ML20244E272
| ML20244E272 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 05/30/1989 |
| From: | Eapen P, Moy D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20244E269 | List: |
| References | |
| 50-317-89-08, 50-317-89-8, 50-318-89-08, 50-318-89-8, GL-87-12, GL-88-17, IEIN-88-036, IEIN-88-36, NUDOCS 8906200241 | |
| Download: ML20244E272 (9) | |
See also: IR 05000317/1989008
Text
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U.S. NUCLEAR REGULATORY COMMISSION
REGION
Report Nos.
50-317/89-08
50-318/89-08
Docket Nos.
50-317
50-318
License Nos.
DPR-69
Licensee:
Balttnore Gas and Electric Company
Calvert Cliffs Nuclear Power Plant
MD Routes 2 &3
Lusbv. Maryland 20657
Facility Name:
Calvert Cliffs Units 1&2
Inspection At:
Lusi v. Maryland
Inspection Conducted:
Anffl 10-14. 1989
Inspector:
N8k
bN
'
T . M'oy , fRe " tor Engineer
Date
D.
Q' K " &J
IAD /f(9
Approved By:
Dr.
P.
K. Eapen,' Chief, Special
' ate /
Test Programs Section, DRS
Inspection Summary:
Routine unannounced safety inspection on
April 10-14, 1989. (Inspection Report Nos. 50-317/89-08 and
50-318/89-08.)
Areas Inspected:
Review of licensee actions in response to the
" expeditious actions" described in Generic Letter 88-17, " Loss of
Decay Heat Removal" during non-power operation.
The inspection
reviewed instrumentation, training, procedures and plant staff
awareness as related to mid-loop operation.
Results:
All " expeditious actions" described in Generic Letter 88-17 were implemented at the Calvert Cliffs Nuclear Station
prior to drain down to mid-loop operation.
The management
involvement, training and staff awareness to problems related to
mid-loop operation were adequate.
Procedures, instrumentation,
and systems required to support mid-loop operation were found to
be consistent with the licensees response to Generic Letter 88-
17.
No violations were identified.
8906200241 890607
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ADOCK 05000317
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DETAILS
1.0
Persons Contacted
1.1
Baltimore Gas and Electric Company
Charlie Cruse, Manager, Nuclear Eng. Service Dept.
- J. E. Gilbert, Supervisor, Procedure Development and
Modification Acceptance Unit
- John R. Hill, Supervisor, Operation Training
- P.
E.
Katz, General Supervisor, Design Engineering
- 'L.
S.
Larragoite, Engineer, Licensing Unit
Keven Nietmann, Supervisor, Nuclear Training
!
- Lee Russell, Manager, Calvert Cliffs
Don Shaw, Licensing Engineer
Alan Thornton, Project Engineer
- J.
E. Thorp, Senior Engineer
- Don Ward, Principal Engineer
- Raymond Wenderlich, General Supervisor, Nuclear.Oper.
i-
1.2
U.S.
Nuclear Reaulatory Commission
- H.
Eichenholz, Senior Resident Inspector, Calvert
Cliffs
- V.
Pritchett, Resident Inspector, Calvert Cliffs
- Denotes those attending the exit meeting on 4/13/89.
The
inspectors also contacted other administrative and-
technical personnel during the inspection.
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2.0
Review of Licensee Action in Response to Generic Letter No.
88-17. Loss of Decay Heat Removal (TI2515/101)
Loss of decay heat removal (DHR) during non-power operation
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and the consequences of such a loss have been of increasing
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concern to the NRC.
Many events of loss of DHR have
occurred while the reactor coolant system has been drained
down for mid-loop activities such as steam generator
inspection or repair of reactor coolant pumps.
The
possibility exists to bypass two fission product barriers as
the ~9 actor coolant system and containment may both be open
while the.mid loop activities are in progress.
GL 87-12, " Loss of Residual Heat Removal (RHR) while the
Reactor Coolant System (RCS) is partially filled" was issued
to all licensees of operating PWR's and holders of
construction permits on July 9,
1987.
