IR 05000317/1998003

Insp Repts 50-317/98-03 & 50-318/98-03 on 980119-27. Violations Noted.Major Areas inspected:follow-up & Verification of Corrective Actions from & Violations Noted in NOV

ML20217C169
Person / Time
Site:	Calvert Cliffs
Issue date:	03/20/1998
From:	Jason White NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20217C114	List: IR 05000317/1998003 ML20217C109 ML20217C142
References
50-317-98-03, 50-317-98-3, 50-318-98-03, 50-318-98-3, NUDOCS 9803260342
Download: ML20217C169 (16)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

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Docket Nos:

50-317;50-318

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License Nos:

DPR-53; DPR-69 Report Nos:

50-317/98-03;50-318/98-03 I

Licensee:

Baltimore Gas and Electric Company Post Office Box 1475 Baltimore, Maryland 21203 Facility:

Calvert Cliffs Nuclear Power Plant, Units 1 and 2

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Location:

Lusby, Maryland Dates:

Onsite - January 19-27,1998 (in office review through February 17,1998)

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Inspector:

R. L. Nimitz, CHP, Senior Radiation Specialist Approved byi John R. White, Chief '

Radiation Safety Branch i

Division of Reactor Safety i

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9803260342 900320 PDR ADOCK 05000317 G

PDR

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EXECUTIVE SUMMARY

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Calvert Cliffs Nuclear Power Plant, Units 1 and 3 -

Inspection Report Nos. 50-317/98-03and 50-318/98-03 Plant Sunoort

. BG&E took action to review, evaluate, and upgrade the radiation protection o

program following the April 3,1997, Unit 2 spent fuel pool diving event and the series of high radiation area access control events that occurred in early/mid-1997.

- e-1There was generally a high level of management attention and oversight directed to resumption of diving activities and enhancement of the radiation protection program.

BG&E was developing and implementing its Radiation Protection improvement e

Plan.

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NRC review indicated generally good planning and preparation for the up coming

- o April 1998 Unit 1 outage. BG&E was reviewing planned work activities from a radiological risk perspective and generally good efforts were ongoing to identify a

risk significant areas / issues before the Unit 1 outage and modify program areas and procedures, as appropriate.

l e-Weaknesses in evaluation of radiological conditions in the reactor cavity during the previous Unit 2 outage and the subsequent planning and conduct of -

- radiological work in the area,.were identified. - Although no personnel exposures in excess of regulatory limits was apparent, the failure to adequately evaluate the existing radiological conditions was identified as a violation of 10 CFR 20.1501(a).

'In addition, an airborne radioac'ivity area generated by the event was not posted t

as required by 10 CFR 20.1902.

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Report Details

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R1 Radiological Protection and Chemistry (RP&C) Controls l

R1.1 Radioloaical Controls for Divina G

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tInsnection Seone (92904; 83750)

I The inspector selectively reviewed BG&E's planning and preparations for L

resumption of diving in radiological _ environments. BG&E had previously l

suspended radiological diving operations following the loss of control and the

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' subsequent unplanned exposure of a diver on April 3,- 1997. The event was the l-subject of NRC Enforcement Action (Reference NRC Inspection 50-317;

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50-318/97 02,EA 97-192). The inspector reviewed the diving activities relative

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to guidance provided in Technical Specifications,10 CFR 20, and applicable NRC

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~ information and guidance documents. The inspector also reviewed the planning and preparation relative to corrective actions outlined in BG&E's September 11, 199~7, letter to the NRC.

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IObservations and Findinas The inspector attended pre-job briefings for diving activities and observed -

resumption of diving activities on January 19,1998. BG&E implemented generally good efforts to revise and improve the radiological controls program for diving activities. The inspector noted significantly improved diver and support

' personnel training, communications, surveys and monitoring, exposure and access control,~ t 3d supervisor and management involvement. BG&E developed special.

dive proceans which detailed explicit responsibilities of personnelinvolved with diving activities. The inspector observed that pre-job briefing and post-dive

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critiques acted to improve over'all performance. BG&E's efforts to consolidate a'nd enhance procedures for diving in radiological environments provided for improved controls for diving activities.

