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| ~Jlili$1illliiI!1<rlivi<piivl ROCHESTER GAS AND ELECTRIC CORPORATION rl~~Oos~O o 89 EAST AVENUE, ROCHESTER, N.Y.L4649 LEON O.iVHITE, JR.VICC PRCSIDCNT TCLCPHOVC iPCi COOC ria 546-2700 April 30, 1980 Mr.Boyce H.Grier, Director U.S.Nuclear Regulatory Commission Office of Inspection and Enforcement Region I 651 Park Avenue King of Prussia, Pennsylvania 19406 | | I/ ix/'Qllj~gslÃl/I |
| | ~Jlili$1illliiI !1<rlivi<piivl ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N.Y. L4649 LEON O. iVHITE, JR. TCLCPHOVC VICC PRCSIDCNT iPCi COOC ria 546-2700 April 30, 1980 Mr. Boyce H. Grier, Director U.S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region I 651 Park Avenue King of Prussia, Pennsylvania 19406 |
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| ==Subject:== | | ==Subject:== |
| Response to IE Bulletin No.'80-04 o | | Response to IE Bulletin No. '80-04 o |
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| ==Dear Mr.Grier.:== | | ==Dear Mr. Grier.:== |
| Enclosed is our response to IE Bulletin No.80-04.This bulletin was received on February 8, 1980.Sincerely, XC: Office of Inspection and Enforcement Division of Reactor Operations Inspection U.S.Nuclear Regulatory Commission Washington, D.C.20555
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| Res onse to IE Bulletin 80-04 Request 1.I Response 1.Review the containment pressure response analysis to determine if the potential for containment overpressure for a main steam line break inside containment included the impact of runout flow from the auxiliary feedwater system and the impact of other energy sources, such as continuation of feedwater or condensate flow.In your review, consider.your ability to detect and isolate the damaged steam generator from these sources and the ability of the pumps to remain operable after extended operation at, runout flow.Although the Ginna post-steamline break containment pressure analysis in the FSAR did not include the effects of auxiliary feedwater flow to the affected steam generator, it is important to recognize that the evaluation also did not include the benefits of passive and active heat sinks inside containment.
| | Enclosed is our response to IE Bulletin No. 80-04. This bulletin was received on February 8, 1980. |
| Continued feedwater/condensate addition to the steam generator will not occur, since, the safety injection signal (generated by a variety of process parameters, including high steam line flow, high containment pressure, and low pressurizer pressure)will close the feedwater control valves and stop the feedwater pumps.The addition of maximum auxiliary feedwater flow to the broken steam generator will eventually require operator action to 1)realign flow to the intact generator, 2)terminate auxiliary feedwater flow to the broken generators.
| | Sincerely, XC: Office of Inspection and Enforcement Division of Reactor Operations Inspection U. S. Nuclear Regulatory Commission Washington, D. C. 20555 |
| Positive information is available to the operator to determine which is the affected steam generator.
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| Steam generator level instrumentation is located inside containment and steam generator pressure is located outside contain-ment where it.would not be affected by the accident environment, inside containment.
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| It is expected that, through proper training and by use of the emergency procedures, the operators will be capable of quickly recognizing the steam line break, and will perform the proper operations.
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| There is substantial time available for the operator to perform the two safety functions noted above.The SEP Safe Shutdown review concluded following their site visit in June 1978 that one steam generator would not boil dry for over thirty minutes.Thus there is substantial time to align flow to the intact steam generator.
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| The termination of auxiliary feedwater flow to the affected steam generator, under the pessimistic circumstances, would require more rapid action (but still easily within the capability of the operators) to maintain containment
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| pressure below design pressure.The analysis presented in Attachment 1 concludes that, assuming minimum safeguards for.containment cooling, auxiliary feedwater flow would have to be terminated in about 26 minutes.With maximum safeguards, this time would be extended to about 44 minutes.There is no need to consider the operation of the auxiliary feedwater pumps at runout flow.The turbine-driven pumps are controlled by a'overnor, and will not, exceed about 400 gpm.The motor driven pump flow is controlled by the AFW control valves, which receive an automatic throttle signal to 200 gpm from their flow controllers. | | Res onse to IE Bulletin 80-04 Request 1. Review the containment pressure response analysis to determine if the potential for containment overpressure for a main steam line break inside containment included the impact of runout flow from the auxiliary feedwater system and the impact of other energy sources, such as continuation of feedwater or condensate flow. In your review, consider. your ability to detect and isolate the damaged steam generator from these sources and the ability of the pumps to remain operable after extended operation at, runout flow. |
| A potential single failure of the flow controller to control.flow to 200 gpm is not considered a worst-case single failure in terms of net energy addition to the containment, since the operation of all containment cooling safeguards (vs.the minimum safeguards assumed in this evaluation) would result in a substantial increase in energy removal from containment.
| | I Response 1. Although the Ginna post-steamline break containment pressure analysis in the FSAR did not include the effects of auxiliary feedwater flow to the affected steam generator, it is important to recognize that the evaluation also did not include the benefits of passive and active heat sinks inside containment. |
| Request 2.Review your analysis of the reactivity increase which results from a main steam line break inside or" outside containment.
