IR 05000445/1983034
| ML20198H914 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 09/10/1983 |
| From: | Hunnicutt D, William Jones, Kelley D, Madsen G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20198H622 | List: |
| References | |
| FOIA-85-59 50-445-83-34-01, 50-445-83-34-1, 50-446-83-18, NUDOCS 8605300594 | |
| Download: ML20198H914 (5) | |
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APPENDIX U. S. NUCLEAR REGULATORY COMMISSION
REGION IV
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NRC Inspection Report:
50-445/83-34 50-446/83-18
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Dockets: 50-445; 50-446 Construction Permits: CPPR-126 CPPR-127
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Licensee: Texas Utilities Generating Company (TUGCO)
2001 Bryan Tower Dallas, Texas 75201
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Facility Name:
Comanche Peak Steam Electric Station (CPSES) Units 1 and 2 Inspection At: Glen Rose, Texas
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Inspection Conducted: August 12-19, 1983
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Inspectors:
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e-e-D. M. Hunnicutt, Chief Dgte '
Reactor Project Section A
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9hoffI D. L. Kelley, Senior Resident Inspector-Operations Date f
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'W. B. Jones, Engineering Aide Date
Approved:
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' G. L. Macsen, Chief Date Reactor Project Branch 1 F01A-85-59 8605300594 860513 g
GARDE 85-59 PDR
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i Insoection Summary
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f Inspection Conducted Aucust 12-19, 1983 (Report 50-445/83-34: 50-446/83-18)
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Areas Inspected: Special, unannounced inspection of (1) detailed review of nonconformance reports for the period October 12, 1981, through October 12,
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1982; (2) inspection to determine that the reactor vessel outer wall did not contact the containment vessel shield wall at any point; and (3) determining
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that no " secret meetings" related to the reactor vessel wall contacting the containment vessel shield wall at any point had occurred. The inspection
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involved 14 inspector-hours onsite by three NRC inspectors.
Results: Within the three areas inspected, no violations or deviations were
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identified.
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Details
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1.
Persons Contacted Principal Licensee personnel
"J. T. Merritt, Manager of Startup
"R. G. Tolson, Site Quality Assurance Supervisor
' Denotes ext interview on August 19, 1983.
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" Denotes exi interview on August 17, 1983.
Review of Nonconformance Reoorts (NCR)
2.
An allegation was received by the NRC that the dismissal of(Mr. Atch
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g, g; may hav~e had a negative (chilling) affect on the preparation and/orT issurance of NCR's at CPSES.
all NCR's for the period between October 12, 1981, and October 12, 1982.
This review was performed to determine whether there w 12, 1982. The been affected by the dismissal of Mr. Atchinson7en April
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7 c c 12, 1981, NRC inspector reviewed the NCR's between the dates of October 12, 1982.
and April 12, 1982, and from April 12, 1982, through October All NCR's issued (six were issued) on April 12, 1982, were purposely The results obtained omitted from this review to remove any data bias.
from this review are summarized below:
NCR's Issued in Non ASME Areas October 12, 1981, through April 11, 1982 = 637 April 13, 1982, through October 12, 1982 = 1342 There was an increase of about 110 percent (210 percent of the prior
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6 months) in the NCR's issued during the subsequent 6 month period.
NCR's Issued in ASME Areas There was an increase of about 70 percent (170 percent of the prior 6 months) in the NCR's issued during the subsequent 6 month period.
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l Due to the large increase in the nt.mber of NCR's (both in the ASME areas and in the non-ASME areas) in the 6 months immediately following Mr. Atchinson's dismissal, there appears to be no negative (chilling)
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ffect on the issuance of NCR's at CPSES due to the dismissal of
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Mr.Atchinson]
The NRC inspector found during this review that the CPSES issued NCR's had been issued in all areas related to CPSES construction and for the sam general and specific activities (deficient areas of items) both prior to l
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and subsequent to the dismissal of
)nApril 12, 1982.
[f 7(c Construction activities and the ndsber o cons ruction type employees remained approximately at the same level during this 1 yeartperiod (October 1981 through October 1982).
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This inspection was perf part a result of a letter dated July 14, 1983, fromLHr.
o Mr. Peter B. Block, Chairman. [, 7 (e ASLB, in the matter of TUGCO, et. a.
SES, Units 1 and 2).
The allegation was found to be unsubstantiated and without merit.
3.
Reactor Vessel Outer Wall Clearance From the containment Vessel Shield Wall An allegation had been received that the reactor vessel outer wall had h )
been or was in contact with the containment vessel shield wall.