Although radiological controls for diving were considered significantly improved, the following was noted prior to initial diving activities on' January 19,1998.

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Due to the size of the survey meter to be used by the diver, the inspector

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questioned th(capability of the diver to survey all areas to be potentially accessed during welding activities.

BG&E evaluated the adequacy of the capabilities'of the diver to survey

' areas to be welded and concluded that the meter was adequate to survey the locations of interest.

i BG&E noted that the diver could potentially enter an unsurveyed area near

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the upender, suspended dive preparations, and performed surveys in the

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BG&E ~ subsequently resumed diving in the Unit 1 spent fuel pool and, as of February 6,1998, had completed eight dives with no diver receiving any measurable exposure.

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Conclusion

BG&E implemented generally good efforts to revise and improve the radiological controls program for diving activities. Applied radiological controls for diving were significantly improved including supervisor and management oversight of this activity. Eight dives were successfully completed.

R1.2 Radioloaical Controls for Reactor Vessel Flanae Cleanina a.

Scope (83750)

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As part of post Unit 2 refueling activities conducted in May 1997, prior to replacing the reactor vessel head onto the reactor, BG&E initiated actions to clean the reactor vessel flange. The ALARA group performed a pre-work ALARA review which was used to develop the special work permit (SWP) for the task. The inspector discussed the May 5,1997, flange cleaning activity _ with cognizant personnel and reviewed the radiological controls provided for this work.

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' Observations and Findinas l

BG&E ALARA personnel indicated that for. initial planning purposes, the ALARA

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review for initial work in the cavity, including surveying of the reactor cavity in -

~ support of cavity decontamination and reactor flange cleaningi used radiological survey information from the previous (1995) Unit 2 outage for the same task. At that time, the documentation indicated loose, removable beta / gamma'

'I contamination of approximately.60 millirad /hr, and removable alpha contamination

of about 1200 disintegrations per 100 centimeters squared (dpm/100cm ).

Radiological controls, including respiratory protection equipment were specified for the May 5,1997, flange cleaning task based on these expected radiological

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I conditions. These contamination levels were expected to result in a beta /camma airborne contamination level of about 7 DAC8, and an alpha airborne contamination level on the order of 0.2 DAC.

8 The derived air concentration (DAC) means the concentration of a given radionuclide

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in air which, if breathed by the reference man for a working year of 2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> under

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conditions of light work (inhalation rate 1.2 cubic meters of air per hour), results in an intake of one All. An annual limit of intake (ALI) means the derived limit for the amount of radioactive material taken into the body of an adult worker by inhalation or ingestion in a

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year._ ALI is the smaller value of intake by reference man that would result in a committed

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effective dose equivalent of 5 rems or a committed dose equivalent of 50 rems to any j

individual organ or tissue.

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BG&E initially required the use of respiratory protection equipment (full-face negative pressure facepiece with HEPA filters) for personnel entering the cavity (after cavity drain down) to perform pre-work radiological surveys and perform work. The inspector noted, that results of the initial radiological survey, made-upon entry into the reactor cavity on May 5,1997, at about 10:40 a.m. indicated significantly different radiological conditions than expected. Specifically, removable beta / gamma contamination measuring up to 1200 millirem /hr was identified, as compared to the 60 millirad /hr expected. In addition, removable alpha contamination up to 150,000dpm/100cm alpha contamination was identified, as compared to approximately 1200 dpm/100cm" expected. However, no additional pre-work reviews were conducted to evaluate the adequacy of planned radiological controls, even though significantly increased radiological conditions were identified by the survey. No significant airborne radioactivity was identified based on a analysis of a general area air sample collected in the cavity.

The inspector noted decontamination personnel entered the reactor cavity about 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> later (May 5,1997 at about 6:15 p.m.) and that as of the time of entry, no additional evaluation or change in radiation protection requirements had been made to accommodate the known radiological conditions that were determined

- from the most recent survey. The inspector noted that the reactor cavity was drying out during the time peried which would.esult in potentially much higher airborne radioactivity levels during work activities in the reactor cavity.

