ML20154E648
| ML20154E648 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 05/05/1988 |
| From: | Cooper R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20154E617 | List: |
| References | |
| 50-341-87-48, NUDOCS 8805200277 | |
| Download: ML20154E648 (9) | |
See also: IR 05000341/1987048
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-341/87048(DRP)
Docket No. 50-341
Operating License No. NPF-43
Licensee: Detroit Edison Company
2000 Second Avenue
Detroit, MI
48226
facility Name:
Fermi 2
Inspection At:
Fenni Site, Newport, Michigan
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Inspection Conducted: October 18, 1987 through March 31, 1988
Inspectors:
W. G. Rogers, Senior Resident Inspector
M. E. Parker, Resident Inspector
Approved By:
R. Cooper, Chief
k8 $1'
S/S//J
Reactor Projects ection 3B
Datt '
Inspection Summary
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Inspection on October 18, 1987 through March 31, 1988 (Report
No. 50-341/87048(DRP))
Areas Inspected:
Special unannounced inspection by a resident inspector of
the events surrounding the failure of licensed operators to comply with the
Technical Specification action requirements associated with a reactor
protection system instrument channel and of the isolation design configuration
of the primary containment radiation monitor.
Results: Three violations were identified (failure to comply with a Technical
Specification action statement, failure to provide adequate procedure content
and inadeouate primary containment isolation capability).
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8805200277 880509
ADOCK 05000341
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DETAILS
1.
Persons Contacted
a.
Detroit Edison Company
F. Abramson, Operations Engineer
- D. Gipson, Plant Manager
- P. Anthony, Compliance
- S. Catola, Vice President, Nuclear Engineering
- L. Goodman, Licensing Supervisor
J. Green, Systems Engineering
R. Laubenstein, Nuclear Assistant Shift Supervisor
J. Leman, Director, Plant Safety, Nuclear Production
- R. Lenart, General Director, Nuclear Engineering
L. Lessor, Advisor to Plant Manager
R. Lightfoot, Nuclear Shift Supervisor
- W. Omer, Vice President, Nuclear Operations / Plant Manager
J. Plona, Operations Support Engineer
E. Preston, Assistant Director, Plant Safety
B. Sheffel, Nuclear Production, Technical Engineering ISI
F. Svetkovich, Technical Engineer, Nuclear Production
- B. R. Sylvia, Group Vice President, Nuclear Operations
- L. Wooden, Supervisor, I&C
- L. Fron, Supervisor, Mechanical and Fluid Systems
- P. Marquart, General Attorney
- W. Tucker, Superintendent, Operations
b.
U.S. Nuclear Regulatory Commission
- M. Parker, Resident inspector
- P. Pelke, Project inspector
- W. Rogers, Senior Resident Inspector
- C. Anderson, Enforcement Specialist
- R. Cooper, Chief, Projects Section 3B
- H. Wong, Sr. Enforcement Specialist, OE
- Dr. C. J. Paperiello, Deputy Regional Administrator
- R. Knop, Chief, Projects Branch 3
- M. Virgilio, Deputy Director, ORP
- T. Quay, Licensing Project Manager, NRR
- Denotes those attending March 30, 1988 exit meeting.
- Denotes those attending the April 28, 1988 enforcement conference.
2.
Review of Drywell Pressure Surveillance Testing
a.
Background
in Inspection Report 50-341/87007(DRP), a violation of a limiting
Condition for Operation (LCO) action statement associated with the
high pressure coolant injection / reactor core iso'ation cooling
(HPC1/RCIC) systems was identified.
That violation was very
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similar to the LC0 violation which occurred during this inspection
period. The previous violation also involved performance of an
I&C surveillance in which licensed operators failed to recognize
the Technical Specification implicaSons,
in response to the HPC1/RCIC violation, the licensee stated that
the violation was attributable to a lack of understanding of plant
conditions and a lack of an impact statement in the procedure.
The licensee further stated that an I&C surveillance procedures
improvement program was in place to upgrade the I&C surveillance
procedures as part of the corrective steps that would be taken to
avoid further violations. This effort included the addition of
impact statements documenting the ramifications, legal and physical,
of perforring a particular surveillance test and to specify the
plant operational conditions under which the test may be performed.
The I&C surveillance improvement program was to be completed by
January 31, 1988.
To nrevent another violation before the completion of the I&C
surveillance improvement procram, the licensee began generating
interim impact statements. The interim impact statements were
being generated as a surveillance became due but were not formally
incorporated into the procedure or approved by the Onsite Review
Connittee (0SRO).
