ML20203P335

Insp Repts 50-424/86-11 & 50-425/86-06 on 860224-28. Violation Noted:Inadequate Measures to Assure Correction of Design Criteria Documents.Deviation Noted:Failure to Inspect Supports

ML20203P335
Person / Time
Site:	Vogtle
Issue date:	04/14/1986
From:	Blake J, Girard E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20203P330	List: ML20203P329 ML20203P331 ML20203P332 ML20203P335
References
50-424-86-11, 50-425-86-06, 50-425-86-6, NUDOCS 8605070006
Download: ML20203P335 (25)

See also: IR 05000424/1986011

Text

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pa tico UNITED $TATES ,

p NUCLEAR REGULATORY COMMisslON +

y' >o,d, REGION 11,

g j 101 MARIETT A STREET, N.W.

• g ATLANTA, GEORGI A 30323

\..../

Report Nos.: 50-424/86-11 and 50-425/86-06

Licensee: Georgia Power Company

P. O. Box 4545

Atlanta, GA .30302

Docket Nos.: 50-424 and 50-425 License Nos.: CPPR-108 and CPPR-109

Facility Name: Vogtle 1 and 2

Inspection Co . Fe r 24-28, 1986

Inspect  : -

s' v fl 6

E. d Dat Si ned

Appr ,ed by: - 'I N 66

.A lake, Section Chief

~

Date Signed

deering Branch

sion of Reactor Safety

SUMMARY

Scope: This special announced inspection involved 41 inspector-hours on site in

the areas of licensee action on previous enforcement matters identified in

inspection of Readiness Review Module 4, review of construction deficiency

reports, and inspector followup items identified in inspection of Readiness

Review Module 4.

Results: Three violations and one deviation were identified - (1) Violation -

Inadequate measures to assure correction of design criteria documents, para-

graph 3.e. (2) Violation -

Failure to promptly identify undersize welds,

paragraph 3.J. (3) Violation - Removal of temporary pipe supports, paragraph 3.k.

(4) Deviation - Failure to inspect supports, paragraph 5. .

.

8605070006 860424 4

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REPORT DETAILS

1. Persons Contacted

Licensee Employees

• M. H. Googe, Project Construction Manager

• R. E. Folker, Project Quality Assurance (QA) Engineer

E. D. Groover, QA Site Manager, Construction

R. W. McManus, Readiness Review (RR) Discipline Manager, Construction

Other licensee employees contacted included construction cra ftsmen,

engineers, technicians, operators, mechanics, security force members, and

office personnel.

Other Organizations

• W.' C. Ramsey, Southern Company Services (SCS), RR Project Manager

• G. R. Trudeau, Bechtel Power Corporation (BPC), RR Special Assistant

R. D. Andrews, BPC RR Team Member

• C. R. Myer, BPC, RR Mechanical Design Team Leader

R. C. Somerfeld, BPC, RR Mechanical Construction Team Leader

J. Steele, Pullman Power Products (PPP) Quality Assurance Manager - Unit 1

W.M. Wright,SCS,RRDisciplineManager(Design)

R. A. Keidel, Bechtel National Incorporated (BNI), Manager - Material and

Quality Services (M and QS)

K. W. Caruso, BNI, M and QS Lead Welding Engineer

D. L. Carlson, BNI, M and QS Coordinator

NRC Resident Inspectors

• J. Rogge, Senior Resident Inspector (Operations)

R. Schepens, Resident Inspector

Other NRC Personnel

M. A. Miller, Vogtle Licensing Project Manager

H. L. Brammer, Reviewer (Pipe Break Criteria)

F. J. Witt, Reviewer (Post Accident Sampling System)

• Attended exit interview

2. Exit Interview

The inspection scope and findings were summarized on February 28, 1986, with

those persons indicated in paragraph 1 above. The inspector described the

areas inspected and discussed in detail the inspection findings listed below

except for the deviation. No dissenting comments were received from the

]

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licensee. Subsequently, in a telephone call on March 19, 1986, the

inspector informed the licensee of the below listed deviation.

a. Violation 424/86-11-01, 425/86-06-01, Inadequate measures to assure

correction of design criteria documents, paragraph 3.e.

b. Violation . 424/86-11-02, 425/86-06-02, Failure to promptly identify

undersize welds, paragraph 3.j.

c. Violation 424/86-11-03, Removal of temporary pipe supports, para-

graph 3.k.

d. Deviation 424/86-11-04, 425/86-06-04, Failure to inspect supports,

paragraph 5.

-The licensee did not -identify as proprietary any of the materials provided

to or reviewed by the inspector during this inspection.

3. Licensee Action on Previous Enforcement Matters

References: (a) Letter dated February 7, 1986, from D. O. Foster

(Georgia Power Company) to J. N. Grace (NRC Region II)

responding to unresolved and inspector followup items

described in NRC Inspection Report 424/85-35.

(b) Letter dated April 26, 1984, from D. O. Foster (Georgia

Power Company) to H. R. Denton (NRC) providing technical

information to justify a request for approval of

alternate pipe break criteria.

(c) Letter dated June 28, 1984, f rom T. M. Novak (NRC

Division of Licensing) to D. 0. Foster (Georgia Power

Company) providing an evaluation and acceptance of

alternate pipe break criteria.

(d) Telecopy dated January 22, 1986 from T. Bennet (Bechtel

LA) to C. Meyer (Readiness Review Design Team Leader)

stating that the Vogtle design contains no moaerate

energy Class I lines.

a. (Closed) Unresolved Item (424/85-35-01): Assurance of Adequate

Readiness Review Coverage of Module 4.

The concern expressed in this item was that the extent of activities

and commitments that should be covered by Module 4 was unclear and, as

consequence, important activities and commitments might be totally

omitted from the Readiness Review 12 activities and/or connitments were

identified which the inspectors stated they believed should be verified

as adequately addressed.

4

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3

In Reference (a) the licensee responded to the concern expressed in

this item, describing the review coverage intended for Module 4. In

addition, they briefly described where and how the Readiness Review

addressed the 12 activities /comitments that the NRC inspectors had

identified for specific verification of adequate coverage.

