ML20198H722

Partially Withheld & marked-up Package of B Caffell Comments on Miscellaneous Category Ssers

ML20198H722
Person / Time
Site:	Comanche Peak
Issue date:	05/13/1986
From:	NRC
To:
Shared Package
ML20198H622	List: IR 05000445/1979022 IR 05000445/1983034 ML20198H615 ML20198H722 ML20198H762 ML20198H787 ML20198H807 ML20198H829 ML20198H850 ML20198H941
References
FOIA-85-59 NUDOCS 8605300534
Download: ML20198H722 (45)

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, Draft 4 9/14/84 AM-5 CP5a-SSER

1. Allegation Group: Miscellaneous No. 5

2. Allegation Number: AM-5

3. Characterization: It is alleged, based on hearsay, that there is a

" possibility" that someone " threw something radioactive in the lake,"

that is, in the Comanche Peak Reservoir, sometime between September and November 1978, and that this material may have been tritium.

4. Assessmer.t of Safety Significance: The NRC Region IV inspector interviewed both the Radiation Protection Engineer (RPE) and the Radiation Protection Supervisor (RPS) about the allegation and about receipt of radioactive material at the Comanche Peak Steam Electric Station (CPSES). The inspector also reviewed their radioactive source log. The RPS stated, and the radioactive source log documented, that CPSES first receipt of radioactive material was on January 10, 1980, and that the quantity of strontium yt'r E 90 received was exempt from licensing requirements.

The alleger stated that his turbine book indicated that tritium was Fo/A- ts - 51 B/t

y used in a leak-detection procedure for the turbine-generator unit. He did not not provide the data for this test. The RPS stated that although a leak-test procedure utilizing tritium could be employed for the turbine-generator unit, the turbine generator at CPSES was hydro tested during hot functional testing, in lieu of the tritium leak-test procedure outlined in the manufacturer's instruction manual.

The NRC inspector reviewed a Texas Utilities Generating Company (TUGCO) startup test log, which documented that the turbine generator primary coolant system and components were pru sure tested, not tritium-tested, in accordance with test procedure CPM 6.9I, " Main Generator Primary Water and Seal 011."

These tests began on December 4, 1982, and concluded on October 11, 1983.

i The RPS stated that CPSES first took receipt of tritium on January 31, 1983, a shipment authorized under a state of Texas radioactive material license (No. 5-2892) issued October 9, 1980.

l

. The CPSES radioactive source log documented this receipt. The RPS also stated that a tfetitiumstandardwasused to prepare a standard to calibrate the tritium monitors located l on the turbine-generator unit. The aliquot of tritium L

N used to prepare the calibration standard was recovered and CPSES verified its original radioactivity. The RPS stated that when the plant returns to hot functional testing after fuel load, the primary coolant of the turbine ge'nerator will be " spiked" with tritium.

During operation of the turbine generator, a small amount of hydrogen gas will be extracted and measured by a tritium monito there leak within the turbine generator cooling system, tritium would be detected in the hydrogen gas.

CPSES established documentated programs for controlling radioactive source material, as outlined in health physics Administrative Procedure (HPA)-105; for receipt of radioactive material, as outlined in health physics instruction (HPI)-202; and, for shipment of radioactive materials, as outlined in (HPI)-203. The Region IV staff conducted extensive reviews of these programs as part of the preparation inspection program, and documented the results of these inspections in Region IV Inspection Reports 50-445/83-16, 83-35, 84-02, and 84-25.

5. Conclusion and Staff Positions: Based on reviews of CPSES procedures, radioactive source 10g sheets, startup test logs and startup test data sheets, and on interviews with the RPE and the RPS, the NRC staff concludes that CPSES received no radioactive material prior to January l 1980 and received no tritium prior to January 1983. The NRC staff

l .

found that no evidence to support the allegation that radioactive material was " dumped" into the Comanche Peak reservoir by CPSES during the period from September 1978 to November 1978. Accordingly, this allegation has neither safety nor generic implications.

6. Actions Required: None.

5

8. Attachments: None.

9. Reference Documents:

1. HPA-105, Radioactive Source Control Procedure.

2. HPI-202, Receipt of Radioactive Materials Procedure.

3. HPI-203, Shipment of Radioactive Materials Procedure.

4. CPSE'S Radioactive Source Log.

5. State of Texas By-Product Material License No. 5-2892.

6. TUGC0 Test Log.

7. Allis-Chalmers Steam Turbine-Generator Instru.ction Manual, pages 2.1-0910.

f

10. This statement prepared by:

Russell Wise Date Reviewed by:

i Group Leader Date i

J,-

1 I

Approved by:

Project Director Date 1

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h4 1'7 Draft 1 / CP5A AM-15, AM-16 etv/A I

September 15, 1984 SSER 7 i

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1. Allegation Group:

b1 tsc tl M0 h 3 gory 13 %Maneoup > w tlJ h

2. Allegation Number: AM-15, AM-16 g

3. Characterization: It is alleged that ft lure to follow procedures and generally poor workmanship on the polar crane resulted in improper shinning and installation.

4. Assessment of Safety Significance: On August 14, 1984, two NRC inspectors made a visual inspection of the polar crane shims from the polar crane. One rotation was made to observe the radial restraint brackets and the seismic restraint brackets from the platform above the operator's level. Several places appeared to have gaps in excess of J {h W These observations were previously documented by NRC N' IR #50-445/84-08, " Notice of Violation."

