ML20040A368
| ML20040A368 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 05/13/1976 |
| From: | Israel S Office of Nuclear Reactor Regulation |
| To: | Deyoung R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201200817 | |
| Download: ML20040A368 (15) | |
Text
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UNITED STATES C [4::,
4 NUCLEAR REGULATORY COMMISSION G
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j WASWNGTON, D. C. 20666 N1 1976 l' l
Docket No. STN 50-561 i
I Richard C. DeYoung, Jr., Assistant Director for LWR's, DPM g
THRU: Thomas M. Novak, Chief, Reactor Systems Branch, DSS B-SAR-205 ROUND ONE QUESTIONS Plant Name: B-SAR-205 Docket No.: STN 50-561 Milestone No.: 05-21 Licensing Stage:
PDA i
Responsible Branch & Project Leader:
LWR-1, T. Cox Bystems Safety Branch Involved:
Reactor Systems Branch Description of Review:
Round One Questions Requested Completion Date: April 26, 1976 Review Status:
Complete
/ '
Adequate responses to the enclosed list of questions and comments are
(
required to assist in our review of the B&W standard plant B-SAR-205.
Reactor Systems. Branch review included Sections 1.5, 3.5, 4.2.3, 5.2.2, 5.2.5, 5.4.7, 6.3, and Chapter 15.0.
The following items are worthy of special note:
1.
The complete response to Acceptance Review Question 212.1 pertaining to the Chapter 15.0 failure modes and effects analysis has not yet been received. We have yet to resolve our problems with the sample analysis submitted and request a meeting or conference call with B&W to reach agreement on the ground rules, format, terminology, and goals of the remaining analyses. The specifics of our comments with regard to the submitted sample analysis were informally transmitted to the B-SAR-205 LPM, T. Cox.
2.
We have reviewed B-SAR-205 on the basis of 4-pump operation only.
The appropriateness of partial loop operation will be pursued at the Operating License stage.
3.
Technical Specifications have not been reviewed.
The initial conditions utilized in Chapter 15.0 must be reflected in the utility-user's Tech:ilcal Specifications as Limiting Conditions of Operation.
For example, we would expect from the 6.55% Ak/k initial suberiticality margin used in the boron dilution event during refueling (Section 15.1.4),
that the utility-ut;er would adopt the 6.55% Ak/k as the minimum margin to be allowed in Technical Specifications.
This aspect of our review will be performed at the Operating License stage.
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I R. C. DeYoung, Jr. '
4 The B&W code " TRAP" used to analyze steam and feedwater line breaks is new and has not yet been reviewed by the staff.
The Analysis Branch is conducting this review.
5.
We are continuing to review the interface criteria provided by B&W.
Since the utility applicant will be ultimately responsible for instituting all safety-related design requirements, the Reactor Systems Brar.ch will review these matters in more detail on a case-by-case basis. A proper response to Item 1 above is an important inter-face requirement in that we are expecting that the protection sequence diagrams submitted by B&W will identify all BOP equipment and systems that are essential to mitigating the consequences of each event in Chapter 15.0.
Also, a proper definition of staff interface require-ments is currently underway as requested in a letter from Walter P. Haass to Assistant Directors dated April 8,1976.
6.
The potential for reactor coolant pump overspeed during a LOCA is under generic review.
Completion of staff review of pump over-speec calculations from various vendors is pending additional two-phase flow and pump scaling data from ongoing research programs.
9 ' t, (Gk Sanford L. Israel Section Leader Reactor Systems Branch Division of Systems Safety Enclosure" Round One Questions s
cc:
S. Hanauer R. Heineman D. Ross J. Stolz T. Cox T. Novak S. Israel J. Watt G. Mazetis (3) 1 1
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212.163 BAW-10099 is referenced on page 15.1-2 as:providing discussicas of (15.1)
ATWS events.
The events in this teport should be resubmitted at the B-SAR-205 power level. Also,jthe staff " Status Report on p
3 g V Anticipated Transients Without Scram for Babcock & Wilcox Reactors N l
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^ dated December 9,1975, specifies the additional analyses and
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fjdtd f [ Testgn-e b gn g' to meet the, safety' objective of WASH-1270.
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O to. provide h additional analyses T.,
specified in the Status Report and an identification of the proposed I
I design changes based on the analyses performed.
212.164 Table 15.1-6:
The version of CRAIT 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:
1 (15.1)
[l' (1)
Event 15.1.2........... 1.71 (2) Event 15.1.5........... 1.73 1
(3)
Event 15.1.8........... 1.76
]
(4)
Event 15.1.10.......... 1.75 g
(5)
Event 15.1.14...........' 1. 72 g
(6)
Event 15.1.17.......... 1.84 Please explain why the initial assumed value varied and specify k
the correct value..Should changes in assumed values be appropriate, specify tha. impact upon minimum DNBR.
