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{{#Wiki_filter: | {{#Wiki_filter:- | ||
l. | |||
, | |||
#pesog.,og UNITED STATES | |||
. 8 p, NUCLEAR REGULATORY COMMISSION | |||
5 <j WASHsNGTON, D. C. 206H | |||
,o | |||
o,..+ February 8, 1968 | |||
Docket No. 50-482 | |||
Wolf Creek Nuclear Operating Corporation | |||
ATTN: Bart D. Withers, President | |||
and Chief Executive Officer | |||
P. O. Box 411 Burlington, Kansas 66839 | |||
Gentlenen: | |||
SUB4ECT: SAFETY SYSTEMS OUTAGE MODIFICATIONS INSPECTION | |||
50-482/87032 | |||
This letter forwards the results and conclusions of the Safety Systems Outage | |||
Modifications Inspection (550MI) at the Wolf Creek nuclear power station | |||
conducted by the NRC's Office of Nuclear Reactor Regulation. The inspection | |||
team was composed of NRC personnel and consultants. The design and procurement | |||
portion of the inspection was conducted November 2-13, 1987, and the installa- | |||
tion and testing portion of the inspection was condt:cted November 9-20, 1987. | |||
The purpose of the design and procurement portion of the SS0MI was to deter- | |||
mine, through an examination of specific work packages, that the design, | |||
engineering, and procurement control was adequate to support the safety-related | |||
modifications and to determine whether services or products acquired to support | |||
the outage were in accordance with your comitments and regulatory | |||
requirements. | |||
The purpose of the installation and test portion of the SSOMI was to determine. l | |||
through an examination of specific work packages, that installetion of the t | |||
selected modifications conformed to design and installation requirements, and to i | |||
verify that the repaired or modified components and systems have the required { | |||
operating configurations and have been adequately tested to ensure that they i | |||
are capable of safely performing their intended functions. i | |||
The inspection team identified significant weaknesses in the areas reviewed | |||
relating to the the adequacy of management control and oversight, engineering | |||
support and engineering evaluations, and corrective actions. Those weaknesses I | |||
are discussed in Appendix A to this letter. | |||
At the conclusion of the inspection, a number of equipment operability concerns I | |||
remained to be resolved prior to unit startup from the refueling outage. | |||
Specifically, the operability of the Control Room Ventilation Isolation System | |||
had not been demonstrated in all anticipated modes of operation, and the single | |||
failure det.1gn of the system had been compromised by an equipment modification; | |||
the operability of the pressurizer safety valves had not been adequately | |||
demonstrated by periodic testing; the pressurizer spray valve had been incor- | |||
rectly modified; several loose and missing piping supports were identified; and | |||
inadequacies were identified with respect to the design of the anti-pumping | |||
logic for the diesel generater output breakers. | |||
8802160005 880208 | |||
PDR ADOCK 05000482 | |||
G PDR | |||
_ _ _ _ _ _ _ _ _ | |||
t t | |||
' | |||
Bart D. Withers -2- February 8, 1988 | |||
The NRC Region IV staff monitored your corrective actions, and detemined that | |||
adequate corrective resolution of the identified concerns was achieved prior to | |||
restart of the plant. Some of the items identified by the team may be poten- | |||
tial enforcement findings. Any enforcement actions will be identified by | |||
Region IV in separate correspondence. | |||
In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure | |||
will be placed in the NRC Public Document Room. | |||
You are requested to respond to this oifice within 60 days regarding the | |||
concerns and weaknesses identified in the enclosed inspection report. Your | |||
response should include a discussion of the role of the Wolf Creek organization | |||
responsible for the assurance of quality. | |||
Should you have any questions concerning this inspection, please contact me or | |||
Mr. J. E. Konklin (301-492-0953) . | |||
Sincerely, | |||
' | |||
><.$'7 | |||
Dennis M. C >rutchfield | |||
e | |||
, Director Division of l | |||
Reactor Projects, III/IV/V and Special l | |||
Projects | |||
Office of Nuclear Reactor Regulation j | |||
Enclosure: Inspection Report No. 50-842/87032 | |||
cc w/ enclosure: See next page | |||
; | |||