Two NRC inspectors visually inspected the Unit I reactor vessel / containment vessel for clearances between the two valls.
The Containment Building and all systems were at ambient temperature (less than 100*F).
The reactor vessel is covered with a mirror shield (this shield is approximately 6-inches thick and is insulation to reduce the loss of heat from the reactor vessel
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during operation).
The NRC inspectors visually inspected the periphery of the reactor vessel mirror shield and the inner surface of the containment vessel shield wall.
There appeared to be about 4 inches of clearance at all points on the reactor vessel outer wal.1 (covered with about 6 inches of mirror shield) and the inner surface'of the c,ontainment vessel shield wall. There were no visible points of contact nor any indications of previous contact. This visual inspection was accomplished by shining a light 6eam between the mirror shield and the shield wall and s
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reflected light from the reactor vessel hot and cold legs.
An NRC inspector repeated this inspection process on the Unit 2 reactor vessel mirror shield and the containment vessel shield wall.
These visual inspections revealed that the reactor vessel mirror shields for both reactor vessels (Units 1 and 2) were in place.
There were no areas of contact between the mirror shield and the containment vessel shield wall in either Units 1 or 2.
This allegation.was found to be without merit and was unsubstantiated.
4.
Allegation That a " Secret Meeting" Related to the Reactor Vessel Wall Contacted the Containment Vessel Shield Wall The Senior Resident Inspector-Operations (SRIO) contacted licensee manage-ment and requested information on " secret meeting (s)" related to a reactor vessel outer wall contacting the containment vessel shielding inner wall.
Licensee management denied that any " secret meeting (s)" had been held.
Licensee management stated that a meeting had been held that involved several corporate level personnel and that this meeting was not " secret" nor had any attempts been made to keep this meeting " secret." /Apparpntly
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the reactor vessel mirror shield did touch the shield wall during hot
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functional testing (HFT) and a related potential _10 CFR Part 50.55(e) _
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report was transmitted to the NRC Senior Resident Inspector-Construction.
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The subject of this potential 50.55(e) was that there was insufficient heat removal capacity in the reactor vessel cavity. This condition was identified during the seneduled HFT.
I The SRIO reviewed the_HFT Ing for any notation on shield wall / reactor (
vessel interface. There were no entries related to this specific subject in the log. There was a notation that _PT-45-06 " Containment Ventilation" failed to meet its acceptance criteria because the following areas were too hot (thermally):
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(1) _All vessel s,u_pports (2) Neutron Instrument Detector Wells (3) Pressurizer Room (4) All Steam Generator Compartments
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The licensee's management reported that Westinghouse (contractor personnel)
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is presently evaluating the heat removal problem that has been identified in the reactor vessel cavity -
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The allegation was found to be without merit and generally inaccurata.
However, the mirror shielding did touch the shield wall during the HFT.
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The occurrence was identified by the licensee and reported to the NRC and i
corrective action is in progress.
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5.
Exit Interviews Exit interviews were held on August 17 and 19,1983, with licensee repre-
sentatives (denoted in paragraph 1).
The NRC inspectors summarized the purpose and scope of the inspection and discussed the inspection findings.
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U. S. NUCLEAR REGULATORY COMMISSION i
OTFICE OF INSPECTION AND ENFORCIMENT
REGION IV
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Report No. 50-445/79-27; 50-446/79-26 Docket No. 50-445; 50-4'46 Category A2 l
Lictnsee: Texas Utilities Generating Company t
2001 Bryan Tower Dallas, Texas 75201
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Facility Name: Comanche Peak, Units 1 & 2 Inspection at: Cotznche Peak Steam Electric Station, Glen Rose. Texas Inspection conducted: November 1979 Inspector:
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p.J. Taylor,RasidentReactorInspector, Projects Section
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Approved:
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JD W. A. Crossman, Chief, Projects Section Data
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Inspection Summary:
h Inspection During November 1979 (Report No. 50-445/79-27; 50 446/79-26)_
Areas Inspected: Routine inspection by the Resident Reactor Inspector (RRI)
j of construction progress and practices; follow up on previously identified
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inspection findings; Quality Assurance procedures; electrical cable instal-j
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lation specifications and procedures; piping system supports; welding of
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reactor coolant and other safety-related piping systems; and a review of the verification program for veld quality in selected Class III piping systems.
The inspection involved eighty-three inspector-hours by one NRC inspector.