Q's

.d This IR did not, however, address the shims for the (28) crane girder to girder support brackets. Another rotation of the crane was made to visually inspect these shims from the platform at the operator's level.

During this rotation, excessive gaps, particularly on the inside edge (looking from the inside of the containment), were observed. gfp-$f-D lb/3

2 A meeting was held with the Texas Utilities Electric Company (TUEC) project civil engineer, the Brown & Root (B&R) project control manager, the B&R subcontracts supervisor, and a representative of Chicago Bridge and Iron to determine the gap tolerance between the pearing plate "A" (Dwg. 2323-SI-0515) and the girder s,upport acket. Gibbs & Hill (G&H) e - s-

. specification SS-14 does not address this issue, nor does the Crane r

Manufacturers Association of America, Manual CMAA-70. The meeting failed to produce a specific answer; however, copies of 2 letters were provided.

The first was a B&R Letter No. BRF-7404, dated November 8,1977, which contained "as built" measurements of gaps at all shim locations, along with a request to G&H to evaluate this information and provide direction.

At 28 locations, the "as builts" show the gaps to range from .000-inch to

.581-inch. In the second letter, G&H (#GHF-2207, dated November 28, 1977, responded as follows:

" Girder Seat Connections These seated connections will not require shimming since the area in bearing is at least the width of the botton flange of the crane girder. The gap dimensions indicated in the Brown & Root survey exist only at the extreme edges of plate A, Section 3-3,

. Dwg. 2323-S1-0515."

It should be noted that the width of the bottom flange of the girder (the bearing surface) is 20 inches.

3 On August 30, 1984, the NRC inspector was accompanied by a TUEC QC inspector to inspect the 28 crane girder to girder support bracket shims.

Nine girders had gaps in excess of 1/16-inch, which extended under the bottom flange and violated the specified minimum of not less than 20 inches of bearing surface. The (9) girders are A7-6 right end (RE),

A7-8 (RE), A7-12 (RE), A7-14 (RE), A7-18 (LE), A7-19 (RE), A7-20 (RE),

A7-24 (RE), and A7-25 (RE).

This condition is contrary to the " Girder Seat Connections" engineering evaluation and required bearing surface statement in the Gibbs & Hill letter referenced above. It should be noted that A7-20 (RE) was closely observed as the crane wheels passed directly over the support bracket with no visible compression (closure) of the gap noted.

Additionally, a visual inspection of the complete rail system revealed that the rail has moved or is moving circumferentially. This is A, supported by the fact that some of the 1-inch diameter stabilizing rods ,

Also, the 3/8-inch design gap #

are bent from the force of this movement. ,*

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between the ends of the rail sections varies from .000-inches to '

.875-inches, measured at the inside edge of the rail. Three of the' rail-to-rail ground wires and two cadwelds are broken. Also, at least two rail shim plates have partially worked out from under the rail.

4

- The polar crane operator was asked if he knew of any existing problems with the crane or it's operation. He replied the crare is operating I

u b satisfactorily with no apparent problems. He also stated there are no

" dead spots," i.e., no loss of energy at spots.

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• This allegation has both safety significance and generic implications.

5 r

Conclusian and Staff Positions: Based on the above inspections, the NRC

! 5.

• t staff concluded that this allegation is substantiated. This allegation E:

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is safety-significant and has generic implications.

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6. Actions Required: Remedial steps to correct the deficiencies.

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8. Attachments: None _

9. Reference Documents:

j NRC IR #50-445/84-08  :

B&R Letter e8RF-7404 G&H Letter LGHF-2207 ,

(2) Conversion Records

J 5

10. This statement prepared by:

J. Corbett 9/12/84 Reviewed by:

i Group Leader Date l Approved by:

Project Director Date i

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Draft 1 / CP6

~M-20 A

~&1 Sept. 16, 1984 (Hunnicutt) p SSER '

• a 1.. Allegation Group: Miscellaneous No. 17

2. Allegation Number: AM-20

3. Characterization: It is alleged that material false statements were made_

by plant management to the Atomic Safety and Licensing Board (ASLB).

4. Assessment of Safety Significance: This allegation is one of 26 allegations made by the Government Accountability Project (GAP) in a letter to R. C. DeYoung dated March 19, 1984, and in affidavits filed by GAP with the NRC's Office of Investigation (01). During an August 10, 1984, telephone conversation with a GAP representative, an 01 investigator learned that the alleoetion referred to testimony of the Brown and Root Project Control Manager before the ASLB on Februaryr _21, g The NRC staff reviewed the transcript of the February 21, 1984, hearing session, and OI Report of Inquiry Q4-84-14, " Comanche Peak Steam Electric Station (CPSES): ASLB Request to Preserve Testimony of Witness," dated March 8, 1984. As a result of the review of this material, the NRC technical staff determined that the most appropriate issue to address was that of the Independent Assessment Program (IAP) -

protocol governing communications between Texas Utilities Generating Fo/A-es-si 3/+

2 Company (TUGCO) and CYGNA. If protocol were not properly established,

-m _.

the independence of the audit could be compromised to the point that the audit findings and conclusions could be seriously criticized. Issnes directly related to the determination of material false statements do not fall within the scope of responsibility of the NRC technical staff.