The NRC inspector examined this unresolved item during the current

inspection through review of the licensee's response and through

discussions with responsible licensee and NRC personnel who had

completed or evaluated the Readiness Review activities (Module reviews,

Independent Design Review, or reviews of Appendices) that the licensee

indicated would cover the 12 activities /comitments. The inspector

found that evaluation of many of the reviews that the licensee

indicated would cover the 12 commitments / activities had, as yet, not

been undertaken by the NRC. Examples included Modules 16 and 20;

Appendices C, F and J, and the Independent Design Review. The

inspector selected 3 of 12 commitments / activities which he found had

been examined by NRC personnel and examined the adequacy of their

coverage. The commitments / activities considered by the inspector were

as follows:

Number (from Report 85-35) Subject

(4) Welding Procedure and Welder Qualifica-

tions

(8) Document Control

(11) Piping Material Controls and Equipment

Maintenance (Construction Maintenance)

Based on his examination of the above, the inspector noted only one

apparently minor area which did not appear to be covered in the

Readiness Review. That area was the development of equipment

maintenance requirements. Performance of equipment maintenance was

covered in the review and the inspector considers that any serious

deficiencies in development of equipment maintenance requirements

should have been detected in the review. (Note: Equipment maintenance

as addressed here is not maintenance for plant operation, but

maintenance prior to operation.)

The inspector's examination of this item found that both licensee

Readiness Review and NRC personnel are sensitive to the potential for

omissions of significant matters from the review and that any serious

omission is likely to be detected and resolved. Therefore, the

inspector considers that it is not necessary to have this matter

separately identified for separate evaluation. The unresolved item is

considered closed.

.

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b. (0 pen) Unresolved Item (424/85-35-03): Design Control of Intermediate

Pipe Breaks.

The concern expressed in identification of this item was that it

appeared that the licensee had not implemented a particular design

commitment they made in obtaining NRC acceptance of changes to design

criteria for postulation of intermediate pipe breaks. As understood by

the inspectors, this design commitment was a provision to assure that

welded attachments lay at least five pipe diameters from any postulated

pipe break locations that would be eliminated in accordance with the

revised criteria.

The licensee's response to this unresolved item, provided in Reference

(a), contended that the commitment questioned by the inspectors had not

been a commitment. They indicated that the NRC acceptance of their

proposed change to criteria for postulation of arbitrary intermediate

pipe breaks was based instead on their compliance with ASME

Section III, Subsections NC/ND-3645, generalized requirements that the

design appropriately consider the effects of local welded attachments.

During the NRC inspection described by this report the NRC inspector

discussed and reviewed the licensee's Reference (a) response with the

Module 4 Readiness Review Mechanical Design Team Leader. In addition,

the inspector reviewed the licensee's Reference (b) submittal that

contained the apparent commitments related to the proposed change in

criteria for postulation of intermediate pipe breaks and reviewed the

Reference (c) NRC evaluation and acceptance letter. Subsequently, the

inspector asked the NRC Vogtle licensing project manager to contact the

cognizant NRC review personnel to determine the validity of the

licensee's contention that acceptance . had been based on their

compliance with NC/ND-3645 rather than on a specific commitment that

welded attachments would not be closer than five piping diameters to

postulated break locations. The NRC Project Manager informed the

inspector that the NRC reviewer stated that the licensee's response in

Reference (a) was incorrect. The licensee's statements in their

Reference (b) letter, including a statement indicating that welded

attachment would be at least five pipe diameters from postulated pipe

break locations, were considered commitments and served, in part, as

the basis for NRC acceptance of alternate criteria for postulation of

intermediate pipe breaks. This was confirmed to Region II in writing.

The inspector informed the licensee of the response obtained from the

NRC reviewer and stated that the criteria that had been and was being

used for postulation of intermediate pipe breaks should be provided for

review in a subsequent NRC inspection.

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The licensee informed the inspector on March 18, 1986, that they plan

to contact NRC licensing personnel to resolve this item. Region II

will verify the licensee's prompt attention to this matter. The item

will remain open pending Region Ils review of the licensee's actions in

resolving this item.

c. (Closed) Unresolved Item (424/85-35-04): Design Control of Moderate

Energy Class 1 Piping.

This unresolved item identified NRC inspector's concerns that the

'

licensee's design criteria documents did not contain or reference a

FSAR described commitment giving criteria for postulation of

'

through-wall leakage cracks in moderate energy Class 1 piping. The

Module 4 Readiness Review report listed the FSAR commitment as being

implemented in Design Criteria (DC) 1018. However, the inspectors

determined that the criteria were not in the DC-1018 document.

Although the licensee initially stated that they did have moderate

energy Class 1 piping to which the commitment criteria would apply.

They have since stated, in Reference (a), that they determined they

have no moderate energy Class 1 piping. They also acknowledged that

the subject commitment criteria should have been in DC-1018 and stated

that the DC was revised to include the criteria.

I

In the current inspection the NRC inspector verified the licensee's

internal communication, Reference (d), from their engineering organiza-

tion stating that they had no moderate energy Class 1 piping. The

matter is considered closed,

d. (Closed) Unresolved Item (424/85-35-05): Implementation of ANSI

N45.2.11.

This unresolved item was identified to expressed NRC inspector's

concerns that the licensee's Module 4 Readiness Review had not clearly

identified ANSI N45.2.11 as a commitment and it was not clear that

their design program adequately implemented this ANSI standard.

In the current inspection the inspector reviewed the licensee's

response to this item as described in Reference (a) and discussed the

matter with the Module 4 Readiness Review Design Team Leader. The

response indicated that ANSI N45.2.11 had not been identified as a l

commitment because the statements in the FSAR regarding its use were

general and the standard was applicable to other modules as well as

Module 4.

The inspector noted that general requirements to comply with ASME

Section III had been identified and addressed in Module 4 as commitment l

880, indicating an apparent inconsistency in the licensee's determina- '

tion of commitments. However, based on his discussions with the

Mechanical Design Team Leader, his examination of checklists and his

further examination of the ANSI N45.2.11 requirements, the inspector is

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satisfied that implementation of ANSI N45.2.11 was being adequately

addressed.

This item is considered closed.

e. (Closed) Unresolved Item (424/85-35-06): Adequacy of Preparation and

Revision of Design Criteria.

This item identified NRC inspector's concerns that the licensee's

Design Criteria documents contained errors and omissions indicating a

possible generic problem with the process of review and revision of

these documents. DC errors and omissions had been discovered by

Readiness Review personnel, as described in their findings 4-66, 4-67

and 4-75. In addition, the licensee had previously identified problems

with DCs not being updated in their INP0 Construction Project Self -

Initiated Evaluation (Finding DC.3-10) completed in 1983. Further, the

inspectors discovered four apparent omissions and errors in DCs 1018,

2702 and 1204 which had not been identified or corrected.

