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| {{#Wiki_filter:* VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 January 17, 1990 United States Nuclear Regulatory Commission Attention: | | {{#Wiki_filter:* |
| Document Control Desk Washington, D. C. 20555 Gentlemen: | | VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 January 17, 1990 United States Nuclear Regulatory Commission Serial No. 89-860 Attention: Document Control Desk PES/ISI :vlh Washington, D. C. 20555 Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 Gentlemen: |
| VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 ASME SECTION XI INSERVICE PUMP TESTING RESPONSE TO NRC QUESTIONS AND COMMENTS Serial No. PES/ISI :vlh Docket Nos. License Nos. 89-860 50-280 50-281 DPR-32 DPR-37 On November 21, 1989, Virginia Electric and Power Company and the NRC discussed by telephone the Surry Units 1 and 2 lnservice Testing Program Plan submitted to the NRC on September 30, 1988. The NRG reviewer had several questions and comments which need resolution or further clarification before the NRC can issue the Safety Evaluation Report on pump and valve testing for Surry Power Station. Attachment 1 contains the questions and comments posed by the NRC reviewer and the responses by Virginia Electric and Power Company. If you have any questions regarding this revision to our Relief Request, please contact us. W. L. Stewart Senior Vice President
| | VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 ASME SECTION XI INSERVICE PUMP TESTING RESPONSE TO NRC QUESTIONS AND COMMENTS On November 21, 1989, Virginia Electric and Power Company and the NRC discussed by telephone the Surry Units 1 and 2 lnservice Testing Program Plan submitted to the NRC on September 30, 1988. The NRG reviewer had several questions and comments which need resolution or further clarification before the NRC can issue the Safety Evaluation Report on pump and valve testing for Surry Power Station. contains the questions and comments posed by the NRC reviewer and the responses by Virginia Electric and Power Company. |
| -Nuclear Attachments r 900 l. 24-0::::~c~! | | If you have any questions regarding this revision to our Relief Request, please contact us. |
| '::'0011 7~ PDR ADOG~ 05000~~0 \ p PDC cc: U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W. Suite 2900 Atlanta, Georgia 30323 Mr. W. E. Holland NRC Senior Resident Inspector Surry Power Station Mr. Clair Ransom EG&G Idaho P.O. Box 1625 Idaho Falls, Idaho 83415 ATTACHMENT 1 1. Question -Is there some type of flow criterion associated with the diesel fuel oil transfer pump and pump discharge check valve tests? Refer to Relief Requests P-12 and V-38. Response -Yes. While the diesel is running, the pump is started and operated until the high level alarm sounds in the day tank. This test verifies that the check valves open and that the pump is providing more fuel than the diesel can consume. The establishment of a crude criterion based on the time to fill the day tank from the low level mark (level at which the pump automatically starts) to the high level mark (level at which the pump automatically stops) will be investigated. | | W. L. Stewart Senior Vice President - Nuclear Attachments r 900 l. 24-0::::~c~! '::'0011 7 ~ |
| : 2. Question -If a diesel air start bank fails to discharge as described in Relief Request V-37, will the other bank discharge in time to satisfy Technical Specification Requirement 4.6.A.1.b for diesel start time (10 seconds)? | | PDR ADOG~ 05000~~0 p PDC |
| Response -Yes. The automatic transfer from one bank to the other occurs within 2 seconds, leaving time to start the diesel within 10 seconds. A trouble start alarm will sound in this case. 3. Question -Are the valves described in Relief Request V-33 verified closed to the full closed position and are they verified open? Response -These valves are verified to the full closed position during the Appendix J, Type C leak tests. The current test procedure does not verify the open position. | | \ |
| However, the test procedure will be modified to verify the open position by opening the valve after the test volume has been pressurized and observing the pressure decrease. | | |
| Based on the use of this test method as a positive means for verifying valve position, Relief Request V-33 is no longer necessary and is hereby withdrawn. | | cc: U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W. |
| : 4. Question -How are the group leakage criteria established for valves described in Relief Request V-39 and can excess leakage for the smallest valve in a group be masked by criteria based on the larger valve Page 1 ATTACHMENT 1 diameters? | | Suite 2900 Atlanta, Georgia 30323 Mr. W. E. Holland NRC Senior Resident Inspector Surry Power Station Mr. Clair Ransom EG&G Idaho P.O. Box 1625 Idaho Falls, Idaho 83415 |
| Response -The group leakage criteria are determined by summing the valve diameters in the group and multiplying the sum by 0.32 SCFH which corresponds to the guideline criterion of 7.5 SCFD per inch of valve diameter given in IWV-3426. | | |
| Based on this method, the ratios of the smallest valve to the sum of the valve diameters for cases where the valve diameters differ are given below. Valve l-SI-150 l-SI-MOV-1867C 1-SI-MOV-1867D l-SI-174 1-SI-MOV-1869A 1-VS-MOV-101 1-VS-MOV-lOOC 1-VS-MOV-lOOD Valve Diameter 111 3 II 3 II 111 3 II 811 3 6 II 36 11 Smallest Valve Diameter/ | | ATTACHMENT 1 |
| Sum of Diameters 0.14 0.25 0.10 The ratios given above establish that the smallest diameter valve in a given group provides a significant contribution to the group leakage. We believe that these ratios provide reasonable assurance that no valve will be returned to service with excessive leakage. The leak test procedure also has an administrative leak limit which is based on 0.16 SCFH per inch of valve diameter.
