IR 05000346/1990012
| ML20043C250 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 05/25/1990 |
| From: | Byron P, Miller D, Schumacher M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20043C243 | List: |
| References | |
| 50-346-90-12, NUDOCS 9006040278 | |
| Download: ML20043C250 (10) | |
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- U.S. NUCLEAR REGULATORY COMMISSIONE
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Licensee: 1 Toledo' Edison' Company Edison' Plaza
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300 Madison'Avenu'e:
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- Facility Name
Davis-BesselNuclear Power Station y
ilnspection~At: Davis-Besse Site, Oak Harbor,; 0hio
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ilnspection'Conductedt May.1-17,.1990
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zInspectors: D.-E.LMiller.
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$)lhMY ppg Approved:By: -M. C. Schumacher,ichief Radiological Controls and
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Chemistry Section
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. Inspection Summary.
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l Inspection on May 1-17, 1990 (Report No. 50-346/90012(DRSS))
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(AreasInspected: -Special announced inspections of circumstances surrounding aunexpected radiation doses. received during transport-of the-core support
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e, assembly on A i May 1, :1990 _(pril ?25,)1990,- and of.the refueling-canal draining event on IP-93702.
Results: ;The review-of the' April 25,,1990, core support assembly.(CSA)'
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. incident and =the May 1,1990, refueling canal draining incident ' indicated
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S apparent violations.of regulatory requirements associated with the e incidents
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.in_that the potential for receipt of significantly greater personal doses a
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existed :for: loss of decay heat'. cooling during-the. refueling car ' drai
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i incident (section4). These two incidents appear to have common caus.
Lfactors. The ~ appropriate enforcement action will be determined n d
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communicatcd tolthe licensee by separate correspondence.
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t DETAILS 1.
Persons Contacted
T. Anderson, Manager, Maintenance Planning
- M. Bezilla, Operations Superintendent
- R. Brandt, Manager, Plant Operations (A)
- R. Coad, Supervisor,' Radiological Protection
- T. _Cobbledick, Quality Assurance F. Currence, B&W Task Leader
- J. Doiron, Senior Radiological Assessor
- R. Gaston, Licensing Technologist
- L. Harder, General-Supervisor, RC-RW
- G. lionma, Compliance Supervisor, Nuclear Licensing
"*J. Lash, Manager, ISE
- S. Love, B&W Site Representative
- J. Polyak, Manager, Radiological Control
- R. Ramsdell, Performance Engineering
- A. Reynolds, B&W Site Manager L. Richter, Performance Engineering
'R. Rishel, Quality Assurance Supervisor
- D. Schreiner, Operations Assessment Supervisor
'D. Shelton, Vice President, Nuclear
- B. Sningleton, Licensing Engineer
- L. Storz, Plant Manager
- P. Byron, NRC Senior Resident Inspector
- D. Kosloff, NRC Resident Inspector
- K. Walton, NRC_ Resident Inspector
- Denotes those present at the May 4, 1990, exit meeting.
Denotes those present at the May 17, 1990, exit meeting.
The inspectors also contacted other licensee and contractor employees.
2.
General This inspection was conducted to review the circumstances surrounding unexpected radiation doses received during transport of'the core support assembly (also described as a Core Barrel Lift (CBL)) on April 25, 1990, and of the refueling canal draining event on May 1, 1990. The inspectors reviewed documentation of the licensee's assessments of the incidents, visited the sites of the incidents, interviewed appropriate persons, and independently reviewed procedures and logs. Apparent violations of regulatory requirements are indicated in 1e details section of this report.
3.
Core Support Assmbly Incident Summar' c' Inspector's Review of Core Support Assembly (CSA) Incident On May 3 and 4, 1990, the inspectors reviewed the circumstances surrounding the CSA incident on April 25, 1990, whereby persons
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performing the CSA move received significantly higher than expected radiation doses. The inspectors reviewed the licensee's assessment team's report and packet of associated accumulated materials; visited the work site with a person present during the CSA move; interviewed some of the principal participants in the move; and reviewed a video tape of.the CSA move to the' storage position on March 4, 1990, and a video tape of obstructions in the shcilow end refueling canal taken-after the April 25, 1990 CSA move. The following abbreviated description of the incident is based on the inspector's review.
