ML20202G307
| ML20202G307 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 02/13/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20202G241 | List: |
| References | |
| 50-254-97-22, 50-265-97-22, NUDOCS 9802200105 | |
| Download: ML20202G307 (41) | |
See also: IR 05000254/1997022
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U. S. NUCLEAR REGULATORY COMMISSION
REGIONlli
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' Docket Nos: 50-254; 50-265
Report Nos: 50 254/97022(DRS); 50 265/97022(DRS)-
. Lloonsee: . Commonwealth Edison Company
Facility: Quad 9ities Nuclear Power Station -
Location: 22710 206th Avenue North '
Cordova,IL. 61242
Dates: October 27 through November 21,1997
Inspectors: ' D. St..',or, Team Leader, Rill
. J.- Noisier, Reactor Engineer, Rlli
J. Guzman, Reactor Engineer, Rlll'
- D. Muller, Reactor Engineer, Rill'-
R. Pulsifer, Licensing Project Manager, NRR
J. Mallanda, NRC Contractor
S. Khabir, NRC Contractor :
Approved by: J. Jacobson, Chief, Lead Engineers Branch
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O k o M 54 '
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EXECUTIVE SUMMARY
Quad Cities Nuclear Power Station
NRC Inspection Report 50 254/g7022(DRS); 50 265/g7022(DRS)
This inspection assessed the HPCI system operational performance. In addition, the inspection
assessed the effectiveness of licensee controls in identifying and resolving problems. The
inspectors concluded that the HPCI system was capable of performing its safety function and
that, with exceptions, licensee controls were adequately identifying and resolving problems.
Qoerations
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During walkdowns on the Unit 1 ar . 2 High Pressure Coolant injection (HPCI) systems,
minor problems with material condition were identified. These minor problems did not
affect the operability of either units' HPCI systems. Minor procedure errors were also
l Identified with both the Unit 1 and 2 HPCI system checklists used to locally verify valve
positions. (Section 02.1)
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Some problems with HPCI system procedures were identified. One procedure,
concerned with local operation of the HPCI system, was inadequate. This was
considered a violation of NRC requirements. (Section 03.1)
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The level of knowledge of two operators intervi9wed concerning the HPCI system was
Good. (Secilon 04.1)
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Overall quality of the HPCI lesson plan used for training licensed operators was good.
Two minor nontechnical problems with the lesson plan were identified. HPCI system
training observed during this inspection was technically accurate and adequate overall.
(Section 05.1 and 05.2)
Mal 0lenaDCQ
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Licenst., maintenance procedures were technically adequate, sufficient to perform the
required maintenance and inspection tasks and had the necessary provisions to identify
and evaluate deficiencies. (Section M3.1)
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The performance by an instrument technician during an observed surveillance was
good. The technician precisely followed the procedure and demonstrated a good level
of skillin the use of the test equipment involved. (Section M4.1)
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The mainteriance training program was adaquate to assure qualified maintenance
technicians. The training facilities were very good and were considered a strength.
(Section MS.1) l
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Maintenance activities were well controlled. The assignment of work week managers to I
coordinate activities was considered a strength. (Section MB.V '
Engineering
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The inspectors noted that important calculations that form part of the Quad Cities HPCI
design basis were not easily retrievable, or did not exist. For example, at the inspecdon
onset, the licensee lacked a calculation to ensure that the HPCI design basis flow of
5000 gpm could be delivered against reactor pressure to support the acceptance criteria
in the Technical Specification surveillance procedure. The inspectors also noted that
the design basis for varicus safety related systems was not clearly established.
However, the licensee had initiated actions such as the Design Basis initiative UFSAR
review and was plannirig generation of approximately 15 missing analyses. (Section
E1.1)
+ While overall, the HPCI system mechanical calculations reviewed were found to be
acceptable, weaknesses were noted with nonconservative assumptions in an Initial
" white paper" analysis and with not consistently accounting for instrument inaccuracles.
(Section E1.1)
- The HPCI system electrical calculations reviewed were generally acceptable, but
numerous examples of inattention to detall and weaknesses in the design verification
review process affected the quality of the analyses. Also, for some of the calculations
reviewed, the tracking of assumptions and of results which may impact other
calculations or procedures was weak. (Section E1.2)
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The sample of HPCI system modifications reviewed was acceptable; however, a
violation of design control was identified. Other minor issues raised by the inspectors
were satisfactorily addressed by the licensee.
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Overall, the inservice testing and Technical Specification (TS) surveillance testing
specifically related to the HPCI system were satisfactory. Based on a recent trend of TS
surveillance noncompliance and potential programmatic testing inadequacies, the
licensee had undertaken a root cause investigation to evaluate the trend and
recommend corrective actions to prevent recurrence. The effectiveness of these actions
could not yet be determined. (Section E2.2)
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The inspectors concluded that the HPCI toom cooler was being adequately cicaned and
inspected pursuant to GL 89-13 commitments. Flow md differential pressure were
trended and monitored for degradation and cleaning v. 1 scheduled on a regular basis.
(Section E2.3)
+ Contrary to procedural requirements, the 50.59 screenings for two temporary alterations
failed to evaluate the physicalinsta!!ation of allinstrumentation installed by the
alteration. (Section E2.4)
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- While the 50.59 safety evaluations reviewed were adequate with supportable
conclusions, weaknesses were identified with the overall 50.59 program. These
weaknesses included poorly written safety evaluations, incomplete summary report
submittals to the NRC, difficult to retrieve screenings, and incomplete corrective actions
to identified deficiencies. The incomplete summary report submittals were considered a
violation of procedural requirements. The Off Site review group, however, was
providing good assessments and comments. (Section E3.1)
+ Inaccurate figures and text in the UFSAR were Identified but the Inspectors also noted
ongoing licensee efforts to improve UFSAR accuracy such as line by line reviews of the
UFSAR design information and an initiative to ensure all facility changes had been
incorporated into the UFSAR. (Section 3.2)
+ Ouad Cities design basis information weaknesses were also exhibited with numerous
errors identified with the HPCI system design basis document (DBD). However, the
licensee was aware of the DBD shortcomings and had designated the DBDs as
"information only" pending completion of a validation process. (Section E3.3)
+ The inspectors reviewed commitments from Comed's March 28,1997, response to the
NRC's request for information pursuant to 10 CFR 50.54(f). The inspectors concluded
that eighteen 10 CFR 50.54(f) commitments were closed. (Section E7.1)
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The inspectors reviewed the actions taken by Quad Cities staff for eight Systematic
Evaluation Program (SEP) topics and concluded that the actions taken were sufficient
for niosure of these items. NRC review of nine remaining SEP ltems was ongoing.
(Section E8.2)
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Report Detalla
1. Ooerations l
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02 Operational Status of Facilities and Equipment
02.1 High Pressure Coolant inlaction (HPCI) Sys'.em Walkdown
a. insoection Semne ,
The inspectors conducted walkdowns of the Unit 1 and 2 HPCI systems, which included '
- observations concerning the status of major HPCI components, piping, valves, and ,
associated electrical switchgear. During the walkdowns, the following procedures were !
checked for adequacy:
. QOM 123001, " Unit 1 HPCI Checkoff List," Revision 4
+- QOM 2 23001, * Unit 2 HPCI Checkoff List," Revision 7
. QOM 12300-02,"HPCI System Fuse and Breaker Checklist," Revision 2
in addition, the above listed checkoff sheets and the as found statue of both units' HPCI
systems were compared to mechanical drawings M 46 for Unit i HPCI and M 87 for
Unit 2 HPCl.
b. Observatic,ns and Findings
Checkoff lists were used, in part, by the licensee to ensure that the HPCI system was
operable. Numerous mirar deficiencies were identified with checklist QOM 123001.
Checklist QOM 12300-1 incorrectly referred to valves 1239916 and 17 as HPCI
Booster pum' dMcharge vent valves, As determined by drawing M 46 and the label
tags attached so the valves, valves 1239916 and 17 were, in fact, HPCI Main pump
discharge vent valves. Checklist QOM 12300-1 incorrectly checked the HPCI steam
line drain line steam trap inlet valve 1230154. As determined by drawing M-46 and the
system walkdown, valve 12301 54 does not exist. It was later confirmed that this valve
had been removed by a system modification in 1994. Checklist QOM 12300-1 did not
indicate that the HPCI steam line drain line steam trap outlet valve 1230155, was to be
locked in position. As determined by drawing M-46 and the system walkdown, valve 1-
. 230155 was locked in position. Checklist QOM 12300-1 did not indicate that the HPCI
cooling water pump discharge valve 1230181, was to be locked in position. As
determined by drawing M 46 and the system walkdown, valve 1230181 was locked in
position.
With the exception of the 12399-16 and 17 valves nomenclature discrepancies, all of
the other discrr :ncies had been previously identified by the licensee and had been
incorporated into a procedure revision. At the time of this inspection, this procedure
revision to QOM 1-2300-1 was awaiting final reviews prior to implementation. Licensee
staff initiated action to conect the NRC identified nomenclature dlscrepancies
associated with the 1239916 and 17 valves.
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An additional discrepancy was also noted common to both units' checklists (OOM
123001 and 2 2300-1). Neither of these checklists required a check on keep fillline
test valves 1(2)-2399 78 and 79. If these valves were inadvertently in the wrong
position, there would have been the potential that keep fill would not have been
established, as required, if the operators manually selected the torus as the HPCI
suction source. These valves, however, were checked in the proper position quarterly
as part of surveillance procedure CCOS 2300 22," Quarterly HPCI Keep Fill Supply
Check Valve Closure Test," Revision 1. Control of these valves' positions was thus
never in doubt; however, the licensee agreed that these valves should be added to the
checklists,
During the system walkdowns, minor problems with material conditions were also noted.
The Unit i HPCI system had a small amount of oil on the floor in the vicinity of the oil
reservoir. Two valves for the Unit 1 HPCI system had no valve label tags, and one
valve for the Unit 2 HPCI system had no label tag. Two gages in the Unit i HPCI oil
system were not properly labeled. The licensee initiated action to correct these issues,
c. Conclusions
There were no issues identified during the walkdowns that affected either units' HPCI
system operability. Minor errors with the system checklists were identified (some of
these errors were previously identified by the licensee). Minor problems with material
condition were identified.
03 Operations Procedures and Documentation
03.1 HPCl System Procedures Reviews
a. Insoection Scogg
The inspectors reviewed the HPCI systems' (Unit i and 2) normal, abnormal,
surveillance, and annunciator response procedures for edequacy. In addition, several of
the above procedures were evaluated using real-time control room simulator exercises.
