IR 05000341/1986005

Insp Rept 50-341/86-05 on 860205-0729.No Violations Noted. Major Areas Inspected:Review of Allegations

ML20212P173
Person / Time
Site:	Fermi
Issue date:	08/25/1986
From:	Wright G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20212P164	List: IR 05000341/1986005 ML20212P161 ML20212P181
References
50-341-86-05, 50-341-86-5, IEB-79-02, IEB-79-2, NUDOCS 8609030019
Download: ML20212P173 (16)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-341/86005(DRS/DRP)

Docket No. 50-341 License No. NPF-43 Licensee: Detroit Edison Company 2000 Second Avenue l Detroit, MI 48224 Facility Name: Fermi Nuclear Power Plant Inspection At: Fermi 2 Site, Newport, MI Inspection Conducted: February 5 through July 29, 1986 Inspectors: J. F. Schapker l

R. W. DeFayette r

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Approved By: . .

t) (, hief 767 8 ' W-8d Reactor Projects Section 2C Date Inspection Summary Inspection on February 5 through July 29, 1986 (Report No. 50-341/86005(DRS/0RP))

Areas Inspected: Special, unannounced safety inspection-to review allegation Results: Of the areas inspected, no violations were identifie PUR ADOCK 05000341 O PDR

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DETAILS Persons Contacted

. Detroit Edison Company (DECO)

• • J. Leman, Superintendent, Maintenance / Modifications

• J. F. Malaric, Assistant Modification Engineer

• J. Conen, Licensing Engineer

• J. E. Contoni, Lead Engineer, Mechanical

• W.-V. Lipton, Senior Engineer

• T. A. Young, Modification / Maintenance Engineer

• J. L. Martin, Quality Assurance Engineer M. Shields, Startup Engineer H. Higgens, HP/0PS Supervisor

• • R. Lenart, Plant Manager US NRC

• • M. E. Parker, Resident Inspector P. D. Kaufman, Reactor Inspector

• • R. W. DeFayette

• • W. G. Rogers

• J. F. Schapker The inspectors also contacted and interviewed other licensee and contractor personne * Denotes those attending Exit Meeting on March 18, 198 .

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• • Denotes those attending Exit Meeting on July 29, 198 . (Closed) Allegation RIII-85-A-0165: On September 25, 1985, the Region III Staff received a telephone call concerning alleged deficiencies at Fermi 2. The NRC Staff contacted the individual on several occasions and obtained the following concerns as a result of these conversation Concern (1)

The Fab shop was responsible for cutting a pipe at a 45' angl However,

"the kid cutting the pipe was eating a pizza and let it go to. . ." pipe repaired. The installation was so bad that two 25 ton bottle jacks had to be used to move the pipe into plac (previously reported to the NRC)

NRC Review The NRC inspector reviewed NRC Inspection Reports No. 50-341/79-04; 50-341/79-26 which addressed this concern. Within this concern the alleger identified that the pipe was installed in the " steam tunnel" and had been previously identified to the NRC. This concern appears to be identical to one addressed in Allegation No. 12 of NRC Inspection Report No. 50-341/79-04 and Paragraph 3, " Investigation of Allegations" of NRC Inspection Report No. 50-341/79-26. These two-inspections addressed the concer .

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For further assurance that the installation was adequate, the NRC inspector reviewed the final radiographs which were referenced in NRC Inspection Report No. 50-341/79-04, Allegation No. 12. The weld was found to be acceptable in accordance with the ASME Section III, NB 5000 requirement Conclusion This concern was previously addressed in the referenced NRC inspection report To reaffirm that the subject pipe was adequate the NRC inspector reviewed the applicable radiographs and found them to be acceptabl This concern is considered close Concern (2)

The wrong weld rod was used on nozzles in the turbine building. Welds were accepted after specifications were changed. (previously identified to the NRC)

NRC Findings This allegation was addressed in NRC Inspection Report No. 50-341/79-04, Allegation No. 9. The results of that inspection determined that the piping / nozzles were non safety related and therefore not a construction safety issue. To address the implied concern of use of the proper weld rod on safety-related components / hardware during and previous to this time period the NRC inspector reviewed NRC Inspection Reports for the years 1973 through 1980. The following NRC Inspection Reports identified weld rod control was satisfactory within the areas surveyed:

NRC Inspection Reports No. 50-341/80-09; No. 50-341/79-12; No. 50-341/79-08; No. 50-341/79-03; No. 50-341/78-20; No. 50-341/78-18; No. 50-341/77-12; No. 50-341/76-06; No. 50-341/74-05; No. 50-341/73-09; No. 50-341/73-06; No. 50-341/73-03; No. 50-341/73-01 Conclusion The initial NRC review referenced in the preceding paragraph adequately addressed the immediate concern as to the safety implications of the subjectwelds. Those systems that are important to assure the safety of the public are described within the Preliminary and Final Safety Analysis Report (PSAR/FSAR) submitted to the NRC by the licensee. These submittals are reviewed and approved by the NRC prior to issue of a license to construct / operat The systems in question do not fall into this categor ,

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Therefore, by virtue of the designation as a non safety-related system, which was reviewed and approved by the NRC, this finding was not addressed further; however, the licensee was informed of this concern so that they could take appropriate action, if require To assure the licensee's safety-related systems do not experience this type of discrepancy the NRC regularly performed inspections which verified that weld rod control was adequate as demonstrated in the referenced NRC inspection report Based on the data reviewed, there appeared to be adequate weld rod control within safety-related application Concern (3)

A qualified welder at Fermi II failed welder qualification tests at the Monroe Station Fossil Plant. The alleger contends that nuclear qualified individual should have been able to pass qualification for the Fossil Uni NRC Review The alleger was unable to identify the welder. Due to the lack of identity of the welder, the NRC inspector could not verify the incident described in his concer The NRC inspector reviewed past NRC reports which addressed welder qualifications at the Fermi II Nuclear Power Plant. In the following list of Inspection Reports welder qualifications were identified as being satisfactory:

NRC Inspection Reports No. 50-341/73-01; No. 50-341/73-05; No. 50-341/73-06; No. 50-341/73-09; No. 50-341/76-06; No. 50-341/78-18; No. 50-341/79-08; No. 50-341/79-12; No. 50-341/79-20; No. 50-341/79-25; No. 50-341/80-09 This review was inclusive of all NRC Inspection Reports for Fermi II for the time period from 1973 through 1980 and indicates that welder qualifications were reviewed throughout construction with no adverse finding .

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Conclusion Because the identity of the wohfer was not known, the task of verifying the validity of this concern was not achievable. However, the NRC inspector reviewed a significant number of NRC Inspection Reports (1973 through 1980)

which did not identify a concern with welder qualification Fossil plants and nuclear plants, require the welder to qualify to the applicable code for which the welder will be performing work. The fossil application may be as or more stringent as the nuclear application depending on the applicable code, i.e., AWS vs. ASM The codes recognize a welder can fail a test but still be able to weld adequatel This is demonstrated in the ASME and AWS Codes which permit a welder to perform two additional tests immediately after failing the initial attempt to qualify, or one additional test if further training is given to the welde In addition, there are several different types of welding processes, such as: Manual Shielded ARC Welding (SMAW); Tungsten Inert Gas Welding (TIG);

and Gas Metal Arc Welding (MIG). These different types of welding processes require skills unique to the process. Therefore a welder who is proficient in one type of welding process may not be able to perform well in anothe For this reason the codes require a welder to qualify in each type of process he/she is to weld in. In the stated concern the welding process for which the welder was Nuclear qualified was not identified, nor was the weld process which he failed at the fossil fuel plan Therefore he could have failed to pass a welding test in a process for which he had not qualified as a nuclear welde Welder qualifications is not the final step to assure sound welds in Nuclear construction. All welds require a nondestructive test (NDE) to be performed by a certified inspector. The type of NDE required is dictated by the applicable codes and standards accepted by the NRC in the PSAR/FSAR. Thus, additional tests were performed to assure the quality of the welds at the Fermi facilit Concern (4)

" Steam whips on the second floor at the headers had torn out of the embeds." The alleger questioned whether " Red Head" anchor bolts were utilized for this installatio NRC Review The above incident apparently involved a non safety-related steam line manifold. The failure reported resulted from transients caused by the opening and closing of the non-safety steam bypass valves. This also caused failure of the bypass steam lines. The bypass steam line failure was reviewed by the NRC and NRC Inspection Repor.t No. 50-341/85049 documents the findings. However, this report does not address the failure of concrete anchors. The inspector's review of referenced embed failure revealed that no anchor bolts (red heads or otherwise) were

