ML20210R378
| ML20210R378 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 08/28/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20210R359 | List: |
| References | |
| 50-382-97-15, NUDOCS 9709030108 | |
| Download: ML20210R378 (27) | |
See also: IR 05000382/1997015
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ENGLQSURE.2
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket No.- 50 382 j
License No.- NPF-38
Report No.: 50 382/97-15
Lico,see: Entergy Operations, Inc.
Facility: Waterford Steam Electric Station, Unit 3
Location: Hwy.18
Killona, Louisiana
Dates: June 29 through August 9,1997
Inspectors: L. A. Keller, Senior Resident inspector
G. A. Pick, Senior Project Engineer
Approved By: P. H. Harrell, Chief, Project Branch D
Attachment: Supplemental Information
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9709030108 970828
PDR ADOCK 05000302
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EXECUTIVE SUMMARY
Waterford Steam Electric Station. Unit 3
NRC inspection Report 50 382/97-15
This routine, announced inspection included aspects of operations, maintenance, l
engineering, event response, and plant support. The report covers a 6-week period of
resident inspection.
Ooerations
Several Technical Specification (TS) amendments were not implemented prior to
startup from the refueling outage (RFO). Administrative controls instituted in lieu of
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proposed TS changes were adequate pending issuance of the TS amendments .
(Section 01.3).
Operators identified a subtle single f ailure vulnerability in the control room air
conditioning (CRAC) system. This design deficiency is identified as a noncited
violation (Section O2.1) (EA 97 348).
Operations provided good leadership in prioritizing outage work activities and making
, sure the plant was ready for mode changes (Section O2.2).
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Throughout the outage, the inspectors observed a commitment by plant
management to correct equipment deficiencies prior to startup. Despite the large
amount of emergent work that evolved during the 108-day outage, the licensee
maintained this commitment (Section 02,2).
- The dilution to criticality was well planned and executed. Use of the simulator to
practice for this evolution was identified as a good practice (Section 04.1).
Maintenance
In general, the conduct of maintenance and surveillance activities was good
(Section M1.1).
The f ailure to include the emergency lighting system in the Maintenance Rule
Program in a timely manner was identified as a noncited violation (Section M1.2)
(EA 97-403).
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The f ailure to establish performance goals for the emergency lighting system is a
violation of 10 CFR 50.65 (Section M1.2) (EA 97-403).
A containment walkdown after entering Mode 3 revealed good material condition.
Prior Quality Assurance (QA) inspections were effective in identifying and removing
potential safety injection sump clogging hazards (Section M2.1).
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Engineenng support in resolving the CRAC single f ailure vulnerability was good
(Section 02.1).
Throughout the RFO, Engineering provided very good support in implementing the
large number of design changes and provided effective resolution for a large number
of emerging issues (Section 02.2).
Reactor engineering provided good support during the dilution to criticality
(Section 04.1).
The licensee ef fectively implemented design changes and completed operability
analyses to address component cooling water (CCW) relief valves lif ting due to
system pressure spikes (Section E2.1).
Engineering performance was good in identifying a Static Uninterruptible Power
System (SUPS)/ Appendix R common mode f ailure vulnerability: however,
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Engineering support in resolving this issue the unf amiliarity with the design basis
(Section E4.1).
The licensee's decision to align Chiller AB to replace Chiller B in order to temporarily
resolve Appendix R concerns demonstrated a lack of understanding of the
procedural requirements to isolate Train AB equipment following an Appendix R fire.
This episode further demonstrated a lack of rigor in resolving the SUPS / Appendix R
issue (Section E4.1).
Licensee Event Report (LER)96-007 reported low auxiliary component cooling
water (ACCW) flow to the essential chillers because of fouling and inadequate
design. The failure to establish appropriate design basis flows for the ultimate heat
sink is identified as a noncited violation (Section E8.1),
The licensee identified that certain ACCW system configurations result in exceeding
the design pressure of the dry cooling tower (DCT) tube bundles. The failure to
appropriately establish design criteria for the DCT tube bundles is identified as a
noncited violation (Section E8.3).
Plant Sucoort
Timely and conservative decisions were made regarding event classification and
reducing reactor coolant temperature in response to Hurricane Danny
(Section 01.2),
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- A security guard was found reading unauthorized material while on watch in the
secondary alarm station (SAS). Inattention to duty by security personnelis an
ongoing concern (Section S1.11.
- The f ailure to provide required emer0ency lighting with at least an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> battery
supply for an access route to the essential chiller area is identified as a violation of
10 CFF) Part 50, Appendix R (Section F2.1),
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Bepart Details :
SummaLy of Plant Status
The plant remained shut down for RFO 8 until July.26,1997, when the reactor achieved
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criticality. RFO 8 was completed on July 29, which resulted in an outage duration of I
108 days, when the main generator was placed on the grid. On August 2, power was
increased to 100 percent. The plant operated at essentially 100 percent power throughout
the remainder of this inspection period.
L_Qnerations
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Conduct of Operations
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O 1.1 General Comments (71707)
- The inspectors performed frequent reviews of ongoing plant operations, control
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l room board walkdowns, and plant tours. Observed activities were generally
performed in a manner consistent with safe operation of the f acility. The inspectors
observed good operator performance during plant startup. Operator calculations of
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shutdown margin were done in accordance with procedure and demonstrated
acceptable results. Plant management was observed in the control room on
numerous occasions. Shif t turnover meetings were professional and informative.
01.2 Resoonse to Hurricane Danny
a. insoection Sco_ c e (93702. 71707. 71750)
At 11 a.m. on July 17, the control room was notified that Tropical Depression
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Danny had been upgraded to tropical storm status and that St. Charles Parish was
under a hurricane watch / tropical storm warning. The inspectors observed the
licensee's preparations for a possible hurricane and subsequent response during the
storm's progress,
b. Qbervations and Findinos
The plant was in Mode 4, conducting surveillance testing in preparation for entering
Mode 3 prior to the tropical storm notification. The storm was approximately
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165' miles southwest of the plant moving northeast at 5 mph -The shift entered
Procedure OP-901-521, " Severe Weather and Flooding." Although projected wind
speed onsite was less than 30 mph, the licensee conducted extensive preparaticas, s
which included cleanup from outage-related activities. Numerous portable tanks,
pallets, cranes, and other items were moved indoors or away fiom the site. A
two-section emergency response watchbill was developed as a contingency. The
personnel on the watchbill were contacted to confirm their availability and to place
them on standby for immediate call out if necessary. The inspectors toured the
pla7t and determined that hurricane preparations were good.