The interim impact statement would then be
formalized into that particular surveillance procedure when the
procedure was revised under the total 1&C surveillance procedure
improvement effort.
b.
1.imiting Condition for Operation
On October 24, 1987, the Nuclear Assistant Shift Supervisor (NASS)
signed on Plant Operations Fanual (P0M) 44.020.015, Revision 3,
"NSSS - Drywell Pressure, Division I, Channel A Response Time fest;
C71-N650A and C71-N050A," to allow the instrument and control (I&C)
repairman to perform a response time test of transmitter C71-N050A
and master analog trip unit C71-N650A for the nuclear steam supply
shutoff system (NSSSS) drywell pressure input. Transmitter
C71-N050A is one of two Division I NSSSS high drynll pressure
channels for the reactor protection system instrunantation and the
isolation actuation instrumentation required by Technical
Specification (T.S.) 3.3.1 and 3.3.2, respectively.
At 12:25 p.m. EDT, on October 24, 1987, the I&C repairman defeated
the high drywell pressure instrumentation for channel A, as part of
the response time test. This action rendered the channel inoperable
and the licensee entered into the Technical Specification two-hour
action statement.
T.S. Table 3.3.1-1 and Table 3.3.2-1, Table
Notation (a) allows a channel to be placed in an inoperable status
for up to two hours for required surveillance testing without
placing the trip system in the tripped condition provided at least
one operable channel in the same trip system is monitoring that
parameter.
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At 2:25 p.m. EDT, the two hour Technical Specification allowance
for having the drywell pressure channel inoperable was exceeded
and went unnoticed by all personnel involved with the surveillance.
It was not until 3:15 p.m. EDT, when the I&C repairman reported
to the operating shift that they were having trouble with the
surveillance, that personnel recognized a possible time constraint
problem with the Technical Specifications.
The drywell pressure
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transmitter was subsequently returned to operable status at
3:37 p.m. EDT.
This resulted in the high drywell pressure channel
being inoperable from 12:25 p.m. to 3:37 p.m., a total of three
hours and twelve minutes.
This is considered an apparent violation of Technical Specification 3.3.1 and 3.3.2 for failing to place the drywell pressure Channel A
in the tripped condition or to return the inoperable channel to
operable status within the two hour period (50-341/87048-01(DRP)).
c.
Inspector Followup
The inspector reviewed the procedures and discussed the event with
the licensee personnel involved.
From these reviews the inspector
ascertained that:
(1) An interim impact statement had been generated for this
surveillance procedure. However, in preparing the interim
impact statement for incorporation into the surveillance test
package, the Shif t Operations Advisor (50A) made an error in
that he incorrectly assumed that 'echnical Specifications
allowed the trip channel to be out of service for surveillance
testing for three hours when Technical Specifications only
allows the channel to be out of service for testing for two
hours. This error was further compounded when the operations
engineer concurred with this impact statement.
(2) The incorrect infomation in the non-0SR0 approved interim
impact statement was in contradiction to OSR0 approved
Procedure P0M 44.020.015, Step 4.13.
(3) Even the incorrect time restraints in the interim impact
statement were not adhered to since the chonnel was not placed
in the tripped condition after the three hour time limit had
expired.
(4) No formal means existed for tracking short-tem LCOs.
Procedure
POM 21.000.18, "Out-of-Specification Log," does not re% ire
short-term LCOs to be placed in the out-of-specifica* ion log.
(5)
I&C personnel were aware that response time testing has
typically taken one shift to complete and they were aware that
this particular response time test would take longer than the
two-hour LC0 limit, but this knowledge was not transmitted to
the operating authority.
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(6) During the surveillance, the licensee only performed Sections 6.1,
6.2 and 6.3 of POM 44.020.015. This consisted of performing the
response time test for transmitter C71-N050A and master Analog
Trip Unit C71-N650A. This would not have resulted in any
isolations or actuations. Had this been fully understood prior
to performing the surveillance, drywell pressure Channel A could
have been placed in the tripped condition with no adverse
consequences and full compliance with Technical Specifications
would have been achieved,
d.
Impact Statement Review
On February 17, 1988, the inspector reviewed POM 44.020.015 to
ensure that the interim impact statement discussed previously had
been corrected to adequately reflect Technical Specification
requirements and had been formally incorporated into the procedure
thereby receiving OSR0 approval.
During this review it was observed
that the impact statement had been incorporated into the procedure
and approved by OSRO. However, during incorporation of the impact
statement into the procedure the licensee failed to incorporate the
correct impact statement.