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involved in the embed failure. The embed was attached to the concrete by Nelson studs which are welded to the embed plate prior to concrete placement. The inspector also reviewed the alleger's concerns addressed in NRC Inspection Report No. 50-341/79-04, Allegation No. 15 whic identified improper installation of concrete anchors. It was alleged

"that concrete anchors were cut and welded to the base plates to give the appearance of properly installed concrete anchors. In addition some anchors were sometimes substituted with shorter anchors because of rebar interference." NRC-Inspection Report No.- 50-341/79-04 addressed this item as unresolved (341/79-04-03) pending satisfactory completion of testing of the anchors by the license This item was closed in NRC Inspection Report No. 50-341/80-11 referencing the licensee's corrective action per IE Bulletin 79-0 IE Bulletin 79-02 addressed pipe support base plate design using concrete expansion anchors. Inspections performed by Region III specialists concerning this bulletin are documented in NRC Inspection Reports No. 50-341/84-30, No. 50-341/84-59, and No. 50-341/85-1 NRC Inspection Report No. 50-341/84-59, Paragraph 4c, references the test and inspections performed that were documented in DECO letter FE4-079 The inspector requested the inspection data referenced on this letter be furnished for review. Review of the inspection data revealed the following:

• The base plate was removed from the wall for the testing of the anchor. bolt * All anchors were tested for each base plat * Dimensional inspection of stud length and threaded stud extension was mad * Anchor, size was recorde Other pertinent parameters were also inspected; however, they did not apply to this concern. A review of this data indicated none of the anchor bolts inspected were cut short or improperly installed. In addition to this review the NRC inspector performed an independent inspection utilizing Ultrasonics (UT) to verify the anchor bolts were of the required lengt The following hangers were verified to be supported by the correct size and length anchor bolt / stu Required No. of Anchor Hanger Length bolts examined Actual (UT)

P41-2295-G68, Revision A 8.5" < 9" 4 8.5" to 8.75" E21-3150-G09, Revision 0 10" < 11" 4 10.5" to 10.7" E21-3149-G13, Revision A 10" < 11" 6 10.3" to 10.7" E21-3149-G06, Revision A 10" < 11" 5 10.4" to 10.9" P34-8213-G01, Revision 0 8.5" < 9" 4 8.5" P44-3345-G01, Revision A 16" < 17" 2 16" and 17" GWI-P44-7033-G05 8.5" < 9" 2 9" GI721-2113-G327 10" < 11" 4 10.5" to 10.9" E11-3164-G18, Revision A 10" < 11" 4 10.6" E11-3164-G19, Revision A 12" < 13" 4 12" to 12.8"

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i Required No. of Anchor Hanger Length bolts examined Actual (UT)

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T48-2097-G32, Revision A 8.5" < 9" 4 8.5" T71-12837-55-G02, Revision G 10" < 11" 4 10.1" to 10. 9" T50-7115-G01, Revision 0 8.5" < 9" 4 8.5" to 8.75" G41-3356-G23, Revision 0 10" < 11" 4 10" to 11" DD-098-0C (EC), No. 4 8.5" < 9" 4 8.5" to 8.8"-

DD-038-2C (EC) 8.5" < 9" 4 8.5" to 9" W-P44-3348-G18, Revision A 5.5" < 6" 2 6" W-P44-3348-G18, Revision A 8.5" < 9" 2 8.5" N30-3638-G05, Revision A 4.75" (FS) * 4 5" N30-3265-G09, Revision 0 15" < 16" 5 16" N30-3619-G17, Revision A 8.5" < 9" 4 9" N30-3377-G09, Revision A 10" < 11" 1 11" N30-3377-G09, Revision A 12" < 13" 3 12" to 13"

R5030-A5841-G903 10" < 11" 4 10" to 11" N30-3119-G14, Revision 0 8.5" < 9" 4 8.5" to 9" GC721-2537-EDP-3672 12" < 13" 4 12" FE-N30-7073C-(S-185) 8.5" < 9" 4 8.5" to 9" FXE-N30-N561C (S-185) 8.5" < 9" 4 8.5" to 9" FE-N30-N0738 (S-185) 8.5" < 9" 4 8.5" to 9" FXE-N30-N561B (S-185) 8.5" < 9" 4 8.5" to 9"