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At 2:16 a.m. on July 18, St. Charles Parish declared a Hurricane Warning per the
National Weather Service when maximum wind speeds within the storm exceeded
74 mph. Winds onsite were still below 20 mph at this time. At 2:20 a.m., an
Unusual Event was declared per Procedure EP-001-001, " Recognition and
Classification of Emergency Conditions," due to St. Charles Parish having declared a <
Hurricano Warning. Although wind speeds were still projected to remain below l
30 mph onsite, as a conservative measure, the licensee reduced reactor coolant '
system temperature from 330 to 205*F. This action was taken to facilitate entry
into Mode 5 if the storm changed direction and placed the plant at hazard.
The Unusual Event was terminated at 4:23 p.m. on July 18 when St. Charles Parish
rescinded the Hurricane Warning. Hurricane Danny's closest point of approach to
Waterford 3 was 60 miles. Maximum winds experienced onsite were less than
25 mph.
c. Conclusions
Licensee preparations for Hurricane Danny were good. Timely and conservative
decisions were made regarding event classification and reducing reactor coolant
temperature.
01.3 Administrative Controls in Lieu of TS (71707)
Prior to and during RFO 8, the licensee identified several TS that were
nonconservative and, therefore, submitted TS amendment requests to NRC
Headquarters. As of the end of this inspection period, several of these TS change
j requests (TSCR) had not been approved. The licensee implemented administrative
controls in order to meet the intent of the proposed TS, pending NRC approval. The
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implementation of these administrative controls was reviewed by personnelin NRC
Region IV and Headquarters prior to plant restart from the refueling outage. The
following is a summary of the TS issues:
TSCR NPF-38196 modifies TS 3.1.1.1, 3.1.1.2, and 3.10.1 and
Figure 3.1.1 by removing the cycle dependent boron concentration and
boration flow rate from the Action Statements and removing the "RWSP at
1720 ppm" curve from the figure.
TSCR NPF-38197 modifies TS 3/4.7.4, " Ultimate Heat Sink," Table 3.7-3,
by incorporating more restrictive DCT f an operating requirements and
changes the wet cooling tower water consumption inihe TS bases.
TSCR NPF-38-198 modifies the action requirements for TS 3/4.3.2 for the
safety injection system sump recirculation actuation signal to require a plant
shutdown if a recirculation actuation signal channel is mai.itained in a tripped
condition.
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TSCR NPF 38199 modifies TS Tables 3.71 and 3.7-2. Table 3.7-1 is
revised to indicate the correct steam line safety valve orifice size.
Table 3.7 2 is revised by deleting the provision that allows continued plant
operation with three main steam safety valves inoperable.
TSCR NPF-38179 modfies TS 3.7.1.3 by requiring the condensate storage
pool to contain a minimum volume of at least 91 percent indicated level.
TSCR NPF-38-190 modifies TS 3.7.1.3 by removing Action b, which allowed
use of the wet cooling tower basins as a source of emergency feedwater,
All of the proposed TS requirements were incorporated into Administrative
Procedure OP 100-014, " Technical Specification And Technical Requirements
l Compliance," pending approval of the submitted TS. The licensee implemented
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these actions to ensure that they operated the plant within the newly established
limits. The inspectors reviewed the administrative guidance incorporated in
Procedure OP-100-14 and concluded that it provided adequate administrative
controls. The inspectors noted that in all cases the proposed TS were more
conservative than the existing TS. The inspectors noted that all operators received
training on these administrative controls prior to startup.
02 Operational Status of Facilities and Equipment i
02.1 CRAC System Sinale Failure Vulnerabilitv
a. Insoection Scone 171707.37551)
The inspectors reviewed the circumstances and corrective actions associated with
the licensee discovery of a single f ailure vulnerability in the CRAC system,
b. Observations and Findinas
The CRAC system is designed to establish and maintain a habitable atmosphere
inside the control room envelope in the event of a toxic chemical accident or a
design basis accident with a resulting radioactive environment. The CRAC system
includes, in part, two full-capacity redundant air handling units, AH-12A(BI, two
full-capacity redundant control room emergency filtration units (CREFU), S 8A(B),
normalintake and exhausts, and emergency air intakes. The CREFUs consist of
medium efficiency filters, high efficiency filters, charcoal absorbers, electric heating
coils, and a centrifugal f an. The CREFU share a common suction duct with
separation / isolation being provided by Dampers HVC-205A(B) and -213A(B).
During normal operation, outside air enters through the normal air intake, combines
with recirculation air flow, and is distributed throughout the control room envelope.
The CREFU are isolated in the normal mode of operation. The CRAC system
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maintains a 1/8 inches of water gauge (iwg) positive pressure, relative to the outside
atmosphere, within the envelope during normal operations to prevent any outside air
from bypassing the sasety-related toxic monitoring instrumentation located in the
normal air intake path. - The north and south emergency air intakes provide air for
pressurizt , ion of the envelope during radiological accident conditions. On a high
radiation or safety injection signal, the normal ventilation intakes am automatically
isolated and manual action is required to bring in outside air from the emergency air
intakes in order to maintain a positive pressure inside the envelope.
TS 3.7.6.1 (Modes 1, 2,3, and 4) and 3.7.6.2 (Modes 5 and 6) requires two CREFU
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to be operable. The operability of the CREFU ensures that, following a design-basis '
accident, the radiation exposure to personnei occupying the control room is limited
to 5 rem or less whole body, or its equivalent. This limitation is consistent with the
requirements of General Design Criterion 19 of Appendix A,10 CFR Part 50.
TS Surveillance Requirement 4.7.6.5.a states that the control room envelope
isolation and pressurization boundaries shall be demonstrated operable at least onca
per 18 months by verifying that the control room envelope can be maintained at a
positive pressure of 21/8 iwg relative to the outside atmosphere with a make-up air i
flowrate s200 cfm during system operation. The implementing procedure for
TS 4,7.6.5.a is Surveillance Procedure OP 903-123, " Control Room Envelope
Pressure Test."
On May 16,1997, Waterford 3 was in Mode 6 and a SUPS Train A maintenance
outage was in progress. Because of the loss of SUPS power,' CREFU Train A was
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inoperable and its inlet damper, HVC-205A, and recirculation damper, HVC-213A,
were deenergized. HVC-205A(B) and -213A(B) fail open on a loss of power.
