ML20039E133
| ML20039E133 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 05/21/1976 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Suhrke K BABCOCK & WILCOX CO. |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201060567 | |
| Download: ML20039E133 (89) | |
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R 11artin t1AY 211976 Docket File W Gammill, DSE flRC/PDR W licDonald,111PC.
L/PDR 0 ELD LWR-1 File IE (3)
Docket ilo:
STil 50-561 R DeYoung R I4accary, SS F Williams R Tedesco, SS Babcock I, Wilcox Company J Stolz D Ross, SS ATTri: fir. Kenneth E. Suhrke T Cox ACRS 14anager, Licensing H Smith (2)
T ilovak tiuclear Power Generation R Heineman, SS P. O. Box 1260 H.Denton, DSE bcc:
Lynchburg, Virginia 24505 V Moore, DSE J Buchanan, NSIC R Vollmer, DSE T Abernathy, TIC Gentlemen:
As a result of our continuing review of your Standard Safety Analysis Report BSAR-205, we find that we need additional infomation to complete our evaluation.
The specific infomation required is listed in the enclosure, and pertains primarily to Chapter 15 of the BSAR-205. With regard to our review of the Chapter 15 material, several items are worthy of special note:
1.
The complete response to Acceptance Review Question 212.1 pertaining to the Chapter 15 failure modes and effects analysis has not yet been received.
We feel that a conference should be held as soon as pos-sible to resolve issures still pending concerning the format, methods, teminology and goals of the remaining analyses, and to set a schedule for the submittal of the remaining analyses.
2.
He have reviewed BSAR-205 on the basis of nomal operation with four pumps only.
It is our intent to review the safety implications.of partial loop operation during the review of your application for a I
Final Design Approval of BSAR-205.
3.
The initial conditions utilized in your analyses of Chapter 15 events must be reflected in the utility user's Technical Specifications as Limiting Conditions of Operation, and should be specified in BSAR-205 i
as interface information. For example, mwould expect from the 6.55%Ak/k initial subcriticality margin used in the boron dilution event during refueling (Section 15.1.4), that the utility-user would i
adopt the 6.55%Ak/k as the minimum margin to be allowed in Technical i
Specifications. This aspect of our review will be conducted during review of your application for a Final Design Approval for BSAR-205.
4.
Your analysis code " TRAP" as used to analyze steam and feedwater line breaks, has not been adequately described to the staff in the BSAR-205 document and we understand that a topical report is planned.
I urge that your complete description of this method be made available to the staff as soon as possible.
i pa ws b-un Aic.ns inn.
- m ArcM 8201060 567 810 40$ ' """"""""'" *"' '"""""
MADDEN 80-515 PDR r.
'1 Babcock & llilcox TAM EI 5
We are continuing to review the interface criteria provided by 83W.
Since the utility applicant will be ultimately responsible for insti-tuting all safety-related design requirements, we will review applicant's compliance with B&W interface specifications at the appropriate time.
A completely adequate response to Item 1 above, constitutes important interface information in that the protection sequence diagrams submitted by B&W util identify all BOP equipment and systems that are essential to mitigating the consequences of each event in Chapter 15 of BSAR-205.
To maintain our licensing review schedule for your SSAR, we need your responses to the enclosure items by June 28, 1976.
If you cannot meet this date, please inform us within seven days after receipt of this letter of the date you plan to meet so that we may revise our schedules accordingly.
Picase contact us if you have any questions about the infomation requested.
Sincerely, Original Signed by John F. Stolz John F. Stolz, Chief Light Water Reactors Branch flo.1 Division of Project Management
Enclosure:
Request for Additional Information Washington Public Powar Supply System cc:
ATTH: Mr. J. J. Stein Mr. A. H. Monteith llanaging Director Ohio Edison Company P. O. Box 968 47 North Main Street Akron, Ohio 44308 3000 George Washington Way Richland, Washington 99352
!!r. W. E. Kessler Mr. Robert Dorsum Comonwealth Associates. Inc.
Bethesda Representative 209 East Washington Babcock & Wilcox Jackson, Michigan 49201 Huclear Power Generation Division Sulte 5515, 7735 Old Georgetown Road Bethesda, Maryland 20014 Mr. B. G. Shultz Project Engineer Stone & Webster Engineering Corporation P. O. Box 2325 Boston, Massach9setts 02107 LWR-1 LWR-1 JCoklUh/h 5b
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212-1
- 212.0 REACTOR SYSTEMS 212.163 RAW-10099 is referenced on page 15.1-2 as providing discussions of (15.1)
ATWS events. The events in this report should be resubmitted at the B-SAR-205' power level.
Also, the staff " Status Report on Anticipated Transients Without Scram for Babcock & Wilcox Reactors,"
dated December 9,1975, specifies the additional analyses and design changes needed 'to meet the safety objective of WASH-1270.
B&W was requested, in a letter to K. Suhrke dated Apr11 7, 1976, to i
provide additional analyses specified in the Status Report and an identification of the proposed design changes based on the analyses perforned.
212.164 Table 15.1-6: The version of CRAFT referenced in this table for (15.1) the LOCA is not the same utilized in BAW-10102. Please explain this discrepancy.
212.165 The assumed initial DNBR values on the following events were:
(15.1)
(1) Event 15.1.2........... 1.71 (2) Event 15.1.5........... 1.73 i
(3) Event 15.1.8........... 1.76 l
(4)
Event 15.1.10.......... 1.75 (5) Event 15.1.14:......... 1.72 (6) Event 15.1.17.......... 1.84 Please explain why the initial assumed value varied and specify.
the correct value.
Should changes in assumed values be appropriate, specify the impact upon minimum DNBR.
212.166 Identify the equipment in Table 15.1-4 which provide a Safety Action (15.1)
(CORE COOLING, REACTOR TRIP, etc.) not meeting both of. the following criteria:
(1)
Each Safety Action shall not be vulnerable to a single active component f ailure.
(2) As equipment important to safety, each system or component provided to meet the redundancy requirement of Item 1 above shall also conform to the requirements of General Design Criteria 1 'through 5.
212.167 Deleted (15.1) 212.168 An acceptance criterion which is not presented in Chapter 15.0 is (15.1) the pressure-temperature NDT curves provided in technical specifications.
Branch Technical Position HIEB No. 5-2 of Standard Review Plan 5.3.2 states that these curves are applicable during upset conditions. Along these lines, provide a cocplete assessment of anticipated operational occurrences (see definition in Appendix A to 10 CFR 50) which could occur during a startup or shutdown of the reactor and discuss for each one the ability to r.aintain the pressure-temperature limits.
i 212-2 212.169 Submit an analysis of the worst-case overpressure transient during (15.1) a startup or shutdown. Provide all assumptions. Plots should
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include pressure versus time, reactor coolant temperature versus time, and safety valve flows versus time.
Show the times of all trips, actuations, or operator actions on the plots.
Rationalize the time of occurrence of this event as being worst-case (with N2 blanket, with steam bubble, etc.).
Show that the, pressure.
temperature limits in technical specifications are not exceeded.
j 212.170 Indicate which events in Chapter 15.0 would produce more severe (15.1) consequences for higher initial core flows (e.g.,110% versus 100%).
f Indicate if. any events would be worst for lower initial core inlet temperature, or higher _ initial RCS pressure.
212.171 Submit a schedule for providing the failure modes and effects (15.1) analyses requested in question 212.1 (2/1/76). It would be preferable to receive each event as it is completed, with the steam line break as the next submittal.
212.172 Confirm that no Chapter.15.0 event would be more severe for the (15.1) 3600 MWt plant than the 3800 MWt design.
Verify that the Station Technical Specifications which are based on Chapter 15.0 events vill not change between the two plant sizes.
212.173 Table 15.1-5 provides a list o'f support systems necessary for the (15.1) operation of each primary safety system.
Although it is recognized that the pressurizer relief system performs a role similar to th'a RC system safety valves, it is the staff's understanding that the relief valves should not be interpreted as a required support system in order for the safety valves to perform their safety function.
If so, the pressurizer relief system should be deleted from the j
table as an essential support system.
If not, B&W should upgrade these components to the same qualifications as the safety valves.
212.174 Section 15.1 identifies a reactor coolant pump seizure as a " fault (15.1) of moderate frequency" and the inadvertent operation of ECCS as an
" infrequent incident." This would appear to be incongruous since operating experience has shown that inadvertent ECCS actuations have occurred in the reactor industry.
In addition, page 15.1.5-1 states that the frequency of occurrence of the pump seizure accident.
is expected to be the same as a LOCA (gross mechanical failure of primary system). Rationalize the apparently inconsistent classifi-cation of these events.
212.175
. Page 15.1-3 states, " Chapter 15 analysis of non-limiting or non-(15.1) design basis..ents may take credit for non-safety grade equipment because if this were not done, the conditions would lead to a limiting or design basis event which is already considered." With regard to this statement, provide the following additional information:
(1)
For each event which takes credit for non-safety grade equip-ment, justify the statement that the consequences would otherwise be limited to a design basis event...
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212-3 212.175 (2) For each event in Chapter 15 which takes credit for non-(15.1) safety grade equipment to mitigate the consequences of the (contd.)
event, provide a list of the essential non-safety grade systems or components.
(3)
Rationalize the above statement in light of the published staff position in Regulatory Guide 1.29 dated August 1973:
"Those portions of structures, systems, or components whose continued function is not required but whose failure could reduce the functioning of any plant feature included in items 1.a. through 1.r. above to an unacceptable safety level should be designed and constructed so that the SSE would not cause such failure."
212.176 The response to question 212.7 (2/1/76) is not adequata to allow (15.1) a sufficient evaluation.
It is clear that for the analysis of each event in Chapter 15.0, maximum instrument uncertainties and allowable operating ranges specified in the Technical Specifications must be accounted for by conservative initial assumptions of such parameters as RCS flow, RCS pressure', power level, RCS-inlet temperature, stcaa generator inventory, ppm boron, and various tank inventories.
It is less clear that B&W has properly accounted for these matters in their analyses in Chapter 15. Discuss how B&W intends to relate the initial conditions in Chapter 15.0 to the limiting conditions of operation in typical Technical Specifications.
212.177 Table 15.1-4:
Credit for several types of secondary side-relieving (15.1) devices is sought by B&W to mitigate the consequences of certain events.
The concern is that such components important to safety may not be proper safety grade. The terminology used in Section 10.5 to discuss available relieving devices is not clear. Identify on Figure 10.1-1 by number the following components:
(1) Spring-Loaded Safety Valves (2)
Secondary Safety Valves (3) Atmospheric Dump Valves (Modulating)
(4) Atmospheric Dump Valves (No6-Modulating)
(5)
Condenser Dump Valves (Modulating)
(6) Condenser Dump Valves (Non-Modulating) l (7) Turbine Bypass System Valves (8)
Safety Grade Steam Dump Valves (Table 15.1-4)
212-4 212.177 State c1carly which of the above valves are taken credit for in (15.1) mitigating the consequences of the events in Chapter 15.0.
Confirm (contd.)
that the appropriate application of safety grade quality of these safety components will be a part of the secondary side interface criteria to the applicant.
212.178 The CVCS dilution event during refueling was initiated from an (15.1.4) initial suberiticality margin of 6.55% Ak/k.
Confirm that this value vill be the minimum allowed suberiticality margin allowed by Technical Specifications during refueling.
212.1'9 The response to Question 212.78 (4/1/76) indicates that a nominal (15.1.4) reactor coolant volume was assumed.
This is not appropriate and should be changed to reflect minimum volumes.
List the alarms which would be expected to alert the operator (consider each operating state).
212.180 The minimum shutdown margin exists at EOL (1% Ak/k with most (15.1.4) reactive rod held out of core). Comment in more detail on the.
effect of a dilution event at cols time when the boron concentration is at its udnimum.
I 212.181 Page 15.1.4-2 states that control rod insertion will cause the feed (15.1.4) block valve to close, yet the top of this same page specifies the failure of the feed block valv.es as an input assumption. Explain this inconsistency and discuss the sequence of events if automatic termination of the deborated water does not occur.
212.182 The bottom of page 15.1.4-1 indicates that with both makeup pumps (15.1.4) operating, a makeup rate much higher than normal could result.
Why w'ould this be limited to 700 gpm (design flow rate of only one HPI pump)?
212.183 Page 15.1.5-1 states that the frequency of occurrence of a single (15.1.5) reactor coolant pump locked rotor is expected to be the same as any gross mechanical failure of the pri=ary system.
The concern is that the potential for this event could be high in more than one reactor coolant pump during a scismic event.
Table 3.2-4 identifies the pressure-retaining portion and the flywheel as seismic Category I.
Examine the non-Category I components (pump motor, impeller, shaf t, etc.) and discuss the potential for seizure of pump / motor components during a seismic event. Alternatively, demonstrate that ~ the consequences of simultaneous pump or motor com-ponent seizures in several pumps (as a result of an earthquake) are acceptable.
212.184 Table 15.1.5-3 indicates that the initial pressure and inlet (15.1.5) temperature are assumed to be nominal values for the locked rotor event analysis.
This is not acceptable and is inconsistent with
. Table 15.1.5-1.
Table 15.1.5-3 also indicates that nominal design conditions were assumed for thermal conditions.
This is also not acceptable.
Reanalyze the locked rotor event using offset conditions.
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212-5 0
212.185 Table 15.1.5-1 indicates that a value of +3 F is added to inlet (15.1.5) temperature to account for control band and instrument errors.
0 Page 15.1-4a indicates that +2 F is a good n mber. Justify why +4 F is not a more appropriate value.
Table 15.1.5-1 also indicates that a value of -45 psi is imposed r
upon the initial system pressure to account for control band and instrument errors.
Expand on what i=provc=ent in design is being proposed which allows a greater confidence in this value compared to Davis-Besse Unit No.1 (-65 psi).
Explain how this added confidence was translated into +20 psi.
212.186 Page 15.1.5-3 states that the natural circulation characteristics (15.1.5) of the RC system have been calculated with conservative values for all resistance and form loss factors.
Confirm that these values were at least as conservative as those shown in Table 4-3 of BAW-10102. Justify the appropriateness of these values. Will startup testing confirm these values?
212.187 Submit a plot of RCS pressure versus time. The maxitu a allowable (15.1.5)
RCS pressure limit should also be an acceptance crittrion for this event.
212.188 An incident of moderate frequency in combination with any single (15.1.5) active component failure, or single operator error, should not cause loss of function of any barrier other than 'the fuel cladding.
A limited number of fuel rod cladding perforations is acceptable.
Submit such an analysis which considers the worst-case single active component failure with each loss of flow event (4-pump trip and locked rotor) and show that any fuel damage calculated to occur is of sufficiently limited extent that the core geometry -will' remain intact with no loss of cooling capability.
212.189 Indicate, or reference, the commitment to confirm the assumed (15.1.5) reactor coolant pump flow coastdown characteristics during pre-operational testing.
212.190 The response to question 212.80 (4/1/76) clearly shows that a (15.1.6) reactor coolant pump startup time of 15 seconds is not an appropriate assumption to utilize in a worst-case Chapter 15.0 cvent; in fact, the reported Oconee Unit No. I tests show that 10 seconds is equally as realistic as 15 seconds. None of these observed values are necessarily conservative.
Resubmit the analysis using an appropriately conservative startup time or show that the consequences of the event would not be sensitive to this parameter.
Also, the response to part 5 of the question has not yet been received.
212.191 With regard to the loss of load event, page 15.1.7-3 states that (15.1.7) decay heat can be removed by the steam generators with feedwater flow supplied by the main or auxiliary feedwater pumps. When would the actuation signal for auxiliary feedwater occur and what paraneter would initiate its operation (in each case discussed)?
212-6 212.191 Confirm that Figure 15.1.7-1 reflects the time histories for the (15.1.7) most severe of the three cases presented.
If not, provide similar (contd.)
plots for the worst case.
212.192 With regard to all events expected to occur with moderate frequency (15.1. 7) which result in a decrease in heat re= oval by the secondary system (loss of load, turbine trip, loss of condenser vacuum, etc.),
identify the most limiting with regard to core thermal margins and pressure within the reactor coolant and main steam system.
Justify this contention and provide, or reference, the analysis of the worst case.
212.193 In the sequence of events for the feedwater line break on page 15.1.8-3, (15.1.8) identify which of the eight steps would require operator action.
Discuss this action with regard to time available, physical movements needed, for how long must action be performed, etc.
Discuss the necessity for these or other manual actions across the bret.
spectrum through long-term decay heat removal.
Address the same question with regard to the seven steps on page 15.1.8-4.
212.194 Page 15.1.8-2 indicates that a complete loss of all feedwater (15.1.8) would be more severe than a feedwater line rupture upstream of the first check valve (with a worst-case single active component failure and loss of offsite power). Please justify this i
contention.
