ML20137X540
| ML20137X540 | |
| Person / Time | |
|---|---|
| Site: | South Texas, 05000000 |
| Issue date: | 05/16/1980 |
| From: | Rodgers S HOUSTON LIGHTING & POWER CO. |
| To: | |
| Shared Package | |
| ML20137X502 | List: |
| References | |
| FOIA-85-519 PEP-11-02, PEP-11-2, NUDOCS 8603100153 | |
| Download: ML20137X540 (58) | |
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FEF-11 REPORi!NG DESIGN AND CONSTR': J TION DEFICIEN:!ES TO NRC
SUMMARY
OF REVISIONS REVISION
SUMMARY
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. g," n rvisio= ossen e n ow 0 Initial Issuance 1 Tnis procedure formalizes the process of notifying the NRC of potentially reportable items. Initial notification for site deficiencies is by the Project QA Manager, Initial netification for all other deficiencies is by the Team Leader of Nuclear Safety & Licensing. A site Incident Review Cort-ittee has been estabitshed to handle site deficiencies. This procedure will re:;uire additional revision once a Licensing engineer is assigned to the site.
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REVISIOil AUTHORIZATION plL
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I HOUSTON LIGHTING & PO'n'ER C0"PANY l SOUTH TEXAS PROJECT l PROJECT ENGINEERING PROCEDURE l PEP-11 REPORTING DESIGN AND CONSTRUCTION DEFICIENCIES TO NRC 1.0 PU:. POSE Tne purpose of this pro:edure is to des: ribe the South Texas Dreje:: progra- for co plying with the re:uiremer.ts of both 10 CFR 21 and 10 CFP 50.55(e).
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} This peo:edure establishes the recuirements for reviewing, 2
evaluating and re:orting defects, non-complian:es and I deficien:ies Mich could potentially affe:: the safety fun:tions of the South Texas Proje: as de'ined by 10 CFR 21 or 10 CFR 50.55(e).
j 3.0 REFEP.ENCE DOCUvENTS 10 CFP 21 10 CFR 50.55(e)
' NSC !&E Inspection Manual, Guidan:e - 10 CFR 50.F5(e)
Deficien:y Reporting, 7-1-76.
4.0 RESDONSIPILITY 4.1 Proje:t GA Manager l The Proje:t GA Mana:er is responsible for pe-fer-ing a preliminary analysis of a site identified in:ident to determine if it is possibly repor+9ble.
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l The Proje:t 0A P.anager is responsible for perfeming a preliminary analysis of a site identified in:ident i to detemine if further analysis of the in:ident is recaired by the In:ident Review Ccmittee.
The Project GA Manager is responsible for providing preliminary notice to the Resident Rea: toe Insee:ter of ea:n possibly reportable deficiency identified on site.
The Proje:t GA Manager is responsible for notifying l the Marager, South Texas Project and the NR Region 1
IV office of incidents deternined to be reportable ruasuant to 10 CFR 21 or 10 CFR 50.55 (e).
(' 4.2 Team Leader, Nu: lear Safety and Licensing l The Team Leader, NS&L is rescensible for perfomin; a prei1Mnary analysis of a home office identified j in:ident to detemine if it is possibly reportable.
[ The Team Leader, P5&L is responsible for perfor-ine e e prelininary analysis of a here office identified
) in:ident to detemine if further analysis of the in:ident is recuired by the In:ident Review l Co?ittee.
The Tea
- Leader, NS&L is responsible for previding prelininary notice to the Resident Reacte* InsDe: toe of each possibly reportable deficiency identified by
- the home office.
The Tean Leader, NS&L is responsible for n0tifyin9 the Panager, South Texas Proje:t ar.d the N0* Pegion
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IV office of incidents detemined to be re ortable
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! The Tea : Leader, NS8L or his designee is rescorsible for chairing all In:1 dent Review Comittee neetings. l
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The Tean Leader, NS8L or his designee is responsible l for preparing the ninutes of all In:ident Review Co : .ittee reetings.
The Tea 9 Leader, NS8L is responsible for pre:aring ard sub-itting all written correscendence to the
, hu: lear Regulatory Comission as recuired by 10 CFR j 21 cr 10 CFR 50.55(e).
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The . Team Leader, NSLL is resDonsible for reinteiring
] I a file of all in:idents censidered by the Ir:1de*t i Peview Comittee.
4.3 Manage *, Snuth Texas Proje:t l The Manage *, So:.:th Texas Proje:t is responsible fc*
rotifying the Vice Presid**.t - Ccnstru tien and Te:nnical Services, of all incidents detemieed by the In:1 dent Review Comittee to be re:crtarle under 10 CFR So.55(e) er 10 CFR 21.
4.1 Ir:ident Review Co :-ittee The In:1 d ea.: Peview Co r-ittee is res:Orsible fo*
deteminin; reDertability of de'icien:ies per 10 CFR s 50.55(e) or defects and non:enpliances per 10 CFP 21.
Houston Lichting A Power has the ultimate res:omsibility for detemining reportability O' deficiencies per In CFR 50.55(e).
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5.1 Initial In:ident hetification j Defects, non:enpliances or deficien:1es which could
! potentially affect the safety functions of the j nuclear power plant can be identified by numerous 1 individuals and from various procedures or 1
interfaces. Regardless of the individual.
l organization or means of identifying such incidents, i it is irperative that either the Proje:t CA Managee l l or the Team Leader, NSA'. he notified ir9ediately so i
he can review the f;0R (or ADR, if a ;rocriate) ard initiate the evaluation Process of the rossibly repertable cefe:t, nen enfomance or def t: ten:y.
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1 The Proje:t Oa Panager or Team Leader, P:S$' . shall i
n:tify the Desident Reactor Insre: tor within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> j er re:eict of ea:h incident censidereti to be possibly reportable.
5.2 Safety Evaluation Ea:h incident identified as being rescetable to the f;R: per 10 CFR 21 or 10 CFR Sn.55(e) will receive i
beth a technical evaluation and a safety evaluation
.to detemine w* ether the in:1 dent could, if un:cere:ted, create a substantial safety hazard.
Te:hnical evaluetions are cerfomed by the ecgnizant engineering organization and safety evaluations are perfomed by the Incident Review Crea.ittee.
For the purpose of initial re:ertability deterrina-tion, the safety evaluation may be waived when sa extensive sa'ety evaluation is coesidered ne:essary and the disposition of the in:1 dent will result in the retention of the original design configuration
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. TITLE PROJE CT ENGINEERING PROCEDURE I PAGE C or ?1 sumateT REPORTIN3 DESIGN AND CONSTRUCTION DEFICIENCIES TO NRC oatt issuto et nc ien and safety criteria. In such cases, the incident shall be considered reportable. The final report shall then consist of a technical evaluation. A safety evaluation shall be performed in all cases where the original design configuration or safety criteria will not be retained.
5.3 Deficiency Determination The following factors shall be considered in evaluating an incident to determine if it is a deficiency reportable to' the NRC under the provisiens of 10 CFR 50.55(e).
- 1. The deficiency shall be evaluated as affecting or
( have the potential to affect the safety of the operation of the plant at any time throughout its expected life. This includes deficiencies which could affect the safety-related function of any structure, systen er ennponent for which credit is taken in the FSAR in evaluating their capability to acconnodate the effects of and to be compatible with environnental conditions associated with norr.al operation, maintenance, testing and postulated accidents. Careful consideration shall be given to the effects of the deficiency to assure that it does not indirectly affect the safety of operation of the pl ant.
- 2. The deficiency shall be related to the design or construction phases. This includes activities of the Nuclear Steam System Supplier, 8Pchitect engineers, consultants, contractors, or suppliers.
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- 3. The deficiency shall fall within one of the following four categories:
(a) A breakdown in the Quality Assurance Program related to any criterion in 10 CFR 50, Appendix R, applied to any design or construction activity affecting the safety to plant operation.
(b) A deficiency in final design as approved and
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released for construction such that the design does not comply to the design criteria and bases stated in the FSAR or Construction Permit. Final design denotes
. those drawings, specifications and other I engineering documents that have been reviewed, approved and released for fabrication, installation or construction.
No deficiency exists if a design stated in the FSAR is changed by approved procedures after receiving proper evaluation and review.
(c) A deficiency in construction of or damage to a structure, system or component which will recuire extensive evaluation, extensive repair to establish the adeouscy of the structure, system or component.
(d) A deviation from performance specifications defined in the functional testing require-ments which will reovire extensive evalue-tion, extensive redesign, or extensive re-pair to establish the adequacy of the struc-ture, systen or component.
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- 4. The significance of the deficiency should be evaluated relative to operational safety. The significance of a deficiency is a subjective evaluation and as such the significance of a given deficiency may not be clear. Therefore, if the significance of deficiency is in doubt, it should be treated as reportable to the NRC.
5.4 Defect Detemination The following factors shall be considered in evalue-ting an incident to detemine if it is defect report.
able to the NRC under the provisions of 10 CFR 21.
- 1. The incident identified shall involve either the I.
software or hardware recuirements associated with a basic component. A basic component is a Safety Class 1, 2, or 3 Seismic Category I structure, system, component or part thereof necessary to assure:
(a) The integrity of the reactor coolant pres-sure boundary, (b) The capability to shutdown the reactor and maintain it in a safe shutdown condition, or (c) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to those referred to in 10 CFR 200.11
- 2. The basic component shall contain a defect which is defined by the following categories:
(a) A deviation in a basic component delivered to a purchaser or user where on the basis of
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REPORTING DESIGN AND CONSTRUCTION DEFICIENCIES TO NRC cavr issure 05/16/80 an evaluation, the deviation could create a substantial safety hazard; or (b) The installation, use or operation of a basic camponent containing a defect as defined in paragraph 2 (a) above; or (c) A deviation in a portion of the nuclear power plant subject to the construction permit requirements of 10 CFR 50 provided the deviation could, on the basis of an evaluation, create a substantial safety hazPrd and the portion of the facility containing the deviation has been offered to the purchaser for acceptance; or I
(d) A condition or circunstance involving a basic component that could contribute to the exceeding of a safety limit, as defined in the technical specifications of a license for operation issued pursuant to 10 CFR 50.
