ML17278A708
| ML17278A708 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 04/02/1986 |
| From: | Phelan P, Qualls P, Thomas Young NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17278A707 | List: |
| References | |
| 50-397-86-05, 50-397-86-5, NUDOCS 8604210272 | |
| Download: ML17278A708 (30) | |
See also: IR 05000397/1986005
Text
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U.S.
NUCLEAR REGULATORY COMMISSION
REGION V
Report No. 50-397/86-05
Docket No. 50-397
Iicense
No. NPF-21
Iicensee:
Washington Public Power Supply System
P.
O. Box 968
Richland,
99352
Facility Name: Washington Nuclear Project No.
2
(WNP-2)
Inspection at:
WNP-2 Site,
Benton County, Washington,
Inspection Conduct d:
Inspectors:
p.
Quails,
Rea to
Inspector
Date Si ned
P. H..Phelan,
Reactor Inspecto
Approved by:
T. Young Jr.,
Chi
Engineering Sectio
Date Signed
Date Signed
~Summar
u
Ins ection on March 3-24
1986
(Re ort No. 50-397/86-05)
Areas Ins ected:,
Announced inspection by a four member
team consisting of two
NRC 'inspectors
and two consultants
from Brookhaven National Laboratory of the
licensee's
compliance with their Fire Protection License Condition.
The team
was
on site March 3-7 and subsequent'elephone
conversations
concerning the
findings were held on March 14, 20,
21 and 24,
1986.
IE Manual Chapter,
TI 2515/62 Rev.
2 was used for, this i'nspection.
ra
Results:
Of the areas
examined,
four unresolved
items
(see paragraphs
4.3,
5
and
10) were, identified.
Enforcement action related to this inspection will
be the subject of separate
correspondence.
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DETAILS
1.
Pers'ons
.Contacted
a.
WNP-2 Staff
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Powers - Plant Manager
Baker - Assistant Plant Manager
Martin
Assistant
Manager, Director
Operations
Bell - Manager
XS 6 FP
Eggen - Senior FP Engineer
Feldman - Plant
QA Manager
Harrold - Engineering
Manager
Davidson - Plant Compliance Engineer
Merhar - Control Room Supervisor
Shaeffer - Shift Manager
Beardsley - Assistant Operations
Manager
Brastad - Electrical Engineer
Harmon - Maintenance
Manager
Powell - Plant Lice'nsing
Porter - Manager Electrical - ISC Systems
Sorensen
- Manager Regulatory Programs
Freeman - Technical Staff
Aeschliman - Sr. Licensing Engineer
b.
Contractors
+A. Jones - F. P. Engineer
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+A. Rapacz - Plant Representative
+
Denotes
those attending exit meeting
on 3/7.
Denotes
those attending special meeting
on 3/6.
2.
Facilit
Desi n - BTP
CMEB 9.5.1
Com liance
The licensee is required by his license to meet the technical
requirements
of BTP
CMEB 9.5.1 in the facility's design.
The criteria
specified,
requires that the fire protection features
provided in the
facility's design be 'capable of limiting fire damage
so that one train of
systems
necessary
to achieve
and maintain hot shutdown conditi'ons from
either the control room or remote
shutdown station be free of fire damage
and that systems
necessary
to achieve
and maintain cold shutdown
conditions are required to be repairable in order to achieve
cold
shutdown within 72 h'ours from either the control room or remote
shutdown
station.
i
A collective assessment
of selected
areas
containing structures,
systems
and components
important to safe
shutdown
was
made by the inspection
team.
The following conclusions
were ascertained:
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2.1
Safe
Shutdown Methodolo
The operating license for WNP-2 contains condition
$/14 which states,
"(14) Fire Protection
Program,
(Section 9.5.1,
SER,
SSER 83,
SSER
84).
The licensee shall maintain in effect all provisions of the
approved fire protection program."
Section 9.5.1 of the
SER and
Supplements
//3 and
84 to the
SER describe
how the fire protection
program meets
the guidelines of the Standard
Review Plan,
CMEB
9.5-1.
