IR 05000219/1996010

Insp Rept 50-219/96-10 on 960923-27.No Violations Noted. Major Areas Inspected:Radiological Source Term Assessment, Results of Exposure Reduction Initiatives & Evidence of Station ALARA Program

ML20138H624
Person / Time
Site:	Oyster Creek
Issue date:	09/27/1996
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20138H617	List: IR 05000219/1996010 ML20138H615
References
50-219-96-10, NUDOCS 9701060145
Download: ML20138H624 (29)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket N ,

License N DPR-16 l

Report N l Licensee: GPU Nuclear Corporation i

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Facility Name: Oyster Creek Nuclear Generating Station l Locatior : Forked River, New Jersey l

l l Dates: September 23 - 27,1996 l

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j Inspectors: James D. Noggle, Senior Radiation Specialist, RI

Arnold J. Lee, Mechanical Engineering, NRR Thomas J. Kenny, Senior Reactor Engineer, RI Charles S. Hinson, Health Physicist, NRR Thomas N. Cerovski, Civil Engineering, NRR Approved By: John R. White, Chief

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Radiation Safety Branch l Division of Reactor Safety l l

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9701060145 961224 PDR ADOCK 05000219 f

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EXECUTIVE SUMMARY Oyster Creek Nuclear Generating Station Inspection Report No. 96-10 During this outage inspection, inspectors reviewed the licensee's temporary shielding program and evaluated the adequacy of the licensee's structural calculations for dynamic and static loads, and for proper placement of temporary shielding with respect to plant and personnel safety. The inspectors reviewed the interactions between the Maintenance, Operations, and Engineering departments with the Radeon department at Oyster Creek and reviewed various aspects of the licensee's ALARA program. These areas included: Oyster Creek's dose history and evaluation of doses associated with high dose jobs, source term assessment, source term reduction program, shielding program performance, and other ALARA initiatives which have resulted in significant dose savings. The ALARA program processes were also reviewed.

The review of available industry data suggests that Oyster Creek has a relatively high radioactive source term as do other non-zinc injection BWRs. Oyster Creek appears to experience higher outage exposures than other similar high source term BWRs. The higher aggregate exposure appears to be attributable, in part, to long duration outages and manpower loading of tasks.

The licensee is making a concerted effort to reduce doses at Oyster Creek through the use of ALARA dose reduction initiatives. As a result of an aggressive source term reduction program, the licensee estimates that it will have removed nearly 60 percent of the sources of cobalt in the reactor coolant system as compared to 1990 levels by the end of the current outage. The collective dose of 90 person rem at Oyster Creek in 1995 (a non-outage year) was the lowest annual dose recorded at Oyster Creek since its first year of ,

operation in 1970. The licensee's goal is for Oyster Creek to be in the top dose quartile for '

BWRs within the next few years.

The results of the licensee's cobalt reduction efforts can be seen in the decreasing trend in the recirculation loop contact dose rate readings (BRAC point readings) over the years (Oyster Creek's BRAC readings are less than half of their mid-1980's value). However, current licensee commitments for implementing exposure reduction initiatives may not i significantly affect Oyster Creek's high exposure perforrnance in the near future.

The licensee has implemented numerous ALARA features / techniques at Oyster Creek, such as, significant temporary shielding and installation of some permanent drywell scaffolding.

Some ALARA program areas were identified where enhancements may provide improved exposure performance are provided below for evaluation. For example, better definition of comprehensive exposure spent by location may provide for a better focus for the licensee's shielding efforts to maximize exposure reduction benefits. The job order exposure tracking number system provices this capability, however, inconsistent use of work location specificity currently limits its usefulness. Continued emphasis on improving the drywell shielding should be evaluated. Also, increased emphasis on using remote surveillance techniques in high dose areas should also be evaluate i

A review of the ALARA program process indicated that the practice of screening new work requests early in the job order planning process for radiologically significant jobs is a positive practice which enables the Radcon department to review the job scope and provide HP input well before the job is actually performed. The RER appears to be a good vehicle for evaluating individual jobs (aside from computer network inaccessibility to outside RP departments), and it appears tc be a good tool for on-line maintenance task reviews. However, cornprehensive task revies for exposure intensive areas of the plant during outages are not as effectively reviewed by this process. The inspector also noted that high task exposure estimates do not require a review by a station exposure committee or higher levels of management. Finally, the exposure estimating and tracking systems reside on different incompatible databases and limit the ability to gauge radiological work performance against its estimate and limits the use of compiling exposure estimates for exposure goal determinations.

Management is involved with the ALARA program and dose reduction efforts at Oyster Creek Station. The creation of an independent exposure assessor position, the addition of ALARA to the system engineers performance evaluation, the frequent communications with the staff through newspaper articles and memos to all departments indicate there is constant vigilance in the ALARA area.

The Operations department is committed to ALARA through making changes in the daily activities and in the organization of a committee for purposes of dose reduction.

GPU has incorporated ALARA practices into the major maintenance procedures and there is evidence of time, distance and shielding concepts being used to reduce radiation dose.

The Engineering department has incorporated the ALARA concept into their procedures for the purpose of dose reduction at Oyster Creek. The inspector also verified that system engineers are implementing ALARA in initiatives and designs to reduce dose.

GPU is active in the self assessment of ALARA and dose reductions through the use of audits and individual monitoring of radiological activities. The results of these efforts have led to improvements to procedures for controlling radiological activities.

The licensee has an appropriate engineering program to resolve shielding requests adequately. The shielding installations reviewed were generally sound, and the integrity of the affected equipment, piping and components was generally demonstrated under pertinent design loadings. The licensee calculations and assumptions for the drywell shielding installation and the static load analysis for the turbine were determined to be acceptabl There were areas, however, where additional engineering review is required. These areas include: (1) establishing a vibration monitoring program for the concrete shield walls at both the HP and LP turbine ends, (2) a followup assessment for the potential piping overstress condition of the condensate return line in the feedwater pump room, (3) a ii l

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clarification concerning the classification of the temporary shielding on the drain line in the -

reactor building north wall (elevation 23 ft.), and (4) demonstration of the structural adequacy of the demineralized air cleanup system piping and submitting the design records for staff review.

In summary, licensee management is supportive of dose reduction efforts at Oyster Creek.