, FE-N30-N073A (S-185) 8.5" < 9" 4 8.5" to 9" FXE-N30-N561A (S-185) 8.5" < 9" 4 8.5" to 9" P95-F006-L4028 (S-385) 8.5" < 9" 4 8.5" LTil-P95-L403B (S-185) 8.5" < 9" 4 8.5" IW-P50-7505-G08 2.625" 2 (FS) 2.6" to 2.7" IW-P50-7505-G11 2.625" 2 (FS) 2.69" IW-P50-7507-G18 2.5" 4 (FS) 2.5" to 2.6" IW-P50-7508-G14 2.5" 4 (FS) 2.5" to 2.6" IW-P50-7515-G19 2.625" 2 (FS) 2.7"

IW-P50-7515-G22 5.5" < 6" 2 5.5" IW-P50-7515-G24 2.65" 2 (FS) 2.7" IW-P50-7515-G25 2.625" 2 (FS) 2.7"

IW-P50-7600-G06 2.5" 2 (FS) 2.6" IW-P50-7604-G08 3.375" 4 (FS) 3.35" to 3.39" IW-P50-7604-G23 2.5" 2 (FS) 2.6" i' IW-P50-7604-G30 2.562" 4 (FS) 2.6" P44-3362-G16 2.5" 4 (FS) 2.5" P44-3362-G17 5.5" < 6" 4 5.9"

• (FS) indicated flush shell anchors, all others were wedge anchor Conclusion The inspector performed independent inspection of 168 anchor bolts and verified proper bolt / stud length. IE Bulletin 79-02 concerning anchor bolts was closed in NRC Inspection Report No. 50-341/85-11. Review of the licensee's inspection / testing program referenced above showed it to be satisfactory. The combined data referenced in these inspections give adequate assurance that there is no evidence to support the referenced t 7

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concern The failure of the Nelson studs on the non-safety embed was investigated by the licensee to assure that embeds supplied by this vendor are adequate for the design function. Further evaluation of the number and application of these embeds was performed by the licensee. This item was reviewed by the NRC and closed in NRC Inspection Report No. 50-341/8601 Concern (5)

Fermi sacrificial shield was delaminating (Previously identified to the NRC).

NRC Review This concern was addressed in NRC Inspection Report No. 50-341/79-04 Allegation No. 16. Several areas of voids in the sacrificial shield were identified by the licensee as documented in the review performed for this allegatio The NRC inspectors also identified other areas of voids through interviews of selected craft personne These voids were not identified as nonconforming by the licensee at the time of the inspection. Subsequently these voids were documented on Deviation Disposition Report (DDR) No. 2610 by the licensee. The NRC issued a Notice of Violation citing 10 CFR 50, Appendix B, Criterion XVI. The licensee submitted its corrective action to the NRC on August 31, 197 On February 28, 1979, Deviation Disposition Request (DDR) No. (C)2610 was generated by Daniel Q.C. department to identify the subject void Disposition of DDR No. (C)2610 was to fill the voids and to visually inspect all outside weepholes in the sacrificial shield. This inspection was completed and DDR No. (C)3061 was generated on June 26, 1979 to identify additional void The additional voids were at approximate azimuth 242 to 320 , elevation 609 feet to 613 feet. Within this general area, six (6) separate void areas were found. DDR No. (C)3610 was subsequently rewritten as a new DDR - No. (C)3245. This was done to clarify the initial dispositio On several occasions during placement of the grout mechanical failure of the grout pumps caused lengthy delays in the placement process while backup pumps were hooked up and pipelines were cleaned of hardened grou In several pours, the individual access holes to the compartments within the sacrificial shield wall became blocked with grout which hardened during installation of the backup pump. This problem and an inadequate monitoring of every weephole were the cause of the problems outlined in Appendix A, Paragraph 1, of NRC Inspection Report No. 50-341/79-0 Corrective action to avoid further noncompliance for pump failure is not applicable because all pumping of grout from mechanical pumps has been complete Prior to repair grouting, craft and Q.C. personnel involved in this activity attended a training session covering the repair method including monitoring of grout placemen _ _ . _ . - ._ . .

