ML20213A453
| ML20213A453 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 01/23/1987 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20213A334 | List: |
| References | |
| 207.2(B), 207.2(B)-R01, 207.2(B)-R1, NUDOCS 8702030285 | |
| Download: ML20213A453 (34) | |
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TVA EMPLOYEL CONCERNS REPORT NUMBER: 207.2 (B)
- SPECIAL PROGRAM REPORT TYPE: SEQUOYAH ELEMENT REVISION NUMBER: 1 TITLE: DEVIATION DOCUMENTATION Safety & Licensing Evaluations PAGE 1 0F 32 REASON FOR REVISION:
- 1. Revised to incorporate SRP/TAS comments and add chronology.
PREPARATION PREPARED BY:
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TVA EMPLOYEE CONCERNS REPORT MlMBER: 207.2 (B)
SPECIAL PROGRAM REVISION NUMBER: 1 PAGE 2 0F 32
- 1. CHARACTERIZATION OF ISSUE (S):
Concern: Issues:
WI-85-100-027 a. Safety and licensing evaluations by "TVA Safety and Licensing EN DES (including NEB) are inadequate.
Evaluations by EN DES (including NEB) are inadequate and appear b. Safety and licensing evaluations by too much in a cover up mode. EN DES (including NEB) are in a CI has no further infomation. cover up mode.
Anonymous concern via letter."
- 2. HAVE ISSUES BEEN IDENTIFIED IN ANOTHER SYSTEMATIC ANALYSIS? YES N0 X Identified by None Date None Documentation Identifiers:
None
- 3. DOClMENT NOS., TAG NOS., LOCATIONS, OR OTHER SPECIFIC DESCRIPTIVE IDENTIFICATIONS STATED IN ELEMENT.
None
- 4. INTERVIEW FILES REVIEWED:
Expurgated file for WI-85-100 was reviewed and no additional l unreviewed infomation was found.
l 5. 00CtMENTS REVIEWED RELATED TO THE ELEMENT:
See Appendix A.
- 6. WHAT REGULATIONS, LICENSING COM4ITMENTS, DESIGN REQUIREMENTS, OR OTHER APPLY OR CONTROL IN THIS AREA?
See Appendix A.
- 7. LIST REQUESTS FOR INFORMATION, MEETINGS, TELEPHONE CALLS, AND OTHER DISCUSSIONS RELATED TO ELEMENT.
See Appendix A.
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SPECIAL PROGRAM REVISION NUMBER: 1 PAGE 3 0F 32
- 8. EVALUATION PROCESS:
- a. Reviewed engineering responsibilities relating to the licensing process in general, and safety evaluations in particular, to detennine types and nisnbers of docisnents to be examined.
- b. Reviewed available transcripts of NRC investigative interviews to gain additional infonnation regarding the Concern.
- c. Reviewed past and present TVA engineering department procedures and practices for safety and licensing evaluations.
- d. Reviewed samples of output docisnentation packages which included safety and licensing evaluations, to detennine compliance with procedures and appropriateness of content.
- 9. DISCUSSION, FINDINGS AND CONCLUSIONS Chronology:
Late 60s: SQN design begins 10/15/68: SQN PSAR issued Mid-69: Draft 10CFR50 Appendix B issued 01 /70 - SQN-QA Manual, through R12, Constitutes Engineering
( 12/73: Design Procedures in effect 06/27/70: 10CFR50 Appendix B fonnally issued 10/73- EN DES Engineering Procedures (EPs), (Red Book) in 06/85: effect. Superseded SQN-QA Manual 1976: TVA commits to ANSI N45.2.11-1974 in SQN FSAR i
l 09/27/80: Operating License (0L) issued for SQN Unit 1 03/18/81: Nuclear Engineering Branch procedure NEB EP25.45 on Safety Evaluations issued 09/15/81: OL issued for SQN Unit 2 03/11/83: Instruction SQN AI-19, R0 issued to cover Plant j Modifications after License received l
1 l
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SPECIAL PROGRAM REVISION NLMBER: 1 PAGE 4 0F 32 04/09/85: TVA Topical Report TVA-TR75-1A, Rev. 8, issued 06/85- Office of Engineering Procedures (0EPs) in effect.
06/86 Superseded EN DES-EPs 08/85: SQN voluntarily shut down by TVA because of questions about the enviromental quality of electrical equipment.
09/17/85: NRC requests TVA submit plans for correcting SQN problems 09/85 to present: Sequoyah Project Manual (SQEPs) in effect 11 /01 /85: TVA presents Sequoyah Nuclear Performance Plan (Draft) and their Corporate Nuclear Performance Plan (initial issue) to NRC 11/17/85: NRC receives employee letter regarding concerns about TVA's nuclear program 12/07/85: TVA receives concern WI-85-100-027 02/18/86: NRC letter forwards copy of Concerned Employee's letter to TVA for review and response 02/21/86: NRC investigative interview with concerned individual 03/10/86: TVA Corporate Nuclear Perfonnance Plan (CNPP) Volme I, R0, issued 06/23/86: NRC forwards transcript of interview with concerned individual to TVA 07/86 to Nuclear Engineering Procedures (NEPs) in effect.
present: Superseded OEPs Discussion:
9.1 BACKGROUND
t l
The Sequoyah Plant Design was begun in the late 1960s when regulatory requirements were not required to be well docisnented. The majority of the SQi1 licensing commitments were made in the late 1960s and early 1970s and many of the present day regulatory documents are not Sequoyah requirements. It is not necessary to update the Sequoyah Plant to include all of the new guidance published by NRC, unless there are specific requirements to do so.
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- TVA 04PLOYEE CONCERNS REPORT NW4BER
- 207.2 (B)
SPECIAL PROGRAM REVISION NUMBER: 1 PAGE 5 0F 32 This concern was originally raised at Watts Bar Nuclear Plant (WBN). It was considered generically applicable to Sequoyah Nuclear Plant (SQN) since the organization and procedures that governed SQN design were essentially identical to WBN.
WBN is a Construction Pennit (CP) Plant while SQN is an Operating License (0L) Plant. Construction is not yet complete at WBN, while SQN received Operating Licenses on Units 1 and 2, in 1980 and 1981, respectively.