1 | |||
l | |||
l | |||
l | |||
. | |||
. . | |||
I 1 | |||
* | |||
Bart D. Withers -3- February 8, 1988 | |||
l | |||
cc: | |||
Otto L. Maynard, Manager of Licensing | |||
Wolf Creek Nuclear Operating Corporation , | |||
P. O. Box 411 | |||
Burlington, Kansas 66839 | |||
Gary Boyer, Plant Manager | |||
Wolf Creek Nuclear Operating Corporation ' | |||
P. O. Box 411 | |||
Surlington, Kansas 66839 | |||
Mr. Robert D. Elliott, Chief Engineer | |||
Kansas Corporation Commission | |||
Fourth Floor, Docking State Office Building | |||
(opeka, Kansas 66612-1571 | |||
Kansas Radiation Control Program Director | |||
1 | |||
> , | |||
l | |||
l | |||
- | |||
l | |||
! | |||
! ' | |||
i | |||
1 | |||
l | |||
. . _ . . -- _ , | |||
~ ~ | |||
* | |||
^l , l' L] | |||
. | |||
3 '9' | |||
~Bart D.' Withers o- -4- : ;. . February 8, 1988 | |||
t i | |||
Distribution: | |||
DCS | |||
POR | |||
LPDR | |||
RSIB R/F | |||
DRIS R/F. | |||
CHaughney ' | |||
JKonklin | |||
' | |||
! | |||
LNorrholm | |||
JPartlow | |||
BGrimes ' | |||
JSniezek | |||
P0' Conners | |||
DCrutchfield | |||
PNoonan | |||
ABeach | |||
-JCalvo | |||
Regional Administrators | |||
Regional Division Directors | |||
CVandenburgh | |||
ACRS (10) | |||
GPA (3) | |||
15 Dist. (All Utility Licensees) | |||
SRI, Wolf Creek | |||
TMartin, ED0 | |||
, | |||
- | |||
A 2 | |||
0FC :RSIB:DRIS:NR :RSIB:DRIS:NRR:RSI 5:NRR:DD RR:D:DR : 4RR :U:DR | |||
, | |||
................. | |||
.:... g....... ....... .......a ).....;....P4455 3 | |||
....... . 6'.:........ | |||
NAME :CVandenbur :J klin : ney :B :JPar 1 w :DCrutchfield: | |||
... ........... | |||
.....:........ | |||
. . . . .01/li/88 | |||
DATE . . . . . . . . .$01/A0/88 | |||
. . . . . . . . . .$01/*#/88 | |||
. . . . . . . . $. .61/2/88 | |||
. . . . . . . $b | |||
. . . . . . . . . :@/ . . . . .88. . . .: . . . k / 3 / | |||
i | |||
6 | |||
F | |||
. | |||
. i | |||
, | |||
APPENDIX A | |||
. | |||
EXECUTIVE SUMMARY ! | |||
l | |||
An announced NRC Safety Systems Outage Modifications Inspection (SSOMI) was | |||
conducted at the Wolf Creek Nuclear Operating Corporation's Wolf Creek Generat- | |||
ing Station during the period of November 2-13, and November 9-20, 1987. | |||
In addition to the inspection of activities involved in this specific outage, | |||
the SS0MI team also reviewed recent Wolf Creek operational events in order to | |||
evaluate the root causes as they relata to the performance of safety system | |||
modifications. The results and conclusions of this review were discussed with | |||
NRC regional management and will be utilized in Region IV's review of the | |||
events. | |||
Overall Conclusions | |||
The modifications activities inspected by the SSOMI team during the Wolf Creek | |||
outage, including procedures, installed equipment and materials, and workman- | |||
ship by crafts, were generally in accordance with NRC requirements and licensee | |||
commitments. The SSOMI team noted specific strengths related to the acquisi- | |||
tion and control of equipment and materials, the trend analysis of quality | |||
findings and reported deficiencies, and workmanship by maintenance personnel. | |||
However, the team also identified weaknesses in the following areas: | |||
1. Manacement Controls | |||
In a number of cases, management failed to implement the appropriate opera- | |||
tional procedures for the removal and return to service of equipment. The | |||
outage management controls specified in Administrative Procedure ADM 01-108, | |||
"Outage Planning," which provided definitive guidance on the planning, schedul- | |||
ing and performance of major outages were not implemented. Identified defi- | |||
ciencies which impacted the ability of maintenance crafts to perform quality | |||
work, such as wiring discrepancies between "as-built" and vendor's wiring | |||
diagrams, were not promptly resolved. Inadequate maintenance management | |||
involvement was provided for complex tasks such as safety valve bench testing | |||
and was partially responsible for a Quality Assurance Work Hold issued during | |||
repairs to piping in the Essential Service Water System. In addition, modifi- | |||
cation of the Pressurizer Spray Valve for liquid sealant injection, a temporary | |||
modification of the Control Room Ventilation Isolation System (CRVIS) which | |||
defeated the single failure design of the system, and the failure to perform | |||
