IR 05000293/1987032
| ML20236T579 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 11/06/1987 |
| From: | Blumberg N, Caphton D, Hunter J, Thomas Koshy, Prell J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236T524 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.18, TASK-TM 50-293-87-32, NUDOCS 8712010371 | |
| Download: ML20236T579 (37) | |
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U.S. NUCLEAR REGULATORY COMMISSION
. REGION I Report N /87-32 Docket No.' 50-293 License N DPR-35 Licensee: Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199 Facility Name: Pilgrim Nuclear Power Station, Boston Edison Engineering Offices, and Pilgrim Training Center Inspection At: Plymouth, Braintree, and Chiltonville, Massachusetts Inspection Conducted: August 3-7, 10-14 and 20, 1987 Inspectors: u#
D - CaphYoh, Senior Technical Reviewer
- / datef//7 I d//
rell, RFhetor Engineer
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T. Koshy, RMictor Engineer
/M47 date Rf ? 7 yHunter,ReactorEngineer dat:e Approved by: -
- ll i ~O j h.Blumberg,Chieff0perationalProg[ams date Section (DRS), Team Leader Inspection Summary: Special, Announced Team Inspection on August 3-7, 10-14, and 20, 1987, Report No. 50-293/87-3 Areas Inspected: Special, announced inspection of five modifications in progress, design concept, design engineering process, 10 CFR 50.59 process, modification procurement process, post modification testing, system restoration, post modification drawing and procedure updating, post-modification training, post-modification administrative controls, QA/QC interfaces, environmental qualification concerns, and licensee action on previous inspection finding Results: One violation was identified: Failure to qualify equipment in con- j tainment to highest potential accident temperature profil Paragraph 6. In )
addition, a concern over the loss of QA engineers from the QA audit group was i discussed in Paragraph !
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Table of Contents
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Page Table of Contents . . . . . . . . . . . . . . . . . . . . . . . . 2 1.0 Introduction and Inspection Overview . . . . . . . . . . . . 3 2.0 Inspection Criteria . ... .............. . 5 3.0 Design / Modification Processes ............... 7 3.1 Engineering / Design Process .............. 7 3.2 10 CFR 50.59 Safety Evaluation Proces ........ 8 3.3 Procurement for Modifications . . . . . . . . . . . . . 9 3.4 Modification Post Installation Testing. . . . . . . . 10 3.5 Restoration Process . . . . . . . . . . . . .... . 11 3.6 Post Modification Drawing and Procedure Update . . . 11 ;
3.7 Post Modification Training Process . . . . . . . . . 12 3.8 Administrative Controls for Modifications . . . . . . 12 l 4.0 Modifications Reviewed . . . . . . . . . . . . . . . . . . 13 l
4.1 Modification 85-07, Reactor Vessel Water Level Instrumentation Modification . ...... . . 13 4.2 Modification 86-51, Direct Torus Vent Installation . . . . . . . . . . . . . . . . . . 16
4.3 Modification 86-52A, Replacement of Containment Spray Header Caps / Nozzles . . . . . . . . 18 4.4 Modification 87-30, ATWS Recirculation Pump Trip . ......... . . . . . 21 4.5 Modification 86-73, Automatic Blowdown System Logic . .... ..........23 5.0 QA/QC Interfaces . . . ... ......... .....25 6.0 Environmental Qualification Findings . ....... . . 28 7.0 Licensee Actions on Previous Inspection Findings . . . . . 30 8.0 Unresolved Items ..... . . . . . . . . . . . . . . . 32 9.0 Management Meeting . . . . . . . . . . . . . . . . . . . 32 Attachment A Persons Contacted . . . . . . . . . . . . . . . . . .A-1 Attachment B Documents Reviewed . . . . . . . . . . . . . . . . B-1 l
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DETAILS 1.0 Introduction and Inspection Overview NRC Region I performed a team inspection of the licensee's engineering and modification process from August 3-14, 1987, exclusive of weekend dates of August 8-9. The team consisted of four inspectors and a supervisor as team leader. One team member participated in the inspection only from the dates of August 3-7, 1987. Inspection activities were performed at the licensee's Pilgrim Station, Nuclear Engineering Department (NED) offices at Braintree, Massachusetts, and training center at Chiltonville, Massachusetts. The overall period of the inspection was extended to August 20, 1987 to include a final exit meeting conducted with the licensee on that dat The inspection covered on a sampling basis, all aspects of the design ,
process from design concept to final turnover for operation by plant {
personnel. To accomplish this, the following four modifications and
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portion of a fifth modification were specifically selected for review: -
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-- 85-07 REACTOR WATER LEVEL INSTRUMENTATION MODIFICATION - Replacement of the existing Yarway heated reference columns in the drywell with new cold reference legs outside the drywell to reduce level indication errors caused by high drywell temperature DIRECT TOROUS VENT - This modification adds an 8" torus vent line connecting the torus to the main stack which will provide a !
flow path for relieving primary containment pressure to prevent ,
overpressurization and subsequent failure of the primary :
containment during a severe accident. The vent line is being ;
installed but will be blanked off until a final safety l evaluation and decision can be made on the use of this vent !
pat A REPLACEMENT OF CONTAINMENT SPRAY HEADER CAPS /N0ZZLES - The !
modification includes replacement of 104 upper and 104 lower containment spray header nozzles. The replacement nozzles are '
identical to the existing nozzles except that the replacement i fogjet nozzles have one open spray cap and six blanked off spray caps, whereas the existing fogjet nozzles have all seven caps open. The reduced spray flow will decrease the possibility of damaging the containment structure by sudden depressurizatio ATWS RECIRC PUMP TRIP MODIFICATION - This modification provides an ATWS trip of the recirc pump M/G drive motor breaker to increase the reliability of RPT-ATWS. The current design trips the recirculation M/G field breaker after receipt of high reactor pressure or low reactor water level. The drive motor trip will use the existing initiating signals for the field i breaker trip, without the current time delay (10 sec.) before initiation of the pump trip on low reactor water leve _ _ _ - _ - - . I
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-- 86-73 AUTOMATIC DEPRESSURIZATION/ BLOWDOWN SYSTEM LOGIC - This modification includes a timed bypass (11 min.) of the high dyrwell pressure initiation signal and a manual inhibit of existing ADS actuation logic to provide automatic ADS initiation, if required, for events such as a break external to the drywell or a stuck open SRV. Manual inhibit switches allow the operator to inhibit ADS operation without repeatedly pressing the reset pushbuttons. Review of this modification consisted solely of witnessing some post modification testin The inspectors reviewed on a sampling basis, each modification package :
to determine the adequacy of implementation. The following areas were reviewed for each modification: Design concept, engineering, 10 CFR 50.59 safety evaluations, procurement, engineering, design and plant organiza- i tion interfaces, installation, post-modification testing, system restora-tion, procedure and drawing updates, post-modifications training, QA/QC interfaces with the modification process and the overall administrative controls. In addition to reviewing the above areas in context of the specific modification, each area was reviewed for general application to all modification Each modification inspected (and many others) were still in process. All, except one, had been installed although the total process had not been completed. In each instance the inspectors attempted to determine that l for portions not yet started the process, procedures and instructions were l sufficiently detailed and in place to provide reasonable assurance that l-items like post-modification testing, procedures and drawing updating, system turnover, etc. would be satisfactorily accomplishe Basic modification engineering and design are performed by the Nuclear Engineering Department (NED) located at Braintree. In many cases specific designs and engineering calculations are contracted to outside organiza-tions and overseen by NED Engineers. Performance of the modification at the site is the responsibility of the Nuclear Operations Department (NOD)
which is headed by the Station Manager. Modifications are installed under the cognizance of the Maintenance Group and Construction Management Group (CMG). However, most modificat ion work is actually performed by an on-site staff of contractor personnel or in some cases by outside contractor A Modification Management Group, also part of N0D, oversees post modifica-tion testing and the post modification review and completion. The Nuclear training Section reports to the Vice President Nuclear Operations and i evaluates the training required for each modification. It is their re-l sponsibility to assure that all appropriate personnel, including reactor operators, are familiar with the modification before it is placed in operation The Training Center is located at Chiltonv111e, Massachusett _ _ _ - _ - _ - _ _ _ _
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The Quality Assurance Department is independent of both NED and N00 and reports to the Vice President of Nuclear Engineering and Quality Assur-anc The Quality Assurance Department is headquartered at the Braintree offices where the QA audit and quality engineering groups are also located. Personnel who perform QA surveillance and QC inspections are located at Pilgrim Station. Modifications are subject to QA audits, surveillance and QC inspection The results of this inspection indicate that the licensee's modification process is acceptable. The modification process assures that the modifi-cations will be properly designed, installed and teste The post modi-fication process provides assurance that testing, training and procedure and drawing updates will be accomplished in a timely manner. The various organizations examined interfaced well, and there was no indication that appropriate license organizations were unaware of modifications in progres There was one environmental qualificatP:n finding regarding the operabi-lity of equipment located in the drywell against the highest temperature that equipment might experience during some accident situations. When EQ considerations were evaluated for the containment Spray modification, previous licensee positions on EQ were found to have been incorrect. (See Paragraph 6.)
