IR 05000219/1979009
| ML19248D205 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/20/1979 |
| From: | Bettenhausen L, Caphton D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19248D202 | List: |
| References | |
| 50-219-79-09, 50-219-79-9, NUDOCS 7908130422 | |
| Download: ML19248D205 (6) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I Report No.
50-719/79-00 Docket No. 50-219 License No. DPR-16 Priority Category C
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Licensee:
Jersey Central Power and Licht Comoany Madison Avenue at Punch Bowl Road Morristown, New Jersey 07960 Facility Name:
Oyster Creek Nuclear Generating Station Inspection at:
Forked River, New Jersey Inspection conducted:
April 24-26, 1979 Inspectors:
AfC N7 L.
M. Bettenhausen, Reactor Inspector date signed date signed
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date signed Approved by:
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d D. L. CaphYo'n,' Chief, Nuclear Support Section
'date' signed No. 1, R0&NS Branch Inspection Summary:
Inspection on April 24-26, 1979 (Recort 50-219/79-09)
Areas Inspected:
Routine, unannounced inspection by regional based inspector of post-refueling startup testing for Cycle 8 initiated in December,1978 including control rod sequence and reactivity checks, determination of reactivity anomaly and shutdown margin, core power distribution, average power range monitor and local power range monitor calibrations.
The inspection involved 19 inspector-hours on site by one NRC regional based inspector.
Results:
No items of noncompliance were identified in the six areas inspected.
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Region I Form 12 (Rev. April 77)
7908130'/JS
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DETAILS 1.
Persons Contacted
- K. Fickeissen, Technical Engineer A. Rone, Technical Supervisor
- D. Ross, Manager, Generating Stations - Nuclear
- J. Sullivan, Chief Engineer
- present at exit interview conducted April 26, 1979 2.
Licensee Action on Previous Inspection Findings (Closed) Deficiency (219/78-03-02): The Training Administrator presented results of a training audit conducted April 4, 1979 which showed that all
" Job Qualification Checksheets" were in place and current as of the audit date.
Full compliance has been achieved.
(Closed) Unresolved Item (219/79-02-01):
Revision 6 of the Oyster Creek Nuclear Generating Station Training Manual, dated March 30, 1979, revised the chemical technician training program, Section 432, to reflect actual conduct of training.
This item is closed.
(Closed) Unresolved Item (219/79-02-03): The Training Manual revision noted above defines the training audit function to be an informal management review and describes methods for resolution of audit findings.
This item is closed.
3.
Analysis and Documentation of Reload Information for Oyster Creek Cycle 8 Cycle 8 uses fuel essentially the same as that used in Cycle 5; slight mechanical differences are described below.
Analyses by General Public Utilities Service Corporation and by the fuel vendor, Exxon Nuclear Company showed that no changes to Technical Specifications were necessary.
The following documents were reviewed by the inspector:
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a.
" Reload Information and Safety Evaluation Report for Oyster Creek Cycle 8 Reload", GPU Service Corp., August 3, 1978.
b.
Addendum 1 to above, dated November 28, 1978, describing four pre-pressurized fuel rods, pressured with four atmospheres of helium and a safety evaluation of Cycle 8 operation with four prepressurized rods.
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c.
Addendum 2 to above, dated November 28, 1978, describing the safety evaluation using final cycle 7 exposure and as-loaded cycle 8 fuel assemblies.
d.
Minutes of Meeting, Plant Operations Review Committee, August 28, 1978, Oyster Creek.
e.
Minutes of Meeting, Plant Operations Review Committee, November 30, 1978, Oyster Creek.
The slight mechanical changes to the fuel design described in reference a.
were (1) simplified tie-rod-to-upper tie plate locking components which did not modify tie plate flow area or affect mechanical integrity; (2) simpli-fied inert non-fueled spacer capture rod using a continuous zircaloy rod rather than a segmented rod; (3) changed specifications on diameter of zircaloy filler rod in the non-fueled spacer capture rod to improve manu-facturability with no changes in bounds on as-built rods.
The changes were first reported in cycle 7 and have had no effect in the operation or behavior of the fuel assemblies.
The fuel assembly containing the four prepressurized fuel rods was loaded in location 41-20 for cycle 8.
No items of non-canaliance or deviations were identified.
4.
Startup Testing Following Refueling The inspector reviewed the following tests and checks to verify that the testing was done in accordance with technically adequate procedures and as required by Technical Specification:
a.
Control Rod Scram Insertion Time Test, Procedure 617.4.003, Revision 1, dated November 14, 1977 was initiated on November 18, 1978 with the following results:r Average Measured TS Average Times to be Insertion Scram Time No Greater Than 5%
0.37 sec.
0.375 sec.
20%
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0.80 0.900 50%
1.75 2.00 90%
3.01 5.00 n>%
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The procedure was again performed for eight selected rods subsequent to a manual scram on December 7, 1978.
Results were consistent with the November 18, 1978 results.
Twenty-five rod drives were replaced with rebuilt drives during the refueling outage.
