IR 05000213/1999002

Insp Rept 50-213/99-02 on 990420-0719.No Violations Noted. Major Areas Inspected:Decommissioning Operations & Sf Safety Maint & Surveillance,Plant Support & Rediological Controls

ML20211E814
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	08/20/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20211E810	List: IR 05000213/1999002 ML20211E805
References
50-213-99-02, NUDOCS 9908300128
Download: ML20211E814 (19)

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i U.S. NUCLEAR REGULATORY COMMISSION

REGION I

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l Docket No.: 50-213 License No.: DPR-61

Report No.: 50-213/99-02 Licensee: Connecticut Yankee Atomic Power Company P. O. Box 270 Hartford, CT 06141-0270 Facility: Haddam Neck Station Location: Haddam, Connecticut Dates: April 20,1999 to July 19,1999 Inspectors: Lonny Eckert , Radiation Specialist Tom Frederichs, Project Manager, NRR Marie Miller, Senior Health Physicist Joseph Nick, Decommissioning Health Physicist John Wray, Decommissioning Health Physicist Approved by: Ronald Bellamy, Chief, Decommissioning and Laboratory Branch Division of Nuclear Materials Safety i

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EXECUTIVE SUMMARY Haddam Neck Station i

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NRC Inspection Report No. 50-213/99-02 This routine inspection included aspects of licensee activities in preparation for dismantlement and decommissioning of the facility. The report covers a three-month period ofinspection by regional and headquarters NRC personnel, and includes reviews and assessments of spent fuel safety, engineering and plant support activities, and management effectivenes Decommissioning Operations and Spent Fuel Safety:

The licensee was generally conducting activities in accordance with license requirements and commitments. The inspectors observed good performance by the facility staff to control and conduct facility decommissionin The licensee maintained an effective corrective action program and performed very good audits and assessments of the decommissioning programs to help self-identify and correct issues and

problems. No significant safety concems were identifie The safety evaluations performed by the HN staff for the control room modification to remove the

"J" bottles and Updated Final Safety Analysis Report (UFSAR) update used incorrect information supplied by the Northeast Utilities Radiation Assessment Branch (NURAB) staf However, the changes had no effect on the dose consequences to a member of the public as a result of a resin accident, in addition, the changes did not result in a significant increase in occupational dose to personnel in the control room during and following a resin acciden The licensee discovered and corrected several errors in its July 1997 calculation of the dose to control room personnel due to a resin accident. The recalculation revised the model used to calculate the doses and included information on actual source term activity that was unavailable at the time the erroneous calculation was performe The documentation presented by the licensee for Design Change Request (DCR) CY-98001, dated April 24,1998, demonstrated a lack of control of design change documentation. The licensee adequately identified unresolved safety questions and the need to revise the TS in its design proces Maintenance and Surveillance:

Required surveillances important to safe decommissioning and storage of spent reactor fuel are being scheduled / planned. For those systems designated operable or available, surveillances are being conducted in a timely and effective manner. The Maintenance Rule Program is thorough with enough detail to ensure effective compliance with 10 CFR 50.65. The Transition Plan to the Decommissioning Operations Contractor (DOC) appears to be very good in that all structures, systems and components (SSCs) have been addressed in the tumover plan. Spent Fuel Pool (SFP) Building Ventilation System modifications appear to be progressing satisfactorily. No concems were identifie il

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i Plant Support and Radiological Controls:

The licensee established representative sampling of effluents from the Bus-10 containment HEPA filter. The licensee provided very good controls for radioactive materials and contamination, surveys and monitoring during decommiss~mning work activitie The licensee established, implemented, and maintained an effective quality assurance program for the radioactive effluent control program with respect to self-assessments, chemistry laboratory quality control, and adverse condition resolutio The licensee completed their corrective action plan to address Spent Fuel Building (SFB)

ventilation system issues. The inspector identdied no further issues pertaining to the SFB ventilation system.