| | Continued feedwater/condensate addition to the steam generator will not occur, since, the safety injection signal (generated by a variety of process parameters, including high steam line flow, high containment pressure, and low pressurizer pressure) will close the feedwater control valves and stop the feedwater pumps. The addition of maximum auxiliary feedwater flow to the broken steam generator will eventually require operator action to 1) realign flow to the intact generator, 2) terminate auxiliary feedwater flow to the broken generators. Positive information is available to the operator to determine which is the affected steam generator. Steam generator level instrumentation is located inside containment and steam generator pressure is located outside contain-ment where it. would not be affected by the accident environment, inside containment. It is expected that, through proper training and by use of the emergency procedures, the operators will be capable of quickly recognizing the steam line break, and will perform the proper operations. |
| This review should consider the reactor cooldown rate and the potential for the reactor to return to power with the most reactive control rod in the fully withdrawn position.If your previous analysis did not consider all potential water sources (such as those listed in 1 above)and if the reactivity increase is greater than previous analysis indicated the report of th'is review should include: a I The boundary conditions for the analysis, e.g., the end of life shutdown margin, the moderator temperature coefficient, power level and the net effect of the associated steam generator water inventory on the reactor system cooling, etc., b.The most restrictive single active failure in the safety injection system and the effect of that failure on delaying the delivery of high concentration boric acid solution to the reactor coolant system, c The effect of extended water supply to the affected steam generator on the core criticality and return to power,
| | There is substantial time available for the operator to perform the two safety functions noted above. |
| | The SEP Safe Shutdown review concluded following their site visit in June 1978 that one steam generator would not boil dry for over thirty minutes. Thus there is substantial time to align flow to the intact steam generator. The termination of auxiliary feedwater flow to the affected steam generator, under the pessimistic circumstances, would require more rapid action (but still easily within the capability of the operators) to maintain containment |
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| d.The hot channel factors corresponding to the most reactive rod in the fully withdrawn'osition at the end of life, and the Minimum Departure from Nucleate Boiling Ratio (MDNBR)values for the analyzed transient. | | pressure below design pressure. The analysis presented in Attachment 1 concludes that, assuming minimum safeguards for .containment cooling, auxiliary feedwater flow would have to be terminated in about 26 minutes. |
| Response 2.Westinghouse Electric Corporation performed the original steam break analysis for Ginna as reported in the FSAR and a reanalysis submitted to the NRC in September 1975.Westinghouse has reviewed the assumptions made for main and auxiliary feedwater flow as they apply to licensing basis steam line break transients. | | With maximum safeguards, this time would be extended to about 44 minutes. |
| Several of the relevant assumptions used in all core transient analyses follow, and are further explained in the Ginna FSAR.1.The reactor is assumed initially to be at hot shutdown conditions, at the minimum allowable shutdown margin.2.For the Condition IV breaks, i.e., double-ended rupture of a main steam pipe, full main feedwater is assumed from the beginning of the transient at a very conservative cold temperature. | | There is no need to consider the operation of the auxiliary feedwater pumps at runout flow. The turbine-driven pumps are controlled by a'overnor, and will not, exceed about 400 gpm. The motor driven pump flow is controlled by the AFW control valves, which receive an automatic throttle signal to 200 gpm from their flow controllers. A potential single failure of the flow controller to control. flow to 200 gpm is not considered a worst-case single failure in terms of net energy addition to the containment, since the operation of all containment cooling safeguards (vs. the minimum safeguards assumed in this evaluation) would result in a substantial increase in energy removal from containment. |
| 3.All auxiliary feedwater pumps are initially assumed to be operating, in addition to the main feedwater. | | Request 2. Review your analysis of the reactivity increase which results from a main steam line break inside or outside containment. This review should consider the reactor cooldown rate and the potential for the reactor to return to power with the most reactive control rod in the fully withdrawn position. If your previous analysis did not consider all potential water sources ( such as those listed in 1 above) and if the reactivity increase is greater than previous analysis indicated the report of th'is review should include: |
| The flow is equivalent to the rated flow of all pumps at the steam generator design pressure.4 Feedwater is assumed to continue at its initial flow rate until feedwater isolation is complete, approximately 10 seconds after the break occurs, while auxiliary feedwater is assumed to continue at its initial flow'rate.5.Main feedwater flow is completely terminated following feedwater isolation. | | aI The boundary conditions for the analysis, e.g., |
| Based on the manner in which the analysis is performed~for Westinghouse plants, the core transient results are very insensitive to auxiliary feedwater flow.The first minute of the transient is dominated entirely by the steam flow contribution to primary-secondary heat transfer, which is the forcing function for both the reactivity and thermal-hydraulic transients in the core.The effect of auxiliary feedwater runout (or failure of runout protection where applicable) is minimal.Greater feedwater flow during the large steamline breaks serves to reduce secondary pressures, accelerating the automatic safeguards actions, i.e.steamline isolation, feedwater isolation and safety injection. | | the end of life shutdown margin, the moderator temperature coefficient, power level and the net effect of the associated steam generator water inventory on the reactor system cooling, etc., |
| The assumptions described above are | | : b. The most restrictive single active failure in the safety injection system and the effect of that failure on delaying the delivery of high concentration boric acid solution to the reactor coolant system, c The effect of extended water supply to the affected steam generator on the core criticality and return to power, |
| | : d. The hot channel factors corresponding to the most reactive rod in the fully withdrawn'osition at the end of life, and the Minimum Departure from Nucleate Boiling Ratio ( MDNBR) values for the analyzed transient. |
| | Response 2. Westinghouse Electric Corporation performed the original steam break analysis for Ginna as reported in the FSAR and a reanalysis submitted to the NRC in September 1975. Westinghouse has reviewed the assumptions made for main and auxiliary feedwater flow as they apply to licensing basis steam line break transients. Several of the relevant assumptions used in all core transient analyses follow, and are further explained in the Ginna FSAR. |
| | : 1. The reactor is assumed initially to be at hot shutdown conditions, at the minimum allowable shutdown margin. |
| | : 2. For the Condition IV breaks, i.e., double-ended rupture of a main steam pipe, full main feedwater is assumed from the beginning of the transient at a very conservative cold temperature. |