| | : 1. Question - Is there some type of flow criterion associated with the diesel fuel oil transfer pump and pump discharge check valve tests? Refer to Relief Requests P-12 and V-38. |
| If the leakage exceeds this limit, the valves will be reworked at the discretion of the Type C test coordinator. | | Response - Yes. While the diesel is running, the pump is started and operated until the high level alarm sounds in the day tank. This test verifies that the check valves open and that the pump is providing more fuel than the diesel can consume. |
| : 5. Comment -Relief Request V-30 does not adequately explain why the valves cannot be tested and how the alternate testing is equivalent to Section XI leak testing. Response -Article IWV-3421 states, "Category A valves shall be leak tested except that valves which function in the course of plant operation in a manner that demonstrates functionally adequate seat tightness need not be leak tested. In such cases, the valve record shall provide the basis for the conclusion that operational observations constitute satisfactory demonstration". | | The establishment of a crude criterion based on the time to fill the day tank from the low level mark (level at which the pump automatically starts) to the high level mark (level at which the pump automatically stops) will be investigated. |
| Page 2 ATTACHMENT 1 The intent of Relief Request V-30 was not to identify valves which cannot be leak tested per Section XI, but to present valves which are part of a leakage detection system that constantly monitors the leakage integrity of the RCS boundary and thus "demonstrates adequate seat tightness" for these valves. The RCS boundary is limited to 1 GPM of unidentified leakage and 10 GPM of identified leakage as required by Technical Specification 3.1.C. RCS leakage is calculated every day. Several parameters are used to determine leakage including | | : 2. Question - If a diesel air start bank fails to discharge as described in Relief Request V-37, will the other bank discharge in time to satisfy Technical Specification Requirement 4.6.A.1.b for diesel start time (10 seconds)? |
| -increased charging flow required to maintain normal level in the pressurizer, -increasing level in the safety injection accumulators and -increasing level in the pressurizer relief tank. 6. Comment -The basis for the request in Relief Request V-5 needs more explanation as to why cold shutdown testing of the valves using flow to verify closure is inconclusive due to the low differential pressure across the valve discs. | | Response - Yes. The automatic transfer from one bank to the other occurs within 2 seconds, leaving time to start the diesel within 10 seconds. A trouble start alarm will sound in this case. |
| * Response -A test was conducted in an effort to verify whether closure of these valves can be determined using flow. Because there is no isolation boundary between the steam generators and the valves, the test volume must include the steam generators. | | : 3. Question - Are the valves described in Relief Request V-33 verified closed to the full closed position and are they verified open? |
| A steam generator was pressurized with a nitrogen blanket to approximately 5 PSIG. The 0.75 inch drain valve just upstream of the check valve was opened and flow was observed. | | Response - These valves are verified to the full closed position during the Appendix J, Type C leak tests. The current test procedure does not verify the open position. However, the test procedure will be modified to verify the open position by opening the valve after the test volume has been pressurized and observing the pressure decrease. Based on the use of this test method as a positive means for verifying valve position, Relief Request V-33 is no longer necessary and is hereby withdrawn. |
| The 14 inch check valve did not stop the back flow through the vent. It was concluded that the flow was inadequate to seat the check valve completely. | | : 4. Question - How are the group leakage criteria established for valves described in Relief Request V-39 and can excess leakage for the smallest valve in a group be masked by criteria based on the larger valve Page 1 |
| Just a small gap between the disc and the seat was sufficient to create a flow area equal to or greater than the flow area through the drain. Therefore, the pressure differential associated with the back flow is being created across the drain valve and not the disc of the check valve. The above test proved to be inconclusive because of the inability to establish a sufficient differential pressure across the disc. The only way to increase the Page 3 | | |
| ' ,, ~-* ATTACHMENT l differential pressure is to increase the flow area from the test volume. However, this is not achievable for the existing configuration. | | ATTACHMENT 1 diameters? |
