ML20202G489
| ML20202G489 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 04/16/1986 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2409, NUDOCS 8607150403 | |
| Download: ML20202G489 (39) | |
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.s' DATE ISSUED: 4/16/86 El IM ACRS COMBINED DHRS/ECCS SUBCOMMITTEE MEETING MINUTES MARCH 26, 1986 WASHINGTON, DC PURPOSE: The purpose of the meeting was to: (1) continue review of the Duke Power Company request to delete the ECCS UHI system at McGuire, Units 1 and 2, and (2) discuss the proposed NRR resolution position for Generic Issue 124: " Auxiliary Feedwater System Reliability."
ATTENDEES: Principle meeting attendees included: ACRS NRC D. Ward, Chairman D. Hood, NRR G. Reed, Member W. Minners, NRR I. Catton, Consultant S. Diab, NRR P. Davis, Consultant J. Watt, NRR V. Schrock, Consultant J. Wilson, NRR H. Sullivan, Consultant B. Sheron, NRR C. Tien, Consultant N. Lauben, NRR P. Boehnert, Staff SNL Westinghouse L. Buxton B. McIntyre D. Shimeck Duke Power W. Reckley Meeting Highlights, Agreements and Requests
- 1. Mr. Ward noted that the Subcommittee was first briefed on the UHI deletion request over one year ago. Given the Utility's shutdown schedule for McGuire Unit 2, NRR's review process has been rushed.
The Subcommittee will accommodate this schedule to the extent we 1 can but he said we owe ourselves a careful review of this item. He
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* .' s DHRS/ECCS Meeting Minutes March 26, 1986 also noted that we are subject to current Appendix K requirements and that we must carefully consider the LB LOCA phenomena in that light.
Mr. Reed asked that during the day's presentation on UHI, a careful delineation is made between equipment inside and outside contain-ment. He also noted that an analysis (by SNL) showed a benefit from malfunctioning (and nitrogen injection) of the UHI. He said he is concerned about the use of accumulators on PWRs, believing them to be a potential safety risk.
- 2. Mr. Darl Hood (NRR-LPM - McGuire Plant) revhwed the purpose of the meeting on UHI - namely to obtain ACRS comment on the Duke Power proposal to initially disable, then eventually delete, the upper head injection system on McGuire, Unit 2. Unit 2 is currently shutdown for refuelir.g until N mid-May. In response to Dr. Catton, Mr. Lauben (NRR) said that the deletion of UHI and associated hardware changes (see below) will make McGuire like a standard W plant vis-a-vis the ECCS/LOCA analysis.
Mr. Ward said the question of disabling / deletion is irrelevant for the Subcommittees concerns; i.e., the system should be considered unavailable as far as we are concerned.
- 3. Mr. W. Reckley (Duke Power) reviewed the Duke Power UHI celetion request. He reviewed the licensing history of this request (Fig-ures I and 2). In response to Mr. Ward, Mr. McIntyre indicated that the BASH ECCS code has had problems with oscillations. Mr.
Ward said that he was concerned that Duke and W have not been straight forward in acknowledging problems with the BASH code. Duke discussed how the LOCA linear heat rate limits impact plant operation and fuel cycle design (Figure 3). Mr. Reed questioned
a $ DHRS/FCCS Meeting Minutes March 26, 1986 the assumption that additional waste is generated due to soluble boron concentration. The Duke Power representative noted that there was a problem with the thermal regenerator in the baron resin system and that Duke was in the process of remedying the situation. The actions required for UHI isolation include: (1) replace flow restricting orifices in cold leg accumulator discharge piping; (2) charge nominal water level in cold leg accumulators and modify level instrumentation and increase nominal cover gas pressure in cold leg accumulators; (3) close UHI isolation valves and maintain in the closed position; and (4) modify plant procedures. The actions required for ultimate removal of UHI (scheduled for a 1987/88 outage) are shown in Figure 4. In response to a previous Subcommittee question, Duke Power has not identified any viable alternate use(s) for the UHI system. The only UHI-related component with a planned alternate use is the containment penetration. It will be used as a liquid sampling line. As a result of further discussion, it was noted that Duke Power believes UHI is an operational burden with no real safety benefit.