The content of the employee concern is without specifics as to identification of individual problen areas. Therefore, it will be addressed on a broad scope basis.
9.2 GOVERNING DOCUMENTS AND DEFINITIONS Engineering implementation procedures for the Sequoyah Nuclear Plant have been docunented in the following:
o Sequoyah Nuclear Plant Quality Assurance Manual (SQN QA Manual) (early 1970 through late 1973) o Division of Engineering Design Engineering Procedures Manual (EN DES EPs) (late 1973 through mid-1985) o Engineering Program Directives, Office of Engineering Procedures Manual (0EPs) (mid-1985 through mid-1986) o Nuclear Engineering Procedures Manual (NEPs) (mid-1986 until present) o Sequoyah Project Manual (SQEPs and SQEP-AI) (late 1985 until present)
The current governing procedures are the NEPs with NEP 2.1 addressing licensing support. From NEP - 2.1, Section 1.0, Definitions :
" Licensing: Those activities required to demonstrate compliance with Federal, state, and local requirements in the design, construction, and operation of TVA facilities.
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" Licensing Su) port: Regulatory engineering activities perfomed wit 1in DNE (Division of Nuclear Engineering) in support of licensing."
There are various types of Safety Evaluations depending on the level of review and approval required by the various procedures and the nature of the subject and reports being examined. The discussion section addresses the details of the various safety and licensing evaluations.
Fomal TVA Safety Evaluation Reports (SERs) are prepared by members of the DNE Nuclear Engineering Branch (NEB). They follow the philosophy originally described in NEB-EP25.4.5 and carried on through NEP 2.1 and 3.1. These SERs are discussed in Section 9.4 of this report.
9.3 DESCRIPTION
OF INVESTIGATION The employee concern is not specific as to individual problem areas, except that the CI does refer to the Engineering Design (EN DES) organization. The NRC Investigative Interview (App. A, 5.bb) was reviewed but no additional infomation was found.
The evaluation team first reviewed the past and present engineering procedures (EPs, OEPs, and NEPs) to detemine which ones dealt with safety evaluations and the licensing process and also defined the scope of Engineering's responsibilities in these areas. The procedures reviewed covered the period of time from the development of the EN DES i EPs through to the current NEPs initially issued July 1986.
l The procedures so identified are listed in Appendix A, 5.u l l
through 5.y.
The SQN QA Manual provided the original procedures for Engineering to use in the licensing support effort. These were in effect prior to creation of EN DES organization. The CI stated that the concern was with the EN DES organization.
l In October 1973, an organization change at TVA resulted in the establishment of the Division of Engineering Design. At this time a project system was established that assigned i engineers from the various engineering branches (Civil, i Electrical, Mechanical and Nuclear) to work exclusively on a specific project, such as Sequoyah. New division-wide procedures, called Division of Engineering Design, i
Engineering Procedures (EN DES EPs), were established and l
were in effect from 1974 to mid-1985.
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SPECIAL PROGRAM REVISION NLMBER: 1 PAGE 7 0F 32 The EPs were quite detailed and prescriptive, and included many adninistrative controls in addition to addressing quality assurance requirements. The EPs that were reviewed in detail for this evaluation are listed in Appendix A, 5.u. l Also in use from early 1981 was the Nuclear Engineering Branch (NEB) Procedure NEB EP25.4.5 " Safety Evaluations -
Preparation and Handling" (App. A, 5.y).
In June 1985, new procedures, OEPs, became effective. The OEPs reviewed in detail for this evaluation are also listed in Appendix A, 5.v. The new OEPs reduced or eliminated most ,
l of the adninistrative details previously found in the EPs.
In place of a stand-alone procedure with many instructions, the OEP referred the reader directly to other procedures that controlled these individual activities. The end result was a series of single sentences that provided anyone unfamiliar with the process rather minimal direction on how to l accomplish a given activity.
In July 1986, the reorganized Division of Nuclear Engineering procedures, NEPs, became effective. The NEPs reviewed in detail for this evaluation are listed in Appendix A, 5.w.
These procedures closely resemble the content, fonnat, and brevity of the OEPs issued a year earlier, and require the active use of other referenced NEPs for full understanding of what is to be done. At this point NEP 2.1 replaced OEP-14 on the subject of licensing support. Also, NEP 3.1 is now used in place of the superseded NEB EP25.4.5 for preparation of fonnal TVA SERs.
The nine types of reports relating to the licensing process that were within the scope of Engineering responsibility were selected from the procedures and are addressed in this evaluation. The applicable procedures are listed in each section of the discussion. The docunentation packages that were reviewed for compliance with procedures and appropriateness of technical content are also listed. in Appendix A, 5.a through h.
The nine types of reports examined are:
o TVA Safety Evaluation Report (SER) o Significant Condition Report (SCR) o Failure Evaluation / Engineering Report (FE/ER) l 1064d (01/09/87)
TVA EMPLOYEE CONCERNS REPORT NLMBER: 207.2 (B) 2 SPECIAL PROGRAM REVISION NLMBER: 1 PAGE 8 0F 32 o Unimplemented Design Item Evaluation Report (UDIE) o Unreviewed Safety Question Detemination Report (USQD) o Preoperational Test Deficiency Report (PT) o NRC Violation Report o 10CFR50.55(e) Report
- o 10CFR50 Part 21 Report Each of the nine report types examined will be discussed in i the following sections and will include the following items unless noted otherwise.
- a. Purpose and Scope of Report
- b. Organization Responsible and Report Review Process l J
- c. Documentation Review Results and Comments l Since the concern lacked specifics, the evaluation team considered a random sample of ten docisnentation packages of each type approoriate to provide an adequate initial look at the previously conducted TVA activities. In the event anomalies were identified, expanded sampling was initiated to the extent necessary to properly assess the problem areas, i
9.4 TVA SAFETY EVALUATION REPORT (SER)
- a. Purpose and Scope of Report Procedural requirements are defined in EN DES-EP 3.03, NEB EP 25.4.5, OEP-7, and NEP 3.1.
l, l As described in procedure NEB-EP25.4.5, ( App. A, 5.y) l fomal safety evaluations produced by Engineering are prepared only by members of the NEB, and these evaluations shall apply to only those areas described in paragraph 2.1 (a through d) of the procedure, which are briefly sumarized below:
o Evaluation of plant conditions or deficiencies to infom Division of Nuclear Power of the i consequences o Ability of systems / subsystems / components to perfom safety functions o Evaluation of proposed changes as a part of a USQD l
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SPECIAL PROGRAM REVISION NLMBER: 1 PAGE 9 0F 32 These safety evaluations constitute either a design verification, an independent review, or an input into an operating plant decision.