| Immediately upstream of the drain valve is another 14 inch check valve, so the only available flow area from the test volume is the drain valve. 7. Question -Referring to Relief Request V-23, could the charging pumps be stopped long enough during cold shutdown to allow for the stroking of the valves? Response -Technical Specification 3.2.B.1 states that one charging pump from the plant in cold shutdown must be available for operation if the other plant is operating. | | Response - The group leakage criteria are determined by summing the valve diameters in the group and multiplying the sum by 0.32 SCFH which corresponds to the guideline criterion of 7.5 SCFD per inch of valve diameter given in IWV-3426. Based on this method, the ratios of the smallest valve to the sum of the valve diameters for cases where the valve diameters differ are given below. |
| Further review of the system revealed that the valves could be stroked during cold shutdown. | | Valve Smallest Valve Diameter/ |
| | Valve Diameter Sum of Diameters l-SI-150 111 0.14 l-SI-MOV-1867C 3 II 1-SI-MOV-1867D 3 II l-SI-174 111 0.25 1-SI-MOV-1869A 3 II 1-VS-MOV-101 811 0.10 1-VS-MOV-lOOC 3 6 II 1-VS-MOV-lOOD 36 11 The ratios given above establish that the smallest diameter valve in a given group provides a significant contribution to the group leakage. We believe that these ratios provide reasonable assurance that no valve will be returned to service with excessive leakage. |
| | The leak test procedure also has an administrative leak limit which is based on 0.16 SCFH per inch of valve diameter. If the leakage exceeds this limit, the valves will be reworked at the discretion of the Type C test coordinator. |
| | : 5. Comment - Relief Request V-30 does not adequately explain why the valves cannot be tested and how the alternate testing is equivalent to Section XI leak testing. |
| | Response - Article IWV-3421 states, "Category A valves shall be leak tested except that valves which function in the course of plant operation in a manner that demonstrates functionally adequate seat tightness need not be leak tested. In such cases, the valve record shall provide the basis for the conclusion that operational observations constitute satisfactory demonstration". |
| | Page 2 |
| | |
| | ATTACHMENT 1 The intent of Relief Request V-30 was not to identify valves which cannot be leak tested per Section XI, but to present valves which are part of a leakage detection system that constantly monitors the leakage integrity of the RCS boundary and thus "demonstrates adequate seat tightness" for these valves. The RCS boundary is limited to 1 GPM of unidentified leakage and 10 GPM of identified leakage as required by Technical Specification 3.1.C. |
| | RCS leakage is calculated every day. Several parameters are used to determine leakage including |
| | - increased charging flow required to maintain normal level in the pressurizer, |
| | - increasing level in the safety injection accumulators and |
| | - increasing level in the pressurizer relief tank. |
| | : 6. Comment - The basis for the request in Relief Request V-5 needs more explanation as to why cold shutdown testing of the valves using flow to verify closure is inconclusive due to the low differential pressure across the valve discs. |
| | * Response - A test was conducted in an effort to verify whether closure of these valves can be determined using flow. Because there is no isolation boundary between the steam generators and the valves, the test volume must include the steam generators. A steam generator was pressurized with a nitrogen blanket to approximately 5 PSIG. The 0.75 inch drain valve just upstream of the check valve was opened and flow was observed. The 14 inch check valve did not stop the back flow through the vent. It was concluded that the flow was inadequate to seat the check valve completely. |
| | Just a small gap between the disc and the seat was sufficient to create a flow area equal to or greater than the flow area through the drain. Therefore, the pressure differential associated with the back flow is being created across the drain valve and not the disc of the check valve. |
| | The above test proved to be inconclusive because of the inability to establish a sufficient differential pressure across the disc. The only way to increase the Page 3 |
| | |
| | ' ,, ~- |
| | ATTACHMENT l |
| | * differential pressure is to increase the flow area from the test volume. However, this is not achievable for the existing configuration. Immediately upstream of the drain valve is another 14 inch check valve, so the only available flow area from the test volume is the drain valve. |
| | : 7. Question - Referring to Relief Request V-23, could the charging pumps be stopped long enough during cold shutdown to allow for the stroking of the valves? |
| | Response - Technical Specification 3.2.B.1 states that one charging pump from the plant in cold shutdown must be available for operation if the other plant is operating. Further review of the system revealed that the valves could be stroked during cold shutdown. |