The inspectors also reviewed the sections of the Technical Specifications (TS)
corresponding to the HPCI system.
b. ObseIyfitions and Findings
Various procedures were exercised and system responses were observed using the
control room simulator. The exercises conducted included: auto initiation of the HPCI
system, HPCI turbine trip, auto isolation of the HPCI system, testing various HPCI valvo
interlocks, performing the quarterly HPCI full-flow surveillance, and a HPCI system oil
leak with the HPCI system injecting into the reactor vessel. There was no difficulty
encountered with any of the procedures used, and the HPCI system responded as
indicated in the procedures.
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During the procedure reviews, QCOP 2300-08, THPCI Local Manual Opera:lon,"
Revision 10, was identified by the inspectors as inadequate. Step F.2.b.(5)(b) of OCOP
2300-08, used to rapidly open HPCI steam isolation valve MO 1(2) 2301-4, instructed
operators b place jumpers between terminals FF 9 and FF 29 at incorrect panels. The
panel designations per this step were 90139 3E for Unit 1 and 902 39 iW for Unit 2. If
an operator had to rapidly open valve MO 1(2) 2301-4 per step F.2.b.(5)(b), he would
have:
(a) for Unit 1, realized that panel 90139 3E does not exist, or,
(b) for Unit 2 proceeded to panel 902 39-1W and realized that the jumper could not
be installed, since terminals FF 9 and FF 29 do not exist at panel 902 391W.
l In either case, the jumper would not have been installed. The correct panels would
have to be determined and authorization would have to be granted to proceed contrary
to the procedure as written. The licensee determined by conducting panel walkdowns
and reviews of the electrical prints that the correct panels for step F.2.b.(5)(b) were
90133 3E for Unit i and 902-331W for Unit 2.
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In addition QCOP 2300-08 Attachment A. "HPCI Local Manual Operation Restoration
Verification Sheet," has the operators verify a jumper removed for vacuum breaker
i isolation valve 2 2399-40, at panel 902 39 2W. The correct panel for this jumper was
902 391E. Additionally, Attachment A has the operators verify a jumper removed for
steam isolation valve MO 2 2301-5, at panel 902 39-2W. The correct panel for this
jumper was 902 39-1E. The licerisee initiated action via a PlF to correct these panel
deficiencies with a revision to procedure QCOP 2300-08.
10 CFR 50, Appendix B Critorion V," Instructions, Procedures, and Drawings," requires,
l In part, that act;vities affecting quality be prescribed by cocumented procedures and
shall be accomplished in accordance with these procedures. The inaccurate procedure
steps in QCOP 2300 08 which would not allow for rapidly opening the HPCI steam
isolation valve were considered a violation of 10 CFR, Appendix B. Criterion V (VIO
50 254/265/97022 01(DRS)).
During the review of HPCI system TS,it was discovered that all of the reactor water
level instrument setpoints (not just for the HPCI system) have different values than
those found in the normal, abnormal, surveillance, and annunciator response
procedures. This was due to a different choice of reference points for reactor water
level between TS and procedures, TS setpoints were referenced to the level in inches
above the top of active fuel (TAF). Setpcints in the procedures were referenced to TAF
being equivalent to 143 inches, which corresponds to the set up of the control room
water level indicators. For example, TS lists the HPCI low reactor water level auto
initiation setpoint as 84 inches above TAF, whereas the procedures list this setpoint as
59 inches. These setpoints are equivalent since TAF was equal to -143 inches
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Indicated water level. While this setpoint value difference was not considered a
significant issue and operators interviewed were aware of the distinction, the inspectors
coricluded it unnecessarily added confusion concoming the level setpoints,
c. Conclusions
While there was no difficulty encountered with any of the procedures exercised nor with
HPCI system response using the control room simulator, some problems with HPCI
system procedures were identl'ied. inspectors identified a violation of procedural
requirements in that the HPCI local manual operation procedure had three sections
where the procedure would have directed operators to incorrect panels, and therefore
this procedure could not have been performed as written.
04 Operator Knowledge sad Performance
04.1 ljPCI System Knowledge '
a. . Inspection Scone
The inspectors interviewed a licensed senior reactor operator and a licensed reactor
operator conceming the HPCI system.
b. Observations and Findings
The licensed operators interviewed were normally assigned positions on an operating
shift crew The licensed senior reactor operator typically was assigned the position of
shift supervisor and the licensed reactor operator typically was assigned a position at
the controls of one of the units. Without the use of reference material (except for
mechanical prints to explain system response), the licensed operators were asked the
following questions conceming the HPCI system:
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What is the purpose of the HPCI system?
+ When is the HPCI system required to be operable?
+ What measures are taken to ensure that HPCI la operable?
+ Describe HPCI OGA support proceoures.
+ Describe the HPCI auto initiation sequence. What setpoints cause an auto
initiation?
+ Describe the HPCI auto isolation sequence. What setpoints cause an auto
isolation?
Some additional questions were also asked conceming valve interlocks and recovery
from an auto isolation. Both operators interviewed provided accurate and detailed
answers to the above questions.
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c, . Conclusions
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The level cf knowledge of the two operators interviewed concerning the HPCI system
was good Both operators had no difficulty in answering the above questions
concerning the HPCI system.
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05 Operator Training and Qualification
05.1 HPCl System Lesson Plan
a. InspaciloDJacope
The inspectors reviewed the lesson plan used to train licensed operators on the HPCI
system. The inspectors also reviewed a sampling of modifications that have been
performed on the HPCI system, to check on their inclusion in operator training.
b. Observations and Findinas
Based on a comparison between the lesson plan and various other references
(procedures, TS, Updated Final Safety Analysis Report (UFSAR), mechanical and
electrical prints), no technical errors in the lesson plan were discovered, in addition, the j
lesson plan appeared to correctly incorporate modifications that have occurred to the ,
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HPCI syt, tem.
Hewover, two minor nontechnical problems were discovered with the lesson plan. The
first problem was that setpoints were not consistently presented in the lesson plan. For
example, in one location of the lesson plan, the actual setpoint for the reactor low
pressure isolation signal was listed (125 psig). In another location the TS required
setpoint for this function was listed (100 psig). The lesson plan by itself was unclear
about this setpoint, and the inspectors had to obtain clarification from the licensee. The
second minor problem discovered was that the section on HPCI auto initiation was
incomplete, in this section of the lesson plan, there was no mention of the repositioning
of three air-operated valves. However, the lesson plan previously mentioned these
effects when each valve was individually discussed. The auto initiation section,
therefore, would have been more complete if these three air-operated valves had been
included in the discussion within this section.
c. Conclusions
Overall quality of the HPCIlesson plan was good. The lesson plan was fairly extensive,
detailed, and accurato. Two minor nontechnical problems were discovered: the
treatment of setpoints and the completeness of the auto initiation section. One
additional measure of the effectiveness of the lesson plan was the previously discussed
(Section 04.1) good level of knowledge displayed by the operators conceming the HPCI
system.
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05.2 Trainina Effectiveness and Periodicity
a. Insoection Sggt
The inspectors observed the conduct of a HPCI system training presentatiori given to a
group of non-licensed operators. Additionally, the inspectors reviewed training
documents which Indicated the periodicity and scope of training which had been given--
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i b. Observations and Findinos
A training department instructor presented a lecture on the HPCI system to a group of
approximately 15 non licensed operators. The training session lasted approximately two
hours. The instructor followed the non-licensed operator lesson plan, which was similar
to the lesson plan previously discussed in section 05.1. The instructor presented
essentially factualinformation during the murse of the lecture. In some cases, the level
'of detail of the presentation was low. Hr. .sver, this appeared consistent with the fact
that this was a part of the licensee's continuing re-tralning program.
The HPCI system was trained on as part of the licensee's continuing re. -
training /requalification program. Additional training on the HPCI system had also
occurred (on a non-scheduled basis) when procedure changes and/or modifications to
} the HPCI system have occurred and during portions of control room simulator exercises,
c. Conclusions
The observed HPCI system training, presented by a training department instructor to a
group of non licensed operators, was technically accurate and adequate overall. The
periodicity of training on the HPCI system also appeared to be adequate. One measure
of the effectiveness of training on the HPCI system was the previously discussed
- (Section 04,1) good level of knowledge displayed by the operators conceming the HPCI
system.
II. Maintenance
M3 - Maintenance Procedures and Documentation
M3.1 Bay [gw of Maintenance Procedutta
a. - Insoection Scone
The inspectors reviewed selected maintenance and surveillance procedures for the
systems selected for inspection. The reviews were for technical adequacy and
satisfaction of vendor requirements and recommendations.
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b. Observations and Findinas ~
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The licensee procedures reviewed during this inspection appeared to be technically
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adequate to perform the specific maintenance task and provide for the identification and i
' evaluation of equipment and work deficiencies. Procedures included in the review were
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current. Modifications to equipment or systems had been included in the procedures !
reviewed. .
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Maintenance and surveillance procedure content was compared against vendor's
' recommendations for the HPCI components. The procedures appeared to satisfy the
vendor's maintenance and inspection requirements.
- The licensee recently initiated the practice of conducting user reviews of maintenance ,
procedures. The inspector reviewed two of the procedures that had been reviewed and -
rearranged by the technicians who would be using the procedures and found the i
procedures were much Pnproved when compared to the original procedures.
c. Conclusions -
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The inspectors concluded that the licensee's procedures were technically adequate,
sufficient to perform the required maintenance and Inspection tasks and had the
- necessary provisions to identify and evaluate deficiencies.
M4.1 Performance of a HPCI System Surveillance
a. IDaggetion Scope
- The inspectors observed the performance of a non-licensed instrument technician
during the conduct of HPCI system surveillance, QCIS 2300-02, "HPCI Reactor Low
Pressure Analog Trip System Calibration and Functional Test," Revision 4. .
b. Observations and Findinas
QCIS 2300 02 was conducted on the Unit 2 HPCI system. The Unit 2 HPCI system was l
not required to be operable during the conduct of this survolliance, due e Unit 2 being in
cold shutdown. This surveillance consisted of the instrument technician hiserting
various test currents into the analog trip unit associated with each channel of the HPCI
reactor low pressure isolation function. The status of the trip unit and associated relays
were then observed to datermine if the trip and reset setpoints of each channel were -
within tolerance. Phone communications were utilized between the technician and
those personnel who observed the associated relays. During the performance of this
surveillance, the inspectors observed the following: ,
. The test equipment used appeared to be within calibration.
- The procedure was precisely followed.
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The technician demonstrated a strong familiarity with the procadure and use of
the test equipment.
The technician understood the effects that procedure steps had on the system.
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The as found conditions of all of the trip units'setpoints were within tolerance. ,
No further calibrations were required. I
c. Conclusions
The overall performance of this surveillance was good. The technician had no difficulty
in following the procedure. The technician demonstrated a good understanding of what
each step in the procedule accomplished, in addition, the technician possessed a good
level of skillin the use of the test equipment.
