ML20215J710
| ML20215J710 | |
| Person / Time | |
|---|---|
| Issue date: | 10/06/1986 |
| From: | Correia R, Jocelyn Craig NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| Shared Package | |
| ML20215J668 | List: |
| References | |
| REF-QA-99901058 99901058-86-01, NUDOCS 8610270160 | |
| Download: ML20215J710 (30) | |
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ORGANIZATION: SOUTHERN COMPANY SERVICES BIRMINGHAM, ALABAMA REPORT INSPECTION INSPECTION NO.: 99901058/86'01 DATE: 5/19-23/86 ON-SITE HOURS- 205 CORRESPONDENCE ADDRESS: Southern Company Services ATTN: Doug Dutton Vice President Post Office Box 2625 Birmingham, Alabama 35202 ORGANIZATIONAL CONTACT: Doug Dutton, Vice President TELEPHONE NUMBER: (205) 870-6011 NUCLEAR INDUSTRY ACTIVITY: Design and engineering services for operating plants and plants under construction within the Southern Company Organization.
ASSIGNED INSPECTOR: Ow R. P. Correid, Special Projects Inspection Section Date
/0 2-8 6 OTHERINSPECTOR(S): P. D. Milano, SPIS K. C. Leu, SPIS S. V. Athavale. Quality Assurance Branch T. Del Gaizo, Consultant
/iD. Golden,gonsultant APPROVED BY:
% / ( [a
/Juhn W. Craig, Chief. SPIS, Wendor Program Branch ~Date INSPECTION BASES AND SCOPE: _
A. BASES: 10 CFR Part 21, 10 CFR Part 50 B. SCOPE: The inspection consisted of an evaluation of design and engineering activities performed for the E. I. Hatch plants and J. M. Farley plants.
1 PLANT SITE APPLICABILITY: E. I. Hatch plants (50-321, 366) and J. M. Farley (50-348,364) 8610270160 861010 PDR GA999 EUTSCSI 99901058 PDR
. ORGANIZATION: SOUTHERN COMPANY SERVICES BIRMINGHAM, ALABAMA REPORT INSPECTIOM NO.: 99901058/86-01 RESULTS: PAGE 2 of 18 A. VIOLATIONS:
No violations were identified during the inspection.
B. NONCONFORMANCES:
- 1. Contrary to 10 CFR Part 50, Appendix B, Criterien III, and SCS procedures 10502.7-5 and 220100.7-5 for identification of structures, systems and components, procedures for calculation and domestic drawing preparation, 10502.4-4, 4-5 (Hatch) and 220100.4-4, 4-5 (Farley), do not adequately provide for the identification of safety classifications of equipment / materials. Equipment / material safety classifications are shown on equipment lists only and not on applica-ble drawings, calculations and other supporting documentation.
(86-01-01)
- 2. Contrary to 10 CFR 50, Appendix E, Criterion III, and SCS procedures 10502.4-4 (Hatch) and 220100.4-4 (Farley) calculations SNE-86-002, Rev. O., 85082-MP, Rev. 3 and SNC-85-098, Rev. O, do not consistently contain required information such as the calculation number, page and sheet numbering listed on each page, the signatures of the preparer and reviewer, and applicable codes and star:dards and sources of input data. (86-01-02)
- 3. Contrary to 10 CFR 50, Appendix B, Criterion III and IEEE-344-1975, Section 4, the SCS safety evaluation for modification package DCP-85-165 and calculation SNC-85-95 did not include a seismic qualification for the fuse and fuse holder. (86-01-03)
- 4. Contrary to 10 CFR 50, Appendix E, Criterion III and IEEE-279-1971, Section 3.9, SCS modification packages for plant Farley 82-0-1346, 83-2-2387 and 84-0-2642 did not include an analysis to establish margins for replacement instrument settings although the new instruments characteristics (accuracy, drif t, minimum resolution, response time, etc.) are different from the instruments which were replaced. (86-01-04)
- 5. Contrary to 10 CFR 50, Appendix B, Criterion III, SCS procedures 10502 and 220100, Section 4.4, and ANSI N45-2.11, Section 4.2, supporting oocumentation for modification packages PCR 82-1-1306, DCR 85-051, PCR 82-0-1345, DCR 79-134-1, DCR 83-218, DCR 83-262 i
. ORGANIZATION: SOUTHERN COMPANY SERVICES BIRMINGHAM, ALABAMA PEPORT INSPECTION RESULTS: PAGE 3 of 18 NO.: 99901058/86-01 and DCR 85-215 did not contain sufficient documentation such that verification of the adequacy of the design could be completed without recourse to the originator. Also, field changes were not subjected to the same design control measures consnensurate with those applied to the original design. (86-01-05)
- 6. Contrary to Criterion III of 10 CFP 50, Appendix B, and Section 4.2 of ANSI N45.2.11, Hatch I cable tray calculations for DCR 83-236, SNC 85-083, Cable Tray Support RB-087-A-2 did not provide a reference list, did not specify the analytical methods and did not contain sufficient detail to permit understanding the contents. Farley 1 calculation No. SC-84-1-3052-001, " Missile Door Operator Bracket,"
did not separate criteria from assumptions; and data which was critical to the analysis used in calculating the clearance to the New Fuel Bridge Crane was obtained verbally without written confir-ma tion. (86-01-06)
C. UNRESOLVED ITEM No unresolved items were identified during this inspection.
D. STATUS OF PREVIOUS FINDINGS None. This is the first NRC Vendor Program Branch inspection of the Southern Company Services.
