L-87-371, Responds to Violations Noted in Insp Repts 50-250/87-33 & 50-251/87-33.Corrective Actions:Reducer & Branch Line to Isolation Valves & Steam Trap Socolet Replaced & Procedure ONOP-12308.2 Re Power Range Nuclear Instrumentation Updated
| ML17342A892 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 09/08/1987 |
| From: | Woody C FLORIDA POWER & LIGHT CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| L-87-371, NUDOCS 8709110269 | |
| Download: ML17342A892 (11) | |
Text
I REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR: 8709110269 DOC. DATE: 87/09/08.
NOTARIZED:
NO DOCKET 4 ACIL:50-250 Turkeg Point Planti Unit 3i Florida Power and Light C
05000250 50-251 Turkeg Point Planti Unit 4i Florida Power and Light C
05000251 AUTH. NAME AUTHOR AFFILIATION WOODY> C. O.
Florida Power Rc Light Co.
RECIP. NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Responds to violations noted in Insp Repts 50-250/87-33 5
50-251/87-33. Corrective actions:reducer 8c branch line to isolation valves 8c steam trap socolet replaced 5 Procedure ONOP-12308. 2 re power range nuclear instrumentation updated.
DISTRIBUTION CODE:
IEOID COPIES RECEIVED: LTR ENCL I SIZE:
TITLE: General (50 Dkt)-Insp Rept/Notice of Violation Response NOTES:
RECIPIENT ID CODE/NAME PD2-2 PD INTERNAL:= AEOD NRR MOR ISSEAUi D NRR/DREP/EPB NRR/DR IS DIR MANiJ t
REG FIL R
2 FILE 01 EXTERNAL:
LPDR NSIC COPIES LTTR ENCL 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 RECIPIENT ID CODE/NAME McDONALD' DEDRO NRR/DQEA DIR NRR/DREP/RPB NRR/PMAS/ ILRB OGC/HDS2 RES DEPY GI NRC PDR COPIES LTTR ENCL 2
2 1
1 1
1 2
2 1
1 1
1 1
1 1
1 TOTAL NUMBER QF COPIES REQUIRED:
LTTR 20 ENCL 20
P. O. BOX 14000, JUNO BEACH, FL 33408 0420 "0
SEPTEMBER 0 8
>987.
L-87-37I I 0 CFR 2.20 I U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gen tlemen:
Re:
Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-25 I Ins ection Re ort 87-33 Florida Power
- 6. Light Company has reviewed the subject inspection report and a response is attached.
There is no proprietary information in the report.
Very truly yours,
.O. Wo Group V'resident Nuclear nergy Department COW/SDF/gp Attachment cc:
Dr. J. Nelson Grace, Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, Turkey Point Plant 8709ii0269 870908 PDR ADOCK 05000250 9
PDR g(vI SDF I /056/ I an FPL Group company
ATTACHMENT RE:
TURKEY POINT UNITS 3 AND 4 DOCKET NOo 50-250, 50-251 IE INSPECTION REPORT 250-87-33
& 251-87-33 FINDING A:
10 CFR 50, Appendix B, Criterion XVI, as implemented by Florida Power and Light Topical Quality Assurance Report FPLTQAR 1-76A, Revision 10, and TQR 16.0, Revision 5, entitled Corrective Action, requires in part, that measures be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.
FPLTQAR 1-76A defines significant conditions adverse to quality as failures, malfunctions, deficiencies or deviations in material and equipment and other nonconformances which require engineering evaluation and/or evaluation for reportability as required by 10 CFR 50.55(e),
reportable occurrences (LERs) or 10 t
CFR 21 deficiencies.
Administrative procedure O-ADM-913, entitled Corrective Action for Conditions Adverse to Quality, revision dated July 15, 1986, specifies in section 5.3 that supervisors shall be alert to significant conditions adverse to quality when recommending or approving changes based on observed or reported discrepancies.
Contrary to the above, between July 14 and July 17, 1987, the licensee failed to report to appropriate levels of management the identification of a significant condition adverse to quality, the cause of the condition, and the corrective action taken.
Consequently, the safety significance of pinhole leaks in a weld on the train 1
AFW steam line was not promptly evaluated.
RESPONSE
1)
FPL concurs with the finding.
2)
The reason for the finding was due to several interrelated factors.
The Auxilary Feedwater Steam Supply drain line valve (AFSS-4-043) was identified in the Plant Work Order (PWO) that was issued to repair the leak.
The computerized PWO system, Nuclear Job Planning System (NJPS), fills in several fields on the PWO including the safety classification.
For the valve identified, that field was blank.
We also have a computerized list of Quality Safety Relapsed components known as the Q-List.
Valve AFSS-4-043 did not appear on the Q-List ~
Based on this, the valve was determined to be Non Nuclear Safety (NNS) and the PWO
"0 was treated as non safety related; not requiring engineering evaluation.
The PWO identified the leak to be a valve packing or piping leak.