Specifically, the impact statement still
allowed the channel to be inoperable for up to three hours without
placing the channel in the tripped condition instead of the two hours
recuired by Technical Specifications 3.3.1 and 3.3.2.
Further review identified that the licensee had taken the same
action with the associated drywell pressure response time
procedures:
POM 44.020.016, Revision 20, Drywell Pressure, Division II, Channel B,
Response Time Test
POM 44.020.017, Revision 20, Drywell Pressure, Division I, Channel C,
Response Time Test
P0M 44.020.018, Revision 20, Drywell Pressure, Division II, Channel D,
Response Time Test
This is considered an apparent violation (50-341/87048-02(DRP)) of
Technical Specification 6.8.1.d for failing to properly implement
the requirements of T.S. 3.3.1 and 3.3.2 into surveillance
procedures,
e.
Sunmary
Although this LC0 violation was caused by inadequate communications
between I&C and Operations, this appears to be indicative of a
breakdown in the overall understanding and appreciation of Technical
Specifications. These sare elements were apparent in the
50-341/87002 violation and as such the corrective actions taken in
response to that violation were inadequate to preclude the current
violation.
Also, this event pointed out the weaknesses in the
licensee's system for not tracking short-term Limiting Conditions
for Operation (LCOs).
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3.
Primary Containment Radiation Monitoring System (PCRM) Design Configuration
a.
Background
The primary containment radiation monitoring (PCRM) system is
configured in a parallel arrangement with the drywell hydrogen /
oxygen sample panel.
Both systems no mally operate continuously
during reactor operation and sample the drywell atmosphere from
five zones through primary containment penetrations X-48a through
X-48e.
Each of these five penetrations has an air-operated remote
manual isolation valve (T50-F401A, F402A, F403A, F404A, and F405A)
and an associated local manual valve (T50-F033A, F034A, F035A,
F036A, and F037A).
The original isolation design for the PCRM system and the drywell
hydrogen / oxygen sampling system was an acceptable alternative to
GDC 56 described in the FSAR Section 6.2.4.
Containment isolation
requirements were achieved using a single isolation valve (T50-F401A
through F405A) and this was based on a closed system outside the
containment. The basis of a single remote manuhl isolation valve
is described in the UFSAR Table 6.2.2 and Section 6.2.4.2.2.3.2.
This design assumed that the PCRM system would operate following a
loss-of-coolant accident (LOCA) and the PCRM system would be in
compliance with the closed system requirements.
In January 1984, the licensee determined that the PCRM system did
not comply with the specific closed system requirements.
Specifi-
cally, the system was not qualified for containment design pressure
and problems were noted with the sei."nic and material certifications
provided by the vendor. The licensee subsequently determined that
the PCRM was a non-essential system following a LOCA ard should be
automatically isolated upon receipt of a LOCA signal. As such, in
early 1984, the two automatic isolation valves (T5P-Fl50 and F451)
and the two manual isolation valves (T50-F063 and ' N4) were added
to isolate the PCRM on a LOCA signal (high drywell pressure).
The automatic isolation valves were added to provide the isolation
of the nob r.:n-essential PCRM system and was intended to return the
system to that of a closed system configuration.
The licensee
believed this configuration was an acceptable alternative to GDC 56.
This configuration resulted in providing two barriers in the event
of a LOCA and failure of the PCRM boundary, the first barrier being
the newly installed automatic isolation valves (T50-F450 and F451)
and the second barrier being the remote manual isolation valves
(T50-F401A through F405A).
b.
Licensee Event Report (LER)
On October 17, 1987, during implementation of Engineering Design
Package (EDP) 1786 on the primary containment radiation monitoring
(PCRM) system, the primary containment isolation was questioned as
to the adequacy of containment integrity.
The isolation boundary
utilized was two solenoid-operated valves (T50-F450 and F451) which
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are not primary containment isolation valves. The containment
isolation valves for this penetration (X48a throu
manual isolation valves (T50-F401A through F405A)gh X48e) are remote
and receive no
automatic isolation signals. The primary containment isolation
valves were open dur'ng implementation of this EDP.
It was further
detennined that valves T50-F450 and F451 were not properly qualified
to satisfy containment integrity requirements.
On October 17, 1987, at 8:50 p.m. EDT, the licensee determined this
was a potential loss of the primary containment integrity.