2.0 Inspection Criteria 2.1 Regulatory Criteria The inspection cot sisted of detailed review of four specifically selected modificet n packages and the modification / engineering process for conformance to the following regulatory requirements:
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10 CFR 50, Appendix B Quality Assurance Criteria for Nuclear Power Plant CFR 50.59 Changes, Tests and Experi-ments (Evaluation for Unreviewed Safety Questions).
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10 CFR 50.49 Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plant Regulatory Guide Personnel Selection Revision 1-R, 1975 and Training (ANSI N18.1-1971/ANS3.1)
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Regulatory Guide 1.33, QA Program Requirements Revision 2, 1978 (Operational)
(ANSI N18.7-1976/ANS 3.2)
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Regulatory Guide 1.64 QA Program Requirements Revision 1, 1975 for the Design of Nuclear (ANSI N45.2.11-1974) Power Plants
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Regulatory Guide 1.123 QA Requirements for Control Rev. 2, 1976 of Procurement of Equipment, (ANSI N45.2.13-1976) Materials, and Services for Nuclear Power Plants
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Regulatory Guide 1.144 QA program Auditing Rev. 1, 1980 Requirements for (ANSI N45.2.12-1977) Nuclear Power Plants
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Regulatory Guide 1.58 Qualification of Nuclear Rev. 1 (9/80) Power Plant Inspection, (ANSI N45.2.6-1978 and Examination and Testing ANSI-SNT-TC-1A-1975) Personnel 2.2 Areas Reviewed Nuclear Operations Department Procedures, Nuclear Operations !
Procedures, Nuclear Engineering Department Procedures, Nuclear Engineering Department Work Instructions, Technical Quality Memorandums, Training Section Procedures, Quality Assurance Procedures, Test Procedures, Drawings, Design Packages, and other documents were reviewed on sampling basis for conformance to the regulatory requirements, regulatory guides, and ANSI Standards listed in paragraph 2.1 above. In addition, discussions were ,
conducted with design engineers, technicians, operators, training j instructors, test personnel, managers, QA auditors, QC inspectors, j contractor and construction personnel as to their involvement with specific modifications and the general engineering / modification proces Inspections were conducted to review the following areas:
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Design packages were complete and comprehensive
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Administrative controls were in place to control the engineering, design, and modification processes
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All procedures were properly reviewed and approved
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Modifications were installed in accordance to design drawings 1
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A design change process was in use
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Proper 10 CFR 50.59 safety evaluations for unreviewed safety 1 questions were performed I i
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Procurement system provided for proper purchase of Q and EQ items
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Systems were properly restored
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Modifications were adequately tested
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Post modification test procedures were technically adequate j
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Where applicable, training on modifications was given to plant I operators
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QA and QC were involved and adequately reviewed and audited the design and modification processes
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Personnel performing engineering, installation, training, testing, and inspection tasks were knowledgeable '
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Environmental qualification for accident conditions were considere !
2.3 Persons Contacted and Documents Reviewed Persons contacted are listed in Attachment A to this repor Procedures, drawings, design packages and other licensee documents reviewed during this inspection are listed in Attachment .0 Design / Modification Processes 3.1 Engineering / Design Process The Nuclear Organization Procedure NOP83E1, " Control of Modifications To Pilgrim Station," is the governing procedure that establishes interdepartmental responsibilities for performing modifications to the Pilgrim Station. This procedure addresses from the scope and requirements of the modification to documentation, implementation and acceptance of the modification. The detailed instructions for generating the complete design package are covered under Nuclear Engineering Department Procedure 3.02, " Preparation, Review, Veri-fication, Approval and Revision of Design Documents for Plant Design Changes." The Nuclear Engineering Manager is responsible for imple-mentation of this procedur The assigned discipline engineer is responsible for developing design J input / criteria. The preliminary approval for the modification is I done in the form of an engineering service request for scope and I justification approval and may include conceptual design document l The cognizant engineer may then develop the conceptual design, if necessary, or proceed to the detailed design. The prescriptive steps are different for major modification's and minor plant design changes. All safety related and fire protection modifications are I
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subjected to an independent review. The reviewer uses a twenty step questionnaire to assure the technical adequacy of the design change. The licensee has adopted the design input requirements from ANSI N45.2.11 - 1974, and has developed a matrix which assigns departmental responsibility for generating the inputs. The appen-dixes to the procedure provide prescriptive instructions for con-structability review, including ALARA consideration, materials and equipment, plant impact on design and procedures. Plant design change packages contain a narrative overview section which discusses and provide an overview of the problem, design change description, safety classification, procurement requirements, analysis of design adequacy, installation instructions, acceptance criteria and tech- !
nical specification changes. The team had no significant concerns in the modification process and no violations were identifie .2 10 CFR F0.59 Process The Nuclear Engineering Department procedure 3.07, "10 CFR 50.59 Safety Evaluation" establishes responsibilies, review guidelines and the control of the safety evaluation The NED group which performs safety evaluations for modifications consists of 12 Engineers with an ;
average of ten years nuclear experience. Two members of this group '
have an operational background, one as senior reactor operator and the other as a reactor operator. The licensee has provided control room simulator training for selected individuals in this group. The group contained specialists in heat transfer, health physics and other engineering disciplines. The inspector reviewed the training records of the department engineers. The department training program involves completing a study program where the supervisor has assigned the required student level of knowledge for documents such as the FSAR; Emergency Plan; Nuclear Station procedures; Technical Specifi-cation; selected industry standards; 10 CFR 20, 21, 50.36, 50.47, 50.48, 50.55a, 50.59, Part 50 Appendixes A, B, and C; 10 CFR 100 Engineering Specifications; and some selected regulatory guides. The j group members read all the NRC Information Notices and Bulletins,
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generic letters, the published data on NRC violations and the owners group correspondence to remain knowledgeable of current safety issue Procedure 3.07 contains detailed instructions on preparing a safety evaluation. Safety evaluation inputs are received from other departments when special expertise is required for evaluating modifications. These inputs are processed in the format suggested in procedure 3.07. The guidelines section, an attachment to this procedure, provides descriptive sections, with examples, helpful in ;
performing adequate safety evaluations. Within the sections are '
descriptions of the change, purpose of the change, systems, subsys-tems and components affected, including directly and indirectly affected components, safety functions of affected components, effect on safety functions and an analysis section. In the analysis section, each safety function is analyzed against inherent
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properties, expected values, in process parameters, previous experience on similar designs, etc. to conclude that there is no potential unreviewed safety question. The completed evaluation package is then presented to the Operations Review Committee (0RC)
for final approva The inspector reviewed the following safety evaluations:
-- 2161 - Raising of RCIC turbine exhaust pressure from 25 psig to 46 psi ATWS auto trip on high reactor pressure at 1175 or low reactor level -46 inche Increase of fuse size from 0.75A to 1.5A for IRM power suppl These evaluations contained sufficient analysis and supporting facts to conclude that there are no unreviewed safety questions. The inspector had no further question No violations or deficiencies were identifie .3 Procurement Process The licensee has established a formalized program governed by NED procedures 4.01, 4.05 and 6.03 for identifying and controlling parts used in safety-related systems or in systems required to operate in adverse environments. All Plant Design Changes (PDCs) are reviewed and approved by a Safety and Systems Analysis (S&SA) group in the Nuclear Engineering Department. S&SA verifies that all parts of the modification have been properly classified as a non-safety related item, a safety related item, a commercial quality item, or an environ-mentally qualified item. In addition to the S&SA group review, the I&C and Power System group review all modification packages to iden-tify items which must be environmentally qualified (EQ). After construction of the modification, the Nuclear Engineering Department (NED) performs walkdowns of the modification to verify all EQ requirements have been satisfie A program has been established for controlling stock replacement of safety related items (Q), commercial quality items (CQI), and environmentally qualified EQ items. This program required NED review to determine if replacement items are Q, CQI, or EQ items. The Q-List lists Q and CQI item All CQI items ordered on purchase orders are identified with a l specific PDC or system. Therefore,-if a 10 CFR Part 21 concern arises for a particular part, that part can be traced to specific systems. S&SA reviews identifies and approves all Stock Material Authorizations prepared for reorders of CQI _ _ - _ _ _ -
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The inspector reviewed the appropriate administrative and department level procurement procedures which, are listed in Attachment 8 and the Licensee's Q-list and held discussions with both engineering and operational personnel regarding their procurement program. At the time of this inspection, the Q-List was being revised to identify items down to the component level. In addition, a detailed review of the items procured under PDC 87-30 was made and it was verified all procurement review controls had been properly administere No violations or other deficiencies were observe .4 Modification Post Installation Testing The Nuclear Engineering Department (NED) procedures 3.02, and 3.06 specify that testing and acceptance criteria will be identified in the design output documentation for each Design Change Notice (DCN)
and Field Revision Notice (FRN).