Core Support ?;sembly (CSA) Incident Details
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As part of the required ten-year in-service-inspection, the reactor vessel internals were removed so that inspections, of components including core barrel bolts, could be performed.
On March 4, 1990, the CSA was removed fram the reactor vessel and placed in the deep end of the refueling canal. The CSA was moved without incident. An RWP had been written, ALARA reviews performed, and pre-job ALARA and operations meetings conducted with all major participants in attendance. The general area exposure rates on the catwalks beside the canal were 400 to 700 mrem /hr during the transfer.
The total dose received by persons performing the CSA move was 0.438 person-rem.
During the move, the CSA was raised so that about eight feet of it was out of the water as predicted during pre-job meetings.
The CSA must be raised with about eight feet out of tne water so that it clears fixed obstructions on the refueling canal shallow end when it is moved to storage.
The move was made under the direction of a vendor lift specialist.
Participating in the move were the lift specialist and his supervisor, contractor craftsman, and Radiological Control Technicians (RCTs). The lift specialist has been involved in most CSA moves at B&W stations, including at least one previous U984) two-way move at Davis-Besse Station.
On April 25, 1990, the CSA was ready to be returned to the reactor vessel ~. The vendor task leader and a TECo operation supervisor walked-down the planned move to look for interferences; they apparently noted that there were some obstructions on the shallow end of the refueling canal floor (hoses, and a " trash can" projecting above a fuel storage rack adjacent to the shallow end to deep end wall). The lift specialist, who had flown in to direct the CSA move, was at a motel and did not participate in this walkdown or luter ALARA and operations pre-job meetings.
The operations supervisor also did not attend the pre-job meetings. During these meetings, the obstructions were not specifically discussed; discussed, apparently, were the circumstances, dose-rates, personnel needs, etc., based on the March 4, 1990 CSA move.
According to ALARA personnel, they were not informed that this CSA move would be different than the past move.
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According to the lift specialist, when he got to the job site he performed a walkdown of the area to be traversed in reinstalling the CSA in the vessel.
He stated that he saw the obstruction and believed he must raise the CSA 12 to 18 inches higher than during the March 4, 1990 CSA' move. He stated that he did not relate this-information to ALARA persons or to RCTs assigi,ed to monitor the move. When the job began, difficulties were experienced in latching the CSA to the rigging assembly; about three hours elapsed before the latching was accomplished and lifting began. During the CSA lift and trolley toward the reactor vessel, the lift specialist directed the CSA to be lifted about 18 inches higher (estimated by the lift specialist) out of the water than the previous lift to clear the obstructions not present during the previous CSA transfer.
As the CSA was lifted higher than before, the general area. dose rate about eight feet from one side of the CSA rose to about 30 R/hr where it had been about 0.7 R/hr during the previous lift. A craftsman, manning a rope used to help stabilize the CSA, was at or near the location of the 30 R/hr reading.
RCTs monitoring the lift directed all persons near the CSA to move away and get behind shielding; the RCTs then yelled to the lead RCT, who was several feet away trom them with the lift specialist informing him of the elevated readings.
The exposure rate at the lead RCTs location was six R/hr. The lead RCT instructed the lift specialist, to lower the CSA to reduce the dose rate; the CSA was being trolleyed over the obstruction at that time. As soon as the CSA was clear of the obstruction, the lift specialist directed the CSA to be lowered. The total time the elevated (unpredicted) exposure rates existed was about two minutes as indicated by plots of direct radiation monitor readings.
The CSA was trolleyed the remainder of the way and lowered into the reactor vessel without further incident. The total dose received by persons performing the CSA move was about 2.8 person-rem; the highest to an individual was about 0.45 rem.
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Inspectors Assessment of the CSA Incident The root cause of the event appears to be inadequate pre-planning and evaluation of the job of returning the CSA to the reactor vessel on April 25, 1990.
