IR 05000482/1993014
| ML20056E704 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 08/19/1993 |
| From: | Beach A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Carns N WOLF CREEK NUCLEAR OPERATING CORP. |
| References | |
| NUDOCS 9308250085 | |
| Download: ML20056E704 (4) | |
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NUCLEAR REGULATORY COMMISSION o
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L I 9 1993 Docket: STN 50-482 License: NPF-42 Wolf Creek Nuclear Operating Corporation ATTN: Neil S. Carns, President and Chief Executive Officer P.O. Box 411 Burlington, Kansas 66839 SUBJECT: NRC INSPECTION REPORT 50-482/P3-14 Thank you for your letter dated August 13, 1993, in response to our letter and Notice of Violation dated July 14, 1993. We have reviewed your reply and find it responsive to the concerns raised in our Notice of Violation. Ve will review the implementation of your corrective actions during a future inspection to determine that full compliance has been achieved and will be maintained.
Sincerely, t-ire)to A. Bill Beach, Division of Re et P j, cts cc:
Wolf Creek Nuclear Operating Corp.
ATTN: Otto Maynard, Vice President Plant Operations P.O. Box 411 Burlington, Kansas 66839 Shaw, Pittman, Potts & Trowbridge ATTN: Jay Silberg, Esq.
2300 N Street, NW Washington, D.C.
20037 okohdgp l
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Wolf Creek Nuclear Operating-2-Corporation i
Public Service Commission j
ATIN:
C. John Renken Policy & Federal Department P.O. Box 360 Jefferson City, Missouri 65102 U.S. Nuclear Regulatory Commission ATTN:
Regional Administrator, Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137
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Wolf Creek Nuclear Operating Corp.
ATTN: Kevin J. Moles Manager Regulatory Services P.O. Box 411 Burlington, Kansas 66839 Kansas Corporation Commission ATTN: Robert Elliot, Chief Engineer Utilities Division i
1500 SW Arrowhead Rd.
l Topeka, Kansas 66604-4027 Office of the Governor State of Kansas Topeka, Kansas 66612 j
Attorney General 1st Floor - The Statehouse Topeka, Kansas 66612 Chairman, Coffey County Commission Coffey County Courthouse Burlington, Kansas 66839-1798
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Kansas Department of Health and Environment Bureau of Air Quality & Radiation Control
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I ATIN: Gerald Allen, Public Health Physicist Division of Environment Forbes Field Building 321 Topeka, Kansas 66620 l
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J. L. Milhoan Resident Inspector Section Chief (DRP/A)
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a WOLF CREEK
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Ned S " Surf Ca'ns August 13, 1993 P evoent ana io, Cn ef Executwe 0"icer c 10
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U. S. Nuclear Regulatory Commission
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ATTN: Document Control Desk Mail Station F1-137 Washington, D.
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20555 Reference:
Letter dated July 14, 1993 from A.
B.
Beach, NRC, to B. D. Withers, WCNOC Subject:
Docket No. 50-482.
Response to Violation 482/9314-02 Gentle:ren:
Atta0ned is Wolf Creek Nuclear Operating Corporation's (WCNOC) " Reply to a Notice of Violation" which was documented in the Reference.
Violation 482/9314-02 concerns two examples of failure to correctly transfer design data into the appropriate documents and computer datz bases.
If you have any questions concerning this matter, please contact me at (316)
364-8831 extension 4000 or Mr. Kevin ~. Moles at extension 4565 Very truly yours,
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Neil S.
Carns President and Chief Executive Officer NSC/jan Attachment cc:
W.
D. Johnson (lac), w/a J.
L. Milhoan (NRC), w/a G. A.
Pick (NRC), w/a W.
D.
Reckley (IGC), w/a l
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DDU %h P O Box 411/ Burlington. KS 66839 / Phone (316) 364 %31 An Eaual Opportunity Ernp6 oyer tWF/HC/ VET
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l Attachment to WM 93-0101
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Page 1 of 4 f
Reply to a Notice of Violation Violation (4B2/9314-02) :
Failure to Correctiv Transfer Desien Data Into the Arrrerriate Documents and C0meuter Data Bases
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Findinas:
"10 CFR 50, Appendix B, Criterion III, " Design Control," specifies, in part, that the design basis for those structures, systems, and components to which j
this appendix applies are correctly translated into specifications, drawings,
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procedures, and instructions.
Two examples of a failure to correctly transfer design data into the appropriate documents and computer data bases were identified.
1.
