ML17333B038
| ML17333B038 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/15/1997 |
| From: | Fitzpatrick E INDIANA MICHIGAN POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-315-97-09, 50-315-97-9, 50-316-97-09, 50-316-97-9, AEP:NRC:1260H, NUDOCS 9709230028 | |
| Download: ML17333B038 (19) | |
See also: IR 05000315/1997009
Text
CATEGORY l REGULATORY
INFORMATION
DISTRIBUTION
SYSTEM (RIDS)DOCKET 05000315 05000316 NOTES: CCESSION NBR:9709230028
DOC.DATE: 97/09/15 NOTARIZED:
YES FACIL:50-315
Donald C.Cook Nuclear Power Plant, Unit 1, Indiana M , 50-316 Donald C.Cook Nuclear Power Plant, Unit 2, Indiana M AUTH.NAME AUTHOR AFFILIATION
FITZPATRICK,E.
Indiana Michigan Power Co.RECIP.NAME
RECIPIENT AFFILIATION
Document Control Branch (Document Control Desk)SUBJECT: Responds to NRC 970815 ltr re violations
noted in insp repts 50-315/97-09
a 50-316/97-09
on 970505-23.Corrective
actions: calculation
DC-D-1-SI-F101
revised a revised safety review incorporating
correct MDAFW pump start time approved.DISTRIBUTION
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0
indiana Michigan Power Company 500 Circle Orive Buchanan, Mi 491071395 INDIANA ItrIICHIGAN
PWER September 15, 1997 AEP:NRC:1260H
10'CFR 2.201 Docket Nos.: 50-315 50-316 U.S.Nuclear Regulatory
Commission
ATTN: Document Control Desk Washington, D.C.20555 Gentlemen:
Donald C.Cook Nuclear Plant Units 1 and 2 NRC INSPECTION
REPORTS NO.50-315/97009 (DRP)AND 50-316/97009 (DRP)REPLY TO NOTICE OF VIOLATION This letter is in response to a letter from G~E.Grant, dated August 15, 1997, that forwarded a notice of two violations
of NRC requirements
to Cook Nuclear Plant.The violations
were identified
during the operational
safety team inspection (OSTI)conducted by the NRC from May 5, 1997, to May 23, 1997.The attachment
contains our response to these violations.
Commitments
were made by Cook Nuclear Plant personnel to the NRC OSTI.The inspectors
identified
concerns related to valve descriptions
on labels, drawings, and in procedures, and they had concerns related to our program for sealed valves.The characterization
and detail of these commitments
in the inspection
report reflects our intent.when the commitments
were made.The second violation relates to the issue of design control.While reviewing the OSTI report and preparing this response, Cook Nuclear Plant underwent an NRC architect engineering (AE)team inspection.
The AE inspection
identified
design control issues, some that are similar to those cited in this violation.
Resolution
of the overall design control issue will require action beyond that which is committed to here.Those actions will be defined in the course of addressing
the AE team inspection
issues.The nuclear engineering
organization, along with our entire nuclear generation
group, understands
the importance
of error free human performance
and attention to detail, and to having a design basis that is clear, understandable, and retrievable.
We believe that a first step was taken on September 2, 1997, when standards for technical information
exchange and use in the nuclear engineering
9'709'230028
9709i5 PDR ADQCK 050003%5 8 PDR r.~~0,<iQ ,,s llllllllilllliIIIJlllllllllllliillilf
ill
U.S.Nuclear Regulatory
Commission
Page 2 AEP: NRC: 1260H organization
were formally established.
The objective of the guidance is to provide assurances
that technical information
is accurate, based on sound engineering
principles, properly conveyed, and properly, documented.
Sincerely, E.E.Fitzpatrick
Vice President SWORN TO AND SUBSCRIBED
BEFORE-ME THIS I'5 DAY OF~LV', 1997 Notary Public My Commission
Expires vlb Attachments
JAN WA%0N CAARYPQSC,BBNKNcoWn, e MYCOMMSSOM
EXPtRES FEL 10, 1999 c: A.A.Blind A.B.Beach MDEQ" DW&RPD NRC Resident Inspector J.R.Padgett
ATTACHMENT
TO AEP:NRC: 1260H REPLY TO NOTICE OF VIOLATION:
NRC INSPECTION
REPORT NOS.50-315/97009 (DRP)AND 50-316/97009 (DRP)
Attachment
to AEP:NRC:1260H
Page 1 On May 23, 1997, the NRC completed an operational
safety team inspection (OSTI)of Cook Nuclear Plant units 1 and 2 reactor facilities.