Some other safety evaluations may be prepared by the various plants or others. In these cases, Engineering is contacted for support only when deemed necessary by the originators.
- b. Organization Responsible and Report Review Process Reports which NEB generates may be initiated by any l person in the NEB, such as a staff specialist or a group head, but the assigrunent of the person who is to investigate the problen and prepare the SER is handled by the Section Supervisor of the Safety Analyses Section. The review process follows procedures for calculations, wherein the calculation cover sheet and sign-off system are also applied to the SERs.
The previous SER procedure NEB-EP-25.4.5 (App. A, 5.y) is no longer an official TVA document. The decision to use the NEP 3.1 cover sheet and fonnat for future SERs was made by the NEB and concurred with by the other branches.
- c. Doctanentation Review Results and Comments Ten SER packages (App. A, 5.a) were reviewed covering a range of subjects including: stean valves; instrumentation and electrical components; an emergency diesel generator; and piping. Infonnation was obtained, tabulated and data developed as necessary to provide feedback to necessary organizations.
No abnonnal conditions were observed in the ten sets of documents reviewed that would cause the evaluation team to suspect that the TVA safety and licensing evaluations were inadequate, or appeared to cover up any fact or circumstance. Hereafter in this report, this thought will be abbreviated as: No anomalies were observed.
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SPECIAL PROGRAM REVISION NLMBER: 1 PAGE 10 0F 32 9.5 SIGNIFICANT CONDITION REPORT (SCR)
- a. Purpose and Scope of Report SCRs document any condition that renders an item unable to perfonn its required function, or creates uncertainty concerning its ability to meet design requirements.
Procedural requirements are defined in EN DES-EP 1.26, OEP-17, and NEP-9.1. For a Construction Pennit (CP)
Plant, a condition adverse to quality (CAQ) is considered to be significant when, if left uncorrected, any of the following conditions could result:
o Compromise of the reactor coolant pressure boundary integrity o Degradation of the ability to shut down and l maintain the reactor in safe shutdown condition o Degradation of the ability to prevent or mitigate accidents that could cause offsite exposures o Requirement for extensive redesign, repair, or rework o Substantial generic implications to other structures, systems, or components in the identified plant or other TVA plants o Falsification of records o Deliberate failure to follow procedures o Repeated occurrences which reasonably indicate an adverse progrannatic trend o Failure to comply with NRC requirements relating to a substantial safety hazard The above conditions have been abstracted from the noted engineering procedures.
During the period when OEP-17 was in effect, a CAQ on an OL nuclear plant was conservatively deemed as a significant quality program issue to ensure compliance with 10CFR50, Appendix B.
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SPECIAL PROGRAM REVISION NIMBER: 1 PAGE 11 0F 32
- b. Organization Responsible and Report Review Process The discussion in this section applies to Engineering generated SCRs only. The plants have their own procedures and contact Engineering for assistance only when they deem it necessary. Contact with Engineering would be through & Design Change Request (DCR).
It is the responsibility of anyone in the Office of Engineering (0E/DNE) who identifies what they perceive to be a significant condition to take the necessary steps to prepare, or have prepared, an SCR. This report is reviewed / verified by an imediate supervisor and forwarded to the appropriate branch / project staff and management review team for review and action.
- c. Doceentation Review Results and Comments Ten randomly selected Sequoyah SCRs (App. A, 5.b) were obtained from the records located in the Sequoyah jobsite engineering office files.
These reports covered subjects such as: correct materials and product qualification; proper product operation; and design and analyses of piping components / assemblies. The results of the evaluations
! were unifonnly and well docunented, and were signed off by appropriate ascending levels of technical and supervisory personnel.
Corrective actions for these cases included: use-as-is as a result of analyses and study results; replace or modify with proper or qualified material / equipment; and corrective action to be handled by another organization
( A study will be initiated and tracked to completion).
Various tracking systems have been used to track the l progress of the many and varied outstanding open items to assure their appropriate and timely close-out. The Corporate Comitment Tracking System (CCTS) is now used for this purpose if the docisnents address a commitment made to the NRC.
No anomalies were observed.
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TVA EMPLOYEE CONCERNS REPORT NLM8ER: 207.2 (8)
SPECIAL PROGRAM REVISION NLMBER: 1 PAGE 12 0F 32 9.6 FAILURE EVALUATION / ENGINEERING REPORT (FE/ER)
- a. Purpose and Scope of Report FE/ERs are used to document results of evaluations of significant deficient design conditions identified on an OE (or other equivalent) Nonconfonnance Report (NCR), or CAQ, that affect a safety-related structure, coliponent, or system.
Procedural requirements are defined in EN DES-EP 1.48, OEP-17 and NEP 9.1.
Typical examples include: errors; omissions; test failures; incorrect or inadequate doctnentation; and deviations from prescribed inspection or test procedures.
In the past the results of these evaluations provided '
the NEB Nuclear Licensing Staff with necessary infomation for their use in evaluating reportability to the NRC, and for detemining operational requirements regarding deficient design conditions for plants under construction. The NSL Site Licensing Manager now detemines reportability, while DNE continues to provide the evaluations.
- b. Organization Responsible and Report Review Process Upon the receipt of an NCR as described above, it is the responsibility of the Branch Chief (BC) or the Project Engineer (PE) to initiate the preparation of an FE/ER. The lead engineer, after assigning a preparer, advises appropriate jobsite personnel and provides a copy to the jobsite. After a failure evaluation has been conducted to detemine any affect on the system / structure /or component's capability to perfom its required design function, and it is documented, the lead engineer perfoms a final review for approval and signs the report. The report is then sent to Nuclear Lead Engineer for concurrence. The report is then given to the PE who reviews, signs and forwards it for distribution.
