IR 05000302/1976008
| ML19317G362 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/13/1976 |
| From: | Jape F, Robert Lewis NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19317G349 | List: |
| References | |
| FOIA-80-250 50-302-76-08, 50-302-76-8, NUDOCS 8002280960 | |
| Download: ML19317G362 (18) | |
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IE Inspection Report No. 50-302/76-8 Licensee: Florida Power Corporation 3201 34th Street, South P. O. Box 14042
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St. Petersburg, Florida 33733 Facility Name: Crystal River 3 Docket No.:
50-302 e
License No.:
CPPR-51
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Category:
B1 Location: Crystal River, Florida
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Type of License:
f Type of Inspection: Routine; Unannounced
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Dates of Inspection:
.ril **.6 and 20-23, 1976
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Dates of Previous Inspection: March 10- a, 13-19, 25-26 and 30-31, 1976 Principal Inspector:
F. Jape, Reactor Inspector Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch Accompanying Inspectors:
F. Bower, Reactor Inspector
Engineering Support Section No. 1 Reactor Construction and Engineering Support Branch l
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W.
B.. Swan, Reactor Inspector l
Engineering Support Section No. 1 Reactor Construction and Engineering
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i Support Branch Other Accompanying Personne : None Prin.:1 pal Inspector:
AM 4d!Lf b
F. Jd fe, Reactor Insp)(ftor/
Date t
l Reactor Projects Section No. 2 Reactor Operations and Nuclear q/
Support.1 ranch t
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2-l IE Rpt. No. 50-302/76-8-
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C-./. t.< M 6 3/7(',
Reviewed By:
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R. C. Lewis, Chief Data Reactor Projects Section No. 2 i
Reactor Operations and Nuclear Support Branch
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.II Rpt. No. 50-302/76-8-3-
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SUMMARY OF FINDINGS I.
Enforcement Matters A.
Infractions 1.
Contrary to 10 CFR 50, Appendix B, Criterion II, in-doctrination of personnel in the quality assurance measures and administrative controls, as described in Section 1.7.6.7.1.b. of the FSAR, was not conducted of
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those personnel performing repairs to reactor coolant pump 3Al on April 4-9, 1976, as determined by interview with several of the individuals involved in the work.
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(Details I, paragraph 2.b. (1))
2.
Contrary to 10 CFR 50, Appendix B, Criterion V, the maintenance procedure for reactor coolant pump 3Al was not folicwed during disassembly of the pump on April 4-9, 1976, as described in Section 1.7.6.7.1.e. of the FSAR
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and Adminis*rative Instructions 400 and 600 in that a procedure wc changed without the required approval, clean room sta:us was not maintained and a clearance
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order was not obtained for removal of an instrument line.
(Details I, paragraph 2.b. (2))
i II.
Licensee Action on Previously Identified Enforcement Matters
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The licensee's reponse to the notice of noncompliance items in IE Report 50-302/76-3, dated March 24, 1976, was received. Followup
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of corrective actions will be conducted on a subsequent inspection.
III. New Unresolved Items Nor
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IV.
Status of Previously Identified Unresolved Items 75-13/1 Incorrect FSAR Drawing Drawing was corrected in Amendment 48 of the FSAR. Item is closed.
(Details I, paragraph 4)
75-15/3 Management Administrative instructions for handling NRC correspondence were reviewed. Item is closed.
(Details I, paragraph 5)
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IE Rpt. No. 50-302/76-8-4-
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I'6-1/1 Performance of Sodium Thiosulfate - Reactor Building Spray System
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A retest of the modified system is planned.
Item is closed.
(Details I, paragraph 6)
76-1/2 Valve Stem Discrepancy Licensee's report, dated April 23, 1976, was received.
Item is closed.
(Details I, paragraph 7)
76-2/1 Separation of Ir.ctrumentation Sensing Lines Rework of the cabinets and sensor lines has been completed.
Item is closed.
(Details II, paragraph 3)
76-3/2 Fire Water Diesel System A note has been added to the system operating procedure to require revision when construction has been completed.
O Item remains open, pending revision of the procedure.
kj (Details I, paragraph 8)
V.
Unusual Occurrences None VI.
Other Significant Findings A.
