ML17158B861
| ML17158B861 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 11/12/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17158B859 | List: |
| References | |
| 50-387-96-10, 50-388-96-10, NUDOCS 9611190029 | |
| Download: ML17158B861 (48) | |
See also: IR 05000387/1996010
Text
0
TABLE OF CONTENTS (Continued)
IV. Plant Support............
R1
Radiological Protection and Chemistry (RP&.C) Controls..........
R1.1
ALARA.... ~ ..
R1.2
Control of Radioactive Material and Contamination ........
R1.3
External Exposure Controls........................
~
R1.4
Use of Flashing Lights for Access Control to High Radiation
Areas
%1
~
~
~
~
~
~
~
~
~
~
~
~
~
~
~
~
~
~
~
~ \\
~
~
~
~
~
~
~
~
~
~
~
R1.5
Response
to Alarming Dosimetry....................
~
R2
Status of RP&C Facilities and Equipment ....................
R5
Staff Training and Performance
in RP&C
R5.1
Staff Training
R6
RP&C Organization and Administration
R7
Quality Assurance
in RP&C Activities
R8
Miscellaneous
RP&C Issues
R8.1
Transverse
Incore Probe (TIP) Work ln December 1995.....
R8.2
Reactor Water Clean-up (RWCU) Pump Work ln January 1996
R8.3
(Closed) Violation 50-387;388/96-04-02 ....
~ . ~.......
~
R8.4
(Closed) Unresolved Item 50-387;388/96-09-02
R8.5
(Opened) Violation 50-387;388/96-10-03,
"Radiological
Posting Deficiencies" ..... ~....... ~....
~ ~........
~
S8
Miscellaneous Security and Safeguards
Issues ................
S8.1
(Update)
EA 95-250, Security Chilling Effect
(Update)
EA 94-212, Security Chilling Effect............
18
18
18
19
20
21
22
24
25
25
26
27
29
29
30
34
34
35
36
36
V. Management Meetings.................. ~.... ~......
~
~
~
~
~
~
~ .
X1
UFSAR Review
X2
Exit Meeting Summary.... ~...... ~..................
36
36
37
VI
9biii90029 96iii2
ADOCK 05000387
8
Re ort Details
Summar
of Plant Status
Unit 1 began this inspection period in the refueling condition.
This condition was reached
on September
9, when the reactor head was detensioned.
Between October 19 and 20
the unit was restarted
and then returned to a shutdown condition in order to repair
acoustic flow indication equipment on the 'L'afety relief valve (SRV).
Unit 2 began this inspection period at 100% percent power.
With the exception of limited
decreases
in power, based on power control center (PCC) requests
and a period of
approximately one day at 80% power to support cooling tower chemistry limitations, the
unit operated throughout the period at 100% power.
I. 0 erations
02
Operational Status of Facilities and Equipment
02.1
S stem Ali nment
a.
Ins ection Sco
e 71707
Susquehanna
Steam Electric Station (SSES) routinely aligns the RHR System to send
post accident injection flow through the RHR heat exchanger
in parallel with the
RHR heat exchanger
bypass line.
b.
Observations
and Findin s
The system alignment was verified and reviewed by the inspector.
The Final Safety
Analysis Report (FSAR) is silent on the desired normal alignment.
It was determined
that the RHR heat exchangers
are designed to accept 10,000 gpm and site
operating procedures,
including emergency operating procedures,
limitthe flow
through the RHR heat exchangers to less than 10,000 gpm.
The practice of
aligning the RHR heat exchangers
in the normal injection path was determined to be
supported
by design basis calculations.
c.
Conclusion
In its normal 100% injection mode lineup, the RHR system is operated
in a
configuration that is consistent with the design basis and the design basis is
supported
by design basis calculations.
02.2
Unit 2 Turbine Combined Intermediate Valves
a.
Ins ection Sco
e 71707
The inspector reviewed a plant upset condition:
while testing the ¹3 CIV, the ¹1
CIV closed and then immediately opened over a 5 to 10 second period.
Observations
and Findin s
The operators responded
appropriately to the upset condition and the site
adequately
performed two operability determinations
in accordance
with the
guidance supplied in NRC Generic Letter 91-18.
The licensee initiated a corrective
action document,
CR 96-1680, and determined that the ability of the CIVs to
perform their intended turbine overspeed
protection design function was not
affected.
The inspectors reviewed operator actions and operator logs, discussed
the issue
with the operators
on shift at the time of the event, reviewed the operability
determinations
and evaluated the corrective actions,
The inspector determined that the ability of the turbine to be tested at power was
degraded
(note: this was in addition to a concurrent degraded
condition involving
the automatic voltage regulator).
However, the symptoms identified by the licensee
do not appear to impact the overspeed
protection function of the CIVs as the
symptoms are limited to test circuitry logic.
C.
Conclusion
OZ3
Operators adequately
responded to an unexpected
upset condition involving the
Unit 2 Turbine Combined Intermediate Valves.
Transient Material Stora
e in the Unit 1 and 2 Reactor Buildin s
Ins ection Sco
e 71707
SSES routinely stores transient material in assigned
areas within the Unit 1 and Unit
2 Reactor Buildings.
The inspector reviewed the process
used by the licensee to
control the material in these areas and inspected the material stored in the transient
areas to determine if there was the potential for an impact on the operation of
safety related equipment.
Observations
and Findin s
The licensee's
program for the control of transient material is addressed
in NDAP-
QA-552,Transient Equipment Controls.
The areas identified in this procedure were
originally supported with an engineering
evaluation and an expectation of what type
of material was to be stored in each of the areas.
The inspector viewed one area
near safety related high pressure coolout injection (HPCI) instrumentation at position
Q-29 in the Unit 2 reactor building.
When requested,
the licensee was able to
perform an engineering
evaluation (through Engineering Work Request M60310)
that the material presently stored in the area presented
no risk to the safety related
equipment.
However, the material that was actually placed in the storage
area was
not considered
in the original engineering
evaluation that was used to designate
the
area.
c.
Conclusions
Power 5 Light (PPSL) management
response to this issue was good
and identified no present impacts on the safe operation of plant equipment from the
storage of transient equipment.
The failure to fully control transient equipment near
'afety related equipment constitutes
a violation of minor significance and is being
treated as a Non-Cited Violation consistent with Section IV of the NRC Enforcement
Policy.
02.4
Unit 1
Restart between October 17 and 22
1996
Ins ection Sco
e 71707
The inspector reviewed the Unit 1 restart on October 22 which resulted in condition
1 operation at approximately 10:00 a.m.
The activities were well performed and
supervised.
Restart activities between October 17 and 18, 1996 resulted in the
Unit returning to condition 3 to repair the acoustic monitor associated
with the
'L'afety
relief valve.
During these earlier startup activities, several weaknesses
occurred which require further NRC inspection.
These weaknesses
are:
The operability of HPCI during the transient through 1504 psig in the RCS.
The operability of LPCI during the mode change into condition 2.
The operability of the acoustic monitors associated
with the A, G, L, and
R
SRVs.
The licensee's review and long term corrective actions for these weaknesses
have
not been completely characterized.
As a result, the NRC review of these
weaknesses
will be tracked under Unresolved Item, (URI 387/96-10-01).
04
Operator Knowledge and Performance
04.1
Control Room Emer enc
Outside Air su
I
S stem
Fan 'B'utlet
Dam er Failure
Ins ection Sco
e 71707
During a routine control room tour the inspector noted that the 'B'REOASS fan
outlet damper was not functioning properly.
The inspector reviewed the licensee's
initial and subsequent
actions for the degraded
safety-related
component,
the
applicable system technical specification (TS), and similar issues previously
identified by the NRC,
0
b.
Observations
and Findin s
Unit 1 TS 3.7.2 requires two subsystems
of CREOASS to be operable with the Unit
in Operational Condition 5. With one subsystem
inoperable, the inoperable
subsystem
must be restored to operable status within 7 days or the remaining
subsystem
must be placed in service.
NDAP-QA-0326, provides guidance for operations with the potential for draining the
reactor vessel (OPDRV). Attachment A of this procedure provides the requirements
for entering an OPDRV condition, including one that requires that certain Limiting
Conditions for Operation are met without reliance on Action St'atements.
Step 1.2
requires both CREOASS subsystems
to be operable, prior to entry into an OPDRV
condition.
SSES procedure
NDAP-QA-0702, Condition Report,
requires the identification,
reporting, evaluation and correction of conditions adverse to safety or quality. This
procedure
provides a mechanism for the types of corrective action required by 10
CFR Appendix B, Criterion XVI.
On September
22, the 'B'REOASS subsystem
was running in support of a
procedure to transfer the reactor protection system power supply.
When the
'B'REOASS
fan was shut down, its outlet damper (HD-07811B) remained open,
when it should have closed.
A NSE representative
was available at the time of the
failure and determined that the 'B'ubsystem was still operable.
A work
authorization was initiated to correct the condition.
However, no CR was initiated
and the basis for operability was not documented.
On September
24, control rod drive mechanism
(CRDM) replacement work began
which required the Unit Supervisor (US) to complete Attachment A of
NDAP-QA-0326 for the OPDRV.
On September
25, the inspector learned that CREOASS damper HD-07811B was
mechanically bound (failed) in the open position, contrary to the expected
as-
designed fail-closed position (shown in FSAR Figure 9.4-1).
The inspector noted
that there was no TS Action Statement entered in the LCO log for the 'B'REOASS
subsystem
due to the damper problem.
The inspector questioned
the US on shift
about his rational for operability of the 'B'REOASS subsystem.
The US stated
that he considered the damper operable because
it was in the position required for
the fan to function, but that the basis for this determination
had not been
documented.
The CRDM replacement work was placed on hold and the damper
was repaired prior to the CRDM work being allowed to restart.
On September
25, CR 96-1645 was issued to document the damper failure and an
The inspector found the justification provided by NSE a
reasonable
basis for operability.
Previous examples where the licensee failed to identify conditions adverse to quality
on safety-related ventilation system components
were documented
in NRC
Inspection Report 50-387/96-09.
5
The inspector concluded that the degraded safety related damper constituted
a
condition adverse to quality and that the licensee had not implemented their
corrective action process
in a timely manner.