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TABLE OF CONTENTS EXECUTIVE S UM MARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1i TABLE OF CONTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv REPORT DETAI LS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 i l

Summary of Facility Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1. Decommissionina Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 i 01 Cond uct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.1 Self-assessment. Auditina. and Corrective Action . . . . . . . . . . . . . . . . . . . . . . 1 01.2 Decommissionina Performance and Status Review . . . . . . . . . . . . . . . . . . . . . 2 08 Miscella neous issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 08.1 Control Room Habitability Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 08.2 Review of 10 CFR 50.59 Evaluations Performed in 1998. . . . . . . . . . . . . . . . . 7 M1 Maintenance and Surveillance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 11. Plant SuDDort and Radioloalcal Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 R1 Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . . . . . . . . 9

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R1.1 Radioloaical and Effluent Controls for the Bus-10 Removal Proiect. . . . . . . . . 9 R1.2 R adioloalcal Su rveys . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 R7 Quality Assurance (QA) in RP&C Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 R8 Miscellaneous RP&C lssues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 R8.1 Soent Fuel Buildina Ventilation System (URI 97-09-01) . . . . . . . . . . . . . . . . . 11 111. M a naaement Meetinas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

X1 Exit M ee tin g S u m ma ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 PARTIAL LIST OF PERSONS CONTACTED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 INSPECTION PROCEDURES USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 1 i

i ITEMS OPEN, CLOSED, AND DISCUSS ED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 LIST OF AC RO!!YfdS U S ED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 IV

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REPORT DETAILS Summarv of Facility Activities The plant was maintained in a permanently shutdown condition during this inspection perio The licensee continued priority activities to prepare the plant and supporting functions for tumover to a decommissioning operations contractor (DOC). These activities included spent fuel cooling system modifications, processing of resins from the reactor coolant system (RCS)

decontamination, radiological remediation of the Bus-10 area, and site radiological characterizatio I. Decommissioninn Operations 01 Conduct of Operations 01.1 Self-assessment. Auditina. and Corrective Action Inspection Scope (40801)

A review was performed to evaluate the effectiveness of licensee controls in identifying, resolving, and preventing issues that degrade safety or the quality of decommissioning. The inspector evaluated the licensee's self-assessment, auditing, corrective actions, and root cause evaluations through a review of licensee documents and interviews with licensee personne Findinas and Observations The inspector reviewed the recent audits and surveillances performed by the Oversight (Quality Assurance) Group. The audits and surveillances were performed by qualified individuals who were independent from the organization performing the work. The inspector found that there was a very good level of detail and appropriate critical review of the areas reviewed by the licensee's Oversight Group. For most concems or findings, effective corrective actions were immediately implemented and longer term corrective actions were timely and appropriat Very good administrative controls were used to track, trend, and implement corrective action The licensee continued to use a Condition Report (CR) system to identify concems and ensure timely, appropriate corrective actions. The inspector noted that the CR system was effectively used by all levels within the organization. CRs were discussed at Management Review Team (MRT) meetings, which ensured that each problem or concem was discussed by an interdisciplinary team with representatives from all major site departments. The inspector attended several MRT meetings and noted that there was a very good level of peer review and discussion for each issue. The inspector also reviewed various CR's to determine how the issues were resolved. Root cause evaluations were performed for higher safety significant issues and repetitious events or occurrences were handled appropriatel Other programs and meetings were used for self-assessment by the licensee's staff. The Nuclear Corporate Assessment Team (NCAT) provides management oversight and reports directly to the Northeast Utilities' Board of Directors. The Joint Oversite Committee (JOC)

performs assessments at both Haddam Neck and Maine Yankee and reports to the Connecticut

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Yankee Atomic Power Company (CYCAPCo) and Maine Yankee Atomic Power Company (MYAPCo) Boards of Directors. An annual site visit is performed by the Independent Management Assessment Committee (IMAC) to provide oversight of decommissioning management. The Nuclear Safety Assessment Board (NSAB) reviews specific areas of decommissioning and makes recommendations to CYCAPCo management. The Employee Concerns Program provides a confidential program for employees to raise concems to site managemen Conclusions The licensee maintained an effective corrective action program and performed very good audits and assessments of the decommissioning programs to help self-identify and correct issues and

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problems. No significant safety concoms were identifie .2 Decommissioning Performance and Status Review Inspection Scoos (71801)

