IR 05000315/1988024
| ML17325B059 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 11/29/1988 |
| From: | Burgess B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17325B057 | List: |
| References | |
| 50-315-88-24, 50-316-88-28, NUDOCS 8812140244 | |
| Download: ML17325B059 (30) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION III
Reports No. 50-315/88024(DRP);
50-316/88028(DRP)
Docket Nos. 50-315; 50-316 Licenses No. DPR-58; DPR-74 Licensee:
American Electric Power Service Corporation Indiana and Michigan Electric Company 1 Riverside Plaza Columbus, OH 43216 Facility Name:
Donald C.
Cook Nuclear Power Plant, Units 1 and
Inspection At:
Donald C.
Cook Site, 'Bridgman, Michigan Inspection Conducted:
October 20 through November 16, 1988 Inspectors:
B. L. Jorgensen M. A. Kunowski D.
G. Passehl Approved By:
B. L. Bur ss, Chief Projects Section 2A
/c p R5 a
e Ins ection Summar Ins ection on October 20 throu h November 16, 1988 (Re orts Nos.
50-315/88024(DRP);
p d
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by dd d ddb inspectors of:
actions on previously identified items; operational safety; radiological controls; maintenance; surveillance; fire protection; emergency preparedness; security; outages; safety assessment and quality verification; allegations, and NRC Region III requests.
No Safety Issues Management System (SIMS) items were reviewed, Results:
Of the 12 areas inspected, no violations or deviations were identified
~sn areas.
One violation was identified (Level IV fire protection procedure not followed - Paragraph 7) in the remaining area.
The inspection disclosed weaknesses in the results of licensee's recent fire prevention/control activities, on the Unit 2 generator repair project, with the cited Violation being only one example (Paragraph 7).
The inspection noted strengths in the licensee's maintenance planning to accomplish several jobs during a'ingle safety-related equipment outage, d
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and the maintenance controls governing use of injection-type leak sealant were found to be'particularly thorough (Paragraphs 5 and 13).
8812140244 881201 PDR ADOCK 05000815 PDC
DETAILS 1.
Persons Contacted
- W. Smith, Jr., Plant Manager A. Blind, Assistant Plant Manager, Administration J. Rutkowski, Assistant Plant Manager, Production
- L. Gibson, Assistant Plant Manager, Technical Support B. Svensson, Licensing Activity Coordinator K. Baker, Operations Superintendent
- J. Sampson, Safety and Assessment Superintendent E. Morse, gC/NDE Gener'al Supervisor T. Beilman, ISC/Planning Superintendent J. Droste, Maintenance Superintendent T. Postlewait, Technical Superintendent, Engineering L. Matthias, Administrative Superintendent
- M. Horvath, guality Assurance Supervisor D. Loope, Radiation Protection Supervisor M. Castiglione, Safety Director, M-K Ferguson
- J. White, SGRP Project Manager R. Krieger, Group Manager, Electrical Systems, AEPSC S. Brewer, Manager, Nuclear Safety and Licensing, AEPSC The inspector also contacted a number of other licensee and contract employees and informally interviewed operations, maintenance, and technical personnel.
- Denotes some of the personnel attending Management Interview on November 17, 1988.
2.
Actions on Previousl Identified Items (92701, 92702)
a ~
(Closed) Violation (316/83004-01):
Containment spray system eductor testing ai e
to incorporate R requirements and assure a proper demonstration of operability.
As indicated in recent reports (including Inspection Reports No. 316/88003 and No. 316/88023),
the licensee had performed special testing to collect data deemed necessary for a demonstration that the eductors, as installed, exhibit "classical" eductor behavior and can therefore be calculationally modeled.
Calculation of acceptable performance from limited or possibly nonrepresentative data was the core of this item.
Representatives of NRC Region III and NRR, along with the inspector, met with licensee representatives on November 15, 1988, to review the results of the special testing.
The licensee presented a satisfactory demonstration that, within the anticipated operating range, eductor performance can be calculationally modeled.
Actual performance and calculated performance were shown to be the same, assuming a
proportionality constant for the eductors of 0.70.
Prior to empirical derivation of this constant, the licensee had assumed
it to be 0.60 - close enough to validate prior test results.
Further, the NRC concluded the special testing met Unit 2 Technical Specification requirements for 1988.
On these bases, the item is considered closed.
Additional discussions were held concerning the licensee'
intentions to simplify the test again in the future to a single-point test.
This would appear to be valid, but an open question from the NRC Office of Nuclear Reactor Regulation (NRR) needed to be addressed as to the need to change Technical Specifications.
The licensee has withdrawn his request for this change, and agreed to provide the proposed single-point test procedure (to meet current Technical Specifications)
for NRR.review.
(Closed)
Unresolved Item (315/87007-01; 316/87007-01):
Review of e ectric protective re ay setpoint setting contro s by the Corporate office.
The subject area was reviewed during an inspection visit to the Corporate Office in Columbus, Ohio on November 16, 1988.
