05000316/LER-2076-028, Re Pressurizer Pressure Transmitters Change Found by NRC Resident Inspector
| ML18219C341 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 07/01/1976 |
| From: | Jurgensen R Indiana Michigan Power Co |
| To: | James Keppler NRC/RGN-III |
| References | |
| LER 1976-028-00 | |
| Download: ML18219C341 (12) | |
| Event date: | |
|---|---|
| Report date: | |
| 3162076028R00 - NRC Website | |
text
Amerloan INDIANA& MICHIGANPOWER COMPANY DONALD C. COOK NUCLEAR PLANT P.O. Box 458, Bridgman, Michigan 49106 July 1, 1976 Mr. J.G.
- Keppler, Regional Director Office of Inspection and Enforcement United States Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, IL 60137
Dear Mr. Keppler:
Operating License DPR-58 Docket No. 50-315 Pursuant to the requirements of Appendix A Technical Specifications and the United States Nuclear Regulatory Commission Regulatory Guide 1.16, Revision 4, Section 2.a, the following report is submitted:
RO 50-315/76-28 Sincerely, R.W. J gen n
Plant anager
/bab cc:
R.S.
Hunter J.E.
Dolan G.E.
Lien R. Kilburn R.J.
Vollen BPI R.C. Callen MPSC K.R. Baker RO: III P.W, Steketee, Esq.
R. Walsh, Esq.
G. Charnoff, Esq.
G. Olson J.M. Hennigan R.S.
Keith PNSRC Dir., IE (40 copies)
Dir., MIPC (4 copies)
CONTROL BLOCK:
LICENSEE EVENT REPORT (PLEASE PRINT ALLREQUIRED Ii'JFORPRATIOtg 6
LICEhNSEE NAME
~codiN I 0CC10 0 00000 8 9 14 15 LICENSc EVcNT TYPE TYPE 0
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25 26 30 31 80 03 THAT IN APRIL OF 1976 THE RANGES OF THE PRESSURIZER PRESSURE TRANSMITTERS HAO BEEN~CHANGED CQ SO Qp4 FROM 1674. 1 - 2471. 1 PSI TO 1725.9 - 2525.9 PSI.
THIS WAS DONE TO'ECTIFY AN IMPROPER 80 7
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HEAD CORRECTION.
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DESCRIPTION
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8 9 12 P ROBMLTE CONSEQUENCES Q~g NA 7
8 9 I OSS OR DAMAGE TO FACILE TYPE OESCRPTION Q~g
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89 10 PUBLICITY NA 7
8 9 80 80 h!ETHOO, OF OISCOVERY D
44 45 OTHER STATUS OISCOVERY OESCRIPTION NRC INSPECTORS AUDIT SO AhiOUNT OF ACTS/ITY LOCATION OF RELEASE 44 45 h
SO 80 80 80 80 REPORT REPORT CATEGORY TYPE SOURCE OOCKET NUMBcR FVENT OATE REPORT DATE [oa]cov'vLL I
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57 58 59 60 61 68 69 74 75 80 EVENT DFSCRIPTION 0 2 WHILE IN MODE 1 AT 100 PERCENT POWERs DURING AN AUDITs THE NRC RESIDENT INSPECTOR FOUND AODITIONAL FACTORS Q1 g SEE SUPPLEMENT 7
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RISCHLING PHoNE 616 465-5901 68)
SUPPLEMENT TO REPORTABLE OCCURANCE RO-050-0315 On April 29, 1976, during a scheduled plant outage, it was discovered by a Control and Instrument Senior Technician that a 25 psi static head correction to the pressurizer pressure transmitters (NPP-151, NPP-.152, NPP-153, NPS-153) had been subtracted from the transmitter input instead of being added.
This error had the effect of lowering the setpoint for the pressurizer low pressure reactor trip from 1865 psig to 1815 psig (Technical Specification Section 2.2.1) and the pressurizer low pressure portion of the safety injection and feedwater isolation signal from 1815 psig to 1765-psig (Technical Specifications Section 3.3.2. 1).