E. INSPECTION FINDINGS AND OTHER COMMENTS:
Southern Company Services Overview Southern Company Services (SCS) is one of six companies which comprises the Southern Company. The other five entities are Alabama Power Company, Mississippi Power Company, Georgia Power Company, Gulf Power Company and Southern Electric International. Nuclear units in the Southern Electric System are the Edwin I. Hatch Nuclear Plants, the Joseph M. Farley Nuclear Plants and the Vogtle Electric Generating Plants.
SCS' engineering scope of work includes: site evaluation and selection; specification, bid evaluation and procurement of major equipment; construc-tion permits and operating licenses; and design and engineering services for both operating plants and plants under construction. During this inspection, only the design and engineering activities for operating plants were reviewed. Current engineerin projects is approximately fif ty percent 50'e) of(g manpower total dedicated SCS' engineering to nu personnel. The scope of this inspection involved only Hatch and Farley plant activities involving Design Change Recuests (DCRs).
ORGANIZATION: SOUTHERN COMPANY SERVICES BIRMINGHAM ALABAMA REPORT INSPECTION RESULTS: PAGE 4 of 18 NO.: 99901058/86-01 SCS responsibilities for the Hatch project included licensing and engi-neering/ procurement activities for both construction and post operational phases of the project. During the construction of the Hatch plants, SCS was primarily responsible for the balance-of-plant (B0P) design and certain safety-related system design (e.g. Control Building, Diesel Generator Building, Intake Structure and systems within these buildings). SCS was also responsible for administering the Bechtel (A-E) and General Electric (GE) (NSSS) contracts. Currently, SCS is the responsible A-E for engineer-ing and procurement.
The SCS Hatch Project Nuclear Plant Support organization provides managers, utility engineering and discipline directors as well as an on-site staff to support requests from Georgia Power Company. Requests include enoineer.
ing assistance, design changes, as-built notice and drawing maintenance for both permanent and temporary on-site support. DCRs may include design drawings, material specifications, safe 6y evaluations and other supporting documentation. The work package may also include environmental and fire protection evaluations, functional test procedures, craft work instructions and technical specification revisions. In the current DCR system all requests generated from Georgia Power Company are forwarded directly to SCS. Upon receipt of the request. SCS determines if the in-house staff can support the man-power, time and/or expertise required.
During construction phases of the Farley project, SCS responsibilities included licensing, engineering, and procurement. The engineering responsibility during construction was primarily the B0P work for Unit 1.
The Unit 2 BOP responsibilities were transferred to Bechtel. Current SCS' Farley activity includes engineering support for both units primarily for the 80P and other areas as assigned by Alabama Power Company. Mcdifica-tions for the Farley units assigned to SCS by Alabama Power Company may be in the form of a Production Changes Request (PCR) or an Engineering Support Request (ES). These requests are either assigned directly to SCS or Bechtel by Alabama Power Company. Approximately 57% of Alabama Power Company's PCR's and ES's are assigned to SCS. PCR's assigned to SCS are processed by the appropriate disciplines and packaged as Production Change Notices (PCN) and returned to Alabama Power Company for implementation.
PCN packages contain the same type of supporting documentation as the DCR response packages produced by the SCS' Nuclear Plant Support - Hatch.
Upon completion of the modifications, Alabama Power Company develops a Work Completion Notice (WCN) which is sent to SCS Nuclear Plant Support -
Farley. Final documents are then prepared by SCS for official drawing transmittal back to Alabama Power Company.
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ORGANIZATION: SOUTHERN COMPANY SERVICES BIRMINGPAM, ALABAMA REPORT INSPECTION NO.: 99901058/86-01 RESULTS: PAGE 5 of 18 SCS Ouality Assurance programs for the Hatch and Farley project support groups provides quality assurance guidance to the engineering staff and vendor shop quality surveillances and audits. Responsibilities include evaluation of SCS engineering and procurement procedures for compliance with Safety Analysis Report (SAR) commitments, NRC regulations and guides, industry codes and standards as well as SCS corporate policy. The SCS Quality Assurance (QA) program is structured to comply with ANSI N45.2, Quality Aswrance Program Requirements for Nuclear Facilities. Included in the QA program are committments to comply with referenced NRC regu-lations, codes, standards, and guides identified in applicable plant SARs.
