ML20216H191
| ML20216H191 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 04/15/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20216H154 | List: |
| References | |
| 50-219-98-04, 50-219-98-4, NUDOCS 9804210081 | |
| Download: ML20216H191 (22) | |
See also: IR 05000219/1998004
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U.S. NUCLEAR REGULATORY COMMISSION l
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REGION I l
Docket No: 50-219
License No: DPR-16
Report No: 50-219/98-04
Licensee: GPU Nuclear incorporated
Facility Name: Oyster Creek Nuclear Generating Station
Dates: February 26 - March 18,1998
Inspectors: J. Jang, Sr. Radiation Specialist
J. McFadden, Radiation Specialist
S. Pindale, Resident inspector
Approved by: John R. White, Chief
Radiation Safety Branch
Division of Reactor Safety
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9004210081 980415
PDR ADOCK 05000219
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EXECUTIVE SUMMARY
i The purpose of this special inspection was to review the circumstances related to the
l licensee's identification that continuous steaming of one of the two isolation condensers,
l constituted an unmonitored radioactive effluent release path not previously reported in
annual reports.
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OPERATIONS:
The licensee failed to perform and document a safety evaluation in accordance with
10 CFR 50.59 to support the use of radiologically contaminated water (tritium) from the
condensate transfer system for makeup to the shell side of the isolation condenser system ,
during Standby conditions, instead of the demineralized water transfer system (a normally I
non-radioactively contaminated system), as described in the design bases. This is a
violation of the requirements of 10 CFR 50.59.
The licensee was not effective in recognizing that the continued use of the condensate
l transfer system as a makeup source to the isolation condenser system during Standby I
conditions was an operator workaround as defined by a licensee established Work
Performance Standard and should have been documented in the Operator Workaround ,
Tracking system. Though not a violation of NRC requirements, this performance '
demonstates potential weakness in this area.
l The licensee was ineffective in timely problem resolution in that the discrepancy between
actual practice and the description in the design basis documents, relative to the expected
source of makeup water to the shell side of the isolation condenser system in Standby l
conditions, was recognized in March 1997 but not followed-up until the potential I
radioactive release pathway from the isolation condenser was discovered in February
1998.
MAINTENANCE:
The licensee effectively monitored the performance of the IC system in accordance with
maintenance rule requirements as specified in 10 CFR 50.65, and was effective in
maintaining system availability.
PLANT SUPPORT:
As required by Technical Specification 6.8.4.a.3, the licensee's radioactive effluent
controls program did not provide for the monitoring, sampling, and analysis of tritium vapor
released from the isolation condenser system to unrestricted and con. trolled areas in order
to demonstrate compliance with the dose limits for individual members of the public. This
is a violation of NRC regulatory requirements.
! No health and safety consequence to members of the public or onsite workers is expected
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as a result of the tritium release pathway through the isolation condenser. Projected doses
to members of the public was a very small fraction of the limit specified in the applicable
regulatory requirements, including 10 CFR 20.
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The licensee provided effective radiological controls for a radioactive steam leak and
resulting potential contamination in the affected areas involving the isolation condenser
system.
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Report Details
02 Operational Status of Facilities and Equipment (71707)
O2.1 Backoround Information
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There are two isolation condensers (IC) used at Oyster Creek which function as
heat exchangers. Each of the two ICs is filled to a specified level with condensate J
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water on the secondary (shell) side. The IC steam release path to the atmosphere is i
l from the IC vents. IC 'A' has one vent pipe that penetrates the east wall of the
reactor building, and IC 'B' has two vent pipes that penetrate the same wall. The
primary side of each IC is supplied steam from the reactor and has two series
isolation valves. After passing through the tube bundles and transferring heat to i
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the condensate-water-filled shell side, the condensed steam returns to the reactor
(recirculation pump suction) via two additional series isolation valves. In the IC
standby configuration, only the first condensate return valve (DC-operated) is
cicsed.
The ICs are designed to depressurize the reactor and remove residual and decay
! heat from the reactor during conditions when the main steam isolation valves have
l closed. The IC system is not considered part of the emergency core cooling
system.
