IA-86-416, Special Review Team Rept. Related Documentation Encl

Special Review Team Rept. Related Documentation Encl

ML20212N817
Person / Time
Site:	Comanche Peak
Issue date:	07/13/1984
From:	NRC - COMANCHE PEAK PROJECT (TECHNICAL REVIEW TEAM)
To:
Shared Package
ML20212N814	List: IA-86-416, Special Review Team Rept. Related Documentation Encl ML20212N812 ML20212N817
References
FOIA-86-416 NUDOCS 8608290195
Download: ML20212N817 (91)

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TABLE OF CONTENTS I Executive Summary and Conclusion II Background III Review Approach

( IV Review Findings A. - Management Organization B. - Quality Assurance / Quality Control C. - Equipment Turnover and Preoperational Testing C. - Electrical SE E. - Design Activities / Control __

3 F. - Installation of Safety Related Fluid Systems -. .

4: G. - Civil Construction , -

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H. - Heating, Ventilation, and Air Conditioning Systems -

I. - Formai Interviews with QA/QC' Personnel (General) ,

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i 3-I. EXECUTIVE

SUMMARY

NRR in coordination with the Director of IE and the Region II & IV Admini-strators fonned a team to perforn a limited unannounced review of Comanche Peak. The purpose of the review was of 1) evaluate the current implementa-tion of the applicant's management control of the construction, inspection and test programs, 2) provide an indepth understanding and background

- infomation to the NRC new management team established by the Executive Director for Operations memorandum of March 12,1984, and 3) obtain infoma-tion necessary to establish a management plan for resolution of all out-standing licensing actions.

The team consisted of eight reviewers, a team leader and team manager. The reviewers and team leader were selected from the Region II staff. The manager was the NRR Comanche Peak Project Director. The team was assembled in Region II headquarters where it was briefed by NRR, IE and ELO. _

Y. The team conducted its review from April 3 to April 13, 1984 The review"

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consisted of an audit of significant elements and processes of the appli-cant's management control in construction, inspections and testing of systems important to safety. These included:

1. Component and material receipt inspection and control.

2. Structure, systems, and component fabrication and installation.

3. Structure, system, and component acceptance, and preoperational testing.

4. Quality assurance and control documentation and procedures to effect i items 1 through item 3 above. ,

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! The portions of the system evaluated included piping, pipe and component

' supports, instrumentation and control, electrical cable separation and cable tray supports, component qualifications, and allegations relating to these areas. _

The reviews also included briefings from the Applicants' management and interviews with QA/QC, Document Control, and craft personnel. The total effort was conducted with little or no advance notice of areas, personnel or documentation to be reviewed.

l Each member of this team was chosen because he had both many years experi-

! ence in the discipline he was reviewing, and he had performed evaluations at

! a wide range of nuclear facilities. The team spent over 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> per-

forming this review. The following is a list of the special review team members, their positions, and field of expertise

Paul Bemis, Section Chief, Management Organization, Qualification and Training Paul Fredrickson, Project Engineer, Quality Assurance / Quality Control, Bill Orders, Senior Resident, Preoperation and Startup Kim VanDoorn, Senior Resident, HVAC and QC inspector interviews i

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Al Ruff, Reactor Inspector, Electrical Louie Jackson, Reactor Inspector, Quality Assurance / Quality Control Winston Liu, Reactor Inspector, Design Activities / Control Ed Girard, Reactor Inspector, Welding and Metallurgy j Joseph Lenahan, Reactor Inspector, Civil and Structures i The teams findings indicated that the apolicants management control over the i construction, inspection, and testing programs is generally effective and is

! receiving proper management attention. The findings identified three i potential enforcement actions (See Sections B&E); two areas of weakness 2 requiring Applicants management attention; (See Section B) and seven areas where Applicants activities exceeded normal and accepted practice (See

Sections A, B & E). The team also found improvements in the relationship between the current QA/QC management and inspectors which in the past has caused consnunication problems (See Section I). The team believes that the F results of this limited revi,ew reveal the plant is being built in a safe __

y manner. '

c. .

The findings and conclusions of this report of the teams review should not '

be construed as resolving any of the issues identified by the ASLB hearings, allegati::ns, or staff concerns of the design adequacy of the plant.

II. Background On March 17, 1984, the ED0 directed NRR to manage all NRC actions leading to

/l licensing decisions for Comanche Peak and Waterford. The purpose is to assure the overall coordination and integration of the outstanding regula- -

tory actions and achieving their resolution prior to a licensing decision.

This effort is to encompass all licensing, hearing, inspection and allega-tions issues.

1 Soon thereafter, the newly established Comanche Peak project team found that -

< there was a need to 1) obtain current infonnation relative to the management control of the construction, inspection and test programs and 2) obtain information necessary to establish a management plan for resolution of all outstanding licensing actions. To help achieve this objective expeditiously and objectively it was decided that an unannounced review of Comanche Peak i plant was necessary. As a consequence,'NRR in coordination with OIE and the

Region II and IV Administrators fonned a review team. Because of resource

limitations in Region IV, the team was staffed with Region II personnel.

J The team was assembled in Region II Headquarters on April 2, 1984. The team

) was briefed on significant issues raised as a consequence of the licensing l review, the hearing contentions and the allegations. The team leader and the reviewers were not provided with the names of the allegers in order to l assure their confidentiality. The team conducted their review from April 3 t to April 13, 1984.

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5 III. Review Approach i The teams' review approach was to first obtain an understanding"of Comanche Peak management and management control systems. This was accomplished by briefing from the Applicants management.

With this understanding, the team reviewers comenced their efforts. These

- included examination of appropriate documentation, formal and infonnal interviews of plant personnel, and specific technical allegations related to

their areas. The allegations were not reviewed separately but were subsumed in the total review in order to provide further assurance of alleger confidentiality and not compromise any on-going or future investigations.

In addition to the review of the Quality Assurance program, from a program-

matic point of view, each of the reviewers examined the impl,ementation of P the QA/QC program in their individual areas of expertise in an attempt to _

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identify any breakdowns that could exist in a narrow area. .

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'IV Review Findings 4

i The team conducted its review of the following areas:

J A. - Management Organization B. - Quality Assurance / Quality Control C. - Equipment Turnover and Preoperational Testing

. D. - Electrical i E. - Design Activities / Control

F. - Installation of Safety Related Fluid Systems .

G. - Civil Construction i H. - Heating, Ventilation, and Air Conditioning Systems

I. - Fonnal Interviews with QA/QC Personnel The review, f'indings and conclusions in each of these areas are provided below

A. Management Organization The construction and operations organization were reviewed to insure a working relationship between the organizations as well as functional

relationsh.ips within each organization. The qualifications of the individuals in positions of authority were reviewed against regulatory standards and the applicant's comitments. In addition to qualifica-tions, a review was made of the interface between all level: of the comand chain. .

The limited review revealed that in all areas, individual qualifications l

appear to meet requirements, the interface between construction and

- operations appears to be functioning in a workable manner and interface between all levels of the management chain appears to be functioning in an acceptable manner. There appears to have been a comunication ,

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problem in the onsite QA/QC chain in the past, but according to interviews conducted during this review the problem has and is being corrected.

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~ This review found the management and craft at Comartche Peak appear to be competent and management to possess a positive attitude which is a strength at this project. Management exhibited a sufficient level of consciousness for both safety and employee concerns. These management attitudes were confirmed by the attitudes they manifested in their

- employees and the attention to detail in the required quality of work.

B. Quality Assurance / Quality Control The following areas were reviewed primarily from a programmatic point of view: nonconformance control; training, audits; records (maintain-gp ability and retrievability); document control; receipt, storage and ~-

handling of materials; and procurement. ,,

T; 12 Within the areas reviewed, there were several findings identified. The following is a brief description of each according to category:

1. Potential Enforcement Issues.

a) ASME record packages were not being maintained .in a fire proof container.

b) At least two vendor audits had not been performed within the required time period. -

2. Weaknesses a) Certain drawing packages issued to the field contained non-applicable DCAs and/or CMCs, which had been deleted by -

engineering.

b) Many non-ASME Section 3 drawings centained a large number of DCA's and CMC's (over 300 in some cases) outstanding without being incorporated by revision.

3. Strengths a) The QA/QC training program is c4tcasivc and comprchcnsivc.

3 b) The use of a recently established computer system drawing control instead of stamped drawings referencing design changes.

, c) The vendor witnessing program is extensive in its audits and source inspection of purchased materials

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d) The ability to. expeditiously locate and retrieve records, without prior notice, from permanent records vault.

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Overall the current QA/QC program appears to be functioning

- satisfactorily. The recent management changes seems to have corrected past comunication problems.

{ C. Equipment Turnover and Preoperational Testing The p~rocesses of turnover of safety related equipment from construction to startup as well as pre-requisite and pre-operational tests of the equipment were reviewed to determine adequacy of: methodology employed in turnover of equipment to startup, return equipment to construction for rework, and ultimate release of equipment to operations; technical and administrative controls over preoperational testing; and preopera-

.' p tional test procedures, both technical content and administrative control. .{

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This review found the majority of the tests to be performed are retests -

or reperform's and could be conducted in parallel with the remaining initial test. The performance of the remaining test should not impact an October 1984, fuel load date. In addition, the turnover methodology and control of the preoperational test program appears adequate.

D. Electrical The assessment in this area was to determine acceptability of the safety related electrical equipment installed and inspected in accord- .

ance with NRC requirement and applicant comitments. A review was made

! of the overall program to include: drawings, procedures, quality 3

control inspections, and records.

The review found th'at the safety-related electrical equipment is being installed and inspected as required.

E. Design Activities / Control This review focused on the following areas: requirements of IEB 79-02;

,I IEB 79-14: Alternate Analysis for small bore piping system, rigorous 4 analysis for safety related piping systems; review of design calcula-l tions for pipe supports; review of stress analysis for piping systems; field inspection and verification; and the iterative design process.

A potential enforcement action was identified in that certain pipe supports which had been inspected and accepted were not installed in accordance with design drawings. There was also a strength identi-

'. fied in that the applicant was found to have used conservative considerations in many areas of design and analysis for the safety related piping systems and pipe supports.

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The review concluded that the design program a'nd its implemen appear to meet or exceed requirements, except as noted above.

F.

Installation of Safety Related Fluid Systems f i

The review of this area was directed towards a This review contained:

first hand observa-lding and shutdown of the plant. 7.y tion of systems by the reviewer; examining control of we d materials; examination of piping supports, welds and recor re s.

The reviewers concludedcommitments that the applicants program appears to a and good engineering compliance with requirements, '

practice. _

G. Civil Construction Activities '}

Examination of site civil design activities, including i (such it) design cha process, procedures and QA records of completed work activi as the SSI dam, cable tray supports and whip and moment restra lica-and procedures and work; activities for ongoing work (such as ap tion of protective coating) was performed.

' The limited review found that(1)the applicant protective coatings was and meeting requi in these areas. Two areas of note: '

(2) themo lag, appear to be progressing in a manner such . jp will not impact an October fuel load. U H. HVAC i

This effort followed up on previously identified In all areasdiscrepancies t appears k Comanche peak and other sites which used the i HVAC vendor.  !

l sis. The reviewed where discrepancies had been identi .

HVAC system appears to be adecuate.

I. Fonnal Interviews with QA/QC Personnel d Formal interviews of five (5) management / supervisory in perso eight (28) inspectors were conducted to assist the It was team twentying quality of work and management support i quality.

of felt discussions with inspection personnel wo'lld give a conse assess insight into the quality of site construction.

The major thrust of the interviews was to determine

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i idation if; (1) th personnel was experienced; had any (plant

3) training safety(4)orinspectors was t.dequate; quality could concerns freely talk to NRC; (5) management supported problem id

(6) was there feedback on identified problem evaluation, l

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• 9 With the exception of two inspectors who were " unsure" due to lack of knowledge, all personnel interviewed felt the plant was being built in a safety and quality manner. There were some concerns raised which will be forwarded to the Comanch Peak Project Director for evaluation; in some cases, Region IV was already aware of the concerns and performing followup. The major problem in the past appears to have been comunication between inspectors and their supervision, but it is apparent that for the past couple of months and presently, this problem is being addressed properly.

In addition to formal interviews, each reviewer performed numercus infomal interviews to determine problem areas. The overall conclusion frc'n all interviews was that the Comanche Peak Project is being built safely and with quality.

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p V. Conclusion _

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The purpose of the special team review has been met in that (1) an assessment of the applicant's current management control of the construction, inspection and test programs has been made; (2) an in depth understanding has been achieved and (3) information has been obtained to establish a management plan for the resolution of all -

outstanding licensing actions.

With respect to the assessment of the applicant's management control of the construction, inspection and testing programs, the special review team has detemined that based on the number and significance of the strengths vs weaknesses identified in this review, that the applicant's programs are being sufficiently controlled to allow continued plant construction while the NRC completes its review and inspection of the facility. .

Further, the review provided a sufficient understanding of these programs .

and their strength and weakness to assist in the deve!opment of the

" Comanche Peak Plan for the Completion of Outstanding Regulatory Actions."

This plan was approved for implementation on June 5,1984.

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A. Management Organization

1. Entrance Meeting

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The afternoon of April 3 the special review team arrived; onsi.te unannounced. The team spent the afternoon of April 3 and:the morning of April 4th meeting with the applicant's Senior Corporate. Management, Site Management, Site QA Management, and Document Control l Super.vi'.s ion 3

being briefed on the organization, functions, and location .of: areas

!, under their control.

j 2. Management Organization >

. The' nuclear portion of Texas Utilities Generating Company is,org'anized in the following manner for its senior management:

l E a) The highest level executive is the President of the company. The r

! V - President has recently turned over all possible non-nuclear dutie~s ',

T to his Executive Vice President-Plant Operations. The President's'

- primary responsibility is to complete the Comanche Peak Steam ETectric Station as safely and expeditiously as possible.

l b) Reporting directly to the President are the Executive Vice

President Engineering and Construction and the Vice President l Operations. Even though there are fossil plants presently being built in the syitem and the licensing organization reports to the i Executive V.P. Engineering ar.d Construction he spends between j 60-80% of his time at the Comanche Peak Site. He has also

• i delegated his non-nuclear responsibilities in an effort to focus on the nuclear station completion. The Vice President-Operations (V.P. OPS.) spends approximately 80% of his available time on site

- directly observing the operations group preparation to take over the plant upon construction completion. He is also an active. .

participant in construction and startup meetings and the decision making process. A few months ago the V.P.0PS. was moved from his nonnal reporting path to Executive V.P.-Plant operctions, directly reporting to the President.

l c) Reporting to the Executive Vice President Engineering and j construction is the Vice President Engineering and Construction (V.P.E.&C.). The V.P.E&C. has been located on the Comanche Peak I site since 1977 and during the same year he assumed the additional

)' title of Project General Manager for Comanche Peak. In January 1984 he delegated his non-nuclear responsibilities in

! order to devote his full attention to Comanche Peak completion.

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! d) The Assistant Project General Manager (APGM) reports to both the 4'

V.P.E&C and the V.P.0PS. He reports to the V.P.E&C. in the areas

, of construction and onsite engineering and to .the V.P.0PS for

- startup (S/U). This position is where the comon tie between i

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construction and operations is most decisive. The APGM has been on site since 1977.

f e) In addition to the APGM, the V.P.0PS has reporting to him: The L. manager of Nuclear Operations, who is located at the site, and the Manager of Quality Assurance who is located in the corporate office but has a Quality Assurance / Quality Control Manager on site who is resp'onsible for all QA/QC on site.

The current positive management attitude is a strength exhibited at Comanche Peak from both the operations and the engineering and construction sides of the company. This positive attitude appears to manifests itself in the attitudes of the workers, the training, i

and in its consciousness for. quality.

t p One additional strength was noted in that the applicant is using -

operations' maintenance procedures to perfonn periodic maintenance on T j Y..

equipment in the plant, and thea' pplicant is using full Anti-C dressout}-

and respirators for the craft (for training) to perform maintenance.

activities so when the equipment becomes contaminated the workers will l be use to the confining clothes and equipment. This practice shculd

! significantly reduce exposure and therefore dose received by these -

individuals after the plant is operational.

3. Project Management Meeting Every Saturday morning a project management meeting is held, wherein work activities, progress, startup and test problems, and QA/QC -

L coverage is discussed. This meeting is attended by Senior Corporate Management; including the President of Texas Utilities Generating i Company, and the Senior management-from construction and operations; it l is also attended by the site management of construction and startup.

Several members of the review team attended this meeting on April 7, 1984. The meeting appeared to be well managed, with problem areas being openly discussed (even though senior company management and NRC l

were in attendance, the dialogue between individual managers and supervisors was not toned down). An example of an area of concern

- which was discussed was the completion of the applicatica of protective  !

coatings in the containment. It was the general consensus that additional manpower was required to complete the work effort. An additional 100 people were authorized with the expectation they would be available within one week.

During this meeting it was decided to change the concept that was presently being used for plant completion. The applicant had been using a Building completion methodology, but after consultation and reviews by an acknowledged industry expert is was decided to prioritize 4

systems completion, with buildings to follow, or run in parallel where i possible.

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3 The highest level of the Company's management in attendance at this meeting allows for imediate decisions to be made for the next weeks priorities for plant completion. This method of holding project meetings appears to have kept the applicant in position to meet their projected fuel load date.

B. Quality Assurance / Quality Control

1. Nonconformance Control

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References:

CP-QAP-16.1,R20, Control of Nonconfonning Items CP-QP-16.0, R13, Nonconformances -

> CP-QP-16.1,R5, Signif* cant Construction j Deficiencies a CP-QP-17.0, R3, Corrective Action CP-QP-15.7,R2, Tracking of Audit Reports /Correc-yE tive Action Reports ,-

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5 E a. General -, ,

' This portion of the review was performed to verify that:

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- nonconformances are being identified

- items were considered for reportability to NRC

! - corrective action prevented recurrence

- the licensee has an adequate trending program

b. Review Effort -

The reviewer selected NCRs from various safety related systems to verify the following:

- logged numerically for centrol

- maintained even when later cancelled -

- considered for reportability to NRC

- corrective action initiated which prevented recurrence s- - considered in a trending program

! The following NCRs were reviewed:

! C-84-01030 M-83-01162, R2 M-84-00965 M-11678N

}a M-82-01528, R2 M-11660N i M-83-01454, R1 M-11675N f

M-04729, R1 M-11687N

' M-05689, RO E-84-01031

M-06244, R1 M-01695N

' M-09765 M-01692 M-09766 M-098125, R1 1

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The responsibility for closing NCR M-09812 S, R1, has been transferred to TUGC0 startup because these Westinghouse valves are required to be disassembled during system flushing. The. valves are to be reassembled under a startup work authorization (SWA).