The inspector evaluated the licensee's status of decommissioning and verified that the licensee and its contracted workforce are conducting decommissioning activities in accordance with licensed requirements through a review of licensee documents and interviews with licensee personnel. The inspector toured the facility to evaluate the material integrity of structures, systems, and components (SSCs) necessary for the safe decommissioning of the facilit Findinos and Observations Status of decommissioning during the inspection period was obtained through weekly phone conferences with the licensee's staff and contractors, and through attendance at onsite planning meetings. The general activity of decommissioning was low since the licensee's DOC was not fully functional on the site. However, modifications to the spent fuel storage facility continued that will allow independent operation of the spent fuel pool (SFP) from other plant systems (spent fuel pool island concept). Work also continued on relocation of the control room to an area adjacent to the industrial security area access point. Other activities involved preparations for turnover of systems and components to the DOC for decontamination and dismantlemen The inspector toured the control room and observed a shift tumover meeting in the control roo The inspector observed that the systems necessary for safe decommissioning were functional and operational. Plant operator logs were maintained with appropriate information. Shift tumover meetings were held to communicate conditions and status to shift personnel. Control room staffing met the licensee's commitment. Professionalism was maintained in control room communications and action The inspector also toured the spent fuel pool building (SFPB) and other nearby areas to assess work conditions and decommissioning activities. Decontamination and remediation of the Bus 10 concrete pad next to the SFP was completed End work was continuing on the new pad and electrical distribution bus. Modifications were continuing on the SFPB ventilation syste Workers were following approved work instructions. Material condition of SSCs was adequate

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to maintain safe storage of spent fuel. The licensee was controlling the SFPB as a vital area; however, the inspector questioned the process for controlling access to the building. The licensee explained the controls to the inspector and the inspector determined that the controls met the minimum standards for protection of spent fuel. Housekeeping and fire protection in the SFPB were adequate. No other concems were note Conclusions The licensee was generally conducting activities in accordance with license requirements and commitments. The inspectors observed good performance by the facility staff to control and conduct facility decommissionin Miscellaneous issues 08.1 Control Room Habitability Evaluation Insoection Scope (37700)

On March 2,1999, in accordance with the requirements of 10 CFR 50.72(b)(1)(ii)(B), the licensee nobfied the NRC of a potenbal error in the calculation of the dose to control room personnel during a postulated resin fire accident as described in the Updated Final Safety Analysis Report (UFSAR). On March 18,1999, the licensee retracted the notification based on a new calculation using the actual curie content of resins stored on site rather than the postulated curie content assumed in the UFSAR accident analysi The licensee's UFSAR analyses of the dose to control room personnel, its root cause investigation of the calculation error, and its 1999 recalculation of the dose were inspecte Observations and Findinos Control Room Dose Before Permanent Shutdown Section 6.4 of the UFSAR presents the design and analysis of the habitability systems installed at HN to maintain it in a safe mode under normal conditions and during and following a postulated design basis accident. The last UFSAR revision of control room habitability before HN permanently shutdown was dated March 199 The design bases for the habitability systems were:

1) The control room is inhabited at all times. Food and potable water are provided in sufficient quantities to sustain five people for five days. Sanitary facilities and medical supplies are provide ) General Design Criteria 19 for providing adequate radiation protection under accident condition ) Self-contained breathing apparatus sufficient for five operators for five hour !

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4) Regulatory Guide 1.95 for protecting control room operators from a postulated chlorine releas ) Regulatory Guide 1.78 for assumptions for evaluating the habitability of the control room following postulated chemical releas The analysis of the control room shielding showed that the 30-day whole body dose for the control room was 0.624 rom, based on the loss of coolant accident (LOCA) source term conforming to NUREG-0588 and Technical Information Document, TID-14844. The whole body dose was calculated from direct shine from the containment and immersion in the noble gas plume from a postulated LOCA. Whole body doses from airbome activity within the control room were not included, and no doses other than to the whole body were include Control Room Dose After Permanent Shutdown Section 6.4 was updated in January 1998 to reflect changes in the facility due to permanent shutdown and defuelin The design bases were revised to remove the reliance on self. contained breathing apparatus (basis 3, above) and other sources of supplied air for the cordrol room (a bank of "J" bottles filled with oxygen). In addition, the design basis accidents changed. The LOCA was no longer possible. In its place, the shielding analysis considered a fuel handling accident and a resin fir The revised shielding evalumbon expanded the number of exposure pathways analyzed. The results presented in the revised UFSAR were:

CONTROL ROOM DOSES (REM)

