IR 05000354/1985045
| ML20137B876 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 11/13/1985 |
| From: | Blough A, Fuhrmeister R, Lyash J, Strosnider J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20137B826 | List: |
| References | |
| TASK-1.A.1.1, TASK-1.A.2.1, TASK-1.A.2.3, TASK-1.A.3.1, TASK-1.B.1.2, TASK-1.C.2, TASK-1.G.1, TASK-2.B.4, TASK-2.E.4.1, TASK-2.E.4.2, TASK-2.K.3.13, TASK-2.K.3.24, TASK-2.K.3.25, TASK-TM 50-354-85-45, IEB-73-02, IEB-73-2, IEB-74-11, IEB-74-12, IEB-83-02, IEB-83-06, IEB-83-2, IEB-83-6, IEB-84-03, IEB-84-3, NUDOCS 8511260373 | |
| Download: ML20137B876 (17) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 50-354/85-45 Docket 50-354<
License CPPR-120 Licensee: Public Service Electric and Gas Company
' Facility: Hope Creek Generating Station Inspection at:
Hancock's Bridge, New Jersey Conducted:
September 24 - October 27, 1985 Inspectors:
A.R. Bl p, senior Resident Inspector Date 6Z7W uh4s
't J. J. Lypfrff, Reactor Engineer Date
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Fuhrmeister, Reactor Engineer Date
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. Strosnider, Chief, Projects Section 18 Date September 24 - October 27, 1985 (Report No. 50-354/85-45): A routine onsite resident inspection (208 hours0.00241 days <br />0.0578 hours <br />3.439153e-4 weeks <br />7.9144e-5 months <br />) of work in progress, preoperational testing, and new fuel receipt was conducted. The inspector also made tours of the site-and reviewed licensee action on previous inspection findings, Construction Deficiencies and TMI Action Plan Items.
One violation was noted involving excessive spacing between cable restraints on vertical cable runs.
8511260373 851119 ADOCK050Q4 PDR
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DETAILS
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Persons Contacted Public Service Electric and Gas Company
- A. Barnabei, Principal QC Engineer T. G. Busch, Technical-Engineer G. C. Connor, Operations Manager E. J. Dalton, Site Engineering G. Daves, Senior Engineer, Operations
- A. E.'Giardino, Station QA Engineer R. Griffith, Principal Staff QA Engineer P. Kudless, Maintenance Manager
- S. La Bruna, Assistant General Manager C. Lambert, Site Engineering R. Lovell, Radiation Protection Manager P. Landrieu, Project Manager.
- M. Metcalf, Principal Startup QA Engineer
- J. A. Nichols, Technical Manager J. M. Rucki, Maintenance Engineer
- R. S. Salvesen, General Manager, Hope Creek Operations C. Vondra, Operating Engineer Bechtel
- W. Goebel, QA Engineer
- C. Jaffee, Startup Engineer
- T. Indico, Startup Director D..Long, Field Construction Manager B. Markowitz, Project Manager G. Moulton, QA Manager 2.
Previous Inspection Item Update 2.1 (0 pen) Unresolved Item (85-19-01), various primary containment isolation valve issues. This item was also updated in inspection 85-42. This item remains open and is updated as follows:
(1) Regarding non-Q instrument lines attached to the extended primary containment boundary, the applicant stated that a design change has been initiated to upgrade these lines to Q-listed.
This action, when complete, will resolve this portion of the item.
(2) The inspector's concern regarding selection of isolation valve stroke times is expanded to include HPCI and RCIC steam line
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isolation valves, whose maximum allowable stroke times have been deleted from the FSAR and the Draft Technical Specifications.
2.2 (Closed) Inspector Follow-Up Item (83-18-02), broken torque paint on QC inspected HVAC duct supports. This item discussed the existence of broken torque paint on bolts used to connect HVAC duct support members.
Torque paint is used by QC as an indicator that torqued bolting has not been disturbed after final QC inspection. Bechtel performed a followup inspection to determine if a generic problem existed and determined that only Task Force II hangers were involved.
Of the Task Force II hangers reinspected, 205 bolts were identified with broken torque paint and of these 205, 10 were found loose. The 10 loose bolts with broken torque paint were isolated to hangers 2, 4, and 8 on W-H drawing SM-275.
Task Force II hangers are those 552 hangers listed on NCR WH-376 that were initially fabricated and inspected in total by W-H and were subsequently reinspected, reworked, and final inspected by Bechtel.