The licensee representatives described internal reviews, findings, and corrective or improvement actions over the past two years.
These actions have served to verify the subject settings are known, have been evaluated to be correct, and are accurately documented.
In many cases, new setpoints were established for instantaneous overcurrent using a
new basis (ten times normal amps vice six times plus "tweaking" if spunous trips occurred).
A safety review under
CFR 50.59 was performed for this design change (RFC-12-2862)
as required.
In addition, licensee enhancements, to the degree and specificity of information available at the site, were reviewed.
This included review of procedure EGSP 5.6, "Processing D.C.
Cook Nuclear Plant Relay Setting Sheets and Engineered Setpoint Lists", Revision 0 dated November 19, 1987.
(Closed)
0 en Item (316/87004-01);
Evaluate cause(s)
of leaks in t e nit component coo ing water system.
A lengthy licensee evaluation was unable to pinpoint.a single cause.
Either chemical or biological attack could have occurred.
The known leaks were repaired; an enhanced process of inspection for possible development of future problems was implemented, and as a preventive measure, the system was chemically treated in a manner to address both possible sources of attack.
(Closed)
Unresolved Item (315/87029-01):
Review actions to prevent recurrence o
fai ure to per orm as ound" Type C leak test on a containment isolation valve before doing maintenance.
A review of licensee actions (as documented in his Problem Report No. 87-0846)
was conducted.
Effectiveness of preventive actions was demonstrated by lack of recurrence of this proble e.
(Closed)
Unresolved Item (315/87029-02):
Review actions to prevent recurrence o
as ure o rep ace emergency diesel control circuit fuses after maintenance.
A review of corrective actions documented under licensee Problem Report No. 87-0614 showed the actions were effective in preventing recurrence.
No violations, deviations, unresolved or open items were identified.
3.
0 erational Safet Verification (71707, 71710, 42700)
Routine facility operating activities were observed as conducted in the plant and from the main control rooms.
Plant startup, steady power operation, plant shutdown, and system(s)
lineup and operation were observed, as applicable.
The performance of licensed Reactor Operators and Senior Reactor Operators, of Shift Technical Advisors, 'and of auxiliary equipment operators was observed and evaluated including:
procedure use and adherence, records and logs, communications, shift/duty turnover, and the degree of professionalism of control room activities.
Evaluation, corrective action, and response for off normal conditions or events, if any, were examined; This included compliance to any reporting requirements.
Observations of the control room monitors, indicators, and recorders were made to verify the operability of emergency systems, radiation monitoring systems and nuclear reactor protection systems, as applicable.
Reviews of surveillance, equipment condition, and tagout logs were conducted.
Proper return to service of selected components was verified.
a.
Unit 1 was started up on the first day of the inspection period (some 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after an unexpected reactor trip the previous day)
and operated routinely throughout the remainder of the inspection period.
b.
Unit 2 remained in its scheduled, prolonged steam generator repair outage (see Paragraph 10) throughout the inspection period.
C.
On October 24, 1988, with Unit 1 in routine operation at 90-percent power, the incore flux thimble leak detection system indicated onset of a thimble tube leak.
An inspection inside containment confirmed the leak and identified the leaking tube as Tube Jl.
At the time of discovery, Tube Jl was being used to store one of the flux detectors.
After evaluation of available options, the licensee decided to isolate the leak (which remained quite small - about 0.02 gpm) by closing the tube isolation valve with the detector cable still inserted.
This action, which was accomplished that same day, October 24, eliminated the possibility of a large-scale leak.
The tube was subsequently
cut and capped to completely eliminate the leak and to permit withdrawal of the severed detector cable, so as to regain use of the associated 10-path transfer device for other detectors.
d.
The Unit 1 startup of October 20 was specifically observed by the inspector for degree of communication and coordination among operators, adherence to applicable procedures, and general degree of professionalism.
Senior plant management was also present in the main control room throughout startup - a routine practice.
All activities observed by the inspector, were done carefully, smoothly, and effectively.
e.
The following procedures were selected for a partial review:
(i)
1-OHP 4024.107
"Annunciator No.
7 Response
- Reactor Coolant," Revision 3 dated May 8, 1984 through Change Sheet 8 dated June 4, 1987.
The inspector noted references to Procedure OHP 4022.001.001 for reactor trip response (for Annunciators 009 and 010) - the current correct reference is Procedure E.O.
A control operator initiated a procedure change request on the spot.
(ii) 1-OHP 4021.001.002
"Determination of Reactor Shutdown Margin,"
Revision.10 dated August 1, 1984.
This procedure review was limited to evaluation of how reactor coolant system (RCS)
temperature transients following a reactor trip might be addressed.
The Unit 1 reactor trip of October 19, 1988, resulted in a slight overcooling of the RCS compared to the hot no-load setpoint of 547 degrees F.
Isolation of certain steam and/or drain lines by operators was less timely than usual, and contributed to a temperature decrease to about 536 degrees F.