The TUGC0/CYGNA IAP protocol was initially established by the NRC and transmitted from Darrell G. Eisenhut to R. J. Gary of TUGC0 on September 23, 1983. In conducting the IAP, questions arose concerning the adequacy of the protocol's establishment of CYGNA independence during the review. This subject was addressed in the ASLB's Memorandum Clarifying Open Listed Issues Regarding QA Program, dated March 15, 1984, the ASLB hearing session of April 24, 1984, and the ASLB hearing session of April 27, 1984. After correspondence between the NRC and TUGC0 and CYGNA in March and April of 1984, a revised protocol was issued by Darrell G. Eisenhut in his May 31, 1984, letter to L. L. Kammerzell of CYG!'A and M. O. Spence of TUGCO. Mr. Eisenhut stated in his May 31, 1984, letter that the new statement of protocol more completely addressed the variety of communications that occur during the course of an independent review.

5. Conclusion and Staff Positions: The NRC staff concludes that the revised protocol adequately addresses the variety of communications that occur during the course of an independent review. No events that would call into question the adequacy of the revised protocol have been brought to

. 3 the attention of the NRC. The issuance of the revised protocol on May 31, 1984, has eliminated potential safety concerns associated with the independence of CYGNA while conducting the IAP. No generic implications are asscciated with this allegation.

6. Actions Required: None.

5

8. Attachments: Letter from Darrell G. Eisenhut to L. L. Kammerzell and M. O. Spence dated May 31, 1984, subject, " Independent Assessment Program (IAP) Performed by CYGNA"

9. Reference Documents:

1. Letter from Darrell G. Eisenhut to L. L. Kammerzell and M. O. Spence dated March 22, 1984

2. Letter from M. N. Shulman to Darrell G. Eisenhut dated April 10, 1984

3. Letter from M. O. Spence to Darrell G. Eisenhut dated April 18, 1984 I _

4

10. This statement prepared by:

D. M. Hunnicutt Date Reviewed by:

R. L. Bangart Date Approved by:

T. A. Ippolito Date

e i

~4 Draft 2-9/18/84 AM-17 CP5A SSER

1. Alleaation Group: Miscellaneous No. 14

2. Allegation Number: AM-17

3. Characterization: It is alleged that a deficient weld on missile barrier door were accepted.

4. Assessment of Safety Significance: The door referred to in this alle,gation is the tornado missile barrier, which is located at ground level on the west side of the Unit 1 Diesel Generator Room. Both Texas Utilities Electric Company (TVEC) quality assurance personnel, who were present at the time of the allegation, and a NRC staff review of the Atomic Safety and Licensing Board (ASLB) hearing record on intimidation and harassment provided this identification. The welds in question are 18 double-groove welds attaching the lifting lugs to the three missile barriers. The welds were terminated by wrapping the weld around the end

/

of the lug. The alleger mistakenly believed that the welds which were fe made were not those the drawing weld symbol called for. Actual N g 3,

\' +

F f inspection and &R r acceptance for the 18 disputed welds by Brown &oot

,~inspectors. JheNRCstaffreviewedtheinspectionreportsindicating p

thatweldy performed in accordance with procedures Welding Procedure Specification (WPS) 10046 Rev 9 and AWS code requirements.

Fo/A-Is-57 8lS

4 The NRC staff reviewed the construction traveler and the inspection reports for shop fabrication and field-fitting of the missile barriers and found that,although rework occurred j no rework was done on the lifting lug welds.

The NRC staff reviewed the inspection procedure, weld procedure, and inspection reports, referred to in the traveler for applicability and compliance with American Welding Society (AWS Code D1-1, Section 4.6.1 and to determine if the code allowed wrapping the weld around the end of the lug. The code states that the weld shall be terminated in a manner which ensures sound welds and would permit wrapping the weld around the end of the lug as was done.

The NRC staff determined which doors were located in building locations and inspected the four tornado missile barriers (east and west) of the diesel generator room, Units 1 and 2) except in Unit I where the missile barriers had been removed. In Unit 2 (east side) one segment of one

! missile barrier was in place for trial fitting and two were being fabricated on location. NRC staff visually inspected the lug welds on the tornadf missile barrier segments, and determined that the@

showed very good workmanship and wraparound on the ends of the lugs was minimal.

i l

i i

' 5. Conclusion and Staff Positions: The NRC staff found no errors in design or any poor workmanship on welds of the lifting lugs on the tornad/!O missile barriers. In addition, they detennined that both the welding and the inspection of doors was done in accordance with specified procedures.

Accordingly, the NRC staff concludes that this allegation has neither safety significance nor generic implications.

6. Actions Required: None.

b

8. Attachments: None.

9. Reference Documents:

1. Drawing 2323-S1-0635, " Safeguards Building Misc. Sects & Depts,"

Sheet 3, Rev.2.

2. Drawing MSB-0635-004, Tornado Missile Barrier Diesel Gen. Bld #1,"

Sheets 1 through 6, Rev. 1.