The inspector also noted that changes
made to NDAP-QA-702, Revision 1, under PCAF 1-96-6510, deleted the examples
of conditions which warrant issuance
of a CR. Although this was not viewed as
justification for failure to implement the procedure, it may have caused some
confusion.
Conclusion
The licensee did not document an operability determination for a CREOASS damper
that had failed in an open position, despite the fact this position was contrary to its
fail-closed design.
Maintenance activities with the potential for draining the reactor
cavity were allowed to commence
based on the presumed operability of both
CREOASS subsystems.
The operability determination provided after questions from
the inspector provided a reasonable
basis for operability.
In this case, the failure to
implement required administrative procedures for a condition adverse to quality
constitutes
a violation of minor consequence
and is being t'reated as a Non-Cited
Violation consistent with Section IV of the NRC Enforcement Policy.
04.2
Nuclear Plant 0 erator Lo
Review
This section documents
NRC inspection activities of specific nuclear plant operator
(NPO) performance items.
The inspector interviewed several of the non-licensed
NPOs to determine the accuracy of certain operator round logs.
Of particular focus,
were those log items that required entries into high radiation areas to obtain specific
parameters.
the inspector reviewed several completed
NPO logs and interviewed
selected
NPOs.
No discrepancies
were identified relative to the adequacy
or validity
of the selected completed logs.
In addition, all NPOs interviewed were familiar with
the licensee's
requirements that prohibit individuals other than qualified NPOs from
obtaining required readings.
Based upon the log reviews and interview results, the
inspector concluded that the log readings for equipment located in high radiation
areas were properly performed by qualified NPOs.
06
Operations Organization and Administration
06.1
Overtime A
royal Review
Ins ection Sco
e 71707
The inspector reviewed the use of overtime by SSES staff who perform safety
related functions against the requirements of TS 6.2.2.f.
Observations
and Findin s
TS 6.2.2.f.2 provides guidance on the use of overtime hours and states that an
individual should not be permitted to work more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any seven day
period, excluding shift turnover.
SSES administrative procedure
NDAP-QA-0650
4
implements the TS requirements
and requires prior approval of deviation from the
overtime hour guidelines.
The inspector found that overtime limit deviation forms were completed in
accordance
with NDAP-QA-0650 and provided the required approval for several
licensed operators who exceeded
the limitfor 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of work in 7 days during the
refueling outage.
The inspector reviewed a sample of licensed operator time
records and determined that the overtime limit deviation forms were being
generated
when required.
A sample of time records for workers in the Maintenance
Department was also
reviewed.
Three examples were identified where the prior approval for an overtime
deviation was not documented.
In response to the inspector's findings, the
Maintenance
Supervisor reviewed the refueling outage time records for the rest of
his organization and found no other examples.
The Maintenance
Supervisor
stated'hat
the deviation forms would be processed
and that the overtime worked by the
individuals was acceptable.
It was determined that the TS 7-day and 2-day overtime restrictions apply to
consecutive
days, and a review of bi-weekly pay records would not necessarily
verify these requirements
are being met.
Overtime Deviation Reports are distributed
to work group supervisors showing their employee's work hours for rolling 7-day
and 2-day periods.
Based on discussions with several supervisors, it was not
apparent to the inspector that these reports were routinely used.
Based on discussions with the inspector the licensee initiated a Condition Report
and initiated a preliminary review. The licensee determined similar weaknesses
in
the Health Physics work areas and has initiated corrective actions to resolve these
weaknesses.
C.
Conclusion
Working hours of SSES staff who perform safety-related functions were sampled in
the Operations
and Maintenance
areas.
No examples of excessive
use of overtime
were identified.
Three examples were found in the Maintenance organization where
administrative forms were not processed
in a timely manner however, this delay
was inconsequential.
07
Quality Assurance
in Operations
07.1
Effectiveness of Licensee Controls - Problem Resolution
Ins ection Sco
e 71707
The inspector reviewed a sample of approximately 300 licensee corrective action
documents,
Condition Reports (CR), Operability Determinations,
and Independent
Safety Evaluation Services
(ISES) documents.
In addition, the inspector attended
several Plant Operations Review Committee (PORC) meetings and reviewed
a
sample of approximately 40 PORC meeting activity summaries.
The inspector
attended routinely'held CR telephone conferences
in which CRs were discussed,
responsibilities assigned
and the impact on operations
assessed.
Finally, the control
room activities of the ISES were observed
and discussions
held with the ISES
control room auditor.
Observations
and Findin s
The inspector determined that the licensee was generating
a relatively large number
of CRs.
The increase
appeared
to be the result of a decrease
in the threshold for
reporting, personnel
related environmental conditions, and an increase in
management
emphasis
on the routine use of the system by plant personnel.
It was
further determined that the licensee was dedicating adequate
resources to the
process including increased visibility and attention through upper level PPSL
management.
The inspector observed the routine active participation in this proces's
by the Vice President Nuclear Operations.
The participation of upper level PPRL
management
in the CR processes,
visibly displayed
a company emphasis
on speedy
evaluation and problem resolution of safety significant issues.
Because of the
marked increase
in the number of CR documents
and issues,
a potential weakness
lies in the ability of PPSL to identify and separate
the safety significant issues from
the areas of lesser safety significance.
In general, current company efforts in this
potential area of weakness
are very good.
In a few instances,
the inspector identified weaknesses
in either the initiation of
CRs or the evaluation of specific CR deficiencies.
These instances
are assessed
by
the inspector on a continual basis for emerging trends.
One example of this type of
finding is contained in Section 04.1 of this report.
Some of the activities of the ISES control room audits mirror the NRC inspection
module 71707, directly. Other activities were individually developed
by ISES.
On a
whole, the ISES activities appear to support control room activities and the problem
resolution process adequately.
C.
Conclusions
From a programmatic standpoint, the licensee has established
and implemented
an
effective program of controls for the resolution of identified problems.
These
controls include the activities of the Independent
Safety Evaluation Services,
the
Plant Operations
Review Committee, and the visible support of corporate officer
level management.
Based on the selected sample, most CRs were resolved
effectively and were associated
with adequate
evaluations of safety significance
and operability impact.
08
Miscellaneous Operations Issues (92700)
08.1
Closed
LER 50-387 96-007:
A Division 1, primary containment isolation occurred
as the result of a loss of power to the Division
1 Reactor Protection System (RPS).
The loss of power to the RPS was caused
by a bus lockout which actuated during a
maintenance
activity. The maintenance
activity involved the connection
(landing) of
an over current relay on the alternate feeder breaker to a 4 Kv engineered
safeguard
system bus.
This event was discussed
in NRC Inspection Report (IR) 50-387,
388/96-09, and it was determined by the inspector, that the event resulted from an
incorrectly landed lead.
More specifically, it was determined that a human
performance error occurred during the connection of an electrical relay.
IR 387,
388/96-09 contained
an error and incorrectly stated that the power at which the
lockout occurred (30% vice 0% power).
This error did not affect the review
performed by or the conclusions of the inspector.
The licensee's
corrective actions
were determined to be adequate.
This failure constitutes
a violation of minor safety
consequence,
and is being treated as a non-cited violation consistent with Section
IV of the NRC Enforcement Policy.
II. Maintenance
M1
Conduct of Maintenance
M1.1
General Comments
a.
Ins ection Sco
e 62707
The inspectors observed
all or portions of the following work activities (62707):
MT-052-002, High Pressure
Coolant Injection (HPCI) Pump 6-
Year Overhaul, Unit 1
WAP52435
HPCI Repair
CP 031-511
Digital Computer Annual Preventive Maintenance,
Unit 1
WAC60854
4 Kv 1A204 Division 2 Relay Replacement,
Unit 1
WAC60501
4 Kv 1A204 Division 2 Rewire Degraded
Grid Test
Switches, Unit 1
~
TP 059-003
Suppression
Pool Cleaning and Inspection, Unit 1
Unit 1 Suppression
Pool and emergency core cooling system
(ECCS) Suction Strainer Inspection, September
19, 1996
WA P52435
Six-Year HPCI Turbine Overhaul, September
26, 1996
WA S67577
Post Maintenance Testing For Zone III BDD17521,
September
24, 1996
WA S63817
Hydraulic Control. Unit (HCU) 26-31 Charging Water Isolation
Valve 147113-2631
Repair, October 3, 1996
~
WA C60479
Reactor Building Closed Cooling Water (RBCCW) Loop
'B'eactor
Recirculation Pump Containment Isolation Solenoid
Valve Replacement,
October 4, 1996
The inspectors observed
all or portions of the following surveillance activities
(61726):
SE 151-202
Core Spray Pump B, 1000psig Hydrostatic Test, Unit 1
SO 070-018
Standby Gas Treatment System, Unit 2
SO 149-005
Residual Heat Removal System Quarterly Valve
Exercising,
Unit 2
SE 151-003 'A'ore Spray Logic Test and Functional Check, Unit
1
SO 024-001
Monthly Operating Test 'C'mergency
Diesel Generator
SO 150-050
Reactor Core Isolation Cooling Logic Test
Unit 1 Loop A Core Spray System Logic Functional
Surveillance, September
9, 1996
Unit 1 Division I LOCA/LOOP 18-Month Surveillance Test,
October 8, 1996
b.
Observations
and Findin s
One of the observations
conducted
by the inspector, involved portions of the
prerequisites
and the execution of the 18-month Unit 1, Division 1, Loss of Coolant
Accident/Loss of Offsite Power (LOCA/LOOP). This comprehensive
test required
numerous teams of in plant personnel
and extensive coordination.
The inspector
noted an excellent pre-job briefing, effective coordination by the NSE test directors,
and good communication with control room operators.
c.
Conclusions
In general, the majority of the observed
maintenance
and surveillance activities
were well performed.
An exceptionally well performed activity involved the 18-
month Unit 1, Division 1, Loss of Coolant Accident/Loss of Offsite Power
surveillance test.
This comprehensive
test included an excellent pre-job briefing,
effective coordination by the NSE test directors, and good communication with
control room operators.
10
M1.2 Control Rod Drive CRD Mechanism
Re lacement
a.
Ins ection Sco
e 62707
On September 25, 1996 the inspector observed
CRD mechanism
replacement
activities and identified several weaknesses.
b.
Observations
and Findin s
On September
25, 1996 at approximately 4:00 p.m., the inspector communicated
to SSES management
several identified weaknesses
observed
during the process of
CRD mechanism replacement.