At the time of the inspection,
licensee
management
identified, that the program that was reviewed
and approved in the
license condition is the program in the
FSAR, Appendix F, Amendments
19 and 24.
On March 3,
1986 the inspectors
determined that the approved fire
protection program,
FSAR section F.4, identifies that the licensee
will use the
RCIC/HPCS systems
and
RHR system to maintain vessel
inventory and cool down the plant.
The licensee
never implemented
this method, instead,
they employ another
method using the ADS/LPCI
systems.
The method
now used by the licensee
exceeds
the guidelines
of BTP 9.5.1 in that
some core uncovery is postulated.
BTP 9.5.1
section 5.c.2.b states
that a performance
goal for BWRs is to
maintain coolant level above
the top of the core.
The licensee
sent to NRR a letter,
on March 23,
1983, discussing
an
analysis of this method
and enclosing
a copy of this analysis.
No
mention of proposed
implementation of this shutdown method is in the
letter.
NRR never responded
to this letter as
no action was
indicated or requested.
The licensee
has since
done
a plant
specific analysis which the inspection
team thought had significant
differences
from the analysis which was sent to NRR in 1983 (which
was based
on a BWR-4).
WNP-2 is a BWR-5.
At a different licensee
(with a BWR-6) which proposed
using the
ADS/LPCI shutdown methodology,
the
NRR staff required the licensee
to use six ADS valves instead of the three that the licensee
proposed
to blowdown the reactor.
This minimized the amount of core
uncovering.
WNP-2 is using three
ADS valves based
on the
same
BWR-4
analysis.
At WNP-2 no exemption to the licensing requirement,
that
there
be no core uncovering in a
BWR, was ever requested.
The license
was issued to WNP-2 on December
20,
1983.
The licensee
had an opportunity to review this license prior to issuance.
An NRC
inspection
(50-397/83-55),
which was conducted in November
1983,
identified to the licensee this discrepancy
between
the approved
fire protection program and the actual plant systems
being
implemented for safe
shutdown
(SSD).
The licensee at that time
committed to correct the documentation
(50-397/86-05-01).
This is an apparent violation of the WNP-2 operating license.
2.2
Cable
S readin
Room
CMEB 9.5.1 section 7.c addresses
the requirements
for cable
spreading
rooms
(CSR) separation criteria.
This section specifies
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that "A separate
cable spreading
zoom should be provided for each
redundant division.... If this is not possible
a dedicated
system
should be provided."
The licensee's
approved program Section F.3
position F.3(a) states "It should be noted that the Appendix R fire
evaluation for this area is based
upon
a dedicated
safe
shutdown
system".
At WNP-2 the licensee
has only one cable spreading
room
(CSR)
and
chose to separate
redundant
safe
shutdown trains with 20 feet of no
intervening combustibles with detection
and suppression.
Their
,method
had been witnessed
and
a trip report by the
NRR fire
'rotection engineer
was written. It was determined that their
method
was acceptable
and would meet the Appendix R to
Section IIX.G.2 requirements.
However no
SER or formal NRC
acceptance
of this method at WNP-2 was written.
The licensee
has
included the description of this in his draft updated
FSAR.
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2.3
Safe
Shutdown Train Se aration
The WNP-2 electrical distribution system
was inspected
to verify
compliance with the separation criteria given in BTP
CMEB 9.5.1
section 5.b.(2) which state:
"(a) Separation of cables
and equipment
and associated
circuits of
redundant'trains
by a fire barrier having
a 3-hour rating.
Structural steel forming a part of or supporting
such fire
barriers
should be protected to provide fire resistance
equivalent to that required of the barrier;
(b)
Separation of cables
and equipment
and associated
circuits of
redundant trains by a horizontal distance of more than 20 feet
with no intervening combustible or fire hazards.
In addition,
fire detectors
and an automatic fire suppression
system should
be installed in the fire area;
or
(c)
Enclosure of cable
and equipment
and associated
circuits of one
redundant train in a fire barrier having
a 1-hour rating.