Of the seven areas inspected, no items of noncompliance or deviations Results:
p were identitified in five areas. One apparent item of noncompliance was identi-L fied in each of the two other areas (infraction - failure to revise obsolete h
Quality Assurance procedure 2 - paragraph 5; infraction - failure to follow procedures for hoisting safety-related components - paragraph 4.)
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DETAILS
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1.
Persons contacted
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Principal Licensee Employees
- R. C. Tolson, TUGCO, Site QA Supervisor
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- J. R. Merritt, TUSI, Construction and Engineering Manager
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- J. V. Bankins, Brown & Root, QC Supervisor
- J. I. Clarka, Brown & Root, Project QA Manager P. Van Teslaar, Westinghouse Nuclear Services Division, Site Manager
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N RRI also interviewed other licensee and Brown & Root employees during the inspection period.
- Denotes those persons with whom the RRI held co-site management meetings
during the inspection period.
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Action on Previous Inspection Findings (closed) Unresolved Ites (50-445/79-16): Electrical Cable Tray Support
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Fabrication-CB&I. The licenses notified RIV that this iten has been determined to be reportable under the Criteria of 10 CFR 50.55(e) and i
that he had submitted the required report dated November 14, 1979.
Data supportive to the report has been reviewed by the RRI and l
appears complete. h rework of the defective hangers will be inspected by the RRI and other NRC inspectors under the routine inspection programs.
N RRI had no further questions regarding this matter.
(closed) Unresolved Item (50-445/79-23; 50-446/79-22): Two Procedures
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for Operational Travelers. h licensee elected to delete CP-QP-2.3, Revision 0 in favor of the joint Construction-Quality Assurance Admin-I istrative Control Procedure CP-CPM-6.3.
N RRI had no further questions
on this matter.
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(Open) Deficiency (50-445/79-18): Failure to Control Inspection Stamps.
h licensee and Brown & Root use of numbered inspection stamps has been l
discontinued and the related procedure has been deleted from the QA/QC i
h licensee is in the process of reviewing various quality systeem.
records to determine if any of the lost and/or unaccounted for inspection stamps were used to document inspections during the period when controls were ineffective. h RRI will review the results of the document search
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during a future inspection, i
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3.
Site Tours The RRI toured the safsty-related plant areas several times weekly during the inspection period to observe the progress of construction and the general pract.ica4 involved. Due to a very limited scope of construction on the second shif t, where safety-related work was involved.
no inspection effort was devoted to the second shif t activities in this
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inspection period.
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i No items of noncompliance or deviations were identified during these general tours.
4.
Reactor Pressure Boundary Construction Activities
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h The RRI made several observations of the methods of h==A14=g and insealling various reactor pressure boundary components during the
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period. In most instances, the methods were consistent with good
industry practices, but in one other instance the RRI observed that i
the rigging used to hoist and position a large actor operated valve
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I appeared to be uncontrolled. The RRI discussed the rigging with the craf t labor person who appeared to be in charge relative to what
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instructions had been given to him. This person indicated that he had received no instructions and wished that he had some. The site
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" General Piping and Inspection Procedure" (CPM 6.9) requires that the =mnufacture's reconnendations be reflected in the installation
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instruction and further that any hoisting operation involving loads on the building shall of over 2000 pounds be referred to ensi-
nearing for review prior to making the hoist. Reference to the annufacturer's data indicated that the valve weighed in excess of 4000
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pounds.
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The RRI informed the' licensee of the situation which was quickly halted. The absence of any instructions to the craf t and the failure to follow the requirements of CPM 6.9 indicated to the RRI
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The RRI also conducted a short investigation into an allegation which
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was received by the RIV office on or about November 8, 1979, relating to the exact location of the Reactor Pressure Vessel in Unit 1. g l
alletariaa__ vas made by a former field engineer (surveyor) for Brown fr Root who 4=A4 e=ted that the vessel was located 3/10 inch to the west of the_
' north-south design centerline through the cone =1a==aq une ART.
ascertained from Westinghouse personnel that their requirements for i
i locating the vessel relate to azimuth and levelness with a secondary
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concern for elevation. The exact location in terms of the vessel
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centerline, in relation to the coneminment centerlina, was of little j
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or no concern to them, since the components attached to the vessel, via the piping, all have substantial adjustment capability. De RRI also interviewed the Brown & Root QC inspector who had been involved in
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the vessel installation, which occurred in mid-1978, and the current head
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of the field engineers who was then a general foreman in the same group and i
was in direct charge of the survey work. The inspector related that he had verified the location of the vessel from a provided fixture against established bench marks in the containment as required by the installation j
procedura. The engineer subsequently described to the RRI how the bench-marks had been derived. The method used should not have created an error
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amounting to 3/16 inch although either a human error in calculations or l
in measurement could conceivably have happened. The engineer also indicated that the party making the allegation (identified to him by the RRI) had not
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l been involved in the survey work relating to the vessel and could have had no first-hand knowledge of any survey errors.