Valve stroke time testing of these valves will be verified under the SWA. The relief valves listed on NCRs M-09765 and M-09766 were required to be reset because the ser. dor had not beer, i furnished the correct back-pressure information to set the valves.

i c. Conclusion 1

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The limited review found that nonconfonnances were being written when identified, the items were considered for reportability to NRC, that corrective action to prevent recurrance was being initiated, and items were being trended.

t E 2. Quality Assurance / Quality Control Training .--

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References:

CP-QAP-2.1,R10, Personnel Training and Qualifica , ',

. tion QI-QAP-2.1-1,R6, Nondestructive Examination Personnel Certification

QI-QAP-2.1-5,RS, Training and Certification of -

l Mechanical Inspection Personnel

a. General .

The. purpose of this part of the review was to verify that the -

4 licensee has:

- a formal training program

- conducted required training to qualify personnel

- requirements for on-the-job training

• objective evidence of personnel qualifications

- evaluated the candidate's education, experience, and training prior to certification

- reevaluated personnel on a periodic basis

- records of personnel qualifications

b. Review Effort -

A review was made of the documents listed above, and the reviewer held discussions with responsible corporate and site personnel to verify that procedures are consistent with regulatory require-ments. A review was made of General Examination Tests, RT-II-G-A, UT-II-G-B, PT-II-G-B, and MT-II-G-F; also Practical Examinations MT-II-P-04 and PT-II-P-07. These examinations confirmed the tests to meet the requirements of ASNT-TC-1A, Recommended Practice. The records of seven QC inspectors were reviewed. The records contained objective evidence of QC inspectors qualifications by l

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5 general and practical examination, on-the-jeb training', classroom, specialized training, education, and work experience records were available to confirm QC inspectors meet the _ requirements of ASNT-TC-1A and ANSI N45.2.6-1978. Confinnation of annual documented evaluations of qualifications of inspectors was verified.

c. Conclusion The training requirements for QA/QC personnel listed in the procedures appear to be complete. When personnel were questioned as to the training they were actually receiving, they confirmed the depth of training which the procedures required.

3. Audits

References:

,0I-QAP-2.1-4, Auditors Certification '

T_ DQI-CS-4.6,-R6, Conduct of Internal, Prime and

-Subcontractor Audits

a. Ge'neral ,

The TUGC0 QA audit program is based on FSAR Section 17.1.2 which addresses ANSI N45.2.12, Draft 3, Rev. O. TUGC0 Corporate Office is responsible for audits both internal and external. The audits spanned contractors, engineering, construction and corporate.

Audits are listed in five areas, Site Construction / Engineering /

Quality Control, Operations /Startup, Vendor, Pre-award Surveys, -

and Vendor Surveillance. Audits scheduled in the five areas were 107, 158, and 80 during 1982, 1983,'and 1984, respectively.

b. Review Effort _

A review was made of the licensee's implemented audit program to verify whether it meets the requirements of the accepted QA Program and ANSI N45.2.12 (Draft 3, Revision 0 - 1973) as endorsed by the QA Program. The reviewer also verified the following

, aspects of the audit program:

- The scope of the audit program has been defined and is consistent with FSAR comitments

- Responsibilities have been assigned in writing for the overall management of the audit program i

- Methods have been defined for taking corrective action when

' deficiencies are identified during audits

- The audited organization is required to respond in <friting to audit findings t .

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6 audit reports and corrective Dist,-ibution requirements fo - -

action responses have been defined f

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' g audits 69, 70, and 7 B

,.g.5;NM 40, 43, 56, 57, 61,The audits were  :/,MQ The TUGreviewer 22 perfomed selected audits during 1982 TPC and 1983 for rehensive.

review. '5 is d in is; g, h ;

preplanned to cover specific however, properfunctions corrective documented and were ".m$.15 c

aCCordance with ANSI "45.2.12-1977; action had')t been 49@,4'

. 43, 'yh t by memorandum datec August 16, 1983. Review of. the vendor audit S-wf ;m f,.

distributed in a timely manner. @Nhj;{f;

, program is discussed in paragraph 8.7. i wed; ,"~

; .#R7x@-nj The records of four lead auditors ified andto two - auditors ,

b The qualifications of auditors and lead auditors 2 23-1978. were ver 't-i s of be in accordance with the requirements ofy/g.jp ANS auditors were verified.

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' As a result of this limited review, the reviewe -

activities are acceptable.

4. Records Implementation of the .

References:

(a) CP-QP-18.2,R2, Permanent Plant Records g Management System  ;' i n l Pemanent Plant Records ,

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j (b) CP-QP-18.3,R2, System Organization <,

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,' d Pemanent Plant Records (c) CP-QP-18.4,R2, Receipt Control and Storage I

(d) CP-QP-18.5,R2, Automatic Records Management System Implementation (f Mu' 3

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Record Turnover to TUGC0 4Qj (e) CP-QP-18.6,RO, Operations Group 3Cm H-5 and N-3 Code Data Reports &%

(f) CP-QP-18.7,RO, A@h

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Records Verification (g) CP-QP-18.8,R1, h.. M:a

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(h) CP-QAP-11.1,R3, Fabrication and Installation Inspection of Components, Components Supports, and ,

Piping (i) CP-QAP-16.1,R20, Control of Nonconfonning '.

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7 Inspection Criteria and 9.;

(j) CP-QAP-12.1,R8, Documentation Requirements l

F Prior to S.W tem N-S Certificat!oa  !.

e Processing QA Records (k) CP-QAP-18.1,R2, p-(1) CP-QAP-18.2,R4, QA Review of ASME III Documentation ., 7 h{

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' - - a. General The quality assurance records Draft 1, Rev. progO,(ram 1973) is for based the on FSA  !

l (B) which addresses ANSI N45.2.9 The site records G design . and construction of Comanche Peak. The program is managed under the control of the Site QA Manager. (P Pennanent Plant Records Vault (PPRV) houses most of the des Ht construction perfonned.

records for completed work and hav .

of the operations records control system on a regular basis.

Temporary storage of records is also ongoing at several workin

locations at the site utilizing one-hour fireproof cabinets.The Records, where possible, are filed, by system and component.

PPRV uses smoke detectcrs tied into the site fire stati records fire protection; a water hose adjacen extinguishers in the area.

A computer is used to aid record retrievability, Records but is notflow essential, as records are maintained in hard copy.to an ASME path.

.b . Review Effort ,

A review was made of various procedures to verify that provision had been made to maintain various types of

-Records storage procedures were also storage requirements.

reviewed to ensure that they described the storage facilities, the filing systems used, methods of receipt, an l

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Section III records to the PPRV was reviewed. The reviewer also

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verified retrievability of records from the PPRV.. ; ,

To verify general record retrievability, the reiiewer selected several general construction and inspection packages :such as weld data, concrete placements, equipment packages,: and equipment travelers. All records were retrieved in a short. period from the PPRV. During the review, other records were retrieved offspecific design / construction / inspection activities. No significant difficulties were identified during these real-t'ime challenges to the records retrievability system. The ability to expeditiously locate and retrieve records is identified as a strenoth. This ability appears to be primarily due to indexing and. storage of ~

' records by component or material, when possible,,instead of by record type. ,

F .

To review the B&R ASME records flow, the records associated with y.

i safety injection isometric SI-2-RB-13-4; Core Spray CS-1-SB-032; -

Chemical and Volume Centrol ~CT-1-SB-14; Component Cooling CC-2-SB-042; Boron Recycle BR-1-SB-05 Spool 103; BR-1-SB-004 Spool IQ3, BR-1-SB-006, and Main Steam MS-1-SB-050 were reviewed.

These records contained the inspector's identification, the type of inspections, the acceptability, verification of review and approval, and were readily retrievable. Heat numbers on materials installed in the field were recorded during a site tour.

Certified Materials Test Reports (CMTR) were requested and furnished which verified traceability for those items recorded -

during the tour. Also CMTRs, for selected subassemblies were verified to meet ASME code requirements. Review of records for

the subassemblies listed abcve confimed that Desi Authorizations (DCAs) and Component Modifications CMCs) (gn Change were incorporated into the as-built drawings prior to the ASME code '

stamp being applied .to systems. This program of records review, approval and turnover from B&R, the ASME "N" stamp holder, to TUGC0 appears to be very thorough, though complex. Records for work perfomance by B&R are assembled, reviewed, and approved,

- then submitted to the Authorized Nuclear Inspector (ANI) for i review, then submitted to TUGC0 for filing. A task force

} comprised of B&R and TUGC0 personnel, thea make another review of l these records. Any discrepancies noted are then resolved between i B&R and TUGCO. These recosds are then red labeled, and etn not be i removed from the vault without written approval of QA management;

, thereby, preventing loss of QA recceds, t -

l A review was made of the temporfry storage of records in the i

field. Although records are best protected in the PPRV, record storage in adequate fire proof cabinets is allowed based on the record storage equipment qualification in NFPA No. 232-1975, which bases fire protection on exterior fire . load calculations.

t j

Although the reviewer did not ch(ck any fire load calculations t

4

~

9 justifying the use of one-hour fire cabinets, those cabinets observed appeared to be adequately protected. During this review, the observation was made that several completed ASME moment restraint record packages being maintained in a non-fireproof cabinet in the ASME Safeguards Bu,ilding QC trailer. This failure i- to store quality assurance records in a fireproof cabinet is a potential enforcement issue. Prior to conclusion of the review, I

these records were relocated to fireproof' cabinets. Based on the J- above problem, the reviewer noted some confusion at the site on l

the control of " documents" as they progresa through design / con-struction/QC and as to when they become " records." This was evident as little distinction appeared to be made for the storage of " documents" or " records" in the field. Working " documents" were provided equal to or better protection than " records" in some instances. Other than the example stated, no other storage F problem was identified. Comanche Peak had esta'alished, on __

3 March 30,1984, records monitoring teams to reviw the records; .

t flow program. The clarification of the document / records interface:

for storage control is a weakness and is to be addressed by the monitor teams. This weakness is considered part of the potential

~

enforcement issue addressed above.

The physical construction of the PPRV was reviewed. The construc-tion of the PPRV is satisfactory for protection from exterior fire damage. For inside originated fire damage, the PPRV has a fire detection system but does not have the industry standard water or halon automatic fire suppression system. The system for unattended PPRV fire control was reviewed. With the fire detec-tion alarms annunciating in the close-by fire station, the fire -

l station personnel having ready access to the PFRV and the location of a fire hose reel outside the PPRV door, the fire protection

! appears adequate. Verification was made that the operations vault, into which all -the PPRV records will be transferred, -

contains an automatic fire suppression system.

. c. Conclusion The records control of the PPRV appears to meet all requirements,

! with sufficient staff to control the activity. Records flow to the PPRV needs clarification, but appears adequate in implementa-tion. Records personnel appeared (nowledgeable as to PPRV j

cperation.

5. Document Control ,

References:

(a) DCP-3, R17, CPSES Document Control (b) DET-12, RO, DCC/ Task Force Interface i

I

4 10 i

a. General Controlled documents, primarily drawings, specifications, 'and procedures are maintained and controlled by the site Document Control Center (DCC). The predominance of document control within the sphere of the DCC relates to drawing control and changes to those drawings. The DCC has established satellite document control centers which control and distribute most of the working l documents. -These satellites provide controlled document copies to crafts and the Unit 1 Task Force Paper Flow Groups (PFG).

Controlled documents and changes are provided to the satellites from the DCC. The DCC also provides controlled documents to

several " controlled number recipients" directly. The PFG provides controlled documents to craft working in that specific building task force. Revisions to controlled drawings and documents that P affect controlled drawings, such as design change authorizations _

g.

(DCAs) or component modification cards (CMCs) are distributed upqn, 4-~

~

receipt to the satellites and controlled number recipients. For :

drawings, a computer system keeps track of drawings and the DCAs and CMCs that affect those drawings. When new drawings, drawing revisions, DCAs, or CMCs are generated the computer is updated.

When the satellites receive a new drawing revision, CMC or DCA, any controlled drawings checked out to the crafts or under the control of the PFG are updated by the satellite DCC personnel.

This maintains current the controlled drawings in use by insuring that drawing packages contain the correct revision with applicable DCAs and CMCs Drawings checked out to the craft from the PFGs or directly from the satelTites are returned at the end of the working day. Prior to checking out drawings from a satellite l directly to the craft, a computer run is made to insure that

' drawing packages contain the appropriate revision and applicable i CMCs and DCAs. When craft personnel return drawing packages to the satellite or PFG, a drawing, CMC and DCA check is again -

perfonned to verify return of the controlled documents.

b. Re'viewiEffort A review was made of the references listed to verify they met the requirements of the accepted QA Program. The reviewer also i

verified that administrative controls have been established for the control of drawings and that indices are maintained for drawings, manuals, specifications, and procedures which indicate current revisions.

In order to verify the control of drawings, the reviewe'r selected several drawings to detennine if the current drawing revision with applicable DCAs and CMCs located in the DCC, was also onhand in the control ard auxiliary building PFGs. Two drawing discre-pancies were noted. Drawings 2323-El-2011, R8 and 2323-El-0900,

Sheet 1, R6 maintair.ed in the PFG had several DCAs in the package i

1 i

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11 that were missing from the current drawing package computer printout. The verification was perfonned on April 12, 1984, using a current drawing status. This problem appears to be from engineering eliminating CMCs and DCAs from its data base applic-able to particular drawings without informing DCC of the change.

Although the computer change keeps satellite issues current, no

" trigger" device causes satellite personnel to remove the CMCs and DCAs from the PFG drawing package. A review of the engineering

, mechanism for updating the data base founti the procedure satis-factory and a review of having non-applicable CMCs or DCAs in the drawing package revealed that while possibly confusing, the  ;

practice is not a technical: problem. As the working controlled  ;

drawing packages are expected to be current at all times, this mechanism whereby non-applicable CMCs and DCAs remain in con-trolled drawing packages is identified as a weakness. ,

F The computer assisted drawing control program was reviewed. tI s.

t Specifically, with the sole reliance on the current computer . ;

printout to detennine drawing package adequacy, the controls of- ~

computer input and changes were reviewed. Access codes have been established so that a limited number of engineering and DCC

personnel have access to affect their respective data base. A procedure and training exists to define appropriate computer

, changes authorized for each group. The system appears to be .

adequately controlled and use of a computer system versus stamped ,

drawings referencing DCAs and CMCs is identified as a strength.

t During this review, a frequent observation from all reviewers was .

the continued maintenance of a large number of CMCs and DCAs in drawings ~ packages, rather than making a revision to the drawing incorporating the completed changes. Interviews with craft and QC  !

personnel revealed that other than the inconvenience of the sheer volume of a large number of CMCs and DCAs in a package, they had -

not encountered construction errors due to accumulation of DCAs and CMCs. In that no problem appear to be developing, but the potential to lose control is high when drawings are not revised periodically to keep outstanding drawing changes reasonably low,

• the maintenance of working drawings with a large number of completed CMCs and DCAs without a drawing revision is identified as a weakness. The applicant does have a program under way which

began two yea'rs ago to update those drawings identified by

operations as needed for safe operation. This program is l

scheduled for completion by fuel load.

c. Conclusion

The limited review revealed that the current document control i system appears to be functioning satisfactorily. All DCC and i PFG personnel interviewed were aware of their responsibilities and how their job was performed. The DCC, satellites, and PFGs I reviewed appeared to be adequately staffed. l

. h

.- -- , ,n. - , .-

12 The use of the drawing control computer appears to keep craft personnel up-to-date in an expeditious manner.

6. Receipt. Storage, and Handling of Materials

References:

(a) CP-CPM 8.1, R1, Receipt, Storage, and Issuance i of Items (b) CI-CPM 8.1, R1, Color Coding of Piping Materials (c) CI-CPM 8.2, RS, Centrol of Spare Parts (d) MCP-10, R7, Storage and Storage .

P Maintenance of Mechanical _

3 and Electrical E,quipment ..

s: _

(e) ICP-5, R3, Control of Permanent Plant -

Instrumentation Receiving Inspection (for (f) CP-QAP-8.1,R7, ASMEitems)

(g) CP-QP-8.0,R2, Receiving Inspection i

a. General ,

l Warehousing activities are managed under the Project Support Services organization. Safety-related materie.1 is stored in

- several warehouses and also in an outside laydown yard. All

. material is received at one warehouse and then moved to the ~

appropriate storage location. Shipping damage inspections are conducted by warehouse pa"sonnel and receipt inspections are perfonned by QC inspectors. Environmentally sensitive material 'is

. stored in a temperature and humidity controlled storage location.

A preventive maintenance program exists to insure that mechanical and electrical equipment is maintained in an operable condition while in storage.

b. Review Effort

' A review of the licensee's program for the receipt, storage, and handling of equipment and material with respect to selected elements of the licensee's accepted QA Program was performed. The

• review was to verify that administrative controls had been established concerning receipt inspection of safety-related materials, preparation and retention of required documentation, l control of nonconfonning and conditional release items and contro of i+. ems in storage. Implementation of the program was reviewed b

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by selecting several safety-related items in storage and verifying document and item controit to be in accordance with the program.

The reviewer also toured the warehousing locations. Storage discrepancies were not. identified. The QC receipt inspection program was also reviewed:.. OC inspections appeared to be conducted in a satisfactory. manner. ,

c. Conclusion ,

Based on the limited review of the warehousing and receipt inspection program and implementation, both programs appear adequately managed. Storage locations appear adequately staffed.

Warehousing and QC personnel were knowledgeable and professional in their respective areas.

1 ,~

g 7. Procurement  ; -

t' ~

References:

(a) CP-EP-5.0, R7, Procedure for Field Procurement ~

(b) DQP-CS-2,R6, Procurement (c) DQP-CS-4,R9, Procedure to Establish and Apply

• A System of Pre-Award Evaluations, Audits, and Surveillances ,

(d) DQI-CS-4.1,R3, Vendor QA Manual Reviews (e) DQI-CS-4.2,R3, Generating and Maintaining the TUGC0 Approved Vendors List ,

~

(f) DQI-CS-4.3, R4, Vendor Perfomance Evaluation System (g) DQI-CS-4.4,R4, Conduct of Vendor Pre-Award Evaluations (h) DQI-CS-4.5,R6, Conduct of Vendor Audits (1) DQP-VC-1,R7, Final Inspection and Release for TUGC0 ,

(j) DQP-VC-2,R7, Witnessing Trip I - (k) DQP-VC-3,R3, Initiating Yellow Flag Sheets (1) DQP-VC-4,R6, Guidelines for Certifying Vendor Compliance Inspection Personnel I

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14 (m) CP-QP-5.0,R1, Guality Assurance Review of Site, Generated Procurement Documents

a. . General Safety-related purchase requisitions are generated by TUGC0 engineering at the site and are converted to purchase orders by the site procurement and subcontracts section. Technical and QA requirements are detemined by engineering. A QA review of all safety-related purchase orders is conducted on site to verify QA Each purchase order requirements and use of an approved vendor.

requires the vendor to inform TUGC0 when a product is ready to ship. TUGC0 QA determines whether to perform a pre-shipment inspection at the vendor's location or to waive this inspection.

