IR 05000321/1988023
| ML20207J268 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/15/1988 |
| From: | Burnett P, Jape F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20207J259 | List: |
| References | |
| 50-321-88-23, 50-366-88-23, NUDOCS 8808300331 | |
| Download: ML20207J268 (8) | |
Text
d
p Ceco UNITED STATES p
.c
NUCLEAR REGULATORY COMMISSION
[o'
'
REGION 11
,
g, j
101 MARIETTA STREET.N.W.
't ATLANTA, GEORGI A 30323
%...../
Report Nos.:
50-321/88-23 and 50-366/88-23 Licensee:
Georgia Power Company P.-0. Box 4545 Atlanta, GA 30302 Docket Nos.:
50-321 and 50-366 License Nos.:
OPR-57 and NPF-5 Facility Name:
Hatch 1 and 2 Inspection Conduc ed:
ust 1 to 5, 1988 P.7. Burnett
-
/f m/< j 8 [
Inspector:
,
Datef51gned Approved by:
dx4 M
Ph4Y F. Jape, Section Chief
//
Date 51gned Engineering Branch Division of Reactor Safety SUMMARY Scope:
This routine, unannounced inspection addressed the review of completed post-refueling startup tests, thermal limits monitoring, thermal power monitoring, and nuclear instrument calibrations.
Results:
No violations or deviations were identified.
One licensee commitment was obtained
.
Inspector followup item 321,366/88-23-01: Resolve, prior to the start of the next cycle, the differences between the startup testing procedure and the program described in the GPC letter of December 29, 1987, which included a specific comparison of the actual and predicted cold critical eigenvalues and recording the APRM calibra-tions performed as part of the procedure paragraph 2.c.
It was noted that the onsite rt: actor engineering staff has been considerably augmented in numbers and experience level of personnel.
i i
esos3Oo331 seos17 DR ADOCK 0500
..
.
__
.
.
'
.
.
.
.
REPORT DETAILS 1.
. Persons Contacted Licensee Employees
- S. J. Bethay
.J.A.Betsili,NuclearSafetyandComplianceSupervisor
Superintendent of Operations Support
- S. F. Curtis, Senior Shift Technical Advisor
- 0. M. Fraser, Quality Assurance Site Manager R. L. Keck, Superintendent of Reactor Systems
- G. W. Neely, Reactor Engineering Supervisor
- T. R. Powers, Manager of Engineering Support
- D. Read, Plant Su) port Manager
- L. Sumner, Plant rianager
- S. B. Tipps, Nuclear Safety and Compliance Manager Other licensee employees contacted included engineers and office personnel.
,
NRC Resident Inspectors
- J. Menning, Senior Resident Inspector
- R. Musser, Resident Inspector
- Attended exit interview Acronyms and initialisms used throughout this report are listed in the last paragraph.
2.
Post Refueling Startup Tests (72700, 61705, 61707)
The licensee's standard post-refueling startup test sequence is described in their letter to the NRC (SL-365c/1827C/X7GJ17-H600) dated December 29, 1987.
The sequence includes:
a.
Core Verification b.
Control Rod Drive Friction Testing
)
c.
Control Rod Drive Timing
,
d.
Full Core Shutdown Margin Demonstration e.
Cold Critical Eigenvalue Comparison f.
Whole Core LPRM Calibration g.
APRM Calibration h.
Control Rod Scram Time Testing i.
TIP Reproducibility and Asymmetry Calculations j.
Reactivity Anomaly Calculation
-
-
- -
-
.
-
.
.
'
..-
.
.
.
-Additionally, that letter requested relief from the requirement to issue a startup report when four or fewer new-design fuel assemblies are added to the core.
The NRC agreed to that request in a letter dated March 3, 1988.
Hence, no startup report was issued for the addition of four ANF LTAs in Unit 2.
In.the discussion below the procedures identified and/or the completed data packages resulting from performance of the procedures were reviewed by the inspector.
a.
Unit 1 Startup for Cycle 11 during June 1987 The startup was controlled by 3rocedure 42FH-ENG-024-1, Startup Testing.