The Modification Management Group (MMG) located at the plcnt receives testing inputs from NED via a form " Exhibit 3.02Y" for PDCs and FRN The MMG assigns a senior test director for each modification and the test director has the assigned responsibility to prepare or have test procedures prepared to satisfy requirements and criteria specified by NE The MMG has responsibility for tracking the testing and verify-ing the resolution of test deficiencies through completion of the modification testin Post installation testing and operational turnover are performed through Nuclear Operations Procedure (NOP) 83A6, " Modification 4 Management." The cognizant engineer prepares the detailed design package and forwards it to the Modification Management Group. This package is then transmitted to all the designated groups including operations, maintenance, quality assurance, ALARA, Technical Speci- i fication review, etc., for filling out their respective checklists )
and developing plans for the completion of the work items associated with the PDC. The checklists and the comments received are then factored by MMG into the preoperational (post modification) test procedure. The test procedure is verified within the MMG and then forwarded to the cognizant departments for review. This activity is done using procedure 1.3,4, "The Procedure for Preparation and Control of Procedures." Once approved, the test procedure governs the post-modification testing. No test exceptions are taken during performance of the test. Although this approach slows down the test process, it has substantial technical merit In order to resolve the problems encountered during the test, the test procedure is revised and subjected to the same level of review it received ini-tiall It is then reissued for testin The final test data is forwarded to the test director. The test director reviews the test results and resolves the discrepancies, if an No violations or deficiencies were identifie _ _ - - - - _ _ _ _ - . .
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3.5 Restoration Process The Modification Management Group is primarily responsibic for assuring all phases of the modification installation are completed before recommending to the Nuclear Operation Manager (NOM) that the station accept the modification for operational use. The MMG has developed a list of organizations who get copies and review the detailed modification package. Each organization sends to the MMG an organization review form which identifies how the modification impacts their organization (E.G. drawing or procedure update, train-ing, tech spec changes, construction, and pre-op test procedures).
Based on these replies the MMG prepare a simplified impact statement which.is used to help track the status of the design implementatio Test Directors from the MMG are assigned to follow specific PDCs and to develop pre-operational tests. Each month, MMG sends to the NOM and to all departments status update sheets which summarize the status of all PDC's. When the MMG receives memos from all the affected organizations stating the procedural / drawing updates, train-ing, etc. are completed, including pre-op test procedures, then the MMG reviews the package for completeness and forwards to ORC and NOM for their approva The Modifications Management Manual, dated July 30, 1987, was reviewed to verify that the program was clearly delineated. A review of several PDC packages was made to verify the organizational review input statements, impact statements, and monthly status reports had been completed. Discussions were held with the group leader, the MMG coordinator and several senior test directors. A review of the organization chart was made including a sample of test directors resumes and qualifications sheets. Based on this review the in-spector determined that the PDC restoration process was well con-trolled and documente No violations or deficiencies were identifie .6 Drawing and Procedure Updates for Modification NED procedure 6.02 provides the methods utilized to effect control and update of drawings related to the modification process. NED procedure Exhibit 3.02Y inputs to MMG a listing of safety related drawings and procedures that will be affected by a modification. MMG completes an impact statement for each modification and tracks pro-cedures and drawings until the necessary changes and/or additions are completed. Procedure 6.02 provides a mechanism to assure and verify that up-to-date drawings and procedures are placed in use, for ex-ample, in the control room. Based upon selective samples taken, administrative and management controls exist to assure that drawings and procedures relating to modifications are revised, as necessary, and placed in use to meet regulatory commitment No violations or deficiencies were identifie _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ - - - _ _ _ _ _ -
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3.7 Post-Modification Training Process Once a modification has been accepted by the station, the Modifi-cation Management Group (MMG) sends a copy of the modification package to the Training Department. The Training Department performs both an operations and technical review of the modification and determines what training requirements are needed. The operation training group and the technical training group, both list their identified training requirements and send them to the MMG. Training modules and material are then developed or gathered and training conducted and documented during the construction phase of the i modification. Upon completion of the training, the MMG is notified l
by the Training Departmen The Training Department has procedures (See Attachment B) wnich define their responsibilities regarding PDC training, the development of training modules and material, and the documentation of training given. The operational and technical training modules for the fol-lowing PDCs which were the subject of this inspection were reviewed and found to be detailed and thorough: PDC 85-07, Reactor Water Level Instrumentation Modification; PDC 86-51, Direct Torus Vent; PDC 86-52, Automatic Depressurization/ Blowdown System Logic; PDC 86-73, Replace-ment of Containment Spray Header Caps / Nozzles; and PDC 87-30, ATWS Recirculation Pump Trip Modification. Most of the training modules were complete and training started even though the modifications had not been fully installe Of 183 PDC's being installed during refueling outage No. 7, approximately 91 were determined net to require training, 45 had training completed, 17 were in the process of module development and the rest were in the process of presenta-tion, walkdown or exa The Training Department uses an automated system for tracking all required training. The Training Department is also establishing an internal audit program to review those PDCs that were determined not to require training in order to verify that the correct decision had been made. The inspector reviewed the results of the first audit conducted. This audit determined that several modifications originally thought to be "not applicable" will require some limited training. The inspector did not identify any instances when required training was not accomplished or identifie No violations or deficiencies were identifie .8 Administrative Controls for Modifications The upper tier of procedures which control both the Nuclear Engineer-ing Department (NED) activities and the station or Nuclear Operations Department (N0D) activities are the Nuclear Organization Procedures (NDPs). NOP procedures were identified and reviewed to verify they adequately addressed all regulatory and ANSI requirements related to programmatic design and modification contro _ _ _ _ _
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3.8'.1 Nuclear Engineering Department (NED) Procedures The NED has developed procedures which govern all phases of.the design / modification process and those reviewed meet all ANSI requirement These NED procedures are supplemented with NED Work Instructions which provide further clarification and/or details regarding various aspects of the design process. All procedures appeared well controlled and the engineers questioned appeared knowledge-able of their requirement The procedures reviewed were detailed and in most cases presided data sheets for controlling the design process (see Attachment B).
.3.8.2 Nuclear Operations Department (NOD) Procedures
' All station procedures are identified as NOD procedures. Those N0D procedures related to procurement, modification control discrepancy reporting, plant design changes and training were reviewed for adequacy and completeness. The design / modification program was procedurally documented and found to meet all regulatory concer .8.3 Findings Selected design / modification related procedures were reviewed for technical and administrative adequacy. Procedures were l found to adequately cover all aspects of the program and were well controlle No violations or deficiencies were identifie .0 Individual Modifications Reviewed 4.1 Plant Design Change (PDC) No. 85-07, Reactor Water Level Instrumentation Modification 4.1.1 Description of Modification Generic Letter 84-23 dated October 26, 1984 concluded that improvements to plants should be made to reduce reactor water level indication errors caused by high drywell temperature These improvements include prevention of reference leg over-heating or reduction of the vertical (piping) drops in the drywell. The purpose of the Pilgrim modification was to comply with the NRC staffs Generic Letter-to improve reactor water level sensing. The modification removed the existing Yarway l I
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water level heated reference columns in the drywell, and instal-led new cold reference columns outside the drywell. To obtain minimum vertical piping line drops inside the drywell, two new primary containment vessel liner penetrations were installed at approximately the 82 foot elevation to accommodate new reference column piping runs. The modification reduced the reference column vertical drop inside the drywell to approximately one foot, a reduction of approximately 11 feet from the previous t installation. The new piping is installed and supported to Seismic Class I requirement The modification included cutting and capping of pipe, removal of the existing Yarway heated reference columns, removal of l thermocouple and cable, and replacement of the local indicating l Yarway transmitters. A separate PDC (84-70) provided for new
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Rosemount transmitters to be installed but was not addressed by this PD At the time of this inspection, the installation of piping had been completed, with the exception of installation of the pipe insulation in the drywell. Also, radiation shielding had not been installed at the two new drywell penetrations, and, pressure testing of the piping remained to be completed. The biological shielding at the reactor vessel nozzles was scheduled to be installed after the pressure testing of the pipin The modification package will be finalized, including a final rewrite of the safety evaluation, before restart of the reacto Open items for this modification were being tracked by the license .1.2 Modification Field Inspection The piping runs both inside and outside the dry well were selectively inspected to verify that the installed configuration conformed to the installation drawing Slopes of piping runs, piping supports and hanger locations, hanger settings, piping clearances, condensing pot locations, valve and orifice loca-tions, and radiation shie'. ding at penetrations were visually inspected. Welds for the newly installed piping were visually examined. The welding procedures, procedure qualifications, welder qualifications and NDE findings for selected welds were inspecte Both cognizant contractor and BEC0 employees were questioned regarding the installation wor The drywell vessel penetration was verified against ASME Section III 1980 Edition threugh the Winter 1980 addenca Boiler and Pressure Vessel Code requirement . _ _ _ - _ _ _ _ _ - _ .