3. Construction Traveler CE-82-8900.

. 4 Inspection Reports IRMS-0142, 0143, 0144, 0145, and 0174

5. NCRs M-84-100117, M-82-01640, M-82-02158, 14503, 14608, 14607, M83-02625, M-83-01078, and M-81-0098.

E, (

l - _ _ - - - - . _ _ _ -- _. - . - - -

g

6. Weld Procedure Specification WPS-10046, Rev. 9.

7. Quality Assurance Instruction Nos. QI-QP-11.21-1, Rev. 7 QI-QP-11.14-1, Rev. 18 QI-QP-11.14-3, Rev. 6

8. American Welding Society Code D1.1, 1977

10. This statement prepared by:

L. Jones Date S. Phillips

-Reviewed by:

Group Leader Date Approved by:

Project Director Date

Draft 1-9/19/84 AM-8, 9, 10/cp6

_SSER

1. Allegation Group: Miscellaneous No. 8

2. Allegation Number: AM-8, AM-9, AM-10

3. Characterization: It is alleged that: (a) Unit 1 main condenser tubes were beat with air and sledge hammers, were split during belling and flaring, and were improperly rolled; (b) the wrong type condenser for the wrong type steam generator was used; (c) the, tube support sheets had holes that were 3/8 inch off; and (d) misalignment of turbinetocondenserandstresscausedgjackingintoal.ignment.

~~~

\

4. Assessment of Safety Significance: NRC Region IV first identified the 3' potential safety concerns associated with this allegation to Texas -

pg M

Utilities Electric Company in a letter dated June 19, 1984,, cutlined j A t' the monitoring of construction activities as well as- -th.e- -

vario

\

that were conducted.T The NRC staff attempted a visual inspection of Y

the condenser internals but this was possible due to the congestion of the tube bundles, the internal piping, and bracing. This inspection was limited to observation of the bundles from the manway using a hand held light. bservation was difficult because of the distance between the manway and the e bundles and surfacef,hc a ver,-

o irregularities were detected; b b F0/A - T E-S i Bl<.

D % 4 ] hree persons who had direct knowledge of the work were interviewed with respect to part (1) of this allegation. The B&R Millwright superintendent stated that hammers were used but only with a special tool to protect the tube ends and that this is standard practice in condenser assembly. This person disclaimed any knowledge of any misa'ligned tube sheet holes.

The second person interviewed was TUGC0 Operations Results Engineering Supervisor and he was contacted concerning available documents; i.e., inspection reports, deficiency reports, in-progress monitoring reports, NDE reports, procedures, etc. ,

He stated that the work was performed to B&R procedures and he did not keep any official records, as his particular involvement was limited to nonsafety-related surveillance for commercial consideration. When asked about tube rolling problems, he stated that the variance in manufacturing tolerances of the tube outside diameters and the tube sheet hole diameters did present a problem in the beginning but the problem was solved when they properly calibrated the rolling tool.

l He also stated that as the work progressed, his level of confidence in the crafts work reached a point where 100 percent surveillance was not reciuired.

He also stated that during the~ rolling of the tubes, the optimum tube wall reduction for positive tube to sheet seal was in the range of six to nine percent. However, at the beginning when experimenting with the torque on the rolling tool the nine percent reduction

d in wall thickness did occur in sorre cases. This did not result in cracking any of the tubes.- See note on the NRC staffs sketch (Exhibit A) attached for explanation of tube cracking.

The, third' person interviewed, a BIR Millwright Foreman, informed us that he was not involved in the construct-ion of the condenser and disclaimed any knowltoge of the alleged problems. This person told us of a rumored retubing of the condensers, the auxiliary condensers, and the moisture separators. This rumor was confirmed

/

by the Millwright Superintendent and in a'telecon with the TUGC0 Nuclear Engineering Manager. The latter infonned us that his group was commissioned to make a feasibility study of condenser retubing with Unit 2 being done first and Unit I during the first outage.

Tile change would be from chrome-nickled (Cr-Ni) tubes to titanium

, tubes to raise the pH level from the present 9.4 to a level of 9.8-10.0 on the secondary side. This material change may have been under consideration when the a~11eger made reference to

" titanium tubes." The installed tubes are Cr-Ni. The alleger may have been referencing this when he said "its the wrong type condenser forthetypeham= generator." The condenser was provided by Westinghouse to Gibbs & Hill, Inc. (G8H) Specification 2323-MS-23 which specifies the operating conditions / criteria that must be met.

On September 14, 1984, an Allis-Chalmers (A-C) representative was contacted and he supplied information regarding part (4) of this allegation. Installation progress reports show the low pressure

~

(LP) II condenser-expansion joint weld was completed on September 18, 1978. The welding process was monitored during s

the complete weld out with micrometer readings taken at 12 points around the joint approximately every hour. The maximum movement 4.,

during welding was .26 inches (.660 mm) which was acceptable to -

A-C, W, and TUGC0. No NDE was required other than visual inspection. The same progress report shows that LP-I expansion -

joint welding was started'the same date. On October 13, 1978, Deviation Report No.14094 was written against the lower weld on this joint. The movement of .277 inches (7.030 M) exceeded that allowed by A-C. Consequently on October 19, 1978, the removal of that lowerweldwasbegun(some1,500inchesof3/4 inch [ weld). The micrometer logs (unnumbered) kept by TUGC0 Operations Maintenance Foreman show the rewelding started on November 13, 1978, and was completed on Novomber 17, 1984, and was controlled to a maximum movement of .23 inches (.584 mm). All welding was done to B&R A

weld grocedure 10046 andvWestinghouse direction for skip welding.