The inspector noted that two air samples were collected to support the cleaning of the reactor flange cleaning. One was a fixed position air sampler hung from the 69' elevation of the reactor cavity to the flange area. This sampler remained in a fixed position and was not representative of the workers' breathing zones as the workers moved about the reactor vessel.- The workers, due to streaming coming from the reactor vessel, apparently crawled on their hands and knees in close proximity to high level surface contamination. The second air sampler was placed on the reactor refueling bridge.

Upon personnel exit from the reactor cavity, the air sample was field checked,

- determined to indicate elevated airborne radioactivity and sent to the chemistry laboratory for counting. The air sample (when analyzed at about 7:19 p.m. on May 5, 1997) indicated about 7 DAC (beta / gamma) as originally anticipated. Consequently no additional review was performed and no evaluation was performed for potential alpha airborne radioactivity. The air sample was not counted for alpha immediately to

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determine alpha airborne radioactivity.

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(Note: Unknown to BG&E the decontamination activities produced significantly i

elevated alpha airborne radioactivity concentrations. The cavity air sample (44 foot elevation), associated with decontamination of the reactor vessel flange, was not counted for alpha airborne radioactivity determination until about 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> later (2:36 a.m. on May 6,1997). This sample was determined to indicated about 103 DAC alpha.)

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As discussed above, concurrent with the air sample collected in the reactor cavity during flange cleaning, an air sample was collected on the 69 foot elevation of the refueling floor (refueling bridge). This sample, when analyzed at about 7:17 p.m. on May 5,1997, indicated about 0.8 DAC beta / gamma airborne radioactivity concentration levels. As a result of this sample, BG&E initiated additional sampling on the refueling floor area.

. The inspector noted that the air sample collected on the refueling floor (69 foot

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elevation) was not evaluated for alpha contamination until about 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> later (6:00 a.m. on May 6,1997). This sample, when analyzed indicated about 10.6 DAC sipha.

The inspector noted that affected individuals on the refueling floor did not wear respiratory protection equipment and were unaware of the elevated alpha wirborne radioactivity. Title.10 CFR 20 requires, in part, posting of an airborne radioactivity area when airborne radioactivity concentrations exceed.1 DAC. Due to delays in analyzing samples, the refueling floor was not posted as an airborne radioactivity area. However,

'the air sample results for the reactor cavity indicated greater than 1 DAC when analyzed and the reactor cavity was not posted as an airborne radioactivity area.

The inspector also noted that additional personnel entered the reactor cavity at about 1:25 a.m. on May 6,1997, to support placement of the reactor vessel head. Prior to the entry, no air samples were collected and evaluated for elevated alpha airborne radioactivity.

<The inspector noted that 10 CFR 20.1501(a) requires, in part, that each licensee make or cause to be made, surveys that may be necessary to comply with the regulations in this part, and are reasonable under the circumstances to evaluate the extent of i

radiation levels, concentrations or quantities of radioactive material, and the potential I

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radiological hazards that may be present. 10 CFR 20.1703(a)(3) requires in part that the licensee shall implement and maintain a respiratory protection program that includes air sampling sufficient to identify the potential hazard, permit proper equipment selection,' and estimate exposures; arid surveys and bioassays as appropriate to evaluate ' actual exposures. Further,10 CFR 20.1902 (d) requires the licensee to post airborne radioactivity areas as defined in 10 CFR 20.1003.

The inspector noted that BG&E failed to implement the provisions of 10 CFR -

' 20.1703(a)(3)during decontamination of the ' reactor vessel flange and setting of the reactor head on May 5,1997 as evidenced by the following.

i BG&E's air sampling was not sufficient to identify the potential hazard in that air

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sampling was not performed in the workers' breathing zone during.

decontamination of the reactor flange. A fixed position general area air sampler i

_ as used to monitor. workers cleaning the reactor flange.

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BG&E did not consider the actual radiological conditions (the high levels of beta,

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gamma, and alpha contamination present), when making respiratory equipment selection. Further, the reactor vessel flange decontamination was conducted h

approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after initial surveys of the reactor cavity which provided an opportunity for dry out of the high levels of beta, gamma, and alpha contamination.