| | : 3. All auxiliary feedwater pumps are initially assumed to be operating, in addition to the main feedwater. The flow is equivalent to the rated flow of all pumps at the steam generator design pressure. |
| | 4 Feedwater is assumed to continue at its initial flow rate until feedwater isolation is complete, approximately 10 seconds after the break occurs, while auxiliary feedwater is assumed to continue at its initial flow 'rate. |
| | : 5. Main feedwater flow is completely terminated following feedwater isolation. |
| | Based on the manner in which the analysis is performed ~ |
| | for Westinghouse plants, the core transient results are very insensitive to auxiliary feedwater flow. |
| | The first minute of the transient is dominated entirely by the steam flow contribution to primary-secondary heat transfer, which is the forcing function for both the reactivity and thermal-hydraulic transients in the core. The effect of auxiliary feedwater runout (or failure of runout protection where applicable) is minimal. Greater feedwater flow during the large steamline breaks serves to reduce secondary pressures, accelerating the automatic safeguards actions, i.e. |
| | steamline isolation, feedwater isolation and safety injection. The assumptions described above are |
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| therefore appropriate and conservative for the short-term aspect of the steamline break transient. | | therefore appropriate and conservative for the short-term aspect of the steamline break transient. |
| The auxiliary feedwater flow becomes a-dominant factor in determining the duration and,magnitude of the steam flow transient'uring later stages in the transient. | | The auxiliary feedwater flow becomes a-dominant factor in determining the duration and,magnitude of the steam flow transient'uring later stages in the transient. However, the limiting portion of the transient occurs during the first minute, both due to higher steam flows inherently present early in the transient and due to the introduction of boron to the core via the safety injection system. |
| However, the limiting portion of the transient occurs during the first minute, both due to higher steam flows inherently present early in the transient and due to the introduction of boron to the core via the safety injection system.In conclusion, Westinghouse has evaluated the effect of runout auxiliary feedwater flows in the core transient for steamline break, and based on this evaluation, has determined that the assumptions presently made are appropriate for use as a licensing basis.The concerns outlined in the introduction to IE Bulletin 80-04 relative to, 1)limiting core conditions occurring during portions of the transient where auxiliary feedwater flow is a relevant contributor to plant cooldown;and 2)incomplete isolation of main feedwater flow, are not representative of the Westinghouse NSSS designs and associated Balance of Plant requirements. | | In conclusion, Westinghouse has evaluated the effect of runout auxiliary feedwater flows in the core transient for steamline break, and based on this evaluation, has determined that the assumptions presently made are appropriate for use as a licensing basis. The concerns outlined in the introduction to IE Bulletin 80-04 relative to, 1) limiting core conditions occurring during portions of the transient where auxiliary feedwater flow is a relevant contributor to plant cooldown; and 2) incomplete isolation of main feedwater flow, are not representative of the Westinghouse NSSS designs and associated Balance of Plant requirements. |
| The most limiting steam line break determined by Westinghouse was analyzed by Exxon Nuclear Co., Inc.and presented in XN-NF-77-40 Supplement 1,"Plant Transient Analysis for the R.E.Ginna Unit 1 Nuclear Power Plant," March 1980.This transient occurs at hot zero power with outside power available and the break occurring at the exit of the steam generator. | | The most limiting steam line break determined by Westinghouse was analyzed by Exxon Nuclear Co., Inc. |
| The Exxon analysis does not specifically account for auxiliary feedwater. | | and presented in XN-NF-77-40 Supplement 1, "Plant Transient Analysis for the R. E. Ginna Unit 1 Nuclear Power Plant," March 1980. This transient occurs at hot zero power with outside power available and the break occurring at the exit of the steam generator. |
| However, the Steam Generator heat.transfer model, using constant heat transfer coefficients, co'ntinues to calculate heat transfer from the primary to the secondary side after the broken steam generator has been estimated to be empty.If auxiliary flow was specifically accounted for, its effect would be negligible during the initial portion of the transient and would have minimal effect during later portions of the transient since by the time the broken steam generator empties, the total system reactivity is negative and core power is decreasing. | | The Exxon analysis does not specifically account for auxiliary feedwater. However, the Steam Generator heat. transfer model, using constant heat transfer coefficients, co'ntinues to calculate heat transfer from the primary to the secondary side after the broken steam generator has been estimated to be empty. If auxiliary flow was specifically accounted for, its effect would be negligible during the initial portion of the transient and would have minimal effect during later portions of the transient since by the time the broken steam generator empties, the total system reactivity is negative and core power is decreasing. The additional reactivity addition associated with the slight cooldown due to runout flow is more than negated by the boron reactivity inserted by safety injection. Therefore, the severity of the transient is not increased. |
| The additional reactivity addition associated with the slight cooldown due to runout flow is more than negated by the boron reactivity inserted by safety injection. | |
| Therefore, the severity of the transient is not increased. | |
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| Request 3.If the potential for containment overpressure exists: or the reactor-return-to-power response worsens, provide a proposed corrective action.If the unit is operating, provide a description of any interim action that will be taken until the proposed corrective action is completed. | | Request 3. If the potential for containment overpressure exists: |
| Response 3.Since neither the potential for containment over-pressurization nor the reactor-return-to-power response worsens no corrective action is required.Request 4.Within 90 days of the date of this Bulletin, complete the review and evaluation required by this Bulletin and provide a written response describing your reviews and actions taken in response to each item.Response 4.This attachment provides the required 90 day response to IE Bulletin No.80-04. | | or the reactor-return-to-power response worsens, provide a proposed corrective action. If the unit is operating, provide a description of any interim action that will be taken until the proposed corrective action is completed. |
| | Response 3. Since neither the potential for containment over-pressurization nor the reactor-return-to-power response worsens no corrective action is required. |
| | Request 4. Within 90 days of the date of this Bulletin, complete the review and evaluation required by this Bulletin and provide a written response describing your reviews and actions taken in response to each item. |
| | Response 4. This attachment provides the required 90 day response to IE Bulletin No. 80-04. |
