IR 05000302/1974010
| ML19317G375 | |
| Person / Time | |
|---|---|
| Site: | Crystal River, Trojan  |
| Issue date: | 09/11/1974 |
| From: | Bryant J, Swan W, Vallish E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19317G371 | List: |
| References | |
| 50-302-74-10, NUDOCS 8003030779 | |
| Download: ML19317G375 (14) | |
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UNITED STATES
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ATOMIC ENERGY COMMISSION 9a
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DIRECTORATE OF REGULATORY OPERATIONS D I-
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REGloN ll - SUlT E 818 230 P E ACHT R E E ST R E ET. NoMT HwEST
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'dif t AT L. ANT A. GEoRo t A 30303
s RO Inspection Report No. 50-302/74-10
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Licensee:
Florida Power Corporation 3201 34th Street South
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P. O. Box 14042 St. Petersburg, Florida 33733'
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Facility Name:
Crystal River 3 Docket No.:
50-302
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License No.:
CPPR-51
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Category:
A3/B1 Location: Crystal River, Florida Type of License:
B&W, PWR, 2452 Mwt, 855 Mwe-Type of Inspection: Routine, Unannounced, Construction Dates of Inspection: August 13-16, 1974
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Dates o5 Previous Inspection: June 11-14, 1974 Principal Inspector:
E. J. Vallish, Reactor Inspector Facilities Section Facilities Construction Branch Accompanying Inspectors:
W. B. Swan, Reactor Inspector Engineering Section Facilities Construction Branch N. Economos, Reactor Inspector Engineering Section Facilities Construction Branch OtherAccompanyingPersonneJ e
Principal Inspector:
f'/b7[
E. J.# Vallish, Reactor Inspector Date l
Facilities Section Facilities Construction Branch Reviewed By:
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F tion F ilities Construction Branch v
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RO Rpt. No. 50-302/74-10-2-
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SUMMARY OF FINDINGS I.
Enforcement Action
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A.
Violations
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Certain items appear to be in violation of 10 CFR 50, Appendix B,
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" Quality Assurance Criteria for Nuclear Power Plants," as indicated below.
These apparent violations are considered to be of Category II
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severity.
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74-10-Al-(II)
Control of Welding Rods Contrary to Criterion V of Appendix B, welding rod was observed to be stored and available for use in an uncontrolled manner which could compromise weld joint quality. The licensee is committed to
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this~ control by Procedure FPC-W-21 and Section 1 of the FSAR. Prompt corrective action by the licensee resolved this issue.
This item is closed.
(Details III, paragraph 2)
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B.
Safety Items None I
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Licensee Action on Previously Identified Enforcement Matters A.
Violations i
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There were no previously identified enforcement matters.
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Safety Items i
l There were no previously identified safety items.
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III. New Unresolved Items
74-10/1 Valve Deficiencies (DROB No. 74-1)
Directorate of Regulatory Operations Bulletin (DROB) 74-1 relates deficiencies concerning weld failures between the valve yoke i
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and the motor operator mounting plate in valves supplied by the Walworth Company, and backseating disk mislocation problems on two inch Darling valves.
Field inspection indicates that-no
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R0 Rpt. No. 50-302/74-10-3-
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Darling valves are to be used on this site and the Walworth valves that needed repair were reworked ty Babcock and Wilcox Construction Company. This item is closed.
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74-10/2 Failure of Structural or Seismic Support Bolts on Class I
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Components (DROB No. 74-3 and 74-3A)
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This bulletin describes bolt failures found during routine in-
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service inspections at a pressurized water reactor on the steam generator seismic support holdowns.
In view of the
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status of construction and the requirements of this bulletin,
the licensee has committed to scheduling these inspection:s in a timely manner for the future.
This item is closed.
74'-10/3 Defective Westinghouse Type W-2 Control Switch Component.
(DROB No. 74-6)
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Recent inspection findings indicate a possible safety problem
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may exist in the plants using Westinghouse Type W-2 control i
switches. The licensee is investigating this possibility for Crystal River 3.