- 4. Mr. D. Shimeck (W) discussed the LOCA and transient analyses to support UHI deletion at McGuire. He discussed the LOCA analysis requirements per Appendix K. In response to Mr. Ward, W said that maximum PCT is usually seen for the LOCA analysis with a discharge coefficient of 0.6. As a result of'further discussion, it was noted that maximum safety injection results in an increase in PCT because of Appendix K requirements (injection water is " thrown ,
away" during blowdown which aggravates the steam binding problem, thus slowing reflood). For McGuire, the PCT penalty for " maximum SI" is a 230 (Figure 5). l l i t
* . m DHRS/ECCS Meeting Minutes March 26, 1986
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The details of the non-UHI analysis for McGuire were discussed: The PCT is 2132 F (Figure 6). The analysis shows results that are typical for any four-loop E plant (Figures 7-9). Mr. B. McIntyre (E) discussed the details of the UHI analysis. His presentation focused on the history of VHI EM development and the problems encountered. Key concessions (or conservatisms) with the EM centered on the heat transfer model and the modeling of the mixing of the UHI injection water in the upper head (perfect / imperfect mixing model). Dr. Catton raised a number of questions regarding the mixing model. He was attempting to determine how much cooling margin is lost with deletion of UHI. Dr. Tien and Mr. Ward asked what the impact is on PCT for.the perfect mixing vs imperfect mixing assumptions. The consultants noted that imperfect mixing is the more valid assump-tion. E responded that imperfect mixing usually results in lower PCTs. The perfect mixing model is the bounding case (highest PCTs). Dr. Catton asked how an ice condenser plant can exhibit the same LOCA behavior as a large dry plant given the low backpres-sure/ steam binding phenomena. Dr. Sullivan noted that g appears to have improved the EM code but he was unsure just which part of the code is affected. Mr. McIntyre made the following points regarding the UHI model: The benefit of UHI is seen in the initial blowdown portion of the LOCA when the heat transfer is assumed to be very low. In responsetoDr.Catton,ysaidtheheattransfercoefficient is weakly dependent on backpressure vis-a-vis FLECHT experi-ments. The major difference between the early UHI calcu-lations and BART is the more accurate modeling of the heat transfer correlation. Another factor resulting in lower PCT
e $ . DHRS/ECCS Meeting Minutes March 26, 1986 is the specific upper head configuration in McGuire. The penalty associated with the low backpressure prevents McGuire running at an Fq of 2.32 with UHI deleted. (McGuire will- run with an Fg of2.26.) 1 As a result of additional discussion, it was noted by-Dr. Sullivan that the model improvements and plant specific geometry (upper head internals - 4% bypass) result in no significant difference between UHI and non-UHI LOCA results. Since no calculations were run early-on, it is not possible'to quantify the margin given up by deletion of UHI. He also said this appears to be a licensing issue and that no real problem is apparent. Results of LOCA calculations run for the D.C. Cook (non-UHI) ice condenser plant shows little difference from McGuire plant results (Figure 10). A set of BE calculations for W UHI and non-UHI plants (generic UHI plant, not McGuire-specific) shows little difference in PCT's (Figure 11).
- Analysis of non-LOCA transients (SLB, etc.) showed no excee-dance of DNB limits.
Mr. Schrock asked about consideration of SB LOCA vis-a-vis j i UHI. W noted that the NOTRUMP calculation showed PCTs in the 1400'F + 200 F range with and without UHI. For very small breaks, UHI is not activated. Mr. Ward asked about intermedi-ate breaks. Mr. McIntyre indicated that the effect on PCT should be minimal. I I I 1 5
s $ . t OHRS/ECCS Meeting Minutes March 26, 1986 In response to Dr. Sullivan, Mr. Lauben said post-TMI small break calculations showed a lower PCT for larger breaks. Mr. Ward again reiterated that the Subcommittee should assure itself the plant will be operated in a safe manner independent of Appendix K " licensing-space" considerations. Dr. Catton said that it is hard to evaluate plant safety given the non-physical phenomena forced upon us by Appendix K.