The eight IC valves associated with each IC are isolation valves. They receive
signals to automatically close upon a high IC steam or condensate return flow in
order to isolate the system in the event of a line break outside the primary
containment. These valves are not leak rate tested. During the Type A
containment integrated leak rate test, the IC system piping, which is part of the
reactor coolant system pressure boundary and may be open directly to the
containment atmosphere under post-accident conditions, is not opened and drained
because the condensate return line is connected to the reactor coolant system
below the level of the water in the reactor. The valves similarly are not Type C
(local leak rate) tested because the IC system is an extension of the containment
boundary and was designed to be operable following an accident. NRC letter
(10 CFR 50, Appendix J Safety Evaluation) dated March 4,1982, documents the ,
above as acceptable. 1
O2.2 Isolation Condenser Valve Operational History
The IC vent lines have a long history of continuous steam wisping to the
, atmosphere. The cause of the steaming is due to smallleakage past the normally
l closed condensate return isolation valves (V-14-34 for IC 'A' and %14-35 for
IC 'B'), as shown in Figure 1. A small amount of leakage past a condensate return
isolation valve allows the shell side to heat up enough to allow a small amount of
steaming through the IC vents to the atme,ohere. Currently, only the shell side
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water on 'B' IC is warm enough to develop steam. Primary and secondary fluid
temperatures of the 'A' IC are typically less than 100*F. The steam temperature of
l the 'B' IC is 545*F, and the condensate temperature is about 160 F. Operators
normally maintain lC level between 7.3 ft and 7.6 ft, which is about a 1200 gallon
band. They make up to the 'B' IC about every two to four days. The licensee
estimated the steaming losses for the 'B' IC to be approximately 400 galloris per
day,
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Six of the eight IC valves are double-disk gate valves. Originally, all eight of these l
valves were a single gate design, but the six valves located outside the primary
containment were modified in 1990.
Over several years, the licensee has provided substantial attention to the IC valves,
and the normally closed condensate return valves in particular, in response to prior
leakage past the valve seats, the licensee attempted different seating torques for
the disk (soft- and hard-seating) to stop the leakage past the valve seat. Neither -
configuration provided notably improve,d performance. In the 13R refueling outage
(1990), six of the eight IC valves were replaced (all except for the two downstream,
normally open, condensate return valves). In 14R (1992), V-14-34 and V-14-05
were opened and inspected after indications of minor leakage (wisping steami
during the operating cycle. In 15R (1994), a modification was implemented to the
same six valves, which installed a new disc and new stem (valve manufacturer was
present). In 16R (1996), V-14-34 and V-14-35 were opened and inspected with
the manufacturer on site. The liconsee has been in contact with a different
representative from the valve manufacturer. They plan to perform additional
inspection and measurement activities of the valve internals during the upcommg
17R refueling outage (Fall 1998).
The IC valve seat leakage problems, although minor, represent a chronic deficiency
for which corrective maintenance and modifications have not been fully effective in
resolving. Engineering and maintenance efforts are continuing to fully resolve this
long-standing rroblem.
03 Operations Procedures and Documentation (71707)
O3.1 Review of Procedures and Desian Basis for Isolation Condenser Operation
a. Insoection Scope
The inspector reviewed procedure 307, / solation Condenser System, the Facility
Design and Safety Analysis Report (FDSAR), and the Updated Final Safety Analysis
Report (UFSAR) to determine the operating practices for makeup to the IC shell side
and the associated design bases. The inspector also interviewed operations and
engineering personnel.
b. Observations and Findinas
UFSAR Section 6.3.1.1.2 (Update 8, 8/93) states that during normal operation,
when the system is in standby, makeup to the ICs is from the domineralized water
transfer system (DWTS); and, makeup during IC operation is provided from the
Condensate Transfer System (CTS). FDSAR Section IV-3-1 (1967 support submittal
for operating license), similarly stated that the water stored in the shell of the
isolation condensers can be supplemented by makeup from storage tanks or a pond,
pumped by either the condensate transfer pumps or by one uf the two diesel driven
fire pumps; and, that domineralized water will be supplied to the IC shells for fill and
normal makeup. The inspector found that this was not the current practice.