Considering these controls of the welding, it is unlikely in the judgment of the NRC staff that any undue or undesirable stresses were introduced into the welded joints. In addition to visual inspections, the only other nondestructive examinations that were performed were hydrostatic and vacuum tests.

Seven condensers hydrostatic test packages of tests conducted in

?

accordance wit @ test procedures jpecification 2323-M-23 were also reviewed. These tests spanned a cime period from December 1980+o April 1984. The retests of th were due to design changes.

Example: The last test, #IC0-002E performed in September 1983, involved sodium tracer injection and sampling at points through the

,5 -

condenser wall. In all cases the tests were successfully completed.

In May 1984, a vacuum and water box priming-retest 1 procedure No. ICP-AT-27-01-RT-1, was performed to reverify the ability of the main condenser to be evacuated by the vacuum pumps and hold vacuum for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The test was witnessed and accepted by the system test engineer and the results reviewed and accepted by the lead startup engineer; the manager, plant operations; TUSI nuclear engineering manager; and the manager, nuclear operations. This test too, was a successfully completed (see note on Exhibit A) review of these docu-ments, indicated no evidence of an overstress problem with the expansion joint or the piping connections. The conclusive evidence is that the condenser was constructed following approved c,onstruction and testing procedures and as such, will perform its designed function.

The condenser is a nonsafety-related component and is not needed for the safe shutdown of the plant.

I

5. Conclusion and Staff Positions: The NRC staff confirmed that in the beginning a number of fabrication problems were experienced but these l

problems appear to have been solved. It was also confirmed that retubing of the condenser with titanium tubes is planned to improve the design and operation. There was neither evidence that holes in d

the tube support sheets were misalign$et-by 3/8 inch nor evidence that the turbine to condenser was misaligned to the point that excessive stress was introduced.

9 This unit is classified nonsafety-related and is not essential for the safe shutdown of the plant. This allegation has neither safety significance nor generic implication.

6. Actions Required: None

8. Attachments: TUGC0 Office Memo's Condenser Tube Sheet Leakage 50-81051 SU-81081 SU-81134

9. Reference Documents: Repair Process Sheet - WDC Serial No. 693899, G&H Test Procedure Spec. 2323-M-23, Hydrotest Procedures:

IC0-002-1901-H0, IC0-002A-1901-H0, IC0-002B-5402-H0, IC0-002C-1901-HO, IC0-002E-1901-H0. Hydrotest procedure: IC0-002G-1901-H0, IC0-003-1901-H0. Vacuum and water box priming procedure flo. ICP-AT-27-01-RT-1. G&H Condenser Specification 2323-MS-23, six conversation records. A/C product service division deviation report #7136, AC product service field report Nos. 133, 148, 158',

and 170. Expansion joint weld out micrometer readings-unnumbered.

10. This statement prepared by:

Jack G. Corbett Date l

Reviewed by:

Group Leader Date Approved by:

Project Director Date O

e k

r Draft 1 - 9/20/84

'I AM-14/CP5 -

f SSER 1

/ h

. /

1. Allegation Group: Miscellaneous No. 12 V 4:

h //'

2. Allegation Number: AM-14 9 11 y

3. Characterization: It is alleged that one of the two Unit 1 diesel generators was damaged in May 1982. There were no specific provided. f

'l

4. Assessment of Safety Significance: The implied significance of this g allegation is that a damaged generator may not be able to provide the necessary emergency power to safety-related systems when called upon. MNO O

The NRC staff reviewed all of the nonconformance reports (NCRs) issued w

in May 1982 that pertained to the emergency diesel generators. Four J (E-82-005335, E-82-00560, E-82-006065, and E-82-004795) of the eleven NCRs reviewed documented equipment or instrument damage; however, the Y

• necessary corrective actions had been taken and the NCRs were closed appropriately.

Fo/ A-t r-rf Bi7

3 -2:N L

y*

Jt is likely that any damage that occurred in May 19 could affect the operability of the EDGs ia lisi vi ne catensi- n M A 1t 1 and testing u ai-ivuk p12ra in 1981 m icauli ef gener c EDG i M

problems" identified to the NRC by TDI and TDI EDG owners ) On August 12, 1983, the mak crankshaft on one of the three EDG's at Shoreham Nuclear Power Station broke into two pieces during a load test. There were several 10 CFR Part 21 reports issued by TDI reflecting a variety of minor and major defects; i.e., cracks in piston shirts, push rod cracks, governor drive coupling failures, potential failures in fuel lines, and dimensional problems with component fasteners and dowel pins. These defects were generic in nature, even though there are some design differences between EDGs at CPSES and those at other plants. During the course of the evaluation of Shoreham failure and the repairs of the Shoreham EDGs, infonnation related to the operating history of TDI engines and a QA program of the manufacturer was identified which called into question the reliability of all TDI ciesels. As a result of the foregoing and the generic implication involved, "0wners Group" consisting of representatives from affected nuclear power plants was formed for the purpose of investigating all aspects of quality and reliability of the EDG's supplied by TDI.

The Owners Group developed and published a generic inspections program in cooperation with each utility. The program addressed the specific concerns brought about by defects reported to the NRC by TDI EDG owners and 10 CFR Part 21 reports from TDI, as well as other areas, in order to develop adequate. confidence in these EDGs. In addition, TUEC expanded on some inspecticns, i.e., increase sample sizes, and inspected other areas on their own initiation.

TUEC conducted the inspections on both Unit 1 EDGs between February and June 1984. The results were transmitted to the Owners Group for evaluation and incorporation into the recertification process. The inspections included disassembly 3nd ncn-destruction examinations of parts using methods such as radiography, liquid penetrant testing, magnetic particle testing, visual inspections and measurements, eddy current testing, ultrasonic testing, and metal corporator testing.

