IR 05000327/1990029
| ML20058B260 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 10/05/1990 |
| From: | Belisle G, Burnett P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20058A846 | List: |
| References | |
| 50-327-90-29, 50-328-90-29, NUDOCS 9010300065 | |
| Download: ML20058B260 (8) | |
Text
_
.
.
. -
.
' '
[da moog 4 UNIT ED STAf t s
b NUCLEAR REGULATORY COMMissl0N o,'
f*
REGION il
'
g 101 MARitTTA STRitT.N.W.
- ATLANTA. GEORGI A 30323 i
- g*...+
J Report Nos.: 50-327/90-29,50-328/90-29
'
t Licensee: Tennessee Valley Authority
!
6N 38A Lookout Place
'
1101 Market Square Chattanooge, TN 37402-2801 Docket Nos.: 50-327 and 50-328 License Nos: DPR-77 and DPR-79
,
facility Name: Sequoyah Units 1 and 2
.
Inspection Conducted: August 27 - 31, 1990 Inspector:.
ewb-r6fd 7.7. Burnett
~
e Signed Approved by:
M
'
vi
/8 - 5 #
G. A. Belisle, CWief Date Signed Test Programs Section Engineering Branch Division of Reactor Safety SUMMARY Scope:
This routine,- announced inspection addressed the areas.of measurement of incore power distributions, calibration of nuclear instruments, measurement of thermal power, and followup _ of.licensce event reports.
Results:
l
,
Surveillance of incore power distribution and hot channel factors monitoring j
was satisfactory with respect to rescits and frequency for both units.
l
.
Power range nuclear instrument calibrations against thermal' power were being
,
perfonned with acceptable frequency and results for both units.
However,
'
'
. anticipat::ry calibrations of the power range instruments following refueling wt not being-performed correctly, because of an improperly written proce-dure.
The consequence of the improp.r procedure was operation in violation of limiting safety system settings as documented in licensee event report-i 50-327/90-011.
Use of an improper procedme has been identified as a viola-tion (paragraph 3.a)..
,
+
One violation for failure to follow -;,ocedure, with three examples, was
"
identified. -Two examples involved the procedure for the cross calibration of
'
incore and'excore nuclear instruments.- The th.rd example of failure to follow.
!
l procedure was ' observed in a related procedure, which was performed to deter-mine if a cross-calibration' procedure is required (paragraph 3 b). -
,
y
-
9010300065 901011
/il t
'
ADOCK 0500g ? Q DR y
{% ;,
,
.
.
c
,
,
The procedure for performing the reactor heat balance calculation was per-
,
,
fvrmed properly in the examples reviewed.
However, the procedure contains an Appendix 0, " Normalized Primary Side Calorimetric," which allows a secondary side flow venturi fouling factor to be calculated from primary side flow.
The procedure did not indicate if any increased propagation of error in the heat balance had been considered.
This issue was identified as an unresolved item at the exit interview.
Subsequently, the licensee was able to produce an evaluation by Westinghouse of the method of analysis and error propagation.
That report was reviewed in the regional office, but the issued remains unresolved pending further discussions with the licensee (paragraph 4).
As a result of a change in the configuration of the temperature measuring detectors in the Unit I hot legs, higher than expected temperatures were being measured in the hot legs, and the coolant temperature rise in the vessel was also higher than expected.
This could have been a result of reduced flow through the reactor. vessel, but other observables argued against that conclu-sion.
The TVA conclusion was that the flow in the hot legs was not well mixed at the points of temperature measurement.
The conclusion of this inspection, based upon review of licensee measurements, is that neither the true mixed mean hot leg temperature nor the vessel temperature rise are excessive and that reactor thermal power has been correctly measured and is within licensee limits.
However, it is apparent that neither the present nor the past method of measuring hot leg temperature provide a reliable measu ement of the mixed-mean hot-leg temperature (paragraph 5).
The reactor engineering unit has an authorized strength of one supervisor and three engineers.
At the time of the inspection, the supervisor had no previ-oils experience as a reactor engineer or in supervising engineers.
Further-more, the supervisor had no permanently ass'.gned reactor engineers in the unit.
Although the staff is expected to reich full-strength soon, the loss of experience is of concern (paragraph 6).
l l
l l
. - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _
_ _ _ _ _ _ _ _ _ _ _ - _ _ _
__
_ _ _ _ _ _ -
_ _ _ _ _ _ _ _ _ _.
j
-.