Although doses have indeed been dropping at Oyster Creek, and the Radcon department has been consistently working on dose reduction initiatives, the current level of exposure reduction activities is not projected to significantly improve Oyster Creek's exposure performance in the near future. Further improvements may be necessary to improve the overall exposure performance at Oyster Creek, i

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Report Details

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l. Operations  ;

010 Operations 10.1 Ooerations involvement in Radiolooical Dose Reduction (83728)

i Inspection Scope The inspector conducted interviews with operations personnel and reviewed selected documents to assess the operations department commitment to ALAR Observations and Findinas The Director of Operations supports a committee that has been active in reducing dose in the condenser bay area. Condenser bay entries at power has been a chionic source of doses for operators, health physics technicians and maintenance personnel. The committee has accomplished the following: installation of longer lasting live loaded valve packing in applicable valves, decontamination of floor areas to allow the use of street clothes for entry, and reduction in the amount of dose with permanent shie: Jing. TV monitors have been also added to provide the  !

operators observation from remote areas. These efforts have reduced entries into i the area from every 2 days to when require !

Operation supervisors have been recently undergoing training to enhance their '

knowledge of resin transfer and radioactive waste handling that includes emphasis on dose reduction when performing these evolution Two areas of the turbine building were locked to prevent personnel from transversing through them during operation accounting for a reduction of one millirem of radiation each time the short cut not was use The installation of remote cameras in the spent fuel pool heat exchanger area and turbine operating floor were other areas where doses were reduced for the operator The daily and shift rounds by operators were assessed to find the most productive method of making them in order to reduce dos Conclusions The inspector concluded that the Operations department was committed to ALARA through making changes in the daily activities and in the organization of a committee for purposes of dose reductio . _ _ -. . . _ _ _ . _ _ _ _ . .__ __ . _ . _

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11. Maintenance M10 Maintenance M10.1 Maintenance Pertainina to Radioloaical Work (83728) Ln_spection Scoce  :

The inspector conducted interviews and reviewed selected documents to assess GPU's maintenance practices in the areas of ALARA and dose reductio , Observations and Findinas i The inspector reviewed maintenance procedures including 2000-ADM-302.01 "On-Line Maintenance Risk Management" and concluded that ALARA was incorporated into them for 's.e purpose of dose reduction. Requirements were delineated to personnel encouraging pre-planning of maintenance activities by using ALARA concepts (time, distance and shielding). The procedures also encouraged the cooperation and collaboration with radiological control engineering personnel during projects associated with radiatio The inspector reviewed a maintenance activity to rebuild valves ND27-A & ND 28-A in the filter portion of the RWCU system. Various cautions throughout the procedure pointed out ways to keep doses to a minimum, such as delineating the way to transport the valves away from the high radiation area to be rebuilt. The type of shielding was specified as well as the type of cart used to make the transport. The valves were decontaminated and rebuilt in a non radioactive environment resulting in minimizing radiation dose GPU started a program to reduce dose by limiting the number of personnel working on any one component in the interest of ALARA. The program included the cross training of the crafts to aid the other crafts during the performance of work on particular components. During this outage, GPU was using one electrician and one mechanical worker instead of the normal two electricians and two mechanical workers for testing work on MOV Conclusion The inspector concluded that GPU incorporated ALARA practices into the major maintenance procedures and that there was evidence of time, distance and shielding concepts being used to reduce radiation dos _ __ _._ __ _. . - - ._ . ._ __... .

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i Ill. Enaineerina

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b E2 Engineering Support of Tacilities and Equipment i E2.1 Seismic Evaluations of Shieldina installations ,

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, Incoection Scope i

The purpose of the inspection, in the mechanical / structural areas, was to: (1) audit ;

the installatha specification of the shielding, (2) audit the engineering evaluations '

performed for ihe effects of shielding weights on safety-related structures and components, and (3) examine the affected plant structures and components to ensure that the as-installed shielding is structurally sound and consistent with the engineering evaluation The inspection covered the general plant configuration which included the reactor l

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and turbine buildings, as well as the drywell. It was noted that many permanent I shieldings were installed on equipment, piping and components in the reactor and turbine buildings. In addition, there was a substantial amount of temporary shielding installed inside the drywell. The temporary shielding, in the form of lead !

blankets, is to be removed before the plant restarts from the current refueling j outage. The massive reinforced-concrete shield walls, which were installed at both i the low-pressure (LP) and high-pressure (HP) ends of the turbine pedestal deck, I were designed as permanent shielding and will be utilized during normal piant operation Observations and Findinas In order to reduce the ic<iiation exposure of personnel working during an outage at l Oyster Creek plant, the fict.nsee added temporary shielding to selected piping systems, valves, tanks and other components. Permanent shielding was also added to some equipment, piping a1d components to reduce personnel radiation exposure during normal operatio A general observation that was apparent as a result of the inspection of the shielding installations was that the shielding blankets were well secured on the piping and equipment. In addition, based on sampling examination of mechanical plant installations, the staff did not identify design deficiencies for piping and equipment components as a result of the additional weight of shielding. Several equipment and piping components were selected for a general review of the associated analytical evaluations. The review findings for each of these items are as follows:

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(1) Turbine Building Condensate Return Line in the Feedwater Pumo Room

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As indicated in the document reviewed by the inspector, an additional weight of 60 lbs/ft of lead blankets was permanently placed on the piping, resulting in a potential overstress condition for the condensate return line, in the event

of a design basis earthquake. It was not apparent to the staff that a i followup evaluation had been performed by the licensee to assess the potential piping overstress coredition. This item will be reviewed in a future j inspection (IFl 50-219/96-10-01).

i Hiah Pressure Turbine Shield Wall i

A reinforced concrete shield wall, consisting of two "L" shaped sections, ,

was installed at the HP turbine just prior to the start of the current refueling i

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outage. It was a fabricated concrete box structure installed around the HP !

turbine to cover a portion of the main steam piping, and to reduce personnel radiation exposure in the area during plant operations. The wall was of a removable design and is free-standing on the turbine pedesta The concrete shield wall appeared to be well designed and fabricate However, the lack of anchorage to the turbine deck raised some concern The inspector questioned the possibility of interaction between the shield wall and the nearby equipment and components, as a result of potential movement of the shield wall during an earthquake event. The staff was also concerned about the potential adverse effects of vibration and impact between the shield wall and the turbine pedestal foundation during normal l operation. During a conference call on September 30,1996, the licensee l was requested to provide a technical justification for the acceptance of the as-installed configuration, and the licensee made a commitment to establish a periodic monitoring program for the concrete shield wall. On October 2, 1996, the licensee provided the staff with an in-house memorandum (Location: Morris Corp. Ctr., 5320-96-072, dated October 2,1996) which addressed the requested information and commitment, in the memorandum it was stated that there was no anticipated concrete wall movement during ,

normal operation, based on the following: I

• The two "L" shaped walls were interconnected at the top by two steel brackets creating a large mass which was resistant to sliding and overturnin * The walls weighed approximately 96,000 lbs., and sat on a neoprene gasket placed between the floor and bottom of the walls. The gasket acted as a dampening device against vibration and provided a high coefficient of friction to resist slidin l

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• The pedestal itself was a large concrete structure and was believed to have sufficient mass to absorb the level of vibration produced by the i spinning turbine roto In addition, the licensee committed to make provisions in plant procedures ,

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for periodic monitoring of shield wall movement. Should movement be detected during operations, an evaluation is to be performed by the licensee, and a course of action determine The staff considered the licensee's response to be acceptable and requested that a similar vibration monitoring program for the HP turbine shield wall be established for the existing LP turbine shield walls. Based on this, the issue of the turbine concrete shield wallis considered closed.