P Approximately one-third of all safety-related shipments are source _._

s inspected. TUGC0 also maintains a vendor audit program to insure .

5:

that vendors can meet the requirements imposed by the purchase. ;

orders. The vendors that are satisfactorily audited are placed on the approved vendors list. TUGC0 has also initiated an annual reviaw of supplier perfo,mance.

b. Review Effort A review was made of the licensee's procurement program with respect to selected elements of the accepted OA Program. The review was to verify that administrative controls had been - *:

established for the preparation, review, approval and revision of procurement documents. A review of the licensee's p.rocedures to verify that acceptable methods were being used to qualify vendors which provide quality goods or services; that these procedures required the maintenance of records of supplier qualifications and -

audits; and that responsibilities have been assigned to perfonn -

the vendor qualification program was performed. Several purchases orders at the site and at the TUGC0 offices in Dallas were

- reviewed. Purchase orders, based on the limited review, appeared to be handled satisfactorily.

Also reviewed was the source inspection or witnessing program implemented from the TUGC0 QA office. The program is quite extensive and appears to be very effective at performing material inspections at the source and identifying potential proolems difficult to detect by a receiving inspection alone.

A portion of this program, though, needs clarification.' Although, the witnessing procedures describes how to perform the source inspection, criteria is not documented for the decision on what This i

purchase orders are source inspected and which are waived.

' is considered a procedure weakness, but not a program weakness.

l The entire witnessing program is a strength.

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t i 15 Also reviewed was the vendor audit program, which is used to maintain the approved vendors list. The reviewer selected several vendors on the current list and reviewed their most current audits. All audits reviewed were considered satisfactory. Two of the vendor audits, Dresser Industries and Forney Engineering were 4

I I last audited in 1978. The licensee, through the FSAR, utilizes j ANSI N45.2.12, Draft 3, Rev. O to develop the audit program, a i part of which is the vendor audit program. Paragraph 3.4.2 of this standard requires the performance of annual audits or at least one audit during the lifetime of the activity. NRC Regulatory Guide 1.144, Revision 1, Auditing of Quality Assurance Programs for Nuclear Power Plants, which the licensee has not endorsed, clarifies this annual requirement with respect to vendor audits, in that vendors may be audited triennially providing that annual evaluations continue to show the vendor perfor;ning satisfac-E torily. The TUGC0 vendor audit. program doet not provide for an -

r annual, triennial or any periodic vendor audit schedule. Vendors. ;

Y are reaudited primarily on a usage and performance history basis. ; *

. This failure to establish measures to audit vendors at least

- triennially is considered a potential enforcement issue. The inspector found no indication that a failure to audit periodically resulted in maintaining an unsatisfactory vendor on the approved _

vendors list. Also, although the vendor witnessing program does not review the vendor's QA program, and is not a substitute for a '

TUGC0 audit, the large number of source inspections would mitigate the possible consequences of not performing periodic vendor

. . audits. . .. ..

j c. Conclusion i The procurement program appears to be satisfactory. The vendor witnessing program is an asset and appears well managed. Other '

than the missing timetable for the vendor audit program, the -

conduct of audits and vendor annual evaluations appears to be well managed. Personnel in the procurement QA staff appear to be knowledgeable and professional in their work.

C. Equipment Turnover and Preoperational Testing

References:

CP-SAP-3, Custody Transfer of Statiori Components STA-802, Final Acceptance of Station Systems, Structures, and Equipment CP-SAP-21 Conduct of Testing f

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a. General The processes of turnover of safety related equipment fro ,

equipment were reviewed in order to detemine if:

(1)

The method employed for transferring custody of components, partial subsystems, subsystems or systems from construction to startup; the return of equipment to construction for rework or modification; and the ultimate release of custody from startup to ,

operations are technically and administratively adequate. ,

(2) The administrative controls over preoperational testing are technically and administratively adequate.

P (3) The preoperational test procedures both perfomed and yet to be perfonned are technically viable and administratively sufficient..

sg

b. Review Effort (1) Equipment Turnover -

The turnover of safety related equipment from Construction to Startup is administratively controlled by Startup Administrative This Procedure CP-SAP-3, Custody Transfer of Station Components.

procedure establishes the requirements and responsibilities for transferring custody of components, partial subsystems, subsystems *~

or systems from:

(a) Construction to Startup (b) Startup back to Construction for rework or modification ,

(c) Startup to Operations The Startup group determines the turnover boundaries necessary The Completions to Group p(erfom pre-operational testing activities.a subgroup of Sta of easipment, valve, piping and instrument lists, drawing lists such as flow, instrumentation and control, and auxiliary one-line diagrams as required to sufficiently describe the content and

. boundaries of the turnover.

The Completion Group is also responsible for initiating and processing turnovers consistent with established schedules in the

)

turnover package, such as to:

(a) identify the equipment (b) indicate the scope of the turnover f

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(c) assemble the late revisions of the appropriate diagrams /

prints and applicable design change documents (0CA's) _

(d) list deficiencies, including design changes that have not ,

been implemented l The Completion Group coordinates all required pre-turnover walkdowns and punchlist activities for the purpose of establishing the states of remaining work to be done prior to turnover of that equipment to startup.

Startup personnel review the packages and perform a walkdown of i the equipment / system to determine if the equipment identified in the package is ready for turnover. Any deficiencies requiring resolution prior to turnover are resolved prior to transfer; those F deficiencies not requiring pre-turnover resolution are added to -

t the Master Data Base (a computerized tracking system) to facilit- ,

tate future disposition. ' Upon_ completion of the startup walkdown 9

and correction of required deficiencias, custody / turnover of the equipment is transferred to startup.

Custody of station components may be returned to construction for perfonnance of work such as . major modifications, repair or clearing of construction deficiencies. The return of equipment to construction , voids all preoperation testing on said equipment.

After tne completion of applicable prerequisite tests, (construc-tion tests), including initial operation of the equipment, startup may relinquish " operational control" to Operations yet maintains custody of the equipment pending completion of preoperational testing, i The turnover packages for the following systems were reviewed:

j (a) Component Cooling

+ (b) Auxiliary Feedwater (c) Containment Spray f (d) Chemical and Volume Control (e) Residual Heat Removal (f) Safety Injection (g) Hydrogen Recombiners (h) Reactor Protection System '

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18 .

The turnover of equipment from Startup to Operations is detailed in Station Administrative Procedure STA-802 Final Acceptance of Station Systems, Structures and Equipment. Pursuant to that Procedure, Operations initiates a detailed review of the turnover Following package and walks down the applicable equipment. successfu  :

accepts the equipment / area.

At this time all responsibility for ,

that equipment lies with operations. [

I There has been no safety related equipment transferred to ,

' operations, thus the review of the process was in terms of -

programatic sufficiency.

i r.

(2) Preoperational Testing Program E The preoperational test program was reviewed in order to verify that the tests to be performed have been identified and that each ;

lT Y of the identified tests entailed at a minimum, test objectives, ', ~

sumary of the test, necessary prerequisites, and acceptance criteria. -

The test organization was reviewed in order to verify that the lines of authority and responsibilities of test personnel are specified and that where . interfaces exist between organizations involved in the test program, that organizational responsibilities .

are clearly established. .

The administration of the test program was reviewed in order to verify that methods are established to receive (from construction) the jurisdiction over systems before comencement of testing.

The administrative mechanisms established for jurisdiction control -

of systems before, during, and after testing were reviewed in order to verify that those mechanisms adequately provide for:

control of system status before preoperational testing including

- the completion of adequate prerequisite (construction) testing; the return of systems to Construction if necessary to support

' modifications and/or reports; the control of system status subsequent to testing including measures necessary to prevent invalidation of test results; the control of the system during testing; only the assigned System Test Engineer or his designate may conduct system testing.

3 The conduct of testing was reviewed in order to verify methods that a to change adequate administrative measures provide for:

test procedure during the conduct of testing; the criteria for interruption of a test and continuation of an interrupted test; methods to coordinate the conduct of testing; methods to document significant events, unusual conditions or interruptions to testing; methods for identifying deficiencies, documenting their e

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19 resolution and documenting retesting; methods for providing the current test procedure to operations and coordinating test activities with the shift 3upervisor; methods to ensure that the systems test engineer has the appropriate latest revision of the required do:umentation/ references.

The program for evaluation of test results was reviewed in order to detemine that: deficiencies are clearly identified and appropriate corrective action proposed, reviewed and completed; subsequent to corrective actions or modifications have been ecmpleted, tests or portions of test have been rerun as necessary to ensure that tests of tne as-built system are adequate; the results of the evaluations were reviewed by the ' appropriate licensee personnel responsible for approving the original proce-dure.

p ,

(3) Prerequisites Tests '.

, t v _

Selected prerequisite tests were reviewed in order to detemine if

• the tests provide and adequate mechanism of accomplishing vital testing and operation of the associated equipment. The tests ~

reviewed appeared technically and administratively sufficient.

The prerequisite tests when perfomed in ccmpl-iance with Startup Administrative Procedure CP-SAP-21, Conduct of Testing, and, as required by the applicable preoperational tests, appear to provide an adequate mechanism for initial equipment checkout and .

operation.

.(4) Preoperational Tests Selected preoperational test procedures for tests which are yet to be perfortred, were reviewed in order to ascertain adequate -

implementation of the following:

(a) Management review and approval (b) Procedure femat with emphasis on clarity of testing required (c) Clarity of test objectives (d) Pertinent prerequisites identified, e.g.

1) required plant systems are specified

2) proper facility procedures and other references are specified and uniquely identified

3) completion of calibration checks, limit switch setting 5 protective device setting, included where applicable

4) special supplies, and test equipment specified.

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• i ified.

(e)

Special environmental conditions, if any, ident d the procedure Acceptance criteria are clearlytance criteria. an identified c*x (f) requires comparison of results with accepi.e.,

d tified, (^ ;t.

(g) The source of the acceptance criteria is s 5},. ie Initial test conditions are specified [B}f6..c (h) 1)

Valve line-ups y'(;m w

Electrical power and control requirements .;;. Wa

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3) Temporary and piping

) installations (instrumentation

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4)

Temperatures, pressures, flows sg The procedure includes reference to appropriate FSA N, (1) sections, T/S, drawingse CW]

h; requirements. f the proce-sure that test [$p:

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(j) Step-by-step instructions for the dure are complete to the extent necessary to as perfontance o ~

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objectives are met. t all items.

having been per-Provisions are available for documenting tha

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(k) including prerequisites, are verified as formed. h conduct of the Provision is made for recording details of t eh ir resolutio )-

(1) test including observed deficiencies, t e -

retest. disconnections p4 Procedure requires that temporary connections,fers to

• (m) or jumpers be restored to normal or re 3 procedure. el conducting a Procedure provides for identificationt of personnor refe f (n) the testing and evaluating the< test da a

)

procedure, tion of critical j (o) Procedure provides for d to the independent following: ve Th'ese procedures included but were not limite Component Cooling 1-CP-PT-11-03 D/G Control & Functional 1-CP-PT-29-2 r

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1-CP-PT-48-01 Containment Spray 1-CP-PT-49-02-RT-1 CVCS - Seal Water & Letdown Perfomance Retest 1-CP-PT-49-03-RT-1 CVCS - Chemical Control Purification and Makeup Retest 1-CP-PT-57-01-RT-1 51 Pump Perfomance Retest Selected completed preoperational procedures were reviewed in order to ascertain, at a minimum that:

(a) The licensee is perfoming an adequata evaluation of test resul ts.

,(b) All test data are either within previously established acceptance criteria, or that deviations are properly p dispositioned. .**

r i '

(c) The lican'see's methods for correcting deficiencies and for #

- retesting a*e adequate.

(d) The adequacy of the licensee's administrative practices in maintaining proper test discipline concerning test execution, test alteration, and test records. _

(e)~ ' The licensee ~is following his procedures for review, evaluation, and acceptance of test results.

These procedures included, but were not limited to: ,

1-CP-PT-57-06 RHR - ECCS 1-CP-PT-67-01 Hydrogen Recombiner Reactor Protection, System

~

1-CP-PT-64-02 1-CP-PT-57-02 Centrifugal Charging Pump 1-CP-PT-57-01 SI Pump Perfomance

• 1-CP-PT-48-01 Containment Spray 1-CP-PT-29-04 D/G Sequencing 1-CP-PT-02-08 Class I-E Switchgear (5) Systems Status i '

System walkdowns were perfomed in order to detemine the current j status of safety related components / systems. The following l

systems, among others were selectively reviewed in that assess-ment:

?

a{ResidualHeatRemoval

b. Chemical Volume and Control ,

c) Safety Injection d Containment Spray e Auxiliary Feedwater

f Component Cooling ,

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22 ltd PreoperationalThe test status reports were also rev review revealed that of the 198 and remaining testing. 7 J original preoperational test procedures, 45 have yet t to be perfortned; of the 34 preoperational/ retest procedures, 33 ha to be performed; that of the 39 preoperational/reperform proce-Thu n dures, 37 have yet to be performed. It should be noted howeverh  %

115 or 42% have 3et to be p'erformed.

that the " Retests" and "Reperforms" are, as$g; g a

should require less time to complete."Reperforms" Note: The will yb i l Wi should run much smoother than the original tests.7m re -

reworkandstationmodifications.) M There is no preoperational testing currently ongoing, h nor has i";

there been any significant testing in the past 10difi' months, t e "r:

result of the aforementioned electrical rework and oth p cations. t tional testing during the month of April 1984 p f

A statistical analysis of the preoperational-Q[ f te essence the period imediately proceeding a virtuall shutdown o -

testing necessitated oy the modifications as aforementioned, i revealed that in that 11 month period,177 of tne 198 orig na -

tests were perfortned. . This calculates to be an avera i tests completed per month.the total tcsting remaining,115 If, however, one tests mately 10 months to complete the preop program. i l ' 3 4

assumes that rate would apply only to the original preoperat ht t .

on tests, not the retests or reperforms, and a va '

least during the time frame of the preopAssuming tests,preop then ld the 45 remaining original preops can be run in 4 months. l testing resumes in April 1984 as planned, pre '

problem is identified.

( It should be noted that a mechanism /mathod now embrac

! utility to facilitate turnovers, is that of room This is / building cumber- turn-overs in conjunction with the equipment inside. Preoperational some and could impact preoperational testing.

testing is perfortned on a system related basis, thus if a syst is complete, yet the room in which the system is placed is n (i.e., painting, etc.), preoperational testing(Note: may be, and under the current program, delayed until room turnover.

See Section A for changing completion methodology).

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23

c. Conclusion Based on the above limited review, the following conclusions were formed:

(1) The administrative process of custody transfer of systems appears to be adequate.

(2) The preoperational test program appears to be intact, viable and adequate.

(3) Preoperational tests appear to be technically and administratively adequate.

(4) Preoperational testing could conclude by August 1984.

m 0. .

Electrical ..

5 5: .-

References:

QI-QP-11.2-3, Torquing and Spacing of Concrete Anchor-Bolts QI-QP-11.3-23, Class 1E Conduit Raceway Inspection QI-QP-11.3-26, Electrical Cable Installation Inspection _

QI-QP-11.3-27, Class 1E Power Cable Meggering l

QI-QP-11.3-28, Class 1E Cable Terminations QI-QP-11.3-29.1, Verif* Ilectrical Separation l QI-QP-11.3-38.1, Installt. tion of Class IE Electrical Equip'nent

.QI-QP-11.3-40, -

Post Construction Inspection of .

Electrical Equipment and Raceways 3 QI-QP-11.3-42, Electrical Inspection of Seismic ~ ,

Category 1 Instrumentation Rack Assemblies QI-QP-11.10-1, Inspection of Seismic Electrical Support .

and Restraint Systems QI-QP-11.3-50, Cable Grip Support Installation

. Inspection

! a. General J

j The assessment in this area was to detennine if safety-related electrical equipment was being installed and inspected in accordance with NRC requirements and licensee consnitments and to detennine if Texas Utilities Services Inc., (TUSI) programs which includes drawings, procedures, quality control and construction inspections, and quality records are adequate to accomplish work in this activity, j

. Discussions were held with~ craftsmen and other Comanche Peak Steam l

Electric Station (CPSES) project personnel to determine their ability and knowledge to carry out their individual responsibilities and to I evaluate their morale and opinion with regard to the Comanche Peak j nuclear project. No adverse consnents were made by the Comanche Peak l

l

- - r-y- e,--- _

. # _ _._ - w ;m

24 d the project to be of high quality project employees and all considere ower into a Building Management r construction. -.',  ;

t efficient use of project resources. L .v,7 The licensee recently organized his manp Building, Safeguards Building.Each t organ

~ Organization (BMO) Containment to make the mos l Building. N' C There are four main BM0s construction, and QA personnel.

i n in their Auxiliary Building, and Electrical Controis NiK@

hh$

r'f; a This group supports the effort to comp og area of assignment The department under the supervisorsnnel report to the applicable are respons lution ~ Sy n,m Director.

direction of their There.is personnel, anand exchange QC perso of problems andIM reso N$f5 QA Department manager.of problems among the project personn QJE As a room or area is considered

1) is performed f thenearly com on final Electrical the' room WW

- i an tion Verification (QI-QP lly 11.3-29.The triggers the Post Raceways (OI-QP-11.3 Constructiori completion When 40). or near c '

area.

vpWJ@

Separation Verification itemsorusua essentially complete, and/or at Inspection of Electricaltor, Equipment the room and/or and area ding becomes deficien-t US both controlled.

these the discretion of procedures the BM0ding direcwork. are complete, The BM0 Director This '

$hd J",

Ac cies or complete other known i ns outstan Startup and Test Group. ~

detennines when this ctions room ofandQI-QP and/or comrletion 11.3-29.1 areainspect of andmost QI-QP J of l

turnover 11.3 40.

usually follows the inspedefic T

An inspection walk downbewas essentially complete.

clean,performe that electrical / mechanical i s. v the BM0 Director considered tray attachments, identificat to tray fill and cable spacing s on showed thatincluding the rooms cable

/ areas were(Kellen grips or separation, barriers, l

cable of cable trays, conduits, and cab es, le supports i

(where equivalent app)licable) were satisfactory.

in trays, and cab q i g Procedures h@

B

b. Review Effort '

4 Review of Quality Assurance Implement n d to assure tha E (1) complied within the areas i

requirements and comitments were be lfn The relating to the installation an ment and components. nd acceptance criteria for These procedures provided check lists a QC inspector.