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BG&E did not perform timely analysis (i.e., evaluations) of the air samples

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. collected, for purposes of exposure assessment during decontamination of the reactor vessel flange. The air sample from the reactor flange decontamination was not evaluated for alpha airborne radioactivity until about 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the work activity was completed. When analyzed, the sample indicated approximately 103 DAC alpha.

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The refueling floor air sample (69 foot elevation) was not analyzed for alpha airborne radioactivity for purposes of personnel exposure evaluation or posting until about 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the work activity was completed. When analyzed, the sample indicated approximately 10.6 DAC alpha. Personnel worked in the area without the use of respiratory protective equipment. -

Personnel re-entered the reactor cavity at about 1:25 a.m. on May 6,1997, to

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perform reactor vessel head replacement, and air samples were not collected prior to the entry to evaluate airborne radioactivity therein.

The above observations constitute a failure to perform necessary and reasonable surveys, as required by 10 CFR 20.1501, to ensure compliance with 10 CFR 20.1703(a)(3) and was a violation. (VIO 50-317/98-03 01)

The following additional observations were made.

Although anomalous elevated extensive contamination was identified during the

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initial reactor cavity surveys, the reactor cavity job coverage record for the May 5,1997, indicated that contamination control was satisfactory.

Despite elevated airborne radioactivity encountered during decontamination of

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the reactor flange on May 5,~.1997, the reactor cavity job coverage record for the flange decontamination indicated contamination control was satisfactory and that revision of the SWP was unnecessary.

There were no apparent programmatic controls in the radiation work permit

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program or ALARA program to provide for reevaluation of initially prescribed -

radiological controls when significant changes in radiological conditions were; detected. Further it appeared that both the ALARA group and the radiation protection field operations group specified radiological contiols requirements on the RWP. However, it was not apparent what process (e.g., consensus) was used to arrive at the final radiological controls to be specified on the RWP.

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Workers who entered the reactor cavity to perform cleaning of the reactor vessel

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flange on May 5,1997, were provided full face negative pressure respirators.

However, airborne radioactivity concentrations initially indicated about 100 DAC (unidentified alpha) after the work activity. The identified activity exceeded the protection factor (50) of the respiratory protective equipment used. BG&E later prescribe similar respirators for additional work in the cavit W

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BG&E was not able to provide indications that all appropriate individuals had l

' received internal dose evaluations or bioassays, as appropriate. BG&E had L

performed dose assessments for potential maximally exposed workers. BG&E l

initiated a comprehensive review of all personnel who may have been in the area l

to ascertain if any unplanned exposures occurred. The inspector's preliminary

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review indicated approximately 13 individuals were associated with or near the b

reactor flange cleaning task. However, licensee dose assessments did not

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l indicate any one individual sustained any significant unplanned exposure.

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The inspector noted that the licensee's radiological controls program did not

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appear to effectively consider suggested guidance presented in NRC Information Notice No. 92-75, Unplanned intakes of Airborne Radioactive Material By individuals At Nuclear Power Plants, dated November 12,1992. The

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information notice discussed several airborne radioactivity events, including

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inspection and housekeeping activities in the reactor cavity and fuel transfer canal, and highlighted the need for vigilance when conducting activities that could significantly increase airborne radioactivity.

The inspector also noted that 10 CFR 20.1902(d) requires that the licensee post airborne radioactivity areas as defined in 10 CFR 20.1003.10 CFR 20.1003 defines an airborne radioactivity area, in parti as an area in which airborne radioactivity exists in concentrations in excess of the derived airborne concentrations (DACs) specified in

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10 CFR 20, Appendix B.10 CFR 20, Appendir. B, also requires, in footnote 4, that for known mixtures of radionuclides, the limiting values of airborne radioactivity

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concentrations in air is determined by the sum of the ratios of each radionuclide's concentration in air and its derived air concentration specified in Appendix B, and may not exceed unity. The inspector noted that the DAC values are intended to control chronic occupational exposures.

' The inspector noted that BG&E did not post the 44 foot elevation of the reactor cavity as an airborne radioactivity area when it was determined at about 7:15 p.m.~ on May 5,.