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(2) Reactor Building Reactor Buildina (Elevation 23 Ft.) Drain Line North Wall i

Two (2) layers of lead blankets were installed on a 4-inch drain line from the !

l floor to approximately the 10-foot elevation of the verticalline. The I

shielding was installed primarily to reduce personnel exposure during torus l

, work and drywell corrosion mitigation and monitoring. Although designated j l as temporary, the inspector was informed that the lead blankets have been l l there for a long time. The licensee was requested to clarify the residency  ;

' time of the shielding. Should the lead shielding be designated as permanent, l the licensee was informed to demonstrate the structural integrity of the drain )

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line, under all pertinent loading condition ;

1 i Reactor Buildina (Elevation 75 Ft.) Demineralized Air Cleanuo System Pioina  !

During the inspection, the licensee was requested to provide the needed j design documents to substantiate the structural adequacy of the air cleanup  ;

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piping, which is located near the Isolation Valve V-6-142 and serves as an interface between the air and water systems. The licensee was unable to retrieve the requested design documents. This item will be reviewed in a future inspection (IFl 50-219/96-10-02).

(3) Drywell A number of temporary shield blankets inside the drywell were inspected, l

especially those installed on the suction and discharge piping above the i recirculation pumps. No deficiency was identified for the shielding installation on the piping. In addition, permanent scaffolding frames at recirculation loop risers were also inspected with no observed deficiencie To supplement the plant walkdown, the staff also reviewed the design documents for the following temporary shielding installed inside the drywell:

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Recirculation Pioina From Suction to Discharae Valves i

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A tabulation for the weight limit of lead blankets was provided for various l pipe diameters, either in terms of a concentrated weight or uniform

, distributed weight. The applied load was further limited to 4 feet in length !

l along the pipe. The pipe stresses were found to be within allowable limit Recirculation Suction and Discharae Risers  !

Based on the information provided, the risers and the permanent scaffolding l

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frames were capable of carrying the shielding blankets as proposed by the i licensee. Base plate hold downs and bracing systems were shown to be adequat Recirculation Suction Nozzley.grd Pioina (From Elevation 46 Ft. to 54 Ft.) I

4 A table of weight limits for piping loads was utilized. In addition, criteria for equipment shielding loading and pipe supports evaluation were also provide ! The licensee determined that, by limiting a total number of lead blankets to l

30, about 33 lbs, apiece, on piping from elevation 46 ft. to 54 ft., the pipe stresses and support loads would remain within allowables. The staff found i the licensee's approach acceptabl ;

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l Core Sorav Pioina and Penetration

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Based on the same loading criteria as stated in item a, the licensee

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determined that, by installing up to four (4) lead blankets, 33 lbs. apiece, on the core spray system nozzle, the stresses would still be within allowable limits. This was found acceptabl ;

Recirculation Pioino Discharae Nozzles The acceptable number of lead blankets were again determined by the same !

loading criteria as in item Drvwell Cleanu1 Syste,rn at Valves V-16-62 and V-16-63 Up to sixteen (16) 1 ft. x 2 ft. lead blankets on the drywell cleanup system at the valves were installed only if the valves were to be repaired. The blankets were installed using seismically qualified TY-wraps. This was found acceptabl .- -- _ - -- -. - - - . . - - _ . _

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Shutdown Coolina System Lines

For shutdown cooling lines and Valves V-17-19 & V-17-54, shielding j blankets were installed to reduce the general area dose for weld inspection

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work. Based on analyses performed by the licensee, it was determined that an additional weight of forty (40) lead blankets would result in acceptable pipe stresses. This was found acceptabl l Based on the staff observations during the plant walkdowns and on the review of

the installation specification, configuration change documents, design calculations !

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and design verification documents, as provided by the licensee, the inspector '

determined that the licensee has initiated appropriate remedial actions in resolving the engineering concerns raised in the Oyster Creak ALARA issue. With the exception of the unresolved items stated in (1)a and (2)b above, the licensee has

generally confirmed the structural adequacy of equipment, piping and components by referencing to previous calculations or predetermined allowable loads due to lead blanket shielding. Where consideration for earthquake loading was needed,

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especially for permanent shielding installations, dynamic effects were accounted for

in the licensee's evaluation.

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c. Conclusion

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Based on selected sampling of plant shielding installations, and on applicable

engineering evaluation documentation provided by the licensee, the inspector concluded that the licensee has an appropriate program to resolve shielding requests

adequately. The shielding installations were generally sound, and the integrity of j

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, pertinent design loadings. In this inspection, the staff has observed the  !

aggressiveness of the licensee's efforts in reducing personnel radiation exposures at i Oyster Creek by installing substantial amount of shielding on the candidate equipment, piping and components.

l There were areas, however, where continued effort from the licensee was required,

in order to complete the program. These were

(1) a vibration monitoring program i

for the concrete shield walls at both the HP and LP turbine ends, (2) a followup assessment for the potential piping overstress condition of the condensate return line in the feedwater pump room, (3) a clarification concerning the classification of the temporary shielding on the drain line in the reactor building north wall (elevation 23 ft.), and (4) demonstration of the structural adequacy of the demineralized air cleanup system piping and submitting the design records for staff review.