Yearly audits of the SCS Hatch and Farley engineering work activities are performed by SCS QA personnel to assure that applicable elements of the SCS QA program have been effectively developed, documented and implemented.
10 CFR Part 21 Evaluations The NRC inspector reviewed the Plant Hatch Operational Support Policy and Procedures fur the identification, evaluation, and reporting of significant deficiencies. Section 2.4.2 of the procedures manual states the Nuclear Safety and Licensing section (NSL) of the Nuclear Safety and Fuel (NSF) organization is responsible for evaluation of significant deficiencies.
1 In an interview with SCS NSL staff engineers, the NRC inspector was briefed cn SCS policy and procedures for the identification, evaluation and report-ing of deficiencies as applicable to Plant Hatch. Potential deficiencies reportable under 10 CFR Part 21 are received from Georgia Power Company (GPC) via work requests with pertinent information. All work requests received by NSL are logged and tracked by both the manager and engineer responsible for the evaluation. Monthly status reports are written by the engineering staff to report progress on evaluations to SCS managers and GPC.
Three work requests involving potential 10 CFR Part 21 reportable deficien-cies from GPC were reviewed by the NRC inspector. The following is a brief description of each problem and followup evaluations by SCS NSF staff.
- 1. File HII-A.22.44, " Errors in Engineering Computer Program," Hatch Units 1 and 2. GE reported to GPC that potential nonconservative results using engineering computer program RVRIZO2 may occur if GE guidelines (ref. NEDE) were not followed when using the program.
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i ORGANIZATION: SOUTHERN COMPANY SERVICES BIRMINGHAM, ALABAMA l REPORT INSPECTION NO.: 99901058/86-01 RESUI.TS: PAGE 6 of 18 Bechtel, the A-E for the Hatch plants, was contacted and subsequently performed en evaluation of the computer program problem. Bechtel concluded that they had used the program as GE had recommended and, therefore, found no problems. SCS had not used the program and, therefore, considered the problem not applicable to Hatch and not reportable under 10 CFR Part 21. The NRC inspector determined that applicable hSF procedures were followed and the item was properly considered closec out.
- 2. File HII-A.22.28 " Pipe Whip Restraints," Hatch Unit 2. This defi-ciency, discovered by Bechtel on 6/24/83, involved pipe whip restraints which were not designed nor installed on the control rod drive (CRD),
Reactor Water Clean-Up (RWCU), Reactor Core Isolation Cooling (RCIC) and the Auxiliary Steam (AS) systems. The systems were isolated while the required restraints were designed, installed and/or scheduled for installation. SCS conducted an audit of the Bechtel Hatch-project's quality assurance program implementation and concluded that the missing whip restraints incident was an isolated event. Bechtel subsequently changed their QA procedures. Followup documentation was completed by SCS personnel and the deficiency evaluation closed out 11/2/03.
- 3. File HII-A.27.40 "RHR Service Water Pumps." Originally, GPC had requested SCS to review planned modifications to the RHR Service Water (SW) Pumps. SCS contracted Bechtel to perform the RHR-SW pump modifications analysis. Bechtel contacted Johnson Pumps, the original pump vendor, to obtain the required design information.
The modification involved relocation of the pumps support / restraint system to a location out of a potential flood area.
Johnson Pumps, contracted McDcnald Engineering to perform a stress analysis for the proposed pump modifications. Mcdonald's analysis required that the flanges and their bolts be upgraded to higher strength capabilities (i.e. a thicker flange and higher strength bolts). Johnson did not inform Bechtel of the design change require-ments. At a later date, Johnson Pump discovered the deficiency and reported it to SCS. SCS inturn initiated the corrective action.
An SCS audit of Johnson Pump revealed that Johnson Pump had received the design change requirements from Mcdonald Engineering and failed to follow procedures which would have prevented the deficiency. SCS notified GPC of the RHR-SW pump problems. GPC, inturn, instructed SCS to evaluate and initiate corrective action. SCS completed all procedural documentation and closed out the report on 2/19/85.
CRGANIZATION: SOUTHERN COMPANY SERVICES
- BIRMINGHAM, ALABAMA REPORT INSPECTION RESULTS: PAGE 7 of 10 NO.: 99901058/86-01 Based upon the sample of 10 CFR Part 21 evaluations reviewed, the NRC inspector determined that the NSL engineering staff had followed applicable procedures and had maintained completed, orderly files which were accessible, clear and concise. No items of noncompliance or unresolved items were identified in this area.
Quality Assurance Audit Report Review The NRC inspectors reviewed the " Quality Assurance Technical Audit Report on the SCS Hatch Nuclear Operational Support Group, GPC, EWO 3141ZZ, 31482Z" dated 12/5/85. Two DCR's, which were part of the QA audit evalu-ation, 83-30 and 83-262 were requested by the NRC inspectors for review of these design modification packages.
DCR 83-80 was a non-safety related modification package issued to GPC to increase the size of a steam trap to accommodate drain flow to provide sufficient capacity to operate without using the bypass. Six Field C.1ange Requests (FCR) were prepared as part of this modification package. There were no calculations or any other supporting documentation which addressed piping pressure rating, over-pressure protection and instrument set-points.