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Specifically, operators typically make up to the ICs, even while the system is in
standby, from the CTS which is radioactively contaminated, including tritium
contamination; The CTS is hard piped, has about 180 gpm makeup capability, and
can be operated from the control room. This is done in accordance with system
operating procedure 307. The use of the DWTS, which is tritium-free, requires the
use of a hose to connect to a sample line at the shell side of the ICs. This is also
controlled by procedure 307.
Based upon a review of the history file for procedure 307,it appears that CTS was
originally the preferred source of any makeup to the ICs. The original alternate was
the fire protection system. However, in the early 1980s, the licensee
proceduralized the use of the DWTS as a makeup source for standby operation in
the event that the CTS became unavailable. This was done to prevent the use of
fire protection (pond) water in non-emergency conditions because of the non-
radioactive contaminants it would introduce to the IC shell side.
Subsequently, the DWTS had become slightly radioactive!y contaminated. It had
become radiologically contaminated during the 15R outage (October 11,1994),
when the system was used during core shroud inspection activities for underwater
equipment. Although it became contaminated with some amount of tritium,
subsequent flushes of the system reduced the tritium level significantly.
Revision 63 to procedure 307, which was in effect at the time of this inspection,
allowed makeup to the IC shell sides from all three sources (CTS, DWTS, and fire
protection). Precaution and Limitation 2.2.9 states that the CTS shall be used to
maintain level in the ICs, and that for normal loses, domineralized water can be
used.
The inspector determined that operators nearly always use the CTS to provide
normal makeup to the ICs while in standby because it is operationally convenient.
The operators can remotely initiate a makeup from the control room when using the
CTS. Use of the DWTS requires local operations, including physically connecting a
temporary hose from a DWTS supply connection directly to an IC sample line. The
IC valve seat leakage and associated steaming of the 'B' IC results in frequent
performance of this evolution. Oyster Creek Work Performance Standard OPS-13,
Operator Workarounds, defines an operator workaround as a plant deficiency that
requires operation of a system or component in a condition other than intended by
design or plant procedures. The licensee had not previously considered this activity
to be a workaround and it was not on the associated Operator Workaround Tracking
Form. After this issue was identified (February 24,1998), the licensee
subsequently determ:ned that the activity should be considered a workaround.
The inspector determined that the licensee was operating the IC system (makeup
mode) different than described in UFSAR Section 6.3.1.1.2 and the original FDSAR.
i Title 10 CFR 50.59, Changes, tests and experiments, permits the licensee to make
changes to its facility and procedures as described in the safety analysis report
without prior Commission approval, provided the change does not involve a change
in the technical specifications or an unreviewed safety question (USO). 10 CFR
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50.59 further states that the licensee shall maintain records of changes in the
facility and these records must include a written safety evaluation which provides
the bases for the determination that the change does not involve a USQ. The
licensee's failure to conduct a written safety evaluation to provide the bases for the
determination that the change did not involve a USQ is a violation.
(VIO 50-219/98-04-01)'
While reviewing this issue, the inspector found that in March 1997, the licensee
recognized the discrepancy between the operating practice and the UFSAR, and
initiated Deviation Report 97-162 to document the issue. They initiated a UFSAR
update to address the inconsistency. However, in mid to late 1997, engineering
discovered that an associated safety evaluation had not been developed.
Subsequently, action was taken to initiate a safety evaluation. About one week
before the February 24,1998 discovery (see Section R1.1), the system engineer
reviewing the safety evaluation independently raised questions related to
radiological implications of continuing to use the CTS for all makeup to the ICs.
He returned the safety evaluation to the originator, and it was ::till under
development and review when the February 24 discovery was documented via the
deviation report process. While the inspector noted the licensee's discovery of this
deficiency in March 1997, and their recent recognition of potential radiological
implications of using the CTS for makeup, the licensee's actions to resolve this
issue were too slow to recognize the effort as successful problem identification and
resolution.