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T TVA EMPLOYEE CONCERNS REPORT NUMBER: 207.2 (B)
SPECIAL PROGRAM REVISION NUMBER: 1 PAGE 13 0F 32
- c. Documentation Review Results and Cannents Ten randomly selected Failure Evaluation / Engineering Report documentation packages (App. A, 5.c) were reviewed for general content, to determine that the process required by the governing procedure (NEP-9.1 or an older equivalent) had been followed, and to determine any evidence of a cover-up effort.
Report topics included inadequate wiring temperature ratings and cable tray supports, malfunctioning valve operators and instrumentation, other electrical equipment problems, and improper personnel qualification and engineering analyses.
Resolutions included: no safety impact /use as is; correct existing problems by replacement (replace as soon as parts become available); and replace all nonconfonning wiring so that it will meet the design requirements.
All of the reports reviewed were well evidenced as having been reviewed by appropriate technical and management personnel. The document contents stated the facts and included the corresponding corrective action /
problem resolution.
No anomalies were observed.
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. TVA EMPLOYEE CONCERNS REPORT NIMBER: 207.2 (B)
SPECIAL PROGRAM REVISION NLMBER: 1 PAGE 14 0F 32 9.7 UNIMPLEMENTED DESIGN ITEM EVALUATION REPORT (UDIE)
, a. Purpose and Scope of Report At a specified time during final phases of plant construction, it is required that unimplemented design items be evaluated. The purpose of these evaluations is to assure that any design items that have not yet been implemented would not put the plant into a configuration that would be unsafe for fuel loading.
These evaluations are not intended to verify the adequacy of design, construction, or the quality of the products provided by suppliers.
. Procedural requirements are defined in EN DES-EP 2.13, OEP-14, and NEP 2.1.
All outstanding design items are identified on an Outstanding Work Items List (0WIL) and are systematically evaluated. The results are doctriented in UDIE Reports,
- b. Organization Responsible and Report Review Process The PE assures the development of the 0WIL and the initiation of an UDIE report for each item that is not planned to be completed before fuel load.
The Lead Nuclear Engineer assures preparation, review, and approval of UDIE reports.
- c. Doctanent Review Results and Comments Ten randomly selected UDIE report packages
( App. A, 5.d) were reviewed for completeness, logical infonnation flow and analyses of each situation, and sign-off by appropriate TVA personnel.
Typical topics covered were: repair of solenoids; incomplete tagging and documentation; pipe relocation; weight changes and gap clearances; sewage capacity; and heat exchanger vibration.
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SPECIAL PROGRAM REVISION NtM8ER: 1 l PAE 15 7 32 l Many of the resolutions to the above situations did not have any impact on safety, but some required that action be taken before fuel loading, such as: repair solenoid; complete docmentation and tagging; and restrict specific heat exchanger operation to within certain limits.
No anomalies were observed.
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SPECIAL PROGRAM REVISION NIMBER: 1 PAGE 16 0F 32 9.8 UNREVIEWED SAFETY QUESTION DETERMINATION REPORT (USQD)
- a. Purpose and Scope of Report The NRC requires, through 10CFR50.59, that owners of a licensed nuclear power plant unit have a written safety evaluation that provides the basis for the detennination that a proposed change, test, or experiment does not involve an unreviewed safety question. Should one of these proposed items constitute an unreviewed safety question, NRC approval is required.
Procedural requirements are defined in EN DES-EP 2.03, OEP-14 and NEP 2.1.
The USQD report docunents the specific change, test or experiment. The following is excerpted from 10CFR 50.59:
"A proposed change, test, or experiment shall be deemed to involve an unreviewed safety question (1) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis l of any technical specification is reduced."
- b. Organization Responsible and Report Review Process The Project Engineer (PE) initiates proposed changes for USQD evaluation. The lead Nuclear Engineer perfonns the USQD evaluation for the PE's review and approval.
l l
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- c. Doctanentation Review Results and Coments Ten USQD doctsnents (App. A, 5.e) were randomly selected for review. Typical subject areas covered were: logic changes; alams and monitors; watch towers; concrete and ERCW system modification; polar crane design; and power supplies.
Not all USQDs were imediately dispositioned, in that several of them required evaluation programs to be conducted before conclusions could be reached. Other resolutions were clear-cut, such as: add a component; and detailed instructions were provided to perform a task.
No anomalies were observed.
- d. Some additional comments regarding USQDs should also be noted in this section. During investigative reviews for other Sequoyah Element Reports some items were noted relating to USQDs. NRC Violation Reports (App. A, 5.g) and External Audit Reports (App. A, 5.cc, dd, and ee) noted concerns relating to the timing of issuance of Formal USQDs that are part of the ECH process. The concern was that the USQD form was produced too early in the process with the potential that all pertinent data may not have been properly considered.
The appropriate change control procedures have been revised to require that for those items determined not be contain an unreviewed safety question, the fonnal USQD fom be provided as a final step before an ECN package goes to the Modifications Group for implementation, and that it be re-reviewed at the time of ECN closure. However, a draft USQD fonn will be prepared initially to provide early recognition to the various organizations that contact may be required with NRC. For Sequoyah, these items are being addressed under the Design Baseline and Verification Program. For those items detennined to contain an unreviewed safety question, NRC approval must be obtained before anything further is done.
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TVA EMPLOYEE CONCERNS REPORT NL4WER: 207.2 (B)
SPECIAL PROGRAM REVISION NLSWER: 1 PAGE 18 0F 32 9.9 PREOPERATIONAL TEST DEFICIENCY REPORT (PT)
- a. Purpose and Scope of Report The Preoperational Test Deficiency Report is used to document any significant deviations from established acceptance criteria for structures, systems and cort:ponents important to safety that are being preoperationally tested to determine their ability to accomplish their safety functions.
- b. Organization Responsible and Report Review Process The Plant Manager is ultimately responsible for all testing, but the Preoperational Test Program Coordinator provides direct supervision of work for the test section. A test director is appointed for each preoperational test. Test areas and responsibilities are explained in the TVA Nuclear Quality Assurance Manual (NQAM) Part II, Section 4.1, paragraph 5.2 (App. A, 5.r). Test procedures and responsibilities have previously been defined in the appropriate ID-QAP documents. These ID-CAP have now been incorporated into the NQMt.