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E2.2 Static Enaineerina Evaluations of Shieldina Installations inspection Scone in order to evaluate the effectiveness of the OCNGS ALARA program, Region I staff, v ith the support of NRR engineering, conducted a review of the licensee procenures shielding installation, and engineering calct'ations. .The purpose of the inspechon by the Civil Engineering and Geosciences Branch, NRR was to: (1)

review the licensee civil engineering calculations for temporary and permanent shielding in the turbine building and the drywell to ensure the conclusions of safety evaluations were made using sound engineering judgement and acceptable assumptions, (2) to review the engineering documentation and references to ensure proper review was conducted for each mcdifications, and (3) to review the plant quality assurance program for the installation of shielding, Observations and Findinas General Observations The majority of the shielding installations in the plant were in the form of lead blankets secured to piping and scaffolding. In some cases, the licensee engineering personnel provided approval for several shielding installations on piping and structures based on a general safety evaluation which only referenced calculation In some cases, the supporting calculations were only available at the corporate office in Parsippany, New Jersey and were not readily available for inspectio The eventual review of licensee calculations and a walkdown of the effected components demonstrated that the licensee has considered the effect of the shielding static load adequately and the inspector agrees with the licensee conclusion that the additional loads to the components will not have any adverse effects on plant safet The inspector reiterated that licensee engineering personnel should be cautious in referencing generic safety evaluations and should be proactive in their evaluations of RP shielding proposals by evaluating alternative shield designs or other options for solving dose reduction problems. The inspector recognized that the Oyster Creek engineering staff continues to play an integr:' role in the plant ALARA program by providing ideas for dose reduction activmo i Turbine Shieldina The licensee installed a massive concrete shield wall around the high pressure turbine at elevation 51' on the turbine pedestal. The reinforced concrete shield was robust and designed in accordance with Oyster Creek commitments for Class il structures. The inspector questioned the adequacy of the licensee documentation and consideration of loads on the turbine floor from the concrete shield and all modifications made since the original design of the floo ._ _

Review of licensee calculations and assumptions revealed that the concrete structure placed an additional load to the turbine of approximately 1% of the original dead load. In itself, this is considered an insignificant contribution to the floor loading. The licensee Seismic Qualification Utility Group (SOUG)

documentation and design review process was discussed with a senior engineering manager to verify that all static loads to the turbine floor since original design were properly accounted for and considered in the floor loading calculations. Upon completion of the SQUG documentation review and staff discussions, it was concluded that the Oyster Creek engineering staff properly considered and documented all static floor loadings appropriately.

Drvwell Scaffoldina During the most recent outage, temporary and permanent scaffolding was installed in the Oyster Creek drywell to support lead blanket shadow shielding. The scaffolding supported up to 30 lead blankets which, transmitted through the scaffolding, resulted in a drywell floor loading of approximately 200 pounds per square foot. The lead shielding on the scaffolding was installed with a factor of safety greater than 1.2 while the plant was shutdown and coo'ad down.

Furthermore, the permanent scaffolding was securely bolted to the floor with support plates. After complotion of an outage and the lead shielding was removed, the bare scaffolding loading to the drywell floor was considered minimalin comparison to the original design. The licensee assumptions and calculations concerning the drywell scaffolding during all modes of operation were considered acceptable to the inspector.

The installation of permanent scaffolding for shielding support was recognized by the staff to be a proactive step toward dose reduction because the permanent supports would not be installed and removed during each outage, saving time and exposure in the drywell.

Quality Assurance An engineering review of the quality assurance program was conducted to evaluate the procedures in place to ensure that all shielding designs were properly implemented. Quality assurance documentation was reviewed and both the engineering and the health physics managers interviewed understood their role in the process for assuring the shielding packages are installed and secured properly.

The inspector raised questions as to the role of plant operations personnel in the 1 walkdowns and confirmed that all safety-related modifications for exposure I reduction were reviewed by operations managemen )

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The shielding and scaffolding in the drywell were found to be instav' in )

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accordance with the procedures and engineering evaluations reviewed in the documentation. ihe inspector concluded that the quality assurance oversight of the ,

shielding program was being performed at a level which ensured plant and personnel safet . -, - - -. . - - .. - . . - . . - -_ _ . -. .

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The health physics, engineering, and plant operations personnel communicated effectively on the scope and responsibility of work, and recognize the ALARA benefits of performing a shielding installation correctly the first tim Conclusions it was concluded that the licensee calculations and assumptions for the drywell shielding installation and the static load analysis for the turbine were acceptabl The inspector noted the proactive steps of the licensee staff toward dose reduction in the Oyster Creek ALARA program such as drywell shielding and permanent scaffold installation. Through the installation of several permanent structures and ,

shield supports, and recognizing the benefits of a fundamental quality assurance !

program, the licensee demonstrated an effective ALARA progra I E.10 Engineering  !

E10.1 Enaineerina Pertainina to Radioloaical Controls (IP 37550) Inspection Scope The inspector conducted interviews and reviewed selected documents to assess GPU's engineeririg practices in the areas of ALARA and dose reductio Observations and Findinas All job orders written pass through the radiation engineering group for radiological i assessment. This ensures radiological engineering was aware of emerging i radiological wor All system performance teams (SPT) assigned to systems containing radiological substances had rad con personnel assigned. These teams were assigned to assist ;

the system engineers to evaluate system conditions, aid in configuration changes l and meet on a quarterly basis for an assessment of the system. Several SPTs had a l list of radiological control engineers that had been assigned to systems who will attend SPT meetings for the purpose of contributing to dose reduction effort A dollar per person-rem figure did not exist and all dose reduction jobs were assessed on a case-by-case basis. GPU spent up to $30,000.00 per person-rem for projects to reduce exposur _

Engineering considerations for reducing the source term at Oyster Creek included:

• Zinc injection

• Chemical cleaning e Covering piping with mirror insulation that contained lead, and l doubled as a shield for radiation reduction  ;

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• Water shielding installed as tanks that can be drained or filled as '

needed

• Cobalt 59 reduction j i

The inspector reviewed a configuration change that was installed to reduce dose associated with the reactor water cleanup system. The design of the system logic l prompted a deviation report because of the reactor water clean up filter material '

being left in the inlet piping following the backflush of the filters and the return of j the filters to service under low flow conditions. The filter material was highly ;

radioactive and cleanup expended high doses to plant personnel. The system logic was changed to prevent this problem by a valve manipulation preventing the filter material from getting into the intet lin The inspector reviewed engineering procedures including Engineering Standard ES-007 "ALARA Guidelines for Configuration Changes" and concluded that ALARA was incorporated into them for the purpose of dose reduction. Requirements were delineated to personnel encouraging pre-planning of engineering activities by using ALARA concepts (time, distance and shielding). The procedures also encouraged the cooperation and collaboration with radiological control engineering personnel during projects associated with radiation.

c. Conclusions The inspector concluded that the engineering department incorporated the ALARA concept into their procedures for the purpose of dose reduction at Oyster Cree The inspector also verified that system engineers implemented ALARA initiatives and designs to reduce dose, m

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IV. Plant Sucoort l

R1 Radiological Protection and Chemistry Controls l l

R1.1 Dose Historv and Outaae Radioloaical Controls

] Inspection Scope The inspector reviewed the past dose history for Oyster Creek, reviewed outage status reports comparing actual job doses for the current outage (16R) with job

dose estimates, and reviewed projections for future annual doses at Oyster Cree The inspector also reviewed the dose history of several dose intensive jobs at Oyster Creek over the past 10 years and reviewed the licensee's efforts to lower

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doses associated with these job Observations and Findinas Dose Historv

Oyster Creek has had a history of high collective doses. Over the 5-year span between 1991 and 1995, Oyster Creek's average collective dose of 638 person-

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rem was the highest in the U.S. Oyster Creek's doses peaked in the mid 1980s and )

the 3-year average doses have been gradually declining since that time. In 1995, i

Oyster Creek's annual dose of 90 person-rem was its lowest annual dose since i 1970, Oyster Creek's first year of operation. The licensee predicts that the annual l collective dose for 1996 fan outage year) will be less than 500 person-rem. The

licensee predicts that the annual dose in 1997 (a non-outage year) will be i approximately 60 person-rem.