212.195 The presentation of worst-case situations in this section is not (15.1.8) clear.
Confirm that, for the primary side pressure limit, the worst-case was the complete loss of all feedwater with offsite power available. What ESFAS parameter supplies the closure signals for the main steam line isolation valves or turbine stop valves and identify the time on Figure 15.1.8-2 at which this signal would occur. Provide a plot of this ESEAS parameter versus time.
212.196 Page 15.1.8-2 states that the worst single active component (15.1.8) failure with regard to core heatup and containment pressure is the malfunction of the turbine stop valves.
Confirm that this also refers to dhe RCS pressure limit.
212.197 For a feedwater line rupture with a loss of offsite power, (15.1. 8) page 15.1.8-3 indicates that a loss of offsite power at the time of reactor trip produced a higher RCS pressure than a loss of offsite power at rupture.
Confirm that the indicated higher RCS Pressure is less than that predicted for the complete loss of all feedwater event.
212.198 The response to question 212.9 (4/1/76) indicates that auxiliary (15.1.8) feedwater flow is initiated by low staam line pressure at 58 seconds.
Table 15.1.8-2 shows initiation was assumed at 40 seconds.
Explain this inconsistency.
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212-7 1
212.199 Page 15.1.9-2 states that assumed initial conditions are given in (15.1.9)
Table 15.1-2.
This table should also provide all secondary side assumptions.
212.200 Consider the effects of a worst-case single active component (15.1.9) failure or operator error upon the loss of offsite power event.
Are the stated criteria met?
i i
212.201 With regard to the feedwater temperature decrease at 3876 MWe, (15.1.10) provide a plot of thermal power versus time. Submit the specific value of ninicum DN3R.
With regard to the feedwater flow increase at no load conditions, provide a plot of DNBR versus time or confirm that DNBR does not decrease below its initial value.
Of all the moderate frequency excessive heat removal events (feed-water malfunctions, opening of steam safety valves,. opening of tur-bine bypass valves, etc.), which event produces the worst cooldown consequences? Justify this contention and provide, or reference, a complete analysis of this worst-case event of moderate frequency.
212.202 With regard to External Events, provide, or reference, interface (15.1.12) criteria which would ensure a reactor design capable of achieving and maintaining a safe shutdown condition after an earthquake.
An ensuing accident resulting from the failure of a non-Category I component must be considered in these interface requirements.
212.203 It appears from the response to question 212.75 (2/1/76) Obat the (15.1.14) worst-case main steam line break was improperly represented in the initial submittal.
The detailed discussion on pages 15.1.14-6a anB and accompanying figures must therefore be supplemented with the same information for the worst case.
212.204 Is Attachment C in Amendment 1 dated 3/29/76 incorrectly labeled (15.1.14)
" Question 212.75"?
212.205 Explain on page 15.1.14-1 why a fouled steam generator is a (15.1.14) conservative assumption (since it would appear that greater cool-down rates would take place for an unfouled steam generator). Sub-mit the unfouled inventory. Provide a discussion of the amount of possible SG inventory fluctuations during power operation and the sensitivity of the analysis to this inventory.
212.206 In the analysis of a steam line break, page 15.1.14-3 indicates (15.1.14.2)that a single failure sensitivity study was conducted to identify the worst active component failure. The text discussion of this study is not clear. Provide a s a ry table with the following information:
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212-8 212.206 (1) Active component failed in each analysis.
(15.1.14.1)
(contd,)
(2) Resulting reactivity margin in each analysis.
Specify the break location assumed for this study.
212.207 Page 15.1.14-4 alludes to both turbine stop valves closing and (15.1.14.2) main steam isolation valves closing (top of page). Which assumption was made for minor secondary pipe break situations?
212.208 The actuation of the atmospheric dump valve is not clear. Show (15.1.14.2)the assumed opening and closing of this valve on Figure 15.1.14-1 (Unaffected Steam Generator Pressure).
What is the assumed setpoint?
212.209 Discuss the procedures for long-term core decay heat removal (15.1.14) across the spectrum of steam line breaks.
212.210 For a seismic event which causes the rupture of a non-Category I (15.1.14.2) portion of a steam line, would any components or systems be relied upon to bring the plant to a safe shutdown condition which are not qualified to function through the earthquake?
For the postulated event on page 15.1.14-5, why wasn't the 42-inch line chosen instead of the amaller 28-inch line? It would appear possible that since the maximum cooling rate occurs within the first 10 seconds, the largest mass discharge would occur through a 42-inch line versus the 28-inch line, thereby resulting in a greater cooldown.
212.211 Page 15.1.14-6a states that the minimum suberitical margin for (15.1.14) Case IV will be the same as that for Case II if a single failure of the main steam isolation valve on the affected steam generator is assumed.
Is this still true with the 42-inch line rupture?
If not, please correct.
What is the value of " threshold temperature" referred to on the bottom of this page?
212.212 With regard to the letdown line failure outside containment, consider (15.1.19) the occurrence of a single active component failure or single operator error on the consequences of this event; for example, could a single failure or operator error extend the. time or prevent the RCS pressure from decreasing to the 1600 psig low pressure isolation setpoint? Could a smaller break in this 3-inch line sufficiently extend the time to reach the setpoint such that higher total mass release could occur?
This brgak does not seem to behave in the same manner as the 0.05 ft cold leg break in BAW-10074 (i.e., much less mass released out of the break in the letdown line). Please explain.
212-9 212.213 The staff considers an inadvertent opening of a safety or relief (15.1.33) valve to be an event of moderate frequency.
It must be shown that no fuel damage results and that pressure in the RCS and main steam system is maintained below 110% of the design pressures.
Confirm that these criteria are met.
Justify further the appropriateness of the assumed discharge coefficient (0.75).
212.214 Page 15.1.34-1 states that the margin of core protection (15.1.34) indicated by the DNBR during the reactor coolant pump coastdown is greater for the pump shaft breakage than for the shaft seizure since the flow decrease is not as rapid. Wouldn't the shaf t break event permit a greater reverse flow through the affected loop later during the transient, thereby resulting in a lower core flow
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at that time? Please discuss.
Also, discuss the expected frequency of occurrence of a pump shaf t breakage relative to a rotor seizure.
212.215 With regard to the inadvertent closure of the main steam line (15.1.35) isolation valves, it is not obvious that the secondary side pressure surge would be bounded by the loss-ef-normal feedwater event.
Please justify.
i 212.216 With regard to the ECCS passive failure analysis, provide the (6.3.2) maximum expected leak rate from the sudden failure of a pump shaft seal.
l Provide sufficient detailed information to enable the staff to review independently the calculated leak rate.
Discuss the ability of the operator to detect and isolate such a failure.
212.217 The response to question 212.61 (4/1/76) indicates that a design (6.3.2.10) change has been made in the operating mechanisms of the reactor internals vent valves.
Such a modification clearly relates to the proper status of these components during normal operation and transient events.
Therefore, this change raises a question on the appropriateness of taking credit for the vent valve operating experience in Oconee Class plants. Justify why the previous flow penalty for B&W plants should not be imposed on the basis of a lack of operating experience with the proposed vent valve operating mechanism.
In addition, submit or reference the interface requirements which are to be placed upon vent valve testing on site, vent valve inspections, and continuous surveillance of vent valve performance anomalies utilizing such means as loose parts monitoring.
l 4
212-10 212.218 Based on the credit taken for systens in Chapter 15.0, Table 3.9-5 (3.9.2.4) should also reflect all BOP components and should be considered as typical interface requirements.
Why were all isolation valves not considered ACTIVE, since containment isolation is a required Safety Action? Provide the component numbers in Table 3.9-5 to facilitate their identification on each P&ID. Why are the core flooding tank venting isolation valves not considered? Why are the RC pumps not included (credit needed for safety as shown in Table 15.1-4)?
212.219 With regard to the interface criteria for sizing the reactor (5.5.11) coolant drain tank, what is the worst-case transient referred to on page 5.5-137 Confirm that this transient would be accommodated without failing the rupture disc.
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May 14, 1976 ENCLOSURE Request for Additional Infonnation First-Round BSAR-205 Docket No. STN 50-561 O
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Numbering Format for' Recuests for Additional Information Assistant Branch No.
Directorship Branch Review Area 4
i 010 Plant Systems Aux. & Power Con-All Areas version Systems
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011 Plant Systems a
Section A' 012 Plant Systems Section B 020 Plant Systems Containment Systems All Areas 02)
Plant Systems Containment Systems Section A 022 Plant ~ Systems Containment Systems Section B 030 Plant Systems Elec. Inst. & Con-All Areas trol Systems 031 Plant Systems Section A 032 Plant Systems Section B 033 Plant Systems Section C 110 Engineering Mechani. cal Engr.
All Areas 111 Engineering Section A 112 Engineering Section B 120 Engineering Materials Engr.
All Areas 121 Engineering Material Integrity 122 Engineering Metallurgy 130 Engineering Structural Eng.
All Areas 131 Engineering Section A 132 Engineering Section B 210 Reactor Safety Reactor Systems All Areas 211 Reactor Safety Section A 212 Reactor Safety Section B 220 Reactor Safety Analysis All Areas 221 Reactor Safety Reactor Analysis 222 Reactor Safety Systems Analysis 230 Reactor Safety Core Performance All Areas 231 Reactor Safety Fuels 232 Reactor Safety Physics 300 Environmental Responsible EP Br.
' Miscellaneous Items Projects 301 Environmental Assigned Lab.
8.11 Areas Projects 310 Site Analysis ~
Accident Analysis All Areas 311 Site Analysis Section A 312 Site Analysis Section B l.
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Assistant Branch No.
Directorship Branch Review Area 320 Site Analysis Effluent Treatment All Areas Systems 321 Site Analysis Applications 322 Site Analysis Systems Analysis 330 Site Analysis Radiological Assessment All Areas 331 Site Analysis Radiation Protection 332 Site Analysis Radiologi al Impact 340 Env. Tech.
Env. Specialist All Areas 350 Env. Tech.
Cost-Benefit Analysis All Areas 360 Site Tech.
Seismology-Geology All Areas 361 Site Tech.
Seismology-Geology Geology 362 Site Tech.
Seismology-Geology Foundation Eng.
370 Site Tech.
Hydrology-Meteorology All Areas 371 Site Tech.
Hydrologic Eng.-
372 Site Tech.
Meteorology 400 Proj. Management Responsible PM Branch Miscellaneous Items 410 Quality.Assur.
Fin. Fesiew Group Financial
& Operations 420 Quality Assur.
Quality Assurance All Areas
& Operations 421 Quality Assur.
Quality Assurance
& Operations 422 Quality Assur.
Conduct of Operations
& Operations 423 Quality Assur.
Initial Test & Oprs,
& Operations 430 Quality Assur.
Ind. Sec. & Emerg.
All Areas
& Operations Planning 431 Quality Assur.
Ind. Security
& Operations 432 Quality Assur.
Emergency Planning
& Operations 440 Quality Assur.
Operator Licensing All Areas
& Operations 441 Quality Assur.
Training
& Operations 442 Quality Assur.
Procedures
& Operations
. h 3-1
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8.
INDEX SYSTEM The system to be used for sub-items within a given item will be as indicated below for item 210.17:
210.17 (j) l (2) 1
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i REQUEST FOR ADDITIONAL INFORMATION' I
t BABCOCK & WILCOX COMPANY t
i BSAR-205 h,
DOCKET NO. STN 50-561 000.0 GENERAL 000.1 Revisions to safety analysis report figures and tables are frequently submitted without explanation. A procedure is needed in B-SAR to quickly identify where and why each change is made.
For example, circling or indexing each P&ID revision (accompanied by an explanation) would serve to facilitate the identification and review of such changes.
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tj s 010.0 AUXILIARY AND POWER CONVERSION SYSTEMS I
012.15 We require modifications to and need additional interface infor-General
'mation in the BSAR-205 (PSAR) as follows:
?,1.
Section 1.7.2.1 defines several interface categories included j
in the various interface sections throughout BSAR-205. The definition for category 18 "Related Services" is not included.
Complete section 1.7.2.1 in PSAR accordingly.
2.
Section 1.7.2.2 states, " Minor connections, such as air supply connections to air operated valves.are not identified 1
by key numbers and letters." This is not in accordance with 4
our interface identification policy.
If the air supply is 4
connected to a safety related component, that interface should i
be identified on the P&ID with a unique key number. The air requirements for all safety related air-operated valves which j
are part of the NSSS design scope should be tabulated in Section 9.3.1.
The tabulation should include the P&ID number, i
system where the valve (s) is installed, valve interface iden-tification node, and air flow requirements in scf/ min. The j
information required may alternately be provided in each system interface section.
If the latter method is chosen, provide in Section 9.3.1 references to other locations in BSAR-205 where air requirements may be found for air operated 1
valves in the BSAR-205 design scope.
3.
The P& ids identify both safety related and nonsafety related interfaces; for exemple, Figure 9.3-1.
Change BSAR-205 to identify only safety related interfaces or di%guish between i
safety and nonsafety interfaces.
It is the staff position i
that, prior to the end of the PDA review, all nonsafety related interfaces should be deleted from the BSAR-205 document.
4.
An audit was made of all P&ID System Interface Tables.
The references provided under column " Interface Requirements 4
References" for each uniquely identified interface in some cases were found to provide insufficient specific criteria to enable the balance-of-plant (80P) designer to design his sup-porting systems to make the NSSS perform as intended by B&W; e.g., air requirements for safety related valves, and additional interface requirements for component cooling water.
Revise the P&ID System Interfhce Tables and System " Interface Requirements"
' sections to include design criteria or data omitted. As a minimum, B&W should specify for each safety related interface location specific criteria'; e.g., type of fluid, flow, pressure, temperature, quality, chemistry, radioactivity, environment e
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012.19 requirements, normal and limiting conditions, seismic and cont.
group classification and any other criteria B&W deems necessary for the safety related function of the NSSS System.
In the interest of maintaining present and future review i
l schedules and performing these reviews in a timely manner, we prefer to have all the necessary criteria concentrated in each NSSS system " Interface Requirements" section. An excep-tion to this would be the referencing of tables and P& ids which i
i are scattered throughout the PSAR. Another exception which is acceptable is the referencing of sections where the specific l
requirements included in that section apply to or satisfy all NSSS systems; for example, component cooling water requirements, section 9.2.2, compressed air requirements, section 9.3.1, environmental requirements, section 9.4, and fire protection requirements, section 9.5.1.
012.20 Section 3.5 (Missile Protection) does not provide any discussion (3.5) or evaluation of internally generated missiles outside containment.
Provide as interface information (1) possible missile sources from equipment within the BSAR-205 design scope outside containment, and (2) the equipment within the BSAR-205 design scope that must be protected from internally generated missiles outside containment.
Provide a tabulation for item (1) giving information similar to table 3.5-1.
For item (2) identify each system or components, reference the P&ID and indicate the component identification number included on the P&ID.
012.21 As defined in BTP APCSB 3-1 (Standard Review P identify (3.6) each and all of those systems,ide of containment, and,further or portions of systems w BSAR-205 design scope and outs identify those essential safety-related systems for which protection is required against the effects of piping failures that occur outside containment. Reference the P&ID on which the system or component is shown, the component identification number included on the P&ID, and the location of the system or component.
012.22 Provide a tabulation of all valves in the reactor coolant pressure (9.0) boundary and in other Seismic Category I Systems, as recomended in Regulatory Guide 1.29; e.g., safety valves, relief valves, stop valves, stop check valves, control valves whose operation is relied upon to assure safe plant shutdown or to mitigate the consequences of an accident. The tabulation should include the system in which the valves are installed, the P&ID number, the type and size of valves the valve identification number on the P&ID, the actuation methods and the environmental design criteria to which the valves are qualified. Where air operated valves are used and furnished with air accumulators, include. the time period the accumulator will permit valve operation in the event of air supply failure.
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012-3 012.23 Section 9.1.4 references several " optional" fuel handling items (RSP) such as auxiliary fuel handling bridge, new and spent fuel (9.1.4) storage racks, underwater closed-circuit TV, boroscope and others.
It is our position that no options should be included in the BSAR-205 NSSS application. Revise BSAR-205 to include only fuel
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handling equi design scope.pment which will be an integral part of the B&W 012.24 Provide a description and design criteria for each crane (or (9.1.4) machine) used for handling fuel, and provided by B&W as part of BSAR-205. The information should include the applicable codes and standards which will be used in the design, fabrication, installa-tion and testing of cranes, bridges, trolleys, hoists, cables (ropes) lifting hooks, special handling fixtures and slings. Also include a description of safety features; e.g., electrical /
i mechanical interlocks, controls and redundant features of the above devices.
012.25 Your response to our acceptance review item 020,5 was not complete.
(9.1.4)
Provide additional description in the PSAR relative to center to center spacing of fuel assemblies, materials and type of construc-tion, seismic qualification, and maximum uplift forces and racks can withstand.