In particular, the licensee failed to evalmate before-hand the radiological significance of having to lift the CSA further out of the water than had been necessary when it was removed from the vessel on March 24, 1990. This difference made the evaluation for the March 4 move inapplicable to the April 25 move and resulted in higher than expected radiation doses to the involved workers. The failure to perform an adequate evaluation for the April 25 CSA move was an apparent violation of 10 CFR 20.201(b) which requires performance of such surveys (evaluations) as are reasonable to evaluate the extent of the radiation hazards that may be present (Violation 50-346/90012-01). The licensee also appears to have violated Technical Specification 6.8.1 and Adminis-trative Procedure DB-MH-00006.
" Control of Lifting and Handling Equipment," which require preparation of detailed handling procedures for items that require special handling.
Contrary to that requirement, implementing procedure DB-MN-090 2, " Reactor Vessel Internals Removal
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and' Installation",1does not include a method for determining the
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7 clearance between the bottom of the: CSA and- @c. floor of the ' shallow :
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- end-of the refueling pool during the movement and transport off the CSA (
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In failing'to' perform this evaluation, the' licensee thereby feiled tor anticipate the significant potential for an overexposure:in theLjob'as :
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. performed onLApril 25,=1990.- The dose rate at about 8 feet from the -CSA.
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.on April,25 was. higher; by a factor of about 40 than that-observed on;
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March 4.
Moreover, without' defined limits for'the-lift =, there was no (
assurance that the CSA could not have beca 1;fted even higher from'the'
.y water.with:a further.significant increase :n dose rate and dose.-
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E Both of the above described violations were recognized by the licensee 1
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and steps appear to'.be iniprogress to corrcct-the causes. The~ licensee E
has also: identified and is pursuing corrective actions for other-i contributory weaknesses as noted in the following section.
l Licensee Assessment of the CSA' Incident si Shortly after the CSA incident occurred,-the licensee formed anc
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- assessment team to review.theLincident. The seven-person team issued-
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a report of.their, findings on May 2, 1990. The assessment team identified the following causal. factors (paraphrased):
Procedure DB-MM-09092 does not.specify a verification method to
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determine: elevation of the CSA during. transfer; procedure -
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DB-MN-00006 requires-detailed handling procedures.
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Because there were lighting differences between-the shallow l
and deep ends of the pool,:walkdowns did notLidentify that the y
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obstructions (a trash can-in a fuel rack and trash can chutes j
u arri hoses -in shallow end) presented a problem.for the CSA move; J
- ano.the lift specialist was not present at the initial walkdown.
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No consideration was given at the pre-job meeting to minimizing.
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elevation; of CSA during the move.
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q The radiation work petuit for this job was too general.
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Procedures did'not require the lift specialist to attend the g.g April 25,1990 pre-job meetings.
f The most senior TECo persor at the ALARA briefing was the ALARA
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coordinator; the team considered this inadequate for such a F
significant physical task 'with significant radiological potential.
The team-believed that the' plant manager or his designated alternate should have been present to provide objective questioning of participants in the CSA move.
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The ALARA briefing did not establish control authority for the job or a "stop-work" dose rate.
The licensee's assessment report includes recommendations for each of the
. listed causal factors to preclude recurrence.
After completion of the onsite portion of the inspection of the CSA incident, the licensee performed an IndepeuNnt Safety Engineering (ISE)
Assessment of the incident. The ISE draft report agreed with most original assessment findings but concluded that the CSA was lifted to a height where the CSA's core barrel bolt ring was 12 to 18 inches above the water; the licensee based this height on projections of "uence near-the top.of the core and the resulting activation, radiation monitor reading plots (duration of time of exposure-rate increase), and rate of travel of the polar-crane while lifting the CSA. The ISE then concluded that, based on expected dose rate increases if the CSA was further raised six to twelve inc6,es, a substantial potential for overexposure in excess
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of 10 CFR 20 limits did not exist.
4.
Refueling Canal Draining Incident Details On April 30, 1990, at 1:20 a.m. the licensee completed refueling the reactor and at 11:30 a.m. installed the indexing fixture in preparation for the insertion of the plenum assembly. The plenum assembly was then installed; the indexing fixture was not removed as it was to be used later for shielding during the cleaning and preparation of the reactor vessel head stud holes.
