IR 05000322/1982010
| ML20054F688 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 05/21/1982 |
| From: | Bettenhausen L, Pullani S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20054F679 | List: |
| References | |
| 50-322-82-10, NUDOCS 8206170219 | |
| Download: ML20054F688 (11) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 50-322/82-10 Docket No. 50-322 License No. CPFR-95 Priority Category
.B Licensee: Long Island Lighting Company 175 East Old Launtry Road Hicksville, New York 11801 j
Facility Name: Shoreham Nuclear Power Station Inspection At: Shoreham, New York Inspection Conducted:
April 26-29, 1982 Inspectors:
5/ao/B2 S. VT'Fu1}6ni, Reactor Inspector da'te
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Approved by:
Ch/
S/2//tI2_.
A L.H. B5ttenhausen, Chief, Test date
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Program Section, Engineering Programs Branch Inspection Summary:
Inspection on April 26-29, 1982 (Report No. 50-322/82-10)
Areas Inspected:
Routine, unannounced inspection of licensee action on previous inspection findings, Primary Containment Local Leak Rate Test witnessing, overall startup test program review, startup test procedure review, startup procedure verification, and tours of the facility.
The inspection involved 18 inspector hours in office and 40 inspector hours onsite by one region based NRC inspector.
Results: One violation was identified (Measuring and Test Equipment not properly calibrated - Paragraph 3.4).
l 8206170219 820524 PDR ADOCK 05000322 (D PDR
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DETAILS 1.
Personnel Contacted 1.1 Long Island Lighting Company (LILCO)
- J. Hawrylak, Nuclear Engineer
- J. Rivello, Plant Manager
- T. Rose, 0QA Engineer
- E. Youngling, Startup Manager 1.2 General Electric (GE)
- H. Upton, STO&A Lead Engineer 1.3 Stone & Webster (S&W)
- S. Aitken, Startup Test Engineer
- W. Matejek, Lead Advisory Engineer
- R. Thomson, Startup Test Engineer 1.4 U. S. Nuclear Regulatory Commission J. Higgins, Senior Resident Inspector
- L. Bettenhausen, Chief, Test Program Section The inspector also interviewed other licensee and contractor employees during the course of the inspection.
- denotes those present at the exit interview on April 29, 1982.
2.
Licensee Action on Previous Inspection Findings (Closed) Unresolved Item (50-322/82-06-01) RPS MG Sets not Preoperationally Tested for Time to Come to Speed Test Change Notice 1 on March 26, 1982 includes this test in Section 8.5.2.2 and Table 7 of PT.312.001A-1, Revision 1, and PT.312.0018-1, Revision 1, 120 V AC RPS MG Sets Preoperational Tests.
The Test Change Notice was approved by the Joint Test Group on March 29, 1982 and by Operational Quality Assurance on March 30, 1982.
This item is closed.
3.
Primary Containment Local Leak Rate Testing 3.1 Documents Reviewed
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ANSI /ANS-56.8-1981, Containment System Leakage Testing Requirements
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FSAR Section 14.1.3.7.25, Primary Containment Local Leak Rate Test Types B and C
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PT.654.003, Revision 1, Primary Containment Leak Rate Test -
Type C
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Handwritten (not approved as of April 28,1982) Procedure used for " Accuracy Check" of Flow Measuring and Test Equipment Panels 42-1, 2, 3, and 4 Manufactured by Brooks Instrument Division of Emerson Electric Company
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Records of " Accuracy Checks" of the Flow Measuring and Test Equipment Panels
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Records of Calibration of Flow Standard, M&TE-15, Model 11, Manufactured by Volumetric, Used for the " Accuracy Check" of the Flow Measuring and Test Equipment Panels.
3.2 Scope of Review The inspector reviewed the above documents to ascertain compliance with regulatory requirements of 10 CFR 50, Appendices B and J, and licensee commitments in FSAR.
3.3 Test Witnessing On April 28, 1982, the inspector witnessed a Type C Local Leak Rate Test of Primary Containment Isolation Valve MOV-031C in the Residual Heat Removal System suction line from the suppression pool (Penetration X-9C).