The error also resulted in the operating band for pressurizer pressure being maintained at a pressure 50 psi lower than the normal operating pressure as specified in Technical Specifications Section 3.2.5.
As soon as the error was discovered, all four pressurizer pressure transmitters were recalibrated correctly and temporary change sheets to the calibration procedures were prepared.
Also, a complete review of all calibration procedures containing pressure head corrections was made with no other errors found.
On January 19, 1974, the initial calibration of the pressurizer pressure transmitters was performed.
Prior to this date, discussions were held to determine if the 25 psi static head correction should be added or subtracted.
As a result of these discussions, it was erroneously decided that this static head should be subtracted from the transmitter span to allow for proper head correction.
The calibrations were repeated in October 1974, and again in April 1976, at which time the error was found.
Although the error was promptly corrected, a condition report was inadvertently not prepared which would have initiated NRC notification.
A review has been made of the effects on safety of the plant with incorrect pressurizer pressure transmitter setpoints.
Operation in this manner has three effects with respect to the safety analysis, which are, as mentioned above
( 1) an effective lowering of the trip points for low pressurizer pressure reactor trip from 1865 to 1815 psig, (2) lowering the trip point for safety injection actuation and feedwater isolation from 1815 psig to 1765 psig, and (3) actual operation of the plant could have been as low as 50 psi below that pressure specified in the technical specifications which is the initial pressure assumed in the safety analysis.
yl Operation of the unit with the lower reactor coolant pressure over the time period involved did not cause any conditions which,had an adverse affect on the health and safety of the public.
,The effects on the safety analyses of the lower pressurizer pressure trip 'setpoint and lower reactor coolant pressure are discussed below.
U; J ill I
Page Two In Section 14 of the Donald C.
Cook Nuclear Plant FSAR, the safety analyses assume reactor trip to occur at 1685 psig rather than the low pressure setpoint of 1865 psig.
The reason for this dif'ference is to allow for setpoint error.
In practice, the actual gages were calibrated with weights traceable to the National Bureau of Standards,,and the maximum difference between actual pressure and,,indicated, gage pressure is conser-vatively estimated to be no greater than 10 psi.
Thus, alow pressurizer pressure reactor trip setpoint of 1865 psi,would reflect an actual pres-surizer pressure reactor trip at no less than 1805 psi, assuming calibration errors resulted in the lowest possible system pressures.
Thus, during operation with this incorrect setpoint, the low pressure setpoint would still have resulted in reactor trip at a level above the minimum analyzed in the safety analysis.
Operation with this incorrect setpoint resulted in the plant being slightly outside the initial pressure bounds assumed in both the loss of coolant accident analysis (LOCA) and plant transient analysis.
With respect to LOCA analysis, operating at an initial system pressure 50 psi lower than the pressure assumed in the analysis has a minor effect in the subcooled blowdown portion of the analysis.
This effect is in the con-servative direction in that ECCS analyses are based on high initial system pressures in order to obtain conservatively high peak clad temperatures.
The short term containment peak pressures would have also been lowered by some very small amount since subcooled blowdown decreases with decreasing system pressure, thus lowering the energy deposition into the containment.
Long term pressure would not be affected by this small change in system pr'essure.
Lowering of system pressure and safety system setpoints by 50 psi had the effect of lowering the departure from nucleate boiling ratio (DNBR) associated with various plant transients analyzed in Section 14.1 of the FSAR.
An assessment of the effect of lowering system pressure by 50 psi was made by evaluating the effect on the correlation used to evaluate DNBR (the W-3 correlation).
As a result of this evaluation, in conjunction with a review of DNB data, it is concluded that the worst effect on system transients would have been to lower the calculated DNBR by approximately 3 percent.
A review of the limiting DNBR for the various transients and accidents included in the safety analysis in both Section 14 of the Donald C.
