ML18152B118
| ML18152B118 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 08/24/1988 |
| From: | Burnett P, Jape F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18152B116 | List: |
| References | |
| 50-280-88-29, 50-281-88-29, NUDOCS 8809070170 | |
| Download: ML18152B118 (6) | |
See also: IR 05000280/1988029
Text
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA ST., N.W.
ATLANTA, GEORGIA 30323
Report Nos.:
50-280/88-29 and 50-281/88-29
Licensee:
Virginia Electric and Power Company
Richmond, VA
23261
Docket Nos.:
50-280 and 50-281
Facility Name:
Surry 1 and 2
License Nos.: DPR-32 and DPR-37
Inspection Conducted:
July 18 - 22, 1988
Inspector: ~
P.
. Burne
Approved bcd.£k~~
Jtt/ F. Jape, Section Chief
t
Engineering Branch
Division of Reactor Safety
SUMMARY
y-11-f<?"
Date Signed
Scope:
This routine, unannounced inspection addressed the areas of
post-refueling startup tests (Unit 1), thennal power monitoring, post-trip
reviews, and surveillance of RCS leakage.
Results:
One violation was identified, incorrect constants were used in the
surveillance procedure (1/2-PT-10) for RCS leakage measurement - paragraph 5.
Post-inspection, in-office review of the approved topical report on rod swap
methodology revealed differences in the post-refueling startup test program as
defined in that report and as actually performed.
The issue will be tracked as
an inspector followup item for further inspection - paragraph 2.
The licensee has installed new transient recorders (GETARs). The GETARs monitor
a total of 450 points on the two units. Since the installation, the post-trip
reviews have become more quantitative and more detailed, reflecting the
improved quality of transient data.
8809070170 80~8~~80
ADOC~
PNU
G.
I.
Persons Contacted
Licensee Employees
REPORT DETAILS
- D. L. Benson, Station Manager
- R. H. Blount, II, Superintendent of Technical Services
T. R. Brookmire, Nuclear Engineer, Nuclear Fuel Operations
- R. W. Cross, Nuclear Specialist
C. A. Ford, Nuclear Engineer, Nuclear Fuel Operations
- G.D. Miller, Licensing Coordinator, Surry
- H. L. Miller, Assistant Station Manager
- J. A. Price, Manager, Quality Assurance
Other licensee employees contacted included engineers, operators, and
office personnel.
NRC Resident Inspector
- W. E. Holland, Senior Resident Inspector
- Attended exit interview
Acronyms and initialisms used throughout this report are defined in the
final paragraph.
2.
Post-Refueling Startup Tests - Unit I (72700, 61708, 61710)
Unit I was made critical for Cycle IO on July 14, 1988, using procedure
I-OP-IC.
.
Startup testing was performed under the guidance and control of 1-PT-28.11
(Issue of July 4, 1988 ), Startup Physics Testing and the approved Surry
1, Cycle 10, Refueling Physics Test Schedule.
The ARO C was 1567 ppmB,
which was less than the predicted 1610 pp~B, but within ~he performance
criterion of+/- 50 ppmB.
At ARO, both the ITC and MTC were negative, -3.2
pcm/°F and -1.52 pcm/°F, respectively.
Prior to rod worth measurements
by rod swap, control bank B was designated the reference bank and its
worth measured against boron dilution. The measured worth of control bank
B was 1134 pcm, which was within acceptable agreement with the predicted
worth of 1242 pcm.
All of the remaining control rod banks were within
acceptable agreement with prediction when measured by rod swap methods.
The sum of the measured worths and the predicted sum agreed within the
performance criterion.
2
PT-28.11 controls only the collection of data for the measurements.
The
analysis of test data is performed off site by the Nuclear Fuels Operation
Subsection.
That work is not controlled by procedures approved by the
SNSOC, but by task instructions approved by departmental management.