At the end of this inspection, the licensee was evaluating the available sources and
procedures for IC Standby makeup. For the interim, they plan to use the DWTS
since is contains less radioactivity than the CTS and is tritium-free. However, they
plan to further address the potential to cross-contaminate the DWTS as per NRC
Bulletin 80-10 since it would be temporarily connected to the contents of the shell
side of the ICs (CTS water, which contains tritium). The licensee will monitor and
record the amount and source of any makeup to the ICs to assist in dose
assessment.
c. Conclusion
The licensee failed to perform and document a safety evaluation in accordance with
10 CFR 50.59 to support the use of radiologically contaminated water (tritium) from
the condensate transfer system for makeup to the shell side of the isolation
condenser system during Standby conditions, instead of the domineralized water
transfer system (a normally non-radioactively contaminated system), as described in
the design bases. This is a violation of the requirements of 10 CFR 50.59.
The licensee was not effective in recognizing that the continued use of the
condensate transfer system as a makeup source to the isolation condenser system
during Standby conditions was an c,)erator workaround as defined by a licensee
established Work Performance Standard and should have been documented in the
Operator Workaround Tracking system. Though not a violation of NRC
- requirements, this performance demonstates potential weakness in this area.
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The licensee was ineffective in timely problem resolution in that the discrepancy
between actual practice and the description in the design basis documents, relative
to the expected source of makeup water to the shell side of the isolation condenser
system in Standby conditions, was recognized in March 1997 but not followed-up
until the potential radioactive release pathway from the isolation condenser was
discovered in February 1998.
M2 Maintenance and Material Condition of Facilities and Equipment (02707)
M2.1 Maintenance Rule Aoolicability and imolementation
a. Inspection Scooe
The inspector reviewed the IC (isolation condenser) system and components as
related to the maintenance rule (10 CFR 50.65). The IC systems are within the
scope of the maintenance rule and are risk significant,
b. Observations and Findinas
The licensee monitors the ICs on a system level and a plant level. In accordance
with the licensee's evaluation, the potential for a single maintenance preventable
functional failure on a component level requires the system to be monitored in
accordance with 10 CFR 50.65(a)(1).
The inspector determined that the licensee appropriately monitors this system in
accordance with the requirements of 10 CFR 50.65. Based on the licensee's
analysis, the unavailability for each IC system (two-year average) was ler,s than
0.10%, which is well below the licensee's established 1.0% unavailability criterion.
Based on discussion with system engineer and maintenance rule coordinator, the
inspeMor determined that the IC system ic considered functional when a flow path
between the reactor vessel and IC is intact, the shell side is intact and open to the -
atmosphere with a source of makeup water available, and all valves are capable of
performing their initiation and isolation function. The small amount of seat leakage
past the normally closed condensate return valves 09es not render the valve or
system non-functional based upon the low leakage rate and isolation redundancy.
c. Conclusion
The licensee effectively monitored the performance of the IC system in accordance
with maintenance rule requirements as specified in 10 CFR 50.65, and was
effective in maintaining system availability.
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R1 Radiological Protection and Chemistry (RP&C) Controls (84724)
R1.1 Imolementation of the Tritium Effluent Control Proaram
a. Insoection Scone
On February 24,1998, the licensee identified that tritium was being released to the
. environment through the 'B' isolation condenser vent and documented the finding in -
a Deviation Report (DR 98-0180). The scope of this inspection was to determine
the following:
> Conformance with the requirements of TS Section 6.8.4.a.3, relative to the
establishment of an effective program to monitor, sample, and analyze
radiological effluents that may be generated from the isolation condenser
system;
- The capability of the licensee to calculate the projected dose to the public due
to tritium release through the isolation condenser;
> The capability of the licensee to quantify tritium release for the isolation
condensers and for the other release point (stack) relative to the reporting
requirements specified in TS Section 6.9.1.d; ;
- Dose consequence to members of the public and onsite workers as a result of
the tritium release pathway through the isolation condenser system; and
> Corrective actions initiated by the licensee,
b. Observations and Findinas
The licensee had been using the condensate water for makeup to the shell sides of
the 'A' and 'B' ICs (isolation condensers), as described in Section O2.1 of this
report. Water from the condensate storage tank typically contains tritium and very
low amounts of other radionuclides. The inspector, therefore, reviewed
radioanalytical measurement results for the condensate storage tank water, data
related to evaporative losses, and their impact on the effluents program.