012.26 In Section 9.1.4.2.2, under Item 13 of " Refueling Equipment,"
(9.1.4) provide additional description of the transfer canal storage rack materials and type of construction, center to center spacing of fuel assemblies, seismic qualification requirements, arl Mc-torial drawing. Provide interface criteria the B0P needs u,.6311 this rack in the structure.
012.27 In Section 9.1.4.2.2, Item 15, "New Fuel Elevator," provide addi-(9.1.4) tional description of the new fuel handling tool and new fuel elevator to demonstrate that no damage to spent fuel will occur during an SSE. The information should include materials and type of construction, seismic qualification requirements and pictorial drawings. Provide the interface criteria the BOP designer needs to install the new fuel elevator in the spent fuel pool structure.
012.28 In Section 9.1.4.2.2, Item 19. " Failed Fuel Containers," provide (9.1.4) additional description of the failed fuel containers design, materials and type of construction, and seismic qualification requirements.
Provide a pictorial drawing.
012.29 Expand Section 9.1.5, " Interface Requirements." to provide (9.1.5) interface criteria for the fuel transfer carriage and spent fuel handling bridge required by the B0P designer to permit installa-tion of this equipment in the spent fuel pool.
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.' -d 012.30 In Section 9.1.5, " Interface Requirements," Category 8. " Component (9.1.5)
Cooling and Heating," item 1, provide interface requirements the BOP designer needs to design the spent fuel pool cooling system, as follows:
1.
Decay heat load from a refueling batch (1/3 core) and within 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after reactor shutdown.
2.
Decay heat load from an emergency unload of the reactor (1 core) 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown plus the decay heat load of i
1/3 core placed in the pool 30 days prior to the emergency core unload. Assume that the 1/3 core was placed in the pool 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after reactor shutdown. This heat load should be 1
the maximum heat load attainable which must be considered by the BOP designer in the design of the spent fuel pool cooling i
system.
In calculating the information requested under items 1 and 2 above, use the method set forth in Branch Technical Position APCSB 9-2,
" Residual Decay Energy for Light Water Reactors for Long Term Cooling," which is a part of the staff's Standard Review Plan 9.2.5 (Ultimate Heat Sink).
012.31 In Section 9.1.5, " Interface Requirements," Category 15. " System /
(9.1.5)
Component Arrangement," item 3, you have specified a minimum storage capacity for new fuel.
Provide a similar interface require-ment for minimum spent fuel storage capacity. Also, under this.
section provide an interface requirement that specifies a minimum water depth above a fuel rod assembly lying horizontally across the top of the spent fuel array that will result in doses above the water level that are within 10 CFR Part 20 limits.
012.32 In Section 9.1.5, " Interface Requirements," Category 18, "Related (9.1.5)
Services," provide the following additional information:
Item 2 states "The maximum uplift forces that may be exerted a.
on any fuel assembly shall be limited to 375 pounds." This does not appear compatible with the fact that one fuel assembly weighs approximately 1600 pounds.
Clarify this interface requirement.
b.
Spent Fuel Pool (i) Provide embedment and foundation supports interface criteria for new and. spent fuel racks.
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a 012-5 012.32 (ii) Provide interface requirements for underwater storage Cont.
of tools and equipment in the spent fuel pool to prevent damage to spent fuel in the event of a safe shutdown earthquake.
I Provide a statement that fire protection requirements are c.
included under section 9.5.1.
012.33 In Section 9.1.5, " Interface Requirements," Category 19. " Environ-(9.1.5) mental," refer to Section 9.4 for equipment and buildings heating, 1
i cooling and ventilation environmental requirements.
012.34 Your response to Acceptance Review item 020.7 (Section 9.2.2) is I
(9.2.2) not complete.
In Table 9.2-2, provide interface requirements for cooling water system pressure and temperature. Give design basis values.
i 012.35 In Section 9.2.2, Table 9.2-1, " Individual Component Cooling Water (9.2.2)
Heat Loads," indicate the maximum, or design basis, heat loads to be used by the B0P designer.
Provide curves showing integrated heat loads for NSSS design scope only over a 30-acy period as follows:
(a) Sensible heat rejected (integrated over a 30-day period)
(b) Decay heat release from the reactor (integrated over a 30-day f
. period).
l (c) The sum of items (a) and (b) above over a 30-day period.
i Also provide the following information in tabular or graphical form:
1 (d) The fission products decay heat.
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(e) Theheavyelement(capture)decayheat.
(f) The sum of items (d) and (e).
In calculating the information requested under all items above, use the method set forth in the attached Branch Technical Position APCSB 9-2.
012.37 In Section 9.2.6, " Condensate Storage Facilities," provide the mini-(9.2.6) mum amount of auxiliary feedwater required by B&W at the intact steam generator (s) to bring the reactor to cold shutdown conditions assuming zero time at hot standby.
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012-6 012.38 The Chemical, Volurn Control and Liquid Poison Systems include (9.3.4) the Makeup and Purification (MU&P) Systein and Chemical Addition and Boron Recovery (CA&BR) systems shown on Figures 9.3-1 through 9.3-4.
Provide a description of the portions of these systems required for reactor shutdown under accident conditions.
Provide i
information on seismic classification, ability to meet single l
active failure and ability to withstand postulated pipe breaks in
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accordance with BTP-APCSB 3-1.
In your description, reference the P&lDs and identify the components or portions of the safety related systems that are required to shut down the reactor.
I 012.39 In Section 9.3.4.1.4, " Interface Requirements," Category 8, "Com-(9.3.4) ponent Heating and Cooling," provide the minimum fluid temperature or line heating requirements and identify which portions of the MU&P system (components and associated piping) must be protected to prevent precipitation of boron in that system.
Provide similar information for the CA&BR system in Section 9.3.4.2.6.
012.40 in Section 9.3.4.1.4, " Interface Requirements," Category 16, (9.3.4)
" Radiological Waste " provide a statement to the effect that an equipment and floor drainage system should be provided to direct the maximum leakage rates from the MU&P systems equipment to a radiological liquid waste processing system.
Identify the com-ponents or portions of systems where these leakages can occur and provide the leakage rate criteria the B0P designer should con-sider in his design of the drainage system.
Reference the section(s) where this information is provided. Provide similar information in Section 9.3.4.2.6 for the CA&BR system.
012.41 In Section 9.3.4 Table 9.3-10. " Makeup and Purification System (9.3.4)
Failure Analysis," provide a reference figure notation on the table and a new column, " Component Identification," after the first column on the table marked " Component." The component identification number should be that contained on the system P&ID.
This also applies to the Table 9.3-11 failure analysis for the CA&BR system.
The failure analyses provided for the MU&P and CA&BR systems are difficult to follow and do not cover the entire systems.
The analysis should be performed to greater depth and should analyze the entire systems as shown on Figures 9.3-1, 9.3-2, 9.3-3 and 9.3-4.
Provide such additional and revised failure analyses.
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1 012-7 012.42 Section 9.5.1 contains the B&W fire protection interface re-(9.5.1) quirements on the 80P designer.
Provide a tabulation of the items of equipment in the BSAR-205 design scope that constitute a significant fire hazard; e.g., pumps with lube oil supply, electrical cabinets, electrical cables, or other equipment containing combustibles.
The table should identify the item or component, the P&ID where shown (where possible) and the material or fluids which constitute the fire hazard. Provide quantities of combustible fluids (e.g., 'ubricating oil in reactor coolant pumps) and types and fire rating of cables used by B&W as part of the NSSS scope of supply.
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020 0 CONTAINMENT SYSTEMS i
022.1 To pennit an evaluation of the minimum containment backpressure, provide the mass and energy release to the containment as a function of time used in the ECCS evaluation. Also, specify the containment backpressure that was assumed.
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022.2 Figure 6.2-4 shows the containment vapor temperature as a function of time.
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Since this figure is not rnentioned in the text, discuss the purpose of the s
i figure, and describe the analysis that was done to generate the curve.
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022.3 The information provided in Section 6.2-4, " Containment Isolation Systems."
is inadequate to complete the review of the containment isolation provisions j
within the scope of B-SAR-205.
Provide the following i
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Discuss the bases for the design of the containment isolation system 3
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including:
a.
The governing conditions under which containment isolation I
becomes mandatory; b.
The criteria used to establish the isolation provisions for
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fluid systems penetrating the containment; I
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The criteria used to establish the isolation provisions for c
fluid instrument lines penetrating the containment; and.
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The design requirements for containment _ isolation barriers.
2.
Provide a table of design information regarding the containment isola-l tion provisions for fluid system lines and fluid instrument lines penetrad i
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020-2 022.3 -
ting the containment.
Include the following infonnation in this table:
Cont.
a.
Containment penetration number; b.
General design criteria or regulatory guide recomendations that -
have been met or other defined bases for acceptability; ^
c.
System name; d.
Fluid contained; e.
Line size (inches);
f.
Engineered safety feature system (yes or no);
g.
Reference to figure in SAR showing arrangement of containment isolation barriers; h.
Isolation valve number; 1.
Location of valve (inside or outside containment);
- j. Type C leakage test (yes or no);
k.
Length of pipe from containment to outermost isolation valve; 1.
Valve type and operator; t
Primary mode of valve actuation; m.
n.
Secondary mode of valve actuation; o.
Normal valve position; p.
Shutdown valve position; q.
Postaccident valve position; r.
Power failure valve position; s.
Containment isolation signals; t.
Valve closure time; and,
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Power source.
Cont.
3.
Specify the plant protection system signals that initiate closure of the containment isolation valves or refer to the section in the SAR where this information can be found.
4.
Provide justification for any containment isolation provisions that differ from the explicit requirements of General Design Criteria 55, i
l 56, and 57.
5.
Discuss the bases for the containment isolation valve closure times.
6.
Discuss the design requirements for the containment isolation barriers regarding:
The extent to which the quality standards and seismic design a.
classification of the containment isolation provisions follow the recomendations of Regulatory Guides 1.26 " Quality Group Classifi-cations and Standards for Water, Steam, and Radioactive-Waste-Containing Comoonents of Nuclear Power Plants," and 1.29. " Seismic i
Design Classification";
b.
Assurance of protection against loss of function from missiles, jet forces, pipe whip, and earthquakes; Assurance of the operability of valves and valve operators in the c.
containment atmosphere under normal plant operating conditions and postulated accident conditions; d.
Qualification of closed systems inside and outside the containment as isolation barriers; l
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020-4 I
022.3 e.
Qualification of a valve as an isolation barrier; Cont.
f.
Required isolation valve closure times; g.
Mechanical and electrical redundancy to preclude comon mode failures; h.
Primary and secondary modes of valve actuation.
8.
Discuss the provisions for detecting leakage from a remote manually controlled system (such as an engineered safety feature system) for the purpose of determining when to isolate the affected system or system train.
9.
Discuss the design provisions for testing the operability of the isolation valves.
- 10. Describe the environmental qualification tests that have been or will be performed on the mechanical and electrical components that may be i
exposed to the accident environment inside the containment. Discuss the test results.
Demonstrate that the environmental test conditions i
(temperature, pressure, humidity, and radiation) are represuntative
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of conditions that would be expected to prevail inside the containment following an accident. Graphically show the environmental test condi-tions as functions of time or refer to the section in the SAR where this information can be found.
11.
Identify the codes, standards, and guides applied in the design of the containment isolation
- system and system components.
020-5 022.3
- 12. Provide an evaluation of the functional capability of the contain-Cont.
ment isolation system in conjunction with a failure mode and effects analysis of the system.
- 13. Provide evaluations of the functional capability of isolation valve l
l seal systems and of fluid-filled systems that serve as seal systems.
022.4 General Design Criteria 53 and 54 require that the containment penetrations and associated containment isolation barriers be designed to permit periodic
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leakage rate testing.
For those containment penetrations within the scope of B-SAR-205, provide the following information:
1.
Discuss the design provisions for leakage rate testing of the contain-ment isolation barriers, and show on system drawings the design provision for testing. Discuss the design and functional capability of associated containment isolation systems (such as isolation valve seal systems) l that provide a sealing fluid or vacuum between isolation barriers and of fluid-filled systems that serve as seal systems.
I 2.
Provide a listing of all containment penetrations.
Identify the contain-ment penetrations that are exempt from leakage rate testing and provide justification.
3.
Provide a listing of all containment isolation valves.
Identify the containment isolation valves that are exempt from leakage rate testing and provide justificat. ion.
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020-6 022.5 The following questions pertain to Section 6.2-5, " Combustible Gas Control in Containment":
1.
Discuss the analysis, including assumptions, that was done to determine the radiolytic hydrogen generation rate shown in Figure 6.2-1.
Compare parameter values to the guideline valves of Branch Technical Position CSB 6-2, " Control of Combustible Gas Concentrations in Containment Following a LOCA."
2.
Figure 6.2.2 shows the total hydrogen generated for three aluminum inven-tories. Clarify how the balance of plant designer is to use this figure.
Discuss the components that are included in these aluminum inventories.
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3.
Discuss the contraints on the balance of plant design for the following i
sources of hydrogen:
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a.
aluminum corrosion; and, b.
zine corrosion (including paints).
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l 033.0 ELECTRICAL, INSTRlNENTATION AND CONTROL SYSTEMS 4
533.16 Provid.e your basis for concluding that a total lifetime (3.11) 4 integrated dose of 2 x 10 rads is sufficiently low enough to g.
preclude the requirement.to radiation type test Class IE equip-a l
ment.
Identify all such equipment that.is within the BSAR-205 scope of supply.
033.17 Recently, the staff has re-examined containment temperatures (3.11) following a postulated steam line break accident and has determined that transient temperatures may exceed 400 F for approximately 70 seconds.
(Exact temperatures and durations P
l are containment dependent.) Table 3.11-2 of B-SAR-205 estab-j lishes the envelope of environmental design conditions that the
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BOP containment must meet. It is not apparent from the infor-mation presently in Section 3.11 and specifically in Table 3.11-2 l
that such temperature transients have been properly addressed.
l We request that this section be amended to reflect this concern 4
by providing appropriate interface criteria requirements.
1 033.18 In order to conduct our review of Section 3.11 in accordance l
(3.11) i I
with the Standard Review Plan, we require that the following l
interface criteria requirements be added to B-SAR-205:
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For the equipment located within the containment, a statement as to whether or not physical location within c
l the missile shield will affect any of the B-SAR environmental i
qualification assumptions, 2.
An envelope of acceptable non-seismic vibration, and l
3.
A statement regarding submergence of the equipment.
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033,2 033.19 In order for the staff to conclude in our safety evaluation (7.0)
(8.0) report (SER).that there is reasonable assurance B-SAR-205 meets the Commission's regulatiions, there must be specific interface criteria requirements provided for the BOP design. We therefore i
j request that Table 7.1-1 be augmented and a table in Section 8.0 i
be added such that the resulting tables are equivalent to those presented in the NRC's Safet Evaluation Reports for other standard I
PWR nuclear steam systems. Table 7.1-1 should include the BSAR 1
supplied systems and all interfacing and/or auxiliary supporting systems.
033.20 "Ihe B 4 W 1etter from Kenneth E. Suhrke to John F. Stolz of (7.1) i (8.0) the NRC dated January 23, 1976, set forth B 6 W's physical i
separation criteria in order to demonstrate compliance with Regulatory Guide 1.7S.
We have found this document to be more complete in terms of electrical separation and more responsive to BOP interface criteria requirements than B-SAR-205. We request that the contents of this letter be made part of the B-SAR docket to become part of the B-SAR review.. In order to faci-litate an expeditious review, the following comments should be addressed where this information is incorporated into the docket.
a.
Ensure that all descriptions pertain to the B-SAR-205 design.
b.
Where references are made to "Balarice of Plant Criteria documents" or statements such as "B 4 W will stipulat.a":
specific interface criteria requirements should be l
substituted.
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033-3 33.20 c.
Expand Section V to include all types of plant protection gont.
system and SRCI isolation devices. It is the staff's position that al.1 such isolation devices be qualified l
in accordance with IEEE Std 279-1971, Section 4.7.2 i
and IEEE Std 323-1974. Your response should include or reference the test plans, test set ups, test durations, acceptability goals, and the schedule for submittal of the results of both the electrical and environmental i
qualifica' tion tests.
In addition, supplement the electrical interface criteria requirements already I
.i present with environmental interface criteria require-cents to ensure that the isolated circuits will never 1
be associated with or subject to environmental condi-tions more severe than those for which the isolators I
have been qualified.
Environmental conditions should include electro-magnetic interference.
d.
The final paragraph of Section VIII should be expanded
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and tabularized in Section 8.0 of B-SAR.