The
following task instructions pertinent to startup testing were reviewed:
1.1
Preparation for Physics Test (Revision 4, 10/7/87)
1.10
Readiness Checklist for Startup Physics Testing (Revision 3,
6/12/85)
2.1
Analysis of Hot Zero Power Rod Worth Data (Revision 5, 12/18/85)
2.2A
Rod Swap Design Data Generation (Revision 4, 12/18/85)
2.2B
Analysis of Hot Zero Power Rod Worth Data (Rod Swap) (Revision 4,
12/18/85)
2.3
Analysis of Boron Endpoint Data (Revision 6, 12/18/85)
2.4
Analysis of Boron Worth Coefficient Data (Revision 3, 12/18/85)
2.5
Analysis of Hot Zero Power Isothermal/Moderator Temperature
Coefficient Data (Revision 2, 12/12/84)
These instructions appeared to provide the level of detail and guidance
necessary for an appropriately trained ~uclear engineer to perform the
subject activities.
The instructions require peer review of data
production or analysis activities.
Subsequent to the on-site portion of the inspection, the licensee provided
a copy of their NRR-approved topical report on Control Rod Reactivity
Worth Determination by the Rod Swap Technique,
VEP-FRD-36 A,
December
1980.
During the in-office review of that report, it was noted that the
test program conducted for Surry 1, Cycle 10 did not contain all* of the
elements described in the report. Specifically, there was no measurement
of ITC with the reference bank inserted as called for in Section 2.2 item
3) and in Table A.2.
Equation 1 on page 7 is used to correct observed
data for differences from a reference calculation plant conditions.
The
adjustment for temperature changes utilizes the measured ITC obtained with
the reference bank inserted.
Table A.2 also calls for flux distribution
map to be made at zero power under rodded conditions.
No zero power flux
map was made under any conditions.
Neither of these deviations from the test program described in the topical
report have a significant effect on the measurement of rod worth.
There
would be a small error in rod worth if the RCS average temperature did
vary a few degrees F during the rod swap measurements and went
u.ncorrected*.
In practice, procedure 2.2B uses the predicted ITC.
The
zero power flux map has no bearing on rod worth results.
In fact few
licensees perform zero power flux maps after refueling, because, in
general, they have proved to be uninformative.
Nevertheless, the licensee
should have formally evaluated and possibly reported their decision to
deviate from the commitments in the approved topical report.
Actual
licensee action will be addressed in a future inspection (Inspector
Followup Item 280, 281/88-29-01).
Otherwise, comparison of NFOS ~rocedures 2.2A and 2.2B with the topical
report showed the approved methodology was being implemented.
3
3.
Core Thermal Power Evaluation (61706)
(Closed)Inspector followup item 280,281/86-40-02:
Review steam and
feedwater flow venturi calibration and reliability. This item stems from
the observation that steam flow was used as the basis for thermal power
measurement rather th1-2,t the more commonly used feedwater fl ow.
The
licensee
1s evaluation of the relative merits and reliability of the two
measurements is presented in their internal report, Surry Power Station
Secondary Plant Performance Evaluation, Phase 2,. issued in August 1984.
Review of that report indicates the selection of steam flow for power
measurement appears to be well founded.
This item is closed.
The report makes numerous recommendations for improvements in pl ant
performance, and two of them are specific to the use of steam flow mea-
surement.
The first was to write a new computer program for steam flow
that wou 1 d determine the average of the square roots of the fl ow dPs
rather than taking th'e square root of the average dP and would solve for
the thermodynamic state of the steam at the steam generator outlet. The
second was to install steam pressure taps 10 to 20 feet upstream of the
flow venturis in place of those 60 feet down stream. This was to enhance
the measurement of the thermodynamic state of the steam.
Discussion with
plant personnel revealed that the computer program had been written and
installed* as discussed above and that in addition it contained a correc-
tion for the steam pressure drop between the point of measurement and the
steam generator outlet.
The latter correction is in lieu of installing
new pressure taps.
The relative precision of the calibrations of the steam and feedwater flow
venturis was not addressed in this inspection.
No violations or deviations were identified.
4.
Review of Reactor Trips (6i707)
Three recent reactor trip reports were reviewed: Unit 1 on May 16, 1987
and February 16, 1988 and Unit 2 on May 16, 1988.
In all cases, the
average RCS temperature cooled significantly below the no-load setpoint of
547°F, the expected post-trip temperature.
The actual minimum
temperatures were 533, 539, and 525°F, respectively.
