IR 05000245/1979024
| ML19294B587 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 12/03/1979 |
| From: | Caphton D, Graham P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19294B563 | List: |
| References | |
| 50-245-79-24, 50-336-79-25, NUDOCS 8003050112 | |
| Download: ML19294B587 (11) | |
Text
.
.
U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I 50-245/79-24 Report No. 50-336/79-25 50-245 Docket No. 50-336 DPR-21
--
C Category C
License No. DPR-65 Priority
--
Licensee:
flortheast Nuclear Energy Company P. O. Box 2'/?
Hartford, Connteticut 06101 Facility Name:
Millstone f. iclear Power Station, Units 1 and 2 Inspection at:
Waterford, f.
- necticut Inspection conducted: October 2-5, 1979 b b.
//- Lc 79 Inspectors:
P. D. Graham, Reactor Inspector date signed
date signed
,e date signed
,
Approved by:
)
G
/4/6
D'.L'.i,a%t'on, Chief,NuclearSupport date signed Section No. 1, RO&kS Branch Inspection Summary:
Inspection on October 2-5,1979 (Report flos. 50-245/79-24; 50-336/79-25)
Areas Inspected:
Routine, unannounced inspection of post refueling startup testing including verification of core power distribution, core thermal power, shutdown margin, reactivity coefficients (Unit 2), and LPRM/APRM calibration (Unit 1).
The inspection involved 34 inspector-hours on site by one regional based NRC inspector.
Results: Of the six areas inspected, one item of noncompliance was identified (Deficiency - failure to initiate and review a temporary procedure change - Unit 2).
003050/f'
Region I Form 12 (Rev. April 77)
.
.
DETAILS 1.
Persons Contacted
- J. Black, Unit 3 Superintendent
- E. Farrell, Unit 2 Superintendent
- R. Herbert, Unit 1 Superintendent.
- J. Opeka, Station Superintendent
- T. Piascik, Reactor Engineer, Unit 1
- P. Przekop, Engineering Supervisor, Unit 1
- S. Scace, Engineering Supervisor, Unit 2 S. Sudigala, Reactor Engineer, Unit 2 The inspector also talked with and interviewed other licensee personnel during the course of this insoection.
- denotes those present at the exit interview.
2.
Post Refueling Startup Testing-Unit 1 The inspector reviewed tests, checks, and documents described below to verify that startup testing was conducted in accordance with technically adequate procedures ar.d as required by Technical Specifications (TS).
Criticality predictions and preiicted physics parameters are contained in Millstone Cycle 7 Management Report dated July 10, 1979, Millstone 1 Cycle 7 Management Report Supplement dated September 17, 1979, and Millstone 1 Reload 6, Nuclear Design Report dated February 2,1979.
a.
Post Refueling Startup Report Review The inspector discussed with the licensee the Cycle 6 Startup Report.
The report was reviewed to verify the following:
The report included the required information; and,
--
The test results were consistent with design predictions and
--
performance specifications.
Except as noted below, the inspector had no further questions in this area.
Paragraph 3 of the Startup Report stated that the TS limit for shutdown margin is 0.47% delta k/k.
The inspector pointed out that this value takes in to account the R +.33% value discussed
.
in the bases section of TS, and that the actual TS limit is 0.33%
delta k/k.
The licensee acknowledged the discrepancy, and a change to the Startup Report will be made.
b.
Control Rod Drive Scram Insertion Times The licensee verified that the control rod drive scram insertion times for all control rods were within the requirements of TS 3.3c by the performance of procedure SP-1051, Control Rod Scram Time Test, Revision 2 on June 28, 1979.
The inspector reviewed the completed procedure and associated strip recorder chart traces, and with the exception of the following items had no further questions.
Procedure SP-1051 requires that the rod selected for testing, as provided on data form SP-1051-5, be verified b/ a licensed operator.
The inspector noted that this verification was not performed by a licensed operator.
A licensee representative commented that there is no license requirement for a licensed operator to indepen-dently verify the selected rod, and a change to the procedure was made to delete the requirement before the inspection was completed.
Procedure SP-1051 step 7.10 requires that reactor power be logged in percent on data form SP-1051-2.
The operators completing SP-1051-2 log power in megawatts electric.
The inspector discussed the discrepancy with the cognizant licensee representative who stated that a procedural cnange would be made to allow more leeway in recording reactor power.
This item is unresolved pending the procedural change and inspector followup (245/79-24-01).
c.