FCRs accompanying various checklists to describe possible impacts of the changes on 10 CFR Part 50 Appendix P, Equipment Environmental Qualifications (EEQ) and 10 C'1 Part 50.59 safety evaluations did not contain any techni-cal justificai ens of the approved changes.
Modification package 83-262, involved the installation of spray water piping and mixing cones to moisture separator reheater drain tanks. In both cases, the audits were found to have been of appropriate scope and definition. The audit findings were documented and deficiencies subse-quently corrected.
Nonconformance 86-01-05 was identified in this area of the inspection.
10 CFR Part 50.59 Procedure Review SCS procedures and interpretations of the requirements of 10 CFR Part 50.59,
" Changes, tests, and experiments" were reviewed. Appendix 0 of the SCS procedure manuals " Plant Hatch Operational Support Policy and Procedures,"
parts 0-5 and D-6, and " Plant Farley Operational Support Policy and Procedures," parts 0-8 and D-9, describe the requirements in determining whether safety-related,10 CFR Part 50.59 or unreviewed safety question
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ORGANIZATION: SOUTHERN COMPANY SERVICES BIRMINGHAM, ALABAMA REPORT INSPECTION NO.: 99901058/86-01 RESULTS: PAGE 8 cf 18 applicability are involved in a design change. Both the Hatch and Farley procedures incorporate a checklist used to make these determinations. In the case of Hatch modifications, all Nuclear Safety Evaluation Checklists are reviewed and signed by a Nuclear Safety and Fuels engineer including those changes permitted in 10 CFR 50.59. In the case of Farley, only those evaluations which required further evaluation as a result of 10 CFR Part 50.59 are reviewed and signed by Nuclear Safety and Fuels engineers.
Written det611ed evaluations are only required when one or more of the answers to the questions on the checklist are "YES." In cases-where check-lists contained only "N0" responses, a written evaluation as to why these answers were "N0" was not provided.
Changes to the Hatch and Farley procedures manual covering the design input to a modification, design verification, and safety evaluation were reviewed by the NRC inspection team. The Hatch procedure revisions had been imple-mented. The revised safety evaluation checklist requires a review in all cases by Nuclear Safety and Fuels staff engineers and supporting analysis by them to be provided and maintained for each modification. These analyses are maintained in separate files from the modification package.
Drawing and Calculation Procedures Review The NRC team reviewed several electrical, structural and mechanical modifi-cation packages for Hatch and Farley. The team had general concerr.s with drawings and calculations which were part of the modification packages not identifying of safety-related) thethe quality item s)(classification being modified. (i.e.,SCS' safety-related, non-procedures 10502.7-5 (Hatch) and 220100.7-5 (Farley), "huclear Cuality Classification System for Design, Equipment and Materials" require that the quality classifi-cation cude be placed on applicable documents in accordance with the corresponding documents preparation procedure. Procedures for the preparation and approval of domestic drawings (SCS procedures 10502.4-5 (Hatch)and 220100.4-5 (Farley)) require the quality classification of the equipment / material and its Q or non-Q icentity be placed on the drawing if the drawing is a bill of material. Procedures for the preparation and review of design calculations (SCS procedures 10502.4-4 (Hatch) and 220100.4-4 (Farley) do not provide a method cf identification of the system, structure or basic component's quality classification [or its Q or non-Q identity].
Nonconformance 86-01-01 was identified in this area of the inspection.
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ORGANIZATION: SOUTHERN COMPANY SERVICES BIRMINGHAM, ALABAMA REPORT INSPECTION RESULTS: PAGE 9 of 18 NO.: 99901058/86-01 Electrical Modification Package Review The NRC team reviewed three electrical modification packages for Hatch Unit 1 and one modification package for Farley Unit I nuclear stations.
These packages were reviewed for (1) consistency with the original design basis requirements, (2) technical adequacy of the chosen design approach, (3) conformance with the applicable regulatory criteria and FSAR commit-ments, and (4) completeness of design details.
The findings identified weaknesses in the system for design control and independent verification of the completed design. The system in place does not effectively address the proper application of safety evaluation reviews for modifications as required by 10 CFR Part 50.59.
- 1. PCR 82-1-1306: Service Water Dilution Flow Loop This Farley Unit 1 PCR was prepared to relocate service water dilution loop flow switch, FS-580. The modification was necessitated by a technical specification item which required the switch to be operated while observing the operation of steam generator blowdown valve, RC-00238. Additionally, in the state-ment of work to be completed, a change to the 24 vdc power supply for the instrument loop, including this flow switch, was included because it was operating near the limit of its capability. Thus, it was being replaced by a separate 48 vdc supply.
In the original design, the flow switch energized an auxiliary relay which closed contacts in the steam generator blowdown processing system. PCR 82-1-1306 changed the circuit such that the flow switch is now the line contact for the steam generator blowdown proce hing circuit allowing the elimination of the auxiliary relay.
