IR 05000354/1985062
| ML20137E728 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 01/08/1986 |
| From: | Marilyn Evans, Galla J, Jerrica Johnson, Petrone C, Petronic C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20137E677 | List: |
| References | |
| 50-354-85-62, NUDOCS 8601170305 | |
| Download: ML20137E728 (10) | |
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U.S. NUCLEAR REGULATORY C0!iMISSION i
REGION I
Report No.
50-354/85-62 Docket No.
50-354 License No.
CPPR-120 Priority Category
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Licensee:
Public Service Electric and Gas Company 80 Park Plaza - 27C Newark, New Jersey 07101 Facility Name:
Hope Creek Generating Station, Unit 1 Inspection At:
Hancocks Bridge, New Jersey Inspection Corducted:
December 9-18, 1985
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W Inspectors:. C.~Petrone, L ad Reactor Engineer date g
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M. Evins, Reactof Engineer I/date
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yGolla', Reactor Engineer
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Approved by:
J. Johnson, Chief, Operational Programs date Section Inspection Summary:
Inspection on December 9-18, 1985 (Report No. 50-354/85-62)
Areas Inspected:
Routine unannounced inspection by region-based inspectors of preope. ational testing, staffing, training and qualification of personnel, and local leak rate testing.
The inspection involved 83 inspection hours onsite and 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> at the Nuclear Training Center.
Results: No violations were identified.
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8601170305 960110 PDR ADOCK 05000354 G
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DETAILS 1.0 Persons Contacted Public Service Electric and Gas Company (PSE&G)
- A. Giardino, QA Manager - Engineering and Construction
~ R. Griffith, Principal QA Engineer
- S. LaBruna, Assistant General Manager - Operations
- M. Metcalf, QA Startup Engineer
- R. Salvesen, General Manager - Operations
- P. J. Kudless, Maintenance Manager - HC0
- C. W. Lambert, Site Engineering
- S. Ketcham, Training Supervisor - HC0
- R. Edmonds, Assistant Manager, NTC
- D. Parks, Training Supervisor, Radiation-Protection
- J. R. Lovell, Radiation Protection / Chemistry Manager J. Forcier, System Test Engineer (STE)
F. Heinz, STE C. Jaffee, Startup Engineer H. Snyder, STE W. Brammeier, QA Startup Engineer G. Duncan, Sr. ISI Supervisor E. Maloney, ISI Supervisor B. Preston, Licensing Manager NRC
- C. Petrone, Lead Reactor Engineer M. Evans, Reactor Engineer J. Golla, Reactor Engineer
- L. Bettenhausen, Chief, Operations Branch
- R. Borchardt, Senior Resident Inspector
- Denotes.those present at the exit meeting on December 18, 1985.
2.0' Preoperational and Detailed Test Procedure Review and Verification 2.1 Scope The Preoperational Test Procedures (PTPs) and Detailed Test Procedures (DTPs) listed below were reviewed in preparation for test witnessing, for technical and administrative adequacy and for verification that testing is planned to satisfy regulatory guidance
.-and license commitments. They were also reviewed to verify licensee review and approval, proper format, test objectives, prer.equisites,
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initial conditions, test data recording requirements and system return to normal.
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PTP-SA-1, Redundant Reactivity Control System, Revision 0.
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PTP-SB-2, Response Time Testing, Revision 0.
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DTP-SB-0002, Turbine Stop Valve / Recirculation Pump Trip
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Response Time and RPS Trip Response Time, Revision 0.
DTP-SB-0003, Turbine Control Valve Fast Closure / Recirculation
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Pump Trip and RPS Trip Response Time, Revision 0.
DTP-SB-0007, Reactor Vessel Low Level Scram Response Time Test,
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Revision 0.
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DTP-SB-0010,-Scram Discharge Volume High Level, MSIV Closure /RPS
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Trip Logic Response Time, Revision 0.
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DTP-SB-0011, RPS/ Intermediate Range Monitor Response Time, Revision 0.