Two violations
of NRC requirements
were identified
during this inspection.
In accordance
with the 60 FR 34381,'General Statement of Policy and Procedures
for NRC Enforcement
Actions", dated June 30, 1995, the violations
and our responses are provided below.NRC Violation I"10 CFR 50, Appendix B, Criterion V,'Instructions, Procedures, and Drawings,'equires, in part, that activities
affecting quality be prescribed
by procedures
of a type appropriate
to the circumstances
and be accomplished
in accordance
with these procedures.
Contrary to the above, On May 10, 1997, the inspectors
identified
that safety related Temporary Modification (TM)1-95-1, which did not require an outage for restoration, had been assigned a (administrative)
date of August 27, 1996, but had not been made a permanent installation
through a design change or been removed as of May 10, 1997 as required by Plant Managers Procedure (PMP)5040.MOD.OO1,'Temporary
Modifications,'evision
7.B.On May 7, 1997, the inspectors
identified
that an activity affecting quality, the deenergization
of the DG2AB inverter, was completed without placing 2-DGAB-INV-CB2
and 2-DGAB-INV-
CB1 to off, contrary to steps 2.1.2 and 2.1.3 of procedure 02-OHP 4021.032.008
'Aligning DG2AB Subsystems
For Standby Operation,'evision
4, Attachment
7.This is a Severity Level IV violation (Supplement
I)." Res onse to NRC Violation I The letter from Mr.Grant, and the notice of viola'on, state that Cook Nuclear Plant's staff provided the NRC with satisfactory
information
regarding the reasons for the violation, and the corrective
actions taken and planned to correct both examples of the violation and prevent recurrence.
The letter goes on to say that, unless our corrective
actions or our position are not accurately
reflected in the inspection
report, we are not required to further respond to this violation.
We have reviewed inspection
report no.50-315/316)-97009
and determined
that it reflects the circumstances
of the examples in the violation as cited, as well as the corrective
actions taken for each.Notwithstanding, for consistency
in understanding, there is information
in the discussion
section of the inspection
report that we wish to clarify.Several condition reports (CRs)are listed as having been reviewed in conjunction
with'he 2AB EDG voltage regulator work.In particular, CR 97-1452 is listed with the title"Partial Clearance Addition Determined
As Root Cause for Blown Fuse On 2AB EDG Inverter." We would like to clarify that the failure to deenergize
the EDG inverter using the appropriate
procedure is not
0
Attachment
to AEP:NRC:1260H
Page 2 considered
to be the root cause for the inverter fuse blowing during reenergization.
Subsequent
investigation
has determined
that the failure to follow'the deenergization
procedure was recognized
before actions were taken to reenergize.
Prior to using the procedure to reenergize
the inverter and place it back in service, the inverter circuit alignment was corrected and the integrity of the fuse was verified.The fuse did blow coincident
with performance
of the procedure to reenergize
the inverter.NRC Violation ZZ"10 CFR 50, Appendix B, Criterion ZZZ, requires, in part, that mdasures be established
to assure that the design basis are correctly translated
into specifications, drawings, procedures, and instructions.
Design control measures shall provide for verifying or checking the adequacy of design.Contrary to the above, design control measures were not adequate to assure that the design basis was correctly translated
into design modification
documents:
b.On May 7, 1997, the inspectors
identified
that calculation
DC-D-1-SZ-F101,'Stress Analysis&Load Generation
for System 1-SI-F101 Per 12-MM-590,'sed
the wrong moment arm and had a missing reaction force and moment.On May 13, 1997, the inspectors
identified
that incorrect and non-conservative
design input was used for the motor-driven
auxiliary feedwater pump start time in the Safety Review Memorandum
f or the Setpoint Values f or the Time Delay Pickup Relays in the AFW Flow Retention Circuits, dated January 15, 1997, for design change package 12-DCP-0817,'Revise Aux.Feedwater Flow Retention Circuit.', On May 13, 1997, the inspectors
identified
that a calculation
for the seismic design adequacy of minor modification
12-MM-337 was not per formed.The design package for 12-MM-337 indicated this calculation
existed as DC-D-12-ES-116.