IyJ2g FuelHandling Accident Resin Container Accident Whole Body 2.36 E-4 3.73 E-5 Thyroid 7.23 E-1 1.41 E-2 Skin 7.02 E-1 1.17 E-3 Max Organ 1.06 E-2 3.83 Licensee Root Cause investigation ,

The erroneous calculation of control room dose in the permanently defueled condition was issued July 28,1997 by Northeast Utilities Radiation Assessment Branch (NURAB) and transmitted to Connecticut Yankee. The results of the calculation were used by the HN staff in safety evaluations of a modification that removed the "J" bottles supplying oxygen to the original control room and the January 1998 update of control room habitability information in the UFSAR. The safety evaluations concluded that no unresolved safety question existed because the whole body dose to control room personnel from accidents in the permanently shutdown condition was less than the whole body dose in the operating conditio Subsequently, the licensee began evaluations of a modification to relocate the control room from the turbine building to a location near the plant access point. The new location is not equipped with the specialized ventilation and shielding found in operating plant control rooms. The

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i licensee's review of the design bases for the control room revealed the errors in the July 1997

, calculation of control room dose.

I The licensee initiated its root cause investigation into the calculation discrepancy on March 4,1999, and completed it on May 9,1999. The investigation team contacted CY and Northeast Utilities personnel involved with the erroneous calculation, including personnel who are no longer employed by the licensee or its parent compan The licensee concluded that the primary root cause was the use of an incorrect dose conversion factor (DCF) by NURAB personnel performing the calculation. The calculation depends on inputting DCFs into the computer code developed for control room habitability determinations (HABIT). NURAB used the HABIT computer code and a method of calculation provided in NUREG/CR-6210. However, the person performing the calculation overlooked the guidance on selecting DCFs for alpha-emitting nuclides, which is the appropriate case at HN due to a history of failed fue An important contributing root cause was the failure to apply QA controls to the calculation, contrary to the direction of the licensee's procedure, NUC DCM, Chapter 5, ' Calculations."

As a result of the error, the maximum organ dose, which was dose to the bone surface, was underestimated by about a factor of ten. However, this dose was less than the equivalent dose to an organ (50 rem) allowed by General Design Criterion (GDC) 19 of Appendix A to 10 CFR Part 50, and fell within the design bases for the control roo The licensee identified the following corrective actions in response to the erro ) Perform a control room habitability calculation (for the existing control room) using CY DCM methodology and modify the UFSAR as necessar ) Ensure that future control room dose calculations are performed using CY DCM methodolog ) Ensure adequate management controls are in place which specify the scope, intemal CY review requirements and quality assurance criteria for engineering work performed by outside organization ) Communicate the results of the root cause analysis to NURAB for inclusion in their corrective action progra An intemal memo (CY-KJH-99-015) dated June 8,1999, assigned responsibility for the first three corrective actions to groups in the CY organization. The CR identifying the errors was forwarded to NURAB on March 3,199 Comse-ieon of Doses for Operatina and Decommissionino Conditions A conference call with the licensee and NRC staff was held on April 5,1999, to discuss the appropriate method for comparing the doses to control room personnel in the operating and decommissioning conditions in view of the fact that the exposure pathways of the two UFSAR evaluations differed considerably. The licensee agreed to calculate and compare the total effective dose equivalent (TEDE) of the two conditions to determine if the changes resulted in an increase in dose consequence l l

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Recalculation of Control Room Dose The licensee recalculated the dose to control room personnel to correct the errors discovered in the July 1997 analysis. The licensee also applied an updated model for atmospheric dispersion of gaseous releases and revised the source term to reflect the actual amount of radioactivity in ion exchange resin stored on sit The licensee found and corrected the following errors in the July 1997 calculation:

1) The bone DCF was corrected to bone surface from red bone marrow, conforming to the guidance of NUREG/CR-6210 2) Co-60 DCF was corrected to class Y from class W, conforming to Appendix B of 10 CFR 20 3) Activities for Pu-239/240 and Cm-243/244 were corrected to the combined activity for each pair from individual activity for each nuclide, conforming to the radiochemical protocol for reporting the concentrations of these nuclides 4) The control room ventilation flow rate was corrected to 9500 CFM from 6150 CFM, conforming to the maximum flow rate of the fa I The licensee recalculated the control room dose using the methodology of NUREG/CR-6331, Atmospheric Relative Concentrations in Building Wakes, which uses the ARCON96 computer code. The licensee selected the ARCON96 code on the basis that it better fits the case of atmospheric dispersion from a point source than the HABIT code used previousl The licensee revised the source term to reflect the actual amount of activity in a resin liner. The estimated activity assumed for the accident analysis presented in the January 1998 UFSAR update was one resin container with 448 curies. Based on radiochemical analysis of the contents, the actual maximum amount of activity in a resin container was 110 curie l The licensee recalculated the TEDE dose to control room personnel in the turbine hall location due to a resin accident as 0.093 rem. This was less than the TEDE of 0.624 rem calculated for a LOCA . The TEDE for the proposed control room location near the plant access point was calculated as 0.068 re The licensee also reviewed its dose estimates at the site boundary due to a resin accident. The site boundary estimates had been performed separately. The licensee concluded that no errors had been made in the previous estimates. Based on the actual source term in a resin container, the licensee determined that the actual dose at the site boundary would be less than previously estimate The control room habitability information submitted for the January 1998 UFSAR update to the dose to control room personnel was incorrect. The reported maximum organ dose was calculated with an incorrect DCF, and consequently underestimated the dose given the assumptions presented in the accident analysis. This is a potential violation of 10 CFR 50.9,

' Completeness and accuracy of information." However, the NRC has not yet completed the inspection and review of this potential violation. At the time of the inspection, the licensee had i not assigned responsibility for each of the corrective actions recommended by its root cause !

investigation team. The licensee's completion of corrective actions will be the subject of a future j inspection. URI 99-02-01 i

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, . Conclusions The safety evaluations performed by the HN staff for the control room modification to remove the

"J" bottles and UFSAR update used incorrect information supplied by the NURAB staf However, the changes had no effect on the dose consequences to a member of the public as a result of a resin accident. In addition, the changes did not result in a significant increase in occupational dose to personnel in the control room during and following a resin acciden The licensee discovered and corrected several errors in its July 1997 calculation of the dose to control room personnel due to a resin accident. The recalculation revised the model used to calculate the doses and included information on actual source term activity that was unavailable at the time the erroneous calculation was performe .2 Review of 10 CFR 50.59 Evaluations Performed in 1998 Inspection Scope The following list of safety evaluations were examined:

SY-EV-974045 SY-EV-97-0138 SY-EV-98 0018, Rev 0 & Re SY-EV-97-0094 SY-EV-97 0145 SY-EV-98-0023 SY-EV-97-0095 DCR CY-98001 SY-EV-98-0031 SY-EV-97 0108 SY-EV-98-002 SY-EV-98-0033 SY-EV-97-0137 SY-EV-98-005 SY-EV-98-0072 Observations and Findinos The safety evaluations were adequate in their assessment of unresolved safety questions. The threshold for determining the existence of an unreviewed safety question (USQ) was appropriat '

However, the documentation presented by the licensee for Design Change Request (DCR)

CY-98001, approved by the CY engineering group on April 24,1998, was an uncontrolled copy of the 10 CFR 50.59 applicability review. It had not been approved by the Plant Operations Review Committee (PORC). The change was for implementation of the defueled Security Plan !

at CY, which included several modifications to the physical security system. The April 24,1998, copy erroneously concluded that a change to the operating license was required for the

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l The licensee presented the controlled copy of the safety evaluation for DCR CY-98001 dated May 7,1999, and approved by PORC. The approved copy correctly concluded that no operating license change was required to implement the defueled Security Pla !

The licensee reported that the uncontrolled copy had been obtained from the documents kept in its Administrative offices. These are uncontrolled copies. As corrective action, the licensee labeled the uncontrolled design change documents as * UNCONTROLLED" and gave directions to obtain controlled copies from the Nuclear Documents Services grou .