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| Attachment 1'-Containment Ener Balance The purpose of this evaluation is to determine the length of time available to the operator to terminate'AFW flow to the broken steam generator following a steam line break inside containment,.prior to containment pressure exceeding 60 psig.Initial Conditions and Assum tions a., ZOCA energy release to containment (includes credit for containment heat sinks): 191.3 x 10 BTU (taken from Table'14.3.4-2 of FSAR).Results in peak containment pressure of 53 psig.b.Additional energy to reach 60 psig: 16 x 10 BTU (taken from Fig.14.3.4-3 of FSAR).C.Minimum safeguards heat removal capability (1 spray pump and 2 fan coolers)=55 x 10 BTU/sec=3.3 x 10 6 BTU/min.(taken from Fig.14.3.4-9 of'the FSAR).d.Energy input from 600 gpm to the broken steam I generator=600 gal x 1180 BTU man ibm x 62.6~lb x 1 ft=5.91 x 10 B'fU 6 ft 7.68 gal mz.n e.Energy released to containment from the initial steam line break accident blowdown=140 x 10 BTU (taken from Fig.14.2.5-10 of the FSAR). | | Attachment 1 '- Containment Ener Balance The purpose of this evaluation is to determine the length of time available to the operator to terminate'AFW flow to the broken steam generator following a steam line break inside containment,.prior to containment pressure exceeding 60 psig. |
| | Initial Conditions and Assum tions |
| | : a. , ZOCA energy release to containment (includes credit for containment heat sinks): 191.3 x 10 BTU (taken from Table '14.3.4-2 of FSAR). Results in peak containment pressure of 53 psig. |
| | : b. Additional energy to reach 60 psig: 16 x 10 BTU (taken from Fig. 14.3.4-3 of FSAR). |
| | C. Minimum safeguards heat removal capability (1 spray 6 |
| | pump and 2 fan coolers) = 55 x 10 BTU/sec = 3.3 x 10 BTU/min. (taken from Fig. 14.3.4-9 of'the FSAR). |
| | : d. Energy input from 600 gpm to the broken steam I |
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| | 600 gal x man 1180 BTU ibm x 62.6 ~lb ft x 1 ft 7.68 gal |
| | = 5.91 x 10 6 B'fU mz.n |
| | : e. Energy released to containment from the initial steam line break accident blowdown = 140 x 10 BTU (taken from Fig. 14.2.5-10 of the FSAR). |
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| Calculations l.Additional energy which could be absorbed by containment, taking credit for passive and minimum active containment, heat sinks following a steam line break.(from a+b-2.e above).(191.3+16)x 10 BTU-140 x 10 BTU=67.3 NBTU 6 6 Net energy addition to containment following initial steam line break blowdown (from d-c above): (5.91 x 10 BTU-3.3 x 10 BTU)=2.61 x 10 BTU 6 6 6 man mz.n mz.n 3.Operator Action Time=67.3 NBTU 2.61 NBTU/man f" lg}} | | Calculations |
| | : l. Additional energy which could be absorbed by containment, taking credit for passive and minimum active containment, heat sinks following a steam line break. (from a + b-e above). |
| | (191.3 + 16) x 10 6 BTU - 140 x 10 6 BTU = 67.3 NBTU |
| | : 2. Net energy addition to containment following initial steam line break blowdown (from d-c above): |
| | (5.91 x 10 6 BTU - 3.3 x 10 6 BTU ) = 2.61 x 10 6 BTU man mz.n mz.n |
| | : 3. Operator Action Time = 67.3 NBTU |
| | : 2. 61 NBTU/man |
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Category:CORRESPONDENCE-LETTERS
MONTHYEARML17250B3041999-10-20020 October 1999 Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated IR 05000244/19990081999-10-14014 October 1999 Forwards Insp Rept 50-244/99-08 on 990809-0919.Severity Level IV Violation of NRC Requirements Occurred & Being Treated as non-cited Violation Consistent with App C of Enforcement Policy ML17265A7651999-10-0808 October 1999 Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request ML17265A7641999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel ML17265A7631999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000 ML20212J3801999-09-30030 September 1999 Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21 ML20212J3561999-09-30030 September 1999 Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21 IR 05000244/19992011999-09-24024 September 1999 Forwards Insp Rept 50-244/99-201 (Operational Safeguards Response Evaluation) on 990621-24.No Violations Noted. Primary Purpose of Osre to Assess Licensee Ability to Respond to External Threat.Insp Rept Withheld IR 05000244/19990051999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs ML17265A7571999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs ML17265A7461999-08-31031 August 1999 Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726 ML17265A7411999-08-26026 August 1999 Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded ML17265A7401999-08-26026 August 1999 Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components ML17265A7451999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d) ML17250B3021999-08-23023 August 1999 Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806 ML17265A7271999-07-30030 July 1999 Forwards 10CFR21 Interim Rept Per Reporting of Defects & Noncompliance,Section 21 (a) (2).Interim Rept Prepared Because Evaluation Cannot Be Completed within 60 Days from Discovery of Deviation or Failure to Comply ML17265A7151999-07-23023 July 1999 Forwards LER 99-007-01 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Reactor Trip.Further Investigation of Event Identified Addl Corrective Actions ML17265A7061999-07-22022 July 1999 Forwards LER 98-003-02,re Actuations of CR Emergency Air Treatment Sys.All Creats Actuations,Including Those Originally Believed to Be Valid Actuations,Were,In Fact Invalid Actuations ML17265A7191999-07-21021 July 1999 Forwards Ginna Station ISI Rept for Refueling Outage Conducted in 1999 ML17265A7141999-07-21021 July 1999 Withdraws Relief Request 35 for Plant Inservice Insp Program Section XI Requirements,Submitted on 980806.Licensee Plans to Resubmit Relief Request,Which Includes Addl Level of Detail,In Near Future ML17265A7041999-07-16016 July 1999 Submits Info Re Specific Licensing Actions Which May Be Expected to Generate Complex Reviews,In Response to Administrative Ltr 99-02,dtd 990603 ML17265A6911999-06-30030 June 1999 Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Readiness Disclosure Attached ML17265A6871999-06-22022 June 1999 Forwards Response to RAI Made During 990225 Telcon Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Calculation Encl. Encl ML17265A6841999-06-21021 June 1999 Informs That Util Wishes to Amend Extend of Alternate Exams Provided for Relief Request Re ISI Program ASME Section XI Require Exams for First 10-Yr Interval for Containment ML17265A6741999-06-15015 June 1999 Submits Annual ECCS Rept IAW 10CFR50.46(a)(3)(ii) Requirements.No Changes Have Been Made to Large Break LOCA PCT & Small Break LOCA PCT ML17265A6751999-06-11011 June 1999 Responds to NRC RAI Re Licensee GL 96-05 Program.Encl Info Verifies That Util Is Implementing Provisions of JOG Program on MOV Periodic Verification ML17309A6551999-06-0707 June 1999 Responds to NRC 990310 RAI Re Verification of Seismic Adequacy of Mechanical & Electrical Equipment ML17265A6671999-06-0101 June 1999 Requests Approval of Ginna QA Program for Radioactive Matl Packages,Form 311,approval Number 0019.Ginna QA Program for Station Operation, Was Most Recently Submitted to NRC by Ltr Dtd 981221 & Supplemented on 990301 ML17265A6641999-05-25025 May 1999 Forwards Addl Info on Use of GIP Method a for Re Ginna Nuclear Power Plant.Copy of Re Ginna Station USI A-46 Outlier Resolution Table as Requested ML17265A6611999-05-24024 May 1999 Forwards LER 99-007-00 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Rt.Util Is Planning to Submit Suppl to LER by 990730 ML20206U0921999-05-13013 May 1999 Forwards Four Copies of Rev R to Gnpp Security Plan,Per Provisions of 10CFR50.54(p).Rev Clarifies Armed Response Team Assignments & Does Not Degrade Physical Security Effectiveness.Rev Withheld,Per 10CFR73.21 ML17265A6461999-05-12012 May 1999 Forwards 1998 Annual Radioactive Effluent Release Rept & 1998 Annual Radiological Environ Operating Rept, for Re Ginna NPP ML17265A6411999-05-12012 May 1999 Forwards LER 99-004-00 IAW 10CFR50.73 & 10CFR21.Further Assessment Will Be Provided in Suppl to LER by 990630 ML20206E7221999-04-29029 April 1999 Forwards Four Copies of Rev Q to Re Ginna Nuclear Power Plant Security Plan,Per 10CFR50.54(p).Changes Do Not Degrade Physical Security Effectiveness.Encl Withheld,Per 10CFR73.21 ML20206H6911999-04-22022 April 1999 Forwards Info