Respcnses indicated
that the licensees did not understand the identified
problems, and the problem continued as evidenced by events
at Waterford on May 12, 1988 and Sequoyah on May 23, 1988.
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The seriousness and continuation of this problem has
resulted in the issuance of GL 88-17.
In addition, the
Director of NRR has written to the CEO of each licensee
operating a PWR, in which he said, "We consider this issue
to be of high priority and request that you assure that your
organization addresses it accordingly."
He also wrote to
each licensed operator at all PWR plants on " Operator
Diligence while in Shutdown Conditions," and enclosed a copy
GL 88-17 requires the recipients to respond with two plans
of actions:
A short-term program entitled " Expeditious Actions,"
a.
and
b.
A long-term program entitled " Programmed Enhancements."
This inspection assessed the short-term licensee actions as
outlined in the " Expeditious Actions" Section of GL 88-17.
2.1
Review of Licensee Response to Generic Letter 88.17.
NRC reviewed the licensee's responses dated January 3,
1989 to Generic Letter 88-17.
This review concluded
that the licensee responses met the intent of the
generic letter with respect to expeditious actions,
even though the responses to some items were brief.
The NRC staff reviewed the licensee's mid-loop
operations in 1988 as detailed in the NRC inspection
reports 50-317/88-16 and 50-318/88-16.
The inspector reviewed the licensee response dated
January 3,
1989 to Generic Letter 88-17 to understand
the actions committed to by the licensee.
The licensee
response provided a description of action taken to
address the eight recommended expeditious actions
identified in the Generic Letter.
The inspector
verified that the licensee actions were implemented
during mid-loop operation in accordance with the NRC
guidance in Generic Letter 88-17 as detailed below.
2.2
Core Exit Temperature Indications
The inspector verified that for mid-loop operation, the
licensee has taken adequate administrative and
procedural steps to provide at least two independent,
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continuous coolant temperature indicators that are
representative of the core exit conditions.
The
licensee monitors the core exit temperature using
thermocouple.
At least two thermocouple
are
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maintained operable to monitor core exit temperature
during mid-loop operation.
A third thermocouple is
used as standby.
The inspector reviewed step (A) in
Appendix (5) of Operating Procedure (OP)-5 " Plant
Shutdown from Hot Standby to Cold Shutdown."
This step
requires as an initial condition that at least two RCS
temperature indications be provided in the control
room.
The inspector verified that there are at least two
independent core exit temperature indications in the
control room.
The process computer alarm is 20 degrees
F above the current Core Exit Thermocouple
(CET)
temperature.
Periodic logging of CET temperature is
required whenever CET temperature or process computer
temperature alarm is not functional.
The inspector also verified that the core exit
thermocouple temperature recorders (2TR-131A for CET's
functioning adequately.
The indicated temperatures
were within the acceptance limits specified in OP-5.
Based on the above, the inspector concluded that the
temperature indication system is consistent with the
expeditious actions of Generic Letter 88-17.
2.3
RCS Water Level Indication
The inspector verified that the licensee has procedures
and administrative controls to provide at least two
independent continuous RCS water level indications
whenever the RCS is in a reduced inventory condition.
Water level indication for mid-loop operation is
provided by a permanently installed pressure
transmitter which senses pressure at the bottom of the
piping between steam generator No. 11 and the reactor
vessel.
Indication is provided in the control room by
a digital meter which is calibrated before each.use.
The meter has operator adjustable high and low level
alarms.
The setpoints are controlled by the operating
procedure for a reduced inventory condition.
The inspector reviewed step (F) in Appendix (5) of
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Operating Procedure (OP)-5 " Plant Shutdown from Hot
Standby to Cold Shutdown" for RCS water level
indication.
The water level alarm is set at +0.2 feet
of the refueling level.
This setpoint was chosen based
on the Calvert Cliffs past operating experience.
In addition to the water level indicator in the control
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room a tygon tube provides local indication at the 27
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ft. elevation in the containment.
This level
indication is also taken from the bottom of the piping
between steam generator No. 11 and the reactor vessel.
