ML20040A319
| ML20040A319 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 10/18/1976 |
| From: | Parr O Office of Nuclear Reactor Regulation |
| To: | Suhrke K BABCOCK & WILCOX CO. |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201200758 | |
| Download: ML20040A319 (9) | |
Text
{{#Wiki_filter:I ~ o ~ ~ -r Distribution: .g-Docket File V. A. Moore ~ flRC PDR R. H. Vollmer OCT 181976-Local PDR M. L. Ernst LWR #3 File W. P. Gamill D. B. Vassallo W. Mcdonald, MIPC F. J. Williams ELD
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DocketIlo.STil50-561 x ._:7. Cox ACRS (16) M. Rushbrook R. Heineman Babcock & Wilcox Ccapany D. Ross ATTH: Mr. Kenneth E. Suhrke Manager, Licensing J. Knight bec: J. B'uchanan R. Tedesco T. Abernathy iluclear Power Generation H. Denton P. O. Box 1260 Lynchburg, Virginia 24505 Genticcen: ROUND 2 POSITIONS A:lD REQUEST FOR IllF0DIATIO!! As a result of our continuing review of the Babcock & Wilcox Standard Safety Analysis Repor.t BSAR-205, your response to certain Regulatory staff positions and requests for information is required. The specific information required is detailed in the Enclosure. Regulatory Positions are identified by (P.SP) underneath the position nunber shown in the Enclosure. He are prepared to meet with you to discuss further any of our positions to assure complete understanding of the factors at issue and tco bases for our positions; however, we do not believe extended or iterative debate would be useful. In order to maintain our licensing review schedule, we'need your complete respcnses to the Enclosure itens by flovember 15, 1976. Please infom us within seven days after receipt of this letter of the date on which you plan to respond so that we may revise our schedule if necessary. If you plan to appeal to licensing management s on any of these positions, please advise us of your intentions g within two weeks. -F Please centact us if you have any questions. Sincerely, Original Signed by, O. D. Paa p Olan D. Parr, Chief light Water Reactors Branch !;o. 3 Division of Project Management
Enclosure:
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L.:..-.-....... . : ~ .a. u -..-.._x.. _. _ _ .....a i ENCLOSURE POSITIONS AND REQUESTS FOR ADDITIONAL INFORMATION BSAR-205 BOCKET NO: STN 50-561 212.0 REACTOR SYSTEMS 212.222 With regard to the response to question 212.35, the staf f posi-(5.2.2) tion is that adequate means must exist to preclude an inadvertent (RSP) overpressurization event caused by a single active component failure or single operator error. Unless B&W can show that the current design would prevent exceeding the pressure-temperature operating limitations due to an overpressure event, design modifications must be proposed to provide this assurance.- . 212.223 With regard to question ~212.~109, the response is partially (5.5.7) acceptable. Discuss the capability to bring the plant to a cold shutdown condition from full power operation using only safety grade systems. Provide a protection sequence diagram similar to the FMEA figures in the response to question 212.1. Of the 13 steps shown in Section 5.1 which are followed by the operator during a shutdown operation, state which ones are essential to achiev1ng and maintaining the plant in a cold shutdown condition from full power operation. Similarly, state each function in Table 5.1-10 which is essential to achieving and maintaining a cold shutdown. 212.224 The response to question 212.116 states that closure of an (5.5.7) RHR suction valve would not necessarily cause pump damage. Justify this contention. Some flow is indicated to be main-tained in the pump recirculation line. Where is this cooling water coming from (i.e., pump suction valve is closed) and how long would it last? 212.225 With regard to the response to question 212.127, (6.3.2f (RSP) (1)ReliefvalvesDH-RV5A,-5B: Submit a plot showing the actual pressure transient in the low pressure piping for the worst case (i.e., HPI actuation). Show the time at which the peak pressure is attained and the times at which the DH relief valves are actuated. Indicate the effect, if any, of the suction isolation valves on the peak pressure. The response indicates that the + 10 psig relief valve setpoint uncertainty was not accounted for in the relief valve sizing. The staff position is that such uncertainties must be considered in the sizing and setpoints of all over-pressure protection devices.