Recently, NRC has expressed concern about
"typical" overcooling of 25 degrees F or more at another plant, because shutdown margin is reduced and newer, higher enriched cores (designed for extending fuel cycles)
have stronger reactivity coefficients.
The subject procedure states, at Step 6.3.2, that sufficient conservatism exists in the shutdown margin calculation that the value determined for 547 degrees F
may be considered valid down to 530 degrees F, but a new calculation would be required below 530 degrees F.
It was not determined if the conservatisms are reverified for each new core load.
This was discussed with plant management and at the Management Interview.
No violations, deviations, unresolved or open items were identified.
4.
Radiolo ical Controls (71707)
(83729)
During routine tours of radiologically controlled plant facilities or areas, the inspector observed occupatibnal radiation safety practices by the radiation protection staff and other worker Effluent releases were routinely checked, including examination of on-line recorder traces and proper operation of automatic monitoring equipment.
Independent surveys were performed in various radiologically controlled areas.
On one occasion the inspector observed an individual working in anti-contamination clothing, and the protective hood was open at the neck.
A radiation protection technician, making the same observation without prompting from the inspector, instructed the worker to seal up the hood properly.
Overall, the Unit 1 steam generator repair project has continued nearly on schedule and below estimated radiological dose commitments.
The need to repair welds on the newly installed steam generators (see Inspection Report No. 316/88024)
has resulted in a final dose total for welding activities (performed under the RWPs written for installation of the steam generators)
that is 2.5 to three times higher than the estimated total; however, the final dose total of approximately 200 person-rem for all activities performed under the welding activity RWPs is well-below the original 1986 estimate of 520 person-rem, and below the 1988 revised estimate of 245 person-rem.
For the'ntire project, as of November 7, 1988, approximately 714 person-rem had been accrued, compared to the revised estimate of 982 person-rem (dose totals are based on pocket ion chamber dosimeter readings).
The licensee has reduced the estimated final dose-total for the project from 1733 to 1031 person-rem.
The inspectors also reviewed the licensee's evaluation of circumstances where a radiographer, working under NRC Byproduct Naterials License No. 34-24757-01, received approximately 400 mrem more than another radiographer and an RP technician during radiography of a steam generator cold leg weld.
The 400-mrem dose was received during the time period from just prior to the last radiograph of that particular job, when the subject radiographer's pocket dosimeter was read, and 20 minutes later when the subject radiographer exited containment and his dosimeter was read again.
(After completion of the last radiograph, the subject radiographer left the presence of the other radiographer and the RP technician and attended to other duties before exiting containment.)
Licensee surveys of the areas in containment in which the subject radiographer worked after leaving the other individuals did not identify a source of radiation of sufficient intensity to account for the 400-mrem exposure in the 20-minute time period.
Therefore, the licensee concluded that the exposure occurred during the last radiograph; the cause was not determined.
The radiography licensee also evaluated the radiographer's exposure and likewise did not determine a specific cause.
The subject radiographer's quarterly exposure total did not exceed quarterly limits.
No problems were identified with the licensee s
(D.C. Cook) evaluation of the incident.
The NRC Region III Byproduct Haterials inspection staff has been apprised of the exposure and will followup with the radiography licensee as warranted.
No deviations or violations were identified.
No violations, deviations, unresolved or open items were identified.
l
5.
Maintenance (62703, 42700, 71714)
Maintenance activities in the plant were routinely inspected, including both corrective maintenance (repairs)
and preventive maintenance.
Mechanical, electrical, and instrument and control group maintenance activities were included as available.
The focus of the inspection was to assure the maintenance activities reviewed were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with Technical Specifications.
The following items were considered during this review:
the Limiting Conditions for Operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures; and post maintenance testing was performed as applicable.
The following activities were inspected:
a ~
Job Order No. 761533:
Preventive maintenance on 1E centrifugal charging pump CCP to include changing oil, lubricating coupling, cleaning vent screens, and inspecting the gearbox.
Radiation Work Permit (RWP)
2089 governed radiological controls on this job.
No procedures were in use at the jobsite.
When the inspector pursued this, it was found no procedures were required.
The oil change and lubrication are straight forward matters governed by a "lubrication card" which is not required to be present during the work.
The gearbox inspection is somewhat more complex, involving as it does the taking of drive pin and drive hole diameter measurements and visually inspecting for cracks, uneven wear, etc.
The workmen knew the acceptance criteria for the micrometry without reference to instructions, so the inspector ultimately concluded having written instructions at the job site was not necessary.
b.
Job Order No. 735103:
Clean/replace 1E CCP bearing oil sight glass-1077
.
is was conducted concurrently with the preventive maintenance conducted under item a.
above.
This showed prior planning to accomplish multiple repairs during one "outage" of this safety-related equipment.
C.
During observation of activities described in a.
and b. above, the inspector noted Instrument 1-LPI-270 (1E charging pump lube oil filter inlet pressure indicator)
was not the original gauge.