Procedural requirements are defined in the following:
o EN DES-EP6.01 o EN DES Test Representative Handbook o NEB DI-125.03 o NQAM Part I, Section 2.11, ID-QAP-ll .1 o NQAM Part II, Section 4.1 o NQAM Part V, Section 11.1 (ID-QAP-11.1) o OEP-13 o NEP 10.1 through 10.4 o SQN AI-19, Part IV, (Site Procedure)
The completed testing instruction document, along with the data collected, is retained as a part of the historical plant records. Any deficiencies are documented by the test director. At the completion of each test, a field evaluation of the test results is perfomed by the test director and other Division of Nuclear Engineering (DNE) test personnel as necessary.
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TVA EMPLOYEE CONCERNS REPORT NLMBER: 207.2 (B)
SPECIAL PROGRAM REVISION NUMBER: 1 PAGE 19 0F 32 DNE, under the coordination of the Nuclear Engineering Branch (NEB) performs a comprehensive review and analysis of test deficiencies or deviations, in accordance with DNE procedures, to assure that all needed licensing and safety evaluation docunents are addressed.
The test data package is forwarded to DNE management for final review and approval. Upon completion of the DNE review and resolution of all test deficiencies, DNE certifies by memorandum that the product or system tested performed in accordance with the design requirements. The test data package, with memorandum (s), is then transmitted to the Plant Manager for final signature, approval, and permanent storage.
- c. Docunentation Review Results and Cmanents Ten Sequoyah Test Deficiency Report packages (App. A, 5.f) were reviewed covering a variety of testing areas including: vibration; alanns and switches; HVAC; wiring; and mechanical equipment.
Records indicated that detailed test procedures were prepared and reviewed before being used, and that detailed test results were collected, reviewed and approved by various levels of TVA supervisory and management personnel. Outstanding identified test deficiencies, as in the case of docunents PT-237 and PT-606, were subsequently tracked through a series of other docunents to their satisfactory resolution.
The test results docunents were prepared by test section personnel and reviewed by the test director, and after being signed, were forwarded to the Chief Nuclear Engineer for review. Staff review and subsequent signature signified management approval, after which the docunents were returned to the jobsite for permanent storage.
No anomalies were observed.
- d. It should be noted here that the subject of Preoperational Test Deficiencies is also addressed in other Sequoyah Element Reports. Report 207.3 addresses overall procedural issues, while Report 213.2 addresses technical issues in the electrical area.
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SPECIAL PROGRAM REVISION NUMBER: 1 PAGE 20 0F 32 9.10 NRC VIOLATION REPORT Responses to NRC Violation Reports are usually handled by the Plant Operations group, with support from Engineering when requested by the plant.
Reports are initially received by the licensing group, however,they are forwarded directly to the appropriate plant for action. The plant personnel detennine if Engineering input is required. For the two reports noted below Engineering input was requested. They are examples of Engineering participation in such matters.
Procedural requirements are defined in 0EP-14 and NEP 2.1.
The TVA response to two NRC Violation Reports
( App. A, 5.g), prepared as a result of an NRC inspection conducted between February 12 and March 14, 1986 at the TVA Knoxville facility and at the Sequoyah Nuclear Plant by NRC Region II personnel, was reviewed. This TVA response to the NRC Violation Reports was extremely detailed and responded specifically to each topic questioned by the NRC.
No anomalies were observed.
9.11 10CFR50.55(e) AND 10CFR PART 21 REPORTS Procedural requirements are defined in EN DES EP 2.02, OEP-14, and NEP 2.1.
l NEP 2.1, paragraph 4.7, as well as all prior procedures, addresses the preparation and handling of these reports. It provides specific guidance regarding report evaluation and detennination of reportability under both 10CFR50.55(e) and 10CFR21. It defines requirements for appropriate notifications and distribution of packages, coordinating report review with affected organizations and sending of approved report to the Nuclear Safety and Licensing (NSL) Group for submittal to NRC.
Three sets of these documents (App. A, 5.h) were reviewed for general content, and to verify that they were being unifonnly prepared and processed by knowledgeable technical and management personnel, and I
that they achieved an appropriate disposition.
No anomalies were observed.
l 1064d (01/09/87)
TVA EMPLOYEE CONCERNS REPORT NLMBER: 207.2 (B)
SPECIAL PROGRAM REVISION NIMBER: 1 PAGE 20 0F 32 9.10 NRC VIOLATION REPORT Responses to NRC Violation Reports are usually handled by the Plant Operations group, with support from Engineering when requested by the plant.
Reports are initially received by the licensing group, however,they are forwarded directly to the appropriate plant for action. The plant personnel detemine if Engineering input is required. For the two reports noted below Engineering input was requested. They are examples of Engineering participation in such matters.
Procedural requirements are defined in OEP-14 and NEP 2.1.
The TVA response to two NRC Violation Reports
( App. A, 5.g), prepared as a result of an NRC inspection conducted between February 12 and March 14, 1926 at the TVA Knoxville facility and at the Sequoyah Nuclear Plant by NRC Region II personnel, was reviewed. This TVA response to the NRC Violation Reports was extremely detailed and responded specifically to each topic questioned by the NRC.
No anomalies were observed.
9.11 10CFR50.55(e) AND 10CFR PART 21 REPORTS Procedural requirements are defined in EN DES EP 2.02, OEP-14, and NEP 2.1.
NEP 2.1, paragraph 4.7, as well as all prior procedures, addresses the preparation and handling of these reports. It provides specific guidance regarding report evaluation and detemination of reportability under both 10CFR50.55(e) and 10CFR21. It defines
! requirements for appropriate notifications and distribution of packages, coordinating report review with affected organizations and sending of approved report to the Nuclear Safety and Licensing (NSL) Croup for submittal to NRC.
Three sets of these documents (App. A, 5.h) were reviewed for general content, and to verify that they were being unifomly prepared and processed by i knowledgeable technical and management personnel, and that they achieved an appropriate disposition.
No anomalies were observed.