- 5. L. Buxton (SNL) reviewed a set of analyses using TRAC for a generic UHI plant assuming: (1) a stuck-open accumulator valve resulting in nitrogen gas injection into the upper head, and (2) a calcu-lation with the UHI removed and the cold-leg accumulator recon-figured similar to a regular 4-loop W_ plant. The 3-D vessel option was used in these calculations. Figures 12-14 show the noding details and boundary conditions used. Dr. Catton commented that the coarse noding used in the core will result in modeling that is more conservative than best estimate (BE).
The results for the stuck valve calculation were: (1) PCT occurs during blowdown and is the same with or without valve failure; (2) enhanced core cooling during reflood is seen due to prolonged UHI water delivery; and (3) most of the injected nitrogen flows to the break via the hot legs and the cooling jets. Figures 15 and 16 show some of the calculational results. Mr. Reed asked if the nitrogen injection was a problem for a SB LOCA if the leak is isolated. Mr. Buxton could not respond as he only analyzed the LB LOCA case. For the UHI deletion calculation, some model changes were made (Figure 17). The results noted included: (1) clad temperatures increase after blowdown until cold-leg accumulator water refills the lower plenum; and, (2) PCT occurs during reflood instead of
s , DHRS/ECCS Meeting Minutes March 26, 1986 blowdown and the reflood PCT is only 54*F higher than blowdown PCT. Figures 18 and 19 show some if the details of this analysis. Dr. Catton made the point that Appendix K requirements are serious-ly distorting the ECCS/LOCA bardware setpoints to the potential detriment of safety. As a result of additional discussions, it was noted by Mr. Buxton that the major differences in the various calculational results for non-UHI versus VHI are due mostly to the assumptions in the codes. Mr. Schrock and Dr. Catton disagreed with this statement arguing that not enought is known about the UHI phenomena to support this assumption. Dr. Sullivan said he be-lieves that there are no major differences in the results of any of the calculations shown. Conclusions noted by Sandia were that: (1) best estimate TRAC-PF1/ MOD-1 calculations of large break LOCA suggest UHI shutoff valve failure has no large effect, and that removal of the UHI accumulator is not overly detrimental; and, (2) all best estimate PCTs are well below the licensing limit. Mr. Davis noted that a PCT during reflood could result in large radiation releases for a low pressure containment. Further dis-cussion resulted in the statement that the core will void for a plant with UHI deleted just as will be seen in a regular W plant. The reflood rate will be slowed for the ice condenser plant which can result in a higher PCT in reflood. However, NRR said that given the PCTs expected (14-1500 F), little cladding damage should j be seen. j i l
- 6. J. Watt (NRR) discussed the NRR review of the UHI deletion request.
Mr. Watt noted that the constraints of licensing requirements (Appendix K) result in penalizing the UHI plant. NRR noted that l
o DHRS/ECCS Meeting Minutes March 26, 1986 the best estimate calculations show that PCT increases by N 400 F without UHI. Dr. Catton noted that deleting UHI shows an increase in risk as the PCT goes up to the 1400*F range and nears the clad rupture temperature regime. Mr. Ward asked if NRR considered the rupture of cladding a major health risk. Dr. Sheron said the LB LOCA core melt risk is dom-inated by equipment reliability. On an overall risk basis, the LB LOCA is a relatively low risk contributor. Dr. Sullivan commented that the best estimate curve in the NRR SER (Figure 20) is overly conservative given the high F 'sq assumed in the analysis.
- 7. Mr. Ward polled the Subcommittee on the Duke Power request. He indicated that the data seems to show that UHI provides no real safety benefit and is an annoyance operationally. He noted uncer-tainties in SE vs EM calculations. Comments received by the Subcommittee Members and Consultants included:
Mr. Reed - He agrees VHI shwid be deleted due to concerns with industrial safety and the leakage potential of UHI piping. The UHI accumulator, with the potential for nitrogen injection given a SB LOCA, concerns him. He also recommended that the gF for McGuire be reduced and that restrictions similar to those seen on DC Cook be imposed. These restrictions could be lifted after revision of Appendix K. Dr. Catton - Supports VHI removal but recommends NRC drop the Fg to be similar to other W large dry containment plants. He would like to see data on the impact of low pressure on PCT. Remove it, but in a way that is reversible so if Appendix K is revised UHI can be used to its maximum benefit.
