time.
The inspector counted the
cycles of a 60Hz signal also present on the chart to
determine time.
The latter method is consistent with
industry practice and guards against chart speed differing
from nominal.
In the current case, the chart speed appeared
to be faster than nominal, which lead to a conservative
estimate of control rod drop time.
If the chart speed had
been less than nominal, a more likely case with high speed
recorders, the results would have been non-conservative.
The licensee is considering revising the procedure, prior to
the next startup test program, to use the imposed 60Hz
signal to measure rod drop time.
Initial Criticality and Post-Criticality Tests
Initial criticality for Unit 2, cycle 11, was achieved using
standard plant procedures, which provide for some monitoring
of ICRR, but not to the extent usually seen at similar
facilities for the first criticality in a cycle.
No special
effort was made to assure that the SRNis were responding
predominately and proportionately to neutrons.
Many similar
facilities use statistical reliability tests to assure
confidence in the SRNis prior to initial cycle criticality,
during fuel movement in the vessel, and during lowered loop
operations, when they are particularly vulnerable to a
dilution accident.
The licensee has a procedure for
performing the statistical analysis, but has no procedure or
practice for implementing it.
The licensee is reviewing
this issue.
Post-critical testing was guided by 2-PT-28.11 (Revision 1),
Startup Physics Testing.
Several good features of that
procedure were noted:
At the beginning of the procedure, high flux trips and rod
3
stop setpoints were set to 85 and 81 percent of indicated
power, respectively.
Both chambers of the PRNI connected to reactivity computer
were confirmed to be supplying current to the computer.
The acceptable range of reactivity computer application was
defined in terms of the measurement of its performance.
The acceptance criteria required that the ITCs measured
during heatup and cooldown agree within 1 pcm/ Degrees F.
However, that criterion was not repeated on the Test Result
and Evaluation sheet.
Test Result and Evaluation sheets are prepared for most. of
the measured parameters, such as control rod worth and MTC.
The sheets summarize the test results and conditions and
compare the measured values with the predicted values at
design conditions and at test conditions.
The acceptance criterion for the measured DBW worth was
agreement within +/-10 percent of the predicted value.
This
is a more stringent requirement than the +/-15 percent imposed
by ANSI/ANS- 19.6.1-1985.
Weakness noted in the procedure included:
The method for determining the point of adding heat was not
well defined, and no margin between that point and the upper
limit for zero power testing was specified.
-No Test Result and Evaluation sheets were provided for the
point of adding heat, or for the incore-excore nuclear
instrument correlation scheduled by the,procedure.
-The source of data to correct the test control rod bank
predicted worth for the position of the reference is not
given.
One difference between the procedure and ANSI/ANS-19.6.1 was
noted, but is not characterized as either a strength or a
weakness:
In the standard, error is defined as (P-M)/M
where Pis the predicted value of a parameter and Mis the
measured value.
In the procedure, error is defined as (M-
P)/P.
At the limits, one approach can accept results
rejected by the other and vice versa.
In practice, none of the results from this series of tests
was near the limit of acceptability by either error
analysis.
In the performance of the tests one particularly good
4
feature was noted:
The temperature spans for ITC
measurement for both heatup and cooldown met or slightly
exceeded the 5 Degrees F specified by the procedure.
The
reliability of this endpoint-dependent measurement increases
with increasing temperature span.
c.
Testing with One Control Rod Inserted
The first flux map (at about 30 percent RTP) for Surry
Unit 2, cycle 11, revealed that the control rod in position
F-6 was fully inserted.
That raised a question about the
adequacy of the reactivity measurements made at zero power.
All zero power tests (CBC at ARO, ITC/MTC, control rod bank
worths, and DBW) had satisfied numerical acceptance criteria
from predictive calculations, which, naturally, had not
modelled a fully inserted rod.
The licensee then made calculations to predict the test
parameters with F-6 inserted.
Discussions with licensee NAF
personnel confirmed that no shortcuts (interpolations,
extrapolations, or perturbations) had been made to earlier
calculations to make the new predictions.
The new predic-
tions were made using the same methods, with more effort,
since core symmetry could not be assumed, as in the first
predictions.
As shown in Table 1, below, insertion of F-6
did not greatly change the predicted values of any of the
test parameters, and the measured values satisfied the
acceptance criteria for both sets of calculations.