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In order'to prevent future closing of DDR's prior to the disposition being complete, all site contractors with Q.C. inspection responsibility were directed to. notify their inspection personnel of this problem and emphasize the necessity for complete' verification of completed dispositions before closing out DDR's.

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The corrective action referenced above was verified as satisfactory

, resolution to the previously identified noncompliance and was closed out.in NRC Inspection Report No. 50-341/80-01.

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As further assurance that the sacrificial shield and other shielding is adequate, the licensee performs radiation surveys prior to and during

power ascension to assess the exposure rates in predetermined areas'of

the plant to permit the detection of any radiological problems, i.e.,

neutron streaming, etc. To date, three surveys have been accomplished with no significant radiological measurements apparent.

j Conclusion t 1 The licensee's corrective action for void detection and repair of the sacrificial shield was reviewed and considered adequate. Further, review of the radiation surveys performed during power ascension has thus far

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demonstrated the adequacy of the shielding. The radiological surveys will

continue throughout power ascension and during the operations phase to

. assure the radiological status of areas within the plant and to maintain j personnel exposure to As Low As Reasonable Achievable (ALARA). The licensee's radiation survey techniques referenced the following industry standards: ANSI Standard N18.9-1972, " Program for Testing Biological Shielding in Nuclear Reactor Plants" and ANSI Standard ANS-6.3.1-1980 .

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" Program for Testing Radiation Shields in Light Water Reactors (LWR)." '

Based on the above data this item is considered closed.

Concern (6)

j A pipe fitter was concerned that a " block, anchoring pipes" in the ea'st

! end of the steam tunnel was cracking. The individual also stated that

! the block had been " doctored up with stainless steel".

i NRC Review i The NRC attempted to obtain the name of the pipefitter to obtain more l information but was unable to do so. However, other items related to steam tunnel lower pipe whip restraints were-inspected by the NRC beginning in 1978 and discussed in Inspection Reports No. 50-341/78018,

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50-341/79003, 50-341/81016, and 50-341/82017. The issue was closed at that time.

, Other activities related in general to this subject were

(1) An NRC Bulletin was issued in 1979 (IEB 79-14) relating to seismic analyses for safety-related piping systems. This was inspected and closed in Inspection

! Report No. 50-341/81004; (2)Bulletin 79-02, " Pipe Support Base Plate i Designs Using Concrete Expansion Anchor Bolts" was inspected and closed

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in Inspection Report No. 50-341/85011; (3) An extensive inspection was conducted in 1986 on embedded support base plates and welding of studs

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] to embedded plates in other parts of the plant (this was documented in Inspection; Report No. 50-341/86012). .The NRC inspector concluded that j- the licensee's program in these areas was acceptable and procedures were correctly implemented.

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j_ Conclusion

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l The NRC attempted to obtain more specific information to follow upon this

, concern but was unable to get it. Other related activities and inspections

-have beers conducted during the last seven years and the licensee's programs

and construction found to be acceptable. Therefore, given the lack of j additional, specific information this item is considered close .. (Closed) Allegation RIII-85-A-0165 and RIII-86-A-0005, South Reactor l Feedpump Turbine Failure.

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In July 1985, the south reactor feedwater pump turbine failed

due to high vibration As a result of this the NRC received an allegation i that the licensee did not perform adequate trouble shooting, that it only

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replaced parts, and did not search for root causes of problem The

allegation also stated there were similar problems with the recirculation and RHR pumps and the main turbine.

j NRC Review i-1 The NRC response to the first-part of this allegation must be qualified by the fact that it delves into areas for which there is little regulatory

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authority. 'Specifically, the feedwater system is not classified as a safety system and is not required to cause or to support the safe shutdown

of the reactor or to perform in the operation of the reactor safety i features. Therefore the stringent rules and procedures for construction,

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installation, and operation which must be followed by licensees for safety systems do not apply in this case. Nevertheless, the NRC did followup on the allegation in an attempt to determine the history of the south reactor

feedwater pump turbine failure. The investigation was conducted by a  ;

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i February 14, 1986 (enclosed with this report) and which is summarized