033.21 Expand your description of the pump monitor sensors and assoc-(7.2) iated equipment that provide RC pump status to the Reactor Pro-tection System. Your response should include the total I
number of lines, physical separation, and protection; a des-cription of the sensors, and a list of any non-Class IE equip-ment relied upon directly or in a supporting role. Full justi-fication and supporting bases must be provided for all non-Class I
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033-4 033.21 IE equipment so identified. Par those portions of the above Cont.
request that exceed the scope of supply of B-SAR, detailed interface criteria requirements should be provided.
033.22 Provide a more detailed description of the manual blocking i
(7.3) l feature that allows the operator to intervene during the t
i accident. Your response should demonstrate conformance with i
the requirements of IEEE Std 279-1971, Section 4.16 and should i
include the preliminary design drawings supporting this aspect i
i of the design.
l 033.23 Lightning protection for the ESFAS as mentioned in Section (7.3) 7.3.1.2.1 is not listed as an interface criterion requirement.
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Specifically address this requirement in B-SAR-205.
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033.24 Supplement Figure 7.3-1 to illustrate the origin of the Reset (7.3) and Trip pushbuttons fecaing the actuation subsystems.
Include in this description whether these contacts are gangedin some fashion or are all separated and require separate actuation.
l 033.25 It appears that the Trip Portion (CRD main and secondary (7.4) buses) of the Control Rod Drive Control System does not meet the single failure criterion as required by the Commission's regulations. Provide sufficient details of this design to demonstrate that all credible failures and electrical faults will not prevent one of the two trip buses from de-energi ing i
as both are required. to de-energize for reactor trip.
Further-l more, the text refers to "very exacting and restrictive criteria."
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i 033-5 033.25 Confirm that these phrases include IEEE Std 279-1971 and Cont.
IEEE Std 323-1974. Figure 7.7-10 which pertains to this aspect of the design is, issing, please provide.
m 033.26 Sections 7.4 and 8.3 describe what appears to be an installed (7.4)
(8.3) spare High Pressure Injection (llPI) pump and attendant controls.
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No power source has been assigned to pump B.
It would appear that i
Section 8.3.1 is in error when it requires only two independent emergency diesel generator busses, as switching the B HPI l
pump from one emergency power division to the other could unduly jeopardize the independence of the BOP onsite emergency power system. The same concern applies to the description in Section 8.3.1.10e that indicates both ESEAS actuation systems l
are routed to the B HPI pump.
Provide specific, detailed, o
and exhaustive bases, justification, and interface criteria f.
requirements to support this aspect of your design.
033.27 Section 9.3.1 states that " compressed air may be required (9.3) for control of auxiliary feedwater, depending on the arrange-ment of Class IE power supplies.
Interface criteria related I
to this are included in Section 10.1.2."
Provide a detailed 6
explanation of the first sentence. Section 10.1.2 does not i
exist. Correct this.
j 033.28 It appears from Table 7.5-1 that the B-SAR scope of supply (7.5) l for safety related display instrumentation does not conform I
to all of the requirements of Branch Technical Position EICSB 23.
t Compare your design to the position in such a manner that it will i
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033-6 033.28 be clear to the BOP designer what must be done to bring the Cont.
entire design (B-SAR and BOP) into conformance with the position.
033.29 Section 7.6.1.1.2 describes the decay heat section valve inter-1 (7.6)
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f locks. However, motive power for the MOVs is not discussed.
i Verify that the motive power aources are so arranged that the j
required system function and system isolation capability will be performed given a single failure.
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033.30 Your response to Acceptance Review Itt.m 033.15 did not demonstrate (7.7) full compliance with Branch Technical Position EICSB 4.
Specifically, your design does not provide 'an ESFAS signal I
to the valves as required by Section B.1 of the position.
I We require your design to be modified to meet the position f
or your bases and justification be provided in any exception you may wish to take.
i 033.31 Describe the methods that will be used for testing ^the RPS trip and ESF sensor response time. Compare their response times to those allowed by the accident analyses and state the margin in terms of time between design response and that used l
in the analysis for the worst case event. Also provide a description of testing of sensor response time for other safety related systems.
033.32 We require the following design criteria or their equivalent (7.2) l (7.3) be used in establishing the range of instrumentation and in selecting trip setpoints for safety related functions:
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033-7 I-033.32 1.
The range selection for instrumentation shall be such as Cont.
to exceed the expected range of the process variable being monitored.
2.
The accuracy of all the safety trip points will not be numerically larger than the accuracy that was assumed in the accident analysis.
3.
The trip setpoints should be located in that portion of an instrument's range which is most accurate and must be j
located in a region with the required accuracy.
4.
All safety trip points will be chosen to allow for the normal expected instrument system setpoint drifts such that the Technical Specification limit will not be exceeded.
)
5.
Verification of the above criteria shall be demonstrated I
as a part of the qualification test program required by IEEE Std 323-1974.
Provide a detailed description of this aspect of your design.
033.33 The Nucicar Instrumentation (NI) System is an integral part (7.8) of the RPS required to operate under normal as well as ab-i normal conditions that may include a hostile envi.ronment.
This hostile environment may prevail for some time before any plant RPS parameter exceeds its trip setpoint and initiation of the reactor trip occurs.
Discuss the assurance you are providing by design and/or equipment qualification that the RPS will provide.all protective functions it is intended to provide under the worst possible environmental conditions that may develop as a result of an incident such
5' 033-8 I
033.33 as a small high energy fluid line break and/or a Safe Cont.
Shutdown Earthquake (SSE).
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033.34 It is not clear from you'r accident analysis what the require-ments are rc?arding disconnection of the Reactor Coolant pumps from their power sources (because of degraded frequency or fault conditions of the power sources) to allow for full coastdown of the pumps. Therefore, provide the following additional information.
1.
In the event that full pump coastdown is needed to prevent s
exceeding core safety limits, provide an analysis to describe the effects on pump coastdown capability in the event that the pump breakers failed to isolate the power supply during an underfrequency condition. Your response i
should define any limiting underfrequency condition and l
j the maximum allowable frequency decay rate. Your analysis should include all allowable RC pump configurations (e.g.,
i two, three or four pump operation).
l 2.
If reactor coolant pump coastdown is demonstrated to be required for any of the conditions in Part I above, the BOP designer must be made aware that the NRC will require the RC pump breakers and associated underfrequency l
trips to be designed and qualified in accordance with the i
requirements of IEEE Std 279-1971 and IEEE Std 308-1974 l
l l
or that Technical Specification requirements will be imposed that prohibit operation in any of the RC pump l
configurations so identified.
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033.35 Following the termination of the B-SAR 241 review, the NRC (General) issued a letter to BGW from A. Schwencer to James Mallay dated 1
April 4, 1975. 'Ihis let,ter contained seven items felt to be of a generic nature and that the staff wished to pursue directly with B6W rather than on any applicant's docket. We i
would now desire to incorporate these items into the B-SAR-205 review for final resolution. We are in receipt of letters l
from K. Suhrke to A. Schwencer dated May 23, 1975; July 29, 1975; I
August 13, 1975; and December 30, 1975. Please confirm that i
i these letters encompass the complete BGW response to date and that each response is applicable to B-SAR-205. We note that some of the items have been incorporated to various degrees 3
in B-SAR-205 and that others have been reworded and asked
?
l again eleswere in this question list, however, we have mentioned all seven items here for completeness. Please i
include a brief discussion in your response to specifically j
address how each item has been incorporated into the B-SAR-205 docket.
033.36 It appears from Sections 8.1.1 and 8.2 that your design calls (8.1)
(8.2) for four reactor coolant pu=ps rated at 12,500 horsepower all powered from one 6.6 kV bus.. Please confirm that this reflects your intended design.
In conjunction with this request, please provide a typical one line diagram that depicts an acceptable configuration to support the B-SAR-205 power requirements.
l 033.37 In Section 8.3, there should be an explicit statement as (8.3) l to whether or not B-SAR-205 has any direct d-c loads.
If so, l
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033-10 l
033.37 specifications covering ripple and other distortions should Cont.
be added to the interface criteria requirements for the d-c sources.
~i 033.38 In Table 8.3-1 or an additional table, include the following i
(8.3) for each piece of equipment listed:
1.
Earliest time required to be fully operational as per Chapter IS accident analyses.
2.
Maximum time from initiation to fully operational status at the limiting onsite power conditions as described in your interface criteria requirements.
033.39 Please address your conformance, by description or reference to l
Pertinent sections of the BEAR-205, with the following Branch Technical Positions listed in Appendix 7A of the USNRC Standard i
Review Plan.
1.
EICSB 14 - Spurious Withdrawals of Single Control Rods in PWRs 2.
EICSB 20 - Design of Instrumentation and Controls Provided to Accomplish Changeover from Injection to Recirculation Mode 3.
EICSB 21 - Guidance for Application of Regulatory Guide 1.47 4.
EICSB 27 - Design Criteria for Thermal Overload Protection of Motor-Operated Valves 033.40 At various places throughout B-SAR-20S reference is made to BAN-10082 (Equip =ent Qualification) Parts I, II and III. As not all of this information has been submitted for NRC review, we require a commitment similar to that made for BAW-1008S that any generic resolution will be directly factored into the B-SAR design.
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110-1 110.0 MECHANICAL ENGINEERING 1
110.18 Specify the location and number of hydraulic enubbers which will (3.6.3.1) be used as piping or equipment supports and their rated load.
Describe the bases for specifying snubber load capacity, per-formance parameters (locking velocity, spring rate, etc.) and qualification testing to assure operability.
110.19 The response to Question 110.15 refers to Section 3.9.2.4.
(3.9.1.2)
However, Section 3.9.2.4 is only applicable to active pumps and valves. Section 3.9.1.2 states that test procedures for the seismic qualification of mechanical eqt.ipment will be detailed in the applicant's FSAR. This is not acceptable.
If B & W is responsible for the seismic qualification of any safety related ASME Class 1, 2 and 3 mechanical equipment such as fans, hmt exchangers, battery and instrument racks, control consoles, cabinets, panels and cable trays, provide, in the BSAR-205, the information requested in Question 110.15.
If the applicant will be responsible for this information, provide an explicit statement to this effect in Section 3.9.1.2 and change "FSAR" to "PSAR" in that section.
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l 110.20 The response to Question 110.9 is not completely acceptable.
l (3.9.2.4)
There are several references,both direct and indirect, in Sections 3.9.2.4, 3.9.2.4.1 and 3.9.2.4.2 to IEEE Standard 344,
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1971. The 1971 version of IEEE 344 is not completely acceptable for seismic qualification of electrical equipment. Revise the above noted references, including IEEE-382, 1972, to be consistent with NRC Standard Review Plan, Section 3.10.
110.21 The dirst sentence in Paragraph 4 of Section 3.9.2.4 of B-SAR-205 (3.9.2.4) is not completely acceptable. The pump and valve manufacturers should be required to demonstrate that the pump or valve will operate normally when subjected to all loads and other environ-mental conditions associated with a faulted condition. These loading conditions should be clearly defined to the pump or valve manufacturer. Provide additional information in Section 3.9.2.4 which is consistent with Paragraph II.2.c, " Design Specifications" of NRC Standard Review Plan, Section 3.9.3.
110.22 Provide a more detailed explanation of how Tables 3.9.3 and
-(3.9.2.4) 3.9.4 will be used.
Include the technical bases for the coefficients listed in both of these tables.
110.23 In Section 3.9.2:4.1, provide a discussion of the procedures (3.9.2.4) for the seismic qualification of valve operators in the frequency range of 1-4 Hz or justify why this range was omitted.
110.24 Seismie qualification of pump motors by enalysis alone. as discus-(3.9.2.4) sed in Section 3.9.2.4.2 is not sufficient.
The staff will require
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110-2 110.24 that operability of all active pumps and valves in the B-SAR-205 Cont.
be assured by tesing complex active devices such as pump motors, valve operators and other electrical, mechanical, pneumatic or hydraulic appurtenances which are vital to the actuation and continued operation of the pumps and valves. The test program should include all applicable loads associated with the faulted condition. An acceptable program is outlined in NRC Standard Review. Plan, Section 3.10.
Provide your commitment to such a program or propose an equivalent alternate for our review.
110.25 In the response to Question 110.13, operability assurance of the I
(3.9.2.4) pressurizer safety relief valve as discussed in Section 3.9.2.4.10 (5.2.2.2) of the PSAR is not entirely acceptable. Clarify the source (OBE, SSE, pressure transients, etc.) of the accelerations to which the factor 1 2 is applied and the basis for the 1.2 factor.
1 Identify the transient forces, describe the analytical model j
and methodology for determining the dynamic response character-l istics (frequencies, raode shapes, displacements and stresses)
I of the safety relief valve and support structure. Discuss how 3
successful actuation within the design requirements of the valve shall be demonstrated.
i 110.26 With respect to the reactor vessel and its supports:
(5.2)
(5.5)
(1) Provide drawings of the reactor support system sufficient to l
show the geometry, dimensions of all principle elements and j
the type of material they are fabricated from.
l (2) Demonstrate that the analytical methods, models and load considerations utilized in obtaining the loads which are i
transmitted to the vessel supports are based upon adequate consideration of the dynamic forces that result from:
(a) Transient asymmetric differential pressure loadings on the reactor internals.
(b) Asymmetric loadings from transient differential pressures that exist around the exterior of the reactor vessel resulting from a postulated pipe rupture in the reactor coolant loop at the reactor vessel nozzle.
110.27 To assure that the postulated pipe rupture information for the (1.7)
BOP supplied piping is accounted for in the NSSS design, provide (3.6.6) the following in either Section 1.7 or 3.6.1 of B-SAR-205:
A commitment to, coordinate the design of the RCS with interfacing BOP-designed piping systems regarding postulated pipe break locations, orientations and configurations and resulting loads to assure compatibility.
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110-3 110.28 To assure the analytical compatibility of NSSS and BOP stress 5'
(1.7) analyses, provide the following in either Section 1.7 or (3.9.1.6) 3.9.1.6 of B-SAR-205:
r A commitment to coordinate the design of the RCS and inter-j facing BOP-designed systems, components and supports when inelastic analysis methods are used by either the NSSS or the BOP to assure compatibility. Areas requiring coordination should include analytical criteria, procedures and results of analyses.
110.29 To assure that all B & W supplied safety related mechanical (1.7) equipment is designed to the BOP defined floor response spectra, l
(3.9.1.2) provide a commitment in either Section 1.7 or 3.9.1.2 to verify i
that all NSSS safety-related mechanical equipment as supported will be adequate for the floor response spectra defined by the 4
. BOP.
110.30 To assure the structural adequancy of NSSS equipment and structural f
(1.7) compatibility with interfacing equipment furnished by the BOP (3.9.1.6) supplier, provide the following in either Section 1.7 or in (5.2.1.5)
Sections 3.9 and 5.2; l
A commitment to coordinate the design of NSSS components and supports with the design of interfacing BOP components and i
supports regarding design loading combinations to assure i
compatibility. Include your commitment to coordinate, with the BOP, the appropriate plant and component operating conditions.
110.31 j
For those NSSS systems designed by B & W but for which safety /
(1.7) relief piping and supports are designed by the B0P supplier, (5.2.2.2) provide the following in either Section 1.7 or 5.2.2.2:
A commitment to provide the BOP with the system overpressure protection parameters.
110.32 To assure that active pumps and valve supports which are supplied (1.7) by the B0P are designed so that they do not deform to the extent (3.9.2.4 that they would impair the operability of the B & W supplied pump or valve, provide the following in either Section 1.7 or 3.9.2.4:
A commitment to provide the BOP with limiting criteria affecting NSSS active component operability.
110.33 To assure that all B & W safety related electrical equipment and (1.7) instrumentation is-designed to the BOP defined floor response (3.10) spectra, provide the following in either Section 1.7 or 3.10:
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110.33 A commitment to verify that all NSSS safety-related instru-Q:i.
Cont.
mentation' and electrical equipment as supported will be-c, adequate for the floor response spectra defined by the BOP.
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121-1 121.0 MATERIALS ENGINEERING - MATERIALS INTEGRITY 121.1 Regulatory Guide 1.14 Revision 1 is applicable to plants (3.1.2.14) docketed subsequent to January 1,1976 and should be addressed for B-SAR-205.
I 121.2 The referenced section suggests that' inlet and ' outlet I
(5.2.4.2f(1)) pressure vessel nozzles may be inspected visually or by l
contact during refueling intervals.
Provice details on method by which inlet nozzles may be inspected without i
removal of internals.
f 121.3 Topical Report BAW 10040 is mentioned in the referenced l
5.2.6.1.7) section but is not' listed in Table 1.6-1 nor as Reference, Table 1.6-1) 20 in Section 5.8 as indicated _by the superscript.
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122-1 122.0 MATERIALS ENGINEERING-METALLURGY 122.8 Your response that no impact toughness requirements are imposed on ASME Class 2 components is unacceptable.