In
addition, fire detectors
and an automatic fire suppression
system should be installed in the fire area."
2.3.1
Inside of the Reactor Building, Fire Area R-I, no automatic
suppression
was installed.
To protect redundant
safe
shutdown
trains the licensee installed
a fire barrier between the trains
using
a Thermo-IAG 330 fire barrier system.
The manufacturer,
in
the manual TSI Technical Note 20684,
Thermo-IAG 330 Fire Barrier
System, Installation Procedure
Manual, Nuclear Plant Applications,
says to install a minimum of 18 inches of thermolag 'between the edge
of the cable tray to be protected
and the exposed steel of its
supporting structure.
On March 4,
1986 the inspector found:
a.
That the licensee
required only 9 inches of thermolag material
to be installed
on the auxiliary support steel which in many
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cases,
inside Fire Area R-I, resulted in less
than
18 inches of
protected steel.
b.
That to be
a rated
3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire assembly
the Thermolag would
have had to be installed in a configuration tested
and approved
by an independent
agency
such
as Underwriters Laboratories
(UL).
UL had tested
and approved
the
18 inch installation
configuration.
The
9 inch criteria is not a UL approved
configuration,
hence the assembly in the Reactor Building did
not meet the criteria to be
a 3-hour rated barrier.
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'That in many cases
on the 501'nd 522'levation not even
9
inches of Thermolag
was installed
on the auxiliary support
steel.
d.
That in two places
on the 501'levation
a cable tray which was
required to be protected
was not completely prot'ected.
e.
In 1983 an analysis,
performed by the AE, was used
as the basis
to install 9 inches of therm'olag
on the auxiliary support
steel.
The analysis
showed that, at
a distance
9 inches
from
the. unprotected
steel,
temperature
in a one hour fire would
reach
800 F.
This is less
than the
1100 F needed
to maintain
the structural integrity, but is well above the
480 F needed
to
cause
degradation of cables in a tray.
The analysis
did not
address protection 'of the cables.
The abo've items a. through e. are based
on protecting the
ADS/LPCI system which is not'the
systems
specified to be
protected in the approved fire protection program.
The
licensee
did not. protect the systems
specified in the approved
program
(RCIC/HPCS) from fire damage.
2.3.2. In the Cable Spreading
Room, Fire Area RC-II, the licensee
chose to 'separate
redundant
safe
shutdown trains by coating all
intervening combustibles in the twenty feet between
redundant
trains, with thermolag
and having sprinklers
and fire detectors
throughout the area'.
This method has
been accepted
by NRR as
meeting criteria 2.3(b) above,
to meet the approved
program
requirement in the FSAR.
However on March 4,
1986 the inspectors
found that
a total of 7
cable trays transversing
the area
across
the 20 ft. separation
area did not have
20 feet of the tray protected.
The inspector
found that in one tray only '17 feet was protected,
another only
19 and another
18.
2.3.3.
The licensee in Section F.2 of the FSAR stated
a
wall one foot higher than instrument racks E-IR-H22/P021
and
E-IR-H22/P027
was installed to shield the racks from an
exposure fire.
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On March 5, 1986, when'easured,
one wall was
7 inches higher
'than the rack, the other wall was only 8 inches higher than the
rack.
2.3.4.
To meet the fire protection requirements in Section F.4 of
this program,
the licensee
performed analysxs,
NES-7.
In it
the licensee identified that cables
going into the instrument
racks in paragraph
2.3.3 should be protected to a distance of
one foot below the top of the shield wall to meet the safe
shutdown requirements
in the
On March 5,
1986 the inspectors
found that cables
going to
instruments
MS-LT-26D and MS-PT-51B were not protected with the
3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> thermolag wrap as required by their analysis.
2.3.5
During his Appendix
R review, to meet his fire protection
program,
the licensee
determined that cable
2NS4232
was
required to be protected by a
3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire barrier.
On March 5,
1986 this cable
was not protected.