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l The troubled history of this turbine dates back to 1972 when the first factory test was aborted due to high vibration. Following shaft balancing, the second factory test was performed satisfactorily and the turbine was

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! shipped to the site where installation was completed about 1978. Following I installation, the turbine was unused until 1983 when four runs were

conducted for preoperational testing. The first run' encountered high

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vibrations and was aborted. Following installation of temporary vibration

! instrumentation a second run was conducted which also was aborted because

! of high vibratio The licensee then developed a special warm-up procedure j and conducted a third run. Vibration was normal throughout the full

! operating range of the turbine. A verification run using this special I warm-up procedure was then conducted satisfactoril ,

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In 1985 the turbine was coupled to its associated pump, to supply water

to the reactor for power ascension testing. During the initial series of tests, vibration alarms were received repeatedly and ultimately the turbine faile .

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The immediate cause of the turbine failure was operations' personnel failure to respond properly to high vibration alarus. When these alarms

were received the personnel checked the plant computer which indicated [

low vibration levels, but they did not dispatch anyone to observe the ,

turbine and locally. verify the validity of the alarm Such an action i undoubtedly would have verified the high vibrations. Contributing to the problem was the fact that.the startup organization, which developed the special warm-up procedure, requested during the preoperational tests in

1983 that the operating procedure be amended to include the special steps, L but neglected to follow through and verify that this was done (which it i was not). Consequently the operations organization was not aware of the 1 special warm-up procedur In performing the inspection, the NRC contractor also examined

! documentation related to the foundation of the turbine. This examination revealed that- the turbine casing pedestals sloped slightly from front to back. The inspector examined the initial installation documentation to l ' determine if this slope was documented at the time of installation. He found that there were three missing signatures.for completion of several steps of the installation procedure including those for acceptability of

the foundation, levelness check of the turbine, and torquing of the <

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foundation bolts. It is probable that because these checks were missed

the slight slope of the turbine occurred during installation but was not i detecte Although it cannot be verified, it is possible that this

slight slope contributed to the turbine failure.

j Contrary to this allegation, the licensee searched for the root cause as evidenced by a Deco internal memorandum dated August 20, 1985,

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"K-T Analysis of the Failure of the SRFPT." The memorandum lists the i

problems encountered, discusses possible causes, and then identifies

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the most probable caus ,

The inspector evaluated the repair program for the turbine to determine its adequacy. The contractor report states that the work performed was

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thorough and that all the required work to prepare the turbine for operational readiness had either been performed or was planne The allegation states that the licensee not only did not search for root causes for the SRFP turbine problems, but that the practice extended to

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, safety-related systems. Since this is a broad allegation with no specific examples, the inspectors evaluated the licensee's general policy with

respect to corrective actions. The licensee has a formal corrective j action policy which is stated in Operational Quality Assurance Policy (0QAP) No. 16. The policy states that " measures shall be established to assure that conditions adverse to quality shall be identified. . .

and. . .the cause of the condition.shall be determined and corrective i-action taken to preclude occurrence."

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Section 5.1 of 0QAP 16 further states that " procedures shall be established that prescribe how conditions adverse to quality are to be promptly reported, documented, and corrected." Section 6.1 of Operational Quality Assurance Program R4quirements (QAPR) 16 requires that procedures shall require that conditions adverse to quality shall be evaluated and that necessary action shall be taken to preclude repetitio This must include consideration of other similar systems or components where similar discrepancies might exis In fulfillment of the requirements stated above, the licensee developed Nuclear Operations Interfacing Procedure No. 11.000.52, " Deviation and Corrective Action Reporting" which defines the corrective action system that controls identifying and evaluating significant conditions adverse to qualit The term " adverse to quality" is defined in the procedure as any problem that affects the quality of safety-related items and includes failures and malfunctions. Section 4.2.3 of the procedure defines the corrective action to prevent recurrence and states that this is directed at determining the root cause of a problem. The procedure also describes the Deviation / Event Reporting system (DER) which is used to document conditions adverse to quality, and the Corrective Action Review Board (CARB) which evaluates the DER Section 7.5.1 states that in performing the evaluation the CARB must determine the root cause of the deviatio Furthermore, the instructions for completing a DER, which are attached to the procedure, state that the evaluation must determine the root cause of the incident or the deviatio Finally, Section 7.8 of the procedure requires the Quality Assurance department to review safety-related DERs and to verify the completeness and accuracy of the documented information before approving the completed DER. Nuclear Quality Assurance Procedure (NQAP) 1607, "NQA Closeout Verification," establishes the basis for NQA closeout verification. Section 6.1.1 of that procedure states that NQA must assure that the root cause stated relates to the identified deficiency and that corrective actions are adequate. To aid in accomplishing this there is an attachment to the procedure in the form of a checklist to determine if the items are or are not acceptabl The first item on the list is " root cause reasonable."