1997, that airborne concentrations of Co-60 were about 5 times its DAC specified in f

Appendix B. Further, the sum of the ratios of each radionuclide's concentration (for

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known beta / gamma emitters) in air and its derived air concentrations, specified in Appendix B, exceeded 1 (i.e., the sum of the ratios was approximately 7.3). This is a

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violation. (VIO 50-317/98-03-02)

l BG&E took the following actions:

Immediate actions when the elevated airborne radioactivity was identified on May 6, 1997:

Upon identification of the elevated airborne radioactivity, a second ALARA

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evaluation was performed. BG&E revised the ALARA review on May 6,1997,

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for SWP 97-2319 to reflect the increased airborne radioactivity levels and to j

' provide administrative controls for air sampling and personnel exposure to alpha

. airborne radioactivity.

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(Note: BG&E's revised ALARA Review for the activity, dated May 6,1997, indicated the contamination was not expected due to previous surveys performed on April 15,1997, after reactor cavity drain down from core off-load.)

Workers performing the cleaning task and support personnel in the vicinity were

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whole body counted.

BG&E selectively collected in-vitro bioassay samples (fecal samples) following

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the event.

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Two air samples and one smear sample of the area were sent to an offsite

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laboratory for isotopic analysis.

BG&E recalculated the highest airborne radioactivity assuming the most

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. restrictive radionuclide. This resulted in an apparent increase of the highest j

measured airborne radioactivity concentrations from 100 DAC to 240 DAC

(unidentified alpha). BG&E subsequently re-evaluated the alpha transmission factor for air sampling media which reduced the highest measured airbome

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radioactivity from 240 DAC to 168 DAC, unidentified alpha, in the reactor l

cavity.

(Note: After the event, BG&E submitted air samples and smear samples to an offsite vendor for isotopic analysis. BG&E subsequently received the laboratory isotopic analysis results for the air samples. The results indicated the majority.

of the alpha radioactivity was attributable to Curium 242. Based on this information, BG&E recalculated maximum airborne radioactivity and concluded

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the maximum measured alpha airborne radioactivity concentration was 24 DAC,

alpha. Notwithstanding, BG&E was unaware of this at the time of the airborne

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event.)

BG&E supp. mented the ALARA review for SWP No. 97-2319 on May 7,1997

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.i to provide for decontamination for the deep ends of the refue!!ng pools.

BG&E issued an issue Report (IR1-067-171)on May 8,1997,to document the j

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unplanned intake of airborne radioactivity by an individual on the 69 foot i

elevation. The issue report was closed in September 1997.

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(Note:- The author of the issue report recommended that a root cause analysis l'

be performed to identify the root causes of the event and in particular review job

planning. However, this recommendation was not implemented. Rather, the recommendation was down graded to a determination of cause and generic

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. implications. The issue report was closed on or about September 4,1997, with limited discussion of causes and corrective actions. In light of this observation, BG&E issued a second issue report on January 26,1998, to include a root cause analysis.)

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BG&E completed a Post Work ALARA Review for SWP 97-2319 on

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September 9,1997.

BG&E used the air sample isotopic analysis results to calculate personnel exposures iattributable to the airborne radioactive materialintakes. BG&E collected fecal samples to support assessment of internal dosos due to intake of transuranics. The maximum

' identified personnel exposure, attributable to intake of airborne radioactivity associated

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with the reactor cavity work, was approximately 1 millirem committed effective dose

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equivalent (CEDE) and 6 millirem committed dose equivalent (CDE) (lung) to the individual working on the 69 foot elevation without a respirator. These doses are a small fraction of annual NRC dose limits (i.e, 5 rem total effective dose equivalent (TEDE) and 50 rem total organ dose equivalent (TODE) as specified in 10 CFR 20.1201.

(Note: TEDE is the sum of the deep dose equivalent (DDE) attributable to external

4 exposure and the CEDE. TODE is the sum of DDE and CDE.)

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The inspector noted that the likelihood of a significant exposure was limited in that:

personnel were in the area a short period of time and had participated in mock-up training for the task, personnel in the reactor cavity wore respiratory protective equipment; and the largest fraction of alpha airborne radioactivity was attributable to relatively short-lived Curium 242.