This item remains open.
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74-10/4 Deficiency in ITE Molded Case Circuit Breakers, Type HE-3 (DROB No. 74-8)
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Deficiencies were discovered during preoperational testing of the Trojan nuclear plant, of magnetic trip elements in these circuit breakers. The licensee is investigating this possibility for Crystal River 3.
This item remains open.
i 74-10/5 Deficiency in the General Electric Model 4KV Magne-Blast Circuit Breakers (DROB No. 74-9)
l Recent information indicates that oversized trip bars and in-proper clearances of related guide holes and linkages in these breakers could negate operation of the engineered safety system j
components which utilize this model breaker.
The licensee is investigating the possible use of this model breaker at Crystal River 3.
This item remains opea.
IV.
Status of Previously Reported Unresolved Items 74-7/2 Containment Building Dome Concrete ('10 CFR 50.55(e))
l Early results of concrete cylinder tests indicated low strength.
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FPC reported as required by 10 CFR 50.55(e).
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m Review of this report and subsequent field inspection of the details resolved this item..This item is closed.
(Details I, paragraph 2)
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RO Rpt. No. 50-302/74-10-4-
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73-5/3 Valve Wall Thickness Verification,~
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a.
Thickness measurements were not made of valve bonnets,
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cover plates or nozzles adjacent to the weldend pre-paration. This item remains open.
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The engineerine evaluation.of the valves' thin walls
was not accept L:.c.-
FPC stated the valves would be
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repaired by weldit.g to meet the requirements of the
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original purchase order.
This item remains open.
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FPC is waiting for an engineering evaluation from
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Babcock and Wilcox Company justifying the wall thick-ness at the weld joint of the pilot operated relief valve. This item remains open.
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73-5/14 Reactor Internals Modifications, i
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l Modification of the internals is being accomplished in the vendor shops and verification will be made after the internals
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are delivered to the site.
This item remains open.
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V.
Design Changes None VI.
Unusual Occurrences None VII. Other Significant Findings Project Status The licensee estimated construction completion at 93%. Placement of Class 1 concrete is virtually complete and placement of concrete facing on the Gulf side of the berm was nearing completion.
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Stressing of the containment tendons is scheduled to start Auy,ust 26, 1974.
VIII. Management Interview An exit interview was held with H. L. Bennett, Director, Generation
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Construction, and other members of the site staff.
The inspectors i
related the results of their respective inspections. One violation of 10 CFR 50, Appendix B was apparent.
It was explained that the Crystal
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River FSAR commits the licensee to control special processes such as
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welding, brazing and heat treating by procedures.
FPC procedure, FPC W-21, spells out the requirement for control of welding rod and
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electrodes and is consistent with the requirements of Criterion V of Appendix B, " Instructions, Procedures and Drawings." Hence, the finding of stores of unattended E 7018 type weld rod in different
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parts of the buildings appears to be in violation of 10 CFR 50 Appendix B.
The immediate action taken by management to correct
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this item and to prevent recurrence was deemed adequate and the item was therewith closed.
(Details III, paragraph 2)
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RO Rp t. No. 50-302/74-10 I-l
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DETAILS I Prepared by:
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- W. B. Swan, Reactorginspector
'Date Engineering Section
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Facilities Construction Branch
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Dates of Inspection: Augus t 13-16, 1974 Reviewed by: DN bud 9'
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L. L. Beratan, Senior Inspector
'Ddte Engineering Section
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Facilities Construction Branch
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Persons Contacted a.
Florida Power Corporation (FPC)
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H. L. Bennett - Director, Generation Construction C. E. Jackson - Construction Superintendent n
R. W. Slater - Quality Engineer U)
T. L. Baker - Supervisor, Quality Records (
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.R. Dorrie - Civil Engineer, by telephone J. C. Hobbs - Superintendant, Mechanical and Electrical Systems
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J. A. Hicks - Engineer
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Contractor Organizations Pittsburgh Testing Laboratories (PTL)
G. B. Browne - Chief Inspector
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Gilbert Associater, Inc. (GAI)
S. R. Buckingham - Site QA Coordinator
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2.
