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{{Adams | |||
| number = ML20197F286 | |||
| issue date = 12/19/1997 | |||
| title = Insp Repts 50-413/97-14 & 50-414/97-14 on 971012-1122. Violations Noted.Major Areas Inspected:Operations,Maint, Engineering & Plant Support | |||
| author name = | |||
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) | |||
| addressee name = | |||
| addressee affiliation = | |||
| docket = 05000413, 05000414 | |||
| license number = | |||
| contact person = | |||
| document report number = 50-413-97-14, 50-414-97-14, NUDOCS 9712300176 | |||
| package number = ML20197F239 | |||
| document type = INSPECTION REPORT, NRC-GENERATED, TEXT-INSPECTION & AUDIT & I&E CIRCULARS | |||
| page count = 34 | |||
}} | |||
See also: [[see also::IR 05000413/1997014]] | |||
=Text= | |||
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, | |||
. | |||
U.S. NUCLEAR REGULATORY COMMISSION | |||
REGION 11 | |||
Docket Nos: 50 413. 50 414 | |||
License Nos: NPt'-35. NPF 52 | |||
Report Nos.: 50 413/97 14. 50 414/97-14 | |||
Licensee: Duke Energy Corporation | |||
Facility: Catawba Nuclear Station. Units 1 and 2 | |||
' Location: 422 South Church Street | |||
Charlotte, NC 28242 | |||
Dates: October 12 - November 22. 1997 | |||
Inspectors: D. Roberts. Senior Resident Inspector | |||
R. Franovich, ResidMt Inspector | |||
M. Giles, Resident inspector (in Training) | |||
D.Forbes.RadiationSpecialist.RegionII(SectionsR1.2. | |||
R1.3. RI.4, R3.1. Re.l. and R8.1) | |||
Approved by: C. Ogle. Chief | |||
Reactor Projects Branch 1 | |||
Division of Reactor Projects | |||
Enclosure 3 | |||
?N | |||
G | |||
00b N y" | |||
. . | |||
EXECUTIVE SUMMARY | |||
Catawba Nuclear Station. Units 1 and 2 | |||
NRC Inspection Report 50-413/97-14, 50 414/97 14 | |||
This integrated inspection included aspects of licensee operations | |||
maintenance, engineering and plant support. Thereportcoversa5 week | |||
period of resident inspection. It also includes the results of an announced | |||
inspection by a regional radiation specialist. | |||
Doerations | |||
. In general, the conduct of operations was professional and safety | |||
conscious. (Section 01.1) 1 | |||
e A minor overpower excursion resulted in the 15 minute running average | |||
for reactor thermal | |||
an extended aeriod. power slightly | |||
The power exceeding | |||
excursion licensed power | |||
was contained limits for | |||
within criteria | |||
established Jy previous NRC guidance. (Section 01.2) | |||
e Control room o)erators failed to detect an extinguished 'DC Power On" | |||
light for the Unit 1 turbine-driven auxiliary feedwater pump for almost | |||
three days. The impact on Jump operability of the blown fuse which | |||
caused the extinguished 1191t. will be reviewed during closeout of the | |||
URI. (Section 01.3) | |||
. Operations personnel inappropriately entered the Technical Specification | |||
action statement more than an hour after a reactor trip system logic | |||
function failed to meet surveillance test acceptance criteria. However, | |||
the failed function was repaired, successfully retested, and returned to | |||
service before Technical Specification actions were required. (Section | |||
01.4) | |||
* Nuclear Systerr Directive 317 provided structure and delineated | |||
responsibilities for freeze protection. Proceduralized activities were | |||
initiated and completed in a timely manner, and work requests were | |||
initiated to resolve identified discrepancies. The licensee's efforts | |||
to effectively protect plant equipment and systems from freezing | |||
conditions improved since the previous cold weather season. (Section | |||
02.1) | |||
* Four unreltted non emergency events were reported to the NRC in | |||
accordance with Title 10 Code of Federal Regulations. Part 50.72 during | |||
the period. All of the events were properly reported with sufficient | |||
information provided. (Section 02.2) | |||
* Examples of poor performance were noted concerning activities | |||
surrounding the inappropriate tagout of a residual heat removal system | |||
miniflow valve during planned maintenance. (Section 04.1) | |||
. A deviation, with two examples, from NRC commitments was identified. | |||
Both examples involved administrativc errors resulting in commitments | |||
Enclosure 3 | |||
. | |||
_ . _ - . | |||
. . | |||
i | |||
2 | |||
being changed internally without proper notification of the NRC. | |||
(Sections 08.1 and 08.2) | |||
licintenance | |||
* Surveillance activities observed by the inspectors involved good | |||
workmanship, proper use of procedures, good radiological practices. and | |||
) roper management of Technical Specification action statements. (Section | |||
11.1) | |||
* New fuel movement activities to support the upcoming Unit I refueling | |||
outage were performed well. (Section M1.2) | |||
Enoineerinq | |||
* An unresolved item was identified concerning containment penetrations | |||
associated with stea < cupply lines to both units' turbine driven | |||
auxiliary feedwater ,o ms, which were not in compliance with Title 10 | |||
Code of federal Regulations. Part 50. Appendix A. General Design | |||
Criterion 57. The licensee had submitted an exemption request | |||
concerning this issue to the NRC during the previous inspection report | |||
period. (Section El.1) | |||
. Remote manual closure capability existed for dual function containment | |||
isolation valves; however, the action involved resetting the emergency | |||
diesel generator load sequencer. an action requiring further evaluation | |||
to be conducted under the above unresolved item. (Section El.1) | |||
. A non cited violation was identified concerning the use of aluminum | |||
separators in high efficiency particulate air 111ters located inside | |||
containment. (Section E8.2) | |||
ElantSunnort | |||
* An example of poor oerformance was identified related to a radiological | |||
control area boundary beinc) compromised. This minor discrepancy was | |||
immediately corrected by plant aersonnel and properly addressed by | |||
licensee management. (Section 11.1) | |||
. The licensee effectively imalemented a program for shipping radioactive | |||
materials required by the NRC and Department of Transportation | |||
regulations. (Section R1.2) | |||
. The licensee was meeting established goals for radioactive waste | |||
generated. Radiological facility conditions and housekeeping in | |||
radioactive waste storage areas were observed to be good. (Section | |||
R1.3) | |||
Enclosure 3 | |||
____- _-__ _ | |||
_ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ | |||
l | |||
. | |||
. | |||
3 l | |||
l | |||
* One violation was identified for failure to provide current dose rate i | |||
information on a radioactive inaterial label as required by Ti'.ie 10 Code | |||
of Federal Regulations. Part 20.1904(a). (Section Rl.3) | |||
* The licensee's water chemistry control program for monitoring primary | |||
and secondary water quality had been implemented for those parameters | |||
reviewed. in accordance with the Technical Specification requirements | |||
and the Station Chemistry Manual for pressurized water reactor water | |||
chemistry. (Section RI.4) | |||
. The licensee had properly implemented procedures to maintain an | |||
effective program to monitor and control liquid and gaseous radioactive | |||
effluents to limit doses to members o the public. Theprojected | |||
offsite doses resulting from those effluents were well within the limits | |||
specified in the Technical Specifications, the Offsite Dose Calculation | |||
Manual. and Title 40 Code of Federal Regulations. Part 190. (Section | |||
R3.1) | |||
. The licensee was effectively conducting formal radiation protection and | |||
chemistry audits as required by Technical Specifications and was | |||
completing corrective actions in a timely manner. (Section R7.1) | |||
. Tha Emergency Jperations facility located in downtown Charlotte. North | |||
Carolina and its associated equipment were in good repair and condition. | |||
Energency communication and plant computer equipment in the Technical | |||
Support Center was in good working order. (Section P2,1) | |||
Enclosure 3 | |||
_ _ _ _ - _ _ _ ____ . _ _ _ _ __ _ _ ____ _ _ _ ___ | |||
. . | |||
Renort Details ; | |||
; | |||
Sumary of Plant Status | |||
Unit 1 operated at or near 100% )ower until November 21, when it began its | |||
end of-cycle 10 coast down for 11e upcoming refueling outage. The unit ended | |||
the inspection period at 98 percent power, | |||
i | |||
Unit 2 operated at or near 100 percent power until October 20, when a power | |||
redu: tion was initiated to comply with Technical Specification (TS) 3.6.3 :' | |||
following a nitrogen leak associated with the accumulator for main feedwater * | |||
isolation valve 2CF 33. Power was reduced to approximately 15 percent power, | |||
at which the time the valve was gagged shut a.1d repairs commenced. _Upon , | |||
completion of the leak repair and valve post naintenance testing activities. ' | |||
the unit was returned to 100 percent power on October 21. On November 21. a , | |||
' power reduction to 50 percent was initiated to allow a control circuit card | |||
associated with main turbine control valve C h1 to be replaced. Licensee , | |||
personnel also replaced a solenoid valve ano cleaned instrument air lines | |||
associated with main generator power circuit breaker (PCB) 28. These | |||
* | |||
activities were completed on November 22 and power was increased to 97 percent ! | |||
by the end of the inspection period. ; | |||
Review of Vodated Final Safety Analysis Reoort (UFSAR) Commitments | |||
While performing inspections discussed in this repart, the inspectors reviewed , | |||
the applicable portions of the UFSAR that were related to the areas inspected. + | |||
The inspectors verified that the UFSAR wording was consistent with the | |||
observed plant practices, procedures, and parameters. | |||
I. Operations | |||
01 Conduct of Operations | |||
01.1 General Comments (71707) | |||
The inspectors conducted frequent control room tours to verify proper | |||
staffing, operator attentiveness and communications, and adherence to | |||
approved procedures. The inspectors attended opf.ations turnovers and | |||
site direction meetings to maintain awareness of overall plant | |||
operations. Operator logs were reviewed to verify operational safety | |||
and compliance with TS. Instrumentation, computer indications, and | |||
safety system lineups were periodically reviewed from the control room | |||
to assess o)erability. Plant tours were conducted to observe equipment > | |||
status and lousekeeping. Problem Identification Process (PIP) reports | |||
were routinely reviewed to ensure that potential safety concerns and | |||
equipment problems were reported and resolved. | |||
In general, the conduct of operations was professional and safety- . | |||
conscious. .The Unit 2 power reduction associated with feedwater | |||
isolation valve 2CF-33 was conducted safely. Good plant equipment | |||
material conditions and housekeeping were noted throughout the report | |||
Enclosure 3 | |||
. | |||
, w----e-,-._..r _-m,-.e-uw- ..v- w - 4me-v y -.i.- eye ._ ,. ,v,m, | |||
. .- - - - - . . . - - - . | |||
, | |||
1 | |||
l | |||
. . | |||
l | |||
: | |||
2 1 | |||
, | |||
period. However as addressed below, several human performance related ! | |||
deficiencies wele identified. | |||
01.2 H1nor Excursion Over Licensed Power Limits for Unit 1 | |||
a. .lmeection Scone (71707) | |||
The inspectors reviewed the circumstances associated with a minor power i | |||
excuision on Unit 1. f | |||
b. Observations and findinas | |||
P | |||
On October 21, 1997. the inspector noted during a review of control room | |||
' | |||
logs that the Unit 1.15 minute running average for reactor power, as , | |||
indicated by the Operator Aid Computer (OAC), had exceeded 100 percent. | |||
< | |||
The Unit 1 operator noticed this at 3:39 a.m. and reduced turbine load | |||
by 5 megawatts and inserted control rodt. two steps to bring power aelow | |||
100 percent. Operations personnel later generated station PIP l-C97- | |||
3382 to document and investigate the power excursion. | |||
The inspectors reviewed the PIP and noted that the Unit 1 OAC 15 minute : | |||
, | |||
average was stated as having been in alarm for 15 minutes. The | |||
inspectors reviewed 0AC trend reports for reactor power and noted that . | |||
the maximum instantaneous reactor power level, according to secondary | |||
heat balance best estimates (computer point C1P1445), was ap3roximately 1 | |||
100.6 percent recorded just before 3:15 a.m. According to tie trend | |||
report.powercontinuallyspikedbetween99.7and100.3percentpower | |||
for the next 20 25 minutes before operators noticed the 15 minute | |||
average and reduced power. Computer trends indicated that the 15 minute | |||
average | |||
However,it peaked at 100.05 | |||
never .'eached the percent | |||
alarm setand was | |||
point in for yercent). | |||
(100.1 about 20 minutes. | |||
Further 1 | |||
discussions with plant personnel and review of alarm listory data | |||
indicated that the statement in the PIP concerning the alarm being in | |||
for 15 minutes was in error. Later, this :tatement was corrected in the | |||
PIP documentation. ' | |||
Further investigation by the inspectors determined that routine reactor | |||
cooldnt system Doron dilutions were )erformed earlier in the shift. | |||
However, this was last done 2 hours aefore the noted power excursion. | |||
The inspector interviewed control room personnel who indicated that | |||
several activities were occurring at the time of the minor over power. | |||
including those associated with a Unit.2 down power (see Section 08.2 of ' | |||
this report). The operator indicated that these activities may have | |||
been a distraction and | |||
minute average earlier,possibly prevented him from noticing the 15- i | |||
Discussions with operators 31d plant management indicated that operators | |||
were expected to maintair eactor p& ar at licensed power levels. | |||
Operators were expected to contiv.' ly monitor power and immediately | |||
take actions to keep it within i yt. Plant management discussed this- - | |||
Enclosure 3 | |||
__ __ _ . . _ _ _ __ _ _. _ _ _ | |||
-_ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ - _ | |||
. . | |||
3 | |||
l | |||
excursion with those personnel involved in the event and emphasized the | |||
need for iiereased diligence when monitoring power levels. | |||
The inspectors reviewed NRC guidance on minor power excursions and noted | |||
that the power level did not exceed previously established criteria, | |||
c. Conclusion | |||
The inspectors concluded from their review that the )ower excursion was | |||
minor and was contained within criteria established )y previous NRC | |||
. guidance. | |||
01.3 Control Power Unavailable to the Unit 1 Turbine Oriven Auxiliarv | |||
Mger (Af W) Pumo Irio And lhrottle Valve | |||
a. Insocction Scone C/110D. | |||
The inspectors reviewed the circumstances associated with a loss of | |||
control power to the Unit 1 turbine driven AFW pump trip and throttle | |||
valve. | |||
b. Observations and Findinni | |||
During a control room tour on November 17, the inspectors noted that the | |||
~0C (Direct Current) Power On" li | |||
4 driven AFW pump was extinguished.ght associated | |||
The inspectors with | |||
informed the the Unit | |||
o)erator 1 turb | |||
at the controls of this observation. The operator replaced the wio and | |||
the light was still not lit. The inspectors mentioned that they had | |||
3reviously observed the light to be out 3 days earlier on November 14 | |||
)ut had assumed then that the extinguished light was related to ongoing | |||
maintenance involving a 72-hour LC0 on the system. Theoperatorstated | |||
that this ? ulb was in the control circuit for the AFW pump turbine trip | |||
and throttle valve. Subsequent licensee troubleshooting determined that | |||
fuse FU-2 in control panel 1ELCP0245 was blown. A review of several | |||
electrical drawings indicated that control aower and electrical | |||
overspeed trip functions for the trip and tirottle valve were powered | |||
through this fuse. The trip and throttle valve and the turbine driven | |||
AFW pump were declared inoperable shortly after 10:00 a.m. and the fuse | |||
was subsequently replaced. | |||
The inspectors discussed aspects of this incident with plant personnel | |||
to determine whether operators may have missed opportunities to identify | |||
the deficiency earlier, and to determine the true impact of the blown | |||
fuse on the system's capability to perform its safety functions. The | |||
inspectors noted that t1e blown fuse also caused control power | |||
indication to be extinguished at a local control panel, and that if the | |||
fuse was indeed blown for more than 3 days, plant personnel may have | |||
missed additional opportunities to identify a problem while on-field | |||
tours. The ins)ector noted that there were no formal checks in licensee | |||
procedures of tle *DC Power On" light in the control room.. There were | |||
Enclosure 3 | |||
_ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ ____ _ _ ____ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ | |||
l | |||
. . | |||
4 i | |||
: | |||
also no control room alarms indicating control power unavailability for | |||
the trip and throttle valve. | |||
Engineering personnel were still evaluating whether or not losing | |||
control power or the electrical overspeed trip function rendered the , | |||
sy. tem inoperable. According to Section 20.4.1.1. " Auxiliary feedwater , | |||
Pump Turbine," of s)ecification CNS-1593.SA-00 0001. Design Basis | |||
Specification for tie Main Steam to Auxiliary Equipment System (SA) and , | |||
feedwater Pump Turbine Exhaust System (TE). Revision 11: at least one of | |||
the overspeed trip devices (mechanical or electrical) must be operable < | |||
for the turbine driven auxiliary feedwater pump to be operable. The t' | |||
mechanical overspeed trip function was not affected by the blown fuse. | |||
The inspectors concluded that further review of this incioent and its ' | |||
impact on the turbine driven AFW pump was necessary. Pending further - | |||
NRC review, this item is characterized as Unresolved item (URI) 50- | |||
413/97-14 01: Control Power Unavailable to the Unit 1 Turbine Driven i | |||
AFW Pump's Trip and Throttle Valve. | |||
c. Conclusion | |||
Control room o)erators failed to detect an extinguished *DC Power On" | |||
light for the Jnit 1 turbine driven AFW pump for more than three days. | |||
The impact of the blown fuse on pump operdbility will be reviewed during | |||
closeout of the URI. t | |||
01.4 Hanaaement of Technical Soecification (TS) Limitina conditions for | |||
Operation | |||
' | |||
a. InspectionStone(71707) | |||
During a surveillance test of the Unit 2 reactor trip system ; | |||
instrumentation on October 10. 1997, a problem associated with the | |||
overpower differential temperature (0PDT) reactor trip logic was | |||
identified. The inspector discussed the test failure with operations | |||
shift i | |||
3271, personnel, read the associated TS, and reviewed station PIP 2 C97- | |||
b. Observations and Findinas | |||
During the performance of IP/2/A/3200/002A Solid State Protection | |||
System (SSPa) Train A Periodic Testing, Revision 21. on October 10. | |||
1997, the OPDT reactor trip logic test acceptance criterion was not met. | |||
A red lamp illuminated to indicate that a malfunction of the logic | |||
testing was detected (a green lamp would have illuminated if the logic | |||
test had been acceptable). The surveillance test began at 9:56 a.m., | |||
and the failure was identified some time before noon. Test technicians | |||
backed out of the test, and the reactor trip system was removed from the | |||
TS Action item List at 12:10 p.m. Engineering aersonnel were_ involved | |||
to assist operations personnel in determining tie extent of the | |||
operability concern (i.e., was the problem limited to OPDT trip logic or | |||
Enclosure 3 - | |||
- ._ - | |||
- _ | |||
. _ ___ - ._ _ _ . _ _ _ _ _ _ _ _ | |||
' | |||
, . | |||
f | |||
5 | |||
, | |||
' | |||
did it affect all of the solid state protection system). Engineerin | |||
personnel concluded that the problem was limited to the OPDT trip lo ic | |||
and communicated their conclusion to operations per.vnnel at around :00 | |||
p.m. The A train of Automatic Trip and Interlo: 's ' unctional Unit | |||
