ML18094A469
| ML18094A469 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 05/12/1989 |
| From: | Eapen P, Lopez A, James Trapp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18094A468 | List: |
| References | |
| 50-272-89-07, 50-272-89-7, 50-311-89-06, 50-311-89-6, GL-87-12, GL-88-17, NUDOCS 8906020145 | |
| Download: ML18094A469 (7) | |
See also: IR 05000272/1989007
Text
.
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
89-07
89-06
Docket No.
50-272
50-311
License No.
Licensee:
Public Service Electric and Gas Company
P. 0. Box 236
Hancocks Bridge, New Jersey 08038
Facility Name:
Salem Units No. 1 & 2
Inspection At:
Hancocks Bridge, NJ
Inspection Conducted:
April 17-21, 1989
Inspectors:
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M. Trapp, Reactor Engineer, DRS, EB
?f';:t:_~lC X-
cA?"E. Lopez, Reactor'~ineer, DRS, EB
Approved by:
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P. K. Eapen, Chi e~
Test Programs
Section, DRS, EB
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'date
Inspection Summary:
Routine Unannounced Inspection* on April 17-21, 1989
(Inspection Report Nos. 50-272/89-07 (Unit 1), 50-311/89-06 (Unit 2)
Areas Inspected:
Review of licensee actions in response to the "expeditious
enhancements" described in Generic Letter No 88-17, "Loss of Decay Heat Removal. 11
The inspection reviewed supporting instrumentation, training, procedures and
staff awareness as related to mid-loop operation.
Results:
The inspectors found that all "expeditious enhancements" described in
Generic Letter 88-17 were implemented at Salem prior to drain down to mid-loop
operation.
The inspectors found the management involvement, training, and
staff awareness to problems related to mid-loop operation to be highly effective.
Procedures, instrumentation, and systems required to support mid-loop operation
were consistent with the licensees response to Generic Letter 88-17.
No
unresolved items or violations were identified during this inspection .
DETAILS
1.0 Persons Contacted
1.1 Public Service Electric and Gas Company
- R. Dulee, -Quality Assurance Principal Engineer
J. Gueller, Operations Manager
- C. Lashkari, Senior Staff Engineer
J. Lloyd, Principal Training Supervisor
- L. Miller, Station Operation Manager
M. Reese, Nuclear Training Coordinator
- G. Raggio, Station Licensing Engineer
T. Worrell, Quality Assurance Lead Engineer
1.2
U. S. Nuclear Regulatory Commission
- ~-
K. Eapen, Section Chief, Special Programs Section
K. Gibson, Sr. Resident, Salem
- A. Lopez, Reactor Engineer
- J. Trapp, Reactor Engineer
- Denotes presence at exit meeting held on April 21, 1989.
2.0 Review of Licensee Actions in Response to Generic Letter (GL) No. 88-17,
Loss of Decay Heat Removal (2515/101)
Loss of decay heat removal (OHR) during non-power operation and the
consequences of such a loss have been of increasing concern to the NRC.
Many events of loss of OHR 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
that two fission product barriers could be breached while these activities
are in progress, since the reactor coolant system and containment will
both be open.
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 1 s and holders of construction permits on July 9, 1987.
Responses
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.
The seriousness and continuation of this problem has resulted in the
issuance of GL 88-17.
In addition, the Di rector of NRR has written
1 to
the CEO of each licensee operating a PWR, in which he said,
11We consider
this issue to be of high priority and request that you assure that your
organization *addresses it accordingly. 11
He also wrote to each licensed
operator at all PWR plants on
110perator Diligence while in Shutdown
Conditions, 11 and enclosed a copy of Generic Letter 88-17.
3
GL 88-17 requires the recipients to respond with two plans of actions:
a.
A short-term program entitled
11 Expeditious Actions," and
b.
A long-term program entitled 11 Programmed Enhancements."
This inspection addressed the short-term licensee actions as outlined
under 11 Expeditious Actions, 11 of GL 88-17.
The inspectors reviewed the licensee response dated January 6, 1989
The licensee response provided a detailed
description of action taken to address the eight recommended expeditious
actions identified in the Generic Letter.
The inspectors verified
that the licensee actions are consistent with the NRC guidance
provided in Generic Letter 88-17.
The NRC reviewed the licensee's mid-loop operations in 1987 as
detailed in the NRC inspection reports 50-272/87-28 and
50-311/87-30.
For the details of the NRC review of the licensee's
preparations and conduct of mid-loop operation during the current
Unit 1 -refueling outage (No. 8) see NRC inspection report
50-272/89-03.