The crew demonstrating
the weaknesses
was
composed of a foreman directing the activities by headset
and two workers located
under the reactor vessel.
In addition, there was an operations engineer that was
performing an advisory function.
The following activities were observed:
~
The foreman was not correctly using the appropriate'rocedure,
MT-055-
015, to direct the activities of the workers under the vessel.
In two instances the foreman directed the workers to perform activities that
were not addressed
by the procedure.
These two activities included removal
of roller pins, and manually wrestling the CRD back and forth to unlodge the
CRD mechanism.
When questioned
by the inspector, the foreman stated that he was not
required to use the procedure or have it in front of him because
he had it
memorized.
~
When the CRD activities impacted an intermediate range monitor cable, the
occurrence was not initially documented
in the work log by the supervising
foreman and therefore, the need for an operability review was not
documented
(note: the operations engineer independently
phoned the control
room and notified them that the cable had been impacted, while the
inspector discussed
the issue with PPSL first line management).
After the
'nspector
had a discussion with SSES first line management
a work log entry
was made.
~
A bent retaining bracket was noted by the inspector and the CRD work crew
independently.
The bent bracket appeared
to be the result of the observed
work activities under the vessel that were intended to free the lodged CRD
mechanism
(see the first bullet above).
After the inspector discussed
the above issues with the first line supervisor, the
supervisor chose to stop work.
In addition, when he was notified, the control room
Shift Supervisor reiterated the stop work direction.
Work activities were
recommenced
after the crew had been instructed in proper procedure
usage
and the
bracket was replaced with a bracket from the SSES training center.
11
On September
26, at approximately 11:00 a.m., a second stop work was ordered
by the Shift Supervisor when it was discovered that six of eight bolts were removed
from an incorrect CRD (CRD 34-51 vice the correct 38-55).
One of the workers
under the vessel identified the condition, notified the foreman on his headset
and
reinstalled the incorrectly removed six bolts.
The licensee documented
the
condition on CR 96-1663, paneled
an event review team (ERT), implemented
a
number of immediate corrective actions including second check verifications, and
submitted the results of its review and corrective action recommendations
to the
Plant Operations Review Committee (PORC) for approval prior to the release of the
stop work order.
The inspector determined that the licensee's
response
and corrective actions
following the September
26 event, were strong, technically sound, and outstanding.
The PORC activities and actions were also determined to be outstanding
and
introspective.
However, the initial response
by site Maintenance
Management to
the weaknesses
identified by the inspector on September 25, was inadequate
in
that it did not fully respond to the concerns
expressed
by the inspector and did not
prevent the September
26 event.
TS 6.8.1 states that the licensee shall establish and implement procedures
recommended
SSES Nuclear Department Administrative
Procedure
(NDAP)-QA-0500, establishes
approved practices for maintenance
procedures
and work plans.
NDAP-QA-0500 refers to Maintenance
Procedure
MT-
AD-501, which establishes
the procedural adherence
requirements for different
types of maintenance
procedures.
Section 6.2 of MT-AD-501, Maintenance
Procedure
Program, states that a step-by-
step conditional procedure provides specific detailed direction.
It further states that
strict adherence to the procedure exactly as written and in its entirety is required.
Finally, it states that the procedure must be in the field and on the job.
Maintenance
Procedure MT-055-001, CRD Removal, is a step-by-step
conditional
procedure that controls the removal and replacement of the CRD mechanisms
including the identification of the correct mechanism
and controls to second party
verify the correct mechanism.
Contrary to the requirements discussed
above, MT-055-001 was not used in a step-
by-step fashion in the field by the foreman directing the CRD removal activities.
Therefore, the controls to identify and verify the correct mechanism were not
effective.
This failure resulted in the wrong CRD being partially disassembled
with
the potential to negatively affect the cooling of the fuel assemblies
in the core and
spent fuel pool, and/or affect local reactivity conditions.
This is a violation, (VIO
387/96-1 0-02).
The inspector determined that the following root causes
contributed to the event:
Training - The workers were directed to memorize applicable portions of MT-
055-001 during their training at the SSES training center.
This training
12
activity left the workers with the impression that they did not have to use
the procedure
because they had it memorized.
Training - One of the foreman observed
by the inspector had been given
credit for previous work at other utilities and therefore, had not been required
to complete
a full course of training at the SSES training facility.
Work Experience - Crews that had been previously hired and trained by SSES
to perform CRD maintenance
were impacted by a recent personnel
realignment.
This resulted in the loss of approximately 17 out of 24
personnel from the CRD/HCU maintenance
crews.
The reassigned
personnel
were replaced with less experienced
contracted workers.
QA Oversight - During the periods observed
by the inspector, no QA
oversight of the work activities was observed.
Line Management
Oversight - During the periods observed
by the inspector,
there was a lack of direct line management
oversight of the foreman
directing under vessel activities.
Direct line management
oversight was
relegated to an Operations
Engineer, who was assigned to the crew without
a clear written statement of responsibilities or authority.
The incorrect CRD mechanism
did not have the control blade withdrawn and backseated
to
provide a leak seal.
If the final two bolts were removed or sheared off, allowing the CRD
mechanism flange to separate
from the bottom of the vessel,
a leak would have occurred.
The consequences
of the flange failure would
be an uncontrolled leak and an uncontrolled
reactivity addition.
The uncontrolled leak was projected by the licensee to be
approximately 300 gpm.
The potential to affect local reactivity conditions in the core
would result from the dropping blade movement.
The leak would not have been isolable
until the rod was either disassembled
from the blade or the flange was reseated
by the
hydraulic unit. If the flange was not reseated,
and if the control rod could be disassembled
from the blade, then there was a special tool (slide flange) available at the work scene that
could have been used to stop the leak.
The following factors mitigated the significance of this specific event:
Secondary containment (ventilation zones
1 and 3) was set by procedure.
In the
event of a leak, this alignment would significantly reduce
a potential release to the
public.
One Core Spray pump was operable,
as required by licensee controls for operations
with the potential to drain the reactor vessel/cavity (OPDRVC). This pump was
capable of supplying enough cooling medium to the reactor vessel to keep the core
covered and cooled during a 300 gpm leak.
The incorrect CRD mechanism
did not have the control rod uncoupled or the
position indicating probe (PIP) removed.
The licensee contended that the installed
PIP would have affected the alignment of the hydraulic removal tool enough to alert
0
13
the workers not to remove the final two bolts.
This argument does nothing to
mitigate the potential of the final two bolts shearing.
c.
Conclusion
The licensee failed to adequately
control CRD mechanism
repair activities in
accordance
with established
SSES procedures.
As a result bolts were removed
from an incorrect CRD mechanism flange.
The resulting plant condition was
identified by an SSES worker and no leak actually occurred.
There was sufficient
pump capacity and onsite emergency power available to supply water and cooling
to the core and spent fuel pool if the leak had occurred.
was operable
and would have allowed a filtered vent path through SBGTS to the
environment.
M1.3
Reactor Feedwater
Pum
Re air
a.
Ins ection Sco
e 62707
On September 20, 1996, and October 1, 1996, the inspector observed
repair
activities on the 'C'nd 'A'eactor Feedwater Pumps (RFPs), respectively.
The
following procedures
were reviewed:
Weld Traveler P52338
MT-048-001, RFP Disassembly Inspection and Reassembly
WAP52338, Pump Stage Repair
S63183,
RFP Rotor Removal
b.
Observations
and Findin s
The activities were observed to be well controlled, were performed in accordance
with the applicable work plans, and were affected by knowledgeable
maintenance
and welding technicians.
The welding technicians performing the weld buildup and
milling operations demonstrated
considerable
technical skills under less than
optimum field conditions.
. c.
Conclusion
The activities involved in the weld repair and milling of the Unit
1 reactor feed
pumps were well controlled by first line management,
and were performed in an
excellent manner.
M1.4
Core Shroud Examination and Re air
a.
Ins ection Sco
e 62707
On September
11, 1996, the licensee completed
a visual and ultrasonic examination
of the core shroud assembly welds.
The inspector observed portions of the
examination and evaluation processes,
and reviewed the following documentation:
14
UT-SUS-503V5, Procedure for automated
Ultrasonic Examination of the
Shroud Assembly Welds.
EC-062-1036, SSES Unit
1 Shroud Defect Calculations, dated October
8, 1996.
Structural Integrity Associates
Inc. Report, dated October 7, 1996.
On October 14, 1996, the inspector attended
a PORC meeting that reviewed the
results of the Inservice Inspection program including the core shroud examination
results.
In addition, the following corrective action documents
were reviewed to
assess
the quality of the licensee's
corrective actions and operability
determinations:
CR-96-1486
CR-96-1565
CR-96-1567
CR-961588
CR-96-1637
CR-96-1639
CR-96-1650
b.
Observations
and Findin s
next opera
c.
Conclusion
The inspector determined that the methods used by the licensee were conservative
and the conclusions arrived at by the licensee were supported
by industry standard
analysis.
The inspector was not able to verify that Unit 1 operation beyond the
ting cycle was supported
by the current test data.
The activities involved in the core shroud evaluation and inspection were well
controlled by first line management,
met current industry standards
and were
performed in an excellent manner.
M1.5
~Scaffoldin
aa
Ins ection Sco
e 62?07
The inspector observed/evaluated
the construction of several scaffolds near and/or
surrounding safety related equipment.
The licensee process for managing scaffolds
is contained
in Maintenance
Procedure
MTAD-504, Scaffolds.
b.
Observations
and Findin s
It was determined that several of the scaffolds were in place for long periods of
time (months).
In addition, the need for the particular scaffolds had not diminished
and was expected to continue.
Because of the permanent
nature of certain
scaffolds, the inspector determined that they were acting as defacto plant
modifications.
It was further determined that MTAD-504 ensured that erected
scaffolding was seismically qualified, but did not fully meet 10 CFR 50.59, in that
the MTAD did not fully ensure that the erected scaffolding would not negatively
0
15
impact nearby safety related equipment during each previously analyzed
Chapter 15 design basis accident.
These issues were discussed with the licensee who agreed with the finding and
implemented
a number of corrective actions.
These actions included:
1)
The establishment
of a task force to address the issue.
2)
A walkdown of existing scaffolds by civil engineers.
3)
Submission of project plans to an SSES Project Review Team
screening
on or before October 17, 1996.
