IR 05000219/1989004

Insp Rept 50-219/89-04 on 890115-0225.Violations Noted.Major Areas Inspected:Activities in Progress,Including Air Accumulator Testing,Environ Qualification,Reactor Pressure Vessel Testing,Operations & Physical Security

ML20246H002
Person / Time
Site:	Oyster Creek
Issue date:	04/14/1989
From:	Cowgill C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20246G980	List: IR 05000219/1989004 ML20246G973 ML20246G987
References
50-219-89-04, 50-219-89-4, NUDOCS 8905150326
Download: ML20246H002 (21)

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U.S. NUCLEAR REGULATORY COMMISSION

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REGION I

Report N ~50-219/89-04 Docket N ' License N OPR-16 Priority -- Category C Licen'see: GPU Nuclear Corporation 1 Upper Pond Road Parsippany, New Jersey 07054 Facility Name: Oyster Creek Nuclear Generating Station Inspection Conducted: January 15 - February 25, 1989 Participating Inspectors: W. Baunack E. Collins K. Kolaczyk T. Koshy D. Lew J. Wechselberger'

Approved By: Edif /V!f7

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C. CowgilY, Chig, Reactor Projects Section 1A Date Inspection Summary: Inspection January 15 - February 25, 1989 (Report N /89-04)

Areas Inspected: Routine inspections by resident and region-based inspectors (272 hours0.00315 days <br />0.0756 hours <br />4.497354e-4 weeks <br />1.03496e-4 months <br />) were conducted on activities in progress, including: air accumulator testing, environmental qualification, reactor pressure vessel testing, operations, radiation protection, and physical security. . Fourteen previous inspection findings, an un-resolved item regarding feedwater flow nozzles, and outstanding questions concern-ing the recent diesel generator cable failure were reviewed. Inspectors witnessed surveillance testing and followed troubleshooting on the core spray booster pumps and a reactor building closed cooling water system pum Results: The licensee is pursuing testing of its air operated valve accumulators and underground electrical cables. These efforts have identified some deficiencies-which are being corrected. Periodic testing would improve the licensee capability to' ensure quality of these components (Paragraphs 1.0 and 6.0). A weakness in licensee control of work was identified in that some Environmental Qualification requirements were not specified. The licensee is reviewing and implementing more thorough requirements. This item is considered unresolved (Paragraph 2.0). One

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violation was identified in that a plant modification was installed without imple-mentation of licensee modification controls (Paragraph 4.0). Fifteen previous inspection findings were closed, the unresolved item regarding feedwater flow was

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updated, and outstanding questions concerning diesel generator cable testing were resolved (Paragraphs 5.0, 6.0, and 7.0).

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TABLE OF CONTENTS PAGE 1.0 Air Accumulator Testing (71707, 93702)............................... I Introduction.................................................... I 1.2 Licensee Activities..... ....................................... I 1.3 Air Accumulator Inspection Results.............................. I 1.4 Summary......................................................... 2 2.0 Environmental Juilification (71707, 93702)........................... 2 2.1 Introdut*.2v,..................................... ............ 2 2.2 Generation Mai ntenance System (GMS-2) . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.3 S umm a ry . . . . . . . . . . . . . . . . . . . ..................................... 3 3.0 Monthly Maintenance Observation (62703).......................... ... 4 3.1 Reactor Pressure Vessel Hydrostatic Test................... .... 4 3.2 Alternate Rod Injection System.................................. 6 4.0 Unresolved Items 86-02-03 and 88-16-02 (92701, 71707)................ 6 5.0 Feedwater Flow Nozzles - (0 pen) Unresolved Item 50-219/88-33-04 (71707)........... ........ ....................................... 7 6.0 Die sel Generator Cabl e Fail ure ( 71707) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 7.0 Licensee Action on Previous Inspection Findings (92701, 71707). . . . . . . 8 8.0 Plant Operational Review (71707, 93702).............................. 12 8.1 Core Spray Booster Pump Trips................................... 12 8.2 Reactor Building Closed Cooling Water (RBCCW) Pump Trip. . . . . . . . . 12 8.3 Tours of Control Room........................................... 13 8.4 Tours of Facility............................................... 13 9.0 Radiation Protection (71707)......................................... 14 10.0 Observation of Physical Security (71707)............................. 14 11.0 Monthly Surveillance Observation (61726)............................. 15 12.0 Enforcement Conference (30703)....................................... 14 12.1 Meeting Attendees............................................... 14 12.2 Meeting Summary............ ..... .............................. 15 i