M5 Malntenance Training and Qualification
M5.1 Maintenance Trainina and Qualification
a, laspection Scope
The team interviewed supervisors, workers and training staff. The team also reviewed
training records and toured the Quad Cities maintenance training facilities.
b. Observations and Findings
The team reviewed training records and interviewed training department personnel
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relative to mtsintenance training for department personnel. The licensee had a
comprehensive training program for the Quad Cities maintenance staff both for Initial
- training and qualification and for continuing training to malt"
- In profielency.
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Th sam toured the Quad Cities maintenance training facility. The facility had
adt,quate space for separate facilities for conducting simultaneous training of
mechanical, electrical and instrument and control technicians. Each facility was well
equipped with training aids that either simulated or were identical to plant systems or
components.
The maintenance shope and supervisors were provided with a matrix t! ot detailed the
training and task qualification of each technician. These matrices were used to assure
that the Individual assigned to perform a task was trained in the performance of that
task,
c. Conclusions
The inspectors concluded that the licensee's maintenance training program was
adequate to assure qualified maintenance technicians. The training facilities were very
good and considered a strength.
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M8 Miscellaneous Maintenance issues
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M8.1 Maintenance Work Control
a, lasgention Ecoce i
The team reviewed work planning orocedures, work requests, planning processes and
interviewed work planners.
b. Observations and Findinga
Quad Cities had a well coordinated and organized control system for maintenance
activities. A work analyst was assigned to each maintenance team.
The NRC team observed a work week planning meeting for work scheduled two weeks
in advance. Work week managers were essigned for each week. Five of these
managers rotate through a three month maintenance cycle. Each is responsible for
coordinating and expediting work activities during their assigned weeks.
Review of work packages and discussion with preparers indicated that sufficient detail
was included in the packages to enable the technician to perform the required tasks.
c. Conclusions
The team concluded that maintenance activities were well controlled. The assignment
of work week managers to coordinate activities was considered a strength.
Ill. Engineerina
E1 Conduct of Engineering
Ei,1 Mechanical Design Calculations
a. insoection Scoon
The inspectors reviewed mechanical calculations to determine if the purpose, scope,
assumptions, analysis methodology, acceptance criteria, and conclusions were
acceptable, in addition, numerous supporting documents were reviewed as reference '
in the calculations. This included design basis calculations for HPCI system thermal-
hydraulic, piping stress, and equipmont sizing,
b. ' Observations and Findings
The inspectors noted that important design basis parameter calculations for the HPCI
systems were not readily retrievable or simply not available. For example, a calculation
to ensure that the HPCI design basis flow of 5000 gpm could be delivered at varying
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reactor pressures was not available. Subsequent interviews with the licensee revealed
that Quad Cities staff had identified this design basis and configuration control concern
and were in the process of scheduling and generating approximately 15 new analyses
and calculations in various systems as part of a corporate wide effort.
During the inspection, minor concerns identified by the inspectors were addressed via '
PIFs, more significant technical or design basis concerns are listed below:
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Calculation ODC 2300 M-0486,' Verification of HPCI Pump Discharge Flow to
Reactor,' Revision O.
This calculation was not available prior to the inspection and was generated end
completed during the inspection. The inspectors were concerned that the
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information from this anclysis, which is needed to appropriately determine test '
acceptance criteria, was unavailable in the Quad Olties design basis. However, .
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the calculation successfully confirmed that after accounting for line losses, the l
TS Surveillance test capacity and discharge pressure acceptance criteria, for the
HPCI pump, assured operability by delivering greater than 5000 gpm against '
reactor pressure as high as 1120 psig. The calculation determined that rated
flow of 5,036 gpm can be delivered at the HPCI pump discharge pressure of
i 1189.4 psig and corresponding reactor pressure of 1120 psig, exceeding the
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design basis assumptions. The calculation provided adequate assurance of
HPCI system operability and its capability to deliver rated flow.
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Calculation ODC 2300-M-0489, " Air Entraining Vortices for the HPCI Pumps,"
Revision O.
As part of NRC's initial Inspection information requests and subsequent
l questions on the contaminated condensate storage tanks (CCST) usable
1 volume, the licensee generated this calculation to ensure that air would not be
introduced into the HPCI pump. The engineering staff initially stated in a white
paper analysis that air entraining vortices would D01 be predicted because the
mlnlmum submergence of one pipe diameter (11.7 In.) existed at Quad Cities
CCST. However, in response to inspector and licensee identified concerns with
the white paper's nonconservatism, engineering prepared a formal calculation,
QDC 2300-M-0489 Rev. O, to comprehensively address vortexing and minimum
usable water issues. The calculation addressed: (1) the usable volume
available to the HPCI system and the Reactor Core Isolation Cooling (RCIC)
system from the CCST, (2) adequacy of water supply to preclude vortexing in the
CCST and determine the ability of HPCI and RCIC suctions to switch over to the
torus. The maximum acceptable time for the switchover to prevent entrained air
from reaching the pumps was 115.4 sec. While the inspectors determined the
calculation to be acceptable, the inspectors noted that the elevation of the low-
level switch activation setpoint was determined to be 40 inches above the bottom
of the tank and did not include any inaccuracy. Additionally, the CCST low level
switches 1.S 1(2)-2350A/B/C/D have 3 9/16 inch (switch actuation rising level)
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and 4 7/16 inch (switch actuation falling level) setpoints with +/ 1/4 inch
accuracles which were also not accounted for in the calculation. ;
Calculation QDC 2300 M-0189, *NPSH Available For the HPCI Pumps," Rev. O
and Rev.1 (PIFs Q1997-03985 & 04087).
The inspectors noted that under LOCA conditions, revision 0 of this calculation
did not include friction losses due to Installation of new torus strainers, The
licensee revlsed the calculation and the inspectors determined that sufficient
NPSH Available margin existed.
c. Conclusions
The insp)ctors noted that important calculations that form part of the Quad Cities HPCI
design basis were not enslly retrievable, or did not exist. In response to NRC questions,
and to clarlfy HPCI operability requirements, the licensee generated various calculations
while the team was onsite. While overall, the HPCI system mechanical calculations
reviewed were found to be acceptable. Weaknesses were noted with non conservative
assumptions In an Initial * white paper" analysis and in not consistently accounting for
Instrument inaccuracles.
l
E1.2 E[ect:.:al Deslan Calculatiqat
a. Insoection Scone
The inspectors reviewed the following electrical calculations to determine if the purpose,
scope, assumptions, analysis methodology, acceptance criteria, and conclusions were
, acceptable, in addition, numerous supporting documents were reviewed as referenced
in the calculations,
b. Observations and Findings
.
Calculation PMED 891377 01," Development Of a Duty Cycle Based On a More
Conservative Application Of Colncident Starting Currents For The 250Vdc
Battery Sizing," Rev 11.
The purpose of this calculation was to assess the impact on battery sizing of a
more conservative assumption ccncerning coincidence of starting currents for
loads actuated by separate relays or devices. In general, electrical Calculation
PMED 891377-01, Rev.11, was adequate but attention to detail affected the
quality of the calculation and weaknesses in the design verification review
process were noted. Specifically:
- The control power for valves 1/ 2 2399 41 was not modeled in the
calculation. This load was estimated to be less than an ampere and did
not significantly affect the results. PIF 4329 was generated to address
this issue.
15
_ _ _ _ _ _-_ ______-_- - -
,
t
.
Revision 11 of the calculation was performed to evaluate the abnormal
lineup condition when the 250Vdc buses from Unit 2 were connected to
Unit i because the battery was being tested on Unit 2. The conclusions
stated that a review of the previous test results should be performed by
the station to ensure that the battery capacity was close to 100% prior to
configuring the plant in the abnormal configuration. The licensee
generated PlFs 4143 and 4354 to correct plant procedures that do not
contain any prerequisites to require a review of the previous capacity
tests,
i
.
The Inspectors determined that the cable lengths utilized in Attachment 6
of Calculation 004 E-043 were nonconservative for the application due to
the 1.2 multiplier of each estimated cable length. Attachment 7,
Calculation 004 E-044, for Unit 1 was also impacted. The licensee
preliminarily determined and the inspectors concurref " at using the
correct cable lengths will have a minor impact on the battery loading.
(PlF 4302)
l
+
Calculation 8250 50101, "250 Vdc System Short Circuit Current," Revision 1
and Revision 2.
The purpose of these calculations was to determine the Quad Cities Station
250Vdc system available short circuit current at each system bus for use in
studying system coordination and for con.parison with overcurrent device
interrupting ratings.
- Revision 2 incorporated the overload heater resistance in the circuits with
[ combination starters and reduced the short circuit currents below the
breaker ratings that had been identified. This reduction was
accomplished F 'ertain minimum resistance overload heaters were
installeo. Nine Dreakers had not been identified during walkdowns and
the calculation recommended obtaining the ratings of these breakers.
The inspector questioned the followup of the nine unidentified breakers in
the calculation and the licensee stated that the followup had been-
performed in accordance with Letter 209332,250Vdc Circuit Breaker
identification and Interrupting Capacity, but the calculation had not been
- revised. PlF 4453 was generated to address this issue. The licensee
indicated that the breakers or their use in combination starters had
sufficient interrupting capacity to interrupt the maximum short circuit
currents without damage.
- The calculation did not include a short circuit analysis for the abnormal
alignment identified in Calculation PMED 891377-01. PlF 4399 was
written to address this issue.
.
The inspectors identified discrepancies lo horsepower ratings for the
HPCI Turbine Gland Steam Condenter Exhauster and the Drain Pump
16
. _ -
4
.
i
between the calculation and applicable drawings. The discrepancies did
not appear to have a significant effect on the calculation results. PlF
4366 was written to address this issue.
.- Study SL-4501, 'Overcurrent Protective Device Coordination Study,' Volume 4.
The purpose of this study was to respond to Generic Letter 8815. Technical
issue 5, which discussed the necessity to ensure that circuit breakers and
protective devices within the onsite electrical distribution system were properly
coordinated. The licensee identified that the UFSAR has no specific statement
for 250Vdc overcurrent protection devios coordination. The Quad Cities
Appendix R Fire Protection Program was based primarily on the procedural
- tripping of a particular list of associated circuits for a fire in a particuler fire zone
and, therefore, did not take orodit for protective devios coordination. The study
also noted that the current IEEE Standard 9461985 requires protective device
coordination. The study concluded that certain breakers without starters should
be replaced and somo upstream breakers required replacement with breakers
that have long-time and short time (no Instantaneous) trip units or time delay
fuses or bus rearrangement if system coordination was to be achieved. The
licensee statea tha' r * allure modes and offects analysis dated September 14,
1990, and an evaluation of operability determination checklist, ENC-QE 40.1,
Rev 0, had been completed to document justification for continued operation.