The licensee
upon learning of the thermolag concern with the
auxiliary support steel immediately placed the reactor building
under
a one hour fire tour as required by Technical
Specification 3/4.7.7 for an inoperable fire assembly.
The
licensee is evaluating the most acceptable
means to correct'he
problem.
Subsequent
to the site visit by the team,
the licensee
has
thermolagged all of the questioned
cables
and has
thermolagged
the separation
distance in the
CSR to ensure that there is
twenty feet separation.
The licensee plans to pursue with NRR the adequacy of the
design of the shield walls that are 7" and 8" above the racks
to get NRR's acceptance
of the design.
The above items are
an apparent violation of the WNP-2 license
condition (50-397/86-05-02).
The Standard
Review Plan details
QA/QC measures
that should be applied to
The licensee
stated that this is met by
applying the appropriate portions of his normal Appendix
B QA program to
During this review, the inspection
team
verified in plant
a relatively small number of fire protection features
and found that
a significant percentage
did not meet requirements
and/or
commitments.
Based
on this, the team observed that the licensee
did not
appear to have applied appropriate
QA/QC to the fire protection program.
The licensee
stated that he would consider this comment. It will be
examined at
a later inspection
(50-397/86-05-03).
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4.
PROTECTION FOR ASSOCIATED CIRCUITS
4.1
Discussion
The WNP-2 plant was inspected for conformance with the associated
circuit provisions of the Standard
Review Plan,
CMEB 9.5-1 and
guidance
documents
issued
by the
NRC as generic letters
and
information notices.
Paragraph
C.5.c-(7) of BTP
CMEB 9.5-1 states
that
SSD equipment
and systems
should be known to be isolated from
associated
circuits such that hot shorts,
open circuits, or shorts
to ground in associated
circuits will not prevent operation of the
SSD equipment.
The concern is that circuits within the fire area
will receive fire damage
which can affect shutdown capability and
thereby prevent post-fire safe
shutdown.
Associated circuits of
concern are those
cables
(safety-related,
non-safety related,
Class
IE, or non-Class
IE) that have physical separation
less
than
specified
and have:
1.
a
common bus with shutdown equipment,
or
2.
a connection to equipment
whose
s urious
o eration would affect
shutdown equipment,
or
3.
a
common enclosure with shutdown cables.
As described
below, the area of associated
circuit protection
remains
open.
4.2
Common Bus
The
common bus concern is found in circuits, either nonsafety
related or safety related,
where there is
a
common power source with
shutdown equipment,
and the power source is not electrically
protected
from the circuit of concern.
In order to ensure proper
electrical protection was provided, the inspection
team reviewed the
licensee's
coordination program f'r circuit breakers,
protective
relays
and fuses.
No discrepancies
were identified.
The team also
reviewed the periodic relay calibration and maintenance
program.
Relay calibration records for the following equipment circuit
breakers
were reviewed:
RHR pump 2A, 2B and
2C and the service
water pumps.
Records
showed calibrations
were verified to be within
specification annually.
The team also witnessed
the calibration and
testing of the Westinghouse protective relay for CRD pump
1B circuit
breaker.
The work was performed professionally
and found
acceptable.
4.3
Common Enclosure
An example of a
common enclosure situation is when there is
a
circuit associated
with a Division 1 circuit in a Division 1 fire
area
and this associated
circuit then leaves
the fire area
and
enters
a
common enclosure with Division 2 circuits.
Common
enclosures
could be raceways,
switchgear,
motor control centers,
etc.
A particular c'oncern here is hot shorts created
by the fire on
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the'ssociated
circuit which could cause
damage in the
common
enclosure.
The licensee
has partially addressed
this concern by
having coordinated circuit breaker
and relay protection and by
providing fire stops
and seals.
However,
one particular aspect of
this concern
has not been addressed.
The licensee
stated
that.
higher voltage cables
are often run with lower voltage cables (e.g.,
125
VDC with 250 VDC). If there were
a fire in a Division 1 fire
area
and
a higher voltage cable were to hot short onto
a lower
voltage Division 1 associated
circuit, which then goes into a
common
enclosure with Division 2 cables,
damage to both Divisions
associated
circuits could occur.