2.1 Temperature Indication
The inspectors verified that for mid-loop conditions, the licensee
has taken adequate administrative and procedural steps to provide at
least two independent, continuous coolant temperature indicators
that are representative of the core exit conditions.
The licensee
monitors the core exit temperature using fifty-eight bottom mounted
thermocouples.
Step 2.11.2 of Procedure II-1.3.6 "Draining of the
Reactor Coolant System," requires as an initial condition that at
least two bottom mounted temperature indicators are providing RCS
temperature indication in the control room or control room racks.
Steps 5.1.l(f)&(g) of Procedure II-1.3.6 requires the thermocouples
be displayed in the control room on the plant process computer, or if
local indication is used constant surveillance and communication with
the control room be established.
The procedure requires that a high
temperature alarm be operational with a set point of 200°F.
This
alarm is initiated by the plant process computer and provides an
audible alarm in the control room, in addition to a print out on the
computer alarm printer.
The inspectors found the temperature indica-
tion system to be consistent with the expeditious actions of Generic
Letter 88-17.
The inspectors had no further questions concerning the
core exit temperature monitoring ?YStem .
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2.2
RCS Water Level Indication
The inspectors 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. The licensee uses two RCS flow trans-
mitters (F0441A, F0400A), recalibrated to indicate RCS level, to
provide indication of RCS level in the control room.
Each transmitter
provides indication in the control room with an alarm set at 97 1-6
11 *
The alarm is audible and lights an overhead annunciator in the control
room.
In addition to the two level indicators in the control room, a
tygon tube is connected to the No. 13 intermediate leg and provides
local indication in the containment.
The licensee requires level
indications with an operable low level alarm in the control room in
Procedure II-1.3.6 Steps 2.11.3 and 2.11.5. Appendix 1 of Procedure
II-1.3.6 provides detailed guidance on how to install the tygon tub~
so that it will provide an accurate level indication.
The inspectors
verified that the tygon tube level indication was installed in
accordance with this procedure.
The inspectors noticed that the
licensee had provided accurate elevation marks so that the temporary
level indicating scale for the tygon tubing could be properly placed.
The inspectors verified that the scales for the RCS flow indications
and alarm description on the overhead annunciators had been changed.*
Procedures were reviewed to assure that precautions were provided for
possible variations in indicated RCS level.
The level indicating system was found to be consistent with the
expeditious actions described in Generic Letter 88-17.
The
inspectors had no further questions concerning RCS level indication
during mid-loop operation.
2.3
RCS Inventory Control
The inspectors verified that the licensee has procedures and
administrative controls to provide at least two available or operable
means of adding inventory to the RCS, in addition to pumps that are a
part of the normal OHR systems.
One. -source of inventory makeup is
from the charging pumps.
Technical Specification 3/4.1.2 requires
one charging pump to be operable in modes 5 and 6.
The second
independent source of makeup comes from either of the two safety
injection pumps.
Step 2.11.7 of Procedure II-1.3.6 requires one
charging pump be operable and one safety injection pump to be avail-
able.
A hot leg injection path for the safety injection pump is also
required to be available by this step.
The inspectors found this to
be consistent with expeditious action described in Generic Letter and
had no further questions concerning this issue.
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2.4 - RCS Perturbations
The inspectors verified that the licensee has implemented procedures
and administrative controls to avoid operations that could perturb the
RCS.
Step 5.1.l(h) of Procedure 11-1.3.6 requires work activities
which could effect RCS inventory be minimized during mid-loop operation.
During each shift the Containment Coordinator and the Senior/Nuclear
Shift Supervisor meet and discuss work activities* which could perturb
RCS inventory.
The content of these meetings is documented.
In addition, during each shift the Nuclear Shift Supervisor will
review shift work activities for impact on RCS inventory.
The
licensee issued a supervisory letter SL-37 "Salem Primary Systems
Loss of Decay Heat Removal" to all supervisory personnel.
Attached
to this letter were posters which describe
1100
11 and
1100 Not" activi-
ties which should be performed to prevent perturbations during
mid-loop operation.
These were displayed thoughout the plant.
The
inspectors found that the licensee had taken positive actions to
avoid perturbations of the RCS during mid-loop operation.
The pro-
cedures and controls were found to be consistent with the applicable
requirements described in Generic Letter 88-17. *The inspectors had
no.further questions concerning this issue.
2.5 Hot Leg Flow Paths
The inspectors 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 vent path is provided
to prevent pressurization of the upper plenum of the reactor vessel.
The licensee uses steam generator nozzle dams during mid-loop opera-
tion.
The installation procedures require the cold leg nozzle dams
to be installed prior to the hot leg dams and the hot leg dams to be
removed prior to the cold leg dams.