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PAGE 13.0 Management Meeting (30703)........................................... 16 14.0BackshiftInspections(71707)........................................ 16 15.0 Review of Periodic and Special Reports (71707) . . . . . . . . . . . . . . . . . . . . . . . 17 16.0 Unresolved Items......................-............................... 17 17.0 Exit Interview (30703)...............................................~17 ATTACHMENTS Attachment I: Management Meeting Attendees, February 2, 1989 Attachment II: GPUN Oyster Creek Nuclear Restart Meeting, February 2, 1989

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DETAILS 1.0 Air Accumulator Testing 1.1 Introduction Inspection Report 50-219/88-38 described the licensee's activities in the testing of air accumulators in response to Oyster Creek Emergency Operating Procedure (EOP) Inspection (IR 50-219/88-200) and Generic Letter 88-14. The initial scope of this testing encompassed the air accumulators and piping associated with eighteen valves. These valves were selected because they either affected primary containment isolation or Standby Gas Treatment System (SGTS) operatio As described in Inspection Report 50-219/88-38, the licensee has tested sixteen of the eighteen air accumulators in the original

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scope. The as-found results of this testing showed that eight of the accumulators failed the acceptance criteria for operability. As a result of the high failure rate of these accumulators, the inspector raised the question to the licensee for-the need to expand their original inspection scope and left this issue as an unresolved item (50-219/88-38-03).

1.2 Licensee Activities De ing the present inspection period, the licensee has effected re-pairs to the eight accumulators identified in the previous inspection period. The leakage on the SGTS valves was reduced to nearly zero and procedures for providing temporary supply air to SGTS valves were

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developed. The accumulators and associated piping of the Main Steam Isolation Valves (MSIV) are now seismically qualified and the piping replaced to reduce leakage. A meeting between the NRC and the licen-see took place at NRC Headquarters on February 13, 1989, to discuss the integrated leakage rate of the MSIVs. The licensee committed to documenting their position for NRC review in the futur The licensee has expanded their inspection scope to include all air accumulators. This added scope encompasses the 22 air accumulators which operate the secondary containment ventilation dampers. Addi-tionally, the licensee and a contractor have performed a seismic de-sign review of all air accumulators, j 1.3 Air Accumulator Inspection Results l

The licensee completed the initial testing of twelve secondary con- ,

tainment isolation valve accumulators. The results of this air test-  !

ing showed that all twelve valves failed their acceptance criteri The secondary containment isolation valves fail "as is" upon loss of instrument air (instrument air is a non-safety related system) and failure of the accumulators. It was noted in the licensee's safety

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y evaluation that secondary containment negative _ pressure can not be maintained when any set of secondary containment isolation valves are not closed.

I The check valves on all twelve accumulators were replaced. Addition-ally, many air operators were replaced and/or repaired. It'was also noted that water was found in two accumulators (2 oz and 8 oz, re-spectively), and a pin hole was found in another accumulato EQE Engineering was contracte'd by the licensee to verify the seismic adequacy of,the accumulators. The preliminary results of their walk-down indicated that there were ten accumulators which required work to ensure seismic adequacy. The licensee has not yet received nor reviewed EQE Engineering's final results. The licensee intends to review EQE Engineering's findings and to upgrade the accumulators at the earliest outage of opportunity during cycle 1 .4 Summary The results of the air accumulator testi n have indicated an unac-ceptably high failure rat The 50% failure rate of the initial

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_ scope and the 100% failure rate of the added scope indicates a need for a maintenance and surveillance program for these accumulator The. licensee does intend to test and correct all the deficiencies of all the accumulators prior to startup. Additionally, the licensee plans to assess the need to establish a maintenance and/or surveil-lance program after receiving all the results from their accumulator testing. The inspectors will continue to follow licensee activities in this area. -This is'an update to unresolved item 50-219/88-38-0 .0 Environmental Qualification 2.1 Introduction On January 13, 1989, the licensee identified that motor control cen-ter (MCC) breakers were installed without implementing environmental qualification (EQ) requirements. A review conducted by the licensee ]

indicated that weaknesses in the Generation Maintenance System '