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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 followins:
criteria:
(1)
Each Safety Action shall not be vulnerabic to a single active l
component failure.
l (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 The total NSS thermal power, including 20 MWt for the reactor (15.1) coolant pumps, is 3820 MWt.
Why wouldn't 1.02 x 3820 = 3896 MWt
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be more appropriate as an initial condition for power level (versus l
I 1.02 x 3800 = 3876 MWt)?
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212.168 An acceptance criterion which is not presented in Chapter 15.0 is-l (15.1) the pressure-temperature NDT curves provided in technical (N
specificat. ions.
Branch Technical Position hTEB No. 5-2 of Stcndard s
Review Plan 5.3.2 states that thc sc curves are applicable during N
upset conditions. Along these lines, provide a complete 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 caintain the pressure-temperature limits.
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212.169 Submit an analysis of the worst-case overpressure transient during (15.1) a startup or shutdown. Provide all assumptions. Plots should include pressure versus time, reactor coolant temperature versus l
time, and safety valve flows versus time.
Show the times of all trips, actuations, or operator actions on the picts.
Rationalize the time of occurrence of this event as being worst-case (with i
N2 blanket, with steam bubble, etc.).
Show that the pressure-temperature limits in technical specifications are not exceeded.
c 212.170 Indicate which events in CF. apter 15.0 would produce more severe
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(15.1) consequences for higher initiJil core flows (e.g.,110% versus 100%).
4 Indicate if any events would be worst for lower' initial core inlet l
f temperature, or higher initial RCS pressure.
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212.171 Submit a schedule for providing the failure modes and effects f
(15.1) analyses requested in question 212.1 (2/1/76).
It would be preferaU e l
to receive each event as it is completed, with the steam line break \\
i as the next submittal.
-t 212.172 Confirm that no Chapter 15.0 event would be more severe for the
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(15.1) 3600 MWt plant than the 3800 MWt design. Verify that the Station s
Technical Specifications which are based on Chapter 15.0 events L
will not change between the two plant sizes.
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I v 12.173 Table 15.1-5 provides a list of support systems necessary for the (15.1) operation of each primary safety system. Although it is recognized I
that the pressurizer relief system performs a role similar to the RC system safety valves, it is the staff's understanding that the i
relief valves should not be interpreted as a required support system l
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in order for the safety valves to perform their safety function.
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t If so, the pressurizer relief system should be deleted from the s
table as an essential support system.
If not, B W should upSrade f
e these components to the same qualifications as the safety valvet.,
212.174 Section 15.1 identifies a reactor coolant pump seizure as a " fault l
(15.1) of moderate frequency" and the inadvertent operation of ECCS as an I
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" infrequent incident." Tbis would appear to be incongruous siner.-
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operating experience has shown that inadvertent ECCS actuations '
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have occurred in the reactor industry.
In ition, page 15.1.5-1
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states that the frequency of occurrence of recident{pumpseizurel j
is expected to be the same CA (gross mechanical, failtfre of pri-
)
i mary system).
Rationalize e4
'nsistentclassificationoftheseeventr..l 5
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212.175 Page 15.1-3 states, " Chapter 15 analysis of non-limiting or non-
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i (15.1) desien basis events may take credit for non-safety crade equipment t
because if this were not done, the conditions would lead to a l
l limiting or desien basis event which is already considered." With regard to this statement, provide the follcring additional information:
i-6 1
l (1)
For each event which takes credit for ncn-safety grade equip-t J
ment, juctify thg statement that the consequences would otherwise l
l be limited to % design basis event.
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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 esaential non-safety grade l
systems or components.
0 (3)
Rationalize the above statement in light of the published staff position in Regulatory Guide 1.29 dated August 1973:
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"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 I
i 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 adequate 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 I
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, steam generator invettory, 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 conditio s,in Chapter 15.0 to the limitingconditionsofoperation/in n 2 Technical E
Specifications.
I 212.177 Table 15.1-4:
Credit for several types of secondary side-relieving (15.1) devices is sought by B6W to mitigate the consequences of certain l
events.
The concern is that such components inporrant to safety y
may not be proper safety grade.
The terminology used in Section 10.5
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g to discuss available relieving devices is not clear.
Identify on,
Figure 10.1-1 by number the following components:
i (1)
Spring-Loaded Safety Valves i
(2)
Secondary Safety Valves (3)
Atmospheric Dump Valves (Modulating)
(4) Atmospheric Dump Valves (Non-Modulating)
I
' (5)
Condenser Dump Valves (Modulating)
(6)
Condenser Dump Valves (Non-Modulating) h (7)
Turbine Bypass System Valves (8)
Safety Grade Steam Du=p Valves (Table 15.1-4)
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212.177 State clearly which of the above valves are taken credit for in
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(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 4
l 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 will be the minimum allowed suberiticality margin allowed by Technical Specifications d,uring refueling.