Requested During Informal Telcon on 980408 Concerning Upcoming Osre at Ginna Station.Info Requested Listed.Without Encls ML17265A6271999-04-19019 April 1999 Forwards Rev 0 & Rev 1 to Colr,Cycle 28 for Re Ginna NPP, Per TS 5.6.5 ML17309A6501999-04-14014 April 1999 Forwards Revised Ginna Station EOPs & Procedures Index ML17265A6121999-03-29029 March 1999 Forwards Rept on Status of Decommissioning Funding for Re Ginna Npp,For Which Rg&E Is Sole Owner,Per 10CFR50.75. Data Presented Herein,Current as of 981231 ML17265A6041999-03-24024 March 1999 Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(ii)(B) & 10CFR21.Addl Analyses Are Being Performed to Support Future Cycle Operation & Supplemental LER Is Scheduled to Be Submitted by 990618 ML17265A5671999-03-0101 March 1999 Forwards Application for Amend to License DPR-18,to Revise TSs Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6).Supporting Tss,Encl ML17265A5641999-03-0101 March 1999 Forwards Response to NRC 990217 RAI Concerning Changes to QA Program for Re Ginna Station Operation.Rg&E Is Modifying Changes Requested in 981221 Submittal.Modified QA Program,Encl ML17265A5551999-02-25025 February 1999 Informs That Util Is in Process of Revising fitness-for-duty Program,Developed in Accordance with 10CFR26.Util Will Continue to Use Dept of Health & Human Svcs Certified Test Facility for Majority of Tests During Yr ML17265A5561999-02-22022 February 1999 Forwards FFD Performance Data Rept for Six Months Ending 981231,per 10CFR26.71(d) ML17309A6491999-02-12012 February 1999 Forwards Ginna Station EOPs ML17265A5451999-02-12012 February 1999 Forwards Simulator Four Year Certification Rept,Per 10CFR55.45(b)(5)(ii) ML17265A5431999-02-0909 February 1999 Supplements 980806 Relief Request with Attached Table.Util Third 10-Yr ISI of Reactor Vessel Being Performed During 1999 Refueling Outage,Beginning on 990301 ML17265A5361999-02-0202 February 1999 Forwards Response to NRC 981203 RAI Re Resolution of Unresolved Safety Issue USI A-46.Util Does Not Agree with NRCs Interpretation.Detailed Bases,Encl ML17311A0691999-01-25025 January 1999 Forwards Revs to Ginna Station Emergency Plan Implementing Procedures (Epips).Previous Rev Had Incorrect Effective Date ML17311A0671999-01-14014 January 1999 Forwards Revised Emergency Operating Procedures for Re Ginna NPP ML17265A5131999-01-12012 January 1999 Forwards Revised Cover Page for Ginna Station Technical Requirements Manual (Trm), Rev 7,correcting Error in Effective Date 1999-09-30
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML17250B3041999-10-20020 October 1999 Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated ML17265A7651999-10-0808 October 1999 Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request ML17265A7641999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel ML17265A7631999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000 ML20212J3801999-09-30030 September 1999 Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21 ML20212J3561999-09-30030 September 1999 Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21 IR 05000244/19990051999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs ML17265A7571999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs ML17265A7461999-08-31031 August 1999 Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726 ML17265A7411999-08-26026 August 1999 Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded ML17265A7401999-08-26026 August 1999 Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components ML17265A7451999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d) ML17250B3021999-08-23023 August 1999 Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806 ML17265A7271999-07-30030 July 1999 Forwards 10CFR21 Interim Rept Per Reporting of Defects & Noncompliance,Section 21 (a) (2).Interim Rept Prepared Because Evaluation Cannot Be Completed within 60 Days from Discovery of Deviation or Failure to Comply ML17265A7151999-07-23023 July 1999 Forwards LER 99-007-01 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Reactor Trip.Further Investigation of Event Identified Addl Corrective Actions ML17265A7061999-07-22022 July 1999 Forwards LER 98-003-02,re Actuations of CR Emergency Air Treatment Sys.All Creats Actuations,Including Those Originally Believed to Be Valid Actuations,Were,In Fact Invalid Actuations ML17265A7191999-07-21021 July 1999 Forwards Ginna Station ISI Rept for Refueling Outage Conducted in 1999 ML17265A7141999-07-21021 July 1999 Withdraws Relief Request 35 for Plant Inservice Insp Program Section XI Requirements,Submitted on 980806.Licensee Plans to Resubmit Relief Request,Which Includes Addl Level of Detail,In Near Future ML17265A7041999-07-16016 July 1999 Submits Info Re Specific Licensing Actions Which May Be Expected to Generate Complex Reviews,In Response to Administrative Ltr 99-02,dtd 990603 ML17265A6911999-06-30030 June 1999 Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Readiness Disclosure Attached ML17265A6871999-06-22022 June 1999 Forwards Response to RAI Made During 990225 Telcon Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Calculation Encl. Encl ML17265A6841999-06-21021 June 1999 Informs That Util Wishes to Amend Extend of Alternate Exams Provided for Relief Request Re ISI Program ASME Section XI Require Exams for First 10-Yr Interval for Containment ML17265A6741999-06-15015 June 1999 Submits Annual ECCS Rept IAW 10CFR50.46(a)(3)(ii) Requirements.No Changes Have Been Made to Large Break LOCA PCT & Small Break LOCA PCT ML17265A6751999-06-11011 June 1999 Responds to NRC RAI Re Licensee GL 96-05 Program.Encl Info Verifies That Util Is Implementing Provisions of JOG Program on MOV Periodic Verification ML17309A6551999-06-0707 June 1999 Responds to NRC 990310 RAI Re Verification of Seismic Adequacy of Mechanical & Electrical Equipment ML17265A6671999-06-0101 June 1999 Requests Approval of Ginna QA Program for Radioactive Matl Packages,Form 311,approval Number 0019.Ginna QA Program for Station Operation, Was Most Recently Submitted to NRC by Ltr Dtd 981221 & Supplemented on 990301 ML17265A6641999-05-25025 May 1999 Forwards Addl Info on Use of GIP Method a for Re Ginna Nuclear Power Plant.Copy of Re Ginna Station USI A-46 Outlier Resolution Table as Requested ML17265A6611999-05-24024 May 1999 Forwards LER 99-007-00 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Rt.Util Is Planning to Submit Suppl to LER by 990730 ML20206U0921999-05-13013 May 1999 Forwards Four Copies of Rev R to Gnpp Security Plan,Per Provisions of 10CFR50.54(p).Rev Clarifies Armed Response Team Assignments & Does Not Degrade Physical Security Effectiveness.Rev Withheld,Per 10CFR73.21 ML17265A6461999-05-12012 May 1999 Forwards 1998 Annual Radioactive Effluent Release Rept & 1998 Annual Radiological Environ Operating Rept, for Re Ginna NPP ML17265A6411999-05-12012 May 1999 Forwards LER 99-004-00 IAW 10CFR50.73 & 10CFR21.Further Assessment Will Be Provided in Suppl to LER by 990630 ML20206E7221999-04-29029 April 1999 Forwards Four Copies of Rev Q to Re Ginna Nuclear Power Plant Security Plan,Per 10CFR50.54(p).Changes Do Not Degrade Physical Security Effectiveness.Encl Withheld,Per 10CFR73.21 ML20206H6911999-04-22022 April 1999 Forwards Info Requested During Informal Telcon on 980408 Concerning Upcoming Osre at Ginna Station.Info Requested Listed.Without Encls ML17265A6271999-04-19019 April 1999 Forwards Rev 0 & Rev 1 to Colr,Cycle 28 for Re Ginna NPP, Per TS 5.6.5 ML17309A6501999-04-14014 April 1999 Forwards Revised Ginna Station EOPs & Procedures Index ML17265A6121999-03-29029 March 1999 Forwards Rept on Status of Decommissioning Funding for Re Ginna Npp,For Which Rg&E Is Sole Owner,Per 10CFR50.75. Data Presented Herein,Current as of 981231 ML17265A6041999-03-24024 March 1999 Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(ii)(B) & 10CFR21.Addl Analyses Are Being Performed to Support Future Cycle Operation & Supplemental LER Is Scheduled to Be Submitted by 990618 ML17265A5671999-03-0101 March 1999 Forwards Application