ML20199G030
| ML20199G030 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 11/18/1997 |
| From: | Leach M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Graesser K COMMONWEALTH EDISON CO. |
| References | |
| 50-454-97-16, 50-455-97-16, NUDOCS 9711250075 | |
| Download: ML20199G030 (2) | |
See also: IR 05000454/1997016
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November 18 1997
Mr. K. Graesser
Site Vice President
Byron Nuclear Power Station
Commonwealth Edison Company
4450 North German Cnurch Road
Byron,IL 61010
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SUBJECT:
NOTICE OF VIOLATION (NRC INSPECTION REPORTS 50-454/97016(DRS);
50-455/97016(DRS))
Dear Mr. Graesser:
This will acknowledge receipt of your letter dated October 29,1997, in response to our letter
dated September 30,1997, transmitting a Notice of Violation associated with the failure to notify
the NRC conceming proceduralinadequacies with Byron emergency operating procedure (EOP)
BEP-3," Steam Generator Tube Rupture," at the Byron Nuclear Power Station. We have reviewed
your corrective actions and have no further questions at this time. These corrective actions will be
examined during future inspections.
Sincerely,
Original Signed b.y Melv.vn Leach
Melvyn Leach, Chief
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Operator Licensing Branch
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Docket Nos. 50-454; 50-456
License Nos. NPF-37; NPF 66
Enclosure:
Ltr dtd 10/29/97 K. L. Graesser
Byron to USNRC
See Attached Distribution
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OFFICIAL RECORD COPY
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K. Graesser
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November 18, 1997
cc w/o encl:
O. Kingsley, Nuclear Generation Group
President & Chief Nuclear Officer
M. Wallace, Senior Vice President,
Corporate Services
H. G. Stanley, Vice President,
PWR Operations
Liaison Officer, NOC-BOD
D. A. Sager, Vice President,
Generation Support
D. Farrar, Nuclear Regulatory
Services Manager
1. Johnson, Licensing Operations Manager
Document Control Desk - Licensing
K. Kofron, Station Manager
D. Brindle, Regulatory Assurance
Supervisor
cc w/ encl:
Richard Hubbard
Nathan Schloss, Economist,
Office of the Attomey General
State Liaison Officer
State Liaison Officer, Wisconsin
Chairman, Illinois Commerco Commission
Dhkik.utiQD:
Docket File w/o enct
SRI, Byron w/o enci
. TSS w/o enci
PUBLIC IE 01 w/o encl
LPM, NRR w/o enct
CAA1 w/o enci
OCFO/LFARB w/o enci
A. B. Beach, Rlll w/o encI
DOCDESK w/o enci
DRP w/o enci
J. L. Caldwell, Rlli w/o encl
DRS w/o enct
Rlli Enf. Coordinator w/o enct
Rill PRR w/o enci
R. A. Capra, NRR w/o enci
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Commonwealth Ihon Company
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thwn Generating Station
4 441i Ninh t errnun C.hurch Ibud
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October 29,1997
LTR:
FILE:
1.10.0101
U.S. Nuclear Regulator,
.nmission
Washington, DC 20555
Attention:
Document Control Desk
Subject:
Byron Nuclear Power Station Units 1 and 2
Response to Notice of Violation
Inspection Report No. 50-454/97016; 50-455/97016
NRC Docket Numbers 50-454, 50-455
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Reference:
John A. Grobe letter to Mr. Graesser dated
September 30,1997, transmitting NRC Inspection
Report 50454/97016; 50-455/97016
Enclosed is Commonwealth Edison Company's response to the Notice of Violation (NOV)
which was transmitted with the referenced letter and Inspection Report. The NOV cited
one (1) Severity Level IV violation requiring a written response. Comed's response is
provided in the attachment.
This letter contains the following commitments:
1)
An LER will >e written in accordance with 10CFR50.73(a)(2)(v) & (vi)
documenting the concern
2)
The Reportability Manual will be revised to clarify the requirements for
reporting ofprocedural problems.