Shortly af ter deenergizing SUPS Train A on May 16, control room operators
attempted to place the CRAC system in the high radiation mode of operation in
order to support ongoing maintenance activities, in order to plars the CRAC system
in the high radiation mode of operation, the operators performed the section of
Surveillance Procedure OP-903-123, " Control Room Envelope Pressure Test," which
simulated a high radiation condition. During this activity, the operators noted the
inability to obtain 1/8 iwg or greater control room envelope pressure. Condition
Report (CR) 97-1255 was written to investigate the inability to achieve 1/8 iwg or
greater control room envelope pressure.
Subsequent investigation found that the inability to achieve the required control
room pressure was due to a design flaw in that, with the opposite train'u
recirculation damper (HVC-213A) f ailed open, a short circuit path was created from
CREFU Train A to Train B via the emergency outside air intake duct. Thus, a single
f ailure of a recirculation damper in either train results in the inability to meet TS
Surveillance Requirement 4.7.6.5.a. This was confirmed during subsequent testing
that demonstrated that either recirculation damper. HVC-213A(B), f ailed open
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results in the inability of either CREFU to pressurize the control room envelcpe to
1/8 iwg with s200 cfm makeup flow. This design deficiency is a violation of
10 CFR Part 50, Appendix B, Criterion lit. This norirepetitive, licensee-identified and
corrected violation is being treated as a noncited violation, consistent with (
Section Vll.B.1 of the NRC Enforcement Policy. Specifically, the viciation was I
identified by the licensee, was not willful, actions taken as a result of a previous
violation should not have corrected this problem, and appropriate corrective actions
were completed by the licensee (50-382H3715 01).
As an interim corrective action, the CREFU Train A emergency air intake was
isolated. This effectively eliminated the common suction path for both units,-
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thereby maintaining CREFU Train B operable and allowing CREFU Train A to remain l
available for recirculation only, As permanent corrective action, a system design i
change installed backdraf t dampers that would prevent this condition. ;
The inspectors reviewed the design change package, discussed it with the system
engineer, and accompanied him on a system walkthrough. As described in the
design package, the position of the backdraft dampers in the system prevented back
flow in the system if a single failure occurs with a damper. The inspectors found
that the proposed design change appeared adequate to preclude a single f ailure from
causing the CREFU to be inoperable. Postmodification testing demonstrated
acceptable results.
The licensee indicated that other ventilation systems would be evaluated to
determine if similar conditions existed. Review of the evaluation of other ventilation
systern potential single failure vulnerabilities will be perforrned when followup of
LER 50-382/97 017 is completed.
The inspectors reviewed Off-normal Procedure OP-901-401, i'High Airborne Activity
in Control Room " Procedure OP-901401 directs the control room operators to
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obtain approximately 200 cfm outside air intake by adjusting the CREFU
recirculation damper (HVC-213A(B)) controller. The inspectors concluded that the
of f normal procedure would probably result in the prompt identification of the single
f ailure at issue and that there were reasonable compensatory actions available
(i.e., splitting out the CREFU trains) to maintain control room envelope integrity.
The inspectors noted that control room design-basis accident exposure is based on'a
' 30-day time period, whereas the ability to mitigate this problem should only require
a few hours. During this time frame, the control room envelope would be in the
recirculation mode (same as for a toxic gas event) with negligible inleakage, as
demonstrated by previous testing / operation in the isolation mode. For these
reasons, the inspectors concluded that the single f ailure vulnerability was of low
safety significance.
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c. Conclusions.
The CREFU single f ailure vulnerability was subtle in that surveillance testing
coincident with a SUPS outage demonstrated the inability to maintain control room
envelope pressure with a recirculation damper failed open. Immediate corrective
actions were effective in restoring one train of CREFU to service. Engineering
support was good in providing a permanent fix for the single failure vulnerability. i
The safety significance was low due to the ability of operators to recognize the
f ailure and take compensatory actions prior to adverse dose consequences.
02.2 Readiness for Restan
a. Insocction Scooe (71707. 37551, 71711) ,
The inspectors reviewed the status of open CRs, applicable modo entry checklists,
and the Combined Operations Deficiency List to determine whether issues of
concern had been adequately resolved prior to startup.
b. Observations and Findinos
The inspectors noted that most of the open issues going into RFO 8 were
appropriately resolved. Only a few issues had to be deferred for future resolution
(i.e., CCW hydraulic transients) and these issues had adequate compensatory
measures implemented to allow for restart. items reviewed by the inspectors and/or
discussed with the licensee included:
Turbine-driven emergency feedwater pump steam line heat trace and steam
trap improvements
Rerouting of the refueling water storage pool (RWSP) levelinstrument
reference legs
Installation of essentialinstrument air stations to enable closure of certain
containment isolation valves
Reduction in the number of operator workarounds from 62 going into the
outage to 27 after the outage
Significant reduction in the amount of long-term scaffolding throughout the
plant
Successful completion of modifications to the ACCW system to prevent
waterhammer
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Addition of additional venting capability in low and high pressure safety
injection system piping to address waterhammer concerns in those systems
Throughout the outage, the inspectors observed a commitment by plant
management to correct equipment deficiencies prior to startup. Despite a large
amount of emergent work that evolved during the 108-day outage, the licensee
rnaintained this commitment. Operations provided good leadership in prioritizing
outage work activities and making sure the plant was ready for mode changes.
Engineering provided very good support in implementing the large number of design
changes and provided effective resolution for a large number of emerging issues.
c. CDDchnion:
The licensee made substantial improvements in the material condition of the plant.
The majority of significant hardware discrepancies were corrected. Severalissues
were deferred but adequate compensatory measures were implemented. No restart
issues were identified.
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04 Operator Knowledge and Performance
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l 04.1 Dilution to Criticality (717071
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On July 26. the inspectors observed the dilution to criticality following RFO 8. An
infrequently performed test or evolution briefing was held prior to pulling control
rods. The Operations Manager was present in the control room throughout the
evolution. The licensee withdrew all control rods prior to diluting to criticality. The
inspectors independently confirmed adequate shutdown margin prior to the dilution.
The inspectors noted that the shift performing this evolution, including the reactor
engineer present in the control room, had practiced the procedure on the simulator
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the actual evolution. The inspectors considered this a good
practice. All observed activities were performed in accordance with procedure and
all observed parameters were within established limits. Good command and control
and communication practices were observed. Reactor engineering personnel were
knowledgeable and provided good support.
08 Miscellaneous Operations issues (92901)
08.1 (Closed) Unresolved item 50-382/9702-01: Review adequacy of vital 125-Vac
breaker coordination. This item involved the lack of coordination between the SUPS
units and the downstream individualloai feeder breakers. The condition involved
the momentary loss of SUPS power die (o the f ast acting shutdown characteristics
of the SUPS unit in that the SUPS unit shuts down before the individualload breaker
clears the f ault. The inspectors reviewed the applicable institute of Electrical and
Electronics Engineers standards and design criteria applicable to this issue. The
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inspectors concluded that, as long as a single f ailure did not result in a loss of
function, the lack of coordination was acceptable.