10Hd (01/09/87) l L
TVA EMPLOYEE CONCERNS REPORT NLMBER: 207.2 (B)
SPECIAL PROGRAM REVISION NLMBER: 1 PAGE 21 0F 32 Findings:
l
- a. There was no evidence obtained or observed that indicated inadequate perfomance by EN DES (including NEB) in the area of safety and/or licensing evaluations.
- b. There was no evioence obtained or observed that indicated any
) effort by EN DES (including NEB) to cover up or hide any infomation related to areas examined.
Conclusions:
Af ter reviewing the contents of the approximately 70 docuentation packages, the evaluation team concluded that the concern is not -
valid. Based on the docuents examined, the problems, concerns, and [
incidents that had occurred had baen well docuented and reviewed by (
successively more responsible and knowledgeable TVA technical, i supervisory and management personnel. These randely selected doc uentation packages were c mplete, legible, and adequate.
There was no evidence of a cover up by EN DES.
- 10. CORRECTIVE ACTION:
No corrective action is required.
v 1064d (01 /09/87)
I TVA EMPLOYEE CONCERNS REPORT NLMBER: 207.2 (B)
SPECIAL PROGRAM REVISION NLSEER: 1 PAGE 22 0F 32 APPENDIX A
- 5. DOClMENTS REVIEWED RELATED TO THE ELEMENT:
Ba5 851104 224 "Detemine proper solenoid engagement under specific accidents / data developed" B45 851219 220 " Containment sump level instrument accuracy / accuracies and levels provided" B45 660715 218 "Superheated steam in valve vaults -
line break long term cooling can be provided" NEB 800811 275 " Safety evaluation IE Bulletin 79-14 piping / safety impact - plant would not meet design basis" NEB 810917 250 " Pressurizer pressure during RHR event / data provided" NEB 820311 251 " Evaluate impact on plant safety due to toiler rupture - Aux.
Bldg./ evaluation prepared - no safety irppact on safe shutdown" NEB 820326 319 "SG level reference error correctional system as designed is acceptable" NEB 820412 252 " List equipment powered by Emergency Generator to mitigate 2 inch diameter location / tabulated date provided" NEB 820505 224 " Heat treat tracing safety related function / investigation perfomed and info tabulated" NEB 820804 226 " Provide equipment categorization -
NUREG 0588 for IE components / info tabulated and provided" 1064d (01/09/87)
. TVA EMPLOYEE CONCERNS REPORT NtMBER: 207.2 (B)
SPECIAL PROGRAM REVISION NLMBER: 1 PAGE 23 0F 32 APPENDIX A (Cont'd)
- b. Documents Reviewed for Section 9.5 - Significant Condition Reports (SCRs)
SQNMEB8201 R7 " Motor Operators W/ Class F Insulation
- Should be Class B, Replace with qualified operators" SQNMEB8206 " Failure of valve actuator yokes &
keys / install replacement yoke & key modification kits" SQNMEB8302 " Automatic closing valves / add closing valves and modify logic" SONMEB8403 R1 " Screen wash system - Class C or G/ corrective action to be handled by NUC PR" SQNMEB8410 "NAMC0 Limit Switches not qualified / Replace w/ qualified switches 10CFR50.49" SQNMEB8411 " Stress Corrosion of SS pipe w/ foam insulation; OK as-is, not evaluated temp" SONCEB8301 " Diesel Generating Building exhaust lines - Seismic adequacy / reanalyze for adequate support scheme" SONCEB8303 R1 " Analysis of piping problems w/o correct technique /do it over but w/ccrrect technique" SONCEB8404 R1 " Base plate flexibility and expansion anchors / design is more than adequate based on study" SONCEB8410 " Seismically Qualified Conduit Supports /TVA to search to determine if installed conduit is adequate" 1064d (01/09/87)
TVA EMPLOYEE CONCERNS REPORT NIDEER: 207.2 (B) l SPECIAL PROGRAM l REVISION hlMBER: 1 !
PAGE 24 0F 32 i
APPENDIX A (Cont'd)
- c. Documents Reviewed for Section 9.6 - Failure Evaluation / Engineering Reports B25 850408 001 " Inadequate wire temperature rating /all wiring will perfonn design function" B45 850625 258 " Baseplate flexibility & construction tolerances not considered for Anchor bolt loads / sampling program no impact on safe operations" SQP 841004 001 " Valve operation slowed due to seal swelling / stroke valves 3-4 times /2 weeks until seals are replaced" S0P 841019 002 " Electrical Cable tray risers not st!pported as specified/ quality existing risers, fix others, add supports as necessary" SCP 850109 010 " Radiation monitors mislocated or ID tags interchanged / correct wiring and nameplates verified" S01 860211 970 " Main Steam Line Break temperature increase invalidate equipment qualification / changed valves flow solenoid - to control" S01 860218 815 "No Engineering Evaluation to justify seismic supported Il systems /no
, failure or safety impact" l
l NEB 840515 253 " Spent Fuel Pool Cooling pump motors
/ NUREG 0588/use as-is, plan to replace in 15 years" NEB 840622 262 " Personnel Qualification /Tng/
certification - no safety impact" NEB 840705 282 "No alarm for low temperature /no safety impact" l 1064d (01/09/87)
TVA EMPLOYEE CONCERNS REPORT NUMBER: 207.2 (B)
SPECIAL PROGRAM REVISION NUMBER: 1 PAGE 25 0F 32 APPENDIX A (Cont'd)
- d. Documents Reviewed for Section 9.7 - Unimplemented Design Item Evaluation Report (UDIEs)
SWP 810401 090 " Tagging condensate demineralizer/not required" SWP 810406 052 " Repair solenoid Main Steam Header / repair before heatup following fuel load" SWP 810407 063 " Relocate connection - pipe /no safety impact," (ECN 2984)
SWP 810601 035 "Added sewerage plant capacity /no safety impact" SWP 810514 051 " Heat exchanger vibration / operation restriction" SWP 810616 015 " Apply red marking - concrete /N.R. no safety impact" SWP 810618 009 " Correct vent valve installation / add support" SWP 810622 082 " Correct pipe gap clearance /no safety impact," (ECN 3030)
SWP 810622 089 " Evaluate pipe weight changes due to actual insulation addition (vs) piping analyses input /no safety impact," (ECN L5271R1)
SWP 840504 019 " Complete documentation / accomplish prior to heat up" l
106dd (01/09/87) l
. TVA EMPLOYEE CONCERNS REPORT NLM ER: 207.2 (B) l SPECIAL PROGRAM REVISION NUMBER: 1 PAGE 26 0F 32 APPENDIX A (Cont'd)
- e. Documents Reviewed for Section 9.8 - Unreviewed Safety Question Determination Reports (USQDs)
SUP 810610 500 " Modification of ac IE ac power supply / detailed instruction provided" SWP 820513 536 " Suggested changes to drawings / schematics to operate as designed" SWP 830325 800 " Generator neutral transfonner resister and bus / transfer unit from Unit 2 to Unit 1" SWP 830408 818 " Logic change - to assure cooling water flow /added to logic of valves (xyz,etc.)"