The correct core load ng was verified by procedure 42FH-ERP-014-05, which was completed on June 8,1987.
All LPRMs wore pre-calibrated using procedure 57CP-C51-007-1, which was completed on June 24, 1987.
Precritical control rod surveillances included procedures 42FH-ENG-21-1, Friction Tests, completed on June 9,1987 for 12 rods and 42FH-ENG-022-1, Drive Time Testing, on all rods, completed on June 18, 1987.
Rods not meeting the acceptance criteria were adjusted and retested.
Those that continued to show minor deviations from the criteria were subjected to engineering review and, based upon that review, accepted for service.
Other precritical activities included verifying the correct MCPR and LHGR limits and the MAPLHGR curves had been installed in the plant.
computer.
At startup, procedure 42FH-ENG-019-1, Shutdown Margin, was completed with satisfactory result.
The inspector independently calculated the shutdown margin using the data obtained in the procedure and obtained the same value.
It was noted there was no documented evidence of performing item E. Cold Critical Eigenvalue Comparison, which is listed above.
However, it could be inferred from the SDM calculation that the calculated and actual critical configurations agreed within less than 0.1% dK/K.
As power escalation progressed, all LPRMs, except 2 of 124 were observed to respond to control rod motion.
The LPRMs were recali-brated by performing the 00-1 program at nominal power levels of 25, 49, 73, and 100% RTP.
Scram time testing of all control rods was completed on June 27, 1987 prior to exceeding 40% RTP using 42FH-ENG-020-1.
The Reactivity Anomaly Calculation, 42CC-ERP-007-1, was completed on July 7,1987 at 9 EFPD..The agreement between rated and predicted rod notches inserted in the core was 196 notches out of an allowance of 375 notches, the reactivity equivalent of 1% dK#.
>
- - - - - - -
.
.
.
. -
.
.
.
-
.
.
.
TIP reproducibility and asymmetry calculations were completed using procedure 42FH-ENG-023-1 on July 15, 1987.
The controlling procedure, 42FH-ENG-024-1, does not address APRM calibrations, however they were performed under the routine surveil-lance program.
b.
-Unit 2 Startup for Cycle 8, March to June 1988 The startup was controlled by procedure 42FH-ENG-024-2, Startup Testing.
The correct core loading was verified by procedure 42FHERP-014-05, which was completed on March 3,1988.
All LPRMS were pre-calibrated using procedure 57CP-C51-007-2, which was completed on March 15, 1988.
Precritical control rod surveillances included procedures 42IT-C11-001-05, Friction Tests co 16, 1988 and 42FH-ENG-022-2, Drive Time Testing,mpleted on March completed on March 15, 1988.
Rods not meeting the acceptance criteria were adjusted and retested.
Those that continued to show minor deviations from the criteria were subjected to engineering review and, based upon that review, accepted for service.
Other precritical activities included verifying the correct MCPR and LHGR limits and the MAPLHGR curves had been installed in the plant computer.
At startup, procedure 42FH-ENG-019-25, Shutdown Margin, was completed with satisfactory result.
The 800 SDM of 1.78% dK/K was comfortably in excess of the required value of 0.38% dK/K.
The inspector independently calculated the shutdown margin using the data obtained in the procedure and obtained the same value.
It was noted there was no documented evidence of performing item E. Cold Critical Eigenvalue Comparison, which is listed above.
However, it could be inferred from the SDM calculation that the calculated and actual critical configurations agreed within less than 0.1% dK/K.
As power escalation progressed, all LPRMs, except 5 of 124 were observed to respond to control rod motion.
The LPRMs were recalibrated by performing the 00-1 program at nominal power levels of'27, 49, 75, and 100% RTP.
Scram time testing of Unit 2 control rods was completed on April 18, 1988 using procedure 42SV-C11-001-25, Control Rod Scram Testing,ia which was aoproved for validation use only.
All acceptance criter as defined in Technical Specification 3/4.1.3 were satisfied.