x.._-. a. =..w--..-.- x ~ \\ (. < ~, i (2) Relief valves CF-RVlA, -1B: A nitrogen overfill situation was not addressed. Also, pressure relief should. account for setpoint uncertainty. 212.226 With regard to question 212.141, the response is insufficient (6.3.2) to allow an adequate evaluation. Provide the safety design basis of these recirculation lines. 212.227 With regard to the response to question 212.147, the low seal (6.3.2) injection flow signal should be included in Table 7.3-2 as an ESFAS trip parameter. Also, the staff position with regard to pipe breaks is that credit for operator action is not allowed until 30 minutes after the first alarm. In addition, operator action outside the control room to mitigate the consequences of pipe breaks is generally not permitted. It appears from the evaluation that, in addition to detecting and diagnosing the event, the operator must; (2)) 1 Isolate the break (from the control room). ( Place the spare makeup /HPI pump in operation (outside the , control room. (3) Isolate letdown (from the contral room). While the desirability.of isolating letdown is obvious, it is not clear that this action is essential. Please comment. Also, ~ discuss the feasibility of adopting the capability of placing the spare pump in operation from the control room. Finally, reassess the plants' response to this event,not taking credit for operator action before 30 minutes. 212.228 With regard to question 212.1, the response is partially (15.1) acceptable. Provide the following additional information: {l} Figure 15C.13-2: " Reactor Trip" should be replaced by "CRDM."' \\21 Figure 15C.13-3: "CRDM Trip" should be shown as "CRDM." (3) Provide a protection sequence diagram for the event described in question 212.147 since it is important to i depict the different protection sequences required for makeup line breaks. (4) Provide a protection sequence diagram for a CFT line break since it is important to depict the different protection sequences required for this type of LOCA. (5) Submit the protection sequence diagram requested in question 212.153 for HPI line breaks. (6) ' Figure 15C.14-1: The need for Secondary Steam Dump valves disagrees with Table 15.1-4 and should be corrected. 4
f t i i s (7) Provide a protection sequence diagram for the event analyzed in question 212.169. l (8) All protection sequence diagrams should again be reviewed by B&W for consistency with Chapter 15 analyses. For example, Figure 15C.4 is met' complete since this event during re-fueling (Mode 4) is analyzed in Chapter 15 and should be reflected by a different protection sequence than shown in this submitted diagram. (9) All events must be reviewed by B&W for applicability to the four operating modes in Table 15C-2. Provide a summary table eNewing (for each mode) the applicability of the event. If an event is not applicable for any given mode, a brief explanation should be provided. It would be expected that each event could occur in all operating modes, with few exceptions. 212.229 The response to question 212.169 with regard to the worst-case (15.1) overpressure transient is insufficient to allow an adequate evaluation. Show that the stated time available for operator action (before exceeding the pressure-temperature limits) would also be available at cooler reactor coolant temperatures. Provide a discussion with regard to a normal startup (or coeldown) and the expected pressure-temperature combinations; special emphasis should be given to the typical high pressure-low temperature conditions normally observed during a startup or cooldown. While it is recognized that one combination of initial pressure and temperature could produce a faster and higher pressure rise, it would appear that a slower pressure rise could be worst-case at the cooler temperatures (in terms of time available before the pressure-temperature limits are exceeded). On a P-T plot, show the initial and final conditions of each P-T combination considered. Also, provide the makeup flow rate assumed with a justification for its value. 212.230 The response t6 question 212.172 did not answer whether any (15.1) Chapter 15 event would be more severe for the 3600 MWt design than for the 3800 MWt design. Therefore, unless it can be demonstrated that the most limiting transients and accidents for the 3800 MWt design would bound the 3600 MWe design, a reanalyses of Chapter 15 will be required to support _the_ _ granting of a Preliminary Design Approval for 3600 Mwt. 212.231 witn regarc to question 212.192, the response is insufficient to (15.1) a':ow an adequate evaluation. Confirm that the most limiting (with regard to pressure) of the events that result in a decrease in heat r'emo ral by the secondary sy~sted are' the' loss of ~ ~ feedwater ano tne turbine ~ trip'. State the most li=iting event with regard to core thermal margins. Provide or reference the ar..alysis for eacn worst case. s. = e \\ g