- 3. The defective basic component should be evaluated to determine whether the defect could, if uncorrected, create a substantial safety hazard.
4 A commercial grade iten is not part of a basic conponent until after dedication which occurs after receipt when that iten is designated for use as a basic component.
5.5 Noncompliance Determination The following factors should be considered in evaluating an incident to determine if it is a noncompliance reportable to the NPC under the provisions of 10 CFR 21.
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- 1. The incident identified should involve either the software or hardware requirements associated with a basic component.
- 2. The basic component shall fail to comply with the Atomic Energy Act of 1954, as amended, or any applicable rule, regulation, order or license of the NRC relating to substantial safety hazards.
This includes failure to conply with design criteria and bases stated in the FSAR or Construction Pemit.
- 3. The basic consonent containing the noncompliance should he evaluated to detemint whether the noncompliance could, if uncorrected, create a substantial safety hazard.
5.6 Incittent Motifications Five primary notifications occur during the process of reporting defects, noncor pliances and deficiencies in accordance with 10 CFR 21 or 10 CFR 50.55(e).
These notifications are:
- 1. Project OA Manager or the Team Leader, NSAL notifying the Resident Reactor Inspector of an incident detemined to be potentially recortable pursuant to 10 CFR 50.55 (e) or 10 CFR 21.
- 2. Any individual notifying the Project GA Manager or the Team Leader, NS&L of an expected incirtent requiring consideration pursuant to 10 CFR 50.55 (e) or 10 CFR 21.
- 3. The Project GA Manager notifying the Team Leader, huclear Safety & Licensing of an incident l
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requiring evaluation by the Incident Review Committee.
- 4. The Team Leader, Nuclear Safety & Licensing notifying the cognizant engineering discipline and others required to call a meeting of the Incident Review Committee.
- 5. The Manager, South Texas Project notifying the Vice President - Power Plant Construction and Technical Services of an incident determined to be reportable or potentially reportable.
5.7 Written Incident Report
( A written repcrt is required to be submitted to the NRC Director of Regional Office and the NRC Director, Office of Inspection and Enforcement, on each defect, noncompliance or deficiency reported to the NRC per 10 CFR 21 or 10 CFR 50.55(e). An outline for these written reports is provided in Attachment PEP-11-01 to this procedure. If information required for the written report is incomplete when the report is prepared, an interim report shall be prepared describing that information which is available and a schedule for completing the remaining information.
5.8 Inci' dent Review Committee A. The Incident Review Committee for home office identified incidents shall consist of the following members, or their designees, as a minimum:
- 1. Engineering Team Leader, NSSL - Chairman
- 2. Project QA Supervisor
- 3. Supervising Project Engineer, Design Engineering l
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,s B. The Incident Review Comittee for site identified incidents shall consist of the following members, or their designees, as a minimun:
- 1. Engineering Team Leader NS&L - Chaiman
- 2. Project GA Manager
- 3. Supervising Project Engineer, Site Engineering Support
, 5.9 Deadline Recuirements The following deadlines shall be observed.
A. The NRC shall be notified by telephone:
(1) Within 2 days pursuant to 10 CFR 21
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(2) Within F4 hours pursuant to 10 CFR SC.55(e).
, following datemination by the Project GA Manager i s or the Team Leader, NSAL that the incident is l
possibly reportable.
B. The NRC shall be sent a written report:
! (1) Within 5 days pursuant to 10 CFR 21 l ,
(2) Within 30 days pursuant to 10 CFR 50.55(e).
following detemination by the Incident Review Comittee that the incident is reportable or detemination by the Project GA Mananer or Team Leader, NSAL that an item is potentially ,
reportable. Should a potentially reportable item 3 be subsequently detemined to be not reportable ,
' 'o then the NCP Region IV office shall be notified in writing of this disposition. e a
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5.11 Posting Requirements Attachment PEP-11-03 to this procedure provides a standard poster which must be displayed in conspicuous locations at both the Houston hone office -
I and the construction site. It is the responsibility of the STP Team Leader, Nuclear Safety and Licensing to ensure that adequate posters are displayed in 4' appropriate locations. This poster identifies the individual to whom initial notification of a defect or noncompliance which would potentially affect the ,
safety functions of the STP.
5.12 Procedure Resoonsibility Action Any Individual 1. Reports to the Project '
QA Manager any incident which he feels should be evaluated pursuant to 10 CFR 50.55(e) or 10 CFP 21.
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Nuclear Safety & ,
Licensing Project QA Manager 3. Determines whether it or Team Leader, is potentially Nuclear Safety & reportable.
Licensing Project QA Manager 4. If not potentially or Project Quality reportable, then Assurance Supervisor handles incident
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Project QA Manager 5. If potentially report-or Team Leader, able, then the Nuclear Safety Resident Reactor
& Licensing Inspector and the Team Leader, Nuclear Safety & Licensing is notified.
Team Leader, 6. Reviews the incident Nuclear Safety report and discusses
& Licensing it with the Project QA Supervisor.
Team Leader, 7. Arranges for review Nuclear Safety of the incident by the
& Licensing appropriate Engineer- g ing discipline.
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& Licensing incident.
Team Leader, 9. If potentially report-Nuclear Safety able then incident is 8 Licensing referred to the Incident Review Comi ttee.
Incident Review 10. The incident Review I Comittee Comittee considers the incident.
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Encineering Huclear 11. If deened not Safety A Licensing reportable, prepares a written report notifying NRC Region IV.
Incident Review 12. If deemed not Comittee reportable, then the incident is resolved through nomal OA Channels.
Incident Review 13. If deemed report-Comittee able, then the incident is reported to the NRC by the Team Leader, Nuclear Safety A Licensing. ;
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Project OA Manager or 15. Notifies the Manager, Team Leader Nuclear South Texas Project Safety & Licensing and the Manager, Quality Assurance.
Team Leader, Nuclear 16. Notifies the Super-Safety & Licensing vising Engineer, Design and the Supervising Engineer
( Nuclear Safety &
Licensing.
Team Leader Nuclear 17. Prepares a citten Safety & Licensing report to the t?C.
Region IV within the time frame noted by section 5.9.
Team Leader, Nuclear 18. Prepares a documenta-Safety & Licensing tion package to be maintained by the Licensing group and RMS.
6.0 00CW.ENTATION Each defect, noncompliance or deficiency evaluated by the Incident Review Comittee shall be recorded in fornal e
meeting ninutes by the Tean Leader Nuclear Safety &
Licensing. A copy of these netting ninutes copies all telephone correspondence and copies of all NRC
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DEFICIENCIES TO NRC Dave issur0 05/16/80 o,
correpondence will be maintained by the Tean Leader, Nuclear Safety & Licensing; in addition to the permanent record copy to be maintained by R"S. It is the responsiblity of each individual originating such documentation to ensure that file copies are properly identified and transmitted to the Team Leader, NSAL and the Record Management System (RMS) File.
7.0 ATTACH"ENTS PEP-11-01 Incident Report Outline
{ PEP-11-02 Procurement Document Statement PEP-11-03 Posting Notification I
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SUMMARY
This section should provide a synopsis of the incident and its resolution. The following information should be contained in this section:
- a. A synopsis of the incident
- b. Postulated cause of incident I
- c. A synopsis of the corrective action taken
- d. Synopsis of the results of the safety evaluation.
!!. DESCRIPTION OF INCIDEt:T This section should provide a clear and cumplete description of the incident and the circumstances surrounding it.- The following information shnuld be contained in the section:
- a. Identification of the facilit component (including vendor)supplied y or activity or basic to such facility or activity involved in the incident.
- b. Source and extent of incident.
- c. The date and means by which the incident infornation was obtained.
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- d. Unusual circumstances pertaining to the incident, such as weather conditions, test procedures underway and/or abnormal conditions at site.
- e. Status of construction at time of incident
- f. Procedures in effect to avoid incident (if any)
!!!. CDPRECTIVE ACTION This section should provide a clear description of the' corrective action taken to rectify the incident and action taken to prevent recurrence. The following information should be contained in this section as applicable:
( a. The innediate response by Ouality Assurance in response to the incident.
- b. A complete description of all innediate actions taken to correct the incident.
- c. A complete description of all long-range actions te be taken to correct the incident and implementation schedule.
- d. A delineation of the testing nethods w51ch will be utilized to ensure that repairs have been conducted properly.
- e. Action taken to prevent the recurrence of the incident during the remaining construction phase of the facility.
I V. SAFETY ANALYS!$
This section should provide sufficient information to fully analyze and evaluate all the possibly safety
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- a. Nature of the incident and the safety hazard which is created or could be created.
- b. Identification of applicable Quality Assurance dccunentation.
- c. A record on all incidents noted in the Duality Assurance documentation and their resolution where applicable to the reported incident.
- d. A complete record of the incident and the results of all investigations,
- e. Postulated cause of incident.
- f. A comolete description of all innediate actions taken to mitigate the consequences of the incident.
- g. In the case of a basic component which contains a defect or noncompliance originated for evaluation by B&R, the number and location of all such components in use at, supplied for, being suoplied for other f aellities or activities with B&R as the engineer or constructor.
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. wousTow ucciNs a Powta cow'INv SOUTH TEXA5 PROJECT PEP-11 PROCEDURE MANU AL raoc No afv wo 1 1 TIT LE PROJECT ENGINEERING PROCEDURE PACE 20 08 21 susact REPORTlfG DESIGN AND CONSTRUCTION DEFICIENCIES TO NRC oatt issuto 05 M 983 ATTACHMENT PEP-11-02 PROCUREE NT DOCUPENT STATEE NT 10 CFR 21 " Reporting of Defects and Noncompliances" The work to be perfomed under this purchase order (subcontract) is considered to involve a " basic component" as defined by Title 10 Code of Federal Reculations Part 21 (published 6/6/77 in Federal Register). Therefore,10 CFR 21 is applicable.
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10 CFR Part 21 ,oS M @fcmc.