., c.

! " ..a' . p A' eg, ,. j

~Mb g.i.j g 1.q h y

25 (2) Electrical Cable Installation The following installed safety-related (S/R) electrical cables that had been accepted as satisfactory by. site construction QC inspectors were examined. A physical examination was made'to detemine compliance with applicable design and installation identification ta criteria relativepoints, at termination to type, location minimum / routing,(where bend radius applicable)gs ,

cable color compatible with designated raceways and separation of trains, excluding barriers, which are perfomed prior to or concurrent with QI-QP-11.29-1, " Verify Electrical Separation."

The routing was checked by using a signal generating device.

~

Type From o To, Cable No.

b EG100483 3/C No. 10AWG MCCIEB2-1 MCC1EB2-1 MOV 1HV5540 -

CP1ECPRTC06 -

T' EG113626 9/C No. -12AWG 9/C No. 12AWG MCC1EB2-1 CPIECPRTC05 i EG113646 MOV IHV4759 CPIECPRTCOS

- EG112219 2/C No. 12AWG MCCIEB2-1 MOV 1HV4759 EG100497 3/C No. 8AWG 5/C No. 12AWG MCCIEB2-1 MOV 1HV4759 EG112216 SWGR1EA-1 TBXCSAPCH01 E0100009 1/C No. 4/0AWG 5/C No. 12AWG MCCIEB3-1 MOV 1HV4758 E0112206 7/C No. 12AWG MCCIEB3-1 CPIECPRTC04 E0112207 2/C No. 12AWG CPIECPRTC04 MOV IHV4758 E0112209 The cabTe identification is accomplished by an alphanumeric coded tag and by the color of the cable jacket. The first character of the alphanumeric code indicates whether the cable is safety ,or

' channel oriented (E), associated train (A) or non-safety (N). The second character identifies the color of the cable jacket and with respect to safety-related (S/R) applications they are "0" "Y" '

(Orange), "G" (Green), "W" (White), "B" (Blue), "R" (Red) and (Yellow). All cables are to be tagged with their unique t

alphanumeric number at termination points in equipment and junction boxes. Cables that enter and leave a junction box but t

are not teminated in that junction bcx are not required to be identified in that box with their alphanumeric number. All of the above cable were properly identified.

l t

j The routing of the above cables was checked with signal tracers.

Using this method, function box covers, cable tray covers, fire

[

barriers and other items did not have to be removed. This check l

i showed that cable tray systems and conduits appeared. to be L properly installed with proper attachments and supports, that these systems were properly identified, and that the cables i

travelled the. route indicated on the cable pull cards.

QC records showed applicable inspections were trade in accordance with the following procedures:

k e4 pg' F (- ~ '-~~~

' ~ * -*-f *- e a .r _ ,_g --

i:

26 QI-QP-11.3-26, Electrical Cable Installation Inepection QI-QP-11.3-27, Class 1E Power Cable Meggering t QI-QP-11.3-28, Class 1E Cable Terminations .

' [e (3) Electrical Cable Temination E A physical examination was made on teminations of selected class r Sj 1E of theelectrical cables unit 1 safeguards in the Hot Shutdown Panel onRe building.

teminations were in compliance with requirements, including proper lug material and size, accurate location, The cable and identifica-wiring diagram

,E i

tion of was terminal used block and to determine conductor.

the proper termination points and conductors identification. Cable Terminations that were checked were for L cables EG104556, EG111148, EG104551, EG139204, E0104791, ,r E010 p

~

- e =

I E0122101, E0104742, E0130596, and E0122103. '

K The QC records showad that inspections were made on these temina Y

tions in accordance with QI-QP-11.3-28, " Class IE Cable Temina ~ .

tions." _0 QI-QP-11.3-40, " Post Construction Inspection of Electrical

" Separation between field run [v Equipment and Raceways" states: redundant Class 1E cables and [t a cabinet shall be maintained in accordance with the equ specification. ,

ments, the minimum separation distance between redundant Class IE and Class 1E/Non-Class 1E cables shall be greater than or equal to ,

In cases where the above separation criteria cannot be 6 inches.

maintained, barrier shall be installed between the cables."

Acceptable barriers include the following:

Metallic conduit, including Servicair Company FC 33 flexible (a) conduit Two sheets of fire retardant material separated by a minimum (b) of 1" of eir space or thermal insulating material (c) A single barrier with a 1" maintained air space barrier During the cable temination inspection in the Hot Shutdown Panel it was noted that barriers were installed but there still existe l some separation problems.that QI-QP-11.3-40 inspection had not b and that the remaining barriers would be installed as needed to meet the separation criteria before QI-QP-11.3-40 was signed off for that room or panel.

l

~~w .-~. 3 3 :.9.g;.;;Q

rs

a. ,.t-" ' -

27 in panels was being P 11.3-40 was essentially To insure thatThese internal panels wereelectrical located in the cable separationadh ,

complete were examined. The panel examined includedRelay P spreading room and control room. tisfactory ,

termination S.

cabinets TC-22, 23, AuxiliaryThese pan f these panels.

O even though work was still in process i in in some the above o N4 During the inspection for electrical separat onin ts the panels were

+M N-panels it was noted that some cablesThis 525 " Wire was was detemi ...

6~

spliced. field cables that FSAR coanitments which state in paragraph binets or racks.

The 8.1. . . ., : 0;:

splices are used in limited applications on length ,3 teminate in certain Class 1E panels, ithout ca the usel of wire -

! 3 31 nomal design is tosplices The wire teminate

~

arefield onlycables used where w additionad reason P.M splices.

is required for theThe field esewire andwire of such it was splices not hasjudge been-- ll<,y The crimping pull a new' field cable.The wire splices are butt splices.plices are identical .

minimized. t panel. The wire b-technique, device and materialsi used tions for such the as s control .W to those used for the teminal lugs in tha t rials N splices are only allowed on low dpower the required appl caSinc "

cables. l * . (l. .

are used, the splices are limited to low powfiel .,

9 wire separation and wire bundles support conducted b'y Interviews with CPSES project t d that there may be personnel a which were other problem members with cable of this review terminations to We tecm ction.indica The eidmuller These teminal blocks employi al screw h ire clamp is stripped conneblocks manufacturer's literature for dthese clamped withouttem any na s -

screw clamp" of its insulation refers toi length to a recomended a connection an in wh

" As inspectors e further preparation.

direct contact on theinsure the connection; wire, damage is would prevented. and since ;;th

" they tug p were making inspections for QI-QP-11.3- g Inspection of Electrical Equipmenttoand h slightlyRaceways,onnecti spread and flex the conductor to insure w clamp that tec connection. gl This action caused the conductor wire Since these connections were previously veri fied. as satis factory p i

e now calls for a  !

per QI-QP-11.3-28 " Class 1E Cable lemina fact that equipment may be energized, the license 3 40 te visual inspection with regard to QI-QP-11. -

qualified per The Weidmueller Terminal Blocks used at CPSES arepp the manufacturer's literature for nuclear aTests for this qualific environmental qualification.

Y ,

28 performed by Franklin Research Center and are documented on their j

reports FC 4959 and 5205.

(4) Electrical Conduit and Cable Tray Installation Conduit and cable tray raceway systems were inspected in rooms and/or areas in which both QI-QP 11.3-29.1, " Electrical Separation '

Verification" and QI-QP 11.3-40, " Post Construction Inspection of Electrical Equipment and Raceways," were essentially completed and This access to these rooms and/or areas were controlled.

inspection was to verify completeness of work in the electrical area, including electrical separation, power cable spacing inAll  !

trays and cable supports on vertical runs of cable systems.

items were condiderea to meet construction criteria.

I Specific Conduit System checked including support and spacing - -, -

- were: '

TL. . .

Location Remarks ' '

Conduit No. .

C13005319 Safeguards Bldg #1, Elev 773, Room 565 Access Controlled _

C1304036 Safeguards Bldg #1 Access Controlled Elev 773, Room 565

~

C13012998 Safeguards Bldg #1 Access Controlled -

Eley 773, Room 565 C13010777 Safeguards Bldg #1 Access Controlled Elev 773, Room 565 C14013679 Safeguards Bldg. "I Access Controlled _

Elev 773, Rocm 54 Aux. Bldg. Eley. 790 Only Room 170 was C22G08188 Various Rooms Access Controlled Aux. Bldg. Elev. 790 Only Room 170 was C22G08189 Various Rooms Access Controlled i

The inspection of these conduits showed that they were installed

! to the construction requirement and that electrical separation was satisfactory. QC records for these conduit systems showed that applicable inspection were made in accordance with the following procedures:

QI-QP- 11.3-23, Class 1E Conduit Raceway Inspection QI-QP- 11.2-3, Torquing and Spacing of Concrete Anenor Bolts k ,

~

b l

• "' IT-' _ _ . ,

" ' ' ' ~ . ,  ; __g ,. , _, * [-' }_

~~; % v -' y ;; ; ~ ~- n' E. -

.2 ._. _ _

l ~.._.__m. .

29 Inspection of Seismic Electrical Support and QI-QP 11.10-1, Restraint Systems QI-QP 11.3-29.1, Verify Electrical Sep(aration [For Room 565,I.R for' Room Inspection Reportfor Room 54, IR# E-1-0013480/3-8 170,IR#E-1-0017514/1-84] 'r .,

ld to .-c' Several additional conduit runs were examined p; i

'y.h verify electrical separation.the cable spreading area and are u-

, ~

orange safety train, goes under ladder d separa-(a) Conduit C12019632, tray T16GCCM02, green safety train, el at tion is approximately 6 inches.

over ladder tray T14GCDH41, green safety the- train, and sepa _

tion is approximately 2" with a barrier installed between r

~

tWO. _ d

Conduit C15R10537, red protection channel, d separa- at one poin (b)

' under ladder tray T13GCCM15, green safety train, an tion is approximately 2 inches. ;4 Conduit C15B11396, blue protection channel, d separa- at one c.po (c) under ladder tray T130CCMO, orange safety train, an n

. tion in approximately 2 inches. S

!~

Conduit C12G21191, green ' safety train, goes under3'- so (d) T140CDJ31, orange safety train, and separation is a I' '

mately 3 inches. i l '

11.3-29.1 " Verify Electr ca Section

,.3 R

The above are acceptable per QI'-QPSeparation U 4.11.3.2. f hich Spacing of power cables in trays is to 4follow bles shallrequireme Gibbs and Hill Specification 2323-ES-100 section t cable. 4.2.1. .,(

in. essence, states that minimum spacing between powerrI be a minimum of one quarter of the diame I considered to meet this requirement: j

- Electrical Separation Verification location _ per OI-QP-11.3-29 Tray Numbers -

Not complete Room 174, Aux. Bldg. Not complete T120ABA05-12 Room 174, Aux. Bldg. Not complete T120ABB01 Room 174, Aux. Bldg. Complete T110AA01-05 Room 54, Safeguards Bldg.

T1105AA30 l

i 4

. e O"

" a" i - =

~

30 l

T120ABA96* Room 219, Aux. Bldg. Approx. 90%

complete T11GAAB11* Room 214, Aux. Bldg. Complete T120ABA98* Room 241, Aux. Bldg. Approx. 90%

complete ~ -

4 T120ABA47-50 Room 241, Aux. Bldg. Approx. 90%

complete

.' T120ABB93 Room 219, Aux. Bldg. Approx. 90%

' complete ,

• Asterisked trays contained vertical runs of cable. Cables were supported properly by Kellem Grips in accord:nce with QI-QP-11.3-50, " Cable Grip Support Installation Inspection."

4 A review of some of licensee Inspection Reports (irs) that were E perfonned for QI-QP-11.3-29.1 " Verify Electrical Separation" -

.r showed that I.R E-1-0024985 of 2/28/84 and IRE-1-0036072 of '.-

T ' 4/12/84 applied to the same room (room 219) in the auxiliary i

. building. Neither of these reports indicated that they were -

perfonned as a result of a specific job or Inspection Item Removal Notice (IRN). Both were designated as final inspections. It is recognized that the licensee can perform re-inspection as deemed 1 . necessary; however, it is considered that there should be only one final inspection for post construction work. If additional final i inspections are required in this area for IRN's, Design Change i Authorizations (DCA), etc., they should be referenced in the 4

remarks section of the IR. The one " final" electrical separation

inspection, which could be perfonned concurrent or before

QI-QP-11.3-40 " Class IE Electrical Post Construction Verifica-l tion," would indicate that electrical work in this area in almost I complete and would aid in t-iggering the performance of -

QI-QP-11.3-40. The licensee stated that this area would be reviewed to see if the " final" inspection in this area could be clarified.

c. Observation and Conclusions There appears to be a good working interface between construction inspectors and the craft. For the most part the electrical construc-tion inspectors appear to be knowledgeable and conscientious in their work areas. The inspector encountered no cas'es of hostility or j harassment with the Comanche Peak Project employees.

[d E. Design Activities / Design Control 1

j

References:

QI-QAP-11.1-28 Rev. 23, Fabrication, Installation Inspections of ASME Component Supports, Class 1, 2, and 3 1

_- f-

'o - ,

31 QI-QAP-11.1-28A, Rev. 5, Installation Inspections of ASME Class 1, 2, and 3 Snubbers Procedure AB-5, Rev. 5 A Simplified Method for Design and Analysis of Small Size Piping J

TUSI Engineering Guidelne, Section IV, Base Plates,

• Rev. 11

!- CPSES, XCP-ME-10, Rev. 1, Pipe Support Adjustments I TUSI CP-EI-4.5-1, Rev. 9, General Program for As-Built Piping Verification p Section II, General Engineering ;

TUSI Engineering Guideline, T. -

5ectopm II - Criteria for Pipe Support -

Design, Rev. 8 Specification 2323-MS-46A, Nuclear Safety Class Rev. 5 Pipe Hangers and Supports Construction Drocedure Field Surveys 35-1195-CCP-9, Rev. 4, TUSI CP-EI-4.6-9, Rev. 1, Performance Instruction for Piping Analysis by SSAG TUSI Engineering Guideline, Hilti Concrete Anchor l BoltsSection V, Rev. 3 _

ADLPIPE, Static and Dynamic Pipe Design and Stress Analysis, Arthur D. Little, Inc., May 1981

a. General The organization of the general site engineering, construction, and procurement efforts were defined in procedure CP-EP-3.0. By this procedure, the Project Manager is responsible for the Comanche Peak Steam Electric Station (CPSES) design and engineering. These activi-ties are normally delegated to Gibbs and Hill, Westinghouse, and other J(

i organizations. However, the licensee, (TUGCO) retains overall respon-I sibility for design activities and performs design functions as j necessary. The TUGC0 Engineering Manager is responsible for the i general direction of engineering activities.

i FSAR Chapter 3 provided the licensee's requirements for the design of 1

structures, components, equipment, and systems. The reviewer selected

?

l samples in pipe support design, piping stress analysis, and design i

~ ' r-r3 .,.n..,. ., .

~'%L'

,, _ S"'W'** T - ,  ; 7_-- _

T 32 to ensure site f procedures procedure applications to verify program implem satisfy NRC requirements and licensee comitments.  ::

1.1

b. Review Effort l in fly The reviewer held discussions with the design t od the to engineering person d *r the pioe support group to determine whether they ble verify un ers o i applicable design control procedures; whetheri they d/orwere a '

review design parameters that were within the applicable The criteria an -

design specifications; and whether theh person design. doing el in the the des gn  ?

e-was independent from the individual who h established perfonned  ?

i piping system Site Stress Analysis The seismic response Group d safe '- -

(SSAG) ^F instructions, procedures, and specifications. the responsible sRectra with respect to operating (SSE) were discussed with basis earthquake (OBE) ,n anl ,

earthquake shutdown It was noted that these seismic response h spectra wero {-

ite engireers. Thehome followingoffice to t e s furnished by the A/E's (Gibbs and Hill, Inc.) l n

major areas were reviewed to deternine t a conclu b IE Bulletin 79-02, Pipe Support Base Plate Designs UsinggC (1) Expansion Anchor Bolts Requirements 1 Factor of Safety for Concrete Expansion Anchor Bolts Des

~

(a) '

q A review of the Pipe Support Engineering TE h b

E conservative value) has been used for'establisfi lt gte allowable loads (tension and shear)factor for the wedge calculation. t manual and the NRC IE Bulletin As noted above, 79-02 requ

- installations use Hilti wedge bolt only).the safe Pipe Support Base Plate Design i (t)

IE Bulletin 79-02 states that pipe support base plate flexibility beDiscussions accounted for in the calculation of anchor with the responsible engineers bolt loads. that the pipe support group personnel do cons indicated into their design calculations.

base plate flexibility (base plate flexibility consideration)

Finite element methodhas been used for non-typical (o in one plate) base plate analysis.

has been utilized for all The typical FUB II(fourprogramanchor holts in.one generally plate) base plate analysis.

i l

9 i- i s * - - - - .

^~ -

2.,,___ _, _ g. - ,

m_ _ _ _

~

33 -

produces loads which are aboutIn25% higher fact, many basethan the loads generated by the Firrite Element Method.

plates were analyzed by the more conservative This program, F computer application (developed by ITT Grinnell Corp.). ,

n approach exceeds the NRC requirements. "

(c) Anchor 30lt' Tension - Shear Interaction IE Bulletin 79-02 pemits a formula to be used for calcula-This fo tion of bolt tension-shear interaction.

interpreted from a linear distribution to an ellipticalC '

distribution.

to use a linear distribution (a conservative approach) for ."

all concrete expansion anchor bolt calculations.

IE Bulletin 79-14, Seismic Analysis for As-Built Safety-Related ';

(2)

Piping Systems, Requirements. _

4 This bulletin states that the seismic analysis input infomation ?

c6nforms to the actual configuration of safety-related pipingpipe h Licensees are requested to verify:

systems. ,

support and restraint design, locations, function and clearance k embedments; pipe attachments; and To accomplish the valve above and valve requirements, thaoperator g locations and weights. o site pipe support group and the site stress analysis groups are The This g

. responsible for verification based on as-built configuration. p 1

as-built configuration is identified by a field survey team.  !

field survey team, which consistshtof threej s

! such as transits, levels, theodolities, etc.the infomation o for implementing the IE Bulleting 79-14 requirements.

l Alternate An'alysis for Small Bore Piping Systems (3)

The reviewer examined portions of procedure AB-5, A Simpl Method for Designs and Analysis of Small Size Piping, Rev. 5 1982.