The inspector noted that at the time, no apparent effort was initiated, to perform a-comprehensive evaluation of the event and develop comprehensive corrective actions

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to address' apparent deficiencies involving the adequacy of applied radiological controls when the high levels of contamination were identified, the performance of breathing zone air sampling for personnel performing the' task, and the lack of timely air' sample analysis for alpha airborne radioactivity.

The inspector did note that BG&E was extensively involved in development and implementation of its broad based Radiation Protection improvement Plan (RPIP) which was reviewing both industry events and events at Calvert Cliffs for areas for improvement. At the. time of the inspection, the licensee was continuing to review the events. The inspector noted that BG&E developed and implemented a broad based Radiation Protection improvement Plan as a result of a Unit 2 diving event which occurred one month preceding the May 5,1997, flange event.

BG&E subsequently initiated development of a Transuranic Activity Assessment and

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Programmatic Action Plan to address apparent programmatic weaknesses identified by the inspector. This action plan is expected to be completed on or by March 15,1998, c.

Conclusions BG&E did not establish radiological controls for work in the reactor cavity on May 5, 1997, commensurate with the _ actual conditions indicated by area surveys. As a result, the adequacy of work planning and process, personnel respiratory protective

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requirements, and monitoring of personnel for potential exposure to airborne radioactivity were not evaluated in view of radiological conditions that were

significantly different than originally anticipated. In addition, monitoring of airborne radiological conditions was ineffective in that air sampling was not representative of the workers' breathing zone; and analysis of air samples was not timely.-

At the time of the event, BG&E did not implement a comprehensive effort to assess

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root causes and determine effective corrective actions for the event. However, a broad ~

based program evaluation was under way as a result of a previous event (See Section R3 of this report). A violation, associated with implementation of 10 CFR 20.1501 and 10 CFR 20.1703, was identified. A violation involving failure to post an airborne

radioactivity as required by 10 CFR 20.1902 was also identified.

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R1.3 Unit 1 Refuelino Outaae Radioloaical Controls (Plannina and Preoaration)

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Inspection Scope (83750)

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The inspector selectively reviewed the planning and preparation for the Unit 1 refueling outage. The inspector reviewed records, discussed outage planning with BG&E rep-resentatives, and observed activities to verify necessary planning and preparations and management support for radiation protection planning.

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Observations and Findinas BG&E performed an evaluation of its previous outage and published a 1997 Outage Report for the Unit 2 refueling outage. BG&E met its refueling outage goals. An

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estimated 35 person-rem was saved by use of early boration and peroxide addition to

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enhance crud removal. The report summarized radiation safety events and -

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recommendations for ALARA program enhancement during further outages. The report recommendations were under review for application to the upco ning April 1997 Unit 1 outage. Inspector discussions with BG&E ALARA personnel indicated improved outage planr;ing techniques / computer programs were being used for work planning and control.

j Radiological controls personnel were aware of planned work activities, including -

i possible emergent work for the Unit 1 outage.

I BG&E was reviewing and planning for addition of peroxide to the reactor coolant system to enhance system crud removal to reduce personnel radiation exposure to as

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low as is reasonably achievable. Further, BG&E planned to install additional high radiation area controls and real-time radiation monitors to monitor and control access to

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piping systems expected to exhibit elevated radiation levels during the peroxide addition.

BG&E had not yet issued its Ur it 1 1998 pre-outage report which would discuss

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ALARA actions / initiatives for the outage. However, BG&E had developed 1998 site l

person-rem goals including outage exposure goals and provided a draft pre-outage report for information. The pre-outage report contained recommendations for

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ALARA/ radiological controls enhancements for planned work. At the time of the inspection, the draft pre-outage report and the Unit 21997 outage report contained limited information on the extensive contamination encountered during final drain down j

of the Unit 2 reactor cavity.

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Conclusions y

BG&E was implementing overall good ALARA planning for the Unit 1 ' refueling outage.