This PCR also deleted the function of the flow switch in the waste processing system. This portion of the change was not addressed in the statement of work for PCR 82-1-13C6. The original design for the waste gas discharge valve solenoid had a contact frcm the vent stack radiation monitor which shut the valve on a radiation alarm.
In a September 1977 design change, the auxiliary relay contact from the dilution loop flow switch was added in parallel to the radiation monitor contacts altering the circuit design. Thus, a vent stack radiation monitor alarm would not cause the waste gas discharge valve to shut unless the dilution loop flow switch auxiliary relay contacts were also open. The design change made in 1977 also failed to address the addition of a signal from the steam generator blevdown processing circuit to the vaste gas discharge valve logic.
l ORGANIZATION: SOUTHERN COMPANY SERVICES PIRMINGHAM, ALABAMA REPORT INSPECTION NO.: 99901058/86-01 RESULTS: PAGE 10 of 18 Although PCR 82-1-1306 deleted the contact in the waste gas system, the original design criteria for this circuit was violated by the September 1977 modification. As a result of questions asked during this inspection, SCS personnel and Bechtel personnel reviewed drawings and made an initial determination that PCR 82-1-1306 and the September 1977 change had been completed. Subsequently, plant personnel at Farley Unit 1 found that while the relays had been installed, the electrical termination changes had not been made.
The normal practice for PCR's is that following receipt, Farley Unit 1 completes the change and returns the as-built information to SCS. There was an apparent failure to ensure that a change had been properly completed in accordance with PCR 82-1-1306 and the September 1977 modification.
Questions concerning the failure to install the modifications and the as-found installation of this equipment which is inconsistent with the as-built drawings reviewed during this inspection will be forwarded to the NRC Region 11 office for information and appropriate action.
For the portion of the design change which described the 24 vdc power supply, the capability of the instruments to operate with a 48 vdc supply was reviewed. The SCS personnel stated that the instruments had dual voltage input terminals. However, this information was not included in the documentation, and would not be available for the design reviewer / checker.
Nonconformance 86-01-05 was identified in this area of the inspection.
- 2. DCR 85-051: Jumper Cell 36 in 250 volt de Station Service Battery This Hatch Unit 2 design change request was provided to SCS for review. The development of the information for the change and its justification was done by Bechtel.
The documentation package provided during the inspection included only a technical and licensing evaluation of the proposed modifi-cation. The documentation package did not include information for other areas which would be necessary for a complete review and verification of the change request. While.the technical justi-fication stated that the jumpering of a battery cell (cell 36) would not reduce the battery voltage under load below technical specification limits, it did not specify any service test to verify the load capacity requirements. Further, the DCR did not
ORGANIZATION: SOUTHERN COMPANY SERVICES
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BIRMINGHAM, ALABAMA REPCRT INSPECTION N0.: 99901058/86-01 RESULTS: PAGE 11 of 18 address any changes to the battery charger equalizer or float voltage settings nor did it discuss the selection of a new pilot cell since cell 36 was a pilot cell. Finally, the package cid not contain the specification requirements for the jumper cables.
In discussion with SCS personnel at the plant to cbtain information on these concerns, the inspectors were told that the DCR was not implemented. Instead, the battery had been completely replaced.
The DCR, however, had not been cancelled as of the date of the inspection.
Nonconformance 85-01-05 was identified in this area of the inspection.
- 3. PCR 82-0-1345: Lube Oil Temperature Control for 1C and 2C Diesels This Farley PCR modified the temperature sensing and control for the 1C and 2C emergency diesel " keep warm" lube oil system. Since the original design had the heater temperature switches on the discharge line in close proximity to the heater, the heater was subjected to excessive cycling. The PCR provided for a resistance temperature detector (RTO), transmitter, and controller circuit with the RTD located in the lube oil sump.
The design change was reviewed and several areas were identified in bhich the documentation was not complete. Since the new control system was composed of a loop of instruments versus the original switch, a failure modes and effects analysis (FMEA) should have been performed to assure that no decrease in reliability would result from the mcdification. Also, since this system would now be performing the low lube oil temperature alarm function, the consequences of an open RTD failure should have been addressed.
Because of the changes to the low temperature alarm circuit, the procedures and setpoint calculation sheet were requested for the ins trumen t. It was found that procedures are not available and a revised setpoint calculation had not been performed. SCS felt that the design change provided a system of better accuracy such that a change to the lube oil low temperature alarm was not necessary.
In a letter dated July 31, 1985, the diesel manufacturer, Colt, provided a response to Alabama Power Company's request for review of the proposed modification. Colt responded that the new design
. ORGANIZATION: SOUTHERN COMPANY SERVICES BIRMINGHAM, ALABAMA REPORT INSPECTION RESULTS: PAGE 12 of 18 N0.: 99901058/86-01 was more complicated than necessary and provided an alter'nate design.