It was noted that the procedure was developed by Gibbs a Hill, Inc., in a very conservative Furthermore, manner in30%

approximately termsof of therma and small seismic boreload (2calculations.

inches and under) low energy pipe d lines in Un and 10% in Unit 2 are analyzed by the Alternate Analysis Me (i.e.,)a piping .

simplified The balance of small method bore pipingfor design is analyzed by the and ana computer application.

(4) Rigorous Analysis for Safety-Rehted Piping Systems i

Most of the safety-related piping systems are analyzed j rigorous analysis method. analysis is one of the typical progr

_,_y_____ ,,

s"- , Q E ;'

~ _ m . -- %

34 industry. This computer program, ADLPIPE Static and Dynamic Pipe Design and Stress Analysis, has been developed and updated by Arthur D. Little, Inc., since the early 1960s.

(5) Iterative Design Process The reviewer held discussions with responsible licensee represent- '

atives in the area of safety-related pipe supports and piping systems. It was noted that the Iterative Design Process was utilized for implementing the design of pipe supports and the In accordance with 'the licensee's anglysis of description: piping systems."the process for the design of piping and supports is iterative in nature. It is unrealistic to expect to design piping and supports to satisfy allSuch applicable requirements an iterative the design approach first time through the process. -

is employed throughout the nuclear industry, and The is utilized reviewer in 1- -

the design of other nuclear components as well."

C noted that the praccices at C manche Peak are not unusual compared; to practices at other nuclear facilities in terms of using the-iterative design process in the area of designing pipe supports ~

- and piping systems.

(6) Review of Design Calculations for Pipe Support Pipe Size Piping System ,

Support No.

AF-1-002-705-S33K, Rev. 3 10" dia. Auxiliary Feedwater CC-1-158-701-A43R, Rev. 2 16" dia. Component Cooling SI-1-031-709-A32R, Rev. 2 12" dia. Safety Injection .

SI-1-029-702-S32R, Rev. 2 24" dia. Safety Injection 2" dia. Boron Recycle BR-1-AB-001-005-3, Rev. 1 .

The above design calculations were randomly . selected and were partially reviewed for conformance to analysis criteria, applic-able codes, NRC requirements, and the licensee comitments.

Furtharmore, these calculations were evaluated during the review Deflection for thoroughness, clarity, consistency, and accuracy.

criteria used for support design were discussedWeld withsize the respon-calcula-sible engineers and were partially verifiel.

tion and snubber size detemination were also verified for In general, the design calculations appeared to be adequacy. reference, units adequate in terms of using ) design input,(dimension, force, a l

! l l

1.

-~~__ . --m .

_ ~

~ ; ;. . ~ ]

35 y

(7) Review of Stress Analysis for Piping Systems

~

Calculation No. Piping System AB-1-19A Safety Injection AB-l-30 Containment Spray AB-1-69 Residual Heat Removal and Safety Injection AB-1-135E Auxiliary Steam and Main Steam i AF-1-SB-006 Auxiliary Feedwater II 'AF-1-SB-007 Auxiliary Feedwater r s:  ;

ji: The above piping stress analyses were partially reviewed for conformance to design specification, applicable code, NRC require-

ments, and the licensee commitmerits. These analyses were also -

- evaluated for thoroughness, clarity, consistency, and accuracy.

The NRC reviewer examined portions of the seismic inputs to be

, used in the stress analysis. These seismic inputs in terms of periods versus accelerations from the corresponding floor response spectra curves under OBE and SSE conditions were partially verified for accuracy. Furthermore, the reviewer held discussions with the responsible engineers to ensure that seismic anchor movement, nozzle thermal movement, and valve orientations were properly considered in the stress analysis.

During the review the reviewer examined piping system AF-1-SB-006. -

This 3/4" dianeter vent and drain pipe was analyzed for support requirements. Results from the analysis revealed that no pipe

- supports were needed for the pipe. However, the reviewar noted that a Component Modification Card (CMC) No. 90567 was issued to the pipe in trat a piece of tee' (pipe) was added to the vent and

. drain system. The pipe support group accepted this CMC without

, performing detailed evaluation. The responsible engineer stated that this CMC was reviewed by a well qualified engineer. Based on his engineering judgement, no detailed calculations were required.

2 The inspector indicated that a detailed evaluation for this CMC was needed. In addition, a sampling program should be initiated ,

to ensure that no other similar CMCs were accepted without performing detailed evaluation. The responsible licensee repre-sentative took immediate action to perform detailed calculations for the vent and drain piping system due to the addition of the CMC (No. 90567). Furthermore, a sampling program was immediately initiated to review 53 other similar packages. This matter will be identified to the Comanche Peak Project Director for followup.

p h

I 1

J5 l A,.C G l &_ l __T Z'

_ ,7 3 ' ** ? -

'k

r~~, .<

i . .

I 36 l

i Results from the detailed calculations revealed that no pipe supports were required for the vent and drain piping system as the original evaluation indicated. Results from the sampling program showed that no discrepancies were identified for the 50 other similar packages. ,

( Piping system AF-1-SB-007 was partially reviewed. It was noted that portions of the calculations were not performed in accordance with established procedures. Some minor mathematical errors were noted. One CMC was not addressed properly by the licensee reviewer. The pipe support group reanalyzed this 3/4 inch piping j system by hand calculations (alternate analysis) and also by

< computer application (rigorous analysis). Results from t'e n two

! analyses were consistent and conservative. Four pipe supports p were required by the analysis. Loads used for support design were verified and were found conservative. This matter will be .

forwarded to the Comanche Peak Project Director for followup. *}

N. .

l - (3) Field Inspection / Verification

- The NRC reviewer performed a field walkdown at the Unit 1 containment building area and noted the fcilowing discrepancies:

, Support No. Status CC-1-218-012-C53K Snubber connection cotter keys missing t

CC-1-295-005-C53R Sway strut installed over 5' tolerance CT-1-038-436-C62K Snubber connection cotter keys missing; *

'j no washers in rear bracket _

t CT-1-117-405-C62K Snubbar connection cotter key missing ,

CT-1-117-415-C62K Snubber safety wire broken l ,

) CT-1-053-444-C62K The south snubber was installed j improperly j ,

DD-1-046-020-C65R Snubber cotter keys missing j.

FW-1-096-705-C62K Snubber safety wire broken FW-1-102-002-C62k Snubber cotter key missing; needs j

relative adjustment on snubber I FW-1-102-003-C62K Snubber cotter keys not bent MS-1-151-025-C52K Snubber installed over 5' tolerance 3

CC-1-RB-066-008-3 Snubber cold setting over the limit

.m s. -m --

37 Snubber cold setting over the limit ,

CC-1-RB-066-007-3 Spring hanger cold setting incorrect  ;

CC-1-RB-068-007-3 (15 lbs. versus 11 lbs.)

f The above pipe supports discrepancies were verified with theAll F licensee's QC inspector in accordance with detailed drawings. 5 the above pipe supports were vendor certified The and wererepreseata-licensee previously  !

f inspected by the licensee QC inspectors.

tives stated that a final walkdown inspection / verification for all i pipe supports is to be implemented in accordance with procedure j CP-QAP-12.1, Inspection Criteria and Documentation Requirenents  !

Prior to System N-5 Certification.

The majority of the discrepancies appeared to be minor problems  !

3 which _ could be easily repaired during the final inspection prior -

Two of the discrepancies were more ,

i to the system pressure test. '

- l:

serious in that rework or reanalysisThese of thesupports support would are be -

required prior to acceptance.

MS-1-151-025-C52K, Rev. 3 and CC-1-295-005-C53R, Rev. 4, which The were not installed in accordance with the detailed drt. wings.

fact that these two supports were inspected by QC is considered as a potential enforcement item.

(9) Design Consideration for Piping Systems Between Safety-Related and Non Safety-Related Buildings The NRC reviewer held discussions eith the licensee representa- ~

tives in the area of plaing stress analysis and pipe support -

design. Stress Analysis No. AB-1-135 E for the Auxiliary Steam and Main iteam System was partially reviewed and discussed with respect to design considerationsThe between piping system safety-related was and classified as non safety-related buildings. The pipe run starts from the high energy line and safety-related. Since i Turbine Building into the Electrical Control building.

seismic classifications for the two buildings are different, the criteria used for the piping system analysis should also be different. The failure of the pipe in the Turbine Building may impose a damage to the pipe inside the Electrical Control Building The if the piping system was not properly analyzed and designed.

responsible licensee representatives agreed to perfcrmed furtherThis mat evaluation with regard to the above concerns.

identified to the Comanche Peak Project Director for resolution. .

(10) Interpretation of Tolerance for Snubber Installation three reviewers interviewed the During the field review, licensee's CC inspectors with respect to their in five degrees tolerance requirements for strut and snubber These QC inspectors appeared to be confused with installation.

the interpretation of the tolerances on the detailed drawings.

r x

,4.; - @_pp g

.. i.

38 i l

The reviewers held discussions with the licensee representatives with regard to the above concerns. It was detemined that the licensee will revise the inspection procedure to clarify the strut / snubber installation tolerance and will conduct a training for all QC inspectors who are involved pipe support inspections.

This matter will be identified to the Comanche Peak Project Director for followup. ,

(11) Final Adjustments for Spring Hangers and Snubber Settings The reviewers held discussions with the responsible licensee representatives with regard to implementing the final adjustments for spring hangers.and snubber settings. It was determined that, after the fuel loading, the licensee QA startup group will perfom the final walkdcwn inspection to ensure that all spring hangers I. and snubbers be adjusted to proper position. This matter will be -

%' brought to the attention of Comanche Peak Project Director fori -#

~

followup. --

(12) Technical Trd ning The reviewer held discussions with the responsible pipe support engineering (PSE) personnel to detertnine whether they performed their work activities in accordance with established proceddres and specifications, and whether the design e~ngineering personnel received proper training with respect to technical applications and NRC requirements.

~

A review of the training record revealed that since 1980, the PSE personnel have received extensive training activities in terms of -

3 technical applications and code interpretations.

9 l

Portions of the training courses are listed as follows:

1 Date Course Attendance (Engineers) l (a) 06/16/80 Introduction to Nuclear All Codes and Standards, QA for Engineers (b) 10/13/80 ASME Code Seminar All 10/14/80 (NFDesignPhilosphy)

(c) 04/13/81 Alternate Analysis Method 26 .

for Small Size Piping (d) 06/21/81 Vent and Drain Piping 8 l Seismic Qualification

~

(a) 05/11/82 Design Verification 34 05/13/82 Process i

- ~ _

39 (f) 07/14/82 Pipe Support Snubber 24 07/15/82 Installation (Instructed by Manufacturer)

(g) 07/27/82 Analysis of ASME Class 16 2 and 3 piping (h) 11/12/82 Seis,mic Analysis of 65 11/16/82 Pipe Supports 11/17/82 (i) 06/14/83 Finite Element Method 19 thru (including ASME 1, 2 & 3 08/06/83 pipinanalysis)

E (j) 06/29/83 Current Version of ADLPIPE 9 ,-

1

-- Computer Code (Stress Analysis)

(k) 11/17/83 Quality - It's Your Job All .

(1) 03/08/84 Snubber Reduction Program 6 (m) 03/19/84 Stability Problem in the 26 Design of Pipe Supports The above training activities in the area of pipe support designs appeared to, be effective and well administered. This observation was supported by the extensive discussions with the responsible -

engineering personnel and by reviewing the procedures and results of the design calculations. ,

c. Conclusion

Discussions with the responsible personnel revealed that the enginee-ring personnel involved in the area of stress analysis' for piping systems and pipe supports appeared to be knowledgeable. A review of ,

portions of the alternate analysis criteria and related documents was perfonned. It was noted that the methods and procedures used in the criteria were conservative. A review of the eleven calculation packages indicated that computer applications were extensively used in the. stress analyses, pipe support designs and, base plate and concrete expansion anchor bolt calculations. Design calculations, in general, were good.

During the review, the NRC reviewer noted that conservative considera-tions were found in many areas of design and analysis. These conserva-tive considerations included: factor .of safety used for concrete expansion anchor bolt calculation, computer program (FUB II) used for base plate analysis, weld stress allowables for welding connections, l i _. c ,

40 alternate analysis for small bore piping, and seismic loads used in design and analysis. These consecutive design considerations are censidered strengths in'the applicants program. Finally, the reviewer noted that the geographic location of Comanche Peak site has the lowest seismic risk -in the United States in accordance with the criteria specified in Uniform Building Code.

4 A field walkdown inspection performed by the reviewer has resulted in various discrepancies for 14 pipe supports that had been previously inspected by the licensee's QC inspectors. This item will be referred to the Comanche Peak Project Director to perform subsequent followup to ensure that safety-related pipe supports are installed in accordance with design drawings and to verify that corrective actions with respect to the aforementioned discrepancies are adequately implemented in accordance with established procedures.

7, __

v; 14 Installation of Safety-Related Fluid Systems =

1: _

References:

(a) QA-QAP-11.1-26, Rev. 14, "ASME Pipe Fabrication and Installation Inspec-tions" (b) QI-0AP-11.1-28, Rev. 23 .

" Fabrication, Installation Inspection of ASME

[ ,

i Component Supports, Class 1, 2, and 3" i

(c) QI-QAP-11.1-28A, Rev. 5, " Installation Inspections l

of ASME Class 1, 2, and 3 Snubbers" (d) CP-QAP-12.3, Rev. 3, " Testing Phase Quality Assurance Functions Prior to ASME Code Certification and Stamping"

~

(e) CP-QAP-12.2, Rev. 7 " Inspection Procedure and Acceptance Criteria for ASME Pressure Testing" l a. General The review of this area was directed to assessing the adequacy of the licensee's construction program as it pertained to installation of safety-related fluid systems required for safe operation and shutdown of the plant. The assessment was undertaken through selective examina-tion of installed systems and installation related activities to determine whether they were accomplished in accordance with good engineering practice and with licensee commitments and NRC require-ments - including the requirements of the applicable code, ASME

- ' ' ' ' _, , . -e.-- _ __ , _ _

• • a W= - __ e _.,__ _

__ , g , ,

i.izil_mAm.

41 luate the The review in this area did not undertake to evapiping in ..

Section licensee's III. final checks and analysis d did of notsystem examine attach-ment of the fluid systems to concreteInspection QCa  :.,

ys: :C t

d System Components 1 YN ;

b. Review Effort Tour of Areas Containing Safety-Related Flui

!M(3h $,

(1) iliary, and Reactor (piW The reviewers toured the Safeguards, Aux Station to observeisua (j

..;f@A related fluid system components for any v W Buildings and the Service Water Pumpingin p$pdff apparent signs o ible support components visual weld defects, undersize welds, fasteners and was ,

C1 supported pf ping, damage to more Only one item of concern, requiring follow-up,fcund suscept(e s

[O,4 i

spacers, etc. A spring can piping support wasinside the ca -

hich was (

identified Theduring thewas tour.to have of a this significant buildup of rust licensee infomed spring The rusting in this item did notimpair-: ican, w spring. identified Serial No. 942-12.  ;

appear to be so severe as to significantlyifi ance to the func c

the course of the rusting and its signshould be e Control of Welding Materials l of the welding (2) ft related piping system h code The reviewers examined the licensee components at the issuance stations to ver II requirements and good practice.to the adequacy of t f filler metals, 1 segregation, identification, and control o h including consumable inserts to limit moisture lj oven storage of low hydrogen electrodes f pick-up preparation of issuance records handling of returned filler metals i limitations documentation of current welder qualificat onthe plan The reviewers also observed areas toured trolled int red for spe cribed in (2) above, and plant areas en e fillertrolled inspections for evidence No evidence of inadequately of uncontrolled con or improperly conTh materials. welding materials was observed.

s

. ~

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t 42

$ controls observed by the reviewers met or exceeded code require-3 ments and good practice.

(3) Piping and Supports The reviewers visually examined examples of installed runs of '.

d .

S safety-related piping and associated supports to verify they were '

in accordance with good engineering practice and that they were in

,' compliance with code requirements and with licensee drawing and

-i procedure requirements. Three runs were selected which had most or all of their final acceptance inspections completed. Two of

- these were nearly ready for the final code review required for l ASME certification (referred to as N-5 certification) that the 1 installations were in accordance with the code. The third had the certification complete.

y The licensee contracted the piping and support installation work. -

to Brown and Root, Inc. This- contractor was responsible for 1

- assuring compliance with code requirements, including obtaining' code inspector certification therefor (on N-5 Data Reports).

~

Licensee procedures applicable to and utilized by the reviewers in d the examination of piping and supports were examined for compli-ance with code requirements. The procedures were as follows:

(a) QA-QAP-11.1-26, Rev. 14, "ASME Pipe Fabrication and Installation Inspections" 11 (b) QI-QAP-11.1-28, Rev. 23 " Fabrication, Installation

? Inspection of ASME Component d

Supports, Class 1, 2, and 3" _

(c) QI-QAP-11.1-28A, Rev. 5 " Installation Inspections of ASME Class 1, 2, and 3 Snubbers" (d) CP-QAP-12.3, Rev. 3, " Testing Phase Quality Assur-4 ance Functions Prior to ASME

' Code Certification and

' Stamping" d (e) CP-QAP-12.2, Rev. 7, " Inspection Procedure and Acceptance Criteria for ASME a Pressure Testing" i

4

.I 1 I l

~

'TEP -T , y , ee , - g - _s ,-

4. .

43

~

The runs of piping and supports installed that were examined by l

the reviewers were described on isometric drawings. The runs examined, identified by the drawing numbers, and the examination checks made by the reviewers are as follows: .

Run: 3" Containment Spray (ASME Section III, Class 3), Drawing BRP-CT-1-SB-019, Rev. 6 The reviewers visually selectively examined the ' installed safety-related piping to verify the following in accordance with the drawing, code, procedures; and good engineering

. practice:

configuration

- apparent pipe size E -

valve identification '

5:

- visual appearance of welds  ;

':~'

- - heat numbers on pieces 2,10, and 18 and serial number '.

on valve piece 14 were traceable through installation ~

records to original receipt and acceptance records -

The reviewers examined the records for the above piping to verify the following in accordance with code and procedural requirements:

- proper installation and inspection steps completed for all components

- - mill test reports for all materials hydrostitic testing, l Run: 2" Reactor Coolant (ASME Section III, Class 1), Drawings -

BRP-RC-1-RB-10, Rev. 8 and BRHL-RC-1-RB-10, Rev. 2.