Provide information indicating the extent to which your equip-ments conform to the following:
Pressure retaining material of I
which ASME Class 2 components are constructed shall be tested for fracture toughness according to the procedures, requirements j
and acceptance standards of the ASME Code,~ Paragraphs NC-2320 l
through NC-2360.
For materials with thickness exceeding 2h inch, the lowest service temperature shall not be lower than RTNDT + 100 i
(rather than RTNDT + 30 as stated in NC-2332d) unless a lower i
temperature is Justified by methods similar to those contained j
in Appendix G.
Materials'describedinNC-2311b,(1)through(7)andotherswhich i
can be shown to have adequate fracture toughness at all service temperatures may be exempted from these requirements.
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i 130-0 STRUCTURAL ENGINEERING
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132.19 Different break points of frequency from those of Regulatory (3.7.1)
Guide 1.60 are used in the design response spectra for vertical motion.
Provide an explanation for such deviation.
132.20 The nominal values of maximum ground acceleration as stated l
(3.7.1) on page 3.7-1 are 0 3g for the SSE and 0.159 for the OBE, while
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the zero period accelerations shown in Figures 3.7-1 and 3.7-3 i
are 0.69 and 0.39 respectively.
Explain how you obtained the r
spectral values, as noted above, for the base mat.
132.21 Clarify that the equivalent static load method is used only for (3.7.2) components that exhibit a dominant single degree of freedom (R3P) response and a fundamental frequency of 33 Hz or greater.
For frequencies less than 33 Hz, a factor of 1.5 should be applied i
to the peak of the floor response spectrum to account for the contribution of higher modes. A factor of less than 1.5 may be used if adequate justification is provided, i
132.22.
Provide information for the following items in analytical modeling:
(3.7.2) j (a
Designation of systems versus subsystems.
,I (b
Decoupling criteria for subsystem.
(c Three-dimensional modeling and decoupling criteria.
132.23 Indicate which NSS components are vulnerable to external missiles.
(3.5.3)
Furthermore, indicate which components need to be protected assuming that they are in the paths.of the external missiles.
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211.0 C'LASSIFICATION, CODES & STANDARDS F
In Tables 3.2-4 and 5.1-2, and in Figure 5.5-8 your Quality Group D classification and non-seismic Category 1 classification of the reactor l'
coolant pump auxiliaries that are an integral part of each reactor I
l coolant pump assembly is not in conformance with Regulatory Guides l
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l 1.26 and 1.29 and is unacceptable.
i In Table 3.2-4 and in Section 5.5.1 of the SSAR provide a detailed I
list and description of those components that are generally identified i,
in Table 3.2-4 as reactor coolant pump auxiliaries.
It is our position that those components that are an integral part of the reactor coolant pump, such as the oil lubrication system, oil I
coolers, seal coolers and cooling water lines to the interface point with the camponent cooling water system should be classified Quality Group C, constructed to ASME Section III, Class 3 and designed to seismic Category I requirements.
It is also our position that the following criteria shall apply to that I
portion of the component cooling water system which interfaces with the reactor coolant pumps to supply cooling water to pump seals and bearings during normal operation, anticipated transients, and following accidents:
(1) A moderate energy leakage crack or a single failure in the com-l l
ponent cooling water system shall not result in fuel damage
211-2 211.2 or damage to the reactor coolant system pressure boundary Cont.
caused by an extended loss of cooling to the reactor coolant pumps. Single failures include operator error, spurious I
actuation of motor operated valves, and loss of component cooling water pumps. Moderate leakage cracks should be l
determined in accordance with the guidelines of Branch Technical i
Position APCSB 3-1.
l (2)
An accident that is initiated from a failure in the component cooling water system piping shall not result in excessive fuel damage or a breach of the reactor 4
coolant system pressure boundary when an extended loss of cooling i
to the reactor coolant pumps occur. A single active failure shall t
j be considered when evaluating the consequences of this accident.
Moderate leakage cracks should be determined in accordance with B anch Technical Position APCSB 3-1.
In order to meet the criteria established above, a BSAR-205 interface l
requirement should be imposed on the balance of plant component cooling I
water system that provides cooling water to the reactor coolant. pump seals and bearings so that the system will meet the following conditions:
(1) A period of 20 minutes is considered acceptable within which an operator can trip the reactor coolant pumps and initiate a safe plant shutdown. Therefore, in order to provide an adequate
7 211-3 211.2 margin of safety, unless it can be demonstrated that the reactor Cont.
coolant pumps are designed so that they can operate without i.
cooling water for a minimum period of 30 minutes without loss of ii i;
function or the need for reactor operator action, the com-l:
i ponent cooling water system should be designed to meet the I
following:
(a) Seismic Classification: Seismic Category 1 (b) Quality Classification: Quality Group C Component Code: ASME Section III, Class 3
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(c) Single Failure: Should be capable of withstanding a single active failure and should be capable of withstanding a l
moderate energy leakage crack in accordance with Branch Tech-4 nical Position APCSB 3-1, with respect to cooling reactor coolant pumps or else, item (1)(d) must be implemented. A t
single failure includes malfunctioning of any valve or pump in the component cooling water lines to the reactor coolant pumps.
(d)
Instrumentation and Controls: Safety grade instrumentation
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is required to detect loss of component cooling water and initiate automatic protection of the plant.
l (e) Containment Isolation of Systems: Only when reactor coolant j
pumps are not functioning.
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211-4 211.2 (2) For a reactor coolant ' pump design that can operate without Cont.
cooling water for longer than 30 minutes without loss of function or the need for reactor operator action, the component cooling i
water system should be designed to meet the following:
(a) Seismic Classification: Non-seismic Category 1, except for that portion of the component cooling water rystem that forms an extension of the containment boundary.
(b) Quality Classification: Quality Group 0, except for that portion of the component cooling water system that forms i
an extension of the containment boundary.
(c) Single Failure: The system should be capable of withstanding a single active failure and must be capable of withstanding a moderate energy leakage crack in accordance with Branch Technical Position APCSB 3-1, with respect to cooling' reactor coolant pumps, or else, item (2)(d.) must be t'
implemented. A single failure includes _ malfunctioning of any valves or pumps in the component cooling water lines to the reactor coolant pumps.
(d)
Instrumentation and Controls: Safety grade instrumentation is required to detect the loss of component cooling water and provide an alarm in the control room to satisfy item (2)(c).
(e) Containment Isolation o'f Systems: Only when reactor coolant pumps are not functioning.
211-5 211.2 The reactor coolant pumps are within the B-SAR-205 scope of supply; Cont.
therefore, in order to demonstrate that a reactor coolant pump design can operate with loss of component cooling water for longer than 30 I'
minutes without loss of function or the need for reactor operator action, provide the following:
(1) A detailed description of the events following the loss of com-ponent cooling water to the reactor coolant pumps and an analysis demonstrating that there are no consequences important to safety which may result from this event.
Include a discussion of the effect that the loss of cooling water to the seal coolers has on the reactor coolant pump seals. Show that the loss of cooling water does not result in a loss-of-coolant accident due to seal failure.
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(2) A detailed analysis to show that loss of cooling water to 'the l
reactor coolant pumps and motors will not cause a loss of the flow coastdown characteristic or cause seizure of the pumps,
assuming no administrative action is taken. The response should include e detailed description of the calculation procedure including:
(a) The equations used.
(b) The parameters used in the equations, such as the design para-meters for the motor bearings, motor, pump and any other equip-
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ment entering into the calculation, and material property values for the oil and g,etal parts.
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1 211-6 i
(c) A discussion of the effects of possible variations in part dimensions and material properties, such as', bearing clearance' tolerances and misalignment.
l (d) A description of the cooling and lubricating systems-(with appropriate figures) associated with the reactor coolant pump and motor and the design criteria and standards therefore.
(e)
Information to verify the applicability of the equations and
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material properties chosen for the analysis (i.e., references should be listed, and if empirical relations are used, provide -
8 a comparison of their range of application to the range used in the analysis).
Should an analysis be provided to demonstrate that loss of. component cooling water to the reactor coolant pumps and motor assembly is acceptable, we will require certain modifications to the plant technical specifications and a reactor coolant pump test conducted under operating conditions and with component cooling water terminated for a specified period of time to verify the analysis.
211.3 (3.2)
In Table 3.2-4, the fuel transfer tube gate valve is incorrectly identified as Safety Class III-2. This component should be classified Safety Class III-4.
Revise Table 3.2-4 accordingly.
211.4 In Table 3.2-4, the deborating demineralizer of the chemical addition (3.2) and boron recovery system is incorrectly identified as Quality Group D.
This component should be classified Quality Group C.
Revise Table 3.2-4 accordingly.
211.5 (5.2.1.3)
In Table 5.2-1, delete the last serrtence of the footnote and add a new sentence as follows: "The code edition and addenda identified -
above for each component are minimum requirements and will be
211-7 superseded as applicable based on a utility applicant's docket date of his PSAR or the actual purchase order date for a specific component in compliance with the rules of 10 CFR Part 50, Section 50.55a."
211.6 In Table 5.2-2, the use of Code Case 1448-2, "Use of Case Interpretations (5.2.1.4) of ANSI B31 Code for Pressure Piping,Section III, (8/14/72)", is l
acceptable, except for Piping Code Case 80, referenced therein, which s
has not been approved for use by NRC. Clarify the use of Code Case 1448-2 in Table 5.2-2.
211.7 Code Case 1572, " Fracture Toughness, Class 1 Components,Section III",
(5.2.1.4) was annulled 3/8/74 and the provisions of the code case were published in the Winter 1973 Addenda to Section III.
Therefore, Code Case 1572 should be deleted from Table 5.2-2 unless it is your intent to utilize this code case in conjunction with the 1971 Edition of Section III.
211.8 (5.2.1.4)
The following code cases have been superseded by a later code case revision than those identified in Table 5.2-2.
Review these code cases and revise Table 5.2-2 where applicable.
CODE CASE IDENTIFIED CODE CASE - LATEST IN TABLE 5.2-2 REVISION 1337-9 1337-10 1484-1 1484-2 1644 1644-2 1618 (Supp. 1) 1618-1 1335-8 1335-9 1553 1553-1
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211-8 I
t 211.9 Code Case 1603, " Toughness Tests When Cross-Section Limits; (5.2.1.4) j-Orientation and Location of Specimens,Section III" was annulled t
5/10/74 and the provisions of the code case were published in the Sumer 1974 Addenda to Section III. Therefore, Code Case 1603 should be deleted from Table 5.2-2..
I 211.10 In. Table 5.2-1, the 1971 Edition of Section III of the ASME Boiler (5.2.1.3) and Pressure Vessel Code that you have identified as applicable to:
(1) Reactor Vessel, (2) Steam Generators, and (3) Pressurizer, is not in conformance with the Codes and Standards Rule, Section 50.55a of 10 CFR Part 50.
l Based on a docket date of 3/1/76 for this application, these components as a minimum should be constructed to the 1974 Edition of Section III.
Revise Table'5.2-1 accordingly.
q 211.11 In Table 9.3-2, the component code for the Makeup Tank of the Makeup (9.3) an'd Purification System is incorrectly identified as ASME Section III, Class 3.
This component should be constructed to ASME Section III, Class 2.
211.12 In Tables 9.2-3 and 9.3-5 add the seismic classification of the (9.3) components identified therein.
211.13 In Table 9.3-5, the Qualit'y Group and component code for the Deborating (9.3) c l
Demineralizer of the Chemical Addition and Boron Recovery System is l
l
211-9 l
incorrectly identified as Quality Group D and the component code as' ASME Section VIII. This component should be classified Quality Group C and constructed to ASME Section III, Class 3.
i 211.14 In Tables 3.2-4 and 9.3-5, the Quality Group, component code and Seismic (9.3)
Classification for the Reactor Coolant Degasifier Package of the l
i Chemical Addition and Boron Recovery System is incorrectly identified
)
as Quality Group D, the component code as ASME Section VIII and non-seismic Category 1.
This component should be classified Quality Group C, constructed to ASME Section III, Class 3, and designed to seismic Category 1 requirements.
211.15 Add the following components of the Chemical Addition and Boron Recovery (3.2)
System that are shown in Figure 9.2-2, Sheet 2 to Table 3.2-4.
These i
components are:
(1) R.C. Bleed Holdup Tank, and (2) R.C. Disti.11 ate Storage Tank.
211.16 In Figure 10.1-1, Sheet 2, Secondary Plant System, identify the changes (10.1) in Quality Group classifications at the appropriate valves in the main steam, feedwater and auxiliary feedwater lines.
211.17 in Section 10.5.2, item 6, the reference to isolation valve V20 in (10.5.2)
Figure 10.0-1, is incorrect. The correct reference is Figure 10.1-1.
f
'All In addition, revise the last sentence of item 6 as follows:
components and piping in the condpnsate storage facilities for the auxiliary feedwater system shall be Seismic Category I and designed to comply with ASME Section III, Class 3 requirements.
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l 212.0 REACTOR SYSTEMS 212.83 (1.1.3)
Deleted P
e 212.84 Table 1.3-1 shows that several design features are unique to -
i (1.3)
B-SAR.
Discuss all changes from previous 205-fuel' assembly plants and explain the reason for each change.
212.85 Provide your schedule for submittal of the 205-fuel assembly vessel (1.5.1) model flow test results.
l 212.86 The statement is made that the Mark C fuel assembly research program i
(1.5.2) results will be completed in 1975. Provide this material.
i 1
212.87 Amendment 1 substantially revised this section. Since the reference (3.5) documentation did not change, submit the new information on which-l these recent revisions were bt. sed..
i 212.88 Explain deleting the statement on page 3.5 4 that the CRD assembly is the worst-case missile for which the shielding above the CRD (3.5.2) assemblies is designed.
212.89 Justify each de.letion in Table 3.5-1 (for example, the pressurizer (3.5.2) heater bundle stud). Every component deleted must be addressed.
212.90 For the worst case missile (CRD assembly), provide the specific (3.5.2) calculations and assumptions upon whth the Table 3.5-1 values are 2 missile impact' area and indicate the based. Justify the 64 in sensitivity of damage to Impact Area.
212.91 Consideration of potential missiles from reactor coolant pumps is (3.5.2) not sufficient to allow an adequate evaluation. Although mention was made of the flywheel,'all other rotating components, such as pump shaft, motor, and impeller must be addressed.
i 212-2 i
212.92 Provide a list of those category I structures, systems and (3.5.2) co=ponents in Table 3.2-4 which are inside containment and iden-tify which ones will be provided with missile protection.
212.93 Submit the CRDM mechanical design analysis report referred to on (4.2.3.3) page 4.2-35.
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212.94 Table 4.2-6:
Describe in greater detail the potential effect l
(4.2.3.3) of a loss of coolant water to the CRDM.
Discuss the maximum i
temperatures which could occur in the CRDM and relate to the thermal affects upon each control rod component essential to the F
trip function (such as bindage of the roller nut assembly or increased friction in the guide tubes).
Quantify the maximum I
expected affect of the high te=peratures upon each clearance
[
critical to the trip function. Reference and sue =arize available test data which would show that the performance of the trip
{
function would remain unaffected by these elevated temperatures.
212.95 B&W proposes exceptions to Regulatory Guide 1.68 by deleting rod (4.2.3.4) drop time measurements with no recirculation flow and reducing the required number of rod drop seasure=ents for each of the fastest and slowest rods from ten to four. Neither proposal is acceptable and the requirements of Regulatory Guide 1.68 must be observed, or sufficient justification presented for an alternative proposal.
212.96 Table 9.3-2:
The rated head of the makeup pump for HPI require-(4.2.3.2) ments (2600 ft) does not appear to agree with Figure 6.3-2 (3400 ft).
Clarify the apparent inconsistency.
212.97 Show that placing trip setpoints at maximum tolerances produces (5.2.2) identical or conservative peak pressures compared to analyses which also assume worst case initial values for trip parameters. Discuss the rationale for not assuming both conservatisms; for example, I
placing the pressure trip setpoint at maximum tolerances to reflect setpoint uncertainty plus assuming the worst case initial pressure allowed by technical specifications, also including instrument uncertainties.
212.98 The staff review of topical report RAW-10043, Supplement 1 is not (5.2.2) complete. Should modifications to these analyses be required from the staff review, they must be considered for applicability to BSAR-205. State your commitment to modified analyses if required.
212.99 NB-7300 of the ASME Code,Section III requires that an analysis be (5.2.2) conducted to indicate the redundancy of pressure-relief devices-employed to preclude a loss of overpressure protection in the event of a failure of a pressure-relief device. BAW-10043, Supple-cent 1 indicates that a 4% capacity margin is available after the worst-case steam side transient.
Since up to 6 capacity is provided on a single steam safety valve, it appears that insufficient capacity is available to support compliance with the ASME Code.
Justify this position.
f W
212-3 t
212.100 B&W states that the analytical results would not change signi-(5.2.2) ficantly as a result of the small disparity in power level between B-SAR and BAW-10043, Supplement 1 (3800 MWe versus 3760 MWe, respectively).