for Amend to License DPR-18,to Revise TSs Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6).Supporting Tss,Encl ML17265A5641999-03-0101 March 1999 Forwards Response to NRC 990217 RAI Concerning Changes to QA Program for Re Ginna Station Operation.Rg&E Is Modifying Changes Requested in 981221 Submittal.Modified QA Program,Encl ML17265A5551999-02-25025 February 1999 Informs That Util Is in Process of Revising fitness-for-duty Program,Developed in Accordance with 10CFR26.Util Will Continue to Use Dept of Health & Human Svcs Certified Test Facility for Majority of Tests During Yr ML17265A5561999-02-22022 February 1999 Forwards FFD Performance Data Rept for Six Months Ending 981231,per 10CFR26.71(d) ML17309A6491999-02-12012 February 1999 Forwards Ginna Station EOPs ML17265A5451999-02-12012 February 1999 Forwards Simulator Four Year Certification Rept,Per 10CFR55.45(b)(5)(ii) ML17265A5431999-02-0909 February 1999 Supplements 980806 Relief Request with Attached Table.Util Third 10-Yr ISI of Reactor Vessel Being Performed During 1999 Refueling Outage,Beginning on 990301 ML17265A5361999-02-0202 February 1999 Forwards Response to NRC 981203 RAI Re Resolution of Unresolved Safety Issue USI A-46.Util Does Not Agree with NRCs Interpretation.Detailed Bases,Encl ML17311A0691999-01-25025 January 1999 Forwards Revs to Ginna Station Emergency Plan Implementing Procedures (Epips).Previous Rev Had Incorrect Effective Date ML17311A0671999-01-14014 January 1999 Forwards Revised Emergency Operating Procedures for Re Ginna NPP ML17265A5131999-01-12012 January 1999 Forwards Revised Cover Page for Ginna Station Technical Requirements Manual (Trm), Rev 7,correcting Error in Effective Date ML17265A5091999-01-11011 January 1999 Requests Relief Per 10CFR50.55a(a)(3)(ii) from Certain Requirements of Section XI of ASME B&PV Code for ISI Program.Relief Request 41 Encl ML17265A5101999-01-11011 January 1999 Requests Relief Per to 10CFR50.55a(a)(3)(ii) from Certain Requirements of Section XI of ASME B&PV Code for ISI Program.Relief Request 40 Encl 1999-09-30
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059L4481990-09-18018 September 1990 Forwards Responses to Rer 900709-13 Team Visit Findings, Per .Responses Withheld (Ref 10CFR73.21) ML17309A4491990-09-13013 September 1990 Forwards Slides Presented by Facility & Westinghouse to NRC in 900724 Meeting.Encl Withheld ML17262A1331990-09-11011 September 1990 Responds to Violations & Several Unresolved Items Noted in SSFI Rept 50-244/89-81.Update of Appropriate Unresolved Items Encl.Specific Actions Re All NRC Unresolved Items Being Tracked to Completion ML17261B1511990-08-29029 August 1990 Forwards Semiannual Radioactive Effluent Release Rept Jan- June 1990 & Rev 3 to Process Control Program for Ginna Station, Per Tech Specs 6.9.1.4 & 6.16,respectively ML17261B1481990-08-28028 August 1990 Lists Understanding of Issues Util Planning to Address Re Containment Integrity,Per 900718 Telcon.Any Concerns or Action Items Different from Listed Submittal Should Be Provided to Util Prior to NRC 900905 & 06 Visit to Plant ML17261B1501990-08-20020 August 1990 Submits Update on Util Plans to Provide Appropriate Ventilation to Intermediate Bldg Clean Side,Per Violation Noted in Insp Rept 50-244/88-26.Extensive Program to Manually Close All External Openings Implemented IR 05000244/19880261990-08-20020 August 1990 Submits Update on Util Plans to Provide Appropriate Ventilation to Intermediate Bldg Clean Side,Per Violation Noted in Insp Rept 50-244/88-26.Extensive Program to Manually Close All External Openings Implemented ML17261B1371990-08-17017 August 1990 Forwards Rev 0 to, Inservice Insp Rept for Third Interval (1990-1999),First Period,First Outage (1990) at Re Ginna Nuclear Power Plant. ML17309A4481990-08-16016 August 1990 Responds to NRC 900717 Ltr Re Violations Noted in Insp Rept 50-244/90-80.Corrective Actions:Westinghouse Drawing E-2508 Approved & Issued to Central Records ML17261B1351990-08-15015 August 1990 Provides Info & Assessment on Integrity of Connection of Containment Bldg to Foundation,Per 900808 Telcon.Util Believes That Existing Condition Not Safety Concern & No Reason Exists to Suspect Joint Will Not Perform Function ML17261B1361990-08-14014 August 1990 Responds to Commitment Tracking Concerns Noted in Insp Rept 50-244/90-09 & Planned Corrective Actions.Util Confirms Commitments Dates for Implementation of Effective Shelf Life Program & Comprehensive Preventive Maint Program for Parts ML17261B1331990-08-13013 August 1990 Recommits to Performing Enhanced primary-to-secondary Leak Rate Monitoring,Per NRC Bulletin 88-002 ML17261B1261990-07-30030 July 1990 Clarifies Commitment Made in 900316 RO Re Restoring Inoperable Fire Damper I-411-21-P.Util Plans to Design Removable Track Which Will Allow Charcoal Drawer to Be Manipulated.Definitive Schedule Will Be Provided in 60 Days ML20055H9291990-07-23023 July 1990 Forwards Revised Page 6:4 of Plant Security Plan.Page Withheld Per 10CFR73.21 ML17261B1091990-07-20020 July 1990 Advises That Structural Evaluations of Containment Sys Being Performed in Response to Containment Integrity Insp ML17309A4471990-07-17017 July 1990 Forwards Decommissioning Rept, for Plant,Per 10CFR50.33(k) & 50.75(b) ML17261B0881990-07-13013 July 1990 Provides Update to Util 860616 Ltr Re Implementation of NUREG-0737,Item 6.2,Suppl 1, Emergency Response Capability. ML17261B0921990-07-11011 July 1990 Responds to 900709 Request for Addl Info Re Containment Integrity Insp.Util Will Provide Preliminary Results by 900716 Re Where Groundwater Entering Annular Access Area. Meeting Proposed for Wk of 900723 ML20044B0481990-07-10010 July 1990 Discusses Review of Station Blackout Documentation,Per 10CFR50.63.Licensee Will Complete Enhancements to Station Blackout Documentation Identified in Attachment 1 as Indicated & Other Items Will Be Completed within 2 Yrs ML17309A4461990-07-0909 July 1990 Responds to Generic Ltr 90-04 Re Status of Licensee Implementation of Generic Safety Issues Resolved W/ Imposition of Requirements or Corrective Actions ML17261B0901990-07-0909 July 1990 Advises That Licensee Will Install Containment Isolation Signal Going to Valve AOV 745 by End of 1992 Refueling Outage,Per Util 900608 Ltr Notifying of Condition Outside Design Basis of Plant Under 10CFR50.72 Criterion ML17261B0931990-07-0909 July 1990 Responds to Generic Ltr 88-14, Instrument Air Supply Sys Problems Affecting Safety-Related Equipment. Util Completed All Actions Requested in Generic Ltr & Will Retain Documentation Verification for Min of 2 Yrs ML17250B2151990-06-29029 June 1990 Forwards Application for Amend to License DPR-18, Reformatting Auxiliary Electrical Sys Tech Specs & Action Statements for Offsite & Onsite Power Sources Available for Plant Auxiliaries ML17261B0851990-06-28028 June 1990 Forwards LER 89-016-02 Re 891117 Failure of Safety Injection Block/Unblock Switch Which Could Render Both Trains of Safety Injection Sys Inoperable.Also Reported Per Part 21 ML17250B1991990-06-0808 June 1990 Responds to NRC 900509 Ltr Re Violations Noted in Insp Rept 50-244/89-81.Corrective Actions:Improved Battery Load Profile Developed Incorporating Calculational Improvements Contained in Current Industry Std IEEE 485-1983 ML17250B1951990-06-0505 June 1990 Responds to Generic Ltr 89-08, Erosion/Corrosion Induced Pipe Wall Thinning. Util Developed Erosion/Corrosion Program for Single & Two Phase Sys Consistent W/Requirements of NUREG-1344 & NUMARC 870611 Rept ML17261B0691990-06-0101 June 1990 Discusses Testing Frequency for Insp of Incore Neutron Monitoring Sys Thimble Tubes,Per NRC Bulletin 88-009.Thimble Tube Indicating Greatest Wear Recently Repositioned in Effort to Minimize Future Wear ML17250B1851990-06-0101 June 1990 Forwards Application for Amend to License DPR-18,providing Guidance for Action Statements Associated W/Power Distribution Limit Specs ML17250B1811990-05-31031 May 1990 Forwards Addl Info Re Response to Notice of Violation from Insp Rept 50-244/90-04.Info Withheld (Ref 