(1) A failure to incorporate revised intermediate pipe break criteria

and controls in DC-1018 (partially identified by the Readiness

Review Team (RRT) in Readiness Review Finding 4-75). This

specifically involves a failure to include the commitment that was

made in the licensee's April 26, 1984, letter (Reference (b)) in

requesting approval of alternate criteria for selection of

intermediate pipe breaks. Not included was a commitment that

arbitrary intermediate pipe break locations eliminated thru use of

the revised criteria be no closer to supports than five pipe

diameters.

(2) A failure to include criteria for postulation of through-wall

leakage cracks in moderate energy Class 1 piping in DC-1018.

(3) A failure to revise DC-2702 to reflect changed criteria for the

location of Residual Heat Removal (RHR) and Containment Spray (CS)

system containment sump post-accident sampling system lines.

(4) A failure to revise DC-1204, Section 6.0.2, to reduce the require-

ment for containment isolation valves in the lines from the RHR

sumps from two to one.

With regard to the above DC errors and omissions, the licensee

responded respectively:

(1) DC-1018 (Pipe Break Criteria - Interdiscipline) did not include

the provision that there be a five pipe diameter distance between

postulated breaks and pipe supports because this was not a

commitment. NRC acceptance required no changes other than those

addressed by finding 4-75.

7

(2) DC-1018 did not include provisions for postulating cracks in

moderate energy Class I lines because there are no lines in that

category.

(3) DC-2702 (Post Accident Sampling System) is nonsafety-related.

Drawings and actual installation correctly reflect the intended

design. The design indicated in DC-2702 is a former design that

was revised. DC-2702 was not updated to reflect the correct

design due to an oversite.

(4) DC-1204 (Safety Injection System), Section 4.0.0.1 reflects the

correct design, which was incorporated in a past revision.

DC-1204, Section 6.0.2, was not corrected due to an oversite.

In the current inspection, the NRC inspector reviewed the above

licensee explanations and agrees with all except (1). The basis for

his disagreement with the licensee's response is that a cognizant NRC

reviewer indicated that statements in the licensee's April 26, 1984,

letter (with regard to the distance of postulated breaks from supports)

were considered commitments. The inspector finds that the licensee has

not maintained his Design Criteria documents (documents used to provide

the primary bases for the licensee's design) up-to-date and correct.

The licensee had been aware of and taken actions to correct and assure

proper updating in response to their Self-Initiated Evaluating Finding

(DC.3-10) and their later Readiness Review Findings (4-66 -67 and ,

-75). However, their corrective action measures apparently did not

assure that the Design Criteria were up-to-date and correct as

indicated by the inspectors discovery of the errors and omission

described in (1), (3) and (4) above. This is considered a violation of

10 CFR 50, Appendix B, Criterion XVI, which requires that the licensee

establish measures which assure prompt correction of conditions adverse

to quality, such as the errors and omissions noted in the DCs. This

violation is identified 424/86-11-01, 425/86-06-01, Inadequate Measures

to Assure Correction of Design Criteria Documents.

f. (0 pen) Unresolved Item (424/85-35-09): Adequacy of Drawing and DCN

Reviews.

This unresolved item identified NRC inspector's concerns that the

licensee's Readiness Review of drawings and drawing change notices

(DCNs) was unsatisfactory in that, for the examples of review

performance examined by the inspectors, the following evidence of

unsatisfactory review was noted:

(1) For the licensee's drawing reviews the checklists used had few

items verified and the verifications performed did not appear to

be significant or thorough.

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(2) For the DCN reviews the inspectors found that, in one of four they

examined, the Readiness Review reviewer failed to note a

significant discrepancy between the DCN and the applicable DC even

though the reviewer indicated a specific verification that the DCN

change was in accordance with the DC.

During the current inspection the NRC inspector addressed this item

through review of the licensee's Reference (a) response, discussions

with the Module 4 Readiness Review Mechanical Design Team Leader and

other involved personnel, and examination of additional checklists.

The licensee's general response for this unresolved item an: for

Unresolved Items 85-35-11 and -12 below (which also involve Readiness

Review checklists) was that:

(1) The checklists were developed for application to several design

areas and inherently included items that were not applicable to

all of the areas.

l

l (2) The checklist items were not intended as absolute check require-

l ments, but rather as guidance to be used by experienced reviewers.

NOTE: The licensee concluded that their checklists adequately

served this purpose.

(3) The reviewers did not always clearly describe the reasoning behind

their checks and acceptances of checklist items making it

! difficult to verify the exact check performed.

NOTE: The licensee stated that their personnel have been given

additional training to assure they better document their

performance in subsequent modules.

In addition to the above, the licensee provided a specific response to

the NRC findings in Unresolved Item 85-35-09 giving a logical

explanation of how the checklists were performed. However, it was

still not clear to the inspector that the actual review performed had

been thorough or accurate due, particularly, to the lack of reviewer

documentation of his efforts and to the reviewer's failure to note that

a drawing change he reviewed was not in accordance with the DC

(DC-2702), a check the review specifically verified on the checklist.

The licensee's response to the latter discrepancy was that DC-2702 is

not safety-related. The inspector found this explanation inadequate as

the DC-2702 system is clearly important to safety, the concern was for l

its interface with the safety-related RHR system, and the reviewer

specifically recorded DC-2702 as having been verified in his notation

on the involved checklist.

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The Mechanical Design Team Leader commented to the inspector that many

of the drawing and DCN reviews that the inspector had not examined, had

• been more complete and well documented than those which the inspector

!

had previously examined. The inspector verified that this was the

.cause through examination of the following auxiliary feedwater system

(AFW) checklists:

Drawing Checklists (Module 4, Figure 6.1-2) for:

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Piping and Instrumentation Diagrams (P and ID)

1X4DB161-1, Revision 13

- P and ID 1X4DB161-2, Revision 12

DCN Checklists (Module 4, Figure 6.1-3) for:

- P and ID IX4DB161-1, Revision 13, DCNs 1 thru 38

- P and ID 1X4DB161-2, Revision 13, DCNs 1 thru 32

The inspector noted, however, that the important check for compliance

with the DC specified on each of there checklists lacked the detail to

provide evidence of a thorough review.

The licensee stated they would further verify the adequacy of the

reviews they had performed by re-reviewing several drawings and DCNs

and documenting the re-review in detail for further NRC evaluation.

Pending NRC evaluation of this re-review the unresolved item will

remain open.

g. (0 pen) Unresolved Item (424/85-35-11): Inadequate Review of Procure-

ment Specifications.