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, i Conclusions The documentation presented by the licensee for DCR CY-98001, dated April 24,1998, demonstrated a lack of control of design change documentation. The licensee adequately identified unresolved safety questions and the need to revise the TS in its design proces M1 Maintenance and Surveillance Inspection Scope (62801)

A review was performed of the licensee's program to ensure that maintenance and surveillance of SSCs important to the safe storage of spent fuel and proper operation of radiation monitoring and effluent control equipment are being effectively conducted. The inspector examined the licensee's planned surveillances, work on the SFPB Ventilation System, and the licensee's Maintenance Rule Program for 10 CFR 50.6 Findinos and Observations The inspector reviewed the following four surveillance packages for effectiveness and detail:

Sealed Sources Test for Leakage Contamination Quarterly Meteorological Tower Preventive Maintenance Activities Emergency Diesel Generator Preventive Maintenance Tests Polar Crane inspections and Tests in all cases, work orders were examined and found to be timely written and thorough in scop The inspector observed field work performed while completing the Quarterly Meteorological Tower surveillances. The inspector noted that the licensee has a system for identifying those systems as operable, available, in lay-up, or abandoned. Surveillances are still planned and conducted on systems designated operable or available. Surveillances for systems in lay-up have been rescheduled for a later time. Surveillances for abandoned systems have been cancele The surveillance and tests of the polar crane were examined in detail since this equipment will play a major role in the removal of the heavy loads from containment. The inspector reviewed available documentation including a letter dated March 29,1999, from the Project Manager entitled ' Polar Crane inspections, Refurbishment and Load Test Closure Summary". The inspector noted that the licensee's inspection revealed that refurbishment of the polar crane was needed prior to heavy load lifting in containment. The refurbishment was completed in December 1998 and the crane successfully load tested in March 1999. The test included a dynamic load test at 100% capacity and a static load test at 125% capacity. The licensee concluded that the containment polar crane is capable of performing to its normal rated capacity of 175 tone and occasionalloading of 227.5 ton The inspector reviewed available documents regarding the licensee's Maintenance Rule Program including the Connecticut Yankee Decommissioning / Nuclear Island Maintenance Rule

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Manual, the 1998 Fourth Quarter Maintenance Rule Summary Report, and meeting notes from l the Connecticut Yankee Expert Panel where decisions are made as to the applicability of the ,

Maintenance Rule for certain plant systems. The inspector verified that the licensee's program j met the requirements of 10 CFR 50.65 and that application of the Maintenance Rule is thorough and effective to ensure safe storage of spent fuel and proper operation of necessary equipmen The inspector reviewed Minor Modification Package No. CY-98535 for the first phase of replacing the ventilation and radiation monitoring equipment in the Spent Fuel Building. This will allow decommissioning of other structures and systems without affecting the operation of the Spent Fuel Building. The inspector also observed work in progress and discussed the project in detail with cognizant engineers. The tie in of the new ventilation system to the existing system is covered under DCR No.98-402 (not completed at the time of this review). The inspector noted adequate precautions were being applied on the SFP operating floor while construction work was being conducted to prevent inadvertent dropping of articles into the pool containing spent fuel. The path for isokinetic sampling lines from the stack to the new monitor was also walked down. Adequate room appeared to be designed into the system to meet the requirements for c,0&elling line losses. The system will be inspected in more detail following completion of all installation ectivitie Conclusions Required surveillances important to safe decommissioning and storage of spent reactor fuel are being scheduled / planned. For those systems designated operable or available, surveillances are being conducted in a timely and effective manner. The Maintenance Rule Program is thorough with enough detail to ensure effective compliance with 10 CFR 50.65. The Transition Plan to the DOC appears to be very good in that all SSCs have been addressed in the tumover plan. SFP Building Ventilation System modifications appear to be progressing satisfactorily. No conoems were identifie . Plant Supcott and Radioloalcal Controls R1 Radiological Protection and Chemistry (RP&C) Controls R Radioloaical and Effluent Controls for the Bus-10 Removal Prolect < Inspection Scope (84750)

The inspection consisted of a review of licensee technical support documents to support isokinetic sampling of Bus-10 high efficiency particulate air (HEPA) filter effluents. The licensee had completed concrete scabbling work, removed the lead sheeting and had started removing soils from the Bus-10 area at the time of the inspectio Findinos and Observations As mentioned in NRC inspection Report 50-213/99-01, HEPA effluent air filter sample results taken from the Bus-10 containment were less than the lower limit of detection. The licensee established isokinetic sampling in accordance with NRC Regulatory Guide 1.2 _

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. Conclusions The licensee established representative sampling of effluents from the Bus-10 containment HEPA filte R1.2 Radioloalcal Surveya Inspection Scope (83750)