10CFR73.21) ML17261B0961990-05-29029 May 1990 Responds to NRC 900427 Ltr Re Violations Noted in Insp Rept 50-244/90-07.Corrective Actions:Worker Records Immediately Corrected Indicating That Worker Received No Significant Exposure for Time While Error Occurred ML17250B1801990-05-29029 May 1990 Responds to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount. Util Does Not Need to Develop Enhanced Surveillance Program to Monitor Currently Installed Transmitter Based Upon Limited Installed Quantity ML17250B1841990-05-29029 May 1990 Responds to NRC Bulletin 89-002, Stress Corrosion Cracking of High Hardness Type 410 Stainless Steel (SS) Internal Pre-Load Bolting.... Valves Disassembled & 410 SS Studs Removed & Visually & Liquid Penetrant Examined ML17250B1721990-05-18018 May 1990 Forwards Rev 1 to Summary Exam Rept for 1990 Steam Generator Eddy Current Insp. Rept Summarizes Observations & Corrective Actions Resulting from Insps Performed During 1990 Outage ML20042G8931990-05-0808 May 1990 Forwards Revised Emergency Operating Procedures,Including Rev 3 to AP-RHR.2,Rev 1 to ES-0.3 & Rev 11 to ES-1.3 ML17261B0611990-04-26026 April 1990 Advises of Changes to 890201 Commitments Made Under Programmed Enhancement Response to Generic Ltr 88-17.Changes Will Not Reduce Capability to Operate Safely in Reduced Inventory Condition or Scope of Programmed Enhancements ML17261B0591990-04-17017 April 1990 Advises That No Interlocks Required for RHR motor-operated Valves 701 & 720 Based on Present Arrangement.Listed Failures Would Have to Occur in Order for Potential Overpressurization of RHR Sys to Occur ML17261B0461990-04-12012 April 1990 Responds to Issues Discussed During 900323 Telcon Re Inservice Testing Program Status & Relief Request.Current Test Methodology for Seat Leakage Acceptable Based on Application of Direct Measurement Sys ML17262A1431990-04-12012 April 1990 Responds to NRC 900202 Ltr Re Weaknesses Noted in Insp Rept 50-244/89-80.Corrective Actions:Procedure ES-0.3 Modified to Provide Guidance for Rapid Cooldown & Depressurization W/ & W/O Reactor & Vessel Instrumentation Sys ML17250B1451990-04-0606 April 1990 Discusses Impact of SER Issuance for Inservice Insp & Inservice Testing Programs & Advises That Timing Will Not Affect Util Implementation Plans for Programs,Except as Listed ML17250B1441990-04-0505 April 1990 Responds to NRC 900309 Ltr Re Violations Noted in Insp Rept 50-244/90-02.Corrective Actions:Positive Indicator Will Be Installed for All Check Valves to Enhance Visible Positive Position Verification Ability & to Avoid Confusion ML17261B0401990-03-30030 March 1990 Advises That Final Results of Station Blackout Documentation Review Will Be Submitted to NRC on or About 900501 ML17261B0321990-03-28028 March 1990 Forwards Annual Rept of ECCS Model Revs as Applicable to Facility,Per 10CFR50.46.Mods to Model for Small Break LOCA Do Not Affect Calculated Peak Clad Temp ML17261B0331990-03-28028 March 1990 Forwards Info Re Reactor Vessel Issues,Per 900305 Telcon Concerning Change of License Expiration Date ML17261B0231990-03-26026 March 1990 Responds to NRC 890222 Ltr Re Violations Noted in Insp Rept 50-244/89-17.Corrective Actions:Personnel Verified Safety Injection Block/Unblock Switch in Proper Position & Operator Procedure 0-1.1 Changed as Indicated ML17261B0281990-03-23023 March 1990 Forwards Addl Info Re Proposed Tech Spec Amend Concerning Use of Reconstituted Fuel,Per 900315 Telcon.During Fuel Assembly Reconstitution,Failed Fuel Rods Will Be Placed W/ Filler Rods ML17261B0221990-03-22022 March 1990 Provides Revised Test Schedule for motor-operated Valve Diagnostic Test Program,Per IE Bulletin 85-003.NRC Notification of Changes to Valve Operability Program Required by Generic Ltr 89-10,dtd 890828 ML17261B0411990-03-20020 March 1990 Forwards Summary of Onsite Property Damage Coverage Currently in Force at Plant,Per 10CFR50.54(w)(4) ML17261B0151990-03-19019 March 1990 Responds to Generic Ltr 89-19, Safety Implication of Control Sys in LWR Nuclear Plants (USI A-47). Overfill Protection Provided Through Trip Bistables in Reactor Protection Racks Powered from 120-volt Instrument Buses ML17261B0101990-03-15015 March 1990 Forwards Application for Amend to License DPR-18,allowing Use of Reconstituted Fuel Assemblies ML20012D2831990-03-14014 March 1990 Responds to Issues Discussed During 900307 Telcon W/Util & Eg&G Re Inservice Testing Program Status & Relief Request. Valves 5960A & 5960B Will Be Disassembled to Verify Forward Flow.Relief Requests PR-8 & VR-25 Encl 1990-09-18
[Table view] |
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~Jlili$1illliiI !1<rlivi<piivl ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N.Y. L4649 LEON O. iVHITE, JR. TCLCPHOVC VICC PRCSIDCNT iPCi COOC ria 546-2700 April 30, 1980 Mr. Boyce H. Grier, Director U.S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region I 651 Park Avenue King of Prussia, Pennsylvania 19406
Subject:
Response to IE Bulletin No. '80-04 o
Dear Mr. Grier.:
Enclosed is our response to IE Bulletin No. 80-04. This bulletin was received on February 8, 1980.
Sincerely, XC: Office of Inspection and Enforcement Division of Reactor Operations Inspection U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Res onse to IE Bulletin 80-04 Request 1. Review the containment pressure response analysis to determine if the potential for containment overpressure for a main steam line break inside containment included the impact of runout flow from the auxiliary feedwater system and the impact of other energy sources, such as continuation of feedwater or condensate flow. In your review, consider. your ability to detect and isolate the damaged steam generator from these sources and the ability of the pumps to remain operable after extended operation at, runout flow.
I Response 1. Although the Ginna post-steamline break containment pressure analysis in the FSAR did not include the effects of auxiliary feedwater flow to the affected steam generator, it is important to recognize that the evaluation also did not include the benefits of passive and active heat sinks inside containment.
Continued feedwater/condensate addition to the steam generator will not occur, since, the safety injection signal (generated by a variety of process parameters, including high steam line flow, high containment pressure, and low pressurizer pressure) will close the feedwater control valves and stop the feedwater pumps. The addition of maximum auxiliary feedwater flow to the broken steam generator will eventually require operator action to 1) realign flow to the intact generator, 2) terminate auxiliary feedwater flow to the broken generators. Positive information is available to the operator to determine which is the affected steam generator. Steam generator level instrumentation is located inside containment and steam generator pressure is located outside contain-ment where it. would not be affected by the accident environment, inside containment. It is expected that, through proper training and by use of the emergency procedures, the operators will be capable of quickly recognizing the steam line break, and will perform the proper operations.
There is substantial time available for the operator to perform the two safety functions noted above.
The SEP Safe Shutdown review concluded following their site visit in June 1978 that one steam generator would not boil dry for over thirty minutes. Thus there is substantial time to align flow to the intact steam generator. The termination of auxiliary feedwater flow to the affected steam generator, under the pessimistic circumstances, would require more rapid action (but still easily within the capability of the operators) to maintain containment
pressure below design pressure. The analysis presented in Attachment 1 concludes that, assuming minimum safeguards for .containment cooling, auxiliary feedwater flow would have to be terminated in about 26 minutes.
With maximum safeguards, this time would be extended to about 44 minutes.