' Reporting of Safety-Related Defects and
- Non-Compliances ,,0c. .. ,,,.3, m.
PAGE 210F 21 DATE ISSUED 05/16/80 The Nuclear Reguletory Commission requires directors and responsible officers of eartein firms and
- ort inisatione to report defects in oomponents and failures to comply with regulator) requirements that may l
roe att in a subetontief eefety heaerd. The new regulations are identifeed as: TMe 10 Chapter 1 Code of Federet
- Aepu/atieris - Energy - Perr 21. They apply to firms that.
e Build. operate. or own NRC licensed facilities or sonduct NRC.lisensed or reguteted actrvrtees.
e Supply safety related components for NRC lecensed facilities.
- Supply eefety related design, testing. inspectine or sensutione services for NRC lecensed facilitees.
The following occuments provide information relative to the reporting of eefety related defects and non. ,
, conformenee.
A COPY OF 10 CFR PART 21 is LOCATED ...........................................................
' A COPY OF THE PROCEDURE FOR IMPLEMENTING ,
! 10 CFR PART 21 IS LOCATED ...........................................................
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ANY DEFECTS OR NONCOMPLIANCES WHICH COULD POTENTIAILY AFFECT THE SAFETY FUNCTIONS OF THE NUCLEAR *0WER PLANT SHOULD BE REPORTED TO ...........................................................
Pene o' tee faserei neve one topwiei.ee senseenens in.e toewnement to rese*t eefety tomed eefects one nea eessies.emse ere i
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OFFICE MEMOR ANDUM C. G. Robertson June 12,1981 <
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ST-HL-19394 f,,, D. G. Barker SFN: V-0530 ' 7p gy, PROCEDURE DEv! ATIONS CONCERNING REPOP.TA!!LE DEFICIENCIES l
The intent of this memorandum is to outline those specific are s in which .urrent practice deviates from PEP !!, Rev. I " Reporting Design Construction Deficiencies to NRC" due to instructions from either Mr.
Goldberg, yourself nr I. It is our understanding that a corporate procedure is being prepared and the project procedure will be revised. During the interim this memo, with your written concurrence, will justify the use of the deviations. Note that all of the deviations are considered minor in that they typically involve a change in the HL&P responsible individual or the elimination of preliminary reviews prior to a full review.
Attachnent 1 provides a list of the deviations. Attachment 2 is a copy of Rev. 2 of PEP 11. Attachment 3 is a copy of your memo of May ll,1981 discussing scet of the deviations.
Your written concurrence of these deviations is recuested by June 22, 1991.
JW/f1j cc: J. G. White L. R. Jacobi M. E. Powell R. R. Hernandez STP RMS 4
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. V ATTACME NT I PEP 11 KY. I SECTION CffRRENT PRACTICE ,
4.1 Projectleonaoer~
Site related incidents are betag referred to the IRC chairman through several sources as opposed to only the 1 Project QA Manaser. Thus, the
" preliminary analysis" and "prelfmfnary notice" are not occurrine for each incident. In additlan, the ProjM t GA Panaeer is not notifying the Project ;
Manager nor Reafon IV office. The IK chairmae is doing this. The IRC chairmar assions the notification of the RRI to the OA'representattwe at the IRC !
meeting. ,
l 4.2 Team Leader Nuclear As wfth s1te related 1tems, there is'no Safety and L1 censing loaner a "preliainary analysis".
Instead all of the incidents are belne .i referred to the IRC. As noted above, the RRI notification is asstened to the l OA representative at the IRC meeting.
Initial report preparation is the responsfbfIfty of the EngIneerfng ,
l Coordinator. The Team Leader, N54L i revfews and submits these reports to manaaemert for approval.
l l
i 4.3 Meneger, STP The IRC chairman nettfles the Manager.
tIcensIn3 who Is respons1ble for keeping ;
l l
the ESP esecutives informed. i I !
I i i l D 5.1 Initial lacfdent Per the May 11, 1981 meme (ST-E-19074)
F Notification the Sepervf sfno Project Enotneer is the
- site contact point. Anain the notifica- i tion of the RR1 is the resoonstbfifty of '
the 04 representative at the IRC 7 meettne. l
. _ . _ __._____ _ , --_ ~ _ _ _ . -- -- _ __ _,_ __ _ _ _ _ - _ _ . _ _ _ _
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PEP 11 REV. I SECTION CURRENT PRACTICE -
t 5.6 Incident Notifications The current practice for these :
notifications that are different are: -
- a. Team Leader NS+L contacts Region IV i
office
- b. 0A representative at IRC meeting is responsible for notification of the RRI
- c. SPE at the site is a contact point
- d. The Manager Licensing informs the esecutive management af ter notification from the IRC chairman. -
5.8 Incident Review Committees A site IRC has not been inplemented.
Site problems are handled by the home office IRC.
5.9 DeadIfne Requirements The IRC is now meetine and makine determinations to comply with the .
deadlines as opposed to depending on the Proj. QA Manager or Team Leader, NS*L 5.11 Posting Requirements The Manager, Licensing is responsible for ensuring that posters and other written notifications are laplemented.
5.12 Procedure Relative to 10CFR 21. the Supervising Project Engineer's office at the site is the site contact point that an individual may utilize to empress a
, concern relative to defects 8
- non-compliances. The procedure for l D foplementing this is out11med in ST-la.-1907 (5/11/R1); a copy of which is 2
t'
- provided as Attachneet # 3.
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P.- G Im) an.3oos July 16. 1981 ST-BR-HL-40321 Houston Lig'hting & Power Company 5FN: A-0220 P. O. Box 1700 Houston. Texas 77001 ATTENTION: D. G. Barker Manager
SUBJECT:
STP HL&P/Quadrex Engineering Review
REFERENCES:
- 1. '5TP Quadrex Engineering Review of May 8.1981
- 2. HL&P Letter of May 6.1931
- 3. B&R Letter ST-BR-HL-38718 of May 8.1981
- 4. BAR Letter Rice to Stello of May 12. 1981
Dear Mr. Barker:
Brown & Root has reviewed Reference 1 forwarded by Reference 2. As a result of this review and ensuing discussions with you and your staff.
Brown & Root has provided you with the results of our review to deter-mine if any items were considered to be reportable as 50.55(e) (Refer-ence 3). B&R has subsequently reported one item in accordance with 10CFR21 in Reference 4.
In addition. Brown & Root has evalusted which of the Quadrex findings are of most significance. These findings and plans for responding are outlined in Enclosure 1. The related Quadrex findings are noted after each item. Several of these findings had been identified by Brown &
Root previously and corrective action programs were already in place or planned. Implementation of these programs will be accelerated. Schedules
- for the accomplishnent of the required action are contained in Enclosure 1 or hcve been integrated into the Project Management Plan and Project Schedule.
Response to the most significant findings contained in Enclosure 1 is our first priority. Response to the remainder of the findings will be the subject of future discussions after completion of these initial tasks.
Continued discussions between our respective technical personnel should help expedite resolution of these areas.
Very truly yours. '
OWN & ROOT. INC.
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JUL ! 7151 '-
n E. A. Saltare111 J.R. SUMPTER 13 EAS/ve Project General Manager
- File No. A0100 $I3 Enclosure ec: W. M. Rice F. E. Mue11ner J. A. Signorelli 5. Dew J.H.Goldberg
, C. L. Buck.Jr. J. L. Hawks G. G. Magnuson J.R. Sumpter J.L.Blau
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Attachrent to ST-BR-ML-40321 ENCLOSURE 1
- 1. Comuter codes. A computer code assessment program was initiated as a result of findings from a Brown 8 Root internal audit performed in Decem-ber 1980. These audit findings were consistent with the findings of the .
Quadrex review. The computer code assessment program is continuing. -
The program includes CPVR review to judge the qua ification of the codes used on STP and review of calculations for appropriate application of computer codes. This is a placed on nuclear analysis. phased program Assistance with for this immediate effort emphasis being by a consultant has been arranged. Our plan is that all computer codes used on STP in all applications will be verified. This pro Quadrex finding
- 3.1(d) 4.2.2(a). (b), and (g) gram .
will respond The estimated to comple-tion date is December 15. 1981.
In addition. STP procedures will be modified so that:
- a. Prior to future use, significant options of a code utilized on STP will be verified and identified in the CPVR-4.2.2(c).
- b. Guidance will.be provided on the priority of options available for computer code benchmarking and on the selection of benchmarks -
4.2.2(f).(h).
- c. Computer codes will be qualified for the intended STP application ifthan thethat code isbenchmark of the to be used in anwith provided application the code - significantly(different 4.2.2 1).
- d. All technical changes to the computer codes will necessitate re-verification. .
- e. Output format will be revised toincorporate a page numbering feature for future code applications - 4.2.2(f). .
The revised procedure. STP-DC-017 covering computer program use and the "
TRD covering the computer codes aIlowable for use on the South Texas Project will be issued by August 31,1981.
~
- 2. HVAC. Accel'erated review of pertinent STP HVAC systems has been under-waLv with NUS and Westinghouse consulting assistance - 4.4.2(a) (b).
3.1(d).(1). The effects of failure of non-safety-related HVAC on safety-related components will be reviewed along with separation require-ments as part of our Appendix R effort - 4.4.2(d). This review is sched-w1ed for completion by February 1.1982.
- 3. Pine Breaks outside Containment. Efforts are being accelerated with con-su' ting assistance being solicEted from Westinghouse and NUS - 4.5.3(a).
Cabletraysupportdesignsarealsobeingre-reviewed-4.1.2(1). As-suming timely subcontract approval actions, these reviews are scheduled for completion in late 1983.
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- inclosure 1 Attacheent to
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- Fage 2 ST-8R-HL-40321 i e . .. ;
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- 4. LARA. An ALARA program has been established with joint Houston Lighting
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1 I rower NUS, and Brown & Root participation, phase I, Model Review, has i been completed - 4.8.2. The phase II Review is scheduled for co pletion !
by May 1, 1982. ;
- 5. Nuclear Analysis. A comprehensive Nuclear Analysis program'has been in-i corporated into the current reforecast plan - 4.6.2(a), (b). This is an '- f ongoing program which will be completed in 1984. Intermediate task cem- i pletion dates are as shown in the plan.