They were tested satisfactorily for dirferential pressure and drive time in the period October 24-26, 1978.
b.
Rod Worth Minimizer Sequence, Procedure 1001.5, Revision 1 dated January 31, 1978 was used to verify that Control Rod' Sequence VIII-A-I was properly imput to the Rod Worth Minimizer. The RWM printout was compared to the Master Sheet for Control Rod Withdrawal on December 3,1978 and reviewed and approved by the Technical Supervisor on December 4, 1978.
c.
Power Distribution Measurement, Procedure 1001.12, Revision 4, dated November 21, 1977 describes the data collection and analysis used to ascertain core power distribution and verify that thermal-hydraulic parameters are within limitations of Technical Specifications. The inspector reviewed the data and results of power distribution measure-ments numbered 8002 through 8009 and performed between December 9,1978 and January 26, 1979.
d.
Shutdown Margin Measurement Test, Procedure 1001.27, Revision 4, dated June 26, 1978 was performed on November 14-15, 1978.
The results are summarized in a memo dated November 30, 1978, subject:
Oyster Creek BOC 8 Shutdown Margin Calculation and XTRA Cold Bias Results.
The test was required to show a shutdown margin of 1.185% with the highest worth rod withdrawn.
The test showed that the reactor was subcritical with the strongest rod, 26-07, withdrawn and with rods 22-11 and 30-11, worth 1.21% withdrawn, demonstrating adequate shutdown margin.
Another test, shutdown Margin Demonstration, Procedure 1001.26, Revision 2 dated June 26, 1978 demonstrates by the reactor remaining subcritical that adequate shutdown margin is available when withdrawing rods according to the Control Rod Schedule Master Sheet.
This test was done on November 24, 1978.
e.
Calibration of Local Power Range Monitors (LPRM), Procedure 1001.12, Power Distribution Measurements contains provisions for LPRM calibration without specifying a time period for the calibration.
The data sheet for Procedure 1001.12 and worksheets attached indicate that post-startup LPRM calibrations were done on December 11, 1978 and January 9, 1979.
The inspector found no procedure for doing LPRM calibrations.
Licensee h
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representatives stated that LPRM calibration at approximately monthly
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intervals was an adequate practice.
A commitment to develop a pro-cedure for method (s) and frequency of LPRM calibration by August 1, 1979 was made by licensee representatives.
This item is unresolved pending NRC review of the procedure developed (219/79-09-01).
f.
Average Power Range Monitor (APRM) Calibration and Core Thermal Power Evaluations.
These calibrations and checks are done as Procedure 1001.6, Core Heat Balance - Power Range, Revision 7, dated June 26, 1978.
Technical Specifications require the APRM Level to be calibra-ted once per three days.
A review of data sheets from Procedure 1001.6 performed in the period December 9,1978 - January 31, 1979 showed that the heat balance and APRM calibrationF were typically perfomed once per shift (three times per day) w; more often during power ascensions.
The satisfactory gain Laustment of the APRMs is noted in the Control Room Operators '.ogbook each time the procedure is perfomed.
The inspector examined the data sheets for core thermal power, verified that the physica' properties used were accurate and independently calculated results for several selected cases.
g.
Determination of Reactivity Anomalies.
The predicted curve of re-activity insertion versus core exposure is developed in accordance with Procedure 1001.16, Reactivity Anomaly Curve Development, Revision 0, dated December 5, 1975.
In this procedure, XTRA calculations performed at a nominal value near rated core thermal power and re-circulation flow for various burnup steps give resultant reactivity values.
These values are plotted, along with + 0.5% and 1.0% warning and control deviations as a function of exposur_e, specifically for c/cle 8.
The credicted values are then compared with a plot of the actual reactivity insertion as determined by actual rod notch insertion, corrected as determined by Procedure 1001.17, Reactivity Anomaly Check,' Revision 2, dated April 4,1979.
Reactivity anomaly checks were made for Cycle 8 on December 28, 1978, January 9,1979, January 27, 1979, February 11, 1979, February 24,1979, March 9,1979, March 22, 1979 and April 18, 1979.
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The procedure as written was not appropriate in that the XTRA calcula-tions were based on 1900 MWt operation, so the reactivity anomaly check was based upon 1900 MWt instead of the licensed power of 1930 MWt.
In addition, Procedure 1001.17 contained several editorial errors.
The licensee representatives acknowledged that the written procedure did not reflect present practice and agreed to revise the procedure by June 1, 1979.
This item is enresolved pending NRC review of the revised procedure (219/79-09-02).
No items of noncompliance or deviations were identified.
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5.
Unresolved Items Unresolved items are those items for which further information is required to ascertain whether the items are acceptable or items of noncompliance.
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Unresolved items are identified in paragraph 4 of this report.
6.
Exit Interview A management meeting was held with licensee personnel (denoted in paragraph 1) at the conclusion of the inspection on April 26, 1979.
The purpose, scope and findings of this inspection were discussed at the exit interview.
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