| Therefore, Relief Request V-23 will be withdrawn and replaced by a cold shutdown justification. | | Therefore, Relief Request V-23 will be withdrawn and replaced by a cold shutdown justification. |
| Page 4}} | | Page 4}} |
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ML18152B6671999-09-17017 September 1999 Forwards Two NRC Forms 536,containing Info on Proposed Site Specific Operator Licensing Exam Schedules & Estimated Number of Applicants Planning to Take Exams And/Or Gfes,In Response to NRC Administrative Ltr 99-03 ML18152B3341999-09-16016 September 1999 Requests Relief from Specific Requirements of Subsection IWE of 1992 Edition with 1992 Addenda of ASME Section XI Re Containment Liner Examination Requirements,For North Anna & Surry Power Stations ML18152B4521999-09-14014 September 1999 Forwards Comments on Review of Preliminary Accident Sequence Precursor Analysis of Operational Event That Occurred at Plant,On 980508,as Reported in LER 98-009 ML18152B4501999-09-0808 September 1999 Submits in Triplicate,Application for Renewal of License for Rd Scherer,Iaw 10CFR55.57.Requests That Certification of Medical Exam by Facility Licensee,Nrc Form 396,be Withheld from Public Disclosure,Per 10CFR2.790(a)(6) ML18152B4471999-09-0101 September 1999 Requests That NRC Remove Listed Labels from Distribution ML18152B3711999-08-27027 August 1999 Forwards LER 99-005-00,per Plant TS Table 3.7.6.Rept Has Been Reviewed by Station Nuclear Safety & Operating Committee.Commitment Made by Util,Listed ML18152B3851999-08-23023 August 1999 Forwards Revised TS Basis Pages for TS 3.1.B,deleting Reactor Vessel Toughness Data Duplicated in Ufsar.Ref to Applicable UFSAR Section Included in TS Basis ML18152B3651999-08-20020 August 1999 Requests Removal of License Condition from Sh Wightman Operator License SOP-21538.Updated NRC Form 396 Is Encl.Form NRC 396 Withheld,Per 10CFR2.790(a)(6) ML18152B3681999-08-20020 August 1999 Submits 30-day Rept Re Two Instances in Which Conditions of Approval in Coc Were Not Observed in Making Shipment.Two Type B Shipments Using Model CNS 8-120B Package Were Made After Expiration of QA Program Approval Between 990531-0628 ML18152B3661999-08-20020 August 1999 Provides Medical Status Rept for E Washington,As Required by License Conditions.Summary of E Washington Current Physical Exam & Pertinent Lab Data Attached.Encl Withheld,Per 10CFR2.790(a)(6) ML18152B3801999-08-18018 August 1999 Forwards Technical Rept NE-1206,Rev 0, Surry Unit 2,Cycle 16 Startup Physics Tests Rept, Summarizing Results of Physics Testing Program Performed After Initial Criticality on 990525 ML18152B3781999-08-13013 August 1999 Forwards ISI Summary Rept for Surry Power Station,Unit 2 for 1999 Refueling Outage.Rept Provides Summary of Examination Performed During Outage for Third ISI Interval.No New Commitments Were Made ML18152B3751999-08-13013 August 1999 Forwards LER 99-004-00,IAW 10CFR50.73.Commitment Made by Util,Listed ML18152B4081999-08-0606 August 1999 Forwards Response to NRC 990520 & 0525 RAIs Re North Anna & Surry Responses to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves. ML20210Q7661999-08-0606 August 1999 Requests Exception to 10CFR50.4 Requirement to Provide Total of Twelve Paper Copies When Submitting Surry & North Anna UFSAR Updates.Seek Approval to Submit Only Signed Original & One CD-ROM Version,Per Conversation with J Skoczlas ML18152B4091999-08-0505 August 1999 Forwards Vepc semi-annual fitness-for-duty Program Performance Data Rept for 990101-990630,IAW 10CFR26.71(d) ML18152B4001999-07-29029 July 1999 Requests Relief from Certain Impractical Requirements of ASME Section XI Code Associated with Partial Exams Conducted During 1998 Surry Unit 1 Refueling Outage.Relief Request SR-020 Encl ML18152B3981999-07-28028 July 1999 Forwards 60-day Response to GL 99-02, Lab Testing of Nuclear-Grade Activated Charcoal. Commitments Contained in Ltr,Listed ML18152B3971999-07-26026 July 1999 Provides Estimates of Licensing Actions Expected to Be Submitted in Fys 2000 & 2001,in Response to NRC AL 99-02 ML18151A6281999-07-23023 July 1999 Forwards Revised Epips,Including Rev 19 to EPIP-4.02,rev 14 to EPIP-4.16,rev 8 to EPIP-4.21 & Rev 7 to EPIP-4.30.EP & EPIPs Continue to Meet Stds of 10CFR50.47(b) ML18152B3991999-07-23023 July 1999 Requests That License for Jz Laplante Be Canceled as License Is No Longer Required ML18152B3961999-07-23023 July 1999 Forwards Preliminary,Uncertified License Application & Medical Certification for License to Operate Surry Power Station Units 1 & 2 for Ds Cobb.Encl Withheld,Per 10CFR2.790 (a)(6) ML18152B3931999-07-16016 July 1999 Forwards Updated NRC Form 396 & Ltr,Which Documents Medical Status of Mb Gross,License SOP-20476-2.Encl Withheld from Public Disclosure,Per 10CFR2.790(a)(6) ML18152B4401999-07-0101 July 1999 Informs NRC That on 990511,Dominion Resources,Inc,Executed Amended & Restated Agreement & Plan of Merger with Consolidated Natural Gas Co