19 of TS 3.3.1. Table 3.3-1) was declared inopc'e 30 p.m. | |||
placing the unit in a six hour action statement .. n the function | |||
or be in Hot Standby (Mode 3) in the following six v v | |||
The inspectors questioned operaticns shift personnel about the decision | |||
to enter the required action at 1:00 p.m. rather than when the OPDT | |||
reactor trip logic ttst failure occurred. The response was that | |||
engineering involvement was needed to determine the scope of the )roblem | |||
(and inoperability) so that the appropriate TS action could be tacen. | |||
' | |||
Af ter the inspectors discussed the issue with the operations shift | |||
personnel, they recognized that determining the scope of the | |||
inoperability was independent of the time after which actions were | |||
required. . | |||
Engineering personnel determined that a failed circuit card caused the | |||
test failure. The circuit card was replaced, and testing was com)leted | |||
successfully. The action statement was terminated at 4:30 p.m. tlat | |||
same day. | |||
c. Conclusions | |||
The inspectors concluded that operations personnel inappro)riately | |||
entered the TS action statement more than one hour after t1e test | |||
failure of a reactor trip system logic function. The failed function | |||
was repaired, successfully retested and returned to service before TS | |||
: actions were required. | |||
02 Operational Status of Facilities and Equipment | |||
02.1 Cold Weather Protection Preoarations | |||
a. Insnection Scone (71714) | |||
.The inspectors reviewed Nuclear System Directive (NSD) 317. Freeze | |||
fruiection Program. Revision 1: interviewed the freeze protection | |||
coordinator: reviewed procedures and work orders to determine what | |||
actions had been taken to prepare for cold weather; and independently- | |||
inspected some vulnerable equipment exposed to the environment for | |||
freeze protection, | |||
b. Observations and Findinas | |||
The licensee completed NSD 317 in March 1997. The NSD governs the | |||
freeze protection alans at all three Duke nuclear stations. During the | |||
previous cold weatler season, the NSD had not been finalized and a | |||
formal program was not in place for ensuring that effective measures | |||
" | |||
Enclosure 3 | |||
- | |||
. __. _ _ _ _ -___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ | |||
. . | |||
6 : | |||
were in place to protect plant equipment and systems from sub freezing j | |||
conditions. | |||
i | |||
The station assigned a freeze protection coordinator to monitor the | |||
status of preparation activities. An equipment freeze protection , | |||
program was developed to identify operating plant systems, structures ! | |||
and components (SSCs) that may be subjected to freezing temperatures | |||
during the cold weather season. An engineering support program was , | |||
initiated to ensure that specific freeze protection measures for | |||
vulnerable SSCs were identified to facilitate the preparation and | |||
completion of a pre-seasonal eneckout. Pre seasonal checkouts were ' | |||
executed via various model work orders for inspection and testing of | |||
electrical heat trace and instrument box heaters. The freeze protection | |||
plan includes surveillance procedures to inspect SSCs considered to be | |||
critical to plant operation on a monthly interval and as necessary | |||
during extreme cold weather. | |||
The inspectors discussed the status of freeze protection preparations | |||
.with the freeze protection coordinator. According to the coordinator, | |||
the annual preventive maintenance activities had been completed by the | |||
end of the inspection report period, and work-orders or work requests - | |||
had been generated to address identificd discrepancies. The freeze . | |||
protection coordinator had performed inspections of vulnerable areas and | |||
' | |||
submitted a list of discrepancies to the maintenance orcanization. Most | |||
ofthesediscrepancieswereresolvedbytheendoftheInspection , | |||
period. | |||
The inspectors conducted inspections of equipment that historically had | |||
been vulnerable to cold or freezing temperatures. The inspectors > | |||
notified the freeze protection coordinator of a few minor discrepancies. | |||
The inspectors also reviewed the work orders associated with the annual , | |||
preventive maintenance (PM) and verified that work had been completed. | |||
c. Conclusions | |||
Nuclear System Directive 317 provided structure and delineated | |||
responsibilities for freeze protection. Proceduralized activities were | |||
initiated and completed in a timely manner, and work orders or work | |||
requests were initiated to resolve identified discrepancies. The | |||
inspector concluded that the licensee's efforts to effectively protect | |||
plant equipment and systems from freezing conditions had improved since | |||
the previous cold weather season. | |||
02.2 Prompt Onsite Response to Events (93702) | |||
. | |||
The licensee reported four unrelated events to the NRC Headquarters | |||
- | |||
Operations Officer via the Emergency Notification System in accordance | |||
with 10 CFR 50.72. The following events were all reported in a timely | |||
fashion with sufficient information being provided, | |||
Enclosure 3 | |||
. - . . - . - .. -- .- . _- | |||
-- _. - -- | |||
- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. - _ _ . _ _ _ | |||
h | |||
. . ; | |||
! | |||
! | |||
7 ! | |||
Oil Sheen on Lake Wylie on October 15 i | |||
! | |||
On October 15. the inspectors were notified of a thin oil sheen that was | |||
discovered on Lake Wylie during a main fire pump test. The source of : | |||
the oil was determined to be an overflowing pump bearing reservoir which | |||
caused oil to spill around the fire pump motor and eventually into the | |||
lake. The oil sheen was contained by a boom beneath the pum) structure. i | |||
The licensee notified the South Carolina Department of Healt1 and | |||
Environmental Controls and the National Response Center, which in turn | |||
required notification of the NRC 'q accordance with 10 CFR | |||
50.72(b)(2)(vi). | |||
Plant Shutdown Reauired By TS on October 20 | |||
As discussed in Section 08.2 of this report. the licensee initiated a > | |||
' | |||
Unit 2 shutdown on October 20 when it entered TS Limiting Condition for | |||
Operation 3.6.3 action statement following the inoperability of the 2A , | |||
' | |||
steam generator main feedwater isolation valve 2CF-33. The unit was | |||
helti at 15 percent power after the valve was deactivated and gagged | |||
shut. The valve was repaired and a forced unit shutdown was avoided. | |||
This item was reported to the NRC in accordance with 10 CFR | |||
50.72(b)(1)(1)(A). | |||
' | |||
)otential Non Conservatism in a Calculation used to Distinauist Between | |||
Reactor Coolant System F' ow Versus leactor Power Restricted anc | |||
)rohibited 02eratino Rea1ons | |||
On October 23. the licensee reported a potential nonconservatism in each ! | |||
units' 15 3/4.2.5. Departure from Nucleate Boiling (DNB) Parameters. | |||
Figure 3.2 1. Reactor Coolant System Total Flow Rate Versus Rated | |||
Thermal Power - Four Loops in Operation. Essentially, licensee | |||
personnel determined that the curve provided in Figure 3.21 for each | |||
unit permitted potential plant operation at reduced power levels with | |||
reactor coolant system flow rates that could possibly challenge DNB | |||
ratio design limits for certain analyzed transients. As a precaution, | |||
until this condition could be resolved, the licensee implemented | |||
administrative restrictions requiring reactor coolant system flow rates | |||
to be maintained above those specified as the permissible operation | |||
region for 100 percent power. These restrictions were verified to be in - | |||
place by the resident inspectors. Long term corrective actions included | |||
completing an analysis to allow a revision to the TS requirements to | |||
eliminate the non-conservatism. This item was reported in accordance | |||
with 10 CFR 50.72(b)(2)(111)(D). The licensee documented this issue in | |||
a 30 day written follow up Licensee Event Re) ort (LER 50-413/97-007) | |||
near the end of the inspection period. Furtier inspector review of this | |||
issue will.be conducted and tracked under the LER in subsequent' | |||
: inspection reports. | |||
Enclosure 3 | |||
_~ , | |||
_ _ _ _ _ _ . _ - _ _ _ - - _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ | |||
. . | |||
8 | |||
i | |||
Potential for Overfilli.Da Steam Generator Durina a Postulated Accident | |||
' | |||
On November 18. the licensee re)orted a single failure vulnerability | |||
involving the loss of 125 Volt )C vital instrument and control | |||
distribution center EDE or ELF during a postulated steam generator tube | |||
rupture event coincident with a loss of offsite power. The licensee | |||
determined, following a detailed analysis that the loss of either of | |||
these busses would result in the inability to isolate turbine driven | |||
auxiliary feedwater pump flow to a ruptured steam generator. The steam | |||
generator would be pntentially overfilled, resulting in uncontrolled | |||
releases of radioactivity to the atmosphere. | |||
Because of this potential, and until further corrective actions are | |||
determined, the licensee implemented conservative administrative | |||
controls limiting the amount of dose equivalent iodine in the reactor | |||
coolant system to ensure the consequences of the Chapter 15 steam | |||
generator tube rupture analysis remain bounding. These restrictions | |||
were contained in procedure CMP 3.4.17.1. Primary Chemistry. Revision 28 | |||
and verified by the inspectors. At the close of the inspection period, | |||
the licensee was evaluating several o)tions for long-term corrective | |||
actions. This item was reported to tie NRC in accordance with 10 CFR | |||
50.72(b)(1)(11)(B). | |||
04 Operator Knowledge and Performance | |||
04.1 Residual Heat Removal (RHR) System Potentially Placed in An Unanalyzed | |||
Condition | |||
a. Insoection Scope (71707) | |||
The inspectors reviewed the circumstances involving an August 20. 1997, | |||
tagout in which the RHR system was potentially ) laced in an unanalyzed | |||
condition. The inspector reviewed the Catawba Jesign Basis Document | |||
(DBD) CNS 1561.ND 00-0001: the UFSAR. Section 6.3 and Chapter 15: and | |||
PIP 2-C97 2722. The inspectors also ruiewed the licensee's root cause | |||
investigation, completed during this inspection period. and discussed | |||
this issue with engineering and operations personnel. | |||
b. Observations and Findings | |||
Residual heat removal system valve ND59B 1s a motor 0)erated globe valve | |||
located in the minimum flow lines of the 18 and 2B RH1 pumps. Valve | |||
N059B and its associated miniflow line normally protect either B train | |||
pump from cavitation at low flow conditions or following a complete loss | |||
of suction during the decay heat removc1 or emergency core cooling modes | |||
of operation. | |||
On August 20. 1997, at 3:38 a.m. operations issued removal and | |||
restoration (R&R) tagout 27-1498 to support work on the Unit 2 Train B | |||
RHR miniflow loop. Unit 2 train B RHR was declared inoperable and | |||
Enclosure 3 | |||
._ | |||
- . | |||
_ | |||
l . | |||
. | |||
9 | |||
entered into the Technical Specification Action Item Log (TSAIL). The | |||
planned work included a miniflow valve controlling set point | |||
modification, a gauge replacement, and an instrument calibration. The | |||
R&R tagged valve ND59B open with power removed. Approximctely 2-1/2 | |||
hours later at 6:00 a.m., work control personnel realized that the | |||
tagout was in conflict with Catawba Design Basis and Criteria, | |||
Specification CNS-1561.ND-00-0001. Revision 5, which stated that "with | |||
ND59B stuck open and incapable of closing, the resulting diversion of | |||
RHR ) ump fluid to the recirculation loop is an unanalyzed condition." | |||
At tiat time, operations personnel cleared the tagout and closed the | |||
valve. Station PIP 2-C97-2722 was initiated and the licensee later | |||
determined that a past operability evaluation was required. | |||
Engineering | |||
September 1997, is, personnel | |||
and concluded completedthat the the | |||
RHRpast | |||
systemo)erability evaluation | |||
was operable during on | |||
the time the miniflow valve was tagged open. The inspectors discussed | |||
this conclusion with licensee personnel and upon reviewing UFSAR Table | |||
6-7, Catawba Nuc1 car Station Emergency Core Cooling System Flow Rates, | |||
arrived at the same conclusion. This was based on the fact that the | |||
RdR flow capacity (approximately 500 gallons per minute) normally | |||
diverted from the reactor coolant system recirculation loop by miniflow | |||
valve ND59B, when subtracted from the total RHR flow ca)acity, still | |||
resulted in sufficient RHR flow being delivered to the RCS during the | |||
post-accident recirculation mode. However, the inspectors considered | |||
the tagging ciscrepancy to represent a problem that could have had | |||
adverse plant impact. | |||
A root cause investigation of the improper tagging incident was | |||
completed by the licensee during this inspection period which concluded | |||
that engineering persornel improperly communicated a 1993 DBD revision | |||
to affected groups. The RHR DBD had been revised then to provide a | |||
discussion of the "unanalyzed condition." However, this analysis did | |||
not take into consideration lesser flow requirements assumed in UFSAR | |||
Table 6-7 for the post-accident long-term recirculation mode of | |||
operation, the time at which the RHR system alignment would be changed | |||
and the miniflow valve would become a diversion flow path. | |||
The inspectors considered other human performance weaknesses contributed | |||
tc the tagging error. When the calibration work order from which the | |||
tagout was generated (PM 95054445) was developed in July 1995, a note | |||
was added for operations personnel to tag the tr niflow valve open. i | |||
Although personnel involved in planning the set point change | |||
modification were aware of the DBD statement, and verbiage was included | |||
in the modification package to ensure the tagout was correct and would | |||
not place the RHR system in an unanalyzed condition, the set point | |||
modification was performed under an existing tagout for the preventive | |||
maintenance work, which had the valve opened on August 20. | |||
The inspectors noted that the DBD had not been consulted when the tagout | |||
associated with the August 20, 1997, activities was developed a week | |||
Enclosure 3 I | |||
i | |||
_ _- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ | |||
i . . | |||
10 | |||
earlier. The inspectors, discussed this with licensee management, who | |||
stated that the 1995 PM work order note likely contributed to an | |||
overr'. ding operating philosophy that tagging the valve open during | |||
maintenance was appropriate. Several corrective actions were generated | |||
for PIP-2-C97-2722, including developing a policy for communicating | |||
engineerits document revisions to affected groups and designating a | |||
specific work management mtem panel to document engineering | |||
recommendations and spe notes. The inspector asked whether or not | |||
the DBD reference to t b analyzed condition" would be deleted to | |||
reflect the engineering e.idlysis discussed above. Licensee management | |||
indicated they would evaluate changing the DBD. | |||
c. Conclusions | |||
The inspectors concluded that although having the Unit 2 train B RHR | |||
pump out of service with valve ND598 de-energized open did not place the | |||
plent in an unanalyzed condition, examples of poor performance were | |||
identified concerning activities leading up to the valve inappropriately | |||
being tagged open during plaaned maintenance. | |||
08 Miscellaneous Operations Issue (92901) | |||
08.1 (Closed) LER 50-414/95-01: Reactor Trip Due to Closure of a Main Steam | |||
isolation Valve | |||
The event described in this LER involved an automatic reactor trip due | |||
to the failure of a digital optical isolator (D01) in the B main steam | |||
isolation valve control circuit that caused the valve to close. This | |||
LEP was discussed in NRC Inspection Report 50-413.414/97-12 and remained | |||
opea pending further NRC review, | |||
Planned corrective action 2 was to develop a PM program to periodically | |||
monitor continuously enc gized E-max 00ls with model numbers 175C156 and | |||
175C157 in critical applications. Instead, the licensee initiated a PM | |||
program to re) lace DOIs that perform a control function and that have AC | |||
voltage for t1eir inaut )ower su) ply every twelve years. The inspectors | |||
determined that the 4RC 1ad not )een apprised of the change. | |||
In light of recent DOI failures that resulted in manual reactor trips in | |||
July and August 1997, the inspectars asked the licensee if monitcring | |||
the D01s could have revealed the root cause (degraded resistors) of the | |||
r eent DOI failures. The licensee indicated that the test methodology | |||
that would have been used to periodically monitor the D01s would not | |||
have revealed degraded resistors (the cause of the 1997 failures). The | |||
inspectors concluded that, while testing the DOIs had the potential to | |||
reveal degraded D01s during periodic testing, the likelihood that it | |||
would have done so was low. Therefore, tl. commitment change did not | |||
substantially reduce the opportunity to identify degraded DOI resistor:: | |||
and take subsequent actions to prevent the 1997 001 failures. | |||
Enclosure 3 | |||
i | |||
l | |||
- _ _ _ _ _ _ _ _ _ _ . | |||
__ - _ - - - - _ --_ - _____-___ - __ - - _ ___ | |||
-i | |||
_ | |||
.- -.- | |||
:, | |||
11- . | |||
According to Nuclear System Directive (NSD) 214. Comitment Management: | |||
ts are a source of NRC- | |||
EProgram Revision | |||
comitments. 2.-214.8.4i | |||
Section Licensee Event | |||
Remove or-Rep" Change a.Comitment. stated .' | |||
:that the regulatory compliance-(RGC) group should be notified if- a | |||
comitment change is needed, and that RGC will determine, in part. if | |||
' | |||
- | |||
the NRC- should be notified.- The NSD'incor> orates a 1994 draft document | |||
prepared by the Nuclear Energy Institute (4EI). entitled " Guideline for | |||
Managing NRC Commitments." | |||
- | |||
. | |||
Acc0rding to NSD 214, when a comitment is changed, the original | |||
comitment-will be modified with a description of the change in the | |||
appropriate section of the PIP database (which is used to track NRC | |||
comitments-to resolution). The NSD further stated that if the change ; | |||
is determined to be significant enough, a new commitment may be- | |||
generated. However, proper cross-references shall be provided to link | |||
-the original commitment to the revised commitment. The licensee | |||
determined that the NRC was not apprised of the commitment change | |||
' | |||
because the corrective actions representing the comitment were | |||
improperly cross-referenced. As a result, the changed corrective action | |||
was not identified as an-NRC commitment, and RGC was not notified.- The | |||
inspector concluded that the licensee failed to notify the NRC of a | |||
- | |||
comitment change regarding planned corrective actions delineated in LER | |||
50-414/95-01. This issue is charact rized as one example of Deviation | |||
50-413.414/97-14-02: Changing NRC Comitments Without Properly | |||
Notifying the NRC.o , | |||
4 | |||
' | |||
This item is closed. | |||
. | |||
08.2 (Ocen) Violation R0-413/97-08-01: Inadequa'.e Alarm Response Results in | |||
Inadequate and untimely Correctite Actions for Valve Operability | |||
' Determination | |||
.The inspectors reviewed Violation 50-413/97-08-01 for an April- 3.1997. | |||
. | |||
-incident following a similar occurrence on October 20, 1997. where a | |||
feedwater isolation valve became inoperable after a nitrogen leak | |||
developed on its accumulator. | |||
On the morning of October 20, 1997. just before shift turnover, the Unit | |||
2-control room operators received a computer alarm indicating low-- | |||
nitrogen gas pressure in the accumulator associated with the 2A steam | |||
, | |||
generator main feedwater isolation' valve. 2CF-33. The valve was- | |||
declared inoperable and TS Limiting Condition for Operation (LCO) 3.6.3 | |||
was imediately entered. Nitrogen pressure was checked and found to be | |||
