ML17202U745
| ML17202U745 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 07/12/1990 |
| From: | Gardner R, Kopp M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17202U744 | List: |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 50-237-90-16, 50-249-90-15, NUDOCS 9007170347 | |
| Download: ML17202U745 (9) | |
See also: IR 05000237/1990016
Text
\\
U. S. NUCLEAR REGULATORY COMMISSION
REGION I I I
. R~port Nos. 50-237/90-016(DRS); 50-249/90-015(DRS)
Docket Nos. 50-237; 50-249
License No. DPR-19; No. DPR-25
Licensee:
Commonwealth Edison Company
Post Office Box 767
Chicago, IL
60690
Facility Name:
Dresden Nuclear Power Station~ Units 2 and 3
Inspection At:
Dresden Site, Morris, Illinois
Inspection Conducted: May 21 through June 21, 1990
~.fl.~
Inspector: M. lf Kopp
Also participatirig in the inspection
and contributing to the report:
G. Hausman, NRC Region III
.~~~~
Approved By: Ronald N. Gardner, Chief
Plant Systems Section
Inspection Summary*
"7/j~/f=
Date
Inspection on May 21 through June 21, *1990 (Report No. 50-237/90016(DRSl;
50-249/90015CDRSll.
.
.
.
Areas Inspected: Special announced safety inspection of licensee actioris
regarding previously identified EQ concerns and implementation of
Regulatory Guide 1.97 commitments (Modules 25587,30703 and 92701) SIMS
67 .3.3 *(open).
Results: Of two are~s inspected, two apparent EQ violations were
identified (Paragraphs 3.a, 3.b) for failure of the EQ program to
identify and ensure qualification of certain electrical* components within
the scope of 10 CFR 50.49, and for failure to identify and establish a
record of qualification for certain terminal blocks within the scope of
10 CFR 50.49; two unresolved. items were ideritified (Pagragraphs 4,
5.a(3)(a)) in the RG 1.97 area concerning resolution of the qualification
of RG 1.97 neutron flux monitoring, and th~ qualification of isolation
devices .
9007170347-90071~
POR
ADOCK
n~nnn~*~7
Q
.
. *-* -~5{
- .
.,/
DETAILS
- * ** 1. * * Persons Contacted
a..
Commonwealth Edison Company (CECo)
~+E. D. Eenigenburg, Station. Manage~, Dresden
+K.
Kociuba~Nuclear Quality Programs
+L. F. Gerner, Technical Superintendent
J. Kotowski,Production Superintendent
+G. Smith, Assistant Superintendent Operations
- D. Van Pelt, Assistant Superintendent Maintenance
+K'. Peterman, Regulatory Assurance Supervisor
- +B. Viehl, Engineering Design Supervisor
- J. Silady, Licensing Administrator
+M. C'. Strait, Technical Staff Supervisor
- +R .. Falbo, Regulatory Assurance *Assistant
+M. Bablitch, Instrument Maintenance
+J. Harrington, Quality Assurance
+D .. Gulati, Instrument Maintenance
+C. M. Allen, Nuclear Operations
- +K. Yates, Administrator
- Z. Boxer, EQ.Engineer
- +K. Polson, EQ Coordinator
- +A. K.Behera, EQ Engineer
+J. Krass,' EQ Engineer
R. Johnson, Technical Staff
. B. Zank, Operating Engineer
+ Denotes those attending the interim site exit on May 24, 1990
- Denotes those attending final exit conducted by telephone on
June 21,1990
2.
Licensee Actions on Previously Identified EO Findings
a.
(Closed) Unresolved Item (50-237/89010-0l(DRS);
50-249/89009-0l(DRS)
This item concerned the lack of weep holes in EQ related
termi na 1 boxes. Based upon NRC concerns,_ the 1 i censee
performed a walkdown and found eight EQ designated boxes in
Unit 3 without the required weep holes.