The licensee had been sampling the shell sides of the 'A' and 'B' ICs and analyzing j
the samples for gamma emitters to track tube integrity as a normal surveillance
activity (Figure 2 pertains). The inspectors reviewed the gamma measurement
results for the condensate storage tank water for the period between January 1997
and February 1998. The average total gamma measurement results for 1997 and
1998 were 1.64E-6pCi/cc and 4.79E-7pCi/cc, respectively. The lower limit of
. detection for the principal gamma emitters for radioactive liquid release, established
in the ODCM, is 1 E-6 pCi/cc.
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The inspector determined that the licensee had not discharged radioactive liquid
routinely since 1990. As a result, the buildup of tritium activity in the reactor
coolant and spent fuel pool water continued to increase over time. As shown in
Figures 3 and 4, tritium activity has been increasing in the reactor coolant and spent
fuel pool water during since mid-1996. Licensee analyses indicate that tritium
activity in condensate storage tank water is similar to the tritium activity measured
in the reactor coolant. Tritium activity of the reactor coolant was 6.76E-2pCi/cc in
February 1998.
As a normal surveillance activity, the licensee also tracks the shell side water
temperatures of the 'A' and 'B' ICs. The inspectors reviewed the water
temperature trendings from November 9,1996 to March 16,1998. The
temperature of the shell side water of the 'B' IC is generally higher than the 'A' IC
for much of this period. For example, the water temperatures of 'A' and 'B' ICs l
were 91 *F and 160*F , respectively, on March 16,1998. The te nperature of the l
shell side of 'B' IC was elevated due to the leak through valve V-14-35, shown in
Figure 1.
As a result, water from the shell side of 'B' IC steamed and was released to the
environment through the 'B' IC vents. The licensee estimated that about
400 gallons per day was steamed off from the 'B' IC during 1997.
The inspector noted that water temperatures of the 'A' IC had elevated slightly
(from about 100*F, to a range between 114 *F and 131 *F) during May and June
1997. The licensee estimated that about 300 gallons of water (approximately
O.8 gallons per day) had been steamed off from the 'A' IC during 1997, including
the May and June 1997 time period.
The inspector determined that, though the licensee routinely measured tritium ,
activities for the reactor coolant (and by inference, condensate storage tank water), !
no specific monitoring, sampling, or analysis had been conducted to demonstrate
compliance with regulatory requirements relative to the tritium vapor released from 2
the isolation condenser system to the environment. The inspector noted that the )
licensee did not consider the projected dose to the public from this release pathway, j
until it was recognized on February 24,1998. I
Section 6.8.4.a.3 of Technical Specifications (TS) requires that a radioactive
effluent controls program shall be provided and include " Monitoring, sampling, and
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analysis of radioactive liquid and gaseous effluent in accordance with 10 CFR
20.1302 and with the methodology and parameters in the ODCM". Section l
20.1302 of 10 CFR specifies that, "The licensee shall make or cause to be made, I
as appropriate, surveys of radiation levels in unrestricted and controlled areas and
radioactive materials in effluents released to unrestricted and controlled areas to ;
demonstrate compliance with the dose limits for individual members of the public in
Section 20.1301." Section 20.1301 of 10 CFR specifies dose limits for individual
members of the public. Failure of the licensee to conform to this regulatory
requirement constitutes a violation. (VIO 50-219/98-04-02)
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The inspector examined and verified the licensee's capability to calculate the
projected dose to the public due to this IC vent release pathway. The inspector
independently performed a projected dose calculation using the NRC PCDOSE code
for the airborne tritium released through the isolation condenser vents during 1997.
The licensee also performed a projected dose calculation using the licensee's
EFFECTS code. Calculation results of the NRC and the licensee were
0.030 mrsm/ year and 0.029 mrem / year, respectively. The TS limit as specified in
the ODCM is 15 mrem /yr. The inspector determined that the licensee had the
capability to effectively calculate the projected doses to the public, and that dose to
the public from this release pathway constituted a small fraction of the regulatory
limit. Based on this analysis, the inspector also determined that dose consequence
to onsite workers was negligible relative to the regulatory limit of 5000 millirem per
year, total effective dose equivalent.
Relative to the licensee's capability to quantify the total trit;um release for all release
pathways, the inspectors reviewed the licensee's calculated annual airborne tritium j
release from the spent fuel pool (SFP) water evaporation and the gland seal
exhauster to the main stack for the period between 1994 and 1997. The inspectors
determined that the licensee's calculation methodology was good and that results
were within the expected values. The inspector concluded that the licensee had the
capability to effectively quantify airborne tritium releases from the plant.