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Based upon the interviews and upon first-hand knowledge of the reactor
coolant system installation, the RRI advised RIV that it was improbable
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hfYi-b that the vessel was actually mislocated, but that it if were, it would have no safety or operational consequence.
No further questions in the matter were raised either by the RRI or other-RIV personnel.
5.
Quality Assurance Procedures i
As noted in Inspection Report 50-445/79-18, the licensee has made sub-stantive changes in his sita (Nality Assurance organization and other lika changes have occurred in the Brown & Root organf u tion. As also g
noted in that report, the RRI was informed of each of these changes in j
advance and had no insediata concern since most of the changes appeared to enhance the overall effectivenass of QA/QC.
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During this inspection period it came to the attention of the RRI that
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the licensee and Brown & Root had failed to revise the organ 4ution
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control procedures to reflect the changes and that there were other procedures in the====1=
which assigned functional requirements to
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personnel by titles which no longer existed. The RRI identified at least nine procedures in the licenses and/or Brown & Root mamsels
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that were known to be obsolete for two or more months since they no longer represantad tha organization in place nor did they describe
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how certain activities were being accomplished in practica.
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The licenses was advised that the practice of = miring substantive changes
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without 1smediate addressnerc in appropriate procedures placed them in noncompliance with Criterien V of Appendix B to 10 CFR 50. The general condition and the nine procedures identified by the RRI were identified to the licensee in a Notir.e of Violation forwarded on November 21, 1979.
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Safety Related Electrical Cable Installation The licensee's cessation of work in this area continued throughout the
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inspection period. During the period, the REI reviewed a revised portion of the Project Electrical Installation Specification, ES-100, which deals with the actual installation efforts and with results to be achieved.
The basis for the RRI review was the various standards published by the Institute of Electrical and Electronic Engineers (IEEE) to which the licenses has coimitted to comply with in Chapters 7 and 8 of the FSAR.
The RRI also reviewed the production and Quality Control procedures for compatibility to the specification and to each other.
No items of noncompliance or deviations were identified.
7.
Piping System Supoorts The RRI observed the work related to making two modificiations required by doctamented engineering changes to hanger CI-1-097-404-C52R. The welder observed was determined to have been properly qualified for the work in accordance with ASME,Section II as was the welding procedure being followed. The RRI also reviewed the weld filler metal certified material test reports for consistency with ASME,Section II requirements.
The RRI observed, during a plant tour, several hanger drawings which reflected that the described hangers were in Class 5.
Amendment 7 of the FSAR in Qiapter 3.2 defines Class 5 as a s=4-4-=My supported pipe
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having no safety role, but whose failure could reduce the effectiveness of some other safety-related component. The FSAR indicates that
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certain lines, two inches and smaller, would not be classified as Class 5.
The hanger drawings involved lines under two inches classified as
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Class 5 but designed such as to provide little or no movement restrainc
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to s=1==1e excitation. The RRI initiated discussions with licensee personnel only to be informed that an investigation had just been initiated into the entire Class 5 support design and Quality Assurance areas.
This matter vill be considered an unresolved issue pending completion of the licensee's investigation and clarification of the FSAR definition.
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Reactor Coolant and other Safety System Welding
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The RRI observed portions of three piping systen velds being nada during the period. These were:
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Field Wald W-6 as shown on isometric drawing SI-1-SB-08 in Safety Injection system line 6-S1-1-070-15112
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Fiald Wald W-6 as shown on isometric MS-1-RB-004 in Main Stesa line 32-MS-1-02-1303-2-5-
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Field Weld F9 2 as shown on isometric RH-1-RB-001 in the Rasetor
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Coolant System Pressure Boundary in the Residual Haat Removal system
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The line designation is 12-RE-1-001-250111.
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The three welding procedures and five valders involved were found to have
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l been qualified in accordance with ASME.Section II.
The veld filler metals
and components being joined by each of the welds were verified by documen-
tation review to be consistant with the requirements of Sections II and j
III as appropriate.
