IR 05000320/1989004
| ML20245E643 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/20/1989 |
| From: | Cowgill C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20245E634 | List: |
| References | |
| 50-320-89-04, 50-320-89-4, NUDOCS 8906270355 | |
| Download: ML20245E643 (10) | |
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l U.S. NUCLEAR REGULATORY COMMISSION
REGION I
i Report No.
50-320/89-04 i
Docket No.
50-320 j
License No.
DPR-73 Priority
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Category C
Licensee:
GPU Nuclear Corporation j
P. O. Box 480 Middletown, Pennsylvania 17057 i
Facility Name: Three Mile Island Nuclear Station, Unit 2 Inspection At: Middletown, Pennsylvania Inspection Conducted: April 22 - June 2,1989 i
Inspectors:
F. Young, Senior Resident Inspector D. Johnson, Resident Inspector T. Mo lak, Jesident Inspector i
Approved by:
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1 N gill, Ch
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ctor Projects Section 1A
' Date '
Inspection Summary:
Areas Inspected: Routine safety inspection by site inspectors of defueling and decontamination activities, including the proper implementation of radiological controls, housekeeping measures, follow-up on a plant event and licensee at-tions on previous inspection findings.
Results: Licensee personnel continue to conduct defueling activities in a safe manner. There was one instance, however, where defueling operations were not conducted in accordance with Technical Specifications.
A core debris sample was transferred without the required notifications being made.
Review of this licensee-identified violation indicated timely identification and response by
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management. Ten previous inspection findings were closed based on inspector review and licensee actions to resolve the issues.
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' TABLE OF CONTENTS
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l 1.0 0verview.............................................................
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1.1. Licensee Activities.............................................
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1.2 NRC Staff Activities............................................
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1.3 Persons Contacted................................................
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'2.0 Defueling/ Decontamination Activities (NIP.71707).....................
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t 2.1 Scope of Review..................................................
'2 2.2 Unauthorized Sample Transfer...........
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2.3 Ge n e ra l F i n d i n g s................................................
3.0 Licensee Actions on Previous Inspection Findings -(NIP. 92703).........
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- 3.1 (Closed) Inspector Follow Item (320/83-02-01):: NRC Review of Licensee Analysi s of Crane Power Supply Cable Strength........
i 3.2 (Closed) Inspector Follow Item (520/83-12-02): Lack of Pro--
cedure to Control Design Interface Between Licensee On-Site Engineering and Of f-site Contractor. Engineering Group.........
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3.3 (Closed) Inspector Follow Item (320/83-16-05): NRC Staff to
Review Corrective Actions to Resolve Radiological-Deficiency Report (RDR) No.83-043.......................................
3.4 (Closed) Inspector Follow Item (320/84-04-03): NRC Staff to Re-view Organizational Responsibilities' and.. Approval Process for -
Implementing the As low As Reasonably Achievable ALARA P r o g r a m........................................... (...... ).....
3.5 (Closed) Inspector Follow Item (320/84-04-05):.NRC Staff Will Review the ALARA Program Decision-Making Process..............
3.6 (Closed) Inspector Follow Item (320/84-12-02): Inconsistent.
Methoc; for Dedicati ng Commercial Grade Components.............
3.7 (Closed) Inspector Follow Item (320/85-12-01): NRC' Review of Quality Assurance Resolution for Material Non-Conformance Report (MNCR) for Reactor Building Service Crane..............
3.8 (Closed) Inspector Follow Item (320/85-21-05): Licensee to
. Provide Evaluation of Unmonitored Release Pathways............
'8 3.9-(Closed) Inspector'FollowItem(320/86-10-04): Licensee to Establish Formal Communication Channels for Inter-De ALARA Coordination...............................partmental.
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3.30 (Closed) Inspector Follow Item (320/87-03-03):. Revisions to Dose Assessment Procedure..........................................
4 '. 0 Management Meeting (NIP 30703).......................................
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DETAILS 1.0- Overview 1.1: Licensee Activities Following removal of the elliptical flow distributor,:defueling crews
. used pick and~ place techniques and air lifting equipment to remove.
large quantities of core debris from the lower heads During the-month of May, over 30,000-pounds of material were' loaded into canis-ters in preparation for~ shipment.
