ML20198N303
| ML20198N303 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 10/24/1997 |
| From: | Chris Miller NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20198N273 | List: |
| References | |
| 50-010-97-13, 50-10-97-13, 50-237-97-13, 50-249-97-13, NUDOCS 9711030260 | |
| Download: ML20198N303 (29) | |
See also: IR 05000010/1997013
Text
~
.
U.S. NUCLEAR REGULATORY COMMISSION .
-
!
REGION lli
l
'
.
Docket Nos: 50-10;50 237;50 249
License Nos: DPR 2; DPR 19; DPR 25
Report No: 50-010N7013(DRP); 50 237/97013(DRP); ,
'
50 249/97013(DRP)
Licensee: Commonweahh Edison Company
Facility: Dresden Nuclear Station Units 1,2 and 3
I
Location: 6500 North D,esden Road
Morris, IL 60450
Dates: July 15 though August 27,1997
Inspectors: K. Riemer, Senior Resident inspector
D. Roth, Resident inspector
C. Settles, Illinois Department of Nuclear Safety Resident
inspector
B. Dickson, Resident inspector in Training
C. Brown, Resident inspector, Big Rock Point
R. Lerch, Project Engineer
Approved by: C. G. Miller, Acting Chief
Reactor Projects Branch 1
.
I
9711030260 971034
PDR ADOCK 050000t0
0 PDR
_
. - - . - - - - - - - - - - - - -
--
. - . - - _ - - - - - - . - - - - - - . - - - - - - - - -
-
EXECUTIVE SUMMARY
Dresden Nuclear Station Units 1,2 and 3
NRC Inspection Report No. 5010/97013(DRP); 50 237/97013(DRP); 50 249/97013(DRP)
'
i
This inspection included aspects of licensee operations, maintenance, engineering, and plant
i
support. The report covers the period from July 15 though August 27,1997, of resident
, inspection augmented by NRC staff from other sites and from the Region til office. '
.
Operations
.
' The inspectors determined that the Unit 2 arid Unit 3 core spray systems were aligned
~ according to procedures. Operators followed procedures during core spray system
testing. (Section 02.1)
i
.
While day le-day operations activities were generally performed satisfactorily, there were
several performance deficiencies by operations personnel during this inspection period.
These included a shift manager's failure to maintain adequate control room staffing, a unit
supervisors failure to assess Technical Specification requirements prior to a surveillance
test, a nuclear station operators inability to control reactor water level, and a non-licensed
operators failure to comply with radiation protection requirements. Some of these issues
were indicative of inadequate corrective actions. (Section 04.1)
.
The inspectors did not identify any performance deficiencies during the Unit 2 startup
from a forced outage. (Section 04.1)
.
The licensee identified that control room staffing was not maintained in accordance with
Dresden Technical Specification 6.2.B. on August 17,1997, The shift manager, while
acting as the Unit 2 unit supervisor, left the control room for one minute without relief
.
while working on an emergent issue. This event was similar to an event on February 26,
1997, and was considered an example of a violation for inadequate corrective action.
(Section 04.2)
.
The licensee did not implement adequate corrective actions for a January 1997 error in
applying a Technical Specification limiting condition for operation (LCO). As a res Jit, a
4
similar event occurred on August 21,1997, when licensed control room operators missed
a required entry into a Technical Specification LCO action statement. This was
.
'
.
considered an example of a violation for inadequate corrective action. (Section O4.3)
.
Operator error while controllir,g reactor vessel water level resulted in a manual reactor trip. The primary root cause of the loss of re tctor vessel level control was an inability of
,
the operating crew to use (esJwater flow, steam flow, and reactor vessel level together to
' stabilize the reactor. Contribuilng causes included the material condition of the reactor
feedwater pump recirculation valve and the feedwater control system. (Section 04.4)
.
Non-licensed operator performance was mixed; a non-licensed operator showed good
initiative in questioning the area temperatures for the hydrogen analyzers. However, nos
licensed operators did not identify unsecured breakere nor changed radiation survey
maps. A violation for failing to review survey maps and a violation for inadequate
corrective actions regarding an unsecured breaker were identified. (Section 04.5)
2
,
!
._ - , , . . _ ~ . . _ _ . - . _ _ ._.. , _.._ -,..__ _
. _ _ _ . . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _
.
Maintenanog !
+
The maintenanos work packages reviewed by the inspectors were completed in
l!
accordance with procedures. Nevertheless, inadequate performance of maintenance and
degraded material condition caused abnormal system alignments and componont
,
malfunctions in the condensate and feedwate system that challenged the operations
staff. (Section M2,1)
i
+ ,
,
During a review of the onuse for high pressure coolant ir$ection system problems, the
!
licensee identified that one valve was binding and not seating property and that another ;
valve had improper seat loading. The licensee also identified several examples of poor
performance by maintenance personnel, Although each example was of minor
significance, collectively, the poor performance of maintenance personnel resulted in - '
inoperable safety-related equipment which had an impact on plant operations.
- (Section M4.1)
.
The standby gas treatment system failed during a routine surveillance tes.. Although the
system was restored to an operable status within the time iimits allowed by Technical
Specifications, various personnelissues slowed troubleshooting and restoration efforts.
Electrical maintenance personnel could .1ot be called into work and the electrical
maintenance supervisor and work analyst assigned to work in place of the electrical
maintenance personnd did not take correct readings. As a result, the licensee did not
diagnose the t. . - "se of the standby gas treatm ent system failure in a timely manner.
The material conditi . of the standby gas treatment system, combined with personnel
error, impacted plant operations and the availability of safety-related equipment. (Section
M4.2)
Enaineerina
.
The licensee missed prior opportunities to identify aad msolve a 250 V-dc battery loading
issue. Engineering parsonnel did not adequately review the battery calculation when .
responding to previously identified technical concems. (Sec':on E4.1)
+
Information used to track the performance indicator of * Engineering Requests" and
" Temporary Alterations" was potentially unrollable and not always an accurate indicator of
performance. The licensee modified the way the information was tracked to make it a
more reliable indicator of performance. (Section E8.1)
Plant Support
.
+-
There were no formal controls over the waming flags (on radiation survey forms) that the
,
radiation protection staff uved to lndicate that radiological conditions had changed. This
resulted in confusion about why a flag was hung and caused thm licensee to re-survey sn
area. (Section R4.1)
,
'
-+-
The licensee performed pooriy in response to a perceived thseat from a toalc gas release.
Security personnel performed poorly in the verification of information, communications,
and execution associated with a shelter order for the perceived threat. Many
stationersonnel either chore not to obey an order from the control room to stay indoors or
.
3
,
i
q e-, w ,. s - y-----a,-e,m.,,9 ,a. , , mv,,-.n---.,m, ,y .a ., , -e--w- - -- ,,,c,n,. , ,- , -enens...-r-,--
.
did not hear the order. The investigation into the cause of the event was weak because it
did not address all of the observed deficiencies. (Section S4.1)
4
l
l
4
k
!
a
I
J
.
$
4
~ - - , , - . . . . , . - . . . - . . . - . . . - - _ - . -
=
Report Details
Summaryof Plant Gtatus
-
Unit 1 activities to support SAFSTOR were performed dudng tn!s ped:4. The fuel stored ni te
Unit i Nel pool was visually characterized. The licensee identified various d%radat!ons of the
fuel assemblies, including some fuel rod failures. Other Un!t 1 activkos induded preparations for
processing of old liquid radwaste.
At the beginning of this inspection pe, lod, Unit 2 was at full thermal power. The licensee
performed planned power decioases during the period from July 15 (o July 21 to attempt to
comply with limits on river discharge temperatures. On July 27, the licensee reduced Unit 2
power to about 680 megawatts electric (MWe) in response to a level transient caused by a
cycling mactor feedwater pump minimum fiow valve. On July 28, the licensee manually tripped
Unit 2 in response to a level transient that occurred during operators' attempt to change
feedwater pump lineup. On July 31, the licensee started Unit 2. On August 1, the licensee
synchronized Unit 2 to the grid. On August 12, the licensee rapidly decreased pov/er from
816 MWe to 635 MWe due to failed reactor feedwater pump ventitation. Power was restored and
the licensee maintained Unit 2 at full power, or in compliance with load dispatcher requests, for
the remainder of the inspection period.