Cook Nuclear Plant FSAR and the Fuel Densification Report (Reference
- 1) was made.
This review indicated that these analyses included the effect of a densifi-cation power spike, and was completed prior to submission of Reference 2,
which indicates that the densification power spike is not appropriate to include as part of DNB calculation.
If the accident analysis in the FSAR and Reference 1 had been redone on the basis of information presented in Reference 2, the result would have been to raise the calculated DNBR by approximately 5 percent, thus completely offsettin'g the effect of operating the plant at 50 psi below the pressure specified in the Technical Specifications.
As a result of the above, it is concluded that if any of the postulated accidents analyzed in the FSAR should have occurred with this error in effect the health. and safety of the public would not have been adversely affected.
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Page Three To prevent reoccurrence of this error, the instrument maintenance procedures governing the calibration of these transmitters have been revised.
Also, control and instrument personnel were reminded to insure that errors of this type are promptly reported, and a program has been initiated to ensure instrument maintenance procedures include a signoff on the data sheets indicating that any out-of-specification data has been reported.
In April, 1976, the temporary sheet to the'ressurizer pressure transmitter calibration procedure was reviewed and approved by the Plant Nuclear Safety Review Committee (PNSRC).
However, the PNSRC overlooked the fact that the setpoint error had: been a technical specification violation.
To give more'ssurance that temporary sheets are reviewed for their impact on the technical specifications, a check list will be used by the PNSRC which will include an 'item requiring evaluating the,,effect of,the change on technical specifications.
A check list has been prepared by the plant and is under PNSRC review.
Use 'of a final check list by the PNSRC will be initiated by July 30, 1976.
REFERENCES:
1.
Westinghouse Report WCAP-8164, "Fuel Densification Donald 'C.
Cook Nuclear Plant Units 1 and 2", dated June, 1973.
2.
We'stinghouse Report WCAP-8218, "Fuel Densification Experimental Results and Nodel for Reactor Application",, dated October, 1973.
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UNI TO'0 STATI!S NUCLEAR REGULATORY COMMISSION ACGION III 700 AOOSCVCLT AOAO GLLN CLLYN, II.LINOIS 00137 dd d
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(Licensee 5 Docket Ho. (or License No.))
Transmittal Date:
Distribution:
June 28 1976 Distribution:
IE Chief, FOSB IE Chief, FC8EB IE:HO(4)
Licensing(4)
DR Central Files IE Files IE Chief, FOSB IE Chief, FCEEB IE:Hq(4)':D/D for Fuels 8 ltaterials DR Central Files IE Files
1 tesatrla
~y f IIVDIAIVA 8 MlICHIGAIV PO O'E'R COMPAIVV DONALD C. COOK NUCLEAR PLANT P.O. Box 458, Bridgman, Michigan 49100 June 17, 1976 Mr. J.G.
- Keppler, Regional Director Office of Inspection and Enforcement United States Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, IL 60137 Operating License DPR-58 Docket No. 50-315
Dear Mr. Keppler:
This letter is to confirm our verbal report to NRC Inspector, K.R. Baker, on June 17, 1976.
On June 16,
- 1976, Mr. Baker was reviewing Instrument Maintenance Procedure data and determined that we had violated Appendix A Technical Specification Section 2.2.
1 and Table 2.2-1 Item 9.
Due to procedural error the correction for the static pressure leg was improperly applied in the calibration of the pressurizer pressure transmitters.
The error had been found and corrected on April 12, 1976, but had not been reported to the NRC.
Plant Nuclear Safety Review Committee review of this incident on June 17, 1976, revealed that we had also violated Appendix A Technical Specification Sections 3.3.F 1, Table 3.3-4 Item ld and 3.2.5.
The PNSRC's analysis of this error indicates that the health and safety of the public was not in jeopardy.
Our investigation is continuing and a formal report will be submitted as required by NRC Regulatory Guide 1.16, Revision 4, by July 1, 1976; Sincerely, R.W. Ju nse Plant Manager
/bab cc:
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