ML18152B4371999-06-24024 June 1999 Forwards Response to NRC Request for Clarification of Relief Requests Submitted on 990212,requesting Relief from Performing Hydrostatic Testing for Certain Small Diameter Class 1,RCS Pressure Boundary Connections ML18152B4361999-06-22022 June 1999 Forwards Response to RAI Re Surry & North Anna Power Stations,Units 1 & 2,per GL 96-06 ML18152B4301999-06-0303 June 1999 Informs of Util Intention to Revise Schedule for Submittal of License Renewal Applications for Surry & North Anna Power Stations to March 2002 ML18152B4261999-05-28028 May 1999 Provides Formal Notification of Effect of Recent Organizational Restructuring on OLs of North Anna & Surry Power Stations,Per NRC 990513 Telcon Request ML18152B4221999-05-27027 May 1999 Forwards Info Concerning Changes to ECCS Evaluation Models & Application in Existing Licensing Analyses for Surry & North Anna Power Stations,Units 1 & 2 ML18152B4231999-05-26026 May 1999 Informs That Vepc Will Revise 180 Day Response to NRC GL 96-05,within 120 Days of Date of Ltr to Incorporate Commitment to Participate in Joint Owners Group Program as Applicable ML18152B4211999-05-25025 May 1999 Forwards Rev 1 to Relief Request P-11 to Clarify Original Intent of Request by Specifically Requesting Relief from Requirements of Section 6.1 of OM-6 ML18152B4171999-05-17017 May 1999 Provides Notification of Number of Steam Generator Tubes That Were Plugged During Spring 1999 Refueling Outage Planned ISI ML18152A3701999-05-13013 May 1999 Submits Proposal to Use Provisions of ASME Section XI Code Case N-597 for Analytical Evaluation of Class 1,2 & 3 Carbon & Low Alloy Steel Piping Components Subjected to Wall Thinning as Result of Flow Accelerated or Other Corrosion ML18152B4121999-05-0303 May 1999 Forwards Application for Renewal of License for SV Ross. Encl Withheld Per 10CFR2.790(a)(6) ML18152B4101999-04-29029 April 1999 Forwards Scope & Objectives for 990803 Surry Power Station Emergency Exercise.Without Encls ML18152B6561999-04-23023 April 1999 Forwards Annual Radioactive Effluent Release Rept for Surry Power Station,Jan-Dec 1998, Which Includes Summary of Quantities of Radioactive Liquid & Gaseous Effluents & Solid Waste Released During CY98 ML18152B6491999-04-13013 April 1999 Forwards MOR for Mar 1999 for Surry Power Station,Units 1 & 2.MOR for Feb 1999 Incorrectly Stated Gross Electrical Energy Generated (Mwh) for Unit 2.Rept Should Have Stated Monthly Figure as 568965.0 ML18151A5851999-03-31031 March 1999 Forwards Rept on Status of Decommissioning Funding for Each of Four Nuclear Power Reactors,Per 10CFR50.75(f)(1) ML18152A2801999-03-30030 March 1999 Forwards Summary of Structural Integrity Evaluation of Thermally Induced Over Pressurization of Containment Penetration Piping During DBA for SPS & Naps,Units 1 & 2,per GL 96-06.Draft Proposed UFSAR Revised Pages,Encl ML18153A2721999-03-29029 March 1999 Forwards LER 99-002-00 Per 10CFR50.73.Listed Commitments Contained in Ltr ML18153A3421999-03-26026 March 1999 Provides Updated Medical Status Rept for Wb Gross in Accordance with License SOP-20476-02,Docket 55-5228,as Amended by 980320 License Amend.Informs That Gross Exhibits No Performance Problems & Will Continue on Current Medicine ML20204H0331999-03-17017 March 1999 Forwards Rev 5 to PSP for Surry & North Anna Power Stations & Associated Isfsis.Description & Justification for Changes Included with Plan Rev.Rev 5 to PSP Withheld Per 10CFR73.21 ML18153A3411999-03-15015 March 1999 Forwards Signed Applications & Medical Certificates for Initial License at Surry Power Station Units 1 & 2 for Listed Individuals.Without Encls 1999-09-08
[Table view] Category:UTILITY TO NRC
MONTHYEARML18153C3661990-09-20020 September 1990 Forwards Topical Rept VEP-NE-3-A, Qualification of WRB-1 CHF Correlation in VEPCO Cobra Code. ML18153C3701990-09-18018 September 1990 Forwards Addl Info Re Facility Containment Isolation Valve Type C Test,Per 900914 10CFR50,App J Exemption Request ML18152A2341990-09-14014 September 1990 Requests Exemption from 10CFR50,App J Section III.D.3 Re Local Leak Rate Testing During Every Reactor Shutdown. Basis & Justification for Exemption Encl ML20059J3341990-09-13013 September 1990 Forwards Rev 15 to Nuclear Security Personnel Training & Qualification Plan.Rev Withheld ML18151A2901990-08-31031 August 1990 Forwards Rev 10 to Updated FSAR for Surry Power Station Units 1 & 2,representing Second Updated FSAR Submitted This Yr ML18153C3451990-08-29029 August 1990 Forwards Proprietary Semiannual Fitness for Duty Program Performance Data Rept for 900103-0630.Rept Includes Summaries of Mgt Sanctions Imposed,Actions Taken to Correct Program Weaknesses & Events Reported to Nrc.Encl Withheld ML18153C3381990-08-22022 August 1990 Responds to NRC 900723 Ltr Re Violations Noted in Insp Rept 50-280/90-21 & 50-281/90-21.Corrective actions:as-found-as- Left Conditions of Auxiliary Feedwater Evaluated & Found Operable ML18153C3391990-08-22022 August 1990 Requests