At the beginning of this inspection period, Unit 3 was near full thermal power. Full thermal power
on Unit 3 was not achieved because the main turbine control valve pos~tions wero limited to an
average of 85% open with no greater than 90% open on any one control valve, and feedwater
flow was limited to 9.73 M!bm/h (instead of the approximately 9.8 Mim/h at full power) as a
result of a review of the fuol cycle analysis performed by engineering personnel. These limits
remained in effect until the end of the inspection period. During the pedod from July 15 to July
21, the licensee performed planned power decreases to attempt to comply with limits on river
discharge temperature. On July 25, the licensee reduced reactor load due to the effect of control
room humidity on the B recirculation pump controller. Humidity in the control room was high
because of a blockage of the B control room refrigeration drain line. On July 28, the licensee
reduced reactor power to 95% in accordance w,th the procedure for isolation of the 3D2
feedwater heater. The licensee isolated the feedwater heater due to repeated level problems
that were caused by a flashing reference leg. On August 13, the licensee rapidly decreased
power from 814 MWe to 700 MWe due to failed reactor feedwater pump ventilation. Power was
restored and the licensee maintained Unit 3 at full (limited) power, or in compliance with load
dispatcher requests, for the remainder of the inspection period,
l. Operations
01 Conduct of Operations
O1.1 General Comments (71707)
The inspectors conducted froquent reviews of ongoing plant operations. Overall, the
conduct of operations was safe and in accordance with procedures. During the
inspection period, events occurred or were discovered for which the licensee was
required by 10 CFR Part 50.72 or 10 CFR Part 50.73 to notify the NRC. Some of the
events and the notification dates are listed below:
5
- _ __ _ . - - - . _ - _ . _ . - . . _ _ _ _ ._ _ . _ _ _ _ _ . _ _ _ _ _
~
July 28
(Unit 2) A manual reactor trip due to reactor vessel level transient during
an attempt to change reactor fecdwater pump lineup.
Based on a preliminary assessment of the licensee's response to the event, the
inspectors determined that the licensee's response to the event was adequate.
Additional review was performed under inspection Report No. 50-237/97016.
02 Operational Status of Facilities and Equipment
02.1 (Units 2. 3) Enaineered Safety Feature System Walkdowns
a. Inspection Scope (71707)
The inspectors conducted a detallad walkdown of the Unit 2 and Unit 3 core spray
systems to verify operability of the systems, assess material condition, and to ensure that
the alignment procedures, piping and instrumentation diagrams (P&lD), and the as-built
configurations were current. The inspectors reviewed:
DOP 1400-M1 Unit 2 Core Spray System (mechanical), Rev.15
+
DOP 1400-M1 Unit 3 Core Spray System (mechanical), Rev.15
- .
COS 1400-05, Unit 3 Core Spray System Pump Test With Torus Available,
Rev 15
+
P&lD M 27, Core Spray Piping Unit 2, Rev. YU
+
P&lD M-358, Core Spray Piping Unit 3, Rev. BR
b. Observations and Findinas
During the walkdown of the core spray systems, the inspectors determined that the
material condition of both systems was adequate. The aligninents of the systems were
found to be in accordance with the operating procedures.
The inspectors also observed the performance of Dresden Operebility Surveillance
(DOS) 1400-05, " Core Spray System Pump Test With Torus Available," for the 3A Core
Spray Pump. The inspectors watched portions of the surveillance and found that the
operators were following the procedure and had the most recent revision av.ailable.
When questioned about procedural steps, the operators demonstrated a detailed
knowled0e of the core spray system and surveillance steps.
c. Conclusion
.
>
The inspectors determined that the Unit 2 and Unit 3 core spray systems were aligned
according to procedures. Operators followed procedures during core spray system
testing.
i
6
-.- --- . . - _ . - .-_ .- .
.. _._.. .._ _ ___ . .-__. _ _ . . _____ _ _ _ -
_ ___ . _ . _ -.
!
~ l
l
04 Operater Knowledge and Performance -
l
'
04.1 [WDR?J1.QDMBU90EEEf9BDA091
$ a. Insoection W (71707)
,
.
The inspectors conducted frequent reviews of operations departms et 6s,
procedure adherence, and swersness of plant conditions durine dx 4 routine operations of
Unit 3 and the serem response and startup operations of Unit 2. Prooedures and
documents reviewed included DGP 01-01, " Unit Startup," "U. tit 2 84.artup Plan (D2F29
-
July 1997)," DGP 02 03, " Reactor Scram," and the non licensed operator sounds .
'
procedures. The inspectors also interviewed oporators regarding self identified and
, inspector-identified operator errors.
- b. Observations and Findinas '
i Unit 2 Startup Activities
a
!
! The Unit 2 startup activities from 02F29 witnessed by the inspectors were performed
1
well. The inspectors observed operators enforce three-way communications, follow plant
procedures, and conduct informative tumovers.
>
- - Routine Activities
' During this inspection period, there were several performance deficiencies exhibited by
the operat'ons staff;
A shift manager who was acting as a temporary unit supervisor left the control
room.
,
A unit supervisor and a peer-check unit supervisor did not successfully determine
] what Technical Specifications applied before authorizing a surveillance.
.
A nuclear station operator was unable to controllevel during the start of a third
i reactor feedwater pump, causing the unit supervisor to order a manual reactor ;
.
trip.
' A non-licensed operator did not review survoy maps or otherwise consult with
radiation protection (RP) personnel before performing rounds and was unaware -
,
a
that RP personnel had determined that radiological conditions had changed.
.
>
+
A unit supervisor did not assure prompt resolation of a breaker that was left
- unsecured in front of safety-related Bus 33-1.
l Several of these issues were similar to previously identified issues, s
c. ' Conclusion
,
While day-to-day operations activities were generally performed satisfactorily, there were "
several performance deficiencies by operations personnel during this inspection period,
i
.
F
7
I ,
a
.w.4. , ,n, , . *_, ,.w,.,.e.w, ,w...s ,,,.e ,.m, ,,-nw,,y--,,,,.. .,,,,,,.,,.m,m.m., ,,,.e_-..,. ,, - w., f_,,p.eg.gw, ,, .ww 96,,, m.,7,-_7--2._yey $
.e-t ,- %.-.-, , _ . . ..
._. _ _ _ _ _
\
o
Deficiencies included a shift manager's failure to maintain adequato control room staffing, ,
a unit supervisors failure to assess Technical Specification requirements prior to a ;
surveillance test, a nuclear station operators inability to control reactor water level, and a
non-licensed operators failure to comply with radiation protection requirements. Some of
these issues were Indicative of inadequate corrective actions.
The inspectors did not identify any perfortnance deficiencies during the Unit 2 startup
from a forced outage.
04.2 runit 2) Unit Staffino Below Technical Specification Reaulrements
a. hLsaggtlon Scop '71707)
.
On Au 3ust 17, the licenseu identified that unit staffing for Unit 2 was inadequate for about
one m Sute because the acJ,(y unit supervisor had left the control room. The inspectors
intervie. ved the involved licensed senior reactor operator (SRO) and independently
reviewe 1 the licensee's investigation.
b. Observations and Findinas
On August 17, the rhift manager (a licensed SRO) provide'd relief to the Unit 2 unit
supervisor (also a licensed SRO), and the unit supervisor left the control room for a short
break. A clip was normally placed on the badge of the SRO who was 19 unit supervisor
in order to remind the SRO not to leave the control room. The ur it supervisor and shift
manager failed to transfer the clip from the unit supervisor's badge to the shift managers
badge during the tumover.
Before providing relief to the unit supervisor, the shift manager had been discussing an
emergent issue with the acting maintenance manager. Soon after assuming the role of
the Unit 2 unit supervbor, the shift mariager again started working on the emergent issue
and left the control room.