Containment Dome Concrete Separation
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j A separation in the concrete dcne of the reactor building was identified when cinch anchors were being installed. The i
l separation is believed to exterd over three-fourths of the dome and is approximately 1-3/4 inchen vide. The affected area is about one inch below the surface at the outer edge and 15 inches at the apex of the dome. FPC has reported the event as a "A.55(e) item.
(Details III, paragraph 2)
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B.
Reactor Coolant Pump 3Al Investigation of cause of excessive vibration on RCP 3Al
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revealed a 2 inch (approximate) crack on the pump impeller.
The impeller has been returned to the vendor's shop for repair.
I Hot functional testing has been delayed until the pump is returned to service. '(Details I, paragraph 2)
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IE Re t. No. 50-302/76-8-5-
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VII. Management Interview A management interview was held on April 16, 1976, with J. Alberdi, and members of his staff. Closeout of the unresolved item on separation of sensing lines was discussed and the inspector expressed areas of concern regarding maintenance activities on RCP 3A1.
A second management interview was held on April 23, 1976, with P. F. McKee, acting for G. P. Beatty, Jr., and J. C. Hobbs, Jr. acting for J. Alberdi. The items of noncompliance were discussed. Followup effort on the concern of test results documentation and the status of previously identified unresolved items were discussed. The inspectors review of the separation in the concrete dome of the reactor building was also discussed.
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b IE Rpt. No. 50-302/76-8 I-1
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DETAILS I Prepared by; F. Jape, Reactor Irispector Date Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch Dates of Inspection: April 20-23, 1976
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M/76 Reviewed by:
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R. C. Lewis, Chief Date Reactor Projects Section No. 2 Reactor Operations and Nuclear
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Support Branch 1.
Individuals Contacted Florida Power Corporation (FPC)
G. P. Beatty, Jr. - Nuclear Plant Superintendent S
P. F. McKee - Assistant Nuclear Plant Superintendent
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T. C. Lutkehaus - Maintenance Engineer
V D. W. Pedrick, IV - Compliance Engineer S. W. Johnson - Plant Engineer H. B. Lucas - Compliance Auditor R. A. Parker - Chief Nuclear Operator i
J. F. Heilman - Nuclear Operator B. L. Chastian - Data Collection Engineer
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E. E. Froats ~ Manager, Site Surveillance J. C. liobbs, Jr. - Manager Ceneration Testing A. P. Vogt - Testing Superintendent G. J. Walker - Manager Field Testing J. Somsel - Assistant Data Collection Technician J. Alberdi - Project Manager D. W. Bienkouski - Quality Engineer J. A. Jones R. T. Carter - Test Data Coordinator J. C. Milnor - Data Collection Technician Babcock and Wilcox Company (B&W)
E. L. Logan - Site Operations Manager 2.
Reactor Coolant Pump 3Al Investigative activities to determine the cause of excessive vibration experienced on RCP 3Al were initiated following plant
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~ IE Rpt. No. 50-302/76-8 I-2 cooldown from hot functional testing on April 2, 1976. Control of activities associated with the disassembly of the RCP was trans-ferred from construction to operations through issuance of an Operations Rework Control form, ORC-342. Work request WR0239 and maintenance procedure MP-115, "RC Pump Inspection and Replacement,"
wera isseed to accomplish the required activities.
On April 4, examination of RCP 3Al impeller revealed a defect that appeared to be a crack. Disassembly began on April 4 and was completed on Ap.-il 9.
The impeller was returned to Byron-Jackson (vendor) in Vernen, California for examination and repair.
a.
Review of Administrative Controls A review of Mdnistrative controls and related work activities was
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conducted by interviewing personnel involved with the work and examination of the below listed documents and records. Two items of noncompliance were identified.
(1) MP-115, "RC Pump Inspection and Replacement" (2) CP-115, "In-Plant Clearance and Switching Orders" (3) Maintenance Progress Reports
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(4) CP-118, " Cutting, Welding and Grinding Permits"
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(5) Operations Rework Control 342 (6) Work Request 0239 (7) AI-400, " Plant Operating Quality Assurance Manual Control Document" (8) AI-600, " Conduct of Maintenance" (9) FSAR Section 1.7.6 "FPC Quality Program Plan"
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(10) CP-ll6, " Standard Cleanliness Specification" b.