Some of the jobs which have historically contributed to the high doses at Oyster Creek include in-service inspections (ISI), scaffolding and insulation work, refueling, I

and control rod drive work. The doses associated with these jobs at Oyster Creek have been gradually trending downwards over the past 10 years.

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Drvwell Shieldina and Scaffoldina Since approximately 80% of the outage dose is accrued in the drywell, the licensee installed 72,000 pounds of temporary lead shielding in the drywell during the current outage to lower personnel dose The quantity of lead installed above each recirculation pump has been increased from 600 pounds (last outage) to 2000 pounds (current outage). The temporary shielding over one of the recirculation lines adjacent to an access ladder in the drywell has been increased from one layer of lead blankets to two layer i l

13 This temporary shielding served to reduce personnel doses during ISI work. During the previous outage, personnel were permitted to move lead shielding in the drywell

. without a shielding modification authorization form. This resulted in increased personnel doses in those instances where the shielding was not replaced once the ISI on a component or pipe was completed. During the current outage, shielding technicians, who are present at the drywell 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day, were the only ones authorized to move shielding for the worker In order to reduce the doses associated with the installation and teardown of temporary scaffolding during outages, the licensee converted approximately half of the temporary scaffolding in the drywell to permanent scaffolding during the current outage. The licensee added permanent scaffolding in the condenser bay and in <

other areas of the plant. The licensee utilized a scaffolding control program which specified the size and location of scaffolding needed for any particular work location in the plant. The use of such a program facilitated the erection of temporary !

scaffolding in the plant. The licensee also used an Exposure Tracking Number system which tracked each job by location, component, system, and type of wor The licensee used this system to identify jobs that may be adjacent to each other so that the same temporary scaffolding and shielding can be erected to service multiple job Control Rod Drive Imorovements The licensee now ships its control rod drives (CRDs) offsite for rebuilding, instead of rebuilding them onsite. The licensee used an automated CRD changer and changed its CRDs every other outage instead of every outage. During CRD changeouts, the licensee reduced the number of people working uno sr the vessel from three to tw These initiatives reduced the doses associated with CRD changeout at Oyster Creek. For example, in 1990 the exposure cost to replace one CRD was person-rem. During the previous refueling outage in 1994 this had dropped to person-rem per CRD replacemen Minimize Occupancy in another effort to reduce outage doses, the licensee restricted the number of people permitted in the drywell during the current outage. Casual visits to the drywell were eliminated. Cyclic training giten to operations and maintenance personnel now included instructions on controlling crew size for jobs performed in high dose rate area c. Conclusions High dose history has improved. In particular, plant operation doses and outage exposures and associated high dose jobs have trended down over the past 10 years. Significant drywell shielding was used during recent outages and during the current outage, some permanent drywell scaffolding was installed. Training was offered to operations and maintenance personnel on controlling crew size for exposure significant areas of the plan I

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R1.2 Source Term Reduction Initiatives Insoection Scone The inspector met with the Manager of Radioiogical Engineering, the ALARA coordinator, and other members of the Radcon group to discuss what initiatives Oyster Creek were taken to reduce the radioactive source term at the plant. The I inspector also reviewed several licensee reports on this subjec ! Observations and Findinus Oyster Creek': high dose rates have been due, in part, to its high source term. The licensee estimated that cobalt-60 accounted for approximately 84 percent of the dose incurred in the drywell, and implemented many initiatives to lower the source term at Oyster Cree j The licensee established a Source Reduction Committee which coordinated the Oyster Creek Source Reduction Program. One of the activities of this committee has been to identify all valves which contain stellite (cobalt) in the plant. The licensee used this list to evaluate the feasibility of replacing these valves with non- i stellite valves when these valves require replacement. Oyster Creek replaced the !

feedwater regulator valves and valves in other systems with non-stellite valve When machining / lapping valves containing stellite, the licensee follows a procedure to ensure that stellite fines were not introduced into the reactor coolant syste The valve seats in the main steam isolation valves (MSIVs) at Oyster Creek contain stellite. Oyster Creek approved a procedure for installation of NOREM valve seats (NOREM contains no stellite)in the MSIVs when the valve seats currently in these valves require replacement (This procedure has not yet been utilized).

The licensee estimated the low pressure (LP) turbines contributed approximately 34 percent of the cobalt to the reactor coolant system. The licensee began replacing the three LP turbines during the 1991 outage and, by the end of the current outage, will have replaced all three. The replacement turbines included lower cobalt containing components. Another major source of cobalt were the stellite pins and rollers in the control roo (CR) blades. The licensee began replacing the stellite containing CR blades in the 1988 outage and by the end of the current outage expects to replace all but 53 of the 137 CR blades in the core. As a result of the above initiatives, the licensee believed they will have removed nearly 60 percent of the sources of cobalt in the reactor coolant system (as compared to 1990 levels) by the end of the current outage. Actuallowering of source term would be realized over several years as the preexisting cobalt-60 decays (5.7 year half life) and the lower level of cobalt input reaches equilibriu The licensee performed 2 system decontaminations in the last 10 years. Oyster Creek decontaminated the reactor recirculation system (RRS) during the 11R outage (1986) and decontaminated the RRS and reactor water cleanup system during the 13R outage (1991). Each of these system decontaminations resulted in lower drywell dose rates for approximately 3 years. Another effort, which the licensee

has recently taken to lower drywell dose rates, has been to reduce the iron concentration in the feedwater from 6 ppb to between 1 and 2 ppb. Since iron in the feedwater enhances cobalt transport, the licensee predicted that this reduction in iron concentration should result in a significant reduction in drywell dose rates after several year Based on the present level of exposure reduction initiative commitments, and with reference to the licensee's engineering study dated March 7,1996, " Cobalt Reduction Evaluation", the licensee estimated that given current cobalt reduction initiatives to replace control rod blades with non-stellite materials and repairing future MSIVs with non-stellite valve seat materials, it would take approximately 12 years to reach the projected first quartile dose goal for BWRs (assumes a 3% per year decrease in BWR exposures over that time period).