Nevertheless, the analysis wouldinot be appropriate to meet the Code requirements for a 3800 MWt reactor unit. Also, the BAW-10043 analyses should have been performed at 1.02'x 3800 = 3876 MWt.
Furthermore, this B&W statement conflicts with the B-SAR loss of feedwater event which shows a peak pressure of about 2740 psia compared to the 2665 psia depicted in BAW-10043.
This means that B-SAR shows smaller margins to the ASME Code limit
'(110% of design pressure) than the so-called sizing analysis. Please address these comments.
212.101 Verify that the lower power units (i.e., 3600 MWr) will use the same (5.2.2) number and capacity of pressure-relieving devices as the 3800 MWR units.
If not, a complete new sizing analysis is required.
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212 102 Table 5.1-2 states that the operating pressure in the pressurizer (5.2.2) is 2195 psig. Confirm that this value is the highest relative to other locations in the reactor coolant system'(the transients in
{
Chapter 15.0 start at 2235 psig).
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l 212.103 The response to question 212.36 (2/1/76) regarding the potential (5.2.2) for,a solid pressurizer is insufficient to allow an adequate I
evaluation. The inquiry was directed at all Chapter 15.0 events.
4 Also, please restate and clarify the last sentence in your response.
212.104 With regard to all overpressure transients in Chapter 15.0, identify (5.2.2) the operating conditions and setpoint assumptions which'are different from those used in BAW-10043, Supplecent 1.
212.105 Why was considerr tion of the loss of ac power as a sizing transient (5.2.2) deleted in Amendment 1 (page 5.2-8)?
212.106 The response to question 212.42 (4/1/76) is partially acceptable.
(5.2.2)
Explain why the backpressure limit is based on the bellows design pressure if a failure of this component has no affect on valve capacity.
212.107 Page 5.2-22 states that monitoring makeup tank level would take (5.2.7) approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to detect a 1 gpm leak.
Regulatory Guide 1.45 states that all detector systems for unidentified leakage should respond to a 1 gpm leakage increase in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or less; therefore, the makeup tank level response time is not adequate to serve as a qualified leakage detection system. To clarify the BSAR presentation, you should include a statement that the makeup tank level is not con-sidered one of the design basis leakage detection parameters.
1 i
212.108 Conformance to Regulatory Guide 1.45 should be presented as an l
(5.2.7) interface require =ent.
l 212.109 Enclosures 1 and 2 present the~ staff position on the design of decay
)
l (5.5.7) heat removal systems. Address these positions and show that the l
B&W design meets these criteria.-
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212-4 212.109 If B&W has selected option 2d in Enclosure 1 for their design, (cont'd) the procedure to be used to confim the integrity of the ch'eck valves inside containment must be perfomed at or near normal reactor coolant operating pressure. The procedure described in the response to question 212.59 is not adequate. Label the inboard check valve in Figure 9.3-5.
212.110 Confim that the decay heat removal sizing data shown in Table 9.3-8 l
(5.5.7)
M 11 be adopted on both the 3600 MWe and 3800 MWt plants.
I 212.111 Note #4 on Figure 9.3-5 states that manual (local) valves MV-24A l
(5.5.7) and MV-24B must be locked closed before opening the DHR suction valves DH-12A or DH-12B when placing the' system in the nomal DH l
mode. These valves (MV-24A and MV-24B) must be remote-actuated j
from the control room.
212.112 Explain the reason for note #6 on Figure 9.3-5 which states that (5.5.7) valves V14A and V14B must be locked closed during nor=al reactor 4
operation.
212.113 Using the syste=s identified in Table 5.1-10, conduct a failure (5.5.7) modes and effects analysis of all equipment and components necessary to achieve the cold shutdown condition. Show that the' plant could achieve and maintain a cold shutdown from the control room, in spite of a single active cc=ponent failure.
212.114 Provide a typical step-by-step procedure that the operator would (5.5.7) follow to achieve a cold shutdown condition from full power operation.
Include estimates of the time at which each step is initiated (i.e., + 1 hr, +3 hrs, etc.).
212.115 With regard to Table 5.1-10, provide the following additional (5.5.7) infomation:
(1) How long are the RC pu=ps needed for heat removal & chemical mixing (i.e., are they not needed for Tavg< l50 F7).
(2) What " chemical mixing" is necessary during shutdown?
(3) Why is " purging the MU tank" required during shutdown?
(4) Aren't the pressurizer heaters necessary for pressure control?
The response to question 222.6 (4/1/76) indicates that the i
pressurizer heaters are required to achieve and maintain a safe shutdown condition.
(5) What is the difference between purging the MU tank and degassing?
(6)
Is water purity control used only for refueling?
I i
212-5 212.115 (7) What operations are performed outside the control room?
(5.5.7)
(cont'd)
(8) Which systems which are listed as not having an alternative method to perform the needed functions are vulnerable to a single active component failure? For the systems vulnerable to a single failure, what operator actions are possible to attain a cold shutdown condition and how much time is the action pre-dicted to consume in arriving at a cold shutdown condition?-
1 (9) For each system which is vulnerable to a single active com-l ponent failure, but for which there is an alternate method listed for performing the required function, provide an estimate i
of the total time to achieve a cold shutdown condition with the l
single failure.
j
' 212.116 Assuming shutdown cooling is using only one train, provide an esti-(5.5.7) mate of the time to decay heat pump malfunction (overheating) af ter a spurious closure of a motor-operated suction line valve.
Specify the alarms which would alert the operator of this situation.
i 212.117 The responses to questions 212.51 (4/1/76) and 212.60 (4/1/76) j (5.5.13) are partially acceptable. All of the safety and relief valves covered under NB-7000 of the ASME Code must be addressed. Expand the discussion in subsection 5 5.13 to provide typical design l
diagrams for each safety and relief valve type within the nuclear j
power plant.
Include such information as typical valve components and typical valve sizes. Attach a list of the major components in each valve type with a sum =ary of the component function.
Provide an explanation of the operation of each valve type.
Include a presentation on the reliability of these pressure-relieving devices including a summary of all related abnormal occurrences.
212.118 Include a discussion on tests and inspections of safety and relief (5.5.13) valves.
Specifically address Regulatory Guide 1.68 and ASME Code,Section III requirements.
212.119 Specify the back ressure assumptions for all safety and relief (5.5.13) valves (except p' ssurizer) and provide the bases for these values.
212.120 Page 6.3-1 references BAW-10102 (currently under review by the staff)
(6.3) as documentation of ECCS performance for B-SAR-205 plants.
Section 10 of this report describes a multi-mode design concept to prevent boron precipitation during the long term after a LOCA. In a letter from John F. Stolz to Kenneth E Suhrke dated January 30, 1976, B&W was requested to provide alternate means to prevent boron precipitation for the 205-FA plants. Submit this design proposal, including the expected operator action needed to actuate the system from the control room.
212-6 212.121 In Figure 9.3-1, the HPIS suction label apparently should be (6.3) changed to read "FROM BWST" instead of "FROM BWST VIA DH SYS."
Please clarify.
212.122 Figure 6.3-7:
Re-submit this figure in three diagrams (6.3-7a (6.3) 6.3-7b and 6.3-7c). These diagrams are to depict the positions i
of all valves during a) normal reactor power operation, b) ECCS injection mode, c) recirculation mode.
Include line sizes and valve numbers on these simplified diagrams.
212.123 Page 6.3-3 states that all ECCS pump components are designed for (6.3.2) an accumulated radiation dose of 10' rads without degradation.
What is the total expected accu =ulated dose over the 40 year plant life?
212.124 Provide a list of industry valve failures (comparable to the. B&W l
(6.3.2) valve designs) attributed to boric, acid crystallization. Discuss whether the double-lantern gland, seals reduced valve leakages (based i
upon operating experience).
Describe the role of the gland seal leak-off provision in reducing the potential for boron crystallization.
212.125 Page 6.3-4 states that all ECCS pump motors are air cooled. Provide, (6.3.2) or reference, the sizing design bases for these air cooling provisions, including a rationale for all input assumptions to the sizing analysis.
Submit an estimate of the time availab3e without motor cooling for the worst-case motor loading after a LOCA. Discuss, or reference, the impact of single failures in air cooling systems during the short term and long term after a LOCA. Also, indicate whether the heat load from these motor coolers will be factored into the sizing of the room coolers.
212.126 Explain the need for the new low steam generator pressure ESFAS Gable actuation signal shown in BSAR-205, but which was not included in
- 1. 3-1 )
previous NSSS designs.
& 1. 3-2 )
212.127 With regard to the ECCS relief valves discussed in Table 6.3-6 (6.3.2) and section 5.5.7, provide the following additional information:
(1) Relief Valves DH-RVSA & SB: Provide plots of pressure versus time for each of the transients considered.
Reference the figure from which the 238 psi shutoff head was selected (Figure 6.3-3 shows about 260 psi shutoff head). Specify the uncertainty (i psi) in this value.
The maximum allowable DH system transient pressure is 110% x 675 psig = 742 psig.
The conditions on page 9.3-26 produce 738 psig, a margin of only 0.54%.
What are the uncertainties associated with these relief valves (i psig)?
4
212-7 i
. 127 (2) Relief valves DH-RV4A, 4B, 7A, & 7B: Provide more detail
.3.2) on the sizing basis for these valves.
Include a discussion j
- ont'd) of pressures and specify relief valve uncertainty. What ambient temperature changes are considered and what technical specification leakage limit is associated with the closed DHR valves referred to in Table 6.3-62 (3) Relief valves CF-RV 1A & IB: Provide the peak pressure
{
attained for the sizing event (maximum makeup fill rate).
Consider also nitrogen overfill situations. Show that the peak pressure would not exceed 110% of the system design pressure.
212.128 Discuss the application of NB-7421 of the ASME Code to the relief (6.3.2) valves in ECCS systems.
212.129 Discuss how backpressure effects were considered for each of the (6. 3. 2)
ECCS relief valves.
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212.130 The LPI capacity on Figure 6.3-5 and Table 6.3-4 reflects an LPI (6. 3.2) flow rate of 5000 gpm. Table 9.3-8 states 5125 gpm. Explain the difference.
t r
e
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212-8 212.131 The response to question 212.56 (4/1/76) is not sufficient to (6.3.2) allow an adequate evaluation. Provide the c'alculational methods routinely employed to determine available NPSH.
Provide all equations and definitions of terms. Justify all assumptions basic to a typical NPSH calculation. Provide a sample calculation clearly showing the implementation of the methods. Also, a commitment to conformance to Regulatory Guide 1.1 should be provided.
Pages 9.3-31 'nd 9.3-33 indicate that adequate NPSH will be 212.132 a
8 (6.3.2) available at the pump suction to permit runout ECCS flow. This j
is an acceptable interface requirement; however, Table'6.3-8 l
implies that the criterion should be design flows, which is not
{
consistent with pages 9s3-31 and 9.3-33.
Provide the NPSH requirements for ECCS pumps using runout flow conditions.
212.133 Since it is desirable to minimize the need for operator actions I
(6.3.2) during a LOCA, it is preferred that manual 1hrottling or re-positioning of valves to ensure that sufficient NPSH is available to ECCS not be necessary. Confirm that such operator action would not be necessary, or modify your design accordingly.
212.134 Venturi testing: Provide a simplified diagram of the test setup (6. 3. 2) used to obtain the data shown on page 6.3-4a.
Show how the data confirm that the limiting flow rate through the venturi varies as the square root of the difference between the inlet pressure and the fluid vapor pressure.
212.135 Confirm that the minimum CFT water volume allowed by Technical:
(6.3.2)
Specifications is the same value assumed in the BAW-10102 ECCS calculations.
212.136 Why is the CFT water at 1800 ppm boron (Table 6.3-1) and the BWST (6.3.2) water at 2270 ppm (page 9.3-33)?
212.137 Page 6.3-5 states that the sizing criteria for-che BWST are given (6.3.2) in Section 9.3.6.7, which does not exist. Provide the missing info rmation.
212.138 Is Figure 6.3-7 applicable to both size B-SAR-205 plants (3800 MWt (6.3.2) and 3600 MWt)?
212.139 Include the CFT line break in Table 6.3-4.
(6.3.2) 212.140 Page 6.3-7:
Explain how the spurious opening of V5A or B, or (6. 3. 2)
V2A or B, could compromise the availability of a CFT.
212.141 Discuss and identify on Figure 6.3-7 the operation of a' mini' um m
(6.3.2) recirculation flow bypass line on the discharge of each EPI pump.
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212-9 i
212.142 Page 9.3-14: The statement is made that, "the piping losses (6. 3. 2) from the HPI pump to the RCS, as seen by one HPI pump, shall not exceed 500 psid, including all valves." What is the basis for this value?
f.
212.143 Question 212.57 (,4/1/76): Page 9.3-14 states that, "To preclude (6.3.2) water hammer, the suction lines to and discharge lines from the ECCS pumps shall be filled prior to startup." To further minimize the potential for a water hammer, the staff requires that a continuous supply of water (static head from storage tank or dynamic means) be provided to the ECCS lines to preclude water discharging into a dry line.
In addition, venting provisions in ECCS line high points and pump. casings shall be provided. Provide i
i your commitment to these design provisions.
212.144 Table 6.3-7 identified a CFT relief valve malfunction as a credible (6.3.2) single active component failure, yet does not postulate the con-l sequences should the failure occur at the time of a LOCA.
Rationalize this position. Discuss the possible failure mechanisms which could cause a failure (open).
Provide a discussion of observed industry failures Copen) of this type of relief valve i
and address these failure mechanisms. Confirm that these valves j
cannot be manually actuated.
212.145 Page 6.3-2 states that SEPARATE and REDUNDANT flow paths are (6. 3. 2) provided in each subsystem of the ECCS, yet page 6.3-6 states that the HPIS lines are joined by crossconnects inside the reactor building. Justify how such common junctions do not compromise the l
physical independence of the two HPI trains. Why were these crossconnects-left out of Figure 6.3-77 I
212 146 The response to question 212.19 (2/1/76) is insufficient to allow (6.3.2) an adequate evaluation. The statement is made that calculational results based on analytical methods presented in topical report BAW-10074 demonstrate that no large pressure differences would exist among the HPI points (between broken and unbroken cold legs) and that the pressure difference is small compared to the flow resistance in the HPI line. BAW-10074 does not adequately demonstrate that no large pressure differences would exist among the HPI points.
Provideanalysestoshowtimehistorigsofthe2, pressures at the four HP injection points for a 0.5 ft, 0.2 ft dnd 0.04 ft2 break at the reactor coolant pump discharge where an HPI line joins a cold leg.
~
Did the ECCS design assumed in BAW-10074 consist of the same size ECCS and same high pressure injection configuration (non-isolatable crossovers)?
Submit or refer to test results which support B&W's predictions that appropriate flow splits would occur across the small break spectrum.
Relative to 177-FA plants, explain the necessity for the change in ECCS design (non-isolatable crossovers).
212-10 t
212.147 It appears that a break in the normally pressurized makeup line (6.3.2)
(4-inch) could result in the unavailability of all core cooling td capability for an unknown period of time.
If normai letdown flow is not stopped quickly, the condition could be made worse in terms of available water mass over the core.
Loss of all normal makeup (break between valves MV77A and MV778) and the unavailability of all HPI due to a single active component failure (pump PIC) requires the operator to take complete control of the event to bring the plant to a safe shutdown condition.
Should such an event occur, what is the first alarm or trip available to leave little doubt in the operator's mind as to the seriousness of the situation? What additional alarus and trips g
would follow the initial alarm, assuming no operator action?
Provide estimates of the times of each of the preceding ~ alarms and trips after the pipe break.
I Provide a table depicting the sequence of events (and estimated times) involved in bringing the plant to a safe shutdown condition (operator action beginning with first alarm)..
Submit time histories of RCS pressure and RCS vater level.
Examine this event for a potentially more severe single active component failure (in lieu of the HPI pump malfunction).
For example, is there an active component malfunction or single operator error which would allow RCS pressure to increase above the shutoff head of the HPI pump, thereby rendering HPI unavailable during an RCS pressure excursion?
212.148 Table 6.3-7:
Provide the component numbers to facilitate their (6.3.2) location on the submitted diagrams.
212.149 TLa single passive failure analysis in Table 6.3-9 cannot be i
(6.3.2) easily followed.
To facilitate an understanding of B&W's effort, the following additional information is requested:
(1)
Correct the reference to Figure 9.3-6 in the footnotes.
l (2)
Docs this evaluation also cover more probable small break LOCA's?
(3) To ease a correlation of Table 6.3-9 to the ECCS design, provide a simplified sketch of the piping location or these assumed passive failures (similar to Figure 6.3-7).
1 (4) Table 6.3-9:
Valve HV9 could not be found on Figure 9.3-1 nor 9.3-5.