The licensee had provided accurate elevation marks
(from 27 ft. elevation) so that the temporary level
indicating scale for the tygon tubing could be properly
placed.
According to step (F) in Appendix (5) of
Operating Procedure (OP)-5, the indication between this
tygon tubing water level and centrol room water level
should agree within +0.50 ft. reading.
The indicated
water levels were within the acceptance range specified
in OP-5.
The inspectors observed that the water level monitor
cabinet in the control room was not secured.
This
unsecured cabinet has the potential to damage other
safety related equipments in the control room in the
event of Design Base Earthquake (DBE).
The inspector
identified this matter to the licensee and the licensee
issued Non-Conformance Report (NCR) No. 7882 to
evaluate this problem and take corrective and
preventive action, as necessary.
On April 14, 1989, the licensee issued revision 2 to
the " Electrical and Controls Section Standard Practice
No. 36, Storage of Transient / Semi-portable Equipment
in Safety-Related Areas," to include the control room
as one of the safety-related areas to be covered by the
procedure.
The inspector reviewed a copy of tnis
revision to standard practice No. 36 and had no further
questions in this regard.
Based on the above, the inspector concluded that the
licensee's level indicating system was consistent with
the expeditious actions described in Generic Letter 88-
17.
2.4
RCS Inventory Control
The inspector verified that the licensee has procedures
and administrative controls to provide at least two
means of adding inventory to the RCS, in addition to
pumps that are a part of the normal DHR systems.
One
source of inventory makeup is from the High Pressure
Safety Injection (HPSI) pump.
The second independent
source of makeup comes from the Containment Spray (CS)
pumps.
Containment spray pumps are aligned to pump
water from the Refueling Water Tank (RWT) to the
Each injection flow path
will provide sufficient flow to keep the core covered.
The inspector verified that the water flow path for
these systems does not bypass the reactor vessel before
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exiting any opening in the RCS.
Based on the above,
the inspector concluded that the licensee actions in
this regard are consistent with expeditious action
described in Generic Letter 88-17 and had no further
questions concerning this issue.
2.5
RCS Perturbations
The inspector verified that the licensee has
implemented procedures and administrative controls
to
avoid operations that cause level perturbations in the
RCS.
Based on interviews and system walkdowns with the
reactor operators, the inspector concluded that the
operators are adequately trained to preclude
unnecessary RCS/SDC perturbations.
Furthermore, step
(D) in Appendix (5) of OP-5 procedures, requires the
reactor operator to log each perturbation in the
control room log.
The Control Room Supervisor (CRS) is
required to evaluate each such logged perturbation.
Inspector verified that the level of detail for each of
the RCS perturbation log entries is adequate for
tracking and evaluating RCS perturbation problems.
Based on the above, the inspector concluded that the
licensee procedures and controls were consistent with
the requirements described in the Generic Letter
concerning this issue.
2.6
Steam Generator Nozzle Dams
The inspector verified that the licensee has
implemented procedures and administrative controls to
assure that all hot legs are not blocked simultaneously
by nozzle dams unless a large enough vent path to
prevent pressurization of the upper plenum of the
reactor vessel is provided.
The licensee uses steam
generator nozzle dams during mid-loop operation.
The
recommendation of NRC Information Notice 88-36 had been
incorporated into the licensee's maintenance procedure
SG-19, " Installation Use and Removal of Steam Generator
Primary Nozzle Dams."
This procedure provides the
proper sequence of steam generator manway and nozzle
dams installation and removal along with appropriate
caution notes and the reasons for the required
sequence.
The steam generator nozzle dam maintenance
procedure requires an adequate reactor vessel head vent
prior to installing all steam generator hot leg nozzle
dams.
Based on the above, the inspector concluded that the
licensee established adequate measures to control
activities related to nozzle dam installation and
removal during mid loop operations.
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2.7
Loon Ston Valves
Loop stop valves are not part of the Calvert Cliffs
Units'1&2 system design.
2.8
Containment Closure
The inspector verified that the licensee has prepared
procedures and administrative controls to reasonably
assure containment closure before core uncovery
following'a loss of DHR event.