at 1640.psig, which was below the low operability limit of 2050 psig. | |||
The' accumulator-was recharged to 2760 psig and the TS'LC0 was exited. | |||
- Approximately 2-3 hours'later at 9:55 a.m., another low aressure alarm | |||
~ | |||
- | |||
was received, ard operators again entered the 4-hour TS _C0 action- | |||
requirement to either return the valve to operable status, de-energize | |||
(gag) it shut, or initiate plans to be in Hot Standby in the following 6 | |||
hours. After the second alarm. the accumulator was found to be at 1810 | |||
Enclosure 3 | |||
_ _ _ _ , _ _ . - - _ _ .. _ _ | |||
. _ . _ - _ _ _ _ _ _ _ _ _ _ _ | |||
_ _ _ _ . _ _ _ . _ _ _ | |||
.., . | |||
] . | |||
, | |||
12 | |||
spsig and a-leak was detected from a solenoid valve at the actuator. The i | |||
nitrogen: accumulator was1again recharged but could not be maintained , | |||
above the low pressure limit. Technical Specification 3.6.3 required ! | |||
that= the plant to be in Hot Standby (Mode _3) by _7:55 p.m. l | |||
Plant management decided that a power reductior, would be initiated- | |||
: shortly after 1:00 ).m.- .The unit was reduced to approximately 15 | |||
rpercent power and t1e valve was gagged shut just before the TS LCO | |||
Laction to be in Hot Standby was required, thus avoiding a forced. | |||
shutdown. A leaking 0 ring at a solenoid-to tube connection was | |||
detected. The solenoid was re) laced and the valve was tested | |||
successfully. Unit 2 exited tie LCO action statement and was returned- | |||
. to 100 percent-power on October 21. | |||
- | |||
. | |||
e | |||
' | |||
The inspectors reviewed Violation 50-413/97-08-01 which documented a. . | |||
timilar occurrence on April 3.1997. Involving feedwater isolation valve ) | |||
1CF-51. Following the April 3. 1997, incident. plant personnel | |||
determined that the control room 0AC alarm set point was set at or near | |||
- the pressure at which the valve became inoperable. One of the planned | |||
corrective actions-documented in-the licensee's written response to the. | |||
violation: dated July 22, 1997. was for engineering personnel to evaluate | |||
whether'the alarm set-point could be raised to provide more margin | |||
- | |||
between it and the operability limit thereby allowing operators more < | |||
time to react to an actuator leak. According to the licensee's letter. | |||
>' ~ | |||
4 | |||
this action was to be completed by September 30, 1997. Following the. J | |||
'' - | |||
.0ctober 20 ;1997.: occurrence, the inspectors inquired about the status - i | |||
i ' | |||
-of the engineering evaluation. Licensee personnel indicated that it had | |||
not been performed and that engineering personnel had been internally | |||
granted an extension of the due date from the safety assurance group to | |||
October 31. | |||
The inspectors noted that the NRC had not been notified of this | |||
commitment change and upon inquiring furt'ner, were told that an | |||
administrative error in the data-entry process for the PIP associated | |||
with the- April 3.1997, incident allowed engineering to be granted an | |||
extension without evaluating the impact of changing this commitment. | |||
Upon discovery of the error, licensee personnel corrected it in the PIP | |||
database and an engineering evaluation was completed by the new | |||
; deadline. A modification was subsequently initiated to raise the | |||
-accumulator alarm set points for-all of the feedwater isolation valves | |||
and provide greater margin above their operability' limits. | |||
The inspectors determined that the failure to perform the engineering | |||
evaluation in a timely manner further increased the chances of a | |||
feedwater isolation valve becoming inoperable prior to the control room | |||
: receiving the alarm. 'The inspettors reviewed the documents associated | |||
- | |||
with NRC commitment management programs described in Section 08.1 above | |||
and determined that the failure to perform this evaluation by September | |||
30. 1997. constituted a Deviation from NRC commitments. This issue is- | |||
- | |||
characterized-as the second example of Deviation 50-413.414/97-14-02: | |||
Enclosure 3 | |||
0 | |||
= .. | |||
_ _ _ _ _ _ _ _ _ - _ _ | |||
_ . _ _ _ | |||
. . | |||
13 | |||
Changing NRC Commitments Without Properly Notifying the NRC. Violatir | |||
50-413/97-08-01 will remain open pending completion of all of the | |||
licensee's corrective actions and further review by the inspectors. | |||
Maintenance | |||
M1 Conduct of Maintenance | |||
[ M1.1 General Comments (61726) | |||
The inspectors observed portions of the fol h ing surveillance and | |||
inspection activities: | |||
. NPP-312. Nuclear Fuel And Core Component Receipt Inspections. | |||
. PT/1/A/4200/09A. Auxiliary Safeguards Test Cabinet Periodic Test. | |||
* PT/1/A/4400/06A. Nuclear Spray (NS) Heat Exchanger 1A Heat | |||
Capacity Test. | |||
. PT/1/A/4400/09. Cooling Water Flow Monitoring For Asiatic Clams | |||
And Mussels Quarte'ly Test. | |||
. PT/1/A/4200/04B. Containment Spray Pump 1A Performance Test. | |||
. PT/1/A/4350/0028. Diesel Generator 18 Operability Test, | |||
s Retype No. 28 | |||
\ During these activities. the ins ectors noted proper use of procedures, | |||
properly calibrated measuring and test equipment effective radiological | |||
controls, and adequate communication between personnel performing the | |||
tests. | |||
M1.2 New Fuel Movements (62707) | |||
The inspector observed movement of new fuel from the dry storage racks | |||
to the spent fuel pool in pre]aration for the upcoming Unit 1 end-of- | |||
cycle 10 refueling outage. T11s activity was conducted under Work Order | |||
97063472-01. Move New Fuel from New Fuel Vault to Spent Fuel Pool. The | |||
technicians used procedures OP/1/A/6550/011. Retype 21. Internal | |||
Transfer of Fuel Assemblies and Components: and OP/1/A/6550/006. Retype | |||
11. Transferring Fuel with the Spent Fuel Manipulator Crane. The | |||
inspector noted. for the fuel assemblies observed, that they were placed | |||
correctly in locations referenced by the procedure attachment. Proper | |||
radiological controls were observed. Crane chNklist prerequisites had | |||
been completed as required. This work activity was conducted well. | |||
M8 Miscellaneous Maintenance Issues (92902) | |||
M8.1 (Closed) Inspector Follow UD Item (IFI) 50-413.414/97-08-04: | |||
Reportability of Nuclear Service Water (NSW) System Actuations. | |||
This item was opened to determine the reportability of NSW system | |||
actuations. The licensee generated station PIP 0-C97-1715 to document | |||
the clarification. The licensee determined that the NSW system is | |||
Enclosure 3 | |||
t | |||
__ ___ _ _ . - _ _ _ _ _ _ _ _ - _ - _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _-__- - | |||
. | |||
. | |||
. _ . . | |||
.. . | |||
. | |||
. - - _ _ | |||
. . | |||
14 | |||
required for support of the Engineered Safety Features (ESF). As such. | |||
the NSW system is characterized as an ESF support system in the UFSAR. | |||
Section 7.3.1.1.5. ESF Support Systems. The licensee concluded that, | |||
since the NSW system is not an ESF and since 10 CFR 50.72 and 50.73 | |||
require licensee's to report any event or condition that results in a | |||
manual or automatic actuation of any ESF. actuations of the NSW system | |||
were not reportable. | |||
The inspector reviewed ap)licable sections of NUREG 1022. Event | |||
Reporting Guidelines 10 C R 50.72 and 50.73: NUREG 0800, the Standard | |||
Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear | |||
Power Plants. Light Water Reactor Edition. June 1987: the UFSAR: and | |||
Nuclear System Directive 202. Reportability. Revision 8. The | |||
characterization of the NSW system as an ESF support system was in | |||
agreement with the SRP which referred to service water systems as | |||
auxiliary systems that directly support ESF systems. However. Chapter 6 | |||
of the UFSAR. Engineered Safety Features, does not contain a listing of | |||
ESF systems: a listing, which does not include the NSW system, is | |||
located in Nuclear System Directive 202. Reportability. Revision 8 | |||
Appendix A. Engineered Safety Features. Chapter 7 of the UFSAR. | |||
Instrumentation and Controls. Section 7.3.1.1.1 lists ESF functions | |||
initiated by the Engineered Safety Features Actuation System (ESFAS): | |||
the NSW pumps which provide cooling water to the component cooling | |||
sy> tem heat exchangers and are thus the heat sink for containment | |||
cooling, are listed. | |||
Based on this r;"tiew, the inspectors determined that NSW system | |||
. | |||
actuations are not reportable. This item is closed. | |||
III. Enaineerina | |||
El Conduct of Engineering | |||
El.1 Operation Of Dual Function Containment isolation Valves-Temocrarv | |||
Instruction (TI) 2515/136 (Closed) | |||
a. Inspection Scoce | |||
The inspectors used TI 2515/136. Operation of Dual Function Containment | |||
Isolation Valves, to determine if the licensee had procedures in place | |||
to remotely close containment isolation valves when required while a | |||
safety injection or a containment spray signal was present. The | |||
inspector discussed this issue with engineering personnel, and reviewed | |||
the UFSAR and design basis documentation. | |||
b. Observations and Findinas | |||
The Tl included a questionnaire survey with four items. Item 1 | |||
requested that the inspectors identify the dual function valves as | |||
listed in the UFSAR and determine whether differences existed in the | |||
Enclosure 3 | |||
l | |||
_ | |||
- - _ - - - - _ - _ - _ __ - | |||
, , | |||
15 | |||
plant. Licensee personnel provided a list of containment isolation | |||
valves, which included dual function valves. The inspector compared the | |||
valve list to the valves shown on UFSAR Table 6-77, Containment | |||
Isolation Valve Data. All valves identified by the licensee were found | |||
to exist in Table 6-77. | |||
During the ins)ectors' review, it was noted that two of the valves | |||
listed SA-1 ()enetration M 261. B Main Steam to Auxiliary Feedwater | |||
Pump Turbine) and SA-4 (Penetration M-393,~ C Main Steam to Auxiliary | |||
Feedwater Pump Turbine), did not comply with 10 CFR 50. Appendix A. | |||
General Design Criterion 57. General Design Criterion (GDC) 57. Closed | |||
System Isolation Valves, specifies that each line that penetrates | |||
primary reactor containment and is neither part of the reactor coolant | |||
pressure boundary nor connected directly to the containment atmosphere | |||
shall have at least one containment isolation valve which shall be | |||
either automatic, locked closed, or capable of remote manual operation. | |||
Valves SA-1 and SA-4 are manual gate valves and normally in the locked | |||
open )osition. These valves and containment penetrations exist in both | |||
Catawaa Units 1 and 2. The GDC 57 noncompliance had been previously | |||
identified and an exemption request (from GDC 57) was submitted on | |||
September 2, 1997. This item is being tracked as Unresolved Item 50- | |||
413.414/97-14-03: Noncompliance With 10 CFR 50. Appendix A. General | |||
Design Criterion 57 Closed System Isolation Valves. | |||
Item 2 asked whether or not a safety-related dual function valve could | |||
be closed from the control room with a switch and remain closed in the | |||
presence of a containment spray or safety injection signal. As | |||
indicated by the licensee's list, reset and closure capability existed | |||
with remote. manual control on all safety-related dual function valves | |||
with the exception of SA-1 and SA-4, which were locked open. Some | |||
valves, as indicated on the licensee's list, would require the emergency | |||
diesel generator (EDG) load sequencer be reset in addition to normally | |||
resetting the ESF (or Safety Injection) signal. The EDG Load Sequencer | |||
system engineer indicated that resetting the ESF signal would not affect | |||
the configuration or operating status of any safety-related equipment, | |||
and that resetting the EDG Load Sequencer would not affect the EDG or | |||
any com)onents being powered from the safety-related 4160 volt busses. | |||
While tie inspectors were familiar with the reset capability for the | |||
safety injection signal, further NRC inspection was necessary to verify | |||
thet resetting the EDG Load Sequencer during an accident would not | |||
adversely impact the operation of safety-related plant equipment. This | |||
review effort will be conducted under URI 50-413,414/97-14-03 discussed | |||
above. | |||
Item 3 requested, for valves that do not have a switch for remote | |||
closure [i.e. , SA-1 and SA-4]. if any proceduralized method existed | |||
(such as deenergizing circuits or lifting leads or installing leads) | |||
that would facilitate remote closure. Since valves SA-1 and SA-4 are | |||
locally operated manual valves, no remote method of closure existed. | |||
Enclosure 3 | |||
. _ . . . _ . _ . . _ _ _ . _ _ _ _ . . . _ _ . . . . _ _ . _ _ | |||
i | |||
. . | |||
-16- , a | |||
' | |||
: Item 4 requested,lfor valves that do not have'any remote method of- ; | |||
closure:available [i.e.. -SA 1 and SA-4], whether there were any other | |||
means that the licensee had to close the. isolation valve. The licensee - | |||
. | |||
provided a list of eight emergency procedures that contain provisions to | |||
isolate Penetration M-261 or M 393'as required. Two isolation options ; | |||
were provided. - The first-option utilizes the SA-1 or SA-4 valve as | |||
required located in the plant doghouses. The second option-isolates the | |||
> | |||
penetration by closing valves SA-3 or SA 6 located downstream of:SA-1- | |||
and SA-4 =in the Penetration Area if SA-1 and SA-4 were inaccessible. | |||
j | |||
. | |||
The inspectors reviewed these procedures and found that the procedural | |||
guidance to' establish containment isolation manually for penetrations.M- | |||
261 and M-393 was available to operators when needed. , | |||
-c. " Conclusions | |||
=An unresolved item was identified concerning containment penetrations | |||
-associated with steam supply lines-to both units' turbine driven | |||
i auxiliary feedwater h h in compliance with 10 CFR 50, | |||
1A>pendix A, GDC 57. The' pumps, | |||
licensee had w icsubmitted | |||
were notan exemption request to | |||
t1e NRC for this-issue. Remote manual closure capability. existed for | |||
- | |||
* dual function containment isolation valves; however, the action involved | |||
resetting the emergency diesel generator load sequencer, an action | |||
requiring further evaluation to be conducted under the above-mentioned | |||
. unresolved item, , | |||
, | |||
E2 Engineering-Support of Facilities and Equipment i | |||
p | |||
E2.1 Solid State Protection System (SSPS) Testino Deficiency | |||
a. Insoection Scooe (37551) | |||
The inspectors reviewed the licensee's discovery of a logic testing | |||
. deficiency associated with both trains of each unit's SSPS - November | |||
11, 1997. | |||
b. Observations and Findinas , | |||
The test deficiency-involved the failure to perform adequate testing of | |||
two universal-cards associated with feedwater isolation functions and | |||
the P-10 source range. nuclear instrumentation reactor trip block | |||
permissive. . Theiuniversal cards contained previously unidentified | |||
parallel: circuit paths which were not being isolated and independently | |||
verif ted to actuate the logic circuitry associated with each function. | |||
sBoth units entered TS 4.0.3 after identifying the missed surveillance | |||
testing; The procedures were revised and the testing conducted- | |||
, | |||
' satisfactorily before each unit exited TS 4.0.3. | |||
The test anomaly was identified by personnel in the licensee's General- | |||
Office and.was immediately communicated to the SSPS vendor and to | |||
various other nuclear power facilities via operating experience data | |||
-Enclosure 3 | |||
, | |||
e _ | |||
^. - - | |||
' | |||
, . | |||
. , . , . , | |||
. .. .- . - - . . . -. . | |||
-. - - - . - - . . | |||
: | |||
' | |||
. | |||
. .- . | |||
- | |||
, | |||
: | |||
17- ) | |||
-bankst Several facilities have since identified the same or.similar; | |||
deficiencies in their SSPS logic testing procedures, | |||
~ | |||
c. Conclusion | |||
-The licensee has issued LER 50;413/97-08 to document the' missed TS- | |||
surveillances and discuss the safety consequences and corrective actions- 1 | |||
- taken for the deficiency. Further NRC-review will-be conducted during | |||
closecut of the LER. ; | |||
Miscellaneous Engineering Issues (92903) | |||
' | |||
E8 | |||
E8.1-- (Closed) Insoector Follow UD Item 50-413.414/96-18-04: Quam. fication | |||
, | |||
' of_ Refueling Water-Storage Tank (FWST) Heat Losses Through Tank Roof | |||
--Including a Wind Velocity Factor. | |||
This-item involved minor modification CNCE-8309 to de-energize one.of > | |||
four-FWST heater clusters. The licensee performed an evaluation to | |||
demonstrate that minimum required tank tem)erature of 70-degrees | |||
- Fahrenheit (*F) could be maintained with t1e three remaining: heaters. | |||
-The evaluation involved a calculation. CNC 1249.00-00-0065. Operability | |||
; Determination for PIP 1-C96-1870 - Heater Sizing for the FWST, that . | |||
' | |||
quantified-heat losses from the tank assuming a minimum temperature of - | |||
5*F and wind velocity of up to 20 miles per hour (mph). The calculation " - | |||
- | |||
>1ndicated that'the total FWST~ heat lossiat was-81.88 KW. r - | |||
,t | |||
The inspector;noted that-the calculation: accounted for wind-induced heat- | |||
losses from the tank walls, but not from the tank roof. To address this | |||
observation the licensee completed Revision 1 of calculation CNC- | |||
1249.00-00-0065 and concluded that, accounting for heat losses from the | |||
FWST roof assuming a 5 m3h average wind velocity the total FWST heat | |||
< loss was 93.46 KW at an WST temperature of-75*F and environmental | |||
temperature of -5*F. With one heater cluster inoperable and-de- | |||
-energized, the total heater capacity available is 90 KW. The-licensee | |||
indicated that the environmental tem ereture selected for design | |||
comparison in the calculation was be ow the coldest temperature ever | |||
recorded at the site. it was unlikely that temperatures would drop to | |||
that temperature. The licensee also indicated that the heat loss would | |||
t 'be 87.42 KW if the tank wall temperature were assumed to be 70 F (the TS | |||
value)'. and therefore within the heating capacity of the three remaining | |||
heater clusters. Based on these and other conservative heat loss- | |||
assumptions applied to the calculation, the licensee asserted that the | |||
- | |||
remaining heater capacity was marginal to maintain the FWST at 75 F. but | |||
that it was adequate to prevent a temperature drop below the TS-required | |||
value of 70*F. | |||
. | |||
. Refueling water storage tank temperature-indications are available in | |||
' | |||
the control room. In addition, a low | |||
a | |||
. | |||
temperature alarm will be | |||
. generated at 74 F. The alarm response would be to dispatch an operator | |||
to verify heater operation. A Lo-Lo temperature alarm would be | |||
. Enclosure 3 | |||
. | |||
a | |||
r -- Wwy wr,y,u | |||
- . - _ - . ~ -- - - - - - - - - - - . _ | |||
, | |||
. :. | |||
- | |||
, | |||
' | |||
. | |||
c18_- ; | |||
' | |||
+ - generatsd when. tank temperature reaches 70*Fi The response then would. | |||
- | |||
= be to declare the FWST inoperable per the appropriate TS, Based on the; i | |||
heat loss calculation, monitoring-capabilities and response procedures. . . | |||
the ins >ector concluded:that-FWST temperature was not likely to-_ drop 4 | |||
, | |||
. | |||
below t1e TS required value of 70*F as:a result of this minor 1 . . 7 | |||
modification. -Shouldsa-low temperature alarm be ger.erated, effective- | |||
measures were in place to ensure that action will be-taken to correct | |||
the. low temaerature condition or. place the unit in a safe condition; In , | |||
. addition.- tie licensee planned to correct the heater leakage, re- | |||
energize the heater and return' it to service during the upcoming end-of- | |||
icycle 10 refueling outage, scheduled to begin in late November.- | |||
. The ins)ector noted that the wind velocity assumed for heat _ losses from | |||
the tan ( walls was 20 mph.-whereas 1t was assumed to-be only 5 mph.for- | |||
: heat losses from the tank roof.- While no explanation for this | |||
