IR 05000289/1976029
| ML19261E859 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/24/1977 |
| From: | Hurd R, Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19261E854 | List: |
| References | |
| 50-289-76-29, NUDOCS 7910170900 | |
| Download: ML19261E859 (10) | |
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II:I Form 12 (Jan 75) (Rev)
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U. S. NUCLEAR REGULATORY COMMISSION OFFICE a' INSPECTION AND ENFORCDdE'TT
REGION I
IE Inspection Report No:
50-289/76-29 Docket No:
50-289 Licensee:
Metropolitan Edison Company License No:
DPR-50 P.O. Box 542 Priority:
Reading, Pennsylvania 19603 C
':a tegory:
Safeguards Loca tio n:
Three Mile Island Unit 1, Middletown, Pennsylvania Type of Licensee:
PWR, 2535 MWt, B&W u ine, nann unced Type _of Inspection:
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Dates of Inspection:
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Dates of Previous Inspection:
November 29-December 1, 1976
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, O b teh ib/77 Reporting Inspector:
R. O. Hurd, Reactor Inspector Accompanying Inspectors:
DATE DATE DATE Other Accc=panying Personnel:
DATE I
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/ 7. i 7 "7 Reviewed By:
n E. C. McCabe, Section Chief, Nuclear Support Section No. 1 DATE Reactor Operations and Nuclear Support Branch 1482 353
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SUMMARY OF FINDINGS
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Enforcement Items None Licensee Action on Previously Identified Enforcement Items Not inspected.
Design Changes Not inspected.
Other Significant Findings A.
Current Findings 1.
Acceptable Items The following items were inspected on a sampling basis and findings did not involve an item of noncompliance, deviation or an unresolved item.
's a.
Power Distribu.'
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(Detail 2)
b.
Incore/Excore Comparison.
(Detail 3)
c.
Core Thermal Power Calculation.
(Detail 5)
d.
Shutdown Margin Calculation.
(Detail 6)
e.
Emergency Medical Drill. (Detail 8)
2.
Unresolved Items The following items were inspected and more information is required to determine item status, a.
76-20-01, Control of Computer Changes.
(Detail 4)
b.
76-29-02, Calibration of Computer Inputs.
(Detail 7)
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Status of Previously Unresolved Items Not inspected.
Management Interview An exit meeting was held onsite on December 30, 1976, at the conclusion
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of the inspection to discuss the findings of the inspection as detailed in this report.
The following licensee personnel were in attendance:
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Mr. W. W. Cotter, Supervisor-Quality Control Mr. J. F. Hilbish, Lead Nuclear Engineer Mr. G. P. Miller, Unit No. 2 Superintendent (Acting for J. Colitz, Unit No. 1 Superintendent)
Mr. J. P. O 'Hanlon, Engineer Senior-Nuclear I
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Mr. V. P. Orlandi, Lead I&C Engineer
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DETAILS 1.
Persons Contacted Discussions were held with the following persons during the conduct of the inspection activities documented in this report:
Mr. M. Benson, Engineer, Nuclear Mr. D. Curry, Engineer, Nuclear
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Mr. D. Dubiel, Radiation Protection Supervisor Mr. G. Hetz, Shift Foreman
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Mr. J. Hilbish, Lead Nuclear Engineer
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Mr. J. O'Hanlon, Engineer, Senior-Nuclear I Mr. V. Orlandi, Lead Instrument and Control Engineer 2.
Power Distribution Determination As required by Technical Specification 3.5.2.7 the licensee per-forms a determination of the power distribution using the plant
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online computer based on data from Self Powered Neutron Detectors
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(SPND) and plant process instrumentation.
The computer calculation
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includes corrections to the SPND currents for background, current fuel burnup, adjacent control rod position, boron concentration,
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fuel enrichment, temperature and xenon concentration.
Based on the
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corrected SPND signal and the ccmputer calculated va.lue of core
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thermal power, the computer then determines the core power dis-tribution.
a.
Performance Data Summary
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The inspector reviewed the " Performance Data Output" for 12/23/76 with plant at 100% of rated power.
This review included:
(1)
Normalized segment powers.
(2)
Uncorrected SPND signals.
(3)
SPND substitute values for locations F-7(1), F-8(1), F-8(4), G-11(3), M-9 (7), 0-10(1) and 0-12 (7).
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SPND correction factors.
(5)
Corrected SPND signals.
(6)
Segment exposure.
The inspector's review indicated the values were reasonable and the segment exposures properly reflected the affect of the refueling for Cycle 2.
b.
Core Limiting Conditions The inspector observed the cognizant licensee engineer request computer calculation and printout of the following items:
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(1)
Group 55, Tilt / Imbalance / Insertion.
