IR 05000373/1986036
| ML20211M850 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 12/10/1986 |
| From: | Azab B, Rescheske P, Ring M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20211M807 | List: |
| References | |
| 50-373-86-36, NUDOCS 8612180065 | |
| Download: ML20211M850 (9) | |
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q U. S. NUCLEAR REGULATORY COMMISSION
REGION III
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Report No. 50-373/86036(DRS)
Docket No. 50-373 License No. NPF-11 s
Licensee: Commonwealth Edison Company P. 0.~ Box 767 Chictgo, IL 60690 Facility Name: LaSalle County Station, Unit 1 Inspection At: LaSalle Site, Marseilles, Illinois Inspection Conducted:
September 16 through November 20, 1986 l
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P.R.Reschesfe
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Inspectors:
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B. A.
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T/MN Approved By:
M.A. Ring,Chfef
/7/[qhs Test Programs Section Date Inspection Summary No. 50-373, on_ Sep(bYS))tember 16 through_ Novemb_er_ 20,19_86_JReport l
Inspection
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i~eis~ Tn~s~p[ected :
86036
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Routine unannounced, safety inspection of surveillance of core power dTstribution limits (61702), calibration of the local power range monitoring system (61705), APRM calibration and core thermal power evaluation (61705and61706),shutdownmargindeterminationandreactivitychecks (61707), control rod drive performance testing (72700), and additional startup testing (72700).
Results: No violations or deviations were identified.
86121c30065 861210 PDR ADOCK 05000373 G
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DETAILS 1.
Persons Contacted
- G. J. Diederich, Station Manager
- R. D. Bishop, Services Superintendent
- J. C. Renwick, Production Superintendent
- M. H. Richter, Assistant Technical Staff Supervisor
- R. W. Stobert, Station Quality Assurance Superintendent The inspectors also interviewed other licensee employees including members of the technical and operating staff.
- Denotes persons attending the exit meeting of November 20, 1986.
2.
Surveillance of Core Power Distribution Limits The inspectors reviewed a number of the licensee's surveillances of thermal limits and core power distribution limits including P-1 computer printouts. A sample of " Operators Shiftly Surveillances," LOS-AA-S1 (Revision 18), Attachment A, performed between September 26 and October 27, 1986, was examined.
The surveillances were verified to be per# armed with the proper frequency, which is at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, l
per Technical Specification 3/4.2. The surveillances are performed more l
frequently following a power increase or when the plant is operating near l
a thermal limit.
The inspectors also reviewed a sample of " Nuclear
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Engineer's Daily Surveillances," LTS-1200-4 (Revision 4), from approximately the same time period; and verified the data was recorded properly and at the recommended frequency of about once per day.
The inspectors determined that the following core power distribution limits satisfied the acceptance criteria and that proper corrective actions were initiated when necessary:
a.
Core Thermal Power (CTP) is not to exceed the rated thermal power of 3323 MWt as specified in the Unit 1 Technical Specifications.
b.
CMAPR (core maximum MAPRAT) is less than or equal to 0.98, where l
MAPRAT is the ratio of the Maximum Average Planar Linear Heat I
Generation Rate (MAPLHGR) to the MAPLHGR limit. This insures that Technical Specification 3.2.1 is satisfied, c.
Core Maxfmum Fraction of Limiting Power Density (CMFLPD) is less than or equ61 to 0.98, which insures that the Linear Heat Generation Rate (LHGR) does not exceed 13.4 kw/ft as required by Technical Specification 3.2.4.
d.
Fraction of Rated Thermal Power (FRTP) is greater than CMFLPD which insures compliance with Technical Specification 3.2.2.
When CMFLPD exceeded FRTP, appropriate actions were initiated by the licensee.
For example, at lower power levels, the APRM gains were set conservatively high.
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e.
The Minimum Critical Power Ratio (MCPR) is greater than or equal to the MCPR limit, to insure cladding integrity as specified in Technical Specifications 4.2.3'.
No violations or deviations were identified.
3.
Calibration of t_he Local P_ower_ _ Range Monitoring Sy_s_ tem The inspectors reviewed information related to two separate calibrations of the local power range monitoring (LPRM) system and the traversing incore probe (TIP) system. The licensee completed the calibrations i
during startup testing on September 20, 1986, at 50% CTP, and on October 9, 1986, at 88% CTP. The following procedures were reviewed and found to l
be adequate for performing the required adjustments and calibrations:
LTP-1600-7, "Whole Core LPRM Calibration," Revision 4.