It was smaller in size and was not attached to or supported by the mounting bracket sized for the larger original.
The Facility Data-Base section had a photo dated October 1984 showing the larger gauge was in place at that time.
The maintenance history files of the Instrument and Control (I&C)
group indicated the original 1-LPI-270 (an Ashcroft 0-30 psi pressure gauge)
was replaced under Job Order 115028 on January 9,
1987.
A Moore Brand 0-30 psi gauge was installed instea Job Order 115028 had been written nearly a year earlier (February 7,
1986)
on a Job Order form which was superseded in about mid-1986.
The Job Order was rewritten from the old form onto a
new one (numbered 009094)
on October 10, 1986, but the old form was not disposed of and, in fact, was used in the "repair" involving installation of a different gauge as noted.
The work crew documented the work done as follows: "installed 30 lb. gauge but not mounted to bracket-need back input Ashcroft gauge."
This information appears to have been disregarded thereafter, as another Job Order (714118)
was written (April 27, 1987)
and executed (June 17 and July 18, 1987) to repair and calibrate the Moore brand gauge; then Job Order 009094 was closed (August 4, 1987)
on the basis the gauge had already been repaired.
As a consequence of this series of actions, the Moore gauge'emained installed but not mounted to the support bracket, and the records were in error - showing an Ashcroft gauge.
The function and potential failure of the subject gauge are not described in the FSAR. Thus, changing gauges is not significant in the senses of concern to 10 CFR 50.59; it creates no new or increased accident potentials and affects no Technical Specifications bases.
This means no Violation of 10 CFR 50.59 occurred.
Other gauge substitutions, however, could be more significant.
Current procedural requirements restrict repairs under a Job Order to a specified scope, and would prohibit substitution of a non-identical part unless design change evaluations are done.
Thus, while the intent of procedural controls for Job Orders and design changes may have been violated, subsequent actions on the part of the licensee have served to prevent a recurrence.
The only remaining action is to reconcile the records and the actual installation.
This was discussed at the Management Interview.
Job Order No. 705838:
In the Unit 2 4KV switchgear room and trans ormer areas, install duct work, hangers, supports and dampers, and remove existing duct as required for design change RFC 12-2868.
The engineering review performed prior to the start of the job appear ed weak as evidenced by the number of design change deviation requests that had to be submitted after work was begun.
The most persistent problem dealt with hangers that coul'dn'-t be installed as the original design drawings called for due to'nterferences or the configuration of the room.
The same design change is planned for the Unit 1 swi tchgear room in the future.
Job Order No. 705868:
In the Unit 2 4KV switchgear room,* remove ca e No. 8221R-2 from conduit and install cable No. 8445R-2 for design change RFC 12-2868.
Cable is to feed MCC-2-E7C-A breaker No. 21A9.
The new cable has a higher. current rating to accommodate the electrical modifications made during the outage.
8,
f.
The inspector reviewed Procedure MHI-5030 "Preventive Maintenance and Environmental gualification Program," Revision 12, dated July I, 1988.
The review was confined to a brief overview, but included a specific look at Task Sheet No. 30, "Plant Winterization and Dewinterization."
General winterizing tasks normally are accomplished by October 31, of each year; however, this year the date has been deferred to November 30 because the licensee is manufacturing better covers used for the various enclosures.
The tasks themselves are completed using data sheets which record inf'ormation such as building or enclosure name, person performing the winterization, date performed, and a remarks section.
The inspector reviewed these documents for their completeness and found no problems.
No violations, deviations, unresolved or open items were identified.
6.
Surveillance 61726, 42700 The inspector reviewed Technical Specifications required surveillance testing as described below and verified that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that Limiting Conditions for Operation were met, that removal and restoration of the affected components were properly accomplished, that test results conformed with Technical Specifications and procedure requirements and were reviewed by personnel other than the individual di recting the test, and that deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.
The following activities were inspected:
a.
- 2 IHP 6030 IMP.256 "Reactor Coolant Wide Range Temperature Calibration," Revision 4 dated October 7, 1988.
b.
- 1 THP 4030 STP.018
"Steam Generator Water Level Protection Set IV Surveillance Test - Monthly," Revision 6 dated December 13, 1985.
c.
- I IHP 4030 STP.032
"Reactor Coolant Pump No. I Undervoltage Bus 1B Surveillance Test (Monthly)."
d.
- 1 IHP 4030 STP.033
"Reactor Coolant Pump No.
2 Undervoltage Bus 1C Surveillance Test (Monthly)."
e.
- 1 IHP 4030 STP.034
"Reactor Coolant Pump No.
3 Undervoltage Bus 1D Surveillance Test (Monthly)."
f.
- 1 IHP 4030 STP.035
"Reactor Coolant Pump No. 4 Undervoltage Bus 1A Surveillance Test (Monthly)."
g.
- I THP 4030 STP.018."Steam Generator Water Level Protection Set IV Surveillance Test (Monthly)."
h.
