ML20078A616
| ML20078A616 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 09/16/1983 |
| From: | Lundvall A BALTIMORE GAS & ELECTRIC CO. |
| To: | John Miller Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20078A619 | List: |
| References | |
| RTR-NUREG-0660, RTR-NUREG-0737, RTR-NUREG-660, RTR-NUREG-737, TASK-2.E.1.2, TASK-TM NUDOCS 8309230345 | |
| Download: ML20078A616 (9) | |
Text
{{#Wiki_filter:.. BALTIM ORE GAS AND ELECTRIC CHARLES CENTER.P.O. BOX 1475 BALTIMORE, MARYLAND 21203 ARTHUR E. LuP4DvALL, JR. vice PREseDEM sumv September 16,1983 Director of Nuclear Reactor Regulation Attention: Mr. 3. R. Miller, Chief Operating Reactors Branch #3 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Subject:
Calvert Cliffs Nuclear Power Plant Units Nos.1 & 2; Dockets Nos. 50-317 and 50-318 Supplement 2 to Unit 1 Cycle 7 License Amendment
References:
(a) BG&E letter from Mr. A. E. Lundvall, Jr. to Mr. R. A. Clark (NRC), dated August 22,1983. (b) BG&E letter from Mr. A. E. Lundvall, Jr. to Mr. R. A. Clark (NRC), dated December 13,1979. (c) BG&E letter from Mr. A. E. Lundvall, Jr. to Mr. R. A. Clark (NRC), dated November 18,1980. (d) BG&E letter from R. F. Ash to Mr. D. H. Jaffe (NRC), dated September 20,1982. (e) BG&E letter from Mr. A. E. Lundvall, Jr. to Mr. R. A. Clark (NRC), dated November 5,1983. (f) Letter from Mr. D. H. Jaffe (NRC), to Mr. A. E. Lundvall, Jr., dated January 10,1983. (g) Letter from Mr. D. H. Jaffe (NRC), to Mr. A. E. Lundvall, Jr., dated 3anuary 18,1983. (h) BG&E letter from Mr. A. E. Lundvall, 3r., to Mr. 3. R. Miller (NRC), dated September 1,1983, Supplement I to Seventh Cycle License Application. (i) Letter from Mr. D. H 3affe (NRC) to Mr. A. E. Lundvall, Jr., dated January 3,1983. Gentlemen: In Referenu (a) we provided an advance request for amendment to Operating Licenses Nos. DPR-53 and DPR-69 to reflect planned modifications to the Auxiliary Feedwater System (AFWS) and certain post-accident and remote monitoring instrumentation. We indicated that we would submit the proposed technical specification
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Mr. 3. R. Mill::r Septsmber 16,1983 changes and a fir.ai determination with regard to significant hazards considerations by September 6,1983. Advance negotiations with our NRC Project Manager regarding the content of this request has resulted in a delay in providing our submittal, but were felt necessary to ensure that all appropriate information was included. By this letter we are also withdrawing our application in Reference (a) for amendment of our License No. DPR-69. A request for amendment to the Unit 2 AFWS technical specifications will be submitted at a later date. Consequently, the information that follows and all previously submitted information concerning our application in Reference (a) pertains to Unit 1 only. PROPOSED CHANGES Delete the following pages of the Technical Specifications and add new pages as indicated (proposed replacement pages are attached): Auxiliary Feedwater System (a) delete page 7-5 and add new page 7-5 (b) delete page 7-Sa and add new page 7-Sa (c) delete page B7-2 and aM r.cw page B7-2 (d) delete page 7-5b Remote Shutdown Monitoring Instrumentation (a) delete page 3-38 and add new page 3-38 (b) delete page 3-39 and add new page 3-39 (c) delete page 341 and add new page 341 (d) delete page 342 and add new page 342 DISCUSSION Auxiliary Feedwater System (FCR 83-1042) The NRC Action Plan developed as a result of the TMI-2 Accident, (NUREG-0660, Vol.1), specifies the requirement for reliability analysis and design modifications for PWR auxiliary feedwater systems. NUREG-0737, Item II.E.1.2 specifies the requirements for auxiliary feedwater system automatic initiation and flow Indication. As a result of these requirements we committed in Reference (b) to further l improving the reliability of the existing AFWS (consisting of two steam-driven pump trains) by adding a third train powered by a motor-driven pump. Installation of the AFWS third train at Unit I will be completed during the fall 1983 refueling outage. The physical work to be performed consists of adding the motor-driven pump and associated piping, valves, and controls, and permanent cross-connect piping between the discharge piping of the Unit I and Unit 2 motor-driven pumps. A motor-driven auxiliary feedwater - train was installed at Unit 2 during the previous refueling outage (Cycle 5) and is r, --