DHRS/ECCS Meeting Minutes March 26, 1986 Dr. Schrock - Calculations are flawed and in general a sound decision is hard to reach. The lack of SB LOCA calculations further confuses matters. NRC is shaky ground to argue "because it meets licensing specifications it OK". We are smarter than we were 10 years ago and we should do calculations to show it. He is not in a position to carefully evaluate this choice. Dr. Sullivan - Does McGuire UHI deletion request comply with 10 CFR 50.46 and Appendix K? Yes. What does the BE model show and does UHI removal impact plant operation? Removal will improve operation j greatly and helps transients situations (no nitrogen gas problem). BE analyses are confusing, there is no consistent picture. Based on data shown though-removal is not a concern. t j Dr. Tien - He found the presentations confusing and does not feel comfortable with the documentation and data shown; it is not consistent. He can't make a sound recommendation; operationally it's probably OK to remove UHI but he would like to see more consistent data. Dr. Davis - need to consider more than the ECCS problem. He shares Mr. Reed's SB LOCA concern and believes the interfacing system LOCA risk is increased as well. Overall, he favors removal of the system but he is not a thermal hydraulics expert. Mr. Ward - Subcommittee consensus seems to be that removal is ! compelled by reduction in CM risk. He noted that neither NRR or Duke Power made this point. P. Abraham noted that a reduction in Fq negates the advantage of deleting UHI. He said that McGuire plans to stay at their current Fqvalue.
DHRS/ECCS Meeting Minutes March 26, 1986 Mr. Rossi noted that NRR is limited by regulation as to what they can proscribe for licensees (i.e., if regulations are met they must allow deletion of UHI).
- 8. S. Diab (NRR) reviewed the status of the resolution effort for Generic Issue 124: " Auxiliary Feedwater System Reliability."
Figure 21 shows an outline of the NRR presentation. Figures 22-23 details the history and status of this issue. NRR noted that AFW system failures are still concurring at a rate that gives concern. Most of these failures are considered precur-sors to a total loss of AFW. An INP0 review of operational perfor-mance in US PWRs from 1980-1984 concluded that industry-wide AFW failures increased each year from 1980-1984. Mr. Reed cautioned that the root cause of many of these events is, in his opinion, poor design. An AE00 study has also shown that AFW reliability can show large variations over time at a given plant. Mr. Ward noted that AFW systems are not subject to consistent demand rates given PWR design differences. This can reflect on the reliability of a given plant and should be factored into the AE0D work. Mr. Ward indicated that AE00 should be more aggressive in obtaining operating data to determine AFW reliability. The seven operating PWRs with AFW systems deemed less reliable than the rest of the PWR OLs are shown in Figure 24. Given the above, NRR has concluded that three-train AFW systems are more reliable than 2-train systems. Increased interdependence among three-train systems does however reduce the advantage over 2 j trains. i
DHRS/ECCS Meeting Minutes March 26, 1986 NRR proposes the following requirements for resolution of GI 124:
- All PWR Licensees perform a rigorous PRA of their AFWs to demonstrate an acceptable reliability (10 10-5/ demand).
All relevant parameters should be considered in the analysis (bleed and feed capability, etc.).
- Relevant failure data should be reevaluated by each PWR licensee at least once every 5 years to assure AFW reliability remains acceptable.
- Plants that cannot satisfy the reliability criterion would propose suitable modifications (hardware, procedures, etc.) to comply.