timely evaluations of operational piping systems with potential wall thickness | |||
problems were further examples of inadequate management support and control of | |||
outage activities. ; | |||
With regard to the recent operational events, the SSOMI team noted that, during | |||
the removal of Vital Bus NB02 from service for scheduled maintenance on | |||
October 14, 1987, the system operating procedures which specified the require- | |||
ments and precautions for system operation and isolation, including the maximum ; | |||
, time the isolated buses could be supplied by the station batteries, were not ! | |||
used. The failure to utilize the operational procedures and to intorporate the l | |||
precautions and requirements of the procedures for the removal and return of | |||
equipment from service in accordance with the requirements of the Technical | |||
! | |||
A-1 l | |||
, | |||
l | |||
l | |||
, | |||
,,e | |||
. i | |||
, | |||
, Specifications and 10 CFR 50, Appendix A, resulted in a chain of events which | |||
culminated in the injection of lake water into the steam generators. | |||
2. Engineering Support and Evaluations | |||
The engineering support provided for a number of recent modifications and , | |||
maintenance activities was found to be inaccurate or lacking in thoroughness. | |||
The SS0MI team identified a number of cases in which engineering evaluations | |||
failed to correctly determine the effects of proposed modifications. Examples | |||
include a wrong estimate of the time of discharge of Battery NK12 prior to the | |||
loss of Vital Bus NB02, failure to prepare for the loss by providing alternate | |||
supplies, failure to provide adequate overpressure protection for the Reactor | |||
Coolant Drain Tank, failure to recognize the single failure design criteria of ; | |||
the CRVIS system, inadequate design of a diesel generator output breaker - | |||
anti pumping logic which prevented the breaker from closing onto a cleared, | |||
deenergized bus, and a number of inadequately justified or documented engineer- | |||
ing evaluations in modification packages. | |||
3. Corrective Actions | |||
Although, as noted above, the licensee has a good trend analysis program, the | |||
SSOMI team identified a weakness involving the adequacy of corrective actions | |||
for identified deficiencies, including the identification of root causes, | |||
evaluation of related areas for similar deficiencies, and actions to prevent | |||
recurrence. One significant example was identified during the evaluation of | |||
PMR 1903, which involved Essential Service Water pipe wall thinning and | |||
through-wall corrosion. After the deficiency was identified, there appeared to | |||
be no attempt to check for similar deficiencies on the other train, there was | |||
no immediate evaluation of the deficiency to determine whether the thinner | |||
walls violated the Updated Safety Analysis Report commitments, and the correc- | |||
tion of the deficiency did not include a determination of a root cause and | |||
specification of actions to prevent recurrence. | |||
l | |||
, | |||
l | |||
1 | |||
j | |||
l | |||
l | |||
r | |||
. | |||
l | |||
A-2 | |||
. | |||
---_-w---w i- wyF- | |||
}} | }} |
Latest revision as of 08:01, 13 November 2020
ML20196C178 | |
Person / Time | |
---|---|
Site: | Wolf Creek ![]() |
Issue date: | 02/08/1988 |
From: | Crutchfield D Office of Nuclear Reactor Regulation |
To: | Withers B WOLF CREEK NUCLEAR OPERATING CORP. |
Shared Package | |
ML20196C181 | List: |
References | |
NUDOCS 8802160005 | |
Download: ML20196C178 (6) | |
See also: IR 05000482/1987032
Text
-
l.
,
- pesog.,og UNITED STATES
. 8 p, NUCLEAR REGULATORY COMMISSION
5 <j WASHsNGTON, D. C. 206H
,o
o,..+ February 8, 1968
Docket No. 50-482
Wolf Creek Nuclear Operating Corporation
ATTN: Bart D. Withers, President
and Chief Executive Officer
P. O. Box 411 Burlington, Kansas 66839
Gentlenen:
SUB4ECT: SAFETY SYSTEMS OUTAGE MODIFICATIONS INSPECTION
50-482/87032
This letter forwards the results and conclusions of the Safety Systems Outage
Modifications Inspection (550MI) at the Wolf Creek nuclear power station
conducted by the NRC's Office of Nuclear Reactor Regulation. The inspection
team was composed of NRC personnel and consultants. The design and procurement
portion of the inspection was conducted November 2-13, 1987, and the installa-
tion and testing portion of the inspection was condt:cted November 9-20, 1987.