The test was being conducted in accordance with approved procedure PT.654.003.
Instrument air was used to pressurize the test volume between the two disks of gate valve MOV-031C through a valve body test connection.
The test pressure was 46.5 psig (44 psig, Pa, the accident pressure + 2.5 psig to compensate for the static head of water on the suppression pool side of MOV-031C).
The test volume could not be pressurized to the required test pressure due to gross leakage through valve seat (s). The test was declared a failure and the valve was to be turned over to maintenance for adjustment or repair.
The inspector observed the performance of the test to ascertain that prerequisites were met, proper precautions were taken, measuring and test equipment was properly calibrated, the test was conducted in accordance with the procedure, test crew actions were correct and timely, and the required data were recorde.
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3.4 Findings 10 CFR 50, Appendix B, Criterion XII, requires measures to be established to assure that Measuring and Test Equipment used in activities affecting quality is properly calibrated to maintain accuracy within necessary limits.
LILCO Startup Manual, Section 4.4.3, requires individual instrument calibrations to be performed using either approved Station Procedures or Manufacturer's Technical Manual, and, if this cannot be done, the calibration is to be performed by the manufacturer or an approved testing laboratory whose standards are traceable to-the National Bureau of Standards.
Contrary to this, on April 28, 1982, Flow Measuring and Test Eauipment Panel 42-4, used for the Local Leak Rate Testing of Primary Cor?.ainment Isolation Valve MOV-031C, was not properly calibrated in that:
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Original records of calibration either by the licensee, the manufacturer, or an approved testing laboratory were not available at the site as of the date of the test, 2.
Instead of the required calibration, an " accuracy check" was performed periodically using a hand-written procedure which was signed by the Test Engineer but not reviewed or approved as of the date of the test.
3.
PT.654.003, Revision 1, Primary Containment Leak Rate Test - Type C, the procedure used for the testing of MOV-031C, requires in Steps 3.3 and 8.6 that certain corrections be applied to the test results, if the temperature or pressure of air used for the test differs from the temperature or pressure of air used for the calibration.
The " accuracy check" procedure had no provisions for recording the calibration temperature.
The inspector determined that all other Flow Measuring and Test Equipment Panels, used for Local Leak Rate Tests performed prior to April 28, 1982, were not properly calibrated in that Items 1, 2, and 3 above also apply to these panels.
This is a violation (50-322/82-10-01).
4.
Overall Startup Test Program Review 4.1 Documents Reviewed (
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FSAR Chapter 14, Initial Tests and Operatiors,
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RG 1.68, November 1973, Preoperational and Initial Startup Test l
Program for Water-Cooled Power Reactors, l
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LILC0 Startup Manual,
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LILC0 Operational Quality Assurance Manual,
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Station Procedure 12.075.01, Revision 2, Administration of Startup Testing,
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Station Procedure 12.006.01, Revision 13, Station Procedures.-
Preperation, Review, Approval, Change, Revision & Cancellation,
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Plant Procedure Status Listing, January 13, 1982,
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Startup Procedure Status Listing, April 16, 1982,
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Startup Procedures Manual (compilation of Startup Procedures issued to date),.
NUREG-0420, April 1981, Safety Evaluation Report Related to the
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Operation of Shoreham Nuclear Power Station, Unit No. 1,
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FSAR Volumes 14 and 15, NRC Requests and Responses, Section 413,.
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General Electric Report GEZ-6410, Shoreham Nuclear Power Station Unit 1 Transient Analysis Design Report Supplement 1, and
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General Electric Document 386HA923, MPL No. A12-3021, Revision 0, Startup Transient Test Specification.
4.2 Scope of Review
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The inspector reviewed the above documents to ascertain that the licensee has developed administration controls which will assure that the startup test program will be prepared, performed, and evaluated in accordance with regulatory requirements and licensee commitments in FSAR.
In FSAR Section 14.1 and Appendix 3a, the license commited to the November 1973 issue of RG 1.68, Preoperational and Initial Startup Test Programs for Water-Cooled Power Reactors.
The inspector reviewed i
the licensee's startup test program against the requirements of this Regulatory Guide.
