IR 05000338/1987035
| ML20236F805 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 10/23/1987 |
| From: | Blake J, Economos N NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236F745 | List: |
| References | |
| 50-338-87-35, 50-339-87-35, NUDOCS 8711020301 | |
| Download: ML20236F805 (10) | |
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.' UNITED STATES -
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' NUCLEAR REGULATORV COMMISSION -
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REGION 11,
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101 MARIETTA STHEET, N.W.
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' ATLANTA, GEORGI A 30323 a
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Report Nos.:
50-338/87-35 an'd 50-339/87 35
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Licensee': ! Virginia Electric and Power Company
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Ricimond, VA. 23261:
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. Docket Nos.:
50-338 and 50-339 License Nos.: 'NPF-4 and NPF-7 L
Facility Name:
North Anna 1 and 2
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- Inspection n
te : ' September 28
_0ctober 1, 1987
Inspecto f
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emw Date Signed:
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J. 4 Take, Section. Chief; Date Signed
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H ials and Processes'Section
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i'ision'of' Reactor Safety SUMMARY.
Scope:.This routine, announced inspection was in the-areas of installation of'
downcomer flow resistance plates in steam generators (S/G);" installation of'
reactor, vessel head. shielding (RVHS);. Pressurizer safety valve loop seal insulation -oven insta11ation;. licensee ~ action Lon Temporary Instruction (TI)
2515/84.
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I Results:
No violations or deviations were identified.
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.-t 8711020301 871027-L ADOCK 05000338 PDR PDR.
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REPORT. DETAILS
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1.
Persons Contacted Licensee Employees
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E. W. Harrell, Station Manager i
- M. L. Bowling, Jr., Assistant' Station Manager
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- J. A. Stall, Superintendent: Technical Services i
L. N. Hartz, Engineering Supervisor, Inservice Inspection (ISI)
- J. E. Wroniewicz, Supervisor Site Nuclear Engineer'ng G. Harkness,' Licensing Coordinator
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- J. Leberstien, Licensing Engineering j
- F. T. Termine11a, Quality Centrol Supervisor
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E. Taylor, Project Engineer, Steam Generator Downcomer Flow Resistance.
H. W.-Burruss, Project. Engineer, Vessel Head Shield Other licensee employees contacted included construction craftsmen, J
engineers, technicians, operators, mechanics, security force members, and'
office personnel.
l Other Organizations Westinghouse, (W) Nuclear Services Integration Division F. T. Swartz, Technical Director J. Tepley, Outage Coordinator J. Marburger, QA Lead Engineer Other licensee employees contacted included QC/QA inspectors, technicians,
and office personnel.
I NRC Resident Inspectors'
J. Caldwell, Senior Resident Inspector L. King, Resident Inspector
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- Attended exit interview 2.
Exit Interview
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The inspection scope and findings were summarized on October 1, 1987, with j
those persons indicated in paragraph 1 above.
The ins)ector described the areas inspected and discussed in detail the inspection findings.
No dissenting comments were received from the licensee.
The licensee did not identify as proprietary any of the materials provided to or reviewed bv the inspector during this inspection.
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Licensee Action on Previous Enforcement Matters This subject was not addressed in the inspection.
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Unresolved Items
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Unresolved' items were not identified during this inspection.
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Temporary Instructions (TI) Units 1 and 2 (25584)
j TI 2515/84,. Verification of Compliance With Order for Modification of i{
Licensee Primary Coolant System Pressure Isolation (Event) Valves the,
Reactor Safety Study (RSS), WASH-1400, identified in a PWR an intersystem l
loss.of coolant accident (LOCA) that a significant contributor to risk of
core melt accidents (Event V).
The design examined in the RSS contained in-series check valves isolating the high pressure primary coolant system i
(PCS) from the; low pressure injection system-(LPIS) piping.
The scenario-l which leads to the Event V accident _is initiated _ by the failure of these check valves to function as a pressure isolation barrier 'against reactor coolant system (RCS) pressure.
This 'causes an overpressurization and rupture of - the LPIS low pressure piping which results in a LOCA that bypasses containment.
To better define the Event V concern, all light water reactor licensees were requested by letter dated February 23, 1980, to provide specific information on such valve configurations in accordance with 10 CFR 50.54(f).
In addition, licensees were asked to perform individual check valve leak testing before plant startup after the'next scheduled outage.