Bechtel QC indicated that, during final inspection of Task Force II hangers, no attempt was made to determine the extent of rework on a given hanger.
Each hanger was reinspected 100?s for conformance to design, and at least 10?s of the bolts were checked for minimum torque and torque paint reapplied. Bolts having broken torque paint were not necessarily selected, nor was broken torque paint removed from reworked bolts.
The reinspection conducted to address this concern identified 10 loose bolts with broken torque paint in a series of duct supports adjacent to one another. These supports were either reworked without proper authorization or the 10% sample inspection missed them.
In any case they were subsequently torqued. A memorandum has been placed in the permanent file with NCR WH-376 which states that a reinspection has been performed and that torque paint on Task Force II hangers from pervious inspections is not indicative of unauthorized rework.
2.3 (Closed) Unresolved Item (84-23-02), verification that polyurethane is not used in scram back-up valves and scram discharge volume pilot valves. The inspector reviewed General Electric Company documentation dated January 21, 1985, that confirmed no polyurethane is used in the subject valves. The inspector observed the subject valves in plant and verified that they were not of the T-ASCO type.
2.4 (Closed) Inspector Follow Item (83-14-08), errors in construction PM program computerized listing. This item is a duplicate of 81-10-02, which was closed in inspection 84-29.
2.5 (Closed) Inspector Follow Item (84-19-01), lack of formal documentation on scope of planned preoperational program audit 84-50.
In inspection 85-33, an NRC specialist inspector (1) reviewed QA audit schedules, procedures, and checklists and found them to
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acceptable; and (2) reviewed completed audit 84-50 and found it to be especially comprehensive.
2.6 (Closed) Unresolved Item (85-27-01), excessive length of unsupported Class IE cable. This item was updated in inspection report 85-42.
In response to the inspectors questions concerning the effect of the
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weight of unsupported cable on the last rung of the cable tray, the applicant provided calculations demonstrating the acceptability of this configuration. The calculations assumed the maximum permissable tray fill, and that the mass of cable acted as a point load on the center of the last cable tray rung.
Subsequent to the applicant's "use-as-is" disposition of the unauthorized raceway addition, the inspector noted the absence of main raceway to extension grounding straps. The applicant provided the inspector with Bechtel drawing E1401-0, Revision 21, Grounding
Notes, Symbols and Details. This is a non-Q procedure and thus presence of grounding straps is not a QC raceway inspection attribute. The applicant stated that grounding straps are primarily for personnel safety and that appropriate straps would be added to the extension.
Based on the calculations and drawings provide by the applicant, this item is closed.
2.7 (0 pen) Violation (84-29-01), control of measuring and test equipment
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(MTE). The inspector reviewed changes made to Startup Administrative Procedure 19 (SAP-19):
(1) to permit the controlled transfer of MTE between certified Test Engineers / Technicians; (2) to permit issuance of MTE in the absence of the Startup. Test Equipment Coordinator, and (3) to ensure that fixed M TE used for "information data" beyond its calibration due date is evaluated for impact on previous tests.
The inspector sampled both fixed and portable MTE in use in the
. field to determine that instruments were within the calibration due date in good condition, and in the case of portable MTE, were in the possession of the responsible individual and accompanied by usage logs.
The inspector-also reviewed Corrective Action Request HCS-149, actions taken in response to HCS-149, and the various surveillance activities conducted by QA/QC to determine the effectiveness of the corrective actions.
No' unacceptable items were identified in the above areas and corrective measures in place appear acceptable.
In assessing the applicant's actions to correct routinely delinquent and improperly documented M TE Log Cards the inspector examined the applicant's delinquent instrument list. The instrument usage records examined indicate that weekly usage log card submittal, timely notification of management in the case of delinquent usage log cards, and proper documentation required on the usage log cards continues to be a problem. The applicant is currently performing a number of
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audits designed to ensure that instrument usage is accurately
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tracked, independent of the usage log cards. These audits, as well as surveillances conducted by QC, indicate that while adherence to usage log card procedural requirements continues to be a problem, traceability of M&TE usage has been maintained.
This item remains open pending further inspector review of the applicant's M&TE ongoing audit program and verification that M&TE usage traceability has been maintained.
3.
IE Bulletin and IE Circular Followup 3.1 (Closed) IE Bulletin 73-02, Containment Purge Valve Switch Failure.