Containment Building Dome Concrete (10 CFR 50.55(e))
Previously Unresolved Item 74-7 /2 By a letter dated July 11, 1974, FPC submitted a final report concerning concrete placed in dome ring L-9 by pour 1004RB on April 4,1974. This 1etter states that the FSAR was to be amended. The inspector found that Amendment No. 4 to the CR #3 FSAR, page 3-15, was issued on July 3, 1974.
This provides for the use of ACI-318-71, paragraph 4.3, to evaluate a limited number of samples, instead of the previously referenced ACI-318-63.
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The licensee had taken 12 core cylinders from the concrete in ring sV, L-9.
These were taken as close as possible to the calculated locations
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RO Rpt. No. 50-302/74-10 I-2
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of the placement of the questioned concrete in the ring while avoiding the cutting of rebar or tendon conduits. Three of these cylinders were tested in the PTL laboratory on July 3,1974, under the surveillance of the GAI/QA structural engineer. These cylinders had compressive
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strengths of 6620 psi, 5960 psi and 5780 psi, an average of 6120 psi.
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t The indicated gain in compressive strength from 4570 psi average at 28 days to 6120 psi average at 90 days showed that the questioned con-
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i crete is now acceptable.
The inspector found that the licensee had been notified on April 18, 1974, I
by PTL that sample cubes made from the cement received on March 22,1974,:
had failed to reach design compressive strength of 5200 psi at 28 days.
Three cubes had yielded 3590, 3930 and 3410 psi. Use of the cement was stopped immediately and all of it was removed from the site.
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mixing was resumed only after a cement from a different supplier had been
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tested and approved.
No sample of the rejected cement was retained, so no additional tests could be made to determine the cause of its slow strength development.
s This item is resolved.
3.
Containment Pos.t Stressing Tendon Installations Most of the tendons had been pulled into the conduit sheaths. Those which had been in place for nearly seven months had been protected by
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filling the conduit with grease. Caps were found on the ends of all conduits except those ending at the top of the tendon gallery. The lower ends of the vertical tendons were being button headed. Formation and dimensional testing of the heads was observed. No discrepancy was noted.
One vertical tendon had been damaged during insertion into its conduit.
It had been rejected. Its replacement was on order but its fabrication required reactivation of the Prescon winding shop. The licensee does not expect that its delivery delay will negatively impact the stressing schedule.
The licensee had scheduled start of tendon stressing by August 26, 1974.
However, the site procedure for tendon stressing, including the sequence, was found to be still in the draft stage.
The licensee stated that telephonic notification will be given to RO:II
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prior to start of stressing.
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The inspector reviewed the QC records for receipt, storage and installation of tendons. The replacement of wires and modification or addition of wires had been documented.
No QC records deficiency was found.
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4.
Cleaning Procedure for Reactor Vessel and Internals F
B&W FCP 084, Rev. O, " Procedure for Hand Cleaning the Reactor Coolant
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System" was approved June 13, 1974.
It covers hand cleaning of surfaces
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exposed to primary coolant in the steam generators, reactor vessel, piping, pump volutes, pressurizer and core flood tank.
Clean room use
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is to be per FCP 083.
Specified cleaniness level is Class B, per FS-II-2.
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Final cleaning is to be done with Grade 1 water (normally demineralized water) using white, low lint cloths. The sequence of cleaning is to be controlled by " Upgraded Check List per 9A-147."
The reactor vessel and heads were received on site in Class C cleaniness condition.
B&W Construction Company (BWCC) plans no field modifications to
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these items beyond the. completed replacement of instrument guide tubes on the head with schedule 160 tubes.
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The RV head assembly was found stored in its storage spot in the contain-ment, above and to the side of the RV pit.
It was covered and protected against damage. A clean room with internal lighting had been installed on top of it.