However, outstanding coordination issues were pending resolution. Resolution -
of these protective device coordination issues was considered an unresolved
item pending further review of licensee actions (URI 50 254/265/97022-02),
e
Calculation 8913 7719-1, *250Vdc Battery Interconnecting Jumper Ampacity,"
Revision 0,
The existing Unit 1250Vdc battery was being increased from 116 oells to 120
l
- cells. The purpose of this calculation we9 to determine the jumper installation -
ampacity for the connection to the additional four cells and a comparison of the -
calculated ampacity to the battery duty cycle loadin0. Whlie overah the
calculation was adequate, the inspectors noted that at the time that the
calculation was performed the maximum discharge current in the load profile was
'1001 amperes for the first one minute load. <The continuous duty rating of the
jumper cab's was calculated to be 980 amperes. Calculation PMED 891377 01, i
Rev 11, shows 1276.4 amperes for the ce minute load for the abnormal
alignment case. Even though this overload condition of the jumper cables was of
short duration, calculation 8913 7719-1 had not been updated to address this
higher current. PlF 4397 was generated to address this issue.
.-
Calculation QDC-8350-E-0074, ' Input and Output Cable and Circuit Breaker
Sizing for the #2 250Vdc Battery Charger,' Revision 1.
The purpose of this calculation was to size the AC supply and de output cables
for the Unit 2 250Vdc battery charger. The settings of the input and output
,
_ . _ _ _ . _ m _ __ - - - - - - - - - - - - -
.
__
.
I
breakers were also determined. Overall, the calculation was considered
acceptable, but the inspectors noted that the recommendation that the charger
output be limited to 300 amperes DC (120% of rated output) had not been
incorporated into plad procedures,
c. Conclu11ons
The HPCI system electrical calculations reviewed were generally found adequate but
numerous examples of inattention to detMI and weaknesses in the design verification
review process affected the quality of the analyses. Also, for some of the calculations
'
reviewed, the tracking of assumptions and of calculation results which may impact other
calculations or procedures was weak.
E2 Engineering Support of Facilities and Equipment
E2.1 Modifications
a, laspection Scopa
The team reviewed mechanical modification packages to ensure the licensee's
effectiveness in proper implementation of design basis documentation. The review
included the modificaL9 recommendations,10 CFR 50.59, UFSAR, and Technical
Specification changes. r 0st modification test requirements were reviewed to determine
8
if testing was sufficient to ensure that the equipment would perform its intended function,
and determine if plant procedures were properly updated to reflect the modification,
b. Qhservation and Findings
The sample of HPCI modifications reviewed was found to be acceptable with the
exception of the following issues:
- In response to the SOPl, the licensee identified that the HPCI steamline high
flow 310-9 second time delay to initiate closure of the HPCI AC-inboard-
isolation valve did not factor in the additional 10 second EDG loading time for a
loss of offsite power event concurrent with a HPCI steamline break. PlF 4344
was initiated on November 12,1997, to track this concern. The overall HPCI -
DC-outboard steamline isolation valve closure time was unaffected. The AC and
DC valve close logics were changed by modification M04-1(2)-91-013, * Modify
Break Detection Logic to Prevent Spurious isolation of HPCl"in response to
NUREG-0737, item II.K.3.15. The licensee initiated an issue Screening Form,
dated November 14,1997, to determine if an operability concern existed. The
longest overs AC inboard isolation time was determined to be 69 seconds. The
licensee concluded that the increased AC isolation time was within the time
assumed in the Updated Final Safety Analysis Report (72 seconds including
valve closure time). However, failure of the modification package to address the
additional AC Inboard isolation valve closure time during a loss of offsite power
event was considered a violation of 10 CFR 50, Apoendix B, Criteria Ill, * Design
18
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.
!
Control." The inspectors reviewed the lasue Screaning Form and concluded that
the increased AC inboard Isolation valve closure time was within the design
basis. (VIO 50-254/97022-03; 50 265/97022 03)
+
HPCI Booster Pump 4 Vane Impeller Replacement.
The inspectors reviewed the HPCI booster pump 4 vane impeller replacement
with 5-vano impeller modification. The impeller was replaced to address
vibration problems. The pump curves for the new 5 vane Impeller were based
on the known hydraulle performance of the 4-vane Impeller originally supplied.
The added vane reduced the outlet area of the Impeller by a small amount and
l
resulted in a 2% decrease in the impeller tip velocity. Thus, the pump head was
reduced by 2% near the design point. To compensate for the loss in head, the
Impeller diameter was increased from 22.375 to 23 inches. The inspectors noted
that any adverse effects of a potentially higher pump shutoff head resulting from
the larger Impeller had not been addressed, in response, the licensee
demonstrated that the changed shutoff head lik,-ly would be minimal and
generated PlF#Q1997-04459 to address this issue.
.
Modification M04 2-01034, Addition of four Cells to Unit 2 250 Vdc Battery.
Overall this modification was acceptable but the inspectors noted that licensee
had waived the operability testing of the battery since the four cells had
successfully passed an acceptance test by the manufacturer prior to shipping
and the unmodified 250Vdc system had successfully passed a service test eight
months earlier. However, the factory acceptance test was a capacity test and
', did not verify the ability of the new cells to meet their required load profile. The
licensee wrote PlF 4193 to address this issue.
.
Modification M04-0-82-025, " Replace the Unit 0 250Vdc Battery Charger."
On Site Review Checklist QAP 1270-S?.7 indicates that no setpoint changes
were required for this modification. However, the Release for Testing Checklist
QAP 1270-S28 indicated that setpoint changes were required. The inspectors ~
concluded that the setpoints were computer point additions. Further review
indicated that the setpoints had been tested. PlF 4359 was generated to confirm
that the computer points were properly tested during the modification test,
c. Conclusion
Overall, the sample of HPCI system modifications reviewed were acceptable, however a
violation of design control was identified. Other minor issues raised by the inspectors
were satisfactorily addressed by the licensee.
)
19
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,
.
E2.2 BPCI System IST and TS Surveillance Testlna Review
a. Insoection Scoce
The inspectors reviewed HPCI surveillance and inservice testing (IST) documentation
including test procedures UFSAR, Technical Specifications, calculations and other
design basis documents.
b. Qbservations and Findinas
The team's reviews of the licensee's HPCI system surveillance and IST testing program
Indicated that ths HPCI system test related procedures generally contained clear and
sufficient preparation and alignment steps, acceptance criteria, and verification steps.
The inspectors noted that the TS surveillance test procedures' acceptance criteria
adequately demonstrated continued operability.
IST records reviewed showed that the licensee monitored and trended test results for
each component to detect degradation; reconfirmed or established reference values for
pump vibration, differential pressure, and flow following maintenance or replacement; .
and verified that the new reference values represented acceptable operation. When
testing results indicated component performance was outside the acceptance range, the
licensee took appropriate corrective action as directed by ASME Section XI and the
, licensee corrective action process.
I
Based on reviews of the completed test procedures and followup documentation, the
inspectors noted that engineering staff adequately supported and contributed to
surveillance test evolutions and were involved in review of test results and test
anomalies.
However, as noted elsewhere in this report, an area of concern was raised in the weak
, design basis availability or retrievability. While ultimately the HPCI system test
l acceptance criterla were considered consistent with the design basis, this had not been
I
fully confirmed until completion of all analyses and calculations. For example, the HPCI
pump's differentiel pressure acceptance criteria included an additional discharge
pressure of 100 psl to address frictional line losses between the pump and the reactor.
The licensee identified that no readily recoverable Quad Cities calculation existed to
confirm that the 100 psl was appropriate and had a technical basis. To address this,
during the inspection, the licensee wmpleted an analysis (Calcu.ation QDC-2300 M-
0486) detalling the piping system resistances. As noted in Section E1.1 of this report,
1% talculation satisfactorily documented that the test acceptance criteria for IST and TS
surveillance testing were adequate. The calculation concluded that the HPCI pump
must produce a minimum of 75 psi above reactor pressure when delivering design ,
flowrates therefore the 100 psi pressure drop value used in the test procedures had (
been acceptable.
Based or, a recent trend of licensee and NRC identified TS surveillance non-
compliances, the licensee had also initiated a root cause investigation to evaluate the
_ __
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,
,
trend of testing deficiencies indicating a potential programmatic concern. Numerous
licensee and NRC findings of deficiencies in this area including NOVs, LERs, PIFs, and
other corrective action records were assessed under Root Cause Investigation Report
(NTS 254 200 97 SCA000088). The investigation identified root causes to be
programmatic weaknesses in the testing predefine process, inadequacy of proceowes,
inadequacy of work-in-progress testing, and human errors. Corrective actions that were
in progress of being implemented to prevent recurrence included reviewing all
implementing procedures to verify TS requirements were satisfied, validation of the
EWCS data for all TS related predefinC, F':engthening of interfcces related to TS
changes, procedure revision and modifics.,on processes with the surveillance program.
While the inspectors did not identify further testing concerns, the team concluded that
insufficient time had passed to gauge the effectiveness of these corrective actions,
c. Condus10DR
i
l
Overall, the IST and TS surveillance testing specifically related to the HPCI system were
satisfactory. Based on a recent trend of TS surveillance noncompliance and potential
programmatic testing inadequacles, the licensee had undertaken a roct cause
investigation to evaluate the trend and recommend corrective actions to pmvLt
recurrence. The effectiveness of these actions could not yet be determlu ..
E2.3 HPCI Room Coolers / GL 89-13
l a. Insoection Scopa
!
l
'
The team reviewed the inspection and cleaning program for the HPCI toom coolers
developed in response to Generic Letter 8913. Procedures reviewed included QCOS
5750-09, ECCS Room Cooler Monthly Surveillance and OCTP 111012, 'ECCS Room
Cooler Trending Program," Rev. 2 OCOS 5750-05," Quarterly Testing of SW supply
HPCI Room Cooler Check Valves," Rev. 3, QCCP 1005-05, * Equipment inspection
Program,' Rev.1. Additionally, the inspectors reviewed calculations on the heat
removal capacity of the HPCI room coolers versus design basis heat loads as well as
calculation NED-H MSD 26 which determined the minimum flows to the room coolers at
design river temperatures and, for future operability determinations, at varying river
temperatures and flow rates.
b. Observations and Findings
Quad Citics has implemented an inspection program for the HPCI room coolers, with
,
cleaning as necessary, in response to Generic Letter 8913. Tne inspection and i
cleaning on a regular interval was implemented in lieu of thermal performance
monitoring as these coolers were not conducive to thermal performance monitoring.