These comments were not accepted by SCS and no documentation of the evaluation of Colt's response was available.
Nonconformance 86-01-05 was identified in this area of the inspection.
- 4. DCR 85-054: RPS Scram Relay Replacement This Hatch Unit 2 design change request was developed to replace the original GE Model 105 relay contactors with a new Model 305 relay. Since the original units are not available, the vendor was providing the new equipment as an acceptable and improved alternate.
The inspectors reviewed the new relay's ability to meet the seismic requirements for each location in the control panel. The Model 105 relays were tested seismically to functional failure. However, the Model 305 relays were tested to a lower saismic input. The documen-tation package contained a statement that it was " felt" that the new relays are acceptable seismically. Further documentation was requested and reviewed which showed that the calculation met the seismic criteria.
- 5. DCR 85-218: AC breakers Overcurrent Trip This Hatch modification package for replacing the existing EC-2A type overcurrent trips with %icro versa" trip units for safety related AK series breakers, and contained the related calculations SNE-86-002, Rev.0, 85082-MP, Rev. 3, and SNC-85-098, Rev 0. These. calculations were part of the final modification package. The team noteo that the calculations did nct have proper page and sheet numbers. Some pages did not have a calculation number, page number, or signatures of the preparer and reviewer. Calculations did not have references to the applicable codes and standards, and sources of input data were not identified. Based upon these findings, these documents had not been independently verified as required by SCS procedures.
Nonconformance 86-01-02 was identified in this area of the inspection.
- 6. DCR 85-165: Seismic Qualifications of Fuses and Box This Hatch DCP added 30 amp fuses to the shunt connected de ammeter circuit of the 125 volts de station distribution system. The team reviewed the safety evaluation for this modification which states the l
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ORGANIZATION: SOUTHERN COMPANY SERVICES BIPMINGHAM, ALABAMA REPORT INSPECTION N0.: 99901058/86-01 RESULTS: PAGE 13 of 18 equipment being added is not required to function after a seismic event. Based on this safety evaluation, the civil department calcu-lation SNC-85-95 dated 12/19/85 concluded that seismic qualification of the fuse and the fuse holder was not required.
Although the fuse and fuse holder will not be required to function af ter a seismic event, the nonqualified fuses and fuseholders were installed in a class 1E panel and may become dislodged during a seismic event and act as a missile and damage other class.1E components and wiring inside the class 1E panel or may fail such that a fault in the 1E circuit is generated. The team reviewed the analysis of this system by the licensee for 10 CFR Part 50 Appendix R requirements and noted that a fault developed in the circuit can lead to unavailability of class 1E,125/250 v de switchgears 1A and 18.
Nonconformance 86-01-03 was identified in this area of the inspection.
- 7. PCN-S-85-1: Replacement Of The level Switch On The Reactor Makeup Water Tank This PCN involved the replacement of an existing instrument in the safety system with a new instrument or a string of instruments consisting of a primary sensor, signal conditioner and bi-stables.
Tnese new instruments were set to the settings of the replaced instruments. The applicability of the old instrument settings to the new instrument was not documented. This analysis is required since the characteristics (accuracy, drif t, response time, etc.)
of the new instruments are different than the old ones. The team was informed by SCS personnel that the individual instruments settings were provided by the vendors of the instruments. SCS does not have a procedure by which an analysis of a system may be performed as a result of instrumentation changes.
Nonconformance 86-01-04 was identified in this area of the inspection.
- 8. DCR 84-80: Fire Wrapping of Cables The team reviewed DCR-84-80 Rev. O issued for the addition of fire wrapping on cables on elevation 130. The team found that the fire wrap used was TSI, Thermo-lac type and was rated for one hour.
10 CFR Part 50 Appendix R, Section III G-2-C requires that the fire area where one hour fire wrap is used shall have fire detectors and an automatic fire suppression system. The team was informed that all the fire areas were not equipped with the fire detection and protection equipment but exemptions were granted by the NRC via a
ORGANIZATION: SOUTHERN COMPANY SERVICES BIRMINGHAM, ALABAMA REPORT INSPECTION NO.: 99901058/86-01 RESULTS: PAGE 14 of 18 letter, dated April 18, 1984. The team reviewed two calculations for derating of the cable current capacities due to use of fire wrap and found that the correct derating factor was used. The calculations did not address other derating factors and factors for temperature corrections. However, the team was informed that complete ampacity calculations were performed under a different calculation number which was reviewed and found to be acceptable.
Seismic Modification Package Review Modification packages for seismic modifications at Farley 1, 2 and Hatch 1, 2 were reviewed.