This is a Severity Level IV violation (Sur,v>lement
I)." Res onse to NRC Violation ZZ 1.Admission or Denial of the Violation We admit to violation ZI as cited in the NRC notice of violation.
Reasons for the Violation , The examples cited in the violation represent issues in the area of design control.Concurrent
with the review of the OSTI report and preparation
of this response, Cook Nuclear Plant underwent an NRC architect engineering (AE)team inspection.
This inspection
identified
design control issues, some that are similar to those cited in the violation.
Zt is recognized
that the three examples cited in this notice of violation must be
0
Attachment
to AEP:NRC: 1260H Page 3 considered
along with any new issues identified
by the AE team inspection, relative to the overall issue of design control.The circumstances
of each issue cited in this notice of violation are discussed below.Calculation
DC-D-1-SI-F101,"Stress Analysis and Load Generation
for System 1-SI-F101 per 12-MM-590", was performed in support of adding a permanent vent line to the safety injection system (SIS)piping.The inspector's
review of the design change package identified
errors made by the authors and overlooked
by the calculation
reviewers.
The errors are characterized
as insufficient
attention to detail on the part of the engineers performing
and reviewing the design change package.When calculating
reaction forces, the length value of a piping span, used as a moment arm in the calculation, was transcribed
from the input data presentation
to the actual algebraic presentation
incorrectly, from 37-5/8" to 35-5/8".This discrepancy
caused the maximum reaction force result to be incorrect, but in a conservative
direction.
In another section of the package, values for reaction force and moment were omitted from a summary format.These discrepancies
consisted of numbers correctly derived in the body of the calculation
on one page, but omitted from the summary on the following page.This problem was administrative
in nature;no incorrect information
was presented or used as a result.A third discrepancy
related to this design change package was noted in the body of the inspection
report, but not specified in the notice of violation.
The inspector made an observation
that incorrect design information
was stated in the safety review documentation.
Our investigation
concluded there was no discrepancy
in the safety review input information.
The values of design temperature
and pressure used by the safety reviewer were correct for the specific location where the new vent valve was to be installed.
The engineer performing
the design calculation
conservatively
used the highest bounding design temperature
and pressure for the SIS as a whole.This approach is often adopted when the inherent safety margin of a design is such that the more stringent design requirements
can be accommodated.
The thought process involved in taking this approach was not clearly documented
in the calculation
package.Design change 12-DCP-0817
was developed to add a time delay relay to the auxiliary feedwater (AFW)flow retention actuation circuit to prevent spurious actuation from momentary outlet pressure spikes, especially
those that occur when the AFW pumps
0
Attachment
to AEP:NRC:1260H
Page 4 automatically
start.A safety review was performed by the design engineering
organization
for the addition of the time delay pick-up relay to the AFW system circuits.The nuclear safety and analysis section was asked to perform an evaluation
of the setpoint value for the time delay relay.The review performed by this group was intended to demonstrate
that the magnitude of the time delay in the flow retention circuits would not adversely impact related accident analysis assumptions
or safety margins.In order to complete this review, the engineer needed to know how quickly the motor driven AFW pumps would start.An'incorrect
value of thirty seconds was used, based on a telephone conversation
with the AFW system engineer at the plant site.The system engineer communicated
that the turbine driven AFW (TDAFW)pumps start and come up to speed within thirty seconds.Surveillance
data on the TDAFW pumps was available on the system engineers desk at the time.What the system engineer intended was that thirty seconds would bound the start time on the motor driven pumps.Most often, in relation to safety analysis or T/S surveillance, the information
of concern is a time which bounds the pump start, time.However, the safety reviewer understood
that the thirty seconds would characterize
the start time for the motor driven auxiliary feedwater (MDAFW)pumps.Based on surveillance
measurements, the correct start time for the motor driven pumps is three seconds.Investigation
into the reason for this incorrect input to the safety review concluded that it was poor communications
between the involved engineers, and an incomplete
understanding
on the part of the system engineer as to the intended use of the information.
One engineer believed the bounding start time was needed, while the other was trying to determine the shortest start time for the pumps.Minor modification
12-MM-337 was performed to replace the emergency diesel generator (EDG)starting air system safety valves.The NRC inspector indicated that the calculation
for the seismic adequacy of the new valve type was not performed.
The design change package referenced
calculation
DC-D-12-ES-116.
This referenced
calculation
was not intended to follow the typical format in what was then the calculation
procedure.