The assigned unit superwsor retumed from the break about one minute later and
immediately noted that the shift manager was absent and that control room staffing was
not in compliance with the Technical Specifications. Dresden Technical
Specification 0.2.B, " Unit Staff," requires that "while the unit is in MODE (s) 1,2, 3 or 4 at
least one licensed Senior Reactor Operator shall be present in the control room." The
unit supervisor made the appropriate notificatior,s oflicensee staff and entered the error
into the ILMsee's integrated reporting process.
The failure to maintain adequate unit shffing in the control room was a repeat event from
February 26,1997. censee Event Report (LER) 50-237/97 006 c'iscussed the absence
of r.n SRO from the control room due to loss of focus on interim duties. The February
event was similar to the August 17 event because: (1) the SRO who left the control room
was providing relief for the assigned unit supervisor; (2) the relief SRO was working on an
emergent issue at the time that the relief SRO left the control room; (3) tne absence of
the relief SRO was discovered by other operations staff (not the ones who left); and
(4) the event was of short duration (sN minutes).
8
. . .
.
Criterion XVI of Appendix B to 10 CFR Part 50 states, in part, that measures shall be
established to assure that conditions adverse to quality are promptly identified and
corrected, in the case of significant conditions adverse to quality, the measures shall
assure that the cause of the condition is determined and corrective action taken to
preclude repetition.
Contrary to this requirement, the measures taken in response to inadequate control room
,
staffing identified on February 26,1997, did not preclude repetition of the condition on
August 17,1997. The failure to establish adequate corrective actions was a violation of
Criterion XVI of Appendix B to 10 CFR Part 50 (VIO 50 23797013 01 A(DRP);
50 249/97013-01 A(DRP)).
c. Conclusions
The licensee identified that control room staffing was not maintained in accordance with
Dresden Technical Specification 6.2.B. on August 17. The shift manager, while acting as
the Unit 2 unit sapervisor, left the control room for one minute without relief while working
on an emergent issue. This event was similar to an event on February 26,1997, and was
considered an example of a violation for inadequate corrective action.
O4.3 Missed Unit 2 Technical Specification Reauirements
a. Inspection Scope f71707)
On August 21,1997, control room operators failed to enter a required Technical
' Specification limiting condition for operation (LCO) action statement during the
performance of a maintenance surveillance test. The inspectors reviewed the licensee's
investigation into the issue and compared the circumstances to past similar occurrences.
b. Observations and Findinos
The licensee identified that the control room unit supervisors failed to enter required
Technical Specification LCO action statements subsequent to authorizing the
performance of Dresden Instrument Surveillance (DIS) 1600-16 (Drywall High Radiation
Monitor Group 21 solation Functional and Calibration Tests). The reference section of the
surveillance test listed specific Technical Specifications. The operators entered
Technical Specification 3.2.F (accident monitoring instrumentation), but failed to enter
Technical Specification 3.2.A (isolation actuation). Technical Specification 3.2.A has a
more restrictive LCO tinn clock (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) than Technical Specification 3.2.F (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />). .
Upon discovery, the licensee completed the surveillance test to a point where the
instrumentation was operable then halted the proceduna pending further review of the
situation. The licensee documented the occurrence via problem identification form
(PlF) D1997-06406 and performed a prompt investigation (" Prompt Investigation Report
for Failure to enter Tech. Spec. LCO 3.2.A," dated August 25,1997).
Dresden Administrative Procedure 07 01," Operations Department Organization,"
Rev. 24, Step F.8.k, requires that the unit supervisor review planned activities to
determine if special considerations or precautions are warranted. On August 21,1997,
the Unit 2 Unit Supervisor reviewed DIS 1600-16 and identified contingency actions
associated with Technical Specification 3.2.F; the Unit 3 Unit Supervisor concurred with
9
>
_____ ---
.
the identified LCOs. The next day, during shift tumover, the oncoming Unit 2 Unit
Supervisor determined that the more restrictive LCO of Technical Specification 3.2.A also
applied. The failure of the unit supervisors to identify the correct precautions (entry into
4
the correct Technical Specificatlan) was a violation of DAP 07-01.
On January 23,1997, a similar situation occurred and was documented in NRC
Inspection Report No. 50-010/97004(DRP); 50-237/97004(DRP); 50 249/97004(DRP). In
that case, sperators did not apply a Technical Specification LCO time clock during the
calibiation of main steam line high flow switches The licensee documented the
occurrence and corrective actions teken in Root Cause report number 237 200-97-00700.
The licensee's response and corrective actions for the January 1997 issue did not
prevent a recurrence of the same issue. Criterion XVI of Appendix B to 10 CFR Part 50
states, in part, that measures shall be established to assure that conditions adverse to
quality are promptly identified and corrected. Criterion XVI also states that Iri the case of
significant conditions adverse to quality, the measures shall assure that the cause of the
condition is determined and corrective action taken to preclude repetition. The inspectors
concluded that the licensee's failure to take corrective actions sufficient to preclude
repetition of a missed Technical Specification LCO entry constituted a violation of
10 CFR Part 50, Appendix, Criterion XVI," Corrective Action"
(VIO 50 237/97013 01B(DRP); 50-249/97013-01B(DRP)).
c. Conclusions
The licensee did not implement adequate corrective actiona for a Janua" 1997 error in
applying a Technical Specification limiting condition for operation (LCO). As a result, a
similar event occurred on August 21,1997, when licensed control room operators missed
a required entry into a Technical Specification LCO action statement. This was
considered an example of a violation for inadequate corrective action.
04.4 (Unit 2) Feedwater and Level Transleju
a. [nspection Scope (71707. 93702)
I
'
The inspectors promptly performed on site inspection after being notified by the licensee
that a manual trip (scram) of Unit 2 was performed on July 28,1997. The unit supervisor
determined that the operating crew did not have control of reactor vessel level while
attempting to shift running feedwater pumps. The inspection was to assess what
happened and the current reactor status. Note that NRC Inspection Report
No. 50-237/97016 documented additional review of the event's root cause and
contributary causes.
b. Observations and Findinas
On the attemoon of July 27,1997, the Unit 2 reactor feedwater pump recirculation valve
went from full closed to midway-open, back to full closed, then to full open in about two
minutes. The operators responded by lowering power to within the capability of the
feedwater system and taking manual control of the feedwater regulating valves. The
operators diagnosed the failing reactor feedwater pump recirculation valve and took
posiiive control of the valve, then stabilized reactor vessel level using manual control of
the feedwater regulating valve and automatic control of the low flow feedwater regulating
10
l
9
___.._.____ _____ _ .__ _ _______ __.______
,-
3
'
valve. The licensee concluded that the 2B reactor feedwater pump needed to be
secured, and the 2C leactor feedwater pump started, so that the failed reactor feedwater
pump recirculation valve could be repaired.
The midnight crew attempted to start and align the 2C reactor foodwater pump earty on
i
July 28. The reactor vessellevelincreased beyond what the crew had expected and the
i
nuclear station operator took manual control of the feedwater regulating valve based only
on reactor vessel level. The nuclear station operator then rapidly closed the feedwater ,
- regulating valve while monitoring only reactor vessellevel. A second operator alerted
i
the nuclear station opers. tor that feedwater flow had been significantly reduced. The _ >
i
nuclear station operator then rapidly opened the feedwater regulating valve. Because the
nuclear station operator opened the feedwater regulating valve too much, the reactor
feedwater pump e+edenced low suction pressure and automatically tripped.
,
Coincidentally, the reactor vessellow level alarm annunciated as the effect of closing the
'
feedwater regulating valve impacted the reactor vessellevel. The u74t supervisor
I observed the rapid sequence of the low feedwater flow, trip of a reactoridedwater pump,
and low level alarm, and concluded that the crew did not have control of the reactor. The
unit supervisor therefore ordered that the reactor be tripped and the reactor was shut
j down. The licensee formed a response team and began the process of interviewing the
- operators and determining the causes for the transient.
.
A.
3
- c. Conclusions
t
The July 27 aftemoon operating crew responded well to the erratic operation of the
reactor feedwater pump recirculating valve. The nuclear station operator on the midnight
crew was not successful at restoring reactor vessel level because the operator *
concentrated only on reacting to reactor vessellevel rather than restoring the reactor
system to a balanced and stabie condition. Thus operator error during performance of a
fundamental task resulted in the need for a manual reactor trip. By contrast, the
aftemoon crew was successful by balancing steam fiow and feedwater flow.