Additionally, inlet, outlet, differential pressure and flow were monitored monthly to trend
cooler fouling and plugging. Inspections were being performed every refueling outage
or more often depending on the trend information.
21
_ _
_ _ _ _ - _ _ _ _ _ _ - _ _ - -
.
.
OCOS 5750-09, Rev.10, 'ECCS Room Cooler Monthly Surveillance," adequately
provided instructions for performing the differential pressure and flow test for the HPCI
.
room coolers. Features were provided in inspection procedures ud process for timely j
detection of flow degradation. Cooler cleaning and inspecting was conducted as a
regular Mechanical Maintenance PM every outage. OCCP 1000-05, Rev.1
- Equipment inspection Program," provided sailsfactory instructions for inspecting heat
exchangers, condensers and other equipment for corrosion, foullng and biological
growth.
System engineering used QCTP 111012, *ECCS Room Cooler Trending Program," to
analyze and trend the differential pressure and flow information taken monthly using
QCOS 5760 09. The procedure adequately directed the engineers to set up low and
high dP criteria and gave appropriate criteria on when to write a PlF for adverse trends,
c. ConcluslQD1
The inspectors concluded that the HPCI room cooler was being adequately cleaned and
inspected pursuant te GL 8913 commitments. Flow and differential pressure were
trended and monitored for degradation and cleaning was scheduled on a regular basis.
E2.4 Temocrarv Alterations (TALT)
a. Insoection Scone
The Inspectors reviewed several temporary alterations to ensure the process followed
,
applicable procedures. The documents reviev.ed included the following:
l
QAP 0300-12, Revision 31, ' System Temporary Alterations"
NSWP A-04, Revision 0,"10 CFR Safety Evaluation Process'
TALT No. 96-1005, ' Fine Tune HPCI Flow Controller at Direction of System Engineer"
TALT No. 961063, " Connect Chart Recorder to HPCI Oil Pressure Switches to
Monitor Pressure during Q1R14"
!
TALT No. 961 116, ' Connect Recorder / Monitor HPCI Parameters during 7tartup"
TALT No. 97-2-043, " Block Signals to Prevent Control Room HVAC isolation"
b. Observations and Findings
b1. Temporary Alteration No. 961005 installed test equipment on the Unit 1 HPCI system
to obtain performance data. The 50.59 screening approved on January 26,1996,
described the preposed change as the connection of a strip chart recorder to various
banana Jacks to monitor six (6) parameters in order to fine tune the control system. The
screening stated that the recorder would be the only interface with the affected
instrument loops. Recorder failure modes (open/ shorts) were addressed in the
screening. The inspectors noted that Work Request Task 950119950-01 also installed a
movement transducer on the turbine linear variable differential transformer (LVDT) lever,
a movement transducer on the secondary operating lever, and three (3) pressure
gauges. However, the original screening did not evaluate the physicalinstallation and
22
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4
.
failure modes of the additionalinstrumentation per Procedure NSWP A 04. Section
' 2.5 of this procedure stated, in part, that previously performed screenings can fulfill the
.
scre.,ning requirement provided they meet the validation criteria of Exhibit H. *Valiaation
>f Previously Performed Safety Evaluations and Screenings," to determine if an existing
poening remains valid. The work control process did not identify that the installation of
additional instrumentation to the HPCI system altered the previous screening evaluation.
This example was considered an example of a violation of 10 CFR 50, Appendix B,
Criteria V, for not following NSWP A-04. The licensee initiated PlF 4500 to track this
item. The safety consequences were minimal since the unit was shut down and HPCI
was not required to be operable (VIO 50 254/97022 048).
b.2 TALT No. 961063 installed test equipment on the Unit i HPCI system to record
various parameters during surveillance testing. The screening approved on
r ebruary 28,1996, described the change as the connection of a strip chart recorder.
The screening stated that the recorder would be the only interface with the affected
instrument loops. Recorder failure modes (open/ shorts) were addressed in the
screening. The inspectors noted that Work Regeest Task 950084936 01 also installed
three (3) Valldyne pressure sensors. However, the original screening did not evaluate
the physicalinstallation and failure modes of the additionalinstrumentation per
l Procedure NSWP A-04. Section 6.2.5 of this procedure stated,in part, that previously
performed screenings can fulfill the screening requirement provided they meet the
validation criteria of Exhibit H, Validation of Previously Performed Safety Evaluations
and Screenings," to determine if an existing screening remains valid. The work control
process did not identify that the instal;ation of additional instrumentation to the HPCI
system altered the previous screening evaluation. This example was conslaered an
example of a violation of 10 CFR 50, Appendix b, Criteria V, for not following piocedure
NSWP A 04. The licensee initiated PIF No. Q1997-03981 to track this item. The safety
consequences were minimal since the unit was shut down and HPCI was not required to
be operable (VIO 50 254/97022-04b).
b.3 TALT No. 97 2-43 (Design Change Package 9700126) Installed relay blocks on relays
2 595102A/B/C/D to maintain the relays in a simulated energized state to prevent the
control room ventilat;on system from isolating. These relays Initiate main steam line
Isolation and control room ventilation system isolation for a main steam line high flow
condition. The relays were blocked for instrument mechanic Out-of Service No. 21122
walkdowns. The main steam flow channels were required to be operable during
operational Modes 1,2 and 3, Since Unit 2 was shutdown, the 50.59 screening
preparer determined that only UFSAR chapters for the current operating Mode were
applicable. As a result, NSWP A-04, Exhibit E,"10 CFR Screening For Facility
Change," Question 2, may not have listed all applicable UFSAR sections. While t;ie
Inspectors determined thct the safety consequences were minimal since the unit was
shutdown, this was of concern since the requirement that a full safety evaluation be
completed may have been circumvented. The licenroe initiated PIF 4394 to track this
itern. Pending NRC review of the affect of application of all modes of operations, the
inspectors considered this an unresolved item. (URI 254/265/97022-05)
23
e - -,y
y '
<
\c. Conclualana
s,
.,
" Contrary to procedure NSWP-A-04 requirements,50.59 screenings for two (2) -
~ Temporary Alterations reviewed by the inspectors did.not evaluate the physical ~
Installation of allinstrumentation installed by the alteration.-
'
,
E3 ' Engineering Procedures and Documentation
f E3.1 ~ 10 CFR 50.59 2 * 'W Evah a+' -7 Prae === ,
-
-
. a.1 inapaction Scone
~
The inspector reviewed a sarhpie of HPCI 50.59 safety evaluation reports with
4
- associated PlFs and NTS reports and the following procedures and documents.;
'
.
, ." Changes, Teste, arid Experiments Completed," LWP 96-048, dated July 08 .
.- 1996.
,
'
- -
" Summary Report of Changes, Tests and Experiments Completed,' SVP-97-251,
" dated October 31,1997
- -
NSWP-A iO4, Rev. O, "10 CFR Safety Evaluation Report"'
'
Y "50.54(f) Task Assignment, Configuration Control Process"
a
.
s '" Conduct of Off-Site Reviews," N.O.-16, Revision 9, dated October 8,1997
4 ,s )
- -
" '
. Updated Final Safety Analysis Report (UFSAR) Update," ESK-96-113, dated
f LJune 7,' 1996 -
E .
.
_
a
[ + l " Updated Final Safety Analysis Report (UFSAR) Update," SVP-97-242, dated
. 1 October 22,1997a ' '
b .
-
b.1 Findinga and Observations
. .
N -:b;11 General Weakna==== in 50 59 Proaram --
" -
,
lThe 10 CFR 50.59 safety evaluations (SE) reviewed were adequate and conclusions
- were supported; however, several general weaknesses were identified.
i Safety Evaluation (SE)96-061 was approved in July 1996 to correct the i
L op;ating pressure range for HPCI in the UFSAR from 1150 psig to 150 psig to
7".
the value 1120 psig to 150 psig. A mathenutical error was introduced during a
previous UFSAR update when the values were converted from absolute
pressure to gage pressure. On September 24,1996, SE 96-061 was checked to
confirm its accuracy and conclusions against UFSAR, Rev. 3 olus approved l
UFSAR change packages because the subject SE may have been initially ;
l
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,,-
evaluated using UFSAR information which did not contain Rev. 3. ;On June 28,
1996, the NRC approved the new Technical Specmcations (TS) which were -
' implemented on Sopiomber 23,1996. These new TS included the same error in
its Bases of TS Section 3/4.5.A and B; however, it '.vas not recognized by the
- reviewer. This change was made in UFSAR, Rev. 4 issued by letter dated
October 22,1997, and PIF 4147 vias initiated on October 30,1997, to address
the TS Bases error. The development of SE 96-061 and the subsequent review
- should have included a review o' Be new TS Bases in the evaluation.
- i
NSWP-A-04, Rev. O, Attachment G, Step 13, requests the preparer to check one -
of four appropriate conditions to determine the effect on Margin of Safety. One
of the conditions reads: "The change does not affect any parameters upon
- which Technical Specifications are based; therefore, there was no reductm in
- the margin of safety r proceed to Step 15." The SE does not require an -
- explanation as to the rationale for that choice as it does fcr the other three ~ ,
- conditions.- This condition was chosen for the majority of the 50.59 summaries
provided in the 50.59 report issued October 31,1997, and the 50.59 SEs
reviewed. Several of these SEs involved changes to' safety related equipment
and, without a rationale for the conclusion of "no effect on any parameter upon
<
' which the TS are based," It was not intuitive to conclude that there was no -
reduction in the margin of safety.
- -
The 50.59 report," Summary Report of Changes,' Tests and Experiments
Completed," SVP-97-251, dated October 31,1997, included a summary of SE
1
96 043 which evaluated changing the UFSAR to address a new TS concoming -
'
heater power for the am filtration unit heater. The summary states in step 3 that,
_ "The margin of safety as defined for any Technical Specification, is reduced
because the actual Tech Spec requirement for heater-power is not given in the .
UFSAR and would be reviewed by the NRC in the Tech Spec change SER." If
there was a reduction in safety,10 CFR 50.59 requires that this change be an-
-
Unreviewed Safety Question (USQ). There was no discussion of a USQ in the
0 summary. Upon review of SE 96-043 it was determined that the conclusion of
>
the SE was that there was not a reduction in the margin of safety.L This was -
considered an example where an insufficient review of information that was
- provided to the NRC. The staff relies on information provided by the licensees to
be accurate.
N 10 CFR 50.59 screenings which should be documents that are controlled under
! the quality control program, were not easily retrievable when a screening's
_ _
. conclusion was that a full safety evaluation was not required.