The licensee has also been considering the use of depleted zinc injection at Oyster Creek. The same engineering study indicates that implementation of a depleted zinc injection program would enable the licensee to reach their first quartile goal in approximately 6 years. However, the licensee had only partially funded this exposure reduction initiative recommendation with additional funding being solicited outside the utility. Another option reviewed estimated that if recirculation system chemical decontamination was performed followed by depleted zinc injection, Oyster Creek would realize the first quartile BWR exposure goal in approximate ly 2 year Conclusions )

i The licensee was making a concerted effort to reduce the source term at Oyster Creek by removing cobalt-60 from the system and replacing components in the reactor coolant system which contain cobalt-59. The results of the licensee's efforts were seen in the decreasing trend in the recirculation loop contact dose rate ,

readings (BRAC point readings) over the years (Oyster Creek's BRAC readings are j less than half of their mid-1980's value). However, current licensee commitments !

for implernenting exposure reduction initiatives may not significantly affect Oyster Creek's high exposure performance in the near futur I R1.3 Ovster Creek Source Term Assessment Insoection Scope The inspector conducted independent dose rate measurements and observations in the plant, reviewed licensee survey records, and made comparisons with other BWRs in order to assess Oyster Creek's source ter .. _ -.-_ - . - . . - .- . --

I Observations and Findinas BRAC Survev Comparisons EPRI Report, NP-6787, March 1990, entitled, "BWR Radiation-Field Assessment: i

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1986-1988" compiled contact dose rate readings from recirculation system piping ,

for 17 BWRs during 1986-1988 and determined that there were two groups of BWRs: those that were natural zinc plants (brass tube condensers) or those that added zine to the feedwater system, and that these constituted low source term BWRs with an average contact dose rate with recirculation system piping of 113 mR/hr. The other group of plants constituted those that were non-zine plants with an average contact dose rate of 268 mR/hr reported. A more recent EPRI report, j EPRI TR-104606, December 1994, entitled, " Evaluation of Recent Experience Using i Zinc Addition to Reduce BWR Primary System Radiation Buildup" utilizing data through 1994, indicates that among 13 BWRs, the average recirculation pipe contact dose rate for non-zinc injection BWRs was approximately 350 mR/hr, whereas the average contact dose rate for zinc injection BWRs was 135 mR/h ,

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Oyster Creek (a non-zinc injection BWR) reported average recirculation pipe readings of 390 mR/hr in this outage, as compared to an average of 235 mR/hr during the previous outage. Although the contact dose rates were higher this outage, the

preshielded general area dose rates in the drywell were the same as the previous

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outage or actually lower in some areas. Oyster Creek is a BWR-2 plant with five j recirculation loops as compared to later BWRs with 2 recirculation loops. Nine Mile Point Unit 1 is a comparable BWR-2 plant and Pilgrim Station is representative of a

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2 loop BWR plant. Nine Mile Point Unit 1 and Pilgrim Station are similar non-zinc BWRs with recent recirculation pipe readings of 300 mR/hr and 330 mR/hr, i respectively.

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In order to compare Oyster Creek's source term result on work area dose rates, the inspector compared the primary containment (drywell) general area dose rates of Oyster Creek, Nine Mile Point Unit 1 and Pilgrim Station (similar non-zinc BWRs) as provided below.

j UNSHIELDED DRYWELL GENERAL AREA READINGS l

Rgttom level Entry level 2nd level oyster Creek (1996) 50-150 mR/hr 40-150 mR/hr 40-80 mR/hr a

q Nine Mile Point, U1 (1995)80-200 mR/hr 40-80 mR/hr 40-120 mR/hr j

! Pilgrim (1995) 100-120 mR/hr 30-70 mR/hr 40-120 mR/hr Recire. Nozzle mezz 150-250 mR/hr

• Conclusions Based on the review of industry data, Oyster Creek's source term and radiological challenge appears to be similar to other non-zine BWR .

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Outaae Dose Comoarisons The inspector compared outage exposure data for the same three plant Outsae Exposure outaae Lenath Rate (Person-rem / day)

oyster Creek (1994) 705 person-rem 97 days (1996 partial) 287 person-rem (partial) 20 days (partial) 1 Nine Mile Point, U1 (1995) 311 person-rem 55 days Pilgrim (1995) 410 person-rem 73 days Although source term comparisons appear to be similar, Oyster Creek appears to accumulate exposure at a faster rate than similar BWRs. Further data comparisons of outage tasks and manpower loading of tasks were not available. Based on discussions with licensee personnel, historically, work craft lines of responsibility have dictated large work teams. The licensee indicated that recent negotiations  !

with the workers' unions have allowed some cross training of some workers to i reduce the size of some work teams (see Section M10.1 of this report).

The review of available industry data suggests that Oyster Creek has a typical high radioactive source term as do other non-zinc injection BWRs. Oyster Creek experiences higher outage exposures than other similar BWRs that may be attributable to long duration outages and higher manpower loading of tasks. This represents an area of opportunity for significant exposure reduction R1.4 Source Term Mitiaation - Shieldina Insoection Scooe The inspector conducted independent surveys in the drywell, reviewed licensee survey documents, and conducted interviews with applicable personnel to ascertain the actions taken to reduce the effects of the radioactive source term to the outage workforce, Observations and Findinas The licensee reported the use of approximately 72,000 pounds of shielding in the drywell. This was up from 54,000 pounds utilized during the previous refueling outage. Additional shielding was used on providing " wings" on the vertical reactor recirculation piping on the entry level to cover the sides of the pipes that previously were shielded on the front surfaces up to a height of approximately 8 feet. The bottom elevation was not shielded as no significant work was scheduled on the recirculation pumps. The second elevation (46') contained some shielding on l

celected recirculation pipe suction elbows and others were not shielded. Additional

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shielding was provided at other miscellaneous locations in the drywell. Based on a review of pre-shielding and post-shielding surveys, and the inspector's own shielding surveys, the licensee had effectively reduced the general area dose rates