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Mr. 3. R. Mill:;r Septe.mber 16,1983 currently operational. A description of the modified AFWS and the new safety-grade Auxiliary Feedwater Actuation System (AFAS) was provided in References (c) and (d). To support plant operations following completion of these modifications we propose to modify the Unit 1 Technical Specifications to include appropriate limiting conditions for operation (LCO) and surveillance requirements for the third train. Thcse changes are related to those that were requested for Unit 2 in Reference (e) and approved by the NRC in References (f) and (g). A detailed description of the proposed changes is provided below. The definition of an operable Unit 1 AFWS (Technical Specification 3.7.1.2) is being revised to be consistent with that for Unit 2. Specifically, an operable AFWS would consist of one steam-driven and one motor-driven train. An operable steam-driven train consists of one pump aligned for automatic flow initiation and one pump aligned in standby. The standby pump will be available for operation but aligned so that automatic flow initiation is defeated upon AFAS actuation. Figure 3.7-1, which specifies the flow limits for the AFWS, is being deleted. In the existing configuration, auxiliary feedwater flow is controlled via pre-set flow control valves. These valves are being replaced with air-operated modulating valves which will automatically control flow within the system design limits. Since the performance of the AFWS flow control valves will be verified by periodic surveillance, the flow limits specified in Figure 3.7-1 are no longer required. A special test exception is proposed for T.S. 3.7.1.2 (first footnote) which will allow the automatic initiation features of the Unit I auxiliary feedwater system to be inoperable for a single period, following startup from the Cycle 7 refueling, of up to 30 days after entering MODE 3. This special test exception notwithstanding, automatic actuation of the AFWS is now a requirement for MODE 3; thus, the existing footnote to T.S. 3.7.1.2 is being deleted. A two-part remedial action statement (T.S. 3.7.1.2.a.1) is being proposed to reflect the addition of the motor-driven pump. In the event that the motor-driven pump became inoperable, the plant operators would be required to align the standby steam-driven pump to automatic initiating status within 72 hours or be in hot shutdown with in the next 12 hours, and to restore the inoperable motor-driven pump to operable status within the next 14 days or be in hot shutdown within the next 12 hours. The Unit 2 Technical Specifications currently require that the plant be shutdown if the motor-driven pump cannot be restored to operable status within 72 hours. The proposed two-part action st tement would extend this period for Unit I to a maximum of 14 days by taking credit for the continued ability to align two pumps to automatic initiating status. The remedial action associated with an inoperable steam-driven pump (T.S. 3.7.1.2.a.2) is being modified to extend the allowable period of continued operations from 72 hours to 30 days. This additional flexibility is balanced against a requirement to place the remaining operable steam-driven into automatic initiating shtus within 72 hours.
Mr. 3. R. Mill::r September 16,1983 In .the event that any two of the three AFWS pumps were to become inoperable at the same time, a new reraedial action statement (T.S. 3.7.1.2.b) is being added that would require the operators to verify within one hour that the remaining pump is aligned for automatic initiation and that the cross connect between the Unit I and Unit 2 motor-driven trains is operable and capable of delivering AFW flow to the affected unit upon manual initiation. In addition, the operators would be. required to restore a second pump to automatic initiating status within 72 hours. If these actions could not be successfully completed, the operators would be required to place the unit in hot shutdown within the following 12 hours. T. S. 3.7.1.2c is being modified to reflect the fact that during monthly logic tests both Auxiliary Feedwater subsystems are Inoperable. To avoid pumping Auxiliary
- Feedwater to the steam generators during power operations, the steam inlet valves to the steam driven pump and the motor-driven pump outlet valve are shut. At present the technical specification implies that only one subsystem may be secured at the same time and must be revised to reflect the addition of the motor-driven subsystem.
- To ensure that plant operational flexibility is not unnecessarily restricted in the event- that a motor-driven or a single steam-driven pump becomes inoperable, we propose a statement (T.S. 3.7.1.2.d) that deletes applicability of T.S. 3.0.4 as long as any two auxiliary feedwater trains can be aligned for automatic initiation. This provision will ensure that the intent of the LCO will continue to be satisfied.
i Various changes to the Surveillance Requirements of T.S. 4.7.1.2 are proposed to reflect changes'to the auxiliary feedwater system. Since the existing pre-set flow
- - . control valves will be replaced with air-operated modulating valves, we propose to delete the existing T.S. 4.7.1.2.d requirement to verify flow in accordance with Figure 3.7-1.