As a result, the
. *licensee committed to perform additional inspections and
found that the cable pull box for the High Pressure Coolant
Injection (HPCl)'~otor Operated Valve (MOV) M0-2301-5
contained two terminal blocks but not the required weep
holes.
Iri addition, the licensee discovered that the two
terminal blocks were not Environmentally Qualified (EQ) and
consequently declared the HPCI system inoperable. Licensee
2
b.
Event' Report 89-005 was initiat~d on May 12, 1989 and
attributed this event to a management defi~iency during the
implementation of the.initial EQ program in that a partial
walkdown of EQ circuits was performed instead of a total
walkdown:~nd physical inspection. Also, the licensee
indicated that the initial identification of EQ components
was based on a review of* electrical schematics and w1~ing
diagrams and that the terminal blocks identified in the HPCI
system were not shown on these drawings.* The litensee's
corrective actions consisted of*replacing the unqualified
terminal blocks. with qualified taped splices and drilling a
weep hole in the cable*pull box.
Due to the discovery of the unqualified components, the
licensee committed to inspect all EQ equipment terminations*
located in junction boxes, conduit fittings and pull boxes in
order to perform a review of splices, terminal blocks, and
weep holes.
The licensee stated that the EQ terminations
located at the equipment had previously been inspected and
therefore a re-inspection of these terminations were not
included as part of the corrective action.
The licensee
committed_ to perform the inspections in accessible areas of
the plant during normal plant operations and during the
ref~el outage~ for the areas not accessible during normal
operations.
During this inspection, .the inspector reviewed the results of
the licensee's EQ inspection program.
The licensee has
completed inspections of EQ terminations in Unit 3, and plans
to complete the Unit 2 inspections during the upcoming Fall
1990 refuel outage.
The results of the completed inspections
indicate that deficiencies in the licensee's EQ program have
existed since the November 30, 1985 EQ deadline, in that the
licensee did not know that certain terminal blocks and
splices.were installed in EQ circuits and that certain EQ
equipment was installed in. a configuration contrary to the EQ
test requirements.
Based upon the EQ deficiencies described
in Section 3 of this report, it appears that the licensee
was in apparent violation of 10 CFR 50.49.
The apparent EQ
deficiencies will be tracked as a separate item, therefore,
this Unresolved Item is considered closed.
(Closed) Unresolved Item (50-237/89022CDRPll:
This item concerned the i~stallation of model FlOO United
Electric Temperature switches, for main steamline and HPCI
steamline leak detection and automatic isolation.
Previous
engineering analysis by the licensee indicated that the FlOO
- switches ~ere ~ualified, however, the licensee discovered
that the model FlOO switches were not referenced in the
environmental qualification (EQ) binder.
The binder listed
Model F7 as the qualified switch .
3
r~ ' *
During this inspect~o~,~ the in~pector reviewed the licensee's
Suitability Evaluation* 89-151, EQ .Variation 89-023, and
Sargent and Lundy document CQ0'#046118, "Justification for
Continued Operation, United Electric Temperature Switches."
The inspector concluded, based upon EQ test data and
similarity between the Model F7 and FlOO switches, that the
switches were qual.ified and that the licensee's EQ
binde~
needed to be updated to include the FlOO j~itches. The
licensee committed to revise the EQ. bi1Jder.
No further NRC concerns were identified.
3.
Requirements and Apparent Violations
Based on NRC concerns, and the results of the litensee~i
inspections '(described in Section 2.a ~f this report), the licensee
committed to walkdown a11* EQ terminations located in junction
boxes, conduit fi~tlngs, and pull boxes to ensure that these
components were qualified, and that weep.holes were installed in
the electrical enclosures as required.
.
.
Ouri~g this inspection, the i~spector revi~wed the results of the
licensee's completed inspections.
As a result of the inspector's
review, the following apparent violations* were identified although
the specific number of examples-may change based on the licensee's
ongoing review.
a.