Approval for Use of Plugs Fabricated of nickel- chromium-iron Uns N-06690 Matl (Alloy 690) to Plug Tubes in Steam Generators for Mechanical & Welded Applications ML18153C3161990-08-0101 August 1990 Provides Supplemental Response to NRC 900629 Ltr Re Electrical Crossties,Load Shedding on Nonblackout Unit & Emergency Diesel Generator Reliability.Emergency Diesel Generator Reliability Program in Place,Per Reg Guide 1.155 ML18153C3171990-08-0101 August 1990 Resubmits Synopsis of Changes to Updated Operational QA Program Topical Rept Vep 1-5A ML18153C3091990-07-30030 July 1990 Provides Outline of Plan to Meet 10CFR50 App G Requirements, for Low Upper Shelf Energy Matls,Per NRC 900521 Request ML18153C3061990-07-30030 July 1990 Forwards Revised Tech Spec Pages,Addressing Constitution of Quorum & Timeliness of Mgt Safety Review Committe Meeting Minutes,Per NRC Request ML18153C3051990-07-26026 July 1990 Advises That Util Submitted Decommissioning Funding Plan & Financial Assurance Info W/Isfsi License Application ML18153C3041990-07-26026 July 1990 Responds to NRC 900626 Ltr Re Violations Noted in Insp Rept 50-281/90-20.Corrective Actions:Leaking Drain Plug & Upper Drain Plug on Motor Replaced W/Oil Drain Assemblies Composed of Piping & Valves ML18153C3031990-07-26026 July 1990 Advises of Withdrawal of Request for NRC Review & Approval of Engineering Evaluation 8.Revised Evaluation Will Be Maintained Onsite for NRC Audit During Future Insps,Per Generic Ltr 86-10 ML18153C3101990-07-26026 July 1990 Forwards Decommissioning Financial Assurance Certification Rept..., Nuclear Decommissioning Trust Agreement & Nonqualified Nuclear Decommissioning Trust Amended & Restated Trust Agreement, Per 10CFR50.75 ML18153C2901990-07-18018 July 1990 Responds to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount. Listed Transmitters Compiled.Transmitters Found Installed within Reactor Protection or ESFAS Have Been Replaced ML18153C2861990-07-12012 July 1990 Requests Cancellation of Operator Licenses for Listed Individuals.Licenses No Longer Required ML18153C2871990-07-11011 July 1990 Responds to Violations Noted in Insp Repts 50-280/90-18 & 50-281/90-18.Corrective Actions:Permanent Drain Line Installed & Matrix Which Describes Proper Ventilation Alignment for Plant Conditions Provided for Personnel Use ML18153C2851990-07-0606 July 1990 Forwards Response to Generic Ltr 90-04 Re Status of Generic Safety Issues ML18153C2831990-07-0303 July 1990 Advises That MW Hotchkiss No Longer Needs Operator License SOP-20548-1.Cancellation of License Requested ML18153C2821990-07-0303 July 1990 Requests Exemption from 10CFR50,App J,Paragraph III.A.6(b), Which Requires That When Two Consecutive Periodic Type a Tests Fail to Meet Applicable Acceptance Criteria Type a Test Shall Be Performed at Plant ML18153C3591990-06-28028 June 1990 Responds to SALP Repts 50-280/90-16 & 50-281/90-16 for Period 890701-900331.Corrective Actions Focus on Issues of Maint Backlog,Maint planning,post-maint Testing,Staffing & Procurement ML18153C2611990-06-21021 June 1990 Responds to NRC 900522 Ltr Re Violations Noted in Insp Repts 50-280/90-07 & 50-281/90-07.Corrective actions:as-built Configurations of 120-volt Ac & Dc Vital & Semivital Panel Breaker Installations Verified to Be Acceptable ML18153C2581990-06-18018 June 1990 Forwards Reissued Semiannual Radioactive Effluent Release Rept,Jul-Dec 1989. Rept Contains Info Re SR-89,Sr-90 & Fe-55 Analytical Results for Liquid Composite Samples ML18153C2591990-06-18018 June 1990 Forwards Response to NRC 900524 Request for Addl Info Re NRC Bulletin 88-004, Potential Safety-Related Pump Loss. Engineering Will Initiate Study to Evaluate Enhancements of Cooldown & Possible Heatup Operation ML18153C2541990-06-15015 June 1990 Forwards Corrected Tech Specs Page 3.1-3,per Identification of Typo in 900522 Application for Amends to Licenses DPR-32 & DPR-37 ML18153C2521990-06-14014 June 1990 Responds to NRC 900515 Ltr Re Violations Noted in Insp Repts 50-280/90-09 & 50-281/90-09.Corrective Actions:Abnormal Procedures & Fire Contingency Action Procedures Being Upgraded Via Technical Procedure Upgrade Program ML18153C2501990-06-0808 June 1990 Confirms That Primary Policy Re Onsite Property Damage Insurance,Provided by Nuclear Mutual Limited ML18153C2361990-05-29029 May 1990 Responds to NRC 900427 Ltr Re Violations Noted in Insp Repts 50-280/90-14 & 50-281/90-14.Corrective Actions:Personnel Involved Counseled as to Importance of Properly Recording & Reporting Surveillance Data ML18153C1971990-04-24024 April 1990 Responds to Unresolved Items Noted in Insp Repts 50-338/89-12,50-339/89-12,50-280/88-19 & 50-281/88-19 Re Secondary Sys Containment Leakage & Concludes Leakage Need Not Be Quantified & Not Included in as-found Leakage ML18153C1961990-04-20020 April 1990 Forwards Facility Previous Tests & Projected Leakage Totals for Type C Testing for Valves & Penetrations,Per 900108 Ltr ML18153C1901990-04-18018 April 1990 Requests That