Items of Noncompliance (1) Failure to Indoctrinate Personnel Personnel involved in the disassembly of RCP 3Al did not appear to be knowledgeable with QA practices and established administrative controls for performing quality related activ-ities. This finding was determined during interviews of several personnel directly involved with the work, who stated that they had not received instructions as to purpose, scope
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and implementation of QA manuals and procedures.
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'II 1pt. No. 50-302/76-8 1-3
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Failure to instruct workmen in the prescribed activity as specified in FSAR Section 1.7.6.7.1.b, is considered.to be an infraction of 10 CFR 50, Appendix B, Criterion II, "QA Program.
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i (1) Failure to Follow Procedures
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The portion of MP-115 that had been used for disassembly of
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the RCP was reviewed, and as a result of this. review, three
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identified. These are described below and are collectively considered to be an infraction of 10 CFR 50, Appendix B,
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I Criterion V, " Instructions, Procedures, and Drawings."
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(a) Removal of Sensing Line 3APTl Without Written Approval
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MP-115 prescribes required initial conditions for removal
of the reactor coolant motor and disassembly of the pump
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in Part 7.0.
It is stated in Part 7.5 that removal of any obstruction requires an equipment clearance order as
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prescribed by CP-115 and furthermore the clearance order
must be approved as per AI-400 for those cases when the-
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operating procedure does not cover removal of safety-
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related equipment.
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When sensing line 3APTl was found to be obstructing
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removal of the RCP, verbal permission to remove the line was obtained from instrument personnel. No clearance
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order was obtained. Further, the QA Program described in Section 1.7.6.7.1.e, and implemented by AI-600 part 4.2.3
states that safety-related equipment shall not be changed without written approval. Removal of the line was by cutting and has therefore been modified. Also, Part 4.2.4 of AI 600 states that work shall not be performed without written maintenance procedures and administrative control.
(b) Procedure Changes Without Approval During disassembly of the reactor coolant pump, it became apparent that the stud tensioner would not loosen the nuts. A method for heating the stuiu using an i
acetylene torch was determined and used successfully to loosen and remove all of the nuts.
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IE Rpt. No. 50-302/76-8 I-4
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The method developed was not reviewed and approved as
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required by Part 8.0, " Changes to POQAM," of AI-400. The details.of.the method were written and turned over to the on-coming shif t for their use to complete the job. The
inspector could not find any record or evidence that the change to the maintenance procedure was reviewed or approved as per AI-400. During discussions licensee management acknowledged that the change had not been processed as per established procedure.
(c) Failure to Maintain Clean Room Status A clean work area for work on the internal parts of the RCP was established within the reactor building. On April 20, the inspector toured the work area, including the clean room located at the 160 foot elevation, and
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found the clean room unlocked, unlabled and food on a work bench. No work was going on at the time, but component parts were in the room. Part 7.6 of MP-115 states that access to the clean room is to be controlled and all objects taken into the room must be accounted for. HP-ll5 also requires that the cleanliness require-
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ments of CP-ll6 be followed. Eating or bringing food I
into a Class A clean room is prohibited by CP-ll6.
Corrective action was initiated immediately when this-item was discussed with. licensee management. The clean room was locked and food removed to re' establish Class A con-dition within the clean room.
3.
Documentation of Test Results During a previous inspection, 1/ ocumentation of test results was d
identf*ied as an area of concern. Followup on this item was I
continued during this inspection.
Interviews were conducted w1th nine people who are involved with the test program. The interview discussions centered around the program as described in Section 13.2 of the FSAR and the Test Program Guide (TPG).
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IE Inspection Report 50-32/76-6, Details I, paragraph 6.
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IE Rpt.;No.- 50-302/76-8 I-5-
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l All of the people interviewed stated that_the program is working as described in the FSAR and the TPG, and that there is no pressure
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being applied to reduce the thoroughness of review in favor of higher productivity. The reviewers stated that a review-of some l
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g3 test packages can be completed in one day whereas others may require as long as one month before they release the package to the Test Working-Croup (TWG).