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The RRI also reviewed the following radiographs pertaining to safety Class I piping system velds for compliance to ASME.Section III require-ments:
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4 Wald Id.
Isometric Line Safety Class
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BRP-CS-1-R.5-23 3-CS-1-076-250111
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11A BRP-CS-1-RB-26 3-CS-1-019-2501R1
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BRP-CS-1-RB-38C 3-CS-1-235-250111
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i 9-BRP-SI-1-RB-56 6-SI-1-089-250111 k
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BRP-RC-1-RB-05 6-RC-1-008-2501R1
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BRP-RC-1-RS-08 3-RC-1-052-250111
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BRP-RC-1-RB-16 6-RC-1-147-250111
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BRP-SI-1-RB-16 6-51-1-101-25012.1
BRP-SI-1-RS-33 3-51-1-033-250111
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No items of noncompliance or deviations were identified.
9.
Safety Class III Weld Quality verification i
The REI reviewed the licensee's implementation of his commitment to
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radiograph and repair as required the field welds in the Cv=_t
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Cooling Water system and in the steam generator A=414=ry Feedwater system. See Inspection Reports 50-445/79-12 and 79-17 for discussions of this commitment. The RRI reviewed program control records main-t tained by Brown & Root Welding Engineering which reflect the number of welds examined to date and the action taken on each veld. The RRI i
selected eleven welds at random from those indiencing initial acceptance,
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acceptance after repair and those shown as still in repair processing to j
obtain an overall view of implementation. Seven of the eleven welds had been determined by welding engineering to be acceptable and the radio-
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graphs were reviewed by the RRI. The balance were verified to be in repair status.
No deviation to the commitment was identified.
10. Unresolved Items I
i Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of non-compliance or deviations. One such item is discussed in paragraph 7 of this report and will be hereafter referenced as " Class 5 Piping System Supports."
11. Mansaament Interviews
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The RRI met with one or more of the persons identified in paragraph 1 I
on November 6, 7, 15, 16, 20, 21 and 30, 1979, to discuss various f
inspection findings and to discuss licensee actions and positions.
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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TEXAS UTILITIES ELECTRIC
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Docket Nos. 50-445 COMPANY, et_ al,.
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50-446
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(Comanche Peak Steam Electric
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Station, Units 1 and 2)
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AFFIDAVIT OF 00YLE HUNNICUTT CLARIFYING NRC INSPECTION REPORT 83/34-83-18 REGARDING REACTOR VESSEL MIRROR SHIELD
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I, Doyle Hunnicutt, do depose and state:
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Q1.
Mr. Hunnicutt, please state your name and position with the NRC.
A1.
I am Doyle Hunnicutt.
I an employed by the U. S. Nuclear Regulatory Commission ("NRC") as Chief, Reactor Project Section B. Reactor Project Branch 2, Ragion IV Arlington, Texas.
knthisposition,I review, approve, and perform inspections of nuclear facilities under Region IV jurisdiction.
I am responsible for the supervision of the
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NRC inspectors who inspected the Comanche Peak Steam Electric Station
("CPSES") during July 1983 to the present.
Q2. Have you prepared a statement of professional qualifications?
A2.
Yes, a statement of my professional qualifications is attached to my affidavit.
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What are your responsibilii,'les with regard to CPSES?
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I am responsible for the direction of inspection personnel.
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review and approve the scope of inspection and investigation reports.
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I participate in direct evaluation of activities related to reactor
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construction, various testing and verification activities, and
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various supervisory responsibilities.
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What is the purpose of your affidavit?
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t My affidavit responds to the Atomic Safety and Licensing Board's f
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83-34/83-18 i
(" Board's") inquiry regarding NRC Inspection Report
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(September 12, 1983).
I understand that the Board requested clari-fication of paragraphs 3 and 4 (pp. 4-5) of this Inspection Report, which discusses the reactor vessel outer wall, reactor vessel mirror The insulation shield, and the containment vessel shield wall.
referred to in Inspection Report 83-34/83-18 is " reflective insula-i Over the years this material has been known by a number of tion".
names, including " mirror shielding." Although the term " mirror
'i 83-34/83-18, reflective insula-shield" was used in Inspection Report tion is a more accurate term, and will be used in this affidavit. ~
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I QS. Please briefly describe this insulation.
Reflective insulation (mirror shielding) is manufactured to size AS.
and shape specifications for particular locations within a plant Reflective such as piping, components, and/or reactor vessels.
insulation consists of several highly polished individual metal
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The sheets of sheets that have been formed into the proper shapes.