2.2 Discussion PTP-SA-1
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During the review of PTP-SA-1, Redundant Reactivity Control System-(RRCS), the inspector compared the test's accsptance criteria to the acceptance criteria of G.E. Preoperational Test Specification 22A2271 AZ Section 833, Revision 2, HCGS FSAR Chapter 14 and HCGS FSAR Question 640.20 Startup Item 9, ATWS Test. At the completion of the review, the inspector noted that G.E. Specification Section B33.3.2.6, " Verify the capability of adding boron : solution to the SLCS storage tank via the mixing tank," and Section 833.3.4.2, " Verify that a high vessel pres-sure in:tiation signal trips the MG sets after a 25 second time delay if reactor power has not been reduced," were not tested in PTP-SA-1.
After discussion with the licensee, it was determined that these sections of the G.E. specifications had been deleted.
Section 833.3.2.6 was deleted in FDDR No. KT1-1646, since there is no mixing tank installed in the Hope Creek SLCS.
Section 833.3.4.2 was deleted in FDDR No. KT1-1576.
This testing was deleted because the total time delay between actuation of the reactor vessel pressure sensor and RPT breakers arc suppression is tested in DTP-SB-0004.
The time delay is s230 millisecoids as required by G.E. Specification Section 88.5.12.8.
PTP-SB-2
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Response Time Testing is accomplished by means of 14 Detailed Test Procedures. At the time of this inspection, 5 of the 14 DTPs had been approved for testing. The inspector reviewed DTP-SB-0002, DTP-SB-0003, DTP-SB-0007, DTP-SB-0010 and DTP-SB-0011 and compared the DTPs acceptance criteria to the acceptance
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. criteria of G.E. Preoperational-Test Specification 22A2271 AZ, Revision 2.
Ali G.E. acceptance criteria were included in the p_rocedures reviewed by the inspector.
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lNo unacceptable conditions weie identified within the scope of this-review.
3.0 Preoperational Test Witnessing.
3.1 Scope
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Testing witnessed by;the inspector included'the observations of overall crew performance stated in Paragraph 3.0 of Inspection Report'
-No. 50-354/85-18.
Portions of the following PTPs were witnessed:
PTP-GU-1, Filtration, Recirculation and Ventilation System
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PTP-KJ-3, Emergency Diesel CG 400
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-PTP-SV-1, Remote Shutdown Panel
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Discussion
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'The inspector observed Section 8.23 of PTP-GU-1, operation of HPCI
. pipe chase room isolation dampers.
Testing was conducted in accordance with the criteria of Paragraph 3.1 above with full QA coverage during the portions witnessed by the inspector.
3.3' PTP-KJ-3 The inspector witnessed a portion of the Lockout Relay testing being conducted under PTP-KJ-3, Emergency Diesel CG 400. All testing observed satisfied the criteria of Paragraph 3.1 above.
3.4 PTP-SV-1
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Several times during the inspection, the inspector observed steps-8.2.20, 8.2.21, 8.2.39 and 8.2.40 being performed for various valves. These steps involve placing certain transfer switches into the emergency position and verifying all operable devices (listed in Appendix G, Table II, of PTP-SV-1) can be operated from the Remote q
Shutdown Panel (RSP) both in the open and close direction or ste*t/
stop for the pumps and proper light indications. A'eso,'nonopeu.
bility of these devices from the Main Control Room (MCR) is verified.
The transfer switches are then placed into the normal position and
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the devices :(Appendix G, Table II) are verified: operable from the
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.RSP. :The inspector-noted that all testing was being conducted in faccordance with the criteria of Paragraph 3.1 above.
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3.5 Findings ~
No unacceptable conditions were observed.
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4.0 = Plant Tours The, inspector made several _ tours of the various areas of the facility to observe work in progress, housekeeping, cleanliness controls and status of construction and preoperational test activities.
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No unacceptable conditions were observed.
5.0_ 0perational Staffing and Training
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5.1; Requirements and References Technical Specificaticn Section 6.
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ANSI /ANS-3.1-1981, Selection, Qualification and Tr'aining of
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Personnel for Nuclear Power Plants.
FSAR Chapter 13.
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Regulatory Guide 1.8-1975, Personnel Selection and Training.