Members of the TWG stated that they have not relaxed their review i
in any manner and intend to continue with a hard attitude toward
acceptance of test results. This same view was expressed by members of the Plant Review Commaittee (PRC).
An interview with the employee who expressed the concern revealed that he felt that conduct of the programs has improved.
In his
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i view, it appeared that test packages are now receiving an adequate review.
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In summary, the people interviewed stated that:
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All persons involved in the test program are cooperating better and have gone through a learning period, resulting in improved understanding of how to conduct and document the testing activities.
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b.
The in-depth reviews conducted by both the Administrative Manager and Manager Site Surveillance have resulted in an improvement in test documentation.
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c.
The TWC and PRC have both maintained a hard attitude toward acceptance of test results.
Following the inspection on April 30, 1976, FPC informed NRC that
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the employes identified himself to FPC management and discussed his concern with them. The inspector contacted the employee by phone on May 3, 1976, to determine his position regarding documentation of test results. He stated that his concern has been resolved.
Based on the telephone conversation and the result of our review, this item is considered closed. Review of test results will con-tinue as part of the NRC's normal inspection program.
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IE Inspection Report 50-302/76-6, Details I, paragraph 6.
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" E Rp t. No. 50-302/76-8 I-6
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Incorrect FSAR Drawing
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During Inspection 50-302/76-13, Figure 6-1 of the FSAR was identi-
fied as being incorrect. Low pressure injection system valves were
shown as manually operated valves whereas these valves are motor operated. The licensee submitted Amendment 48 to the FSAR to correct the error. Unresolved Item 75-13/1 is closed.
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Administrative Instruction for NRC Correspondence
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t During Inspection 50-302/75-15, unresolved item 75-15/3 was identi-i l
fied, which dealt with administrative instructions for handling NRC correspondence. The instructions for processing NRC correspondence were explained by licensee management. The method appears to work satisfactorily as evidenced by recent enforcement correspondence received from FPC. Inspector had no further questions on this
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item. Unresolved Item 75-15/3 is closed.
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Retesting of BWST Drawdown Test
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i During Inspection 50-302/76-1 licensee management stated that a retest of the sodium thiosulfate and sodium hydroxide tanks was not planned following modification of the discharge lines. The modifi-cation was required to adjust the chemical injection rate to
correspond with the drawdown rate of the borated water storage tank.
During a meeting on April 8, 1976, with representatives from FPC, B&W, GAI and NRC, FPC stated that a retest of the revised system would be run. This understanding was confirmed at the management
- 4eeting. The results of the retest will be reviewed during a
subsequent inspection. Unresolved Item 76-1/1 is closed.
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7.
valve Stem Discrepancy j
I During Inspection 50-302/76-1, Unresolved Item 76-1/2 was identified due to incomplete information provided to Region II by FPC, relating to the number of valves with bent stems. A letter report listing i
all known valves with bent stems was received from FPC on April 23, 1976. Unresolved Ites 76-1/2 is closed.
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IT Ipc. No. 50-302/76-8 I-7
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Fire Water Diesel System During Inspection 50-302/76-3, Unresolved Item 76-3/2 was iden-tified which involved incomplete operating procedure OP-707, l
" Receiving Diesel Fuel."
Followup on this item revealed that a note has been added to the procedure to require revision upon acceptance of the fire water system by Operations. Construction of the diesel fuel storage tanks and fire water pumps is currentiv underway. This item remains open, pending revision of the proe cedure.
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LE krt. No. 50-302/76-8 II-1
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DETAILS II Prepared by:
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A 4' 1,/ 70 y79 F. U. Bower, Reactor Inspector Data
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j Engineering Support Section No. 1 i
Reactor Construction and Engineering Support Branch Dates of Inspection: April 13-16, 1976
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Reviewed by:
S. D. Ebneter, Acting Section Chief Date Engineering Support Section No. 1 Reactor Construction and Engineering i
3upport Branch 1.
Persons Contacted
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Florida Power Corporation (FPC)
E. Froats - QA Supervisor
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D. Olsen - QA Engineer. Instrumentation P. McKee - Assistant Nuclear Plant Superintendent
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l D. Pedrick - Compliance Engineer T. Baker - Records Supervisor R. C. Bonner - Electrical Construction Supervisor 2.