' ' | |||
discrepancy was- 3rovided in the calculation, the inspector concluded | |||
that, since the 1 eater was:to be returned to service in December 1997, ; | |||
' | |||
this discrepancy did not pose a safety concern. This item is closed. | |||
b E8.2_ (Closed) Unresolved Item-50-413.414/97-11-04: Use of Aluminum High | |||
Efficiency Particulate Air (HEPA) Filter Separators Inside Containment. | |||
' | |||
- This item involved the licensee's ~1dentification of aluminum HEPA filter | |||
i | |||
separators in the containment ventilation system's containment auxiliary ' | |||
w a- charcoal filter units?(CACFUs) that had~not been accounted for:in the | |||
~ | |||
t : | |||
- ~ | |||
station's aluminum-inventory records'. :The licensee initiated an > | |||
l -- ; | |||
evaluation _to determine the-root cause'of the inappropriate material. | |||
, | |||
usage. | |||
. | |||
The licensee's evaluation revealed that the HEPA filters had contained | |||
- | |||
aluminum since 1986 or before. Design Specification CNS-1211.00-3, | |||
Containment Auxiliary Charcoal Filter Units, Section 5.5, High | |||
, | |||
Efficiency Filter Section, states that " Separators, if used, shall be | |||
304 stainless-steel." The licensee determined that the original HEPA | |||
. | |||
filters were a separatorless, nuclear grade filter without aluminum. | |||
However, at some undetermined point in time, the station began to use a | |||
: different HEPA filter, containing aluminum. ,in the CACFUs. The licensee. | |||
- could not locate any documentation to support the change in filters and | |||
' | |||
terminated the ioot cause evaluation, which was not likely to reveal the | |||
origination of the discrepancy. | |||
- The. inspector concluded that, although the error leading-to the. | |||
discrepancy had occurred over. ten years ago, the licensee has since | |||
" | |||
established a 3rocess that would prevent a similar oversight from | |||
, | |||
occurring _ at tae:present time. - A changeLin filter components (or other - | |||
. components:inside containment) would involve the-modification process. | |||
- Essentially, NSO.301. Nuclear Station-Modifications.: dated September 30, ^ | |||
1997? required that a Technical Issues Checklist be completed for- any | |||
temporary.: minor, or nuclear station;(permanent-and major) modification. | |||
; _ The Technical 7 Issues Checklist, located in Appendix A of the NSO. 1 | |||
' Enclosure 3- | |||
, | |||
. - | |||
sm'? ,---,,--,,,.,,.,,--,,1%.' .,em'---w, | |||
' ' | |||
. , . - | |||
- | |||
# -.- --r ,, - | |||
y , s.e.- . . - - , - , | |||
-.m ,, , | |||
- - | |||
.. . | |||
-19 | |||
addressed containment. issues and hydrogen control. The question "Does | |||
the change add aluminum or zinc that could potentially increase the | |||
amount of hydrogen gerierated inside the containment post accident?" | |||
would likely prompt a review for this potential during the current | |||
modification process. | |||
The licensee re-evaluated the original hydrogen generation calculation | |||
and determined that the amount of hydrogen generated inside containment | |||
following a design basis accident that would oe produced by the | |||
additional aluminum did not exceed revised allowable limits. | |||
Therefore, the safety consequences were minor. However, measures were | |||
not effective in preventing the selection of thesc filters for use in an | |||
unsuitable application as required by 10 CFR Part 50. Appendix B. | |||
Criterion III. This constitutes a violation of minor significance and | |||
is characterized as a Non-Cited Vielation (NCV). consistent with Section | |||
IV of the NRC Enforcement Policy. This item is identified as NCV 50- | |||
413.414/97-14-04: Failure to Control Use of Aluminum Inside The | |||
Containment Building. | |||
The ins)ector determined that the licensee hed been informed by | |||
Westing louse of the potential that certain HEPA filters were being | |||
manufactured with aluminum separators. The information was conveyed via | |||
Vendor Information Letter 96-30 in September 1996. The licensee's | |||
response to the information was to consult the design specification | |||
(CNS-1211.00 3) to determine if aluminum was specified.' Upon finding | |||
that the specification required the.use of 304 stainless steel, the | |||
licensee concluded that the CACFU's " EPA filters did not contain _. | |||
aluminum. The inspector concluded 4 at the original review in response | |||
to the Westinghouse information letter was cursory and ineffective in | |||
revealing this discrepancy. The inspector reviewed the revised hydrogen | |||
generation calculation: no concerns or discrepancies were identified. | |||
This item is closed. | |||
IV. Plant Sucoort | |||
R1 Radiological Protection (RP) and Chemistry Contro' | |||
R1.1 Tours of the Radiolooical Control Area (RCA) | |||
a. Insoection Scoce (71750) | |||
The inspectors periodically toured the RCA during the inspection period. | |||
Radiological control practices were observed and discussed with | |||
radiation protection personnel, including RCA entry and exit controls, | |||
survey postings, and radiological area material conditions. | |||
Enclosure 3 | |||
- . _ . . | |||
_ - . . .. .. m .. _ _ _ _ _ . _ _ _ _ _ . . . ._ | |||
a c | |||
. .- | |||
d | |||
' | |||
- 20 | |||
b'. : Observations and Findinas' | |||
, | |||
100 November.17. the inspectors noticed an RCA exit door propped wide- | |||
_open with a brick. -The doorway was on the 594 foot elevation of the- | |||
- | |||
. | |||
auxiliary building at the end of corridor number 517 and provided RCA'- | |||
, | |||
-access from the outside. - Two stanchions with a roped sign hanging | |||
-between them ncrmally blocked 6ccess past the door into tns: RCA but the | |||
- -stanchions and' sign had been moved to the side and out of viewt -The | |||
: sign was intended to warn personnel that they were about to enter the 3 | |||
- RCA and directed them to contact radiation protection, personnel for | |||
_ | |||
assistance. At the-time of the ins)ector's obstrvation. no personnel | |||
were present to control access at t11s RCA entry point. 1 | |||
' | |||
The inspectors notified radiation protection (RP) 3ersonnel who- - | |||
'immediately responded to the location and closed tie-door. Later. the | |||
' | |||
same: sign was attached to a swing gate which was placed at the entrance. | |||
, | |||
-The gate would close after allowing personnel pre-appraved access across- | |||
.the boundary. -The inspectors were in. formed by RP personnel that a: | |||
maintenance crew-had been using the door to bring scaffolding into the | |||
- plant in preparation.for the upcoming Unit I refueling outage. The- | |||
maintenance crew had received permission from RP to use the door. The | |||
crew had moved the, sign blocked the door open, and temporarily left the | |||
drea to Conduct other activities. | |||
.~ "'' | |||
r :The inspectors discussed with" licensee personnel the need to properly- * | |||
- | |||
. | |||
* | |||
control access:to the RCA. Licensee personnel generated PIP 0-C-97-3670: | |||
> | |||
^ | |||
to document this deficiency. The incident was discussed in a-subsequent: - | |||
c daily management meeting. In addition to the immediate corrective | |||
4 actions above. RP management discussed this incident with scaffolding | |||
supervisors who later discussed it with their crew members to reinforce | |||
proper procedures for entering the RCA. | |||
-The inspectors later observed that general access to this area from the | |||
outside was limited to the scaffold crew because of a second external | |||
n . barrier that had been _placed outside to control personnel traffic. | |||
While this barrier was not intended for RCA access control it reduced | |||
the significance:of the inspectors' finding. | |||
c. Conclusions- | |||
An example of poor performance was identified related to an RCA boundary | |||
. | |||
-being_ compromised. -This' minor _ discrepancy was immediately corrected by_ | |||
- | |||
plant personnel and properly addressed by licensee management. | |||
s3 | |||
, | |||
RI.2 Transoortation of Radioactive Materials | |||
' | |||
-ai -Insoection Scone (86750) - | |||
L ' | |||
The inspectors evaluated the licensee's transportation of radioactive | |||
materials programs for implementing the revised Department of | |||
Enclosure 3 | |||
. | |||
., , - , , , , - ... .,. - , , . ~ . - - 4 , .. ,---- | |||
.v. . * e. ---.w | |||
- | |||
' | |||
. . | |||
21 | |||
Transportation (00T) and. NRC trans)ortation regulations for shipment of | |||
radioactive materials as required )y 10 Code of Federal Regulations | |||
(CFR) 71.5 and 49 CFR Parts 100 through 177. | |||
:b. Qb.servations and Findinos | |||
The inspectors reviewed procedures and determined that they adequately | |||
addressed the following: assuring ' hat the receiver has a license to | |||
receive the material being shipped; assigning the form. quantity type, | |||
and proper shipping name of the material to be shipped: classifying | |||
waste destined for burial; selecting the type of package required: | |||
assuring that the radiation and contaminatis : limits are met: and | |||
preparing shipping papers. | |||
Licensee's records for the six shipments of radioactive material | |||
performed in 1997 were reviewed and the inspectors determined the | |||
shipping papers contained the required information. The inspectors also | |||
determined the licensee had maintained records of shipments of licensed | |||
material for a period of three years after shipment as required by | |||
10 CFR_71.91(a). In addition, the licensee )ossessed a current | |||
certificate of approval (NRC Form 311) for t1eir " Quality Assurance | |||
Program Description for Radioactive Material Shipping Packages Licensed | |||
Under 10 CFR 71." The licensee had also maintained current NRC | |||
certificate of compliance for the NRC approved cask in use. | |||
The inspectors reviewed the training records for selected individuals | |||
- | |||
authorized to sign shipping papers and: handle radioactive waste which | |||
included a w area su)ervisc who was assigned to the area of | |||
transportati .. the weet of the inspection. The training specifically | |||
addressed the new rules for the following to)ics: low specific activity | |||
(LSA) and surface contaminated object (SCO) idzards, definitions, and | |||
requirements: placarding, labeling, and marking of vehicles and | |||
packages: use of Systems Internationals (SI) units on shipping papers, | |||
labels, and emergency response instructions after April 1.1997: package | |||
selection: waste classification: shipping papers; and receipt procedures | |||
and surveys. The inspectors concluded that personnel involved with | |||
radioactive material shipping were maintaining current training | |||
qualifications. | |||
c. Conclusions | |||
The licensee had effectively im)1emented a program for shipping | |||
radioactive materials required ]y NRC and DOT regulations. | |||
R1.3 Radiolooical Protection and Chemistry Controls | |||
a. Insoection Scone (84750) | |||
The inspectors reviewed implementation of-selected elements of the | |||
licensee *s radiation protection and chemistry program. The review | |||
Enclosure 3 | |||
. | |||
_ . .. . . _ , . . _ . . _ _ _ . . . _ _ _. _ _ . _ _ _ . | |||
V | |||
. | |||
. - | |||
._ | |||
' | |||
.o ^ | |||
, | |||
, | |||
- | |||
22 | |||
1 | |||
included observation of radiological protection activities for the | |||
: control of. radioactive material as required by 10 CFR Parts 20,1801. | |||
~ | |||
- | |||
_ __ | |||
:1802. 1902. and 1904. | |||
b.-' Observations and Findinos , | |||
i The inspectors reviewed licensee goals for waste generated and buried | |||
- | |||
, | |||
and determined the licensee was meeting these goals. During tours of | |||
- | |||
- the auxiliary building and radwaste building facilities, the inspectors | |||
reviewed survey _ data and performed selected independent radiation and , | |||
contamination surveys of radioactive material storage areas. During a | |||
' | |||
tour of the hot tool issue room on November 19, 1997, the inspectors | |||
found a vacuum cleaner with radiation dose rates higher than indicated | |||
' | |||
on the radioactive material label.. dated 1995, affixed to the vacuum 1 | |||
4 | |||
cleaner. The tag stated radiation levels to be 1.5 millirem per hour on | |||
-contact and 0.5 millirem at 30 centimeters. However, the inspectors | |||
determined and the licensee confirmed radiation levels to be up to 40 | |||
< millirem per hour contact and 2-3 millirem at 30 centimeters. Also, the | |||
vacuum cleaner hose was not taped or capped on the end as required by- | |||
' | |||
licensee procedure for vacuum cleaners in storage. Licensee procedure | |||
' | |||
required vacuum cleaners to be surveyed after use.and that current | |||
- survey information was to be included on the radioactive material label | |||
" | |||
(yellow tag). - The licensee taped over the vacuum hose and performed | |||
independent radiation-and contamination surveys of the vacuum cleaner | |||
and the general area'. :The licensee determined contamination hadinot. - | |||
- | |||
* - | |||
- | |||
- | |||
1been spread as:a result of the open hose. The licensee also relabeled r | |||
the vacuum cleaner to include current' survey information. . | |||
4 | |||
^ | |||
Ouke Power Company. System Radiation Protection Manual. Procedure No. | |||
III-18. titled Use of Vacuum Cleaners In Radiologically Controlled | |||
Areas. Revision 3. dated August 1. 2996, states that vacuum cleaners | |||
should be surveyed during and after use and update dose rates on yellow | |||
' | |||
_ tags, if applicable, each time a radiation survey is performed. | |||
10 CFR 20.1904(a) recuires, that the licensee shall ensure that each | |||
container of licensec material bears a durable. clearly visible label | |||
~ | |||
bearing the radiation symbol and the words CAUTION RADI0 ACTIVE MATERIAL | |||
or DANGER RADI0 ACTIVE MATERIAL. The label must also provide sufficient | |||
information-(such as radionuclides f the quantity | |||
i .of radioactivity, radiation-levels.present, kinds ofan estimate | |||
materials, and o mass | |||
enrichment) to permit individuals-handling or using the containers or | |||
working in the vicinity of the containers. to take precautions to avoid | |||
or minimize exposures. | |||
:The-inspector informed the licensee that failure to provide current | |||
-survey.information on the radioactive material label constituted a | |||
; violation of licensee procedure Use of Vacuum Cleaners In Radiologically | |||
Controlled Areas. III-18. Revision 3 and a violation of 10 CFR | |||
- | |||
: | |||
" - | |||
20.1904(a). This item is identified as Violation 50-413.414/97-14-05: | |||
1 Failure to Label Radioactive Material As Required by 10 CFR 20.1904. | |||
Enclosure 3 | |||
g - | |||
-1 | |||
. | |||
t + -- W- w -e-vip---r r- w - ,->N,-,r ,-cv-m ,- - | |||
w | |||
- .. .. . _ - . - . . | |||
. | |||
. | |||
23 | |||
c. Conclusions | |||
The licensee was meeting established goals for radioactive waste | |||
generation. During plant tours, radiological facility conditions and | |||
housekeeping in radioactive waste storage areas were observed to be - | |||
good. One violation was identified for failure to provide current dose | |||
rate information on a radioactive material label as required by licensee | |||
procedure and 10 CFR 20.1904(a). | |||
P1.4 Water Chemistry Controls | |||
a. Insoection Scoce (847501 | |||
The inspectors reviewed implementation of selected elements of the | |||
licensee's water chemistry control program for monitoring primary and | |||
secondary water quality as described in the TS limits, the Station | |||
Chemistry Manual, and the UFSAR. The review included examination of | |||
program guidance and implementing procedures and analytical results for | |||
selected chemistry parameters, | |||
b. Observations and Findinos | |||
The inspectors reviewed selected analytical results recorded for Units 1 | |||
and 2 reactor coolant primary water chemistry samples taken between May, | |||
= | |||
1997 and November, 1997, and secondary system water chemistry samples | |||
taken between August, 1997 and November, 1997. The selected parameters | |||
reviewed for primary water chemistry included dissolved oxygen, | |||
chloride, pH. and fluoride. The selected parameters reviewed for | |||
secondary water chemistry included hydrazine, dissolved o.xygen sodium, | |||
copper, and chloride. Those primary system parameters reviewed were | |||
maintained well within the relevant TS limits for power operations. | |||
Those secondary system parameters reviewed were maintained according to | |||
station procedures. | |||
The inspectors reviewed and discussed the licensee's system for tracking | |||
performance indicators in the areas of primary and secondary water | |||
chemistry. The inspectors noted the licensee had maintained a high | |||
level of success in human performance and equipment reliability in 1997 | |||
based on performance indicators for these areas which included no missed | |||
surveillances and no mispositioning of components. | |||
c. Conclusions | |||
Based en the above reviews, it was concluded that the licensee's water | |||
chemistry control program for monitoring primary and secondary water | |||
quality had been implemented, for those parameters reviewed in. | |||
accordance with TS requirements and the Station Chemistry Manual for | |||
pressurized water reactor water chemistry. The licensee had maintained | |||
a high level of success in human performance and equipment reliability | |||
-in 1997. | |||
Enclosure 3 | |||
. . | |||
- | |||
24 | |||
R3 Radiation Protection and Chemistry Procedures and Documentation | |||
R3.1 Radiation Protection and Chemistry Procedures and Documentation | |||
a. Insoection Scone (84750) | |||
The inspectors reviewed licensee effluent release limits and pathways as | |||
described in the licensee's Offsite Dose Calculation Manual and in | |||
Chapter 16 of the Selected License Commitments Manual, | |||
b. Observations and Findinas | |||
The inspectors reviewed annual effluent data for 1996 and compared the | |||
data to previous annual reports back to 1992. Arinual Radioactive | |||
Effluent Release Reports were required to be submitted to the NRC prior | |||
to May 1 of each year. Summaries of the quantities of radioactive | |||
-materials in liquid and gaseous effluents released from the facility and | |||
an assessment of the radiation doses due to those releases were required | |||
to be included in the reports. The inspectors reviewed the supporting | |||
data for the effluent release report covering 1996. The amount of | |||
activity released during 1996 as dissolved gases in liquid effiuents and | |||
fission gases, and that released as iodines and particulates in gaseous | |||
effluents was generally within the ranges observed in past years. The | |||
annual average per unit radiation doses for an individual from the | |||
- | |||
liquids and gaseous effluents were only a small percentage of their | |||
respective annual limits. The total body dose as calculated by | |||
environmental sampling data, was 0.902 millirem for 1996. There were no | |||
abnormal releases reported in 1996. | |||
c. Conclusions | |||
Based on the above reviews. it was concluded that the licensee had | |||
maintained an effective program to monitor and control liquid and | |||
gaseous radioactive effluents, thereby limiting dose to members of the | |||
public. The )rojected offsite doses resulting from those effluents were | |||
well within tie limits specified in the TS. Offsite Dose C61culation | |||
Manual, and 40 CFR 190. | |||
R7 Quality Assurance in Radiation Protection and Chemistry Activities | |||
R7.1 Ouality Assurance in Radiation Protection (RP) and Chemistry | |||
a. Insoection Scooe (84750) | |||
Licensee activities and self assessment programs were reviewed to | |||
determine t.he adequacy of corrective action programs for identified | |||
deficiencies in the areas of RP and chemistry. | |||
Enclosure 3 | |||
- . _ . __ . _ . _. _ _ _ _ _ _ _ _ - | |||
. 4 | |||
o | |||
25 -j | |||
. | |||
bl Ilc trvations and Findinos | |||
-Reviews by the= inspectors: determined that Quality AssuranceLaudits-and | |||
self assessments-in the RP and chemistry areas were accomplished _by | |||
. | |||
reviewing-procedures, observing work, reviewing industry documentation, | |||
and performing plant walkdowns to include surveillance of work areas by | |||