(GMS-2) to properly classify the breakers in seven MCC's (DC-1, DC-2, IAB2, 1A21A, 1A21B, 1821A AND 1B218) resulted in the potential to I affect their environmental ~ qualification. Additionally, the licen-see's review indicated that six breakers and two fuses in the MCCs did not have adequate environmental qualification documentatio These breakers and fuses were subsequently replace , 2.2 Generation Maintenance System (GMS-2)

GMS-2 is a computerized maintenance data system utilized by Mainten-ance, Construction and Facilities (MCF) to generate preventative and corrective maintenance work packages. This system contains a sepa-rate data base for components and preventive maintenance (PM) task l

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ing information from the EQ Master List. Input to the PM tasks' data is made by Plant Materiel utilizing the information in the component data in GMS- '

- The GMS-2 component' data base does not correlate'to the EQ Master List component 'for component. This data base assigns tag. numbers for MCC's'as well as to individual breakers within the MCC's. The EQ

Master List, on the other hand, lists only the MCC's. The breakers within'these EQ MCC's are understood to be also EQ, Additionally,-

not'all the information fields within'the GMS-2 data base for each .

breaker is complete. The inform'ation on some are only partially filled in and verified. As a result _of this situation, several poten-H tial causes for failing te classify these breakers as EQ existe These potential causes include:

(a) Personnel verifying the'EQ classification failed to make the correlation between the breaker component level in GMS-2 and the MCC component level in the EQ Master List. Personnel did not recognize that the breaker was located inside the.EQ MCC and/or that the' breakers would also be E (b) Personnel did not recognize that GMS-2 component data was not-verified nor complet Failure to recognize this fact may poten-tially lead personnel to conclude erroneously'that the component is not E (c) . Engineering component data base may have potentially been er-roneous because the engineers inputting data to the component data base for the breakers failed to correlate the breakers to the MCC's listed in the EQ Master Lis The licensee has taken several corrective measures. The engineering component data base is being updated to ensure component data is com-plete and verified. PM tasks have been reviewed to ensure t.he cor-rect EQ classification. A review of corrective maintenance to iden-tify the impact on EQ equipment was performed, and all components adversely affected were replaced. The EQ Master List has been re- ,

viewed to ensure that there are no other components similar to the MCCs which can potentially result in erroneous EQ classification All classifications in the GMS-2 will require verification by the job planner until resolution of GMS-2 weaknesses is complete Although PM tasks were reviewed to ensure correct EQ classification, inspector review identified that the licensee did not consider the impact of improperly classified PM tasks on equipment qualificatio The licensee initiated this evaluation upon inspector identificatio _ - - _ _ _ - _ _ _ _ _ - _ - _ _ _ - - _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _

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2.3 Summary The licensee's review of this event identified over 10 PM tasks and 50 breakers which were not properly classified or controlled as E The licensee was able to show through existing documentation that all breakers were qualifiable with the exception of six. The breakers were subsequently replaced. Work on four of the six breakers had taken place during the current outage. The other two breakers (as-sociated with valves V-37-11 and V-5-166) were replaced in the 11R outage, thus, they were installed throughout operating cycle 11. The

"A" recirculation pump bypass valve, V-37-11, is required to be EQ because it is required not to be inadvertently actuated or reposi-tioned during an accident scenario. The reactor building closed cooling water (RBCCW) system isolation valve, V-5-166, is required to be EQ since it is tied on to the same bus as Core Spray System I fill pump (the core spray fill pump has been determined by the licensee to be required for accident mitigation).

The weaknesses in GMS-2 had the potential to degrade safety related components. These two breakers have been sent to the licensee's laboratory in Reading, PA to determine if they can be environmentally qualifie Pending the results of the environmental qualification of the two breakers and the review of the licensee's completed correc-tive actions, this item will be unresolved (50-219/89-04-02).