The EDGs were reassembled after cleaning and the inspections and non-destruction tests were completed. This effort was closely controlled by procedure and QC surveillance. Parts which were previously identified as possibly generically defective were repl. aced.

Also parts found to have defects were replaced.

t

-4 e Upon completion of assembly, each EDG was retested by performing the entire portion of the preoperational test program which involved operation of the EDGs. The following tests were conducted, and satisfactory results were achieved:

ICP-PT-29-01, RF1 " Diesel Generator Auxiliary Systems, Retest 1" 1CP-PT-29-02 " Diesel Generator Control Circuit Functional and Start Test" 1CP-PT-29-03 " Diesel Generator Load Tests" 1CP-PT-29-04 " Diesel Generator Sequencing and Operational Stability Test" ICP-PT-29-05 " Diesel Generator Reliability Tests" The NRC Region IV Resident Inspector for Operations conducted inspections on nearly a daily basis starting with disassembly in February 1984 and ending with the witnessing of the testing in August 1984. NRC inspection effort included (but were not limited to) observation of the work and testing in progress, review of procedures used and compliance thereto, and tracking the work to ensure the Owners Group plan was followed and adequately document. This inspection effort was documented in NRC Inspection Reports 50-445/84-07, -15, -17, -18, and -20. The overall conclusion of these reports was that the recertification program was satisfactory completed.

[

[.

1 s

-5 g

5. Conclusion and Staff Positions: The NRC staff -found documented evidence supporting the alleger's concerns over damage to the EDGs in the four NCRs listed above. However, since appropriate corrective action was taken and also documented, the damage no longer exists.

Any damage affecting the relability and operation of the EDGs that may not have been documented would have been discovered and corrected v4 during the comprehensipfr recertification program described above.

Satisfactory completion of the above retests confirmed that the EDGs will perform in accordance with design.

The NRC staff, therefore, does not consider this allegation to have safety significance nor generic implications at this time.

6. Actions Required: None.

8. Attachments: None.

g

9. Reference Documents:

1. NCR's E-82-005335, E-82-00560 E-82-006065, and E-82-004795 regarding to damage to Emergency Diesel Generators during installation.

2. EDG Train A and Train B Owners Group Inspection Plans and Results, 1984.

3. 10 CFR Part 21, " Reporting of Defects and Noncompliance"

4. ICP-PT-29-01, RT-1, " Diesel Generator Auxiliary Systems, Retest 1"

5. 1-CP-PT-29-02, " Diesel Generator Control Circuit Functional and Start Test" 6: ICP-PT-29-03, " Diesel Generator Load Tests"

7. ICP-PT-29-04, " Diesel Generator Sequencing and Operating Stability Test"

8. ICP-PT-29-05, " Diesel Generator Reliability Tests"

10. This statement prepared by:

Name Date Reviewed by:

Group Leader Date Approved by:

Project Director Date

A /w q n .-r

.' U.S.Dsprrtmsnt cf Lcbar Employmtnt Sttndirds Administration g

(214)767-6294 Wage cnd Hour Division f' g' '.

i 1607 Main Street, Suite 200  ;  ;,

8 Dallas, Texas 75201 e September 26, 1984 Thomas Westennan Director of Enforcement Nuclear Regulatory Comission 611 Ryan Plaza Drive Arlington, TX 76011

. y o

Dear Mr. Westerman:

Enclosed please find our letter of detennination in the above-captioned matter.

Sincerely, s' . ) h, Curtis L. Poer Area Director Enclosure Fo/A-r5- c'1 8/ 8

l U. S. Dspartm2nt of Lcbar Employmsnt Stend rds Administrction "d " "' U# " '

(214)767-6294 1607 Main Street, Suite 200 h ,'

-) 1 i Dallas, Texas 75201

,,,] ./ 1 September 26, 1984 1 Stephen L. Hoech CERTIFIED Mall Manager of Employee Relations / Compliance P 423 878 265 Brown & Root, Inc. .

P.O. Box 3 Houston, TX 77001 Re: "Ko)

Brown & Root. Inc.

Dear Mr. Hoech:

This letter s of our compliance actions in the

\

above case. iled a complaint with the Secretary of -](o

Labor under the Energy Reorgaa zation on August 27, 1984. A copy of the complaint, a copy of Regulations, 29 CFR Part 24 and a copy of the pertinent section of the statute were furnished in a previous letter from this office.

Our initial efforts to conciliate the matter revealed that the parties would not at that time reach a mutually agreeable settlement. An investigation was then conduct ou investigation, the weight of evidence to date as a protected employee engaging in a protected '? Co indicates activity within the am to he Energy Reorganization Act, and that discrimina-tion as defined and prohibi ed by the statute was a factor in the actions which comprise M eomplaint. The following disclosures were persuasive in this determination:

The employees who were to be affected by the Reduction of Force were allegedly determined by the application of three criteria. These were:

(1) whether the employee had ever been denied unescorted access to nuclear fuel areas;(2) level of qualification; 3) dependability (i.e., absences of ld .

more than 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> for reason s other than vacation). Brown and Root maintains these criteria were objectively applied. However, least one other employee who'had precisely'the same qualificatior cate-gories (1) and (2), and who had more absences was not laid off, and this employee had not, to our know edge,g ven any infonna-tion to the NRC or any other regulatory agency.

This letter will notify you that the following actions are required to abate the violation and provide appropriate relief:

Reinstatement to former position, with salary, be'nefits, and '/ coj seniority uninterrupted; payment of lost wages; and payment of fees, expenses and damages as appropriate.

This letter will also notify you that if you wish to appeal the above findings and remedy, you have a right to a formal hearing on the record. To exercise

