ML19329B983
| ML19329B983 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 04/18/1975 |
| From: | Schwnencer A Office of Nuclear Reactor Regulation |
| To: | Roe L TOLEDO EDISON CO. |
| References | |
| NUDOCS 8002110766 | |
| Download: ML19329B983 (29) | |
Text
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9 DISTRIBUTION: w/o encls.
NRC PDR w/encls.
L PDR w/encls.
Docket File w/encis.
Docket No.: 50-346 LWR 2-3 Rdg w/encls.
LWR ads Aya 1 S R5 FSchroeder AKenneke LWR TCs The Toledo Edison Company
-ELD ATTN: Mr. Lowell E. Roe IE (3)
Vice President Facilities Development LEngle w/encis.
EGoulbourne 300 Edison Plaza TR BCs Toledo, Ohio 43652 LWR BCs ACRS (16) w/encls.
Gentlemen:
Our continuing review of the Final Safety Analysis Report (FSAR) and your responses to our first-round requests for the Davis-Besse Nuclear Power Station Unit 1, have resulted in the need for additional infor=ation and the development of NRC staff positions.
Since an acceptable ECCS perforr.ance analyses does not exist at this time for Davis-Besse, Unit 1, our review remains inconplete through the second round review. Areas needing further discussion have been provided in Enclosura 1 and should be addressed in your subnittal for ECCS analysis.
We are providing Enclosure 1 at this time in order to assist you in preparing an expiditous response concerning these matters.
The NRC positions are listed in Enclosure 2 and requests for information are listed in Enclosure 3.
We request that you amend your FSAR to conform with the enclosed positions and to provide the additional information requested.
In order to maintain our licensing schedule, we will need your responses to Enclosures 2 and 3 by June 9, 1975. If you cannot =eet this date, please inform us within saven days af ter receipt of this letter so that we may revise our scheduling.
If you plan to appeal to Licensing Management on the positions specified in Enclosure 2, please advise us of your intentions within two weeks.
Please be advised that those requests for information provided in Enclosure 3 concerning Technical Specifications are applicable to the B&W Standard Technical Specifications which will be adopted for Davis-Besse, Unit 1.
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" Please contact us if you have any questions regarding the enclosures provided.
Sincerely, Cri;!nal Qead ty w
A. Schwencer, Chief Light Water Reactors Branch 2-3 Division of Reactor Licensing
Enclosures:
1.
ECCS Performance 2.
NRC Positions 3.
Requests for Additional Information ecs: Donald H. Hauser. Esquire Mr. G. M. Olds The Cleveland Electric Illuminating Co.
Senior Project Manager P. O. Box 5000, Rootu 610 Babcock' '& Wilcox Co.
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Cleveland, Ohio 44101 P. O. Box 1260 Nuclear Power Generation Cerald Charnoff, Esquire Lynchburg,. Virginia 24505 Shaw, Pittman, Potts, Throwbridge and Madden Mr. Pete Anis 910 - 17th Street, N.W.
Licensing Engineer Washington, D. C.
20006 Bechtel Corporation Shady Grove Road Leslie Henry, Esquire Gaithersburg, Maryland Fuller, Seney, Henry & Hodge 800 Owens-Illinois Bldg.
405 Madison Avenue Toledo, Ohio 43604 becs:
J. R. Buchanan, ORNL w/o encls.
T..B. Abernathy, DTIE w/o encls.
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ECCS PERF0P.:*.ANCE_
TOLEDO EDISO: CnMP.4:!Y DAVIS-BESSE NUCLEAR STNIIO:1, U.'!IT 1 DOCKET No. 50-346 19, 1974, and in a In a letter to 11r. A. Schwencer dated August recent revision to the Davis'-Besse FSAR, Toledo Edison indicated that the completion of the LOCA analysis and ECCS design by Babccck &
With regard Wilcox is expected in Decenber of 1974.to the LOCA analysi ECCS Evaluation Model Conformance in the Matter of Babcock & Wilcox to 10 CFR 50, Appendix K " and Supolcment 1 indicate that the applica-bility of the generic medel to Davis-3 esse has yet to be verified.
Justify the appropriateness of each section in Apoendix K to the Davi,-Besse design, and include the folicwing additional informa s
Describe the major design differences between Davis-1.
Besse and the Oconee Class reactors.
I Provide detailed coccarisen tables between Davis-Besse and the generic analysis of key parameters (s'uch as r. umber of vent 2.