3.0 Monthly Maintenance Observation 3.1 Reactor Pressure Vessel Hydrostatic Test The licensee performed a hydrostatic test of the reactor pressure vessel in accordance with ASME Code Section XI, 1980 on February 15-16, 1989. The inspector witnessed portions of this test and veri-fied the administration and adequacy of the procedur During the performance of the hydrostatic test inspections, two ,

through wall leaks were identifie One leak was located at a one inch capped vent pipe which is welded to the "C" recirculation pump discharge valve body. The licensee observed a six to eight drops per minute leakrate f rom the interface between the weld and the discharge valve body. The other leak was located 1/32 of an inch asove the weld on the two to six inch reducer on the Reactor Head Cooling (RHC)

System. The licensee observed a leakrate of 20 to 30 drops per minut The leak on the recirculation discharge valve was repaired by remov-ing the entire weld and rewelding the vent pipe to the valve. The l RHC weld, including one half inch of piping on either side, was cut out to facilitate failure analysis by General Electric. The reducer

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was then repaired by weld build up of both halves of the reducer an then welding these halves together. Retest of both repairs in ac-

cordance with ASME Code Section'XI, 1980 was performed satisfactoril Analysis of the RHC reducer weld by General Electric indicated that

.the cause of failure was Intergranular Stress Corrosion Cracking (IGSCC). As a result the licensee performed in'spections on a 50%

sample of the family of welds which include the RHC reducer weld. No additional' indications of IGSCC were identified by the -license Regional and resident inspectors following the licensee's corrective-actions had no further concern '

While. reviewing the weld repairs, the inspector noted that'the RHC reducer weld was not identified by the license _e's-ISI drawings. When the inspector questioned the licensee on this deficiency, it was evi -

dent that the licensee was already aware of and in the process of correcting this deficiency. The licensee indicated that.there were typically a couple of welds identified every outage which were not reflected on the IS1 drawings. These welds were discovered after removai of lagging for ISI. ISI drawings were developed by General Physics in 1978 and 1979. The failure by General. Physics.to identify some welds may have been inability to walkdown all systems as a re-sult of lagging. It was noted, however, that the-RHC weld which was not reflected in the ISI drawing was reflected on a General Electric drawing. The licensee intends to compare GE drawings 'to ISI drawings to ensure that all welds on'GE drawings are included on ISI drawing The inspector has no further concern While reviewing the hydrostatic procedure to ensure conformance with TS requirements, the inspector raised a question on core spray oper-ability. TS 3.4.A.10 requires that the core spray system be operable when the reactor coolant system is maintained at less than 212 de-grees F or vented. Exception to this requirement is allowed during Reactor Vessel Pressure Testing. TS 1.39 defines Reactor Pressure Vessel Testing as " system pressure testing required by ASME Code Sec-tion XI, Article IWA-5000, including system leakage and hydrostatic tests, witn the reactor vessel completely wator solid, core not cri-tical and section 3.2.A satisfied." As part of the licensee's pro-cedure, scram time testing is directed to be performed with the reac-tor at normal system operating pressure. The inspector concluded, however, that although scram time testing is a part of the hydro-static test procedure, it is not part of Reactor Pressure Vessel Testing and therefore not an exception to TS 3.4.A.10. During scram time testing, the core spray system was in reduced availability vice operable. The licensee stated that they interpreted the meaning of Reactor Pressure Vessel Testing to include scram time testing since it was a part of their procedure and that it was the intent of the TS. The licensee stated that the wording is ambiguous and that they

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will consider a change to the Technical Specifications to clarify that scram time testing is also an exception. The inspector had no further questions regarding this matte .2 Alternate Rod Injection System (ARIS)

The inspector observed the post modification testing of the Alternate Rod Injection System (ARIS) on January 19, 1989. Items inspected in-cluded a review of the procedural adequacy to test required system functions, a review of the acceptance criteria test data to meet de-sign requirements and the verification that the required administra-tive approvals were obtained. No concerns were note .0 (Closed) Unresolved Items 86-02-03 and 88-16-02 To prevent entry into IRM Range 10 prior to establishing the entry con-ditions required by procedures and possibly violating a safety limit, a mechanical modification was to be installed.