The recirculation pump was also found to have been placed in a clean room structure for checking by Southwest Research Institute.
No cleaning was in process.
5.
Storage and Storage Inspection of Reactor Vessel Internals The controlling procedure is procedure No. B&W 082, Rev. O, approved April 19, 1974, and entitled "A Procedure for Unloading and Storage of Reactor Vessel Internals."
The plenum assembly and core support shield are stored in a sealed steel container mounted on a tilt up handling dolly and housed in a wood fram6d canvas covered shelter.
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The core barrel, thermal shield, lower grid assembly and flow dome are stored in a similar container and shelter.
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Both containers are kept under low inert gas pressure and have'a moisture indicator.
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The B&WCC inspector makes daily inspections of these items. His record of daily findings was reviewed.
No deficiencies were found in the storage conditions or in the records.
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Control rod drive mechaninms and other reactor internals are stored
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in a warehouse in their shipping boxes which have been. resealed. The inspector found that these have attached QC cards noting " Release for
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Construction."
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Southwest Research Institute had set up a plastic covered clean room
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in this warehouse for visual and NDT inspection of welds on CRDM's and other RV internal components. Use of the room is to be per B & W.
- Pro edure 083 " Procedure for Control, Maintenance and Use of Clean Rooms." No NDT work had started.
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RO Rpt. No. 50-302/74-10 II-l
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gM9%dev
/f DETAILS II'
Prepared by:
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N.'E(onomo V Metallurgical Engineer
/ gate Engineering Section
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Facilities Construction Branch
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Dates of Inspection: August 13-16, 1974 Reviewed by:
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L. L. Beratan, Senior Inspector Dat6
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Engineering Section
,i Facilities Construction Branch
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Persons Contacted
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Florida Power Corporation (FPC)
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C. E. Jackson - Construction Superintendent
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l R. W. Slater - Quality Engineer _
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i C. Hicks - Mechanical, NDE, Level III b.
Babcock and Wilcox Construction Company (BWCC)
J. R. McGill - QC Supervisor
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J. A. Jones Construction Company (JAJ)
J. Amundson - Site QC Manager W. Fleckenstein - QC Mechanical
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2.
Fabrication of field welds on safety related piping is continuing.
At-
the time of this inspection JAJ indicated that approximately 306 welds remained to be done while BWCC had only seven.
In response to questioning BWCC stated that three of these welds involved. nozzles on the pressurizer
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that were cut-out because of interference during installation. For these
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nozzles, BWCC generated field construction procedure (FCP) B&W 088. The
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procedure provides step-by-step instructions for welding, NDE and QC inspections with appropriate sign-offs. The inspector reviewed this
procedure and had no questions. Post-weld heat treatment of field welds was continuing; however, time did not permit a review of strip charts.
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RO Rpt. No. 50-302/74-10 II-2
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Primary Coolant Piping, Hydrostatic and Cleanliness QC Record Review
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Requ'irements for maintaining cleanliness during erection were established l
by BWCC field specification FS-II-2.
Records of cleanliness inspections l
during construction are on file at the site.
For primary pipe spool
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sections, certificates of compliance showed that no shop hydro had baen
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performed; this will be done at the time the system is tested
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following erection.
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Reactor Coolant Piping Hanger Installation
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For the seismic restraints on the primary piping, the inspector performed
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i a review of QC records which included: material certification; vendor
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inspection reports; certificates of compliance; and receipt insjection reports. Design engineering was performed by Gilbert Associates Incorporated (GAI). The restraints were fabricated by Northern Steel
' Corporation under purchase order No. PR3-3688Q.
The. controlling document.
was GAI drawing No. S-521-016.
Installation and inspection is being performed by BWCC in accordance with field procedure (FCP 091) which references the aforementioned GAI drawing. Adj ustments and final inspection
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remained to be completed.