supervisors and technicians during normal work coverage. Documentation | |||
of problems by licensee representatives was included in Quality | |||
Assurance Audits and self assessment _ reports. Corrective actions-were | |||
> included in the licensee's-PIPS and were being completed in a timely | |||
manner. | |||
' | |||
c. Conclusions- | |||
' | |||
4 | |||
The ins | |||
RP andchemistry | |||
pectors-determined the-licensee | |||
audits as required was effectively | |||
by the TS-and was completing conducting formal , | |||
corrective actions in a timely manner. * | |||
R8 Miscellaneous Radiation Protection and' Chemistry Issues (92904) , | |||
, | |||
R8,1 (Closed) URI 50-413.414/97-04-Q2; Determine the y plicability of | |||
' | |||
-Monitoring Requirements of Criterion 64 of 10 CFR s0 A)pendix A: and | |||
Reporting Requirenents of 40 CFR 190 and 10 CFR 50.36a legarding | |||
Potential Unmonitored Release Pothways. . | |||
. | |||
This item was-closed using guidance from Regulatory Guide 1.109, , | |||
Calculation of Annual Doses to Man from Routine Releases of Reactor | |||
Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50. | |||
Appendix I. The specific guidance was found in Appendix 01. No | |||
violation of regulatory requirements was identified. This item is | |||
closed. | |||
P2 Status of Emergency Protection Facilities, Equipment, and Resources | |||
e | |||
P2.1 General Comments (71750) | |||
. | |||
The inspectors toured the Eme'gency Operations Facility located in | |||
downtown Charlotte. North Carolina on November 18, 1997. The inspectors | |||
observed that the facility and associated. equipment, including emergency | |||
communication telephones and plant computer screens and controls were | |||
. functioning and in good repair. During tours of the Technical Support | |||
. | |||
Center, facility equipment was also noted to be in working order and of | |||
good condition and repair. | |||
. | |||
Enclosure 3 | |||
d | |||
,. , . - ~ _ | |||
. . | |||
26 | |||
V. Manaaement Meetinoq | |||
X1 Exit Heeting Sumary | |||
The inspector presented the inspection results to members of licensee | |||
management at the conclusion of the inspection on December 3, 1997. The | |||
licensee acknowledged the findings presented. No proprietary | |||
information was identified. | |||
Enclosure 3 | |||
- | |||
. . | |||
27. | |||
PARTIAL LIST OF PERSONS CONTACTED _ | |||
Licensee | |||
. | |||
M. Birch. Safety Assuranco Manager | |||
M. Boyle. Radiation Protection Manager | |||
R. Glover. Operations Superintendent | |||
J. Forbes. Engineering Manager | |||
R. Jones. Station Manager | |||
K. Nicholson, Compliance Specialist | |||
M. Kitlan Regulatory Compliance Manager | |||
G.-Peterson Catawba Site Vice-President | |||
R. Propst. Chemistry Manager | |||
Enclosure 3 | |||
. . | |||
28 | |||
INSPECTION PROCEDURES USED | |||
IP 37551: -Onsite Engineering | |||
IP 61726: Surveillance | |||
IP 62707: Maintenance Observation | |||
IP 71707: Plant Operations | |||
IP 71714: Cold Weather Preparations | |||
IP 71750: Plant Support Activities | |||
IP 84750: Radioactive Waste Treatment, and Effluent and Environmental | |||
Monitoring | |||
IP 86750: Solid Radioactive Waste Management and Transportation of | |||
Radioactive Materials | |||
IP 92901: Follow up - Operations | |||
IP 92902: Follow up - Maintenance | |||
IP 92903: Follow up - Engineering | |||
IP 92904: Follow up - Plant Support | |||
IP 93702: Prompt Onsite Response to Events | |||
TI-2515/136: Operation of Dual Function Containment Isolation Valves | |||
ITEMS OPENED, CLOSED, AND DISCUSSED | |||
i Opened | |||
50-413/97-14-01 URI Control Power Unavailable to the Unit 1 | |||
Turbine-Driven AFW Pump's Trip and | |||
Throttle Valve (Section 01.3) | |||
50-413.414/97-14-02 DEV Changing NRC Commitments Without Properly | |||
Notifying the NRC (Section 08.1 and 08.2) | |||
50-413,414/97-14-03 URI Noncompliance With 10 CFR 50 Appendix A | |||
General Design Criterion 57 (Section | |||
El.1) | |||
50-413.414/97-14-04 NCV Failure to Control Use of Aluminum Inside | |||
the Containment Building (Section E8.2) | |||
50-413,414/97-14-05 VIO Failure to label Radioactive Material As | |||
Required by 10 CFR 20.1904 (Section R1,3) | |||
Closed | |||
50-414/95-01 LER Reactor Trip Due to Closure of a Main | |||
Steam Isolation Valve (Section 08.1) | |||
50-413.414/97-08-04 IFI Reportability of Nuclear Service Water | |||
System Actuations (Section M8.1) | |||
Enclosure 3- | |||
. . | |||
29 | |||
50-413.414/96-18-04 IFI Quantification of Refueling Water Storage | |||
Tank Heat Losses Through Tank Roof | |||
Including a Wind velocity Factor (Section | |||
E8.1) | |||
50 413.414/97-11-04 URI Use of Aluminum HEPA Filter Separators | |||
Inside Containment (Section E8.2) | |||
50 413.414/97-05-02 URI Determine the Applicability of Monitoring | |||
Requirements of Criterion 64 of 10 CFR 50. | |||
Appendix A: and Reporting Requirements of | |||
40 CFR 190 and 10 CFR 50.36a Regarding | |||
Potential of Unmonitored Release Pathways | |||
(Section RF, 1) | |||
TI 2515/136 TI Operation of Dual Function Containment | |||
Isolation Valves (Section El.1) | |||
Discussed | |||
50-413/97-08-01 VIO Inadequate Alarm Response Results in | |||
Inadequate add Untimely Corrective Actions | |||
for Valve Operability Determination | |||
. (Section 08.2) - | |||
LIST OF ACRONYMS USED , | |||
AFW - Auxiliary Feedsater | |||
CACFU - Containment Auxiliary Charcoal Filter Units | |||
CFR - | |||
Code of Federal Regulations | |||
DC - | |||
Direct Current | |||
DBD - | |||
Design Basis Documents | |||
DEV - | |||
Deviation | |||
001 - Digital Optical Isolator | |||
DOT - | |||
Department of Transportation | |||
DNB - | |||
Departure From Nucleate Boiling | |||
EDG - | |||
Emergency Diesel Generator | |||
ESF - | |||
Engineered Safety Features | |||
ESFAS - Engineered Safety Features Actuation System | |||
FWST - | |||
Refueling Water Storage Tank | |||
GDC - General Design Criterion | |||
HEPA - High Efficiency Particulate Air | |||
KW - | |||
Kilowatt- | |||
LC0 - | |||
Limiting Condition for Operation | |||
LER - | |||
Licer.see Event Report | |||
LSA - | |||
Low Specific Activity | |||
MPH - | |||
Miles Per Hour | |||
NEI - | |||
Nuclear Energy Institute | |||
NRC - | |||
Nuclear Regulatory Commission | |||
Enclosure 3 | |||
o .- | |||
30 | |||
NS - Nucicar Spray | |||
Nuclear System Directive | |||
~ | |||
NSD - | |||
NSW -- | |||
Nuclear Service Water | |||
0AC- - Operator Aid Computer | |||
00CM - | |||
Offsite Dose Calculation Manual | |||
0PDT - 0,erpower Differential Temperature | |||
PCB - | |||
Power Circuit Breaker | |||
PDR - | |||
Public Document Room | |||
PIP - Problem Investigation Report | |||
PM - | |||
-Preventive Maintenance | |||
PORVS - Power Operated Relief Valves | |||
PSIG - Pounds per Square Inch Gauge | |||
RCA - Radiological Control Area | |||
RGC - | |||
Regulatory Comaliance . | |||
"RHR - | |||
Residual Heat Removal | |||
RP - | |||
Radiation Protection | |||
R&R - | |||
Repair and Restor 6 tion | |||
SCO - Surface Contaminated Object | |||
SI - System Internationale | |||
SRP - | |||
Standard Review Plan | |||
SSC - Structures. Systems, and Components | |||
SSPS - | |||
Solid State Protection System | |||
TS -- | |||
Technical S ecification | |||
TSAll - | |||
UFSAR - | |||
.. Technical S ecification Action Items List | |||
Updated Fin 1 Safety Analysis Report r: | |||
-URI- - | |||
Unresolved Item | |||
VIO - | |||
-Violation - | |||
WO - | |||
Work Order | |||
Enclosure 3- | |||
9 | |||
}} |
Latest revision as of 10:19, 15 December 2020
ML20197F286 | |
Person / Time | |
---|---|
Site: | Catawba ![]() |
Issue date: | 12/19/1997 |
From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
To: | |
Shared Package | |
ML20197F239 | List: |
References | |
50-413-97-14, 50-414-97-14, NUDOCS 9712300176 | |
Download: ML20197F286 (34) | |
See also: IR 05000413/1997014
Text
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _
,
.
U.S. NUCLEAR REGULATORY COMMISSION
REGION 11
Docket Nos: 50 413. 50 414
License Nos: NPt'-35. NPF 52
Report Nos.: 50 413/97 14. 50 414/97-14
Licensee: Duke Energy Corporation
Facility: Catawba Nuclear Station. Units 1 and 2
' Location: 422 South Church Street
Charlotte, NC 28242
Dates: October 12 - November 22. 1997
Inspectors: D. Roberts. Senior Resident Inspector
R. Franovich, ResidMt Inspector
M. Giles, Resident inspector (in Training)
D.Forbes.RadiationSpecialist.RegionII(SectionsR1.2.
R1.3. RI.4, R3.1. Re.l. and R8.1)
Approved by: C. Ogle. Chief
Reactor Projects Branch 1
Division of Reactor Projects
Enclosure 3
?N
G
00b N y"
. .
EXECUTIVE SUMMARY
Catawba Nuclear Station. Units 1 and 2
NRC Inspection Report 50-413/97-14, 50 414/97 14
This integrated inspection included aspects of licensee operations
maintenance, engineering and plant support. Thereportcoversa5 week
period of resident inspection. It also includes the results of an announced
inspection by a regional radiation specialist.
Doerations
. In general, the conduct of operations was professional and safety
conscious. (Section 01.1) 1
e A minor overpower excursion resulted in the 15 minute running average
for reactor thermal
an extended aeriod. power slightly
The power exceeding
excursion licensed power
was contained limits for
within criteria
established Jy previous NRC guidance. (Section 01.2)
e Control room o)erators failed to detect an extinguished 'DC Power On"
light for the Unit 1 turbine-driven auxiliary feedwater pump for almost
three days. The impact on Jump operability of the blown fuse which
caused the extinguished 1191t. will be reviewed during closeout of the
URI. (Section 01.3)
. Operations personnel inappropriately entered the Technical Specification
action statement more than an hour after a reactor trip system logic
function failed to meet surveillance test acceptance criteria. However,
the failed function was repaired, successfully retested, and returned to
service before Technical Specification actions were required. (Section
01.4)
- Nuclear Systerr Directive 317 provided structure and delineated
responsibilities for freeze protection. Proceduralized activities were
initiated and completed in a timely manner, and work requests were
initiated to resolve identified discrepancies. The licensee's efforts
to effectively protect plant equipment and systems from freezing
conditions improved since the previous cold weather season. (Section
02.1)
- Four unreltted non emergency events were reported to the NRC in
accordance with Title 10 Code of Federal Regulations. Part 50.72 during
the period. All of the events were properly reported with sufficient
information provided. (Section 02.2)
- Examples of poor performance were noted concerning activities
surrounding the inappropriate tagout of a residual heat removal system
miniflow valve during planned maintenance. (Section 04.1)
. A deviation, with two examples, from NRC commitments was identified.
Both examples involved administrativc errors resulting in commitments
Enclosure 3
.
_ . _ - .
. .
i
2
being changed internally without proper notification of the NRC.
(Sections 08.1 and 08.2)
licintenance
- Surveillance activities observed by the inspectors involved good
workmanship, proper use of procedures, good radiological practices. and
) roper management of Technical Specification action statements. (Section
11.1)
- New fuel movement activities to support the upcoming Unit I refueling
outage were performed well. (Section M1.2)
Enoineerinq
- An unresolved item was identified concerning containment penetrations
associated with stea < cupply lines to both units' turbine driven
auxiliary feedwater ,o ms, which were not in compliance with Title 10
Code of federal Regulations. Part 50. Appendix A. General Design
Criterion 57. The licensee had submitted an exemption request
concerning this issue to the NRC during the previous inspection report
period. (Section El.1)
. Remote manual closure capability existed for dual function containment
isolation valves; however, the action involved resetting the emergency
diesel generator load sequencer. an action requiring further evaluation
to be conducted under the above unresolved item. (Section El.1)
. A non cited violation was identified concerning the use of aluminum
separators in high efficiency particulate air 111ters located inside
containment. (Section E8.2)
ElantSunnort
- An example of poor oerformance was identified related to a radiological
control area boundary beinc) compromised. This minor discrepancy was
immediately corrected by plant aersonnel and properly addressed by
licensee management. (Section 11.1)
. The licensee effectively imalemented a program for shipping radioactive
materials required by the NRC and Department of Transportation
regulations. (Section R1.2)
. The licensee was meeting established goals for radioactive waste
generated. Radiological facility conditions and housekeeping in
radioactive waste storage areas were observed to be good. (Section
R1.3)
Enclosure 3
____- _-__ _
_ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
l
.
.
3 l
l
- One violation was identified for failure to provide current dose rate i
information on a radioactive inaterial label as required by Ti'.ie 10 Code
of Federal Regulations. Part 20.1904(a). (Section Rl.3)
- The licensee's water chemistry control program for monitoring primary
and secondary water quality had been implemented for those parameters
reviewed. in accordance with the Technical Specification requirements
and the Station Chemistry Manual for pressurized water reactor water
chemistry. (Section RI.4)
. The licensee had properly implemented procedures to maintain an
effective program to monitor and control liquid and gaseous radioactive
effluents to limit doses to members o the public. Theprojected
offsite doses resulting from those effluents were well within the limits
specified in the Technical Specifications, the Offsite Dose Calculation
Manual. and Title 40 Code of Federal Regulations. Part 190. (Section
R3.1)
. The licensee was effectively conducting formal radiation protection and
chemistry audits as required by Technical Specifications and was
completing corrective actions in a timely manner. (Section R7.1)
. Tha Emergency Jperations facility located in downtown Charlotte. North
Carolina and its associated equipment were in good repair and condition.
Energency communication and plant computer equipment in the Technical
Support Center was in good working order. (Section P2,1)
Enclosure 3
_ _ _ _ - _ _ _ ____ . _ _ _ _ __ _ _ ____ _ _ _ ___
. .
Renort Details ;
Sumary of Plant Status
Unit 1 operated at or near 100% )ower until November 21, when it began its
end of-cycle 10 coast down for 11e upcoming refueling outage. The unit ended
the inspection period at 98 percent power,
i
Unit 2 operated at or near 100 percent power until October 20, when a power
redu: tion was initiated to comply with Technical Specification (TS) 3.6.3 :'
following a nitrogen leak associated with the accumulator for main feedwater *
isolation valve 2CF 33. Power was reduced to approximately 15 percent power,
at which the time the valve was gagged shut a.1d repairs commenced. _Upon ,
completion of the leak repair and valve post naintenance testing activities. '
the unit was returned to 100 percent power on October 21. On November 21. a ,
' power reduction to 50 percent was initiated to allow a control circuit card
associated with main turbine control valve C h1 to be replaced. Licensee ,
personnel also replaced a solenoid valve ano cleaned instrument air lines
associated with main generator power circuit breaker (PCB) 28. These
activities were completed on November 22 and power was increased to 97 percent !
by the end of the inspection period. ;
Review of Vodated Final Safety Analysis Reoort (UFSAR) Commitments
While performing inspections discussed in this repart, the inspectors reviewed ,
the applicable portions of the UFSAR that were related to the areas inspected. +
The inspectors verified that the UFSAR wording was consistent with the
observed plant practices, procedures, and parameters.
I. Operations
01 Conduct of Operations
01.1 General Comments (71707)
The inspectors conducted frequent control room tours to verify proper
staffing, operator attentiveness and communications, and adherence to
approved procedures. The inspectors attended opf.ations turnovers and
site direction meetings to maintain awareness of overall plant
operations. Operator logs were reviewed to verify operational safety
and compliance with TS. Instrumentation, computer indications, and
safety system lineups were periodically reviewed from the control room
to assess o)erability. Plant tours were conducted to observe equipment >
status and lousekeeping. Problem Identification Process (PIP) reports
were routinely reviewed to ensure that potential safety concerns and
equipment problems were reported and resolved.
In general, the conduct of operations was professional and safety- .
conscious. .The Unit 2 power reduction associated with feedwater
isolation valve 2CF-33 was conducted safely. Good plant equipment
material conditions and housekeeping were noted throughout the report
Enclosure 3
.
, w----e-,-._..r _-m,-.e-uw- ..v- w - 4me-v y -.i.- eye ._ ,. ,v,m,
. .- - - - - . . . - - - .
,
1
l
. .
l
2 1
,
period. However as addressed below, several human performance related !
deficiencies wele identified.
01.2 H1nor Excursion Over Licensed Power Limits for Unit 1
a. .lmeection Scone (71707)
The inspectors reviewed the circumstances associated with a minor power i
excuision on Unit 1. f
b. Observations and findinas
P
On October 21, 1997. the inspector noted during a review of control room
'
logs that the Unit 1.15 minute running average for reactor power, as ,
indicated by the Operator Aid Computer (OAC), had exceeded 100 percent.
<
The Unit 1 operator noticed this at 3:39 a.m. and reduced turbine load
by 5 megawatts and inserted control rodt. two steps to bring power aelow
100 percent. Operations personnel later generated station PIP l-C97-
3382 to document and investigate the power excursion.
The inspectors reviewed the PIP and noted that the Unit 1 OAC 15 minute :
,
average was stated as having been in alarm for 15 minutes. The
inspectors reviewed 0AC trend reports for reactor power and noted that .
the maximum instantaneous reactor power level, according to secondary
heat balance best estimates (computer point C1P1445), was ap3roximately 1
100.6 percent recorded just before 3:15 a.m. According to tie trend
report.powercontinuallyspikedbetween99.7and100.3percentpower
for the next 20 25 minutes before operators noticed the 15 minute
average and reduced power. Computer trends indicated that the 15 minute
average
However,it peaked at 100.05
never .'eached the percent
alarm setand was
point in for yercent).
(100.1 about 20 minutes.
Further 1
discussions with plant personnel and review of alarm listory data
indicated that the statement in the PIP concerning the alarm being in
for 15 minutes was in error. Later, this :tatement was corrected in the
PIP documentation. '
Further investigation by the inspectors determined that routine reactor
cooldnt system Doron dilutions were )erformed earlier in the shift.
However, this was last done 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> aefore the noted power excursion.
The inspector interviewed control room personnel who indicated that
several activities were occurring at the time of the minor over power.
including those associated with a Unit.2 down power (see Section 08.2 of '
this report). The operator indicated that these activities may have
been a distraction and
minute average earlier,possibly prevented him from noticing the 15- i
Discussions with operators 31d plant management indicated that operators
were expected to maintair eactor p& ar at licensed power levels.
Operators were expected to contiv.' ly monitor power and immediately
take actions to keep it within i yt. Plant management discussed this- -
Enclosure 3
__ __ _ . . _ _ _ __ _ _. _ _ _
-_ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ - _
. .
3
l
excursion with those personnel involved in the event and emphasized the
need for iiereased diligence when monitoring power levels.
The inspectors reviewed NRC guidance on minor power excursions and noted
that the power level did not exceed previously established criteria,
c. Conclusion
The inspectors concluded from their review that the )ower excursion was
minor and was contained within criteria established )y previous NRC
. guidance.
01.3 Control Power Unavailable to the Unit 1 Turbine Oriven Auxiliarv
Mger (Af W) Pumo Irio And lhrottle Valve
a. Insocction Scone C/110D.