The inspector audited a selected sampling of modification packages at Farley 1, 2 and Hatch 1, 2 to determine whether the applicable regulatory, technical, and QA requirements were included. These documents are listed below:
Farley Production Change Request (PCR), PCR 84-2-3048 (2113PB)
Farley Missile Door Operator Bracket Mod., SC-84-1-3052-001 Farley Unit 2 PCR, PCR-84-2-2544 UP (2115)
Hatch Field Deviation Requests, DCR 86-165 Hatch Support Documents, DCR-83-236 Hatch 1 Reactor Bldg Support Documentation (4 Volumes), DCR-85-144 Hatch 1 Reactor Bldg Support Documentation (4 Volumes), DCR-85-117 Hatch 1, AppaMix R Reactor Bldg Wrap Torus, DCR-83-236, SNC-84-083 Hatch 1, Appendix R Reactor Bldg, EL.87 and EL.130 Re: Qualification i of cable tray and conducts, Books No. 1 and No. 4 Summary of DCR-85-117 Cable tray and Conduct Support Modification
' Hatch Field Deviation Request, DCR-86-148 l Hatch Field Deviation Request, DCR-86-114 The following are findings with respect to Hatch 1: (DCR 236, SNC 85-083, cable tray support RB-087-A-2, Book 1 of 8) l J
l a. References used in the analysis and calculations were not specified.
- b. Methods used in obtaining the seismic loads were not specified.
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- c. Analysis and calculation results were not in sufficient detail
! to facilitate the verification of results.
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ORGANIZATION: SOUTHERN COMPANY SERVICES
- BIRMINGHAM, ALABAMA REPORT INSPECTION RESULTS: PAGE 15 of 18 N0.: 99901058/86-01 The following are findings with respect to Farley 1: (Cal. No.
SC-84-1-3052-001, " Missile Door Operator Bracket Modification")
- a. Criteria and assumptions were not separated;
- b. A dimension used in calculating the clearance to the New Fuel Bridge Crane was obtained verbally without written confi rma tion.
Nonconformance 86-01-06 was identified in this area of the inspection.
The NRC inspector also reviewed procurement procedures and checked appli-cable 10 CFR Part 21 implementation. One document related to Farley 2 purchase order SS85-1859 (PCR 84-2-3048) concerning modification of the Missile Door Bracket (dated 9/12/85 from SCS to Overly Manufacturing Company) did not specify the requirements of 10 CFR Part 21 in the P0.
Af ter a discussion with SCS personel involved with this P0, a document containing the Part 21 specification was produced from another source.
Mechanical Systems Modification Reviews The NRC team reviewed several mechanical modification packages for Hatch Units 1 and 2. The packages were reviewed for (1) consistency with the original design basis requirements, (2) technical adequacy of the design approach, (3) conformance with applicable regulatory criteria and FSAR commitments, and (4) completeness of design details.
In design packages compiled in the late 1970's and early 1980's, a number of concerns were identified, largely related to documentation. Specifi-cally, the number and type of calculations or analysis r.eeded to support the calculations themselves did not always clearly identify the source of input, assumptions for later verification, and acceptance criteria. More recent packages reviewed (e.g.86-055) indicate that many of the prior l
problems have been addressed and corrected. Draft revisions of project
! procedures on design input and design verification were reviewed and address the documentation and quality of the packages.
- 1. DCR 79-134-1: Modification to Containment Penetrations X-26 and X-220 This package consisted of the design sketches and a number of field change requests. It did not contain: 10 CFR Part 50.59 review, a l safety evaluation, a design input sheet, a list of installation instruments required, or purchase specifications.
Nonconformance 86-01-05 were identified in this area of the inspection.
i
. ORGANIZATION: SOUTHERN COMPANY SERVICES BIRMINGHAM, ALABAMA REPORT INSPECTION NO.: 99901058/86-01 RESULTS: PAGE 16 of le
- 2. DCR 83-218: Motor Cooling Modifications to PSW and RHR SW Pumps Calculation SNH-83-004 (determination of PSW flow to RHR SW pump motors with failed sightglasses downstream of PSW motor coolers) did not demonstrate minimum cooling flow still existed following the piping change. Similarly, calculations were not performed to demon-strate the effects of removing relief valves from the PHR SW piping.
Where engineering judgements were used in lieu of calculations, the judgements were not documented. The calculation did not state the acceptance criteria and therefore there was no way of determining whether or not the results were acceptable. Similarly, calculation assumptions #1 and #7 were not justified. The calculation did not contain necessary informat. ion to permit independent review.
Nor.conformance 86-01-05 was identified in this area of the inspection.
- 3. DCR 83-262: Modifications to Heater Drain Tank Piping The package only contained FCRs accompanied by various checklists used to describe possible impacts of the modification on 10 CFR Part 50, Appendix R, Electrical Equipment Qualification (EEQ), and Safety Evaluations. There was no indication of a technical justifi-cation of the approved field changes.
Nonconformance 86-01-05 was identified in this area of the inspection.
- 4. DCR 85-215 and 86-063: Celetion of Head Spray, Plant Hatch Units 1 and 2 The head spray system on boiling water reactors is intended to cool down the reactor. pressure vessel (RVP) head upon plant shutdown.