I valves, vent valve "K", loop resistance factors, ECCS cesign com-I<
parison, core inlet fluid temoerature, core ficw, reactor pressure, and other relevant parameters) employed in the ECCS analysis, especially those parameters of greatest effect on peak clad tem-perature and metal-water reaction.
Confirm that the volumetric average fuel temperature ut the maximum power lo. cation assumed. in the Davis-Besse analysis is 3.
equal to or greater than that calculated in the approved version.
of TAFY.
With regard to question 4.2.7 on vent valve design, the state-4.
ment is made that the number and size of the vent valves areThis is under shown to be acceptable by ECCS analyses.
however, also of concern are any variations in the manr.er in which these venting areas are applied in the ECCS analyses.
Also, it is noted that valve venting areas available on Davis-Besse are less than on any other Babcock & WilcoxVerify reactor, including other plants with raised lo:ps.
that the methods of calculating and apolying such venting areas have not changed for Davis-Besse.
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-2 5.
Discuss the probability of unacceptably high concentrations of boric acid in the core region during long term ccoling due to continuous evapcration.
Describe any analyses you have performed to indicate that precipitation will not occur. The description should clearly identify the criteria, method of analyses, and equip-ment required.
6.
List and describe any other' computer codes, such as
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SAVER, which are used in the Davis-Besse analysis, but are not refe,renced in the generic model description.
7.
Examine, and provide new analyses for Davis-Besse where appropriate, each sensitivity study used as a basis for the ECCS model in BAW-10091. Completely justify any study which is not re-analyzed for Davis-Bes'se. Compare each study to those design aspects of Davis-Besse and the Oconee Class reactor of greatest influence on the outcome.
Should the result of any sensitivity study en Davis-Besse be in opposition to those studies in BAW-10091, provide a thorough quantitative explanation of the phenomenological differences produced and the reasons for these differences.
8.
Provide, oi reference in Chapter 16.0, the linear heat generation rate limit as a function of axial elevation.
9.
With regard to the response to question 6.3.12 (small breaks), BAW-10075 is currently under review by the Regulatory staff. The conclusions of this review will be applicable to Davis-Besse. Also, compare the results, using 10 CFR 50, Appendix K, of a 0.5 ft2 break with the large break analytical medel and the small break analytical model. Also, include a plot of peak clad temperature versus break size.using the small break model up to, and including, the 0.5 ft4 break.
- 10. With regard to question 6.3.3,(CFT line break), the response is insufficient to allow an adeq' ate evaluation.
Resubmit the core flooding tank (CFT) line break considering the re-quirements of 10 CFR 50, Append {x K.
Explain why the peak clad temoerature for the 0.3 ft breakjreferencedinBAW-10075) was 1090' F. but for the 0.44 ft break (CFT line) was only 932* F (FSAR page 6-81d).
For the Cl-T line break, FSAR page 6-81c states that the reactor trips at a primary system pressure of 2050 psig.
Isn't the icw pressure trip
-at 1900 psig? Also, if the analytical techniques and as-sueptions were similar to BAW-10064 (as stated en FSAR page 6-81b), why did the higher power plant (Davis-Besse) proquee the lower peak clad temperature?
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- 11. With regard to question 6.3.4, describe the design of the CFT relief valve and, if this valve can be remotely actuated, analyze an inadvertant opening of this valve occurring during a loss-of-coolant accident (in lieu of a diesel failure).
- 12. Describe the conse.,uences of a loss-of-coolant accident during a startup or shutdown (whichever is the worst-case) while the CFT tanks are routinely isolated, and the coolant system pressure is assumed to be at the maximum expected during the time of CFT isolation.
- 13. For a break in a reactor coolant line, it appears that a single failure of an HPI train could result in all available HPI ' water flowing through the break. When the break is small enough such that MPI flow is needed to cool the core.(reactor
. coolant pressure too high for LPI flow), it is of concern whether sufficient ECCS flow exists.
Examine a variety of small and intermediate break sizes to assure that a sufficient flow split occurs in the rema'ining HPI train to cool the core. Credit for operator action is not acceptable.unless ana4yses show that I
sufficient time exists for the operator to recognize the initiating event, ascertain the proper response, perform the appropriate manual action (s), and that the required actions are clearly defined in the operating procedures.
Identify the exact manual operations required by the operator during the short tem and long term.
Speci fy: (1) the information available to the operator, (2) the time delay during which his failure to act properly will have no unsafe consequences, and (3) the consequences if the action is not performed at all.