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Based on licensee responses and the ongoing unsatisfactory operational j
experience at several plants, the NRC staff concluded that a valve-
configuration of concern existed; meaning that, when pressure isolation was provided by two check valves in-series, and when failure of one valve in the pair can go undetected for a substantial length of time, verifica-tion of valve integrity was aquired.
The staff concluded that, since these valves are safety-related, they needed to be tested periodically to
ensure low probability of gross failure, as a result, the staff determined l
that periodic examination of check valves was required to be undertaken by i
the licensees to verify that each valve was seated properly and func-l tioning as a pressure isolation device.
Such testing was intended to reduce the overall risk of an inter-system LOCA.
On April 20, 1987, the Commission issued an Order for Modification of License Concerning Primary Coolant System Pressure Isolation Valves, to i
specified PWR and BWR plants, requiring that the above described testing be implemented.
This Order included a Safety Evaluation Report (SER) and i
Technical Specification insert pages to require leak rate testing of Event V pressure isolation valves.
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TI 2515/56 was issued on - June 1,1981, for followup inspection of
implementation of the Event V orders.
It expired December 1983.
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present TI 2515/34, is. intended to verify satisfactory completion of licensee actions to implement the periodic testing of Event V valves as required by the aforementioned Order.
i Within these areas, the below listed valves were selected at random for a review of test records and compliance with actions required by the subject'
Order and TI 2515/84.
The valves are listed in Table 3.4-1 of North Anna's TS.
Reactor Coolant System Pressure. Isolation Valves
Unit 1 t
System Valve No.
j Low Head Safety Injection to Cold Leg j
. Loop-1 1-SI-83/1-SI-195 i
Loop-2 1-SI-86/1-SI-197'
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Loop-3 1-SI-86/1-SI-199 l
Unit 2 Low' Head Safety Injection to Cold Legs 2-SI-91/2-SI-99/2-SI-105 Low Head Safety Injection to Hot Leg 2-SI-128 J
Accumulator Discharge Check Valves 2-SI-170
Residual Heat Removal Isolation Valves MOV-2700/MOV-2701/M0V-2720A Low Head Safety Injection to Cold Legs and Hot Legs MOV-2890A, B, C, & D This inspection effort included review of: ' the plant's Technical Specifi-cation (TS) 3.4.6.2.2 Reactor Coolant Operational leakape rates and
limiting conditions for operation; a copy of the licensee s Event V Order I
and the original SER, procedure changes and records of hardware modifica-
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tions as applicable.
I The procedure for this leak test is PT-61.4, "RCS Pressure Isolation Valve Leakage Test."
This procedure was reviewed to verify that it is consis-l tent with TS requirements including:
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An acceptable test method is used.
This would include a direct volumetric leakage rate measurement or other equivalent means capable of-demonstrating that leakage rate limits given in the TS are not exceeded.
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The test procedure ensures that leakage rates obtcined are for l
individual valves rather than for combined components.
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The test' procedure requires that leakage rates received at ' test
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pressures.1ess than the maximum potential pressure differential-across the valve be adjusted by assuming leakage to be directly.
proportional to the pressure dit'ferential to the-one-half power as noted in the SER which accornpanies the Order, d.'
Acceptance criteria stated in the test procedure are in accordance with the TS.
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Verify that the procedure identifies required corrective actions-in l
event leakage rate results are unacceptable.
l For the valves: listed above, the inspector reviewed records of test completed during previous outages' for each of the two units as listed below.
Unit 1 Date Tests and Leak Rates in Gallons Per Minute (GPM)
Va1ve 6/14/87 9/10/86 12/19/85 9/15/84
1-SI-195
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1-SI-83 0.027
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1-51-86 0.032
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1-51-89 0.125
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1-SI-209
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Unit 2 5/30/87 3/26/86 1985 4/6/84 10/25/84 11/1/84 12/13/84 2-SI-170
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0.699 0.0114
0.345 MOV-2720A 2-SI-91
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2-SI-99
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2-SI-105
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2-51-128
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- No Test Performed
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I 5/30/87 3/25-31/86 1985 10/24/84 4/5/84 l
MOV-2890A 0.00076
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MOV-28900
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MOV-28900
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2-MOV-2700'
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' Discussions with the cognizant engineer disclosed that none of the subject isolation valves failed leak rate acceptance criteria of I gpm stipulated
,1 in the TS and, therefore, no repair-work has been performed during the l
time frame inspected.
In addition, test data was reviewed and verify that:
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Test records contain major test data including upstream and down-i stream pressures, leak volume per unit time (or equivalent), leakage rate adjustment calculations when required, and leakage rate accept-'
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ance criteria based on trending from previous tests where applicable.