Administrative
procedure SUADM-0-02 ( Issued 11/5/87), Post Trip Review, requires in
section X.b(l) that RCS temperature, be analyzed and evaluated by compari-
son with expected values.
The reviews clearly identified the excessive
cool downs to be at variance with expected performance and some evalua-
tions of actions or modifications to limit cool down to 547°F were initi-
ated.
The reviews also referenced an October 25, 1985 study of Unit 1
trips on August 8, and September 9, 1985.
In that review, the first trip,
with steam dumps available, had excessive cool down, but the second
without steam dumps
did not cool below 540°F.
However, none of the
reviews specifically correlated the reduced temperatures with reductions
in shutdown margin or addressed shutdown margin in any way.
A member of
the plant staff did state that cool dpwn effects on shutdown margin had
4
been considered, but had not been significant enough to include in the
trip reports.
He also stated that, in the cycle design process, shutdown
margin was evaluated at 522°F, the minimum temperature for criticality
specified in TS 3.1.E.4.
Typical Surry cores have EOL shutdown margins of 3200 pcm. The TS required
shutdown margin is 1770 pcm.
Typical MTCs at EOL are about -35 pcm/°F.
Using these nominal values, the excess shutdown margin would be expended
by a post-trip cool down to 481°F, which is well below the experience to
date.
Currently the plant staff expect to receive an engineering evaluation of
means to limit cool down in time to implement corrective action during the
forthcoming Unit 2 refueling outage.
In December 1986, the licensee installed new transient recorders (GETARs)
on a trial basis and installed them permanently in May 1987.
The GETARs
monitor a total of 450 points on the two units.
Since the installation,
the post-trip review have become more quantitative and more detailed,
reflecting the improved quality of transient data.
No violations or deviations were identified.
5.
Reactor Coolant System Leakage Measurement (61728)
(Closed) Inspector Foll owup Item 280, 281/86-40-01: Revise procedures
1/2-PT-10 to incorporate all of the parameters required for surveillance.
The procedures were re-issued in October 1987, and now contain all of the
required parameters. This item is closed.
However, review of the genera ti on of some of the required parameters
revealed significant errors.
a.
The constant used for making a correction for a change in average RCS
temperature from the beginning to end of the RCS leakage measurement
is not based upon the change in the mass inventory in the solid RCS
volume.
In fact, no basis could be determined for the calculations
on the engineering work sheet used in generating the constant.
b.
The constant for equating changes in pressurizer level with changes
in RCS inventory was calculated on an engineering work sheet which
indicates the wrong distance between pressurizer level taps was used
in the calculation of monitored pressurizer volume.
There is no
correction in the calculation for the density difference in the
pressurizer and VCT fluids.
The two errors are not fully
compensating.
These errors have been identified as violation 280, 281/88-29-02: The
procedures for RCS leakage calculations contained errors in constants that
could lead to under estimating the leakage.
It was also noted that the
5
procedure does not require summing all sources of identified leakage for
comparison with the 10 gpm limit.
The inspector reviewed 19 examples of 1-PT-10 performed in January 1988,
and 5 of them had changes in either pressurizer level or temperature or
both, which required use of the constants discussed above.
Completed
copies of 2-PT-10 completed in May and June 1988 included four cases in
which one or both of the constants were used.
In none of the cases reviewed, would use of the correct constants have led
to a calculation of unidentifed leakage in excess of one gallon per
minute.
7.
Exit Interview
The inspection scope and findings were summarized on July 22, 1988, 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 materi-
als were provided to and reviewed by the inspector during this inspection,
but are not included in this report.
The licensee.was informed by tele-
phone on August 12, 1988 of the inspector followup item discussed in
paragraph 2.
8.
Acronyms and Initialisms Used in This Report
ARO
-
~~
EOL
-
GETAR-
-
MTC
-
NFOS -
-
OP
pcm -
ppmB -
-
SNSOC-
TS
-
a 11 rods out
boron concentration
differential pressure
end of life
General Electric transient recorder
isothermal temperature coefficient
moderator temperature coefficient
Nuclear Fuel Operations Subsection
Office of Nuclear Reactor Regulation
operating procedure
percent millirho (unit of reactivity)
parts per million boron
periodic test
Station Nuclear Safety and Operating Committee
Technical Specification
volume control tank