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In addition, the inspector reviewed the installation of Rosemount Transmitters covered under modification 84-7 Installation drawing M-8372, Revision D, " Analog trip system modification Rosemount Model 1153, Series B Transmitter" and the work instructions contained the torque requirements for mounting bolts, electrical termination and the side covers. The inspector reviewed four selected installation records and confirmed compliance to the instructions and adequacy of QC coverag The field inspection of the instruments did not identify any discrepancie .1.3 Modification _ Testing .
Modification management issued a testing status memorandum to the NOM (Nuclear Operations Manager) on August 6, 1987, in which the intended pre-op testing was listed. The following items were listed:
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Recalibration of the associated instrumentatio Hydro pressure test of the new piping; l
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Inspection of welds that could not be isolated (during the l' vessel hydro);
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The excess flow check (valve) procedure (performed during vessel hydro);
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A procedure to verify all the water level indications l (Feedwater and Reactor Protection System) read the same l when reactor level is in the normal range; and,
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A flush test to verify new sensing lines are free l and clea In addition to the above and resulting from questions raised l during this inspection, a office memorandum was written from the NOM to the NED manager on August 10, 1987, requesting that design verification testing be specified for confirmation of the radiation shield design for PDC 85-07.
l Hydro testing during construction of a portion of the new l piping has been completed. The remaining hydro testing of piping adjacent to the reactor vessel is scheduled to be completed during the reactor vessel hydro. The senior test director stated that the only test procedure completed was a procedure for recalibration of the associated instrumentatio .1.4 Findings The piping and equipment was found to be installed in accordance with a selected sample of installation drawings. Piping insula-tion and penetration shielding inside the drywell were not installed at the time of this inspection due to the scheduled hydro pressure test for the piping. The hanger drawing did not
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specify whether or not the piping hangers were to be set with the piping full or empty. The contractor stated that the hangers had been set with the piping empty and the design engineer stated that for the small pipe size the large hanger tolerances provided were adequate to cover the small weight involved, whether full or empty. The gamma shielding at the new penetrations was not installed; however, it was scheduled to be installed. The neutron shielding inside the new drywell pene-tration was noted to contain designed clearances between the neutron shielding and the sensing line pipe where it exited the penetration on the exterior wall of the drywel The amount of this clearance provided a potential path for neutron streaming through the penetration. This concern was discussed further with the cognizant design engineers and a memo from K. P. Roberts to R. N. Swanson dated August 10, 1987, was written to request testing to confirm the adequacy of the shielding desig No violations were identifie .2 PDC 86-51 Direct Torus Vent Modification 4.2.1 Description of Modification and Status The purpose of this modification was to address a severe ac-cident concern - overpressurization and subsequent failure of the primary containment during a severe accident. The process of venting the containment directly is part of the overall actions included in the BWR owners Group Emergency Procedure Guidelines (EPG) Revision 4 and is being considered for the plant specific Emergency Operating Procedures (EOPs). Such venting will not become part of the E0Ps unless a new safety evaluation is performed and NRC approval is obtaine The modification adds an 8 inch torus vent line connecting the torus to the main stack which will provide a flow path for relieving primary containment pressure to prevent containment failure. Additional work includes the installation of a butter-fly valve and rupture disc in the 8 inch vent line, replacement of a portion of the 20 inch ductwork of the Standby Gas Treat-ment System (SGTS) with 20 inch safety class piping and field runs of 1 inch nitrogen piping to supply backup nitrogen to the air operated butterfly valves associated with the vent lin The licensee is in the process of completing the modification; however, the licensee stated that the system is blanked off and will not be declared operable and ready for use until the appro-priate safety evaluation has been written and approved. An inspection of the physical installation was not performe _- _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ - _ _ _ - _ _
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4.2.2 Findings The licensee's Architect Engineer (AE) was responsible for generating the draft narrative and design specifications associated with the modification. The licensee's System and Safety Analysis, Nuclear Analysis, Civil / Structural, Fluid Systems and Mechanical Components, Power Systems, Instruments and Controls, Environmental and Radiological Health Services, Construction Managen;ent and various onsite groups provided design change criteria and inputs. In addition to the inputs from the various disciplines, the modification package was also reviewed by the Nuclear Engineering Department's (NED) Design Review Board (DRB) and approved with comments to be incorporate The package was also evaluated for an unreviewed safety question by the Operations Review Committee and approved for implemen-tation by the Nuclear Operations Manager. The inspector determined that the design considerations and requirements were adequately stated and approved in accordance with licensee procedural control The design outputs incorporated from the design considerations and requirements were adequately developed into installation instruction and drawings; maintenance requirements; welding and nondestructive examination requirements; electrical, mechanical and functional testing requirements and acceptance criteria; and requirements for updating drawings and procedures and performing trainin A safety evaluation was properly conducted, documented and reviewed by the ORC in accordance with 10 CFR 50.59 and the Technical Specifications (TS). The evaluation considered the applicable areas of changes to the SAR or TS and the existence of an unreviewed safety question. The evaluation demonstrated that the installation (not operation) of the Direct Torus Vent does not pose an unreviewed safety question. The actual use of the Direct Torus Vent w111 be addressed in another safety evalua-tion which will include the criteria for using the vent as well as an assessment of the consequences of using i The design verification process utilized for the modification, to meet the requirements for ANSI N45.2.11, consisted of an independent review performed by the AE. The modification ,
package also received supplemental multidisciplinary reviews from the licensee's cognizant engineer, the cognizant group 3 leader, and the DRB members. The multidisciplinary technical i review comments were dispositioned and incorporated into the design package, as applicabl )
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The inspector reviewed selected Field Revision Notices (FRNs),
included in the package to date, associated with the modifica-tion and determined that they were generally minor.in nature or i issued for clarification. The inspector determined that FRN 86-51-26, incorporating a rupture disc in the 8" vent line, was processed in accordance with the licensee's procedural controls for handling major changes. The level of review afforded the change was commensurate with the initial design review and the change was evaluated in a SE. The dispositions of the FRNs adequately addressed the identified concerns. The minor scope of the changes to the design package indicates that the origina design was well develope The design change package reviewed was well organized. The design. inputs generated by the AE and licensee were effectively developed as evidenced by the lack of major FRNs. The safety evaluations adequately demonstrated the bases for the conclusions drawn and were reviewed as required. Overall the implementation of the Plant Design Change program reviewed by the inspector was effective and interfaces between the AE and cognizant licensee engineering groups were maintaine No violations were identifie .3 PDC 86-52A Replacement of Containment Spray Header 4.3.1 Description of Modification The inspector reviewed plant design change 86-52A. This design change involved replacing the 104 upper and 100 lower contain-ment spray header nozzles inside the drywell. The replacement nozzles were identical to.the existing nozzles except that six ]
of the seven spray openings in each nozzle would be capped off leaving only one opening for spray. This replacement reduces the capacity of the drywell spray system. General Electric has performed an analysis to confirm that the new spray capability is adequate to maintain drywell structural and atmospheric temperatures below the design limits of 281 F and 340 F during LOCA or a small steam line break accident. The reduced flow will minimize the possibility of damaging the containment by sudden depressurization following the initiation of containment spra During the implementation of this modification the licensee observed rust formation on the upper header. The system ,
engineering group and the corporate engineering staff were j developing a design change to adequately drain the lines and '
prevent any rust formatio The proposed design change to install a drain connection between the containment isolation
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valves was still under review. The adequacy of this corrective action is being addressed by a previous NRC inspection findin (50-293/87-26-01)
The inspector reviewed the design' inputs from the various engineering disciplines and verified that the inputs were adequate and in compliance with NED procedure 3.02, "Prepara-tion, Verification, Approval and Revision of Design Documents for Plant Design Changes ~/ModiHeations."