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5.2 Program Review The following procedures were reviewed in accordance with the requirements referenced in Paragraph 5.1.
Hope Creek Training Procedures Manual includiag thirty-r e of
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the enclosed training plans.
Nuclear Department Training Center Administrative Policies.
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Instructor Development Manual.
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Hope Creek Operations Organization and Manning Charts for
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Operations, Chemistry, Radiation Protectica, Technical and Maintenan:e Departments.
TE-TI.ZZ-001(Q), Trainina Records Management.
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p This review verified that:
-ThelicenseehiscommittedtoANSI/ANS-3.1-1981 per Technical
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Specification 6.4 in the selection and training of their staff-.
Procedures were in place to provide for. updating lesson plans.
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'A Nuclear Training'0versight Committee and Training Review
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Groups had.been established to assure fo'rmal periodic reviews are conducted and documented.
The responsibilities for administering the training programs
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have been assigned, including scheduling, assignment of instructors, examining and record keeping.
A program and procedures were in place to train and qualify
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training instructors.
a 5.3 Program Implementation Implementation of the program was verified by the following:
The department managers were interviewed and organization
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charts were reviewed, as well as personnel qualifications, experience, and staffing levels.
Personnel from various departments were also selected for interviews. An examin-ation of their personnel and training folders were reviewed to confirm their qualifications.
The General Employee Training lesson plans were reviewed,
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one GET classroom lesson was witnessed, and the GET exam was reviewed.
The administration of the exam was also witnessed.
The GET training was effecti_vely implemented and met regulatory requirements and commitments.
The training records of at least two individuals in each of the
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following job classifications were reviewed:
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Principal staff members Maintenance craftsmen
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Instrumentation and Control technicians
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Chemistry technicians
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Radiation Protection technicians
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QA/QC inspectors
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These records contained the information required by the depart-mental training programs and regulatory requirements.
They were accessible, properly filed, legible and curren..
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Ac least one individual in each of the preceding job classif t-
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cations was interviewed. The individuals were qualified and had.
received the training listed in their training records. All individuals interviewed believed that the training program was good. Some thought the material was presented too fast, while others (with more experience or education) thought it was too slow. All thought the emphasis on practical knowledge was good.
The Nuclear training facility was excellent. The labs were extensively equipped with equipment for hands-on-training. The classes witnessed by the inspector were taught by experienced instructors. The lectures followed the detailed lesson plans.
A new hire was interviewed.
He expressed satisfaction with the
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training. program provided to him.
The licensee has established a Nuclear Training Oversight Com-
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mittee (NTOC) to assure formal periodic review and documentation of the responsiveness'and program requirements of the Hope Creek Operations training program. A training review group has been established for the Operations, Maintenance, Chemistry, Instru-mentation and Controls, and Radiation Protection departments.
Members of each group include the station department head and the nuclear training department head.
Each group is required to meet at least quarterly and recommend changes or additions to the training program in their area.
They are required to review
' and respond to the regulatory and industry requirements and standards for each job classification in their department. They recommend improvements to the departmental training programs when problem areas arise or weaknesses become apparent. They also review and propose topic content for annual requalification and continuing training.
The inspector reviewed the charter for the Nuclear Training Review Group and the Nuclear Training Oversight Committee (NTOC)
Meeting Minutes for the meetings held September 18, June 4, and
'i March 26, 1985. The topics discussed at these meetings appeared to provide meaningful and useful feedback to the training program.
Review of the stiffing levels for the following departments
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indicated the following (approximate) staffing levels.
Authorized Actual Maintenance /ICC 158 150 Operations
90 (Includes 7 contractors)
Chemistry
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Authorized Actual Radiation Protection
43 (Includes 6 contractors)
Technical
36 (Includes 3 contractors)
Nuclear Training Department (Includes Salem)
115 103 The department managers are in the process of hiring personnel to fill the remaining vacancies.
5.4 Findings No violations were identified. The station appears to be nearly fully staffed, personnel are qualified, and an effective training program is in place.
However, further inspections of training programs for licensed operators and other personnel will be per-formed during future inspections.