4)
The establishment
of additional controls on long term scaffolding
requests.
There were no conditions identified by the inspector, that clearly impacted the
present operation of safety related equipment.
Subsequent
to the inspector
identifying the issue to the licensee, the licensee performed
a walkdown of erected
scaffolds, removed some unneeded
scaffolds and determined that there were no
present instances of impact on the operation of safety related equipment.
c.
Conclusions
PPSL mana gement response
was comprehensive
and identified no present impacts
on the safe operation of plant equipment from existing scaffolds.
Therefore, this
failure to consider the effects of plant modifications in accordance
with 10 CFR
50.59, constitutes
a violation of minor consequence
and is being treated as a Non-
Cited Violation consistent with Section IV of the NRC Enforcement Policy.
M1.6
Overall Conclusions
on Conduct of Maintenance
Maintenance activities were generally well controlled.
The Unit 1 LOCA/LOOP
testing observed was well coordinated
and executed.
One violation and several
weaknesses
were identified during control rod drive mechanism maintenance
activities.
Programmatic weaknesses
were identified in non-cited violations for the
control of scaffolds and the control of temporary storage of materials.
M2
Maintenance and Material Condition of Facilities and Equipment
M2.1
Safet
Relief Valve Set Screw Lockwire
a.
Ins ection Sco
e 62707
During the Unit 1 refueling outage the inspector performed walkdown inspections,
inside containment, of ongoing maintenance
activities and general equipment
condition.
16
b.
Observations
and Findin s
The inspector found that the set screw lockwire and the lead seal for the 'S'afety
relief valve (SRV) were broken.
Following the repair, testing, and adjustment of an
SRV, the refurbishment company installs a lockwire and lead seal to indicate the
final adjustment
has been made.
The 'S'RV was last refurbished during the Unit
1 Spring 1995 outage.
The inspector questioned
whether the settings of either
SRV set screw had been changed
since its adjustment.
The torque of the 'S'RV set screws was verified by the licensee to be the required
250 ft-Ibs and the lockwire/seal was replaced.
The set screws are used to maintain
the SRV nozzle ring and adjusting ring positions following the calibration of the SRV
setpoint and during the installation process.
Adjustment of these rings requires
access through the valve's outlet flange and would require removal of the SRV
tailpipe.
Based on the fact that the set screws were found to be torqued by the
licensee, the inspector considered it unlikely that vibration during the last operating
cycle could have affected these adjustments.
The licensee, concluded that work
activities conducted
between and around the SRVs was the cause of the. broken
lockwire. The licensee's inspection of set screw lockwires on the remaining 15
SRVs did not identify any additional problems.
c.
Conclusion
The licensee's followup in response
to a broken lockwire and seal discovered
on set
screws for the 'S'afety relief valve was good.
M8
Miscellaneous Maintenance Issues (92903)
M8.1
Closed
Unresolved Item
URI 50-388 95-12-01
Hi h Pressure
Coolant In'ection
HPCI On-line Maintenance
On June 26, 1995, the licensee commenced
a four day on-line maintenance
work
window for the Unit 2 HPCI system.
The work scope of the outage increased the
risk consequences
for analyzed events and increased the core damage frequency.
The licensee evaluated the increase
in risk and risk consequence
and provided
adequate
compensatory
measures.
Four issues were identified in NRC IR 50-
388/95-12 that occurred during the performance of the maintenance
window. This
URI was opened to review the licensee's corrective actions for these four issues.
The inspector reviewed the licensee's completed corrective actions which included
improvements
in scheduling techniques,
personnel training, and procedural
upgrades.
The licensee's corrective actions were determined to be adequate.
In
this case, the maintenance
procedure quality and adherence
concerns constitute
a
violation of minor consequence
and are being treated as a Non-Cited Violation
consistent with Section IV of the NRC Enforcement Policy.
17
M8.2
Closed
URI 50-387 95-05-01:
Observation Of Activities On The Refuel Floor
The inspector observed the movement of new fuel from the fuel vault to the spent
fuel pool in preparation for the Unit 1 8th refueling outage.
The inspector found
that a procedure step for second verification of the bundle serial number was not
being performed.
After identification, an individual was assigned to verify the bundle serial numbers
as required by the licensee's
procedure.
A video tape of the fuel pool was used to
verify all bundles were moved to their proper location.
In addition to the CR process
review, an investigation of this issue was performed by the ISES group.
The
licensee's corrective actions included ISES recommendations
and established
periodic training, better definition of the supervisory assignments
and enhancement
of procedural controls.
The inspectors did not identify additional problems of this
type during the remainder of the Unit 1 8th Refueling Outage.
New and irradiated
fuel movements were observed during this inspection period (Unit 1 9th Refueling
Outage) and the recording of bundle serial numbers was observed.
Based on the
licensee's corrective actions, the failure to implement required procedure steps
constitutes
a violation of minor consequence
and is being treated as a Non-Cited
Violation consistent with Section IV of the NRC Enforcement Policy.
I
III. En ineerin
Conduct of Engineering
U date - URI 387 388 96-06-01
Containment Secondar
B
ass Leaka
e
~73051
The adequacy containment secondary
bypass leakage was identified as an issue in
several licensee corrective action documents
(CR). The need for resolution of
potential design discrepancies
was detailed in NRC Inspection 50-387, 388/96-06.
The licensee's
plan for resolving CR-96-46, 310, 356, 1359, 504, 522, 1038,
1360, and 1407 was included in an action plan dated October 10, 1996.
The
inspector reviewed the conclusions
and evaluations proposed
by site engineering to
the PORC on October 14, 1996.
In addition, a copy of the action plan was
provided to NRC Licensing (NRR) for review under TAC numbers 96641 and 96642.
The inspector determined that the plan appeared
to be complete and was
thoroughly reviewed by PORC.
The plan included proposed
TS changes to account
for a potential unreviewed safety question.
The inspector was not able to
determine if the increase from 5 to 9 SCFH of leakage was appropriate without a
prior review of the methodology by NRR, which will be conducted
under the TAC
numbers listed above.
This issue is being tracked by unresolved
item 387, 388/96-
06-01.
IV. Plant Su
ort
Radiological Protection and Chemistry (RP8cC) Controls
The inspector performed
a review of the radiological controls program, with special
emphasis
on the controls implemented for the Unit 1 refueling outage.
Specific
areas reviewed included:
maintaining radiation exposures
as low as is reasonably
achievable
(ALARA);control of radioactive material and contamination; external
exposure controls; facilities and equipment; staff training; organization and
administration; and program audits and appraisals.
In addition, the inspector
reviewed the radiological controls implemented for historical work including Unit 2
reactor water clean-up (RWCU) pump work performed in January 1996 and
transverse
incore probe (TIP) drive box work performed in December 1995.
The
inspector also reviewed licensee response
to a previous NRC violation and an NRC
unresolved item, and an evaluation of facility condition versus the UFSAR was
performed.
lns ection Sco
e 83750
The inspector performed
a review of the program to maintain radiation exposures
ALARA,including the use of radiation exposure
goals and ALARAreviews.
The
inspector gathered information by a review of a handout used at the September
25,
1996, Station ALARACouncil meeting; ALARAreviews performed for in-service
inspections
and control rod drive (CRD) exchanges;
inspections
in the plant; and
through discussions with cognizant personnel.
Observations
and Findin s
A station goal had been set to maintain total station dose to less than 265 person-
rem for 1996.
This included 140 person-rem for the Unit 1 fall refueling outage.
The inspector noted that as of September 26, 1996, at 07:30, the total outage
dose was 90.95 person-rem.
Major contributors to this total were CRDs, drywell
drywell scaffolding, reactor pressure
vessel disassembly,
drywell
shielding, and drywell in-service inspection activities. At the time of the inspection,
the dose accumulation rate had increased
due to CRD exchanges.
The inspector
noted that this had been accounted for in the dose goal.
Person-rem
dose for
individuals, radiation work permits, plant areas,
and station totals were being
closely tracked as evidenced
by a review of a daily and cumulative dose graph, daily
printouts of RWP totals, a weekly evaluation of challenges
and successes,
and a
weekly review of person-rem
breakdown by work group.
The inspector also selectively reviewed ALARApre-job reviews included in RWP
packages.
ALARAreviews included person-rem estimates,
work planning
information, external and internal exposure controls, health physics operational
concerns, dosimetry and radiological monitoring, anticipated dose rates, additional
comments,
and a work flow synopsis.
ALARApackages
were detailed, included
19
specific instructions from lessons learned from previous jobs, and were overall very
good.
Through discussions with members of the ALARAstaff, the inspector observed that
extensive planning had been performed for outage activities, and ALARAmeasures
such as radiation work permit controls, temporary shielding, and system flushes
were effectively implemented.
C.
Conclusions
The inspector concluded that the ALARAprogram was effectively implemented.
This was evidenced
by excellent use of ALARAdose goals, extensive planning, use
of RWP controls, use of temporary shielding, system flushes, and postings and
verbal communications.
R1.2
Control of Radioactive Material and Contamination
ae
Ins ection Sco
e 83750
The inspector performed
a review of the control of radioactive material and
contamination.
Information was gathered
by direct observation of personnel
and
equipment contamination monitoring practices during tours through the facility.
Observations
and Findin s
During tours of the facility, the inspector examined radioactive material packages,
and noted that all packages
(e.g., bags, containers,
or boxes) of radioactive material
were appropriately labeled as radioactive material.
All contamination monitoring equipment observed to be in-use including friskers,
personnel contamination monitors (PCMs), tool contamination monitors (TCMs), and
continuous air monitors (CAMs) appeared to be in good condition, and records
showed that they were within calibration.
The inspector noted that the number of access/egress
points to the radiologically
controlled area had been reduced to two locations; the Unit 2 turbine (South)
access,
and the control enclosure access.
This change
had been implemented to
better control radioactive material and contamination.
All personnel were authorized
to use the Unit 2 turbine access
and the control enclosure access was limited to
personnel from operations,
chemistry, health physics, and security.
In addition, a
policy had recently been implemented to allow station personnel to monitor personal
items such as note paper and pens, using tool contamination monitors (TCMs).
Health physics personnel
provided direct oversight at the Unit 2 access point, and
monitored the control enclosure access
by video and video tape.