] T.S. 4.7.1.2.c (proposed) requires the performance of a flow verification test every 18 months to assure that each auxiliary feedwater pump delivers flow to each flow leg upon automatic initiation. Paragraph 4.7.1.2.a.1 currently states that the secondary steam pressure must be greater than 800 psig whe! the steam-driven pump performance verification test is conducted. This requirement was only an operational consideration to ensure a minimum steam supply pressure to the steam driven pumps. In reviewing the Updated Final Safety t
- Analysis Report, steam supply pressure is defined by the minimum pressure to maintain L constant pump speed. The actual value referenced in ' the FSAR is below the corresponding saturation pressure for the minimum temperature limit required in MODE ,
' 3 and higher MODES; therefore, we have deleted reference to this limit since we always operate well above the minimum design bases limit. Instead, we propose a minimum RCS e temperature limit of 300 F if pump operability is to be demonstrated during startup. r A m*nimum dynamic head of 3100 ft is included in the proposed motor-driven L ~ pump operability verification test (T.S 4.7.1.2.a.2). Finally, the bases for T. S. 3/4 7.1.2, " Auxiliary Feedwater System," is being revised as a result of the modifications to the AFWS. 'Ihe existing TS 3/4 7.1.2 is based upon two steam-driven auxiliary feedwater pumps which are automatically started with
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Mr. 3. R. Milizr Septembtr 16,1983 flow controlling valves (1/2-CV-4512) preset to supply between 100 and 130 gpm. This configuration allows up to 20 minutes for the reactor operator to terminate auxiliary feedwater flow in the event of an overcooling transient,' or to increase auxiliary feedwater flow to maintain an adequate heat sink. The revised TS 3/4 7.1.2 is based upon , automatic actuation of one steam-driven auxiliary feedwater pump and the motor-driven auxiliary feedwater pump with the AFAS providing a modulated flow of 160 gpm (t,10 gpm). l The transients affected by AFWS performance, and the assumptions used the 4 the analysis of these transients, are being listed in the basis for T.S. 3/4 7.1.2. Licensing grade analyses have demonstrated that for the first 10 minutes no flow is needed for undercooling transients, and that the maximum Auxiliary Feedwater suction leg flow is acceptable for overcooling transients. Although the operational nominal flow setpoint of 160 gpm is discussed, but it is made clear that flow fluctuations beyond the discussed band are allowable. Post-Accident and Remote Shutdown Monitoring Instrumentation (FCR 83-1042 and 83-1053) Associated with the redesign of the auxiliary feedwater system are changes to the new safe shutdown panel which is being installed and located to meet the requirements of Appendix R (Fire Protection Rule) of 10 CFR 50. Als panel, which is located on the 45' elevation of the safety-related switchgear rocm (immediately adjacent to the Control Room), replaces the existing shutdown panel located in the steam driven auxiliary feedwater pump room. The results of the Appendix R review necessitate l various changes in Remote Shutdown and Post Accident Instrumentation provided on the alternate safe shtudown panel and in the control room as listed in Technical Specification i Tables 3.3-9,~ 3,3-10, 4.3-6, and 4.3-10. Specifically,the Power Range Flux and the Reactor Coolant Flow monitors are being deleted from T. S. Tables 3.3-10 and 4.3-10. The function of post-accident flux monitoring will be performed by the Wide Range Logarithmic Neutron Flux monitor in the event that an evacuation of the control room is necessitated by fire. The Wida . Range -Logarithmic Neutron Flux monitor is listed on Tables 3.3-10 and 4.3-10. he measurement range of the Wide Range Neutron Flux monitor is currently identified on T. S. Table 3.3-9 as being from 0.1 cps to 150% power. %e upper limit of this range is i being increased to 200% power. Reactor coolant system total flow is considered a non-essential channel of instrumentation in the post-trip condition because of the availability of Reactor Coolant System (RCS) temperature indication, RCS subcooled margin and Reactor Coolant Pumps status. In the absence of adequate core flow, RCS temperature and subcooled margin are i indirect indicators of core flow conditions and provide adequate display to ensure , appropriate actions are initiated by operations personnel to recover from any abnormal post-trip conditions. Neither the Total Reactor Coolant Flow nor the Power Range Nuclear Flux monitor is referenced in the Calvert Cliffs emergency operating procedures. I _. . - _ _ . . _ _ _ _ _..__,___.________.~,.,,____..,_______,s
'Mr. 3. R.' Mill:r Septembtr 16,1983 Reactor Coolant Cold Leg Temperature measurement as shown on T.S. Table
' 3.3-9 is being changed to indicate a range of 212 F to 705 F. Reactor Coolant Cold Leg
- Temperature output is utilized by the subcooled margin monitors. The lower range limit of 212 F corresponds to the boiling point of water at atmospheric pressure; thus temperature measurement below this lower limit would provide no additional useful input to the subcooled margin monitor. Similarly, the upper limit of 705 F corresponds to the critical point of water, above which temperature measurements are not meaningful.