To date,.approximately'90%.of the core debris has been removed'from the vessel.
Upon removing'the accessible material from the' bottom-area ofLthe
. reactor vessel, airlifting' operations were stopped and a new in-vessel water filtration system (IVFS) was installed in the essel.
The IVFS is aesigned to. augment the defueling water clean-up.. system andl improve water clarity during futur'e defueling operations; Fol-lowing installation of.the IVFS, defueling crews began breaking up the solid slab of material located at the bottom of the' reactor vessel. The slab weighs about 7.5 tons, measures'about 5 feet in diameter and consists of resolidified material that flowed into the
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lower head region during the accident.
The crust impact tool has been very effective in breaking.up the slab.
The resolidified material consists of a relatively thin crust with underlying loose gravel-like material. ' Following completion of the first phase of breaking up the slab, defuelers found. that a hard -
.i mass, about 24 inches in diameter and 8 inches high, remains under
the broken pieces. Approximately 80 percent of the original rock-like mass has been broken es into small easily. removable pieces.
Much of this material is suitable for airlifting. -At the close of the inspection period, the licensee was still. in the process of air-lifting material from.the bottom of the reactor. vessel.
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1.2 NRC Staff Activities The purpose of this inspection was to assess licensee activities dur-ing defueling and decontamination activities:.
This assessment was
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made through observations of licensee activities, interviews with-licensee personnel, and review of applicable documents.
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q The inspectors reviewed licensee's procedures implementing control.'
i on several interfacing systems with the reactor vessel to assure ade-
quate controls were in place to prevent uncontrolled boron dilution.
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The. inspectors also reviewed instrument calibration, switching and l
tagging of valves, approval authority, and activities of operations.
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NRC staff inspections use the acceptance criteria and guidance of NRC Inspection Procedures (NIP's).
These NIP's were annotated in the Table of Contents to this report.
1.3 Persons Contacted During this inspection, the following key licensee personnel provided l
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information in the development of the inspectors' findings.
- J. Byrne, Manager, TMI-2 Licensing
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l L. Edwards, Quality Assurance (QA) Auditor
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- J. Frew, Defueling Director
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- G. Kuehn, Site Operations Director, TMI-2
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- S. Levin, Defueling Operations Director
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- M. Roche, Director, TMI-2
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- R. Rogan, Director, Licensing & Nuclear Safety, TMI-2
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- E. Schrull, TMI-2 Licensing Engineer
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- D. Turner, Director, Radiological Controls
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l R. Wells, Licensing Engineer
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- Attended the final management meeting.
2.0 Defueling/ Decontamination Activities
2.1 Scope of Review
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The inspector observed and/or reviewed licensee defueling/ decontamination activities to: (1) ascertain factual status of such activities and (2) as-sure proper adherence to applicable procedures.
The inspector also made observations in facility spaces with respect to proper housekeeping, fire protection, and radiological controls.
The general acceptance criteria for this review was Section 6 of the TMI-2 Technical Specifications (TS).
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In performing the above inspections, the inspectors focused on the fol-lowing areas of licensee performance:
control of operations in progress by supervisory personnel;
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knowledge of the task by technicians and support personnel; appropriateness of governing documents, including procedures and
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Radiation Work Permits (RWP's);
alertness of various controlling station personnel;
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assess the quality of implementation of selected evolutions wit-
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nessed; and, assess the material condition of the plant.
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1 2.2 Unauthorized Sample Transfer Overview l
During the course of transferring small sample; receptacles containing core debris from the reactor vessel to shielded casks, in preparation
for shipment, the licensee identified an event in Pich.one of.the '
i sample containers was transferred without the authorization of a lic-l ensed Senior ReactorL0perator (SRO).or Fuel Handling limited SRO I
( FHSRO)..The details of this event were reported to the on-site NRC-l staff by the licensee.
-Details of the Event i
On May 9, 1989'at approximately 9:55 a.m., a small (2.5" 0.D. X_7.69" j
long) sample receptacle containing about 5.2-pounds of core _ debris j
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was transferred from the reactor vessel-and:placed in a shielded con-tainer. This assembly was then placed. into a type 2R-container for i
storage in preparation for eventual shipment off-site. The activity I
was performed successfully and safely with the sample assembly placed-l in. temporary storage on the 305' elevation of the' reactor building.