The Reactivity Anomaly Calculation, 42CC-ERP-007-2, was completed on March 28, 1988 at 3 EFPD.
The agreement between rated anrl predicted rod notches inserted in the core was 24 notches out of an allowance of ~400 notches, the reactivity equivalent of 1% dK/K.
.
.
.
'
-
.
.
TIPz reproducibility and asymmetry calculations were completed using procedure 42FH-ENG-023-2 on June 21, 1988.
The controlling procedure, 42FH-ENG-024-2,. does_ not address APRM-calibrations, however, they,were performed under the routine surveil-lance program.
.c.
Improvements in Procedures In performing independent calculations of shutdown margin and reac-tivity anomaly, the ins)ector noted the procedures in use were difficult to follow in logic, ions, and did not preserve all of the did not require entry-of all of the data required for the calculat calculations performed.
Discussions with licensee personnel revealed those deficiencies had been identified and that extensive procedure revisions were nearly completed.
The Shutdown Margin Demonstration Procedure has also been renumbered to 42CC-ERP-101-05.
In reviewing the draft, the-inspector noted that the R term is described in words as in the Technical Specifications as the maximum positive difference in core reactivity at any time in the cycle from the reactivity at 800.
However, the actual calcula-tion of R is the difference between the cold SOM at 800 and.the minimum cold SDM at any time in the cycle.
Based upon case studies, the licensee is confident the the equatinn is conservative with respect to the definition.
The revision of the Reactivity Anomaly Calculation has reached the verification phase of procedure development and appears to address of the concerns identified by.the inspector.
The licensee made a commitment to resolve the differences between the startup testir.g procedures and.the program described in the GPC letter of December 29, 1987 which included a specific comparison of theactualandpredictedcoldcriticaleigenvaluesandrecordingthe APRM calibrations performed 'as part of the procedure (inspector followup item 321,366/88-23-01).
d.
Other Documents Reviewed
The following documents important to the startup testing process were reviewed:
(1) Cycle Management Report for Unit 1, Cycle 11; (2) Cycle Management Report for Unit 2, Cycle 8; (3) Supplemental Reload Licensing Submittal for Unit 1, Reload 10; and, (4) Supplemental Reload Licensing Submittal for Unit 2, Reload.-
.
.
.
.
.
,
,
'
.
e.
Other' Reactor Engineering Surveillance Activities Surveillance of the reactivity anomaly is required every 31 EFPD.
A review of the completed copies of 42CC-ERP-007-1/2 confirmed that the procedure had been performed with acceptable frequency and results
.for Unit 1-from December 1987 to preser.t 'and for Unit 2 during the present cycle.
The inspector also observed an engineer perform the calculation for Unit 2 for August 1988.
The requirements for surveillance of control rod exposure are addressed in IEB 79-26.
The licensee performs the surveillance using procedure 42CC-ERP-005-1/2 (Revision 2).
The surveillance instrument is the digital computer program BLADELIM.
Fluence to the blades is continuously calculated and updated by the plant computer.
BLADELIM, which operates on the VAX com] uter, accesses the plant computer exposure data base.
Exposure 'imits are a function of blade design.
BLADELIM calculates the percentage of exposure lia.it for each blade as a function of fluence and type and reports by location in the core the blade having the highest percentage of allowable exposure and all
.
J blades over 80% of limiting exposure.
The program was originally written to accommodate the three types of blades manufactured by General Electric Company.
Four new blades from ASEA-ATOM were installed in Unit 2 for Cycle 8.
The inspector confirmed the program had been revised to account for the fourth blade type.
The inspector confirmed that the surveillance procedure was being performed with monthly frequency for the current operating cycles for both units.
In Unit 1, one blade is at 82.3% of its limiting exposure and three additforal blades exceed 80% of their limits.
In Unit 2, the limiting blade is at 53% of allowed exposure.
No violations or deviations were faentified.
3.