The bulletin was issued due to the failure of both inboard and outboard valves to close. This failure was the result of a slipped cam in a rotary switch in the control circuit for both inboard and outboard valves. The condition does not exist at Hope Creek, as each valve is controlled by a separate momentary contact pushbutton switch.
In addition, at HCGS, a containment isolation signal to the valves will override an opening signal. This item is closed.
3.2 (Closed) IE Bulletin 74-11, Improper Wiring of Safety Injection Logic. The bulletin was issued as the result of a miswired relay which remained undetected during several years of operation. At HCGS, Technical Specification 4.3.3.2 requires a logic system functional test every 18 months for all ECCS systems.
In addition, SA-AP.ZZ-009(Q), Control of Station Maintenance, Rev. 3 dated 8/4/85, requires, in Section 3.5, post-maintenance retesting per Technical Specification Surveillance Procedures whenever a system is to be
- returned to service following maintenance.
These requirements would ensure that should such a condition exist, it_would identified. This item is closed.
3.3 (Closed) IE Bulletin 74-12, Incorrect Coils.in Westinghouse Type SG Relays. This bulletin was issued as a result of 48VDC coils being installed in places of, and marked as, 125VDC coils.
Hope Creek switchgear utilizes General Electric type HGA relays, as indicated in the E-109 Switchgear Vendor Manual and Bechtel drawing E-1416.
The Hope Creek Switchgear were delivered in April 1980 and September
- 1981, well removed from the 1974 tin,e frame of the reported event.
These conditions, combined with the preliminary and preoperational testing would prevent a recurrence. This item is closed.
3.4 (Closed) IE Bulletin 83-02, Intergranular Stress Corrosion Cracking (IGSCC) of Large Diameter Piping. The issue was covered in the licensing process.
Reactor coolant pressure boundary materials
- selection and measures to mitigate IGSCC were accepted in SER Section 5.2.3.
The Preservice Inspection (PSI) Program adequacy was originally an SER Open Item but has been resolved (reference NRC:NRR letter to the applicant of 10/2/85, entitled, "Preservice Inspection Program").
Implementation of the PSI program has been verified by NRC during the construction phase inspection program.
The most notable inspector concern was inspectability of
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corrosion-resistant-clad welds; this concern was resolved in inspection 85-37.
Technical Specification 4.0.5 will require, throughout. plant life, an inservice inspection program. After reviewing the above-mentioned documents, the inspector had no further questions.
3.5 (Closed) IE Bulletin 83-06, Non-Conforming Materials Supplied by Tube-Line Corporation.
This bulletin was issued to alert licensees to the fact that Tube-Line Corporation supplied non qualified material to several sites while representing the material as qualified. A review of records shows that over 90% of the 578 fittings supplied to Hope Creek have been located.
Twenty-four were carbon-steel flanges installed in the Standby Diesel Generators which have been determined to be acceptable per ASME Code Case N-242. The heat numbers for the remaining unlocated fittings were to be added to Exhibit XIII of SWP/P-P-109 Control and Application of ASME Code Stamp and Preparation of ASME Code Data Reports.
This exhibit has been superseded by 10855-FSK-P-386, by adding the subject heat numbers to the checklist all ASME Code Stamp systems will be
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inspected for these materials.
Field Specification for N-5 Nuclear Piping Control Check List.
This item is closed.
.3.6 (Closed) IE Bulletin 84-03 Reactor Cavity Water Seal Failure. This bulletin described the failure of a pneumatic seal and the subsequent draining of the Reactor Cavity into the drywell.
The HCGS design incorporates a metallic bellows-type seal with leakage indication connections. An engineering evaluation by Bechtel has revealed no credible failure mechanism for this seal. Procedure OP-AB.ZZ-144(Q),
Loss of Fuel Pool Inventory, Rev. O, does include steps which address this. This item is closed.
3.7 (Closed) IE Circular 81-02,' Performance of NRC Licensed. Personnel on Duty. This circular was considered, and appropriate guidance incorporated, by the applicant in developing Administrative Procedure SA-AP.ZZ-002(Q), Station Organization and Operating Practices.
3.8 (Closed) IE Circular 81-12, Inadequate Periodic Test of PWR Protection System. The Hope Creek design is significantly different
'from the one discussed in the circular. Comprehensive protection system testing is specified in Technical Specifications 4.3.1 and 4.1.3.
4.
Construction Deficiency Reports (CDR)
4.1: (Closed) CDR (84-00-10), diesel generator control panel economizing-resistor deficiencies.
The resistors were replaced to correct mounting problems and to provide the correct resistance.