LIS-NR-111, " Unit 1 LPRM Flux Amplifier Gain Adjustment," Revision 1.
LTP-1600-6, "TIP System Calibration," Revision 4.
The inspectors reviewed the associated data sheets and process computer printouts (i.e., P-1 and 00-10 Option 7 edits), and verified that the acceptance criteria were met and the results were properly signed-off and reviewed. The inspectors also verified that the as-found and as-left data was properly recorded, the calculations were correct, and the APRMs I
and LPRMs were bypassed and returned to service as required.
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No violations or deviations were identified.
4.
APRM Calibrations and Core Thermal Power Evaluation LTP-1600-8, " Nuclear Engineer's Method For APRM Calibration," Revision 2, was used by the licensee to calibrate (adjust) the APRMs at both low and high powers during startup. Three different methods are used, depending on the reactor power level and the accuracy of the process computer. The inspectors reviewed the procedure and found the following methods and results to be acceptable:
a.
Step F.1 outlines the procedure to be used when the APRMs first begin to have on-scale readings following a refueling outage. The APRMs are adjusted to either the percerit bypass valves open or the average power from the IRMs (intermediate range monitors), whichever is most conservative. The inspectors reviewed the calculations performed on September 18, 1986, at power less than 5% CTP. No adjustments were necessary due to the conservatism of the as-found APRM readings.
b.
Step F.2 describes the method used to calibrate the APRMs at powers high enough to obtain an accurate heat balance.
The licensee determines CTP using LTP-1600-10, Attachment A, " Heat
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Balance Calculation Sheet for Unit 1."
A process computer heat I
balance edit OD-3 Option 2 is obtained for comparison, although the computer was not accepted for use until Step F.3.
The inspectors reviewed the heat balance calculation performed by the licensee on September 22, 1986, at 24% CTP, and the subsequent APRM gain adjustments per LIS-NR-109, " Unit 1 APRM Gain Adjustment."
c.
Step F.3 is used following verification of the accuracy of the process computer heat balance 0D-3 Option 2.
The licensee performed an APRM adjustment on September 27, 1986, at 51% CTP, and a check on the APRMs (no adjustment necessary) on October 8, 1986, at 89% CTP.
At power levels greater than 25% CTP, Technical Specification 4.3.1.1 sets limits on the allowable APRM gain adjustment factor (AGAF).
f The inspectors reviewed the as-left 0D-3 edits and verified that the acceptance criteria was met.
i No violations or deviations were identified.
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5.
_ Shutdown Margin De_ termination and Re_ activity Checks l
a.
LTS-1100-14, " Shutdown Margin Subcritical Demonstration," Revision 0, l
was performed by the licensee prior to startup to demonstrate that core loading was limited such that the required shutdown margin was met. The licensee used the two-rod diagonally adjacent subcritical method to demonstrate that Technical Specification 3/4.1.1. was satisfied. The inspectors reviewed the procedure and the test results from the test performed on September 17, 1986; and verified that the
methodology was technically correct and that the reactor remained subcritical with a sufficient shutdown margin, with the analytically determined strongest rod fully withdrawn and a diagonally adjacent rod withdrawn to position 24.
In addition, the vendor-supplied data (including rod worths and the moderator temperature correction),
the control rod sequence, and SRM (source range monitor) response were also reviewed during this inspection.
b.
LTS-1100-1, " Shutdown Margin Test," Revision 1, was performed by the licensee during the normal in-sequence pull to critical on September 17, 1986. The inspectors reviewed the procedure, the test results, and the control rod sequence sheets; and verified that the methodology was technically correct and that the shutdown margin determined was within the required limits. The following vendor-supplied information required to perform the shutdown margin calculation was also reviewed:
reactivity worth of the strongest control rod, worths of the rods withdrawn to obtain criticality, and temperature and period corrections.
c.
LTS-1100-2, " Checking for Reactivity Ancmalies," Revision 8, was used by the licensee during unit startup subsequent to the refueling outage to compare the actual and predicted critical control rod configurations (Technical Specification 3/4.1.2). The inspectors
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reviewed the procedure and data sheets (Attachment B), the vendor-supplied data, and the calculations perfonned on September 17, 1986; and verified that the methodology was technically correct, and that a reactivity anomaly did not exist.
In addition, during power operations, the licensee performs a reactivity anomaly surveillance every 31 effective full power days. The inspectors reviewed the process computer edits (00-3 Option 2) and the calculations (Attachment A) performed on October 9, 1986,(90% power)andonNovember 10,1986,(95% power), and verified that no unexpected reactivity trends were observed during power operation.