04.5 Unit 2. 3 Non-l.icensed Operator Rounds
'
a. Inspection Scope (71707)
The inspectors assessed the effectiveness of non-licensed operator required rounds.
The review included walking down the rounds performed in accordance with the
appendices to Dresden Administrative Procedure (DAP) 07-40, ' Equipment Operator and
Attendant Rounds." When the inspectors observed equipment parameters outside of the
tolerance bands provided by the DAP, the inspectors then determined if the condition had
,,
been identified by operations staff,
b. Observations and Findinos
in general, the operations staff identified equipment parameters that did not f6!! within
tolerances. The inspectors verified that rounds data taken by the non-licensed operators
reflected plant conditions.
11
'
1
. _ _ _ _ . . . , _ ~ . ; u_ .. _ , _ _ _ . _ _ ~ , _ _ . _ _ _ . . . _ . _ _ . . _ _ _ _.._ . . _ . . _ , _ _ _ . _ _ . , . . . .
-
The Inspectors had concems with the following:
Monitoring hydrogen analyzer area temperature,
a
Monitoring Unit 2/3 emergency diesel generator govemor oil sight glass level,
a
Using radi Jon survey maps, and
'
Detecting unsecured breakers.
Each of these concems is discussed below
2/3 Emergency Diesel Generator Governor Oil Sight Glass Level
The inspectors waiked down DAP 7 40, Appendix D, "HVO Inside Round Lo0 sheet," and
identified one issue regarding the 2/3 emergency diesel generator govemor oil sight glass
level. The appendix required the user to verify that "sightglass level (is] between lines."
The inspectors noted that only one line was visible on the sightglass. By contrast, the
Unit 2 and Unit 3 diesel generators' sight glasses did have two lines. The inspectors
' noted that the oillevel visible in the 2/3 govemor sight glass was similar to the levels in
the Unit 2 and Unit 3 diesels.
Operator verification of oil levels in sight glar.ses for condensate and condensate booster
pumps was identified as weak in NRC inspection Report 50-010/97009(DRP);
50-237/97009(DRP); 50 249/97009(DRP) Section 04.1, Inspection Follow Up item (IFI)
50-237/97004-01(DRP); 50-249/07004-01(DRP) w 3 opened to review the effectiveness
of the equipment operator tours in identifying material condition deficiencies, and the
failure to identify the lack of two marks on the sight glass mark is considered an additional
example of issues encompassed by IFl 50-237/97004 01(DRP); 50-249/97004-ODRP).
Unsecured 4-kV Breakers / Portable Equipment
The inspectors noted an adverse trend in instances where portsbit, equipment was found
unrestrained. From May 1996 through December 1996, there was >nly one instance
where portat's equipment was found unrestrained. However, during 1997 there has
been a total of seventsthree instances identified by the licensee. Starting from April of
1997, an average of nearly thirteen instances per month has been identified, with the
peak months being April, May and June (15,15, and 21 instances, respectively).
On July 22,1997, the inspectors discovered that a spare 4160-V breaker was not
properly restrained. Dresden Operating Procedure (DOP) 6500-04, " Racking Out
4160 Volt Manually Operated Air Circuit Breaker," stated, "the breaker must be fully
removed from the cubicle AND the wheels restrained to prevent rolling." This breaker
was found directly in front of safety-related 4 kV switchgear (Bus 33-1). The inspectors
informed the Unit supervisor of the breaker situation.
On July 28,1997, the inspectors discovered that the same spare breaker was still
improperly secured. The inspectors again notified the Unit Supervisor. The Unit 3 Unit
Supervisor stated that actions would be taken to secure the breaker. The inspectors
subsequently verified that the breaker was property secured.
12
.
. _ _ _ _ ___ _ _ _ _ - . _ _ _ _ _ _ _ - __-__- _ _ ___ _ ________
-
O
Inspection Report No.50-010,'95015(DRP); 50-237/95015(DRP); 50-249/95015(DRP)
documented a violation for failing to secure 4-kV breakers. Criterion XVI of Appendix B to
10 CFR Part 50 states, in part, that * Measures shall be established to assure that
conditions adverse to quality . . . and nonconformances are promptly identified and
corrected.' Contrary to the above, the licehsee f8iled to take prompt corrective action to
secure the 4-kV breaker. The inspectors concluded that the licensee's failure to take
{
prompt corrective actions constituted a v%lation of 10 CFR Part 50, Appendix, B, '
Criterion XVI, " Corrective Action"(WO 50 237/97013 01C(DRP);
50 249/97013 01C(DRP)).
Use of Survey Maps
On August 8,1997, the inspectors identified that some non-licerred operators did not
review the radiological survey maps or otherwise discuss area radiological conditions with
the RP staff prior to performance of rounds. The related issue of control of survey map
updates is also discussed in Section R4.1 of this report.
Appendix F, " Unit 2 Turbine Building and Reactor Building Rounds Logsheet," of Dresden
Administrative Procedure (DAP) 7-40 requires the non-licensed operator to go to the
turbine building 601 elevation every shift to verify reactor building ventilation stack sample
pump flow. Pdor to walking down Appendix F during the day shift, the inspectors
reviewed the radiological survey maps and noted that the Unit 2 turbine building
601 elevation map had a waming tag placed over the survey map stating that conditions
,
had changed and that RP personnel must be contacted for the current survey. The
inspectors stopped at access control and RP personnel did not have a new survey for the
Unit 2 turbine building 601 elevation, nor were they aware of why the waming was posted.
The personnel advised the inspectors not to go to the area pnor to a new survey being
done and sent a technician to conduct a survey.
The inspectors cuestioned the non-licensed operator who nad performed the Appendix F
rounds to determine if he had looked at the survey maps. The operator had not and was
unaware of the waming posted over the survey map. The non-licensed opwator who had
performed the Unit 3 Aopendix G rounds also had not reviewed the survey maps prior to
performance of rounds. However, the Unit 3 601 elevation survey map had no waming
posted.
After the day shift was relieved by the aftemoon shift, the inspector noted that the Unit 2
turt ina- building 601 elevation survey map still had a waming posted. The inspectors
questioned the aftemoon shift RP personnel and found that no information about the -
601 elevation was passed from the day shift to the aftemoon shift. The ahemoon shift
located a survey performed by the day shift that, although unreviewed, indicated
radiological conditions had not significantly changed. No non-licensed operator had
informed the aftemoon RP shift about the waming posted over the survey map. This
was an indication that the aftemoon-shift non-Rensed operator performing Appendix F
had not reviewed the survey map prior to work.
Dresden Administrative Procedure (DAP) 12 25, " Radiation Work Permit Program,"
Revision 06, Step F.1.f.(3), requires that personnel, " Verify understanding of work area
dose rates and contamination levels and location af low dose rate waiting areas "
Contrary to the above, on August 8,1997, personnel who performed operator rounds in
13
i
. _
_ _ _ _
,
Appendix F did not verify dose rates for the Unit 2 Turbine Building Elevation 601 general
area prior to performing rounds in that area. Failure to verify dose rates was a victation of
these requirements (VIO 50410/9701342(DRP); 50 237/97013 02(DRP);
50 249/97013 02(DRP)).
Monttoring Hydrogen Analyzer Area Temperature
On August 8,1997, the inspecto.s observed that the temperature in the area of the Unit 2
hydrogen analyze; was 114 degrees Fahrenheit ('F), while the normal values listed in
DAP 7-40, Appendix F, " Unit 2 Turbine Building and Reactor Building Rounds Logsheet"
was from 50 to 110'F. The inspectors' observation was made during a day shift, but
- Appendix F only required recording of the temperature during the midnight sh.ft. A large
space-heaterinstalled under Dresden Administrative Procedure 7-48," Control of Lay
Down, Storage beas, and Equipment in Use,"was heating the area of the hydrogen
analyzer.