The inspector reviewed the completed nonconformance reports and inspected resistors in two of the four diesel panels. No problems were noted.
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4.2 (Closed)CDR(84-00-02), defective diesel generator fuel filter differential pressure switches. The applicant's evaluation of this item concluded that switch failure would cause only a small fuel oil leak and would not affect diesel operability.
Therefore, the CDR was withdrawn on June 29, 1984.
Nonetheless, the switches were replaced.
The inspector reviewed the applicants withdrawal letter and supporting documentation, reviewed fuel oil system drawings and observed the switches in plant. The inspector had no further questions on this item.
4.3 (Closed) CDR (85-00-03), incompatibility of HPCI steam supply valve motor with its power supply circuitry. The applicant replaced the motor with one having a lower locked rotor current to prevent overcurrent trips during normal valve actuation. The inspector reviewed the associated completed documentation, including Start-up Deviation Report (SDR), work procedures, QC Inspection Records, rework cards, and test package.
The inspector also observed the valve in plant and checked nameplate data.
4.4 (Closed) CDR (85-00-02), misapplied diesel generator load sequencer relays. The applicant had been informed by the vendor that the ability of the relays to open contacts in high current applications was indeterminate; and, thus, diesels might be overloaded in LOCA with loss of off-site power.
The relays.were replaced with a different model. The inspector reviewed documentation, including the NCR, the Field Change Request, the QC Inspection Reports, the Startup Deviation Report, and the Design Change Package. The inspector also inspected panels to verify the relay. replacements and a sampling of the wire terminations for the B' and D' diesel load sequencers.
5.
TMI Action Plan (TAP) Items 5.1 (0 pen) TAP Item I.A.I.1, Shift Technical Advisors. The applicant's program allows for certifying STA's via either (1) completion of SRO licensing plus add'itional training, or (2) completion of a comprehensive STA program, described in procedure TP303HC, Shift Technical Advisor Training and Certification, including comprehensive exam.
The formal training and exam are expected to be complete by November 22. The inspector will review course curriculum, the examination and its grading, and final STA certifications in a future inspection. The inspector also reviewed plant administrative procedures SA-AP-002Q and OP-AP-002Q, and found them to adequately address STA position responsibilities and manning requirements.
5.2 (Closed) TAP Item II.K.3.24, HPCI and RCIC Space Cooling.
Space Cooling must provide adequate cooling for at least two hours in event of loss of off-site power.
The Hope Creek design uses two 100 percent capacity safety grade coolers per room, each powered from an emergency (diesel-backed) bus. Cooling water is supplied by the Safety Auxiliaries Cooling System (SACS) which is also safety grade, diesel-backed. To verify these features the inspector reviewed FSAR excerpts, system design specifications (DITS D3.48), control logic
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8 diagrams (H-83-0), Air Handing Unit Specifications (M-711Q--sampling review), airflow diagrams and SACS piping diagram, and various electrical power supply diagrams and load tabulations. No unacceptable conditions were identified; however, the_ inspector identified some concerns, discussed below, unrelated to this TAP Item.
The ECCS room coolers are controlled by temperature switches which sense ambient room temperature.
The inspector expressed the following associated concerns:
(1) Draft Technical Specifications do not require periodic functional testing of the room coolers (In most BWR designs the cooler starts automatically when the associated ECCS pump starts, and, hence, periodic ECCS pump testing also tests the cooler).
(2) Draft Technical Specifications do not require periodic calibration of temperature switches.
(3) The temperature switch setpoint adjustment knobs are easily accessible, unprotected from inadvertent or purposeful re-adjustment.
(4) The temperature switche.s do not appear to be optimally located to sense a representative room temperature.
The applicant addressed the above items as follows:
(1) ECCS pump test procedures will be revised to include room cooler
&(2) functional test. Temperature switches will be periodically cali-brated.
The testing will be documented as an NRC commitment.
(3) Securing devices or covers will be provided for the setpoint idjustment devices.
(4) The applicant-showed the inspector, using appropriate drawings and system specifications, that significant air mixing will occur due to forced ventilation in both normal and accident conditions. Hence the temperature switches should sense a representative room temperature.
In a future inspection, the inspector will verify implementation of the commitments discussed above (85-45-01).
5.3 (Closed) TAP Item II.E.4.1, Dedicated Hydrogen Penetrations. The SER accepted this item based on an-alternate approach, endorsed by NUREG-0737, whereby the hydrogen control design is single failure proof for containment isolation purposes.and single failure proof for recombiner operation, with all components safety grade.