On October 27, 1988, the inspector attended a briefing on a proposed new methodology for accomplishing nuclear instrument incore/excore calibrations.
A standing commitment existed on the part of the licensee to apprise the resident inspector of changes to the existing methodology (embodied in licensee Procedure
- 12 THP 4030 STP.362,
"Incore/Excore Detector Calibration" ) from the time this methodology was developed and implemented in about 1979.
Recently, a new technique has been adopted by some PMR licensees.
At D.C. Cook, the licensee has developed an additional procedure which he intends to utilize for routine incore/excore calibrations.
The existing procedure will not be eliminated but will be preserved for use as needed.
The new methodology was explained and data provided to demonstrate the high degree of consistency in results obtained using the different techniques.
Some advantages of the new approach were described.
Inspector questions were addressed.
The licensee supplied selected data and studies, along with technical and safety evaluations, for inspector review.
In addition, some technical review was performed by an NRC Region III reactor physics inspector.
No Technical Specification or Safety Analysis changes appear to be involved, and the inspector identified no concerns or issues arguing against licensee implementation.
No violations, deviations, unresolved or open items were identified.
7.
Fire Protection (71707, 64704)
Fire protection program activities, including fire prevention and other activities associated with maintaining capability for early detection and suppression of postulated fires, were examined.
Plant cleanliness, with a focus on control of combustibles and on maintaining continuous ready access to fire fighting equipment and materials, was included in the items evaluated.
During or just preceding the inspection period, several problems were encounter ed in fire protection -activities associated with the Unit 2 steam generator repair project (SGRP).
Some of these problems were documented for corrective action in the following Condition/Problem Reports:
Problem Report 88-810 (11/7/88)
SGRP Fire - arc gouging sparks ignited oxy-cyclene hose and electrical cables, damaging four. cables.
Problem Report 88-769 (10/26/88)
Firewatch on SGRP welding activity in crane bay had no operable extinguisher.
Problem Report 88-757 (10/22/88) Fire reported at No.
22 steam generator.
Problem Report 88-746 (10/16/88)
Two firewatch fire extinguishers found empty when use was attempted in fighting small SGRP fire (cadwelding sparks fell about 30 feet off the top of a steam generator enclosure and ignited a filter).
,
In addition, the inspector observed a welding activity involving attachment of a Unistrut for clamping a couple of small instrument lines from the No.
21 steam generator, and no firewatch appeared to be present.
This occurred at about 11:00 a.m.
on October 28, 1988.
Due to the congestion in the area from multiple temporary scaffolds, the sight lines were limited, so the inspector climbed directly to the work area on the east side of the steam generator some ten feet below the girth region.
When questioned as to the whereabouts of his firewatch, the welder said, in effect ".
.
. there should be one around here somewhere,"
but no one was in sight.
The welder stopped work and left the area.
The inspector subsequently pursued additional information via SGRP and Plant management discussions and procedure review.
SGRP Safety Procedure SP-03.00
"Hot Work Permits" states firewatches are to be designated on the hot work permit, they shall have fire extinguishers in their possession, and they must remain in the work area 30 minutes after completion of the operation.
On October 28, 1988, hot work permit No.
1247 for day shift welding in the 658-foot elevation on steam generator No.
specified a fire watch.
Inasmuch as neither a firewatch nor extinguisher were apparent at the time the inspector observed hot work in progress, requirements of Procedure SP-03.00 were violated.
In that implementation of fire protection program procedures is a requirement of Technical Specification 6.8.1.f., this is considered a Violation of the Technical Speci fication (316/88028-01).
One violation and no deviations, unresolved or open items were identified.
Emer enc Pre aredness (82201, 82203)
On November 16, 1988, at about for the steam generator repair and was injured while going to the site indicated the man was Thus, the licensee declared an (EDT).
A subsequent survey at contaminated, and the "Unusual 5:30 a.m.
(EDT), with Unit 2 shutdown project, a licensee contractor slipped work on the Ice Condenser.
A frisk at contaminated with approximately 150 cpm.
"Unusual Event" at 6:09 a.m.
the hospital revealed the man was not Event" was terminated at 7: 18 a.m.
(EDT).
Licensee detection, classification and notification on this event appeared proper.
No violations, deviations, unresolved or open items were identified.
Securit (71707)
Routine facility security measures, including control of access for vehicles, packages and personnel, were observed.
Performance of dedicated physical security equipment was verified during inspections in various plant areas.
The activities of the professional security force in maintaining facility security protection were occasionally examined or r'eviewed, and interviews were occasionally conducted with security force member On October 25, 1988, the inspector met with representatives of the Security Department for a briefing and discussion on an October 20, practice exercise.
The Federal Bureau of Investigation, the licensee, and State and local law enforcement agencies participated in an annual FBI regional drill, responding to a scenario involving armed terrorists, hostages, and threats to plant personnel and property.
More than a
hundred non-licensee participants, including FBI agents from three major metropolitan offices (Detroit, Chicago and Milwaukee) participated, while still others observed.