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polished metal are separated from each other by air geps, and are fabricated into an assembly to be attached to the piping, component,
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or reactor vessel in a manner that assures that the reflective insu-
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1ation will remain attached during service.
Each sheet of the fabri-j
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cated reflective insulation reflects a percentage of the heat back toward the heat source to reduce heat loss through conduction and
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convection. Therefore, several sheets of reflective insulation return a very high percentage of the heat back to the source and t
i reduce the loss or transfer of heat from the heat source to an accep-table (specified) value.
't The relationship between the reactor vessel, the reflective insula-
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tion and the reactor cavity wall is shown on Figure 1 attached.
V The reactor cavity wall was referred to as the " containment vessel l
shield wall" in Inspection Report 83-34/83-18.
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Although the reactor vessel is insulated by the reflective insula-I
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tion, there is some heat loss such that a cooling system must be j
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employed. This cooling system is known as the " reactor cavity /
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neutron detector well cooling system." The cooling system blows chilled air into the 3-inch gap between the mirror insulation and the reactor cavity wall, to remove the residual heat escaping past
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the reflective insulation.
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-4-qs. Paragraph 3 of Inspection Report 83-34/83-18 states that Region IV received an allegation that the reactor vessel outer wall had been
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or was in contact with the containment vessel shield wall.
Can you
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describe the resolution of this allegation by Region IV?
A6.
Following the receipt of this allegation I, together with William Jones, an Engineering Aide with the NRC, inspected the Unit i reactor vessel reflective insulation and the reactor cavity wall on August 12 and 19, 1983, to determine whether the allegation was true. We conducted our inspection by visually inspecting the gap (about 3",) between the reactor vessel reflective insulation and the
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inner surface of the reactor cavity wall around the full periphery of the vessel. Our visual inspection consisted of shining the light from a flashlight up into the gap, from the bottom of the vessel, looking for signs of contact. The reactor vessel was at ambient
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temperature (less than 100*F). We did not see any signs of contact between the reflective insulation and the reactor cavity wall. The Unit 2 reactor vessel reflective insulation was inspected in a similar manner by Mr. Jones, and he also found no evidence of contact..
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Q7.
Paragraph 4 of the Inspection Report states:
Apparently the reactor vessel mirror shield did touch the shield wall during hot functional testing (HFT) and a related potential 10 CFR Part 50.55(e) report was transmitted to the NRC Senior Resident Inspection - Construction. The subject of this potential 50.55(e) report was that there was insufficient heat removal capacity in the reactor vessel cavity.
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The Inspection Report conc
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ld shield did touch the shiewas identified
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stating:
l mirror occurrence RC and correc-However, thewell during the HFT.and reported to the N The f
by the licenseeaction is in tements, in light of you
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regress.
r answer to
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f tive Can you explain these sta observed y
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temperatures up to 314 f
reactor vessel reflec-i ig During hot l'unctional test n, annulus between the unanticipated high f
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by the Applicants in the reactor cavity watt.. T e air flow in the h
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tive insulation 1) restrictions in and the ctor
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attributed to:
he annulus and to the rea
temperatures wereconstruction debris in t heat loss from the reacto r
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annulus due to support channel; and 2)
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vessel insulation originally anticipated.
than vessef'which was higher actions to address
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eral corrective ved, and the insulation
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The Appifcants implemented sevConstruction de age holas.
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the problem.
ified by drilling air passseals, modified
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support channel was mod d several insulation acity of the addition Applicants addereased the heat removal cap
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Appli-
)j existing seals, and inc The Staff will require the
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actions u
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cavity cooling system. hot testing that these to'the grant reactor cants to demonstrate by annulus' temperature prior sufficient to reduce the
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are operating Ilconse.
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The statement on page 5 of Inspection Report 83-34/83-18 that "the j
i mirror shielding did touch the shield wall during HFT" was intended
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to refer to the contact between the reflective insulation and the construction d*bris, which in turn was in contact with the reactor
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cavity wall. The removal of this debris took place before the inspec-i tion discussed on page 4 above. Consequently, during that, inspection we observed no contact between the reactor reflective insulation and
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l the reactor cavity. Nor was there any indication that the construc-tion debris had damaged or crushed the reflective insulation. We
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i concluded that there had been no actual contact between the reactor vessel and the reactor cavity wall.
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~Dof/le M. Hunnicutt
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Subscribed and sworn to before me l
this Sl6i day of May,1984
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