Describe the identification scheme whereby such nomenclature can be related to the diagrams (i.e., explain HV, MV, V, etc.).
1 212-11 j
I 212.149 (5) Passive failure location 3 indicates that the operator is t
(6.3.2) to terminate flow by shutting MV34, yet MV34 is shown to be (cont'd) a stop check without a hand jack operator. Please explain l
inconsistency.
(6) Submit the calculations upon which the crack flow rates are I
based.
i i
(7) Why was CCW flow mentioned in passive failure 17
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212.150 The core flooding system discharge valves (Figure 6.3-1) are i
(6. 3.2) numbered VIA and V1B. The DHR injection valves are also numbered VIA and V1B (Figure 9.3-5).
Such a numbering system is confusing and should be reevaluated and corrected on all P&ID's.
t l
212.151 It is noted that the transfer to the recirculation mode is not 8
2 (6. 3.2) completely automatic (operator must close BWST valves and trip off the HPI pumps). What are the consequences if the operator fails to follow th<se manual procedures? The need for this manual action is not consistent with Figure 15.1.13-1 which shows that all the valve actions are automatic.
212.152 With regard to question 212.23 (2/1/76) part 2, the statement is (6.3.2) made (to justify manual " piggy-back" initiation) that the operator will have at least one hour to make his decision. No mention is made of the small break " window" which could also involve partial low pressure injection flow, thereby not only allowing less time j
for the operator to evaluate the situation, but also providing i
less obvious clues to the proper course of action. As page 6.3-10a makes clear, a failure of the operator to provide a required " piggy-i back" operation could result in an interruption of core cooling at j
the time of automatic switchover to the sump. Comment on the p'ossibility of any adverse effect on core cooling across the break spectrum of including the HPI pumps in the automatic transfer of LPI pu=ps from the BWST to the sump.
212.153 Table 6.3-7 alludes to a situation requiring operator action (6. 3. 2) within 40 minutes or core uncovery will occur.
Provide a more detailed description of this event and required manual actions to facilitate an understanding of the potential consequences. Submit all calculations upon which the 40 minutes are based. List and justify all assumptions. Why isn't this break protection sequence shown in Figure 15.1.13-3 or discussed in Section 6.3.2.177-212.154 Table 6.3-7 (LPIS) should address the remaining LPI valves (6.3.2)
(injection valves, throttle valves, etc.).
Similarly, the CFS should address all other motbr-operated valves. If certain valves have power disconnected, they should be so identified and their spurious actuation should be specified as either not credible or not applicable.
212-12 j
i-1 212.155 Table 6.3-7 OCFS) should address the failure of a discharge (6. 3. 2) isolation valve to close during a' shutdown.
k 212.156L Page 6.3-15 states that no operator action is required'during
~
(6.3.3) the interval between automatic switchover and BWST exhaustion (maximum depletion rate). Comment onLslower depletion rates-I i
and the piggy-back mode (i.e., is this statement applicable across the break spectrum and for a variety of depletion rates?).
212.157 Page 6.3-12 states that, "LOCA analyses have shown that one makeup (6.3.3) pump is sufficient to prevent core damage for the smaller leak 4.
sizes that do not allow the RCS pressure to decrease rapidly to the point at which LPI is initiated." Justify this statement for a variety of break sizes in a CFT line.
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212.158 Page 6.3-16 indicates that the minimum volume of the core' flooding (6. 3. 3)
. tank is given in Table 6.3.1.
Provide, or reference..the minimum-volume. allowed (not NORMAL volume) and submit the criteria upon which this minimum volume is based.
212.159 With regard to the responses to questions 212.68 (4/1/76) and (6.3.4) 212.70 (4/1/76), conduct of a sump test to verify calculated NPSE~
}
and vortex margins on the prototype plant-only is not acceptable.
l Subtle changes in sump geometries, for example, may have an adverse effect on vortex margins. Each applicant referencing BSAR-205 must comnit to testing according to Regulatory Guide 1.79.
1 212.160 Preoper2tional verification of the manual backup (local) _
(6.3.4) capability of each ESF motor-operated valve will be required.
i 212.161 What is the B&W separation criteria in terms of minimum separation (6.3.6) distance requirements (i.e., feet between trains)?
1' 212. 162 Provide, or reference, a specific interface requirement with (6.3.6) regard to installation of equipment below the predicted' post-LOCA flood level inside the containment.
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212-13 ENCLOSURE 1 TO REQUEST 212.109 Regulatory Position on RHR System Isolation Recuirements A.
Background
Two motor-operated valves in series form the boundary between the reactor coolant system (RCS) and the suction side of the low pressure residual heat removal (RHR) system. For many years the NRC staff has required that there be interlocks which automatically close these valves whenever the RCS pressure increases to a value that approaches the design pressure of the RHR system.
I In nearly all plants the RHR system is located outside of containment.
Further, in most of these plants, the ECCS and RHR system share many co=ponents and pipes. In plants with this design the consequences'of over-pressurization of the RHR system would be likely to have a severe conseq'uence because:
l (1) The RHR system usually has a design pressure of about 600 psig.
j If exposed to the reactor coolant normal operating pressure of'
)
about 2,200 psia, rupture of the RHR system is probable.
(2) Since the RHR system is located outside of containment, rupture of the RHR system due to exposure to normal reactor coolant system operating pressure will result in a LOCA outside of containment.
(3) Since the RHR system shares many components and pipes with the ECCS, a rupture of the RHR system could render redundant ECCS trains inoperable.
J J
Because of the potentially severe consequences of this postulated event, i
we require a high degree of assurance that the low pressure RHR be i
j isolated from high pressure in the reactor coolant system. Our position regarding this isolation requirement is presented in Sections B and C.
j below.
5 For the purposes of this position, two types of low pressure RHR systems i
are defined, Type A and Type B.
Type A RHR System To qualify as a Type A system, the RHR system'must satisfy all of the following requirements.
(1) The RHR system is located entirely inside of the primary containment, (2) Overpressurization of the RHR system shall not cause consequential failure of any connecting systems that are located outside of containment.
That is, connecting syste=s located outside of containcent must be designed such that in event that the RHR system fails because of overpressurization:
7
)
212-14 4
The consequential failure of the connecting system a.
will occur only in the portion of the system located within the containment, or b.
The consequential failure of the connecting system
{
will be isolatable so that reactor coolant will t
l not be discharged outside of the containment, despite any postulated single failure of an active fluid system component er an active or passive l
electrical component.
(3) Structural failure of the RHR system, or of any connecting systems located within containment, due to overpressurization of the RHR system shall not reduce ECCS capability to a point l
where the ECCS cannot mitigate the consequences of the postulated RHR system or connecting systems failure, despite any postulated l
single failure, and (4) Structural failure of the RHR system due to overpressurization shall not generate misailes that consequentially cause or induce any system or component failure that would increase the severity of the LOCA event or reduce the performance capability of any other engineered safety feature (e.g., containment spray system, ECCS, containment integrity).
Tyce B RHR System Any low pressure HHR system that does not satisfy the requirements of a Type A system is defined as a Type B system.
B.
Tyce A RHR System Isolation Recuireeents 1.
Isolation of the suction side of the RHR system from the RCS shall be provided by at least two power operated valves in series.
The valve positions shall be indicated in the contral Alarms in the control room shall be provided to alert room.
the operator if either valve is open when the RCS pressure exceeds the RHR system design pressure.
The valves shall have independent diiterse interlocks to prevent the valves from being opened unless the RCS pressure is below the RHR system design pressure. Failure of a power supply shall not cause any valve to change position.
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4 213-15 i
One of the following shall be provided on tne discharge side 2.
of the RHR system to isolate it from the RCS.
The valves, position indicators, interlocks, and alarms a.
described in item B.I.
One or more check valves in series with a normally closed i
b.
I power operated valve.
The power operated valve position If the RHR system shall be indicated in the control room.
j discharge line is used for an ECCS function, the power operated valve is to be opened upon receipt,of a safety injection signal once the reactor coolant pressure has decreased below the ECCS design pressure, I
Three check valves in series, or c.
Two check valves in series, provided that there are design f
d.
provisions to permit periodic testing of the check valves for leak tightness and the testing is performed at least annually.
C.
Tvoe A RHR System Isolation Recuirements 7ne following shall be provided in the suction side of the RHR system to isolate it from the RCS.
Isolation shall be provided by a least two power operated valves 1.
The valve positions shall be indicated in the control in series.
room.
The valves shall have independent diverse interlocks to prevent 2.
the valves from being opened unless the RCS pressure is below the RHR system design pressure.
Failure of a power supply shall not cause any valve to change position.
The vaIves shall have independent diverse interlocks to provide 3
power actuation to automatically close each valve if the pressure in the RCS increases above the design pressure of the RHR system.
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212-16
-4 One of the following shall be provided on the discharge side of the RHR system to isolate it from the RCS.
1.
The valves, position indicators,-and interlocks described in items C.1, C.2, and C.3
}
2.
One or more check valves in series with a normally closed power l
i operated valve.
The power operated valve position shall be indicated in the control room.
If the RHR system discharge l
line is used for an ECCS function, the power operated valve l
is to be opened upon receipt of a safety injection signal once the reactor coolant pressure has decreased below the ECCS design if i
- pressure,
[
3 Three check valves in series, or i
4.
Two check valves in series, provided that there are design provisions to permit periodic testing of the check valves for leak tightness and the testing is performed at least annually.
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212-17 ENCLOSURE 2 TO REQUEST 212.109 Ferulatory' Position on Residual Heat ' Removal The system or systems employed to remove residual heat shall satisfy the following functional requirements:
(1) The system (s) shall be capable of performing the function of transferring heat from the reactor to the environment using only safety grade systems.
These systems shall satisfy the requirements of General Design, Criteria 1 through 5.
}
(2) The system (s) shall have suitable redundancy in components j
and features, and suitable interconnections, leak detection, and isolation capabilities to assure that for onsite-i electrical power system operation (assuming offsite power, is not available) and for offsite electrical power system' operation (assuming onsite power is not available) the system safety function can be accomplished, assusing a single failure.
(3) The system (s) shall be capable of being operated from the I
control roon with either only onsite or only offsite power j
available with an assumed single failure.
In demonstrating l
that the system can perform its function assuming a single failure, limited operator action outside of the control,
room would be considered acceptable if suitably justified.
(4) The system (s) shall be capable of bringing the reactor to a cold shutdown condition *, with only offsite or onsite power available, within a reasonable period of time following shut-down, assuming the most limiting single failure.
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' Cold shutdown is defined by the Standard Technical Specifications; that is, reactor suberitical and reactor coolant temperature no greater than 200 F.
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I 222-1 i
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220.0 ANALYSIS S ^*
., 222.1' With regard to the mass and, energy release calculations for the
~ (6.2.1) containment subcompartment analysis discussed on.page 6.2-1, pro-
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vide the following information.
l
~ ). ?The, version of the CRAFT code referenced for these calculations l
a
'he Moody corre1'ation to calculate the flow at the break.
4
., do not believe that this approach is sufficiently conserva-1 tive for the calculation of subcooled flow rates. We believe l.
the modified Zaloudek correlation used in later versions of l
the CRAFT code is accep' table for this purpose. Using an ac-l ceptable break flow model, provide mass and energy release I
rates in rabular fons,' from time zero to about 1.5 seconds at l
i approximately.05 second Intervals, for the following pipe
\\
failures.
1.
double-ended hot leg; 1
2.
double-ended pump suction break, 3.
double-ended pump discharge break, 4.
double-ended pressur,1zer, surge line, 5.
double-ended pressurizer spray line, 4
~
6.
double-ended flood tank line, 7.
double-ended steam line, s
8.
double-ended feedwater, lines.
b.) Using the above method, provide the results of a sensitivity study in which thd'noding.in the piping adjacent to the break 5
1 4
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222-2
,e 222.1 is increased until a convergent solution of break flow is Cont.
obtained. Justify the break noding of each of the above postulated piping failures.
c.) 1he above analysis should be base'd on a r'edcto'r power, of N
3800 MWt with an additional 2% increase in power level to account for instrument error.
)
1 222.2. For the long term mass and energy release calculations discussed l
(6.2.1) l in section 6.2.1.3, there are two sets of mass and energy release f
information. The first set in tables 6.2-6 through 6.2-18 appear I
to assume complete quenching of steam by the injected ECCS water.
l The second set in tables 6.2-19 through 6.2-22 assu=e no quenching of steam by ECCS water. Discuss which of these data sets are pro-vided as interface information to the BOP for use in sizing the containment safety systems. Discuss the purpose of the other data set.
If the above data without steam quenching is to be the interface 222.3.
(0* *1) information to be used in the BOP design, provide the mass and energy release for a complete spectrum of pump suction breaks.
Also, include the mass and energy release data for A double-ended break at the pump discharge. These analyses should be based on a power level of 3800 MWt with a 2% overpower allowance to account for instrument error.
For the double-ended pump suction break, provide the average core teeperature at the end of blowdown.
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222-3 i
I 222.5.
In the long term r. ass and energy release calculation of section (6.2.1) l 6.2.1.3., the CRArt code is used to calculate mass and energy
(
releases in the reflooding period as well as the blowdown period.
6 3 ~
You state that a containment volu=e of 3.4 x 10 ft is used to compute the containment pressure for the reflooding part of these f
calculations.
'a., For the double-ended pucp suction break, provide the contain=ent pressure calculated for the reflooding analysis as a function of time.
b.
BOP containment designs may have higher contain=ent pressures than the values utilized in your analysis.
During the reflooding period, higher containment pressures will produce less steam binding in the coolant loops which may increase the mass and energy release i
rate to the containment. Provide a sensitivity study showing mass and energy release to the containment as a function of assumed con-tainment pressure and discuss the effect on the containment pressure that would result from these releases.
222.6 During a LOCA when the primary fluid te=perature decreases below (6.2.1) the secondary fluid temperature, reverse heat flow in the steam generators will produce boiling in the. primary system which will provide additional steam to the containment. A reverse heat transfer multiplier of 0.1 was assumed in B-SAR-205. Provide justification for this asse=ption by comparison with the appropriate
~
heat transfer-correlations on the primary and secondary side of the tubes.
Include consideration of, steam condensation in the upper region of the steam generator (secondary side) as a result of this process.
4
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222-4 222.7 Compare the carryout fraction from the core calculated by the (6.2.1)
CRAFTcodeduringtherefloodingperiodtothedatafromthe}ECHT experiments and justify the values calculated by CRAFT.
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222.8 The mass and energy release data in B-SAR-205 for subcompartment (6.2.1) analysis and containment design basis analysis,was, calculated based i
on a core power of 3760 K4e and adjusted to a power.of 3800 E4t.
1 f
Provide a detailed description of how this adjustment was made l
for both subcompartment and DBA analysis.
222.9.
The following questions concern the calculation of mass and energy (15.1.14) release to the containment for a main steam line break described in section 15.1.14.
nese calculations were performed using the l
TRAP code.
i s.)
Provide the assumptions made for steam separation in the secondary steam side of the generator. Discuss the conservatism of these assu=ptions for containment analysis, b.) Table 15.1.14-2 lists the initial energy content of one steam 6
generator at 29.93 x 10 BTU. Table 6.2-23 give's ths initial fluid energy content of two steam generators at 87.9 x 10 BTUs or 44 x 10 BTUs per steam generator. Discusu this apparent inconsistency.
c.) Auxiliary feedwater flow to the ruptured steam generator will provide an additional long-term source of steam to the containment.
Discuss how auxiliary feedwater is' treated in your analysis.
d.)
Provide the input constants and assumptions used to calculate reverse heat flow in the intact. steam generator.
231-1 230.0 CORE PERFORMANCE 231.0 Reactor Fuels 231.1 The fuel handling and shipping design loads are provided.
(4.2.1.1)
State the baser. for these design loads. Describe the extent to which these design loads have been confirmed experimentally.
State wnether the relationship between these design loads and stress-strain limits is such that no design limit:: are
{
exceeded during handling and shipping.
I 231.2 Specify yield and ultimate stress values as a function of (4.2.1.1) temperature and irradiation. Also, specify design limits with regard to fretting wear and deflection.
231.3 Provide the nunerical values used for the Zircaloy cladding (4. 2.1.1) yield strength and ultimate tensile strength mentioned in conjunction with the stress intensity limits. In addition, i
state the cladding thermo-mechanical history, associated tenperature, Last neutron flux and fluence for which the stress limits apply.
l 231.4 The relationship between cladding co=pressive load. coolant (4.2.1.1) temperature, and zirconium hydride precipitation is dis-j cussed in Section 4.2.1.3.2.
In Section 4.2.1.1.2, describe how the specified coolant temperature limits and associated cladding loadings are used in design basis calculations of cladding stresses.