this right you must, within five (5) calendar days of receipt of this letter, file your request for a hearing by telegram to:

The Chief Administrative Law Judge U.S. Department of Labor Suite 700, Vanguard Building lill - 20th Street, NW.

Washington, DC 20036 Unless a telegram request is received by the Chief Administrative Law Judge within the five-day period, this notice of determination and remedial action will become the final o the Secretary of Labor. By copy of this letter I am f the detennination and right to a hearing. A copy o is er and e complaint have also been sent to the Chief 7CO)

Administrative Law Judge. If you decide to t a heart it will be necessary to send copies of the telegram to at 1607 Main Street, Suite 200, Dallas, Texas 5201 (214-767-6294). After I

( receive the copy of your request, appropriate preparations for the hearing can be made. If you have any questions do not hesitate to call me.

It should be made clear to all parties that the role of the Department of Labor is not to represent the parties in any hearing. The Department would be neutral in such a hearing which is simply part of the fact-development process, and only allows the parties an opportunity to present evidence for the record.

If there is a hearing, an Order of the Secretary shall be based upon the record made at said hearing, and shall either provide appropriate relief or deny the complaint.

Sincerely, m ) e rtis L. Poer Area Director g.

cc:

ARA /W1 Admin mi N.0.

Chief ALJ N

/ I

(/f. . h i ,, , - f l '

• ., o v

j SEER WRITEUP DOCUMENT CONTROL / ROUTE SHEET Allegation Numbers db . , ,,,e Subject of A11epation RMf W6fWMMShib Mdwn U1c El Kofbfb47st TRT Group [d1"l4 H ti A/D* // ~ %d/&Mk &D AW/ fab 9&ulbM Author

(' //. No #/r7& r' " '

V This sheet will be initialed by each reviewer. It stays with all revisions to the SSER writeup and serves as a routing a'nd review record. It will be filed in the work package when the writeup is, published.

Draft Number Dra f t- 1 2- 3 4 5 Author N

• fN

, Group Leader o//va rf 7 Wf f Tech. Editor 27 " '

• k)JG k'OR 4 fh<

Wessman/Vietti '

J. Gagliardo T. Ippolito -

Revision Number Final 1 2 3 4 5

Author Tech. Editor Group Leader -

J. Gagliardo y T. Ippolito #

l 'b Administrative ,

I Writeup integrated into SSER Potential Violations to Region IV

<AL s*

U

)b Workpackage File Complete Workpacks e e ned to Group Leader .

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k h;/jW,Q ye 9 1 JT*

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. . Draft 5 - 8/25/84 AC-41 Category 11/CP3 SSER

1. Allegation Group: Civil / Structural No. 11

2. Allegation Number: AC-41

3. Characterization: It is alleged that there was poor workmanship regarding the use of elastic joint filler material ("rotofoam") as a temporary spacer during construction to maintain the required air space between seismic Category I concrete structures.

4 Assessment of Safety Significance: TUEC informed the NRC Region IV o ce on November 23, 1977, of this allegation, which TUEC received anonymously in a telephone call on November 22, 1977. An NRC Region IV IE inspector reviewed the allegation during an inspection conducted between November 28 and December 2, 1977, and concluded, based on the information available to him at the time, that all temporary rotofoam had been removed from the areas identified. The matter was left open pending a Region IV review of the Brown & Root (B&R) QA/QC inspection and documentation program, which was being initiated to assure that the required seismic gap between Category 1 structures was being maintained.

Rotofoam is used as a temporary spacer during construction to maintain this gap. Once the concrete has hardened the rotofoam is removed to eliminate any load transfer or dynamic interaction between buildings. If the relative motion between buildings is small and the presence of

c

, Civil / Structural No. 11 rotofoam is considered in the dynamic analysis of 'the building, leaving r

therotofoaminplacemKpnothaveasignificantimpactonthedynamic performance of the buildings.

During an inspection between January 3 and 13, 1978, the inspector reviewed B&R procedure CP-QCI-2.4-9, " Inspection of Elastic Joint Filler Material Removal," Revision 1 (December 12, 1977) and B&R inspection reports for December 15, 1977 and January 3, 1978, and had no further questions regarding this matter.

The NRC Technical Review Team (TRT) attempted to obtain a fr-tier clarification of the concerns expressed by the-alleger; however, neither TUEC nor the NRC Region IV office has any records of the alleger's telephone conversation other than what is stated above. However, the TRT determined that prior to the time the allegation was made there was a misunderstanding as to whether the rotofoam should remain in place as part of the final construction. A letter from Gibbs & Hill (G&H) of September 6,1977 (GTT-1543) indicates that construction was proceeding on the basis that the rotofoam could be left in place. The letter further stated that this assumption was not made in accordance with the facility design drawings and design concept and .that expansion joints above grade should consist of a clear gap between buildings. As noted in the Gibbs & Hill letter, it was intended that the rotofoam be left in place below grade.

Civil / Structural No. 11 Based on discussions with TUEC and Gibbs & Hill engineers, it is also the TRT's understanding that the rotofoam was to be left in place for the expansion joints between the Safeguards Building and the Reactor Building. Since construction had proceeded above grade, TUEC instructed Brown & Root, in a letter of October 7, 1977 (TUS-5012), to remove the rotofoam. As noted, B&R procedure CP-QCI-2.4-9 was also implemented to verify removal of the rotofoam.

If properly implemented B&R procedure CP-QCI-2.4-9 should have provided an adequate inspection record for demonstrating that the air gap between buildings was adequately maintained. However, the TRT could only find two inspection reports relating to this procedure (the January 3, 1977