Electrical / Instrumentation (E/I) Installation Progress
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Apart from modification and rework, the E/I installation work is nearing completion. It was reported that 80 cables remain to be
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pulled. Cable fire stops are being installed. Termination, testing and certification work is continuing. Work is expected to be complete as presently scheduled barring any unforeseen events.
3.
Containment Pressure Sensors (Ref. i6-2/l)
The several questions raised regarding the subject installation as
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well as the related instrument cabinets have been suitably resolved
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and this item is closed. Rework of the sensor lines was accomplished under Engineering Change Notice (ECN) No. 257 which re-routed the lines to comply with the separation criteria. Work Order No. W-97 l
was issued to authorize the work and it has been completed. The separation criteria (FPC-56) is being revised to include sensor I
piping as well as tubing.
In addition to the foregoing the instrument cabinets have been
refurbished by cleaning and painting. Work is going forward to i
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Ir Rpt. No. 50-302/76-8 II-2
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l install cabinet heaters to assure a suitable environment for the contained instruments. This work has been approved and ECN No.
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2582 has been issued.
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The nonconformance report, NCR 2366 issued by the site QC inspectors
rejecting the original installation has new been signed off,~ approv-ed by the QC unit.
A missing hold down bolt for one of the cabinets has been replaced and the applicable NCR No. 2292 has also been accepted.
4.
Primary System (Reactor Coolant-RC) Pressure Instrument Sensor Lines
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Sensor lines for pressure transmitters 3A-PT1 and 3B-PT1 were examined during a prev'ious inspection (Ref. 50-302/75-17) as a representative sample of safety related instrumentation systems.
i At the time of that inspection, the installation work and related QC
inspection action was incomplete. The same items were examined during this inspection.
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Examination during this inspection revealed that the QC record
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file for the installed systems was signed off as complete and accepted by the QC inspector. Physical inspection of the installed
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hardware revealed that a portion of system
"A" sensor line had been
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cut and removed to accommodate access for the removal of other equip-
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ment. Continued discussion and record examination r&vealed that the rework causing the removal of the sensor line was planned and controlled by several documents, recently prepared, that were still in the paper work pipe line, per se, not yet in the record file.
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The documents in question that were produced for the inspector's ex-j amination included NCR No. 2345, Operational Rework Control (ORC)
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No. 342 and maintenance procedure (MP) 115.
The above listed documents include a " turn-over" of the rework responsibility to the Operations Unit at the site and include
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procedures and a planning program intended to meet the requirements of the applicable QC procedures. Based on the QC procedures gover-ing the construction activity, the action appeared to be in order and there are no further questions regarding the events leading to l
the turn-over of the rework responsibility to the plant Operations l
Unit.
l Potential problems identified by the inspector during the cursory review of the operations unit procedure and related conversations with the staff have been discussed informally with members of the
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IE Rpt. No. 50-302/76-8 II-3
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II:II Reactor Operations Nuclear Support Branch. This unit, now with the lead inspection responsibility at this site, will examine the areas of concern and take any action deemed appropriate.
5.
Class IE Equipment Qualification Program A requirement exists to examine an appropriate sample of Class IE eluipment certification documents to determine the milestone
claracteristics of the qualification program as conceived and i:splemented for the Crystal River Nuclear Project.
Areas of specific interest include design requirements, test plans, test set-up, test procedures and acceptability as they are related to environmental factors, seismic capabilities, suitability, dura-bility, fire resistance and other areas relative to public safety.
Work on this examination was started during this inspection for cables, switchgear and motor control centers; however, the work was not completed and no findings are available at this time. Future inspections will be scheduled wherein this work can be continued.
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IE Rpt. No. 50-302/76-8 III-1-
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DETAILS III Prepared by:
J.-vd7w M
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W. B. Swan, Reaef5~r Inspector Date Engineering Support Section No. 1 kaactor Construction and Engineering Support Branch Dates of Inspection:
r 20-21, 1976
'f!7!M Reviewed by:
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,m S. D. Ebnetier, Acting Section Chief Date Engineering Support Section No. 1 Reactor Construction and Engineering
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Support Branch
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1.