The
licensee took immediate action to isolate the primary containment
boundary. The Division 1 PCMS was subsequently shut down and
isolated. The isolation consisted of closing primary containment
isolation valves T50-F401A through F405A for the Division 1 PCMS.
This resulted in the plant being in a seven days and 30 day Limiting
Condition for Operation as a result of having the primary containment
H2/02 monitoring system and radiation monitoring system, respectively,
out of service. After investigation into the isolation, the licensee
determined the inadeouate isolation was a reportable event and on
October 17,1987, at 9:30 p.m. EDT the licensee made the applicable
notifications per 10 CFR 50.72 for a primary containment integrity
violation.
Local leak rate testing (LLRT) on T50-F450 and T50-F451 was
satisfactorily completed on October 18, 1987, and the applicable
out of service log was subsequently cleared for these valves.
This
allowed the licensee to isolate the PCMS radiation monitor utilizing
T50-F450/F451 and open the designated containment isolation valves
T50-F401A through F405A. This action allowed the H2/02 monitor to
be placed in service and took the plant out of the seven day LC0.
However, a 30 day LC0 was still in place for not having the PCMS
radiation monitor in service.
Discussions with the resident inspector concerning the configuration
resulted in the inspector questioning the current design configuration
of the PCRM system to meet 10 CFR 50, Appendix A, General Design
Criterion (GDC) 56 requirements.
c.
On October 27, 1987, in DECO Letter No. NRC-87-0211, the licensee
requested a temporary exemption from the requirements of 10 CFR 50,
Appendix A, Criterion 56 (GDC 56), Primary Containment Isolation.
This request was a result of review by the NRC in detennining that
the current design configuration, for the Division I primary
containment monitoring system, did not meet the requirements of
This exemption request along with other correspondence
identified the licensee's proposed course of action to return the
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primary containment radiation monitoring system to service utilizing
the current isolation design.
On November 13, 1987, the NRC granted to the licensee an exemption
to General Design Criterion 56 of Appendix A to 10 CFR Part 50.
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This exemption permitted postponement of full compliance with GDC 56
for the primary conteinment radiation monitoring isolation until
startup following the planned local leak rate testing in March
1988.
To support operation pending incorporation of modifications
of the PCRM isolation, while the exemption is in effect, the
licensee comitted to upgrade the effectiveness of the isolation
scheme as described in NRC letter, dated November 13, 1987.
This
action included treating valves T50-F450, F451, F040, F046, F063
and F064 as primary containment isolation valves, in a manner
consistent with any other valve listed in Technical Specifications.
The licensee also comitted to revise the Emergency Operating
Procedures and enhance operator training as an interim compensatory
measure.
On January 29,(License Amendment) change which results from1988, the lice
Specification
modifications to bring the PCRM isolation design up to the standards
set forth in GDC 56.
On March 29, 1988, NRR issued Amendment 17 to
the operating license in response to the January 29, 1988 letter,
d.
Inspector Followup
In reviewing the PCRM design, the licensee was unable to find any
correspondence accepting this configuration as an acceptable
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alternative to GDC 56.
Review of the UFSAR identified that it had
not been updated to reflect the current design configuration and
categorized penetrations X48a through X48e as Engineered Safety
Feature (ESF) or ESF-related system penetrations and that these
penetrations are attached to a closed system.
Technical
Specifications (TS) had also not reflected this modification to
include the additional automatic isolation valves (T50-F450 and
F451).
In addition, the licensee's program in 1984 to install
these valves failed to properly maintain these valves in accordance
with the applicable requirements of GDC 56,10 CFR 50 Appendix J
and other testing requirements (functional testing, logic testing,
positive indicator checks, LLRT testing, etc.).
The PCRM containment isolation system design as described in UFSAR
Section 6.2.4 does not reflect the as-built system as required by
10 CFR 50.34(b). The deviation between the UFSAR and the as-built
system was not evaluated in accordance with 10 CFR 50.59.
This is
an apparent violation (50-341/87048-03(DRP)).
e.
Enforcement Conference
An Enforcement Conference was held in the Region 111 Office on
April 28, 1988 to discuss the PCRM containment isolation system
design.
No new information was provided.
4.
Exit Interview (30703)
The inspectors net with licensee representatives (denoted in Paragraph 1)
on March 30, 1988, and informally throughout the inspection period and
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summarized the scope and findings of the inspection activities. The
inspectors also discussed the likely informational content of the
inspection report with regard to documents or processes reviewed by the
inspectors during the inspection. The licensee did not identify any
such documents / processes as proprietary.
The licensee acknowledged the
findings of the inspection.
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