The licensee
stated 'that there are
numerous
places in the plant where Division 1 associated
circuits
are routed into Division 2 common enclosures
(except
raceways)
and
vice-versa.
The licensee further stated that no analysis
had been
done or protection provided for this type of common enclosure
concern.
This item has been referred to NRR for determining the
requirement,
and safety significance of not performing it.
This item
is unresolved
(50-397/86-05-04).
4.4
S urious Si nals
The spurious signal concern is made
up of two items:
The false motor, control and instrument readings
such
as those
which occurred at the
1975 Browns Ferry fire, that could be
caused
by fire initiated grounds,
shorts
and open circuits.
Spurious operation of safety-related
or nonsafety-related
components
that would adversely affect shutdown capability
(e.g., RHR/recirculation system isolation valves).
The spurious signal concern
was inspected for the following cases:
4.4.1 Diesel Generato'r
Anal sis
Since signals
from and for the Diesel Generator
(DG) go to many
different fire arear'n
the plant, the licensee
performed
a separate
spurious signal analysis for each
DG.
This analysis
was not
finalized, but the team reviewed what material
was available in
draft form.
For DG-1A, the licensee listed all cables exiting in
the
DG room and then tabulated
which fire areas
they ran through.
Any cables entering
a Division 2 fire area
were then analyzed to see
if a fire i'n that Division 2 fire area
would also incapacitate
the
Division 1 DG.
In a short review of this analysis,
the team
identified two errors where cables listed as exiting in the
DG room
were
shown to go through the wrong fire areas.
Reanalysis
of these
cables
during the week showed that they would not incapacitate
the
DG.
The licensee
has been requested
to evaluate this analysis to
determine if these errors are generic or an isolated
case.
(50-397/86-05-11)
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4.4.2 Hi h Low Pressure
Interface
This concern relates
to the situation where fire-induced spurious
signals
could cause interface valves between high pressure
piping
and low pressure piping to open, resulting in a I,oss-of-Coolant
Accident
(LOCA) coincident with the fire.
The licensee's final
analyses
were not available for review.
However,
the team did
review some older material and
some interim results.
The licensee
had determined that three sets of MK system valves
needed analysis
and protection.
FSAR (Amendment 31) Question
and Answer 040-079
stated that protection for one valve from each pair would be
provided.
The reanlysis identified some areas of cables
to these
valves that should have been protected (i.e., thermolagged)
but were
not.
Fire watches
were initiated and cables protected
by
implementing rework,
when plant conditions permitted.
Some cables
still are scheduled for protection during the upcoming refueling
outage.'wo
RHR valves
(53A and
123A) have controls in the control room
where they both could spuriously operate
given
a control room fire.
The licensee identified these valves to NRR who stated that the
valves should be deenergized.
The licensee
has not yet taken action
'n this recommendation..
The
NRC has formally transmitted this
to the licensee
on March 14,
1986.
The licensee's
response willbe
discussed
in a later inspection.
(50-397/86-05-12)
The inspection
team also reviewed flow diagrams to identify any
other sets of high/low pressure
interface valves which should have
been analyzed or protected.
Several potential candidates
were
identified and discussed
with the licensee.
The licensee
was able
to resolve all concerns with the additional valves.
1.4.3
Isolation of Control Room
In the event of a fire in the control room, the licensee
has
a
remote
shutdown panel where he can operate
the
SSD equipment.
Xn
order to clear fire-generated electrical faults, isolation switches
typically must be operated at the remote
shutdown panel.
During his
reanalysis,
the licensee identified (LER 84-31) additional circuits
that must be isolated from the control room.
Modifications are
scheduled for this refueling.
In the interim, the licensee
has
procedures for lifting leads
and install jumpers to clear the
faults.
The licensees
detailed calculations
and final analyses in
this area
were not available for review.
5.