R3 RP&C Procedures and Documentation '

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Scope.192904)

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LThe inspector reviewed efforts to develop and implement a Radiation Protection i

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Improvement Plan (RPIP) and other rediation protection program enhancements. The inspector reviewed documentation and met with BG&E's recently appointed RPIP manager.-

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Observations and Findinas M

BG8 E formally issued its RPIP on December 19,1997. An RPIP manager (current Supervisor-Maintenance Performance Assessment) was named as the plan manager.

The manager, with his staff, has begun consolidation of all radiation protection!

i concerns and findings into a document for ease of tracking and review. BG&E was-

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reviewing previous audits and assessments to ensure allitems have been included in

the plan. - BG&E also initiated review of previous issue reports dealing with radiation

safety matters. The plan manager provides a weekly update of RPIP status to station

'l management.

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j BG&E had established teams to 1) review and upgrade management oversight and

'l communication, 2) integrate radiation protection into site processert, 3) improve site j

radiation protection knowledge, and 4) improve station radiation protection

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assessments.- In addition, a separate team was constituted to review and improve station radiation protection procedures. Each of the five teams had clearly identified.

program improvement deliverables to be done by the outage. BG&E had prioritized'

radiation safety procedures to be reviewed by the outage.

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BG&E had also initiated a procedure upgrade program using senior personnel to review and suggest revisions to procedures for work. A matrix was develop to classify work'

by risk levels. BG&E anticipated identifying all high risk outage work, by the start of -

i the outage, and reviewing and revising applicable procedures, as appropriate. BG&E

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was planning to provide personnel with specific training on expectations before the

outage. The inspector noted draft procedures for various reactor cavity work were under development. The procedures under development exhibited extensive radiation

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protection caution statements and holdpoints.

BG&E also revised its. site ALARA Committee to designate it as the Radiation Protection

. Oversight Committee (RPOC). In addition to reviewing ALARA program performance, the RPOC will also review radiation protection program performance.

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Conclusion J

BG&E took broad based action to review and address radiation protection program

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E weaknesses. BG&E had established and was implementing a radiation protection l-improvement plan. Corrective actions were underway to address identified program

weaknesses.

R8 Miscellaneous RP&C issues R8.1 UFSAR Conformanca A recent discovery of a licensee operating their facility in a manner contrary to the UFSAR description highlighted the need for a special focused review that compares

l plant practices, procedures and/or parameters to the UFSAR description. While -

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performing the inspections discussed in this report, the inspectors reviewed the applicable portions of the UFSAR that related to the areas inspected. No concerns were noted relative to UFSAR conformance.

I R8.2 Qualifications and Responsibilities of the Radiation orotection Manaaer

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BG&E selected a new Radiation Protection Manager in December 1997. The individual was subsequently determined to have insufficient experience to meet Regulatory Guide 1.8 qualification requirements as RPM as required by Technical Specifications. BG&E i

subsequently named an interim RPM but permitted the incumbent to fill the General Supervisor Radiation Safety position. The inspector noted the interim RPM to be participating in the review and exercise of control over practices and decisions potentially impacting worker and public health and safety.

l R8.2 (Closed) (Violation 01022)(eel 50-317&318/97-02-15)

Failure to orovide instructions for a diver in a radioloaically' control area.

l This matter is discussed in Section R1.1 of this report. The licensee implemented the corrective and preventative actions outlined in its September 11,1997, letter to the NRC.

R8.3' (Uodate)(Violation 010102)(eel 50-317&318/97-02-16)

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Failure to control access to a verv hiah radiation arga i

This matter is discussed in Section R1.1 of this report. The licensee was implementing

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the corrective action outlined in its September 11,1997, letter to the NRC for this matter. The inspector did not complete the review of this matter and was continuing to seview access controls to very high radiation areas.

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(Closed) (Violation 01032)(eel 50-317&318/97-02-17)

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j-Failure to complete radioloaical survevs orlor to personnel entry.

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This matter is discussed in Section R1.1 of this report. The licensee implemented the corrective action outlined in its September 11,1997, letter to the NRC for this matter.