If credit for sufficient flow splits using line orifices is proposed provide a complete description of the design basis of this l
3 feature and justify its capability to achieve this design basis.
l I
Describe all testing that has been conducted to confirm the ex-pected flow rates.
In addition, discuss the preoperational tests which are planned to observe the flow splits for Davis-Besse.
- 14. Provide a plot similar to PSAR Figura 14-48, "ECCS Capa-bility to Meet Fuel Clad Temperature Design limit." Comment on any differences between the PSAR and FSAR figures, Show also the capa-
.bility of the HPI pump + l core flooding tank combination (CFT line
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break).
- 15. With regard to the response to quesSn 15.3.1, the state-l ment is made that for a hot. leg break all the fluid injected by the l
core flooding tanks, the HPI pump, and the LEI pump must enter the core before being lost out the break.
For the 14.1 ft2 hot leg break, I
discuss the potential for a portion of the ECIwater to flow into both cold legs, through the steam generators,and out the hot leg hvoak.
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- 16. Provide a plot of peak clad-temperature versus break 1
size for the complete break spectrum.
- 17. Provide a sensitivity study of various axial power i'
shapes to iustify the appropriateness of the shape selected for the LOCA analyses. Specify the cositions of all control rods.
Identify the shape utilized for tt.*' LOCA and discuss the rationale for its selection.
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ENCLOSURE 2 REGUU.TCRY POSITIONS TOLEDO EDISO:1 C0:3*A:.Y DAVIS-BESSE NUCI. EAR PO*.*ER STATION, UNIT 1 DOCKET 50. 50-346 POSITIO' 4.0 REACTOR 4.4.1 With regard to question 4.4.3, unless actual operating infor-C.
mation frem operating B&W plants would indicate that all vent valves are in their normally closed cosition, the staff's position remains unchanged; that is, one vent valve less than the minimum detect able number of stuck open vent yalves shall be assumed to be open for the analyses of the thermal-hydraulic design of the reactor coolant system and core and for all tran-sients.
Either, submit the re-analyses or:
1.
Show that a stuck open vent valve.would be detected by an operator or, 2.
Show that valves are normally. closed on operating reactors (wear on components, inspections, etc). -
Also, the resconse states that the valves can be tested during i
,each refueling.
Submit (or reference) your proposed Technical Speelficatic, which adopts this surveillance requirement, n
5.0 REACTOR COOLAllT SYSTEM With regard to question 5.5.3, the response is insufficient to allow an adequate evaluation.
It is stated that during normal 5.5.1 operation, the ECCS lines are fil. led with fluid but are in a no-flow C*
Describe the system provided to maintain water in the condition.
ECCS lines in spite of any expected leakage back through the provided State this system's safety design basis and provide check valves.
detailed design specifications including line sizes and capacities.
Our position is that the Davis-Besse design must reflect consideration of a water har.=cr being generated when coolant discharges.into an Also describe the design features that are provided empty line.
(venting, etc.) to prevent air entrapment within ECCS pump casings from reducing ECCS pump performance.
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.2-POSITIO;;
6.0 EMINEERED SAFETY FEATURES With regard to question 6.3.21, the Regulatory staff's positions 6.3.1 that the proposed LPI-to-HPI_ crossover and LPI-to-LPI crossover' are I.
not acceptable designs remain unchanged. The Davis-Gesse design must be modified as requested, the FSAR must be revised to reflect these design modifications, and the information requested in question 6.3.21 (part b) must be provided.
Also, the Regulatory staff notes that a break in the CFT line when the assumed diesel failure is the one upon which the normally closed in-jection valve in the intact CFT line is dependent (valve fails to open) renders the LPI crossovers ineffective. The design modifications adopted above must also reflect consideration of this situation.
Re-lying on the operator te repair the failure is not acceptable.
6.3.2 Revision 9 to Figure 6-17 now shows previously closed low pressure injection valves,0HIA and OflIB to be.open. Why?
J.
1.
It is the Regulatory staff's position that the over-pressure protection now in these ECCS lines is insufficient. The number and type of valves used to form the interface between the low pressure ECCS discharge and the reactor coolant system must provide adequate assurance tha't the ECCS will not be subjected to a pressure greater than its design pressure. This may be accomplished by any of the following methods:
a.
One or more check valves in series with a normally
. closed motor operated valve. The motor operated valve is to be opened upon request of a safety injection signal once the reactor coolant pressure has decreased belcw the ECCS design pressure.
b.
