IR 05000280/1989018
| ML18153B808 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 06/28/1989 |
| From: | Belisle G, Julian C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18153B807 | List: |
| References | |
| 50-280-89-18, 50-281-89-18, IEB-79-14, NUDOCS 8907140179 | |
| Download: ML18153B808 (8) | |
Text
Re port Nos. :
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, ATLANTA, GEORGIA 30323 50-280/89-18 and 50-281/89-18 Licensee:
Virginia Electric and Power Company Glen Allen, VA 23060 Docket Nos.:
50-280 and 50-281 Facility Name:
Surry 1 and 2 Inspection Conducted:
May 23-26, 1989 License Nos.: DPR-32 and DPR-37 Inspector:
[j:,I(~,; fr."?
G. A. Belisle~*
>JI_
Approved by: C * A ~
'YI C. A. Julian~
Engineering Branen Division of Reactor Safety SUMMARY Scope:
t67 v:;-
Date igned
~t~~~~
This routine, announced inspection was in the areas of resolving safety system functional inspection (SSFI) findings, reviewing licensee's actions pertaining to component cooling water (CCW) issues, and actions on previous inspection finding *
. Results:
In the areas inspected, violations or deviations were not identifie The licensee continues to take corrective action to resolve the SSFI startup issues. Additional testing is ongoing regarding valve leak tightness. Testing was satisfactorily_ performed and observed by the inspector on the "B" ESW pum The licensee is in the process of modifying the CCW system based on information from the vendor regarding an RCP thermal barrier leak to the CCW syste R ADOCK 0500028 PNU
I
REPORT DETAILS 1~
Persons Contacted Licensee Employees *W. Benthall, Licensing Supervisor R. Calder, Manager, Nuclear Licensing W. Conkin, Supervisor, Project Engineering
- E. Grecheck, Assistant Station Manager, Licensing A. Hall, System Engineer
- D. Hart, Supervisor, Quality Control
- G. Miller, Licensing Coordinator
- T. Sowers, Engineering Other licensee employees contacted during this inspection included engineers, operators, and administrative personne Other Organizations L. Culligan, Engineer, Stone & Webster R. Komron, Engineer, Impell NRC Resident Inspecto *W. Kolland, Senior Resident Inspector
- Attended exit interview Acronyms and initialisms used throughout this report are listed in the last paragrap SSFI Iss~es (92701, 92702)
During the last_ inspection in this area (May 1-5, 1989, NRC Inspection Report Nos. 50-280/89-15 and 50-281/89-15), several issues remained outstandin These included completing special testing on CCW and SW valves, updating calculations based on the special testing results, and updating plant procedures and personnel training due to changes in plant systems and newly*established plant *operating parameter The inspector conducted. interviews with licensee personnel regarding special testing and reviewed the results of the following Special Tests:
I ST-251, Leak Checks of Condenser Inlet/Outlet Valves, dated February 24, 1989, for valves 1-CW-106 A, B, C, D and l-CW-100 A, B, C, D.
1-ST-246, Leak Test of Service Water Isolation Valves, dated April 28, 1989, for valves 1-SW-MOV-lOlA, 101 ST-256, Emergency Service Water Flow Uprating Test, dated May 8, 1989, for 1-SW-P-l ST-255, CCHX Flow Instrumentation Calibration, dated April 28, 1989, for HX 1-CC-E-lA, lB, lC, l ST-243, CCHX SW Flow with Vacuum Priming Insolated, dated January 19, 1989, for CCWHX A and The results of the above tests were found to meet test acceptance criteria and no discrepancies were identifie The inspector observed ST-231, * Emergency Service Water Pumps Performance Run, dated January 21, l989, for 1-SW-P-l During this test, 1-SW-P-lB discharged approximately 17,000 gp A question arose regarding the accuracy of the fl ow instrumentation (annubar).
Licensee personnel provided the manufacturer's specifications which delineated the accuracy to be +/-1 percent and answered the inspector's concern The inspector held discussions with plant training personnel and was informed that all operators attended training on canal inventory issues during the Licensed Operator Requalification Program (RQ-89.Sl-LP-1).