The inspector
observed
health physics personnel providing coaching and counseling to station
personnel when contamination monitoring deficiencies were observed.
In the event
that contamination monitoring deficiencies were identified by video/video tape
review, appropriate followup actions such as interviews and followup surveys were
20
performed to determine if contamination
had been improperly released.
Based on
this review, the inspector concluded that contamination control practices and health
physics monitoring of contamination control practices were very good.
C.
Conclusions
Based on this review, the inspector made the following conclusions.
~
Radioactive material control practices were good as evidenced
by appropriate
labeling of radioactive material packages
including bags, containers,
and
boxes.
~
Contamination monitoring equipment were well maintained as evidenced
by
very good equipment condition, and instrument calibration records.
~
Health physics oversight of contamination monitoring practices was excellent
and showed improvement.
This was evidenced
by the elimination of multiple
radiological control area (RCA) egress points, and very close health physics
oversight of personnel
and equipment contamination monitoring at the Unit 2
RCA access/egress
point.
R1.3
External Ex osure Controls
a.
Ins ection Sco
e 83760
The inspector performed
a review of external exposure controls including use of
radiological boundaries,
drywell shielding, and drywell pipe flushing.
Information
was gathered through tours of the facility, review of results of the effectiveness of
installed shielding and pipe flushes, and through discussions with cognizant
personnel.
b.
Observations
and Findin s
The inspector toured the Unit 1 reactor building including the Unit 1 drywell and
refuel floor. Radiological boundaries
were well defined, and based on a review of
radiological surveys, radiation areas and high radiation areas were properly posted.
The inspector noted good use of informational posting including dose rate range
posting, and Unit 1 drywell maps that identified areas with elevated dose rates and
low dose areas.
During tours of the Unit 1 drywell, the inspector observed multiple temporary
shielding installations.
Approximately 24,000 pounds of temporary lead shielding
and associated
hardware were installed in the Unit 1 drywell. Shielding packages
were well designed
and installed, appeared
neat and orderly, and provided evidence
of detailed planning.
Based on a review of radiation surveys,
a pre- and post-
shielding dose rate summary, and selective radiation dose rate verification surveys
performed by the inspector, the inspector concluded that the drywell shielding
21
program was well planned and managed,
and shielding packages
were effective in
reducing general area dose rates.
The inspector also performed
a selected review of the effectiveness of drywell
system flushes.
The inspector reviewed an outage schedule
and noted that specific
"flushes," designed to reduce radiation dose rates in the drywell, were integrated
into the outage schedule.
The inspector reviewed documentation
of pre- and post-
radiation survey results for the "flushes," and noted significant dose rate reductions
were achieved.
Dose rates at a distance of 30 centimeters were reduced on the B
recirculation pump drain line from 1000 mrem/h to 150 mrem/h; on B core spray
piping from 1000 mrem/h to 80 mrem/h, and dose rates at the N2F recirculation
discharge
nozzle were reduced from 6000 mrem/h to 600 mrem/h.
Based on this
review, the inspector concluded that drywell flushes were well planned
and
effectively implemented.
b.
Conclusion
Based on this review, the inspector made the following con'elusion.
~
External exposure controls were excellent including use of informational
postings, use of temporary shielding, and ALARAsystem flushes.
R1.4
Use of Flashin
Li hts for Access Control to Hi h Radiation Areas
a,
Ins ection Sco
e 83750
The inspector reviewed the use of flashing lights for access control to high radiation
areas.
The inspector gathered information by reviewing Technical Specification 6.12, "High Radiation Area;" Nuclear Department Administrative Procedure
NDAP-
00-0626, "Radiologically Controlled Access and Radiation Work Permit (RWP)
System," Rev. 4; procedure
HP-TP-311, "Locking, Barricading, 5 Key Control," Rev.
11; procedure
HP-TP-310, "Posting and Labeling," Rev. 16; NRC Information Notice 88-79, "Misuse of Flashing Lights For High Radiation Area Controls," 10/7/88;
radiological survey data for the Unit
1 drywell and 704'levation equipment space.
The inspector also performed tours through the Unit 1 drywell, and discussed
high
radiation area access controls with cognizant personnel.
b.
Observations
and Findin s
Technical Specification 6.12, "High Radiation Area," and procedural guidance,
requires areas with dose rates greater than or equal to 1000 mrem per hour to be
controlled with barricades
and locked doors.
In the event the area is located within
a large area where no enclosure could reasonably
be constructed,
then access
control is achieved by roping the area off, conspicuously
posting the area, and
activating a flashing light as a warning device.
i
22
The inspector noted that, in general, access control to high radiation areas with
dose rates greater than 1000 mrem per hour were controlled with barricades
and
locked doors.
However, access to several locations within the Unit 1 drywell and
one location in the Unit 1 reactor building, 704'levation equipment space, with
dose rates in excess of 1000 mrem per hour were controlled by roping the area off,
posting the area as a high radiation area, and activating a flashing light.
For
example, due to hydrolasing/flushing
activities performed on the Unit
1 scram
volume discharge
a high radiation area with dose rates of 1400 mrem per
hour was created in piping located in the Unit 1 704'levation equipment space.
The radiological controls staff initiated plans to flush the piping to remove the
elevated dose rates, and during the interim, controlled access with a rope boundary;
a high radiation area posting; a flashing warning light; a radiation work permit; use
of alarming dosimetry; radiological briefings; and health physics monitoring of
personnel
accessing
the larger equipment space that surrounded
the area.
The inspector was informed that other effective measures
for controlling access to
this area had been considered
(e.g., locking the outer 683'quipment
space doors,
or installing locking ladder blocks).
However, the use of flashing warning lights
combined with additional access controls was thought to be reasonable
while
providing positive access control.
Licensee staff also informed the inspector that a radiological controls information
document
(Rad Safety Note) had been prepared for distribution to all station
personnel
~ This was intended to serve as an additional training method to inform
personnel of the use of flashing warning lights to control access to high radiation
areas.
The inspector noted this as a good initiative.
The inspector concluded that the licensee's
use of flashing warning lights combined
with radiation work permit controls for access to high radiation areas was
reasonable
and represented
positive exposure control.
No examples of improper
entries into high radiation areas controlled with flashing warning lights were
identified.,
C.
Conclusions
Based on this review, the inspector made the following conclusion.
~
The use of flashing warning lights combined with radiation work permit
requirements to control access to high radiation areas, was reasonable,
represented
positive exposure control, and was in compliance with Technical Specification 6.12, "High Radiation Area."
R1.5
Res
onse to Alarmin
Dosimetr
a.
Ins ection Sco
e 83750
The inspector performed
a review of licensee response to alarming personnel
dosimetry.
Information was gathered
by a review of HP-TP-126, "Use of the Real
23
Time Dose Tracking System," Rev. 3; interviews with health physics technicians;
and a review of actions taken in response
to personnel dosimetry that alarmed
during Unit 1 control rod drive exchanges
performed the morning of September 25,
1996.
Observations
and Findin s
During a review of the radiological controls implemented for Unit 1 control rod drive
(CRD) exchanges,
the inspector was informed that on the morning of September
25, 1996, several individuals received alarms on their personal dosimetry while
working on CRD exchanges.
The inspector performed a review to determine if
appropriate actions were taken in response to alarming dosimetry.
Health physics procedure
HP-TP-126, "Use of the Real Time Dose Tracking
System," Rev. 3, requires workers to evacuate their immediate work area and notify
Health Physics whenever their dosimetry alarms.
The procedure
also requires health
physics to determine the cause of the alarm.
'I
On the early morning of September
25, 1996, two maintenance
workers were
located on the Unit 1 drywell subpile room carousel,
and were involved in CRD
exchanges.
A senior health physics technician was providing constant health
physics coverage,
and undervessel
activities were being monitored remotely with
audio and video communications
by several maintenance
personnel,
and a
representative
from the operations department.
Undervessel
maintenance
personnel
were working under radiation work permit (RWP) 1996-1356,
"CRD Exchange:
Undervessel Work," personal alarming dosimetry (PAD) had been relocated to the
head (location of whole body exposed to the highest dose rate), and the PAD dose
limit was automatically set at 400 mrem during RWP sign-in.
According to the health physics technician that provided constant coverage for the
job, on the morning of September 25, 1996, two maintenance
workers had been in
a 200 mrem/h area for approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when one of the individuals PAD
alarmed.
The alarm was immediately detected
by the health physics technician in
the field, and by the maintenance
and operations
CRD support team that were
remotely monitoring activities by video.
The CRD support team was not sure if the
individual heard the PAD alarm, so they informed the individual that he had received
an alarm on his dosimetry, and instructed the individual's to contact health physics.
The health physics technician responded
to the alarming dosimetry by performing a
radiation survey to identify any unusual dose rates in the work area; no unusual
dose rates were found.
The HP technician then checked the individual's dosimetry
and determined that the dosimeter alarmed because
the 400 mR alarm set point had
been exceeded.
The technician then concluded that no unusual conditions existed
and that the 400 mR PAD alarm setpoint had been exceeded
because
the individual
had worked in a 200 mR/h area for approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
The HP technician then
communicated
his findings by radio to HP supervision.
supervision then determined that since they knew the reason for the alarm, the
individual was not in danger of exceeding
an administrative or regulatory dose limit,
and constant health physics coverage was being provided, then the individual
0
should continue working to place his work in a safe condition.
The HP technician
then informed the maintenance
technician to continue working to place the CRD
exchange
machine in a safe condition, and then to leave the work area.
While
exiting the area, the second maintenance
technician received an alarm on his PAD.
The highest PAD exposure recorded was 440 mR.
The health physics technician informed the inspector that the PAD alarm dose
setpoint was used as a tool to identify when workers exceeded
an established
dose,
and not as a dose limit.
The inspector noted that the actions taken in response
to the PAD alarm were in
compliance with procedural guidance
and were appropriate.
The individual with a
dosimeter in an alarm status notified the HP technician; the HP technician
determined the cause of the alarm and informed HP supervision;
HP supervision
determined that the individual was not in danger of exceeding
an administrative or
regulatory dose limit, and authorized the HP technician to allow the workers to
remain in the area in order to place their work in a safe condition; and the
maintenance
workers placed their equipment in a safe condition, and left the area.
C.
Conclusions
Based on this review, the inspector made the following conclusions.