No violations or deviations were identified.
6.
Control Rod Drive Performance Testing The inspectors reviewed the following licensee procedures and test results for startup testing of the control rod drives (CRDs); and found them acceptable unless otherwise noted.
a.
LTS-1100-3 (Revision 4), " Control Rod Following and LPRM Operability Verification," was used by the licensee to verify that the control rods are coupled to their drives, and the LPRMs are operable by moving the control rods and observing a response on adjacent LPRMs.
Technical Specification 3.1.3.6 requires the control rods to be coupled to the drive mechanisms. The procedure requires that at least three of the four LPRM detectors in each LPRM string should be checked to ensure all four detectors are connected in the proper order. The licensee tested more than the minimal amount required.
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All four detectors in each string were checked several times by moving different control rods surrounding the LPRM. The only control rods that were not withdrawn were the periphery rods which had less than a 2% effect on the LPRMs.
b.
LTS-1100-4 (Revision 7), " Scram Insertion Times," was reviewed by the inspectors and found to be adequately performed and comply with the Technical Specifications. The maximum scram insertion time shall not exceed 7.0 seconds per Technical Specification 3.1.3.1.
Also, the average scram insertion times at various positions of rod insertion, shall not exceed those stateo in Technical Specifications 3.1.3.3 and 3.1.3.4 No rod exceeded the scram insertion time of 7.0 seconds or the average insertion limits. The inspectors compared the strip charts with the completed data sheets and computer inputs and found no discrepancies in the data. The inspectors also performed spot checks of the computer calculations averaging the three fastest rods in all two by two arrays, and found no errors.
c.
LTP-700-2 (Revision 5), " Control Rod Friction and Settle Testing,"
was reviewed for adequacy and compliance with administrative requirements. A limit of 15 psid differential pressure (dp) over a full stroke of the rod is the acceptance criteria for friction
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testing. All of the rods had a dp of less than 15 psid except for
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Rod 38-23, which had a dp of 27 psid.. A settle test was required and performed on this rod. The acceptance criteria of at least 30 psid differential settling pressure, with no more than a 10 psid variation in pressure over a full stroke was required and met by the settle test.
'd.
Operating Procedure LOP-RD-04, " Control Rod Drive Timing," Revision'2, was used by the licensee to adjust the insert and withdrawal times of the CRDs as necessary. The inspectors noted that CRD timing is not a Technical Specification requirement, however, it is routinely performed as a startup test at most plants. The licensee requires this test to be completed as specified in LST-86-215, "LaSalle Unit 1-Cycle 2 Startup Test Package". The administrative requirement states that CRD speeds, over the stroke of the rod, should be adjusted to 48 seconds 10%. By design, the speed tolerance is 48 seconds 20%.
A coupling check is also included as part of LOP-RD-04, although coupling checks (Technical Specification 3.1.3.6) are routinely performed in conjunction with other licensee procedures. The licensee.
performed LOP-RD-04 prior to startup, comencing in mid-July, with final review signatures dated September 16, 1986. The data sheet required to be used was LOP-RD-02T, "CRD Speed Timing".
The inspectors reviewed the procedure and test results, and found
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that the incorrect data sheet had been used to perform the test.
According to the licensee, the correct data sheet could not be found at the start of the test; therefore - Attachment C of LOS-AA-W1 was
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modified and used to document the test. The inspectors noted that a procedure change request form was not initiated. Furthermore, strict adherence to the procedure could not be achieved with the use of the modified data sheet. For example: documentation of a coupling check, date/ time, operator initials, and the as-found CRD speeds, were not required on the mcdified data sheet; however, this information was required per procedure. The as-left insert and withdrawal speeds were documented and did satisfy the administrative acceptance criteria.
Identification of the stopwatch used during the test was documented, and review signatures appeared on an attachment (LOS-AA-W1,AttachmentC). Discussions with the licensee revealed that some of the other data was retrievable; i.e., coupling checks were documented in LOS-RD-SR1, which satisfied the Technical
Specifications. The inspectors held discussions with the licensee on October 8,1986, concerning the problems with the performance of the CRD timing test.
On November 17, 1986, the inspectors received a draft procedure from the licensee which will replace LOP-RD-04. The new procedure will be a surveillance procedure, LOS-RD-SRS. The change from an operating procedure to a surveillance should, in effect, place stricter controls on test performance, documentation, and record retention requirements.