After plenum installati:m the licensee planned to insert the incore neutron detectors; to facilitate this the refueling canal was to be drained to an elevation of 578'6" which is 6".above the-bottom of the refueling canal and incore tank.
The operators discussed the draining of the refueling canal and requested that the containment coordinator periodically monitor the refueling canal water level. A dedicated Reactor Operator (RO) was assigned to go.into containment when the refueling canal irs was at five feet. The licensee lined up Decay Heat Pump (DHP) 1-c for decay heat removal, and DHP 1-1 for draining the refueling cand to the Borated Water Storage Tank (BWST) at a rate of approximately 1500 gpm.
Draining started about 10:00 p.m. on April 30, 1990. The assigned R0 in the control roon. was monitoring the refueling canal level indicator L1 1627, the BWST level indicator and reactor vessel level indicator L1 214; LI 214 senses level from the hot leg of Steam Generator (SG) 1-1 to the high point of SC 1 -1.
The high point vents were open. The refueling canal level innecation was observed by the R0 to be decreasing at the rate of one inch every 10 minutes. The operator observed a disparity between the refueling caal le'lel and that shown on L1214.
Subsequent investigation by the licensee determined that the filter on '.I 214 was damaged which affected its venting capability and resulted in erroneous data.
Later the operator observed that reactor vessel 'evel was decreasing while refueling canal level was constant and reduced the drainage rate to about 500 gpm. The containment coordinator notified the control room, about che same time about the reactor vessel draining.
He also mentioned the
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placement of the indexing fixture to the RO who stated that he thought the indexing fixture was in its stored position.
The containment coordinator then notified tt.e control room tt t the plenum was beginning to become uncovered. The RO stopped draining and secured DHP 1-1.
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the containri.ent radiological controls (RC) supervisor observed the uncovered plenum he directed personnel to leave the area and then he performed surveys at tha walkway.
The highest exposure rata measured was 130 mR/hr e+ the west walkway.
The shift supervisor directed that
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the suction to AP l-1 be swapped to the BWST and the reactor vessel was f
gravity filled to the refueling canal level.
The indexina fixture was
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placed in its stored position and the canal draining was completed.
The licensee notified the resident inspectors of the event and initiated a review by the Operations Assr sment Team.
This event and the subsequent investigations are, cumented in PCAQ 90-00369.
The inspectors determined that outage manac'.nent made the decision to e
I leave the indexing fixture in place while draining of the refueling b
canal. Operations input was not solicited as outage management believed that the indexing fixture would have no effect on the draining operation.
This decision was not adequately communicated to the operators who were not aware that the indexing fixture had not been removed.
Review of the outage short interval overview schedule, which is part of tne plan of am the day could lead one te conclude that the indexing fixture would be installed after the refueling canal was drained.
Outage management stated that B&W was aware of its decision to leave the inde; ng fixture in place during the drain.
a The refueling canal draining was accomplished utilizing procedure DB-0P-06023, " Fill, Drain, and Puri fication of Refueling Canal,"
Revision 00, dated February 14, 1990.
Section 3.6 addresses lowering the water level using a DHP.
The inspectors reviewed DB-0P-06023 and
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Jetermined that the procedure is not appropriate to the circumstances in that it does not address the indexing fixture nor does it contain precautions about the loss of decay heat cooling from a partially filled reactor coolant system as described in Generic Letter 87-12.
This is a violation of 10 CFR 50, Appendix B, Criterion V, which requires that activities effecting quality shall be appropriate to the circumstances (346/90012-03).
In addition, Section 3.6.13 requires that an operator be stationed to observe refueling canal levei but does not specify l oca tion.
The procedure also does not address any fluid levels other than that of the refueling canal.
The inspectors consider these to be
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procedural deficiencies.
Had the R0 only monitored the refueling canal leve' w rqui,ed by the procedure he would not have noted the disc ocies between the canal and reactor vessel levels.