The inspectors reviewed the controls for radioactive materials and contamination, surveys and monitoring through observation of work activities, tours of the facility, interviews with personnel and a review oflicensee document Findinas and Observations The inspectors verified that there was an adequate supply of radiation survey and monitoring equipment available to assess radiological conditions in work areas. All equipment checked by the inspector was operable and within the current calibration period. Current radiological surveys of various work locations were reviewed by the inspector. The surveys contained detailed information regarding current radiological dose rates and hazards in the work area Surveys were posted at the main control point for the RCA. Appropriate licensee management personnel had reviewed the radiological survey The inspectors toured various areas of the facility to determine the adequacy of contamination controls. Portal monitors and frisking instruments were located throughout the facility for use by workers as they left radioactive materials areas or contaminated areas. Appropriate instructions were given to workers to ensure that materials taken from an RCA were surveyed for potential radiological contamination. The inspectors noted that the licensee had erected a fence around an area on the Southwest Storage Site Area (also known as the peninsula). This area had been used to store potentially contaminated items in the past. The fence was erected to maintain controls for items that had not yet been surveyed for release for unrestricted use. Signs were posted inside the fenced area to remind workers that permission was required to remove items form the area. The exterior of the fence was posted with no trespassing signs and wamings since the area was accessible from the public areas adjoining the site. The inspectors determined that the controls were adequate to prevent unintentional removal of potentially contaminated items from the are Radiological housekeeping was generally good with appropriate controls established to minimize the spread of contamination. Areas that presented a challenge to the licensee's staff due to changing conditions and ongoing work were kept in good condition. Posting of radioactive material areas and labeling of radioactive materials was very goo * Conclusions I

The licensee provided very good controls for radioactive materials and contamination, surveys and monitoring during decommissioning work activities. No violations or significant safety concems were identifie !

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, s R7 Quality Assurance (QA)in RP&C Activities Insoection Scooe (83750 and 84750)

The inspection consisted of: (1) a review of inter-laboratory measurement comparisons; (2) a review of chemistry laboratory quality control program radioactive liquid and gaseous effluent samples; (3) a review of the 1998 Health Physics Assessment; and (4) a review of responses to adverse condition reports (ACRs) 98-0788,98-0801,98-0819, and 98-084 Observations and Findinas All interlaboratory quality assurance measurement comparisons were within the licensee's acceptance criteria. Anomalous trends identified during reviews of quality control charts for gamma and tritium measurements were investigated and resolve I The 1938 Health Physics Assessment provided an integrated assessment of the Health Physics l program. This document largely took credit for the surveillances and departmental self-assessments conducted in 1998. No significant issues were identifie Licensee responses to the above-noted ACRs were reasonable and timely. These particular issues were identified and corrective actions were initiated soon enough to preclude them frorn having regulatory significanc Conclusions The licensee established, implemented, and maintained an effective quality assurance program for,the radioactive effluent control program with respect to self-assessments, chemistry laboratory quality control, and adverse condition resolutio R8 Miscellaneous RP&C issues R Soent Fuel Buildina Ventilation System (URI 97-09-01) Inspection Scooe (84750)

NRC Inspection Report 50-213/97-09 identified problems relating to the spent fuel building (SFB) ventilation system resulting in a violation pertaining to design control. in addition, an unresolved item was opened to further evaluate the issues pertaining to the SFB ventilation syste The inspection consisted of: (1) a tour and discussion of the SFB ventilation system modifications with the system engineer; (2) a review of flow rate meter electronic alignment results; (3) a review of DCRs CY-97026, "SFB Supply Fan Speed Reduction" and CY-98034,

" Ventilation Flow instrument Replacement"; and (4) a review of pertinent operating procedures to determine whether limitations identified within the DCRs were administratively controlled through incorporation into the operating procedures. The following procedures were reviewe i