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i on the entry level of the drywell from 40-100 mR/hr down to 25-50 mR/hr. The '

lower elevation where no significant work was planned, and on the second  ;

elevation in the drywell, the general area dose rates remained unchanged from the '

unshielded conditio The inspector reviewed the licensee's emphasis on shielding the entry level of the drywell as opposed to the second elevation, as very little entry level drywell work was scheduled (no MSIV work or CRD replacement). The licensee did not have detailed knowledge of estimated drywell doses by elevation or areas within each elevatio The inspector reviewed the capability of the job order computer system that categorizes each job using exposure tracking numbers (ETN). This system is approximately 15 years old and identifies the job location by building (or drywell), i elevation, system, equipment component, and work group. The ETNs can provide  ;

actual exposure data sorted by the above categories when effectively utilized. The

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inspector requested actual exposure data sorted by each elevation of the drywell for the previous 15 R refueling outage and the current (partial) 16R outage in order to determine the percentage of actual dose spent working on the various levels of the drywell. The results are presented belo i l

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Bottom fevel Suboile Entry level 2nd level Too level All levels 15R 1% 8% 2.5 % 10% 1.7 % 77 % 477.591 rem l

16R(partial) 2.3 % 1.2 % 3.7 % 6.7 % 0 86.1 % 226.491 rem

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As can be seen from this data, most of the drywell jobs are not identified by a specific location. Examples are scaffolding and insulation removal. Although these tasks do involve specific locations, the work planning has not carried this i information through to the ETN system to allow an accurate sort of estimated exposures by drywell elevation. Based on the information available, the second .

level of the drywell appears to be the more exposure intensive location and appears  !

to be the area for prioritizing exposure reduction efforts (although this assessment is somewhat speculative since a large portion of the exposure in the drywell is not ,

specified by elevation).  ;

The inspector reviewed the results of the licensee's drywell shielding efforts through independent surveys and reviews of licensee survey data. The results of tilis review are provided belo Preshielded Bottomlevel Entrv level 2nd level Too level 50-150 (100 avg)mR/hr 40-150 (80 avg)mR/hr 40-80 (50 avg)mR/hr 20-60(40 avg)mR/hr Shielded Same as above 25-50 (40 avg)mR/hr same as above same as above

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The licensee's efforts focused on the entry level and achieved an average dose reduction of 50%. All other elevations remained the same. It may be assumed that !

if the entry level shielding were not installed, then the exposures for this level may i have been double and with reference to the ETN data, the entry level and the i second level may represent equal areas of exposure reduction potential. It is of interest that, for the previous 15R outage when a similar entry level shielding priority was pursued (slightly less effective), the ETN data suggests four times as much exposure was accrued on the next higher elevation. The inspector also noted significant dose rate gradients in most drywell work areas indicating continued shielding potential remain l Conclusions ,

I The licensee has implemented numerous ALARA features / techniques at Oyster i C.eek, such as, significant temporary shielding and installation of some permanent ;

drywell scaffolding. Some ALARA program creas were identified where i enhancements may provide improved exposure performance are provided below for evaluation. Better definition of comprehensive exposure spent by location to l provide for a better focus for the licensee's shielding efforts to maximize exposure !

reduction benefits. The job order exposure tracking number system provides this j capability, however, inconsistent use of work location specificity currently limits its usefulness. Continued emphasis on improving the drywell shielding should be ;

evaluate '

R1.5 ALARA Features and Job Plannino inspection Scope The inspector met with the Manager of Radiological Engineering, the ALARA Coordinator, and other members of the Radcon group to discuss what ALARA initiatives the licensee has instituted to lower personnel doses at Oyster Creek. The inspector also toured the reactor, turbine, and radwaste building Observations and Findinos ALARA Features The licensee used a video camera in the condenser bay area during operation to monitor the area for leaks, thereby minimizing the number of personnel entries required in the area for such observations. A video camera was also used in the new radwaste building to remotely observe the filling of radwaste liners. The j inspectors noted that video cameras were not used for remote surveillance in other l high exposure areas, such as the drywel l!

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20 The licensee used telemetric dosimetry to remotely monitor doses of personnel working in high dose rate areas. This dosimetry has also been used by divers working in the spent fuel pool. Four such monitors were permanently used in the drywell to provide readouts of drywell dose rates during operation Oyster Creek utilized a surrogate tour video system. This system can store up to 50,000 digital images of the plant and the images in the system can be updated to 1

reflect equipment changes. Video screens for this system were located in the Radeon office are- near the access point to the reactor building, and at corporate offices in Parsipparo NJ. The licensee used this system for training purpose Personnel requiring a cess to a high dose area to perform maintenance work can use the surrogate tour system to " access" the area beforehand and become familiar

with the location of plant components before actually entering the high dose rate l work locatio Job Plannina

Job orders were reviewed by the Radcon department soon after their inceptio This gave the Radcon department sufficient time to ensure that potentially high dose jobs receive the necessary ALARA reviews to ensure that personnel doses are !

minimized. The job order computer system included the actual RWP information ;

i I and was available for job planning purposes across station departments by local area network, however, the radiological engineering reviews (RERs) were not readily ;

accessibl I Whenever a job was expected to exceed two person-rem, the Radcon department prepared a radiological engineering review (RER) report. The inspector considered

the two person-rem a very good low threshold level for initiating radiological

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engineering reviews. The inspector noted that there was no higher exposure threshold to require a station committee or higher management review (e.g., ALARA committee review).

The inspector reviewed the Radiological Controls Outage Reports for the last three outages for a review of lessons learned from each outage. While the outage reports

, for the 13R and 14R outages contained good descriptions of the jobs performed and lessons learned during these outages, the outage report for the 15R outage contained minimalinformation on lessons learned from the outag Job orders requiring RWPs were processed by the radiation control group and provided with exposure astimates, which in turn (> 2 person-rem) initiated radiological engineering reviews (RERs). RERs result in more refined job order

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exposure estimates. The inspector noted that these individual exposure estimates were not used to determine outage dose goals or for establishing an outage exposure graph timeline in order to gauge outage performance. The licensee indicated that a summary of job order exposure estimates was not accurate and was not used for outage exposure goals or outage tracking purposes. Instead, the actual exposure performance during the previous refueling outage was used as the basis for the current refueling outage exposure estimate, with some adjustments i

21 made for major work differences and to fit the outage duration window. The inspector also noted that the exposure and person-hour estimate inforrnation resided in the job order database system (GMS-2) while the actual RWP exposure data resided in another incompatible database system (REM). RWP actual versus estimate comparisons were not automatic and were made on an as-needed basi Conclusions The inspector noted that the practice of screening new work requests early in the job order planning process for radiologically significant jobs was a positive practice which enabled the Radcon department to review the job scope and provide HP input well before the job was actually performed. The RER appeared to be a good vehicle for evaluating individual jobs (aside from computer network inaccessibility to outside RP departments), and it appeared to be a good tool for on-line maintenance task reviews. However, comprehensive task reviews for exposure intensive areas of the plant during outages were not effectively reviewed by this process. The inspector also noted that high task exposure estimates do not require a review by a station exposure committee or higher levels of management. Finally, the exposure estimating and tracking systems resided on incompatible databases, which limited the ability to gauge radiological work performance against its estimace and also limited its use for compiling exposure estimates for exposure goal determinations.