Procedural guidance for personnel
responding to alarming dosimetry was
adequate.
Personnel
actions taken in response to alarming dosimetry were very good.
This included notification of health physics; performance of radiological
surveys to determine the cause of PAD alarms; evaluation of individuals dose
status after initiation of a PAD alarm; and health physics instructions to
personnel
concerning the need to evacuate work areas after a PAD alarm
occurred.
R2
Status of RP8cC Facilities and Equipment
a.
Ins ection Sco
e
The inspector toured the Unit 1 and 2 reactor buildings, the Unit 1 drywell, and the
Unit 1 turbine building to evaluate radiological control boundaries
and housekeeping.
b.
Observations
and Findin s
In general, housekeeping
was good and radiological boundaries
were clearly
delineated
and well maintained.
Housekeeping
and radiological boundary
deficiencies were identified on 729'levation of Unit 1 turbine deck, and at the
683'ntrance
to the Unit
1 suppression
pool.
Examples included unsecured
cables
and cords draped across contaminated
area boundaries,
materials stored in clean
areas with edges partly in a contaminated
area, and miscellaneous
supplies,
25
materials, and trash lying in aisles and walkways.
Upon notification, radiological
boundary deficiencies were corrected,
and improvements were made in
housekeeping.
Finally, no obvious signs of degraded
material conditions in rooms or
equipment were identified.
C.
Conclusions
Based on this review, the inspector made the following conclusion.
~
Overall conditions of housekeeping,
and radiological boundaries
in the Unit
1
and 2 Reactor buildings, the Unit 1 Drywell, and the Unit 1 Turbine building
were good.
R5
Staff Training and Performance in RP8cC
a.
Ins ection Sco
e 83750
The inspector performed
a review to determine if lessons
learned and industry
events were being incorporated into health physics (radworker) training, and if
contract health physics technicians brought in to support the Unit
1 outage were
appropriately trained and qualified.
Information was gathered
by a review of lesson
plans, resumes
and training records, and through discussions
with cognizant
personnel.
b.
Observations
and Findin s
The inspector reviewed lesson plan HP-002, "Health Physics Level II," Rev. 5, and
noted that multiple examples of lessons
learned and industry events were included
in training provided to workers who were permitted unrestricted access to the
ra8iologically controlled area (RCA).
The inspector was informed that 87 contract health physics technicians,
each with
at least two years experience
had been brought in to support the Unit 1 refueling
outage.
Nine names were randomly selected from a list of contract health physics
technicians,
and the inspector requested
documentation
of qualification and training.
Based on a review of resumes,
each of selected individuals had greater than two
years experience
as a health physics technician.
Records showed that all selected
individuals had passed
the Susquehanna
Health Physics entrance exam or passed
the Northeast Utilities Health Physics exam within two years.
Finally, records of
procedure
and task qualification sign-offs were available and complete.
C.
Conclusions
Based on this review, the inspector made the.following conclusions:
26
~
Radworker training was excellent in that lessons
learned and industry events
were included in training provided to workers permitted unrestricted access
to the RCA.
~
Health physics technicians brought in to support the Unit 1 refueling outage
were qualified and appropriately trained, and training records were excellent.
R6
RPSC Organization and Administration
a.
Ins ection Sco
e 83750
The inspector reviewed the organization and administration of the health physics
organization during the refueling outage.
Information was gathered by a review of a
health physics and ALARAorganization chart, attendance
at the night shift turnover
meeting, observations
in the plant, and through interviews with cognizant
personnel.
b.
Observations
and Findin s
Health physics oversight of work and contractor health physics technician
performance,
were primarily accomplished
by upgrading experienced
Level II Health
Physics Technicians to the position of Assistant Foreman-Health
Physics, and
assigning these individuals to key outage work areas or projects.
For example, day
and night shift area coordinator positions were established for the following: Shift
Supervisor, Drywell, Refuel Floor, Reactor Building, Balance of Plant, Radwaste,
and
Maintenance
Pool. Day shift area coordinator positions were also established for the
Turbine Building and Training.
The inspector attended two shift turnover meetings,
interviewed six area coordinators,
and observed
health physics communications
in
the plant.
Assistant Foremen were extremely knowledgeable of planned work,
available resources,
and radiological conditions and controls for the areas they were
assigned to.
In addition, Assistant Foremen maintained close oversight of contract
health physics technicians
and provided appropriate directions when needed.
The
inspector concluded that the process for upgrading experienced
Level II health
physics technicians to the position of Assistant Foreman-Health
Physics, in order to
allow for better coordination and control of outage work, and better oversight of
contract health physics technician performance was effective, and noted as a
program strength.
Similarly, the ALARAstaff was increased to a total of five positions during the
outage with support from the corporate technical staff, technical training, and
health physics.
ALARAstaff members were assigned
responsibility for ALARA
planning and coordination on the refuel floor, for in-service inspection and valve
work, drywell work and shielding, and for balance of plant work.
Based on
interviews, the inspector concluded that these individuals were knowledgeable
of
planned work, and were effectively coordinating and implementing ALARAcontrols.
27
C.
Conclusions
Based on this review, the inspector made the following conclusion.
~
The process of upgrading Health Physics Level II technicians to Assistant
Foremen for outage work allowed for very good oversight of work and
contract health physics technician performance.
~
ALARAstaff members were knowledgeable
of planned work, and were
effectively coordinating and implementing ALARAcontrols.
R7
Quality Assurance in RP&C Activities
Ins ection Sco
e 83750
The inspector reviewed quality assurance
and self assessment
oversight of the
radiological controls organization.
Information was gathered by a review of Audit
96-007, "Health Physics Program," April 8, 1996, an "Assessment
of 1996 Trends
in Condition Reports," dated September
3, 1996, a selected review of condition
reports, and thorough interviews with cognizant personnel.
Observations
and Findin s
Audit 96-007, "Health Physics Program," dated April 8, 1996, was performed from
February 12- March 11, 1996, and included a review of radiological surveys,
processing
and controls, respiratory protection, internal and external dosimetry,
personnel training, and radiological postings.
The audit was conducted
as a "limited
scope" audit, because
other health physics program areas were being addressed
in
other audits/assessments
(e.g., NAS-Independent
Safety Evaluation Services
Evaluation No.95-041).
The audit concluded that the health physics program was
being effectively implemented.
One finding and ten observations/recommendations
for consideration were identified. The finding noted that six Site Modification Group
(SMG) Engineers
had not received Engineering ALARATraining (course number
EG011).
Resolution of this finding was being addressed
and tracked by Condition
Report No. 96-0219.
The inspector noted that audit observations/recommendations
were insightful and well developed.
Based on a review of the audit results, the
inspector concluded that audits were effectively used to identify program
deficiencies.
The inspector also reviewed an assessment
performed to evaluate trends in
condition reports (CRs) related to radiological controls entitled "Assessment of 1996
Trends in Condition Reports," dated September
3, 1996.
This report indicated that
from March 5, 1995, to December 31, 1995, 789 CRs were written.
Of these, only
12 (1.5%) concerned
radiological protection.
From January
1, 1996, to August 23,
1996, 1294 CRs were written. Of these, [178] (=14%) concerned
radiological
protection.
This indicates
a dramatic increase
in the generation of CRs related to
radiological protection.
The significance of the HP CRs written in 1995, and in
1996 from January to August 23, 1996, were evaluated
using a four-level
0
28
significance model.
The model assigned
significance to CRs using the following
categories:
Level 1-Significant, Level 2-Important, Level 3-Minor, and Level 4-Not a
Deficiency. The licensee reported that the majority of the 1995 CRs would have
been assigned to the Level
1 and Level 2, significance categories.
This was
compared to 178 HP CRs initiated from January to August 23, 1996, with the
majority being in the Level 3 and Level 4 significance categories:
Level 1-3%, Level
2-11%, Level 3-68%, and Level 4-18%.
The license also correlated the time at
which the increase
in the HP CR generation rate occurred to management
efforts to
"lower the threshold" for initiating condition reports.
Based on this data, the
licensee concluded that increases
in the rate of generation of HP condition reports
were attributed to management
efforts to lower the initiation threshold for CRs; the
majority of HP CRs were of lower safety significance; and the volume of HP CRs
was not indicative of an adverse trend in the significance of deficiencies in the
radiological protection program.
Based on a review of 1996 condition reports, and discussions with Health Physics
technicians,
and Quality Assurance
and Health Physics supervision, the inspector
concluded that the overall increase
in numbers of condition'reports appeared
to be
the result of a lowering of the threshold for initiating condition reports.
However,
the number of condition reports related to deficiencies in high radiation area
barricades
and postings had increased
since 1995
~ Based on a review of a list of
condition reports for 1996, the inspector identified approximately 22 condition
reports pertaining to deficiencies in barricading, postings, or locking doors to high
radiation areas, that were written between the period of January
1, 1996 to July
31, 1996.
Although none of these events have had a significant radiological
consequence
(e.g., overexposure
or significant unplanned
exposure), the number of
these events suggest
an increase in the significance of health physics related
condition reports from the year 1995 to the year 1996.
The inspector also selectively reviewed 25 condition reports related to radiological
controls.
The inspector noted that the threshold for initiation of CRs appeared
low,
that causal factors and root causes were being identified, corrective actions were
being evaluated
and identified, and corrective actions were being implemented.
Based on this review, the inspector concluded that the condition reporting system
was very effective.
The inspector did identify one condition report, involving an un-
barricaded
and un-posted
access to a high radiation area, that warranted additional
review (CR-1371).
The NRC review of this condition report appears
in Section R8.5
of this report.
Conclusions
Based on this review, the inspector made the following conclusions.
~
Quality assurance
audits were broad in scope,
and effectively used to
identify program deficiencies.
29
~
The increase
in numbers of condition reports related to radiological controls
appears to be related to management
efforts to lower the threshold for
writing condition reports.
~
Although none of the condition reports that documented
deficiencies in high
radiation area postings,
barricades,
or locked doors had significant
radiological consequences
(e.g., overexposure
or significant unplanned
exposure), the number of these events suggest
an increase
in the
significance of health physics related condition reports from the year 1995 to
the year 1996.
~
The condition reporting system was effectively used to identify, evaluate,
and resolve program deficiencies.
R8
IVliscellaneous RP&C Issues
R8.1
Transverse
Incore Probe
TIP Work In December 1995
a.