231.5 Provide a drawing that shows the details of the s7acer (4.2.1.2) grid at the instrument tube location. Discuss host the spacer sleeves restrict the movement of the spacer grid.
231.6 Provide the deflection design value's (including nominal (4. 2.1. 2) dimensions and spring constants) and experimental observa-tions for the upper and lower plenum springs. Describe any evidence of permanent deflection due to fuel rod and fuel assembly handling. State whether gradual deflection of the lower spring would be expected as a function of irradiation and state the mn4== deflectien, both expected and possible.
Discuss the QC procedures that assure that the proper type of spring is in the lower plenum.
231.7 Provide the design bases for Zircaloy-4 irradiation growth (4.2.1.3) and supply supporting data or references.
o 231-2 4
231.8 List fuel rod deficctions and cladding strain limits and (4.2.1.3), provide justification for their adequacy. Demonstrate how r
these criteria are satisfied during steady state, transient and accident analyses. Your responses should provide a i
comprehensive network for an overview of the design basis in this area.
Include in the discussion a sum =ary of the safety analysis in the form of a stress report for each component under specified loadings.
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,. Discuss how the different loadings categories are combined j
to satisfy the design limit for each ce=ponent of the fuel assembly.
Discuss the design basis load (s) and their justification for each component of the fuel assembly.
231.9-Provide tables of numerical valbes (or equations) of material (4. 2.1. 3) properties of the cladding Snd fuel pellets where specified)as functions of temperature and irradiation.
The following prop-ertf.es should be included:
I (1) Modulus of elasticity (2) Poisson's ratio (3) Thernal expansion coefficient (cladding and fuel pellets)
(4) Yield stress (5) Ultimate stress
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(6) Uniform ultimate strain (7) Thermal conductivity (cladding and fuel pellets)
(8) Specific heat (cladding and fuel pellets)
I 231.10 Describe procedures used for sizing the fuel rod plenum, in-(4. 2.1. 3) cluding any computer codes used and the fission gas release rate assumed.
State whether this volume is adequate for accidents and transients in which the fuel might reach a temperature in excess of the design temperature. Also, de-scribe how creep effects and dimensional stability are accounted for in designing the fuel rod plenum. State whether it is possible for the cladding temperature to become so high in a transient, that cladding swelling would occur due to internal pressura. State the end-of-life internal pressure for these fuel rods (both average burnup and peak burnup). State the temperatures assigned to each of the following void regions when determining the fuel rod pressure:
(a) fuel rod upper end plenum (b) fuel-cladding annulus (c) fuel pellet end dishes (d) fual pellet opep porosity t
231-3 231.11 Discuss the values used for the following parameters in your (NONE) analysis of flow induced vibration, and describe any additional R&D required to support your selection of values:
~
(1) natural frequency limitation of the fuel assembly, (2) natural frequency relative to primary system frequency, and (3) stiffness limitations on the spacer-grid assembly and individual grid spring.
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231.12 Describe the corrective actions indicated in the last j
(4.2.1.3.2) sentence of the paragraph titled, " Potential for Water Logging Rupture."
l 231.13 Discuss all procedures used during fabrication to assure (NONE) that no axial gapa are introduced during the loading of the i
fuel rods, such as weighing, counting pellets, and fluoroscopic examination.
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231.14 Internal and external surfaces of the Zircaloy tubing are I
(NONE) cleaned with dry cotton swabs and acetone saturated cloth, i
respectively. Discuss the effect of residual lint on the internal surfaces upon the fuel performance, and specify acceptable Itaits, analyzed results and ultimate chemical disposition.
231.15 DELETED (NONE) 231.16 Provide steady state, transient and accident responses for (NONE)
'the guide tubes in terms of stress, strain, di=ensional stability, and deflection.
231.17 Evaluate the effects of fuel rod bowing together with spacer (4.2.1.3) grid response.
Include time dependent behavior due to creep in your evaluation.
231.18 Discuss the analytical calculation of fuel / cladding mechan-(4.2.1.3) ical interaction that is used to describe the design basis.
A detailed, complete description is needed, including a general description, assumptions, mathematical equations, sequence of application of equations or a flow chart, all empirical constants used in the equations, 2 sample calcu-lation and a comparison with data.
Include the effects of:
(1) fuel swellivg driven cladding strain (specify fuel swelling assu=ptions used),
(2)
"bambooing" of the cladding due to fuel pellet end effects, and (3) radial and/or axial differential thermal ex-pansion of the fuel and cladding.
_=
231-4 i
231.18 (cont.)Emplain how a transient in which fuel rod power was in-7 (4. 2.1. 3) creased would affect the fuel / cladding mech'anical interac-j tion and how this is taken into account in the model de-scribed above. Discuss operating design limits (transient conditions) or total lifetime limits based on calculated fuel cladding mechanica1' interaction effects.
231.19 Discuss.the fuel assembly seismic model and analysis method.
t In particular, describe a method of obtaining detailed j
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stress and deformation of the fuel rod from the simple spring-mass beam mode response.
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231.20 Identify all velded joints in the assembly and categorize j
(NONE) them by type and importance for safety. Describe both de-l i
structive and non-destractive veld testing, e.g., localized i
corrosion, metallographic examination and dimensional in-spections, and indicate what constitutes an acceptable i
j result.
State whether x-ray or equivalent inspection of fuel rods is part of the QC program.
If not, explain why.
l 231.21 Give the following properties of A1 0 -5 C'as a function 23 4 i
(4. 2. 3. 2) of temperature at various burnups:
[
t (1) swellins (2) thermal expansion (3) melting point r
(4) thermal conductivity (5) specific heat (6) compatibility with Zircaloy
[
i Describe how helium release is accounted for. Describe any reaction between A10 -B C and steam or hot water if 23 4 j
the cladding perforates.
1' 231.22 Discuss operating experience of B&W pressurized water (4. 2. 3. 2) reactors with burnable poison rods, including reactor names.
l.
loadingdates,andirradiatignparameters.'Describehow l
this A1 0 -B C behaves as 31 23 4 is burned. What models does B&W use to predict the behavior of these poison rods?.
231.23 Discuss the potential swelling of fuel rods during a (4. 2. 3. 3) postulated LOCA with respect to interference with control rod guide tubes to the point of hindering control rod move J ment. Describe the bases.for your response to.*.his question, including creep and thermal dimensional changes,' spacer grid / fuel rod interaction and fuel rod bowing.
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231-5 231.24 Supply calculations of the following parameters for an
.j
- (4.4. 2.0) average burnup and peak burnup 17x17 fuel rod as a function of burnup (Nominal values should be used.)
(1) gap conductance (2) hot pellet diameter l
(3) hot gap (4) fuel centerline temperature (5) fuel volumetric average temperature (6) internal gas pressure (7) gas thermal conductance (8) jump distances for fuel pellet and cladding or total (9) cladding inside diameter temperature l
Specify a reference or supply a complete description of the computer program used for these calculations, including all materials properties and models used. Supply the power history assu=ed for these calculations and explain why it is typical of what can be expected in your reactor.
Shpply the axial power shape assumed for this calculation.
Explain how the stored energy obtained from this calcula-tion is used as input to the fuel rod heat up calculation in the ECCS analysie i
231.25 Provide an engineering failure analysis of the design r
(15.1.23.2) basis fuel handling accident.
In particular, provide a j
detailed mechanistic description and calculation of the accident, including assembly drop height and a justification for the selection of the height and the location of the 1
. drop (such.as in fuel storage pool or over the reactor).
Provide experimental data to support your calculation of fuel rod damage.
Also, justify the assumption that damage occurs to only 64 fuel rods due to a drop of the fuel assembly.
231.26 Describe independent checks made at the cumpletion of fuel (NONE) loading to verify the location and orientation of the fuel in the core.
231.27 Describe the effects of blowdown forces on the fuel rods (NONE) during a LOCA. Name and give detailed descriptions of the computer codes used to calculate these forces and ex-plain how the thermal hydraulics calculations are used as input into these c, odes.
231-6 231.28 List the fuel rod stresses caused by lateral differential (NONE) thermal expansion between the and fitting and the spacer grid.
231.29 Provide a detailed description of the fuel fatigue analysis.
(NONE)
Include a sample calculation for a 17x17 rod fuel assembly showing the assunptions that are made for fuel rod tolerances, fuel temperatures and cycling history. Also describe the primary creep rate expression that is used.
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Give safety factors applied in the fatigue design, creep rupture, i
i fatigue creep interaction and instability (buckling) analysis for the fuel assemblies, f
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232-1 232.0 PHYSICS 232.1 APSR maloperation implies that some limits are placed on the (4. 3. 2.~2) location of the APSRs. Please clarify.
If limits are used in performing design analyses will these limits become subjects of Technical Specifications?
232.2 Frovide local (intra-cell) peaking factors for the various
{
(4. 3. 2. 2) types of assemblies used in the core (different enrichments and burnable poison loadings, e.g.).
232.3 Update the discussion of the nuclear uncertainty factor (4.3.2.2) to include the experience gained with operating reactors.
In particular, include any comparisons between calculation and experiment for transient conditions.
1 Sustained azimuthal xenon oscillations are predicted not 232.4 However, it is not clear that a peaking factor.
(4. 3. 2. 2) to occur.
problem does not occur for damped azimuthal oscillations, such as might occur after a dropped rod. Please comment on the behavior of the core in such a case and provisions for avoiding peaking factor problems.
232.5 What is the maximum peaking factor increase caused by (4. 3. 2. 2) misloading fuel in the core? What is the maximum increase that would go undetected by the incore detectors?
232.6 Tables 4.3-9 and 4.3-10 are inconsistent with respect to (4. 3. 2. 4)
Doppler and Moderator temperature deficits at BOL, first cycle. Please clarify.
232.7 Discuss the effect of failure of the ICS to keep the (4. 3. 2. 8) temperature in the-two loops equal.
232.8 Provide additional data on the new fuel storage racks to support the conclusion that k will not exceed 0.95.
At (9.1. 4. 2) least the following informati8kIshould be provided:
(1) Calculation method used and methods used to verify its accuracy.
(2) Assumptions used (fuel enrichment, credit for structural materials, etc.).
(3) Uncertainties used (a) Calculational bias (b) Calculational uncertainty (c) Mechanical uncertainty (4) Nominal,v'alue of k,gg for the racks v
T 232-2 3
-.4; 232.9 Provide assurance that fresh fuel racks will remain sub-(9.1.4.2) critical (keff f 0.98 with uncertainties included) for accident conditions.
Include in these conditions the flooding of the racks with partial density water.
(If the racks are not subcritical for some densities of water each applicant will have to provide assurance that achieving such densities is incredible),
s 232.10 Provide additional data on the spent fuel storage racks (9.1.4.2) to support the conclusion that k,ff will not exceed 0.95.
The same information as that sought in Question 232.8 should be provided.
232.11 Justify the use of a symmetric cosine axial' power shape (15.1) in generating the shutdown reactivity curve, since bottom-and top-peaked distributions are permitted during normal operations.
In particular discuss the effect of the as-sumed shape of the initial axial distribution on the min-imum DNBR achieved during Loss
- of Flow and Increased Heat Removal Transients.
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232.12 Justify the use of the same shutdown reactivity curve for
(
l (15.1)
Loth full power and zero power transients.
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232.13 Revise the SAR to reflect the fact that the additional 6
(15.1) conservatism claimed for the shutdown reactivity curves due to the assumption of minimum tripped worth does not apply at the end of the first cycle where only a 1.0%
shutdown margin is predicted.
232.14 There are references at several places in Chapter 15 (15.1.1)
(e.g., p.15.1-4b and Table 15.1-3) to Table 15.1-2 l
which bear no relation to the contents of the table.
Picase clarify.
j 232.15 Discuss the case of withdrawal of a single CRA at zero (15.1.1) power. Does the increase in peaking factor resulting from this transient cause,the heat generation rate to exceed the thermal limits in any part of the core?
232.16 The sensitivity studies show that the maximum vessel (15.1.2) pressure during the transient increases with more negative values of the moderator temperature and Doppler coefficients.
Discuss the consegnences of this transient at EOL when both of these coefficients have their most negative values.
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232.17 Discuss the consequences of a single CRA withdrawal as a
'f (15.1.2) function of initial power.- Assume operation of the control rod groups at the limit of their insertion band and discuss the effect of a single CRA withdrawal on peaking factors.
232.18 Correct the legend on the " Thermal Power" ordinate of (15.1.2)
Figures 15.1.2-la, Ib and Id.
3 232.19 Discuss the effect on the hot pin results, of the azimuthal tilt (15.1.3) produced by dropping the rod, particularly for the return to 100% FP in the EOL case.
Is there a trade-off between dropped rod worth and location such that a smaller rod worth may have greater hot pin consequences?
232.20 l
(15.1.3)
Discuss the consequences of this accident on DNBR, par-ticularly in view of the reduced pressure at BOL.
232.21 What assumption was made with respect to tripped rod worth (15.1.3) in the analysis of the rod ejection accident? Was a rod assumed to be stu,ck in addition to the postulated ejected rod?
232.22 Describe the manner in which the feedback coefficients, (15.1.18) particularly the Doppler coefficient, are obtained for the point kinetics calculations.
In particular discuss the manner in which "three-dimensional" effects are treated.
232.23 The point kinetics results in Table 15.18-3 are different (15.1.18) from those given in the PSAR for the Greene County Nuclear Power Plant.
Please clarify the discrepancy.
232.24 Justify the use of the design peaking factor in the point (15.1.18) kinetics calculation rather than some larger peaking factor representative of the situation with the ejected rod removed from the core.
232.25 Provide, in addition to the number of rods experiencing cladding (15.1.10) failure, the number of rods in which the fuel reaches the melting temperature. Indicate whether the same rods experience both types of failure.
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s 311-1 310.0 ACCIDENT ANALYSIS 311.6 Appendix 15A, Radiation Sources, should provide a discussion and data on iodine spiking as a result of reactor pressure and camperature changes.
Iodine spiking should be censidered in the.cnviron= ental consequence analyses of the steam line break and steam generator. tube rupture accidents (see SRP 15.1.5 and 15.6.3). Provide the requested data for the appendix and update the appropriate accident sections.
311.7 It is stated that Table 15.1.16-3 su==arizes the assu=ptions used in i
(15.1.16) determining the radiological assu=ptions of a vaste gas decay tank rupture. The table is missing and should be supplied.
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311.8 The last paragraph on page 15.1.18-6 appears to conclude that no fuel
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(15.1.18) melting occurs as a result of the worst rod ejection accident. Verify that this is true. (See questio'n 232.25)
If fuel melting is calculated to occur, modify the outline of the radiological consequence analysis to conform with the guidelines of Regulatory Guide 1.77 regarding fuel melt.
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o 320.0 EFFLUENT TREATMENT SYSTEMS 320.5 Your source terms for normal operation are based on parameters j
(11.1.2) and calculational methods that are not consistent with Regulatory Guide 1.BB, " Calculation of Releases of Radioactive Materials in Liquid and Gaseous Effluents from Pressurized Water Reactors,"
dated September 9, 1975. Provide operating data to justify the l
parameters used in your evaluation.
320.6 Table 11.1-8 indicates that corrosion product activities in the (11.1.2.3) primary coolant are based on extrapolated data from similar operating reactors. Provide these data for our review in accordance I
with the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Revision 1.
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320.7 In your expected releases, you do not consider liquid effluents (11.1.5) from the Boron Recovery System. Reactor operating experience does not justify this assumption.
You should provide justifica-tion that:
i
- 1) The plant water inventory can be maintained over the t
plant lifetime without discharge from these systems;
- 2) The tritium levels in the plant can be controlled to maintain radiation exposure to operating personnel as low as practicable without discharge from these systems; and
- 3) There is sufficient capacity and flexibility or redun-dancy in the system that discharge from these systems will not be necessary as the result of anticipated operational occurrences and equipment downtime.
320.8 Provide the interface requirement. reference for item 15(8) in (9.3.4.1) figure 9.3-1, sheet 1.
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331-1 l
331.0 RADIOLOGICAL /.SSESSMENT 331.8 With regard to the design review program described in Section (12.1. 2) 12.1.2, describe the specific review functions and responsibilities of the person (s) who assures that features for assuring that occupattanal radiation exposures can be maintained as low as is reasonably achievable are included in the design of the plant.
331.0 Describe what specifications are used in the design and selection (12.3.1.1) of materials for the reactor coolant system that are specifically directedtowardthgreductiggoftheproduction,transportand redeposition of Co and Co
-1 331.10 The answer to acceptanco review question 331.3 is incomplete.
(9.3.4)
Describe the design considerations used in selecting a letdown range of 50-200 gpa. Discuss the effect on reactor coolant corrosion and fission product activity of significantly increasing the letdown rate. Discuss your design considerations for the use of a high te=perature, high flow filtration system to minimize the buildup of corrosion products in the reactor coolant and i
auxiliary systems.
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