The reviewers visually examined the installed piping and supports to verify the following, in accordance with the drawings, code, procedures and good engineering practice:

- configuration

- apparent pipe size j - snubber and spring can sizes

- offset for snubber RC-1-015-707-C41X

- spring can settings

- visual appearance of welds

- size of piping welds

- support serial numbers 19050,17791 and 17789 traceable to installation and receiving records

- heat numbers on material pieces 1 and 12 that were traceable to acceptable mill test reports I - serial numbers on valves 1RC-8057A and -8058A that were traceable to installation and acceptable receiving inspection records t  %

t

• =*-*wo-w- , , _ . * * , - _

__ ,, , . , _ _ . -mem p , ._

=.+w-=*==.,

e--w,

44

- visual appearance of fasteners

- snubber pins and washers j - evidence of damage to or deterioration of any components i The reviewer examined the records for the above piping 'and

{

supports to verify the following, in accordance with code and j ,

procedural requirements: .

1

- proper installation and inspection steps completed for piping i

- hydrostatic testing k Run': 8" Auxiliary Feedwater (ASME Section III, Class 3),

10" Drawings BRP-AF-1-SB-006, Rev. 17 and BRHL-AF-1-SB-006, Rev. 3

.? .-

J S, - The reviewers visually examined the installed piping.and ' '. #

supports to verify the following, in accordance with the ,

drawings, code, procedures and good engineering practice:

- configuration '

- apparent pipe size

i - snubber sizes and settings 1

- visual appearance and size of welds

? - serial number on valve 1AF-031 traceable to acceptable receiving records

- snubber pins and washers d - evidence of damage to or' deterioration of any components Note

Heat number traceability could not be checked on the materials and weld quality could not be checked -

entirely satisfactorily as most of the components were painted.

The reviewers examined the records for the above piping and supports to verify 'the following in accordance.with code and procedural requirements:

- proper installation and inspection steps completed for

, piping <

- hydrostatic testing 4

The licensee's procedures and installation appeared to i generally meet or exceed the applicable requirements and were

! in accordance with good engineering practice. Records proved i readily retrievable and complete. Licensee QC inspectors who

! accompanied the NRC reviewers in their examinations of the 1 installations appeared knowledgeable. One item of concern j was noted - it was not clear what tolerance was applied to

• snubbers and sway struts that were installed with offsets or 4

)

't .

I ~. , .-

= .-

45 angles specified by drawing. This concern is discussed in paragraph E.b(10).

(4) Residual Heat Removal Heat Exchangers (RHR Hxs Supports)

The reviewers requested the licensee to identify and provide for review the bolting requirements, the drawings and the installation records for the RHR Hxs. The drawings and some of the installa-

tion records were provided. The bolting requirements were not identified and the welding records were not provided by the completion of the inspector's visit. The records and infomation had been requested about li to 2 days before the end of the visit and licensee personnel indicated insufficient time was allowed to provide all of what was requested.

!P The reviewers examined the RHR Hx supports for visual weld quality-5; (size and location were not checked) and installation of bolting.

5: The weld quality appeared sati.sfactory (in accordance with code requirements). A few nuts were seen to be very loose, with many _

~

threads exposed between the nuts and the surfaces against whrich they would tighten. Also, the threads between the loose nuts and tightening surfaces were noted to have been painted (apparently inadvertently).

The status of the final inspections to be perfomed on the Hxs was i unclear, but the reviewers were informed that a final inspection of welds and to verify that bolting was in place and remained to be pertormed.

As already indicated above, the installation records for the Hxs .

did not appear to be readily retrievable and bolting requirements were not readily identified by the licensee. This appears to be

- contradictory to the findings of the general finding of the team.

I c. Conclusions

Based on their examination and findings described above, the reviewers

! generally concluded that the licensee's program for installation of 4 safety-related fluid system components assures compliance with require-

ments, comitments and good engineering practice. As their assessment

• was incomplete relative installation of ~the Hxs described above, the

reviewers recomend additional evaluation to complete the review J

relative to such components. This will be identified to the Comanche l

Peak Project Director for followup. . .

L t

t h

I b ..m , _ _ .~ 37 -.--__--.m.__-_-.v.._..  :."._,~.>,. ...__

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46

^

e l G. Cisil Construction Activities l

a. General The objective of this portion of the review was to determine the adequacy of the implementation of the licensee's quality control /

- quality assurance program for civil construction activities. During the review selected quality assurance records were examined to verify

- the records were complete and retrievable. Emphasis was also placed on examination of the document control system. The reviewer examined site civil design activities, including the design change process, proce-dures and QA records for completed work activities such as the SSI dam,

! selected cable tray supports, and whip and moment restraints; and

! procedures and work activities for ongoing work including application of protective coatings and testing of Rictmond inserts. The reviewer F also interviewed QC inspection personnel. -

1 ~

b. Review Effort _

(1) Safe-S.1etdown Impoundment Dam, Units 1 and 2 _

(a) Review of Construction and Quality Control Procedures

[

The reviewer examined specifications, drawings, and quality

! contre 1 procedures for construction of the safe-shutdown impoundment (SSI) dam. Acceptance criteria utilized by the

? reviewer appear in FSAR Section 2.5.4.5 and NRC requ'irements.

Construction of the SSI dam was completed in Spring of 1977.

The dam was designed by Freese and Nichols, consulting

' engineers, and was constructed by Brown and Root. The onsite quality control inspection activities were perfonned by ~

Freese and Nichols and the finn of ' Mason-Johnston and Associates. Quality assurance was provided by Brown and Root site quality assurance group and the Texas Utilities Services, Inc., (TUSI) site QA surveillance group. Documents

. examined were as follows:

- Freese and Nichols drawing numbers FN-SSI-3 through FN-SSI-7, Safe Shutdown Impoundment Dam

- Freese and Nichols specification FNSSI-1, Contract

. Specification for Safe Shutdown Impoundment Dam

> - Brown and Root Construction Procedure numbers

' 35-1195-CCP-2 through CCP-8

- Brown and Root Quality Control Procedure CP-QCP-7.1, Surveillance of SSI Dam Activities t

9 4

J n,_ .,

, v. _ _ __

=. . ...:. w .

47 The Mason-Johnston and Associates Cor  %

dures .

Review of Quality Records e (b) 6. d The reviewer examined selected records.which'Ifocument c control inspection and quality assurance act'ivities duringA '

Records -

construction of the SSI dam.the reviewer are the procedures examined were as follows:

Records of QA workshops conducted .by ..

conducted to provide training for field inspection. ,.

personnel. ,

Weekly field correctjve action reports for April - July; -

. 1976 and January - March, 1977. ~

Results of quality control tests perfomed on filter -

materials; and impervious core materials placed betwe L April and July 1976.

Atterberg Limits, field density tests, and proctor lts tests W..

perfomed on the imprevious core materials, relative density and mechanical and resu Q <

t- y.

of field density, analysis tests performed on the Type A a .

materials, C; 0$E

- Stop work orders .%

Brown and Root GA Audit Reports -

Training records of QC inspection personnel i

Design Change / Design Deviation request numbers b FN-82 and FN-84 ht ;i Based on review of the records, the reviewer irements concluded d as i the dam was constructed in accordance wit e l-stipulated in the FSAR.

complete, and retrievable.  ;

Unit 1 Reactor Building Internal Pipe Whip Restraints f (2)

Review of Quality Control and Construction Procedures i (a)

The reviewer examined specifications, drawings, h

and control procedures for construction and inspection of t Acceptance pipe whip restraints in the reactor building.

- _ --. 7 ,

.m . s,: . ,

~ _ _ . .

,x ,

l J

.4 .

4g l l criteria utilized by the reviewer appear in Section 3.8 of the FSAR. The pipe whip restraints are non-ASME since they are not attached to the piping. The restraints are treated as part of the reactor building internal structure and are constructed in accordance with the American Institute of Steel Construction (AISC) Standard Practices, as is all other non-ASME structural steel members (cable tray supports, structural steel building: frames, stairwells, non-ASME equipment supports) in the power block. This is standard industry practice. Tha whip restraints were fabricated by the Chicago Bridge and, Iron (CB&I) Company. Onsite installa-tion was performed by Brown'and Rcot. Documents examined by 4

the reviewer were as follows:

- Gibbs 'and Hill Specification 2323-SS-16B, Structural b Steel (Category I) ,--

t

• i: ~ .. - Gibbs and Hill Drawing numbers 2323-51-0581, 0081-01, #

7, C584, and 0585, Reactor Building Internal Structure,'

Pipe Whip Restraints

~

- T1JGC0 Instruction Number QI-0P-11.14-1, Inspection of

. Site Fabrication and Installation of Structural and Miscellaneous Steel

}

The reviewer also examined the outstanding (unincorporated)

design changes against the above specification and drawings.

Tnere were 29 DCAs against the specification,12 against drawing number 0581, 3 against drawing number 0581-01, 11 i

against drawing number 0584, and 11 against drawing 0585.

The reviewer examined the document packages maintained in DCC Satellite 306 for the above specification and drawings and -

verified that they were complete and contained the latest (current) revisions of the drawing and design changes.

~

- (b) FieldInspectionofWhipRestraints o= The reviewer, accompanied by a QC inspector, examined pipe whip restraint numbers M-22 and M-25 which are located in steam generator compartment numbers 4 and 1, respectively, on i

L elevation 900 of the reactor building. Acceptance criteria etilized by the reviewer are these documents listed above.

Examination of these and other restraints on the 900 eleva-4 tion, and discussions with the QC inspector and design engineers, disclosed the following problem. DCA number i 14,813, Rev. 2, against drawing number 2323-51-0581 revises the erection notes for the whip restraints to require installation of jam nuts (or spoiling of threads) on bolts which have nuts installed hand tight for holes noted on the drawings. Discussions with various design engineers and the o

I e

i

- - .. _ ,gy  ;. g .

g, -

4 N.

i 49 l

inspector disclosed that there was some confusion as to where l 4 the use of jam nuts was required. In addition, the reviewer l

observed several locations where jam nuts had not been )

, installed on anchor bolts where nuts had only been installed hand tight. This item will be turned over to the Comanche Peak Project Director for folle'wup.

, l (c) Review of Quality. Records -

The reviewer examined quality records documenting construc-tion (site erection) and QC inspection of whip restraint numbers M22, M25, and M-37 on elevation 900 of the Unit 1 reactor building. These records included weld travelers, QC inspection of structural steel bolting, QC inspection of welding, and as-built drawings showing as-built dimensions, F elevation and location for the restraints. The reviewer -

,y noted that inspections for installation of jam nuts required.,

9:- per DCA 14813, R2, was not documented in the inspection .

packages. Tnere was no resolution of this item during the -

review, therefore, this item will be refered to the Comanche Peak Project Director for followup and resolution. The reviewer did not examine the CB&I whip restraint fabrication

~

records.

1 (3) Review of Nonconformance (NCR) 10453 The reviewer exa' mined NCR 10453 which was written to document and disposition a problem which developed during field erection of four moment limiting component supports on the feedwater lines in the Unit 1 Safeguards Building. The supports, which are ASPE -

components, are similar to pipe whip restraints. The purpose of the supports, which were ere:ted' around the feedwater lines, is to limit movement of the pipes during pipe break accidents. The

- restraints are constructed from heavy beams and columns which were fabricated offsite by CB&I. During fiel'd erection of the restraints (which was accomplished by Brown and Root) cracks

developed in welds which attached small (6 inch by 9 inch) gussett plates to the colt.mns and beant when the bolts in the beam-column connections were torqued.

The reviewer examined the NCR and discussed the corrective action

! with QC inspection personnel. Review of the NCR disclosed that it had been revised five times. Some of these revisions resulted from changes to the corrective action aft'er further evaluation of l the problem.. Other revisions were as a result of changes to the administrative handling of the NCR, e.g., to repair all four 4- rer,traints under one NCR is lieu of writing a separate NCR for each restraint. These types of revisions are nonnal during disposition of NCRs. Review of the NCR and discussions with responsible inspectors disclosed that the problem was resolved by

removal of the damaged gusset plates (i.e., the plates where welds 1

,, ;m -. -. . . .. .. m ~ r- -

a .

50

. had cracked) from the beam and columns, non-destructive examina-tion (NDE) of the base metal in the beams and columns at the points where the gusset plates had been attached, fabrication of j new gusset plates, and rewelding of the new gusset plates to the

beam columns. The reviewer examined selected quality records associated with repairs of one of the restraints, including weld travelers, PT inspection report number 19059 and 19054 and design documents including CMC 96060 and Brown and Root drawing number MSB-0683-CBI. The corrective action to rasolve this NCR was completed in March 1984.

(4) Unit 1 Cable Tray Supports (a) Review of Quality Control and Construction Procedures F The reviewer examined specifications, drawings, and quality -

1: control procedures for construction and inspection of cabla .

Documents. examined by the reviewer were as !

T: tray supports.

follows:

- G8H Drawing Number 2323-El-0713-01-S, Cable Tray Support c Plan, EL 792'-0"& 790-6", Aux & Elect. Control Bldgs.

- G&H Drawing numbers 2323-5-0901, 0902, and 0903, Cable Tray Support Details, Sheets 1-3 4

- G&H Specification number 2323-SS-16B, Structural Steel j (Category I)

- Brown and Root drawing number FSE-00185, Sheets 1-3, Reference Drawing for Cable Tray Hangers

- Brown and Root drawing number FSE-00159, sheet numbers 527, 537, 557, 2895, 2898, 2904, 2905, 2908, 12580, n 12600, 12608. These are the fabrication drawings for the cable tray hanger supports. The sheet number r corresponds with the hanger number. ,

The reviewer also examined the outstanding (unincorporated) l design changes against the above G&H drawing. There were 344~ CMCs and 19 DCAs against drawing 0713-01-S, 6 CMCs and 9 DCAs against drawing 0901, 4 CMCs and 10 DCAs against  !

t drawing 0902, and 26 CMCs and 29 DCAs against drawing 0903.

The reviewer examined the document packages maintained in DCC Satellite 306 for the above drawings and verified that they 1 were complete and contained the late'st (current) revisions of

- the design changes. During examination of the design changes the reviewer noted that the majority of them were originated ll 1

as a result of minor construction problems. For example, j most of the design changes to drawihg 0713-01-5, which is the I

m .. . z .. a =. .- , ,. x . . - . . _ =

4- .

51

- cable tray support location plan, were as a result of

. interferences encountered during construction and were requested by , construction personnel. These interferences required- relocation of some of the supports shown on this t

drawing. Often the relocated supports were only moved'a few inches.

(b) Field Inspection of Cable Tray Supports d

The reviewer, accompanied by a QC inspector, examined .

randomly selected cable tray supports located on eleva-1 tions 790'-5" and 792"-0" of the electrical control building.

The supports and the acceptance criteria utilized by the reviewer appear in the table below.

'F TABLE -

~

%: ~ Support .- Applicable Support Tyoe Design Change

. . Number

• l 527 B-2(Dwg0901) CMC 8250 j 537 .D-1 W/ Brace -

i (Dwg 0901) 557 A-1(Dwg0901) CMC 94628 DCA 1946 DCA 2687 2895 SP-2 (Dwg 0903) CMC 50474 2898 SP-2 (Dwg 0903) CMC 4521 CMC 2646

2904 SP-2(Dwg0903) CMC 52473, R2 DCA 3494 i

I 2905 SP-2 (Dwg 0903) DCA 6299-R7 CMC 2646

! 2908 B-2 (Dwg 0903) -

12580 B (Dwg 0601-015) CMC 61731 i

A (Dwg 0500-04-5) 12600 CMC 67033 12608 SP-7 (Dwg 0903) CMC 68393 CMC 1969 DCA 19973

• Support number and location shown on B&R drawing number j FSE-00185

-M _ eeeem -- _ _ .

pp .e _

w .

52 i

During the; field inspection, the reviewer verified the following w.ere in 'accordance with requirements specified or

': method of attachment to wall and/or design ceiling, drawings: dimensions, elevation of support, proper size of structura l..' steel members, joint connection details, and ,'

configuration of support.

The reviewer 'als'oTwalked down other areas in the auxiliary  ?

and electrical control building and examir.ed cable tray

~

supports for . general configuration and quality of workman-ship. During ' examination of supports in the Unit I cable spreading' room, 'the reviewer noted that six and eight Theinch c siderails had .been added to four inch deep trays.

' practice of 3ncreasing the height of siderails on cable trays -

and its effect 'on the design of cable tray supports was -

examined by. the reviewer. Details of this review are '

discussed in paragraph G.b.(7).c below.

. ~ ' -

(c) Review of Quality Records -

The reviewer examined quality records documenting construc-tion and QC inspection of the cable tray supports listed in the paragraph above. These records included construction travelers, weld filler material logs, and cable tray  !

inspection reports for installation of cable tray hangers, cable tray clamps, and installation of expansion anchors or Based on review of the records, and the Richmond Inserts. ,

walkdown inspection discussed above, the reviewer concluded .'

~ that the cable tray supports were constructed and inspected in accordance with the requirements of the construction .

drawings. The records were neat, legible, complete, and retrievable.

(5) Inspection and Testing of Richmond Inserts (a)

Review of Program for Verification of Installation of Richmond Insert Bolts During review of records, the licensee detennined th In order to verify that tions of Richmond Insert bolts.

bolts of the proper length were installed in the' Richmond Insert sleeves, the licensee carried out a reinspection The reviewer examined program for the Richmond Insert bolts.

TUGC0 procedure number QI-QP-11.14-8, Verification of Installation of Pichmond Insert Bolts, which During thewas used to reinspection control .the reinspection program.

a 53 l

1 program, QC inspectors verified the length of the bolts l either through ultrasonic testing or physical measurement, and checked bolt diameter, minimum embedment length, 'and

" snug tight" condition of the bolts. The reviewer discussed the reinspection program with mechanical QC inspectors responsible for its implementation in the electrical control building. Based on review of the procedures and discussions with the QC personnel, the reviewer concluded that the reinspection program to verify installation of the Richmond Insert bolts was comprehensive.

(b) Observation of Testing of Richmond Inserts The licensee is perforring extensive onsite testing of the Richmond Inserts to ce;1 firm the strength values used in y ' design of . structures using this type of anchorage. The _

reviewer examined -TUGC0 Engineering Instruction number -

T- -

CP-EI-13.0-13 which specifies the method of installation of ';

ttst specimens, and describes the test apparatus and -

_ specifies the technique used in application of the test -

1 loads. The reviewer examined the testing apparatus and verified that the test equipment had current calibration stickers. The reviewer observed the tension test of specimen 28, a 1 inch EC-2W Richmond Insert, and the shear-tension test of specimen 6, a li inch EC-6W Richmond Insert. During the tests, the reviewer verified that application of the test load was accomplished in accordance with the procedure requirements and that the test data was accurately recorded.