4.3 Findings As a result of the above review, the inspector discussed several questions with licensee personnel.
Some of these. questions were satisfactorily resolved by the licensee personnel and were further verified by the inspector by review of pertinent records. Other
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questions need further licensee actions for their resolution. A brief discussion of these questions follows:
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RG 1.68, Appendix A, Paragraphs B.2.c and C.2.a, requires evaluation of control rod sequences to verify safety criteria and check operation of control rod inhibit functions during approach to critical and up to power levels at which the inhibit functions may be bypassed. The inspector inquired which startup test procedure includes these tests.
Licensee personnel stated that this is done using Station Procedure 22.001.01,_ Startup-Cold Shutdown to 20 Percent Power, as Pre-requisite 6.1 to Startup Test Procedure STP-4, Revision 1, Full Core Shutdown Margin. The inspector verified this by review of above procedures.
RG 1.68, Appendix A, Paragraph C.2.d, requires comparison of actual critical control rod configuration against predicted configuration during low Power Testing.
The inspector inquired which startup test procedure includes this test.
Licensee personnel stated that the intent of this test is met by performing a shutdown margin demonstration test as described in Step 8.2.4.2 and Table 4.8.2-7 of STP-4, Revision 1, Full Core Shutdown Margin. The inspector verified this by review of this procedure.
RG 1.68,. Appendix A, Paragraph C.2.f, requires verification of response-of effluent radiation monitors using known sources during Low Power Testing. The inspector inquired which startup test procedure includes these tests.
Licensee personnel stated that these tests are performed as Prerequisites (Section 6.0) to STP-1, Revision 0, Chemical and Radiochemical Tests, using Station Procedures:
SP-74.025.01, Airborne Process Radiation Monitor Calibration and Sensor Check; SP-74.025.03, Liquid Radioactive Discharge Radiation Monitor Calibration and Sensor check; SP-74.631.08, Offgas Vent Discharge Radiation Monitor Calibration and Sensor Check. The inspector verified that the above procedures are scheduled to be written by review of Plant Procedure Status Listing, January 13, 1982.
RG 1.68, Appendix A, Paragraph C.2.g, requires a final leak test of-Reactor Coolant System during Low Power Testing. The inspector inquired how this requirement would be satisfied.
Licensee personnel stated that this test will be performed using Station Procedure SP-22.009.01, Inservice Reactor Pressure Boundary Leak Test.
The inspector verified that this procedure is written by reviewing Plant Procedure Status Listing, January 13, 1982.
RG 1.68, Appendix A, Paragraph C.2.1, requires a functional test to be performed for the Reactor Vessel Head Cooling System at operating temperature and pressure during Low Power Testing.
The inspector inquired which startup test procedure includes this test.
The licensee personnel stated that this test is included in STP-71, Revision 0, Residual Heat Removal System, Section 8.3.2, Procedure Step 5.
The inspector verified this by review of this procedure.
RG 1.68, Appendix A, Paragrpah D.2.g, requires chemical analysis of
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fluid systems at 25, 50, 75, and 100 percent power plateaus during
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Power Ascension Testing. The inspector inquired which startup test procedure includes this test.
The licensee personnel stated that-this requirement will be met by the tests described in STP-1, Rev. O, Chemical and Radiochemical Tests. The inspector verified this by review of the above test procedure.
RG 1.68, Appendix A, Paragraph D.2.o, requires demonstration of rod pattern exchange during Power Ascension Testing, at the maximum power level that rod exchange will be permitted during operation. The-inspector inquired at what power level that this test would be performed.
The licensee personnel stated that this test is scheduled to be performed at Test Condition 5 which corresponds to approximately by 60 percent power. The exact power level would be determined by the Reactor Engineer in accordance with Pre-Conditioning Interim Operating Management Recommendations (PCIOMR), a General Electric recommendation-to minimize fuel failures by Pellet Clad Interaction. This is described in STP-8, Revision 1, Control Rod Sequence Exchange, Paragraph 2.1.
The inspector verified this by review of this procedure.