Hope Creek satisfies this requirement with redundant
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recombiners, which are mechanically and electrically independent, each also having redundant containment isolation valves. To verify
.these features, the inspector reviewed the FSAR, the system design specification (DITS D3.40), portions of the control logic diagram (J-58-0), and selected piping diagrams. The inspector also walked-down po-tions of the system, including piping from the containment penetrations to the isolation valves.
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5.4 '(Closed) TAP Item II. K.1.22, Proper Functioning of Auxiliary Heat Removal Syste s When Feedwater is not Functioning. The applicant's design was described and accepted in SER Section 15.9.2.
Two commitments for modifications are pertinent to this item:
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Automatic RCIC restart after a high level trip. This modification was TAP Item II.K.3.13, which is discussed in Detail 5.6 below.
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Modify ADS logic.
Completion of this item is being tracked by TAP Item II.K.3.18, which is still open.
In reviewing this TAP item the inspector noted the importance of proper automatic functioning of HPCI and RCIC. The inspector reviewed a sampling of HPCI and RCIC operating and test procedures to verify that they provided for (1) proper restoration of system to standby readiness, and (2) proper sequencing of steam line isolation valves during warm-up.
5.5 (Closed) TAP Item II.K.3.25, Loss of Power to Pump Seal Coolers. The applicant endorsed a BWR Owners Group resolution of this item, which was based on analyses and actual testing of a representative pump.
This was accepted in the SER. The inspector reviewed the following documents to verify the Owners' Group reports apply to Hope Creek recirculation pump seal and cooler design:
(1) BWR Owners Group letters dated May 1981, and September 21, 1981; (2) Design Specification for Reactor Auxiliaries Cooling System, DITS D3.11 (excerpts);
(3) N3SS. vendor bill of materials for recirculaticn pumps (order
- 205-AD-522, Item 711-S-0765), pump mechanical seal detail drawing; and (4). Reactor Recirculation System Drawing, M-43.
No discrepancies or inadequacies were noted.
5.6 (Closed) TAP Item II.K.3.13, Separation of HPCI and RCIC Initiation Levels.
In the SER, the NRC:NRR staff accepted the applicant's commitment to the BWR Owners' Group position that (1) initiation level separation was unnecessary, but (2) RCIC should be modified to restart automatically on low water level after a high level RCIC
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trip. The applicant implemented this item through Design Change Package (DCP) #66. The inspector interviewed cognizant engineers; reviewed system drawings, elementary diagrams, the SDR (work control document), and the DCP; and observed some of the in plant equipment.
The inspector also verified that the RCIC system preoperational test, as written, tests the automatic restart feature.
The inspector has no further questions.
5.7 (Closed) TAP Item I.A.2.1, Immediate Upgrading of Reactor Operator and Senior Operator Exams.
The applicant's program was reviewed and approved in SER Supplement 2.
Proper implementation _of the training program was verified by NRC Region I during license examinations on July 8-19, 1985. The inspector has no further questions.
5.8 (Closed) TAP Item I.A.2.3, Administration of Training Programs. This item primarily involved certification of instructors.
Each
instructor is to have been either SR0 licensed or certified at a similar BWR.
The inspector reviewed a sampling of instructor certifications to verify this item. Also, once the requalification program starts, permanent instructors are to participate. The inspector verified that training procedures require instructor particioation in requalification.
5.9 (Closed) TAP Item I. A.3.1, Revised Scope and Criteria for Licensing Exams.
This item included commitments relative to both (1) initial license examinations, and (2) requalification programs.
Initial license examination requirements have been implemented in the operator licensing process.
Regarding requalification program, applicant commitments (which were accepted by NRC;NRR in SER Supplement 2) included the following:
(1) Fundamentals review lectures, including heat transfer, fluid flow, thermodynamics, and mitigating core damage; (2) More rigorous grading criteria for annual written examinations; (3) Accelerated training programs for those found deficient in either the written exams or annual oral exams; and (4) Performance of all reactivity manipulations and evolutions specified by the H. R. Denton letter of March 28, 1980.
The inspector reviewed the applicant's requalification program procedure, TP-305HC, Revision 2, and found that it appropriately incorporates the commitments.
5.10 (Closed) TAP Item II.B.4, Training for Mitigating Core Damage. The applicant committed to provide training on mitigating core damage to operations personnel (from plant manager through the operations chain to licensed operators) and appropriate technicians and managers in the I&C, Health Physics, and Chemistry Departments.