1 6. E--*' aad nuhraa+ metar centro 1. The following actions will be under- '
i I
taken in th's area: '
i a. Future specifications or requirements documents associated with '
i direct Brown & Root subcontracts will contain guidelines as to what ,
l methods for analysis or testing will be considered acceptable or, t i alternatively, the potential subcontractor will be requested to '
- describe his proposed methods for Brown & Root concurrence - 4.3.2(q).
! b. Brown & Root is in the process of reviewing the requirements for
- Westinghouse review and concurrence of 80p-N555 interfaces and will i l advise Houston Lighting & power of the status anf completion date of j this program by July 30,1981 - 3.1(g). '
t
! 7. Desian Verification. Brown & Root will revise its design verification :
! procedore to require that an executed copy of the verification checklist I j be arpended to the document that was verified and to require the verifier i
4 to include a sumary statement describing the scope of his verification
.r actic.s and tha general bases on which he concludes the documents are correct - 3.1(j). The procedure modification will be completed by October 1,1981. .
i
! 8. Integrated Systens level Reviow. An integrated systems level review is !
j underway on 5TP. Items in th's review include: ;
l a. SpecificguidanceonperformingFMEAs-3.1(e). !
) b. Required support systems and operations to be performed at remote panels under bounding plant conditions - 3.1(d). ,
I l
! c. Systematic review of systems interactions under the Safety System i Hazard Analysis and protection Program (fires, missiles !
ment, impact of non-safety component failures, etc.) I. 1(d).
jet irpinge-1
- d. Comprehensive review of all environmental quellfication requirements !
- under Special Problem No. 24 - 3.1(g) (1).
i i
l e. Identification of access provisions for main.tenance, testing, and !$1 -
3.1(a).
- f. Under the Design Assurance Program, identification of plant operating l modes including startup, shutdown, pre-op and startup testing, norrral l i An j l . l t i 4 -- __ .-- -
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1;nclosure 1 Attachment to
, Page 3 ST-BR-HL-40321 l
. I plant testing, refueling, maintenance, transients, and accidents and the pararaters that should be considered under these conditions and the bounding worst case - 3.1(c).
Portions of these tasks will be completed in early 1982 and the rer.ainder by mid-1983. :.
- 9. Interference Review. An Interference Review Program was outlined in 5T-BR-HL-38556 of May 4,1981. ' Augmented internal review procedures are being impleranted to reduce development of new interferences, and a multidiscipline interference review comittee has been established to accelerate interference resolution. The comittee will swet weekly throughout the period of the design process.
- 10. _151 and Accessmaintenance Review. A program is being accelerated through augmented staffing of existing Brown & Root ISI task force. Modified program, staffing, and implementation have been addressed in the Brean &
Root letter to Houston Lighting & Power, 5T-BR-HL-39636, of Jur,e 5,1981.
The TRD covering this program will be issued the week of July 20,1981.
The program will be continuing throughout the period of the design process - 3.2(n).
- 11. Licensino. In response to the licensing concerns expressed in the design review, the following explanations clariffes the in-place activities re-lated to licensing natters. Existing Brown & Root procedures are designed to provide prorpt identification of any necessary FSAR changes and to cause the text of these changes to be prepared concurrently to facilitate periodic FSAR amendrent submittals. B&R document change notice forms (DCli) indicate whether or not an FSAR change must be associated with a proposed design change and an affirmative DCN disposition requires the associated FSAR text to accompany the DCN as a prerequisite for its acceptance. B&R 11-censing maintains the pending FSAR change files and is responsible for the draft arandment preparation. STP-DC-012 outlines the steps in implerenting an FSAR change.
HL&P, as the applicant, has primary licensing responsibility for STP.
- HL&P requests inputs from B&R concerning various NRC licensing regulations and requirerents. B&R maintains a coeputerized tracking syste:. concernir; these requests and the BAR responses. In addition. Br.R independently ad-vises HL&P of important licensing matters or licensing related positions which come to its attention through direct NRC mailings. 3.1(f).
B&R will con'.inue to monitor licensing actions to insure they are per-fonned on t timely basis recognizing the importance, particularly, of FSAR cor ittents.
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- July 16,1981 5T-BR-HL-40321 5FN: A-0220 Houston Lighting & Power Company -
P. O. Box 1700 Houston. Texas 77001 ATTENTION: D. G. Barker. Manager .
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SUBJECT:
STP HLtp/Quadrex Engineering Review jg y
REFERENCES:
- 1. STP Quadrex Engineerin Review of May 8, 1981
- 2. HL&p Letter of May 6, 9S1 #
J. L. BLA.tJ
- 3. B&R Letter ST-BR-HL-38718 of May B,1981 -
- 4. B&R Letter Rice to Stello of May 12. 1981 .
Dear Mr. Barker:
Brown &Roothasrevie$tedReference1forwardedbyReference2. As a result of this review and ensuing discussions with you and your staff, .
Brown & Root has provided you with the results of our review to deter-mine if any items were considered to be reportable as 50.55(e) (Refer-ence3). BER has subsequently reported one item in accordance with 10CFR21 in Reference 4. .
In addition Brown & Root has evaluated which of the Quadrex findings ,
are of most significance. These findings and plans for responding are outlined in Enclosure 1. The related Quadrex findings are noted after each item. Several of these findings had been identified by Brown &
. Root previously and corrective action programs were already in place or
. planned. Implementation of these programs will be accelerated. Schedules
- for the accc:plishment of the required action are contained in Enclosure 1 or have been integrated into the Project Management Plan and Prcject Schedule.
Response to the r.ost significant findings contained in Enclosure 1 is ,
our first priority. Response to the remainder of the findin s will be the subject of future discussions after completion of these !nitial t [
Continued discussions between our respective technical personnel sho g .
help expedite resolution of these areas. n-Very truly yours.
B OWN & ROOT INC.
a ['
.k. ha$ n g
V E. A. Saltar:111
- i' Project General Manage EAS/ve A I 4-
- g., File No. A0100 '
Enclosure 8, cc: W. H. Rice F. E. Mue11ner J. A. Signorelli 5. Dew J.R. Sumpter J.H.Goldberg J.L.tlev o s .4 C. L. Buck.Jr. J. L. Hawks G. G. Magnuson
.I' '
Attachment to ST-8R-Ht.-40321
. ENCL.050RE 1 -
~
- 1. Cosouqer Codes. A computer code assessment program was initiated as a resu14 of findings from a Brown & Root internal audit performed in Decem- '
her 1980. These audit findings were consistent with the findir.gs of the .,
Guadrex review. The computer code assessment propram is continutng. ;
The program includes CPVR review to judge the qua ification of the codes !
used on STP and review of calculations for appropriste application of
- This is a
- computer codes. -
placed on nuclear analysis. phased program with immediate erphasis beincAs '
t has been arranged. Our plan is that all computer codes used on STP -
in all applications will be verified. This ',,.
Quadrex findings 3.1(d), 4.2.2(a), (b), andg). frogram will respond toThe estimate *
. . tion date is Decerber 15, 1981. .
In addition, STP procedures will be modified so that: ,
p Prior to future use, signtficant options of a code utilized on STP .
i
, will be verified and identified in the CPVR-4.2.2(c).
~
i
- b. Guidance will be provided on the priority of options available for .
t
, . . computer code banchmarking and on the selection of benchmarks - . l
'.. 4.2.2(f),(h). , l
- c. Computer codes will be qualified for the intended STP application *
.. . if'the code than that of is thetobench be usedarkinprovtdad an applicationwith thesignificantly(different code - 4.2.2 i). *
., d. All technical changes to the computer co' des will necessitate *re- * -
verification. .
- e. Output format will be revised toincorporate a page numbering feature for future code applications - 4.2.2(f). .
The revised procedure, STp-DC-017 covering computer program use and the i TRD covering the co puter codes allowable for use on the South Texas -
.' '. Project will be issued by August 31,1981. .
l
- way(with 3.1d),(1).HUS and Westinghouse The effects consulting assistance of failure of non-safety-related HVAC- 4.4.2(a),
on (b), : ,..
safety-related components will be reviewed along with separation require * ~J ments as part of our Ap wndix R effort - 4.4.2(d). This review is sched-tied for completion by rebruary 1,1982. -
- 3. Pine Breaks Outside Contatnment. Efforts are being ac'celerated with con-sulting assistance t>eing solicited from Westinghouse and itUS - 4.5.3(a). -
f- Cabletraysupportdesignsarealsobeingre-reviewed-4.1.t(1). As-(g v j suming timely subcontract approval actions, these reytews are scheduled for completion in late 1983. A If NRC RIV QUADRCX REVIEW I l
REFERENCE DOCUMENT .
. . . No. 3 .
' Attachment to Anclosure 1
- ST-SR-ML-40321 -
,. , rege,2
- r. NUS. and been completed - 4.8.2.Brown & Root participation.The Phase !! Review is scheduIed for c s
by MRy 1. 1982.
- 5. Nuclear Analnis. A comprehensive Nuclear Analysis Program has been in-corporated in;o the current reforecast plan - 4.6.t(s). (b). This is an '.
- ongoing program which will be completed in 1984. Intermediate task com .- .
pletion dates are as shown in the plan. .
- 6. Vendor and Subcontractor Control. The following actions will be under-taken in this area:
~
- a. Future specifications or requirements documents associated with '
direct Brown & Root subcontracts will contain guidelines as to what .
- methods for analysis or testing will be considered acceptable er.
alternatively, the potential subcontractor will be requested to .
describe his proposed cathods for Brown & Root concurrence - 4.3.t(q). ,.
- b. Brown & Root is in the process of reviewing the requirements for
~ . .
Westinghouse review and concurrence of 80P-N555 interfaces and will ,
- advise Houston Lighting & Power of the status and completion date of this program hy July 30,1981 - 3.1(g).. *
. ^
- 7. Desian Verification. Brown & Root will mvise its desten verification Procedure to requ' re that an executed copy of the verification checklist
'. De appended to the docueent that was verified and to require the verifier .'