-
Simultaneously but independently, the non-licensed operator performing DAP 07-40,
Appendix G, " Unit 3 Turbine Building and Reactor Building Rounds Logsheet," noted that
the linit 3 hydrogen analyzer sample chamber temperature was above the listed limits,
and that a space heater was also directed towaid the Unit 3 hydrogen analyzers. The
Unit 3 non-licensed operator secured the space heater. The non-licensed operator also
contacted the system engineer to determine the impact of the temperatures on the
hydrogen analyzers. The system engineer responded that the analyzers were operable.
The system engineer stated that the purpose of the space-heaters was to maintain area
temperatures above a minimum temperature, and that the heaters should be off when the
weatheris warm. The non-licensed operator questioned the system engineer about the a
bases for the temperature bands and the system engineer replied that the bases were
taken from vendor technical informrtion.
The Unit 3 non-licensed operator noted that unlike the Unit 2 Appendix F rounds, the
i
Unit 3 Appendix G rounds did not require the operators to record the area temperature for
!
the Unit 3 hydrogen analyzers. The inspectors noted that the recording of area
temperature only on the midnight shift precluded determining the maximum daytime
tempsrature. The inspectors discussed this issue with the system engineer and the
regulatory assurance manager and was informed that both Appendix F and Appendix G
will be changed to require recording area temperature on all shifts.
The off gas hydrogen analyzer operation and maintenance manuallisted the ambient
temperatures for the hydrogen analyzer as a nominal 50'F to 105'F with a maximum of
140*F for eight hours. By contrast, the Appendix F operator rounds listed 50*F to 110'F
as the acceptable temperature. The bases for the temperature limits for the hydrogen
analyzers and the impact on equipment operability and the use of space heaters is an
inspector follow-up item (IFl 50 237/97013-03(DRP); 50-249/97013 03(DRP)).
c. Conclusions
in general, the non-licensed operators correctly assessed equipment performance.
Non-licensed operator performance was mixed; a non-licensed operator showed good
initiative in questioning the area temperatures for the hydrogen analyzers, non-licensed
operators did not identify unsecured breakers or changed radiation survey m :ps. A
14
l
.
.. . .
..
. .-
- -
-_ . - . - - - . _ - __ --. . -. ._ _ - - ..
,
violation for failing to review survey maps anu a violation for inadequate corrective
actions were identified. I
!
'
11.Malaienance
M2 Maintenance and Material Londition of Facilities and Equipment
M2.1 Condensate end Feedwater System
'
a. Inspection Scope
The inspectors reviewed the maintenance and performance history for the condensate
and feedwater system to assess maintenance performance against plant requirements.
The inspectors reviewed th following work packages against the requirements of
DAP,15-06 " Work Analyst Guide to Work Package Preparation":
-
WR 950065101-01 3B Moisture Separator Drain Tank (MSDT) to 3D1 heater level
con'.rol valve replacement.
WR 9500i.5101-01 38 MSDT to 3D3 heaterlevel control valve replacement.
b. Observations and Findinas
4
Overall the work packages, including temporary rigging permits and quality control
inspection checks, were completed according to procedures. When questioned about
activities done while completing the tash, the workers were knowledgeable of proce Jural
steps and radiation safety concems.
In reviewing the control room abnormality log and the work request history, the inspectors
noticed that the feedwater and condensate system on both units had continued to
challenge the operators. The~ challenges were the result of feedwater heater level
control, failed valve logic, and other problems. For example:
The 2A1 Flash Tank Drain Valve would not pass adequate flow and caused b!gh
leve! alarms and trips of the flash tank. On May 19, the 281 feedwater heater
emergency drain was biased open to help maintain the appropriate levc4 in the
2A1 flash tank. This abnormal condition continued throughout the inspection
period.
-
On June 27, during startup following outage D3R14, the positioner arms on the
3-3508A and on the 3-3508C level control valves (LCVs) became unattached from
a Moisture Separator Drain Tank (MSDT) drain valve stem, preventing the LCVs
from opening. The licensee concluded that the valve actuator to volve stem
coupling (clamp) was loose allowing the coupling to rotate. The licensee also
concluded that because the limit switch arm, which is also connected to the stem
coupling, became misaligned, there was no control room annunciation of this
condition. On July 1, the licensee discovered a gross body to bonnet leak on the
3C1 feedwater Heater Extraction Valve (NO 3-3102A).
15
. . . - -. - -. . - -- --. ..- - - .- ---.
,
,
On July 11, the control room received continuous 3B MSDT ncmal drain valves
,
(3 3508 A, B, and C) full closed alarms during operations. Tb i ensee
F
concluded that these alarms were caused by a combination of unequal valve
L positions, instal.ation tolerances of the limit swilches, and a change of valve
- '
design resulting in the referenced valves cycling near the closed position.
.
,
On July 12, the 3D2 feedwater heater tripped on hth level due to the draining and
.
flashing of the reference log of a Barton level trip switch (3 3541-558). This
caused a falso high level signal. The licensee calibrated the level instrument and
i
backfilled the reference leg. Later on July 15, the 3D2 feedwater heater tripped
again on high level. Both incidents caused a load reduction of 50 MWe to occur.
l
The 3D2 feedwater heater was isolated and operations personnel refused to
i
declare the system operable until engineering and maintenance staff fixed the
problem by installing a temporary alteration (see August 3 entry below).
Along with these material condition issues, the inspectors also noted some instances
where the performance of maintenance personnel created situations where operators
were challenged.
4
On May 20, the Unit 3 turbine tripped after the MSDT emergency valve failed
2
opened. The supply at. line to the 3B MSDT emergency level control transmitter
was disconnected. The licensee concluded that the technician who had earlier
,
performed work on the MSDT failed to connect the air line from the regulator to
the emergency level transmitter (reference IR 97012).
l *
On August 3, b aa attempt to alleviate repeated draining and flashing of the 3D2
high pressure heae.ir high level switch (3-3541 558) reference leg, the licensee
j attempted to install a temporary alteration (Reference Temp Alt 111-1597 and
Wt#970082873). The temporary alteration instructed maintenance personnel to
connect the reference leg downstream of an existing condensing pot. Acceptance
1-
criterli ns o it satisfied during the post installation test. The licensee
discoWW v #e reference leg had been connected to the wrong pipe. Results
of inspeck. interviews of maintenance and engineering personnel suggest
! workers became confused due to labeling problems during installation. This
-
demonstrated poor work practices since DAP 05-08 * Control of Temporary
i
System Alterations" stated that "while performing a temporary alteration, if field
.
conditions are different from those shown in a work package, then stop and
contact the preparer." Pending further inspector review of work performed in the
>
field and pracedural requirements, this is an unresolved item .
(URI 504019701344(DRP); 50-237/9701344(DRP); 50 249/9701344(DRP)),
Most events mentioned above have required or will require rework in contaminated and
high radiation areas, and may increase the dose received by maintenance personnel.
'
c. Conclusion
The maintenance work packages reviewed by the inspectors were comple'.ed in
i
accordance with procedures. Nevertheless, inadequate performance of maintenance and
degraded material condition caused ebnormal system alignments and component
malfunctions in the condensate and feedwater system that challenged the operations
.
16
_ , _ _ . _ _ _ . , _ _ ..- _.~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
.
P
staff. Also, required rework had the potential to increast he dose reelved by !
t maintenance personnel.
? ,
M4 Maintenance Staff Kr.r tO and Performance
,
i
M4.1 Hiah Pressure Coolant Irhian 8vstem Drain Vd;= L@lna
g_ . ' a. Inspection Scoce (62703)
>
On June 19, the licensee declared the Unit 3 high pressure coolant injection system
inoperable due to turbine stop valve above seat drain valve leakage. The licensee issued
LER 50-249/97-003 to document this event. The inspectors reviewed the LER, discussed
the LER with the licensee, and examined the valves in the plant.
_ b. Observations and Findinos
l' The in-series steam line drain valves (2301-64 and 65) were leaking. The licensee
~ determinc : that the problem with the "65" valve binding resulted from inadequate
i procedure reviews and, in June of 1997, poor verification of a clearance between two
i valve components. The cause of a low seat loading on the "64" valve could not be
determined. The inspectors noted the following performance issues:
.
- :*
Procedure preparers and reviewers were not thorough in assuring that all the
applicable procedures were revised to address Copes Vulcan revised instructions
for checking that there was clearance between the valve frame and the diaphragm
case.-
Mechanical maintenance workers assembled the "65" valve in June 1997 with the
diaphragm case binding against the valve frame.