.
.
.
-
-
..
!
,
!
!
REPORT DETAILS
,
1.
Persons Contacted i
Licensee Employees
- R. Beecken, Maintenance Manager M. Cooper, Site Licensing Manager
- R. Fortenberry, Program Manager for Plant Manager W. Lagergren, Operations Manager J. Nieri. Quality Assurance Audit Specialist
-
- J. Proffitt, Compliance Licensing Manager
- H. Rogers, Technical Support Program Manager
- M. Skarzinski, Reactor Engineering Manager
- J. Smith, Regulatory Licensing Manager
- R. Thompson, Licensing Engineer P. Trudel Project Engineer, Nuclear Engineering C. Vondra, Plant Manager S. Wilburn, Balance of Plant Section Supervisor C. Wood, Quality Assurance Supervisor Other licensee employees contacted included engineers and office personnel.
NRC Resident Inspectors i
P. E. Harmon, Senior Resident Inspector
- D. P.' Loveless, Resident Inspector
.
'
- S. M. Shaeffer, Resident Inspector
.
NRC/ Region II Staff
- W. Little, Chief, TVA Projects Section 1
- Attended exit interview on August 31, 1990 Acronyms and initialisms used throughout this report are listed in the final paragraph.
2.
Core Power Distribution Monitoring (61702)
a,
' Unit 1 Procedure 1-SI-NXX-004.0 L(Revision 0),' Hot Channel Factor Determina-r tion, has = been performed six times in the current fuel cycle, at
! power levels ranging from 28% to 100% RTP. 'The measured values of
- - -.
-.
.
.
,
.
.
.
F and F satisfied TS 3.2.2 and TS 3.2.3, respectively, in all
'
dH C ses.
0-SI-NXX-000-011.0 (Revision 0), Movable Detector Determination of Quadrant Power Tilt Ratio, was performed once at 99.8% RTP in re-sponse to one inoperable PRNI (N-42).
The surveillance requirements of TS 4.2.4.2 were satisfied.
b.
Unit 2 Cycle 4 surveillances for 1990 were reviewed by the inspector.51-126 (Revision 23), Hot Channel Factors Determination, was per-formed five times, each at a nominal full power.
Surveillance activities were then continued using procedure 2-SI-NXX-004.0 (Revi-sion 0)
Hot Channel Factor Determination, which has been performed thrice.
The measured values of F satisfied TS 3.2.2 and TS 3.2.3, respectively, in all casks.and F dH The frequency of performance satisfied TS 4.2.2.2 and TS 4.2.3.2 for both units for the time periods reviewed.
No violations or deviations were identified.
3.
Nuclear Instrument Calibrations (61705)
a.
Power Calibrations
!
Review of completed copies of 51-78 (Revision 10), Power Range l
Neutron Flux channel Calibrations by Heat Balance Comparison, for the month of August 1990, confirmed that the power range nuclear instrument calibrations against thermal power were being performed with acceptable frequency and results for both units.
However, anticipatory calibrations of the power range instruments following refueling were not being performed correctly.
TI-81, Prestartup Nuclear Instrument System Calibration Following Core Load, was issued in 1984 to account for expected reductions in neutron flux leakage to the excore PRNIs and IRNIs as a result of installing low-leakage cores.
The intent of the procedure was to implement guidance provided by Westinghouse in their letter, 84TV*-G-005, January 6, 1984.
That letter presented a method to calculate the 800, zero-offset, ion-chamber currents by multiplying the zero-offset currents measured in the last incore-excore calibra-tion of the previous cycle by a ratio, R.
R was supposed to be determined by summing the predicted, BOC relative powers of fuel assemblies adjacent to the chambers and dividing by the sum of the
.rolative powers of the same fuel asi,en:bly locations during the last incore-excore calibration.
Instead, the procedure defined the denominator as the sum of the relative fuel assembly powers predicted at BOC for the previous cycle.
This error could have been
.
.
.
.
'
.
.
exposed by dimensional analysis on the part of either the procedure writer o' the reviewer, but was not.
Subsequently, Westin0 ouse letter, 85TV*-G-090, September 27, 1985, h
changed thi number of fuel assemblies to be used to calculate R and changed th) weighting factors of the assemblies.