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During this inspection, the installation of the modification was verifie Observation of the switches in the control room and discussions with operators indicate that the modification has been completed and that it appears to be effective. This installation of the range switch modifi- I cation closes out unresolved item 86-02-0 Further followup concerning the installation of the modification revealed !

that it was installed through the use of a Maintenance, Construction and Facilities (MC&F) Short Form. For this task Short Form 49203, "IRM (Range 10) Range Switch Mod, Install Per MPR Direction" was used to implement the ;

modification on January 14, 1988. Discussions with facility personnel indicate the modification should, at that time, have been installed in accordance with Station Procedure 124, " Plant Modification Control." This procedure is provided for the control and acceptance of modifications and inclodes a detailed check list to assure proper control and acceptanc In this instance, the short form implemented a modification without proper engineering approval and without the proper administrative control Also, QA identified certain documentation deficiencies which were still under review during the time of the inspection. It should be noted that this was a relatively simple modification which did not require a great deal of controls. Nevertheless, installation procedures, drawings and spare parts which were provided by MPR Associates, Inc., the modification designer and fabricator, were involved and should have bee :,ntrolle '

Following the identification of this deficiency, the lice. ,2 red functional and demonstrated parameters within Technical Specification Limiting Condi-tions for Operation.

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Plant housekeeping and cleanliness were in accordance with approved licensee program No inspector concerns were identifie ,

9.0 Radiation Protection During entry to and exit from the RCA, the inspectors verified that proper warning signs were posted, personnel entering were wearing proper dosime-u try, personnel and materials leaving were properly monitored for radio-active contamination, and monitoring instruments were functional and in calibration. Posted extended Radiation Work permits (RWPs) and survey status boards were reviewed to verify that they were current and accurat The inspector observed activities in the RCA to verify that personnel complied with the requirements of applicable RWPs and that workers were aware of the radiological conditions in the area. No unacceptable con-ditions were identifie .0 Observation of Physical Security During daily tours,-the inspectors verified that access controls were in

. accordance with the Security Plan, security posts were properly manned, protected area' gates were locked or guarded and that isolation zones were free of obstructions. The inspectors examined vital area access points to verify that they were properly locked or guarded and that access control was in accordance with the security plan. No unacceptable conditions were identifie .0 Monthly Surveillance Observation The inspector observed the performance of the Core Spray Valve Operability and Inservice Test, Procedure 610.4.003, on January 21, 1989. Review of this surveillance included verifying that the required administrative approvals were received, that no Technical Specification Limiting Condi-tions for Operation was violated, and that the system restoration was complete During this test, the core spray test valve failed to ope Upon inves-tigation by a Startup and Test engineer, the on/off switch in the speed controller for the valve was discovered in the off position. The instal-lation of the speed controller was a modification which was completed during this outage to prevent water hammer during valve operation. At the time of this test, a partial turnover of the modification had been per-formed. The procedures had been revised to reflect the new modification, however, the control of the switch position for the speed controller was not included in either the surveillance procedure nor the system operating procedure electrical switch lineu _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ - _ _ _ - -

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Upon further followup of this event, it was noted that this discrepancy I was not brought to the attention of the Plant Engineer for the modifica- 4 tion nor the Operations Engineer. The inspector pointed out the discre-pancy to the engineers. The licensee intends to revise the procedures to include adequate controls of the switch position. The inspector had no  ;

further concern j 12.0 Enforcement Conference On September 6, 1988, an enforcement conference was held at the NRC Region I' office in King of Prussia. Pennsylvani Topics discussed at this meet-ing were the events described in NRC Region I Inspection Report N /87-2 .1 Meeting Attendees

.GPU Nuclear Corporation J. Barton, Deputy Director Oyster Creek P. Clark, President GPUNC

. P. Czaya, BWR Licensing Engineer E. Fitzpatrick, Vice President and Director Oyster Creek-R. Long, Vice President and Director, Division of Planning and Nuclear Safety S. Polon, Manager Public Information, Oyster Creek J. Sullivan, Director Licensing and Regulatory Affairs NRC l W. Baunack, Project Engineer, Reactor Projects Section 1A L. Bettenhausen, Chief Projects Branch 1 C. Cowgill III, Chief, Reactor Projects Section IA A. Dromerick, Licensing Project Manager J. Gutierrez, Regional Counsel D. Holody Jr., Enforcement Officer W. Kane, Director, Division of Reactor Projects J. Wechselberger, Senior Resident Inspector Oyster Creek State of New Jersey R. Ebright, NJ DEP/BNE - Special Projects M. Jacobs, NJ DEP/BNE - Nuclear Engineer 12.2 Meeting Summary The causes for the safety limit violation were described as were the valve maintenance activity which initiated the leak necessitating the removal of the reactor building closed cooling water system from ser-vice. The causes for the leak as well as the corrective actions which were taken were discussed.