212.179 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
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operating state).
f 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 i
effect of a dilution event at this time when the boron concentration l
is at its minimum.
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 valves as an input assumption.
Explain I
this inconsistency and discuss the sequence of events if automatic termination of the deborated water does not occur.
S 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 would this be limited to 700 gpm '. design flow rate of only one HPI pump)'l i
i 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 primary system.
The concern l
is that the potential for this event could be high in more than one reactor coolant pump during a seismic 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, shaft, etc.) and discuss the potential for seizure l
of pump / motor components during a seismic event.
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212.184 Table 15.1.5-3 indicates that the initial pressure and inlet f
(15.1.5) temperature are assumed to be nominal values for the locked rotor event analysis.
This is not acceptable and is inconsistent with i
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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[12.185 Table 15.1.5-1 indicates that a value of +3 F is added to inlet 0
i (15.1.5) temperature to account for control band and instrument errors.
Page 15.1-4a indicates that +20F is a good number.
Justify why +4*F i
is not a more appropriate value.
3 Table 15.1.3-1 also indicates that a value of -45 psi is imposed upon the initial system pressure to account for control band and instrument errors.
Expand on what improvement in design is being proposed which allows a greater confidence in this value compared to Davis-Besse Unit No.1 (-65 psi).
Explain how thiF edded 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 val.ues were at.letst as conservative as those shown in Table 4-3 of BAW-10102. Justify the appropriateness of these values. Will startup testing confirm these values?
c The maximum allowabl#o -
212.187 Submit a plot of RCS pnasure versus time.
(15.1.5)
RCS pressure limit should also be an acceptance criterion 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 k
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 I
active component failure with each loss of flow event (4-pump trip and locked rotor) and show that a'ny fuel damage calculated o
rTmain occur is of sufficiently limited extp'nt that the cor w
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212.190 The response to question 212.80 (4/1/76) clearly shows that a l
-(15.1. 6) reactor coolant pump startup time of 15 seconds is not an appropriate assumption to utilice in a worst-case Chapter 15.0
(
event; in fact, the reported Oconee Unit No. 1 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 j
consequences of the event would not be sensitive to this parameter.
l Also, the response to part 5 of the question has not yet been i
received.
j
'12.191 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 l
parameter would initiate its operation (in each case discussed)?
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212.191 Confirm that Figure 15.1.7-1 reflects the time histories for the (15.1.7) most se rere 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-I (15.1.7) which result in a decrease in heat removal 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 lin$e break on page 15.1.8-3,
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(15.1.8) identify which of the eight steps would require operator action.
Discuss this action with regard to time available, physical movements
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needed, for how long must action be performed, etc.
Discuss the necessity for these or other manual actions across the break 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 i
first check valve (with a worst-case single active component N-failure and loss of offsite power). Please justify this 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 co=plete 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 ESFAS 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 calfunction of the turbine stop valves.
Confirm that this also refers to the 3CS pressure limit.
212.197 For a feedwater line rupture with a loss of offsite power, I
(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 es.s33jjpanalossof l
Confirm tha
.gner RCS pressare mut n off ije g g grupture.
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iia : P 7e"r the complet;c loss of cil l
1 feedwater event.
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'12.198 The response to question 212.9 (4/1/76) indicates that auxiliary (15.1.8) feedwater flow is initiated by low steam line pressure at 58 i
seconds.
Table 15.1.8-2 shows initiation was assumed at 40 l
l seconds.
Explain this inconsistency.
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1 212.199 Page 15.1.9-2 states that assumed initial conditions are given in 1
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Table 15.l-2.
This table should also provide all secondary side assumpcions.
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?
212.201 With regard to the feedwater temperature decrease at 3876 MWt, (15.1.10) provide a plot of thermal power versus time.
Submit the specific value of minimum DNBR.
With regard to the feedwatec flow increase at n'o load conditions, s
I provide a plot of DNBR versus time or confirm that DNBR does not decrease below its initial value.
y of all the moderate frequency excessive heat removal events (feed-opening of turbine water malfunctions, opening of steam safetz,he worst cooldown bypass valves, etc.), which event produces t 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 e-component must be considered in these interface requirements.
-212.203 It appears from the response to question 212.75 (2/1/76) that tae (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 summary table with the following information:
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g is 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 atmosphetic 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 s
setpoint?
212.209 Discuss ~ the procedures for long-term core decay heat remcval (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
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t line chosen instead of the smaller 28-inch line?