Appendix R lists requirements for cases where cables, including nonsafe;y circuits,
could cause maloperation due to hot shorts, open circuits, of shorts to ground of
redundant trains of systems necessary to achieve and maintain bot shutdown.
These requirements include either separation of cables / circuits ol redundant trains
by fire barrier (s), or separation of cables / circuits of redundant trains by a horizontal
destance of 20 feet, or alternate shutdown capability.
On Apol 23,1997, the licensee documented, in CR 97 0988, a condition where
- Appendix R cables, one from each train, were routed within 20 feet of each other,
i without a barrier between them and without one train being protected by a fire
wrap. The licensee established an hourly firewatch in the rifected area per
Technical Requirements Manual Section 3.7.11. The initial'icensee review of this
condition indicated that it would only affect we train of SUPS. Af ter further
review, on June 4, the licensee identified a potential common modo f ailure
associated with the SUPS units. The common moce failure vulnerability involved
postulated fire damage to cables in various plant locations (i.e., cable vault area)
that do not meet the Appendix R separation / lire barrier criteria, combined with the
lack of SUPS breaker coordination, which could result in the momentary shutdown
of both trains of safety related SUPS. Although the interruption of SUPS power
would probably be of short duration, Operations could potentially be required to
manua!!y reset safe shutdown loads on multiple SUPS trains. This issue was
reported via LER 50 382/97 20.
The inspectors concluded that, if 'ne cables in question were in compliance with
Appendix R criteria, the lack 9 T.,PS coordination would not be an issue. The
Appendix R compliance issues will be evaluated durin0 the closure of LER 97 20.
[L.Malatenancs
M1 Conduct of Maintenance (62707,01726)
M 1.1 General Comments
The inspectors observed all or portions of the following maintenance and
surveillance activities:
- WA 0116.2267 Repair Power Supply Cable for Plant Protection System
Channel D Test Drawer
OP 903 014
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in general, the inspectoss found the conduct of these maintenance and surveillance
activities to be good. All activities observed were performed with the work
authortration package and/or test procedures present and in active use. When
applicable, appropriate radiation control measures were implemented. The
inspectors observed supervisors monitoring job progross and quality control
personnel present whenever required by procedure.
M1.2 EAilure to Evalua10_Eincigency Linhtina System Bulb Failures
a. lulocction Ecoce (62.707. 71750]
The inspectors observed the status of the emergency lighting units (ELU) throughout
the plant and reviewed their recent performance history,
b. Observations and Findings
10 CFR Part 50, Appendix R, requ;tes an ELU with at least an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> battery power
supply in all areas needed for operation of safe shutdown equipment and in access
and egress toutes thereto. Updated Final Safety Analysis Report (UFSAR)
Section 9.5.1.2.5 states that emergency lighting is provided in areas needed for
manual operation of safe shutdown equipment and the access and egress routes
thereto. Prior to 1995, the licensee had a relatively high f ailure rate associated with
ELUs. In 1995. the licensee replaced the old ELUs with new Spectron AS BC/BX
Series ELUs The Spectron ELU design features an integrated circuit chip that
automatically monitors the performance of the ELU. Any malfunction of the battery,
charger, transfer circuit, or emergency lamps is indicated by light emitting diodes on
the display panel.
On July 11, the inspectors noted that an ELU located in the stairwellleading to the
control room indicated an emergency lamp malfunction on its display panel. The
inspectors informed the control room and the malfunctiuning bulb was promptly
replaced. On July 14 and again on July 22, the inspectors noted three additional
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ELUs with display panels indicating emergency lump malfunctions. Again, the f ailed
bulbs were promptly replaced. The inspectors noted that neither CRs nor condition
identifications (Cl) were written for these bulb failures. The inspectors requested
copies of all Cls/CRs written against ELU bulb f ailures since 1995 in order to more
fully assess the issue. The inspectors were informed that there were no Cis/CRs
written for ELU bulb f ailures since bulb replacements had been considered minor
maintenance activities not requiring documentation.
The licensee Indicated that 26 ELU bulbs had been removed from the warehouse,
ostensibly to replace f ailed bulbs, during this time period. Thn licensee had
generated a Cl for ELU f ailures not associated with bulbs (i.e., battery f ailure).
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Approximately 5 percent of the ELUs had Cis for various nonbulb failures since
1995. The inspectors were concerned that the failure to document and evaluate
ELU bulb f ailures resulted in the inability of the licensee to accurately monitor the
performance of this system.
The inspectors reviewed the ELU maintenance program to determine if the system
was being treated consistent with the requirements of 10 CFR 50.65,
" Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power
Plants" (the Maintenance Rule). The inspectors identified the following:
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(1) The emergency lighting system was not included within the scope of the
Maintenance Rule Program until January 1997, which was ef ter the
10 CFR 50.65 required date of July 20,1996 for scoping of systems.
(2) Although the system engineer set a personal goal for the system (90 percent
availability), this performance criteria had not been reviewed or approved by
ths Maintenance Rule Exnert Panel.
(3) The system engineer's informal definition of a functional f ailure for the
emergency lighting system (< 90 percent availability) was based solely on the
results of the 6-month surveillance and therefore would not account for f
f ailures discovered during the previous 6 month period leading up to the
surveillance.
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(4) CRs and/or Cls were not written for ELU bulb f ailures.
(5) Plant personnel were not trained to recognize the ELU display panel
indications of malfunction status; therefore, ELUs could have been inoperable
for prolonged periods without recognition of a problem.
On August 7, CR 97 2038 was written to document end disposition these issues.
10 CFR 50.65 requires the licensee to include appropriate structures, systems, or i
components in the Maintenar.co Rule Program by July 20,1990. The f ailure to
include the emergency lighting system in the Mair.*enance Rule Program is a
violation of 10 CFR 50.65. This nonrepetitive, licensee identified and corrected
violation is being treated as a noncited violation, consistent with Section Vll B.1 of
the NRC Enforcement Policy. Specifically, the violation was not willful, actions
taken as a result of a previous violation should not have corrected this problem, and
appropriate actions were completed by the licensee (382/9715-02).
10 CFR 50.65 also requires the licensee to monitor the performance or condition of
structures, systems, or components against licensee established goals. The f ailure
to create licensee established goals for the emergency lighting system is a violation
of 10 CFR 50.65 (50-382/9715-03).