During one of the inrpe:tions, the system for verifying and checking root causes was demonstrated to be operating. The inspectors received a copy of a DER on which the root cause was determined to be inadequate by the reviewer. The reviewer documented this on the DER with the comment, "the root cause is not acceptable. . .let's find the real root cause."

The concern while general in nature did mention that the potential problems were also related to the Residual Heat Removal (RHR) and Reactor Recirculation pump. While the concerns for the south reactor feed pump were valid, there is no evidence at this time to suspect any problems with the pumps mentioned abov It is noted that one RHR pump has fai'ed; however, inspection determined that the failure was due to a manufacturing defect and was unrelated to installation practice , .

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The NRC verified that there have been problems, most likely related to poor installation and work control and verification, with the south reactor feedwater pump turbine, but that the licensee documented these '

troubles with a Deviation / Event Report at the time of the turbine failur The turbine was properly repaired and a procedure revision is being prepared to change the method of post-maintenance testing. The licensee also has in place and has implemented a system for determining the root causes of conditions adverse to quality and appears to be using the syste Conclusion Based on the information above, we conclude that those parts of the concern dealing with improper physical installation of the south reactor feedpump turbine are most likely substantiated. However, those concerns dealing with people and systems for handling problems were not substantiated. This allegation is close . (Closed) Allegation RIII-86-A-0010, Crack in a Containment Hatch and Torus Shell.

l Concern: The initial allegation, which was received by NRC Region III

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through a Monroe City / County Office of Civil Preparedness official, claimed that there was a crack in the torus shell and a crack in the " containment l hatch" at Fermi 2. The Monroe official stated the alleger would not give

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his telephone number and therefore the NRC could not contact him for further information. Subsequent to this initial notification, Region III received another telephone call stating that the alleger had no personal kncwledge of any specific cracks at Fermi, only that he had heard that a similar reactor had these problems and he was concerned that they may be present at Ferm NRC Review Prior to operation of the reactor the licensee is required to perform a Containment Integrated Leak Rate Test (CILRT) which measures total leakage from all sources from the containment (the regulations strictly limit the amount of this leakage). As part of this test the licensee is required to inspect the containment inner surfaces. If there were cracks in the torus shell or in the containment hatch they would be identified during this tes On February 19, 1985, the licensee submitted a " Reactor Containment Building Integrated Leak Rate Test Report" to the NRC which documented the results of its CILRT. Region III specialist inspectors reviewed the report and determined that the licensee adequately performed, reviewed, and evaluated the preoperational CILRT and no cracks were identified. This review is documented in Inspection Report No. 50-341/85006. Furthermore, no cracks in containment have been identified by the resident inspectors during their routine tours of the facilit _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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Conclusion n

Based on the information stated above this allegation is not substantiated

and therefore is close i l

(Closed) Allegation No. RIII-86-A-0036, Dropped Fuel Bundle.

5. Allegation

A Monroe County, Michigan, Commissioner stated that he had

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heard that Detroit Edison employees had dropped a fuel rod bundle and did l nothing about it. They did not do any of the paperwork, and only " covered it up." He could not provide any additional informatio The individual also stated that Bechtel supervisors and managers of the Fermi plant were.