The safety evaluation (No. 2133) was an in depth analysis of all the potential areas of concern including systems, subsystems, components affected, safety functions of affected systems, effect on safety functions, and analysis of the effects on safety function The safety analysis also included a section on recommended FSAR change The licensee had a program for the procurement of commercial grade items for the safety related application. The inspector reviewed Safety Evaluation No. 236 on the procurement of spray nozzle The evaluation contained inspection requirements, quality requirements tnd post installation testin For post modification testing the licensee utilized an existing plant Operations Department Procedure No. 8.5.3.4, "Drywell and Torus Headers and Nozzles Air Test." This procedure was used by the test group and added into the plant design change package for verifying the function of the modified spray syste .3.2 Modification Field Inspection The inspector reviewed the work instruction for the replacement of the spray nozzles. Torquing requirements and handling were clearly identifie Even though the existing nozzles could have been modified to meet the intent of this modification, the licensee chose to replace the nozzles due to ALARA considera-tions. 'All the 104 nozzles in the upper header and 100 nozzles in the lower header were installed at the time of inspectio The inspector examined the installation and no discrepancies j were observe j
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i 4.3.3 Findings The inspector reviewed the containment spray modification package 85-52A against the design requirements and the Nuclear Engineering, Department procedures. The cognizant-engineer was very knowledgeable of the modification analyses performed by Bechtel Engineering and General Electric. Based on the review I
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of the design package and the discussions with the contributing departments, the following aspects of the modification process were verifie *
The design package was reviewed in sufficient detail by all the relevant engineering disciplines and there were no outstanding concern *
The final package had undergone approval and an independent verification before installatio *
The changes to the FSAR were identified and properly ,
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The affected drawings were adequately revise *
The safety evaluation addressed all the potentially affected areas and the conclusions had sound base *
The equipment procurement had adequate quality requirement During the performance of this modification, the licensee identified the rust formation in the spray headers. The exact cause of the problem was unknown, but licensee representatives suspected that it was caused by water leakage into the contain-ment spray header. The inspector observed the documentation of this problem and reviewed Field Revision Notice 86-52A-06 which was under consideration for rectifying the problem. The cognizant engineer for the modification was kept informed of all the significant observations during the implementation of this ,
modificatio l The post modification test procedure generated by the Modifi- !
cation Management Group was in compliance with the Nuclear Operations Department Procedure 1.3.4. The author of the test procedure was very knowledgeable of the station test and sur-veillance procedures. For this modification the test engineer elected to use an existing Nuclear Operations Procedure 8.5.3.4,
"Drywell and Torus Headers and Nozzles Air Test." The procedure was found to be adequate for testing the modified configuration of the plan No violations or other discrepancies were identi-fied for this modification. A violation was identified concern-ing the qualification of environmentally qualified equipment in the drywell and is further detailed in paragraph i :
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1 4.4 PDC 87-30 Anticipated Transient Without Scram (ATWS)
Recirculation Pump Trip Modification 4.4.1 Description of Modification The recirculation pump trip (RPT) is an important BWR plant
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feature used for the mitigation of ATWS events. Pilgrim Station uses motor / generator (M/G) sets to control recirculation pump flow. For such M/G controlled systems, the trip of the recir-culation pump may be done by tripping the field breaker of the M/G set or the drive motor of the M/G set. Pilgrim presently uses the field breaker trip. In June 1986, however, it was discovered that'with the reactor in cold shutdown an M/G field breaker had failed to trip on demand. Two similar incidents had occurred in 1985. Subsequent investigation determined that the probable cause was improper lubrication and adjustment. To correct this problem, spare breakers are used to replace the
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older units when PM or corrective maintenance is required, and l
l the replaced breakers are sent to the vendor for proper lubrica-tion and adjustment.
l l The purpose of this modification was to add diversity and increased ' reliability to the RPT by adding a trip to the drive mStor of the M/G set. The modification changed the existing l recirculation pump logic to accept an ATWS trip signal. A redundant trip coil was added to the existing trip coil for each 4160 volt drive motor breaker. Diversity was added by having division "1" control power operate one of the coils and division "2" control power operate the other. When installed the modification 'l provide an automatic trip of the drive motor breakers coder high reactor pressure or low reactor water level conditions. The logic system was also separated into the two source divisions with two separate trip channels in each division. For each division, both channels, "Two-out-of-two logic", must initiate to get a trip signa The system uses an energize to trip logic schem At the time of this inspection, NED had completed the modifi-cation pacLa;e and it had been approved and accepted by the station for construction and implementation. Training modules / material had been prepared on the modification by the Training Department and training started for both operations and maintenance personnel. Draft preoperational testing procedures had been prepared and were undergoing staff review. Instal-lation of the modification was about 20 to 30 percent completed.
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4.4.2 Modification Field Inspection The RPT pump trip signal will be initiated from existing level and/or pressure transmitters and will operate existing trip unit Eight spare relays in the logic cabinets near the ATWS pressure and level transmitters will be rewired to operate in parallel with the existing trip units. A second trip coil was added within each 4160v recirculation pump breaker. Thus there were two trip coils for each 4160v recirculation pump breaker, or a total of 4 coil One of the coils on each breakers was powered and actuated by Division 1 RPT-ATWS logic while the second trip coil was actuated by Division 2 RPT-ATWS logi Cables were run from the logic cabinets located in the reactor
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j building to the switchgear rooms located in the turbine build-ing. Each division was routed in separate raceways. The inspector, with the senior construction engineer, walked down those portions of the modification which had been installed as well as those areas where installation work had not yet commence Included in this inspection was a tour of the switchgear room, the cable spreading room and the reactor building. The following work was completed or in progress at the time of this inspection: Conduits were installed in portions of the switchgear room and cable spreading room; seismic bracing for the cable raceways were installed in the cable spreading room; and the logic rack was removed for modi-fication from the logic cabinets in the reactor buildin Based on this tour the inspector determined that:
Good housekeeping practices were in evidence throughout the plant, both in areas where construction activities were going on as well as non-work areas, a
Good radiation protection rantrols were being used to control contamination and .nonitor work activities, and
The senior construction engineer appeared knowledgeable of the modification and his responsibilities as they related to the installation proces .4.3 Findings l The inspector reviewed the ATWS recirculation pump trip l modification design, PDC 87-30, against ANSI N45.2.11 design requirements and licensee procedures. Discussions were held l with the cognizant engineer. Based on the above review and !
discussions, it was verified that:
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The cognizant engineer was knowledgeable of all phases of his design and the program used to assure quality of the design;
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All organizational groups which might be affected by the modification were involved in reviewing both the proposed design and final design package;
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The safety evaluation provided a detailed discussion of the basis for the safety evaluation conclusions; and,
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The licensee had developed good controls over the design inputs given to the contracto The inspector reviewed management control of PDC 87-30 once it was accepted by the station. This consisted of reviewing station procedures against ANSI standards; walking down the modification with a member of the construction group; discussing and reviewing control of the modification package with the Modification Management Group; discussing and reviewing the status of training on the modification with the Training Department; reviewing and discussing the procurement of safety related items, commercial quality items, and environmentally qualified items with the procurement group; and reviewing the draft preoperational test procedure against the design packag No violations or other deficiencies were observe .5 PDC 86-73, Automatic Depressurization/ Blowdown System Modification and Testing 4.5.1 Modification Description and Status This modification was designed to comply with NUREG-0737,
" Clarification of TMI Action Plan", item 11.K.3.18, recommen-dation which states, "The automatic Depressurization system actuation logic should be modified to eliminate the need for manual activation to assure adequate core cooling." The existing automatic depressurization system logic requires a LOCA signal consisting of concurrent high drywell pressure and low reactor water level signals. When both high drywell pressure and low water level signals have been received the logic starts the 120 seconds delay timer. The timer automatically resets if the low water level clears. This timer allows the operator to prevent ADS actuation by resetting the timer. The modification provides a bypass of the drywell pressure signal after a set time delay and a manual inhibit function. The low reactor pressure permissive of the RHR and core spray pump start logic is also bypassed, after the above time delay takes place. The modification automates the ADS system by providing automatic L_________-____
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initiation for events such as a break external to drywell or a stuck open safety relief valv The modification also provides the capability to easily inhibit ADS actuatio The modifica-tion has been installed and was in the process of being tested during this inspectio .5.2 Test Review and Findings The inspector reviewed the post modification test procedure TP87-'148 "Preoperational Test of the Automatic Depressuriza-tion System Following Logic Modification". The procedure was generated in compliance with procedure 1.3.4 which provided the guidance in preparing teraporary test procedure. The inspector witnessed the first section of the test procedure. The inspector independently timed the relay actuations and concurred with the test results. The instruments used were properly calibrate The annunciator display, relay actuation etc. were independently verified with QC personnel posted at the control room and at the tested relay pane In order to assure the adequacy of the test procedure, the inspector randomly selected the 2 relays and 2 switches from the modified portion of the control logic and verified that the test procedure includes functional check of these device Presently, Procedure 1.3.4 addresses reviews for format and other administrative areas but does not state that procedures should be reviewed for technical adequacy. Fven though this activity is performed under the existing interdepartmental review process, the licensee agreed to revise the operations manual to require the existing review for technical adequac In reviewing the test data on post modification testing, the inspector observed that the test directors accepted valve test results even when it is very close to the upper limits of tolerance. Following the inspector's comments on this, the licensee took immediate steps to notify the test directors to make an effort to adjust the setpoints such that any foreseeable drift will not cause the instrument to drift outside the ac-ceptance value between calibration interval The maintenance department also agreed to adopt a trending program to minimize instrument drif t exceeding acceptance value between two consecu-tive calibration interval No violations were identifie !