Relative to the licensee's compliance with the reporting requirement specified in TS Section 6.9.1.d, the inspectors noted the following. From review of pertinent
effluent reports the inspector determined that from the time of initial plant operation
until the beginning of 1996, projected doses to the public due to airborne tritium
were insignificant relative to other gaseous effluents. However, due to improved
fuel integrity, the increasing effective treatment of gaseous effluent through ,
operation of the airborne radioactive material clean-up system (e.g., the Augmented I
Offgas System), and the increase of tritium activity in the condensate water, ;
projected doses to the public due to airborne tritium became a more significant j
fraction of the total annual dose due to effluents. Accordingly, airborne tritium i
releases through the IC system vents became a more significant dose contributor to
the total annual effluent dose to the public. For example, the total 1997 annual
thyroid dose due to particulates/ iodine / tritium releases was 0.046 mrem, which
included the tritium dose due to IC system vent releases. The 1997 annual thyroid
dose due to IC system vent due to tritium alone was 0.030 mrem. The annual
thyroid dose limit is 15 mrem / year, as defined in Section 4.6.1.1.7.A of the ODCM.
The licensee quantified the 1997 tritium release from the IC system, performed a
projected dose calculation, and reported the result in the 1997 Annual Report, as
required by TS. At the time of this inspection, the licensee was continuing review
of the 1995 and 1996 operations log book to determine the total water loss through
the 'A' and 'B' ICs. Subsequently, the licensee intends to quantify the 1995 and
1996 tritium releases from the IC system and calculate the projected doses to the
public. The calculation results will be reviewed by NRC to determine if the
quantities released were reportable. (IFl 50-219/98-04-03)
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y The licensee's initial corrective action plans were described in Deviation Reports
(DRs)98-178,98-179, and 98-180 which were generated on February 24,1998.
DR 98178 was assigned to the Radiation Protection organization with an action
item to update the effluent report for 1997 and for previous years during which
condensate transfer was used for makeup. DR 98-179 was assigned to the
Engineering organization with action items to evaluate the continued use of the
condensate transfer as makeup to the isolation condensers and to establish an
effective method of monitoring the radioactive release from the isolation condensers
dus to steaming. ' DR 98-180 was assigned to the Licensing organization to perform
al. sot cause investigation to determine why the IC system had not been previously
identified as a release po:nt.
c. Conclusions
As required by Technical Specification 6.8.4.a.3, the licensee's radioactive effluent
controls program did not provide for the monitoring, sampling, and analysis of
tritium vapor released from the isolation condenser system to unrestricted and
controlled areas in order to demonstrate compliance with the dose limits for
individual members of the public. This is a violation of NRC regulatory
requirements.
No health and safety consequence to members of the public or onsite workers is
expected as a result of the tritium release pathway through the isolation condenser.
Projected doses to members of the public was a very small fraction of the limit
specified in the applicable regulatory requirements, including 10 CFR 20.
R8 Miscellwoous RP&C issues
R8.1 fLAC Controls
a. Inspection Scooe (IP 84724-01)
The radiological controls for a radioactive steam leak and condensation on the floor
in the immediate vicinity of the isolation condensers were reviewed.
b. Observations and Findinas
The inspector noted that an isolation condenser tube-side isolation valve was
leaking steam and that the resulting condensation was dripping to the floor. In
response, the licensee implemented the following radiological controls for thir,
potential radioactive contamination problem: (1) the affected floor area was
cordoned off and posted as a contaminated area; (2) contamination and radiation
surveys were conducted in the area; (3) a continuous air monitor was located in <
close proximity to the affected area; (4) monitoring of the frequency of personnel I
contaminations due to noble gas daughters was initiated; (5) a noble gas air sample I
was taken in the immediate vicinity; (6) absorbent materials, employed to contain
the condensation on the floor, were checked and replaced on a regular basis; and
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(7) a tritium air sampling program was initiated. No indication of increased airborne
radioactivity or of increased radiation exposure to personnel was found based on
the licensee's m'oasurements and evaluations. Nc personnel contamination was
attributed to the leak. Actions were initiated to effect repair of the valve.
c. Conclusions
The licensee provided effective radiological controls for a radioactive steam leak and
resulting potential contamination in the affected areas involving the isolation
condenser sy:: tam.