Wolf Creek Generating Station Cycle 7 Core Operating Limits Report, Revision 0,
Figure 2,
" Axial Flux Dif ference Limits as a Function of Rated Thermal Power," provided information for operating the reactor within acceptable limits, between 50 and 100 percent rated taermal power, in order to comply with Technical Specification 3.2.1.
When the axial flux difference monitor alarm is inoperable, Technical Specification Surveillance Requirement 4.2.1.1.b specifies that manual logging of the axial flux difference shall be performed each hour for
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l the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Contrary to the above, on June 3,
1993, licensee personnel determined that the NPIS software used to calculate the upper and lower AFD limits for Control Room Annunciator 79D was not updated to correspond with the changes in the limits following the sixth refueling cutage and prior to exceeding 50 percent power.
This action rendered the axial flux difference monitor alarm incperable since the outdated limits were nonconservative.
As a result, the manual logging of axial flux difference required by Technical Specification 4.2.1.1.b was not performed for 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />.
2.
Document WCRX-12, " Control Room Operating Curves and Tables Reference Manual," Revision 3, and the table on page 5.5,
" Integrated Boron Worth
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at BOL With ARI as a Function cf Boron Concentration and Temperature," at a burnup less than 7000 MWD /MTU, were based en 3565 megawatts rated thermal power.
Procedure STS RE-004, "Shutdcwn Margin Determination," Revision 13, provided guidance for performing the
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l shutdown margin determination and referenced tables in Document WCRX-12.
Centrary to the above, on June 16, 1993, licensee personnel determined that Procedure STS RE-004 referenced the table on 5.5 of Document WCRX-12 but failed to adjust the 100 percent rated thermal power to the cperating power level of 3411 megawatts rated thermal power.
Using the incorrect rated thermal power could have resulted in a nonconservative shutdown margin calculation."
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t Attachment to WM 93-0101
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Page 2 of 4 l
Reascn for Violation:
1.
On May 18, 1993, WCGS was coming out of the Sixth Refueling Outage with
the plant in Mode 1 at 49.5 percent reactor power and was commencing a power increase at a rate of 3 percent per hour.
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Amendment 61 to the WCGS Technical Specifications, dated March 30, 1993, discussed changes associated with Cycle 7 and included the use of the i
" Core Operating Limits Report" (COLR).
These changes included revised allowable Axial Flux Difference (AFD) limits as a function of thermal
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power.
Per Technical Specification (T/S) Surveillance Requirement 4.2.1.1 the AFD shall be determined to be within its limits during Power Operation above 50% of rated thermal power by either:
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Monitoring the indicated AFD for each CPERABLE excore channel at least once per 7 days when the AFD Monitor Alarm is OPERABLE and at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter restoring the AFD
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Monitor Alarm to OPERABLE status, or
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Monitoring and logging the indicated AFD for each OPERABLE excore I
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channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is
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In March 1993 reactor engineering received the COLR, which included changes to the AFD limits, for review and implementation. Se AFD limit changes were atypical of cycle changes. The reactor engineer performing the review noted that the Control Room Operating Curves and Tables Reference Manual would need to be revised as a result of the changes to the AFD limits for Cycle 7 that were identified in the COLR.
However, it was not identified that the Nuclear Plant Information System (NPIS)
software was affected and would need revision.
Thus, the root cause of this violation was an inadequate review of the COLR.
On June 3,
1993, a reactor erigineer identified that the NPIS software used to calculate the upper and lower AFD limits for Annunciator 79D was not updated to correspond with the changes in those limits per T/S Amendment 61.
A comparison of the AFD limits determined that the values were identical at 100% reactor thermal power. The limiting condition was calculated to be at 50% reactor thermal power, where the limits differed by 4% AFD on the negative side and 2t AFD on the positive side.
Therefore, the AFD Alarm was ineperable, when at reduced power, since startup from the sixth refueling outage.
From May 18, 1993, through June 3,
1993, with the plant in Mode 1 greater than 50% reactor power, T/S Surveillance Requirement 4.2.1.1.b was not satisfied.
Due to the failure to update the NPIS software, apprcximately 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> were spent at reduced power under this condition during power ascension for Cycle '
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Attachment to WM 93-0101 Page 3 of 4 2.
On June 16, 1993, during discussions between the Wolf Creek Generating Station (WCGS) reactor engineering group and the core design group, it was discovered that all tables and figures, within Curve Book WCRX-12,
" Control Room Operating Curves and Tables Reference Manual Lycle 7",
indicating in percent power, were based on the proposed power rerate value of 3565 MW as 100% reactor thermal power, when in fact the current NRC approved licensed reactor power limit was 3411 MWT.