The licensee's strength in core analysis was also demon-
strated by accurate prediction of the power distribution at
30 percent RTP with F-6 inserted.
Since the purpose of the zero power tests is to assess or
challenge the adequacy of the predictive methods, and all
acceptance criteria for agreement between predictions and
measurements were satisfied; the zero power tests are
acceptable as performed.
The licensee's program for post-refueling startup test
program is in substantial agreement with ANSI/ANS-19.6.1-
1985Property "ANSI code" (as page type) with input value "ANSI/ANS-19.6.1-</br></br>1985" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., Reload Startup Physics Tests for Pressurized Water
Reactors.
No violations or deviations were identified.
3.
Operation with Reconstituted Fuel (61702, 61706)
For purposes of thermal-hydraulic and critical heat flux
analyses, a flow channel is usually defined or constructed
from four fuel pins.
The channel includes one-quarter of
the cladding of each of the four pins plus the enveloped
coolant.
An unheated pin, such as a control rod or instru-
ment thimble introduces a cold wall into the channel,
_,
4.
5
resulting in an equivalent heated diameter less than the
hydraulic diameter.
Cold walls have been demonstrated to
reduce the critical heat flux of a flow channel.
Both units are currently operating with reconstituted fuel
bundles, which, as a result of replacing damaged fuel pins
with solid Zircaloy rods, have more cold wall in some flow
channels than was considered in the development of the
critical heat flux correlations used in the analyses of fuel
performance.
Much of the current concern for operation with
excess cold wall is mitigated by the fact that all
reconstituted fuel assemblies are once-burned, and, hence,
are not peak power assemblies.
Westinghouse is scheduled to submit a topical report on this
subject, for review by NRR, by mid-August 1991.
This issue will continue to be monitored in future
inspections at this and similar facilities.
Nuclear Instrument Calibrations (61705)
A new single point method of correlating incore axial offset
with the excore axial flux difference indicated by the power
range nuclear instruments has been instituted at Surry.
The
statistical analysis and arguments for the single point
method vice the earlier multipoint method appear to be well-
founded.
The licensee has identified those conditions under
which return to the multipoint method will be required.
The
licensee's evaluations of this methodology are documented in
the following reports, which were reviewed by the inspector:
PM-325, Evaluation of Single Point Calibration
Methodology for Surry Power Station, with four addenda,
dated from September 14, to December 6, 1990.
Technical Report NE-815, Evaluation of Excore Channel
Single-Point Calibration Methodology, December 1990.
NUCLEAR CORE DESIGN MANUAL, USER'S COPY, PART VII,
CHAPTER E, Power Range Detector Calibration Versus
Burnup, Revision 1, January 1991.
The licensee's implementation of this methodology is through
procedures l/2-NPT-RX-005 (Revision O), Single Point Power
Range Nuclear Instrument Calibration (Effective May 31,
1991).
Review of completed copies of both procedures
confirmed that the procedure was being performed with
acceptable frequency and results.
No violations or deviations were identified.
6
5.
Thermal Power Monitoring (61706)
New feedwater flow venturis were installed in Unit 2 during
the past outage, and will be used to evaluate the
performance of the steam venturis.
(This licensee is one of
the few to use steam flow venturis for plant calorimetrics.)
Data are being collected by performance of the following
procedures:
-2-ST-302 (Revision O), Calibration of Steam and Feedwater
Flow Transmitters at Power, and
- ,
I
-ENG-35.0 (Revision O), Calculating Reactor*Power, Delta T
Setpoints and RCS Flow.
However, no one test procedure was capturing all of the
information necessary to evaluate the flow and
calorimetrics, and it was not clear that sufficient,
contemporaneous data were being collected to accomplish
those tasks.
The licensee was reminded that industry
experience is that new or cleaned feedwater venturis retain
their precision for only 30 to 60 days of operation.
The
licensee acknowledged that only a limited time remained to
complete the task.
The licensee's activities in this area
will be tracked as inspector followup item 50-281/91-20-0l:
Correlate thermal power and secondary side flow using both
feedwater and steam flow venturis early in cycle 11.
No violations or deviations were identified.
6.