The following are the specific training objectives:
Explain the modifications made to intake canal level instrumentation*
to address the problem of non-safety grade equipmen Explain the modifications made to address the issue of canal level drawdown during a OBA due to a postulated single failur Explain the TS change required to address the issue of a OBA occurring coincident with cooldown or RHR operations on the non-accident uni *
Explain the modifications made to the circulating water pump discharge lines to prevent siphoning the upper intake canal back to the rive Explain the Appendix R modifications being made to the condenser water boxe Explain the modification made to the river water makeup pump suction lin Explain the modification made to maintain upper level intake canal level during DBA The inspector reviewed the exam given after this training and verified that appropriate licensed personnel attended the course and received passing grades for the material presente The inspector was informed that additional training was being develope The inspector conducted interviews with appropriate 1 icensee personnel responsible,for assuring that newly installed equipment and operating parameters are accurately reflected in plant procedure Each specific discipline (operations, maintenance, electrical, etc.) had assigned personnel that were actively involved in updating_ their respective procedure The inspector reviewed the following procedures, several were in dr:aft form:
SW-P-E/SAl SW-P-M/A3 SW-P-M/SA2-OP-4 OP-48. IMP-C-G-3 PT-PT-2.21A 1-PT-2.21A 1-PT..'.2.21 2-PT-2.21 AP-10.00 1F-78(H-8)
1F-77(H-7}
,18-33( E-1)
Preventative Maintenance Procedures For the Emergency Service Water Pump Motor 1-SW-PMO-IA Emergency Service Water Pump Diesel_ Service and Inspection "Safety Related" Emergency Service Water Pump Couplingt Clutch, Right-,
Angle Gear Drive and Pump Lubri~ation, Strainer and Vent Trap Inspection Diesel Driven Emergency Service Water Pump Operation Starting Any Circulating Water Pump-ITT Barton Differential Pressure Indicators Model 288A, 289A, 290A, and 291A With Switch; Model 226, 227, 246, 200, and 247 Without Switch Intake Canal Level Logic Testing Functional Test of Low Canal Level Circuitry Functional Test of Low Canal Level Circuitry Intake Canal Level (L-CW-101)
Intake Canal Level (L-CW-101)
Station Blackout
,_'
Intake C_anal Lo Level CH-2 Intake Canal Lo Level CH-1 Intake Canal Hi-Lo Level Intake Canal Lo Level Trip
1-PT-25.3A Emergency Service Water Pump (2-SW-P-lA)
Minimum Equipment List for Criticality and Power Operations Nuclear Safety Analysis Evaluation of Procedure Modification, Surry Intake Canal Inventory Management Program Al though these procedures have not been approved by pl ant management, appropriate steps/instructions have been* added for newly installed equipment and operating parameter Within the area inspected no violations or deviations were identifie.
CCW System On May 12, 1989, a vendor notified the licensee that recent calculations indicated that a leak from the main coolant pump thermal barrier to the CCW system could be as much as 1,400 gp This flow capacity exceeded the existing relief valve flow* capacity and could possibly rupture the CCW pjping inside containmen The existing relfef valve (RV-150/250) flow capac.ity is 167 gp A memorandum from Mr. J.. 0. Erb to Mr~ R. L. Rasnic, dated May 15, 1989, stated that it was possible to postulate a flow rate for the.thermal barrier tube to have a discharge coeffitient of 1.0 (as opposed to a discharge coefficient of 0.4 for the steam_generator tube rupture at North Anna which lea*d to an approximate 650 gpm leak).
The memorandum al so stated that using the higher discharge coefficient, the maximum theoretical flow rate through a double ended rupture of a reactor coolant pump thermal barrier tube could be approximately 1,500 gp~.