The inspector reviewed various Lessons Plans and quizzes, and interviewed training
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supervisors and instructors.
Each training discipline had an appropriate program, including (1) lesson plans tailored to the responsibilities and knowledge requirements of that discipline, and (2) tracking' systems to ensure course completion by required personnel. The majority of personnel had completed the training, and the remainder were scheduled well before fuel load.
Lesson Plans reviewed by inspector included LP302HC-000-00-100 (series) for operations personnel, LP501HC-157.05 (series) for Chemistry, LP406HC-155.04161000 (series) for Radiation Protection, and 602HC-000.00-MCD(-1) and (-2) for I&C.
5.11 (Closed) TAP Item I.G. I, Training During Low Power Testing. The NRC;NRR staff accepted this item, including the overall test program scope and the involvement of operations personnel, in Chapter 14 of the SER. Proper implementation of committed test programs is the subject of extensive inspections during the preoperational and startup phases. Therefore, to avoid redundancy of inspection effort, this item is closed administratively.
5.12 (Closed) TAP Item I.B.1.2, Organization and Management. Most of this item is incorporated into the licensing process.
Inspection activities remaining are to verify proper staffing and functioning of the following safety review groups described in the FSAR:
Station Operations Review Committee (SORC), Onsite Safety Review Group (SRG),
and Offsite Review Group.
These inspections were started in inspection 85-33, and completion is being tracked via unresolved items 85-33-01 through 85-33-03.
Therefore, to prevent redundancy of inspection effort, this TAP item is administratively closed.
5.13 (Closed) tap Item I.C.2, Shift Relief and Turnover Procedures. The item was reviewed in inspection 85-19, with exception of the applicant's program for verifying effectiveness of shift turnover.
The inspector reviewed procedure OP-AP-ZZ-017Q and found that it includes programmatic internal audits of shift turnover.
5.14 (0 pen) TAP Item II.E.4.2, Containment Isolation Dependability. This item was reviewed by NRC;NRR in the licensing process.
In SER Section 6.2.4 the NRC staff accepted the applicant's classification of essential /non-essential systems, containment isolation initiation design, design of the control for automatic containment isolation resetting, and containment pressure setpoint.
Subsequent to the staff's acceptance of a containment pressure setpoint of 2.0 psig the applicant revised the FSAR and Technical Specifications to reduce the setpoint to 1.68 psig. This change appears to be acceptable in that the setpoint is changed in the conservative direction, and the change is consistent with BWROG recommendations.
In SER Section 6.2.4.1 and SER Supplement 2, Appendix L the staff
+ reviewed the applicant's measures to ensure containment purge and vent valve operability.
The applicant did not specifically qualify the containment purge and vent valves to LOCA conditions.
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Consequently, the applicant committed to lock closed and administratively control the 24-inch and 26-inch purge and vent valves during operational conditions 1, 2, and 3.
The 26-inch inboard purge valve may be opened in conjunction with a 2-inch outboard bypass for drywell pressure control, during power operation. This is allowable because the 2-inch bypass valve is qualified to close against LOCA flows, and the inboard 26-inch valve is qualified to close against the choked LOCA flow through the 2-inch bypass valve. Based on the choked flow assumption described above, the Technical Specification /FSAR closure times for the vent / purge valves were-extended from 5 sec to 15 sec. ~ Technical Specifications have been included addressing closure requirements and necessary leak rate testing. The inspector was informed that no administrative procedures were in place for sealing and maintaining closed the valves in question.
The applicant further stated that valve prototypes have been delivered to a laboratory for qualification testing under LOCA conditions, and that qualification tests would be complete prior to OL. The inspector pointed out that changes to the FSAR and Technical Specifications would be required, including re-evaluation of the TS valve stroke times and leak rate surveillance testing requirements. The inspector also examined installation of the containment purge debris screens and reviewed selected portions of containment isolation logic.
This item remains open pending completion of either (1) containment vent / purge valve qualification testing and applicable Technical Specification changes, or (2) implementation of edministrative controls to ensure the large diameter valves remain locked closed in Modes 1, 2, and 3.
6.
Preoperational Phase Activities 6.1 plant Tour The inspector toured the control room on regular and backshifts.