This item addressed NRC inspector's concern that the licensee's review

of procurement specifications in their Module 4 Readiness Review had

been inadequate and that the specifications might contain significant

unidentified deficiencies. This concern was based on the inspectors'

findings in examining the licensee's Module 4 review of Specification

X4AH04-(shop Fabricated Atmospheric Tanks...) which had been performed

by a licensee reviewer using their checklist 6.1-7. The inspectors had

found that a number of checklist items for the specification example

appeared not to have been performed satisfactorily and that the

specification contained the following apparent deficiencies which had

not been detected in the review.

(1) The specification failed to incorporate (fully) a requirement to

comply with Regulatory Guide 1.44 commitments.

(2) The specification failed to require compliance with ANSI N45.2.11.

(3) Radiation levels specified for some tanks did not include units.

I

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The inspectors examined this unresolved item through a review of the

licensee's Reference (a) response, discussions with the Module 4  ;

Readiness Review Mechanical Design Team Leader and other responsible i

1

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Readiness Review management, and examination of revision of the subject

specification to verify information described in Reference (a)

response.

l The licensee's general response with regard to the adequacy of the

reviews they had performed with their checklist is described in f.

above. With regard to the specific review addressed by this unresolved

'

item, they responded (in Reference (a) and as supplemented by

discussions) as follows:

'

(1) ANSI N45.2.11 does not apply significantly to the items covered by

the specification, as ASME Section III requirements are prescribed

and serve in place of the ANSI document requirements.

'

(2) The only requirement from Regulatory Guide 1.44 not addressed in

the specification was a requirement to aid in assuring that the

welding procedures used did not result in sensitization. This

assurance was obtained through a requirement that the manufacturer

,

'

submit his welding procedures for review and approval. Our review

of the submitted procedure verifies their adequacy.

l NOTE: The NRC inspector reviewed the welding procedures submitted

by one of the manufacturers and verified its apparent accept-

ability. These were the procedures for the valve encapsulation

vessels covered by the specification.

(3) The omission of the radiation level units is an error that

,

occurred in revision of the specification. It is not safety

significant, as proper radiation levels, including units, are

given in the Appendix EA attachment to the specification. The

omission was not detected by the reviewer because it was not an

item he checked.

NOTE: The inspector verified that the radiation level units had

been given in the previous specification revision and that the

current Attachment EA had the correct units.

1

The inspector considered that the licensee's explanation showed that

the subject specification was satisfactory. However, the inspector

considers that revfew performance that the licensee documented for this

j checklist lacks sufficient recorded detail for him to conclude that the

4

review was adequate.

As with item 85-35-09, the licensee stated they would further

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demonstrate the adequacy of their original review by reperforming

3

several of the specification reviews and documenting the reviews in

! detail. Pending NRC evaluation of the reperformed reviews, this item

will remain open.

,

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h. (0 pen) Unresolved Item (424/85-35-12): Inadequate Review of Vendor

Drawings.

This item addresses NRC inspector's concern that the licensee's review

of vendor drawings for Readiness Review Module 4 had beer: unsatis-

factory and that the drawings might contain deficiencies. This concern

was based on the inspector's findings in their examination of the

licensee's Module 4 review of vendor drawing 1X4AII04-23-13 (i.e.,

Revision 13), which had been performed by a licensee reviewer using

their checklist 6.1-8. The NRC inspectors found that it appeared that

few significant checklist items had been performed and the reviewer

failed to note that certain important drawing details (e.g., weld

sizes) were illegible.

In the current inspection the inspector reviewed the licensee's

Reference (a) response to the item, discussed the item with the

Readiness Review Mechanical Design Team Leader and QA personnel, and

reviewed additional vendor drawing submittals (AFW Pump drawings and

Revisions 11 and 12 of the encapsulation vessel drawing). The

licensee's general response with regard to reviews which they performed

with checklists, such as the review of vendor drawings, is described in

f. above. With regard to the drawing illegibility addressed by the

unresolved item, they responded as follows:

Suppliers are required to submit drawings of good microfilm

quality. Revision 13 of the subject drawing was accepted with

portions illegible because the portion revised was legible and

Revision 12 had been determined to have acceptable microfilm

quali ty. Revision 11 of the drawing had been returned to the

vendor because of poor microfilm quality.

In addition to the above, the licensee provided explanation as to ' low

the review checklist had been used, noting in several instances that

the reviewer had simply elected not to perform certain checklist items

and that he instead chose to verify other data, (but did not document

the checks on the checklist).

The NRC inspector verified that the licensee did have a legible copy of

drawing IX4AH04. The Revision 11 that had been rejected by the

licensee for unsatisfactory microfilm quality was reasonably legible

while the copy of Revision 12 shown to the inspector, (that was

accepted by the licensee) was found partly illegible. Based on his

examination of other vendor drawings (the AFW pump drawings) and on

discussions with licensee QA personnel who had checked additional

vendor drawings in response to Unresolved Item 85-35-12, the inspector

was satisfied that the drawing illegibility he had found appeared to be

isolated and of no safety significance.

<

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With regard to the explanation of checklist use provided by the

licensee the inspector informed the licensee that the reviews were

supported by such limited documentation that he could not make a

conclusive determination as to their adequacy. The licensee stated

that they would further verify the adequacy of the reviews they had

performed by re-reviewing several vendor drawings and documenting the

re-reviews in detail. They indicated that they would notify the NRC

when the re-reviews had been completed and they would be available for

NRC evaluation. Pending NRC evaluation ]f the re-reviews this item

will remain open.

1. (0 pen) Unresolved Item (424/85-35-13): Inadequate Resolution of

Readiness Review Design Verification Findings.

This item expressed NRC inspectors' concern that the licensee had not

adequately obtained correction for deficiencies their Readiness Review

Team (RRT) had identified in their Module 4 Readiness Review Design

Verification Findings. The inspectors had examined the licensce's

resolution of two of the RRT findings identified and the inspectors'

observations were as follows:

(1) Finding 4-75

This finding was that a calculation failed to postulate inter-

mediate pipe breaks in accordance with DC-1018. The licensee's

internal project response to the finding, which was accepted by

the Readiness Review Team (RRT), was that:

• -

the calculation was acceptable as it had been performed to a

NRC approved change in the criteria described in DC-1018

due to misplacement of a change notice, DC-1018 had not been

revised, it was corrected in response to Finding 4-75

actions had been taken to prevent recurrence of unincor-

porated DC changes

The inspectors determined that the licensee's RRT failed to

recognize that there had been conditions upon NRC approval of the

change that were not included in the correction of DC-1018. The

licensee's failure to recognize these conditions, which were

commitments made in obtaining NRC acceptance, is addressed in

Unresolved Item 85-35-03 in b. above.