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C. CDHCIUlion5
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The f ailure to include the emergency hghting system in the Maintenance Rule
Program is a timely manner and f ailure to estabbsh performance criteria are
violations of the Maintenance Rule. The violations reflect a lack of a'.ention to
details in the implementation of the Mainter,ance Rule.
M2 Maintanance and Material Condition of Facilities and Equipment
M 2.1 Contatnment31fiMawn.1027071
On July 21, shortly af ter reaching Mode 3, the inspectors toured containment to
determine the effectiveness of cleanup activities following RFO 8 and to assess the
OA inspection conducted earlier in the day. The inspectors noted that there was no
loose material or debris that would potentially ciente missile hazards or clogging
hazards for the emergency sump. Cleanliness and material condition were good. All
scaf folding rigs were disassernbied and the scaffolding material properly stored.
Previous to the inspectors' tour, OA performed inspections and identified several
concerns, including debris in the emergency sump. CR 97-1901 was initiated and
all debris was removed. Tho inspectors concluded that the OA inspections were
e4fective in ensuring containment was ready for power operations.
M8 Miscellaneous Maintenance issues (92902)
1
!
M8.1 LClosed) Violation 50-382/9522-03: Failure to ideritify, schedule, and con tiete
mechanical retests. This item involved the discovery following RFO 7 that not all
required mechanical retests had been completed prior to startup. Corrective action
included the creation of a program within the shif t technical advisor group to ensure
mechanical retests were completed during the RFO. On July 30, the inspectors
were informed that all mechanical retests for safety related equipment had been
completed. The inspectors toured accessible safety related spaces and confirmed
there were no mechanical retest tags in the field.
III. Enointedng I
E1- Conduct of Engineering
E'.1 Containment Vacuum Relief (CVRi Modification (315511
NRC Inspection Report 50-382/9620-03 identified that the actual design
configuration of the CVR system was different from that described in
Amendment 28 to the UFSAR and identified the f ailure to peiform the required
written safety evaluation for the deviation. Specifically, it was identified that the
CVR instrument lines terminated at a location that was not within the controlled
ventilation area system or any other filtration system for postaccident operation.
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Further, the CVR monitoring lines did not meet the design criteria for a closed
system outside of containment and the'; were not seismically qualified.
By letter dated August 21,1996, the licensee submitted a proposed change to
resolve th( design discrepancies regarding the nonessential CVR sensing lines. By
'
letter dated March 17,1997, the licensee responded to an NRC request for
additionalinformation regarding the proposed change and proposed to modify
Attachment 1 to the Operating License regarding position indication for CVR
containment isolation valves for Penetrations 53 and 65. The NRC approved the
proposed change, and Attachment 1 to the operating license was modified to reflect
the proposed change to the CVR system and to incorporate the following statement:
"This attachment identifies items which must be completed to the Commission's
satisf action prior to startup following the refueling outage number 8 "
t
On July 29, the inspectors reviewed Design Change 3502, regarding the proposed
change to the CVR system, and performed a walkdown of the applicable sections of
the system affected. The inspectors concluded that the nonessential CVR sensing
lines were modified consistent with the wording in Attachment i to the Operating
License. '
E2 Engineering Support of Facilities and Equipment
E2.1 CCW Svstem Pressure Soikes
.
a. Scone (92903.37551)
The inspectors evaluated the actions implemented to eliminate the pressure surges
that result in lif ting of thermal relief valves. The inspectors assessed the impact of
the pressure surges on the operability of the CCW system,
b. Observation and Findinos ,
NRC Inspection Report 50 382/95 08 discusses unanticipated lif ting of CCW system
thermal relief valves and the f ailure of the licensee to document abnormal system
.
. operating conditions in the corrective action program. In addition, the report stated
that a special test demonstrated that, during accident conditions, the system -
'
. semained operable and that the loss of volume that would result from all thermal
relief valves lifting could be offset by the capacity of the CCW makeup pump.
. Since that inspection, licensea personnelinitiated a CR each time a CCW system
pressure surge occurred. From November 1995 uritil April 1997, the licensee
,
initiated numerous CRs that documented unanticipated lifting of the reactor coolant
pump oil cooler, containment f an cooler, shutdown heat exchanger, and CCW heat
, exchanger thermal relief valves. Part of the licensee's corrective action involved
1
gagging closed some of the thermal relief valves. The inspectors reviewed the
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10 CFR 50.59 evaluation that supported gaggmg the relief valves and agreed with
the conclusion that no unreviewed safety question existed. The inspectors noted
that the thermal relief valves function to relieve system pressure in the event that
the system was isolated and the entrapped watet heated up.
The licensee performed several operability evaluations, including two root cause
analyses for: (1) CR 96-0429, " Containment Fan Cooler CCW Relief Valve
Actuations," dated May 1,1996, and (2) CR 901088 and CR 961196, "RCP Seal
Cooler CCW Outlet Relief Valve t.if t," dated April 16,1997. The inspectors found
,.
the root cause analyses to be detailed. The licensee concluded that the root cause
l
resulted from poor design. The events occurred during isolation of the nonessential
from the essential portions of the CCW system without a corresponding increase in
flow through the shutdown cooling water heat exchanger. The sudden decrease in
flow momentum resulted in a pressure spike. Specifically, although the butterfly
isolation valves had a 3 second stroke time, the license o estimated that the
1750 gpm flow went to O gpm in approximately 1 second. Similarly, during valve
stroke testing of the shutdown cooling heat exchanger flow control valves, the
closure time for the flow control valve was too short and resulted in pressure surges
that exceeded the system design pressure and the relief valve setpoints. The
licensee concluded that the closed stroke time needed to be lengthened and that a
modification to the actuator would be required to ensure fine control of the close
stroke.