10 CFR 50.49 (a) requires each holder of a license to operate
a nuclear power plant to establish a program for qualifying
safety-related electrical equipment, nonsafety-related elect-
ri ca 1 equi pmerit whose* failure could prevent the satisfactory
fulfillment of a safety function, and certain post-accident
monitoring equipment ..
10 CFR 50.49 (f) requires that each item of electrical
equipment important to safety be qu~lified by testing and/or
analysis of an identical item of equipment under identical
conditions, or a similar item or under similar conditions
with a supporting analysis to show that the equipment to be
qualified is acceptable.
Contrary to the above, equipment important to safety which
the li~ensee determined -had.to be-qualified was not properly
qualified by EQ tests and/or analjses as demonstrated by the
following:
( 1)
Approximately six terminal *blocks installed in EQ
circuits for which the licensee could not determine*the
manufacturer and consequently whether or not the
terminal blocks had been EQ *tested.
4
- )
(2)
(3)'
(4)
(5)
(6)
(7)
(8)
Approximately 16 taped splices installed in EQ circuits
for which the licensee could not determine the
manufacturer and consequently whether or not the
splices had been EQ tested.
Two splices made with wire nuts and_tape that were
installed in EQ circuits for which the licensee could
not determine the manufacturer and consequently whether
or not the splices had been EQ tested.
Four General Electric 2960 series terminal blocks
~nstalled in EQ circuits without EQ test~ and/or
analysis to supp6rt qualification.
Approximat~ly six splices made with Voltrex heat shrink
tubing installed Jn EQ circu~ts without EQ tests arid/or
'analysis to support qualification.
Suppression Pool *Temperature Monitoring cables
installed in.the Toru~*Room which have been submerged
at various .times since their installation in 1983.
The
EQ test for these cabl~s d~d not include submerged
condition~.
Approximately 61 electrical enclosures containing
either terminal blocks or splices and required to have
.weep holes were found installed in the plant without
weep holes.
One Amphenol connector tbat contained a Teflon and
Rexalite insert material without EQ tests and/or
analysis to support qualification. *
.b.
10 CFR 50.49 (f) requires each item of electrical equipment
impo~tant to safety be qualified by testing and/or analysis
of an identical item of equipment under identical conditions,
or a similar item or under similar conditions with a
supporting analysis to show that the equipment to be
qualified is acceptable ..
10 CFR 50.49 (j) requires a re~ord of qualification be
maintained in an auditable *form for the entire period in
which the covered item.is installed in the *plant.
- Contrary to the above, as of April 1989 the licensee's EQ
files did not contain EQ tests and/or analyses which
demonstrated qualification of .63 General Electric terminal
blocks and 59 Marathon 1500 te~minal blocks which were found
installed in EQ related circuits .
5
4.
Licensee Action on *RG 1.97 SER/TER Commitments (SIMS No. 67.3.3)
The NRC inspectors evaluated the implementation of the licensee's
RG 1.97 commitments discussed in the September 1988 Saf~ty
Evaluation Report (SER).
This SER n.oted that the neutron flux
- ~onitoring in~trumentation did.not comply with the Category 1
requ i rem'ents of RG 1. 97, Revision * 2.
The l i cen*see was required to
upgrade their neutron flux instruments to comply with this
- "criteria.
During thi~ current review, the*i*n~pectori noted that the licensee
was requested by NRC letter dated February 14, 1990, to provide a
schedule for the installation of neutron flux monitoring
instrumentation that meets the requirements of RG 1.97, Revision 2.
Th~ BWR Owners' Group responded to the NRC, in a letter dated
February 21, 1990, and stated that the NRC had not provided the
"necessary guidance for the design and implementation of a post
accid~nt neutron moni~oring system.
The BWR Owners' Group proposed
the development of the necessary design criteria in order to a 11 ow
licensees to evaluate the suitability of the currently available
instrumentation.