Calibration of Local Power Range Monitors (LPRM)
The licensee calibrates the LPRM's in accordance with procedure RE-1003, LPRM Calibration and Gain Adjustment, Revision 2, dated July 26, 1979.
The inspector reviewed the procedures completed on July 23 and August 29, 1979, which included the Traversing Incore Probe (TIP) traces and supporting computer data.
The inspector witnessed the operation of the TIP System in accordance with procedure IC-405A, Nuclear Instrumentation (TIP Channels),
Revision 0, on October 3, 1979.
No items of noncompliance were identifie.
d.
Calibration of Average Power Range Monitor (APRM)
The licensee calibrates the APRM's weekly pursuant to the require-ments of TS Table 4.1.2 by the performance of procedure SP-1040, APRM Calibration Using Heat Balance, Revision 1, dated May 4, 1979.
The inspector reviewed the completed procedures performed during the period of July 13 through August 31, 1979.
No items of noncompliance were identified.
e.
Shutdown Margin Determination (SDM)
The licensee complied with the SDM demonstration requirements of TS 4.3.a.1 by determining that the reactor SDM with the strongest control rod fully withdrawn was equal to 1.79% delta k/k vice the required TS SDM value of 0.33% delta k/k.
The licensee used the
"in-sequence critical" method of demonstrating SDM by performing procedure SP-6908, Reactivity Margin - Core Loading, Shutdown Margin Test, Revision 3, on June 27, 1979.
The inspector reviewed the above procedure, and except as noted below, had no further questions in this area at this time.
Procedure SP-690B step 7.2 requires an independent verification of the data supplied by the Reactor Engineer.
The step appeared to be confusing as written.
The cognizant licensee representative stated that a procedural change would be made to clarify the verification requirement.
This item is unresolved pending the procedural change and inspector followup (245/79-24-02).
f.
Core Power Distribution The procedures and methods used by the licensee to verify that the plant is operating within the power distribution limits of TS 3.11 were reviewed and discussed with cognizant licensee personnel.
The TIP System is used to measure the neutron flux distribution in the monitored core locations.
The flux distribution, control rod positions, core flow, and core power are monitored directly by the on-line process computer.
The computer orogram 05-1 normalizes flux measurements from the TIP machines.
The P-1 program performs a core power distribution calculation based on the normalized flux measurements.
The 00-6 program provides calculations and edits of Linear Heat Generation Rates (LHGR) and Minimum Critical Power Ratio (MCPR).
.
.
.
The inspector reviewed the following plant documents and data to verify that the plant is operated in compliance with the limits defined in the TS.
00-1, Whole Core LPRM Calibration and Base Distributions
--
completed July 23 and August 29, 1979.
P1, Pe fodic NSS Core Performance Log performed August 27,
--
1979.
00-6, Option 3, the 12 Bundles Closest to CPR limits performed
--
August 27, 1979 at 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br />.
OD-6, Option 4, the 12 Highest Ratios of a Bundle Maximim
--
Average Planar LHGR (MAPLHGR) to its Limiting LHGR performed August 27, 1979 at 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br />.
SP-1030, Reactor Core Peak Heat Flux Check, Revision 3,
--
dated September 14, 1979.
SP-1031, Local and Average Planar LHGR Surveillance, Revision
--
1, dated July 26, 1978.
SP-1032, MCPR Surveillance, Revision 2, dated August 1,
--
1979.
Daily Surveillance Log for September 1-18, 1979.
--
RE-1017, "BUCLE" Maintenance, Revision 1, dated April 25,
--
1979.
RE-1051, Core Flow Calibration, Revision 3, dated September
--
24, 1979, completed August 30, 1979.
Except as noted below, the inspector had no further questions in this area at this time.
The Daily Surveillance Log step 4 addresses the three types of fuel bundles used in the Cycle 7 core; 7x7, 8x8 and 8x8R.
Step 7 of the Log does not include the 7x7 type bundle.
A licensee representative acknowledged the discrepency and committed to a procedural change that would incorporate 7x7 fu?1 into step 7.
This item is unresolved pending this change and inspector review (245/79-24-03).
.
g.
Core Thermal Power The procedures and methods used by the licensee to determine core thermal power were reviewed and discuned with cognizant personnel.
The inspector observed the cognizant engineer request the performance of an on-demand program, "0D-3, Core Thermal Power and APRM Cali-bration," and receive the resultant printout.