(2) Finding 4-85

This finding was that Project Classes stated on the specification

and technical provisions for the RHR isolation valve incapsulation

vessel were incorrect. The licensee's internal project response

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13

to this finding (which was accepted by the licensee's RRT) was

that, while the incorrect Project Class had been identified on the

documents, the specific requirements given in the text assured

that the proper requirements were met. To assure that proper

Project Classes were indicated on other documents, the response

stated that eight specifications were checked and no other

discrepancies were found. The inspector questioned the adequacy

of the corrective action in the response because the inspectors

had identified cdditional examples of misclassification in a Field

Change Request (FCR), calculation, and a specification proposal.

The licensee's response to unresolved item was as follows with

regard to the inspectors' obscrvations for the two findings and

associated responses:

(1) Finding 4-75

The licensee stated their disagreement with the inspectors'

observations for this item referring to their response for

Unresolved Item 85-35-03 above.

(2) Finding 4-85

The licensee provided explanations for each of the examples

cited by the inspectors.

The NRC inspector reviewed and discussed the licensee's response

with the Readiness Review Mechanical Design Team Leader. The

inspector agreed with the licensee's explanations for Finding

4-85. The licensee's explanation for Finding 4-75 did not appear

correct, as described for the related Unresolved Item 85-35-03

above. Pending resolution of 85-35-03 this item will remain open.

j. (Closed) Unresolved Item (424/85-35-14): Undersized /0verground Welds.

This unresolved item expressed a concern described by an NRC inspector

to licenste management personnel in a meeting dated August 30, 1985,

that RHR isolation valve encapsulation vessel welds were overground and

under size. These welds were on vessel 1-1205-V4-001, drawing

IX4AH04-23-13, and were identified as welds 49 and 50 on the drawing.

In their Reference (a) response to this item the licensee stated:

(1) The hardware involved was not part of the Module 4 Readiness Review

sample.

(2) The inspection of the welds questioned had not been the respon-

sibility of the project, but instead that of the vendor.

(3) The weld condition was being addressed on deviation reports (DRs)

MD-8721 and 8723.

. .

14

During the current inspection the NRC inspector examined this item by

reviewing the deviation reports that the licensee had prepared

following the examination of the subject vessel welds performed in

response to this unresolved item. In addition, the inspector also

reviewed related information on DRs MD-8725 and 2264. The discre-

pancies that were described on the four DRs were as follows:

MD-8721 (dated February 1, 1986)

This DR confirmed and documented the undersize condition for

vessel 1-1205-V4-001, weld 50, that had been identified by the NRC

inspector. In addition, it identified similarly located welds

that were undersize on other :imilar Unit 1 vessels fabricated by

the same manufacturer.

MD-8723 (dated February 4,1986)

This DR confirmed and documented the undersize condition for

vessel 1-1205-V4-001, weld 49, that had been identified by the NRC

inspector. In addition, it identified similarly located welds

that were undersize on other similar Unit i vessels fabricated by

the same manufacturer.

MD-8725 (dated February 5, 1986)

This DR documented the licensee's identification of undersize

conditions on Unit 2 vessels and welds similar to those covered by

DRs MD-8721 and MD-8723.

MD-2264 (dated July 20, 1982 and resolved March 19,1985)

This DR documented the licensee's identification and correction of

six unsatisfactory condition related to welding on vessel

1-1205-V4-001. This included the identification and removal of

slag pockets in weld 49. The licensee failed to recognize the

undersize condition for weld 49 when they inspected it following

the removal of slag pockets from the weld.

Based on the above, the inspector finds that the licensee failed to

comply with 10 CFR 50, Appendix B, Criterion V requirements that

activities affecting quality be accomplished in accordance with

documented procedures, instructions, or drawings, in that vessel

1-1205-V4-001 welds 49 and 50 did not comply with size requirements

given on drawing IX4AH04-23-12 and similarly located welds on similar

Unit 1 and 2 vessels also did not comply with drawing size require-

ments, as described in licensee deficiency reports MD-8721, 8723 and

8725. In addition, the licensee also failed to comply with 10 CFR 50,

Appendix B, Criterion XVI requirements that measures be established to

assure that deficiencies are promptly identified and corrected, in

that:

.

15

(1) Although the licensee identified (July 20, 1982) and corrected

complete March 19, 1985 six welding related deficiencies on vessel

1-1205-V4-001, including a deficiency on weld 49 (which they

visually re-inspected) they failed to detect and correct the

undersize weld condition.

(2) Licensee management was not prompt in identifying the condition in

that they did not identify the condition on a deficiency report

until February 1986, over five months after it was reported to

them by an NRC inspector on August 30, 1985.

The licensee's failure to comply with drawing requirements and their

failure to promptly identify the noncompliance with drawing require-

ments is identified as violation 424/86-11-02,425/86-06-02, Failure to

Promptly Identify Undersize Welds.

k. (Closed) Unresolved Item (424/85-35-15): Undocumented Piping Supports.

This unresolved item expressed NRC inspector's concern that structural

members supporting piping inside vessel 1-1205-V4-001 were not depicted

on the drawing and the basis for and controls on the installation of

these supports could not be readily determined.

The licensee's Reference (u) response to this item was as follows:

The piping and valve supports questioned by the inspectors are

temporary supports for shipping purposes. They are shown on

Bechtel approved vendor drawing number B-81-22. Revision 0 (vendor

document log number AX4A404-90-1). The drawing includes

instructions to remove the supports.

The RHR system has been turned over to Nuclear Operation for

pre-operational testing and the work completion checklist

identifies the vessel as requiring additional work as documented

by work item BC0626 and work order 18513301. The control of work

activities during the pre-operational test phase is described in

Module 3A which was submitted to the NRC on May 1,1985 with a

revision submitted on July 29, 1985.

During the current inspection the NRC inspector reviewed the licensee's

response to the item and questioned licensee personnel as to what

specific controls they had to assure that the subject temporary pipe

supports were removed at an appropriate time.

I

- .

16

The inspector was referred to and reviewed the following documents:

- Nuclear Plant Maintenance Work Order, Control No. 18513301, Work

Item BC-0626

The inspector found that this document only stated, " Secure

Encapsulation Vessel 1-1205-V4-001 and perform test 1-1205-10".