This method of nozzle dam
installation reduces the effects of upper reactor vessel plenum
pressurization.
Prior to draining the.reactor vessel to mid-loop operation,
Procedure 11-1.3.6 Step 5.1.l(b) requires the removal of all three
pressurizer safety valves.
This provides a vent area of approximately
0.5 square foot.
The licensee has calculated that the back pressure
in the reactor vessel with three safety valves removed, at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
after shutdown, would be approximately 3.1 psig.
The licensee stated
that the 3.1 psig back pressure was acceptable as the existing cold
leg openings were sufficiently* high to avoid spilling of the reactor
coolant under this back pressure.
The licensee stated that cold leg
openings for maintenance during mid-loop operation would be considered
on a case by case basis, and the required vent area for each opening
of the told legs would be recalculated.
The reactor vessel back
pressure is important to determine the amount of water which could
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spill out of a cold leg opening.
This would also affect the time to
core uncovery.
The licensee stated at the exit meeting that they
would reassess the back pressure issue as part of the long-term
corrective action plan.
The inspectors determined that this was _
acceptable based on the fact that the licensee is aware of the
importance of the upper plenum back pressure and does considered cold
leg openings on a case by case basis.
The inspectors had no further
questions concerning this issue.
2.6
Loop Stop Valves
Loop stop valves are not part of the Salem Unit No. 1 or 2 system
design.
2.7 Containment Closure
The inspectors verified that the licensee has prepared procedures
and administrative controls to assure containment closure prior to
core uncovery during a loss of OHR event.
Procedure II-1.3.6 Step
2.11.1, requires that the equipment hatch be installed prior to
decreasing reactor vessel level more than three feet below the
reactor vessel flange.
Other penetrations in the containment are
isolated using Abnormal Operating Procedure AOP-CONT~2
11Containment
Closure on Loss of RHR.
11
This procedure is entered when RHR is lost
and the reactor vessel level can not be restored.
In most situations,
the containment could be isolated from the control room using the
manual phase-A initiation.
The inspectors concluded that the contain-
ment building could be isolated prior to core uncovery.
The inspectors
had no further questions concerning containment closure.
2.8 Training
The inspectors verified that training conducted by the licensee, made
the licensee personnel aware of the risks associated with mid-loop
operation.
The inspectors verified the effectiveness of this training
by interviewing operating personnel and reviewing the training material
provided to the trainees.
The operators interviewed were found to
h~ve an indepth knowledge of the issues disc~ssed in Generic Letter 88-17.
All operating personnel were trained just prior to draining
the reactor vessel to mid-loop.
The training material and lesson plans reviewed by the inspectors
did not contain all of the material committed to be part of training
in the licensee response to Generic Letter 88-17.
The inspectors
discussed the topics missi~g from the lesson plans with the licensed
operators. It became apparent to the inspectors that all the topics
committed to be taught in the response to the Generic Letter had in
fact be~n presented during the training sessions.
The inspectors
discussed the weakness in the lesson plan and training material with
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the Plant Operations Manag~r who stated at the exit meeting that the
lesson plans would be improved prior to the next use of this course
material.
The inspectors had no further. questions concerning this
issue.
2.9
QA/QC Involvement
The QA organization was found to have extensive involvement in the
mid-loop operation issue.
The inspectors reviewed licensee surveil-
lance report 88-0633 which was a surveillance conducted during the
last drain down operation at Salem Unit 2.
The surveillance results
were satisfactory.
QA had also performed indepth root cause analysis
of industry events relating to loss of RHR and studied these events
for applicability at Salem.
The inspectors interviewed QA personnel
involved in mid-loop operation concerns and found that they had full
knowledge of industry events, and plant specific details relative to
mid-loop concerns.
2. 10 Summary
The inspectors found the licensee management and staff to be aware
of the types of problems which may occur during mid-loop operation.
The actions taken by the licensee in*installation of instrumentation,
training, staff awareness and management involvement were found to be
highly effective. All the expeditious actions addressed in the
licensee response to Generic Letter 88-17, dated January 6, 1989,
were implemented during mid-loop operation.
The licensee was found
to be pursuing additional improvement, for mid-loop operation, as
part of a long-term corrective action program.
3.0 Exit Meeting
At the conclusion of the site inspection, on April 21, 1989, an exit
interview was conducted with the licensee 1 s senior ~ite representatives
(denoted in Section 1) to discuss the results and conclusions of this
inspection.
At no time during this inspection was written material provided to the
licensee by the inspector.
Based on the NRC Region I review of this
report and 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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