For plants where cooling the RPV head is critical, head spray is beneficial. But for plants for which RPV head cooling time is not critical, such as Hatch, it provides no benefit (ref. GE memorandum MDE-109-1284).
Design packages for this modification were reviewed. GE recommended removal of head spray as part of the Hatch Performance Improvement Program. An analysis performed by GE for removal of head spray provided the justification, a safety evaluation and alternate means for accomplishing removal. SCS developed the sketches to remove the head spray piping between the RPV nozzle flange and a flange at elevation 202 f t. , 4 inches downstream of valve 511-F019, and for the installation of blind flanges. Additionally, pipe support E11-RHR 402 was found to require the addition of a north-south directional restraint. A sketch was provided for this purpose.
ORGANIZATION: SOUTHERN COMPANY SERVICES BIRMINGHAM,' ALABAMA REPORT INSPECTION NO.: 99901058/86-01 RESULTS: PAGE 17 of 18 Technical specification changes were not made to allow locking and racking out the breakers for valves E11-F002 and E11-F023 as described in the modification. Upon discovery of this problem, GPC was notified by SCS not to perform this part of the modification. A hand calcula-tion was performed by SCS to determine if pipe support E11-RHR-402 was adequate to withstand loadings with water in the piping. Previous analyses assumed the pipe would be empty. This calculaiton was found to be unverifiable because the following items were not clearly stated: calculation number, design criteria, assumptions, calculation methodology, input to the calculation, references, or conclusions.
SCS provided a second calculation on the head spray piping analysis.
This calculation was performed during the inspection in response to discussions on the first calculation.
The second calculation, SM04036-060, dated May 22, 1986, was reviewed.
The seismic loading on the pipe supports after removal of portions of the head spray piping by DCR-85-215 was calculated using computer code PIPESD. The calculation was in a format consistent with SCS procedures The calculation appeared to be technically correct.
A pipe support stress calculation in support of an FCR generated as part of the head spray deletion modification on Plant Hatch Unit 1 (DCR 85-215) was not independently verifiable. This calculation was done to confim that the bracing added to pipe support E11-RHR-402 under FCR 85-215-3 would be adequate. This calculation was found to be unverifiable because the calculation did not clearly state:
design criteria, assumptions made, input to the calculation, calcu-lation methodology, references, or conclusions.
Prior to the end of the inspection, SCS engineers completed a computer calculation for the seismic response of the head spray piping as modified by DCR 85-215 and associated FCR's. This calculation was found to be independently verifiable and appeared to be technically correct.
Nonconformance 86-01-05 was identified in this area of the inspection.
ORGANIZATION: SOUTHERN COMPANY SERVICES
- BIRMINGHAM, ALABAMA REPORT INSPECTION RESULTS: PAGE 18 of 18 N0.: 99901058/86-01
- 5. DCR 86-055: Hydrogen Water Chemistry Test, Plant Hatch Unit 1 The addition of hydrogen to the reactor coolant system in BWR's to scavenge oxygen may result in reduced stress corrosion cracking in recirculation piping. The purpose of this test is to verify the efficiency of hydrogen analysis in the water chemistry control system in the plant and to determine the additional shielding requirements due to the increased nitrogen 16 carryover.
The design package fcr the test was reviewed. Since this design package did not include any design calculations by SCS and relied on work by GE, it was reviewed only for its format as compared to earlier design packages. This design package showed a higher level of organi-zation, including logs and checklists for the contents of each section.
The inspector concluded that the method of organization in this design package is an improvement over older design packages which were reviewed during the inspection.
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ITEM TYPE OF DOCUMENT NO. DOCLNENT NO. REV. DE TITLE / SUBJECT 1 DCR DCH 86-06 --
4/86 -
. JetR : aa spray plant Hatch Unit 2 y 2 DCR DCR-85-2r --
11/85 RrtR id spray plant Hatch Unit 1 3 REA REA-HT-45.lC-- --
2/1/84 _+ t eine t. ..nbility of deleting RMt heed spray .
4 Memo MDE-109-1284 --
12/84 (General Electric) Evaluations to justify head spray removal 5 Calc --
Stress calculation for FCR 85-215-3 l 6 DCR DCR-86-055 --
1/86 IWC test plant Hatch Unit 1 l
7 Manual --
3/27/86 Plant hatch operational support policy and procedures i
- 8 DCR DCR-83-080 --
5/17/83 Steam strainer mods, steam packing exhauster f) 9 DWG H-11102 11 12/7/79 Piping trubine drains plant Hatch Unit 1 10 DWG H-11630 0 3/10/83 P&ID turbine drains
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- e 11 Calc SNH-83-007 --
9/14/83 Steam trap sizing iM
1974 Yarway steam trap catalog 14 Catalog 15M-247 --
Yarway steam trap selection guide
,y design verification (4-1 and 4-9) NpSH and NPSF
- c, Type of Doc.sment
DWG - Drawing LTR - Letter Spec - Specification DCR - Design Change Request Package .
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! c' P0 - Purchase Order