The purpose of the design and procurement portion of the SS0MI was to deter-
mine, through an examination of specific work packages, that the design,
engineering, and procurement control was adequate to support the safety-related
modifications and to determine whether services or products acquired to support
the outage were in accordance with your comitments and regulatory
requirements.
The purpose of the installation and test portion of the SSOMI was to determine. l
through an examination of specific work packages, that installetion of the t
selected modifications conformed to design and installation requirements, and to i
verify that the repaired or modified components and systems have the required {
operating configurations and have been adequately tested to ensure that they i
are capable of safely performing their intended functions. i
The inspection team identified significant weaknesses in the areas reviewed
relating to the the adequacy of management control and oversight, engineering
support and engineering evaluations, and corrective actions. Those weaknesses I
are discussed in Appendix A to this letter.
At the conclusion of the inspection, a number of equipment operability concerns I
remained to be resolved prior to unit startup from the refueling outage.
Specifically, the operability of the Control Room Ventilation Isolation System
had not been demonstrated in all anticipated modes of operation, and the single
failure det.1gn of the system had been compromised by an equipment modification;
the operability of the pressurizer safety valves had not been adequately
demonstrated by periodic testing; the pressurizer spray valve had been incor-
rectly modified; several loose and missing piping supports were identified; and
inadequacies were identified with respect to the design of the anti-pumping
logic for the diesel generater output breakers.
8802160005 880208
PDR ADOCK 05000482
G PDR
_ _ _ _ _ _ _ _ _
t t
'
Bart D. Withers -2- February 8, 1988
The NRC Region IV staff monitored your corrective actions, and detemined that
adequate corrective resolution of the identified concerns was achieved prior to
restart of the plant. Some of the items identified by the team may be poten-
tial enforcement findings. Any enforcement actions will be identified by
Region IV in separate correspondence.
In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure
will be placed in the NRC Public Document Room.
You are requested to respond to this oifice within 60 days regarding the
concerns and weaknesses identified in the enclosed inspection report. Your
response should include a discussion of the role of the Wolf Creek organization
responsible for the assurance of quality.
Should you have any questions concerning this inspection, please contact me or
Mr. J. E. Konklin (301-492-0953) .
Sincerely,
'
><.$'7
Dennis M. C >rutchfield
e
, Director Division of l
Reactor Projects, III/IV/V and Special l
Projects
Office of Nuclear Reactor Regulation j
Enclosure: Inspection Report No. 50-842/87032
cc w/ enclosure: See next page
1
l
l
l
.
. .
I 1
Bart D. Withers -3- February 8, 1988
l
cc:
Otto L. Maynard, Manager of Licensing
Wolf Creek Nuclear Operating Corporation ,
P. O. Box 411
Burlington, Kansas 66839
Gary Boyer, Plant Manager
Wolf Creek Nuclear Operating Corporation '
P. O. Box 411
Surlington, Kansas 66839
Mr. Robert D. Elliott, Chief Engineer
Kansas Corporation Commission
Fourth Floor, Docking State Office Building
(opeka, Kansas 66612-1571
Kansas Radiation Control Program Director
1
> ,
l
l
-
l
!
! '
i
1
l
. . _ . . -- _ ,
~ ~
^l , l' L]
.
3 '9'
~Bart D.' Withers o- -4- : ;. . February 8, 1988
t i
Distribution:
POR
LPDR
RSIB R/F
DRIS R/F.
CHaughney '
JKonklin
'
!
LNorrholm
JPartlow
BGrimes '
JSniezek
P0' Conners
DCrutchfield
PNoonan
ABeach
-JCalvo
Regional Administrators
Regional Division Directors
CVandenburgh
ACRS (10)
GPA (3)
15 Dist. (All Utility Licensees)
SRI, Wolf Creek
TMartin, ED0
,
-
A 2
0FC :RSIB:DRIS:NR :RSIB:DRIS:NRR:RSI 5:NRR:DD RR:D:DR : 4RR :U:DR
,
.................
.:... g....... ....... .......a ).....;....P4455 3
....... . 6'.:........
NAME :CVandenbur :J klin : ney :B :JPar 1 w :DCrutchfield:
... ...........
.....:........
. . . . .01/li/88
DATE . . . . . . . . .$01/A0/88
. . . . . . . . . .$01/*#/88
. . . . . . . . $. .61/2/88
. . . . . . . $b
. . . . . . . . . :@/ . . . . .88. . . .: . . . k / 3 /
i
6
F
.