Westinghouse letter 88T\\ *-G-0021, March 30,1988, again described the process as using currents from the previous cycle's incore-excore measurement, the measured X-Y power distribution from the calibration and the predicted X-Y power distribution at HFP for the next cycle.
These letters did not precipiti.te a full review of the procedure.
TI-81 was rep 1; rd by u ilt specific procedures,1/2-PI-NXX-092-001.0, but the review Jid not expose the basic error in the calculation of zero-offset currents et BOC.
Consequently, when the procedure was performed for the sta" tup of Unit 1 for cycle 5, the predicted full power currents were calculated to be 20 to 30% higher than a correct calculation would have yielded.
As reported in LER 50-327/90-011, Unit I was critical with the power range trip setpoints in excess of the LSSS for over 72 hout's until a heat balance revealed the error.
This failure to provide a correct procedure for prestartup calibra-tion of the PRNIs has been identified as a violation (VIO 50-327 and 328/90-29-01).
b.
Incore-Excore Nuclear Instrument Calibrations 0-PI-NXX-092-001.0 (Revision 0), Incore-Excore Detector Calibration, implies, by example, that three full-core flux maps and five quar-ter-core flux maps be performed over the approximately twenty-four hour period of the induced xenon transient.
The procedure does give the test director the latitude to vary the number of flux maps.
Recent examples of the completed procedure indicate that this e
discretion has been abused.
Only two flux maps, both full-core, were obtained in a period of less than twelve hours.
Two maps are insufficient to confirm a linear relationship between incore axial offset and the excore nuclear instrument response.
Licensee person-nel confirmed that recent practice on both units has been to perform only two full-core maps over a time span of ten to twelve hours.
,
This has been identified as a violation for failure to follow
"
procedure (VIO 50-327 and 328/90-29-02).
0-PI-NXX-092-001.0 also requires that the calibration be performed at or above 70% RTP.
The procedure performed for Unit 1 in the period from June 7,1990 to July 26, 1990, was performed at 30% RTP.
A calibration above 70% RTP was not performed until August 4,1990, and that yielded considerably different results in the zero-offset currents and slope of current versus axial flux difference for all l
eight ion chambers.
The differences can be ascribed to both the l
differences in power level and the use of two point correlations.
Failure to perform that required calibration at the proper power l
..
.
.
-_
_ _ _ - _ _
-
,
.
-
,
.
.
level has been identified as a second example of the violation for failure to follow procedure.
A third example of failure to follow procedure was observed in a related procedure 0-SI-NXX-092-079.0, Power Range Monitor Channel Calibration by Incore-Excore Axial Imbalance Comparison, which is performed to determine if a cross-calibration procedure is required.
In two of seven completed procedures reviewed, required independent verification steps were not signed off for procedures performed on June 8, 1990, and June 15-20, 1990.
No additional violations or deviations were identified.
4.
Core Thermal Power Calculations (61706)
51-78, the procedure for performing the reactor heat balance calculation was performed properly in the examples reviewed.
However, the procedure contains an Appendix D, " Normalized Primary Side Calorimetric," which allows a secondary side flow venturi fouling factor to be calculated from primary side flow.
Since primary side flow was measured from the secon-dary side flow, by equating heat balances, there appeared to be some circular reasoning at work.
The procedure did not indicate if any in-creased propagation of error in the heat balance had been considered.
Accidents evaluated in The Final Safety Analysis Report consider a two percent uncertainty in thermal power measurements.
A greater uncertainty would invalidate the safety analyses.
This issue was identified as an unresolved item at the exit interview.
Subsequently, the licensee was able to produce an evaluation by Westinghouse of the error propagation resulting from using a primary side heat balance for thermal power analy-sis.
That report, " EVALUATION OF PERFORMANCE DATA AND METHODS OF RECOVERING THE LOSS IN STATION ELECTRICAL OUTPUT AT SEQV0YAH UNIT 1," was reviewed in the regional office.
Although the report appears to be thorough and follows familiar methodology, no conclusion could be reached without further discussions with the licensee and their contractor on specific points of the error propagation.
Also, certain assumptions made in the report may not be valid in light of recent observations of coolant streaming in the hot leg (see paragraph 5 below).
The assumptions are that changes in the temperature streaming pattern have a minor effect on the measured hot leg temperature and that the observed core differential temperature does not change during the cycle.