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5.0 Quality Assurance and Quality Control (QA/QC)
5.1 Scope of Inspection The inspection focused on the effectiveness of BECOs QA/QC for the modification process, including installation of modification Included in the inspection was the staffing for QA/QC, audits and surveillance, procedural controls and corrective actions to l finding .2 Organizations and Staffing The nuclear quality assurance organization has a manager and four group leaders who report to the manager and have responsibilities for the QA/QC of the modification process. The department manager stated that there were a total of 52 BECO employees in the Nuclear Quality Assurance Departmen The four QA/QC groups are discussed belo The Quality Engineering Group had a group leader and eleven QA engineers, with one position opening unfilled. The group's responsi-bilities include QA program related procedure review, supplier evaluation / surveys, external audits (including on-Site contractor surveillance monitoring), source inspections, procurement documents review, maintenance of BECO's QA approved supplier list, specifica-tion review, review of PDCs and Major FRN The Audit Group had a group leader and four QA engineers, however the staffing plan was stated to be designated for six QA engineer One of the existing four was on extended medical leave and another was transferring to the operations department at the end of August, thus will leave only three active QA engineers for auditing. The Audit Group's resptqsibilities include auditing the Nuclear Organiza-tion Departments, Operations Quality Control, support groups such as purchasing, stores and the general test sectio The Surveillance Monitoring Group had a group leader and six QA engineers. There were two vacancies, however, a new hiree was being brought on board. The Surveillance Group's responsibilities include surveillance monitoring of Pilgrim operation The Operations Quality Control Group had a group leader and ten quality control engineers. There were three vacancies. The group leader stated that up to 25 contractor QC personnel were providing support during the current outage. This group performs first line QC for BECO and second line QC for surveillance of contractor wor .. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __
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5.3 QA/QC Audits and Surveillance 5.3.1 The Quality Engineering Group (QEG) had 84 audits scheduled for 1987, 29 were completed and the others were scheduled and on i
schedule with the planned completion dates. QEG audits the
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contractors that are performing modifications, such as Bechtel
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which is performing certain aspects of the PDC 85-07, " Reactor ,
Water level Modification." For example, QEG conducted audit ;
87-20 of Bechtel and published a audit report of the audit dated May 1987. The purpose of the audit was to verify that Bechtel at Gaithersburg is performing an effective QA program for work that has been contracted for Pilgrim Station. The audit identi-fied items requiring corrective action by Bechtel. Bechtel formerly responded by letters to the Boston Edison QA manager providing corrective action being taken or to be taken to correct identified deficiencie QEG also performs ongoing surveillance of contractor's on-site QA program implementatio These surveillance compliment QEG audits. Interviews with a QEG quality engineer on site verified the implementation of this surveillance program. Contractor monitoring programs for both Bechtel and GE were sample inspected and found to be in place and functionin .3.2 The Audit Group was determined to have audits planned and scheduled. Review of the audits completed against those sche-duled and ones in progress determined that audits were on schedule at the time of the inspection. Review of the 1987 audit schedule showed 24 areas to be audited, 12 audit reports had been issued at the time of the inspection. The audit schedule is approved by the Vice President of Nuclear Engineer-ing and Quality Assurance. The Audit Groups basic focus is on whether or not the Boston Edison Nuclear Organization is imple-menting its procedures and policy. Audits were inspected and discussed with auditors to ascertain audit effectiveness. An audit of plant modification (No. 86-2) conducted January 20, 1986, through January 31, 1986, was reviewe Two deficiency reports (DRs) resulted from the above audit. Licensee correc-tive action was sampled for DR 1498 and the inspector determined that corrective action was satisfactorily completed on June 6, 198 An audit ('2. 86-45) conducted August 29, 1986 to October 14, 1986, covered design control activities. This audit identified 16 deficient conditions, of which, ten were reported as being immediately corrected. Six DRs were issued as a result of the >
audit, four were against the Nuclear Engineering Department (NED), one against Nuclear Management Services and one against j
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Outage Managemen The audit provided eight recommendations and verified that the quality assurance program was in place to cover ANSI N45.2.11-1974. Corrective action for two DRs, 1582 and 1589, were sampled. One QA auditor was interviewed regarding corrective action; both DRs were closed out satisfactorily.
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One audit (No. 84-17) of the RHR system was innovative in the method and utilization of technical specialists to provide for an in depth technical assessment of system operation. This audit identified eight DRs and 14 recommendations. Based upon the audits sampled, the audit group audits were encompassing the area of modification .3.3 The Surveillance Group performs ongoing surveillance of Pilgrim operations. This group does not specifically focus on modifi-cations, however, they do have an involvement such as doing surveillance on post work and modification testing. A monthly surveillance schedule is published which assigns the surveil-lance group engineers to an area. Each engineer prepares a check list for the surveillance to be done in accordance with a work instruction. When unsatisfactory conditions are identi-fied during a surveillance, corrective action is required within five days regarding the matter or it is escalated via the DR process. The Surveillance Group publishes a report monthly of surveillance performe .3.4 The Operations Quality Control group (0QC) performs inspections and surveillance of modification work and testing. Hold points are used to assure that 0QC engineers perform timely inspec-tion Inspections by OQC of the PDC 85-07 modification of the reactor water level instrumentation were inspected. 0QC inspection report No.87-288 dated March 16, 1987, identified a concern regarding hanger settings and was carried as an open item in the report log. A later report No.87-518 performed on April 24, 1987, closed out the above report indicating that the OQC corrective action system was functioning. It was noted that neither QC report questioned whether or not the hangers were being set cold with pipes full or empty. As detailed previously in this report, the design drawings did not specify a full or empty pipe for the hanger settings. OQC reports were noted to cover other testing, core drilling including assuring protection of the containment vessel liner (IRS86-262), and surveillance of Bechtel QC and NDE of Welding performed by Bechtel and GE for the PDC 85-07 modificatio .. _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _
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5.4 Management Controls for QA/QC A " Policy Statement On Quality Assurance," No. B-5, was issued by the Senior Vice President-Nuclear and approved by the Chairman, President and Chief Executive Officer on February 20, 1987. This policy state-ment clearly sets down the overall objectives to establish, maintain and implement a QA program that effectively complies with 10 CFR 50, Appendix The Quality Assurance Manager is provided with "Stop Work" authority whereby he can suspend any quality related activity or process which may, in his opinion, adversely affect the safe operation of the Pilgrim statio Procedure N0P84A10, revision dated August 31, 1984, titled "Stop Work Orders" provides the procedure to effect a stop work order. During interviews with QA engineers at the working level, they were found to be knowledgeable regarding the use of a stop work orde During the review and inspection of QA activities, it was found that procedures and group work instructions were available and were being implemented to carry out the licensee's commitment .5 Findings Based upon the sample taken, QA and QC of the modification process was found to be effectively implemented. A concern was expressed regarding the loss of QA engineers from the Audit Group's staff. The staff as of September 1, 1987 will drop from four to three active quality assurance engineer auditors; however, the full staffed complement is stated to be six quality assurance engineer No violations were identifie .0 Equipment Qualification (EQ) Considerations in Containment as Result of Containment Spray Modification During the inspection of the containment spray modification, PDC 86-52A the inspector reviewed the qualification of environmentally qualified equipment used in the drywell. The System Component Evaluation Work sheets for solenoid valve SV-220-44 and splices assemblies Q102A, B, Q103A and B located inside containment indicated that the equipment was quali-fied only up to 290 F. This temperature is derived from the drywell LOCA analysis which gives a temperature peak of 280 F. The Final Safety Analysis Report (FSAR) review sheets in the subject design package con-tains figure 5.2-3 " containment response to 0.10 square feet steam leak."