. . NOP 2.15-1, "PAB Ventilation System Operation," Revision 17, 4/13/99 NOP 2.15-2, * Reactor Containment Atmospheric Control System,' Revision 19, 4/13/99 NOP 2.15-3, " Spent Fuel Building Ventilation System Operation," Revision 17, 4/13/99 Findinas and Observations Visual observation indicated that the licensee installed the new flow rate indicators and air flow testing ports in accordance with industry standards. Electronic alignment results for the flow rate meters were found to be within the licensee's acceptance criteria. No issues were noted pertaining to licensee acceptance criteria. Procedures were modified to properly address limitations identified by the DCR The licensee reduced the SFB ventilation system supply fan speed. SFB ventilation system air flow testing results indicated that the licensee's modification successfully balanced system air flow rates to be in conformance with the airflow rates described in the UFSA Conclusions The licensee completed their corrective action plan to address SFB ventilation system issue The inspector identified no further issues pertaining to the SFB ventilation system; therefore, URI 97-09-01 is closed. As mentioned previously, enforcement action pertaining to the SFB ventilation system design issues was already issued as VIO 97-09-02. As such, no additional enforcement actions will be taken pertaining to these specific issue . Management Meetings X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management periodically during the inspection, and during a teleconference with Mr. R. Mellor and others at the conclusion of the inspection on July 19,1999. The licensee acknowledged the findings presented by the inspector. The inspector reviewed with the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie s

s s PARTRAL LIST OF PERSONS CONTACTED

• G. Bouchard, Unit Director J. Bourassa, Oversight Manager

• M. Cavanaugh, Communications Manager P. Dadlani, Project QA Manager, Bechtel N. Fetherston, Decommissioning Operations Manager K. Harner, Chemistry Manager

• J. Haseltine, Strategic Planning Director P. Hollenbeck, Site Characterization Supervisor M. Homyak, Supervisor- Corrective Actions D. Hoffeman, Maintenance Manager

• K. Heider, Decommissioning Director K. Jackson, Assistant Project Manager, Bechtel D. Karr, Oversight A. Kelly, Project Manager, Bechtel S. Kumar, Regulatory Affairs

• R. Mellor, Vice President Operations and Decommissioning

• R. Mitchell, Operations and Maintenance Manager ]

G. van Noordonnen, Regulatory Affairs Manager D. Scribner, Project Engineer, Bechtel j

'R. Sexton, Radiation Protection Manager

• J. Tarzia, Radiation Protection Manager, Bechtel T. Troutman, Transition Manger, Bechtel j S. Webster, Licensing, Bechtel i

• Denotes attendance at the telephone exit meeting held on June ig,199 INSPECTION PROCEDURES USED IP 40801: Self-Assessment, Auditing, and Corrective Action at Permanently Shutdown Reactors IP 62801: Maintenance and Surveillance at Permanently Shutdown Reactors IP 71801: Decommissioning Performance and Status Review IP 83750: Occupation Radiation Exposure Controls IP 84750: Radioactive Waste Treatment, and Effluent and Environmental Monitoring

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. b ITEMS OPEN, CLOSED, AND DISCUSSED Open 99-02-01 URI Incorrect Control Room Habitability information Submitted for UFSAR Closed 97-09-01 URI Spent Fuel Building Ventilation System Discussed 97-09-02 VIO Inadequate SFB Ventilation Design Controls s

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..- s LIST OF ACRONYMS USED -

ACR Adverse Condition Report CAL Confirmatory Action Letter CFH- Certified Fuel Handler CFR- Code of Federal Regulations CR Condition Report CYAPCo Connecticut Yankee Atomic Power Company DCF- Dose Conversion Factor DCR Design Change Request DOC Decommissioning Operations Contractor GDC General Design Criterion HEPA High Efficiency Particulate HN Haddam Neck HP Health Physics IFl Inspection Followup item '

IMAC Independent Management Assessment Committee IR inspection Report JOC Joint Oversite Committee LOCA Loss of Coolant Accident mrom milliram MRT Management Review Team MYAPCo Maine Yankee Atomic Power Company NCAT Nuclear Corporate Assessment Team NOV Notice of Violation NRC Nuclear Regulatory Commission NSAB Nuclear Safety Assessment Board NU Northeast Utilities NURAB Noitheast Utilities Radiation Assessment Branch PDR Public Document Room PORC' Plant Operation Review Committee QA Quality Assurance 1 RCS Reactor Coolant System RP Radiation Protection RP&C Radiological Protection and Chemistry RWPs Radiation Work Permhs RWST Reactor Water Storage Tank SAT Systems Approach to Training SFB Spent Fuel Building SFP Spent Fuel Pool SFPB Spent Fuel Pool Building SSCs Structures, Systems, and Components TEDE Total Effective Dose Equivalent TlO Technical Information Document TS Technical Specifications UFSAR Updated Final Safety Analysis Report URI Unresolved item USQ Unreviewed Safety Question

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