Operator License OP-20447-1 for Ja Yourish Be Cancelled ML18153C1861990-04-0505 April 1990 Requests Exemption from 10CFR50,App J,Paragraph III.A.6(b). Util Implemented Corrective Action Program Which Meets Intent of Regulation in Establishing Containment Integrity ML18153C1671990-03-30030 March 1990 Submits Supplemental Response to 10CFR50.63, Loss of All AC Power. Understands That Load Mgt Schemes for Both Blackout & Nonblackout Units Allowed by Station Blackout Rule ML18151A2551990-03-30030 March 1990 Forwards Rev to, Corporate Emergency Response Plan & Rev to, Corporate Plan Implementing Procedures. ML18151A4941990-03-29029 March 1990 Forwards Listed Info Re Licensee Guarantees of Payment of Deferred Premiums,Per 10CFR140.21(e) ML18153C1631990-03-27027 March 1990 Responds to Violations Noted in Insp Repts 50-280/86-05 & 50-281/86-05.Corrective Action:Surveillance Tests Being Performed in Accordance W/Administrative Requirements of Station Procedures & Plans Implementing New Review Process ML18153C1551990-03-20020 March 1990 Clarifies 900108 Request for Exemption from 10CFR50,App J Re Type C Testing Requirements ML18153C1511990-03-19019 March 1990 Responds to Generic Ltr 89-19, Request for Action Re Resolution of USI A-47. Operability & Surveillance Requirements for Steam Generator Overfill Protection Sys Will Be Incorporated in Tech Spec Change ML18153C1661990-03-16016 March 1990 Discusses Waiver of Compliance Re Containment Vacuum Sys Operability,Per 900315 & 16 Telcons ML18153C1471990-03-14014 March 1990 Discusses Functional Test for High Setpoint for PORVs ML18153C1571990-03-12012 March 1990 Forwards List of Emergency Operating Procedures in Preparation for 900402-12 Insp.Vol I Is Emergency Operating Procedure Set & Consists of 47 Notebooks.Vol II Contains Fire Contingency Action (App R) Procedures ML18153C1371990-03-0808 March 1990 Forwards Suppl to 1986 Inservice Insp Summary Rept,Adding Two Missing NIS-2 Forms Containing Info Re Replacement of Bolting Matl on 1-RC-SV-1551C (Flange a) & 1-RC-HCV-1556A ML18153C1281990-03-0101 March 1990 Submits 1989 Annual Steam Generator Inservice Insp Rept Results.No Steam Generator Tubes Plugged in 1989 ML18153C1261990-03-0101 March 1990 Responds to Generic Ltr 90-01 Re NRC Regulatory Impact Survey.Survey Covers Type of Insp,Audit or Evaluation by NRC Resident,Nrc Regional Ofc,Nrc Teams & INPO ML18153C1231990-02-22022 February 1990 Responds to NRC 900123 Ltr Re Violations Noted in Insp Repts 50-280/89-32 & 50-281/89-32.Corrective Actions:Use of Lab Hood Attached to F-2 Fan Suction Prohibited & Contaminated & Radioactive Items Removed from Hood ML18152A4881990-02-0606 February 1990 Responds to NRC 891222 Ltr Re Violations Noted in Insp Repts 50-280/89-34 & 50-281/89-34 on 891029-1125.Corrective Actions:Steps in Operating Procedure 2-OP-1.3 Associated W/ Valve Test Being Evaluated for Inclusion in OP-7.1.1 ML18153C0991990-02-0101 February 1990 Withdraws 891018 Application for Amends to Licenses DPR-32 & DPR-37,increasing Pressurizer Safety Valve Setpoint Tolerance to +/- 3% of Nomical Lift Setpoint.Emergency Tech Spec Change Granted on 891116 Provided Modified Tolerances ML18153C0951990-01-29029 January 1990 Forwards Response to Generic Ltr 89-13, Svc Water Sys Problems Affecting Safety-Related Equipment. Belief in Appropriateness to Address Generic Ltr 89-13 Concerns within Context of Established Programmatic Improvements Noted 1990-09-20
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Text
VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 January 17, 1990 United States Nuclear Regulatory Commission Serial No.89-860 Attention: Document Control Desk PES/ISI :vlh Washington, D. C. 20555 Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 ASME SECTION XI INSERVICE PUMP TESTING RESPONSE TO NRC QUESTIONS AND COMMENTS On November 21, 1989, Virginia Electric and Power Company and the NRC discussed by telephone the Surry Units 1 and 2 lnservice Testing Program Plan submitted to the NRC on September 30, 1988. The NRG reviewer had several questions and comments which need resolution or further clarification before the NRC can issue the Safety Evaluation Report on pump and valve testing for Surry Power Station. contains the questions and comments posed by the NRC reviewer and the responses by Virginia Electric and Power Company.
If you have any questions regarding this revision to our Relief Request, please contact us.
W. L. Stewart Senior Vice President - Nuclear Attachments r 900 l. 24-0::::~c~! '::'0011 7 ~
PDR ADOG~ 05000~~0 p PDC
\
cc: U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W.
Suite 2900 Atlanta, Georgia 30323 Mr. W. E. Holland NRC Senior Resident Inspector Surry Power Station Mr. Clair Ransom EG&G Idaho P.O. Box 1625 Idaho Falls, Idaho 83415
ATTACHMENT 1
- 1. Question - Is there some type of flow criterion associated with the diesel fuel oil transfer pump and pump discharge check valve tests? Refer to Relief Requests P-12 and V-38.