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~ R8.5 Sliaht Contamination in North Outfall'

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NRC Inspection Report 50-317; 318/97-07 noted that BG&E ' analyzed soil at the north outfall and detected trace levels of Cobalt 60 in soil. The outfall is a point of release fer the turbine building sump. The sump is a recognized effluent release point and its

effluent is sampled and analyzed for offsite dose analysis purposes. BG&E performed a

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RESAD calculation for the trace contamination and did not identify any dose -

l consequences associated with the trace contamination. BG&E was continuing to review the matter.

R8.8 Housekeepina l

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The inspector noted that overall housekeeping in the refuel pool areas was good.

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V. Management Meetinas X1 Ealt Meeting Summary The inspector meet with licensee representatives on January 27,1998. The inspector summarized the purpose scope and findings of the inspection. The licensee acknowledged the findings presented.

X3i Management Meeting Summary On February 20,- 1997, Mr. John White, Chief, Radiation Safety Branch, Division of Reactor Safety, NRC Region I, attended BG&E's onsite monthly Radiation Safety Oversight Committee -

(RPOC) meeting and its Radiation Protection improvement Plan (RPIP) Team meeting.

,l The purpose of NRC attendance of the meeting was to gain better understanding of the

- licensee's' efforts relative to improvements in management oversight and involvement,

' integration of radiation protection into site processes, zelf-assessment, job planning, and

. corrective action development and implementation. The meeting included discussion of

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- BG&E's plans and process to prevent potential unplanned exposures. The status and results achieved from radiation protection improvement initiatives were reviewed and assessed by thi

licensee.s BG&E's efforts appeared comprehensive, directed toward weaknesses, and directed L toward determination _of appropriate corrective actions and improvement processes to address program deficiencies. -

Mr. White discussed the apparent violations identified, the significance of the issues, and the

= need for lasting and effective corrective actions.

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PARTIAL LIST OF PERSONS CONTACTED l

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Licensee IR. C. Gradle, Compliance Engineer

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W. Holston, General Supervisor, Mechanical Maintenance

P P. Jones,'ALARA Coordinator L

P. Katz, Plant General Manager l~

K. Nietman', Superintendent Nuclear Operations l

S. Sanders; General Supervisor, Radiation Safety

. L. Smialek, Senior Plant Health Physicist (Interim Radiation Protection Manager)

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- W. Spina, Superintendent Nuclear Maintenance (Acting)

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W. Paulhardt, Radiation Safety Supervisor-Dosimetry M. Rigsby, Supervisor-Radiation Technical Services -

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' R. Wyvill, Radiation Safety Supervisor.

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EBC '

F. Bower,' Resident inspector.

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i INSPECTION PROCEDURES USED lu l;

.lP 83750:

Occupational Exposure

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' IP 92904:

Follow-up - Plant Support L

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ITEMS OPENED, CLOSED, AND DISCUSSED p.

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Opened

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50-317&318/98-03-01 VIO. Failure to Perform Surveys as required by 10 CFR

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20.1501 to ensure compliance with 10 CFR

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20.1703(a)(3)

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50-317&318/98-03-02 VIO Failure to Post an Airborne Radioactivity Area as

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l required by 10 CFR 20.1902(d)

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ry Closed 50-317&318/97-02-15'

eel Failure to provide instructions for a diver in a i

radiologically control area.

.50-317&318/97-02-17 eel Failure to complete radiological surveys prior to I

personnel entry.

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50-317&318/97-02-16 eel Failure to control access to a very high radiation

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area.

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LIST OF ACRONYMS USED ALARA-

. As Low As Reasonably Achievable CDE Committed Dose Equivalent. D CEDE.

Committed Effective Dose Equivalent-

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'DAC

. Derived Air Concentration s

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Escalated Enforcement item '

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. HP-Health Physics l.a,

RCS-Reactor Coolant System

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RPIP..

Radiation Protection improvement Plan RPM

. Radiation Protection Manager RWP Radiation Work Permit

. SWP '.

Special Work Permit TEDE Total Effective Dose Equivalent

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1TLD Thermoluminescent Dosimeter l.;

TODE.

Total Organ Dose Equivalent -

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UFSAR Updated Safety Analysis Report URI Unresolved item l

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