ML20216H191

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Insp Rept 50-219/98-04 on 980226-0318.Violations Noted.Major Areas Inspected:Operations,Maint & Plant Support
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:

'

h WIQT REPORT NUMBER DOCKET NUMBER

A V0 f! 0 TO ~ N /

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REPORT TRANSMITTAL ($1GNATURE) DATE: / / .

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

(v:0 oNLT) (vro ONLT)

UNIT STATUS W N E

A O / /

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RESP / CLOSEOUT ORG: / l h Ihl O! RESP / CLOSEOUT EMP: $! l CONTACT EMP: l I l

l PROCEDURE #: M/ @D E

FUNC1L AREA: 0181$l  ! I l, I  !  ! I I l INTEREST CODE: ,eg n CAUSE CODE: l, I

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NOV ISSUE DATE: / /

NTS:

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SECTION B

ORIGINATING IR NUMBER: -

SEQUENCE NUMBER: 1 I ITEM TYPE CODE: l l  !

OR (IF FOLLOW-UP TO ENFORCEMENT 8CT10W LETTER VIOLATION): 'EA NUMBER and *NOV ID:

UPDATE / CLOSEOUT ACTUAL

E EE [, .D.43 O sk.N<3

A _

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8 _

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C _ / /

  • COIMENTS:

emme: . - w aseita m ema meses. au esm vu n.ser

1

e SEE SACK FOR EAPLANATIONS AfG CODES RI\ f3\ IFS.INSP.Ffpt

'

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.. -

SEQUENCE NBR.: 8 Ic2 l ITEM TYPE CODE: V!IlOl SEVERITY LEVEL: SUPPLEMENT CODE: l, j

(vso oNLT) (V]o oNLY)

CTED (OPTIONAL.PaoVIDE ONLY

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RESP /CLOSEQUT ORG: /![l[!D! RESP /CLOSCOUT EMP: IfI!N CONTACT EMP:  ! ! !

' PROCEDURE #: . 7/r!@ E

FUNCTL AREA: [l6fIldlM!  !,  !  !  ! 1  !  ! INTEREST CODE: E CAUSE CODE: J,  !

NOV ISSUE DATE: / /

NTS: *% / YHi~$  % & CMFJb& O t//~LJ&/ /~~ 5 C.-

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V O V

SEQUENCE NBR.: Ol3 ! ITEM TYPE CODE: Il [!Il SEVERITY LEVEL:  ! SUPPLEMENT CODE: l, j

(vso ONLv) (vio oNov) l

UNIT STATUS 0 WkN15 N Ntf

A ( / /

B _, / /

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  • TITLE: /~/ Y/t-f m WM'/ / YW/57/ /S/7

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PROCEDURE #: INN

FUNCTL AREA: klb!IlI ll[! / !,  !  ! I l l  ! INTEREST CODE: M CAUSE CODE: 1, 1

NOV ISSUE DATE: / /

(vso oNLY)

  • CGINlENTS: YYIMM  %/ +1/ C'~//'f

'

A r Y W /?P / G M G/// V17 CM b

& $ /~~$kt5 kHJ GA/ r/ ffYJ/W / f Vb

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SEQUENCE NBR.: I  ! ITEM TYPE CODE:  ! I l SEVERITY LEVEL:  ! SUPPLEMENT CODE: 1, J

(vso oNLY) (Vlo oNLY)

UNIT STATUS W $15 N Ntf

A _

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8 _

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C _

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  • TITLE:

e. . mas

RESP /CLOSE00T ORG:  !  !  !  ! RESP /CLOSE0VT EMP  !  !  ! CONTACT EMP: l ! !

(OPTIONAL)

PROCEDURE #:

FUNCTL AREA: l I I  !  !  !,  !  !  ! I f 1 INTEREST CCOE: m CAUSE CODE: J, 1

NOV ISSUE DATE: / /

(vio oNLY)

"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)