The document neither provided or referred to any criteria that-

assured removal of the supports.

- Field Process Sheet for Mark No. V-1-1205-V4-001, dated

October 27, 1981, and entitled " Process Sheet for Inspection

and Cleaning of Encapsulation Vessels. Also, to Allow Removal

of Shipping Braces and Installation of Electrical Penetration

by others".

The inspector found that this process sheet had included an

instruction for removal of the temporary supports (support

braces). The instruction for support removal was not performed,

but was accompanied by a modifying note (dated April 22,1982)

that stated the instruction was to be performed after welds were

complete. The step was never performed and there was no evidence

of it having been transferred to any other document for

performance. The field process sheet was considered complete and

had received final QA approval. There had been no requirement for

a QC inspection to verify the support removal.

In his review of the above and in discussions with licensee personnel,

the inspector found no documented criteria that would assure removal of

the subject pipe supports. This is considered noncompliance with

10 CFR 50, Appendix B, Criterion V requirements that documented

instructions, procedures or drawings be provided and used to assure

proper accomplishment of activities affecting quality, such as removal

of the supports. This noncompliance is identified as violation

424/86-11-03, Removal of Temporary Pipe Supports,

l '. (Closed) Unresolved Item (424/85-35-17): Inadequate Resolution of

Readiness Review Findings.

This item expressed NRC inspectors' concern that the licensee had not

adequately resolved Readiness Review Module 4 Construction RRT

findings. The concern was based on the inspectors' determination that

the resolution of finding 4-83 stated in the Module 4 report was

incorrect.

In their Reference (a) response the licensee stated that the response

provided in the Module 4 report had been based on certain correspon-

dence that was subsequently found incorrect. The response indicated

that the determination that the response was incorrect was not

completed until af ter the Module 4 report had been submitted to the

NRC. In the current inspection the NRC inspector verified that the

-_

.

17

licensee had become aware of the error, as stated in their Reference

(a) response. The licensee's resolution of the finding appears

adequate, though belated, and the matter is considered closed.

4. Unresolved Items

Unresolved items were not identified during this inspection.

5. Construction Deficiency Reports (CDRs)

(0 pen) Item 424, 425 CDR 83-41: Embed Plate Base Metal Failure

References: (a) Letter dated July 11, 1983, from R. E. Conway (Georgia

Power Company) to J. P. O'Reilly (NRC Region II)

providing on interim report on investigation of an embed

plate base metal failure.

(b) Letter dated February 15, 1984, from D. O. Foster

(Georgia Power Company) to J. P. O'Reilly (NRC

Region II) stating that the previously described embed

plate base metal failure had been determined not to be

reportable in accordance with 10 CFR 50.55(e).

(c) Georgia Power Company Nonconformance Report MD-4266,

dated April 9,1983, documenting separation of weld from

embed plate that occurred on piping support

V1-1901-043-H019.

(d) V. H. Wadhwani, Bechtel Group, Incorporated,

" Investigation of Weld Failures Between Welded Embed

Plates and Steel Support Tubes at the Vogtle Jobsite",

June 1983.

(e)

Letter dated July) 5,1983,

Power Corporation from M.

to J. A. Bailey MalcolmCompany

(Southern (Bechtel

Services) describing the status of the investigation of

the embed plate base metal failure.

(f) Letter dated August 1,1983, from M. Malcolm (Bechtel

Power Corporation) to D. O. Foster (Georgia Power

Company) describing the results of a meeting to discuss

a field inspection program to be conducted to further

investigate the pipe support weld failure (the embed

plate failure).

(g) Letter dated August 2,1983, from H. H. Gregory (Georgia

Power Company) to M. Malcolm (Bechtel Power Corporation)

describing the results of a survey to identify flare

bevel welds made to embeded and surface mounted plates.

r

e

.

18

(h) Letter dated September 12, 1983, from M. Malcolm

(Bechtel Power Corporation) to H. H. Gregory (Georgia

Power Company) describing the field inspection program

to be performed for further investigation of the pipe

support failure.

(1) Letter dated January 19, 1984, from M. Malcolm (Bechtel

Power Corporation) to J. A. Bailey (Southern Company

Services) describing the completed engineering

evaluation of the potential construction deficiency

involving the pipe support weld failure.

(j) " Commentary on Highly Restrained Welded Connections",

Reprint, AISC Journal, 61-78, Third Quarter /1973.

(k) L. F. Porter, "Lamellar Tearing in Plate Steels (A

Literature Su rvey )", United States Steel Corporation

Research Laboratory Technical Report, August 29, 1975.

(1) U.S. Nuclear Regulatory Commission, " Potential for Low

Fracture Toughness and Lamellar Tearing on PWR Steam

Generator and Reactor Coolant Pump Supports", USNRC

Report NUREG-0577 for comment, October 1979. Available

for purchase from USNRC Division of Technical

Information and Document Control, Washington, DC 20555.

(m) E. J. Kaufmann, A. W. Pense and R. D. Stout, "An

Evaluation of Factors Significant to Lamellar Tearing",

Reprint, Welding Research Supplement to the Welding

Journal, 43s-49s,liarch 1981.

(n) R. D. Stout and A. W. Pense, "Causes and Prevention of

Lamellar Tearing", Civil Engineering - ASCE, 74-75,

April 1982.

I

(o) G. A. Knoroveki, R. D. Krieg and G. C. Allen, Sandia

National Laboratories, " Fracture Toughness of PWR

Components Supports", USNRC Report NUREG/CR-3008,

February 1983. Available for purchase from National

Technical Information Service, Springfield, VA 22161.

(p) U.S. Nuclear Regulatory Commission, " Potential for Low

Fracture Toughness and Lamellar Tearing in PWR Steam

Generator and Reactor Coolant Pump Supports", Rev.1,

October 1983.

(q) J. L. Grover and R. C. Cipolla, Aptech Engineering

Services, incorporated, "The Significance of Lamellar

Tearing ir ".tructural Steels" Electric Power Research

Institute kcaort NP-3570, June 1984.

. - - -

- .

19

On June 6,1983, the licensee notified NRC Region II by telephone of a

support failure which they had determined was potentially reportable as a

construction deficiency in accordance with the requirements of

10 CFR 50.55(e). In their written report, dated July 11,1983 (Reference

(a)), the licensee informed NRC Region II that the potential deficiency

concerned a failure of a support constructed of rectangular tubular steel

welded to an embed plate. The support construction had involved welding one

flat face of a piece of tubular steel against the embed plate face utilizing

flare bevel welds placed along the two rounded edges of the tubular steel

that lay adjacent to the tubular steel / embed plate interface. The failure

occurred when a craf tsman stepped on the support. In a letter dated

February 14, 1984 (Reference (b)), the licensee informed Region II that they

had completed their investigation of the support failure and had determined

that it was not reportable in accordance with 10 CFR 50.55(e).