A non-safeguards information video tape is being assembled from taping done during the all day exercise.
No violations, deviations, unresolved or open items were identified.
Outa es (37701)
The following significant activities involving the Unit 2 steam generator repair project, a number of which were. observed by the inspector, occurred during this inspection:
a.
complete concrete pours - walls and roofs b.
complete post-weld heat treatment of girth welds c.
set main steam pipe whip restraints d.
complete plant ISI activities e.
complete steam dome internal modifications f.
replace primary and secondary system insulation - ongoing g.
set and shim lower generator lateral restraints h.
remove concrete form work fabricate/install small bore pipe restraints
- ongoing j.
reinstall upper containment ventilation units - ongoing k.
hydro-test all steam generators
- secondary side The project has continued to undergo periodic specialized inspections by NRC Region III personnel, in several disciplines.
These inspections are documented separately.
No violations, deviations, unresolved or open items were identified.
Safet Assessment/Oualit Verification (37701, 38702, 40704)
The effectiveness of management controls, verification and oversight activities, in the conduct of jobs observed during this inspection, was evaluated.
The inspector frequently attended management and supervisory meetings involving plant status and plans and focusing on proper co-ordination among Departments.
The results of licensee auditing and corrective action programs were routinely monitored by attendance at Problem Assessment Group (PAG)
meetings and by review of Condition Reports, Problem Reports, Radiological
Deficiency Reports, and security incident reports.
As applicable, corrective action program documents were forwarded to NRC Region III technical specialists for information and possible followup evaluation.
No violations, deviations, unresolved or open items were identified.
12.
Al 1 e ation Fol 1 owu (99014)
Discussed below are several anonymous allegations received by the NRC Region III Office, related to the radiation protection program at D.C. Cook, which were evaluated during this inspection.
The inspection consisted of record and procedure review, job site observations, and discussions with licensee employees.
a ~
Alle ation No. RIII-88-A-108 (0 en)
fly
~AII I: lb g
p fddby I
Pf I
p staff during modification of the Unit 2 steam generator upper assemblies was inadequate.
Discussion:
In late July and August 1988, repairs and dnl d <<1
pp bl1 drums or domes)
of the Unit 2 steam generators in conjunction with replacement of the remainder of the steam generators.
The work was performed in the Unit 2 turbine building trackway.
Based on review of licensee survey records, contact dose rates measured on the domes were generally less than 0.2 mR/hr; removable contamination was less than 1000 dpm/100 cm',
and fixed contamination ranged from less than 100 cpm to 2000 cpm per pancake probe area.
These radiological conditions are quite benign, requiring only minimal radiological precautions.
Reviews of the RP log maintained at the job site and other licensee records, discussions with licensee employees, and observations of the job by the inspectors indicated that RP job coverage appeared to be adequate as noted in a previous NRC inspection (see Inspection Reports No. 315/88020; 316/88023).
The work on the domes did not result in any significant external doses, skin or clothing contaminations, or intakes of radioactive material, as expected due to the minimal hazard.
There were instances of protective clothing shortages, an inoperable PCM-18 whole-body contamination monitor (because of high background),
and malfunctioning portable ventilation equipment; however, these problems were short-term, apparently resulted only in job completion delays, and indicated planning weaknesses rather than inadequate RP coverage.
~Findin:
The allegation that radiation protection job coverage was inadequate was not substantiated.
No deviations or violations were identified.
(2)
Alle ation:
."Dirty" protective clothing was reused during work on t e nest 2 steam generator upper assemblies due to lack of
"clean" protective clothing.
Discussion:
A review of records and discussions with licensee representatives responsible for providing protective clothing indicated that on several occasions during the steam drum work, the supply of various items of protective clothing, e.g., cloth coveralls, rubber gloves, and rubber footwear, was exhausted, but that the only item reused without being laundered was rubber outer footwear.
Report'edly, when the supply of other articles of protective clothing was depleted, work was halted until the supply was replenished.
The occasional reuse of potentially contaminated footwear under these circumstances appears acceptable.
As noted above the radiological conditions associated with the work did not demand stringent radiological controls.
(3)
Findin
Although the allegation that "dirty" protective c ot sng (i.e., protective clothing has not been laundered to remove loose contamination)
was reused due to lack of "clean" clothing was partially substantiated in that potentially contaminated outer footwear was reused, no deviations or violations were identified, nor were the licensee's radiological controls inadequate.
Alle ation:
Frisking and dressing procedures were not followed urging t e overhaul of the Unit 2 steam generator upper as semb 1 ies.
Discussion:
No problems were identified with worker adherence
11 d
dg the job by NRC inspectors.
As stated above, protective clothing shortages occurred on several occasions during the job; however, as best could be determined, work was either stopped or, in the case of the shortage of rubber outer footwear, unlaundered footwear was reused.
Even if there were instances when dress requirements were not fully met, the consequences would have been insignificant because contamination on the domes was very low-level and protective clothing requirements for the job were conservative.