The inspector reviewed the printout comparing the edited process values with the control room indicator readings and performed a verification calculation.
The inspector also witnessed the licensee's calcula-tion of thermal power using an off-line calculator.
The following documents were reviewed.
0D-3, Core Thermal Power and APRM Calibration, performed
--
October 3, 1979 at 0930 hours0.0108 days <br />0.258 hours <br />0.00154 weeks <br />3.53865e-4 months <br />.
RE-1002, Core Heat Balance - Power Range, Revision 5, dated
--
June 14, 1979, completed October 3, 1979.
Except as noted below, the inspector had no further questions in this area at this time.
Procedure RE-1002 requires that data form RE-1002-1 be used to record data.
The inspector noted that the operators were not using this form, but instead, were attaching the off-line computer printed tape.
The tape contains the same data that is required by the data form.
The inspector discussed this with the cognizant licensee representative and a commitment was made to change the procedure to allow the operators the option of attaching the tape.
This item is unresolved pending the above change and inspector review (245/79-24-04).
h.
Reactivity Anomaly The procedures and methods used by the licensee to determine the reactivity anomaly were reviewed and discussed with the cognizant personnel.
Procedure SP-1050, Critical Rod Configuration Comparison (Startu,'), Revision 1, was performed on July 23, 1979 as part of the.ita. tup Test Program to compare predicted versus actual critical rod configuration.
The following data was noted by the inspector.
Critical Position on Startup Predicted Actual Differences TS Limit 1068 notches 1112 notches 44 notches
+1% delta k
.
Procedure SP-1045, Critical Rod Configuration (Operating), Revision 0, dated April 6,1977, is performed every month to insure compliance with TS 4.3.e.
Except as noted below, the inspector had no further questions in this area at this time.
Procedure SP-1045 requires that the actual control rod notches inserted in the core be calculated by using the control rod density value provided by the process computer. The inspector noted that the personnel performing the procedure are manually determining the number of notches inserted by adding control rod positions.
The inspector discussed this with the cognizant licensee representative and a commitment was made to change the procedure to allow this option.
This item is unresolved pending the procedural change and inspector review (245/79-24-05).
3.
Post Refueling Startup Testing - Unit 2 The inspector reviewed tests, checks and documents described below to verify that startup testing was conducted in accordance with technically adequate procedures and as required by TS. The inspector also reviewed the licensee's Startup Report to verify the following:
The report included the required information; and,
--
The test results were consistent with design predictions and per-
--
formance specifications.
The inspector noted that in Section III.B of the Startup Report the TS values for the unrodded integrated radial peaking factor, Fr, and the unrodded planar radial peaking factor, Fxy, were interchanged.
The licensee acknowl'adged the discrepancy, and a change to the Startup Report will be made.
At the time of the inspection, the results of the Startup Testing Procedures had not completed the required review cycle.
This item is unresolved pending completion of the review cycle (336/79-25-01),
a.
Initial Criticality and Low Power Physics '.ests The startup and low power testing for cycle 3 was conducted according to T-79-9, Criticality / Low Power Physics Test - Cycle 3, Revision 0, approved May 10, 1979.
The procedure includes Control Element Assembly (CEA) symmetry checks, rod drop times, initial criticality, zero power isothermal temperature coefficient determinations, and control rod worth determinations.
The inspector determined the fol'owing summary of test results:
.
.
Lower Power Physics Tests Test Predicted Measured CEA Symmetry Check
+1.St from group average all satisfactory Rod Drop Times
<3.1 sec all <3.1 sec All Rods Out (AR0) Critical 1205 + 75 ppm 1221 ppm Boron (Group 7 at 135 steps)
Isothermal Temperatures
+0.372x10-4* +.3x10-4*
+.272x10-4*
Coefficient (AR0)
CEA Group Worth:
.64**++
.64**
.25
.25
.16
.17
.95
.88
.72
.67
1.08 1.15
- Units: numbers are delta k/k/F.
- Units: numbers are % delta k/k.
++ Acceptance criteria is the greater of +15% of predicted or 0.06% delta
k/k.
All tests were performed in accordance with the approved procedure.
The inspector reviewed the procedure, the data obtained and the calculations perforn.ed. All parameters reviewed were within test acceptance criteria and TS limits.
Procedure T-79-9 has not completed the required review cycle.
The incompleteness was identified above as Unresolved Item (336/79-25-01).
b.