. i
,
APPENDIX A
.
EXECUTIVE SUMMARY !
l
An announced NRC Safety Systems Outage Modifications Inspection (SSOMI) was
conducted at the Wolf Creek Nuclear Operating Corporation's Wolf Creek Generat-
ing Station during the period of November 2-13, and November 9-20, 1987.
In addition to the inspection of activities involved in this specific outage,
the SS0MI team also reviewed recent Wolf Creek operational events in order to
evaluate the root causes as they relata to the performance of safety system
modifications. The results and conclusions of this review were discussed with
NRC regional management and will be utilized in Region IV's review of the
events.
Overall Conclusions
The modifications activities inspected by the SSOMI team during the Wolf Creek
outage, including procedures, installed equipment and materials, and workman-
ship by crafts, were generally in accordance with NRC requirements and licensee
commitments. The SSOMI team noted specific strengths related to the acquisi-
tion and control of equipment and materials, the trend analysis of quality
findings and reported deficiencies, and workmanship by maintenance personnel.
However, the team also identified weaknesses in the following areas:
1. Manacement Controls
In a number of cases, management failed to implement the appropriate opera-
tional procedures for the removal and return to service of equipment. The
outage management controls specified in Administrative Procedure ADM 01-108,
"Outage Planning," which provided definitive guidance on the planning, schedul-
ing and performance of major outages were not implemented. Identified defi-
ciencies which impacted the ability of maintenance crafts to perform quality
work, such as wiring discrepancies between "as-built" and vendor's wiring
diagrams, were not promptly resolved. Inadequate maintenance management
involvement was provided for complex tasks such as safety valve bench testing
and was partially responsible for a Quality Assurance Work Hold issued during
repairs to piping in the Essential Service Water System. In addition, modifi-
cation of the Pressurizer Spray Valve for liquid sealant injection, a temporary
modification of the Control Room Ventilation Isolation System (CRVIS) which
defeated the single failure design of the system, and the failure to perform
timely evaluations of operational piping systems with potential wall thickness
problems were further examples of inadequate management support and control of
outage activities. ;
With regard to the recent operational events, the SSOMI team noted that, during
the removal of Vital Bus NB02 from service for scheduled maintenance on
October 14, 1987, the system operating procedures which specified the require-
ments and precautions for system operation and isolation, including the maximum ;
, time the isolated buses could be supplied by the station batteries, were not !
used. The failure to utilize the operational procedures and to intorporate the l
precautions and requirements of the procedures for the removal and return of
equipment from service in accordance with the requirements of the Technical
!
A-1 l
,
l
l
,
,,e
. i
,
, Specifications and 10 CFR 50, Appendix A, resulted in a chain of events which
culminated in the injection of lake water into the steam generators.
2. Engineering Support and Evaluations
The engineering support provided for a number of recent modifications and ,
maintenance activities was found to be inaccurate or lacking in thoroughness.
The SS0MI team identified a number of cases in which engineering evaluations
failed to correctly determine the effects of proposed modifications. Examples
include a wrong estimate of the time of discharge of Battery NK12 prior to the
loss of Vital Bus NB02, failure to prepare for the loss by providing alternate
supplies, failure to provide adequate overpressure protection for the Reactor
Coolant Drain Tank, failure to recognize the single failure design criteria of ;
the CRVIS system, inadequate design of a diesel generator output breaker -
anti pumping logic which prevented the breaker from closing onto a cleared,
deenergized bus, and a number of inadequately justified or documented engineer-
ing evaluations in modification packages.
3. Corrective Actions
Although, as noted above, the licensee has a good trend analysis program, the
SSOMI team identified a weakness involving the adequacy of corrective actions
for identified deficiencies, including the identification of root causes,
evaluation of related areas for similar deficiencies, and actions to prevent
recurrence. One significant example was identified during the evaluation of
PMR 1903, which involved Essential Service Water pipe wall thinning and
through-wall corrosion. After the deficiency was identified, there appeared to
be no attempt to check for similar deficiencies on the other train, there was
no immediate evaluation of the deficiency to determine whether the thinner
walls violated the Updated Safety Analysis Report commitments, and the correc-
tion of the deficiency did not include a determination of a root cause and
specification of actions to prevent recurrence.
l
,
l
1
j
l
l
r
.
l
A-2
.
---_-w---w i- wyF-