WNP-2 Regnal sis
The initial safe
shutdown fire protection analysis for WNP-2 was
performed by Burns
and Roe using the WNP-2 Engineering Criteria, Appendix
4, Criteria for 10 CFR 50, Appendix R Compliance.
After WNP-2 received
its operating license in 1983, the licensee
decided that additional
Appendix R analyses
were needed.
Burns and Roe revised the Engineering
Criteria, Appendix 4, with a Revision
14 on May 25,
1984.
Discussions
with licensee
personnel
indicated that WNP-2 then decided to have the new
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analyses
performed internally by WNP-2 personnel
rather than by Burns and
Roe.
Discussions with personnel performing the analyses
and reviews of
the actual analyses
paperwork and data revealed that the analyses
were
not done to the Burns and Roe document.
WNP-2 found additional analyses
and different methods
were required to adequately
complete the tasks.
The forms described in the Burns and Roe appendix were not used nor were
the approvals
and second
checks, documented
as described.
When asked to
what engineering criteria the analyses
were performed,
the licensee
provided
a draft document
NES-7, Revision 0, Supply System Engineering
Standard,,Safe
Shutdown Analysis.
The licensee's
Operational Quality
Assurance
Program Descri'ption, Revision 5, paragraph
3.2.1 states
that
Organizations participating i.n the preparation,
review, approval
and
verification of design
documents
(including analysis) shall develop
and
implement procedures
that clearly deli'neate actions to be accomplished.
This item will remain unresolved
pending review of the implementing
procedures
that established
the engineering criteria to which WNP-2
performed Appendix R analyses.
(50-397/86-05-05)
The inspection
team noted that the Appendix R reanalysis
that had been
ongoing for the last one to two years
was not performed to approved
procedures
or detailed engineering criteria.
The team also noted that
even after this length of time, final approved
analyses
were not
available for review.
The team did note,
however, that the licensee
was
performing significant additional analyses
to upgrade
the fire protection
features of the plant.
As deficiencies in the fire 'protection features
were found, the licensee
appeared
to take appropriate actions,
such as
establishing fire watches, initiating design
changes
and issuing Licensee
Event Reports
(LERs) to notify the NRC.
The licensee
has
a Draft Amendment
37 to the
FSAR which summarizes
the
results of the reanalysis.
Paragraph F.4.4.1.5 states
that,,"The design
basis fire for the Main Control Room and, the Cable Spreading
Room,
even
though not considered
credible,
can result in generating transients
more
severe
than presently analyzed in the FSAR Chapter
15 if worst case
conditions are applied.
These conditions are not analyzed."
The
inspection
team noted that Appendix R and the Standard
Review Plan both
require analysis to determine
the consequences
of fire in any location in
the plant.
The statement
of the draft FSAR amendment
did not appear
consistent with the need for such
a full analysis.
NRR has been requested,
by Region V, to evaluate
the licensee's
analysis
for the control room fire and cable spreading
room fire to ensure that
the analysis
are adequate
and that no unreviewed or unanalyzed
safety
questions exist.
This matter is unresolved
pending
NRR review
(50-397/86-05-06).
6.
Post-Fire
Safe
Shutdown
Ca abilit
6.1
S stems
Re uired for Safe Shutdown
The licensee
has
chosen not to protect the RCIC system, utilizing
instead
ADS and residual heat removal'system
loop
B (RHRB) for a
rapid depressurization
in going from event initiation directly to
cold shutdown where
a dedicated
low pressure
system
(RHR) is
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utilized in the alternate
shutdown cooling mode.
This approach
differs from the
SER and is documented in draft Amendment
)f37 to the
WNP-2 CESAR.
6.1.3, Reactivit
Control
l.
Vpon detection of a, disabling fire, the control rods
can be inserted
using the scram switch in the control room or they are automatically
inserted 'upon loss of off-site power.
'.1.2 Reactor Coolant Inventor
Control
Sinck,the
RCIC is not used,
the reactor level and pressure
must drop
before the low pressure
system
(RHRB-IPCI) can begin to inject water
into the'essel
for level control.