The inspectors reviewed the circumstances associated with a loss of
control power to the Unit 1 turbine driven AFW pump trip and throttle
valve.
b. Observations and Findinni
During a control room tour on November 17, the inspectors noted that the
~0C (Direct Current) Power On" li
4 driven AFW pump was extinguished.ght associated
The inspectors with
informed the the Unit
o)erator 1 turb
at the controls of this observation. The operator replaced the wio and
the light was still not lit. The inspectors mentioned that they had
3reviously observed the light to be out 3 days earlier on November 14
)ut had assumed then that the extinguished light was related to ongoing
maintenance involving a 72-hour LC0 on the system. Theoperatorstated
that this ? ulb was in the control circuit for the AFW pump turbine trip
and throttle valve. Subsequent licensee troubleshooting determined that
fuse FU-2 in control panel 1ELCP0245 was blown. A review of several
electrical drawings indicated that control aower and electrical
overspeed trip functions for the trip and tirottle valve were powered
through this fuse. The trip and throttle valve and the turbine driven
AFW pump were declared inoperable shortly after 10:00 a.m. and the fuse
was subsequently replaced.
The inspectors discussed aspects of this incident with plant personnel
to determine whether operators may have missed opportunities to identify
the deficiency earlier, and to determine the true impact of the blown
fuse on the system's capability to perform its safety functions. The
inspectors noted that t1e blown fuse also caused control power
indication to be extinguished at a local control panel, and that if the
fuse was indeed blown for more than 3 days, plant personnel may have
missed additional opportunities to identify a problem while on-field
tours. The ins)ector noted that there were no formal checks in licensee
procedures of tle *DC Power On" light in the control room.. There were
Enclosure 3
_ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ ____ _ _ ____ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _
l
. .
4 i
also no control room alarms indicating control power unavailability for
the trip and throttle valve.
Engineering personnel were still evaluating whether or not losing
control power or the electrical overspeed trip function rendered the ,
sy. tem inoperable. According to Section 20.4.1.1. " Auxiliary feedwater ,
Pump Turbine," of s)ecification CNS-1593.SA-00 0001. Design Basis
Specification for tie Main Steam to Auxiliary Equipment System (SA) and ,
feedwater Pump Turbine Exhaust System (TE). Revision 11: at least one of
the overspeed trip devices (mechanical or electrical) must be operable <
for the turbine driven auxiliary feedwater pump to be operable. The t'
mechanical overspeed trip function was not affected by the blown fuse.
The inspectors concluded that further review of this incioent and its '
impact on the turbine driven AFW pump was necessary. Pending further -
NRC review, this item is characterized as Unresolved item (URI) 50-
413/97-14 01: Control Power Unavailable to the Unit 1 Turbine Driven i
AFW Pump's Trip and Throttle Valve.
c. Conclusion
Control room o)erators failed to detect an extinguished *DC Power On"
light for the Jnit 1 turbine driven AFW pump for more than three days.
The impact of the blown fuse on pump operdbility will be reviewed during
closeout of the URI. t
01.4 Hanaaement of Technical Soecification (TS) Limitina conditions for
Operation
'
a. InspectionStone(71707)
During a surveillance test of the Unit 2 reactor trip system ;
instrumentation on October 10. 1997, a problem associated with the
overpower differential temperature (0PDT) reactor trip logic was
identified. The inspector discussed the test failure with operations
shift i
3271, personnel, read the associated TS, and reviewed station PIP 2 C97-
b. Observations and Findinas
During the performance of IP/2/A/3200/002A Solid State Protection
System (SSPa) Train A Periodic Testing, Revision 21. on October 10.
1997, the OPDT reactor trip logic test acceptance criterion was not met.
A red lamp illuminated to indicate that a malfunction of the logic
testing was detected (a green lamp would have illuminated if the logic
test had been acceptable). The surveillance test began at 9:56 a.m.,
and the failure was identified some time before noon. Test technicians
backed out of the test, and the reactor trip system was removed from the
TS Action item List at 12:10 p.m. Engineering aersonnel were_ involved
to assist operations personnel in determining tie extent of the
operability concern (i.e., was the problem limited to OPDT trip logic or
Enclosure 3 -
- ._ -
- _
. _ ___ - ._ _ _ . _ _ _ _ _ _ _ _
'
, .
f
5
,
'
did it affect all of the solid state protection system). Engineerin
personnel concluded that the problem was limited to the OPDT trip lo ic
and communicated their conclusion to operations per.vnnel at around :00
p.m. The A train of Automatic Trip and Interlo: 's ' unctional Unit
19 of TS 3.3.1. Table 3.3-1) was declared inopc'e 30 p.m.
placing the unit in a six hour action statement .. n the function
or be in Hot Standby (Mode 3) in the following six v v
The inspectors questioned operaticns shift personnel about the decision
to enter the required action at 1:00 p.m. rather than when the OPDT
reactor trip logic ttst failure occurred. The response was that
engineering involvement was needed to determine the scope of the )roblem
(and inoperability) so that the appropriate TS action could be tacen.
'
Af ter the inspectors discussed the issue with the operations shift
personnel, they recognized that determining the scope of the
inoperability was independent of the time after which actions were
required. .
Engineering personnel determined that a failed circuit card caused the
test failure. The circuit card was replaced, and testing was com)leted
successfully. The action statement was terminated at 4:30 p.m. tlat
same day.
c. Conclusions
The inspectors concluded that operations personnel inappro)riately
entered the TS action statement more than one hour after t1e test
failure of a reactor trip system logic function. The failed function
was repaired, successfully retested and returned to service before TS
- actions were required.
02 Operational Status of Facilities and Equipment
02.1 Cold Weather Protection Preoarations
a. Insnection Scone (71714)
.The inspectors reviewed Nuclear System Directive (NSD) 317. Freeze
fruiection Program. Revision 1: interviewed the freeze protection
coordinator: reviewed procedures and work orders to determine what
actions had been taken to prepare for cold weather; and independently-
inspected some vulnerable equipment exposed to the environment for
freeze protection,
b. Observations and Findinas
The licensee completed NSD 317 in March 1997. The NSD governs the
freeze protection alans at all three Duke nuclear stations. During the
previous cold weatler season, the NSD had not been finalized and a
formal program was not in place for ensuring that effective measures
"
Enclosure 3
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were in place to protect plant equipment and systems from sub freezing j
conditions.
i
The station assigned a freeze protection coordinator to monitor the
status of preparation activities. An equipment freeze protection ,
program was developed to identify operating plant systems, structures !
and components (SSCs) that may be subjected to freezing temperatures
during the cold weather season. An engineering support program was ,
initiated to ensure that specific freeze protection measures for
vulnerable SSCs were identified to facilitate the preparation and
completion of a pre-seasonal eneckout. Pre seasonal checkouts were '
executed via various model work orders for inspection and testing of
electrical heat trace and instrument box heaters. The freeze protection
plan includes surveillance procedures to inspect SSCs considered to be
critical to plant operation on a monthly interval and as necessary
during extreme cold weather.
The inspectors discussed the status of freeze protection preparations
.with the freeze protection coordinator. According to the coordinator,
the annual preventive maintenance activities had been completed by the
end of the inspection report period, and work-orders or work requests -
had been generated to address identificd discrepancies. The freeze .
protection coordinator had performed inspections of vulnerable areas and
'
submitted a list of discrepancies to the maintenance orcanization. Most
ofthesediscrepancieswereresolvedbytheendoftheInspection ,
period.
The inspectors conducted inspections of equipment that historically had
been vulnerable to cold or freezing temperatures. The inspectors >
notified the freeze protection coordinator of a few minor discrepancies.
The inspectors also reviewed the work orders associated with the annual ,
preventive maintenance (PM) and verified that work had been completed.
c. Conclusions
Nuclear System Directive 317 provided structure and delineated
responsibilities for freeze protection. Proceduralized activities were
initiated and completed in a timely manner, and work orders or work
requests were initiated to resolve identified discrepancies. The
inspector concluded that the licensee's efforts to effectively protect
plant equipment and systems from freezing conditions had improved since
the previous cold weather season.
02.2 Prompt Onsite Response to Events (93702)
.
The licensee reported four unrelated events to the NRC Headquarters
-
Operations Officer via the Emergency Notification System in accordance
with 10 CFR 50.72. The following events were all reported in a timely
fashion with sufficient information being provided,
Enclosure 3
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Oil Sheen on Lake Wylie on October 15 i
!
On October 15. the inspectors were notified of a thin oil sheen that was
discovered on Lake Wylie during a main fire pump test. The source of :
the oil was determined to be an overflowing pump bearing reservoir which
caused oil to spill around the fire pump motor and eventually into the
lake. The oil sheen was contained by a boom beneath the pum) structure. i
The licensee notified the South Carolina Department of Healt1 and
Environmental Controls and the National Response Center, which in turn
required notification of the NRC 'q accordance with 10 CFR
50.72(b)(2)(vi).
Plant Shutdown Reauired By TS on October 20
As discussed in Section 08.2 of this report. the licensee initiated a >
'
Unit 2 shutdown on October 20 when it entered TS Limiting Condition for
Operation 3.6.3 action statement following the inoperability of the 2A ,
'
steam generator main feedwater isolation valve 2CF-33. The unit was
helti at 15 percent power after the valve was deactivated and gagged
shut. The valve was repaired and a forced unit shutdown was avoided.
This item was reported to the NRC in accordance with 10 CFR
50.72(b)(1)(1)(A).
'
)otential Non Conservatism in a Calculation used to Distinauist Between
Reactor Coolant System F' ow Versus leactor Power Restricted anc
)rohibited 02eratino Rea1ons
On October 23. the licensee reported a potential nonconservatism in each !
units' 15 3/4.2.5. Departure from Nucleate Boiling (DNB) Parameters.
Figure 3.2 1. Reactor Coolant System Total Flow Rate Versus Rated
Thermal Power - Four Loops in Operation. Essentially, licensee
personnel determined that the curve provided in Figure 3.21 for each
unit permitted potential plant operation at reduced power levels with
reactor coolant system flow rates that could possibly challenge DNB
ratio design limits for certain analyzed transients. As a precaution,
until this condition could be resolved, the licensee implemented
administrative restrictions requiring reactor coolant system flow rates
to be maintained above those specified as the permissible operation
region for 100 percent power. These restrictions were verified to be in -
place by the resident inspectors. Long term corrective actions included
completing an analysis to allow a revision to the TS requirements to
eliminate the non-conservatism. This item was reported in accordance
with 10 CFR 50.72(b)(2)(111)(D). The licensee documented this issue in
a 30 day written follow up Licensee Event Re) ort (LER 50-413/97-007)
near the end of the inspection period. Furtier inspector review of this
issue will.be conducted and tracked under the LER in subsequent'
- inspection reports.
Enclosure 3
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8
i
Potential for Overfilli.Da Steam Generator Durina a Postulated Accident
'
On November 18. the licensee re)orted a single failure vulnerability
involving the loss of 125 Volt )C vital instrument and control
distribution center EDE or ELF during a postulated steam generator tube
rupture event coincident with a loss of offsite power. The licensee
determined, following a detailed analysis that the loss of either of
these busses would result in the inability to isolate turbine driven
auxiliary feedwater pump flow to a ruptured steam generator. The steam
generator would be pntentially overfilled, resulting in uncontrolled
releases of radioactivity to the atmosphere.
Because of this potential, and until further corrective actions are
determined, the licensee implemented conservative administrative
controls limiting the amount of dose equivalent iodine in the reactor
coolant system to ensure the consequences of the Chapter 15 steam
generator tube rupture analysis remain bounding. These restrictions
were contained in procedure CMP 3.4.17.1. Primary Chemistry. Revision 28
and verified by the inspectors. At the close of the inspection period,
the licensee was evaluating several o)tions for long-term corrective
actions. This item was reported to tie NRC in accordance with 10 CFR
50.72(b)(1)(11)(B).
04 Operator Knowledge and Performance
04.1 Residual Heat Removal (RHR) System Potentially Placed in An Unanalyzed
Condition
a. Insoection Scope (71707)
The inspectors reviewed the circumstances involving an August 20. 1997,
tagout in which the RHR system was potentially ) laced in an unanalyzed
condition. The inspector reviewed the Catawba Jesign Basis Document
(DBD) CNS 1561.ND 00-0001: the UFSAR. Section 6.3 and Chapter 15: and
PIP 2-C97 2722. The inspectors also ruiewed the licensee's root cause
investigation, completed during this inspection period. and discussed
this issue with engineering and operations personnel.
b. Observations and Findings
Residual heat removal system valve ND59B 1s a motor 0)erated globe valve
located in the minimum flow lines of the 18 and 2B RH1 pumps. Valve
N059B and its associated miniflow line normally protect either B train
pump from cavitation at low flow conditions or following a complete loss
of suction during the decay heat removc1 or emergency core cooling modes
of operation.
On August 20. 1997, at 3:38 a.m. operations issued removal and
restoration (R&R) tagout 27-1498 to support work on the Unit 2 Train B
RHR miniflow loop. Unit 2 train B RHR was declared inoperable and
Enclosure 3
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entered into the Technical Specification Action Item Log (TSAIL). The
planned work included a miniflow valve controlling set point
modification, a gauge replacement, and an instrument calibration. The
R&R tagged valve ND59B open with power removed. Approximctely 2-1/2
hours later at 6:00 a.m., work control personnel realized that the
tagout was in conflict with Catawba Design Basis and Criteria,
Specification CNS-1561.ND-00-0001. Revision 5, which stated that "with
ND59B stuck open and incapable of closing, the resulting diversion of
RHR ) ump fluid to the recirculation loop is an unanalyzed condition."
At tiat time, operations personnel cleared the tagout and closed the
valve. Station PIP 2-C97-2722 was initiated and the licensee later
determined that a past operability evaluation was required.
Engineering
September 1997, is, personnel
and concluded completedthat the the
RHRpast
systemo)erability evaluation
was operable during on
the time the miniflow valve was tagged open. The inspectors discussed
this conclusion with licensee personnel and upon reviewing UFSAR Table
6-7, Catawba Nuc1 car Station Emergency Core Cooling System Flow Rates,
arrived at the same conclusion. This was based on the fact that the
RdR flow capacity (approximately 500 gallons per minute) normally
diverted from the reactor coolant system recirculation loop by miniflow
valve ND59B, when subtracted from the total RHR flow ca)acity, still
resulted in sufficient RHR flow being delivered to the RCS during the
post-accident recirculation mode. However, the inspectors considered
the tagging ciscrepancy to represent a problem that could have had
adverse plant impact.
A root cause investigation of the improper tagging incident was
completed by the licensee during this inspection period which concluded
that engineering persornel improperly communicated a 1993 DBD revision
to affected groups. The RHR DBD had been revised then to provide a
discussion of the "unanalyzed condition." However, this analysis did
not take into consideration lesser flow requirements assumed in UFSAR
Table 6-7 for the post-accident long-term recirculation mode of
operation, the time at which the RHR system alignment would be changed
and the miniflow valve would become a diversion flow path.
The inspectors considered other human performance weaknesses contributed
tc the tagging error. When the calibration work order from which the
tagout was generated (PM 95054445) was developed in July 1995, a note
was added for operations personnel to tag the tr niflow valve open. i
Although personnel involved in planning the set point change
modification were aware of the DBD statement, and verbiage was included
in the modification package to ensure the tagout was correct and would
not place the RHR system in an unanalyzed condition, the set point
modification was performed under an existing tagout for the preventive
maintenance work, which had the valve opened on August 20.
The inspectors noted that the DBD had not been consulted when the tagout
associated with the August 20, 1997, activities was developed a week
Enclosure 3 I
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10
earlier. The inspectors, discussed this with licensee management, who
stated that the 1995 PM work order note likely contributed to an
overr'. ding operating philosophy that tagging the valve open during
maintenance was appropriate. Several corrective actions were generated
for PIP-2-C97-2722, including developing a policy for communicating
engineerits document revisions to affected groups and designating a
specific work management mtem panel to document engineering
recommendations and spe notes. The inspector asked whether or not
the DBD reference to t b analyzed condition" would be deleted to
reflect the engineering e.idlysis discussed above. Licensee management
indicated they would evaluate changing the DBD.
c. Conclusions
The inspectors concluded that although having the Unit 2 train B RHR
pump out of service with valve ND598 de-energized open did not place the
plent in an unanalyzed condition, examples of poor performance were
identified concerning activities leading up to the valve inappropriately
being tagged open during plaaned maintenance.
08 Miscellaneous Operations Issue (92901)
08.1 (Closed) LER 50-414/95-01: Reactor Trip Due to Closure of a Main Steam
isolation Valve
The event described in this LER involved an automatic reactor trip due
to the failure of a digital optical isolator (D01) in the B main steam
isolation valve control circuit that caused the valve to close. This
LEP was discussed in NRC Inspection Report 50-413.414/97-12 and remained
opea pending further NRC review,
Planned corrective action 2 was to develop a PM program to periodically
monitor continuously enc gized E-max 00ls with model numbers 175C156 and
175C157 in critical applications. Instead, the licensee initiated a PM
program to re) lace DOIs that perform a control function and that have AC
voltage for t1eir inaut )ower su) ply every twelve years. The inspectors
determined that the 4RC 1ad not )een apprised of the change.
In light of recent DOI failures that resulted in manual reactor trips in
July and August 1997, the inspectars asked the licensee if monitcring
the D01s could have revealed the root cause (degraded resistors) of the
r eent DOI failures. The licensee indicated that the test methodology
that would have been used to periodically monitor the D01s would not
have revealed degraded resistors (the cause of the 1997 failures). The
inspectors concluded that, while testing the DOIs had the potential to
reveal degraded D01s during periodic testing, the likelihood that it
would have done so was low. Therefore, tl. commitment change did not
substantially reduce the opportunity to identify degraded DOI resistor::
and take subsequent actions to prevent the 1997 001 failures.
Enclosure 3
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According to Nuclear System Directive (NSD) 214. Comitment Management:
ts are a source of NRC-
EProgram Revision
comitments. 2.-214.8.4i
Section Licensee Event
Remove or-Rep" Change a.Comitment. stated .'
- that the regulatory compliance-(RGC) group should be notified if- a
comitment change is needed, and that RGC will determine, in part. if
'
-
the NRC- should be notified.- The NSD'incor> orates a 1994 draft document
prepared by the Nuclear Energy Institute (4EI). entitled " Guideline for
Managing NRC Commitments."
-
.
Acc0rding to NSD 214, when a comitment is changed, the original
comitment-will be modified with a description of the change in the
appropriate section of the PIP database (which is used to track NRC
comitments-to resolution). The NSD further stated that if the change ;
is determined to be significant enough, a new commitment may be-
generated. However, proper cross-references shall be provided to link
-the original commitment to the revised commitment. The licensee
determined that the NRC was not apprised of the commitment change
'
because the corrective actions representing the comitment were
improperly cross-referenced. As a result, the changed corrective action
was not identified as an-NRC commitment, and RGC was not notified.- The
inspector concluded that the licensee failed to notify the NRC of a
-
comitment change regarding planned corrective actions delineated in LER
50-414/95-01. This issue is charact rized as one example of Deviation
50-413.414/97-14-02: Changing NRC Comitments Without Properly
Notifying the NRC.o ,
4
'
This item is closed.
.
08.2 (Ocen) Violation R0-413/97-08-01: Inadequa'.e Alarm Response Results in
Inadequate and untimely Correctite Actions for Valve Operability
' Determination
.The inspectors reviewed Violation 50-413/97-08-01 for an April- 3.1997.
.
-incident following a similar occurrence on October 20, 1997. where a
feedwater isolation valve became inoperable after a nitrogen leak
developed on its accumulator.
On the morning of October 20, 1997. just before shift turnover, the Unit
2-control room operators received a computer alarm indicating low--
nitrogen gas pressure in the accumulator associated with the 2A steam
,
generator main feedwater isolation' valve. 2CF-33. The valve was-
declared inoperable and TS Limiting Condition for Operation (LCO) 3.6.3
was imediately entered. Nitrogen pressure was checked and found to be
at 1640.psig, which was below the low operability limit of 2050 psig.