- - to include a sumary state:ent describing the scope of his verification sctions and the general bases on which he concludes the docux.ents are .
correct - 3.1(j). The procedure modification will be completed by -
October 1,1981. ,..- .
- 8. Integrated Systens Level Review. An integrated systems level review is underway on 5TP. 13 ems in this Mview include: ., , ,
., ,' : a.SpecificguidanceonperformingR2ks-3.1(e). -
- b. Required support systems and operations to be perforr.ed at remote -
panels under bounding plant conditions - 3.1(d).
- c. Systematic review of systems interactions under the Safety System Hazard Analysis and Protection Program (fires, missiles $. jet impinge ' >
ment,impactofnon-safetycomponentfa41ures,etc.)- 1(d). .
- d. Comprehensive review of all environmental qualification requirements
- under Special Problem No. 24 - 3.1(g). (1). .
- e. Identification of access provisions for maintenance, testing, and 151 3.1(a).
V f., Under the Design Assurance Program, identifica' tion of plant operating
- sodes including startup, shutdown, pre-os and startun testina, norma NRC Riv QUADREX REVIEW bW
- REFERENCE DOCUMENT NO. b
- e a -
i .wm .u ..
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etu s fJ q1 1 b GJi M Quadrex Corporation
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1700 Dell Avenue Campbell, CA 95008 In January 1981. HL&P initiated a technical review of portions of the engineering design work perfomed by Brown & Root, Inc. for the South Texas Electric Generating Station. The purpose was to ascertain the overall technical adequacy of the STP design. The following 84R technical disciplines were reviewed: Civil / Structural Computer Programs and Codes, Elegtrical/
Instrtsnentation and Control Geotechnical, HVAC, Mechanical. Nuclear Analysis, Piping and Supports and Spe,cial Stress, and Radiological Control.
The report stated that an exhaustive review was not feasible or desired, however, the sample was supposed to provide insight on the adecuacy of technical work perfomed. Further, it was stated HL&P inquire further into specific details.
The Report was composed of the following:
Volume ! Introduction, 2.0 Design Review Program Methodology, 3.0 Generic Findings 4.0 Technical Discipline Adequacy Assessment Voltane !! Questions Responses and Assessments (Civil thru Mechanical)
Volume III Questions, Responses and Assessments (Nuclear thru Radiological Control)
Review of Volume I W Socion ! Consultant Scope of Work p Section 2 Methodology Meetings in:
Jan.1981 - Dr. J. R. Sumpter, HL&P met with Quadrex Mgt.
Feb. 12 - Review of Quadrex Plan by HL&P and B&R '
ed k ees Feb.17 Pre meeting = a hundse+ by 42 BAR Engineers /
Managers,15 HL&P Engineers / Managers, and 11 Quadrex consulting Engineers.
March 5-27 Meetings with B&R discioline. HL&P- 22, Quadrex-12 B&R-75 including 6 NRS personnel. Approx. attendance.
Meetings at EDS ,
EDS-10. B&R-5, HL&P-3 and Quadrex-4 Messrs. Booth Stoddart, Scapini, Essarin and starting of Quadrex studied the STP plant model as related to design review implementation for ALARA, ISI access Maintenance, pipe restraints, separation and related technical issues.
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page 2 The review was based on:
Design criteria adequate for design basis?
Criteria / requirement, adequate for STP7 complete ~{
Analysis methods adequate:
Models and Computer Codes adequate? " '
Analysis factored into STP design?
Technical assumptions for designs / analysis aciquate?
Design input correct and current ?
Environmental conditions *and plant operating stat $t considered?
g Transients / accidents conditions considreed in design and analysis?
Reflected in design STP! i (h) Interface between disciplines / major subcontractors adequate? '
Disciplines assured design inputs properly used by others?
Design compatible other plant systems!
(1) Disciplines closed loop to assure equipment characteristics remain within the design boundaries?
Is as-built factored in?
Maintenance, test, and inspection considered in the design review?
Acceptance Criteria reasonable?
Design verification methods acceptable? Discrepancies resolved?
. Design verifiers appropriately chosen?
(m) Over design / conservatism?
(n) Do reviewers doubt tecnnical adequacy of design / analysis?
Problems during review? Generic or specific?
Is the problem significant for STP7 (o) Based on the results or answers to the above should other areas be investigated? If so where.
The review lasted six weeks and review provid[a clear indication that certain
- practices, policies and procedures adopted by B&R continue to have a generic impact on most, if not all, of the technical disciplines as described in Section 3.0. ;
Section 3.0
- 3. ( Moct lerious (neneric bd%dwhir* a r= avn=r*ad *a h aet elapt iconsacility)l p. i . .
Systems level intergration
- a. There is no indication of an effective systems intergration and over-view function within the 84R design process (multiple examples). ,
for example "also, a multi-disciplinary interpretation of single failure criterion does not exist in controlled documentation."
$R Review of Encineerine Cata f.1 3
- b. Three concerns:
(1) Input data to technical group not reviewed for reasonableness prior to use (Questions C-1 H-1. H-3, H-27, M-28, N-3, & N-7 (2) Calculations error rate af ter review / verification higher than l should ce encountered. (O C-16. H-15, N-1 & N-17)
(3) B&R review of vendor reports is inconsistent - well done it dlk poorly done. No occumented criteria to govern evaluation process for vendor reports.
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page 3 B&R assumes that work done by major subcontractors is correct because they verified design ie. Westinghouse. EDS Nuclear.
84R does not control analysis methods nor do they provide guidance to vendors stipulating acceptable analysis / test methods data and report fonnat required. Quadrex judged this to be inadequate.
(c) Plant Operating Modes Environmental Conditions Analysis I +
. Thorough treatment of these subjects was not evident.
. No written design basis are provided to guide the designer as to what combinations of events / modes must be considered nor was degraded eovipment performance considered.
. Design criteria in Systems Design Descriptions (SDO.)
and Technical References Documents (TRDs) appear adequate for 1973-1975 time frame but do not reflect recent developments (without TMI-2 concerns) such as loss of off site power.
environmental conditions, plant specific areas, and operating conditions (anticipat'ed from degraded from nonnel, full power etc.)
(d) Safety Related Vs Non Safety Distinctions Several instances design activities that affected plant safety were designated as non-5/R.
(1) High energy piping (2) Shielding calculations (3) HVAC re-quirements-abnonnal conditions (4) Comcuter Code CPUR status stems, Qvest, E3, EIS, H-4, H-13. MS, (5) Identified M-25, support N-10 N-17; R-6. sy(6) Operations perfonned if remote panels (7) System interaction No planned effort to review new NRC requirements.
(e) FMEA and Single Failure Criterior Analysis p.3.7
. No written guidelines for conducting failure mode and effects
. analysis (FMEA) and the only FEAs provided were those in the FSAR (not a design document) which are superficial. Such FEAs are not adequate to assure design.
. No guidelines as to what type failures should be considered for different types of equipment. There was no douemented evidence of the single failure criterion being satisfied and violation was noted regarding HVAC and IEC.
. The varied interpretation by different disciplines relative to " direct and consequential failures" resulting from postualted events was a concern and such failure establishes the first condition assumed in evaluating the plant for single failure tolerance. Quadrex "that B&R does not fully understand t* cse implications. (See quest. M-4, H-6,P-20,N-19andE-2).
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. A number of disciplines were requested to give Quadrex a list of postulated single failures considered in B&R design. "None of the disciplines was able to provide such list (Sec. Q. H-6, P-20, N-19
& E4).
(f) FSAR Comitment Tracking . . p.3.7 Documented evidence for assuring that FSAR Consnitments for systems equipment, or calculations were systematically implemented into design.
Many ' inconsistencies were noted between the FSAR and* design / pro-curement documents. There was no assurance that sub-contractor methodology changes were documented in FSAR cosuiitments and especially this was evident in EDS practices versus FSAR cossnitments.
. FSAR changes not timely and FSAR out of date.
. Site personnel are possibly not aware of FSAR comunitments and could cause real problems for field initiated changes.
. B&R Licensing group input to disciplines was not evident in order to assure a consistent understanding and implementation of NRC requi rements. Therefore, there was no effective means to get the NRC to review and concur with desired alternative methods which differ from those in the FSAR.
. B&R interpretation on Codes and Standards was inconsistent and undocumented. Interpretation is up to the individual or vendor /
supplier. ASME Code is a particularily weak area.
The following generic findings mlay, a have serious impact on plant licensability:
(g) PlantDesignBagis p.3.8
. There is little evidence of a well thought out/ consistent basis for design. Too much design is based soley on engineering judgement and rationale for such judgement wr.s not documented in such a manner ,
that it can be retrieved. Personnel turnover makes this approach very questionable if not unacceptable.
. No document exists that describes the interface design information required frvm the other technical disciplines. The lack of verified data, for at least the civil / structural discipline, may have caused conservative design but much of design is based on unverified pre-liminary data that may cause problems if later found inadequate.
The extensive use of preliminary data may have been caused by con-struction pressures controlled the engineering schedule.
. A number of criteria documents which should have been available at the onset of the STP project were missing and BAR did not recognize this until mid 1980. (3 examples were given).
. The professional quality of design review rwunents were noted to vary significantly.
. B&R indicated that W reviewed a portion of initial STP design. The AI+
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. page 5
. quality of the review and completeness was questionable.
. EDS indicated that B&R drawing changes are not reviewed on a routine basis.
. W design basis for the nuclear was carried over into BVP design without assuring appropriateness of such action.
Design from other plants have been applied to STP without assuring theirapplication was correct for STP. =
. A requirement for design margin for each discipline was not evident but individual engineers make such decisions.
. B&R does not require the use of design manuals for acceptable guidance or individual engineer log books to record key bases, assumptions or decisions. Quadrex thought this to be crucial. Without this it is difficult to retrieve fundamental background 'information regarding STP design.
(h) Equipment Reliability p.3.11
. Specific reliability requirement ie. ESF sequences were not established and this may cause problees. B&R's dependence on single failure criterion would be unsatisfactory if such problems were encountered.