Restraint No. RCR-2 had been tagged with reject ticket No.18691 as a result of unauthorized flame cutting " trimming" of stiffener ribs to
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attain minimum clearance between the pipe and restraint. To cover this, occurrance BWCC has issued a nonconformance report and repair procedure No. 089. It was anticipated that further trimming would be required to obtain the necessary clearance.
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5.
Installation and Inspection of Safety Related. Systems - A Record' Review Hangers for the makeup, decay heat, core flood,10" surge line and
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2-1/2" containment spray system are being installed by BWCC under pro-visions of FCP-047 Revision 4 approved by the licensee.
Installation inspections and final settings are subject to QC inspections. Form FCP-047 is used to sign-off. inspection items for,each hanger and provides permanent QC record. ' spection records selected for review were
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as follows:
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Core Flood:
CFH-5
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CFH-8 CFH-9 Decay Heat:
DHH-8 DHH-9 s
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f'~'s RO Rpt. No. 50-302/74-10 II-3 s_-
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DHH-10 DHH-ll
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Make Up:
MUH-20 MUH-21 MUH-22
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For the above items there were no u,nresolved questions.
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Installation of seismic restraints on other safety related systems is
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performed by JAJ under provisions established by FPC procedure W-93 s
Revision 0, dated December 18, 1973. QC inspections for individual supports and their appropriate locations are outlined in FPC procedures i
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Q-23 and Q-35 respectively. These procedures were reviewed and dis-
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cussed in Inspection Report No. 50-302/74-7.
Hangers for these, systems,
were supplied by Power Piping Company under Contract No. PR-3-1403. The
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inspector reviewed reports of materials received, inspection reports, certificates of compliance, vendor inspection reports and drawings of the
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hydraulic snubbers RCH-39 and RCH-37; guides RCH-21, 24 and 25; variable
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supports RCH-7, 13, 17,.19; constant support RCH-14.
Electrical system interference necessitated a change in design for knee support braces for variable supports RCH-7, 13 and 19. An engineering change notice covering this change was received by FPC on July 26, 1974. The design change has been approved 17 the licensee's engineering department and JAJ; the revised drawinge will be issued by Power Piping Company.
In response to questioning JAJ stated that their inspection efforts were presently concentrated in the auxiliary building. Within the areas
examined, there were no unresolved questions.
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bO R0 Rpt. No. 50-302/74-10 III-1
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DETAILS III Prepared by:
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l E. J.VVallish, Reactor Inspector Date
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Facilities Section Facilities Construction Branch t
Dates of Inspection: August 15-16, 1974 Reviewed by: \\ab.* M4A f
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J. C. Bryant, Senior Inspector
'Date
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Facilities Section
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Facilities Construction Branch
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1.
Persons Contacted
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Florida Power Corporation ~ (FPC)
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C. E. Jackson - Construction Superintendent R. W. Slater - Quality Engineer
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2.
Control of Weldine Rods During the observation of construction progress in the field, quantities
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of E 7018 welding rod were found lying around to the extent that weld quality could be compromised. Quantities were found in the reactor building, the auxiliary building and in the turbine building.
Some were in craft gang boxes, and some on a work bench. These supplies of
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welding rods were unattended and not being cared for as required by Procedure FPC W-21 and as stated in Section 1 of the Crystal River 3 FSAR. This lack of control of welding appears to be in violation of the intent of Criterion V, " Instructions, Procedures and Drawings."
Immediate action was taken by the construction superintendent to re-establish control of welding rod on the site and prevent recurrence of this condition. Directives were sent to all subcontractors and crafts on the site. The crafts' general superintendent also directed all his piping general foremen and foremen to take action on this problem.
In view of the nature of this problem, the response was deemed adequate and the item resolved prior to the management exit interview. No response is required from the licensee.
There are no further questions
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on this item.
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Letter to Florida Power Corporation from N. C. Moseley,
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50-302/74 10 dated. g{pj3h]/4
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DISTRIBUTION:
H. D. Thornburg, R0
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RO:HQ (4)
Directorate ol Lice %n " ing (4)
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- Local PDR
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- To be dispatched at a later date.
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