The' accumulator-was recharged to 2760 psig and the TS'LC0 was exited.
- Approximately 2-3 hours'later at 9:55 a.m., another low aressure alarm
~
-
was received, ard operators again entered the 4-hour TS _C0 action-
requirement to either return the valve to operable status, de-energize
(gag) it shut, or initiate plans to be in Hot Standby in the following 6
hours. After the second alarm. the accumulator was found to be at 1810
Enclosure 3
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spsig and a-leak was detected from a solenoid valve at the actuator. The i
nitrogen: accumulator was1again recharged but could not be maintained ,
above the low pressure limit. Technical Specification 3.6.3 required !
that= the plant to be in Hot Standby (Mode _3) by _7:55 p.m. l
Plant management decided that a power reductior, would be initiated-
- shortly after 1:00 ).m.- .The unit was reduced to approximately 15
rpercent power and t1e valve was gagged shut just before the TS LCO
Laction to be in Hot Standby was required, thus avoiding a forced.
shutdown. A leaking 0 ring at a solenoid-to tube connection was
detected. The solenoid was re) laced and the valve was tested
successfully. Unit 2 exited tie LCO action statement and was returned-
. to 100 percent-power on October 21.
-
.
e
'
The inspectors reviewed Violation 50-413/97-08-01 which documented a. .
timilar occurrence on April 3.1997. Involving feedwater isolation valve )
1CF-51. Following the April 3. 1997, incident. plant personnel
determined that the control room 0AC alarm set point was set at or near
- the pressure at which the valve became inoperable. One of the planned
corrective actions-documented in-the licensee's written response to the.
violation: dated July 22, 1997. was for engineering personnel to evaluate
whether'the alarm set-point could be raised to provide more margin
-
between it and the operability limit thereby allowing operators more <
time to react to an actuator leak. According to the licensee's letter.
>' ~
4
this action was to be completed by September 30, 1997. Following the. J
-
.0ctober 20 ;1997.: occurrence, the inspectors inquired about the status - i
i '
-of the engineering evaluation. Licensee personnel indicated that it had
not been performed and that engineering personnel had been internally
granted an extension of the due date from the safety assurance group to
October 31.
The inspectors noted that the NRC had not been notified of this
commitment change and upon inquiring furt'ner, were told that an
administrative error in the data-entry process for the PIP associated
with the- April 3.1997, incident allowed engineering to be granted an
extension without evaluating the impact of changing this commitment.
Upon discovery of the error, licensee personnel corrected it in the PIP
database and an engineering evaluation was completed by the new
- deadline. A modification was subsequently initiated to raise the
-accumulator alarm set points for-all of the feedwater isolation valves
and provide greater margin above their operability' limits.
The inspectors determined that the failure to perform the engineering
evaluation in a timely manner further increased the chances of a
feedwater isolation valve becoming inoperable prior to the control room
- receiving the alarm. 'The inspettors reviewed the documents associated
-
with NRC commitment management programs described in Section 08.1 above
and determined that the failure to perform this evaluation by September
30. 1997. constituted a Deviation from NRC commitments. This issue is-
-
characterized-as the second example of Deviation 50-413.414/97-14-02:
Enclosure 3
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Changing NRC Commitments Without Properly Notifying the NRC. Violatir
50-413/97-08-01 will remain open pending completion of all of the
licensee's corrective actions and further review by the inspectors.
Maintenance
M1 Conduct of Maintenance
[ M1.1 General Comments (61726)
The inspectors observed portions of the fol h ing surveillance and
inspection activities:
. NPP-312. Nuclear Fuel And Core Component Receipt Inspections.
. PT/1/A/4200/09A. Auxiliary Safeguards Test Cabinet Periodic Test.
- PT/1/A/4400/06A. Nuclear Spray (NS) Heat Exchanger 1A Heat
Capacity Test.
. PT/1/A/4400/09. Cooling Water Flow Monitoring For Asiatic Clams
And Mussels Quarte'ly Test.
. PT/1/A/4200/04B. Containment Spray Pump 1A Performance Test.
. PT/1/A/4350/0028. Diesel Generator 18 Operability Test,
s Retype No. 28
\ During these activities. the ins ectors noted proper use of procedures,
properly calibrated measuring and test equipment effective radiological
controls, and adequate communication between personnel performing the
tests.
M1.2 New Fuel Movements (62707)
The inspector observed movement of new fuel from the dry storage racks
to the spent fuel pool in pre]aration for the upcoming Unit 1 end-of-
cycle 10 refueling outage. T11s activity was conducted under Work Order 97063472-01. Move New Fuel from New Fuel Vault to Spent Fuel Pool. The
technicians used procedures OP/1/A/6550/011. Retype 21. Internal
Transfer of Fuel Assemblies and Components: and OP/1/A/6550/006. Retype
11. Transferring Fuel with the Spent Fuel Manipulator Crane. The
inspector noted. for the fuel assemblies observed, that they were placed
correctly in locations referenced by the procedure attachment. Proper
radiological controls were observed. Crane chNklist prerequisites had
been completed as required. This work activity was conducted well.
M8 Miscellaneous Maintenance Issues (92902)
M8.1 (Closed) Inspector Follow UD Item (IFI) 50-413.414/97-08-04:
Reportability of Nuclear Service Water (NSW) System Actuations.
This item was opened to determine the reportability of NSW system
actuations. The licensee generated station PIP 0-C97-1715 to document
the clarification. The licensee determined that the NSW system is
Enclosure 3
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required for support of the Engineered Safety Features (ESF). As such.
the NSW system is characterized as an ESF support system in the UFSAR.
Section 7.3.1.1.5. ESF Support Systems. The licensee concluded that,
since the NSW system is not an ESF and since 10 CFR 50.72 and 50.73
require licensee's to report any event or condition that results in a
manual or automatic actuation of any ESF. actuations of the NSW system
were not reportable.
The inspector reviewed ap)licable sections of NUREG 1022. Event
Reporting Guidelines 10 C R 50.72 and 50.73: NUREG 0800, the Standard
Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear
Power Plants. Light Water Reactor Edition. June 1987: the UFSAR: and
Nuclear System Directive 202. Reportability. Revision 8. The
characterization of the NSW system as an ESF support system was in
agreement with the SRP which referred to service water systems as
auxiliary systems that directly support ESF systems. However. Chapter 6
of the UFSAR. Engineered Safety Features, does not contain a listing of
ESF systems: a listing, which does not include the NSW system, is
located in Nuclear System Directive 202. Reportability. Revision 8
Appendix A. Engineered Safety Features. Chapter 7 of the UFSAR.
Instrumentation and Controls. Section 7.3.1.1.1 lists ESF functions
initiated by the Engineered Safety Features Actuation System (ESFAS):
the NSW pumps which provide cooling water to the component cooling
sy> tem heat exchangers and are thus the heat sink for containment
cooling, are listed.
Based on this r;"tiew, the inspectors determined that NSW system
.
actuations are not reportable. This item is closed.
III. Enaineerina
El Conduct of Engineering
El.1 Operation Of Dual Function Containment isolation Valves-Temocrarv
Instruction (TI) 2515/136 (Closed)
a. Inspection Scoce
The inspectors used TI 2515/136. Operation of Dual Function Containment
Isolation Valves, to determine if the licensee had procedures in place
to remotely close containment isolation valves when required while a
safety injection or a containment spray signal was present. The
inspector discussed this issue with engineering personnel, and reviewed
the UFSAR and design basis documentation.
b. Observations and Findinas
The Tl included a questionnaire survey with four items. Item 1
requested that the inspectors identify the dual function valves as
listed in the UFSAR and determine whether differences existed in the
Enclosure 3
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plant. Licensee personnel provided a list of containment isolation
valves, which included dual function valves. The inspector compared the
valve list to the valves shown on UFSAR Table 6-77, Containment
Isolation Valve Data. All valves identified by the licensee were found
to exist in Table 6-77.
During the ins)ectors' review, it was noted that two of the valves
listed SA-1 ()enetration M 261. B Main Steam to Auxiliary Feedwater
Pump Turbine) and SA-4 (Penetration M-393,~ C Main Steam to Auxiliary
Feedwater Pump Turbine), did not comply with 10 CFR 50. Appendix A.
General Design Criterion 57. General Design Criterion (GDC) 57. Closed
System Isolation Valves, specifies that each line that penetrates
primary reactor containment and is neither part of the reactor coolant
pressure boundary nor connected directly to the containment atmosphere
shall have at least one containment isolation valve which shall be
either automatic, locked closed, or capable of remote manual operation.
Valves SA-1 and SA-4 are manual gate valves and normally in the locked
open )osition. These valves and containment penetrations exist in both
Catawaa Units 1 and 2. The GDC 57 noncompliance had been previously
identified and an exemption request (from GDC 57) was submitted on
September 2, 1997. This item is being tracked as Unresolved Item 50-
413.414/97-14-03: Noncompliance With 10 CFR 50. Appendix A. General
Design Criterion 57 Closed System Isolation Valves.
Item 2 asked whether or not a safety-related dual function valve could
be closed from the control room with a switch and remain closed in the
presence of a containment spray or safety injection signal. As
indicated by the licensee's list, reset and closure capability existed
with remote. manual control on all safety-related dual function valves
with the exception of SA-1 and SA-4, which were locked open. Some
valves, as indicated on the licensee's list, would require the emergency
diesel generator (EDG) load sequencer be reset in addition to normally
resetting the ESF (or Safety Injection) signal. The EDG Load Sequencer
system engineer indicated that resetting the ESF signal would not affect
the configuration or operating status of any safety-related equipment,
and that resetting the EDG Load Sequencer would not affect the EDG or
any com)onents being powered from the safety-related 4160 volt busses.
While tie inspectors were familiar with the reset capability for the
safety injection signal, further NRC inspection was necessary to verify
thet resetting the EDG Load Sequencer during an accident would not
adversely impact the operation of safety-related plant equipment. This
review effort will be conducted under URI 50-413,414/97-14-03 discussed
above.
Item 3 requested, for valves that do not have a switch for remote
closure [i.e. , SA-1 and SA-4]. if any proceduralized method existed
(such as deenergizing circuits or lifting leads or installing leads)
that would facilitate remote closure. Since valves SA-1 and SA-4 are
locally operated manual valves, no remote method of closure existed.
Enclosure 3
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- Item 4 requested,lfor valves that do not have'any remote method of- ;
closure:available [i.e.. -SA 1 and SA-4], whether there were any other
means that the licensee had to close the. isolation valve. The licensee -
.
provided a list of eight emergency procedures that contain provisions to
isolate Penetration M-261 or M 393'as required. Two isolation options ;
were provided. - The first-option utilizes the SA-1 or SA-4 valve as
required located in the plant doghouses. The second option-isolates the
>
penetration by closing valves SA-3 or SA 6 located downstream of:SA-1-
and SA-4 =in the Penetration Area if SA-1 and SA-4 were inaccessible.
j
.
The inspectors reviewed these procedures and found that the procedural
guidance to' establish containment isolation manually for penetrations.M-
261 and M-393 was available to operators when needed. ,
-c. " Conclusions
=An unresolved item was identified concerning containment penetrations
-associated with steam supply lines-to both units' turbine driven
i auxiliary feedwater h h in compliance with 10 CFR 50,
1A>pendix A, GDC 57. The' pumps,
licensee had w icsubmitted
were notan exemption request to
t1e NRC for this-issue. Remote manual closure capability. existed for
-
- dual function containment isolation valves; however, the action involved
resetting the emergency diesel generator load sequencer, an action
requiring further evaluation to be conducted under the above-mentioned
. unresolved item, ,
,
E2 Engineering-Support of Facilities and Equipment i
p
E2.1 Solid State Protection System (SSPS) Testino Deficiency
a. Insoection Scooe (37551)
The inspectors reviewed the licensee's discovery of a logic testing
. deficiency associated with both trains of each unit's SSPS - November
11, 1997.
b. Observations and Findinas ,
The test deficiency-involved the failure to perform adequate testing of
two universal-cards associated with feedwater isolation functions and
the P-10 source range. nuclear instrumentation reactor trip block
permissive. . Theiuniversal cards contained previously unidentified
parallel: circuit paths which were not being isolated and independently
verif ted to actuate the logic circuitry associated with each function.
sBoth units entered TS 4.0.3 after identifying the missed surveillance
testing; The procedures were revised and the testing conducted-
,
' satisfactorily before each unit exited TS 4.0.3.
The test anomaly was identified by personnel in the licensee's General-
Office and.was immediately communicated to the SSPS vendor and to
various other nuclear power facilities via operating experience data
-Enclosure 3
,
e _
^. - -
'
, .
. , . , . ,
. .. .- . - - . . . -. .
-. - - - . - - . .
'
.
. .- .
-
,
17- )
-bankst Several facilities have since identified the same or.similar;
deficiencies in their SSPS logic testing procedures,
~
c. Conclusion
-The licensee has issued LER 50;413/97-08 to document the' missed TS-
surveillances and discuss the safety consequences and corrective actions- 1
- taken for the deficiency. Further NRC-review will-be conducted during
closecut of the LER. ;
Miscellaneous Engineering Issues (92903)
'
E8
E8.1-- (Closed) Insoector Follow UD Item 50-413.414/96-18-04: Quam. fication
,
' of_ Refueling Water-Storage Tank (FWST) Heat Losses Through Tank Roof
--Including a Wind Velocity Factor.
This-item involved minor modification CNCE-8309 to de-energize one.of >
four-FWST heater clusters. The licensee performed an evaluation to
demonstrate that minimum required tank tem)erature of 70-degrees
- Fahrenheit (*F) could be maintained with t1e three remaining: heaters.
-The evaluation involved a calculation. CNC 1249.00-00-0065. Operability
- Determination for PIP 1-C96-1870 - Heater Sizing for the FWST, that .
'
quantified-heat losses from the tank assuming a minimum temperature of -
5*F and wind velocity of up to 20 miles per hour (mph). The calculation " -
-
>1ndicated that'the total FWST~ heat lossiat was-81.88 KW. r -
,t
The inspector;noted that-the calculation: accounted for wind-induced heat-
losses from the tank walls, but not from the tank roof. To address this
observation the licensee completed Revision 1 of calculation CNC-
1249.00-00-0065 and concluded that, accounting for heat losses from the
FWST roof assuming a 5 m3h average wind velocity the total FWST heat
< loss was 93.46 KW at an WST temperature of-75*F and environmental
temperature of -5*F. With one heater cluster inoperable and-de-
-energized, the total heater capacity available is 90 KW. The-licensee
indicated that the environmental tem ereture selected for design
comparison in the calculation was be ow the coldest temperature ever
recorded at the site. it was unlikely that temperatures would drop to
that temperature. The licensee also indicated that the heat loss would
t 'be 87.42 KW if the tank wall temperature were assumed to be 70 F (the TS
value)'. and therefore within the heating capacity of the three remaining
heater clusters. Based on these and other conservative heat loss-
assumptions applied to the calculation, the licensee asserted that the
-
remaining heater capacity was marginal to maintain the FWST at 75 F. but
that it was adequate to prevent a temperature drop below the TS-required
value of 70*F.
.
. Refueling water storage tank temperature-indications are available in
'
the control room. In addition, a low
a
.
temperature alarm will be
. generated at 74 F. The alarm response would be to dispatch an operator
to verify heater operation. A Lo-Lo temperature alarm would be
. Enclosure 3
.
a
r -- Wwy wr,y,u
- . - _ - . ~ -- - - - - - - - - - - . _
,
. :.
-
,
'
.
c18_- ;
'
+ - generatsd when. tank temperature reaches 70*Fi The response then would.
-
= be to declare the FWST inoperable per the appropriate TS, Based on the; i
heat loss calculation, monitoring-capabilities and response procedures. . .
the ins >ector concluded:that-FWST temperature was not likely to-_ drop 4
,
.
below t1e TS required value of 70*F as:a result of this minor 1 . . 7
modification. -Shouldsa-low temperature alarm be ger.erated, effective-
measures were in place to ensure that action will be-taken to correct
the. low temaerature condition or. place the unit in a safe condition; In ,
. addition.- tie licensee planned to correct the heater leakage, re-
energize the heater and return' it to service during the upcoming end-of-
icycle 10 refueling outage, scheduled to begin in late November.-
. The ins)ector noted that the wind velocity assumed for heat _ losses from
the tan ( walls was 20 mph.-whereas 1t was assumed to-be only 5 mph.for-
- heat losses from the tank roof.- While no explanation for this
' '
discrepancy was- 3rovided in the calculation, the inspector concluded
that, since the 1 eater was:to be returned to service in December 1997, ;
'
this discrepancy did not pose a safety concern. This item is closed.
b E8.2_ (Closed) Unresolved Item-50-413.414/97-11-04: Use of Aluminum High
Efficiency Particulate Air (HEPA) Filter Separators Inside Containment.
'
- This item involved the licensee's ~1dentification of aluminum HEPA filter
i
separators in the containment ventilation system's containment auxiliary '
w a- charcoal filter units?(CACFUs) that had~not been accounted for:in the
~
t :
- ~
station's aluminum-inventory records'. :The licensee initiated an >
l -- ;
evaluation _to determine the-root cause'of the inappropriate material.
,
usage.
.
The licensee's evaluation revealed that the HEPA filters had contained
-
aluminum since 1986 or before. Design Specification CNS-1211.00-3,
Containment Auxiliary Charcoal Filter Units, Section 5.5, High
,
Efficiency Filter Section, states that " Separators, if used, shall be
304 stainless-steel." The licensee determined that the original HEPA
.
filters were a separatorless, nuclear grade filter without aluminum.
However, at some undetermined point in time, the station began to use a
- could not locate any documentation to support the change in filters and
'
terminated the ioot cause evaluation, which was not likely to reveal the
origination of the discrepancy.
- The. inspector concluded that, although the error leading-to the.
discrepancy had occurred over. ten years ago, the licensee has since
"
established a 3rocess that would prevent a similar oversight from
,
occurring _ at tae:present time. - A changeLin filter components (or other -
. components:inside containment) would involve the-modification process.
- Essentially, NSO.301. Nuclear Station-Modifications.: dated September 30, ^
1997? required that a Technical Issues Checklist be completed for- any
temporary.: minor, or nuclear station;(permanent-and major) modification.
- _ The Technical 7 Issues Checklist, located in Appendix A of the NSO. 1
' Enclosure 3-
,
. -
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' '
. , . -
-
- -.- --r ,, -
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- -
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-19
addressed containment. issues and hydrogen control. The question "Does
the change add aluminum or zinc that could potentially increase the
amount of hydrogen gerierated inside the containment post accident?"
would likely prompt a review for this potential during the current
modification process.
The licensee re-evaluated the original hydrogen generation calculation
and determined that the amount of hydrogen generated inside containment
following a design basis accident that would oe produced by the
additional aluminum did not exceed revised allowable limits.
Therefore, the safety consequences were minor. However, measures were
not effective in preventing the selection of thesc filters for use in an
unsuitable application as required by 10 CFR Part 50. Appendix B.
Criterion III. This constitutes a violation of minor significance and
is characterized as a Non-Cited Vielation (NCV). consistent with Section
IV of the NRC Enforcement Policy. This item is identified as NCV 50-
413.414/97-14-04: Failure to Control Use of Aluminum Inside The
Containment Building.
The ins)ector determined that the licensee hed been informed by
Westing louse of the potential that certain HEPA filters were being
manufactured with aluminum separators. The information was conveyed via
Vendor Information Letter 96-30 in September 1996. The licensee's
response to the information was to consult the design specification
(CNS-1211.00 3) to determine if aluminum was specified.' Upon finding
that the specification required the.use of 304 stainless steel, the
licensee concluded that the CACFU's " EPA filters did not contain _.
aluminum. The inspector concluded 4 at the original review in response
to the Westinghouse information letter was cursory and ineffective in
revealing this discrepancy. The inspector reviewed the revised hydrogen
generation calculation: no concerns or discrepancies were identified.