DC-D-12-ES-116
was a record-keeping
and retrieval file for a number of individual
reviews prepared for the replacement
of non-identical
valves.The file did contain the final approval letter from the structural
design section documenting
that the valve change had been reviewed.However, we would have expected to find information
in this file related to
Attachment
to AEP:NRC:1260H
Page 5 the decision making process, such as isometric data, weight data, and support location information.
For unknown reasons, this file did not contain the information
that would have been expected pertaining
to the review of the valve replacement
of 12-MM-337.
The information
could not be found.It was reconstructed
and the new review was documented
appropriately.
The file now contains the appropriate
information
and review documentation (performed
in May 1997)that confirmed the conclusion
of the original design approval letter.Whether the file was lost, or the review never documented, this condition is characterized
as insufficient
attention to detail.It resulted in the inability to retrieve design data or design basis related information.
3.Corrective
Action Taken and Results Achieved b.c Calculation
DC-D-1-SI-F101, for the safety injection system stress analysis, was revised and approved on June 2, 1997, to address the cited discrepancies.
On May 29, 1997, a revised safety review, incorporating
the correct MDAFW pump start time was approved by the plant nuclear safety review committee (PNSRC).The conclusions
of the original safety review remained unchanged.
A walkdown and review of the valves installed under 12-MM-337, for the EDG starting air system, was performed on May 13, 1997, and documented
with the related condition report.This review confirmed the original conclusions
of the seismic qualification
review performed in 1992.The review was formally documented
on May 15, 1997.Corrective
Actions Taken to Avoid Further Violations
We understand
the importance
of"attention
to detail", and to having a design basis that is clear, understandable, and retrievable.
Each of the three cited examples in the NRC inspection
report refer to a lack of"attention
to detail", or a lack of clear communication
of design information.
The three examples of design control problems highlighted
in this violation will be considered
again as a part of the larger set of issues identified
by the NRC AE team inspection
of Cook Nuclear Plant.Resolution
of the overall design control issue will require action beyond that which is committed'in this response.Those actions will be defined in the course of addressing
the AE team inspection
issues.The violation examples a.and c.have been characterized
as insufficient
attention to detail.When the errors were identified
by the inspector, discussions
were held with the engineers in the design engineering
organization
who are involved in the development
and
0
Attachment
tq AEP:NRC:1260H
Page 6 documentation
of the calculations.
They were made aware of the inspection
findings and the importance
of attention to detail.This was accomplished
while the OSTI was still in progress.Training will be provided for personnel in the nuclear engineering
organization
who perform, review, and approve engineering
and design calculations.
The session will emphasize the importance
of"attention
to detail" and good calculation
control processes.
This training will be completed by December 31, 1997.In 1990, as a result of design verification
concerns raised during the safety system functional
inspection
of our essential service water system, quality review teams (QRTs)were established
to periodically
review design output documentation
for technical adequacy and procedural
compliance.
These teams were disbanded in 1996.The discrepancies
found under the QRT program had no impact on the conclusions
of the calculations.
Selected calculations
performed during the past year, August 1996, to August 1997, will be reviewed.The review will look for calculation
errors, inconsistencies, proper documentation
of assumptions, and procedure adherence.
Any findings will be addressed and documented
under the corrective
action program.This assessment
will be completed by December 1, 1997.The problem cited in example b.of the violation, incorrect data input to a safety review, has been identified
to be a communication
problem.On August 26, 1997, the nuclear safety and analysis section conducted a tabletop session that discussed the need for precision in the use of technical information
in safety reviews.It stressed that the use of written input is the preferred method, and that if verbal communication
is needed, it must be followed up with a written document.These standards for information
exchange and use were formally established
by procedural
direction issued on September 2, 1997.This document provides requirements
for nuclear engxneering
organization
personnel when providing technical direction.
The objective of the standard is to provide assurances
that the information
is accurate, based on sound engineering
principles, properly conveyed, and properly documented.
5.Date When Full Com liance Was Achieved Relative to the individual
examples cited in the violation, full compliance
was achieved: On June 2, 1997, when calculation
DC-D-1-SI-F101
was reviewed and approved for the safety injection system stress analysis.On May 29, 1997, when the revised safety review for the AFW flow retention time delay relay setpoint was approved by the PNSRC.
I 1
Attachment
to AEP:NRC:1260H
Page 7 On May 15, 1997, when walkdown and review of the seismic qualification
of the EDG starting air system safety valves was documented
and verified.