. Specific reliability requirements in mechanical electrical equipment specifications and the inability to produce standard checklist of postulated failures makes rigor of the safety-related evaluation process doubtful. ,
. The design reviev of specifications showed that spurious operations were not considered. Such ommissions are no longer nonnel in the nuclear industry.
(1) Nuclear-Related Analysis p.,3.11
. Chosen analysis point out a paradox between conventional engineering work and the uniquely nuclear engineering work required for some STP design.. In certain disciplines such as civil / structural and electrical technically adequate methods have been chosen. For nuclear aspects on b* the project of the project, 8&R choice of analysis methods and assumptions have been much less than adequate. In addition, an abnonnally high error rate was observed in these type calculations plus insufficient work has been accomplished for the present state of STP design, pro-curement, and ccnstruction.
. The areas of greatest concern have been with Nuclear Analysis, Piping /
Supports, Special stress, and HVAC. Design review questions in great numbers have been reviewed to these groups for resolution.
These Groups were categorized as. "high risk" in tems of meeting licensing needs.
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page 6 B&R technical guidance on perfoming nuclear related analysis to
' subcontractor, such as EDS and NUS does not appear to be adequate.
Multiple examples i.e. B&R pipe rupture analysis, AFW pump motor qualification, time that S/R HVAC equipment and other instances in Sections 4.4, 4.5, 4.6 and 4.7 of the Report. ,
(j) Final Design verification '
p.3.13 8&R design verification process allows the use of preliminary data up to the point of STP fyel loading.
Final verification for structural will likely occur after con-struction has been completed and for equipment delivery to site.
No documented standards for minimum qualifications for design verifiers.
Ir . The only evidence of a completed design verification is signature g -+ since B&R does not have the use of a checklist. Consequently there is evidence that key design verifications questions are not being adequately) reasonable .considered. (e.g.
This practice is not are typical assumptions of industryvalid, practices are input /inoutput recent years. The observed error rate suggests a need for tighter design review and verification.
3.2 Serious Generic Findings (appear to impact reliable power generation but p.3.14 are not considered a serious threat to plant licensability):
(k) Plant Operation Criteria B&R deficient relative to operational considerations, observed on the Crestpark model 1.e. reliability, maintainability and accessibility.
(1) Plant Symmetry p.3.15 Location chosen for RHR valves, SI valves and accumulator tanks suggest symmetry was not a design objective in RCB
- (m) Valve Opening Closing f' p,3.16 Instance of incomplete specificatfory of valve closure rates. Acceptable parfomance bounds during transcents could not be confimed during this design review.
(n) Access Provisions p. 3.16 Access provisions for maintenance, inspection and test appear to be based soley on engineering judgements rather than from established-and documented requirements.". Analysis methods for assuring adequate access for maintenance and ISI coupled with considerations for ALARA,.
radiation exposures have been inadequate.
l (o) Test Provisions p.3.16 Preop test and resultant test provision were not systematically considered / implemented into design documents.
(pl Local Temperatures during Maintenance Areas in MAB and FHB temps 76-85 degrees F. and certain cubicles may be as high as 103.8 degrees F. during routine operation, HVAC considerations for personnel, if p@.
suited up for radiation may not be adequate to meet HL&P needs.
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Continued use of abnomally low seismic valves for valve and pump end loads (even though valves were a nom in 1975) is a major factor for the number and size of pipe supports currently provided in STP design. Other fims in the industry have changed progressively in recent. years and use higher acceleration limits for valves and their components to reduce the nunber of supports.
4.0 Technical Discipline Assessment p.4.3
- 4.1 Civil /StVuctural Findings and Conclusions
- 1 4.1.1 Basis for Questions: Geotechnic, containment, seismic, design i
criteria, interface, adequacy.
4.1.2 Technical Assessment for Adequacy A number of studies in this area were noted (probably result of Show Cause). Infor1 nation relative to these studies were excluded from this report. Quadrex was unable to detemine the aoecuacy of Civil / Structural based on this review and further analysis was recomunended. Following was noted:
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. design over conservatism
, . personnel turnover i . some degree design inexperience
. over cautious reaction to schedule pcessures
. ' lack experience in code interpretations and industry practices throughout design process 1
. chosen criteria nom with industry
. designer / design verifier misjudged certain areas.
. transient conditions a problem in past but B&R recognizes the affect on design now.
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. Geotechnical Engineering ok *
. Containment Analysis / Design equals industry practice. Analysis of RCB internals ok, models analysis techniques ok. Loading phenomenon and combinations ok.'
4.2.1 Most Serious Findings (which are expected to impact plant p.4.fi licensability):
(a) Input data reasonablemess. Environmental info affecting discipline
- not fixed.
l (b) No evidence that Civil / Structural (C/S) evaluation of internal missiles reasonable or criteria for some describe.d in TRD was implemented in design.
Following may seriously impact:
(c) Turbine bid not analyzed SSE (consnitment)
(d) Significant C/S differences may exist with respect to NRC criteria, methods, and/or FSAR.
(e) C/S not appear responsive to recent NRC requirements.
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. page 8 l
.(f) B&R use of input data furnished by EDS for pipe rupture l'oading may not be adequate.
(g) B&R assumptions MAB dead loads may not represent actural conditions, final design of floor elements MAB/EA8 _may, be adequate but calculations were hard to follow and amp,itfication effects of vertical seismic were not properly considered.
(h) The plan to verify actural loading after the structure has been errected poses a potential licensing risk. ,
4.1.2.2 No Comnent on Serious Finding by Quadrex 4.1.2.3 p.4.7 Noteworthy Findings increased cost / schedule input '
(i) (j) (k) (1) (m) overconservatism, proper input for desing, poor interface between C/S and vendors criticized.
4.1.2.4 Potential Problem Finding (may or may not exist but should be investigated).
(n) - (v) interface or use of data absent between disciplines, MAB coefficient for verticle seismic incorrect, pressure load not correct in duct ring calculation, lateral capacity for vertical wells may not be sufficiently rigorous, equipment loading valves preliminary, torsional effects on generation of response spectra were not con-sidered, B&R did not rigorously consider flexibility of slabs as relates to vertical response spectra generation which may affect equipment qualification, equipment qualification need additional attention /
manpower effort.
4.1.2.5 Other findings minor effect on design (w) - (z) input data docunented informally, (aa) - (dd) penetration design questioned, some aspects of design process not understood by some personnel, B&R questioned finite .
element model conservatism but used it anyway without justification, B&R did not often use mass participation in high frequency range, tornado loading questioned, computer codes not reviewed, containment temp.-
4.2 Computer Codes p.4-12 4.2.1 Basis for Review computer Codes Is B&R being consistent with standard industry practices in their usage of computer programs for STP. Program verification and qualification, per three ASME publications, was defined 4.2.2 Computer Code Technical Adequacy Assessment
. Many B&R Computer codes unverified at present however, design may be adequate.
. Qualification of programs not treated in a systematic manner or separated from verification.
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page 9 Overall desion adeouate only if both verifications and oualifications procedures reconnended by ASME and used in industry followed.
4.2.2.1 Most Serious Findings (which are expected to impact licensability):
(a) - (c) Nianerous programs with no (CPUR), BAR verification non-safety related not required but B&R classification S/R questioned.
Asstanption that entire computer code verified if portion of code verified is not adequate approach.
May have imoact
- p. 4-15 (d) - (f) differences between FSAR and Program Status Sunnary, CPVR procedure deficient, computer calculations not properly con-trolled, five options to perform CPUR lacks guidances, reverification should not be limited to only significant applications.
4.2.2.2 Serious Findings No findings in this category 4.2.2.3 Noteworthy Ditto 4.2.2.4 Potential Problem Findings (may/may not exist)
Control document for computer codes not evident but should be pre-pared since FSAR is not a design document.
.. Guidelines for certifying appropriateness of computer code application do not exist.
Use of recornized program w/o verifying is not industry practice.
Improved control procedures for program modifications in computer Code CPV2 necessary.
Program errors must be brought to attention of appropriate personnel /
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4.3 Electrical p,4 17 4.3.1 Basis of' review Control room human factors, design criteria Actuation sequence status of safety load, testing, operating req / accessibility, conductor sizing,
- design basis alams, design basis instrumentation.
In depth review not perfonned--design incomplete or in process of being redone.
4.3.2.1 Most Serious (would seriously impact):
' No findings but following should be corrected:
(a) Common instrument air line, per FSAR dwg 9.4.2-2, does not meet single failure criterion. This finding suggest B&R is not sufficiently experienced in Facture Mode Effect Analysis.
That crosses discipline boundaries.
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.(b)-(o) TRD does not address plant wide separation, no plan to develop top level TRD's, no methodology or documentation to verify adequate separation or single failure criterion, no doctanentation for defining separation barrier requirements for internal designers / vendors, no TRD for equipment classification, no method to assure FSAR comitments are implemented in design, recent NRC requirements may not be reflected in STP design.
HL&P/B&R interface pre op and operating procedures a problem long standing, designer also the verifier.
4.3.2.2 " Serious Findings
- p.4.24 None 4.3.2.3 Noteworthy Findings p. 4-24 (p)-(q) Technical communication between disciplines Specs, TRDs, SDD inconsistentcies could cause design errors.
4.3.2.4 Potential Problems Findings (r)-(z) Tech comunications, specificiations. TRD SDD inconsistencies may cause design errors, no. procedures for vendor document (aa)-(bb) review, B&R position in use of only fail safe C4 ass IE solenoid valve questioned, Criteria for equipment / system monitors not clear, separation objectives ESF sequences questioned, sensor response time questioned, analysis of actuation logic for FW/AFW questioned, STP apecial raceways questioned.
4.3.2.5 Other Flamemastic and PVS basis for use questioned.
Note Flamemastic caused Brownsferry fire problems.