Thermal Limits Monitoring (61702)
Procedures 42CC-ERP-002-1/25, APLHGR, LHGR, and MCPR Calculation, are performed daily with the plant and VAX computers operational by transcrib-ing P-1 output of the monitored parameters CMFLPD, CMAPR, and CMFCP to a data sheet, and confirming by inspection that those parameters are within the bounds that assure compliance with the corresponding thermal limits.
The procedure data packages for the past two weeks were reviewed for both units and found to have acceptable frequency and results.
The data sheets also confirmed that the GAFs for the APRMs were within the acceptable range.
The inspector also confirmed that the procedures contain acceptable manual methods for performing the surveillance in the absence of the computers.
No violations or deviations were identified.
I
'l O
_ _ _.
_
__
..
..
d
..
.
-
.
4.
Calibration of Nuclear Instruments (61705)
,
Procedure 42CC-ERP-003-0S (Revision 1) LPRM Calibration Current Calcula-tion, is performed on either 1000 EFPH intervals or when the average GAF-exceeds 1.25.
The calibration is based upon a flux distribution measured used TIPS.
Review of plant records confirmed this surveillance had been performed with acceptable frequency and results for both units during the period November 1987 to June 1988.
Procedure 42CC-ERP-006-15 (Revision 1), APRM Adjustment-to Thermal Power, is performed whenever the APRMs are not within 1.9% of indicated thermal-
)ower.
Completed procedures for a recent power reduction and recovery for Jnit I were reviewed and found acceptable.
-
No violations or deviations were identified.
5.
Core Thermal Power Evaluation (61706)
A manual method of calculating core thermal power is provided in proce-dures 42CC-ERP-001-1/2S (Revision 2), Core Heat Balance - Power Range.
The procedure uses instrument readings available from the main control board.
Review of the equations in the procedure confirmed it was an acceptable method for determining core thermal The procedure is performed fortnightly and must agree within 2% power.RTP of the 00-3 (plant computer) result or the reason for the discrepancy must be investigated.
The licensee has installed a program on the VAX computer to continuously monitor reactor power for the current level, average level over the aast hour, and average level over the past eight hours.
For the same per ods, the program also keeps a running sum of the time spent over limits of 100, 100.5,101,101.5, and 102% of RTP, and generates alarms if specified times are exceeded.
The inspector observed the program in action on one of the remote computer terminals.
Procedure 42CC-ERP-009-05, Maintaining Rated Thermal Power, captures the power monitor program output daily.
Review of the data sheets for the period July 23 to August 1,1988 for both units showed that the power levels were rarely over 100.5% and never over 101% RTP.
No violations or deviations were identified.
6.
Exit Interview The inspection scope and findings were summarized on August 5, 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 materials were provided to and reviewed by the inspector during this
inspection, but are not included in this report.
Management confirmed the commitment discussed in inspector followup item 321,366/88-23-01.
.
- - - - - - - - - -
-
,
L..
.
^ *
=
-
.
.
.
7.
Acronyms and Initialisms Used in This Report ANF Advanced Nuclear Fuels (Formerly Exxon Nuclear)
-
APLHGR Average Planar Linear Heat Generation Rate
-
APRM Average Power Range Monitor ( An Array of LPRMs)
-
BOC Beginning of Cycle
-
CMAPR Core Maximum Average Planar Linear Heat Generation Rate
-
CMFCP Core Maximum Fraction of Critical Power
-
CHFLPD Core Maximum Fraction of Limiting Power Density
-
dK/K Reactivity - dimensionless
-
EFPD Effective Full Power Day
-
EFPH Effective Full Power Hour
-
GAF GainAdjustmentFactor
-
IEB Inspection and Enforcement Bulletin
-
LHGR Linear Heat Generation Rate
-
LPRM Local Power Range Monitor
-
-
MAPLHGR Maximum Average Planar Linear Heat Generation Rate
-
MCPR Minimum Critical Power Ratio
-
OD-n On-Demand plant computer program n
-
RTP Rated Thermal Power
-
SDM Shutdownliargin
-
TIP Traveling Incore Probe
-
VAX Minicomputer manefactured by Digital Equipment Corporation
-
&