Ins ection Sco
e 83750
The inspector reviewed the radiological controls implemented for work performed in
December 1995, on the transverse
incore probe (TIP) drive system.
The inspector
reviewed the radiological controls implemented on this job through interviews with
cognizant personnel
and reviews of procedures.
b.
Observations
and Findin s
In late 1995, Susquehanna
Steam Electric Station had entered
a four-hour technical
specification limiting condition of operation
(LCO) due to an inability to isolate
containment.
A traversing incore probe (TIP) drive was stuck and had to be
manually retracted into it's shielded location.
Dose rates from the TIP drives have
the potential to be very high and have created problems in other plants.
The
inspector noted that the radiological work performed on the TIP system was
performed using a general radiation work permit (RWP), and not under a specific
RWP which would include specific radiological requirements.
The inspector noted
that NDAP-00-0626, Rev 5, "Radiologically Controlled Area Access and Radiation
Work Permit (RWP) System," Section 6.2.2.b (2) required a job specific RWP,
approved by the Radiological Operations Supervisor, for access to areas with dose
rates greater than 10,000 mrem/h.
However, Section 6.2.2.b, of the same
procedure,
states the following.
"NOTE:
In unusual conditions as determined by Shift Supervision,
a
Qualified/Certified Level II Health Physics Technician providing
constant coverage with a dose rate meter, may be used in lieu of a
job specific RWP."
Based on interviews, it was determined that shift supervision requested
health
physics to support TIP drive box work during a four hour LCO; a qualified Level II
30
health physics technician was assigned to provide constant coverage for the work;
the health physics technician used
a dose rate meter; and the technician signed-on
to a general RWP.
In addition, it was determined that radiological briefings were
performed, remote radiation monitoring equipment were utilized, and supervisory
personnel provided constant oversight towards the end of the job.
Based on
discussions
with cognizant personnel,
radiological control measures
were
appropriate
and no adverse consequences
occurred.
Conclusions'he
inspector concluded that although advanced
planning and use of a specific
radiation work permit for jobs with the potential for high radiation levels are
desirable, the work performed on the traversing incore probe (TIP) system in late
1995, was performed in accordance
with approved procedures,
and appropriate
radiological control measures
were implemented.
No violations of NRC
requirements were identified.
Reactor Water Clean-u
RWCU Pum
Work In Januar
1996
Ins ection Sco
e 83750
The inspector reviewed the radiological controls implemented on a job performed in
January of 1996, involving the disassembly
and inspection of the Unit 1 reactor
water cleanup (RWCU) pump and discharge check valve.
Specifically, the inspector
reviewed hot particle (discrete radioactive particle) controls, use of respiratory
protection, and air sampling.
Information was gathered by a review of radiation
work permit (RWP) number 1996-0045, revisions 0, 1, 2, 3, and 4; ALARA
reviews; radiological survey data; and interviews with cognizant personnel.
Observations
and Findin
s
The inspector reviewed the following radiation work permit revisions.
Hot Particle
Zone
~Yee No
Rev.
RWP No.
RWP Deecri tion
No
0
1996-0045
Disassemble
RWCU pump motor and discharge
No Pump Breach on this Revision
Yes
No
1
1996-0045
Disassemble
and inspect RWCU motor and pump
2
1996-0045
Disassemble
and inspect RWCU motor and pump
Yes
3
1996-0045
Disassemble
and inspect RWCU motor and pump
Yes
4
1996-0045
Disassemble
and inspect RWCU motor and pump
31
The inspector noted that hot particle control zones were used on RWP 19S6-0045,
revisions 1, 3, and 4; but not on revisions 0 and 2. Health physics procedure
HP-
TP-511, "Hot Particle Controls," Rev. 2, provided guidance for monitoring and
controlling hot particles (discrete radioactive particles).
The procedure was
designed to implement controls to minimize the transfer of hot particles to
personnel,
equipment and work areas,
and to account for exposures
received from
hot particles,
HP-TP-511, Rev. 2, Section 9.1.1 stated that "hot particle control
zones shall be set up to minimize the spread of hot particles...and
should include
primary/secondary
contaminated
system breaches
when there is a potential for Hot
Particle exposure
as directed by HP supervision."
The inspector noted that although
jobs such as RWCU pump breaches
have the potential for hot particle exposure,
HP-
TP-511 did not specifically require a hot particle zone to be established for RWCU
pump breaches;
it required HP supervision to direct when a hot particle control zone
was necessary.
During interviews, the inspector was informed that Rev. 0 of RWP 1996-0045, did
not require hot particle zone controls because it did not involve a pump breach; Rev.
1 required hot particle zone controls because
it involved a pump breach; Rev. 2 did
not require hot particle zone controls because
no hot particles were found during
the initial pump breach; and Revs. 3 and 4 required hot particle zone controls
because
high contamination
and fine, dust-like metal shavings were encountered.
Based on a review of radiological survey data, the inspector noted that although
high contamination was present (e.g., 2,400 mrad per hour per 100 cm'n the
pump flange on January 22, 1996), no discrete hot particles were identified during
the course of the work.
The inspector noted that it is generally prudent to use additional contamination
controls, similar to hot particle controls (e.g., use of contamination buffer zones and
additional contamination monitoring of personnel
and areas), for jobs involving high
contamination, with the presence
of very fine, dust-like metal shavings.
Although
the elimination of hot particle controls on RWP 1996-0045, Rev. 2, showed
a lack
of prudence,
no procedural violations or consequential
radiation exposures
were
identified.
The inspector also reviewed use of respiratory protection for RWCU pump work.
"Disassemble
and inspect RWCU motor and pump," Rev. 1,
required use of decontamination,
wetting areas to suppress
dust generation,
and
use of a high efficiency particulate air (HEPA) filtration unit to minimize airborne
radioactivity, and eliminate the need for respiratory protection.
During removal of
the pump diffuser and impeller, an air sample was obtained which indicated an
airborne radioactivity concentration
up to 0.66 times the derived air concentration
(DAC) of radionuclides for occupational
exposure.
Based on these results, on
January
19, 1996, an "ALARAIn-progress Review" was performed, which
documented
the need for respiratory protection.
A filter respirator with a protection
factor of 50 was selected for RWP 1996-0045, Rev. 2, rather than a powered air
purifying respirator (PAPR) with a protection factor of 1000, because
a protection
factor of 50 was thought to be adequate,
and concerns were raised that the PAPR
32
could "mask" heat stress symptoms.
On January 22, 1996, a 53 4 DAC-hr air
sample was obtained during the job. The RWP was then revised (Rev. 3) to include
the use of PAPRs when contamination was expected to exceed
10 DAC-hrs.
10 CFR 20.1703, "Use of individual respiratory protective equipment," requires that
if respiratory protection equipment is used, the licensee
is to select respiratory
protective equipment that "provides a protection factor greater than the multiple by
which peak concentrations
of airborne radioactive materials in the work area are
expected to exceed the values specified in Appendix B to 10 CFR 20, Table 1,
column 3." In addition,
~.."ifthe exposure
is later found to be greater than
estimated, the corrected value must be used; if the exposure
is later found to be
less than estimated, the corrected value may be used."
Based on a review of RWP and radiological survey records, and discussions with HP
staff members, the inspector concluded that although air sample results exceeded
initial estimates, the application of engineering controls and selection of respiratory
equipment were reasonable
under the circumstances.
'I
The inspector also reviewed personal exposure records to evaluate the airborne
radioactivity exposure assessments.
Computer records showed that each of the
individuals that wore a filter respirator and was exposed to an airborne
concentration of 53.4 DAC-hrs received
a whole body count to evaluate uptake of
radioactive materials, and to assign internal dose.
One individual was assigned
an
internal dose of 3 mrem, two other individuals were assigned
an internal dose of 0
mrem.
The inspector noted that the licensee elected to use the whole body count
data for internal dose assessment.
Finally, the inspector reviewed the adequacy of airborne radioactivity sampling
practices for areas outside of the RWCU pump room.
Through interviews, the
inspector was informed that during Unit 1 RWCU pump work, in January 1996, a
continuous air monitor (CAM) was stationed on 749 foot elevation of the reactor
building.
However, the CAM was located in the eastern north-south hallway, and
not immediately adjacent to the entrance to the Unit 1 B RWCU pump room.
Licensee staff acknowledged that the location of the CAM was not ideal for
verifying that airborne radioactivity areas were limited to areas inside of the RWCU
pump rooms.
However, the staff explained that they had concluded that an
airborne radioactivity area did not exist outside of the entrance to the RWCU pump
room based on the following.
Negative pressure
air flow was maintained by plant ventilation, for the
majority of time, at the entrance to the RWCU pump rooms.
Throughout the
job, negative pressure
air flow was monitored with a hanging streamer.
Note:
Negative air pressure flow was lost for a period of time during non-
work activities.
Upon notification, the operations department performed an
air balance to restore negative air flow.
a 500 cubic feet per minute (CFM) HEPA was used inside the room to clean
airborne radioactivity as it was being generated.
33
High contamination
levels were not found in areas adjacent to the RWCU
pump room egress point.
Several thousand dpm/100 cm'as identified on
the step-off-pad at the egress point, and 1000 dpm/100 cm'as found on a
large area smear taken in clean areas surrounding the egress point.
Licensee
staff attributed this to contamination that was tracked out of the room,
rather than airborne radioactivity migration.
No individuals working in clean areas immediately outside of the RWCU
pump rooms, were found to have clothing contamination.
Based on this information, the inspector concluded the following.
The continuous air monitor located on the Unit 1 reactor building
749'levation,
eastern north-south hallway, was not positioned in a location to
allow immediate notification of an airborne radioactivity condition near the
entrance to the B RWCU pump room.
Considering the low contamination levels found at the egress point to the
Unit 1 749'B" RWCU pump room, it is not likely that an airborne
radioactivity area existed outside of the RWCU pump room entrance.
Based on a review of the ALARApost job review for RWP 1996-0045, and through
discussions with health physics supervision, the inspector was informed that
lessons learned during RWP 1996-0045 were planned to be incorporated into future
RWCU pump jobs planned for Unit 2.
c.
Conclusions
Based on this review, the inspector made the following conclusions.
Although the elimination of hot particle controls on RWP 1996-0045, Rev. 2,
showed
a lack of prudence,
no procedural violations or consequential
radiation exposures
were identified.