4.4 HVAC Findings Conclusions p. 4-26 4.4.1 Basis ~
Adequacy of HVAC for nuclear systems i.e. RCB Fan Cooler, Reactor Cavity Cooling, Control Room and EAB System & Safety Class . , ,
Filtration System: The following considered design input, bases',
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criteria; test / inspection, engineering calculations, design control, details, i
4.4.2 Technical Adequacy Concerns indicate present design inadequate but all would be ok
, if concerns addressed and design changed.
l Engineering effort unorganized and inadequate.
4.4.2.1 Most Serious findings (would impact)
(a) Design bases not well defined.
(b) Safety classification questioned.
Following may seriously impact: $ l4-o
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Page 11 (c)-(g) Calculations for hydrogen questioned, separation requirements to protect against common mode failure not evident, floor drainage impact.on ECW leak detection questioned. FSAR-design documents inconsistent, HVAC design verification questioned.
4.4.2.2 Serious Finding Consideration of actual plant operating condition not considered in HVAC calculations.
4.4.2.3 Noteworthy Findings None 4.4.2.4 Potential Problem Findings (i) HVAC design input marginal (j) basis for ambient conditions not traceable to user requirement.
(1) Fire hazard analysis not a formal control document to be used by HVAC.
(m) HVAC design crtteria scattered thru documents (n) HVAC margins questioned.
(0) No procedure to assure voltane damper pressure drops.
(p) Cooling support skirt adequacy determine.
(q) HVAC support adequacy should be determined.
(r) Calculation refinements rx cavity cooling necessary.
(s) No provisions for damper status (t) HVAC damper designed by vendor not reviewed. ,
(u) Review HVAC near high energy line not performed.
(v)-(z) various.
Other findings documented basis for selecting HVAC alarms
- NUS calculation not available at B&R.
4.5 Mechanical Discipline Findings Conclusion p.4,36 4.5.1 Basis of Review 4.5.2 Pipe Rupture Tech. Adequacy (inside RCB)
EDS efforts would be adequate if the following findings were resolved:
Numerous potential interface problems exist between EDS and B&R p }4, such as:
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. page 12
. Evaluation of high/ moderate energy environmental con.dition's
.inside.RCB.
. Evaluation of pipe whip, jet impingement, and restraint loads or structural components.
. Timely updating interaction matrices and related pipe rupture analysis to show reroutes, design-construction changes. ,
4.5.2.1 Most Serious None where would impact licensability.
Following may: ,
(a) 10 degree half cone to define jet surface not appropriate for sub cooled liquid per ANSI N176.
(b) EDS did not perform design review or verification of preliminary loads transmitted to B&R, however, data was used as basis for plant design.
(c) Secondary effects from pipe rupture not adequately investigated.
(d) Verification of super pipe stress limits for breaks not done.
4.5.2.2 Serious Findings may impact (e) EDS did not apply controlled criteria for ALARA ISI & Maintenance access.
4.5.2.3 Noteworthy findings (f)-(i) Interchange between EDS/B&R, conservative analysis, specification unrealistic, restraint design complexity causing interferences.
4.5.2.4 Potential Problem Findings (may/may not be)
(j) No guidelines amount analysis (k) Evaluation pipe whip impact, jet impingement restraint loads on structural components should have been made sooner.
(1) restraint prevention of.out-of-plane loads not evident.
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(m) Nonconservative stability guidelines for dynamic analysis (computerevaluation).
.; 4.5.2.5 Other findings (n)-(p) Test plan, installation tolerances pipe whip restraints, stability criteria used in evaluation questioned.
4.5.3 Pipe Rupture Tech. Adequate (outside) 1 Pipe ruoture design outside containment does not appear to be adequate.
4.5.3.1 Most Serious Have impact:
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- (c) The current B&R status indicates a general lack of understanding planning preparation and availability of procedures.
(d) More specific guidelines needed for criteria to delete lines from analysis and must consider secondary effects.
(e) 10 degree half angle not appropriate for jet impingement.
(f) dynamic analysis required for super pipe runs.
(g) responses to questions were sometimes inconsistent and con-tradictory indicating a general lack of expertise with pipe rupture considerations.
(h) n'ot all cr1teria stated for "no break zone" appeir to be applicable to STP.
(i) B&R does not plan to look at field installed lines for inter-action analysis.
(j) B&R response on bi linear idealization of restraint stiffnesses was unclear.
(k) Acceptance criteria for pipe whip are not well fomulated.
4.5.3.2 Serious Findings (1) 4.5.3.3 Noteworthy Findings (m)-(s) No analysis perfomed, conservative break EDS/B&R interface weak, pipe rupture evaluation.will be performed as retro fit, adequacy of procured equipment because pipe rupture analysis not yet considered, some over design.
4.5.3.4 Potential Problem Findings.
(t)-(z) Interference resolution slow (u) assumption on pipe whip (aa)-(hh)restrint may not be valid, (v) B&R does not plan to routinely review existing interaction matrices to account for field changes -(w) specific measures to mitigate flooding and com-partment pressurization may be required (x) separation and protective enclosures for environmental effect mitigation .
should be verified for adequacy (aa) Restraint type selection criteria needs to be improved (bb) Load combinations details need to be improved (dd) imprmved material specifications should be improved (gg) existing structural and embed designs may)not (hh The beprocedure adequatefor due to unnecessarily specifying desired high pipelocations restraint whip loads was unclear NOTE: Short Review from this point because of lack of time-------
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- page 14 4.5.4 Basis for Mechanical Review: p.4-46 Design adequacy, interface procurement, knowledge codes.
4.5.5 Technical Adequacy Assessment "ThetechnicaladeQuacyoftheMechanicaldisciplineIsnot presently adequate".
There are two major concerns'(1) Their understanding and implementation of Code and industry requirement and (2) the apparent lack of results from an integrated systems review function. If thesd areas were corrected the mechanical aspects of design would be adequate.
Most Serious Findings:
(a)-(g) Typical: (a) Concern that NRC SRP criteria for active components operability during service not tested or tested / analyzed.
(b) Correlation of transients / plant design may not be valid for systems such as Eces where pipe rupture endition occurs.
(c) Concern over B&R review of vendor design calculations.
. (d) Main Steam-SRV calculation designated non safety related.
4.5.5.2 Serious Findings -
(H)-(i) 4.5.5.3 Noteworthy findings (j)-(k) 4.5.5.4 Potential Problem Findings
~
I (1)-(s) 4.5.5.5 Other Findings
! 4.6 Nuclear Analysis Findings .
4.6.1 Basis for Review Mechanical transient analysis, steady state analysis, thermal hydraulics (fluid flow, heat transfer, environmental (analysis
& systems integration and performance analysis).
4.6.2 Technical Adequacy Assessment p.4-57
. NUS perfonned one adequate environmental analysis.
. Nuclear Analysis performed by B4R to date are either incomplete or are not adequate.
. Considering the stage of construction B&R Nuclear Analysis Group has not produced a significant contribution to the STP design. Licensing delays, retrofits, non cost effective over conservatism in equip-ment specs. may result.
. Technical assumption used in design / analysis are not reasonable for STP
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, page 15 Design inputs used by disciplines were not verified and incorrect design basis inputs were used in the environmental and ECP analysis.
Plant operating states were considered for environmental analysis conditions. These results are not reflected in STP design. All plant operating conditions not for ECP analysis were not considered and ECP initial conditions are not consistent with heavy civil.
Analyses for annulus pressurization, pipe rupture environmental effects. ECP heating and. hydrogen concentration considered transcient and accident conditions however these analyses are either incomplete or incorrect, final results are not yet adequately reflected in STP design.
Nuclear analysis group (eycept for environmental) either could not respond at all or indicated that the analyses were the respon-sibility of another group. (this was the opinion of other groups relative to questions referred to the Nuclear Analysis Group.
Quadrex said there is uncertainty within B&R as to who is responsible for analysis and this could cause real problems.
4.6.2.1 Most Serious Findings (a)-(o) Typical (a) Temp valves for equipment design not controlled nor analytical basis for temp outside containment.
(b) either insufficient analysis or no analysis.
(c) Nuclear Analysis failed to scope, perfonn or have analyses performed that should have been completed give or present state design / construction.
4.6.2.2 Serious Findings (p) B&R plan to analyze all high energy lines in IVC and MAB and support all non safety lines does not appear reasonable or cost effective and this is impacting plant access and main-tainability.
l 4.6.2.3 Noteworthy
- l l (q)-(s) Analysis methods inadequate, needed analysis not perfonned impacting project, B&R response to NRC question i.e. hydrogen concentration not timely.
4.6.2.4 Potential Problem Findings (t)-(u) BAR did not thoroughly review W data (v) analysis of essential cooling ponW failed to consider all design bases.
4.6.3 Basis for Mechanical p. 4-63 ISO Valve qualification, CCW/ECW operation / design, RCW makeup system pump ECCS ptanp room flooding, sump pump head calculations, FW perfonnance analysis; Reactor coolant pump oil system line size.
4.6.4 Mechanical Tech. Adequate Assess b Worked appeared to be adequate
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4.6.4.'1 Most Serious -
i (a)-(c) (a) A complete FMEA (failure mode effects analyais) inclusing effects of pipe rupture, in combination with a
, single active component failure, not performed. (b) 44% non-conservative error ECCS pump room but ok.
(c) Fomal FMEA methodology and sinble failure analysis guide-lines do not exist.
4.6.4.2 Serious Finding No Quadrex connent .
4.6.4.3 Noteworthy Findings i (Lack timeliness FW analysis) q 4.6.4.4 Potential Problem
. (e)-(j) Six potential problems could cause serious problems in operation.
4.6.4.5 Other findinos
. (n)-(1) Potential design interface problems some pump calculation problems.
4.7 Pioine Suooort Findings p. 4-69 4.7.1 Basis for Review How mechanical piping and piping support design and analyses perforined
. 4.7.2 Piping Support Tech. Adequate Assessment In general EDS in-containment design analysis appears adequate, ,
, 4.7.2.1 Most Serious (a)-(f) (a) FSAR incorrect since no modal analysis done.
(b) EDS does not review FSAR for changes (accuracy) -
(d) High stiffness (default valves) for pipe supports unconservative.
j (f.) FSAR incorrect i.e. ED's modified spectra used in
- simplified method.