- :These above concems noted by the inspectors, along with numerous licensee-identified
PlFs conceming 50.59 problems with screenings, with documentation deficiweies, with
inadequate reviaws, with qualifications of reviewers and preparers l and with lack of off-
site reviews, substantiated a programmatic weakness in the Quad Cities 50.59 program.
Corrective actions initiated by Quad Cities staff included establishment of a third level
-
25
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..
k
review of n,afety evaluations by the Engineering Assurance Group and further training.
The effectiveness of these recent actions could not yet be assessed.
- b.2 incomolate 10 CFR 50.59 Summarv Reo_ ort
The inspector reviewed 10 CFR 50.59 summary reports LWP 90-048, dated July 08,
1096, and SVP-97-251, dated October 31,1997, to verify completeness of the reports.
The summaries were being provided pursuant to 10 CFR 50.59(b)(2) and 10 CFR
" 50.71(e). The inspectors noted that each of thew summary reports contained only the
descriptions of 50.59 evaluations tnat actually changed the Updated Final Safety
Analysis Report (UFSAR). It did not Include a description of each changa, test, or ,
, experiment to the fecility as described in the SAR completed since the last report.
Approximately 60 safsty evaluations performsd in accordance with the 50.59 program
were not included in these reports. Listed below are examples of 50.59s SEs reviewed
during the SOPl inspection that were not included in their 50.59 Summary report:
.
SE 96-22; Temporary alteration to discble thermal overload alarm on HPCI '
auxiliary oil pump.
.
SE 97 019; Interim procedure will render the HPCI subsystem unavailable while
the steam isolation valves are closeo. Perform interim procedure to test the
HPCI interlock which runs the motor speed changer to the high speed stop upon ,
a high drywell pressure initiation signal.
.
SE 96-085; Reclassify the HPCI keep-fill lines between valves 1(2)-2381 and
>
1(2)-2399 as Safety Related.
!
.
SE 96 095; install an expandable plug in floor drain to restrict air flow. The floor
drain allows an opening in secondary containment when the reactor building
i'
(Inner) door is opened. The plug will restrict air flow to ensure that the required
negative pressure is maintained in the Reactor Building.
The failure to report a description of each 50.59 safety evaluation was contrary to
procedure NSWP-A-04, Rev. O, "10 CFR 50.59 Safety Evaluation Process, Section
5.4.1.3, which states, in part, that the report shall contain a brief description of each
- change, test, or experiment and a summary of the safety evaluation performed." This
was considered a violation of 10 CFR 50, Apper dix B, Criterion V (VIO
50-254/265/97022-06).
b.3 Off-Site Review
Procedure N.O.-16, Revision 8, " Conduct of Off-Site Review," was reviewed in
conjunction with several Quad Cities Off-Site Monthly Review Reports, and Off-Site
Review Reports on HPCI 50.59 safety evaluations. Procedure N.O.-16 describes the
organization, responsibilities, and duties of the off-site review group including personnel
qualifications and review process. This procedure provided a comprehensive program
for SE review along with reporting requirements on questions and comments on
26
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,
\
individual SEs and the reporting of off-site review activities on a monthly basis.
Questions raised by the off site review group were reported in an Off Site Review l
'
Report where tracking was provideo by the site Nuclear Tracking System (NTS) with
response requested usually within 75 days. Based on the comments provided by the
off site review grouo in the individual HPCI Off-Site Review Reports and the June / July
and August Off-Site Review Monthly Reports the inspectors concluded that the off site
review group was providing good safety focussed feedback when assessing the 50.59
SEs.
b.4 Status of m Mm Task Assignment - Configuration Control Process
.
- In November 1996, the licensee completed a "50.54(f) Task Assignment - Configuration
Control Process," as input into their response to a staff letter * Request For information
Pursuant To 10 CFR 50.54(f) Regarding Adequacy And Availability Of Design Bases
information* dated October 9,1996. This input to the response letter addressed five
corrective acuons after reviewing various inspection and assessment reports. The
inspectors reviewed the status of these corrective actions:
Corrective Action #1 was to perform increased reviews of current 50.59 safety
evaluations and screenings to target trends which show a likely noncompliance,
as stated in the task assignment. It further stated that this was an appropriate
response to the frequency ofinadequate evaluations. PlF 96-03374 was
initiated to address this issue and NTS 254 20196-337401 was assigned to
track this effort. The inspectors noted that this NTS item was still open and had
been passed between various personnel with a due date in January 1998.
Corrective Action #2 was to address inadequacies in offsite reviews. PlF
96-03375 was generated and NTS 254-201-96-337501 was assigned to this
task. This action addressed issues such as 50.59 documentation requirements,
l off-site review location, and adequacy of off-site review requirements. Based on
l reviews of actions taken, the inspectors concluded that this issue was
adequately closed in March 1997.
Corrective Action #3 addressed the adequacy of the current 50.59 safety
evaluation procedure. PIF-03376 was initiated and appropriately closed upon
issuance of a new corporate 50.59 safety evaluation procedure NStNP-A-04,
Rev. O, which was issued January 1997. Although the procedure was under
review by the licensee for further changes to improve written justifications when
addressing a facility change, the procedure was considered a distinct
improvement over the existing site procedures.
Corrective Action #4 was to address the establishment of a site 50.59
coordinator to help maintain consistency in safety evaluations. PIF-03377 under
NTS 254-251-96-08507 was initiated and came to a reasonable conclusion that
a specific 50.59 coordinator was not necessary because of the establishment of
an Engineering Assessment Group that will review and verify the adequacy of
50.59 screenings and safety evaluations.
27
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i
Corrective Action #5 under PIF 03378 and NTS 254 20196-337801 was to
address various PIFs and external processes outside of the 50.59 process to
determine whether 50.59 requirements were being bypassed. The inspectors
noted that this activity was still open, had been passed between various
personnel and was projected to be completed in March 1998. J
c. Conclusions
While the SEs reviewed were adequate with supportable conclusions, weaknesses were i
identified with the overall 50.59 pregram. These included poorly written safety
evaluations, incomplete summary report submittals to the NRC, difficult to retrieve
screenings, and incomplete corrective actions to identified deficiencies. Additional 50.59 i
related concems were discussed in Section E2.4 of this report.- The Off Site review
group, however, was providing good assessments and comments.
E3.2 UFSAR Sections on the HPCI System
a. insoection Scoos
,
The inspectors reviewed UFSAR sections on the HPCI system, and compared the
UFSAR to system prir'ts, TS, and HPCI procedures to check the accuracy of the
bc Observations and Fir dinos
The inspectors identlW that UFSAR Figures 6.3-14 and 6.3-15 did not
accurately represent me HPCI system. Figure 6.3-14 was a simplified overall
- diagram of major HPCI system components, piping and valves. The valve
positions indicated on the Figure 6.3-14 did not represent any operational state
of the system; not the standby lineup, test lineup, injection lineup, or pressure
control lineup of the system. Additionally, the figure indicated that the HPCI
Contaminated Condensate Storage Tank (CCST) suction valve MO 1(2)-2301-6
was open simultaneously with tM torus suction valves MO 1(2)-2301-35 and 36.
This lineup was generally avr )y the licensee as it creates a potential drain
path between the torus and C,,.,. . Additionally, an interlock exists that if the
torus suction valves were both opea, then the CCST suction valve will auto
close. Figure 6.3-15 also did not accurately represent the HPCI system. Figure
6.3-15 was a simplified functional block diagram for key valves in the HPCI
system. This figure incorrectly indicated that steam isolation valves MO 1(2)-
2301-4 and 5 receive some kind of open signal from a HPCI auto initiation. The
licensee generated a PIF to address this issue and correct these identified errors
with figures in the UFSAR
The inspectors identified that UFSAR Section 6.3.2.1.4, " Core Spray Discharge
Line Fill Provisions," stated that pressure switches were provided to indicate and
alarm high or low pressure in the ECCS pump discharge headers to ensure
proper functioning of the fill system. However, there was no pressure switch
28
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,
Installed on the HPCl pump discharae header to indicate loss of fill. The
licensee generated PlF 4050 to address this discrepancy.
.
The licensee reviewed the UFSAR in June 1996, to determine whether the
document reflected all necessary changes in response to a finding at another
facility. The inspectors considered this a good initiative by the licensee to assure
that the most recent update met the intent of 10 CFR 50.71(e). All changes inat ,
were found not to have been incorporated into the UFSAR, some completed as 1
early as 1991, were included in the most recent update, Revision 4, on
October 22,1997. Corrective actions to programmatically ensure that all future
changes will be reflected in the UFSAR in accordance with 10 CFR 50.71(e)
were in progress. NRC review of these plans and of their effectiveness was
considered an inspection follow up item (IFl 50-254/265/97022-07).
.
The inspectors also noted that the Quad Cities Design Basis initiative (DBI)
project was in the process of initiating a line-by-line validation of UFSAR design
basis information.
c. Conclusirms
inaccurate figures and text in the UFSAR were identified but the inspectors also noted
ongoing licensee efforts to improve UFSAR accuracy such as line by line reviews of the .
UFSAR design information ano an initiative to ensure facility changes had been
incorporated into the UFSAR.
E3.3 HPCi System Design Basis Document (DBD)
a. insoection Scone
The inspectors reviewed the HPCI system design basis document DBD, and compared
the HPCI DBD to the UFSAR, TS, end HPCI procedures to check the accuracy of the
DBD.
b. Observations and Findings
At the time of this inspection, the HPCI DBD was in its third revision and was undergoing
further review prior to validation. Numerous discrepancies in the HPCI DBD were
discovered. These included:
.
Inconsistencies within the DBD concerning the response of steam isolation
valves MO 1(2)-2301-4 and 5 to a HPCIinitiation signal, in response to an
initiation signal, page 4-10 of the DBD said that these valves will auto open; page
4-18 of the DBD said that these valves will not auto open. In fact, valves MO
1(2)-2301-4 and 5 will not auto open if an initiation signal was received.
.
The DBD incorrectly omitted HPCI high area temperature as a signal which
would close the steam isolation valves MO 1(2)-2301-4 and 5,
29
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The DBD incorrectly stated that the minimum flow bypass valve MO
1(2)-2301 14, cannot be opened if the HPCI turbine steam suoply valve MO
1(2)-2301-3, was closed.
.. The DBD incorrt,ctly omitted high suppression pool level as a signal which would
'
open torus suction valve MO 1(2) 230136.
c. Conclus10Ds
Quad Cities design basis and configuration controlinformation weaknesses were also
exhibited with numerous errors identified with the HPCI system DBD. However, the
-licensee was aware of the DBD shortcomings and had designated the DBDs as
"information only" pending completion of a validation process.