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Also discussed were the actions by the shift personnel following the-occurrence of the safety'11mit violation. These actions included

, untimely notification of the event to upper management and the de-struction of certain records associated with the violation. The lic-ensee's immediate and long term corrective actions were identifie The low safety significance of the event was noted. The recircula-tion loop requirement was identified as more appropriately being a technical specification limiting condition for operations as opposed to a safety limit violatio Certain mitigating circumstances associated with the maintenance activity and subsequent safety limit violation were presented by the license The meeting concluded with summary discussions by both the licensee and by Region I management. Region I management stated that the lic-ensee would be informed of any enforcement action.

l Due to an administrative oversight the conduction of this enforcement conference was not documented at an earlier dat .0 Management Meeting A management meeting was held at the NRC Region I office on February 2, 1989, to discuss issues related to the restart of the plant from the 12R Outage. Outstanding items were identified for resolution. These included for the licensee: 1) implementation of procedure changes to address the Isolation Condenser steaming phenomena; 2) implementation of procedure changes to provide more restrictive thermal limit requirements under cer-tain conditions of Core Spray System operability; 3) complete installation'

of the cathodic protection system; 4) review and approve work packages associated with Bulletins 79-02 and 79-14; 5) submit a letter stating their commitments to perform testing on Main Steam Isolation Valve stems and 6) submit an updated drywell corrosion report. Prior to startup, the NRC will review the procedure changes of 1) and 2) above, provide concur-rence on licensee core spray sparger inspection and provide approval of licensee weld overlay reports. Attachment I lists the meeting attendee Attachment II summarizes the licensee's presentation to the NRC and the issues discusse .0 Backshift Inspection NRC' inspections of licensee activities on backshifts were conducted on the following dates:

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February 4,1989

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Areas of inspection included the observation of core spray valve oper-ability and inservice test and control activities, and performance of plant tour .0 Review of Periodic and Special Reports The inspectors reviewed the licensee Monthly Operating Report for January 1989 and the Weekly Operating Reports as they were submitted. In addi-tion, reports submitted by the licensee pursuant to Technical Specifica-tion requirements were examined by the inspectors. These reviews included the following considerations: the report includes the information required to be reported to the NRC; the planned corrective actions are adequate for resolution of identified problems; and the reported information is vali .0 Unresolved Items Unresolved items are matters for which more information is required in order to ascertain whether they are acceptable, violations, or deviation Unresolved items are discussed in paragraphs 1.0, 2.0 and 5.0 of this repor .0 Exit Interview A summary of the results of the inspection activities performed during this report period were made at meetings with senior licensee management at the end of this inspection. The licensee stated that, of the subjects discussed at the exit interview, no proprietary information was included.

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ATTACHMENT I .

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ATTENDEES AT MANAGEMENT MEETING

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FEBRUARY 2, 1989

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GPU Nuclear Corporation G. Busch, Dyster. Creek Licensing Manager G. Capodanno, Director, Engineering and Design E. Fitzpatrick, Vice President and Director Oyster Creek M. : Laggart, Licensing 'and Regulatory Affairs A. Rone, Plant Engineering Director J. Sullivan, Licensing and Regulatory Affairs NRC-W. Baunack, Project Engineer, Reactor Projects Section IA L Bettenhausen, Chief Projects. Branch 1 S. Chaudhary,. Senior. Reactor Engineer E. Collins, Senior. Resident Inspector, Qyster Creek S. Collins, Deputy Director of Reactor Projects C. Cowgill, Chief, Reactor Projects Section 1A J. Durr, Chief, Engineering Branch, DRS T. Easlick, Dperations Engineer W. Kane, Director of Reactor Projects K. Kolaczyk, Reactor Engineer R. Mc Brearty, Reactor Engineer J. Strosnider,- Chief, Materials and Processes Section NRR'

L A. Dromerick, Licensing Project Manager S. Guthrie, Sr. Ops. Engineer / Team Leader, NRR/RSIB R. Hernan, Senior Project Manager (TMI-1)

J. Stolz, Project Director, PD 1-4 C. Vandenburgh, Sr. Ops. Engineer / Team Leader, NRR/RSIB State of New Jersey R.:Ebright, NJ DEP/BNE - Special Projects M._Jacobs, NJ DEP/BNE - Nuclear Engineer D. Whito, NJ DEP/BNE - Section Supervisor l

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