The inspectors examined the draft procedure and the associated data sheets, and found them acceptable for
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perfoming and documenting CRD timing. However, since LOS-RD-SR5 has not as yet been approved for use and will not be implemented until Unit 2 startup testing, it will be tracked as an Open Item (373/86036-01(DRS)).
No violations or deviations were identified; however, a portion of this area requires further review and is considered an open item.
7.
Additional Startup_Te_ sting The inspectors reviewed special test LST-86-215 "LaSalle Unit 1 Cycle 2 Startup Test Program," Revision 0, which outlines the method used by the licensee to perform the required startup tests subsequent to the refueling outage. The inspectors verified that the test was approved by the appropriate personnel and that a safety evaluation was performed prior to use. This procedure references the tests that should be completed during startup, including those described in Paragraphs 3 through 6 above. The inspectors reviewed the following additional startup tests and found them to be acceptable:
a.
LTP-1600-22, "SRM Performance Check," Revision 6, was used by the licensee during startup to check SRM operability (count level)
and response to rod movement per Technical Specification 3.3.7.6.
(1) The licensee performed Step F.1 of the procedure and completed Attachment A, "SRM Signal-to-Noise and Count Level Data Sheet,"
on September 16, 1986. The inspectors noted that, when inserted, SRM Channel 8 did not have a count rate of at least three cps; however, the signal-to-noise ratio was calculated to be greater than 2.0, as required by Technical Specifications.
(2) The licensee performed Step F.2 and completed Attachment B,
"SRM Response to Rod Movement," on September 17, 1986. SRM readings were obtained from all-rods-in to critical. The inspectors reviewed the data and verified that the SRM count rate increased as control rods were withdrawn.
b.
LTP-1600-23, " Intermediate Range Monitor Performance Check,"
Revision 4, was used by the licensee to verify that the IRMs provide continuous neutron monitoring, overlapping both the SRMs and the APRMs (Technical Specification 4.3.1).
The inspectors reviewed the data sheets completed by the licensee on September 17-18, 1986; and verified the existence of a SRM-IRM overlap of at least one-half decade, continuity between successive IRM ranges, and a sufficient IRM-APRM overlap. The inspectors noted that, during the performance of this test, IRM Channel F was declared inoperable and Channel B required adjustment to obtain proper continuity as required by Technical Specifications.
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c.
LTP-1600-14 " Process Computer Nuclear Program Check," Revision 3, outlines the method of verifying that the nuclear software constants have been properly updated following a refueling outage, and that the computer is acceptable for use during the new fuel cycle. The licensee completed the package on October 31, 1986. Prior to reactor startup, checks were completed using Attachments A and B; Attachment C was used to document checks during power operation.
In addition, computer program verification was accomplished at power operations; i.e., examination of 0D-3 (0ption 2), 00-1, 00-10 (0ptions 2, 5, 12, 13, 14, 17, 36), P-1, and 00-6 (Option 2). The inspectors reviewed the checklists and a sampling of the computer edits. The licensee informed the inspectors of three discrepancies
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j found during the testing, and discussed the resolutions and corrective actions taken:
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(1) Step F.2.c.l.d.1 in the procedure is a comparison check of the 00-10 Option 2 RC array (LPRM substitute values) and the 0D-1 BASLP array (BASE values at LPRM elevations). These arrays j
l should be equal to each other; however, the procedure step
l states that RC values equal zero. The licensee plans to revise this step to correct the discrepancy.
(2) A U-235 weight correction was found necessary and was i
implemented by the licensee to correct the fuel isotopic data (00-12 edit) in the plant process computer.
l (3) A software error in the P-1 (actually P-1-3 program) was l
identified by the licensee during the comparison of OD-6 Option 2, and P1 edits. According to the vendor, General Electric Company (GE), the error caused incorrect mapping of symetric control cells, resulting in fuel bundle local peaking factor (LPKF) asymmetries. This coding error was not detected earlier in plant life because the affected section of the code was not used for the non-control cell, Cycle 1 Core. The inspectors reviewed the resolution package completed on October 6,
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1986, and determined that it appeared to be adequate.
I No violations were identified.
8.
Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspectors, and which involve some action on the part of the NRC or licensee or both. An open item disclosed during the inspection is discussed in Paragraph 6.d.
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9.
Exit Interview The inspectors net with the licensee representatives (denoted in Paragraph 1) on November 20, 1986. The inspectors summarized the scope and findings of the inspection. The licensee acknowledged the statements made by the inspectors with respect to the open item (denoted in Paragraph 6.d.).
The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspectors during the inspection. The licensee did not
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identify any such documents / processes as proprietary.
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