E7 ' Quality Assurance in Engl9 eering Activities
y
E7.1 10 CFR 50.54(f) Letter Commitment Review
a. Insoection Scoce
The team reviewed the status of commitments pertaining to the licer'see's March 28,
1997, response to the NRC's request for information purruant to 10 CFR 50.54(f). The
following commitments related to engineering and the corrective action program at Quad
Cities were reviewed by the team. The commitment numbers correspond to those used
l by the licensee in their March 28,1997, submittal,
i
b. Observations and Findinas ;
b.1 Commitment 16: "These actions included, in part, establishment of an engineering
'
assurance function at each site and the NOD central offices to further ensure the quality
of design and technical work, commencement of safety system functional inspections, .
(
' review of Technical Specification interpretations, and a review of the top ten risk
significant systems for items that may impact system readiness."
The team determined that the Engineering Assurance Group (EAG) had been
established at Quad Cities. One SSFI-related review and the top ten risk significant
systems reviews have been completed. Technical Specification interpretations were no
longer used at Quad Cities.
_ b.2 Commitment 20: "A nuclear engineering procedure for this effort is being prepared and
will address the review and reconstitution of selected key design basis '
parameters / calculations."
The team verified that procedure NEP-17-08, Revision O, " Design Basis Initiative," May
23,1997, had been implemented at Quad Cities. Design Basis initiative engineering
personnel had been trained on the procedure.
30
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b.3 Commitment 55: "In order to ensure that corrective actions and responses to lessons
learned are consistently and vigorously implemented throughout NOD, a new corrective
action program has been developed by representatives from all six nuclear sites and the
NOD central office."
The team verified that the new corrective action process had been implemented at Quad
Cities on May 12,1997. A common set of performance indicators had been developed
and summarized on a monthly basis,
'
b.4 Commitment 58: "The new process includes several improvements over the current
program. It clearly delineates and standardizes the threshold for problem identification
through Problem Identification Form (PlF) initiation, and establishes common PlF
screening criteria that provide greater ability to analyze PlF data."
The team verified that NSWP A-15, Revision 1, " Problem Identification Form," May 5,
1997, had been implemented at Quad Cities on May 12,1C37. The procedure included
standerdized requirements on when a PlF should be issued and established common
PlF screening criteria,
b.5 Commitment 59: " Groups of these trained individuals will be stationed at each of the
nuclear plant sites and in the NOD central office."
The team verified that the Quad Cities root cause analysis team was trained by an
outside contractor on February 11-14,1997,
b.6 Commitment 61: "The remaining sites have devebped plans to implement this process
during 1997."
The team verified that NSWP-A-15, Revision 1, " Problem Identification Form,"
implementation and initiation of monthly Quality and Safety Assessment Performance
indicator reports satisfy this commitment.
b7 Commitment 63: "The information will be taken monthly and used to evaluate the
effectiveness of corrective action process improvements as well as participation by each
site in the process."
The team concluded that the monthly Quad Cities Quality and Safety Assessment
(Q&SA) reports were adequate to satisfy the commitment,
b.8 Commitment 64: " Performance indicators have also been developed to monitor the
timeliness of implementation, quality of corrective actions, and the number of significant
events which are repeated. These indicators are being tasted at Byron. Site and NOD
central management will take appropriate actions based on performance and results."
Based upon the review of Quad Cities performance indicators in the Q&SA monthly
reports, and the results of corrective action, the team concluded that this commitment
was satisfied.
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b.9 Commitment 71: "In February 1997, a procedure was issued for evaluating and
initiating NOD-wide action in response to operating experience at any of the Comed
nuclear stations. The procedure also covers response to operating experience
materials both from Comed and non-Comed stations. The procedure provides for
review and screening of operating experience items, development of responsive action,
and review and evaluation of the effectiveness of responsive action."
The inspector's review of procedure NSWP-A-06, Revision O," Operating Experience '
Program," February 27,1997, resulted in the determination that this commitment was
satisfied.
- b.10 Commitment 88: "Esch site also has a group that evaluates the severity of events, and
oetermines whether a root cause analysis is warranted. Processes are being
implemented for evaluation of the effectiveness of corrective action." a
The inspector verified that the Event-Screening Committee was in place and performing
those functic~s delineated in the commitmant.
b.11 Commitment 89: " Monitoring of performance against the indicators, Corrective Action
Requests (CAR), and industry experience; and review of site self-assessments will also
be conducted within SQV."
Quad Cities bd implemented Nuclear Overs:ght procedure NO-19, Revision 1,
" Integrated Analysis Process and Routine Reporting," that contains instructions for
l monitoring performance indicators, CARS and Self-assessments,
b.12 Commitment 91: "The SRBs evaluate station safety performance, corrective actions,
and improvement plans. The SRB chairman will also provide input to the NOC of the
board. The site gains outside perspective aM critical review of performance from this
body."
The inspectors verified that Quad Cities received recommendations from the Safety
Review Board (SRB) independent reviews and entered the recommendations into the
site tracking system. The recommendations were then assigned to on-site managers for
analysis and necessary actions.
b.13 Commitment 284: " Design records were transferred from contract design engineering
organizations to Comed."
Design records had been transferred to Quad Cities and stored in the station central
filing system,
b.14 Commitment 287: "A standardized corporate corrective action program, based on a
review of industry programs, is being implemented throughout NOD. The program
incit' des specific performance measures to gauge program effectiveness. A corporate
corrective actions group is being established to ensure the appropriate response to site
and industry events."
32 1
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,
The inspectors verified that a corporate corrective actions group had been established
within the Q&SA organization at Quad Cities.
b.15 Commitment 301: "A NOD-wide formal program for evaluating, sharing, and assessing
the effectiveness of responses to lessons leamed at both Comed and other nuclear
stations is being implemented to assure lessons leamed are being shared and
responded to throughout NOD."
The inspectors verified that the Lessons Leamed Program, a corrective Actions
Program and an Operating Experience Program had been implemented at Quad Cities,
b.16 Commitment 304: " Standardized performance measures are being implemented to
gauge processes and effectiveness of Corrective Actions."
The inspectors verified that standardized performance measures have been
implemented at Quad Cities and were evaluated in Q&SA monthly reports,
b.17 Commitment 323: " Design authority and design records were transferred from contract
design engineering organizations to Comed, on-site design engineering capabilities
were increased, and we are developing a series of common engineering processes and
procedures for the division."
Design records had been transferred and on-site capabilities increased. The common
engineering procedures and processes had been developed,
b.18 Commitment 3_24 "We initiated a broad set of initiatives to ensure that each of our sites
has a high quality engineering support to maintain the facility design bases. Engineering
Assurance groups [EAG) were formed at each site to improve the quality of design and
technical work, with a specific focus on maintaining the design basis."
The EAG at Quad Cities produced monthly reports of results of their monitoring of site
engineering activities. EAG staffing was three full time nnd one full time equivalent and
appeared adequate.
c. Conclusions
Based on procedure and other documentation reviews, and interviews with cognizant
licensee personnel, the inspectors concluded that the preceding 10 CFR 50.54(f)
commitments were closed. The remaining 10 CFR 50.54(f) commitments that remained
open will be reviewed in future NRC inspections.
E8 Miscellaneous Engineering issues
E8.1 Ooen items
(Closed) VIO 50-254/265/97013-01: TS Surveillance Requirements for the RHRSW
pumps were not relocated to the IST program as required by TSUP t.ommitments. The
33
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.
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- lST test procedure (OCOS-1000-4 " Quarterly RHRSW Pump Operability Test," Rev,13)
was updated to include the test criteria. The inspectors verified that the required TS
surveillance requirements were appropriately relocated to the IST program. Further, all
TSUP transfers that involved IST relocations were reviewed to ensure that similar
transfers of test requirements were not missed. No other problems were found, this
violation is closed.
E8.2 Systematic Evaluation Procram (SEP)
a, insoection Scone
Subsequent to the Systematic Evaluation Program (SEP) that was completed for
Dresden 2 in 1930 the licensee contracted Sargent & Lundy (S&L) to review the
_
i
Dresden SEP actions for applicability to Quad Cities. S&L recommended action on 17
SEP topics and in Septe.nber 22,1993, QC engineering issued a report acknow! edging
the S&L recommendations. The NRC's November 1993 Quad Cities Diagnostic .
Evaluation Team (DET) inspection report exp.assed concern that action to address the
Dresden 2 SEP issues at Quad Cities had not been initiated and action plans had not .
. been completed (DET Issue #9). . in subsequent correspondence, the licensee informed l
the NRC that the action plans vere completed and subsequently that all actions were
also completed or had been reassessed as not required. The inspectors reviewed these
. actions.
b. Observations and Findinos
I
b.1 SEP ltem 1. SEP Toolc ll-3.B. Floodina Potential and Protection Reauirements
The team reviewed the licensee's assessment of the probable maximum flood (PMF)
effect on Quad Cities to evaluate the plant's ability to cope with extemal flood'.ng
conditions. Quad Cities initial design was based on a PMF with a 200 year racurrence
interval, and . plant design was shown to have sufficient margin to withstt4nd floods
- with 1000 year recurrence interval. Subsequent to plant co,struction, the NRC adopted
the PMF as defined by the US Army Corps of Engineers as criteria for plant design
purposes. The e'fect of the updated PMF was noted and addressed in the UFSAR.
The updated PMF results in flood levels about eight feet above plant grade. Such a
flood would take place with sufficient warning to allow effective maasures to ensure that
the plant to be placed in a safe shutdown condition.
The inspectors reviewed We licensee's flood emergency procedure and determined that
the procedure was adequate to place the plant in a safe shutdown condition and to
maintain structuralintegrity up to a flood elevation of 603 feet elevation. Based on
reviews of the licensee's PMF calculations, flooding assessments, emergency
procedures and discussions with cognizant licensee personnel, the inspectors
concluded that the licensee had adequately addressed this issue. NRC review of SEP
ltem 1, Topic ll-3.B is considered closed.
34
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6
b.2 SEE ltem 2. SEP Tooic !!-3.B.1. Caoability of Ooerating Plant to Cooe with Desin-
Elosding Conditions
The team reviewed the Quad Cities flood emergency procedures and event
classification and notification and comparison to design basis as stated in the UFSAR.