This figure indicates a temperature peak of 320 _
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The licensee has considered steam line breaks, FSAR chapter 5 section 2.3.2, which included a spectrum of Main Steam Line Break [MSLB] analyses ranging from 0.02 to 0.05 square feet. FSAR figure 5.2-4 indicates a temperature peak of 320 The NRC safety evaluation on equipment qualification transmitted on June 3, 1981, states the followin " Temperature, Pressure, and Humidity Conditions Inside Containment Max Temp ( F) Max Press (psig) Humidity (%)
LOCA 290 44 100 MSLB 320 24 100 The licensee's specified temperature profile for qualification purposes enveloped both the MSLB and LOCA temperature profiles and includes a margin at least as large as would result from the staff's recommendatio Therefore, we conclude that the specif#ed temperature profile is acceptable."
The SER indicated that Boston Edison Company intended to qualify electri-cal equipment inside drywell to a temperature 320 F. On August 12, 1987, the qualification file on solenoid operated valve SV-220-44 and splice assemblies Q102A, B and Q 103A, located in the Drywell established environ-mental qualifications for only up to 290 F, which was based on large break LOCA. The 290 temperature is 30 F lower than the temperature to which drywell equipment will be subjected in case of a small steam line break of 0.10 square feet which is contrary to 10 CFR 50.49(e)(1) since the licensee did not establish the most severe design basis time dependent temperature condition for EQ equipment in drywell (50-293/87-32-01).
i Subsequent to this inspection, an exit meeting was conducted on August 20, 1987, at the NRC Region I office. At this meeting the licensee discussed the various background information that lead to the conclusion for using the LOCA profile for qualifying the EQ equipment inside containment. The licensee explained the lack of specific guidelines for addressing MSLB in Boiling Water Reactors. A Boston Edison letter to the NRC dated July 9, 1984, states that the large break LOCA curve represents the most severe conditions inside containment resulting from a postulated high energy line break and is used as the basis for BECO's EQ evaluation. This conclusion L resulted from various discussions with the NRC staff and other utility ]
groups which provided some implicit concurrence that if MSLB is not i
. intended as the design basis event for the plant licensing basis, the licensee does not have to consider MSLB at al l The supporting documents for using the large break LOCA curve were submitted to the NRC on August 24, 1987. The EQ issue and the above documents will be the subject of further NRC review.
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The licensee ' developed and presented a draft small break LOCA profile for drywell with a peak temperature of 330 F. The representatives from NRR were consulted in this matter and the licensee was requested to submit the newly developed temperature profile for NRR revie Based on the newly developed temperature profile, the licensee reviewed all the EQ equipment in the drywell and stated that their equipment i qualifiable to the new temperature profile. The licensee committed to update the subject qualification records to support. qualification of the subject equipment for the new temperature profile before plant restar This is an unresolved item pending NRC review on the capability of the EQ equipment in drywell to perform its safety function during a small break LOCA (50-293/87-32-02).
7.0 Licensee Action on Previous Inspectors Findings 7.1 (Closed) Unresolved Item (50-293/85-30-03):
The Safety Evaluation (SE) written to support Temporary Modification 85-22 for the removal of demister sections from each of the two redundant Standby Gas Treatment (SGTS) filter trains was determined to be incomplete.in that it did not address the effects of this modi- ;
fication on all accident scenarios in which the SGTS would function.
, This SE did not analyze the effects of the assumed primary contain-I ment leak rate of 0.5% per day during an accident, and the effects on SGTS operability for other events, such as total loss of spent fuel pool cooling anc HPCI system operation during small break LOC The inspector verified that by June 30, 1986 the safety evaluation was expanded to consider the following:
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Main steam line break outside containmen Failure of the Fuel Pool Cooling system HPCI Operation post intermediate break LOCA (gland seal condenser exhaust to SBGTS).
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Design basis accident primary break inside containment to include 0.5% per day primary containment leakag Fuel handling acciden The results of the revised SE indicated that the modification was still acceptable. The inspector observed no deficiencies in the revised S ________
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In addition, training was conducted for engineers in the performance of safety evaluations. The inspector verified that Procedure 3.07,
"10 CFR 50.59 Safety Evaluations" was revised to provide more compre-hensive safety evaluation check sheets and incorporating industry good practice standard for performing safety evaluations. Based on the above, this item is close .2 (Closed) Unresolved Item (85-20-0911 Nitrogen purge capability was added to the HPCI System in May, 1973, by a modificatio This modification allowed nitrogen to be purged into the HPCI turbine steam exhaust line to break a siphon when the turbine stops and to prevent unnecessary water drainage from the torus to the Reactor Building sumps. Operating and test procedures required the steam exhaust line be purged for two or three minutes respectivel The two minutes was based on field experience and there were no design calculations to establish the duration of the Nitrogen purge. This lack of design analysis appeared to be contrary to ANSI N45.2.1 Through review of the applicable documents, the inspection verified that the licensee had accomplished the following actions:
(1) Design calculation M-273 was performed which verified that the minimum purge time is approximately two minute (2) Procedure 2.2.21, "[ Operation of the] High Pressure Coolant Injection System", has been revised to allow nitrogen purge to operate for three minutes for conservatis This is consistent with test procedures 8.5.4.1, "HPC1 Pump Operability Slow Rate and Valve Test at 1000 PSIG", 8.5.4.3, "HPC1 Flow Rate Test at 150 PSIG", and 8.5.4.6, "HPCI Pump and Valve Operability From Alternate Shutdown Station" each of which currently require a three minute purge tim (3) Technical Quality Memo (TQM) No. 77 was issued on November 24, 1986. This TQM instructed all engineers that it is a require-ment that all design analyses be performed in a planned, con-trolled and correct manner and that there exist traceability from design input through design output. Original design basis analysis and calculation ntust be retrievabl (4) Documentation was provided that NED engineers have received additional training in the plant design change process to cor-rect this proble Based on the above, this item is close _____ _
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8.0 Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable, deviations or violations. One unresolved item was identified during this inspection and is detailed in Paragraph 6.0, 4
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9.0 Management Meetings j Licensee management was informed of the scope and purpose of the inspection at an entrance interview conducted on August 3, 1987. The findings of the inspection were periodically discussed with licensee representatives during the course of the inspectio A preliminary exit interview was conducted on site at the conclusion of the inspection on August 14, 1987 at which time the preliminary findings of the inspection were presented. A final exit was conducted with licensee representatives at Region I offices on August 20, 198 Attendees for both exits are listed in Attachment At no time during this inspection was written material concerning inspection findings provided to the licensee by the inspectors. At the conclusion of the exit meeting the. licensee representatives did not identify any inspection items to be considered as proprietary information.