NRR is currently reviewing the Owners' Group
request and pending resolution of this issue, this is considered an
Unresolved Item (50-237/90016-03(DRS); 50-249/90015-03(DRS)).
5.
Program Compliance to RG 1.97 CTI 2515/87; SIMS No. 67.3.3)
Commonwealth Edison Company is committed to meeting the criteria in RG
1.97, Revision 2, for the Dresden Station.
RG 1.97 identifies plant
variables to be measured during and after a design basis accident, and
specifies criteria for assuring the reliability of these instruments
during and after an a~cident.
Generic Letter No~ 82-33 issued December 17, 1987, as Supplement 1 to
NUREG-0737 specified, in P*rt, the application of RG 1.97 to emergency
response facilities, including the control room.
Licensees were
required to implement an instrumentation system within the scope of RG
1.97 and to identify any deviations to NRR for resolution.
NRR reviewed
those deviations identified bY the l i.censee and addressed the re solution
of these deviations in their SER.
The region~l inspectors were then
d.irected by TI 2515/87 to verify details of the licensee's
implementation of RG 1.97 at nuclear.facilities.
The objective of this NRC inspection was to review the licensee's RG
1.97 program to assure that.a reliable instrumentation system had been
implemented by the licensee to allow operators to assess plant
conditions during and following an accident.* The NRC review included a
technical evaluation of the RG 1.97 instrumentation and an examination
df instruments and int~rfaces in the field.
a.
Technical Evaluation of RG 1.97 Instrumentation
The NRC inspectors performed a technical evaluation, including a
6
physical inspection in the control room, of instruments for
Category 1, Type A variables to* verify licensee compliance to
criteria outlined in RG 1.97, Revision 2.
The appropriate
instruments and interfaces were.evaluated for inclusion or
omission to.the Master* Equipment List (MEL); environmental and
seismic qualification.of instruments and interfaces, including
isolation devices; redundancy; isolation from non lE circuits;
integrity of independent power supplies; range; direct m~asurement
of parameters; and frequency of testing, surveillance and
calibration.
The RG 1.97 variables selected for review included:
0 Reactor Coolant Level, Category 1, Type A
0 Reactor Coolant System Pressure, Category 1, Type A
0 Drywell Pressure, Category 1, Type A,
0 Suppression Chamber Pressure, Category 1, Type A
0 Suppression Pool Water Temperature, Category 1, Type A
0 Suppression Pool Water Level, C_ategory 1, Type A
The inspectors reviewed P&ID drawfngs, e]ectric~l schematics,
instrument loop diagrams, power supply diagrams, FSAR and SERs,
and the technical specifications* for the selected variables.
The
results of the RG 1.97 review are listed below:
(1)
Reactor Coolant Level and Reactor Coolant System Pressure
RG 1.97, paragraph 1.3.1 states that qual~fication applies
to the complete instrumentation channel from sensor to
display where the display is a direct indicating meter.
The
inspectors noted that the Master Equipment List (MEL)
identified level indicator LI 2(3)-26.3-106A and pressure
indicator PI 2(3)-263-156 as Regulatory Related (RG). *The
MEL defines RG related equipment as non-safety related.
In
addition, the inspectors reviewed the instrument loop
schematics and noted that the indicators are isolated from
the safety-related part of the instrument channels .
. The licen~ee stated that the RG designation was used because
the indicators could not be classified as safety-related due
to the absence of specific qualification documentation.
The
licensee further stated that this exception was documented
in the Technical Approach section of the Summary Report,
Dresden Station Units 2 and 3, Compliance to Regulatory
. Guide 1.97, Revision 2, dated July 31, 1985.