This item is closed.
IV. Plant Sucoort
R1 Radiological Protection (RP) and Chemistry Contro'
R1.1 Tours of the Radiolooical Control Area (RCA)
a. Insoection Scoce (71750)
The inspectors periodically toured the RCA during the inspection period.
Radiological control practices were observed and discussed with
radiation protection personnel, including RCA entry and exit controls,
survey postings, and radiological area material conditions.
Enclosure 3
- . _ . .
_ - . . .. .. m .. _ _ _ _ _ . _ _ _ _ _ . . . ._
a c
. .-
d
'
- 20
b'. : Observations and Findinas'
,
100 November.17. the inspectors noticed an RCA exit door propped wide-
_open with a brick. -The doorway was on the 594 foot elevation of the-
-
.
auxiliary building at the end of corridor number 517 and provided RCA'-
,
-access from the outside. - Two stanchions with a roped sign hanging
-between them ncrmally blocked 6ccess past the door into tns: RCA but the
- -stanchions and' sign had been moved to the side and out of viewt -The
- sign was intended to warn personnel that they were about to enter the 3
- RCA and directed them to contact radiation protection, personnel for
_
assistance. At the-time of the ins)ector's obstrvation. no personnel
were present to control access at t11s RCA entry point. 1
'
The inspectors notified radiation protection (RP) 3ersonnel who- -
'immediately responded to the location and closed tie-door. Later. the
'
same: sign was attached to a swing gate which was placed at the entrance.
,
-The gate would close after allowing personnel pre-appraved access across-
.the boundary. -The inspectors were in. formed by RP personnel that a:
maintenance crew-had been using the door to bring scaffolding into the
- plant in preparation.for the upcoming Unit I refueling outage. The-
maintenance crew had received permission from RP to use the door. The
crew had moved the, sign blocked the door open, and temporarily left the
drea to Conduct other activities.
.~ "
r :The inspectors discussed with" licensee personnel the need to properly- *
-
.
control access:to the RCA. Licensee personnel generated PIP 0-C-97-3670:
>
^
to document this deficiency. The incident was discussed in a-subsequent: -
c daily management meeting. In addition to the immediate corrective
4 actions above. RP management discussed this incident with scaffolding
supervisors who later discussed it with their crew members to reinforce
proper procedures for entering the RCA.
-The inspectors later observed that general access to this area from the
outside was limited to the scaffold crew because of a second external
n . barrier that had been _placed outside to control personnel traffic.
While this barrier was not intended for RCA access control it reduced
the significance:of the inspectors' finding.
c. Conclusions-
An example of poor performance was identified related to an RCA boundary
.
-being_ compromised. -This' minor _ discrepancy was immediately corrected by_
-
plant personnel and properly addressed by licensee management.
s3
,
RI.2 Transoortation of Radioactive Materials
'
-ai -Insoection Scone (86750) -
L '
The inspectors evaluated the licensee's transportation of radioactive
materials programs for implementing the revised Department of
Enclosure 3
.
., , - , , , , - ... .,. - , , . ~ . - - 4 , .. ,----
.v. . * e. ---.w
-
'
. .
21
Transportation (00T) and. NRC trans)ortation regulations for shipment of
radioactive materials as required )y 10 Code of Federal Regulations
(CFR) 71.5 and 49 CFR Parts 100 through 177.
- b. Qb.servations and Findinos
The inspectors reviewed procedures and determined that they adequately
addressed the following: assuring ' hat the receiver has a license to
receive the material being shipped; assigning the form. quantity type,
and proper shipping name of the material to be shipped: classifying
waste destined for burial; selecting the type of package required:
assuring that the radiation and contaminatis : limits are met: and
preparing shipping papers.
Licensee's records for the six shipments of radioactive material
performed in 1997 were reviewed and the inspectors determined the
shipping papers contained the required information. The inspectors also
determined the licensee had maintained records of shipments of licensed
material for a period of three years after shipment as required by
10 CFR_71.91(a). In addition, the licensee )ossessed a current
certificate of approval (NRC Form 311) for t1eir " Quality Assurance
Program Description for Radioactive Material Shipping Packages Licensed
Under 10 CFR 71." The licensee had also maintained current NRC
certificate of compliance for the NRC approved cask in use.
The inspectors reviewed the training records for selected individuals
-
authorized to sign shipping papers and: handle radioactive waste which
included a w area su)ervisc who was assigned to the area of
transportati .. the weet of the inspection. The training specifically
addressed the new rules for the following to)ics: low specific activity
(LSA) and surface contaminated object (SCO) idzards, definitions, and
requirements: placarding, labeling, and marking of vehicles and
packages: use of Systems Internationals (SI) units on shipping papers,
labels, and emergency response instructions after April 1.1997: package
selection: waste classification: shipping papers; and receipt procedures
and surveys. The inspectors concluded that personnel involved with
radioactive material shipping were maintaining current training
qualifications.
c. Conclusions
The licensee had effectively im)1emented a program for shipping
radioactive materials required ]y NRC and DOT regulations.
R1.3 Radiolooical Protection and Chemistry Controls
a. Insoection Scone (84750)
The inspectors reviewed implementation of-selected elements of the
licensee *s radiation protection and chemistry program. The review
Enclosure 3
.
_ . .. . . _ , . . _ . . _ _ _ . . . _ _ _. _ _ . _ _ _ .
V
.
. -
._
'
.o ^
,
,
-
22
1
included observation of radiological protection activities for the
- control of. radioactive material as required by 10 CFR Parts 20,1801.
~
-
_ __
- 1802. 1902. and 1904.
b.-' Observations and Findinos ,
i The inspectors reviewed licensee goals for waste generated and buried
-
,
and determined the licensee was meeting these goals. During tours of
-
- the auxiliary building and radwaste building facilities, the inspectors
reviewed survey _ data and performed selected independent radiation and ,
contamination surveys of radioactive material storage areas. During a
'
tour of the hot tool issue room on November 19, 1997, the inspectors
found a vacuum cleaner with radiation dose rates higher than indicated
'
on the radioactive material label.. dated 1995, affixed to the vacuum 1
4
cleaner. The tag stated radiation levels to be 1.5 millirem per hour on
-contact and 0.5 millirem at 30 centimeters. However, the inspectors
determined and the licensee confirmed radiation levels to be up to 40
< millirem per hour contact and 2-3 millirem at 30 centimeters. Also, the
vacuum cleaner hose was not taped or capped on the end as required by-
'
licensee procedure for vacuum cleaners in storage. Licensee procedure
'
required vacuum cleaners to be surveyed after use.and that current
- survey information was to be included on the radioactive material label
"
(yellow tag). - The licensee taped over the vacuum hose and performed
independent radiation-and contamination surveys of the vacuum cleaner
and the general area'. :The licensee determined contamination hadinot. -
-
- -
-
-
1been spread as:a result of the open hose. The licensee also relabeled r
the vacuum cleaner to include current' survey information. .
4
^
Ouke Power Company. System Radiation Protection Manual. Procedure No.
III-18. titled Use of Vacuum Cleaners In Radiologically Controlled
Areas. Revision 3. dated August 1. 2996, states that vacuum cleaners
should be surveyed during and after use and update dose rates on yellow
'
_ tags, if applicable, each time a radiation survey is performed.
10 CFR 20.1904(a) recuires, that the licensee shall ensure that each
container of licensec material bears a durable. clearly visible label
~
bearing the radiation symbol and the words CAUTION RADI0 ACTIVE MATERIAL
or DANGER RADI0 ACTIVE MATERIAL. The label must also provide sufficient
information-(such as radionuclides f the quantity
i .of radioactivity, radiation-levels.present, kinds ofan estimate
materials, and o mass
enrichment) to permit individuals-handling or using the containers or
working in the vicinity of the containers. to take precautions to avoid
or minimize exposures.
- The-inspector informed the licensee that failure to provide current
-survey.information on the radioactive material label constituted a
- violation of licensee procedure Use of Vacuum Cleaners In Radiologically
Controlled Areas. III-18. Revision 3 and a violation of 10 CFR
-
" -
20.1904(a). This item is identified as Violation 50-413.414/97-14-05:
1 Failure to Label Radioactive Material As Required by 10 CFR 20.1904.
Enclosure 3
g -
-1
.
t + -- W- w -e-vip---r r- w - ,->N,-,r ,-cv-m ,- -
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- .. .. . _ - . - . .
.
.
23
c. Conclusions
The licensee was meeting established goals for radioactive waste
generation. During plant tours, radiological facility conditions and
housekeeping in radioactive waste storage areas were observed to be -
good. One violation was identified for failure to provide current dose
rate information on a radioactive material label as required by licensee
procedure and 10 CFR 20.1904(a).
P1.4 Water Chemistry Controls
a. Insoection Scoce (847501
The inspectors reviewed implementation of selected elements of the
licensee's water chemistry control program for monitoring primary and
secondary water quality as described in the TS limits, the Station
Chemistry Manual, and the UFSAR. The review included examination of
program guidance and implementing procedures and analytical results for
selected chemistry parameters,
b. Observations and Findinos
The inspectors reviewed selected analytical results recorded for Units 1
and 2 reactor coolant primary water chemistry samples taken between May,
=
1997 and November, 1997, and secondary system water chemistry samples
taken between August, 1997 and November, 1997. The selected parameters
reviewed for primary water chemistry included dissolved oxygen,
chloride, pH. and fluoride. The selected parameters reviewed for
secondary water chemistry included hydrazine, dissolved o.xygen sodium,
copper, and chloride. Those primary system parameters reviewed were
maintained well within the relevant TS limits for power operations.
Those secondary system parameters reviewed were maintained according to
station procedures.
The inspectors reviewed and discussed the licensee's system for tracking
performance indicators in the areas of primary and secondary water
chemistry. The inspectors noted the licensee had maintained a high
level of success in human performance and equipment reliability in 1997
based on performance indicators for these areas which included no missed
surveillances and no mispositioning of components.
c. Conclusions
Based en the above reviews, it was concluded that the licensee's water
chemistry control program for monitoring primary and secondary water
quality had been implemented, for those parameters reviewed in.
accordance with TS requirements and the Station Chemistry Manual for
pressurized water reactor water chemistry. The licensee had maintained
a high level of success in human performance and equipment reliability
-in 1997.
Enclosure 3
. .
-
24
R3 Radiation Protection and Chemistry Procedures and Documentation
R3.1 Radiation Protection and Chemistry Procedures and Documentation
a. Insoection Scone (84750)
The inspectors reviewed licensee effluent release limits and pathways as
described in the licensee's Offsite Dose Calculation Manual and in
Chapter 16 of the Selected License Commitments Manual,
b. Observations and Findinas
The inspectors reviewed annual effluent data for 1996 and compared the
data to previous annual reports back to 1992. Arinual Radioactive
Effluent Release Reports were required to be submitted to the NRC prior
to May 1 of each year. Summaries of the quantities of radioactive
-materials in liquid and gaseous effluents released from the facility and
an assessment of the radiation doses due to those releases were required
to be included in the reports. The inspectors reviewed the supporting
data for the effluent release report covering 1996. The amount of
activity released during 1996 as dissolved gases in liquid effiuents and
fission gases, and that released as iodines and particulates in gaseous
effluents was generally within the ranges observed in past years. The
annual average per unit radiation doses for an individual from the
-
liquids and gaseous effluents were only a small percentage of their
respective annual limits. The total body dose as calculated by
environmental sampling data, was 0.902 millirem for 1996. There were no
abnormal releases reported in 1996.
c. Conclusions
Based on the above reviews. it was concluded that the licensee had
maintained an effective program to monitor and control liquid and
gaseous radioactive effluents, thereby limiting dose to members of the
public. The )rojected offsite doses resulting from those effluents were
well within tie limits specified in the TS. Offsite Dose C61culation
Manual, and 40 CFR 190.
R7 Quality Assurance in Radiation Protection and Chemistry Activities
R7.1 Ouality Assurance in Radiation Protection (RP) and Chemistry
a. Insoection Scooe (84750)
Licensee activities and self assessment programs were reviewed to
determine t.he adequacy of corrective action programs for identified
deficiencies in the areas of RP and chemistry.
Enclosure 3
- . _ . __ . _ . _. _ _ _ _ _ _ _ _ -
. 4
o
25 -j
.
bl Ilc trvations and Findinos
-Reviews by the= inspectors: determined that Quality AssuranceLaudits-and
self assessments-in the RP and chemistry areas were accomplished _by
.
reviewing-procedures, observing work, reviewing industry documentation,
and performing plant walkdowns to include surveillance of work areas by
supervisors and technicians during normal work coverage. Documentation
of problems by licensee representatives was included in Quality
Assurance Audits and self assessment _ reports. Corrective actions-were
> included in the licensee's-PIPS and were being completed in a timely
manner.
'
c. Conclusions-
'
4
The ins
RP andchemistry
pectors-determined the-licensee
audits as required was effectively
by the TS-and was completing conducting formal ,
corrective actions in a timely manner. *
R8 Miscellaneous Radiation Protection and' Chemistry Issues (92904) ,
,
R8,1 (Closed) URI 50-413.414/97-04-Q2; Determine the y plicability of
'
-Monitoring Requirements of Criterion 64 of 10 CFR s0 A)pendix A: and
Reporting Requirenents of 40 CFR 190 and 10 CFR 50.36a legarding
Potential Unmonitored Release Pothways. .
.
This item was-closed using guidance from Regulatory Guide 1.109, ,
Calculation of Annual Doses to Man from Routine Releases of Reactor
Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50.
Appendix I. The specific guidance was found in Appendix 01. No
violation of regulatory requirements was identified. This item is
closed.
P2 Status of Emergency Protection Facilities, Equipment, and Resources
e
P2.1 General Comments (71750)
.
The inspectors toured the Eme'gency Operations Facility located in
downtown Charlotte. North Carolina on November 18, 1997. The inspectors
observed that the facility and associated. equipment, including emergency
communication telephones and plant computer screens and controls were
. functioning and in good repair. During tours of the Technical Support
.
Center, facility equipment was also noted to be in working order and of
good condition and repair.
.
Enclosure 3
d
,. , . - ~ _
. .
26
V. Manaaement Meetinoq
X1 Exit Heeting Sumary
The inspector presented the inspection results to members of licensee
management at the conclusion of the inspection on December 3, 1997. The
licensee acknowledged the findings presented. No proprietary
information was identified.
Enclosure 3
-
. .
27.
PARTIAL LIST OF PERSONS CONTACTED _
Licensee
.
M. Birch. Safety Assuranco Manager
M. Boyle. Radiation Protection Manager
R. Glover. Operations Superintendent
J. Forbes. Engineering Manager
R. Jones. Station Manager
K. Nicholson, Compliance Specialist
M. Kitlan Regulatory Compliance Manager
G.-Peterson Catawba Site Vice-President
R. Propst. Chemistry Manager
Enclosure 3
. .
28
INSPECTION PROCEDURES USED
IP 37551: -Onsite Engineering
IP 61726: Surveillance
IP 62707: Maintenance Observation
IP 71707: Plant Operations
IP 71714: Cold Weather Preparations
IP 71750: Plant Support Activities
IP 84750: Radioactive Waste Treatment, and Effluent and Environmental
Monitoring
IP 86750: Solid Radioactive Waste Management and Transportation of
Radioactive Materials
IP 92901: Follow up - Operations
IP 92902: Follow up - Maintenance
IP 92903: Follow up - Engineering
IP 92904: Follow up - Plant Support
IP 93702: Prompt Onsite Response to Events
TI-2515/136: Operation of Dual Function Containment Isolation Valves
ITEMS OPENED, CLOSED, AND DISCUSSED
i Opened
50-413/97-14-01 URI Control Power Unavailable to the Unit 1
Turbine-Driven AFW Pump's Trip and
Throttle Valve (Section 01.3)
50-413.414/97-14-02 DEV Changing NRC Commitments Without Properly
Notifying the NRC (Section 08.1 and 08.2)
50-413,414/97-14-03 URI Noncompliance With 10 CFR 50 Appendix A
General Design Criterion 57 (Section
El.1)
50-413.414/97-14-04 NCV Failure to Control Use of Aluminum Inside
the Containment Building (Section E8.2)
50-413,414/97-14-05 VIO Failure to label Radioactive Material As
Required by 10 CFR 20.1904 (Section R1,3)
Closed
50-414/95-01 LER Reactor Trip Due to Closure of a Main
Steam Isolation Valve (Section 08.1)
50-413.414/97-08-04 IFI Reportability of Nuclear Service Water
System Actuations (Section M8.1)
Enclosure 3-
. .
29
50-413.414/96-18-04 IFI Quantification of Refueling Water Storage
Tank Heat Losses Through Tank Roof
Including a Wind velocity Factor (Section
E8.1)
50 413.414/97-11-04 URI Use of Aluminum HEPA Filter Separators
Inside Containment (Section E8.2)
50 413.414/97-05-02 URI Determine the Applicability of Monitoring
Requirements of Criterion 64 of 10 CFR 50.
Appendix A: and Reporting Requirements of
40 CFR 190 and 10 CFR 50.36a Regarding
Potential of Unmonitored Release Pathways
(Section RF, 1)
TI 2515/136 TI Operation of Dual Function Containment
Isolation Valves (Section El.1)
Discussed
50-413/97-08-01 VIO Inadequate Alarm Response Results in
Inadequate add Untimely Corrective Actions
for Valve Operability Determination
. (Section 08.2) -
LIST OF ACRONYMS USED ,
AFW - Auxiliary Feedsater
CACFU - Containment Auxiliary Charcoal Filter Units
CFR -
Code of Federal Regulations
DC -
Direct Current
DBD -
Design Basis Documents
DEV -
Deviation
001 - Digital Optical Isolator
DOT -
Department of Transportation
DNB -
Departure From Nucleate Boiling
EDG -
ESF -
Engineered Safety Features
ESFAS - Engineered Safety Features Actuation System
FWST -
Refueling Water Storage Tank
GDC - General Design Criterion
HEPA - High Efficiency Particulate Air
KW -
Kilowatt-
LC0 -
Limiting Condition for Operation
LER -
Licer.see Event Report
LSA -
Low Specific Activity
MPH -
Miles Per Hour
NEI -
Nuclear Energy Institute
NRC -
Nuclear Regulatory Commission
Enclosure 3
o .-
30
NS - Nucicar Spray
Nuclear System Directive
~
NSD -
NSW --
Nuclear Service Water
0AC- - Operator Aid Computer
00CM -
Offsite Dose Calculation Manual
0PDT - 0,erpower Differential Temperature
PCB -
Power Circuit Breaker
PDR -
Public Document Room
PIP - Problem Investigation Report
PM -
-Preventive Maintenance
PORVS - Power Operated Relief Valves
PSIG - Pounds per Square Inch Gauge
RCA - Radiological Control Area
RGC -
Regulatory Comaliance .
"RHR -
RP -
Radiation Protection
R&R -
Repair and Restor 6 tion
SCO - Surface Contaminated Object
SI - System Internationale
SRP -
Standard Review Plan
SSC - Structures. Systems, and Components
SSPS -
Solid State Protection System
TS --
Technical S ecification
TSAll -
UFSAR -
.. Technical S ecification Action Items List
Updated Fin 1 Safety Analysis Report r:
-URI- -
Unresolved Item
VIO -
-Violation -
WO -
Work Order
Enclosure 3-
9