In view of these concerns and questions, use of Appendix D to SI-78 will remain unresolved pending further evaluation by the licensee and further discussions with them (URN 50-327 and 328/90-29-03).
No violations or deviations were identified.
,
.
.
.
-
.
5.
Hot Leg Streaming (61705, 61706)
]
,
By letter dated June 18, 1990, TVA informed NRC that, as a result of a change in the configuration of the temperature measuring detectors in the Unit 1 bot legs, higher than expected temperatures were being measured in the hot legs and that the coolant temperature rise in the vessel was also j
higher than expected.
This could have been a result of reduced flow through the reactor vessel, but other observables argued against that conclusion.
The TVA conclusion was that the flow in the hot legs was not well-mixed at the point of temperature measurement, and that the poor mixing was exacerbated by the low leakage core design, which preferen-
tially streamed the cooler water to the bottom of the cold legs.
The conclusion of this inspection, based upon review of licensee measure-ments, is that neither the true mixed mean hot leg temperature nor the vessel temperature rise are excessive and that reactor thermal power has been correctly measured and is within licensee limits.
However, the theory that the poor mixing in the hot legs is the result of the reduced power level of peripheral fuel assemblics has not been proved and can not be tested until the incore power distribution changes significantly, as expected, later in the cycle.
Based upon the findings of this inspection and discussions held with Westinghouse and NRR on September 11, 1990, in Rockville, Maryland, it appears that neither the manifold system nor the direct immersion system of hot leg temperature measurement have the inherent capability of measuring a mixed mean hot leg temperature.
Therefore, any measurement of thermal power or surveillance of loop flow based upon the core differential temperature is suspect.
No violations or deviations were identified.
,
6.
Staffing The reactor engineering unit has an authorized strength of one supervisor and three engineers.
At the time of the inspection, the supervisor had no previous experience as a reactor engineer or in supervising engineers.
Furthermore, the supervisor had no permanently assigned reactor engineers in the unit.
That situation was scheduled to change as of September 4, l
1990; when a former STA was to be permanently assigned to the unit and an experienced reactor engineer from another facility was to be detailed on a long term assignment.
Additionally, vacancy announcements were being i
'
prepared for two positions.
The unit does have the benefit of the counsel of a previous reactor engineering unit supervisor.
The loss of the previous unit supervisor and a staff of experienced reactor engineers
'
occurred'over a period of a few months.
Although the violations discussed above are the errors of the previous staff, the wholesale dispersal of the staf f and the concomitant loss. of facility specific experience is of concern.
No violations or deviations were identified.
l
_
~
-
..
..
,
.
.
'
I
1
-
l 7.
Reactivity Balance (61707)
Performance of procedure SI-120 (kevision 6), Overall Reactivity Balance, was reviewed for cycle 4 on Unit 1 Both the frequency of perf ormance and the individual result. were satisfactory.
,
J No violations or deviations were iden.ified.
8.
Exit Interview (30703)
The inspection scope and findings were summarized on August 31, 1990 with those persons indicated in paragraph 1 sbove.
The inspector described the areas inspected and discussed in detail the ~ inspection findings
listed below.
Dissenting comments were not received from the licensee.
Proprietary information was reviewed in the course of the inspection, but i
is not contained in this report.
!
Violation 50-327 and 328/90-29-01:
Failure to provide a correct procedure for prestartup calibration of the PRNIs paragraph 3.a.
Violation 50-327 and 328/90-29-02:
Failure to follow procedure - insuffi-cient flux map, power level too low, and verifications steps not signed off paragraph 3.b.
Unresolved item 50-327 and 328/90-29-03:
Further review and evaluation
,
of the use of the primary side heet balance should be performed by the licensee prior to further discussions on the subject.
Acronyms and Initialisms Used in This Report
>
ARO all rods out BOC beginning of cycle dP differential pressure E0C end of cycle F
enthalpy rise hot channel factor dH'
F heat flux hot channel factor HEP hot full. power IRNI intermediate range nuclear instrument
'
i LER license. event report
'
- limiting safety system setting
!
NRR'
USNRC Office of Nuclear Reactor Regulation PRNI power range nuclear instrument-RCS reactor. coolant system RTD resistance temperature device
.
RTP rated thermal power
'
,
L SI surveillance instruction
'
STA'
shift: technical advisor
,
TI technical-instruction.
TS Technical Specification UNR unresolved VIO'
-violation
'
.