PWP 830922 506 " Polar crane welded (vs) bolted keybar/better design authorized" PWP 831004 506 " Modify concrete and steel near main steam check valves /not safety related
- help maintenance effort" PWP 831104 529 " Add watchtowers to complement power concept of security improvement / study and cost benefit analyses provided" PWP 831209 503 " Radiation monitors and air flow switches remove and remount w/ shock absorber mt1/ listed - environmentally qualify sensing tubing" PWP 840307 538 " Waste disposal system high alarm / add an alarm - not safety related" B25 860815 504 " Modify ERCW system / add valve locking devices" 1064d (01/09/87)
TVA EMPLOYEE CONCERNS REPORT NUMBER: 207.2 (B)
SPECIAL PROGRAM REVISION NUMBER: 1 PAGE 27 0F 32 APPENDIX A (Cont'd)
- f. Documents Reviewed for Section 9.9 - Preoperational Test Deficiency Reports PT-221 " Pipe vibration - Support added," (ECN 3371)
PT-222 " Low level pump alarm - setpoint change," (ECN 2450)
PT-237 "High temperature of containment - lower compartment /rebalance lower compartment cooling system" PT-253 " Air Conditioner Air Supply Flow Rate / Adjust damper" PT-297 "Downscale Trip Annunciators testing /not necessary" PT-606 " Pressure switches /setpoint changes" PT-623 " Lack of current flow temperatures control / check for continuity and proper insulation" 2PT-596 " Wiring to drawing but no power / revise drawing" (ECN 2483) 2PT-642 " Damper springs / replace damper assembly," (ECN 3006) 2PT-652 " Ice weight limits / increase per Westinghouse letter" l
l l
\
l 106dd (01/09/87) l
TVA EMPLOYEE CONCERNS REPORT NUMBER: 207.2 (B)
SPECIAL PROGRAM REVISION NtMBER: 1 )
l PAGE 28 0F 32 APPENDIX A (Cont'd)
- g. Documents Reviewed for Section 9.10 - NRC Violation Reports 50-327/86-27 NRC Violation Report - Sequoyah Unit 1 50-328/86-27 NRC Violation Report - Sequoyah Unit 2
- h. Documents Reviewed for Section 9.11 - 1CCFR50.55(e) and 10CFR Part 21 Reports NEB 800124 254 " Diesel Generator Load Sequence Timer Error" NEB 8C0513 250 "Ruskir Fire Dampers" NEB 810416 288 " Environmental Qualification of 480V and Control Cables in EGTS Filter Rooms"
- i. Title 10 of the Code of Federal Regulations, Part 50 (10CFR50), Chapter 1, Apgendix B, " Quality Assurance Criteria For Nuclear Power Plants
- j. Title 10 of the Code of Federal Regulations, Part 21,
" Reporting of Defects and Noncompliance"
- k. Title 10 of the Code of Federal Regulations, Part 50.55,
" Conditions of Construction Permits"
- 1. Title 10 of the Code of Federal Regulations, Part 50.59,
" Changes, Tests, and Experiments"
- m. ANSI N45.2.ll - 1974, " Quality Assurance Requirements for the Design of Nuclear Power Plants"
- n. Regulatory Guide 1.64, " Quality Assurance Requirements for the Design of Nuclear Power Plants," R2, (06/76)
- o. Sequoyah Preliminary Safety Analysis Report, (PSAR)
- p. Sequoyah Final Safety Analysis Report (FSAR) Updated, through Amendment 3, (04/85)
- q. TVA, TR 75-1 A, " Quality Assurance Program Description for the Design, Construction, and Operation of TVA Nuclear Power Plants," R8 1064d (01/09/87)
TVA EMPLOYEE CONCERNS REPORT NUMBER: 207.2 (B)
SPECIAL PROGRAM REVISION NUMBER: 1 PAGE 29 0F 32 APPENDIX A (Cont'd)
- r. TVA, Nuclear Quality Assurance Manual, (NQAM), Parts I, II, and IV (11/14/85)
- s. Sequoyah Nuclear Plant Quality Assurance Manual (SON QAM)
Quality Assurance Procedures (QAPs) employed in the review of this element are from Revision 11 of the SQN QAM (and prior).