Following completion of the' above tests, the reviewer examined the results of tension and shear-tension tests that

( had been previously completed and noted that those results -

were. consistent with the results of the tests witnessed by the reviewer. The majority of the modes of failure resulted i in failure of t:1e high strength bolts, not the concrete or insert sleeve. The reviewer also examined the concrete cylinder unconfined compressive test data to verify the strength of the concrete was recorded for use in evaluation

, of the test results.

(6) Program for Application of Protective Ccatings in the Unit 1 Containment Building (a) Review of Specification and Quality Control Inspection Procedures The reviewer examined specifications and quality control procedures for application and inspection of Service Level I l protective coatings, for steel structures, including the polar crane and liner plate, inside the Unit I reactor

building. Acceptance criteria utilized by the reviewer I

i A-- _ _. _

-. . -. ~

t 54 and FSAR Section appear 3.8.1.6.5.g.in ANSI Standard N101.2-1972 Procedures ex l i

G&H Specification 2323-AS-31, Protective Coating, ,

Inspection of  ;

TUGC0 procedure number CP-QP-11.4, ,

Protective Coatings f.

TUGC0 Procedure number These QI-QP-11.4-1,11.4-5,11.4-17, procedures cover 11.4-22, 11.4-26, and 11.4-28. application inspection of the of storage a materials, surface preparation, primer and finish coa r repairs.

These _  !'

- TUGC0 Procedure Number QI-QP-11.4-23'and 11.4-2

~

procedures cover . reinspection and testing t

ObservationofProtectiveCoatingshorkActivities +*

(b)

The reviewer witnessed application and inspection of  %

protective coatings on steel structure inside the Un f f reactor building. E protective coating application The reviewer verified work in prog j 4-tion for application of coatings.

conditions were being monitored and were 6

environmental acceptable in the reactor building at time of application lo coating.

The reviewer observed that application of the p coatings and QC inspection of the coating  :

(7)

Onsite Civil Design Activities (a) General Onsite civil design activities are perfonned by Gibbs and Hill (G&H) civil-structural engineers who work under the .

direction of the G8H lead civil-structural The engineer onsite who reports to the TUGC0 Nuclear Engineering Manager.

G&H engineers have access to the FSAR, codes, standa designThe criteria, and copies of the original design calcula-bulk of the design work presently being performed tions.

onsite relate to review and approval of design changes (C Many of the design changes are originated at the and DCAs).

request of construction personnel and involve minor chan usually due to construction interferences.

p

- g 7 ear

-- 4.w ww esyme _

c *tr= -

,yy.

. ~ ...

55 (b) Review of the Desigrt Change Program 2 The reviewer examined G&H Project Guide-29, This procedure establishesSite the Review of CMCs, DCAs and S-0910s. {

7 guidelines performed.

under which onsite design change reviews b a ANSI N45.2.11 and NRC requirements (Criteria III to ,

Appendix B, 10 CFR 50). -

The reviewer discussed the design change program with license 2 These discussions disclosed that when a recuest engineers.

I'

~

for design change is made by construction craft or QC -

personnel, the design change is prepared by civil pr engineer.

the civil project engineer usually perfonns some preliminary calculations in order to arrive at a feasible and workableAfte

~

solution to it the problem.is transmitted to the G&H onsite design '

J prepared, Construction personnel C

engineers and to construction. That is, if the G&H [:.

implement the design change "at risk."

^ design engineers do not approve the design change, a removal notice is issued and the work affected by the design charge is either removed or reworked Discussions in order to comply with the f[

with licensee approved design change request. l engineers disclosed that approximately 99 percent of the design changes are approved by the G&H design engineer without revisions and therefore, do not After require rework receiving theafter ,,

1 they are implemented by construction. design change r  ;

review. Approval of the design changes consists of a detailed review by an engineer, followed by an independent If the

,. review l.; another engineer serving as a checker.

design change does not neet the requirements of the designAfter criteria, it is revised es necessary.and approved, the desig requirements.

' The reviewer examined rardomly selected design changes which had been made to drawing number 2323-El-0713-01-S, Cable Tray These included two which were currently being Support Plan.

reviewed by the G&H design engineers, (CMC 8229, R12 and CM 8235. R3), several which had recently been reviewed and approved by the G&H design engir.eers, and several others which had been reviewed by G8H engineer since 1979, the last date drawing 0713-01-5 had been revised.

Based on this limited review of the design change control program implemented at the site, the reviewer concluded that design changes are being properly revie requirements.

v-_ _

s .

56 ,

(c) Review of Cable Tray Loading As discussed in paragraph G.b.(4) above, the reviewer noted during field walkdown inspections that siderails had been raised on some cable trays in order to accommodate additional electrical cables. The reviewer also noted that fire barrier materials, commonly known as.thermolag, were being added to the cable trays (electrical raceways). The reviewer examined the design controls used to verify the structural adequacy of the cable trays from the increase in loadings due to the addition of thermolag and/or addition of cables to the trays.

Details of the review are discussed below.

- Evaluation of Effect of Thermolag Fire Barriers on Structural Adequacy of Cable Trays / Supports F The reviewer examined TUGC0 engineering procedure -

Y- -

CP-EI-4.0-49, $ valuation of Thermolag (TSI) Fire Barrier ' }-

This procedure.

Material on Class 1E Electrical Raceways.

outlines the program to be implemented to verify that cable trays and supports meet seismic design criteria after installation of the thermolag is completed. The program will verify that the combination of the weight of the cables in the trays, the dead weight of the trays, and the weight of the thermolag will not exceed the maximum design allowable load of 35 psf. The procedure outlines steps to be followed when the allowable design load is exceeded. The reviewer discussed this program with licensee engineers who stated that the "as-bu.'iding" of the cable trays to account for the ~

installation of the themolag will begin. in the near future.

After the as-building program is completed, the evaluation of the effect of additional weight of the thermolag on the cable ,

trays will be perfomed per precedure CP-EI-4.0-49 require-ments. This area is being referred to the Comanche Peak 4 Project Director for followup.

- Evaluation of Increases to Height of Cable Tray Side Rails

• 1 During the field walkdown dfscussed above, the reviewer randomly selected for review three four-inch cable trays in the Unit I cable spreading room which had 6 or 8 inch side rails. These were tray numbers T-13-0CC-F7, T-13-GCC-M10, and T-13-GCC-M33. The above trays ; ; 30 inches wide. The reviewer examined sheets 1 'and 12 of drawing number 2323-El-0712, and the 133 DCAs against sheet 1 and 4 DCAs against sheet 12. These drawings

. detail the layout ar.d size / type of the above cable.

- trays. The resiewer also examined the document packages i

- y h~ ... , ,

57 b ve drawings ,.

9  : <

maintained and verified thatinthey DCCwereSatellite complete an 307From for the a od(- cont 3' latest (current) revisions of the thedesign reviewer changes.

h side rails to review of the design change by DCAs. docuaents, For sfg;p>

verifi to cable tray igl$e, the 4 inch deep trays was authorized In

. example, the addition of 8 inch side rails N ;'s

n

- ,, 13-0CC-Q07 was authorized by DCA 15207. the side W j)

~ '

rt design W The reviewer discussed the effect that  ?

%hra U  ?. M load of 35 psf with project civilThese l tended above discussions disclo $l' engineers.

rail depths were increased becauseresult cab eofexTh.is often .occu W[

~? :

r '

the side rails of the 4" deep trays.at intersecti T cable pulling problems. l

'/

,tM MJ it is below  :-

whenever the height The cable of siderails y Schedule, is increaslo MS na the design allowable of 35 psf. [g tray is documented in the G&H Cable i ht ofRacewaVarious "~s4 2323-E-1-1700. J. :4.

identity ofThe each cable in each tray raceway schedule expresses capacity o and the we g -

each cable. Review of the schedules the trays as percent filled. disclosed the data sho Number of Cables-

.P_ercent Filled- W->

c Tray Number ~ _

28 i 198 '

T-13-0CC-Q07. 31 a 288 T-13-GCC-M10 28 $w 217 40 T-13-GCC-M33 bles in tray deter- k6,:

From review of the cable schedule, the reviewer f mined that the average weight of the ca linear 1 T-13-0CC-Q07 was approximately Therefore the cable load in this tray is 0.11 pounds per i foot. d/ft = -

(number of cables)(w +/ cable)_ 2.5 Ft

= (198)(.11) pou N li

=

8.8 PSF

• Width of tray l

This is well below the design allowable i ssions load va ue. '

ncluded that Based on review of with responsible engineers, the reviewer the above l schedules covalu

' tive.

the design adequacy of cable tray supports are conserva ,

-9ee weewee ***** g=*eww=w;+peensw y

- Nd**M* w ~ '~ ~ ~ ~ - _ . _ _ _

{kt v x::74

--a: c.m gmllffy -;;

i

58 4

(8) Personnel Interviews -

The reviewer conducted informal Subjects interviews covered with during the nine civil arid six interviews mechanical QC inspectors. .

were the inspector training program, ability to discuss their  ;;.  :

safety concerns with their management and/or the NRC, cooperation K@

between craft and QC personnel, and availability of technical assistance from engineering personnel. From the interviews, reviewer concluded that the QC inspectors felt freedom to express the I;W i

h@

their. safety concerns to management and/or the NRC, that the J'.

inspectors felt that . craft personnel were aware of the require- .

ments to do the work properly, and that the craft recognized the pi importance of QC inspection activities and cooperated with the F2 inspectors. The inspectors stated that engineering assistance in C~J resolution of problems was available whenever they requested it. .

^

The interviews also disclosed that the licensee has an extensive _.

training program which the. inspectors are required to complete , .

prior to becoming certified and being able to inspect and accept '

work. The training program involves classroom training, on the m.'

U~e job training, and passing written and practical exams (the exams M contain essay type questions, not multiple choice). The training '

4 program for the inspectors performing inspection of structural Mht steel protective coating involved 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of classroom training and 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> of on the job trtining. The inspectors did state  %

?".1

• that the large number of unicorporated design changes against some -

drawings made their jobs more difficult at times, but most said l that after working in an area for a period of time they became familiar with the changes and were able to overcome this problem.

c. Conclusions (1) The licensee has effectively implemented the QA program require- M ments in the areas examined by the reviewer.

(2) QC inspectors are knowledgeable of their inspection requirements and perfonn their inspection in accordance with the licensee's QC

' [h@y D

procedures. Vpf (3) The licensee's QC inspector training program is comprehensive, hy tfi tu Though (4) The licensee's present document control system is good. L4 I

the number of unincorporated design changes against some drawings is large, the availsbility cf a package containing a complete set k((4 of the documents made review of the documents possible without too ,

W. [

much diffeulty to an experienced inspector. The licensee's new unique DCC system (use of computers,) exceeds NRC requirements in 4:$l E.,

l

'the area. Le .

i ' l%2

'y. !

a j .

9 rzy.

' N *.

i h-

- = gg

59 (5)' The quality records examined by the reviewer were neat, legible,

retrievable, and complete.

I (6) One negative point noted by the reviewer is the larger number of l

unincorporated design changes against some drawings. This results L in a cumbersome package to be reviewed when perfonning work or  ;

! inspections. This item allows opportunity for errors and requires l l additional time to be consumed for work to prevent these errors. 1 j The reviewer did not identify any hardware problems resulting from l

the licensee's system,.except for the item identified in paragraph <

1 G.b.(2) above.

I (7) The design change process is controlled and co'mplies with NRC

) requirements. The "at-risk" design change process described in

! paragraph G.b.(7) above is not unique since it has been used on f other nuclear construction projects. The design change program is-j; .

laid out, but could allow.for implementation problems if not k -

i :- meticulously followed. _

(

1 H. Review of Heating Ventilation and Air Conditioning Systems (HVAC) l

References:

Drawings, standards, and specifications applicable to this equipment are as follows:

f Hanger Dwg. SG-790-2J-1R, Rev. O Hanger Dwg. SG-790-2J-1V, Rev. O Hanger Dwg. SG-790-2J-RIB, Rev. O Hanger Dwg. SG-790-1J-R1L, Rev. 1 Hanger Dwg. SG-790-1J-10C, Rev. 0 ,

, Hanger Dwg. SG-790-1H-RIG, Rev. 0 Hanger Dwg. RB-832-1E-1A, Rev. 0 .

Hanger Dwg. RB-832-1E-1L, Rev. 0 .

• Dwg. 2323-M2-0651-HAN, Rev. 2 '

Dwg. 2323-M2-0651-HBSC, Rev. 1 Dwg. 2323-M1-0651-HAN, Rev. 6 Dwg. 2323-M1-0651-BSC, Rev. 6 l l Dwg. 2323-M1-0551-BSC, Rev. 10

• i l Dwg. 2323-M1-0551-HAN, Rev. 9 I Dwg. 2323-M1-0554-BSC, Rev. 12 Dwg. 2323-M1-0554-HAN, Rev. 7 .

Dwg. FCUS-0010-HAN, Rev. 5

, Dwg. 2323-51-0600, Rev. 17 L Dwg. MC-134-680C Dwg. MC-143-689C .

Dwg. DCA 3262, Rev. 1 i

Dwg. ANS D1.1 i Specification 2323-MS-85, Rev. 3

Procedure WP-TUSI-001, Rev. O *

! Procedure DFP-TUSI-003, Rev. 8 l

60 l

a. General The reviewer conducted tours of containment, auxiliary buildirig, safeguards building, and control building for both t; nits to generally observe quality, work in progress, material control, and protection of HVAC equipment, as well as weld rod control. Discussions were held with craft and inspection personnel during these tours relative to plant quality.

b. Review Effort Previous discrepancies identified by NRC regarding HVAC installation served as a driving force for this review effort. A review was made of .

evaluations and calculations performed as a result of the previously identified problems. In addition, the reviewer observed HVAC ducting p- and supports for confonnance to applicable drawings, specifications, __

y and standards. , , ,

c. i The reviewer generally observed ducting in various areas of the -

containments, auxiliary building, safeguards buildings, and control building for both units for proper bolting, proper gaskets, and structural integrity. In addition, the inspector observed duct and equipment supports for confonnance to requirements. Supports reviewed included unit 2 duct hangers 2J-1R, 2J-IV, and 2J-RIB; Unit 1 duct

hangers 1J-RIL, IJ-10C, IE-1A, IE-1L, and 1H-RIG; floor mount of Unit 1

! Train A Containment Spray Pump Room fan coil unit; and the two unit 1 j Safety Injection Pump Room Fan Coil unit hangers. ,

c. Conclusion No significant problems were identified relative to ducting. Only .

minor problems, well within previous discrepancies evaluated, were found in duct supports. Dimensional variations were noted in the ,

hangers for Safety Ir.jection Pump Room Coolers. These deviations were analysed during the review indicating that these hangers were accept-able. Several minor drawing errors were also notad which were corrected during the review. The evaluations and corrective actions perfonned as a result of previously identified problems with HVAC installation appear to be adequate.

I. Fonnal Interviews of QA/QC Personnel ,

a. Fonnal interviews were conducted of QA/QC personnel in order to assist I, in assessing . site quality and management support of site quality. It

! was felt that discussions with inspection personnel would give a good conservative insight into whether or not the plant was being const-ructed properly. Interviews of five management personnel and twenty-eight inspectors were conducted. Inspectors were selected at random with one exception. Electrical inspectors were primarily selected from a group of inspectors which had recently been involved in a personnel

k' f ..  :.

l 7

. V 61 '

1 -

incident involving a dress code "(Tee Shirt)" issue in order to assess whether these persons had significant technical concerns. In addition, i two electrical inspectors indicated a desire to talk to NRC and were

? interviewed. Several additional electrical inspectors were chosen in addition to inspectors in various other disciplines.

The group included inspectors working for eight different supervisors.

Experience of these personnel ranged from persons who had been in QC less than a year, to persons who had been at Comanche Peak from early construction (mid 1970s). Most had some previous experience such as site craft, non-nuclear industry or military experience. Some had worked at other nuclear facilities.

The major thrust of the interviews was to detennine if the personnel had any plant safety or quality concerns. Concerns in these areas were i p- solicited from all those interviewed. Discussions of other subjects .,_ -

were also held with most of the individuals interviewed. These t .;

1l ? ~

subjects included intimidation, support for identifying problems, ability to have problems evaluated and corrected as necessary, feedback'

! on evaluation of problems, adequacy of training program, and relation-

ship with NRC.

! All but two inspectors stated they felt the plant would be safe which

meant they had no significant quality problems which they felt would compromise safe operation. One inspector, who was not sure of the-

plant's safety, stated he was assigned to an' area which was less controlled than he was used to, e.g., non-ASME code work versus ASME -

code work (which has the most stringent requirements), and was uncomfortable with the leeway allowed in this area. This person also indicated he had doubts about QA at nuclear plants in general. The othe: individual who was unsure of plant safety indicate he was satisfied with quality with one exception. Thi.s involved a specific problem which he was not sure was adequately evaluated. This item was described to the NRC:RIV Senior Resident Inspector for followup. Two inspectors who stated they had decided on their own that they wanted to talk to NRC, expressed very strongly that the plant quality was

" excellent" and there was no plant safety concern. Another inspector,

with over twenty-years' experience, who was at his fifth nuclear plant ,

said Comanche Peak was the "best" plant he had seen.

Seven inspectors expressed one or more specific concerns. These concerns involved cuestions on whether a particular procedure require-

ment or whether a particular technical evaluation was appropriate, documentation problems not involving quality of construction., questions whether certain personnel transfers were discriminatory, inaccuracies in some written Nonconformance Report (NCR) evaluations, and concerns

", which had recently been brought up and were yet to be evaluated by the i licensee. All concerns have been forwarded to the Comanche Peak j Project Director for followup for review and evaluation as necessary.

Several concerns were given to NRC

RIV personnel during this inspection j and followup showed that there was no technical problem identified.

1

t . .-

P

62 The NRC Residait Inspector was familiar with one of the concerns and bad already evaluated the conditlon as technically acceptable. Several additional concerns were given to RIV personnel verbally en the last l day of this insper. tion for timely followup.

The special team interviewer reviewed the concern regarding transfers of six of seven individuals mentioned in the personnel transfer concerns. .Ihese transfers appeared to be non-discriminatory. It should be noted that in all cases of concerns involving specific hardware discrepancies these discrepancies had been identified to appropriate licensee personnel and had been or were being evaluated.

All inspectors questioned (21) as to their ability to identify problems

such as via NCRs, indicated no suppression in this area. Several inspectors indicated that NCR written evaluations could be more clear p and complete in some cases. _

Feedback regarding problems, su6 as via explanations of NCR evalua * '; -

i~. tions, was considered good by 19 of the individuals questioned. One -

I- individual indicated he did not always receive complete feedback but I these items did not involve significant t'echnical concerns. Two

! individuals stated they felt uncomfortable with some "use-as-is" NCR i evaluations. One stated that more feedback was needed as to reasons l

for procedure changes.