r 1..._~_._.. ~ .a 4 t s 212.232 with regard to the response to questions 212.178 through 212.182' (15.1.4) pertaining to the CVCS dilution event, the staff position is that alarms shall be available during refueling from audible count rate instrumentation to detect changes in the reactivity condition of the core. The analysis of this dilution event would dictate that the operator have a prompt and definite indication of any boron dilution from the audible count rate instrumentation; therefore, for initial core loading it is prudent to require the more sensitive i channels temporarily installed to provide audible indication. It is not obvious that sufficient sensitivity of the excore source range instrumentation exists during initial core loading to promptly alert the operator of a continuing boron dilution. \\ State whether the operator would have more time from the first alarm or less time if the dilution analysis during refueling assumed a higher initial boron concentration. State the concen-tration values assumed during refueling and the time predicted i available af ter the first alarm until the system goes critical. 212.233 With regard to the main steam line break, the response to several (15.1.14) questions asked during and subsequent to the Acceptance Review show that a systems-level analyses had not been performed well enough to identify the worst-case steam line break (with respect to reactivity margin and DNBR). Lack of clarity continues to exist in the Section 15.1.14 discussion and our position ~~ is that the main steam line break analysis must be rewritten in BSAR-or resubmitted in the form of a topical report. This new submittal must include the following points: (1) A summary table which lists each break location considered (Inside vs outside containment, 28" line vs 42" line, with offsite power vs without offsite power, etc). For each break the table should list the resulting reactivity margin, peak cladding temperature, and selected worst single active component failure. The table should clearly show which break location is worst-case in terms of reactivity margin and which break location is worst-case in terms of peak cladding temperature. (2) For each case represented in the above table, the text should present the analysis with a discussion and justifica-tion of why the selected single active component failure was considered worst-case for that break location. (3) For the worst-case break with regard to reactivity margin and peak cladding temperature, provide the time histories of reactivity margin,~ peak cladding temperature, break flow rate, DNBR, pressurizer' level, st'eam generator levels, and steam generator pressures. I I
t 212.234 The description of a steam pressure regulator malfunction (Section (15.1) 15.1.36) and the inadvertent opening of a steam generator relief or safety valve (Section 15.1.37) indicated that the consequences of these events are bounded by the main steam line break. S&U's criterion for Chapter 15 events to ensure that no fuel damage occurs is that a DNBR greater than 1.30 must be maintained throughout the transient. Since the steam line break analysis shows a DNBR less than 1.30, a reference to this analysis as a bounding calculation is not appropriate. Provide the specific analyses for the cooldown transients to show that DNBR remains greater than 1.30 for each event. 212.235 The response to question 212.71 indicates that the rod worth (15.1) curve shown in the PSAR (figure 15.1-1) was non-conservative. B&W states that they will modify their scram time specs to compensate for this non-conservatism. Provide sensitivity studies to show that the limiting events in Chapter 15.0 would not become more severe with these new assumptions. Also, the original information requested regarding the conservatism of the combination of axial power shpae, rod worth and assumed trip delay times was not provided. In addition, for events that do not involve reactor trip, the variances in axial power shapes must be accounted for - ~ - ~ ~ in addition to consideration of maximum design peaking factors. Show that the assumed cosine shape was conservative or resubmit analyses with the appropriate power shpae of each event which does not involve reactor trip. s { / e, Y+ /kA 1 1 1 e
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..l..--.~.-.-....-. ~ I 3.0 ANALYSIS 222.15 This request relates to the double ended hot leg rupture adjacent f to the steam generator inlet nozzle which you have analyzed in Appendix 6A. Your analysis assumes that during the reflooding period 80% of the fluid which enters the core is ejected from the break. We believe that the carry out fraction may be less than 80% and that separation of the entrained liquid may occur in the upper plenum and the section of hot leg piping leading to the break. If less liquid is ejected from the break the time to reflood will be less and the time when ECCS water first enters the steam generators will be earlier. Perform an analysis showing the effect on containment pressure of zero liquid carryout through the core during the reflooding period after a hot leg break. Provide mass and energy release data for this case. 222.16 Tables 6A.5-5, 6A.5-7 and 6A5-9 appear to be inconsistent. These tables provide energy ba'ance data for the three break locations. Correct or justify these apparent inconsistencies. (1) The core heat generation should be zero BTU's at time zero instead of the finite values indicated in the table. (2) The initial steam generator energy is different for each table. The final equilibrium values also differ in the three tables. (3) The times given for E0B (end of reflood) do not correspond to the values on page 6A-4. ~ w w -- em
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q " 2.17 We understand that for ultimate heat sM calculations, the i energy release rate from the reactor core w.1 be calculated using Branch Technical Position APCSB 9-2. We believe that the long tem continuous mass and energy release calculations for containment design should be perfomed using the same method. Please revise Appendix 6A accordingly. A P m h r -4+ew e.
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