As indicated in MPPSS's letter
of March 21,
1983, this can result in core uncovery and therefore is
not in. conformance with II.L.2.b of Appendix R or the
SRP.
The
licensee
has performed
a plant-specific analysis NE-02-84-30,
Revision 0, for the core uncovery scenario at MNP-2 which should be
reviewed by NRR as it appeared
to differ .from the analysis
submitted
to NRR for review (see paragraph 2.1).
6.1,.3 Deca
Heat Removal
During hot shutdown,
decay heat is transmitted directly to the
suppre'ssion
pool by discharging
steam from the RPV through the SRVs.
The suppression
pool cooling mode in the
RHR system transfers
the
heat to the standby service water system through the
RHR heat
exchangers.
6.1.4 Process
Monitorin
For control room fires that cause
an ev'acuation
and subsequent
use
of the remote
shutdown panel,
the licensee provides the following
process
instrumentation at the xemote panel.
Reactor pressure
Reactor level
RHR flow
Service water flow
Suppression
pool level
Suppression
pool temperature
6.1.5
Su
ort E ui ment
Supporting electrical power, diesel generator
batteries
Service water system
Supporting
HVAC systems
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When the RPV pressure is less than
125 psig, the shutdown cooling
mode of RHRB is initiated and cold shutdown conditions maintained
by
transferring heat to the service water system through the
RHR heat
exchanger.
6.2
Alternate Shutdown
The only area identified that requires alternate
shutdown is the
control room.
The preferred
shutdown path at the remote panel
utilizes Division 2 safe
shutdown equipment
and Division,1 SRVs.
Since the remote
shutdown system cannot be completely isolated from
the main control room until design modifications in progress
are
completed,
the shutdown procedures
call for lifting live wires and
replacing blown fuses to restore functions lost as
a result of
spurious actions.
These are considered
repairs
and are not
permitted under Appendix.R or the
SRP for hot shutdown
(see
following paragraph 7.1).
7.
Procedures/Alternate
For WNP-2 the only area that requires alternate
shutdown capability is
the Main Control Room.
Division 1 and
2 shutdown paths are provided
for'afe
shutdown,
however, Division 2 is the preferred division and is
protected
where necessary.
7.1
Procedures,
The team reviewed Procedure
No. 4.12.1.1,
Rev.
5 entitled, "Control
Room Evacuation" which is written in a single column sequential
'ormat... A total of 6 people are required for the shutdown procedure
grouped in three
teams of two with the shift manager
and
a reactor
operator
statj.oned at the Safe Shutdown Panel.
The procedures
are
cur'rently in use
and are used for operator training.
They include
the necessary
steps
to bring the plant directly to cold shutdown
using the
SRVs for rapid depressurizat'.ion.
Since the control room
~ will not, be isolated until design modifications adding transfer
switches are complete;
the procedure calls for operator actions in
Attachment D during hot shutdown to preclude spurious trips by
lifting wires and replacing fuses.
Under Appendix R, these actions
which are considered
as repairs
are not permitted.
Therefore, until
the transfer switch modifications are completed this remains
an open
item.
This item was identified to the
NRC xn LER 84-31
and its
supplements.
An SER dated December,4,
1985 found the interim
corrective measures
adequate
(50-397/86-05-13).
7.1.1 Procedure
Walkdown
The procedure
(4.12.1.1) for Conrol Room Evacuation, was,walked
down
by the team
on. March 5,
1986 to determine that safe
shutdown could
be accomplished in a timely and orderly fashion with the number of
personnel
assigned
(6) to perform the actions required by the
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procedure.
,Prior to the walkdown the shift manager
was given the
following scenario:
diesel generator
2 has not started
RCXC not available
Xnspection
team members
accompanied
each operator
as
he performed
the necessary
procedural actions.
The team did not identify. any
unacceptable
conditions except
as follows:
The emergency light in the diesel generator
room was blocked off by
a panel
and could not illuminate the diesel panel.