NRR would issue a Generic Letter to delineate the above require-ments. In response to Mr. Reed, Dr. Sheron indicated that NRR policy is to not allow bleed and feed to be used to meet regulatory requirements vis-a-vis DHR. Mr. Reed said that NRR should focus on bleed and feed as a back-up to AFW, should it fail. Mr. Diab showed a list of other licensing issues related to Generic Issue 124 (Figure 25). These Issues will be integrated with 124 as appropriate. Mr. Davis suggested that USI A-44 and the Safety Goals should also be factored into this Item. Dr. Sheron indicated that the AFW is considered a first-line safety system and that a judgment must be made as to whether we want to rely on back-up system (s) for this function.
DHRS/ECCS Meeting Minutes March 26, 1986 The schedule for resolution of this issue was shown (Figure 26). ACRS Subcommittee review is scheduled for May 1986 with full Committee review expected in June 1986. Details of the Value/ Impact analysis supporting the proposed resolution of Generic Issue 124 were discussed. The results of the analysis (Figure 27) show the amount of money that can be expended (justified) for a fix of the issue. Mr. Reed suggested that the analysis should also show the actual core melt risk in order to provide meaningful information. NRR agreed and will provide this data in the next revision.
- 9. The meeting was adjourned at 3:30 p.m.
NOTE: Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room, 1717 H Street, NW, Washington, DC, or can be purchased from ACE-Federal Reporters, 444 North Capital Street, Washington, DC 20001, (202) 347-3700. l
w LICENSING HISTORY JAN/FEB 1985 MEETINGS WITH NRC STAFF AND ACRS ECCS SUBCOMMITTEE
- MAY 9, 1985 SUBMITTAL - TECH SPEC CHANGES REQUIRED FOR REMOVAL OF UHI
- OCTOBER 2, 1985 SUBMITTAL - SAFETY EVALUATION BASH LBLOCA ANALYSIS NOTRUMP SBLOCA ANALYSIS' STEAMLINE BREAK ANALYSIS CONTAINMENT RESPONSE EVALUATION
- OCTOBER 14, 1985 SUBMITTAL - TECH SPEC CHANGES (REV. 1), CLA VOLUME
- DECEMBER 17, 1985 SUBMITTAL - TECH SPEC CHANGES (REV, 2), SPECIFY REQUIREMENTS FOR UHI ISOLATED CONFIGURATION 1
LICENSING HISTORY
- DECEMBER 23, 1985 SUBMITTAL - RADIOLOGICAL ASSESSMENT OF UHI REMOVAL
- JANUARY 14, 1986 SUBMITTAL - CLARIFICATION OF DECEMBER 17, 1985 SUBMITTAL
- MARCH 17, 1986 SUBMITTAL - BART LBLOCA ANALYSIS
- APRIL 1986 SUBMITTAL - TECH SPEC CHANGES FOR UNIT 1 CYCLE 4 NUCLEAR DESIGN
- SUMMER / FALL 1986 SUBMITTAL - BASH LBLOCA ANALYSIS
& TECH SPEC CHiNGES (F-0) i
NUCLEAR DESIGN LIMITS LOCA ANALYSIS ASSUMPTION ON LINEAR HEAT RATE IS INPUT TO NUCLEAR DESIGN NUCLEAR DESIGN MUST VERIFY VARIOUS PARAMETERS SUCH AS CONTROL ROD POSITION, POWER, RADIAL POWER DIS-TRIBUTION, AXIAL POWER DISTRIBUTION, ETC. ARE LIMITED TO PREVENT EXCEEDING LOCA KW/FT LIMIT HIGHER LOCA LIMIT (F-0) ALLOWS MORE FLEXIBILITY IN PLANT OPERATION AND FUEL CYCLE DESIGN LESS SOLUBLE POISON PROCESSING (REDUCE WASTE GENERATION) POSSIBLE CYCLE LENGTH EXTENSION POSSIBLE DESIGN CHANGES SUCH AS LOW LEAKAGE LOADING PATTERNS, AXIAL BLANKETS, ETC. INCREASED ABILITY TO LOAD FOLLOW l l l 5 f/ U Ml
ACTIONS REQUIRED FOR UHI REMOVAL
.- PENETRATIONS TO UPPER HEAD CUT AND CAPPED UHI PIPING AND FILL & VENT PIPING, REMOVED FROM CONTAINMENT CONTAINMENT PENETRATIONS CAPPED OR REUSED COLD LEG ACCUMULATOR CONFIGURATION IS SAME AS UHI ISOLATION MODIFY PLANT PROCEDURES NO FINAL PLANS FOR AUXILIARY BUILDING COMPONENTS (TANKS, PIPING, ETC.)