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Although this activity was planned, it was performed without the Con-trol Room or senior. licensed personnel beino first notified. Since the activity of removing core debris is classified as c " Core Altera-l tion" in accordance with-lMI-2 Technical Specifications (TS) and, accordingly, requires that communications be established with senior licensed personnei prior to commencing.such an activity, the evolu-tion was not in compliance with the Technical Specifications.
The failure to make the required notifications was identified a day j
after the evolution when, following their standard practice, the De-i fueling Operations Support Section reviewed ~the preceding day's log
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entries.
The reviewer noticed that senior licensed personnel had no
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watch stationed during the transfer of the sample.
The -'signi ficance of failure to comply with the notification requirements'was realized
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and the appropriate actions / notifications were taken.
Subsequently, a critique meeting was_ convened with the responsible parties on May
12, 1989 to identify the cause and initiate corrective actions to preclude a recurrence of a similar nature.
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Tnrough the critique, the licensee determined that the event involved a communications error.
Core alteration (the sample transfer) had
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been performed by the Defueling Operations Support Section without
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unauthorized core alteration occurred when core debris, contained in
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a sample receptacle, that was suspended by a lanyard in the reactor vessel, was retrieved and transferred out.
The non-licensed Task Supervisor, who performed this activity, knew this type of operation I
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was' classified as a Core Alteration,' however',, he inadvertently per-formed the task'without the FHSRO being.present in the Command Cen-ter.
The FHSRO was in the control room when the evolution occurred.
He-was aware that sometime during the shift-the' sample' transfer. might be performed.
The root,cause of the' event was evaluated based on the following facts:
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The FHSR0 being aware that a task c1hssified as a Core Altera-
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tion was imminent,-failed to emphasize-to the Task Supervisor.
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that he must.be notifiedibefore proceeding with this operation.
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The Task Supervisor failed to notify' the FHSR0 immediately prior to commencement' of the. fuel debris. sample retrieval.,
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quently,,no.FHSRO watch'was stationed ~for this activity as re-quired by procedures (4730-IMP-3221.02) and TMI-2 Technical Specifications.
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The Task Supervisor and FHSRO failed to discuss planned acti-vities in sufficient detail following shift turnover.
Licensee representatives. conducting the critique concluded that the event could have been prevented if the Task Supervisor had adequately briefed the FHSR0 and maintained close communications with the on-
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duty FHSR0 to ensure that the FHSRO was cognizant of every phase of
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the activities to be performed during the shift.
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Immediate actions-taken to preclude a recurrence included site man-agement discussing the event with all FHSR0s and Task Supervisors on a one-on-one basis, re emphasizing that-future activities which have l
the potential to constitute a-Core Alteration should be conserva-tively approached and that it is : imperative that defueling support j
personnel maintain close communication with the on-duty FHSRO so that'
the FHSRO is cognizant of all activities which are plannedito be per--
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formed in vessel.
The Defueling Operations Director'followed up on
these counseling sessions by issuing a memorandum (4700-89-F-005) to-
supervisors reiterating what procedural prerequisites must he com-pleted prior to performing work in the reactor vessel. Additionally, a revision (Procedure Change Request 89-0356) was made to the sample removal procedure (4730-IMP-3221.02) that would make it more explicit
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in contentito ensure the direct notification of'the FHSRO resulted in-i his " Stationing a Watch" and verifying communications with the re-
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sponsible parties.
l Inspector Findings
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From a review of the circumstances resulting in this event, the'in-
spector determined that the' event resulted from unclear verbal com-
munications between the Task Supervisor and the FHSRO.
Failure of
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the Task. Supervisor to have direct communications with the FHSR0 while performing the sample transfer is contrary to the intent of
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Technical Specifications 3.5 that requires in part.that " Direct com-munication'shall be maintained between the Control Room or.the Com-mand Center and personnel in the' Reactor Building As stated.in Technical Specifications, Table 6.2-1, the additional senior reactor operator (SRO) or SRO ' limited.to fuel handling, notwithstanding;1oca -
tion, will.have' direct communications with personnel;in the Reactor Building performing core alterations."