The inspectors reviewed the Quad Cities flood emergency procedures, appropriate
sections of the UFSAR, Technical Specifications and the probable maximum flood
(PMF) esessment for the site. The procedures controlling site activities necessary to
protect the plant and equipment during a PMF appeared sufficiently comprehensive to
maintain plant component and structuralintegrity up to a flood level of 603 feet mean
sea level elevation. Based on the i spectors' review of procedures, UFSAR and
Technical Specifications, the inspectors concluoed that the licensee had adequately
addressed this issue. NRC review of SEP ltem 2, Topic ill-3.B.1 is considered closed.
b.3 SEP Item 3. SEP Toolc lil_;LC. Inservice insoection of Water Control Structures
This topic assessej the at%quacy of the inservice inspection program of water control
structures for operating plants to assure conformance with the intent of Regulatory
Guide 1,127. The recommended actions were to identify or create procedures to ensure
review and approval of the ISI program by qualified engineering personnel and initiate
inspection after extreme events as required by RG 1.127. The *,ensee identified that
procedures were in existence to address the topic and these procedures were reviewed
by the inspectors. The inspectors reviewed Procedure NEP 17-03, " Structures
Monitoring," Revision 0, Procedure QCMPM 4400-11, "RHRSW Intake Bay inspection,"
Revision 3, and Procedure OCMPM 4400-12, " Circulating Water intake Bay inspection,"
l Revision 2. The inspectors noted that the procedures adequately describe a formal
annual inspection of the intake structure by qualified engineering personnel who would
document the results of the inspection. In addition, the inspection program included
provisions for special inspections immediately after occurrence of extreme events.
These actions satisfied the SEP topic concerns. NRC review of SEP ltem 3. Topic 111-
3.C is considered closed.
b.4 SEP ltem 4. SEP Tooic lil-8.C. Iriadiation Damage. use of Sensitized Stainless Steel
SEP ltem 6. SEP Toolc V-4. Pioing and Safe-End Integrity
These topics assessed the safety aspects of intergranular stress corrosion cracking of
sensitized stainless steel used within the reactor vessel systems. For both topics, th
recommended action plan was to confirm that all sensitized safe-ends had been
removed from service. The inspectors reviewed the licensee's actions taken that
included, historical reviews of the ISI program, of code data material reports, and
reviews of historical modifications and licensing basis documents that have replaced
sensitized components. The inspectors concluded that the licensee's determination that
the Quad Cities reactor vessels do not contain any sensitized safe ends was
acceptable. NRC reviews of SEP ltems 4 & 6 Topic lil-8.C and V-4, respectively, are
considered closed.
35
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Discharge Valves
~ This topic essessed the necessity to modify the control circuit configuration of the .
' recirculation line suction valves to ensure that these valves remain open on a 1
recirculatbn line break LOCA so that LPCI can be successfully initiated. (GE had
- identified the potential for spurious closure of these valves with a LOCA occurring
between the pump suction and discharge valves.)-The S&L action plan recommended
that the station confirm that the breakers to the recirculation line suction valves were-
racked out.' Subsequent to issuance of the S&L action plan, the licensee identified that
racking out the breakers was not required because these valves had been modified in
1978 to reconfigure the closing logic. The inspectors reviewed these actions and-' ;
-
'
confirmed that the modifications had been implemented and concurred that the need for
racking out the valves was no longer applicable. NRC SEP ltem 5, Topic lil 10.C is
considered closed.
b.6 SEP ltem 7. SEP Toole V-5. Reactot. Coolant Pressure Boundarv I ask Detardhn -
This topic assessed the adeqtacy of the reactor primary coolant leakage detection
system. The S&L action plan had recommended revising Technical Specifications
_ Section 3.6 / 4.6 to include additional monitoring requirements and tighter limits on
"
unidentified leakage. The licensee identified that these recommendations had been - ,
t
!
Incorporated into the upgraded Technical Speclilcations.- The inspectors reviewed the -
, latest version of the Technical Specifications (the NRR reviewed and approved TSUP) :
- and concluded that sections 3.6 G/4.6.G. Leakage Detection Systems, and 3.6.H/4.6.H.
L : Operational Leakage satisfactorily addressed the SEP issue with the type and sensitivity
ofleak detection systems. NRC review of SEP ltem 7, Topic V-5 is considered closed,
b.7 ' SEP ltem 17. SEP Toolc XV-16. Radioloalcal Consecuences of FalNre of Small Lines
Carrvina Primarv Coolant Outside Containment
' This topic assessed the radiological consequences of failure of small lines carrying
' primary coolant outside containment and reviewed Technical Specifications associated
with primary coolant radioactivity concentrations. 'Similar to Dresden SEP action, the _
~
action plan recommended changes to the Technical Specification to_ limit reactor coolant
specif:c activity during power operation within the BWR Standard Technical Specification -
limits. The licensee identified that these recommendations had been incorporated into
the upgraded Technical Specifications.; The inspectors reviewed the Technical
Specifications and verified that TS Section 3.6.J/4_.6.J, Specific Activity had been revised
- and reviewed under the TSUP program. The specific activity of the reactor coolant was
now limited to less than 0.2 microcurie / gram dose equivalent 1-131 during modes 1,2
- and three. NRC rieuw of SEP ltem 17, Topic XV-16 is considered closed.
.
c. Concinalon
The inspectors reviewed the action taken by the Quad Cities staff for eight SEP topics -
and concluded that the actions taken were sufficient for closure of these items. NRC
review of nine remaining SEP ltems was ongoing.
36
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3-
V. Management Meetings
-X1 Exit Meeting Summary -
The inspectors presented the liispection results to members of licensee management at
- the conclusion of the inspection on November 21,1997. The licensee acknowledged .
the findings presented and did not identify any documents provided to the inspectors as
' proprietary,
37
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PARTIAL LIST OF PERSONS CONTACTED j
ucense.
- S. Boline, Quad Cities Mechanical Design
D. Brown, S&L
- A. Chemick, Comed RA Supervisor.
D. Cook,-Station Manager
< D. Cook, Comed Maintenance 1
D.' Egan, Comed dbl-:
S. Eldridge, Comed, EAG Supervisor- ,
R. Faltbanks, Engineering Manager
' H.; Gavankar, Comed, Chief Engineer, Mechanical
.
W. - Heinmiller, Comed Site Design Supervisor -
R.- Hoyn, Quad Cities Mechanical Design
J. Hosmer, Vice President, Engineering
G. Klone, Quad Cities Operations--
P. Lawless, Quad Cities SOPI Team Lesder
- H. Palas, Comed Pump Specialist
L. Pearce, Site Vice Presidsnt
- K. Salehl, Comed, Engineering Assurance Group-
.
- R.. Svaleson, Operations Manager
'J. Swales, SystemEngineer
T.-. Thorsell, Comed Chief Engineer, Ell &C -
- F. Tsakeres, Training Manager .
.- M. Wayland, Maintenance Manager.
JJ.1 Williams l Comed Project Manager
NRC
. L. Collins, Quad Cities Resident inspector .
R. Ganser, Illinois Department of Nuclear Safety .
.
R. Gardner, Chief, Engineering Branch No. 2
C, Miller, Quad Cities Senior Resident inspector
i M. Ring, Chief, Projects Branch No.1
K. Walton, Quad Cities Resident inspector
INSPECTION PROCEDURt:$ USED
lP 40500: . Effectiveness cf Licensee Controls in Ident!fying, Resolving, and Preventing
Problems
IP 93801: Safety System Functional Inspection
- IP 37550
- Engineering -
38
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ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
ii . l 50-254/265/97022-01 t VIO - Criterion V, failure to prescribe in procedures
- activities affecting quality (Procedure would not
work)~
-50-254/265/97022-02 . URI. Breaker coordination action plan
'50-254/265/97022-03 VIO Criterion 111, failure to factor additional AC inboard
isolation valve closure time for LOOP concurrent w/
HPCI steamline break .<
50-254/97022-04a VIO Criterion V, failure to fotbw 50.59 procedure
- 50-254/97022-04b - Vio Criterion V, failure to follow 50.59 procedure
50-265/97022 05 URI 50.59 screening did not evaluate al' modes of
operation
5b-254/265/97022 06 VIO Criterion V, fa!!ure to follow 50.59 procedure on
report submittals
i
50-254/265/97022 07 IFl Follow up on corrective action on 10 CFR 50,71(e) i
UFSAR update requirements !
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LIST OF ACRONYMS USED
AOP Abnormal Operating Procedure :
ASME - American Society'of Mechanical Engineers
BWR Bolling Water Reactor
CAR Corrective Action Record
CCST Contaminated Condensate Storage Tank
CFR Code of Federal Regulations
Comed Commonwealth Edison Company-
DBD Design Basis Document
DET Diagnostic Evaluation Team
DCP Design Change Package
DCR Design Change Request
DC/dc Direct Current
EAG Engineerirg Assurance Group _
ECCS Emergency Core Cooling System
EOP Emergency Operating Procedeire
ESW Essential Service Water
EWCS Electronic Work Control System 1
GL Generic Letter
HVAC Heating Ventliation Air Conditioning
HX Heat Exchanger
- . HPCI- High Pressure Coolant injection
l IEEE Institute of Electrical and Electronic Engineering
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IFl Inspection Follow up Item
ISI- Inservice inspection
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-lST Inservice Testing
L HPCI- High Pressure Coolant injection System
KV Kilovolt
LCO Limiting Condition for Operation
LER Licensee Event Report
LOCA Loss of Coolant Accident
LPCI Low Pressure Coolant injection System
LVDT. Linear Variable Differential Transformer
MOV Motor-Operated Valve -
NSO Nuclear Station Operator
NSWP Nuclear Station Work Procedure
NPSH Net Positive Suction Head
NRC Nuclear Regulatory Commission
NRR Office of Nuclear Reactor Regulation
'NTS Nuclear Tracking System
PlF Problem identification Form-
PM: Preventive Maintenance
PMF- Probable Maximum Flood
PMT Post-Maintenance Testing
PSIA Pounds Per Square Inch Absolute
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LIST OF ACRONYMS USED (CONT)
PSIG Pounds Per Square Inch Gauge
QA -Quality Assurance
QC- Quality Control
QCAP Quad Cities Administrative Procedure
QCIS Quad Cit;es instrumen' Rurveillance
QCOP Quad Cities Operating Procedure
QCOS Quad Cities Operating Surveillance Procedure
- QSA- . Quality and Safety Assessment
-Q&SA Quality and Safety Assessment
-QTP Quad Cities Technical Procedure
RCIC _ Reactor Core Isolatio : Cooling
RG Regulatory Guide
-RHR Residual Heat Removal System
RHRSW Residual Heat Removal Service Water
RR Reactor Recirculation
S&L Sargent and Lundy
SAR' Safety Analysis Report
SE Safety Evaluation
SEP. Systematic Evaluation Program
- SIL Service informatior Letter
SOPl System Operational Performance inspection
SRB Safety Review Board
SROL Senior Reactor Operator
L SSFl Safety System Functional inspection
TAF -Top of Active Fuel
TALT Temporary Alteration
TS '- Technical Specification
TSUP Technical Specification Upgrade Project
UFSAR - Updated Final Safety Analysis Report
URI Unresolved item
US - Unit Supervisor
VIO Violation -
WR Work Request
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