There is no need to consider the operation of the auxiliary feedwater pumps at runout flow. The turbine-driven pumps are controlled by a'overnor, and will not, exceed about 400 gpm. The motor driven pump flow is controlled by the AFW control valves, which receive an automatic throttle signal to 200 gpm from their flow controllers. A potential single failure of the flow controller to control. flow to 200 gpm is not considered a worst-case single failure in terms of net energy addition to the containment, since the operation of all containment cooling safeguards (vs. the minimum safeguards assumed in this evaluation) would result in a substantial increase in energy removal from containment.
Request 2. Review your analysis of the reactivity increase which results from a main steam line break inside or outside containment. This review should consider the reactor cooldown rate and the potential for the reactor to return to power with the most reactive control rod in the fully withdrawn position. If your previous analysis did not consider all potential water sources ( such as those listed in 1 above) and if the reactivity increase is greater than previous analysis indicated the report of th'is review should include:
aI The boundary conditions for the analysis, e.g.,
the end of life shutdown margin, the moderator temperature coefficient, power level and the net effect of the associated steam generator water inventory on the reactor system cooling, etc.,
- b. The most restrictive single active failure in the safety injection system and the effect of that failure on delaying the delivery of high concentration boric acid solution to the reactor coolant system, c The effect of extended water supply to the affected steam generator on the core criticality and return to power,
- d. The hot channel factors corresponding to the most reactive rod in the fully withdrawn'osition at the end of life, and the Minimum Departure from Nucleate Boiling Ratio ( MDNBR) values for the analyzed transient.
Response 2. Westinghouse Electric Corporation performed the original steam break analysis for Ginna as reported in the FSAR and a reanalysis submitted to the NRC in September 1975. Westinghouse has reviewed the assumptions made for main and auxiliary feedwater flow as they apply to licensing basis steam line break transients. Several of the relevant assumptions used in all core transient analyses follow, and are further explained in the Ginna FSAR.
- 1. The reactor is assumed initially to be at hot shutdown conditions, at the minimum allowable shutdown margin.
- 2. For the Condition IV breaks, i.e., double-ended rupture of a main steam pipe, full main feedwater is assumed from the beginning of the transient at a very conservative cold temperature.
- 3. All auxiliary feedwater pumps are initially assumed to be operating, in addition to the main feedwater. The flow is equivalent to the rated flow of all pumps at the steam generator design pressure.
4 Feedwater is assumed to continue at its initial flow rate until feedwater isolation is complete, approximately 10 seconds after the break occurs, while auxiliary feedwater is assumed to continue at its initial flow 'rate.
- 5. Main feedwater flow is completely terminated following feedwater isolation.
Based on the manner in which the analysis is performed ~
for Westinghouse plants, the core transient results are very insensitive to auxiliary feedwater flow.
The first minute of the transient is dominated entirely by the steam flow contribution to primary-secondary heat transfer, which is the forcing function for both the reactivity and thermal-hydraulic transients in the core. The effect of auxiliary feedwater runout (or failure of runout protection where applicable) is minimal. Greater feedwater flow during the large steamline breaks serves to reduce secondary pressures, accelerating the automatic safeguards actions, i.e.
steamline isolation, feedwater isolation and safety injection. The assumptions described above are
therefore appropriate and conservative for the short-term aspect of the steamline break transient.
The auxiliary feedwater flow becomes a-dominant factor in determining the duration and,magnitude of the steam flow transient'uring later stages in the transient. However, the limiting portion of the transient occurs during the first minute, both due to higher steam flows inherently present early in the transient and due to the introduction of boron to the core via the safety injection system.
In conclusion, Westinghouse has evaluated the effect of runout auxiliary feedwater flows in the core transient for steamline break, and based on this evaluation, has determined that the assumptions presently made are appropriate for use as a licensing basis. The concerns outlined in the introduction to IE Bulletin 80-04 relative to, 1) limiting core conditions occurring during portions of the transient where auxiliary feedwater flow is a relevant contributor to plant cooldown; and 2) incomplete isolation of main feedwater flow, are not representative of the Westinghouse NSSS designs and associated Balance of Plant requirements.
The most limiting steam line break determined by Westinghouse was analyzed by Exxon Nuclear Co., Inc.
and presented in XN-NF-77-40 Supplement 1, "Plant Transient Analysis for the R. E. Ginna Unit 1 Nuclear Power Plant," March 1980. This transient occurs at hot zero power with outside power available and the break occurring at the exit of the steam generator.
The Exxon analysis does not specifically account for auxiliary feedwater. However, the Steam Generator heat. transfer model, using constant heat transfer coefficients, co'ntinues to calculate heat transfer from the primary to the secondary side after the broken steam generator has been estimated to be empty. If auxiliary flow was specifically accounted for, its effect would be negligible during the initial portion of the transient and would have minimal effect during later portions of the transient since by the time the broken steam generator empties, the total system reactivity is negative and core power is decreasing. The additional reactivity addition associated with the slight cooldown due to runout flow is more than negated by the boron reactivity inserted by safety injection. Therefore, the severity of the transient is not increased.
Request 3. If the potential for containment overpressure exists:
or the reactor-return-to-power response worsens, provide a proposed corrective action. If the unit is operating, provide a description of any interim action that will be taken until the proposed corrective action is completed.
Response 3. Since neither the potential for containment over-pressurization nor the reactor-return-to-power response worsens no corrective action is required.
Request 4. Within 90 days of the date of this Bulletin, complete the review and evaluation required by this Bulletin and provide a written response describing your reviews and actions taken in response to each item.
Response 4. This attachment provides the required 90 day response to IE Bulletin No. 80-04.
Attachment 1 '- Containment Ener Balance The purpose of this evaluation is to determine the length of time available to the operator to terminate'AFW flow to the broken steam generator following a steam line break inside containment,.prior to containment pressure exceeding 60 psig.
Initial Conditions and Assum tions
- a. , ZOCA energy release to containment (includes credit for containment heat sinks): 191.3 x 10 BTU (taken from Table '14.3.4-2 of FSAR). Results in peak containment pressure of 53 psig.
- b. Additional energy to reach 60 psig: 16 x 10 BTU (taken from Fig. 14.3.4-3 of FSAR).
C. Minimum safeguards heat removal capability (1 spray 6
pump and 2 fan coolers) = 55 x 10 BTU/sec = 3.3 x 10 BTU/min. (taken from Fig. 14.3.4-9 of'the FSAR).
- d. Energy input from 600 gpm to the broken steam I
generator =
600 gal x man 1180 BTU ibm x 62.6 ~lb ft x 1 ft 7.68 gal
= 5.91 x 10 6 B'fU mz.n
- e. Energy released to containment from the initial steam line break accident blowdown = 140 x 10 BTU (taken from Fig. 14.2.5-10 of the FSAR).
Calculations
- l. Additional energy which could be absorbed by containment, taking credit for passive and minimum active containment, heat sinks following a steam line break. (from a + b-e above).
(191.3 + 16) x 10 6 BTU - 140 x 10 6 BTU = 67.3 NBTU
- 2. Net energy addition to containment following initial steam line break blowdown (from d-c above):
(5.91 x 10 6 BTU - 3.3 x 10 6 BTU ) = 2.61 x 10 6 BTU man mz.n mz.n
- 3. Operator Action Time = 67.3 NBTU
- 2. 61 NBTU/man
f" lg