6.0 Local Leak Rate Testing 6.1 Documents Reviewed
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Preoperational Test Procedure No. PTP-GP-2, Revision 0, " Local
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Leak Rate Test" 0A/QC Surveillance Plan PTP-GP-1
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Hope Creek Technical Spec 1rication Section 3.6.1.2, " Primary
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Containment Leakage, Limiting Condition for Operation" LLRT Results Running Total
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Calibration Data for the following Leak Rate Monitors (LRM's):
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Serial Nos. 8033, 7535, 115, 687-1, 125, 141, 6981, 687-2, 687-3 6.2 Scope of Review The inspector reviewed the above documentation to ascertain that the licensee's LLRT program was conducted in compliance with the regula-tory requirements of 10 CFR 50, Appendix J, and applicable industry standards. The inspector also witnessed local leakage testing and held discussions with the licensee regarding the documentation of the test results, the repair and retesting of failed tests and the relationship of these items to the "as-left" condition of containment as applied to preoperational ILRT results.
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6.3 Test Witnessing On' December 10, 1985, the inspector witnessed a Type C LLRT of Pene-tration No. PIA, Main Steam Isolation Valve V028.
The test was con-ducted in accordance with approved Procedure No. PTP-GP-2, Local Leak Rate Test," Appendix 4A.
The test volume was pressurized to 5 psig and the leakage was measured with LRM S/N 141 by flowing in.
The valve leakage was measured at 6250 sccm which is a test failure-since this valve has a maximum allowable leakage of 5427 sccm.
This penetration (which is exempt from the.6L, criteria) is protected by a constant pressure seal system which regulates the pressure between the inboard and outboard MSIV's at 5 psid above reactor vessel pres-sure during a LOCA.
'On December 11, 1985, the inspector discussed the results of this test with the licensee. The licensee stated that their plans included the possibility of conducting the ILRT with the valve in the present con-dition as long as it is reworked afterward, local tested again, and the additional leakage added to the preoperational ILRT result. Also, the effect of the seal system on ILRT was discussed. The licensee stated that during ILRT, the seal system will be regulated to 5 psid below containment pressure. The inspector accepted this but question-ed the licensee as to whether or not this seal pressure would be moni-tored during the ILRT. The licensee stated that there wasn't anything presently written into the procedure concerning this but they would make an additional test requirement to periodically check the seal system pressure during the ILRT.
This is needed to ensure no seal system inleakage during the ILRT.
On December 11, 1985, the inspector witnessed a Type C LLRT of
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Penetration No. J38, Valve No. GS31, of the H /0 analyzer system.
2 2 This test was conducted according to Appendix 4A of PTP-GP-2 using LRM No. 8023. This procedure is a flow in test at pressure P, +
.5 psid, - O psid (P, is 48.1 psig). The valve measured 0.3 sccm leakage wnich was a satisfactory test.
The inspector observed the performance of these tests to ensure that
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prerequisites were met such as proper valve lineup and calibration of test equipment and to ascertain the qualifications of the test
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personnel. In all cases observed, the people involved in local test-L ing (technicians and test engineers) were found competent in local leakage testing, and were knowledgeable of requirements and the use l
of the test equipment.
No unacceptable conditions were identified.
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7.0 Integrated Leak' Rate Testing Procedure Review The inspector reviewed Preoperational Test Procedure No. PTP-GP-1, "Pri-e.
mary Containment Integrated Leak Rate Test (CILRT)," Revision 0.
This
' procedure appears to be adequate-(complies with 10 CFR 50, Appendix J, and technical specificatien requirements) with the exception of the one addi-tion discussed in Section 6.3 of this report. No unacceptable conditions were identified.
8.0 QA/QC Involvement QA/QC-involvement in LLRT appeared to be adequate. QA personnel were well informed on LLRT activities and had interest in monitoring specific LLRT's such as penetrations with a history of leakage and otherwise large penetrations.
No unacceptable conditions were identified.
9.0 Management Meetings The licensee's management was informed of the scope and purpose of the inspection at the entrance meeting on December 9, 1985. The findings of the-inspection were discussed with licensee representatives during the course of the inspection and presented to licensee management at the December 18, 1985 exit meeting (see paragraph I for attendees).
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