RG 1.68, Appendix A, Paragraph D.2.r, requires Recirculation Pump Trip (RPT) single pump trip and two pump trip at 100 percent power. The inspector noted that single pump trip is included in STP-30, Revision 1, Recirculation System; and inquired which startup test procedure includes two pump trip. The licensee personnel stated the two pump test is not planned as a separate test but would be done as a part of turbine trip at 100 percent power. This is described in STP-27 Revision 1, Turbine Trip and Generator Load Reject, Paragraph 8.1.4, Procedure Step 5.f.
The inspector verified this by review of th;s procedure.
As stated earlier, the licensee is committed to November 1973 (Revision 0) issue of RG 1.68.
In Appendix A, Paragraphs D.2.s and D.2.t, this requires separate turbine trip and generator trip testing, both at 100 percent power. The inspector noted that the licensee's startup test program includes the turbine trip at 100 percent power but not the generator trip at 100 percent power.
The inspector further stated
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that the licensee did not take any specific exception in FSAR for not
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performing the generator trip at 100 percent power.
The inspector explained that RG 1.68, Revision 1, Appendix A, Paragraphs 1.1, permits
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I combining the turbine trip at 100 percent power with the generator trip at 100 percent power, under specified conditions.
However, the turbine trip alone will not satisfy the requirements of RG 1.68,
Revision 1, either. The licensee personnel stated that the generator trip at 100 percent power is not scheduled due to potential damaging effects on the machine and that the difference in dynamic response of the plant between the turbine trip and the generator trip is not l
significant. The inspector reviewed General Electric Report. GEZ-6410, Shoreham Nuclear Power Station Unit 1 Transient Analysis Design Report, Supplement 1, Tables 5.3.1 and 5.4.1, which shows an increase in reactor vessel dome pressure from an initial dome pressure of 1020
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psia by 136 psi and 138 psi respectively, for the turbine trip and the generator trip.
This shows the generator trip to be a slightly move severe transient.than the turbine trip. The inspector further reviewed General Electric Document 386HA923, MPL No. A12-3021, Startup Transient Test Specification.
Paragraph 4.6.2.2.2 of this document gives the basis for the choice to perform a turbine or ger arator trip at 100 percent power.
The basis is that, at 100 perce : power, there is very little difference in plant response for a partial bypass valve. capacity plant such as Shoreham. The inspector reviewed NUREG-0420, and FSAR Volume 14 and 15, NRC Requests and Responses, Section 413, to determine the licensing status related to these tests. Conclusive justification for not performing the generator trip test was not presented._.The turbine trip / generator trip requirements of RG 1.68, either Revision 0 or Revision 1, do not appear to be satisfied. This is an Unresolved Item (50-322/82-10-02).
No violations or deviations were identified.
5.
Startup Procedure Review 5.1 Documents Reviewed
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STP-5, Revision 1, Control Rod Drive System FSAR Section 14.1.4.8.5, Test Abstract, Control Rod Drive System
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Startup Test
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General Electric Document 22 A 5727, Revision B, Startup Test Specification, Test Number 5 - Control Rod Drive System
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Proposed Technical Specification, February 1, 1982, Shoreham Nuclear Power Station - Unit 1, Section 3/4.1.3, Control Rods
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IE Bulletin 80-17, Failure of 76 of 185 Control Rods to Fully Insert During a Scram at a BWR
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5.2 Scope of Review
The inspector reviewed above documents to ascertain that STP-5, Revision 1, Control Rod Drive System, a heatup phase startup test procedure selected for review on a sampling basis, is consistent with regulatory requirements, FSAR commitments, croposed Technical Specification requirements, regulatory guidance, and applicable codes and standards.
5.3 Findings
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The inspector discussed several questions with licensee personnel.
Some of these questions were satisfactorily resolved by licensee personnel and were further verified by the inspector by review of pertinent records. Other questions need further licensee actions for their resolution. A brief discussion of these questions follows:
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RG 1.68, Appendix A, Paragraph D.2.b, requires friction testing of all control rod drives at rated pressure. A: described in STP-5, Revision 1, Section 2.4, this test is performed only for 4 rods. The inspector stated that this limited sample also does not conform to the test purpose, stated in STP-5, Section 1.0:
"The purposes of the Control Rod Drive System test are:
1.1 To demonstrate that the Control Rod Drive (CRD) System operates properly over the full range of primary coolant temperatures and pressures from ambient to operating.