Valve test personnel were performing the gap verification, although it was not in
the procedure, and no procedure revision or PlF was initiated.
When test personnel verified clearance, they checked one side, but did not check
- for the clearance on the back side of the valve.
A search of the maintenance work history did not reveal when the seat loading on
the "64" valve was physically changed from 12 psi to 3 psi.
Licensee corrective actions included revising the procedures and discussing the event
with the responsible individuals. The failure to have adequate procedures and to
accomplish activities in accordance with those procedures was a violation of
10 CFR Part 50, Appendix B, Criterion V,finstructions, Procedures and Drawings." This
licensee-identified and corrected violation is being treated as a Non-Cited Violation,
consistent with Section Vll.B.1 of the NRC Enforcement Policy
17
_ _ _ . _ _ _ . - _ . _ _ __
,
,
c. Conclusions
The licensee identified that one valve was binding and not seating property and that
another valve had improper seat loading. This self-disclosing event revealed several
examples of poor performance by maintenance personnel. Although each example
rvpresented a minor departure from regulatory requirements, collectively, the poor
performance by maintenance personnel resulted in inoperable safety-related equipment
which had an impact on operations.
M4.2 (Units 2. 3) Failure of the 2/3B Standby Gas Treatment System
1
'
a. Inspection Scope f62707)
The inspectors observed and r(viewed the licensee's response to the failure of the
2/3 8 train of standby gas treatment. The inspectors observed work in the field and
planning meetings and reviewed the licensee's resolution of problems encountered during
maintenance,
b. Observations and Findinas
Event Chronology
On July 19, during a surveillanes test, an operator noted a loss of indication for the heater
for the 2/3 8 train of standby gas treatment. The 2/3 B train of standby gas treatment
failed the surveillance test and was declared inoperable by the unit supervisor. Because
of the failed surveillance test, the operations staff entered a 7-day dual-unit shutdown
.
limiting condition for operation.
The electrical maintenance supervisor attempted to " call out" electrical maintenance
i
personnel, but was unsuccessful. Therefore, an electrical maintenance supervisor, and
an electrical maintenance department work analyst were assigned to perform
troubleshooting.
!
. Early on July 20, the electrical maintenance supervisor and work analyst took heater
current measurements but incorrectly had the test equipment set on voltage instead of
current. As a result, the licensee ceased troubleshooting and commenced trying to
procure a replacement heater.
On July '1, the staff working on the standby gas treatment system questioned the ,
readings, The licensee re-created how the readings were obtained and the electrical
maintenance work analyst identified the error. The licensee then recommenced
troubleshooting of other standby gas treatment system components in addition to the
heater.
'
The licensee had to again remove the hester to take measurements needed to size a
replacement heater. These rc,eaprements were not taken when the heater was removed
the previous day. This had the potential to increase the amount of time the standby gas
treatment system was out-of-service, but did not because breaker work that was being
conducted in parallel, was not yet finished.
.
18
.
'
.
'
.
The work instructions directed that testing on the 480-VAC,60-amp breaker be performed
using Dresden Electrical Surveillance (DES) 7300-08, " inspection and Maintenance of
480 Volt MCC Molded Case Circuit Breakers," and Dresden Administrative
Procedure 15-07, " Electrical / Instrument Maintenance Troubleshooting Procedure."
Subsequently, the work instructions were modified at the request of a plant engineer to
include a hold-in current test by plant engineering personnel. The test was not part of any
approved procedure. The hold-in current test was not written to be consistent with the
.
guidelines of EPRI report NP-7410-V3R1, " Molded Case Circuit Breaker Application and
Maintenance Guide," Rev.1, produced by the Nuclear Maintenance Applications Center
and was not reviewed and approved. Specifically, the tests were performed with cables
shorter tha' e mmended, resulting in less heat dissipation during the test. Procedure
use by engina . rsonnel during surveillance tests was originally discussed as an
unresolved item (dRI 50-237/97006-05(DRP); 50 249/97006-05(DRP)). Pending further
review, this matter will continue to be unresolved.
The breaker failed the informal hold-in test and part of the tests from DES 7300-08. A
spare breaker tt.at had been purchased in the same lot as the failed breaker was sent
from another site. The spare breaker also faigd the informal hold-in test. On August 5,
the licensee's corporate material engineering group informed the licensee that both
.
breakers failed hold-in current tests that were performed in accord ,nce with NEMA AB-4,
1996, Section 5.7,4.
The licensee had no more spare 60-amp breakers, instead, the licensee used
DAP 11-11, "Setpoint Change Control," to replace the 60 amp breaker with a 70-amp
breaker. Although the 70-smp breaker was also tested in accordance with the informal
hold-in test instead of per the industry-standard test, the licensee concluded that the
hold-in test was conservative relative to the recommended industry test. The standby
gas treatment system was declared operable after a successful 10-hour test.
Licensee Response to Standby Gas Trectment System Problems
As each material condition or human performance problem was encountered, the
licensee documented the problem and entered the problem into the licensee's integrated
reporting process. Additionally, the licensee performed and documented an investigation
(Nuclear Tracking System (NTS) report number 237-200-97-0400) to address the issues.
Based on the investigation, the licensee identified three problem areas: the material
condition of the breakers, the testing of the breakers, and the management of
maintenance during the short LCO. Some of the planned corrective actions and the NTS
numbers used to track the actions are discussed below.
Material Condition of Breakers: The licensee planned to retum the breakers to the
manufacturer for failure analysis (NTS number 237-200-97-04001).
Breaker Testing: Engineering and work analyst personnel were to be trained on the
event, including use of correct equipmer i settings, following procedures, use of approved
versus unreviewed informal tests, and completeness of work package instructions (NTS
numbers 237-200-97-04004 and 237-200-97-04005).
Management of Maintenance: The work control department was assigned to enhance
the procedures which govem coordination of work involving Technical Specification
19
-
t
LCOs. The enhancements were to include use of an interdisciplinary team approach.
Work control staff was also to review the corrective actions for effectiveness by January
of 1998 (NTS numbers 237-200-97-04002,237 200-97-04003, and 237-200-97-04006)
The inspectors noted that the licensee's report incorrectly stated that the incorrect
readings were performed by the electrical maintenance department work analysts.
Inspector interviews revealed that actual equipment setup and readings were done by the
electrical maintenance first-Gne supervisor, and the work analyst only recorded the
readings. The involved first-line supervisor was counseled by the electrical maintenance
department supervisor,
c. Conclusions
.
The initial failure of the standby gas treatment system was discovered during a routine
surveillance test and caused the licensee to be in a 7-day dual-unit shutdown LCO. The
test failure was caused by a failed breaker. However, the licensee had found one other
breaker, purchased at the same time, that failed in the same manner. The root cause for
the breaker failures had not been determined as of the end of this inspection period.
The standby gas treatment system failed during a routine survenlance test. Although the
system was restored to an operable status within the time limits allowed by Technical
Specifications, various personnelissues slowed troubleshooting and restoration to
service efforts. Electrical maintenance personnel could not be called into work and the
electrical maintenance supervisor and work enalyst assigned to work in place of the
electrical maintenance personnel did not take correct readings. As a result, the licensee
did not diagnose the true cause of the standby gas treatment system failure in a timely
manner. The material condition of the standby gas treatment system, combined with
personnel error, impacted plant operations and the availability of safety-related
equipment.
The errors in taking current readings, evaluating the heater, and testing the breakers
resulted in inoperable safety-related equipment for a longer period than necessary. The
material condition of the plant, combined with personnel error, impacted plant operations
and the availability of safety-related equipment.
M8
Miscellaneous Maintenance issues (92902)
M8.1 (Closed) LER 50-249/97003-00: High pressure coolant injection system declared
inoperable due to turbine stop valve above seat drain valves leaking steam. The _
corrective actions addressed revising the procedures and discussing the event with the
responsible individuals. See Paragraph M4.1
20
- - -- . _ = - - . - . . _ - .- .- .- . . .