R7.0 Quality Assurance in RP&C Activities (IP 83728) Insoection Scoce The inspector conducted interviews and reviewed selected documents to assess GPUs self assessment in the areas of ALARA and dose reductio Observations and Findinas GPU conducted audits and monitoring of selected work practices in order to evaluate program effectiveness. The inspector reviewed one audit and two QA surveillances that were performed in the last year in the area of radiological controls and ALARA. These reports identified areas of weaknesses regarding ALARA. As a result of these findings, GPU changed certain procedures to address the weaknesses. For example: Audit S-OC-96-10 identified a weakness in the implementation of industry guidelines and the need for increased management and supervisory leadership in implementing the radiological control progra Radiological control procedures 6630-ADM-4110.02 & 04 were changed to bolster pre-job briefings to include the presence of supervision and managemen In addition to the above self assessment ALARA program reviews, the licensee requested an industry ALARA program assist inspection that was conducted in May of 1996. The industry asset inspection findings were similar to the NRC's perception of licensee performance in that there continued to be room for improvement in the ALARA program with a continuing exposure reduction challeng l 22 Conclusions The inspector concluded that GPU was active in the self assessment of ALARA and dose reductions through the use of audits and individual monitoring of radiological activities. The results of these efforts have led to improvements to procedures for controlling radiological activitie R8.0 Miscellaneous RP&C lssues R8.1 Manaaement involvement in Radioloaical issues (837281 Inspection Scoce The inspector conducted interviews and reviewed selected documents to ascertain management commitment to radiological dose reductio Observations and Findinas The inspector assessed GPU's management commitment and involvement in ALARA and noted the following:

e Site management used the site newspaper " Nuclear Power Points" to inform site personnel of ALARA accomplishments and ongoing initiatives to reduce dose through ALAR * Memos to managers with staff distribution, discussed ALARA issues for the purposes of reducing dose (e.g., Memo of July 29,1996, discussed chemical decontamination and zinc injection studies as methods of reducing the source term).

System engineers (SE) have had ALARA items added to their performance evaluation stating, " Ensure source term reduction and ALARA concerns are considered on your System." SE walkdown guidelines included: " Systematically walkdown all accessible areas within your system boundaries consistent with ALARA." An example of ALARA practiced by system engineering was the recent change of fuel pool cooling pump (FPCP) surveillance from monthly to every 3 months, and a design change moving vibration monitor readout to an area outside of the radiation area of the FPCP. These initiatives reduced dose for applicable surveillance activities by a factor of thre GPU management created a " Dose Champion" who reported directly to the Station Director with the following responsibilities: internal auditor / consultant on radiological matters pertaining to radiological dose reduction and for the promotion of ALARA by interaction with alllevels of management and bargaining unit personnel for dose reduction.

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e An example of the dose champion's oversight was identified as follows. A reactor

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water cleanup system filter maintenance job was stopped by the dose champion so

that personnel could regroup and properly prepare for a significant radiological task.

, The dose champion had discovered, during one of his observation tours, that the l

pre-job briefing conducted by the group performing the work was not fully I t

coordinated regarding sequencing of the crafts involved. The groups reconvened I later that day and successfully performed the evolution. Projected doses were

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reduced as a result of more thorough preplannin Conclusions i

j Management was involved with the ALARA program and dose reduction efforts at

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Oyster Creek Station. The creation of the dose champion, the addition of ALARA to the system engineer's performance evaluation, the frequent communications with i

the staff through newspaper articles and memos to all departments indicated there j was constant vigilance in the ALARA are R8.2 Review of UFSAR Commitments i While performing the inspections discussed in this report, the inspectors reviewed Section 12.1 of the UFSAR that related to the ALARA program area that was j inspected. The inspectors verified that the UFSAR wording was consistent with the I observed plant practices, procedures and/or parameter !

V. Manaaement Meetinas ,

l Exit Meetina Summarv  :

The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on September 27,1996. The licensee acknowledged the findings presente ;

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4 I

) PARTIAL LIST OF PERSONS CONTACTED Licensee

F. Applegate, Assessor, Nuclear Safety Assessment

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W. Behrle, Director, Emergency Response W: ting)

. G. Busch, Manager, Regulatory Affairs W. Cooper, Manager, Radiological Engineering

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D. Croneberger, Director, Equipment Reliability l J. DeBlasio, Manager, System Engineering 1 J. Hildebrand, Director, Plant Maintenance

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S. Levin, Director, Operations and Maintenance K. Mulligan, Director, Plant Operations T. Quintenz, Manager, Mechanical / Structural Engineering M. Roche, Vice President and Director, Oyster Creek Station R. Shaw, Director, Radiation Control / Safety ,

D. Slear, Director, Configuration Control l M. Slobodien, Director, Radiological Health and Safety l K. Wolf, Manager, Radiological Field Operations NRC L. Briggs, Senior Resident inspector T. Ceroski, Civil Engineer, Civil Engineering Branch, NRR C. Hinson, Health Physicist, Radiation Protection Branch, NRR T. Kenny, Senior Engineering Inspector, Region i A. Lee, Mechanical Engineer, Mechanical Engineering Branch, NRR J. Noggle, Senior Radiation Specialist, Region i INSPECTION PROCEDURES USED IP 83728: Maintaining Occupational Exposures ALARA ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-219/96-10-01 URI Seismic load evaluation for permanent shielding added to the condensate return line located in the feedwater pump roo /96-10-02 URI Structural design evaluation for the demineralized air cleanup system piping located on the reactor building 75-foot elevatio Closed None

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k l LIST OF ACRONYMS USED .

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ALARA As low as is reasonably achievable BRAC' BWR Radiation Assessment and Control Program CR Control rod CRD Control rod drive ETN Exposure Tracking Number <

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FPCP Fuel pool cooling pump ISI In-service inspection MOV Motor-operated valve MSIV Main steam isolation valve RCA Radiological controlled area RER Radiological Engineering Review ,

RP Radiation Protection RRS Reactor recirculation system j RWCU Reactor water cleanup  ;

RWP Radiation wp-k permit  !

SE System engineer  !

SPT System performance teams

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