A PPR (No.89-046) was written on May 16, 1989, outlining this problem.* A station DR (SI-89-1175) was written on May 16, 1989, based on the PP The torrective action for the DR stated that a design change would be required to provide adequate relief valve capacity to prevent CCW pipe ruptur *
A memorandum from the vendor to the licensee dated May 24, 1989, provided technical information about the reactor coolant pump thermal barriers and indicated that, if a leak were to occur, it would be located in a weld joint and would be very low weepage or very small leakage into the CCW syste This leakage would then be detected by radiation detectors within the CCW syste The memorandum also noted that this type leakage failure would not likely result in.an *instantaneous guillotine type failure mechanism due to external pressurization compressive loads on the weld joint and the du_ctile nature of the 304 stainless steel material The inspector conducted *interviews with licensee and A/E personnel to determine what resolution to this problem is being planne The inspector also walked down portions of the CCW piping in containment.* As more information was obtained, the resolution to the problem also changed.*,. * * *-
The initial resolution*wa5 to install three relief valves on the suction piping and three relief valves on the discharge piping of the CCW lines for each reactor coolant pum This resolution. was discarded when the vendor established that the probable leak rate from the thermal barrier would be approximately 7.5 gp Because the licensee was still evaluating this issue and determining its resolution, the inspector requested, and received from the licensee, a commitment to notify the NRC, in writing, once a final solution was agreed upon and the basis for this solutio On May 9, 1989, a station DR (SI- -89-1122) identified that the penetration CCW piping in containment in Unit 1 does not have seismic support It also states that the same piping is seismically supportep in Unit The Unit 2 work was done during IEB 79-14. It further stated that the Unit 1 piping may need to be seismically supported as wel This issue was discovered by an A/E review of CCW piping in containmen The initial actions/corrective actions/results section of this DR state that it appears that this piping is not required to be operable during the OBA for concrete cooling. It further states that specific references in the UFSAR
_ must be checked and a review of the need for this piping should be further analyze The basis for doing it in Unit 1 and not in Unit 2 must be resolve This issue is being evaluated by the license On March 3, 1989, a station DR (SI-89-687) identified the following:
UFSAR 5.2 lists CC piping within containment as Class III, Level 2, essential lines which are separated from containment atmosphere by a closed valve or membrane barrie According to MKS drawings, many parts of CC inside containment is non~seismically qualified (e.g., CC to RCPs) and the seismic barrier is either at a normally open manual valve or at, the containment penetratio Since some piping is considered-non-seismic, the CC system cannot be considered closed within containment (see UFSAR Table 5.2-1).
The CC containment isolation valves are not Type C tested per PT 1 The CC containment isolation valves are not subject to Type A test pressure per PT 16.3 and therefore are not included in the total integrated leak rate. This may be a violation of 10 CFR 50, Appendix A, GDC 53, 54, 55, 56, and 57 and 10 CFR 50, Appendix J, III.A.1.(d).
This problem was identified by an engineering review of closed Type 1 NP1133 Rev. 1, review of drawings (MKS) and PT The stated corrective action was to submit a station deviatio On May 16, 1989, PES responded to the issue identified by SI-89-68 PES concluded that a combined position which acknowledges that small bore guided pipe is not susceptible to failure and a proactive approach to containment isolation valve leak testing and interim operating procedures form a reasonable basis for justifying restart_ of the plant before permanent modifications are implemente PES a 1 so stated severa 1 recommendations for performing Type C testing of CCW isolation valves, implementing interim operating procedures for ensuring CCW inventory is
- maintained, and performing a Type 2 study for evaluating long term (post startup) modifications..
This issue is being' revi.ewed by the NRC to assure adequate short term and *
long term resolutio Within the area inspected no violations or deviations were identifie.
Actions on Previous Inspection Findings (92702).
. (Closed) Inspector Follow-up Item 280, 281/88-32-16:
Clarify Testing of the ESW Batteries Without the Charger Being Connecte The inspector reviewed PT-23.70, tmergency Service Water Pumps Batteries Weekly Check, dated May 14, 1989.. This pr,ocedure contains steps to disconnect the battery charger prior to recording battery voltage, ce 11 temperature, and specific gravit.
Exit Interview The inspection scope and results were summarized. on May 26, 1989, with those persons indicated in paragraph The inspector described the areas insp~cted and discussed in detail the inspection results listed belo Proprietary information is not contained in this repor Dissenting comments were not received from the license The licensee made an oral commitment to provide to the NRC an action plan resolving licensee actions to contain a thermal barrier heat exchanger leak from the-reactor coolant. pumps into the component cooling water system, paragrap~.
Acronyms and Initialisms A/E cc ccw OBA DR ESW GDC gpm HX IEB N NRC PES PPR PT RCP RHR SSFI Architec~/Engineering Component Cooling Component Cooling Water Design Basis Accident Deviation Report Emergency Service Water General Design Criteria Gallons per minute Heat Exchanger NRC Inspection and Enforcement Bulletin
- Number Nuclear Regulatory Commission Power Engineering Services Potential Problem Report Periodic Test Reactor Coolant Pump Residual Heat Removal Safety System Functional Inspection
,.
Service Water Te~hnical Specification Updated Final Safety Analysis Report
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