! unqualifie NRC Review

'

An NRC inspector interviewed licensee personnel involved with fuel

loading and discussed this issue with them. The personne! stated they

were not aware of any dropped fuel bundles, and volunteered information

on the only two events which they believe possibly could have led to the

allegatio One event relates to Fermi inspection of new fuel when it was receive There were seven bundles for which fuel deviation reports were written and which were discussed by the licensee with GE representatives onsite at the time of receipt of the fue Four of these documented minor deficiencies were resolved immediately. The other three bundles were returned to GE for refurbishing or repai One fuel bundle had a foreign tar-like substance on it, and the other two bundles appeared to have been dropped while in the shipping crates sometime during the shipment because

' the bottoms of the bundles were slightly distorted. GE inspected the i bundles at its Wilmington facility, performed a 100% check of fuel rod characteristics, x-rayed 100% of the end plug welds and the full length

,

of the pellets, repaired the bundles, and returned them to Fermi. The

' NRC inspector reviewed the fuel deviation reports for the three bundles returned to GE and the GE letter to Fermi describing its evaluation and corrective actions. There was no evidence of cover-up and all documentation was proper and correct. The resident inspectors were aware of these l problems at the time they occurre The second event described by licensee representatives occurred as fuel was being moved from the storage pool to the reactor. An operator latched a fuel bundle in the storage pool with the telescoping fuel handling tool and was in the process of raising it when he realized more than one section of the fuel handling-tool was being raised (e.g. - two

sections of the " telescope" were raising simultaneously). He immediately began lowering the bundle to free the stuck section when the power cable, which was caught in the stuck section and which was holding it up,

became unstuc The stuck section of the fuel handling tool then fell

- about one to two feet until it hit a "stop" pin. The fuel bundle itself did not move, however. , Nevertheless, the bundle was lowered into a fuel

inspection rig and examined for damage but none was found.

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- . . _ _ . _ __ __. _ __ - . _ _ _ _ _ . . _ _ _ _ _ , _ _ . _ . . _ _ . _ - , . . . _

.

Later that same day a different fuel bundle had been placed in the reactor core and the fuel handling tool was being raised (after releasing the fuel bundle) when the outer two sections fell full lengt Examination revealed that an air hose had lodged in the sections causing them to stick together and when the air hose broke the sections fell. The bottom section, which actually latches to the fuel, did not mov The fuel handling tool was repaired and fuel loading continue The alleger also stated that he had heard that there were " bullets all over the floor." He had no other information, but fuel pellets are small cylindrical objects which, to an untrained individual, might be mistaken for " bullets." The only way such pellets could be scattered over a wide area would be if a fuel rod were to be broken and the pellets fell ou This is not a probable scenario because fuel rods are not known to break and if they did only a few pellets may fall ou Regardless, the inspectors attempted to determine if fuel pellets may have been on the floor. No evidence of this was found. The rad-waste supervisor at the time stated that he had two people per shift working full time during fuel loading to keep the refueling floor clea Other licensee personnel also were interviewed and none of them had any evidence of fuel pellets scattered on the floo If such an event did occur, rad protection personnel would have set up a temporary rope barrier until the pellets were cleaned up. The inspectors could find no evidence of this being don Finally, the NRC provided 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> coverage during the entire fuel loading operation and, although the inspectors were not on the refueling floor 100% of the time, they could not recall any evidence of either spilled fuel pellets or dropped fuel bundle The former rad-waste supervisor did recall that about the time of refueling a piece of machinery, which had been stored outside, was taken to the refueling floor without being washed of Some gravel had stuck to the bottom and this gravel eventually ended up on the refuel floor and had to be cleaned up. The gravel was not contaminated and caused no problems other than the fact that it should not have been ther Licensee personnel were counselled about cleaning things off before taking them to the refuel floo The second part of this allegation pertained to the lack of qualifications of Bechtel supervisors and managers at the Fermi plan The alleger received a statement from another individual that these personnel were unqualified, but the individual could not say how they were unqualified nor identify specific individuals alleged to be unqualified. Based on the non-specificity of the allegation and NRC guidance, this allegation is closed without further action. It is further noted that there are no NRC regulations dealing with qualifications of licensee contractor supervisory personne .- . . ..,

.

Conclusion The licensee's actions in examining, documenting, and correcting the problems with the new fuel were proper and correct. The NRC could find no evidence of the licensee having dropped a fuel bundle, and for the items cited entries were found in the_ log books. Therefore, there was no " cover-up. "

6. Exit Interview The inspectors met with site representatives (denoted in Persons Contacted Paragraph) at the conclusion of the inspection. The inspectors summarized the scope and findings of the inspection noted in this repor The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection. The licensee did not identify any such documents / processes as proprietary.

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