Project Manual
- u. Division of Engineering Design - Procedures (EN DES EPs)
EP 1.26, R9 "Nonconformances - Reporting and handling by EN DES" (03/15/85)
EP 1.48, R2 " Preparation of Failure Evaluations / Engineering Reports of Deficient Condition for Nuclear Plants," (03/15/85)
EP 2.02, R9 "Handlirg of Conditions Potentially Reportable Under Title 10 of the Code of Federal Regulations, Parts 21, 50.36 and 50.55(e)," (03/15/85)
EP 2.03, R6 "Unreviewed Safety Question Determination -
!!andling and Preparation," (04/24/84)
EP 2.13, R3 " Initial Fuel Loading Safety Evaluation -
Handling," (11/29/84)
EP 3.03, R8 " Design Calculations," (04/24/84)
EP 6.01, R5 "Preoperational Testing Documents -
Processing," (01/12/83) l Test Representative Handbook, R2 (04/83)
- v. Office of Engineering Procedures (0EPs)
OEP-7, R0 " Calculations," (04/26/85)
OEP-13, R0 " Testing," (04/26/85)
OEP-14, R1 " Licensing," (12/02/85) 0EP-17, R2 " Corrective Action," (08/30/85) 1064d (01/09/87)
TVA EMPLOYEE CONCERNS REPORT NUMBER: 207.2 (B)
SPECIAL PROGRAM REVISION NtMBER: 1 PAGE 30 0F 32 APPENDIX A (Cont'd)
- w. Nuclear Engineering Procedures (NEPs)
NEP 2.1, R0 " Licensing Support," (07/01/86)
NEP 3.1, R0 " Calculations," (07/01/86)
NEP 9.1, R0 " Corrective Action," (07/01/86)
NEP 10.1, R0 " Engineering Services Output," (07/01/86)
NEP 10.2, R0 " Control of Measuring and Test Equipment,"
(07/01/86)
NEP 10.3, R0 " Testing," (07/01/86)
NEP 10.4, R0 " Test Scoping Documents and Instructions,"
(07/01/86)
- x. NEB DI-125.03, R0, "Preoperational Testing Documents -
Processing" (07/01/85)
- y. NEB EP25.4.5, RO, " Safety Evaluations - Preparation and Handling," (03/19/81)
- z. SQN AI-19, Part IV, R10, " Plant Modifications After Licensing," (Site Procedure), (06/07/85) aa. Letter from B. J. Youngblood, NRC, to S. A. White, TVA,
" Concerns Regarding TVA Nuclear Programs" (02/18/86)
I bb. Letter from B. J. Youngblood, NRC, to S. A. White, TVA,
" Transcription of Interview of Dallas R. Hicks," (06/23/86) cc. Letter from M. J. Salerno, Impe11, to H. Jones, TVA, " Design Control Program, External Audit Finding Evaluation,"
(06/19/86) dd. Letter from M. J. Salerno, Impell, To H. Jones, TVA, " Design Control Program, External Audit Finding Evaluation,"
(860723 002) (07/10/86) ee. Letter from M. J. Salerno, Impe11, to H. Jones, TVA, " Design Control Program, External Audit Finding Evaluation," 07/23/86) 1064d (01/09/87)
TVA EMPLOYEE CONCERNS REPORT NUMBER: 207.2 (B)
SPECIAL PROGRAM REVISION NUMBER: 1 PAGE 310F 32 APPENDIX A (Cont'd)
- 6. WHAT REGULATIONS, LICENSING COMMITMENTS, DESIGN REQUIREMENTS, OR OTHER APPLY OR CONTROL IN THIS AREA?
- a. Title 10 of the Code of Federal Regulations, Part 50, Chapter 1, Appendix B, " Quality Assurance Criteria for Nuclear Power Plants"
- b. Title 10 of the Code of Federal Regulations, Part 50.59,
" Changes, Tests, and Experiments"
- c. Title 10 of the Code of Federal Regulations, Part 21,
" Reporting of Defects and Noncompliance"
- d. Title 10 of the Code of Federal Regulations, Part 50.55,
" Conditions of Construction Permits"
- e. ANSI N45.2.11 - 1974, " Quality Assurance Requirements for the Design of Nuclear Power Plants"
- f. Regulatory Guide 1.64, " Quality Assurance Requirements for the Design of Nuclear Power Plants," R2, (06/76)
- g. Sequoyah Final Safety Analysis Report (FSAR) Updated, through Amendment 3 (04/85)
- h. TVA, Nuclear Quality Assurance Manual, (NQAM), Parts I, II, and IV (11/14/85)
- 1. Nuclear Engineering Procedures (NEPs)
NEP 2.1, R0 " Licensing Support," (07/01/86)
NEP 3.1, R0 " Calculations," (07/01/86)
NEP 9.1, R0 " Corrective Action," (07/01/86)
NEP 10.1, R0 " Engineering Services Output," (07/01/86)
NEP 10.2, R0 " Control of Measuring and Test Equipment,"
(07/01/86)
NEP 10.3, R0 " Testing" (07/01/86)
NEP 10.4, R0 " Testing Scoping Documents and Instructions" (07/01/86) 1064d (01/09/87)
.~
TVA EMPLOYEE CONCERNS REPORT NUMBER: 207.2 (B)
SPECIAL PROGRAM REVISION NUMBER: 1 PAGE 32 0F 32 APPENDIX A (Cont'd)
- 7. LIST REQUESTS FOR INFORMATION, MEETINGS, TELEPHONE CALLS, AND OTHER DISCUSSIONS RELATED TO ELEMENT.
- a. Trip Report, by R. K. Dann, Bechtel, Knoxville, TN on 09/09/09/10, and 09/12/86, and, Sequoyah Site on 09/11/86, IOM #353, (10/29/86)
- b. RFI Number 672, (10/29/86), covering documents handcarried by R. K. Dann to Bechtel SFHO on 09/15/8E.
- c. Trip Report, Visit by R. Wolters to Knoxville, 10/21-10/23/86 and to SQN jobsite 10/24/86; IOM 439
- d. TTB No.150 (11/13/86) 1064d (01/09/87)
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.- , .: ,1 REFERENCE - ECPS120J-ECPS121C TENNESSEE VALLEY AUTHORITY PAGE -
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. -) ONP - ISSS - RHM '
EMPLOYEE CONCERN PROGRAM SYSTEM (ECPS) RUN DATE - 12/02/86 LIST OF EMPLOYEE CONCERN INFORMATION CATEGORY: EN DES PROCESS & OUTPUT SUBCATEGORY: 20702 SAFETY AND LICENSING EVALUATIONS
) S GENERIC KEYHORD A H APPL QTC/NSRS P KEYHORD B CONCERN SUB R PLT BBSH INVESTIGATION S CONCERN KEYNORD C CAT D LOC FLQB REPORT R DESCRIPTION KEYHORD D
) NUMBER CAT HI 100-027 EN 20702 N HBN YYYY SR TVA SAFETY AND LICENSING EVALUATIONS AUDIT T50212 REPORT BY ENDES (INCLUDING NEB) ARE INADEQ NONCONFORMANCE UATE AND APPEAR T00 MUCH IN A COVER GENERAL l UP MODE. CI HAS NO FURTHER INFORMAT CONST MANAGEMENT ION. ANONYMOUS CONCERN VIA LETTER.
1 CONCERNS FOR CATEGORY EN SUBCATEGORY 20702
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