During RFO 8, the licensee lengthened the valve stroke times for the essential to
nonessential supply and return isolation valves to decrease the rapid change in fluid
momentum, in addition, the licensee lengthened the valve stroke time for the supply
and return valves to the reactor coolant pump seal oil coolers. The licensee slowed
the valve stroke time to 4-4.5 seconds. The slower stroke times did not reduce the
pressure transients to below system design pressure. The licensee was unable to
further increase the stroke time without exceeding accident analysis assumptions,
since the valves performed a containment isolation function. Testing was performed
during RFO 8 to bound the magnitude of the pressure spikes by manipulating the
CCW system h a configuration that results in the most severe transient. An
operability evaluation was performed based on the results of this testing. The
evaluation determined that CCW piping and supports remained within acceptable .
l
. code allowable stresses and system components will function as designed,
By letter dated July 21,1997, the licensee committed to reduce the magnitude of
CCW hydraulic transients to below system design pressure no later than the end of
the next RFO. ,
c. Conclusions
The licensee ef fectively implemented design changes and completed operability
analyses to address CCW relief valves lif ting due to system pressure spikes. l
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E4 Engineering Staff Knowledge and Performance
E4.1 .C_omoonsatorv Actions for Fire Vulnerability
a. lasnection Scoce (37551. 71707. 71711)
The inspectors reviewed the licensee's compensatory actions in response to the
identification of a SUPS common mode failure vulnerability.
f b. .Qhnervations and Findinas
1
As discussed in Section 08.1 of this report, on June 4,1997, the licensee identified
a common modo f ailure vulnerability associated with the SUPS units. The common
mode failure vulnerability involved postulated fire damage to cables in various plant
locations (i.e., cable vault area) that do not meet the Appendix R separation / fire
barrier criteria, combined with the lack of SUPS breaker coordination, which could
result in the momentary shutdown of both trains of safety-related SUPS. Although
the interruption of SUPS power would probably be of short duration, Operations
might be required to manually reset safe shutdown loads on multiple SUPS trains,
i
By letter dated July 3,1997, the licensee described the issue and corrective actions
and provided their assessment that their compensatory actions (firewatches) were
adequate to support a startup from the RFO. The letter indicated that there were j
numerous locations where cable train separation was less than the 20 feet required
by Appendix R; however, a control room / cable vault fire would be the bounding '
scenario. The letter also indicated that the condition was manageable with existing ;
procedures.
>
On July 8,1997, the inspectors met with various licensee personnel to discuss the
common mode f ailure vulnerability, compensatory actions, and the basis for the
licensee's assertion that this was not a startup issue. The inspectors were informed
that the licensee's safe shutdown analysis assumed the unprotected train (Train A)
of safe shutdown equipment was lost following a fire (Appendix R) in the cable vault
or control room areas. The Train 8 safe shutdown circuits were designated as the
protected train and are isolated from the cable vault via design and procedure.-
SUPS Train B powers 47 safety related loads of which 23 are considered " essential"
- and 24 " nonessential." The nonossential cables are unprotected and therefore
susceptible to fire induced f aults. During an Appendix R fire, there could be up to
24 successive shorts as the fire spreads and engulfs the noneesential cables. Due
to the lack of coordination between the SUPS f ast acting automatic shutdown
circuitry and the individual load circuit breakers, these fire induced shorts could
-result in repeated shutdown of SUPS Train B.
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The inspectors noted that Procedure OP 901502, " Evacuation Of The Contre' Room
And Subsequent Plant Shutdown," was written under the assumption that all
Train B equipment would remain available. During the July 8 meeting, the
int.pectors questioned what additional complications could arise due to ternporarily
losing both trains of SUPS, as opposed to the previously assumed loss of only
Train A. The licensee stated that, if SUPS Train B power was interrupted, then
Chiller B would have to be manually reset at its local control panel. Based on further
discussion during the meeting, the inspectors understood that the loss of both trains
of chilled water would not prevent cooldown to Hot Standby. On July 30, the
inspectors noted that the Waterford 3 Safety Evaluation Report, Supplement 8,
!- Section 9.5,1.4(1), " Safe Shutdown Capability." lists the chilled water system as
necessary to achieve Hot Standby. The inspectors requested the licensee clarify
whether chilled water was required to achieve postfire safe shutdown.
On July 31 at approximately 5 p.m., the licensee informed the inspectors that chilled
water was indeed required to achieve Hot Standby due to the need to maintain safe
shutdown equipment room temperatures. The licensee acknowledged that both
trains of chillers would shut down in this scenario, but believed the scenario was
still manageable in that Operations could reset Chiller B before room temperatures
became unacceptably high. The licensee stated that Chiller B would be restored
within approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. However, the licensee could not say how high room
temperatures would go in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, produce any documentation that a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
interruption of chilled water had been evaluated, or otherwise provide reasonable
assurance that room temperatures would remain below acceptable limits. The
inspectors expressed concern that there was not a formal evaluation to confirm the
ability to achieve safe shutdown with both trains of chilled water inoperable for up
to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
At approximately 5:45 p.m., the licensee informed the inspectors that Chiller AB
- was being placed in service instead of Chiller B until the room temperature issue
could be evaluatud, since Chiller AB was unaffected by SUPS perturbations. At
approximately 9 p.m., the licensee realized that they could not take credit for-
Chiller AB since Procedure OP 901502 required isolating all Train AB equipment.
The shif t therefore entered Site Directive W4.101, " Operability /Oualification
Confirmation Process," to evaluate the f act that Chiller B may require repeated
manual restarting due to SUPS Train B repeatedly shutting done in response to fire
ind aced f aults. The licensee subsequently completed this evaluation, which
concluded room temperatures would not exceed required temperatures within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
and that operations personnel could restart the chiller within this time frame,
c. Conclusom
Engineering performance was good in identifying a SUPS / Appendix R common mode
f ailure vulnerability; however, Engineering support in resolving this issue was poor.
The licensee's decision to align Chiller AB to replace Chiller B in order to temporarily
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resolve the Appendix R concerns demonstrated a lack of understanding of the
procedural requirements to isolate Train AB equipment following an Appendix R fire.
This episode further demonstrated a lack of rigor in resolving this issue.
E8 Miscellaneous Engineering issues
E8.1 LCloadLLEBJiga82/96 007: Low ACCW flow to essential chillers because of
foukng and inadequate design.
This LER described a condition that placed the facility outside of the design basis.
As a result of corrective actions for mispositioned throttle valves and corresponding
low flows in she essential chilled water system (refer to LER 50 382/95 007), the
licensee initiated a review of throttle valve positions and performed special flow
balance tests of the ACCW and CCW systems. On April 11,1996, the licensee
identified that they could not achieve the specified ACCW B flows through both the
essential chiller (850 gpm) and the CCW heat exchanger (5000 gpm). As
documented in NRC Inspection Report 50 382/96 202, Section E1.1.1, CCW Train B
remained operable at reduced ACCW flow through the heat exchanger as
demonstrated by the calculations of the heat loads that occur during the accident
time sequence and the lack of fouling in the CCW heat exchanger..
The licensee attributed the root cause to inadequate design and analysis. Originally,
the ultimate heat sink provided cooling to the CCW heat exchanger but did not
include flows to the essential chillers. The final design specifications increased the
- ACCW flow to include the flow required by the essential chillers.