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ATTACHMENT A
. PERSONS CONTACTED Boston Edison Company, Contractors and Consultants
M. Akhtar - Modification management Group Leader (NOD) <
R. Andrew - Senior I&C Engineer (NED)
R. Antonopoulos - Group Leader Nuclear Analysis (NED)
H. Balfour - Licensing and Operations Training Group Leader (NOD)
R. Barasei - Site Structural Engineer (NSD)
J. Bellefev111e - On Site Safety and Performance Group Leader ;
S. Bibo - QA Audit Group Leader '
R. Bird - Senior Vice President Nuclear c
H. Brannan - Quality Assurance Manager
M. Brosee - Outage Manager R. Cook - Operations Training J. Edelhauser - Assistant Director Outage Management (WOD)
R. Fairbank - Licensing and Analysis Section manager (NED)
F. Famulari - Operations Quality Control Group Leader P. Farron - Staff Assistant to Nuclear Operations Manager
R. Grazio - Field Engineering Section Manager J. Hamawi - President, Entech Engineering In *
P. Hamilton - Compliance Management Group Leader K. Harper - Construction Management Group Project Engineer (Quadrex)
S. Hudson - Operations Section Manager D. Hughes - Requalification Training J. Ivens - Engineer, Modification management Group P. Karatzas - Principal Radiological Engineer B. Lunn - Compliance Engineer
J. Mattia - QA Surveillance Group Leader W. McCann - Group Leader, Procurement Group G. Meieris - Senier Mechanical Engineer (NED)
D. Mills - Construction Management Group Leader
F. Mogolesko - Principal Engineer Systems and Safety Analysis
A. Morisi - Test and Turnover Manager i 'J. Neville - Senior Electrical Engineer Power Systems Group (NED)
H. Chem - Deputy Engineering manager J. Peters - Site Consr.ruction Engineer D. Richards - Mechanical Engineer (MED)
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K. Roberts - Nuclear Operations Manager
J. Rogers - Systems and Safety Analysis Group Leader R. Schifone - Quality Assurance J. Seery - Technical Section J. Smith - Technical Training P. Smith - Chief Technical Engineer (NOD)
- D. Sparks - Consultant to Bostor Edison C T. Sullivan - Nuclear Watch Engineer
- R. Swanson - Nuclear Engineering Manager
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L. Vallee - Sight Engineering Office manager
R. Velez - Project Manager P. Voordeporte - Mechanical Engineer (Bechtel)
l J. White - Civil / Structural Engineer (NED)
l V. Whitney - Senior Methods Training and Compliance Engineer l R. Williams - Senior Engineer, System and Safety Analysis
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E. Ziemianski - Nuclear Training Section manager l
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Attachment A A-2
US Nuclear Regulatory Conmission
~ McBride - Senior President Inspection
J. Lyash - Resident Inspector
C. Anderson - Chief, Plant Systems Section (DRS)
J. Durr - Acting Deputy Director, Division of Reactor Safety (DRS)
- L. Bettenhausen-- Chief, Operations Branch (DRS)
J. Strosnider - Acting Chief, Engineering Branch (DRS)
- J. Wiggins - Chief, Reactor Projects Section IB
- R. Fuhrmeister - Reactor Engineer
- D. Holody - Enforcement Specialist
- C. Tinkler - NRR
- J. Craig - NRR i
- J. Kudrick - NRR '
i The inspectors also reviewed numerous other licensee personne *
Denotes those present at preliminary exit meeting conducted at Pilgrim Station on August 14, 198 ** Denoted those present at Final Exit meeting conducted at Region I offices on August 20, 198 *** Denotes those NRC participants by telephone in final exit meeting conducted on August 20, 198 l I
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ATTACHMENT B
.. Documents Reviewed
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Nuclear Operations Department Procedures 1, Procedures, Rev. 32 i 1.3.13 Plant Design Change, Rev.-15 l 1.3; Preparation of Safety Evaluations, Rev. 10 1.3.27 Field Revision Notice,' Rev. 6 1.3.31 _ Procurement of Items and Services,=Rev. 2 1.3.3 Design Walkdownsfor Inspections Conducted at PNPS, Rev.0
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1.3.45 Preparation / Review / Approval / Retention and Control of N00 .,
Calculations, Rev. '
1. Temporary Modifications, Rev. 1 >
' Nuclear Operations Procedures (N0P)
N0P 83A2 Station Configuration and its Control .
NOP'83A6 Modifications Management,J7/31/84 N0P 83El' Control of Modifications to Pilgrim Station, 5/14/87 NOP 84A7 Drawing Control, 7/13/87 NOP 83A16 10 CFR 21 Reporting of Defects and Non-Compliances, 9/19/85-NOP 84A8- Control of Commercial Quality Items, 11/4/85 NOP 84A9 Equipment Qualification Program, 11/29/85 NOP 83E5' 10'CFR 50:59 Safety Evaluations, 7/16/86,
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NOP 84El- Engineering Service Request Process, 2/21/85 Nuclear Engineering Department Procedures (NED)
12.03 ' Indoctrination and Training Program,.Rev. 1, 2/10/83-3.0 Review, Evaluation and. Approval of Contractor Design Documents, Rev. 13, 11/25/85 3.02 Preparation, Verification, Approval cnd Revision of Design Documents 1for. Plant Design Changes and Modifications, Rev. 20,17/22/87 3.03 Field- Revision Notice, Rev. 17, 6/22/87 3.04 Field Control of Materials Parts, and Components Rev. 1, 10/6/87 3.05 Design Calculations, Rev. 8, 11/7/86 3.06 Design Verification, Rev. 1, 5/23/84 !
3.07 10 CFR 50:59 Safety Evaluations, Rev. 4, 4/16/87 3.09 Design Review to Determine Conformance to Regulations, Rev. O, 3/22/82 3.15 Review of Temporary Modifications, Rev. 1, 5/12/86 4.01 Procurement of-Items and/or Services, Rev. 12, 6/27/86 4.05 Evaluation of Commercial Quality Items and Specification, Rev. O, 10/1/84 6.03 Maintaining the Environmental Qualification Master List, Rev.11, 11/29/85 15.01 Non-Conformance Reports and Document Deficiency Notices, Rev. 10, 11/12/86-16.01 Evaluation of Conditions Adverse to Quality, Rev. 8, 7/3/84
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Attachment B B-2
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16.02 Response to Deficiency Reports and Stop Work Orders, Rev. 2, 2/9/83 ;
16.03 Corrective Action Program, Rev. 2, 11/24/86 16.04 Implementation of Regulatory. Requirements Analysis Process, j 5/25/84
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NED WORK INSTRUCTIONS NEDWI 268, S&SA Problem Analysis and Recommendation Reports, Rev. 1 NEDWI 357, Preparation of SJA's, PDC's and FRN's for DRB Review and Presentation, Rev. O TECHNICAL QUALITY MEMORANDUMS (TQM)
TQM-10 Correptual Design Inputs and Review, 1/18/83 TQM-17 Recurring Problems with PDC /FRN Noted During QAD Reviews, 4/29/83 TQM-25 Plant Design Change Procedure NEV 3.02, 9/22/83 TQM-35- Preparation of Design and Design Documents, 4/17/84 TQM-56 PDC. Specification References, 1/11/86 TQM-58 Drawings Checklist, 3/14/86 TQM-62 N0D/CMG Reviews and Sign-offs of PDC's, 6/2/86 TQM-63 Design Criteria Input Sheets, 7/15/86 TQM-70 Standard Presentation Format for Design Review Board, 9/16/86 TQM-72 Design Document Control List, 9/23/86 TQM-74 Guidelines for Conversions of Temporary Modifications to Permanent Modifications, Rev. 1, 11/5/81 TQM-83 FRN Reviews, 2/13/87 TQM-84 Distribution of Material for DRB Review, 2/17/87 TQM-86 Distribution of Material for DRB Review, 5/20/87 Training Section Procedures A-7 Program Attendance Records, Rev. O A-23 Collecting Training Analysis and Evaluation Documentation, Rev. 1 T-1 Preparation of Curriculum, Rev. 0 T-6 Revision of Curriculum, Rev. 1 T-8 Module Development, Rev. 3 i
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Attachment B B-3
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Test Procedures TP 87-126, Pre-Operational Test of The ATWS Recirculation Pump Trip Modification Draft TP 87-148, Preoperational Test of the Automatic Depressurization System following Logic Modification Procedure 8.5.3.4, Drywell and Torus Header Nozzles Air Test Miscellaneous Documents Modification Management Manual (N0D), 7/30/87 Boston Edison Co.'s Purchase Order 64045A, 12/15/86 NED Design Review Board Charter, 12/2/86 Memorandum to File From H. R. Balfour, Dated 8/10/87, Audit of Training Requirements Associated with RF0 #7 PDC' Pilgrim's Nuclear Power Station Q List, 3/30/87 Drawings P&ID M251, Recirculation Pump Instrumentation, Rev. E3 u Schematic E415, Recirculation System Sheet 7 of 9, Rev. E Plant Design Change Packages 85-07 Reactor Water Level Instrumentation Modification 86-51 Director Torus Vent Modification 86-52A Replacement of Containment Spray Header Caps / Nozzles 86-73 Automatic'Depressurization. Blowdown System Logic Modification 87-30 ATWS Recirc. Pump Trip Modification 87-45 Torus Pressure / Containment Water Level Indication Quality Assurance NOP 84 A10 Stop Work Orders, 9/30/84 NOP 83 A13 Deficiency Report Process, 1/13/87 N0P 83 A9 Management Corrective Action Process, 9/30/86 No. B-5 Policy Statement on Ouality Assurance, 2/20/87 (Unnumbered) Contractor Surveillance Monitoring Program, GE, 6/22/87 (Unnumbered) Contractor Surveillance Monitoring Program, Bechtel Construction, Inc., Rev. 2, 5/5/8 <
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