During the current inspection the NRC inspector examined the licensee's

investigation of the support failure through a review of literature on

lamellar tearing, a review of the licensee's documentation of their

investigations of the failure and discussions with cognizant licen3ee

personnel. The literature and licensee documentation reviewed are listed as

references above. Significant information and findings obtained by the

inspector in his review and discussions is summarized as follows:

a. The licensee reported that the failure occurred when an individual

stepped on the tubular steel portion of a pipe support fabrication from

tubular steel that was welded to an embed plate utilizing flare bevel

welds with fillet weld reinforcement.

b. The licensee indicated that failure occurred in the embed plate base

metal. The fracture ran the full length of the welds and pulled out

plate base metal about 1/16 inch deep,

c. The licensee reported that weld size was excessive, greatly exceeding

the size specified by the design drawing. The drawing specified

unreinforced flare bevel welds to join the tubular steel to the plate

(one on each side of the tube). The licensee found that the welds used  ;

in practice had been flare bevels plus fillet weld reinforcements of

over 1/2 inch. The welds joined 3/8 inch thick (6w x 3d) tubular steel

to a inch thick plate.

d. Based on a metallurgical evaluation of the failure, the licensee

concluded that the failure was the result of lamellar tearing caused by

the large shrinkage stresses from heavy welding, compounded by embed

plate inclusions and ferrite banding close to the plate surface.

e. Based on his review of the licensee's metallurgical investigation

report (Reference (d)) and relevant literature (Reference (j) through

(g)) the inspector found that the licensee's conclusions in d. above

were amply supported, except that the literature does rot support

ferrite banding as a factor.

%... .

_

. _ . _ _ . _. __

-

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4

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20

,

f. In their July 11, 1983, written report (Reference (a)) on this item the

licensee presented a summary of the results from their metallorgical

investigation and also. described a test which they performed at the

site to investigate the failure. This test involved four similarly

,

welded (tubular steel welded to plate) supports which were visually

i' inspected, magnetic particle examined, and load tested to destruction.

For the testing, the licensee removed the weld from one tube edge / plate

interface on each of the supports and loaded the supports to failure ,

4

utilizing a hydraulic ram and a wedge. They noted that extensive

effort was required for each destructive failure. One of the four

samples exhibited magnetic particle indications an.d it failed by

lamellar tearing. Licensee personnel estimated that'the force required

to produce the destructive failure was well in excess of that required

! by design,

g. The licensee's July 11, 1983, letter stated that the following

additional investigation would be conducted to verify that lamellar

tearing had not occurred elsewhere:'

4

- A field walkdown of supports which utilize,weldments of similar

size and type will be conducted to confirst that lamellar tearing

i has not occurred elsewhere. Flare bevel' welds with reinforcing

fillet welds 3/8" and larger, on plates one inch or thicker will

'

be identified and evaluated based upon a detailed visual

inspection program. This inspection will include, as a minimum, '.

samples of weldments made to the embed plates fabricated from heat, ".

numbers 7417461 and 7419919. These are the heat numbers

respectively of the embed plates to which the subject support an'd

!

the fourth support tested to destruction were attached. In

,

addition, appropriate samples will be inspected using magnetic

'

particle testing to assure a higher confidence level on the

integrity of the existing installation. The 3/8" fillet

,

reinforcement is considered an appropriate threshold for the

initial scope of this investigation.

'

-

The final scope of this investigation will be determined after a

detailed assessment has been made of the proposed field walkdown

and MT sampling results have been obtained.

i h. In their internal letter of January 19, 1984, (Reference (i)), the

,

licensee described their completed evaluation of the support failure as

'

a potential construction deficiency >fn accordance with 10 CFR 50.55(e).

The investigation that the licensee performed to verify that lamellar

! tearing had not occurred elsewhere was described in that letter.

'

Significant aspects of the investigation, as determined by the NRC

inspector from his review of the letter, were as follows:

-

The investigation was conducted through visual and magnetic

particle examination of a sample of supports for evidence of

lamellar tearing. ,

4

, _ _ , _ , _ - . ._~_.,...,_~_._.~--,,_._,-_._..,y ,._._.y. . , , , _ . . . . . ~ _ . , _ _ . _ , . - , - - _ , . _ _ , , , , . _ _ _ _ . , , _ , _ _ . . . . ~ . , _ _ _ . - . , . - , . . . _ . , . , . _ , . _ . , . . , - - -

. - .

_

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21

-

The population of supports from which the sample for examination

was limited to previously inspected supports involving flare bevel

welds made to embed plates and surface mounted plates. This total

population consisted of 447 supports.

- Examination was performed on a sample consisting of 53 supports

selected at random on a statistical basis from the stated

population.

-

The following finding and conclusions were r.)orted:

No lamellar tear indication was identified in the inspection

of the samples.

There is a 95% confidence level that at least 95% of the

flare bevel welds included in the total population considered

will not have any lamellar tear indications.

There is one half of one percent probability of exceeding 1/2

inch overweld in the field. It should be noted that the

subject support that failed had 1/2 inch fillet reinforcement

which was not called out in the design drawing. It is

therefore concluded that significant overwelding is a rare

occurrence.

- The evaluation of the field inspection program results have

confirmed that the weld failure of the subject support was an

isolated event.

-

Examples of supports fabricated from plate heat numbers 7417461

and/or 7419919 were ot identified and included in the investiga-

tion. As noted in 9. above, the licensee had informed the NRC

that weldments made to embed plates fabricated from those heats

would be inspected as a minimum,

i. The licensee's internal letter of January 19, 1984 (Reference (i))

stated that the conditions that had made the original failed support

plate susceptible to lamellar tearing were considered highly unlikely

to exist uniformly or repeat within a given heat of plate material.

NOTE: No reference or data was provided in support of this statement.

k. Significant related data obtained by the inspector from literature on

lamellar tearing is as follows:

-

Lamellar tearing is a form of cracking that occurs in planes

essentially parallel to the rolled surface of a plate under high

thru-thickness (welding induced) loading. It tends to initiate

thru cracking or decoherence of elongated inclusions. (Reference

(o))

'

,

. .