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MAX SI IMPACT ON 4-LOOP WESTINGHOUSE. PLANT ECCS PERFORMANCE MIN SI MAX SI MAX SI PLANT MODEL PCT PCT PENALTY BYRON 1978 2104 2110 6 WOLF CREEK 1978 2114 2174 60 MILLSTONE 3 1981 2100 2132 32 D.C. COOK #1 1981 1999 2170 171 D.C. COOK #1 1981/BART 1937 2154 217 DUKE, UHI REMOVED 1981/BART 1895 2132 237 a , s
- THE MAGNITUDE OF THE PENALTY IS INVERSELY RELATED TO AVERAGE FLOODING RATE.
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t - -* . MARCH 26, 1986 PRESENTATION TO THE ACRS DHR/ECSS' SUBCOMMITTEE AUXILIARY FEEDWATER SYSTEM (AFWS) RELIABILITY GENERIC ISSUE N0, 124 l 1 THIS IS A STATUS PROGRESS REPORT EVOLUTION OF STAFF REVIEW 0F THE AFWS-R ISSUE 4 POSSIBLE CORRELATION OF CALCULATED AFWS RELIABILITIES AND OPERATING EXPERIENCE
- STAFF OBSERVATIONS
- INP0 REVIEW
- AEOD REVIEW I
- FINDINGS
- PROGRAK l
- ISSUES RELATED TO GI-124 PROPOSED NRC ACTION VALUE/ IMPACT STUDY -
SUMMARY
i SCHEDULE FOR RESOLUTION a
'. . .[ . ,
EVOLUTION OF STAFF REVIEW 0F THE AFWS REllABILITY ISSUE STAFF CRITERIA FOR AFWS DESIGN PRIOR T0 ISSUANCE OF THE SRP (1975) SPECIFIED GOOD ENGINEERING PRACTICE. PLANTS REVIEWED AGANIST THE SRP (1975 TO PRESENT) HAVE PROVIDED A " SAFETY-RELATED" SYSTEM. A REVIEW 0F THE AFWS IN PLANTS LICENSED PRIOR TO THE TM1-2 ACCIDENT (1979) WAS UNDERTAKEN BY THE B&O TASK FORCE. BOTH A DETERMINISTIC AND PROBABILISTIC EVALUATION OF RELIABILITY WAS PERFORMED BY THE STAFF. RESULTS WERE PUBLISHED IN NUREG-0611 AND NUREG-0635 FOR WESTINGHOUSE AND COMBUSTION ENGINEERING PLANTS
- RESPECTIVELY.
B8W PLANTS PERFORMED THEIR OWN AFWS RELIABILITY STUDIES (BAW-1584) CONSISTENT WITH NUREG-0611 AND NUREG-0635. THESE STUDIES WERE REVIEWED BY THE STAFF AND PLANT SPECIFIC SERs ISSUED.
- ALL PLANTS WERE REVIEWED AGAINST THE CRITERIA 0F NUREG-0737, ITEMS II.E.1.1 AND ll.E.1.2 IN ORDER TO CONFIRM OR IMPROVE AFWS RELIABILITY, AND WERE FOUND ACCEPTABLE.