He interviewed operations personnel regarding testing scheduled or in progress, reviewed logs and night orders, and observed alignment and indications of systems undergoing tests. Operators and supervisors were knowledgeable regarding plant status and test plans. The inspector toured areas of the plant, including drywell, reactor building, and the control building. He checked on tests and operations in progress, observed equipment and housekeeping conditions, and interviewed personnel involved in ongoing activities.
Spot-checks on tests in progress included CRD stroking and venting, RCIC and HPCI testing, testing on 'B'
and 'D'
diesels, and isolation valve local leakrate testing (LLRT). While observing applicant set-up for LLRT of containment ventilation valves the inspector noted that the LLRT control box was set up remotely from the system test connection.
The box was connected to the-system by one quarter inch tubing. The test engineer stated that on occasion one hundred feet
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of tubing is used. Test pressure is controlled at the box and nitrogen flows through the tubing to makeup for leakage.
The allowable leakage criterion is based on valve diameter and could be as high as 2000 cc/ min. on large diameter lines. The inspector questioned whether the long tubing runs could cause a head loss, thereby affecting actual test pressure at the test valves. The applicant stated that the loss would be negligible and set-up is standard for the industry. The inspector performed independent calculation which confirmed that the loss would be negligible.
The inspector had no further questions.
6.2 Preoperational Test Procedure (PTP) Review The inspector reviewed the following preoperational test procedure for verification that planned testing fully demonstrates that system operation and response meet regulatory requirements and applicant commitments. The review also verified proper format and content, and applicant review and approval of the subject procedure.
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PTP-BB-1, Revision 1, Nuclear Boiler Process Instrumentation.
The HCGS FSAR, Section 14.2.12.1.3, BB-Nuclear Boiler, states that the test objective is to verify proper response and indication of all nuclear boiler process instruments unique to the reactor vessel.
It further establishes acceptance criteria for system interlock and control functions.
The inspector noted that paragraph 7.2.7 of PTP-BB-1 states that failure of a trip function to occur, or obvious discrepancy in setpoint, does not
constitute a Failure of any acceptance criteria.
This paragraph states that initiation of a Startup Deviation Report will be equivalent to successful completion of the step and permit the Test Engineer to continue test performance.
The inspector informed the applicant that disposition of discrepancies in this manner did not ensure that proper recalibration/ retest would be
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performed, and did not provide the basis for meeting test acceptance criteria.
In response to the inspector's concerns the applicant. approved Change Notice 2 to PTP-BB-1 requiring these discrepancies to be documented and processed as test exceptions. This requires instrument recalibration and loopcheck of the affected instrument be complete prior to signoff of the associated test acceptance criteria.
6.3 Preoperational Test Witnessing The inspector witnessed testing in progress on regular and backshifts and verified that:
1) testing was conducted using the latest revision of -the approved procedure by qualified individuals, 2)
controlled and calibrated measuring and test equipment was available for required data gathering, 3) adequate quality control coverage was provided, 4) proper coordination between test engineers and
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14 operations existed, and 5). test exceptions and changes were documented and.dispositioned properly.
During the report period the inspector witnessed several sections of the following preoperational tests:
(1)
BJ-1, Revision 0, High Pressure Coolant Injection (2) BD-1, Revision 0, Reactor Core Isolation Cooling (3)
KJ-4, Revision 2, Emergency Diesel Generator (DG400)
(4)
KJ-2, Revision 2, Emergency Diesel Generator (BG400)
Sections of the above tests observed included interlock and logic testing, valve functional tests and system integrated operation testing.
In review of the Master Test Copies for PTP-8J-1 and PTP-BJ-2 the inspector noted the large number of outstanding Startup Deviation Reports (SDR) impacting the tests. Of particular interest are those SDRs which involve extensive system design changes performed under Bechtel Design Change Packages (DCP). The inspector discussed this concern with Startup Management and stressed the need for careful evaluation of the scope and type of retest necessary to ensure PTP validity after DCP implementation.
Observation of PTP-KJ-4 included portions of the Diesel overspeed test, load reject test and start reliability test. The inspector noted that a large number of On-the-Spot Changes and Change Notices had been issued, making test readability poor.
Problems arising
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during test sections witnessed by the inspectors were properly documented. The inspector will evaluate test changes and test exception during results review.
7.
Construction -- Plant Tour and Walk-Through Inspection The inspector periodically toured the plant and performed walk-through inspection during this inspection period.
In the walk-through inspection, special emphasis was placed in the areas of drywell, reactor building, torus /wetwell, and diesel generator buildings. These inspections were carried-out to assess the level of general workmanship in the' areas of piping and pipe support; effectiveness of cleanliness and housekeeping program; and general conformance to project procedures in the work in progress and completed work. On two occasions the inspector noted NCR
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tags still hanging on in plant equipment several months after NCR closure.