Calculation MN(Q) 9-050, "ACCW System Resistance," used ACCW Train A to
determine the system resistance and concluded that both ACCW trains could
support 5000 gpm flow to the CCW heat exchanger and 850 gpm flow to the
essential chiller. During review of.the flow deficiency in ACCW Train B, the licensee i
determined ACCW Train B had higher system resistance when achieving the design
'
accident flow rates to the CCW heat exchanger and the essential chiller. The
licensee further determined that preoperational testing, performed in accordance
with Procedure SPO 36-002, " Component Cooling Water Flow Balance and Pump
Performance," had f ailed to confirm a combined flow of 5000 gpm to the CCW heat
exchanger and 850 gpm to the essential chiller for either ACCW train.
The licensee initiated the following as corrective actions to prevent recurrence:
(1) ensured chemistry controls identified and limited biofouling; (2) developed formal
flow balance procedures for the ACCW and the CCW systems; (3) modified '
ACCW B as necessary to provide design accident flow rates to Essential Chilled
Water Train B and to the CCW Train B heat exchanger: (4) ensured that throttled
valves identified during preoperational testing are correctly specified in system
operating procedures; (5) verified periodic testing of throttled valves; and
(6) ensured sufficient ACCW flow to the escential chillers in the event of a tornado
that damages the DCT.
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The inspectors verified that the licensee had established guidance and chemistry
limits in the chemistry procedures for controlling biofouling. The inspectors
determmed that the licensee had developed emergency response resource guides to
ensure an adequate supply of ACCW water under varying plant conditions, including
-
a tornado event.
The inspectors confirmed that the system engineers developed a list of throttled
valves in their systems and identifico diverse methods for determining valve
position. Further, the system engineers identified systems that receive periodic flow
tests and evaluated the flow tests to determine whether the test appropriately
tested the components. Following these reviews, the licensee developed a list of
safety related throttled valves that the systern engineers compared to system
operating procedure valvo lineups to ensure that all throttle valves are periodically
verified and that the valves are in the same position as the startup test procedures.
Because Valves ACC 126A(B), ACCW flow control valves, automatically modulate
to increase or decrease ACCW flow through the CCW heat exchanger to maintain
the CCW outlet temperature below 115'F, the licensee determined that the
ccmbination of CCW heat exchanger fouling and low ACCW Train B flow would
have resulted in six instances of both chilled water trains being inoperable during the
same period. These instances all could have resulted in unknown, but required,
entry into TS 3.0.3 for inoperable essential chilled water systems. The inspectors
reviewed the methodology used by the licensee to determine these weather
conditions and identified no concerns. The inspectors confirmed that the data
indicated that, although entry into TS 3.0,3 was required, the time duration for five
instances did not exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and for one instance did not exceed 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The
inspectors concluded that no TS required shutdown would have been necessary.
The failure of the licensee to establish appropriate design basis flows for the
ultimate heat sink is a violation of 10 CFR Part 50, Appendix B, Criterion 111. This
nonrepetitive, licensee identified and corrected violation is being treated as a
noncited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy.
Specifically, the violation was identified by the licensee and was not willful, actions
taken as a result of a previous violation should not have corrected this problem, and
appropriate corrective actions were completed by the licensee (50 382/9715 04).
E8.2 (Closed) Unresolved item 50 382/9604-02: Review past operability of CCW and/or
essential chilled water.
The inspectors identified this as an unresolved item to ensure review of the root
cause analysis for CR 96 0543 and any past operability concerns. The Safety
System FunctionalInspection that ended on October 10,1996, evaluated, in part,
the results of the flow balance testing. The inspection team had concluded that the
CCW system remained capable of performing its design basis function. This item is
addressed in Section E8.1 by closure of LER 382/96-007.
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E8.3 LClasadLUnresolved item 50 382/9fi20102: CCW system DCT overpressurization
issue.
In October 1996, NRC inspectors questioned operating the CCW system with
system operating pressures that could exceed the design pressure of the DCT,
Specifically, the inspectors identified that the DCT manifold pressure at low flow
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conditions would approach 145 psig while the DCT had a design pressure rating of
130 psig. The licensee initiated CR 961553 that demonstrated the CCW system
remained operable so loog as the flow remained greater than 4000 gpm. The
licensee provided guidance to the operators about the required operating condition
of the CCW system, in addition, the licensee established a long term corrective j
action to rerate the DCT.
During this inspection, the inspectors determined that Calculation MN(019 040,
" Design Pressure of Component Cooling Water System," Revision 0, dated
February 1984, determined the design pressure of the CCW system as 125 psig and
135 psig at the DCT ( 35 foot elevation) for the losest possible expected operating
flow condition. The calculation also demonstrated that the highest possible flow at
- the shutoff head of the CCW pumps would be 148.7 psig. The licensee used this
shutoff head value to demonstrate that the piping in the area of the DCT would !
withstand the highest potential operating pressures. The original calculation did not l
address any effect on the DCT. ,
A self assessment of the CCW system in December 1988 identified that
Calculation MN(O) 9 049 did not rigorously comply with the requirements of ASME
Code, Section 111,1971 specifically, the self assessment identified that:
(1) although the calculation utilized the maximum expected discharge pressure
(lowest flow) during normal operations as specified in the UFSAR, this flow
condition was not necessarily the highest possible operating pressure of the system,
and (2) the version of the code applicable to Waterford 3 does not limit you to the
pressure and temperature limits during normal operation, hence the most
conservative application of the code for determining a maximum operating pressure
would be the pump shutoff head.
Subsequently, the licensee completed Celculation MN(Q) 9-049, Amendment 1,
dated September 11,1989, that specified, for a minimum flow of 150 gpm (shutoff
head of 209 feet), the pressure at the intet to the DCT waterbox was 145 psig. The
licensee concluded that the existing pipe schedules were acceptable since the
maximum operating pressure was less than that used during the pipe wall thickness
determinations (148.7 psig),
Although the original pipe wall thickness determinations identified that the piping
could withstand the highest possible pressure; and the calculation periormed in
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1989 determined that the piping could withstand the highest possible pressure
generated at the DCT waterbox, the licensee did not apply these higher potential
pressures to the design rating of the DCT. The licensee indicated that the DCT had
been ferated to at least 155 psig.
,
The f ailure to apply appropriate system operating pressure design information to all
I applicable components is a violation of 10 CFR Part 50, Appendix B, Criterion Ill.