The licensee's estimate for containment closure time in
the event of a loss of shutdown cooling event are:
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a.
If openings totaling less than one square inch are
present in the RCS cold leg or reactor coolant
pump (RCP), closure will be within 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
b.
If openings totaling greater than one square inch
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are present in the RCS cold leg or RCP, closure
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will be within 45 minutes.
c.
If openings totaling greater than one square inch
are present in the RCS cold leg or RCP and a
sufficient vent path is provided for the reactor
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vessel head to prevent a loss of inventory out the
maintenance opening, closure will be accomplished-
within 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
The inspector reviewed step (E) in Appendix (5) of OP-5
procedure for containment closure.
Containment
integrity is established per containment integrity
verification procedure STP No. 055A-2.
Any closure or
penetration which deviate from the containment
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integrity verification procedure will be logged and
maintained by the shift supervisor in the control room.
Based on the review of STP No. 055A-2 procedure and the
closure deviation log, the inspector concluded that the
licensee has adequatae measures to isolate the
containment prior to core uncovery.
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2.9
Trainina
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The inspector verified that the licensee training made
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the licensee personnel adequately aware of the risks
associated with mid-loop operation.
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The inspector reviewed training program documentation,
held discussions with the training supervisor and
senior reactor operators regarding reduced inventory
operation at Calvert Cliffs Stations.
Both the
supervisor and senior reactor operators were
knowledgeable of the Loss of Decay Heat Removal event
at Diablo Canyon Unit 2 in 1987 and the lessons learned
from it.
In addition, there is adequate procedural
information in the control room to guide the operators
during a loss of shutdown cooling event.
The inspector
reviewed the following:
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Loss of shutdown cooling with the RCS in reduced
inventory condition for licensed operator
requalification training program (lesson plan LOR-
202-3B-89 & 203-5-89).
Pre-shift briefing (GSO instruction 88-5).
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Loss of component cooling water / shutdown cooling
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(lesson plan LOR-202-3B-1002-89).
The inspector interviewed parsonnel responsible for
installing temporary core exit thermocouple jumpers,
containment closure and steam generator nozzle dam
installation.
All these personnel were knowledgeable
of their inspection responsibilities.
Based on the
above, the inspector concluded that the licensee has
established and implemented an effective training
program to provide guidance to personnel during a loss
of shutdown cooling event.
2.10 Summary
Licensee management and supervision were actively
involved in the conduct of reduced inventory operation.
This involvement was particularly evident in the
establishment of procedures for mid loop operations.
These procedures thoroughly addressed the concerns
discussed in Generic Letter 88-17.
The personnel
involved in mid-loop operation were knowledgeable and
adequately trained in their responsibilities.
The
involvement of supervision was particularly evident in
the review of containment closure deviations and core
exit temperature deviations by control room
supervisors.
The licensee was responsive to the
concerns addressed in Generic Letter 88-17.
During the inspection, the inspectors raised a concern
regarding the unsecured level instrumentation cabinet
in the control room for mid loop operation.
The
licensee promptly issued an LER to assess this concern.
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Prior to the conclusion of the inspection, the licensee
revised an existing' procedure for the control of
movable objects in the vital areas to include the
control room.
Licensee's QA involvement for mid-loop operation was
not very apparent.
For example, the inspector observed
no QA/QC coverage for the mid loop operation.
Additionally, the applicable procedures did not have
" hold points" for verification by the line organization
or QA.
3.0
Plant Tours
The inspector made several tours of the plant including
the control room, auxiliary building and turbine
building.to observe work in progress, housekeeping and
cleanliness.
No unacceptable conditions were noted.
4.0
Exit Interview
On April 13, 1989, an exit interview was conducted with
the License's senior site representatives (denoted in
Section 1) to summarize the observations and
conclusions of this inspection.
At no time during this
inspection was written material provide to the licensee
by the inspector.
Based on the NRC Region I review of
this report and the discussions held with licensee
representatives during this inspection, it was
determined that this report does not contain
information subject to 10 CFR 2.790 restrictions.
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