Although air sample results exceeded
the protection factor of the respirator
initially selected for RWCU pump work, the application of engineering
controls and selection of respiratory equipment were reasonable
under the
circumstances.
Internal dose assessments
and assignments,
following the use of respiratory
protection equipment with a protection factor less than the measured
airborne DAC-hr concentration,
were appropriate.
0
Pre-job ALARAand RWP planning for planned work in the Unit 1 B RWCU
pump room, which required the application of engineering controls including
use of water for dust suppression,
use of HEPA ventilation and containment
(e.g., maintenance
of negative pressure
air flow into the room) during RWCU
pump work, represented
a sufficient survey that was reasonable
under the
34
circumstances to conclude that an airborne radioactivity area would not exist
outside of the B RWCU pump room during pump work.
~
Considering the low contamination levels found at the egress point to the
Unit 1 749'B" RWCU pump room, it is not likely that an airborne
radioactivity area existed outside of the RWCU pump room entrance.
R8.3
Closed
Violation 50-387 388 96-04-02
The inspector performed
a review to evaluate licensee response to NRC Violation
50-387;388/96-04-02.
The inspector discussed
corrective actions taken with
various members of the health physics staff, and reviewed the following June 24,
1996, Reply to Notice of Violation 50-387/96-04-02 5 50-388/96-04-02.
The inspector verified that specific corrective actions described
in the licensee's
response
letter, dated June 24, 1996, were complete.
The inspector noted that
appropriate corrective actions were taken when radiological posting deficiencies
were identified, and based on licensee reviews, no unplann'ed radiation exposures
resulted from the radiological posting deficiencies.
This item is closed.
RBA
Closed
Unresolved Item 50-387.388 96-09-02
Note:
During NRC Inspection Nos. 50-387;388/96-09, the inspector noted that four
additional radiological posting deficiencies were identified in the station's condition
reporting system: one radiation area posting deficiency, and three high radiation
area posting/barricading
deficiencies.
An unresolved
item, URI 50-387;388/96-09-
02, was opened pending further NRC review of the licensee's review and evaluation
of these events.
The NRC evaluation of these deficiencies appears
as follows.
An unposted
radiation area was identified on July 22, 1996, at the entrance
to the Unit 2, 645'
residual heat removal (RHR) pump room.
The majority
of the room was controlled as a high radiation area.
However, a small area
inside the northeast door, leading to the high radiation area, with dose rates
as high as 12 mrem/h was not posted as a radiation area.
This condition
was corrected by placing a radiation area posting on the outer entrance door,
and documenting the occurrence
in the station condition reporting system
(CR-0956).
This failure constitutes
a violation of minor significance and is
being treated as a Non-Cited Violation, consistent with Section IV of the NRC
Enforcement Policy. This item is closed.
Upon further review, it was determined that the three high radiation area
posting/barricading
deficiencies had greater significance, and are considered
a violation of NRC requirements.
The review of this issue appears
in Section
R8.5 of this report.
Accordingly, this unresolved item is being
administratively closed.
35
R8.5
0 ened
Violation 50-387 388 96-10-03
"Radiolo ical Postin
Deficiencies"
During NRC Inspection Nos. 50-387;388/96-09, the inspector reviewed the
following condition reports that documented
deficiencies in barricading and posting
the access to high radiation areas.
CR 96-508
Unit 1 Turbine Building 676'. On May 5, 1996, the entrance to the D
Demin Room from the E Demin Room was not posted
as a high
radiation area, and had dose rates of 200 mrem/h.
CR 96-535
Unit 1 Turbine Building 676':
On May 11, 1996, the entrance to the
Steam Jet Air Ejector (SJAE) Room from the spare SJAE room was
not posted
as a high radiation area, and had dose rates of 1,200
mrem/h.
CR 96-1056
Unit 2 Reactor Building 779':
On July 31, 1996, an un-'posted
high
radiation area of 400 mrem/h was found originating from the HV-
24511B resin inlet valve.
An unresolved item, URI 50-387;388/96-09-02,
was opened
pending the licensee's
review and evaluation of these events.
During the followup review of these
condition reports, the inspector noted that an investigation was performed for each
of these events; safety assessments
concluded that each of these events had low
actual safety consequences;
causes
and causal factors were identified; corrective
actions and actions to prevent recurrence were identified; and a review of past
performance was performed.
These investigations were generally good.
However, the inspector noted that corrective actions implemented to address the
failure to post the access to the Unit 1 676'levation
D Demin Room from the E
Demin Room, were not effective in preventing the May 11, 1996, event in which
the entrance from the spare SJAE room was found blocked open with a hand truck.
In addition, the inspector noted that the July 31, 1996, event involving a high
radiation area found originating from the HV-24511 B valve was a repeat
occurrence.
Further, the inspector identified another condition report that
warranted review due to a similarity to these events.
Condition report number CR
96-1371 was initiated on September
6, 1996, when the door to the Unit 2 Turbine
Building 729'oisture separator room, a high radiation area with dose rates as high
as 800 mrem/h inside the room, was found propped open.
This effectively left the
access to the high radiation area un-barricaded
and un-posted.
Although a followup
review concluded that no unauthorized
entries were made into the room during the
time in which the door was left open and un-posted, this event provides an
indication that weaknesses
exist in the radiological posting and access control
program, and that corrective actions implemented to address
previous radiological
posting and barricading deficiencies, have not been effective in preventing
recurrence.
Due to the repetitive nature of these events, and the ineffectiveness of corrective
actions, collectively, these examples of radiological posting and barricading
36
deficiencies are considered
a violation of NRC requirements,
specifically, Technical Specification 6.12, "High Radiation Area," which requires each high radiation area
to be barricaded
and conspicuously
posted as a high radiation area.
{VIO50-
387;388/96-10-03)
S8
Miscellaneous Security and Safeguards
Issues
S8.1
U date
EA 95-250
Securit
Chillin
Effect
U date
EA 94-212
Securit
Chillin
Effect
The inspector reviewed the licensee's corrective actions for the events that led to the NRC
escalated
enforcement and Department of Labor actions.
The corrective actions committed
to by the licensee were in place in the Security Department including: management
reassignments,
mandatory management
and employee training, and improvements
in the
SSES Employee Concerns Program.
In addition, the inspector verified the
existence/conduct
of the Security Issues Team, which is a peer interaction process
intended to identify employee concerns to Security line management
as an alternative to
the PPSL Employee Concerns Program.
The inspector concluded that the implementation
of the corrective actions was acceptable.
The NRC will evaluate the effectiveness of these
measures
to prevent a chilling effect on raising safety concerns to management
during
future inspection and assessment
activities.
Y. !Vlana ement Meetin s
X1
UFSAR Review
A recent discovery of a licensee operating their facility in a manner contrary to the
Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a
special focused review that compares
plant practices, procedures
and/or parameters
to the UFSAR description.
While performing the inspections discussed
in this
report, the inspectors reviewed the applicable portions of the UFSAR that related to
the areas inspected.
The inspector reviewed selected sections of Chapters
12.1 - 12.5, "Radiation
Protection," of the Updated Final Safety Analysis Report (UFSAR), pertaining to
radiological controls, to evaluate the accuracy of the UFSAR regarding existing plant
conditions and practices.
No UFSAR discrepancies
were identified during this
review.
0
37
The inspectors presented
the inspection results to members of licensee management
at the
conclusion of the inspection on October 21, 1996.
The licensee acknowledged the
findings presented.
The inspectors asked the licensee whether any materials examined during the inspection
should be considered
proprietary.
No'roprietary information was identified.
INSPECTION PROCEDURES USED
IP 62707:
IP 71707:
IP 73051:
IP 83750:
IP 92700:
IP 92902:
Maintenance
Observation
Plant Operations
Inservice Inspection - Review of Program
Occupational Radiation Exposure
Onsite Followup of Written Reports of Nonroutine Events at Power Reactor
Facilities
Followup - Engineering
~Oened
50-387;388/96-1 0-01
50-387;388/96-1 0-02
50-387;388/96-1 0-03
Closed
ITEMS OPENED, CLOSED, AND DISCUSSED
Weaknesses
during Unit 1 Restart
Control Rod Drive (CRDI Mechanism Replacement
Failure to post and barricade high radiation areas in
accordance
with Technical Specification 6.12
50-387/96-007-00
50-388/95-1 2-01
50-387/95-05-01
50-387;388/96-04-02
50-387;388/96-09-02
Discussed
LER
Loss of 4 Kv Bus
HPCI On-line Maintenance
Observation Of Activities On The Refueling Floor
High radiation area posting discrepancies
Failure to implement procedural posting requirements
for a radiation area and high radiation areas
50-387;388/96-06-01
EA 95-250
EA 94-212
Containment Bypass Leakage
Security Chilling Effect
Security Chilling Effect
LIST OF ACRONYMS USED
CFR
CFR
CR
dpm
gpm
IR
ISES
LCO
LER
mR/h
mR
mrem/h
mrem
NRC
Ol
PCC
scfh
TCM
TS
US
As Low As Is Reasonably Achievable
Continuous Air Monitor
Cubic Feet per Minute
Code of Federal Regulations
Code of Federal Regulations
Combined. Intermediate Valves
Condition Report
Control Rod Drive
Control Rod Drive Mechanism
Control Room Emergency Outside Air Supply System
Derived Air Concentration
disintegration per minute
Escalated Action
Final Safety Analysis Report
gallons per minute
High Efficiency Particulate Air
Health Physics
High Pressure
Coolant Injection
Inspection Report
Independent
Safety Evaluation Services
Limiting Condition of Operation
Licensee Event Report
milliRoentgen per hour
milliRoentgen
millirem per hour
millirem
Non-Cited Violation
Nuclear Regulatory Commission
NRC Office of Nuclear Reactor Regulation
Office of Investigations
Operations with the Potential for Draining the Reactor Vessel
Personal Alarming Dosimeter
Powered Air Purifying Respirator
power control center
Personnel
Contamination Monitor
Plant Operations Review Committee
Reactor Core Isolation Cooling
Radiation Work Permit
standard cubic foot/feet per hour
Tool Contamination Monitor
Transverse
Incore Probe
Technical Specification
Updated Final Safety Analysis Report
Unresolved Item
Unit Supervisor
Violation