From the. review, the inspector determined that the activity of trans-ferring the sample itself was performed in a safe manner, by the re-quired support personnel in the reactor building, and under the direc-tion of.the Task Supervisor.
The movement of a small quantity of core debris within the confines of the reactor building had minor
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safety or environmental significance.
The licensee's internal ad-ministrative,antrols identified the oversight in a time'y manner.
The licensee acted promptly to identify the cause and to take reme-dial action.
In general, the itcensee's response to the event was timely and thorough.
Based on licensee actions and the limited safety significance, the violation has been classified as a non-cited violation per the cri-teria specified in Section V.G. of the Enforcement Policy of 10 CFR 2 (320/89-04-01).
2.3 General Findings As a result of the routine review noted above, the inspectors iden-tified a non-citable violation.
In general, defueling activities and clean-up operations were conducted in a safe manner.
3.0 Licensee Actions on Previously Identified Items The inspector reviewed licensee action on previous inspection-findings to ensure that the licensee took appropriate action in response to the find-ings or by self-initiative and that the licensee's action was timely.
3.1 (Closed) Inspector Follow Item (89-320/83-02-01h NRC Review of Licensee Analysis of Crane Power Supply Cable Strength i
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s Through review of the safety analysis ~that supported installation of i
a new electric breaker and power cable for the Polar Crane (S-ECM
1017), the inspector determined that the power source and circuit are.
-l Non-IE and are not associated with-IE equipment. The installed power
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feeder (3-1/C #2/0 welding cable) is-attached with messenger hangers to 5/16" diameter messenger wire which is double clamped at the
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crane's walkway.
The feeder is supported by Kellems grips at the l
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In the highly-unlikely event that the 5/16" diameter mes-senger wire and all 3-1C #2/0 cables were to b'reak, the. power cabling -
would not strike any~ safety related equipment or systems.
To date, routine inspections by the licensee indicate'that the crane'.s power cable has not experienced any undue stresses and.is functioning as designed. This item is closed.
3.2 IClosed) Inspector'FollowItem(320/83-12-02): Lack of Procedure to.
Control Design Interface Between Licensee On-Site Engineering and
Off-Site Contractor Engineering Group
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The inspector reviewed the interface between licen'see'on-site engi-neering and the. off-site contractor. erigineering group. The. inspector found the Bechtel Design Engineering Group, located off-site in Gaithersburg', MD and the on-site GPU Nuclear Recovery Engineering.
Section integrated their organization in 1983. 'This integration..
negated the need for a design interface control procedure since all procedures written for'GPU Nuclear directly applied t'o the off-site Design Engineering Group. The inspector had no additional questions on this matter.
This item is closed.
3.3 (Closed) Inspector Follow Item (320/83-16-05): NRC Staff to Review Corrective Actions to Resolve Radiological.Deficienc3. 5p' ort (RDR)
No.83-043 The inspector reviewed Revision 1 to licensee Administrative Proce-L l
dure 1501-ADM-4410.01, " Receipt of Radioactive Material at Three Mile.
Island". Through this review, the inspector determined that the lic-ensee had provided the necessary guidance to. assure. that radioactive
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materials are received in accordance with applicable. regulations and'
l licenses. The procedural. requirement _s.are also applicable to all contractors providing services to the licensee.
The inspector con-cluded that the licensee had adequately resolved RDP No.83-043.
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This item is closed l
i 3.4 (Closed) Inspector Follow Item (320/84-04-03): NRC Staff to Review Organizational Responsibilities and Approval Process for Implement-ing the As Low As Reasonably Achievable (ALARA) Program i
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-i The inspector reviewed TMI-2 Unit Policy / Plan 4000-PLN-4010.01,
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"TMI-2 ALARA Program Plan".
Through this review, the inspector con-cluded that the licensee has established a documented review process identifying the basis of ALARA measures specified in various planning documents including those developed by contractors-This process
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also includes ALARA program effectiveness reviews performed by Radio--
logical Controls Department management.
The inspector had no further questions on this matter.
This item is closed.
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3.5 :(Closed) Inspector Follow Item (320/84-04-05): NRC Staff Will Review the ALARA Program Decision Making Process The inspecto'r reviewed the ;TMI-2 ALARA Policy / Plan (4000-PLN-4010.01)
and TMI-2 ALARA Program (4000-ADM-4010.01) that implemented the unit's ALARA program.