_
.-
v-
t
Ill. Enaineerina
E4 Engineering Staff Knowledge and Performance
E4.1 250-VDC Batterv Load Criculatkm
a. insoection Scope f37551) -
.
The licensee identified that the present 250-VDC loading calculation did not model the
worst case load profile for the battery. The licensee initiated PIF D1997-06167 to
oocument the concem. The inspectors reviewed the licensee's PlF, the licensee's initial
operability determination (OPD 97-87), and prior history and docketed information with
4
- .
respect to the 250-VDC battery,
b. Observations and Findinos
,
in response to inspectors' questions at another Commonwealth Edison station, the
i licensee reviewed the battery calculation and identified that the loading calculation
, (PMED 898230-01, Revision 11) for the 250 Vdc battery did not model the worst case
load profile. The worst case load profile existed when a unit battery was aligned to the
swing charger and a loss of the unit's diesel generator occ irred concurrent with a
-
i-
loss of-coolant accident. This scenario, which added an additional 75 ampere load on
the battery for a four-hour duration, was not verified by the last Technical Specification
battery load profile surveillance test. The licensee's operability determination concluded
that the station batteries were operable so long as they were aligned to their respective
unit chargers. The licensee provided guidance to the operators to declare the battery
inoperable and en'er the appropriate Technical Specification LCO action statement if the
battery was aligned to the swing charger. The inspectors reviewed the licensee's
operability determination end did not identify any substantive concems with the licensee's
,
,
conclusions.
The NRC previously documented concems with the licensee's 250 Vdc battery sizing
,
calculations in inspection report 50-237/96-201; 50-249/96-201 (Independent Safety
,
inspection (ISI) of Dresden Nuclear Power Station, dated December 24,1996). The ISI
report stated, "The licensee initiated a PlF to document the discovery of the overlap of
.
loads for certain MOVs and the need to revise Calculation PMED-898230-01." The ISI
report also documented inat "The licensee stated that the calcu!ation will be formally
revised to incorporate these issueo." Pending further NRC review of the Updated Final
Safety Analysis Report, the licensee's response to the ISI report, and the revised battery _
calculation, this is an Unresolved item (URI 50-237/97015-06(DRP);
i
50-249/97013-06(DRP)).
c. Conclusions
>
The licensee missed prior opportunities to identify and resolve the 250 Vdc battery
Ioading issue. Enginec ing personnel did not adequately review the battery calculation
when responding to previously identified technical concems.
21
.
h
_ ,
- _- - - . - . - . . - - - - .- . _ .
_
=.
v
E8 Miscellaneous Engineering issues
E8.1 Assessment of Performance Indicators
-
a. Insoection Scope f37551. 92902)
The inspedors reviewed two of the licensee's performance indicators submitted by the
licensee in response to the NRC request for information pursuant to
10 CFR Part 50.54(f).1he inspectors reviewed Engineering Requests (ERs) and
Temoorwy Alterations (TAs).
- b. Observations and Findinas
.
The inspectors reviewed ERs and TAs to assess ineir status and the licensee's tracking
of the issues.
.
Encineerina Reauests
The inspectors determined that the quality and safety assessment (Q&SA)
departmen! had identified that the tracking of open and overdue ERs was based
,
on potentially unreliable data that provided artificially optimistic results, initially,
approximately half of the open ERs had no due date and, as such, were not
included in the overdue indicator. Also, the number of overdue ERs was based
'
on ERs generated after January 1,1997. This practice excluded a large
population of ERs created prior to that date. The Q&SA personnel that
investigated the ER data (QAS number 12-97-15, daad May 13,1997)
documented these issues and generated several PlFs to track the licensee's
resolution of the concems.
. Temporary Afterations
The licensee changed the method for tracking TAs at the beginning of the
inspection period. Initially, the tracking of open TAs was based on the number
entered into the TA log. However, by procedure, non safety-related TAs were
able to be accomplished by a nuclear work request or procedure and not logged. ,
,
At the beginning of the inspection period, the licensee was unable to provide to
the inspectors the number of all "open" TAs in the plant. The tracking of TAs was
potentially inconclusive as an indicator of performance. A Q&SA surveillance
report (QAS 12-97-21, dated July 24,1997) that assessed the TA process stated
4
that * The indicator may not accurately reflect the number of TAs in the field." The
Q&SA surveillance did not, however, identify any examples of " missed" TAs By
the end of the inspection period, the licensee had changed the definition of TAs.
c. Conclusions
i
Initially, the data that the licensee used to track the performance indicators was
potentially unreliable and not always an accurate indicator of performance. The licensee
modified the way the data was tracked to make it a more reliable indicator of
performance,
22
i
l
l
1
- . _ _. . _ . _ ____ _
,
-
?
IV. Plant Support
R1 Radiological Protection and Chemistry (RP&C) Controls
R1.1 General Comments (71750)
'
During routine inspections in radiologically controlled areas, the insoectors assessed the
performance of the licensee. Overall, the licensee's radiation protection staff enforced
the plant's radiological control standards. The licensee continued to uqe personnel
functioning as " greeters" to assure that workers entering the radiologically controlled area
were aware of dose rates and administrative protection requirements. However, as
stated in Section 04.1 and Section R4.1 of this report, the inspectors identified that some
non-licensed operators did not review station survey maps before performing work.
R4 Radiological Protection Staff Knowledge and Performance
R4.1 Control of Survey Mao Updates
a. Inspection Scope (71I50)
.
On August 8,1997, the inspectors identified that two shifts of radiation protection
personnel were unaware of the reason for a waming posted over a survey map for an
area that was accessed every shift by non-licensed operators. The inspectors assessed
control of wamings posted over the survey maps by talking with cognizant radiation
protection personnel,
b. Observations and Findinas
As discussed in Section O4.5 of this report, the inspector identified on August 8,1997,
that the survey map for Unit 2 turbine building elevation 601 had a waming flag hung by
' radiation protection personnelindicating that conditions had changed, but neither the day
shift nor the aftemoon shift of radiation protection personnel was aware of why the flag
had been hung or what conditions had changed.
After the inspector asked the day shift radiation protection personnel about the
601 elevation conditions, the radiation protection personnel re-surveyed the area. At the
start of the aftemoon shift, the inspector saw that the survey map and waming flag were
unchanged. The inspectors identified that the aftemoon shift radiation protection
personnel were also unaware of the reason for the waming flag. The day shift had not -
diccusaed the issue with the aftemoon shift.
t
No formal log existed to document by whom, why, and when a waming flag was hung
over a radiat!on survey map. For the case of the turbine building 601 elevation, the old
survey was performed on July 21,1997. with hydrogen addition in effect and Unit 2 at
801 MWe. Unit 2 was at almost the same power (811 MWe) when the survey was
performed on August 7,1997, after the inspectors' inquiry, and the results of the survey
were similar. Since Unit 2 was shut down on July 28, the radiation protection personnel
initially speculated that the flag was hung due to the shutdown and startup of Unit 2. That
speculation was incorrect Subsequent to the inspectors' inquiry, the radiation protection
personnelidentified that the conditions have-changed flag was hung because a
23
.. _ _ . - _ . ._
,
.
?
technician wanted to assure that the status of hydrogen addition was verified. The .
licensee planned on formalizing the use of the waming flags. !
c. Conclusions
No formal control of the conditions-have-changed flags existed. This resulted in
confusion about why a flag was hung and caused the licensee to re-survey an area.
Information about the radiological conditions of the Unit 2 Turbine Building 601' elevation
was rot passed from day shift to aftemoon shift personnel.
S4 Security and Safeguards Staff Knowledge and Performance
S4.1 Response to Perceived Toxic Gas Cloud
a. Inspection Scope (71750)
On August 6,1997, the control room issued an order over the plant public address (PA)
system that personnel remain indoors until further notice. The inspectors assessed the
reasons for the order and station compliance.
b. Observations and Findinos
At 9:10 a.m., security personnel participating in a pre-brief for a generating station
emergency plan drillleamed of a chemical release. A combination of poor
intradepartmental and interdepartmental communications and acting without validated
information caused security personnel to suspend gate activities and order personnel
indoors. At 9:40 a.m., control room personnel became aware of the situation when a
non-licensed operator was not permitted gate access. At 9:58 a.m., control room
personnel, who did not have any confirming information, made the shelter announcement.