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ALL PLANTS WERE REVIEWED AGANIST THE CRITERIA 0F GENERIC LETTER 8]-14 TO IMPROVE AFWS SEISMIC RESISTANCE AND FOUND ACCEPTABLE WITH THE EXCEPTION OF OCONEE, i SRP SECTION 10.4.9 WAS REVISED JULY 1981 TO INCORPORATE I NUMERICAL RELIABILITY CRITERION, 10-4 TO 10-5 UNAVAIL,/ DEMAND EASED ON KNOWLEDGE GAINED DURING B80 TASK FORCE REVIEWS, ALL PWRs NOT REVIEWED IN NUREG-0611 AND NUREG-0635 AND LICENSED AFTER THE TMI-2 ACCIDENT SUBMITTED A RELIABILITY STUDY UTILIZING A COMPARABLE APPROACH, CURPENT OPERATING LICENSE APPLANTS COMPLY WITH THE 1 SRP REllABILITY GOAL. l IE BULLETIN 85-01 WAS ISSUED IN OCTOBER, 1985 REQUIRING PLANTS TO DEMONSTRATE PROTECTION AGANIST STEAM BINDING OF AFWP (GI-93), STAFF REVIEW 0F THIS ISSUE IS NOT COMPLETE. STAFF REVIEW 0F AFWS RELIABILITY HAS CONTINUED SINCE C0FPLET10N OF lHE AB0VE ACTIONS, DIVISION OF SYSTEMS INTEGRATION PREPARED DRAFT REGULATORY ANALYSIS TO REASSESS AFWS REllABILITY, AFTER NRR REORGANIZATION THE AFWS REllABILITY ISSUE WAS ASSIGNED TO DSR0 AS GENERIC ISSUE GI-124 WITH A PRIORITY RANKING OF "NEARLY RESOLVED," FEBURARY 86,
7 PLANTS WITH AFWSs THAT ARE LESS RELIABLE THAN THE REST OF PWRs: ANO-1: ONE MDP, AND ONE TDP AFWS ANO-2: ONE MDP, AND ONE TDP AFWS CRYSTAL RIVER: ONE MDP, AND ONE TDP AFWS FT, CALHOUN: ONE MDP, AND ONE TDP AFWS PRAIRIE ISLAND-1 8 2: ONE MDP, AND ONE TDP AFWS PER UNIT (WITH MANUAL CAPABILITY TO ALIGN THE MDPs TO EITHER UNIT) RANCHO SECO: ONE MDP, AND ONE TDP AFWS f]$-
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ISSUES RELATED TO GI-124 THESE ISSUES ARE: USI A-45, DECAY HEAT REMOVAL REQUIREMENTS GI-68, LOSS OF AFWS DUE TO A HELB GI-93, STEAM BINDING 0F AFW PUMPS (GL 85-0], 10/29/85) GI-122.1A, COMMON MODE FAILURE OF ISOLATION VALVES IB, RECOVERY OF AFW Jc, INTERRUPTION OF AFW FLOW 2, INITIATING FEED AND BLEED ELEMENTS OF Gis-68, 93, AND 122 WILL BE INTEGRATED WITH GI-124, AS APFROPRIATE.
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j,*'o.t . SCHEDULE FOR GI-124 RESOLUTION t ISSUE PROPOSED RESOLUTION FOR NRR COMMENT 3/28/86 PRESENT PROPOSED RESOLUTION TO NRR DIRECTOR 4/21/86 I ACRS MEETING 5/86 l SEND PACKAGE TO CRGR 6/86 l ISSUE PROPOSED RESOLUTION ?0R PUBLIC COMMENT 7/86 INCORPORATE COMMENTS AND PRESENT TO 11/86 NRR DIRECTOR
! NRR DIRECTOR REVIEW COMPLETE, SEND TO CRGR 12/86
- CRGR REVIEW COMPLETE 1/87 ISSUE GENERIC LETTER 2/87 1
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Table 7 Main Conclusions Dollars Per Person-Rem Expenditure Averted $ Including ORE Justified Plant Co re-Mel t ' Per Person and Averted at $1000 Per Name Frequency Rem saved Costs Person-Rem ANO-1 4.4E-5 13K IK 180K ANO-2 2.9E-5 15K SK 150K Crystal River 3 8.1E-5 10K -7K 230K Ft. Calhoun 3.2E-5 34K 20K 70K Prairie Island 1 7.2E-5 9K 3K 260K Prairie Island 2 7.2E-5 8K 3K 270K Rancho Seco 4.1E-5 14K -2K 170K Other PWRs 7.2E-5 5K -1K 480K 1 1
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