4.7.2.2 Serious - No Quadrex Comment 4.7.2.3 Noteworthy -
(g) to (r) EDS may not analyzed adequately belt pull out 4.7.2.4 Potential Problem Findings (1)-(o) Several problems that must be investigated.
4.7.2.5 Other blk
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4.7.3PipingSupport(outs'de)
Technical Adequacy The annount of work done on STP piping design analysis by B&R was very little. Quadrex could not sssess.
4.7.3.1 Most Serious Findings (a)-(r) Several inadequate items - FSAR inconsistent Lack coordinator, NRC EB-6 reconnendations agnored, unconservative seismic assianptions on anchors.
4.7.3.2 Serious Findings (e)-(o) Design margins not assured, unsatisfactory, design IVC inter-face support restraints for main steam /feedwater lines appears possible.
4.7.3.3 Noteworthy findings.
(p) - (q).acceiration limits too low-conservative
. ~ No tolerances for restraint orientation 4.7.3.4 Potential Problems (r)-(y) (r) little dynamic analysis (s) lack doctmentation of hours for selecting valve end loads, (w) formal documentation of B&R Stress Group design process / practices not evident 4.8 Radiological 4.8.1 Basis of Review How design affects personnel radiation exposure to on and off site population.
4.8.2 Radiological Contaol Technical Adequacy. The B&R Radiological Control design program is not currently adequate. Quadrex reconnended innediate corrective action to assure future adequacy. ~
4.8.2.1 Most Serious (a)-(g) . Failure to meet single failure criterion
. Design review inadequate (ALARA Spec. design)
. MAB HVAC filter media eliminated (questioned)
. B&R position - shield calculation not SR questioned
. Radiation zones to shielding not adequate by c'onsidered for ISI
. Radiation zone dwgs not prepared.
. Design basis controlling removable concrete block walls not evident.
4.8.2.2 Serious (h)-(0) (1) Design review inadequate (j) maintenance (test) inspection requirements not incorporated
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, page 18 4.8.2.3 - 4.8.2.4 (p)-(z) Many serious problems of some of these are really problems (aa)-(11) 4.9 ISI and Maintenance Access ,, ,
4.9.1 ISI Review of STP Model Basis (1) does it meet 10 CFR 50.55 a(g) 3 and ASE Code Section XI?
(2) How many exceptions to access engineering criteria processed and cbmpact? *
. Not all criteria have been implemented into drawing. Model revealed noncompliances with criteria. Review of program (TRD draft vs Model showed consistency has been lacking.
4.9.2. Maintenance Access Engineering
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t RIV - HL&P MEETING NOTES (PHILLIPS) 9-8-81 10:00 a.m.
Attendance List Attached Carl S. Introductory Remarks G. Oprea Introductory Remarks. Turned over to G. Goldbers ..
G. Goldberg:
Background., I came in and wanted a bench mark on BAR engi,neering since they were a first time AE. Approach was not to employ an AE to review since they do not do these reviews because of possible conflict of interest. Next decided on a company or consultant to perfom such a review.
Basis of Ouadrex Review 270,250 questions spanning all disciplines. After these questions were answered meetings were held between Quadrex and B&R engineering personnel.
Quadrex conclusions were based on this data.
One significant difft ence relative to this review was no draft would be submitted to HL&P becau.se of the sensitiveness of the STP. There were instructed to stick to the facts, and leave out opinions. Unfortunately this did not happen, that is, opinions and speculations were included.
B&R met to review all questions / conclusions, relative to reportability.
B&R found only one HVAC. HL&P reviewed there and then reported two additiona"1 50.55(e)s. Computer codes and .
HL&P additionally decided that some significant items existed.
- 1. Computer Codes (Verification / Qualification)
- 2. HVAC (Faulted Condition Loads not considered)
- 3. Pipe Break Analysis Outside Containment (had not Perfor1ned analysis)
This may result in redesign.
p 4. B&R ALARA Censideration Inadequate i. e. not being factored routinely 8 into design. A M.
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, Page 2 RIV HL&P Meeting
- 5. Same was true of ISI/ Maintainability !
- 6. Nuclear Analysis'(B&R analysis had gone into a lull for several years). Possible explanation: B&R utilizing NUS personnel.
Haliburton was to bring in EBASCO but did not occur. This dropped thru the crack.
I
, 7. Subcontractor / Vendor -
B&R not giving them adequate guidance.
Analysis /Models etc. not adequate.
- 8. Design Review Process. B&R has had problems. Now have a good process but B&R must use it.
- 9. Systems Integration. "
The first day I came aboard I identified this problem. "
Quadrex identified the same problem. This is a problem in the indur',ry. .
- 10. Interference Control. Model and Inter-Squad Control are used. !
i Quadrex found this to week in teries of implementation.
- 11. FSAR Tracking. "I agree that some of this fin ding is factual but part was subjective.
. B&R does not adequately track FSAR changes etc.
" Disagree with Quadrex finding relative to B&R licensing group not being involved because this finding was subjective. ,
B&R Letter This letter outlines B&R actions.
Quadrex Saw Inexperience HL&P agreed with report that B&R did not have the experienced staff that
[ was needed.
HL&P, with B&R assisting, initiated a recruiting program. This has had l limited sucess.
HL&P had B&R go to industry to contract experienced people that are not fully utilized at different companies. This did not work.
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Page 3 HL&P RIV Meeting HL&P had B&R go to subcontractors to subcontract portions of AE effort.
NdSandWispresentlydoingsomework. Gibbs Hill is being considered.
Goldberg Opened the Meeting for Questions Q. Considering Quadrex identified concerns with control of subcontractors, do you forsee a problem controlling the additional subcontractors?
A. Yes, This is still a problem that we are working to' resolv,e. We have stressed to B&R that the controls must be in place and just getting contractors does not solve the problem.
Q. Will you subcontract the design effort?
A. This is being considered. Gibbs Hall may do some portion of the AE work.
Q. How will you monitor the AE? With your own organization?
A. HL&P has hired some experienced personnel who have systems integration capabilities.
Note: Question,made more specific does HL&P organization review design specs for adequacy on a spot check basis?
Note: Answer Yes, HL&P organization does such reviews.
Note: Question -What about SAR cannitnents implemented into design / work?
Note: Answer-Tracking systems being set up.
Q. Is the control of vendors / subcontractors limited to designs?
A. No ANSI and daughter standard, not a part of contract. A large retrofit effort is has been underway for sometime. ,
Q. Did you review Quadrex Engineers qualifications?
A. Yes, we assured they had appropriate education / experience.
Q. Are you following up effects on hardware?
A. Yes, we are following up on this matter. Our production work has been curtailed and we are primarily doing reexamination and rework.
Q. What steps are you taking to review questionable engineering practices and work already done.
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Page 4 RIV HL&P Meeting O
A. A review is in progress in the areas where such review is needed. For example structural and loading of same was not properly documented.
They must have a balance sheet to see if there is a problem.
KARL SEYFRIT Expressed concern relative to ISI/ maintenance problems be caused by over desig'n'. ALARA considerations, etc.
Q. What makes you think B&R will do it differently the second time?
A. I can't guarantee they will but we will monitor to assure that they do.
If they don't we will get someone else who can.
Q. Did you consider other besides Quadrex?
A. Yes.
Q. Has B&R responded to all 275-300 Quadrex Concerns?
A. No, we are reserving the perogative to look at only the significant problems. We have considered all when reviewing for 50.55(e) Reportability.
Some did not merit further . review.
Q. Construction is ahead of design and it appears it will be built in many instances before design is verified f.e. equipment qualifications or procurement.
A. Yes, we are looking at that and are calling B&R's attention to that by .
using a risk evaluaticn team to assure that equipment is not placed in that cannot be withdrawn.
Q. Sans1xastionxixxaheadxafxdesignxandxi1xappeaxxxitxwilixhexhmildixinxmany instamaanxheferexdesignxis Q. Are you not limiting the review of identified problems when you said you let a draft to come about?
A. We felt-it was only necessary to review problems we felt to be significant.
HL&P does not have to consider their consultant's work " gospel".
KARL SEYFRIT There is a need to sta document why insignificant areas needed no further review or action, gg4
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Page 5 RIV Ht&P Meeting Q. What impact will the Quadrex Report have on ccnstruction? .
A. From our point we have a disaster. We cannot do meaningful construction for a year. We are going to turn the clock back for a few years. This '
is going to cost a great deal of money and our company will pay a huge penalty.
Q. Asked if the breakdown in design would be reported per 10 CFR 50.55(e)?
A. Mr. Oprea 's' aid they can respond to that within the next several weeks.
Q. What about the problems with the use of Codes. Who is doing the review?
A. B&R did not benchmark the revision of ASME Code. B&R is having NUS do the review.
Karl Seyfrit restated his position that broadly speaking there has been a breakdown in design and a 50.55(e) should be reported.
George Oprea stated he was concerned with the timing.
Karl Seyfrit replied this matter should be reported as any other breakdown.
George Oprea replied that timing is essential. Let us go back and assess this matter further.
Goldberg stated that he hopes the Quadrex Report brings about corrective action not sensationalism.
Q. Since this has been caused by management, what is being done to assure. the kind of management needed. .
A. B&R has brought in several performers.
Goldberg--We were concerned that B&R did not have enought race horses in the stable to finish the race. B&R is trying to get the people but many are afraid that B&R is going to get kicked off the job. B&R is not a company that is willing to reorganize. We are currently negotiating with B&R to reorganize / restructure the company or at least restructure a special arm of the company to do the nuclear work. B&R does not understand nuclear I work since most of their efforts are dedicated to petro chemical and fossil work. B&R personnel does not even understand what expertise is g{2h required in the nuclear area.
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. Page 6 ,RIV HL&P Meeting
.1 Karl Seyfrit reemphasized the need to be notified relative to the design.
Golberg stated that B&R was not geared up to build a nuclear plant. B&R AE management is a heck of a lot different from others. B&R is operating like AEs operated 10 years ago.
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