Response - Yes. While the diesel is running, the pump is started and operated until the high level alarm sounds in the day tank. This test verifies that the check valves open and that the pump is providing more fuel than the diesel can consume.
The establishment of a crude criterion based on the time to fill the day tank from the low level mark (level at which the pump automatically starts) to the high level mark (level at which the pump automatically stops) will be investigated.
- 2. Question - If a diesel air start bank fails to discharge as described in Relief Request V-37, will the other bank discharge in time to satisfy Technical Specification Requirement 4.6.A.1.b for diesel start time (10 seconds)?
Response - Yes. The automatic transfer from one bank to the other occurs within 2 seconds, leaving time to start the diesel within 10 seconds. A trouble start alarm will sound in this case.
- 3. Question - Are the valves described in Relief Request V-33 verified closed to the full closed position and are they verified open?
Response - These valves are verified to the full closed position during the Appendix J, Type C leak tests. The current test procedure does not verify the open position. However, the test procedure will be modified to verify the open position by opening the valve after the test volume has been pressurized and observing the pressure decrease. Based on the use of this test method as a positive means for verifying valve position, Relief Request V-33 is no longer necessary and is hereby withdrawn.
- 4. Question - How are the group leakage criteria established for valves described in Relief Request V-39 and can excess leakage for the smallest valve in a group be masked by criteria based on the larger valve Page 1
ATTACHMENT 1 diameters?
Response - The group leakage criteria are determined by summing the valve diameters in the group and multiplying the sum by 0.32 SCFH which corresponds to the guideline criterion of 7.5 SCFD per inch of valve diameter given in IWV-3426. Based on this method, the ratios of the smallest valve to the sum of the valve diameters for cases where the valve diameters differ are given below.
Valve Smallest Valve Diameter/
Valve Diameter Sum of Diameters l-SI-150 111 0.14 l-SI-MOV-1867C 3 II 1-SI-MOV-1867D 3 II l-SI-174 111 0.25 1-SI-MOV-1869A 3 II 1-VS-MOV-101 811 0.10 1-VS-MOV-lOOC 3 6 II 1-VS-MOV-lOOD 36 11 The ratios given above establish that the smallest diameter valve in a given group provides a significant contribution to the group leakage. We believe that these ratios provide reasonable assurance that no valve will be returned to service with excessive leakage.
The leak test procedure also has an administrative leak limit which is based on 0.16 SCFH per inch of valve diameter. If the leakage exceeds this limit, the valves will be reworked at the discretion of the Type C test coordinator.
- 5. Comment - Relief Request V-30 does not adequately explain why the valves cannot be tested and how the alternate testing is equivalent to Section XI leak testing.
Response - Article IWV-3421 states, "Category A valves shall be leak tested except that valves which function in the course of plant operation in a manner that demonstrates functionally adequate seat tightness need not be leak tested. In such cases, the valve record shall provide the basis for the conclusion that operational observations constitute satisfactory demonstration".
Page 2
ATTACHMENT 1 The intent of Relief Request V-30 was not to identify valves which cannot be leak tested per Section XI, but to present valves which are part of a leakage detection system that constantly monitors the leakage integrity of the RCS boundary and thus "demonstrates adequate seat tightness" for these valves. The RCS boundary is limited to 1 GPM of unidentified leakage and 10 GPM of identified leakage as required by Technical Specification 3.1.C.
RCS leakage is calculated every day. Several parameters are used to determine leakage including
- increased charging flow required to maintain normal level in the pressurizer,
- increasing level in the safety injection accumulators and
- increasing level in the pressurizer relief tank.
- 6. Comment - The basis for the request in Relief Request V-5 needs more explanation as to why cold shutdown testing of the valves using flow to verify closure is inconclusive due to the low differential pressure across the valve discs.
- Response - A test was conducted in an effort to verify whether closure of these valves can be determined using flow. Because there is no isolation boundary between the steam generators and the valves, the test volume must include the steam generators. A steam generator was pressurized with a nitrogen blanket to approximately 5 PSIG. The 0.75 inch drain valve just upstream of the check valve was opened and flow was observed. The 14 inch check valve did not stop the back flow through the vent. It was concluded that the flow was inadequate to seat the check valve completely.
Just a small gap between the disc and the seat was sufficient to create a flow area equal to or greater than the flow area through the drain. Therefore, the pressure differential associated with the back flow is being created across the drain valve and not the disc of the check valve.
The above test proved to be inconclusive because of the inability to establish a sufficient differential pressure across the disc. The only way to increase the Page 3
' ,, ~-
ATTACHMENT l
- differential pressure is to increase the flow area from the test volume. However, this is not achievable for the existing configuration. Immediately upstream of the drain valve is another 14 inch check valve, so the only available flow area from the test volume is the drain valve.
- 7. Question - Referring to Relief Request V-23, could the charging pumps be stopped long enough during cold shutdown to allow for the stroking of the valves?
Response - Technical Specification 3.2.B.1 states that one charging pump from the plant in cold shutdown must be available for operation if the other plant is operating. Further review of the system revealed that the valves could be stroked during cold shutdown.
Therefore, Relief Request V-23 will be withdrawn and replaced by a cold shutdown justification.
Page 4