Unit 1, Core Performance surveillance Activities (61702,
61707)
The monthly Unit 1 Core Follow Reports for cycle 11 were
reviewed by the inspector.
The review confirmed that
surveillances of FQ, FdH' and reactivity anomaly were being
performed at the required frequency and with satisfactory
results.
No violations or deviations were identified.
7.
Audit of Reactor Engineering Activities (72700, 61702)
Much of the same subject area addressed in this report was
audited by the licensee in Quality Assurance Audit 91-09,
Nuclear Fuel.
The inspector was provided a copy of the
final audit report after the inspection, for in-office
review .
7
With respect to onsite surveillance and test activities *by
reactor engineering and core design and follow activities by
NAF, the report demonstrates that QA was able to perform a
peer review of these activities and to identify the
significant strengths and weaknesses of the audited
organizations.
The resolutions of issues raised by the audit will be
reviewed in later inspections.
8.
Followup of Open Items (92701)
(Closed) Unresolved item 50-280 and 281/90-29-0l:
The
interval between surveillances of hot channel factors
exceeded 1.25 EFPM.
The Unit 2, cycle 10, records revealed that the interval
between surveillances was 1.44 EFPM (44.8 EFPD) between July
18, 1990 and September 4, 1990.
TS 4.lOB requires that the
hot channel factors of TS 3.12 shall be determined every
EFPM.
TS 4.02 allows a 25 percent tolerance on surveillance
intervals, or a maximum of 1.25 EFPM, in this case.
The
licensee's position was that the language of the
specification requires the surveillance in each full-power
month, but does not limit the interval, which might then be
nearly 60 EFPD.
They further claimed that the NRC has found
this interpretation and implementation of the surveillance
requirement satisfactory in the past, but provided no
documentation of that claim.
Review of the affected TS and their BASES was conducted in
the regional office and by OGC and NRR/OTSB.
The conclusion
reached was that the surveillance should be conducted every
EFPM (31 EFPD), with the grace period of TS 4.02 applying.
Hence, this issue has been identified as a violation, 50-
281/91-20-01;
The interval between surveillances of hot
channel factors exceeded 1.25 EFPM resulting in a delay in
assessing the acceptability of continued operation of the
core.
9.
Exit Interview
The inspection scope and findings were summarized on
July 12, 1991, with those persons indicated in paragraph 1
above.
The inspector described the areas inspected and
discussed in detail the inspection findings.
No dissenting
comments were received from the licensee.
Proprietary
material was reviewed in the course of this inspection, but
is not included in this report.
The licensee was informed
by a telephone call on August 16, 1991, that a violation
would be issued for an excessive surveillance interval for
10.
8
Unit 2 hot channel factors.
During a later telephone
conversation on September 4, 1991, licensee management
stated they disagreed with the violation.
Acronyms and Initialisms used throughout this report
ANSI
ARO
CBC
DBW
EFPM
FdH
Fg
Hz
HZP
ICRR
Ins/Sec
MTC
NAF
OTSB
pcm
ppmB
SRNI
TS
American Nuclear Society
American National Standards Institute
All rods out
Critical boron concentration
Differential boron worth
Effective full power day
Effective full power month
Enthalpy rise hot channel factor
Heat flux hot channel factor
Hertz
Hot zero power
Inverse countrate ratio
Inches per second
Isothermal temperature coefficient
Moderator temperature coefficient
Nuclear Analysis and Fuel Department
Office of Nuclear Reactor Regulation
Technical Specifications Branch
Office of the General Counsel
Percent millirho
Parts per million boron
Power range nuclear instrument
Quality assurance
Rated thermal power
Source range nuclear instruments
Technical Specifications
TABLE 1
Surry 1, Cycle 11, Startup Physics Test Results
Predicted
Parameter
Measured*
Tolerance
CBC (ARO,HZP), ppmB
1926
1930
1912
+/- 50
ITC (ARO,HZP), pcm/°F
-0.53
-1.04
-1. 23
+/- 3.0
DBW (ARO,HZP}, pcm/ppmB -7.51
-7.31
-7.35
+/- 10%
Reference Rod Bank
Worth, pcm
1375
1337
1368
+/- 10%
Total Worth of All
Rod Banks, pcm
5480
5730
5623
+/- 10%
With rod F-6 inserted
Without rod F-6 inserted