The Planning and Q.C.
personnel focused on the valve and failed to realize the potential consequences of the leak being in a safety related section of pipe.
3) 4)
The safety class of the PWO was reevaluated and it was determined that the PWO should have been classified as safety related (SR).
A non conformance report was generated to initiate an engineering evaluation.
The source of leakage was identified as a pinhole in the weld of a 2"xl" reducer between the steam supply header and the steam trap isolation valve.
The reducer was replaced.
Also, the branch line to the isolation valves and the steam trap socolet were replaced.
Planning personnel have been told not to assume a safety class to be NNS when that component is not listed in the Q-List or the safety class field is blank in the NJPS.
The safety class is to be determined by the system which includes the component or by engine'ering evaluation.
When a
PWO addresses more than one possible deficiency, the most serious is to be pursued until proven otherwise.
5) 1)
Full compliance for item 3 above was achieved on July 20, 1987.
2)
Full compliance for item 4 above was achieved on July 18, 1987.
FINDING B:
t Technical Specification 3.2.6.h requires that, if the quadrant to average power tilt exceeds a value of 2 percent and the hot channel factors are not determined within two hours, then reactor power shall be reduced from rated power 2 percent for each percent of quadrant tilt.
Technical Specification 3.2.6.i requires that, if after a further period of 24
- hours, the quadrant to average power tilt is not corrected to less than 2 percent and the hot channel factors are not determined, then:
1) the Nuclear Regulatory Commission shall be notified; and 2) the overpower differential temperature (OPDT) and overtemperature differential temperature (OTDT) trip settings shall be reduced by the equivalent of 2 percent power for every 1 percent, quadrant to average power tilt.
Contrary to the above, between 12:Ol a.m.
on July ll and 7:05 p.m. on July 12,
- 1987, a period of approximately 43 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br />, the quadrant to average power tilt exceeded 2 percent, the hot channel f'actors were not determined and the OPDT and OTDT trip settings were not reduced.
RESPONSE
The reason for this event was personnel error mainly due to insufficient emphasis on the Technical Specification requirements in procedure ONOP-12308.2, Power Range Nuclear Instrumentation Verification of Upper, Lower and Channel Deviatior Alarms.
There was a perception by the operators on shift that the calculated tilt wa>>
erroneous.
This perception was based on the results of the flux map that had been
, run on July 9,1987.
This perception was further strengthened by problems that existec with the power range detector currents at the same time as this event.
(Additional details on this event are available in LER 251-87-018.)
- 3) a)
A flux map was performed which verified the Hot Channel Factors and Incore Power Tilt to be acceptable.
b) The power level was limited based upon the most conservative calculation of the gPTR so that overpower limits would not be violated.
- 4) a) Procedure ONOP-12308.2, Power Range Nuclear Instrumentation Verification of Upper, Lower and Channel Deviation Alarms, will be updated to address the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time limit for determining Hot Channel Factors and reducing the OP/OT DT setpoints.
b) This event will be addressed in the Licensed Operator Requalification classes.
- 5) a) Full compliance for Item 3.a above was achieved on July 12, 1987.
b) Full compliance for Item 3.b above was achieved on July 11, 1987.
c) Full compliance for Item 4.a above will be achieved by Nov. 11, 1987.
d) Full compliance for Item 4.b above will 'be achieved by Dec, 1,
1987.
FINDING C:
Technical Specification 6.8.1 requires that written procedures and administrative policies be established, implemented and maintained that meet or exceed the requirements and recommendations of sections 5.1 and 5.3 of ANSI N18.7-1972 and Appendix A of USNRC Regulatory Guide 1.33.
Appendix A of USNRC Regulatory Guide 1.33 states that procedures should be established for the operation of plant fire protection equipment.
Operating Procedure O-OP-016.1, Fire Protection Water Systems,'evision dated December 9,
1986, provides valve line-ups to control deluge systems.
Contrary to the above, procedure O-OP-016.1 was inadequate,, in that it did not control the positions of pressure switch isolation valves for deluge systems for, the unit 3 and 4 component cooling water pump rooms and the A and B emergency t
diesel generators.
Consequently, the pressure switch isolation valve for the Unit 3 component cooling water room was discovered to be isolated, preventing control room and local area deluge actuation alarms from functioning.
1)
FPL concurs with the finding.
2)
Some of the valves in the Fire Protection Water System have not been assigned specific identification numbers.
These valves are labeled as NNA (No Number Assigned) on the system drawing in the field.
Consequently, some of these valves were inadvertently omitted when the Fire Protection Water System Procedure (O-OP-016.1) was upgraded.
o' 3) 4)
The Fire Protection Water System Procedure (O-OP-016.1) has been reviseg to include the valves that were omitted when the procedure was upgraded.
A letter of instruction will be written to the procedure writers to ensure that all appropriate valves are included in the applicable procedures.
5) a) Full compliance for Item 3 was achieved by August 18, 1987.
b) Full compliance for Item 4 will be achieved by Sept.
15, 1987.