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Conduct of Operations Procedure DB-0P-00000, Revision 01, dated April 2,
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1990, Section 6.'.6 requires that prior to the performance of critical, complicated, unusual, or inf requent operations, a procedure review shal'
be performed, and briefings shall be conducted bv the individual in charge of the evolution.
The licensee has determined that no rmal briefing was held nor was a pre-evolution walkde.a performed.
The shif t supervisor briefed individuals sep6rately, but did not identify
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4-equipment which was involved as reauired by 6.7.1.c nor does it appear that a procedure review was perfe aed. Since the draining of the refueling canal is performed during refueling outages it meets the criteria of infrequent operations. These are a violation of Technical Specification 6.8.1 (346/9U012-04) which requires that procedures be adhered to.
It should be noted that Section 6.7.6 of-procedure DB-0P-00000 was enhanced as corrective action for previous resident inspector concerns regarding the adequacy of pre-evolution briefings which are described in Inspection Report No. 50-346/88039(DRP)..
This event was not. reportable but had the potential to result in loss of decay heat cooling.
Circumstances: leading to the event indicate weaknesses in control.
Had che operators _not taken timely action the water level in the reactor vessel could have decreased to a point where cavitation would have occurred in the DHPs, thus losing decay heat removal capability.
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. Summary of Conclusions of the Inspector Review of Incident The' inspector determined that the operators normally drained the refueling canal using the re; cling canal drain puap in series with the BWST recirculation pump; this utilizes a flow path thst is not through the reactor vessel, with draining rates of 200 gpm.
The location of the indexing fixture has no affect when the normal draining method is used. The normal method was not utilized as the BWST recirculation pump was out of service. The indexing fixture is normally stored during draining.
This event raises several concerns:
The operators were not aware of plant conditions.
Operational decisiens were made by outage management.
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Operators did not perform walkdown of area prior to performing evolution.
An inadequate pre-job briefing was performed.
O g ratorc did not stop evolution to determine actual condition:, and did not assume responsibility, Procedure review appears inadequate es procedure does not address potential conc.arns.
Operations appeared to relegate their responsibilities.
The inspecters discussed their concerns with operations management *
May 3. 1990. The licensee took immediate action by emphasizing to the operators their authorities and responsibilities. The inspectors have observed an immediate and positive response to their concerns and have observed shift supervisors taking a more active role in pre-evolution actions.
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Apparent Common Causes of the Two Incidents a
Neither.the core support assembly event nor the refueling canal drain
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event were reportable, but both events had potentially significant consequences. The two. events were different but had the'following common causes.-
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Inadequate Procedures
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Core barrel lift (CBL): iso detailet instructions or procedures
as required by DB-MN-00006, Section 6.1.8.
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Refueling canal drain (RCD):
Procedure DB-0P-06023 did not address the potential when the indexing fixture is in place, or the possible loss of decay heat removal.
b.
Inadequate Pre-Job Briefings CBL:
Not all principal participants were required to attend the briefings.
_ RCD:
No formal briefing and the requirements of procedure DB-0P-00000 were not met, c.
Inadequate Walkdowns CBL:
Not all principal participants were required to attend wal kdown.
' RCD: No walkdown performed, d.
No One Person 0 Charge cBL: 'Both the lead RCT and the lift specialist believed he was in charge.
RCD: Outage management made operational decisions.
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Inadequate Review of Evolution CBL:
No one questioned the possible radiological consequences caused by the existence of obstructions in the refueling canal shallow end.
RCD:
No. a evaluated the effects of having the indexing fixture inplace while performing draining.
The root causes appeat to be inadequate preplanning; inattention to detail; failure to implement adequate quality controls; uncertain Jelegation of responsibilities, either real or assumed; and the failure to communicate between participating organizations.
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Exit Meeting (IP 30703).
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The inspectors met with licensee representativet (denote'd in Section 1)-
on May 4, 1990, and May 17, 1990, to discuss the-scope and findings of-the-special inspection including the apparent-violations (Sections 3 and 4). The inspectors also discussed the likely. informational content of
~the inspection report with regard to documents or processes reviewed by
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inspectors during the inspection. Licensee representat_ives identified no-such documents oc processes as proprietary, r
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