I i Many of the inspectors indicated that comunications were improving and i the assignment of the new site QA manager was a positive step in

improving comunications. It was clear that some.comunications

problems had existed in the past and rapport between inspectors and -

l their management had been strained previously in some areas. Comuni-i cations in the ASME code construction area appeared to be exceptionally i

positive.

i -

! All but a few inspectors were questioned regarding intimidation by

! craft. No significant problems were identified although two indivi-

! duals mentioned two incidents when the craft were upset with inspectors

! when problems wars found. No threats were made during these incidents.

Generallye the rapport between craft and inspection appeared to be very t

gcod.

Adequacy of the training program was discussed with approximately half i of the inspectors. Several indicated that the fomal training could be

! better, i.e., tougher (not necessarily more extensive) but femal l training, plus on-the-job training was adequate to perfom the j inspection functions. Many stated that the training was excellent.

Twenty inspectors felt no hindrance at all to talking with NRC and L indicated that the freedom to talk with NRC has been continually-stressed by management. Several indicated some apprehension about talking with NRC which appeared to be a natural fear of the position '

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

63 NRC holds. Several were under the impression for a short while that they must have their "act together" if they were going to see the NRC,

but now appear to feel no hindrance. Most indicated they saw NRC inspectors regularly in the field but a majority indicated that they had not talked directly with NRC in the field.

Interviews of management indicated they were very supportive of inspectors and sensitive to inspector concerns. There appeared to be a strong encouragement for personnel to come forward with any concerns, as evidenced by a memorandum dated March 22, 1984, to all QA/QC personnel from the Site QA Manager. Postings indicating management support for inspectors and other personnel in identifying problems were

. prominent 1'y displayed along with NRC Form 3, NRC Information Notice 84-07 and 10 CFR 21 information.

(- In senmary, although some concerns were expressed requiring further --

3; review, these concerns did not appear to be excessive in number or t . _

5 . serious and would be normally expeqted during the interview process. ;

Generally, the most experienced inspectors had a high confidence in the quality of the plant. Past problems in communication and some past apprehension about management support had existed but there seems to have been a marked improvement in this area. No one indicated that past communication problems had caused them to not perform inspections properly or not to identify problems when found. Inspector freedom to identify problems and freedom to talk with NRC has apparently been strongly stressed. Management appeared t'o be sensitive to employee concerns and appeared to be seriously evaluating existing concerns.

b. In addition to formal interviews, numerous informal discussions were l

held between the NRC team personnel and site managers, craft, inspec-( tors, engineers, and office personnel as indicated previously in other i sections of this report. The comments received from these individuals l

were consistent with those received during the formal interviews.

' These discussions covered topics such as plant quality, training, l management support, and document control. ,

Appendix A, which follows, is a sanitized listing of concerns raised by individuals during the interview process. The concerns are only those which will require followup by the Comanche Peak Project Director.

The interviews were sanitized only so far as confidentiality is related. ,

l l

l I

a,...--._.. . . . . _ . ..m. _,.. ,..

7_, ., _ . . . . . . _ _ _ . . . .

64 APPENDIX A Inspector Name: A-1

] Date Interviewed:

General

Background:

Interviewee Coninents:

- Uncomfortable with less structured program for non-ASME versus

ASME; e.g., seem to change dwg. when structure doesn't meet original, can add welds in field and he doesn't think it gets i incorporated into dwg., QC lead can approve changes to travelers

' for non-ASME structures, not much QA involvement in this area.

- Specific: Procedure QIQP 1114-12, electrical mounting backfit,

!T. - craft complair.ed so procedure was revised to reduce number of ' }

inspections, 4 revisions made to delete requirements (bolt tight-ening,<etc.) _

- Has the impression that QA has been generally deficient at nuclear plants and QC has not been supported at Co,nanche Peak in the past.

- Indicated ma'in problem is probably him being able to adjust to non-ASME work: is not aware of code violations taking place.

i i

1 .

4 8

4 s

a

'l Q

f .

l.

e

.e,, - ,,-- - , --- -- ~v - - - - . , _ - - , , - - - - - , -w

65 Inspector Name: A-2 Date Interviewed:

G0neral

Background:

Interviewee Comments:

- Has some concern with use-as-is NCR situations, use-as-is seems particularly prevalent when using Specification ES-100.

~

- Specific Technical Concern: NCR was written when cable damage occurred during Biso Seal removal using a threaded rod. This occurred in Auxiliary Building, elev. 832'. NCR said no damage p was done to cable but some insulation had been scraped off by rod. Feel further evaluation may be in order for these cables and- '"

( there may be similar problems elsewhere.

- Specific: Wrote 2 NCR's regarding traceability of fuse blocks.

Blocks were not marked "Q". NCR said OK as-is because no non-Q blocks were purchased via order MS-605. Feels othee similar -

non-Q blocks have been purchased via different purchase order and could have been installed as Q. Thinks this a possible paperwork problem.

I

- Specific: Wrote recent NCR (not yet evaluated) on GE Motor .

Control Centers. Compression lugs have bends as much as 180 degrees (more than nomally done done by site constriction).

Don't think GE can violate requirements and may be -

a problem elsewhere in GE MCC's. Also have some broken wire strands which we are fixing as we find.

.I

- Specific: Hac previous paperwork conflict problem in solving rework of teminal blo:ks. 6 page RFIC involved and Proc. SAP-6 involved. Wrote 2 NCR's. NRC inspectors Creek and Johnson were aware, Creek told NRC inspector Taylor, Taylor told to have an answer. Never got feedback as to results.

I'

- Specific: Repaired a solenoid, shortly after coming to Comanche Peak in craft, without paperwork. Don't know if it was safety

< related. Not concerned with solenoid technically - did a

? good job.

f Notes: The specific concerns were given verbally to the SRI - Construction on'4/12/84 for further followup. It was indicated during the interview he would get more specifics for SRI. MCC problem was still being evaluated. I suggest allowing the licensee to evaluate and then followup for adequacy of corrective action. ,

L .- _ _. - _ ___ _ __ _ .___ _ _._._._

l 66

, Inspector Name: A-3 .

Date Interviewed:

General

Background:

a i Interviewee Coments:

- Generally concerned with finding numerous problems during past construction inspection and procedure being changed to delete inspection, e.g., loose terminations found in lighting.

~

- Some NCR's are answered simply that the problem is not addressed in Specification ES-100.

- Recent NCR written because restraint cable (lighting) crimp gages -

T -- #ere worn & therefore, inspection was inadequate. This is sti1T**.

'% being evaluated.- ,'

- Wires of two different gages were terminated at some lugs and many .

terminations are loose. ,

• - Have more pressure not to write NCR's during turnover.

- Found loose LB's (elbow termination fittings) 9 East & South ends of Unit 1 Diesel Generators, wrote two NCR's, was accepted as is.

Found cables not trained (routed) in workmanlike manner in Unit 1 Cable Spread Room 9 junction boxes 1058 and 1059. NCR said OK because cable radius was OK but did not admit workmanship problem.

~

- Feels post construction inspectors were transferred to Unit 2 as .

[ retaliation for finding problems.

J - Heard second hand that .IR's (inspection reperts) were being L

written falsely (without reinspection) to clea. IIRN's (discrep-p Heard from lady in Paper Flow Group L

ancy)

(PFG and report) ladyonincable vault.trays.Said he would get back to NRC with more specifics.

]

!! Notes: Some review of the lighting tennination issue and post check j procedure was conducted by tean member Ruff. The site inspector j indicated he had told of most of these issues and QA was

" , evaluating. I forwarded concern relative to 1058 & 1059 junction boxes to RIV: Martin and he indicated he inspected these boxes and sees no technir.a1 problem. Resident Inspector: Smith partic-3 ipated in most of the interview ard indicated he was aware of the D/G loose fittings and sees no technical problem. I evaluated y reasons why 6 personnel including '

were transferred to Unit 2 and this move does not appear to be discriminatory.

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67 Inspector Name: A-4 1 Date Interviewed:

General

Background:

Interviewee Coninents:

cable Uncomfortable with some use-as-is situations, e.g.,

separation problem found in fuel building during walkdown did not meet procedure but was evaluated as use-as-is. He crn shcw someone where it is.

- Wrote NCR on lack of S-thread engagement on a conduit fitting

- poor evaluation in that they simply said that couldn't see it; a second NCR was written on this area for cable damage, seemed to be-looking for a way to buy this area off, took two tries to get ' }

knows about this but didn't get back c.. ~ '

everytning evaluated.

- to him on fact that NCR's were poorly handled, i.e., non-tech- ~

- nical aspects.

~

- Feels discriminated against in that he was transferred to Unit 2 where there is no ?vertime. Got grilled on cable damage NCR at the same time as being counseled on a personnel issue so it appeared that his transfer had something to do with NCR.

Management is aware of this concern.

Note: I did not review this person's transfer situation.

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68 Inspector Name: A-5 Date Interviawed:

General

Background:

i Interviewee Coments: .

- Had problems with post check, e.g., loose ' lighting tenninations and junction boxes. Took lighting out of procecure and made it more difficult to look at junction boxes. Management was made aware of these concerns. (Hasnosignificantsafetyconcern)

- More tendency toward use-as-is when pressure is on (safety p requirements are being met, however) _

~

T. - Has had some fear of talking with NRC, didn't think reporting ' i on-site would ever get off-site, doesn't have NRC RIV phone number

~

- Fiels discriminated against .by being transferred to Unit 2

- Some NCR evaluations are inaccurate or unclear, e.g., statement that workmanship was not compromised when in fact workmanship was poor but the item was technically acceptable r

Notes: I reviewed the transfer situation; appears to be reasonable but not as clear as reasoning on other 5 transfers. NRC Form 3 appears well posted so I'm not sure why he doesn't have the number. He does not appear to fear talking with NRC now.

Although, he stated he does not have significant safety / quality .

concerns, his coment on NCR answers is interesting. Similar general coments were received from other inspectors and this could indicate a need for better answers on NCR's. An example i would be that if a workmanship question was not addressed properly

. then perhaps needed retraining of personnel as preventive action would not get perfonned. Perhapss the licensee needs to improve in this area.

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OQ N 69 Inspector Nane: A-6 Date Interviewed:

f General

Background:

Interviewee Conments:

- Added higher sides to some cable trays to keep cables in trays

- Also there may be cable density / compaction problem in this area 1

- It's tough to keep people off trays to keep from damaging them

- Have had problems with clearance of pipe and cables, have to notch

' I" insulation, place metal between insulation and trays .-

s;: .

.' ':T

- There is alot of rework to get-proper separation .'- ~

Notes: This man was questioned primarily to get input for RIV review of

~

cable spread rsom as to where there could be problems. He personally has little problem with plant quality. RIV - Martin was at the interview and verbal feedback on the first two items indicated that the situations were acceptable.

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Inspector Name: A-7 ,

Date Interviewed: .

General

Background:

Interviewee Comments: ,

- Had problems with Paper Flow Group (PFG), when first implemented, with completeness of packages. Getting.better and does not know

~

of safety problem involved '..'

- Some inaccurate NCR answers .

N. .

5; _

- Site has problem with lost records, 2 people are assigned full k-

! N:- time in the vault, NCR's are n&t written on lost records, reinspect

, . when record is lost but this reinspection may be very difficult or

very impractical. He has no evidence that reinspections are not -

- getting done. This problem could relate to competance of PFG people, i.e., maybe they lost records.

4 Note: Various special team members looked quite extensively at records.

Results are in the team report. -

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UNITED STATES

[!# "n%,5 s NUCLEAR REGULATORY COMMISSION OFFICE OF PUBLIC AFFAIRS, REGION IV

% ,g .o 611 Ryan Plaza Drive, Suite 1000, Arlington, Texas 76012 t

.1 4 .

RIV: 84-71 FOR I!9tEDIATE. RELEASE

Contact:

Clyde E.'Wisner (Thursday, July 19, 1984)

Telephone: 817/860-8128 (Office) -

g} .

817/571-9907 (Home)

4 4

j STATEMENT OF THOMAS IPPOLITO, COMANCHE PEAK PROGRAM MANAGER, AT NEWS CONFERENCE ON THURSDAY, JULY 19, 1984 o

.y The objective of this news conference is to provide an cpportunity to

~ the news media to obtain information relative to (1) the Comanche Peak project's Technical Review Team, and (2) the Comanche Peak Special Review Team Report.

3 '

. The purpose of the review was to (1) evaluate the current implementation .

,s of the applicant's mr.,nagement control of the construction, inspection and test W programs; (2) provide an indepth understanding and background information to the NRC new management team established by Executive Director for Operations

& t" memorandum of March 12,1984; and (3) obtain information necessary to h establish a management plan for resolution of all outstanding licensing is actions. The team spent over 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> performing this review.

$ The goal of the special team review rias been met in that (1) an q assessment of the applicant's current management ~ control of the construction, y- inspection and test programs has been made; (2) an indepth understanding has been achieved; and (3) information has been obtained to establish a management

, plan for the resolution of all outstanding licensing actions. '

With respect to the assessment of the applicant's management control of

, the construction, inspection and testing programs, the Special Review Team has a'

determined that based on the number and significance of the strengths vs.

weaknesses identified in this review, the applicant's programs are being

'sufficiently controlled to allow continued plant construction while the NRC completes its review and inspection of the facility.

Further, the review provided a sufficient understanding of these programs and their strengths and weaknesses to assist in the development of the

.) " Comanche Peak Plan for the Completion of Outstanding Regulatory Actions."

This plan was approved for implementation on June 5, 1984, t

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TECHNICAL REVIEW TEAM (TRT) EFFORT EDO LETTER DIRECTING MARCH 12, 1984 NRR MANAGE ALL NECESSARY NRC ACTIONS LEADING TO LICENSING DECISION FOR COMANCHE PEAK, APPROVAL OF COMANCHE PEAK JUNE 5, 1984 PLAN FOR THE COMPLETION OF OUTSTANDING REGULATORY ACTIONS - ESTABLISHED '

TECHNICAL REVIEW TEAM REVIEW 0F ALLEGATIONS, TECHNICAL REVIEW TEAM ONSITE JULY 8-20, 1984 (5 SESSIONS) JdLY 29-AUGUST 10, 1984 AUGUST 19-31, 1984 SEPTEMBER 9-21, 1984 SEPTEMBER 30-0CTOBER 12, 1984 MEETING AND LETTER TO TEXAS SEPTEMBER 18, 1984 UTILITIES ON PRELIMINARY RESULTS OF TRT REVIEW IN ELECTRICAL, CIVIL AND TEST ~

PROGRAMS AREAS, f -

V. NOONAN REPLACED T. IPPOLITO OCTOBER 22, 1984 AS DIRECTOR COMANCHE PEAK PROJECT, LETTER TO TEXAS UTILITIES NOVEMBER 29, 1984 ON PRELIMINARY STATUS /RESULTS OF TRT REVIEW IN PROTECTIVE C0ATINGS, MECHANICAL AND MISCELLANEOUS AREA, MEETING A'ND LETTER TO TEXAS (LETTER) JANUARY 8, 1985 ON PRELIMINARY RESULTS 0.F (MEETING) JANUARY 17, 1985 i TRT REVIEW IN QA/QC AREA,

- ISSUANCE OF SSER N0, 7 JANUARY 1985 STAFF EVALUATION OF ALLEGATIONS IN THE AREAS OF ELECTRICAL /

INSTRUMENTATION AND TEST PROGRAMS i

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- ISSUANCE OF SSER N0, 8 FEBRUARY 1985 STAFF EVALUATION OF ALLEGATIONS IN THE AREAS OF CIVIL /

STRUCTURAL AND MISCELLANEOUS

- ISSUANCE OF SSER N0, 9 ' MARCH 1985 STAFF EVALUATION OF -

ALLEGATIONS IN THE AREA 0F PROTECTIVE C0ATINGS

- ISSUANCE OF SSER N0, 10 APRIL 1985 STAFF EVALUATION OF ALLEGATIONS IN THE AREA 0F MECHANICAL /

PIPING

- ISSUANCE OF SSER N0, 11 MAY 1985 -

STAFF EVALUATION OF ALLEGATIONS IN THE AREA 0F QA/QC ,

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f*"'.a 3 s NUCLEAR REGULATORY COMMISSION UNh cD STATES ,

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o OFFICE OF PUBLIC AFFAIRS, REGION IV 611 Ryan Plaza Drive, Suite 1000, Arlington, Texas 76012 p  %.....

RIV: 84-65 FOR IMMEDIATE RELEASE Contac.t: Clyde E. Wisner (Monday, July 9, 1984)

Telephone: 817/860-8128 (Office) 817/571-9907 (Home)

NRC ASSEMBLES TECHNICAL REVIEW TEAM FOR COMANCHE PEAK The Nuclear Regulatory Commission staff has assembled a Technical Review Team (TRT) that will be gathering information at, the Texas Utilities Electric Company's Comanche Peak site for the next few weeks as the construction of the facility nears completion at Glen Rose, Texas. Onsite activities will be-managed by Thomas Ippolito, Project Directt r, under the direction of Darrell Eisenhut, Director, Division of Licensing in Washington, D.C.

A number of outstanding technical matters remain to be completed before the Nuclear Regulatory Commission is ready to make its licensing decision. Included j are: (1) completion of review of TUGCo's Final Safety Analysis Report, (2) i completion of the NRC inspection activities related to construction completion j and preoperational testing, and (3) resolution of allegations received by the NRC regarding improper practices during construction of the Comanche Peak f acility.

Mr. Eisenhut, Mr. Ippo.lito, and John T. Collins, Region IV Administrator, _

Arlington, Texas, were dn Granbury, Texas on July 9 to to kick off the TRT

. effort. The team is co'mposed of staff from NRC headquarters in Bethesda, Maryland, Region IV and other NRC Regional Offices, and NRC consultants. The 2 information developed by this team will be used in making licensing decisions on Comanche Peak. Areas to be examined include the installation and testing of

. equipment as well as the' quality assurance aspects of construction. Of specific concern are allegations received by the NRC staff of improper construction practices. For these, the staff intends to determine the accuracy of the t ellegations, assess their overall importance and implications, and evaluate their safety significance.

The results of this team review will be published in an NRC safety evaluation report. In that document the NRC will describe the areas reviewed and the safety significance of any concern identified, and set forth any necessary corrective action.

Questions concerning these efforts should be directed to either Mr. Eisenhut, 4

Mr. Ippolito, or Mr. Collins. Mr. Eisenhut can be reached at (301) 492-7672 i

and Mr. Ippolito can be reached at the Comanche Peak TRT Office at (817)

, 897-3167, or in Washington, D.C. at (301) 492-7014. Mr. Collins can be reached at (817) 860-8225. -

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