Concerning frisking, some weaknesses were observed by an NRC resident inspector and a licensee quality assurance inspector in that individuals were observed performing only cursory frisks for personal contamination (see Inspection Reports No. 315/88020; 316/88023).
(The RP technician assigned to the job was made aware of the poor surveys and directed the workers to resurvey themselves.)
However, these problems appear to have been isolated inasmuch as no other examples of inadequate personal frisking were identified.
As noted above, the PCN-18 whole
.
b.
Alle body contamination monitor installed at the job site was occasionally inoperable for brief periods of time because of high background; however, hand-held friskers were available to perform personal contamination frisks during these periods.
(The licensee eventually fabricated a metal-shielded booth for the PCM-1B to reduce background and allow its use.)
Findin
While significant deviations from frisking and ressing procedures were not identified, some minor problems did occur.
However, such minor problems are practically unavoidable, and radiological conditions did not demand more than rudimentary radiological controls.
The radiological controls instituted were more than adequate for the existing radiological conditions.
No deviations or violations were identified.
ation No. RIII-88-A-129 (Closed)
~kgg 1:
Wld kl A Ul
main coolant pipe connections were not wearing dosimetry in an attempt to lower their dose records.
Supervision is aware of and encouraging the practice.
.Discussion:
Melding of the new steam generator lower assemblies lp
1 lyA d
d into October of 1988.
Host of the welding was performed using a track-mounted, automated welding machine (diametric welder).
Typically, two welders were assigned to operate the diametric welder, working two hours per entry (or "dive") into the lower assembly-coolant pipe controlled area.
After two hours, two other welders would replace the first group, etc.
Dosimetry requirements for the welders usually included a low and a high range pocket ion chamber dosimeter (SRD)
and a Panasonic whole-body TLD.
General work area dose rates were variable, ranging from approximately 10 to 130 mrem/hr.
Dose rates in areas around steam generator cold legs were higher than those around the hot legs because of "crud" in the resistance temperature detectors (RTDs).
Enclosures were constructed surrounding the welding sites to localize airborne radioactivity and surface contamination from the operation.
Viewing windows existed to allow observation of the welders, from outside the enclosures.
Radiation protection coverage was continuous during the welding, with technicians positioned in the staging area between steam generators one and four and'between steam generators two and three.
After diametric welding, non-destructive examination was conducted on the welds.
If indications of imperfections were observed, welders using manual welding equipment performed repairs (see Inspection Report No. 316/88024 for additional discussion on the overall quality of the welding on the Unit 2 steam generators).
NRC resident inspectors had reviewed welding activities at the job site routinely before receipt of this allegation.
After receipt they individually verified that welders were wearing the prescribed dosimetry.
According to licensee representatives, throughout the removal of the old steam generator lower assemblies and the installation of the new assemblies, radiation protection (RP) technicians assigned to the job periodically read workers'ocket dosimeters, and that all workers in containment were, by policy, subject to random checks by RP personnel to verify'ompliance with RWP dosimetry and PC requirements.
Licensee representatives reported no instances of welders working on the steam generator-coolant pipe job without the required dosimetry.
Since early September, access control RP technicians read SRDs of workers entering and leaving the main radiologically controlled access point for containment, in an attempt to reduce discrepancies between TLD and SRD readings whereby the SRD readings were unreasonably higher than the TLD readings.
Such discrepancies are not'ncommon, typically resulting from overly conservative reading of the SRDs by workers.
(See Inspection Reports No. 315/88019; 316/88022, for additional discussion of these discrepancies.)
The inspectors interviewed several welders, welding foremen, and the contractor superintendent responsible for the welding activity.
None of the individuals admitted knowledge of welders not wearing dosimetry while welding, nor of hearing statements made by welding foremen or the superintendent that directed workers to, or implied that workers should, keep recorded doses as low as possible re ardless of how it was accomplished.
.Welding supervisors a so enied making any such statements.
The welding superintendent maintained records, which were reviewed by an NRC inspector listing daily dose information for the welders.
The superintendent explained that because of the strong emphasis by the project ALARA group to keep dose evenly spread among welders and to preclude exceeding administrative dose limits, he kept track of welders'aily dose totals and directed the foremen to schedule work accordingly.
He also stated that on several occasions he noticed that certain workers began to report noticeably lower dose than others.
He stated that he spoke to the individuals and determined that these workers had become proficient enough with the diametric welding equipment to allow them to leave the immediate welding area for a lower dose area once the equipment was set up and running smoothly, unlike other welders with less experience or skill in use of the equipment who would spend more time in the higher dose rate welding area.
RP technicians assigned to the welding job confirmed that some, welders spent more time in the low dose rate area away from the welding equipment than other welders.
The superintendent added that welder daily dose totals also varied because of the variability in dose rates between hot and cold legs on any one steam generator, the variability in dose rates among steam generators,
and the unpredictable but not infrequent operability problems with the diametric welding equipment.