Escalation to Power Escalation to power was conducted in accordance with procedure T-79-10, Power Ascension Test, Cycle 3, Revision 0, dated May 21, 1979 and T-79-16, Power Ascension to Stre'ch Power - Cycle 3, Revision 0, dated June 27, 1979.
These proced_.es required the measurement of the Isothermal Temperature Coefficient (ITC), the Power Coeffi-cient of Reactivity (PC), and the Core Power Distribution measure-ments at the 50%, 94.8% and 100% power plateaus.
Based on review of the above procedures, the inspector developed the following table of data:
.
.
50% Power 94.8% Power 100% Power Value Predicted Measured Predicted Measured Predicted Measured n
n ITC-0.356 ++
-0.502
--
--
--
--
--
--
--
--
T F
<l.776 1.535
<l.637 1.513
<l.63 1.502 r
T F
<l.776 1.64
<1.676 1.552
<l.615 1.532 xy Tq (Azi- <2%
0.88%
<2%
0.39%
<2%
0.2%
muthal Power Tilt)
MLHR
<l5.6 7.204
<l5.6 12.93
<l5.6 13.13 (Maximum Linear Heat Rate kw/ft)
- Units:numbersarex10fdeltak/k/F,
- Units: numbers are x10-delta k/k/% Power.
++The acceptance criteria is +.3x10-4 The inspector noted that the measured value of PC adjusted to 100% power did not meet the acceptance criteria.
The licensee presents a discussion of the evaluation made concerning the PC in the Startup Report.
The inspector had no further questions concerning PC at this time.
All tests were performed according to the approved procedures.
The inspector reviewed the procedures, the data obtained, and the calculations performed.
All parameters reviewed, with the excep-tion of PC, were within test acceptance criteria and TS limits.
The required review cycle has not been completed.
The incomplete-ness was identified above as Unresolved Item (336/79-25-01).
c.
Other Tests and Surveillances The inspector reviewed the following tests and surveillance procedures conducted during and after the Startup Testing Progra.
OP-2208, Reactivity Calculations, Revision 3, dated March
--
28, 1979, Completed May 17, 1979.
EN-21004, Reactivity Measurements, Revision 0, dated March
--
28, 1979.
EN-21010, CEA Drop Times, Revision 0, dated May 10, 1979,
--
Completed May 17, 1979.
RE-21003, Incore Analysis (INCA), Revision 4, dated April
--
18, 1978, Completed May 23, 1979.
EN-21019, Incore Detector Operability, Revision 0, dated
--
January 31, 1979, Completed May 23, 1979.
RE-21002, Heat Balance, Revision 2, dated July 2, 1977,
--
- ompleted July 20, August 7 and September 6, 1979.
--
RE-21006, Core Flow Determination, Revision 1, dated August 9, 1977, Completed May 26, 1979.
EN-21014, Azimuthal Power Tilt - Tq, Revision 0, dated May
--
10, 1979, Completed May 23 through October 1,1979.
SP-26010, Power Range Safety Channel and Delta 1 Power
--
Channel Check, Revision 1, dated March 19, 1979, Completed October 5, 1979.
IC-2401E, Nuclear Instrument Calibration to Incores, Revision
--
2, dated February 16, 1979, Completed October 5, 1979.
EN-21012, Radial Peaking Factors, Revision 0, dated May 13,
--
1979, Completed May 23 through October 1,1979.
EN-21016, Incore Detector Alarms, Revision 0, May 20, 1977,
--
Completed May 28 through October 1, 1979.
EN-210ll, Moderator Temperature Coefficient - MTC, Revision
--
0, dated May 13, 1979, Completed June 7, 1979.
EN-21018, Reactivity Anomalies, Revision 0, dated May 20,
--
1077, Completed July 2, 1979.
.
Except as noted below, the inspector had no further questions in this area at this tim.
Review of Procedure SP-26010 indicated that a change to the procedure had been made by the operators without initiating a temporary change request.
Plant personnel could not identify the date on which the change was made.
This failure to properly initiate, review and approve a temporary procedural change as required by TS 6.8.3 constitutes a Deficiency level item of noncompliance (336/79-25-02).
4.
Unresolved Items Items about which more information is required to determine acceptability are considered unresolved.
Paragraphs 2 and 3 of this report contain unresolved items.
5.
Exit Interview The inspector met with licensee representatives (denoted in f'aragraph 1) at the conclusion of the inspection on October 5, 1979.
The inspector summarized the scope and findings of the inspection as they are detailed in this report