The licensee
agreed
to correct this condition.
This will be examined at
a later
inspection
(50-397/86-05-07).
The voice actuated
phone jack cord at the Remote
Shutdown Panel
was
too short requiring the operator to stoop
down when, communicating.
The licens'ee
has since corr'e'cted this condition.
8.
Communications
Plant communications
radio transmitter is
single fire will not
during the procedural
walkdown appeared
adequate.
A
'separated
from the communications
room so that,
a
disable
communications.
9.
Emer enc
Li htin
The team, during the procedure
walkdown, verified that: emergency lighting
was installed in areas
needed to safely shutdown the plant and in access
and egress
thereto.
The only discrepancy is noted above.
The inspectors
did not have time to check the emergency lighting drawings
ox. ratings.
This will be accomplished at a later inspection (50-397/86-05-08).
10.
Plant Tour
During the tours through the facility the inspectors
noted the following
items.
Housekeeping
appeared
adequate
with no problem areas
noted
Pire protection equipment
was in place
and appeared
to be well
maintained
4
One fire door with an improper label was identified.
The door
(which required
a 3-hour 'rating) had
a 3/4 hour label.
The licensee
has since contacted
the manufacturer
which stated that the door was
a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated door.
The inspector will follow up the licensee's
actions in a later inspection.
This item is unresolved
pending
licensee proof of door rating and label correction
(50-397/86-05-09).
It
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Fire Protection Audit
The inspector
reviewed the licensee's
annual Fire Protection
(FP) Audit
to assure
that it addressed
the following areas of their F.P.
program:
0
0
Compliance with the F.P.
program
.
Compliance with Administrative Controls
Implementation of Quality Assurance Criteria including design
and
procurement,
'in'structions,
procedures
and drawings
Inspection
and test activities
Fire Brigade qualification and training
While tou~ing the Control Room. the audit team observed
a control room
'perator
clear'an
incoming fire alarm off the panel with no further
actions.
The plants F.P. procedures
require
a fire brigade
member to be
sent to investigate all F.P. alarms'hen
questioned,
the control room
operator indicated that the alarm was spurious
and had
come in
sporatically during his shift.
A person
was sent to investigate
the
first time the alarm
came in.
The audit team found this condition unacceptable
and identified it as
a
concern.
Managements
corrective action was to send
a fire bragade
member
to investigate
when
a "continuous" alarm was received.
The aud'it team
found this response
unsatisfactory
due to the undefined nature of
"continuous".
The inspector
was unable to ascertain if a satisfactory corrective action
was achieved.
This will be
a follow-up item during
a subsequent
inspection
(50-397/86-05-10).
Fire Bri ade Trainin
The inspector
reviewed the training and qualification of the fire brigade
members to assure
that each shift had at least the minimum number of
qualified brigade
members present.
PPM 1.3.36, Fire Protection Program
Training, Rev.
1 states
"The fire brigade shall normally consist of the
following individuals:
a.
Shift, Support Supervisor
(Leader)
b.
Health Physics
(1)
c.
Operators
(3)"
In reviewing the Shift Supervisor's
Log, instances
were found where the
Shift Support Supervisor
(SSS)
was not
a qualified brigade leader,
When
questioned,
the licensee
stated that occasionally the
SSS switches with
the Control Room Supervisor
(CRS) in order to maintain their license
status.
When this occurs the
CRS,
who is
a qualified fire brigade
leader, will automatically
assume
the role of leader in the event of a
fire.
The inspector verified that during each shift either the
CRS or
was
a qualified fire brigade leader.
This item has minimal safety
significance,
however, it does represent
a weakness
in the fire
protection program,
and should be addressed.
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R
P
14
An exit meeting
was 'held'with members of the licensee's
staff on March 7,
1986.
Additional, details
from subsequent
telephone
conversations
with
your staff on March 14, 20,
21 and 24,
1986 have been included in this
report.
The ar'eas of concern
and open items in this report were
discussed-with
your staff at: the exit meeting.
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