(2) Group 32, Heat Balance Calculation.
(3)
Group 21, Rod Map.
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(4) Group.20, Worst Case Thermal Conditions.
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(5)
Group 36, Fuel Assembly to Average Fuel Assembly Power Ratio.
(6) Group 35, Fuel Assembly Power Distribution in Megawatts.
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(7) Group 34, 3-D Power Map '. Segment Power Megawatts).
The inspector discussed the content of these printouts with co gnizant licensee personnel and reviewed Surveillance Pro-cedure 1301-9.8, Core Power Map Distribution (Rev. 1, 7/1/76).
Based on the discussions and methods used in SP 1301-9.8, the following table of Core Limiting Conditions were derived by the inspector:
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Parameter Value Limit Core Power (MWT)
2522 2535 Rod Withdrawal Index (180 EFPD)
192.86 168.9 to 204.7 DNBR 3.78 1.3 Maximum Linear Heat Rate (kw/ft)
9.32(1)
15.5 Maximum Linear Heat Rate (kw/ft)
11.16(2)
15.5 The inspector found no inadequacies in this area.
3.
Incore/Excore Comparison Technical Specification 3.5.2.4 requires that the core quadrant tilt be less 4% and monitored every two hours when the core power is above 15% of rated power.
Technical Specification 3.5.2.5.d requires that the core imbalance be within the envelopes defined by T.S. Figures 3.5-2G, 3.5-2H and 3.5-2I and monitored every two hours when core pcwer is above 40% of rated over.
The inspector reviewed the following procedures and documentation relative to the above requirements:
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GP 1301-1, Shift and Daily Surveillance Check Data Sheets November 20, 1976 through December 20, 1976.
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b.
SP 1302-2.1, Power Imbalance Calibration.
c.
SP 1302-2.1, Power Imbalance Calibration Data Sheets dated December 6, 1976.
d.
AP 1203-7, Hand Calculations for Quadrants Power Tilt and Core Power Imbalance.
Based on review of the above documents, the inspector determined the licensee had monitored core power tilt and imbalance at the required frequency and has a procedure to perform manual calcula-tions of the core power tilt and imbalance using incore data.
In addition, the inspector reviewed the plant cocputer " Group 55" printout of 12/29/76 and noted the following _omparison of incore to out of core detector data:
(1) Value from Group 20, Worst Case Ther=al Conditions.
(2) Value determined by methods of SP 1301-98 which includes penalty factors for power spike, hot channel and local peaking fact.or uncertainties.
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POWER IMBALANCE Incore 0 :t of Core N15
.E6, N17 N18-2.61-3.44-1.50-3.08-2.75 (Allowable value -12.91 to +10.82)
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QUADRANT POWER TILT WX WY YZ ZW Incore Values-1.87-0.53 2.13 0.27 Out of Core Values
.027-0.05-0.14-0.08
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Tilt Limit 4.01 The inspector noted that all of the above values were within the
required Technical Specification limitations.
The inspector found no inadequacies in this area.
4.
Control of Computer Changes
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The inspector requested the licensee documentation of changes to the computer sof tware.
The licensee stated that most significant i
sof tware changes or changes in correlation constant,.re requested by che NSSS vendor and are transmitted by " Software Change Notices" or " Field Authorization Change Notices." The licensee also stated that all changes were nor= ally entered in the " Computer Log" main-tained by the cognizant computer engineer or technician.
The inspector reviewed this log for the period of March 1, 1975 through December 20, 1976.
The inspector noted that there were numerous sof tware changes detailed, including those due to the core refuel-ing in March, 1976, SPND correction factor changes on July 12, 1976 and September 27, 1976 and changes in Core Imbalance / Rod Insertion Limits due to the core exposure reaching 152 EFPD.
The insper. tor also reviewed a letter from the NSSS Vendor, dated September 7, 1976 which indicated that following the online computer update for Cycle 2, the following checks had been performed by the NSSS Vender:
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a.
An automated comparison between the IHI-IST (Intec. n System Test) and the B&W in-house IST results showed all c'wec values agreed within 0.05%.
b.
A comparison of the end of Cycle 1 historical data from TMI online computer with historical data generated from the TMI Cycle 1 performance data output showed that the burnur in the un-roded assemblies agreed within 2%.
I The inspector also reviewed the content of the TMI-IST and the sub-
sequent data.
This review indicated that the TMI-IST included all l
significant power distribution calculations and thermal hydraulic performance calculations made by the online computer.
l The inspector expressed concern that the present " computer log" and l
the other documentation of computer changes and cocputer verifica-tion were not required by approved plant procedures.