However, the
inspectors n6ted that this deviation was not addressed in
The Region III inspectors
7
(2)
diScusse.d the licensee's approach with the Instrumentation
Systems Control Branch (ISCB) at NRR to determine if this
deviation to RG' 1. 9.7 had been reviewed and approved by the
SER.reviewer.* NR~ indicated that the use of non-safety
indicators in Category I/Type A instrument channels had -been
used by other licensees and:'was acceptable provided the
non-safety iristruments were isolit~d from.the safety-related
instruments and that the licensee found the instruments to.
'be reliable.* Based upon NRR's review and acceptance of *
this approach, the inspectors had no further concerns.
Physical Separation of RG 1.97 Circuits
During this inspection, the inspectors noted that the
.licensee took exception with the ~equi~ement.to comply with
the RG 1. 75 *:cable separation criteria in the Summary 'Reports
for Station*compliance to Regulatory Guide 1.97, revision 2,
dated July 31, 1985.
Specifically, the licensee stated that
the station was *licensed before RG 1.75 established the
requirements for physical independence of ele~tii~al systems
and. that existing instrumentation used for post accident
monitoring does not follow these separation requirements.
The licensee stated that new instrument loops added to
fulfill Category I requirements would comply with the
separation requirements.
The inspectors noted that thi*S deviation to RG 1.97 had not
been addressed in the licensee's SER.
The Region III
inspectors contacted ISCB at NRR to determine if this
exception to RG 1.97 was evaluated by the SER reviewer and
determined to be acceptable.
NRR informed the inspectors
that aJ~hough RG 1.97 requires compl.iance to the separation
requirements, plants that were licensed prior to RG 1.75 are
not required to meet the separation requirements for
.. existing circuits but must comply.for new circuits which are
added.
Based upon NRR acceptance of this deviation, the
inspectors had no further concerns.
(3)
Electrical. Isolation of RG'l.97 Circuits/ Moore Industries
Signal *conditioners
(a)
The inspectors noted that Moore Indust~ies signal
conditioners are used as isolation devices in RG 1.97
- circuits. The licensee recently initiated maximum
credibl~ fault testing of the Moore isolators, Model
SCT/4'-20mA/4-20mA/117VAC(STD) and MVT/80-160mV/4-
20mA/l l7VAC(STD). per Nutherm test report CWE-#3690R,
dated December, 1989.
Pending further review of the
~esults of this test, this is considered an Unresolved
Item (50-237/90016-04(DRS);50-249/90015-04(DRS)).
(b)
The inspectors also noted that non-safety related
8
~)
\\
5.
Unresolved Items
. '
. ,
. computer points, A~2840, A-3840, A2096 and A-3096,
installed in the.Suppression Pool Water Level
transmitter loops 2(3)-1641-SA and ~B, were not
isolated from the safety-related in.strumentation in
the circuit. The licensee informed the inspectors
that the computer points are not terminated and that
Temporary Alterations 11-72-89, 111-42-89, 11-71-89,
and 111-41-89 prev~nt termination of the* compute~
points until isolators are installed~* The licensee
stated that isolators will be .install~d in the
upcoming Fall 1990 Unit 2 refuel outage .and during the
- February 1991 Unit*3 refuel outage. *Based upon the
licensee commitments t6 install isolators, the NRC
inspectors had no further concerns.
An unresolved item is a matter about which more information is required
in order to ascertain whether it is an acceptable item, an op*en item, a
deviation, or -a violation. *unr.esolved items disclosed during this
inspection are discussed in Paragraphs 4, and 4.i.(3)(a).
6.
Exit Interview
The Region II I inspector met with the 1 i censee' s repr.esentat i ves
.. _(denoted in Paragraph 1) at the conclusion of th~ site inspection on May
24, 1990, and discussed the purpose.and findings of the inspection by
telephone at *the conclusion of the inspection on June 21, 1990.
The
licensee acknowledged this information.
The inspector also discussed
the likely informational content of the inspection report with regard to
documents or processes reviewed by the inspector during the inspection. *
The licensee did not identify any such documents/processes as
proprietary ..
9