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Byron Ltr. 97-0242
October 29,1997
Page 2
Ifyour staff has any questions or comments concerning this letter, please refer t
Don Brindle, Regulatory Assurance Supervisor, at (815) 234-5441 ext. 2280.
Respectfully,
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K. L. Grae
Site Vice President
Byron Nuclear Power Station
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Attachment (s)
A. B. Beach, NRC Regional Administrator - RIII
cc:
G. F. Dick Jr., Byron Project Manager - NRR
Senior Resident inspector, Byron
R. D. Lanksbury, Reactor Projects Chief- RIII
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F. Niziolek, Division of Engineering - IDNS
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ATTACilMENT 1
VIOLATION (454/455 97016-01)
Code of Federal Regulations Title 10 Part 50.72(b)(2)(iii) states, in part, that licensees
shall notify the NRC when practical and in all cases, within four hours of the occurrence,
"Any event or condition that alone could have prevented the fulfillment of the safety
function of structures or systems needed to: (D) httigate the consequences of an
accident."
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Code of Federal Regulations Title 10 Part 50.73(a)(2)(v) states, in part, that licensees shall
report, "Any event or condition that alone could have prevented the fulfillment of the
safety function of structures or systems that are needed to: (D) Mitigate the consequences
of an accident."
Code of Federal Regulations Title 10 Part 50.73(a)(2)(vi) states that, " Events covered in
paragraph (a)(2)(v) of this section may include one or more personnel errors, equipment
failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural
inadequacies "
Contrary to the above, the inspector identified that the licensee on February 19,1996, had
failed to notify the NRC concerning procedural inadequacies with Byron emergency
operating procedure (EOP) BEP-3, " Steam Generator Tube Rupture," and functional
restoration procedure BFR P.1, " Response to Imminent Pressurized Thermal Shock
Condition," which could limit operator response such that the EOP operator response time
limits documented in the Updated Final Safety Analysis Report may not be met.
This is a Severity Level IV violation (Supplement I).
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REASON FOR THE VIOLATION
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The original concern with the EOPs was identified on Feb. 19,1996 in Problem
Identification Form (PIF) 454 201 96-0298 written by the EOP Procedure Writer. The
concern was based on an Emergency Response Guideline (ERG) Direct Work Request
(DW) DW 89-077 that had been submitted to the Westinghouse Owners Group (WOG)
on November 10,1989 by Turkey Point. The EOP Procedure Writer had been reviewing
the WOG response to the DW and felt that the response was inadequate. He also felt that
the issue perteined to Byron Station, and therefore wrote the PIF. The DW postulated
that under some Steam Generator Tube Rupture (SGTR) scenarios it was possible that
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cold Safety Irsection (SI) flow cor'd flow backwards through the Reactor Coolant System
(RCS) loop and out the SG break if the Reactor Coolant Pumps (RCPs) were not running.;
If this were to occur, it was further postulated that the cold SI flow might be such that the
wide range RCS loop thermocouples would indicate a temperature lower than that used as
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. criteria for entry into Byron Functional Response procedure BFR-P.1, Response to-
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~ Imminent Pressurized Thermal Shock Condition. Byron Emergency.*Nocedure BEP 3,
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Steam Generator Tube Rupture, did not provide any guidance to the opomors that the
indication oflow temperature in the ruptured RCS loop was expected and that Pressurized
Thermal Shock (PTS) was not a concern Thus, if the operators exited BEP-3 in order to
follow BRP P.1, it was then postulated that they might use up so much time responding t
BFR P,1 that by the time they came back to BEP 3, the SG would have overfilled before
they had equalized RCS and SG pressurcs. Byron's Design Basis does not include overfill
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of the SG's. Therefore, the issue was one of muting the operator response time to SGTR
assumed in the UFSAR. The DW response stated that "For most plants, it is expected
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that for a SGTR with RCP: tripped, the operator will remain in E 3 to properly respond to
the SGTR."