The foremen interviewed by the inspector confirmed the superintendent's statement about the dose tracking and worker scheduling method and the factors affecting worker dose totals.
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~
~
~
Findin:
The allegation that welders, with the tacit approval of t eir supervisors, were not wearing dosimetry in an attempt to lower their dose records could not be substantiated.
It is recognized that with the strong emphasis on ALARA by project management, there were both personal and contractor incentives to maintain doses low, and that by not wearing their dosimetry workers could make it appear that their doses were being maintained lower than they actually were.
However, observations of welders at the job site by NRC inspectors and licensee radiation protection personnel and the extent of the administrative controls implemented by the licensee's radiation protection group make it improbable that such a practice existed.
The licensee's explanation for different dose totals reported by individual welder s performing similar work is plausible.
The inspector found no information to support the claim that welding supervisors suggested to workers that the workers should violate dosimetry requirements.
No deviations or violations were identified.
Alle ation No. RIII-88-A-0053 (0 en)
(1)
Alle ation:
Water level in the steam generator was not owere as required by the radiation work permit during outage maintenance activities in 1987.
Discussion:
This allegation was originally referred to the Thi bi<<d<<h NRC in a letter dated September 9, 1988, and indicated that they could not substantiate the allegation.
A followup review was conducted by an NRC inspector.
Discussions with licensee personnel and a review of records showed that the requirement to lower steam generator water level for steam generator maintenance has not been specified on radiation work permits for at least four years; rather the requirement is specified in the maintenance procedure for removing the steam generator manway covers and diaphragms.
According to the licensee's September 9,
1988 response and discussions with licensee workers the water level had been lowered as required by the maintenance procedures prior to maintenance on the steam generators in 1987. It was noted, however, that in August 1985, the water had not been completely drained from a steam generator prior to the start of maintenance
activities.
Approximately 3000-5000 gallons of water spilled out of the steam generator as a diaphragm was removed.
NRC resident inspectors reviewed the incident and the corrective actions at the time of the incident.
The licensee's corrective actions apparently have been adequate to prevent recurrence.
Findin
The allegation was not substantiated.
No deviations
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or vio ations were identified.
No violations, deviations, unresolved or open items were identified.
13.
Re ion III Re uests (71707)
a
~
b.
By letter dated August 22, 1988, from E.
G. Greenman, Director, Division of Reactor Projects, the inspector was requested to provide information to establish whether there is a high temperature problem in the plant containment structure.
The inspection guidance is contained in the NRC Inspection Manual Temporary Instruction (TI) 2515/98, dated June 30, 1988.
In order to assess the degree to which the temperatures measured by the licensee inside containment represent actual conditions, the inspector gathered information such as summer peak operating temperatures, temperatures used to calculate equipment environmental qualification lifetime, technical specification 1'imits, design limits specified in the Updated Safety Analysis Report, and the location of containment temperature recorder sensors.
The inspector concluded that average temperature readings taken in the upper and lower containment were fairly representative of actual ambient conditions, except for those temperature readings in the upper volumes of the pressurizer and the four steam generator
"doghouses."
Temperatures in those areas run higher than average, and would be called "local hot spots."
The Plant Systems Branch of the Office of Nuclear Reactor Regulation will review the information further.
Inspection review under Temporary Instruction 2515/98 is considered closed.
By memorandum dated November 2, 1988, from E.
G. Greenman, Director, Division of Reactor Projects, the inspector was requested to discuss issues relating to use of sealing fluids on primary pressure boundary components with licensee management.
This discussion was held November 9, 1988.
In addition, the inspector reviewed the licensee's procedure (**12 MHP 5021.001.051
"Leak Sealing of Plant Components in Pressurized Piping Systems by Furmanite Injection" Revision
dated January 22, 1987) against the concerns identified in an attachment to the referenced memorandum.
This review showed the licensee had not addressed all of the concerns in his procedure.
When this was discussed further with the Maintenance Engineer, he provided an information copy of proposed Revision 3, which was in
the review and approval process.
All of the concerns for proper control of this type activity are well covered in the new Revision.
The Maintenance Engineer indicated the additional controls have been in practice, under a Department memorandum, since March 1988.
No violations, deviations, unresolved or open items were identified.
14.
Mana ement Interview (30703)
The inspectors met with licensee representatives (denoted in Paragraph 1)
on November 17, 1988, to discuss the scope and findings of the inspection.
In addition, the inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection.
The licensee did not identify any such documents/processes as proprietary.
The following items were specifically discussed:
a.
The apparent violation, and other examples of weaknesses in the steam generator repair project fire protection (Paragraph 7);
b.
the need to evaluate procedural shutdown margin bounds for potential post-trip overcooling with each new core load/cycle (Paragraph 3);
c.
the need to reconcile the "as installed" configuration on Gauge 1-LPI-270 with applicable records (Paragraph 5);
d.
the results of NRC review of certain allegations concerning radiological work practices (Paragraph 12).
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