It would appear possible that since the maximum cooling rate occurs within the s
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
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released out of the break in the letdown line).
Please crplain.
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212.213 The staff considers an inadvertent opening of a safety or relief
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.(15.1.33) valve to be en event of moderate frequency.
It must be showa j
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 t'he reactor coolant pump coastdown b
is greater for the pump shaft breakage than for the shaf t 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 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 I
(15.1.35) isolation valves, it is not obvious that the secondary side i
pressure surge would be bounded by the loss-of-normal feedwater event. Please justify.
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.
Provide sufficient detailed information to enable the staff to review independently the calculated leak rate.
I Discuss the ability of the operator to detect and isolate such a failure.
8 212.217 The response to question 212.61 (4/1/76) indicates that a design I
(6.3.2.10) change has been made in the operating mechanisms of the reactor N internals vent valves.
Such a modification clearly relates to th j
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proper status of these components during normal operation and j
transient events.
Therefore, this change raises a question on the appropriateness of taking credit for the vent valve operating l
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,andcontinuoussurveillanceofventvalvefanomalies l
utilicing such means as loose parts monitoring.
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212.218 Based on the credit taken for systems 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.'
i 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)?
i 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 5.5-13?
Confirm that this transient woul.d be accommodated without failing the rupture disc.
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UNITED STATES l
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j WASHINGTON, D. C. 20555 t
NUCLEAR REGULATORY COMMISSION i
c gygg Docket No. STN 50-561 i
RichardC.Dehoung,.Jr.,AssistantDirectorforLWR's,DPM g
THRU:
Thomas M. Novak, Chief, Reactor Systems Branch, DSS B-SAR-205 ROUND ONE QUESTIONS Plant Name:
B-SAR-205 Docket No.:
STN 50-561 Milestone No.:
05-21 Licensing Stage:
PDA Responsible Branch & Project Leader:
LWR-1, T. Cox Systems Safety Branch involved:
Reactor Systems Branch Description of Review: Round One Questions Requested Completion Date: April 26, 1976 Review Status:
Complete Adequate responses to the enclosed list of questions and comments are required to assist in our review of the B&W standard plant B-SAR-205.
Reactor Systems. Branch review included Sections 1.5, 3.5, 4.2.3, 5.2.2, 5.2.5, 5.4.7,_6.3. and_ Chapter _15.0.__The_following items are worthy of special note:
1.
The complete response to Acceptance Review Question 212.1 pertaining to the Chapter 15.0 failure modes and effects analysis has not yet been r;eceived. We have yet to resolve our problems with the sample analysis submitted and request a meeting or conference call with B&W te reach agreement on the ground rules, format, terminology, and goals of the remaining analyses. The specifics of our comments with regard to the submitted sample analysis were informally transmitted to the B-SAR-205 LPM, T. Cox.
2.
We have reviewed B-SAR-205 on the basis of 4-pump operation only. The appropriateness of partial loop operation will be pursued at the Operating License stage.
3.
Technical Specifications have not been reviewed.
The initial conditions utilized in Cha~pter 15.0 must be reflected in the utility-user's Technical Specifications as Limiting Conditions of Operation.
For example, we would expect from the 6.55%Ak/k initial suberiticality margin used in the boron dilution event during refueling (Section 15.1.4),
that the utility-user would adopt the 6.55% Ak/k as the minimum margin to be' allowed in Technical Specifications.
This aspect of our review I
will be performed at the Operating License stage.
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MY l 31976 P
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R. C. DeYoung, Jr. '
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The B&W code " TRAP" used to analyze steam and feedwater s
line breaks is new and has not yet been reviewed by the staff.
The Analysis Branch is conducting this review.
5.
We are continuing to review the interface criteria provided by B&W.
Since the utility applicant will be ultimately responsible for instituting all safety-related design requirements, the Reactor Systems Branch will review these matters in more detail on a case-by-case basis. A proper response to Item 1 above is.an important inter-face requirement in that we are expecting that the protection sequence diagrams submitted by B&W will identify all BOP equipment and systems that are essential to mitigating the consequences of each event in Chapter 15.0.
Also, a proper definition of staff interface require-ments is currently underway as requested in a letter from Walter P. Haass to Assistant Directors dated April 8, 1976.
6.
The potential for reactor coolant pump overspeed during a LOCA is under generic review.
Completion of staff review of pump over-speed calculations from various vendors is pending additional two-phase flow and pump scaling data from ongoing research programs.
Sanford L. Israel Section Leader Reactor Systems Branch Division of Systems Safety Enclosure" Round One Questions cc:
S. Hanauer R. Heineman D. Ross j
J. Stolz T. Cox.
T. Novak S. Israel J. Watt G. Hazetis (3) l
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