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Recorded test frequency is in accordance with TS.
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As found leakage (i.e., prior to valve stroking, modification, adjustments,.etc.) is recorded, d.
Leakage rates trending has been documented and adequately evaluated by the licensee in accordance with the TS requirement, e.
No test data anomalies exist which indicate improper or inaccurate-
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testing.
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f.
Adequate corrective actions were taken for valves not meeting the acceptance criteria.
Within these areas, the inspector noted that although corrective action is taken when designated isolation valves fail to meet leak rate acceptance criteria, subject procedure PT-61.4 does not specifically stipulate that-such action be taken.
This observation was discussed with the cognizant
engineer who took immediate action to revise the procedure to include
corrective action requirements.
This matter was identified as an i
inspector followup item (IFI) for tracking purpu:ct.
Corrective Action for Valves not meeting Acceptance Criteria 338, 339/07-35-01.
Except for this IFI no violations or deviations were identified.
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6.-
. Modifications (37700),
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Installation of Downcomer Flow Resistance Plates (DFRP). in. Steam Generators, Units 1 and 2 (37700)
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The DFRPs are being installed in the steam generators (SGs) in order to reduce mass flow through the SG tube bundle and, therefore, reduce
.the potential for fluid elastic instability of steam generator tubes.
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Model 51 steam generators were originally designed and placed in service with DFRPs which were subsequently removed by W in an attempt.
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to. increase. flow through the tube bundle and ultimately reduce the potential for sludge depositfor, in certain areas of the tube sheet.
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The DFRPs are located in the upper part of the downcomer annulus between the tubes bundle wrapper and the SG shell.
They consist of 20' circular segments, each measuring 27"x13"x3/4" with two attached j
gussets for welding it to the wrapper.
Each plate segment had eight
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(8) drilled holes approximately 1-1/2 inches in diameter.
The plate material was identified as ASTM A285 Grade C.
Controlling documents-included:
EWR 87-572, " Installation of Downcomer Flow Resistance Plates in Steam Generators" i
DCP 87-026-3, " Steam Generator Flow Resistance Plate Installa-
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tion North Anna 1 and 2" Addition of tne DFPRs are regarded as non-safety modifications to q
safety-related equipment as part of the W steam generator improve-
ment.
Westinghouse stress analysis / evaluation used ASME Section III i
Subsection NG as a guide in calculating stress limits.
According to l
this analysis, stress loads resulting from the addition of DFRP j
assembly do not exceed original design load levels.
North Anna's FSAR section impacted by this design change / modification and non-l scheduled for revision include: 5.5.2.3.5 (changing the recirculation
ratio and internals); 15.2.8, 15.2.9, 15.4.3 and 15.4.22 (impact.on i
the results of certain accidents.
The licensee's evaluation showed that there were no unreviewed safety questions involved in this modification.
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.In addition to the aforementioned controlling documents, the inspector reviewed the following W field documents for technical content and compliance with referencing code requirements.
Welder Qualification Specification WQS 1000, Rev. 8 Shielded Metal Arc Welding of Carbon (P-1)
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and Low Alloy (P-3) Steels with Carbon (P-4)
Low Alloy (F-4) Steel Covered Electrodes l
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p Welding Engineering Procedure
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WEP 200,'Rev. A Welding Carbon and Low Alloy Steel by the l
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Shielded Metal Arc Process PQR-053 9/20/82; Ref.'ASME Sections IX and l-II NB WEP 310, Rev. A Weld Filler Metal Control for Product
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. Application
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i NDE-130, Rev. 0 Visual Examination L
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-NDE-310, Rev. O Magnetic Particle Examination Ref. ASME
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y Section III and V (80W81)
Applicable Drawings
N-8726-3-V-800, Rev. 0 NG Units 1 and 2 0FRPs Installation N-8726-3-V-801, Rev. 0 NG Units 1 and 2 DFRPs N-8726-3-V-802, Rev. O W S/G (Vertical), General Assembly and Final Fabrication
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Other quality records reviewed included pert.onnel qualifications and
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material documents.
These recorders were as follows:
i DFRPs:
W QR#14189, Rev. C and 13312, Rev. O, Certificate of
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i Conformafice - Radform Tool Co. Heat #65561, #64706, #50354, i
- U5059.