Pressurizer Pressure measurement as shown on T. S. Table 3.3-9 is being changed to indicate a range of 0 psia to 4000. psia. The existing range of pressurizer pressure indication is 0-1600 psia. A new instrument with the higher range is being installed concurrently with the AFWS modifications in r,rder to satisfy the RCS pressure
- measurement requirements associated with the resolution of ATWS issues.
Steam generator level measurement (also shown on T.S. Tables 4.3-6, 3.3-9, 3.3-10 and 4.3-10) is being modified to provide an extended range of indication. The i_ existing instrumentation indicates level from -116 to +63.5 inches. A new instrument will increase the range to -401 to +63.5 inches. His wide range steam generator level i measurement channel will provide indication at the alternate safe shutdown panel. The addition of wide range level indication provides the operator with more representative information of actual steam generator inventory during post-trip / shutdown conditions.
- DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS Auxiliary Feedwater System We have reviewed the proposed technical _ specification changes and have 5 i determined that they do not involve a significant hazards consideration. The intent of the existing LCO is to ensure that two auxiliary feedwater pumps and associated flowpaths are operable and capable automatically initiating flow. De proposed revision to the LCO maintains this intent by requiring that the motor-driven pump and one steam-driven pump be aligned for automatic initiation.. De remedial action statements' l associated with the loss'of the motor-driven pump or a single steam-driven pump provide for transferring the standby steam-driven pump to automatic initiating status with 72 hours. Once this operation is successfully completed a period of 14 days is allowed for restoring an inoperable motor-driven pump to operating status. De corresponding time -
- period for the. inoperable steam-driven train is 30 days, which is consistent with the current Unit 2 Technical Specifications. A shorter duration is proposed for an inoperable motor-driven pump in recognition of the importance of providing diversity in power ,
source. l De remedial action statement associated with two inoperable AFW p'umpi ' l, maintains the intent of the LCO by requiring that a second pump be aligned for . automatic initiating status within 72 hours or, failing this, that the unit be placed in hot ..'~ shutdown within 12 hours. During this 72-hour period the operators must also verify , (within one hour) that the Unit 2 motor-driven nump is oparable and that the cross' connect valve (2-CV-4550) has been exercised within the previous 31 daysf.or place the unit in hot shutdown within 12 hours.
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- Mr. 3. R. Mill:r September 16,1983 The above remedial actions, when considered with the system reliability Improvements provided by .the addition of the motor-driven train, the cross connect between the Unit I and 2 motor-driven pumps, and the new automatic feedwater actuation system, provide for' an overall increase in the margin of safety currently provided in the_ technical specifications. The maximum period of operation that would be allowed with only one automatically actuated AFW pump in operable status (72 hours) is not being increased, while the proposed specification would require additional remedial actions (verification of cross-connect operability) by the operators beyond that which is currently _ required by the technical specifications.
, The 30-day special test exception is required to accomodate testing of the entire Unit 1 AFWS during MODE 1 cperations following the Cycle 7 refueling outage. This testing is a prerequisite to declaring the, automatic f-edwater actuation system (AFAS) operable.1 Prior to entering MODE 3 adequate testhg will have been performed to confirm that all pumps, valves, and controls will operate manually. This exception is considered justified because: (1) Operators will be aware of the fact that the system is in manual, or being tested and will be particularly alert to any transients requiring the startup or, termination of Auxiliary Feedwater flow;.end V- _ _ m. . (2) Prior to the Unit 2 Cycle 6 reload.the plant was operated at all times 'with a O manually operated Auxi!Iary Feedwater System. ~
~ (3) The safety analyses perforrNd "to evaluate the effects of AFT flow i . deviations show that no flow for' 10 minutes after the m?st severe undercooling trancient, or runout flon for 10 minutes after the most severe
'C overcooling transient, is acceptab!e. Consequently, adequate time would be available to manually initiate and/or control auxiliary feedwater> flow in the event that either of these transients were to occur during the pioposed 30 day period.