V. Maneaement Meetinas
X1 Exit Meeting Summary
The inspector presented the inspection results to members of licensee management at the
conclusion of the inspection on March 18,1998. The licensee acknowledged the findings
presented.
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PARTIAL LIST OF PERSONS CONTACTED
Licensee (in alohabetical order)
F. Applegate, NSA Assessor
G. Busch, Mariager, Nuclear Safety & Licensing
W. Cooper, Manager, Radiological Engineering
B. DeMerchant, Licensing Engineer
R. Hillman, Manager, Radioactive Waste and Chemistry l
S. Levin, Director, Operations and Maintenance
J. Mockridge, Sr. Chemist
K. Mulligan, Director, Plant Operations
M. Roche, Director, Oyster Creek
M. Slobodien, Director, Radiological Health and Safety
P. Thompson, Root Cause Coordinator
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K. Wolf, Manager, Radiation Control Operations
NRC (in alphabetical order) l
J. Jang, Sr. Radiation Specialist
J. McFadden, Radiation Specialist
S. Pindale, Resident inspector
J. Schoppy, Sr. Resident inspector
State of New Jersev
K. Tosch, Bureau of Nuclear Engineering
R. Pinney, Bureau of Nuclear Engineering
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INSPECTION PROCEDURES USED
Procedure No. Title
37551 Onsite Engineering
62707 Maintenance Observation
71707 Plant Operations
84724 Gaseous Waste System (Minimum and Basic)
ITEMS OPENED, CLOSED, AND DISCUSSED
OPENED: (VIO 50-219/98-04-01); failure to conduct a written safety evaluation
to provide the bases for the determination that the IC change did not
involve a USQ.
(VIO 50-219/98-04-02); failure to make a complete survey for IC
release pathway, as required by 6.8.4.a.3 of TS.
(IFl 50-219/98-04-03); tritium monitoring and reporting requirements
of IC release pathway for 1995 and 1996.
CLOSED: NONE
DISCUSSED: NONE
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LIST OF ACRONYMS USED
CTS Condensate Transfer System
DR Deviation Report
DWTS Demineralized Water Transfer System
FDSAR Facility Design and Safety Analysis Report
IC isolation Condenser
ODCM Offsite Dose Calculation Manual
TS Technical Specification
UFSAR Updated Final Safety Analysis Report
l USQ Unreviewed Safety Question
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REACT M 14tSPECTION FIm illGS IFS DATA ENTRY FORM
4 4
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, REACT (Mt/FIEL FACILITY INSPECTHals
PAGE 1 0F
(USE CONTINUAT:ON SHEET IF peJLT:PLE 27085)
SITE: //S #F /~E SC REVIEWED BY:
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i RESPONSIBLE ORG. CODE: / l[! M Di
LEAD INSPECTOR (RITS INITIALS): MII'
ITEMS OPENED BY THIS REPORT (Y/N): Y! IF "Y" COMPLETE SECTION A
ITEMS UPDATED / CLOSED BY TN*$ REPORT (Y/N): "Y"M IF
COMPLETE SECTION B
SECTION A
SEQUENCE NBR.: 81/ l ITEM TYPE CODE: Y!I!Of SEVERITY LEVEL: SUPPLEMENT CODE: [ , j
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UNIT STATUS W N E
A O / /
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ORIGINATING IR NUMBER: -
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REGION I FORM 325
(DCT 1995) ~
1.
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CONTINUATION SHEET - PAGE d 0F M
AEACTOR INSPECTION FIISINGS IFS DATA ENTRY FORM - REACTGt/REL FACILITY INSPECTIONS
SECTION A (CONT NunTiON or :Tsis aPEN)
.n . n.. -
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"C0fetENTS:
c uana a==,u. =was.u en== se== .aasia. as an==in n=an
as gir s\1t s_ CONT .rnN
CONTlWUATION SHEET TO REGION 1 FORM 325 (OCT 1995)