Welding Electrodes:
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- QR#9832, Rev. 9 7018 1/8"0 HT-401C4481 Atom Arc W
7018 3/32"0 HT-412K7107 Atom Arc j
W QR#12546, Rev. 3 7018 3/32"0 HT-20791 Teledyne Mckay Magnetic Particle Powder:
W QR#35.34, Rev. O Batch #E66010, type 8A Red Pcvder
1 Welder Qualifications:
M151, M159, M139, M147, M146 and M83 Visual Examiners:
Vi/ Level II M.J.S., R.J.H., L.D.M., and J.D.E.
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Nonconformance reports (NRC) generated.during this modification which l
were reviewed for completeness, accuracy and adequacy of disposition
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included VRA-87-00026, VRA-87-00027, VRA-87-00028._ Of these, two?
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involved inadvertently generated arc strikes on the transition cone -
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area of the shell of SGs A and B, respectively, these were being
repaired under ASME Section XI repair program.
The third NCR related to some small items dropped through the staging down the S/GLannulus.
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Disposition.of these NCRs was being pursued by-the licensee._. On.
September 30,1987,'the inspector entered _S/Gs A and C and observed the field and shop fabricated welds for appearance and workmanship j
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quality, j
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Within these areas, no violations or deviations were identified.
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Reactor Pressure Vessel (RPV) Head Radiation Shield Units 1 and 2 In an; effort to reduce current high men-rem expenditures resulting from work performed on the RPV head area, the licensee installed
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permanent. reactor vessel head radiation shielding fixtures in each :
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unit.
Each shield is a six foot high cylinder made of one inch: thick.
q ASTM A36 Grade A steel plate material which is permanent 1y' attached ~
i to a circular structural member and supported by special mounting
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devices attached to the three existing RPV head lifting-rig lugs.
j The shields were designated as QA category 1, Group A.
The licensee
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will revise North Anna's FSAR Section 5.1.1 to include the shield.
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By stress analysis calculations W has shown that the additional
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weight on the RPV head lightning rig does not violate NUREG-0612
stress limits.
Moreover W has indicated that their evaluation showed
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that the additional weight of 11,000 lbs. will have a minimal affect on the RPV supports, lugs and control rod mechanism.
The. work was performed under design change DCP-85-06, Rev.1 and installation procedure 43-HWB-20190.
The inspector selected, Unit 2 shield for
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review of quality records.
Documents reviewed purchase order M.N.
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90292-D, Drawing No.1777E43, Rev. 5, the manufacturers certificate
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of compliance, VEPCOs certificate of conformance, certified material test reports and deviations notices.
At the time of this inspection,
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s both shields had been installed and access restrictions imposed by
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health physics precluded a close visual examination at this time.
J Within the areas examined, no violations or deviations were identi-fied, j
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Safety Valve Loop Seal Insulation Ovens Units 1 and 2 This work effort was performed as a followup to that documented in Inspection Report No. 338, 339/87-12 and pertains to modifications i
performed to qualify piping and supports for Design Basis Transient l
and accident conditions.
The modifications became necessary when j
stress analysis on pressurizer safety and relief valves piping showed the original piping analyses did not consider the fluid transient loads associated with a water slug passing through the system.
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Results of a subsequent analysis of safety valve transients showed unacceptable fluid transient loads on piping and supports downstream of the safety valves assuming " cold" loop-seal water temperatures upstream of the safety valves.
This work effort covers the licensee's fix for maintaining loop seal water temperature > 400 F.
In order to reach and maintain the water in the logs seals at the I
aforementioned temperature, the licensee installed metal (stainless I
steel) reflective thermal insulation boxes which enclose the safety valve loop seal piping.
The controlling document for this modifica-l tion was 0C-84-69-1, Pressurizer Safety Valve Loop Seal Insulation
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Ovens.
The insulation boxes were manufactured by Diamond Power Speciality Company per VEPCOs specification NAS-2017, " Metal
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Reflective Thermal Insulation Boxes for Pressurizer Loop Seal l
Piping." The specification provided requirements for the engineering design, fabrication testing and shipment of the insulation.
They were seismically designed and classified as non safety-related category Quality Group E.
VEPC0's engineering review and safety analysis report indicated the design change did not involve an Unreviewed Safety Question" as defined in 10 CFR 50.59.
Proper function of the insulation ovens will be verified through temporarily installed and strategically located thermacouples.
The. inspector reviewed available fabrication quality and installation records for compliance with specification requirements.
A field inspection was performed in Unit 1 to observe the installed ovens and the thermocouple.
Within the areas exanined, no violations or deviations were identi-fied.