Prior to the shelter order, the site security staff was directly ordering personnel to remain
indoors due to an approaching toxic gas cloud. However, the inspectors noted that one
security person who was providing the order then proceeded outside and smoked a
cigarette. Even after the PA system announcement, the inspectors observed rnany
licensee personnel outside.
In reality, there was no toxic gas approaching the station. A chemical plant in a different
town was performing a planned release of nitrous oxide and nitrogen dioxide and the
release was planned to be within Environmental Protection Agency limits. At 10:18 a.m.,
the operations shift manager determined that no hazard existed and no emergency plan
activation was necessary.
Security personnel performed a prompt investigation to determine how the event
occurred. Based on the investigation, " Nitrous Oxide / Nitrous Dioxide release in Seneca,
.
IL," security management concluded that the cause of the event was the failure of
4
security staff to verify that the ope.ations shift manager was notified of the release.
The emergency preparedness coordinator determined that the prompt investigation was
in " gross error," and documented the errors in Problem Identification
Form #D1997-06110. The same security personnel then issueo Revision 1 of the prompt
24
.
. _ . _ - . _ _ - _ . _ _ . _ _ _ _ . _ . _ _ . _ _ _ _ _ _ _ _ . _ . _ _
Y
l
l
investigation However the inspectors noted that both revisions of the prompt
investigation concentrated only on the role of security personnel in the event, and that the
following concems were not discussed:
Security personnel were not following their own orders to shelter (one smoked a
cigarette outside),
Station personnel ignored orders to stay indoors.
No guidance existed to perform a shelter order without a personnel assembly first,
The control room announcements did not state that the shelter was not a drill (a
generating station emergency plan drill was coincident with the announcements, -
and some personnel may have mistaken the announcement as part of the drill),
Not all site personnel were made aware of the shelter order (Unit 1 personnel
documented this in PlF D1997-06052)
The licensee planned to perform drills similar to the event to improve station personnel
response. .
c. Conclusions
The security personnel performed poorly in verification of information, communications,
and execution of a shelter order for a perceived threat. Many station personnel choose
not to obey an order from the control room to stay indoors, or did not hear the order. The
licensee performed pooriy and in an uncoordinated manner in response to a perceived
threat from a toxic gas release, The investigations into the cause of the event were weak
because they did not address all observed deficiencies,
V. Manaaement Meetinas
X1 Management Meetings
Commissioner Nils Diaz, A. Bill Beach, Region lli Administretor, Geoffrey Grant, Region lll
Director of the Division of Reactor Projects, and others visited the site on August 12,
1997. Commissioner Diaz met with NRC personnel and station management, heard
station management presentations on their improvement plans, and toured parts of the
station. The commissioner then held a press conference and discussed his observations.
X2 Exit Meeting Summary
The inspectors presented the inspection results to members oflicense management a
August 27,1997, following the conclusion of the inspection period. The licensee
acknowledged the findings presented. The inspectors asked the licensee whether any
materials examined during the inspection should be considered proprietary. No
proprietary information was identified.
25
i
, - . . . , . . -
-
.
-t
PARTIAL UST OF PERSONS CONTACTED
Licensee
- T, Bezouska, Site VP Staff
E. Carroll, NRC Coordinator
- F Fink, Business Site Business Manager
- J Homey, Station Manager
- B. Holbrook, Training Manager _.
- S. Kuczynski, Shift operations Supervisor
- J. Lewand, Regulatory Performance Administrator
- B. Osgood, Communications
- S. Perry, Site Vice President
- C. Richards, SQV Audit Supervisor
'T, Riley, Reg Assurance Supervisor
- F. Spangenberg, Reg Assurance Manager
- J. Tietz, Plant Engineering Safety System Supervisor
- L. Weir, Design Engineering Superintendent
- R. Wiggins, Maintenance Staff Superintendent
- B. Zank, Unit 1 Operations Manager
- Denotes present at August 28,1997 exit meeting
INSPECTION PROCEDURES USED
IP 40500:
Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing
Problems
IP 62707: Maintenance Observations
IP 61726: Surveillance Observations
IP 71707: Plant Operations
IP 71750: Plant Support Activities
IP 93702: Prompt Onsite Response to Events at Operating Power Reactors
IP 37551: Onsite Engineering ,
IP 92902: Followup - Maintenance
26
. - - . . _ _ - . .
.
1
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
50-010;237;249/97013-01A VIO Inadequate corrective actions resulted in inadequate control
-
room staffing.
!50-010;237;249/97013-018 VIO inadequate corrective actions resulted in missed Technical
Specifications.50-010;237;249/97013-01C VIO Inadequate corrective actions for ar' ansecured breakers.
,50-010;237;249/97013-02 VIO Failure to review radiological surve', maps.50-010;237;249/97013-03 IFl Hydrogen Analyzer area temperature.50-010;237;249/97013-04 URI Inadequate work on feedwater heaters.
50-249/970i 05 NCV High pressure coolant injection system declared inoperable
due to turbine stop-va!ve above-seat drain-valves leaking
. as a result of a failure to have adequate procedures to
accomplish activities.50-010;50-237;249/97013 URI 250 VDC Battery Loading
Discussed
50-010;237;249/97004-01 IFl Effectiveness of the equipment operator rounds
50-237;249/9700G-05 URI Changes to a surveillance procedure
Closed
50-249/97003-00 LER high pressure coolant injection system declared inoperable
due to turbine stop valve above seat drain valves leaking
steam.
50-249/97013-06 URI high pressure coolant injection system declared inoperable
due to turbine stop-valve above-seat drain-valves leaking
as a result of a failure to have adequate procedures to
accomplish activities
27
l
_ __
__ . . _ _ _ __ _ _ . . . _ _ ____ . _ . _ . _ _ _ .. ._ __
q
?~
LIST OF ACRONYMS USED
ACAD Atmospheric Containment Atmosphere Dilution
BRC Business Review Committee
CCST Contaminated Condensate Storage Tank '
CCSW Containment Cooling Service Water
CFR Code of Federal Regulations
CR Control Room
DAP Dresden Administrative Procedure
DATR Dresden Administrative Technical Requirements
DEOP Dresden Emergency Operating Procedure
DES' Dresden Engineering Surveillance
DGP Dresden General Procedure
>
DlS Dresden instrument Surveillance
DOA- Dresden Operating Abnormal
DOE Department of Energy ,
i DOP Dresden Operations Procedure
DOS Dresden Operations Surveillance
i DTS C=tr Technical Surveillance
i ECCS Emergency Core Cooling System
i EDG Emergency Diesel Generator
EMD Electrical Maintenance Department
EOF Emergency Operations Facility
ERO Emergency Response Organization
FHA Fire Hazard Analysis
FME Foreign Material Exclusion
gpm Gallons Per Minute
GSEP Generating Station Emergency Plan
HPCI High Pressure Coolant injection
HVAC Heating, Ventilation, and Air Conditioning
- IFl inspector Followup Item
IMD instrument Maintenance Department
IRB Issues Review Board
kW Kilowatt
kV Kilovolt
LER Licensee Event Report
LOCA Loss Of Coolant Accident
MMD Mechanical Maintenance Department
MW Megawatt .
.
NCAD Nitrogen Containment Atmosphere Dilution
NSO- Nuclear Station Operator
, NTS Nuclear Tracking System
OSC Operational Support Center
OE Operability Evaluations
[ . PIF Problem Identification Form
psig Pounds Square Inch Gage
_
Poly-Vinyl Chloride
'
RPT Radiation Protection Technician
SE Safety Evaluation
4
28
,
i
. . . . . -. . . _. . - . . _ - . - . . . - . - _ . = . - . . - . . _ - . . _ ~ . . . . . . . . . . . . . . ~ . - . - . . . . .
,
-
o
. J. e.
- -
"
SQV Site Quality Verification'
UFSAR Updated Final Safety Analysis Report
URI- Unresolved item
2-
-
4
i
.
4
4
i
!
!
$
k *
i-
i
.
4
.i
l!
,
29
_.. . - . ._