Individuals Contacted a.
Florida Power Corporation (FPC)
J. Alberdi - Project Manager
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E. E. Froats - Manager, Site Surveillance R. S. Dorrie - Engineer, Quality Programs Department
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T. C. Lutkehaus - Maintenance Engineer D. W. Pedrick, IV - Compliance Engineer C. E. Jackson - Construction Superintendent T. L. Baker - Supervisor, Quality Records W. W. Misula - Structural Engineer C. Pachos - Architectural and Structural Superintendent b.
Contractor Organizations (1) Babcock and Wilcox Construction Company (B&W)
E. L. Logan - Site Operations Manager
(2) Gilbert Associates, Incorporated (GAI)
J. C. Herr, P. E. - Structural-Engineer T. J. Waytena - Structural Engineer I
M. J. Wardrop - Civil Engineer
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Review of Documentation Pertaining to Concrete Separation in
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-Containment Dome
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The inspector reviewed six sketches which were in preparation for a f
10 CFR 50.55(e) incident report on a concrete seraration in the containment structural dome. These sketches had been prepared from i
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IE Rpt. No. 50-302/76-8-III-2
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core drillings from the top surface of the concrete down to the craity created by the separation and from soundings made to deter-m;ne approximately the periphery of the cavity. These sketches were included in the licensee's 50.55(e) report dated April 21, 1976.
The inspector reviewed tabulations showing the sequence and dates of stressing of tha dome tendons, approximately two thirds of which j
had been stressed by December 9, 1974. The concrete apparently i
separated on the morning of December 10, 1974 at 7:20 a.m.
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l eight of the tendon sheaths had been packed with grease prior to December 9, 1974. Separation of the concrete was not discovered at i
the time so stressing of the dome tendons was completed soon after.
j the December 10, 1974 incident.
The inspector was shown a report concerning an occurrence on December 10, 1974. As tne first workmen of the morning shif t were entering the containment, a great sonic boom was heard, dust fell from scaffoldings in the containment, and the containment structure and surrounding earth experienced extensive shock waves. The O,
licensee conducted an extensive investigation to determine the
cause of the incident including inquires about blasting in the I
vicinity and earthquakes. The inside of the containment was inspacted for damage and the top surface of the concrete dome was inspected. No damage or cracks were found at that time. The licensee concluded that a significant event had occurred but reported that the cause could not be determined.
With GAI and FPC representatives in attendance, the inspector made a visual inspection of the surface of the dome, the core drilled holes and the sonic investigations under way to more clearly define the extent of the separation.
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GAI and FPC structural engineers discussed with the inspector the probable mode of failure of the concrete (tensile), outlined plans for instrument measurements of relative movements of the surface above and below the cavity, and discussed several possible courses of action. GAI representatives stated that their HQ engineers were running a comprehensive computer analysis of the strength of the residual dome beneath the sepSration. Tentatively, a meeting is to be held with NRR at Bethesda, Maryland, on May 12, 1976, to discuss-possible corrective action by representatives of FPC and GAI.
Inspections will be conducted following selection and implementation of acceptable repair method.
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IE Rpt. No. 50-302/76-8 III-3
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3.
Failure of Impeller Vane and Apparent Rubbing of Moving Parts of Reactor Coolant Pump
The IE inspectors held discussions with B&W representatives concern-
ing the discovery of pump malfunctions-(excess vibration and
. rubbing sounds prior to pump shutdown), and their findings when the
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pump was disassembled. The impeller and other affected components
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had been shipped to the Byron Jackson Pump Company. factory in
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Vernon, California. On April 21, 1976, the inspector notified IE:II supervision concerning the inspection findings and of an impending
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analysis conference to be held at Vernon on April 22, 1976.
The inspectors reviewed B&W photographs of the pump components. An opening of the leading edge of one of the pump impeller vanes had occurred near the base ring. The photographs seemed to indicate that the initial half of the tear had been along a residual flaw in the stainless steel casting and extending through tha flaw into the base metal. The photographs displayed some evidence-af rubbing.
An unrepaired flaw was found in the machined surfaci of a factory repair veld.
- The pump bowl had been secured with protective coverings and was not available for inspection in the containment.
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