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The original PIF was reviewed for reportability by the Shift Manager. The Shift Manager
wrote that the PIF was "Not an operability issue at this time pending further review.
Concern involves response in the emergency procedures and should be addressed by
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WOG." The Event Screening Committee also believed it was a generic WOG issue and
assigned the PlF to the Emergency Procedure writer for resolution as a Level 4 (Apparent
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Cause Evaluation) PIF. The RA Supervisor consulted the Comed Reportability Manual.
Section SAF 1.17 of the manual addresses 10CFR50.72(b)(2)(iii) and
10CFRSO.73(a)(2)(v) & (vi). Procedure problems are covered by the following
paragraph:
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A plant procedure, approved but not yet used, that' has an
error which would cause a safety system to become
inoperable would be reportable. -If the error was
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discovered before the procedure was approved, it would
not be reportable.
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The root cause of the violation was the inappropriate decision made by the Regulatory
Assurance Supervisor. The procedure problem did not "cause a safety system to become
inoperable " It appeared that several circumstances would have had to occur in order to
not meet the desy basis. but none of them included making a system inoperable. The
issue was also discussed with the RA Supenisor at Braidwood, who concurred with the
Byron RA Supenisors position. On 3/15/96, based on another review of the significance
level, the PlFs significance level was raised to Level 3 (Root Cause Repon), however it
was still believed that the event was not reportable.
At that time, Byron Station was not aware of any other utilities that had made an NRC
notification of the concem. Therefore, the Byron RA Supenisor made the decision that
the issue was not reportable.
The Byron Emergency Procedure Writer was not satisfied with the WOG response to
DW 89-077. On May 2,1996, he attended the WOG Operations Subcommittee meeting
and again raised the concem with the DFR P.1 issue. The WOG Ops Subcommittee
requested that he write another DW, which he did (DW 96-028). The DW was issued as
Category 4 (Feedback to provide clarification or improve guidance (NOT to correct an
error)). In early 1997 the WOG authorized a program to investigate operator action
times. On Feb. 28,1997, the WOG Ops Subcommittee issued a letter (OG 97-021) to
applicable utilities asking them to run selected time critical scenarios on the simulator,
video taping them if possible, and to provide the results of the operator response back to
the WOG. Byron and Braidwood both responded to the request. The WOG is reviewing
the various responses and expects to issue a fmal report by the end of 1997. The second
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DW is still open.
CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED
in response to the originalissue with the BEP 3 and DFR P.1 procedures, it was realized
that there might be other operator response times assumed in the UFSAR that had not
been validated. Therefore, a Task Force was chartered to review those operator response
times that were known at that time. The Task Force has reviewed 18 items since, with 16
items being closed out. Most of them have not been determined to be of concern.
liowever, several of them will be periodically validated to ensure that operators continue
to meet the UFSAR assumed times.
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As pan of the Task Force response to the issue of DEP 3 and BFR P.1, a procciute
change was made to BEP-3 that informed the operators that a low temperature condition
was expected on the ruptured loop, and that ifit occurred, to NOT go to BFR P.I. If the
operators do not go to BFR P.1, they have time to equalize RCS and 50 pressures prior
to SG overfill.
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The RA Supenisor has been counseled on conservative decision making with respect to
ENS notifications.
The Byron PA Supenisor has discussed this issue with the Braidwood RA Supervisor.
CORRECTIVE STEPS THAT WILL BE TAKEN 10 AVOID FURTHER VIOLATION
An LER will be written in accordance with 10CFR50.73(a)(2)(v) & (vi) documenting the
concern This is tracked by NTS item # 454 100-97-01601-01.
The Rep artability Manual will be revised to clarify the requirements for reporting of
procedural problems. This is tracked by NTS itera# 454 100 97-01601 02.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED
Full compliance will be achieved 30 days from the date of this letter when the LER will be
submitted to the NRC.
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