The inspector
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i noted that although the Nuclear Engineer was knowledgeable of the sof tware changes listed above, the present documentation did not
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indicate h!s review or concurrence.
The licensee stated that changes would be made to present procedures or new procedures
developed which would require documentation of all computer soft-
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ware changes and changes in stored correlation constants.
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i documentation would include a summary of the change, the reasons
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the change was made and an indication that the Lead Nuclear Engineer had reviewed and concurred with the change.
This is unresolved item 76-29-01.
5.
Core Thermal Power Calculation The inspector reviewed Survei31ance Procedure 1302-2.1 which pro-vides the instruction and documentation for the calibration of Power Range Amplifiers by heat balance which is required to be performed t/ ice weekly by Technical Specifications 4.1, Table 4.1-1 Item 3.
Normally, the core ther=al power is determined by a heat balance calculation perfor=ed by the online computer of the primary side and the secondary side of the plant.
Technical Specificatien 1.5.6 defines the weighting factor to be used in determining the core power based on both the primary side and secondary side values.
The inspector verified, by review of SP 1302-2.1 data sheets for November 20, 1976 to December 20, 1976, that the heat balance calculation had been performed at the required frequency.
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The inspector also requested that the computer power calculations be performed and receive the computer " Group 32" printout.
The inspector then verified this calculation by independently calcu-lating core power using the methods of OP 1103-16, Manual Calcu-lation of Core Thermal Power.
This manual calculation yielded a value of 2517 MWT versus the computer calculated value of 2521 MWT, a difference of less than 0.2%.
The inspector found no Iradequa-cies in this area.
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6.
Shutdown Margin Calculation Technical Specification 3.5.2.1 requires that the available shut-down margin shall not be less than cne percent ak/k with the high-est worth control rod fully withdrawn.
The licensee uses Procedure OP 1103-15, Revision 17, 11/5/76, Reactivity Balance to provide the instructions and documentation for this determination.
The inspec-tor reviewed the procedure and performed the calculation using data i
supplied by the online computer and curves contained in OP 1103-15.
Below is a summary of that calculation:
i Reactivity Factors
%Ak/k
Core Excess Reactivity (All rods out, no
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+10.5 Shutdown and Control Banks (Total rod
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worth)
-9.3 Axial Power Shape Rods (18% withdrawn)
-0.21 Boron Concentration (320 ppm)
-3.3 7.enon (Equilibrium at 100%)
-2.65 Highest Worth Rod
+2.21 Calculated SDM-2.75 The above determination assu=es no samarium poisoning and an average temperature of 532 F (this is no load Tav).
The inspector reviewed
the data sheets for OP 1103-15 dated 10/23/76 which the licensee used to verify SDM during a reactor shutdown.
The inspector also noted that procedure OP 1103-15 included consideration of the following items:
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a.
Misaligned rod considered to be inoperable and not included in
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Rod Worth Calculation.
b.
Boron concentration is checked prior to SDM calculation.
The xenon reactivity worth is based on recent power history c.
tracked by the online computer or man.a1 methods.
The inspector found no inadequacies in this area.
7.
Calibration of Computer Inputs The inspector requested the documentation for the most recent calibration of instrument channels which provide computer inputs used by the computer in the core power distribution and core ther=al power calculations.
Typical of those inputs are Feedwater Flc., Feedwater Temperature, Reactor Coolant Flow, Reactor Coolant Temperatures, Turbine Steam Header Pressure and Turbine Steam Header Temperature. The licensee stated all instru=ent channels in that category had been calibrated during the February-May 1976 refueling outage.
The inspector reviewed completed Work Request No. 13669 and noted that the above listed instrument channels had been calibrated in March 1976.
The inspector expressed concern
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that there was no plant procedure that required calibration of J
these instruments at specific interval.
The licensee stated thet a procedure or a standing work request similiar to WR 13669 would be developed and would be required to be performed at a regular in-terval such as refueling outages.
This is unresolved item 76-29-02.
8.
Emergency Medical Drill During the period of the inspection, the licensee held an annual medical drill.
Following the com?letion of the drill, the in-spector reviewed the drill with the Radiation Frotection Super-visor.
The drill scenario consisted of a si=ulated injury and contamination to two plant personnel.
One injury was of a minor nature and one was serious.
The drill involved the local volunteer a=bulance squad, transportation to the Hershey Medical Center and sinulated treatment and handling of the contaninated injured per-sonnel.
The licensee stated that the drill had been viewed and videotaped by the licensee's medical censultant and the videotape will be used in future training involving the plant staff, the ambulance squad and the hospital staff.
The licensee further stated that the drill was considered a success.
The inspector had no further questions in this area.
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