.and December 15, 1977 reports referenced). These reports do not fulfill the complete inspection requ'irements of CP-QCI-2.4-9. Furthermore, this procedure was deleted on July 18, 1978 (B&R memo IM-14835). A G&B memo of January 30, 1978 (GHF-2390) indicated that an inspection was made on November 23,.1977, and stated that the removal of the rotofoam from the subject areas was acceptable. However, the memo could only relate to construction at that point and did not provide any documented evidence of the inspections that were made.

A B&R interof fice memo of February 19, 1978 (IM-12934) discusses an inspection of the seismic gap between the Auxiliary Building and the Containment Building for Unit 1. It indicates that the removal of

'l Civil / Structural No. 11 rotofoam was not completed and requests further removal and/or engineering evaluation. TUEC engineers apparently did investigate this matter; however, there does not appear to be any formal documentation indicating the resolution of this matter.

Between September 14, 1978, and October 17, 1978, additional inspections of the air gap between seismic Category I structures were made by a B&R QC inspector. Six different areas were inspected. In five out of the six areas the inspector indicated unsatisfactory conditions due to the presence of foreign material in the air gap, such as wood wedges, rocks, clumps of concrete, and rotofoam. These unsatisfactory inspection reports were officially resolved on April 18, 1983, in response to NCR C-83-01067 (April 13, 1983). The disposition of this NCR notes that

" field investigation reveals that most of the material has been removed."

Based on discussions with TUEC engineers, it is the TRT's understanding ,

that field investigations were made but that no permanent records of these investigations were maintained. TUEC engineers did provide the TRT with five pages of field measurements made between March 15 and March 24, 1983, which indicated that investigations of the air gap between the Auxiliary Building and the Fuel Building were conducted. These measurements appear to indicate that the require'd air gap is not provided to the 813 foot, 6-inch elevation (the required evaluation in procedure CP-QCI-2.4-9). Even though the meausurements indicated a nonconforming e

T Civil / Structural No. 11 condition, TUEC could not provide any documentation indicating whether an engineering analysis was performed to justify this nonconformance or whether the material was subsequently removed. The TRT attempted to inspect the air gap between the structures but could not because in most cases the final joint sealer or roof flashing had already been installed.

In several areas between the Auxiliary Building and the Safeguards Building the air gap could be observed and appeared to be clear of any obstructions. In one doorway between the Safeguards Building for Unit 1 and the Auxiliary Building at the 830-foot, 6-inch elevation the air gap was clear to an observer looking up. However, a wooden board and other debris were observed when looked at straight in and down.

5. Conclusion and Staff Positions: Based on the above observations and document reviews, the TRT cannot determine whether an adequate air gap has been provided between the seismic Category I structures. In addition, it is not apparent that the permanent installation of rotofoam between the Safeguards Building and the Reactor Building and below grade for the other Category I structures is consistent with the seismic analysis assumptions and dynamic models used to analyze the buildings as delineated in applicable Final Safety Analysis Report (FSAR) sections.

The TRT, therefore, concludes that TUEC has not adequately demonstrated compliance with FSAR Sections 3.8.1.1.1 and 3.8.4.5.1, which require separation of Seismic Category I buildings to prevent seismic interaction during an earthquake.

i. Civil / Structural No. 11 Depending on the extent of conconformance with FSAR Section 3.8.1.1.1 and 3.8.4.5.1, the allegation is judged to have merit and potential safety

.s.

significance. Prompt remedial actions as delineated under Section 6 of this SSER should be implemented.

6. Actions Required: TUEC shall provide to NRC:

(a) the results of as-built investigations which demonstrate that adequate separation between all seismic Category I structures has been provided.

'~

(b) the results of analyses which demonstrate that the presence of rotofoam between the Safeguards Building and the Reactor Building of both units does not result in any significant increase in seismic response or alter the dynamic response characteristics of Category I ,

structures, components, and piping, when compared with the results of the original analyses.

l

,' (c) the results of analyses which demonstrate that the presence of rotofoam below grade between all Category I structures (as currently installed) does not result in any significant increase in seismic response or alter the dynamic response characteristics of all of the Category I structures, components, and piping, when compared with the results of the original analyses.

i.

Civil / Structural No. 11

_7_

4 .-

, , sJ

8. Attachments: None.

9. Reference documents:

1. IR 77-13, dated December 20, 1977

2. IR 78-01, dated January 30, 1978 --

3. CP-QCI-2.4-9, " Inspection of Elastic Joint Filler Material Removal."

Revision 1, dated December 12, 1977

4. B&R Inspection of Elastic Joint Filler Material Removal, dated January 3, 1978 and December 15, 1977

5. CTT-1543, dated September 16, 1977

6. TUS-5012, dated October 7, 1977

7. IM-14835, dated July 18, 1978

8. IM-12934, dated February 19, 1978

9. IRC-7706, dated October 17, 1978

10. IRC-7707, dated October 11, 1978

11. IRC-0320, dated September 14, 1978

12. IRC-0319, dated September 14, 1978

13. IRC-7705, dated September 20, 1978

14. IRC-7708, dated October 3, 1978

~

~

Civil / Structural No.11

15. NCR C-83-01067 April 13, 1983

16. B&R Field Measurements concerning "As-Built on Concrete Inside l Seismic Gap @A-F," 5 pages, dated between March 15 and 24, 1983

17. GHF-2390, dated January 30, 1978.

10. This statement prepared by:

C. H. Hofmayer

• Date Reviewed by:

L.C. Shao Date Approved by:

T. Ippolito Date