This nontepetitive, licensee identified and corrected violation is being treated as a j
noncited violation, consistent with Section Vll B.1 of the NHC Enforcement Policy. j
Specificall/, the violation was identified by the licensee and was not willful, actions
taken as a result of a previous violation should not have corrected this problem, and
appropriate corrective actions were completed by the licensee (50 382/9715 05).
IV. PlanLSupnott
S1 Conduct of Security and Safeguards Activities
S 1.1 Secur.ity Guard Inattentive To Duties (71750)
On July 25 at approximately 11 p.m., during a plant tour, the inspectors noted that i
the security guard manning the SAS was inattentive to his duties in that he was
reading a novel while on watch. The security shif t supervisor was informed, who in
turn had the SAS officer relieved by another officer. CR 971949 was initiated to
document the incident and was subsequently designated Category 1, "Significant
Adverse Condition," which requires a formal root cause analysis.
The licensee's review of the SAS alarm history indicated that the officer had
acknowledged all required alarms while on watch, which mitigated the direct
significance of this incident. The inspectors noted there was no procedural guidance
on what could be read while performing a security watch. Licensee management
stated that it was their expectation that central alarm station and SAS officers
would not read material not consistent with the performance of their duties while on
watch. The SAS officer reading unofficial material while on watch represented an
example of inappropriate watch standing practices by security personnel.
F2 Status of Fire Protection Facilities and Equipment
F 2.1 Eailure to Provide an ELU in Required Area (71110]
On August 4,1907, the inspectors noted that there was not an ELU in the reactor
auxiliary building stairwellleading from the + 21-foot elevation to the +46 foot
chiller room. The r46 foot chiller room contains the essential chillers which provide
cooling water for air handling units that cool spaces containing safety-related
equipment. Safety Evaluation Report, Supplement 8, Section 9, and UFSAR
Table 9.5.1-4 both indicate that the chilled water system is required for safe -
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shutdown following a fire.10 CFR Part 50, Appendix R, Section ill(J), requires ELU ,
with at least an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> battery power supply shall be provided in all areas needed for '
operation of safe shutdown equipment and in access and egress routes thereto. The
failure to provido required ELU is a violation of 10 CFR Part 50, Appendix R
(50 382/97'5 06).
The licensee initiated CR 97 2003 to evaluate the lack of ELU in the stairwell
leading to the essential chiller units. The fire protection system engineer determined
that, although not battery powered, the florescent lighting in the stairwellis
powered from Motor Control Center 311B, which can be powered from an
emergency diesel generator. The licensee, thereforo, concluded that no immediato
corrective actions woro necessary. As of tno end of this inspection period, the
licensco was still ovaluating this issue generically to try to determino if other areas
did not have the required ELU.
L_Manageninct.Menikuis
X1 Exit Meet ing Summary
The inspectors presentea the inspection results to mernbers of licenseo management
on August 14,1997. The licensee acknowledged the findings presented.
The inspectors asked the licenseo whether any materials examined during the
inspMtion should be considered proprietary. No proprietary information was
identified.
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AIIAC11 met 1I
SL'PPLEMLt{LAkit4E0llMaI10N
EABIIAL LIST OF PEBSQHS_C..ONTACTED
Eltcascu I
F. J. Drummond, Director Site Support
C. M. Dugger, Vice-President, Operations
E. C. Ewing, Director Nuclear Safety & Regulatory Affairs
T. R. Leonard, General Manager, Plant Operations
D. C. Matheny, Manager, Operations
G. D. Pierce, Director of Quality
D. W. Vinci, Superintendent, System Engineering
A. J. Wrape, Director, Design Engineering
INSEECTION PROfEDJ)EES_1). SED l
37551 Onsite Engineering
j
61726 Surveillance Observations
62707 Maintenance Observations
71707 Plant Operations
71711 Plant Startup From hefueling
71750 Plant Support Activities '
92901 Followup Plant Operations
92902 Followup Maintenance
93702 Event Response
ITEMS OPENED. CLOSED. AND DISCUSSED
DD_ened
50-382/9715-01 NCV CRAC System Single Failure Vulnerability (Section O2.1.b)
50-382/9715-02 NCV Failure to include the Emergency Lighting System in the
Maintenance Rule Program (Section M1.2.b)
50-382/9715-03 VIO Failure to Establish Performance Critoria for the Emergency
Lighting System (Section M1.2.b)
50-382/9715-04 NCV Low ACCW flow to essential chillers because of fouling and
inadequate design (Section E8.1)
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50-382/9715 05 NCV CCW system DCT overpressurization issue (Section E8.3)
50 382/9715 06 VIO Failure to provide required ELU (Section F2.1)
C10!ind
50 382/9715-01 NCV CRAC System Single Failure Vu'nerability (Section 02.1)
50-382/9715 04 NCV Low ACCW flow to essential chillers because of fouling and
inadequate design (Section E8.1)
50-382/9715-05 NCV CCW system DCT overpressurization issue (Section E8.3)
50 382/9522-03 VIO Failure to identify, schedule and complete mechanical
retests (Section M8.1)
50 382/9702-01 URI Review adequacy of vital 125 Vac breaker coordination
(Section 08.1)
50 382/96-007 LER Low ACCW f!ow to essential chillers because of fouling and
inadequate design (Section E8.1)
50 382/9604 02 URI Review past operability of CCW and/or essential chilled
water (Section E8.2)
50 382/96202 02 URI CCW system DCT overpressurization issue (Section E8.3)
Dlicuned *'
50-382/97 20 LER Potential safety related SUPS common rnode f ailure
(Section 08.1)
50 382/9620 03 eel Failure to perform required written safety evaluation
(Section E1.1)
50 382/97 17 LER Control room emergency filtration unit common modo
f ailure (Section 02.1)
)
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3-
LISLQfJiCHQ11YldS_USED
ACCW auxiliary component cooling water
l
ASME American Society of Mechanical Engineers
CCW cornponent cooling water
cfm cubic feet per minute
CFR Code of Federal Regulations
Cl condition identification
CR condition report
CRAC control room air conditioning
CREFU control room emergency filtration units
CVR Containment Vacuum Relief
DCT dry cooling tower
ELU emergency !ighting units
gpm gallons per minute
l
iwg inches of water gauge
LER licensee event report i
NRC Nuclear Regulatory Commission
PDR Putlic Document Room
ppm parts per million
psig pounds per square inch gaugo
QA quality assurance
RFO refueling outage
tpm revolutions per minute
RWSP refueling water storage pool
SAS secondary alarm station
SUPS static uninterruptible power supply
TS Technical Specification
TSCR Technical Specification Change Request
UFSAR Updated Final Safety Analysis Report
Vac volts alternating current