At the time of this violation, WCGS was in preparation for a power rerate from 3411 MWT to 3565 MW, during the middle of Cycle 7.
As part of this rerate, the WCGS core design group revised the cure design report tables and curves utilizing 3565 MWT as full power for the entire cycle.
The rated thermal power before 200 effective full power days (EFPD) was anticipated to be 3411 MWT prior to rerate, and 3565 MWT after 200 EFPD.
When using power dependent tables and figures prior to 200 EFFD, 100% reactor power would be measured at 95.7% power.
It was believed by the reactor engineering group that the revised curves and figures of WCRX-12 reflected 3411 MWT at 100% power for the first 200 EFPD and 3565 MWT after 200 EFPD.
This was attributed to a breakdown in communications between the core design group and reactor j
engineering.
The incorrect information was referenced by Procedure STS I
RE-004, " Shutdown Margin Determination."
Because the power rating was interpretted to be 3411 MWT for the first 200 EFPD, the shutdown margin calculations, if performed, would have been incorrect and
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nonconservative.
Corrective Steos That Have Been Taken And Results Achieved:
1.
Upon identification that AFD limits were not updated, Control Room Annunciator 79D was declared inoperable and Control Room Operators commenced T/S Surveillance Requirements 4.2.1.1.b.
This was successfully completed on June 4, 1993.
Work Request 03848-93 and a computer software modification request were initiated to revise the NPIS software to the correct Cycle 7 AFD limits.
This was completed on June 4, 1993.
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Reactor engineering has developed a procedure which provides a checklist for use when determining implementation requirements for non-routine design changes.
The procedure also captures implementation specifics for changes (including the AFD limits) to assist in subsequent reviews.
This procedure was implemented August 12, 1993.
2.
Immediate actions included issuing a temporary change to procedure STS RE-004, " Shutdown Margin Determination," to provide a correction factor for power level.
On July 15, 1993, the Control Room Operating Curves and Tables Reference Manual was revised, such that data prior to power uprate, approximately 200 EFFD, is based on 100% reactor power being 3411 MWT.
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Attachment to WM 93-0101 Page 4 of 4 A revision to Procedure KP-C237, " Reload Design and Safety Evaluation,"
to require formal transmittal of proposed core design changes to reactor engineering for approval prior to implementation was approved on August 5,
1993.
This eliminated the potential for confusion of future changes
to the core design data.
Procedure KP-1537,
" Development and Control of Technical Guidelines /Tcpical Reports", was revised on August 5,
1993 to require
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that assumptiens/ conditions associated with data be explicitly stated.
This will ensure that the user is able to properly use the data.
Corrective Sters That Will Be Taken To Avoid Further Violations.
I All corrective actions associated with this violation have been completed.
Date When Full Comoliance Will Be Achieved:
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1.
Full compliance was achieved on August 12, 1993, with the issuance or i
revision of the associated procedures mentioned above.
Immediate j
corrective actions placed WCGS within the limits of Technical Specification requirements, and correctly updated plant NPIS software.
All design basis information has been correctly translated into as'sociated specifications, drawings, procedures, or instructions, and all planned corrective actions have been completed.
2.
Actions completed to be in full compliance included revising the Control Room Operating Curves and Tables Reference Manual, procedure KP-C237, and procedure KP-1537.
The Control Room Operating Curves and Tables Reference Manual was revised on July 15, 1993.
Full compliance was
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achieved on August 5, 1993 upon revision of procedures KP-C237 and KP-1537.
Actual er Potential Consecuences of this Violation:
1.
Since the AFD limits were correct at full power, Annunciator 79D was operable at full power.
Approximately 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> were spent above 50%
power where alarm limits were nonconservative.
At this time the measured AFD was well within the current required AFD limits.
Therefore, the probability of an accident or malfunction of equipment important to safety was not increased.
2.
If a shutdown margin calculation had been performed, the maximum calculated error would have occurred about 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after a trip from 85% power. At this time, Xenon would have been overestimated by 125 pcm and Samarium would have been overestimated by 2 pcm. Although it is not likely that a shutdown margin would have been performed for peak Xenon conditions, this total error of 127 pcm (20 ppm) is within the 100 ppm margin included in the shutdown margin calculation as a factor of safety.
Therefore, under worse case assumptions, shutdown margin requirements would not have been violated.
Based on the above, there were no adverse consequences to the plant during these conditions.
Plant safety and public health and safety were always assured.