While these particular items were not of significance, the inspector discussed with the applicant the importance for operations phase of ensuring that only valid posted information and tagging remains in place.
During a tour of Reactor Building elevation 102 on September 27, 1985 the inspector noted excessive lengths of unrestrained safety-related cable in vertical cable trays above floor penetration E-4301-4. Cable tray
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12DTMP91 and several adjacent trays had cable spans of six to twelve feet between the penetration fire seal and the first Ty rap to a cable tray ruag. Applicant review indicated that the penetration seal had been
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completed and-QC-accepted in May 1985.
10 CFR 50 Appendix B Criterion V, Instructions, Procedures and Drawings, requires that activities affecting quality shall be accomplished in accordance with instruction, procedures, or drawings, Drawing E-1408, Revision 19, August 26, 1985, Wire and Cable Notes and Details requires cables in cable tr o to be tied to the rungs at maximum intervals of 5 feet for vertical trays.
Failure to meet drawing specifications regarding cable tie spacing is a Violation (85-45-02).. When informed, the applicant's QA group promptly investigated this item and issued Corrective Action Request (CAR) HCS-232. The architect-engineer (AE) issued NCR #8493. AE inspection identified two other penetrations with similar cable tie nonconformances. The Ty-raps were installed and inspected, and the NCR closed on October 15. The inspector verified that.the Ty-raps were in place. Because the inspector have verified immediate corrective actions, the applicant's response to this item need only address measures to identify similar QC inspection error and measures to prevent additional similar occurrences.
No other unacceptable conditions were noted.
8.
Followup on Events Occurring During the Inspection -- Hurricane Near Site, September 26-27 On September 27, 1985, a hurricane passed near the site, creating high winds and torrential rains on-site. On September 26-27 the inspector monitored on-site activities to verify the applicant was taking adequate measures to (1) protect equipment important to safety and (2) identify any potentially damaged equipment.
During the severe weather, the applicant manned his TSC and EOF and had response teams continuously touring the plant. Due to the applicant's extensive preparations, actual damage was slight and was limited to nonsafety-related support facilities.
No
inadequacies in the applicant's response were noted.
9.
Security The inspector reviewed the security plan and procedures for protecting new
. fuel, interviewed the security supervisor, and spot-checked on security measures during new fuel receipt. No inadequacies were noted.
10. Maintenance
The inspector observed portions of preventive maintenance (PMs) on 480-volt breakers. The PMs were being done in conjunction with planned outages of 4-KV emergency buses. The inspector interviews craftsmen and QC inspectors and observed breaker trip testing. No inadequacies were noted.
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11.
Inspection Program Status Preoperational Test Program Inspection completion status is approximately as follows:
Area
% Inspection Complete Overall Program
Procedure Reviews -
Mandatory
Primal 100 Test Witness -
Mandatory
Primal
Results Review -
. Mandatory.
Primal
Inspection status is consistent with applicant test program progress.
Operational readiness inspection status is approximately as follows:
Area
% Inspection Complete OPS-Staffing & Procedure
Tech Spec Review
i Maintenance
Fire Protection
Fuel Receipt
Surveillance
Rad. Controls
Rad. Waste
' Security
Emerg. Planning
Additional inspection will be done in each area to verify readiness for fuel load.
Open Item inspection approximate status is listed below.
Items are considered " backlogged" if the applicant has presented information for closure but the inspector has not begun his review.
Area
% Closed
% Working (NRC)
% Backlogged
- Items Now Open Inspection
2
63
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Findings *
3
20-for 100%
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Area
% Closed
% Working (NRC)
% Backlogged
- Items Now Open t
Bulletins *
0
39
Circulars
0
27 CDRs*
0
12 SER
0
5 Verifications *
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- All data subject to change as new items are opened.
Start date for this accounting system was April 15, 1985 -- Items closed before then are not reflected in percentages.
The inspector considers the. backlogs as minor; NRC inspection status is consistent with applicant progress in closing open items.
Exit Interview The inspectors mat with applicant and contractor personnel periodically
.and at the end of the inspection report to summarize the scope and
-findings of their inspection activities. Written material was.not provided to the applicant.
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Based on Region I review and discussions with the licensee, it was determined that this report does not contain information subject to 10 CFR 2 restrictions.
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