1.2 To determine the-initial operating characteristics of the-entire CR0 system."
The inspector noted the reasoning for not testing all CRDs, as given in STP-5, Paragraph 2.4.3, which states in part, "it can be demonstrated from tests on a small number of CRD's that no significant difference in friction exists between cold and hot operating conditions". The licensee personnel stated that this position is based on previous CRD testing at other plants. The inspector stated that, since CRD instella-tion practices could differ among plants, in the absence of friction testing of all CRDs at rated pressure, the above position should be demonstrated by the Shoreham CRD test data. The licensee personnel agreed to make necessary changes in the test procedure. This is an Inspector Followup Item (50-322/82-10-03).
The inspector inquired why the Level 1 and Level 2 Acceptance Criteria for maximum CRD withdraw speed, shown in.STP-5 Paragraphs 3.1.1 and 3.1.2 as 3.6 and 3.0 + 0.6 inches per second respectively, differs.
The licensee personnel stated that 3.0 inches per second is the nominal value around which they try to adjust the speed and 3.6 inches per second is the ultimate maximum speed permissible, thus the two different criteria.
The inspector inquired as to the difference in Acceptance Criteria for mean scrau time for all operable CRDs and for three fastest CRDs in all groups of 2 x 2 array, as given in STP-5 Paragraphs 3.1.2 and 3.1.3 respectively, from their corresponding values in Technical Specifications Sections 3.1.3.2 cnd 3.1.3.4, since it appears that the TS values are truncated beyond the third significant digit without rounding off (e.g., 3.497 is truncated as 3.49 in TS 3.1.3.2).
The licensee personnel explained that STP-5 values are derived from FSAR analysis and therefore more accurate.
The difference is beyond the accuracy of test instruments and is therefore insignificant. This satisfied the inspector's concern.
The inspector stated that STP-5, Section 8.0, shows that the minimum requirement for successful completion of STP-5 excludes two subtests described in this test procedure, namely, Subtest 8.4, Scram Discharge Volume (SDV) Testing and SDV Valve Testing, and Subtest 8.5, SDV Fill
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Time. The inspector explained that these two subtests are required to be performed in accordance with IE~ Bulletin 80-17.
The licensee agreed to revise STP-5, Section 8.0 to include the subtests as require-ments for successful completion.
This is an Inspector Followup Item (50-322/82-10-04).
6.
Startup Procedure Verification 6.1 Procedures Verified
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STP-8, Revision 1, Control Rod Sequence Exchange
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STP-9, Revision 1, Water Level Measurements
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STP-16, Revision l', Selected Process Temperatures
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STP-21, Revision 1, Core Power - Void Mode Response
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STP-29, Revision 1, Recirculation Flow Control System
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STP-31, Revision 1, Loss of Turbine - Generator and Offsite Power.
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STP-36, Revision 1, Feedwater and Steam System
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STP-37, Revision 2, Reactor Building Closed Loop Cooling and Drywell Cooling
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STP-39, Revision 1, RBSVS and CRAC Chilled Water and Ventilation Systems
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STP-40, Revision 1, Condenser Offgas and Gaseous Radwaste System 6.2 Scope of Review The inspector verified that the above procedures were reviewed and approved in accordance with established administrative procedures and that the test objectives and procedure format were consistent with regulatory guidance and licensee commitments.
6.3 Findings No violations or deviations were identified.
7.
Unresolved Items Unresolved items are matters about which more information is required to ascertain whether they are acceptable or whether they are violations or deviations. Unresolved items are discussed in Paragraph.
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8.
Plant Tours-The inspector made several tours of the facility, including the reactor building, control room and turbine building.
During these tours, the inspector observed and evaluated work in progress, general equipment protection and cleanliness controls, and component tagging for safety and jurisdictional purposes.
No violations or deviations were identified.
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Exit Interview The inspector met with licensee management representatives (see Section 1
- for attendees) at the conclusion of the inspection on April 29, 1982._ The inspector summarized the scope and findings of the inspection at that time.
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