22

- No relationship has been established between lamellar teaYing and

banding. (Reference (j))

- The zone of decohesion responsible for lamellar tearing generally

extends from the lower part of the weld heat af fected zone to 1/4

inch below the plate surface. (Reference (k))

- Most of the time lamellar tearing occurs beneath the plate surface

where it is best detected ,by ultrasonics (Reference (n)).

Ultrasonic examination is the most suitable method for detecting

lamellar tearing in completed weld joints. (Reference (j))

- Plate material susceptibility to lamellar tearing is strongly

dependent on base metal compositions, particularly on sulfur and

carbon content. (Reference (g))

NOTE: Based on licensee analysis cf the failed embed plate if

appears susceptible to lamellar tearing.

-

Only one documented inservice failure can be attributed to

lamellar tearing and this was not in a nuclear application.

(Reference (p))

-

Preliminary tests suggest that incipient lamellar tears buried

from view do not lower static strength of the joint seriously

unless they have propagated sufficiently to be detected 'by

ultrasonic examination. (Reference (m))

In reviewing the licensee's actions on this matter, the NRC inspector

found that the licensee's investigation was generally technically

sound. However, the inspector noted two factors that the licensee did

not appear to adequately consider to assure that they did not have

additional supports that exhibited lamellar tearing: .

- The licensee did not identify and inspect or test any. examples of'

supports fabricated from the heat of plate that originally failed,

such that they might verify that the plate metallurgical

characteristics had not resulted in additional examples of

lamellar tearing. As described in g. above, the licensee had

specifically reported to NRC Region II that additional support

examples from the failed heat of plate would be inspected. (The

licensee did not subsequently netify, Region II of their decision

not to inspect additional examples of the heats of plate that had

exhibited lamellar tearing.)

-

The preferred method for detection of lamellar tearing is

ultrasonic examination. The licensee employed magnetic particle

examination (rectf fied alternating current using prods) in their

investigation.

'

, ,

e

• n.

f

. - . . - - + - , , - , . -. . - - - - - . - - - - - - - - - - - - - - - _ - - - - . -

_

.

23

NOTE: The use of ultrasonic examination may not have been

practical for the support configurations involved in the

investigation but the inspector saw no pending evidence that its

use had even been considered. Pending further NRC examination of

the licensee's embed plate installation records and evaluation of

the examinations performed by the licensee to detect lamellar

tearing in their investigation, item 424,425 CDR 83-41 will remain

open. The licensee's failure to identify and inspect samples of

weldments made to embed plates fabricated from heat numbers

7417461 and 7419919 is considered a deviation from the commitment

made in the licensee's letter of July 11, 1983. This deviation is

identified 424/86-11-04, 425/86-06-03, Failure to Inspect

Supports.

6. Inspector Followup Items (IFIs)

a. (Closed) IFI (424/85-35-02): Omission of Westinghouse Offsite

Activities for the Readiness Review.

This item identified NRC inspector's concerns that the licensee had

omitted the offsite activities of the Nuclear Steam Supply System

suppliar (Westinghouse) from the Readiness Review.

In their response (see paragraph 3 above, Reference (a)) to this IFI,

the licensee acknowledged and described their basis for omitting

Westinghouse offsite activities from the Readiness Review. The

licensee's position on the matter is clear and will be acknowledged in

the final NRC report on Module 4. The inspector considers that no

further inspection of this item is necessary.

b. (Closed) IFI (424/85-35-07): Calculation Corrective Actions.

This item was opened by NRC inspectors for followup to assure that

corrective actions with regard to minor discrepancies in calculations

had receivr;d adequate attention.

During the current inspection the NRC inspector examined this item

through a review of the licensee's written response (Reference (a) in

paragraph 3 above) and discussions with the Module 4 Readiness Review

1

Design Team Leader. The inspector accepted licensee explanations

citing the lack of safety significance in the subject calculation

y discrepancies.

"

c. (Closed) IFI (424/85-35-08): Maximum Design Pressure Discrepancy

This IFI was opened to address an apparently minor discrepancy that the

inspector's observed in comparing a maximum design pressure given on an

isometric drawing with that indicated for the same line on the

licensee's Line Designation List.

,

- _ . - -. - - - - - ----- ._ - .-. _---.

. .

24

'

s

During the current inspection, the NRC inspector followed up on this  ;

item by reviewing the II:ensee's response to the item (Reference (a) in

paragraph 3 above), discussing the item with the Module 4 Readiness

Review Mechanical Design Team Leader, and verifying correction of the -

. design pressure entry given on the isometric drawing. The licensee's

explanation of the discrepancy indicated that the pressure given on the

drawing is not used in stress analysis and had no impact on the system

design. This explanation was accepted by the inspector and the matter

is considered closed.

!

d. (Closed)IFI(424/85-35-10): Review of Construction Specifications and

'

Procedures. i

This item expressed minor concerns with regard to review coverage

provided by the Module 4 Readiness Review of the construction

specification and of construction procedures for receipt, storage, and

maintenance. The inspector's who identified this IFI based their

concern on the fact that there was no indicatten of a review of these

documents for commitment implementation in the Module 4 Readiness

Review report commitment section.

During the current inspection, the NRC inspector examined this item

through review of the licensee's written response (see paragraph 3

Reference (a)) and discussions with cognizant Readiness Review

<

personnel. The licensee's response pointed out that the subject

construction specification and procedures had been addressed in

development of Readiness Review Module 4 report, Section 4 which

provides the description of the licensee's program and includes '

, descriptions of procedures for receipt, storage and maintenance. The

response also noted with regard to the construction specification that

they had identified a significant finding (Finding 4-56) and a -

Readiness Review Information Request (RIR 4-6) which were evidence of

,

their review. The inspector verified the finding and the description i

of the procedures for receipt, storage and maintenance given in the

^

,

Module 4 report and accepts the licensee's explanation. The matter is

! considered closed.

i

! e. (Closed) IFI (424/85-35-16): Inconsistency in the use of low carbon

stainless steel.

.

This IFI was opened by the NRC inspectors to follow-up on an apparent

inconsistency in the composition of piping materials used in the RHR

isolation value encapsulation vessel. The inconsistency was that all

-

of the pipe was a low carbon stainless steel, while the valve itself

was not.

During the current inspection, the NRC inspector followed up on this item I

by verifying that the material specified for the valve in the design

was not low carbon stainless steel. The inspector determined that the

materials were as specified by the designer. The matter is considered

closed.

_

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