Through this review, the' inspector determined that therlicensee has provided definitive guidance for the' inclusion-of the ALARA concept into the decision making process associated with j
management and supervisory level' directives and evaluations that could result in personnel radiation exposure. These procedures pro :
vide the methodology to be used to determine when it is necessary.to-perform a decision analysis of alternative work-options to maintain-radiation exposures ALARA.
General guidance for performing such'de-
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cision analysis is provided.
TheLinspector had no further questions on this matter. :This item is closed.
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3.6 (Closed) Inspector Follow Item (320/84-12-02): Inconsistent Method for Dedicating Commercial Grade Components:
The inspector reviewed Engineering Procedure 4000-ENG-6200.01, " Pre-paration and Review of Procurement Documents" Revision 5.
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this review, the inspector determined that the licensce-revised this procedure in response to this finding to assure that appropriate-i technical and quality. requirements are contained in procurement docu-d ments that are prepared, reviewed,. approved and subm_itted by the lic-ensee and contractor Engineering groups. These' procedural changes-assure consistency for. dedicating commercial grade.' components. The
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inspector had no additional questions'on this matter.
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3.7 (Closed) Inspector Follow Item (320/85-12-01): NRC Review of Quality j
Assurance Resolution for Material Nonconformance Report (MNCR) of
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Reactor Building Service Crane i
.The inspector reviewed licensee documents resolving MNCR 0150-85, addressing load testing of the Reactor Building Service Crane.
Through this review, the inspector _ determined _that the licensee's
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Design Engineering Department contacted the crane manufacturer to
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provide an independent inspection of the overhead crane.
Subsequent to this inspection, the licensee implemented the manufacturer's. recom-i mendations.
Following an operational verification-that-the affected components functioned satisfactorily, the MNCR was resolved by the
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Quality Assurance Department.
The inspector com iders the MNCR was
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appropriately resolved within the licensee's organization. This item-l is closed.
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3.8 '(Closed) Inspector Follow ~ Item (320/85-21-05): Licensee to Provide-
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Evaluation of Unmonitored Release Pathways The. Licensee provided.the NRC with an evaluation of potential unmoni-
' tored release pathways which may have existed during 'and after the-TMI-2 accident.
In the correspondence, (4410-89-L-0005/0216P) dated May 10,.1989, the pathways were evaluated to determine the bounds of
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credible releases and projected dose assessments. This report also serves to update-the 1979 Dose Assessment reports that were pre-viously submitted. This item is closed.
3.9 -(Closed) Inspector Follow Item (320/86-10-04): Licensee to Establish Formal Communication Channels for Interdepartmental ALARA Coordination The' inspector determined that the licensee had established' procedural guidance on' interdepartmental communications for implementing the.
ALARA concept through the TMI-2 ALARA Plan (4000-PLN-4010.01) and.the-TMI-2 ALARA' Program (4000-ADM-4010.01).
To evaluate the effective-
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ness of this guidance, administrative procedure 9200-ADM-1201.09,
" Radiological Controls Internal Assessments", was developed. The inspector had no additional questions on this matter.. This item-is closed.
3.10 (Closed) Inspector Follow Item (320/87-03-03): Revisio'n to Dose Assessment Procedure
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The inspector reviewed Revision 4 to Radiological Controls Procedure 9200-ADM-4330.02, " Personnel Contamination Monitoring and_Decontami-nation". Through this review, the inspector determined that-the pro-l cedure has been adequately revised to explicitly describe ' personnel; contamination monitoring requirements and actions to be taken by various members of the licensees Radiological Controls organization wsen contamination has been detected in personnel. The inspector had no further questions on this matter.
This item is closed.
4.0 Management Meetino
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The inspector discussed the inspection scope and findings with licensee i
management periodica11y'~d0 ring the course of the inspection and at a final j
= meeting conducted June 5,1989.
Licensee management attending the exit meeting are noted in paragraph 1.3.
The inspection results, as discussed at the meeting, are summarized in the l
cover page.of the inspection report.
Licensee representativesLindicated i
that none of the subjects discussed contained proprietary or safeguards
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