- ku. To ensure that the plant operators are sufficientlylaware of the l- status of the Unit 1 AFWS during testing, a warning sign {wiIl be pladed on p 2~ -
-{ the AFWS panel in the control room to clearly indicate when~tha systern is in
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manual. In addition, each oper,ator shift will be briefed once per week on the status of the system until such time as AFWS testing is complete and P y / ' automatic operation is restored. - t i ; %- j-The proposed changes to the suiveil' lance requirements contained in I L T.S. 4.7.1.2 are consistent with the new automatic Ipitiation and flow control features of . i a ;the AFWS and will provide adequate assurance that the system will be maintained with
. sufficient capacity to meet the operability requirements of the LCO (T.S. 3.7.1.2).
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m 'Ihdperformance $1'the modified AFWS has been evaluated in Reference (h) x. l and. in our letter'of NovemMr 17; 1982 (which presented an evaluation of the Unit 2 f AFWS). The three desigt basis events which are affected by the automatic actuation of the modified AFWS; ile., loss of feedwater, feed line break, and main stram line break; E, s .
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5)$ i have been reanalyzed. De results indicate that with system operation in accordance with the proposed technical spe:lfication, all relevant acceptince criteria continue to be 4 met. , , y Thus, we conclude tha' t the proposed technical specifications do not involve a '
- ^ significant hazards consideration in that they would not:
3 $.. a , (1) Involve a significant increase in the probability or consequences of an '~ St 4 , accident previously evaluat-d; or e (2), Create the possibility:-of a new or different kind of accident from any A ' accident previously evaluated; or a (3) Irivolve a significant reduction in the margin of safety currently provided by 7 the technical specifications. , o' , _ s Post-Accident and Remoted Monitoring Instrumentadon The instrumentation on the new alternate safe shutdown panelis provided for remote monitoring purposes 'to assist the operators .in mitigating the effects of an 6 accident and placing the plant in a safe shutdown conditions therefore, deletion of instrumentation from the alternate safe shutdown panel would not create the possibility of an accident tnot previously. considered. The Power Range Flux and Reactor Coolant Flow monitors are not required at the alternate safe shutd_own panel. Neutron flux monitoring in tte power range is accomplished with the Ylde Range Logarithmic Neutron Flux Monitor. Adequate Indication or reactor coolant if$w can obtained by monitoring i . s RCS temperature indication, RCS subcooled margin, and reactor coolant pumps status.
' . _ Consequently, the deletion of the Power Range Flux and Reactor Coolant Fio* rnansrs from the list of remote monitoring instrumentation will not significantly
, , . rease the ' probability or consequences of any accident, nor will the margin of safety
. ; contained in the technical specifications be reduced. These changes are identical to
~ those that were previously reviewed and approved by the NRC for Unit 2 in Reference (1).
l Re proposed change in the Recitor Coolant Cold Leg temperature measurement range in T.S. Table 3.3-9 is justified because indication outside this range will not provide information that is meaningful icr the safety function that is being supported by this variable; 1.e., maintenanc'e'of adequate core cooling.
- j. Q The proposed change to the Pressurize.< Pressure measurement range in T.S.
Table 3.3-9 is justified because the increased range will ensure that indication of RCS
- pressure will be provided to the operator during all postulated RCS overpressurization transients. Consequently, this change improves the quality of the information provided to the operator and represents an increase.,in the margin of safety afforded by' the
- technical specifications. Similarly, the proposed increase in the steam generator. level measurerrent rangeOis appropriate because it provides the operator with more Q representative Information on actual steam generator inventory during post-trip or %tg' un shutdown conditions.
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Mr. 3. R. Mill;r Septemb r 16,1983 SAFETY COMMITTEE REVIEW These proposed changes to the Technical Specifications has been reviewed by our Plant Operations and Safety and Off-Site Review Committees, and they have concluded that implementation of these changes will not result in an undue risk to the health and safety to the public. FEE DETERMINATION The fee for this application was forwarded with Reference (a). Sincerely, j
- f. A AEL/BSM/vf Attachments cc: 3. A. Biddison, Jr., Esq.
G. F. Trowbridge, Esq. Mr. D. H. Jaffe, NRC Mr. R. E. Archit7el, NRC Mr. R. R. Mills, , 'E Mr. R. E. Cochran, DHMH _ - .}}