Three check valves in series.~
c.
Two che~ck 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.
L,Why, are check valves DH 76 and DH 77 shown to be locked open (see Figure 6-17)?;
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. POSITIO:
Regulatory Guide 1.79 provides additional guidance for pre-6.3.3 operational testing of ECCS. Evaluate the planned. Davis-Besse L-testing program with this Guide and itemize the areas of non-conformance. Toledo Edison will be required to conduct a test under ambient conditions that demonstrates the capab.ility of the system to operate in the recirculation mode of ECCS oper-ation. Our scecific concerns are the possibility of inadequate NPSH, air bindage, or vortex formation at the sump screens,any of which could adversely affect ECCS performance.
Discuss how To your proposed test program will address these concerns.
avoid reactor coolant system contamination, the sump water may be discharged to external drains or other syster.s.
Temporary arrangements may be made to provide adequate sump capacity for pump operation.
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ENCLOSURE 3 REQUESTS FOR ADDITIONAL INFORMATION TOLEDO EDISOU COMPANY DAVIS-BESSE NUCLEAR STATION, UNIT l' DOCKET NO. 50-346 4.0 REACTOR C
REQUEST NO.
With regard to the response to question 4.2.2 (Part h),
4.2.2 describe in detail the modification to the prototype Type-A roller nut control rod drive mechanism and show E:.
that this modificaticn represents no unwarranted extra-polation of prototype testing technology. Why is this modification not shown in Table 1-3, "Cemparison of Final and Preliminary Designs?" List the B&W reactors which incorporate the Type-C mechanism.
4.2.8 With regard to question 4.2.8, the response is in-sufficient to allow an adequate evaluation. The discussion A.
indicates that vibration tescing of operating reactor in-ternals has verified that the vent valves do not undergo excessive vibration.
Identify which specific instrument or combination of instruments described in BAN-10039 led to this conclusion.
Justiify this instrucent's (strain gage, accelerometer or pressure sensor) ability to detect a vibrating vent valve.,
Specify the expected exciting frequency of a vent valve and discuss any plans to instrument the Davis-Besse vent valves.
4.4.1 With regard to question 4.4.1, the response is insufficient to allow an adequate evaluation. Page 1-25 B*
states, " Table 1-3 identifies all. the significant changes that have been made in the station design since submittal of the Preliminary Safety Analysis Report (PSAR)." The Regulatory staff previously noted several inaccuracies in this table and requested that the table be corrected.
Since Toledo Edison is apparently in conflict with the staff's interpretation of the word "significant" (vent valves, ECCS cross-overs, new main steam line rupture trips, new feedwater line rupture trips, auxiliary feedwater piping re-design, autcmatic switch to recirculation mode, rod worth changes, etc.),our position is that this table must now reflect All changes specified in. question 4 4.1 since the PSAR.
.t.
REQUEST NO.
4.4.4 Revision 10 to the FSAR (page 4-85) increased the D.
number of control rod assemblies frca 49 to 53, with a resultant increase in total worth from 8.34 k/k to 10.0 4k/k.
Explain why such a significant reactivity control modification is necessary at this time.
5.0 REACTOR COOLANT SYSTEM 5.2.4 With regard to question 5.2.4, the response is insufficient to allow an adequate evaluation. The requested discussion should be A*
provided for pressurizer and steam generator safety valves and relief valves. Also, submit the following material:
1.
A, detailed description to accompany the requested diagrams explaining pressurizer safety and relief valve operation and identifying if and when credit for pressurizer electromagnetic relief. valve operation (2255 psig set point) is assumed in Chapter 15.0.
2.
A discussion of whether consideration of backpressure has been factored into the safety valve sizing analyses.
3.
Do the pressurizer and steam generator safety valve sizing analyses assume the failure of one valve in each instance?
4.
BAW-10043 is not an acceptable reference for Davis-Besse.
The response states that the plant analyzed in BAU-10043 is the same size as Davis-Besse; however, the number of steam safety valves re-quired on Davis-Besse (18 valves) does not reflect tne analytical conclusions in the topical report (22 valves-BAW-10043).
Also,thesteamgeneratordesignpressureis1050psig(FSARpage
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5-68). The peak steam generator pressure for the feedwater tempera-ture decrease transient (page 15-62) is about 1175 psia. Since this transient is more severe than the sizing transient, BAW-10043 could not be applicable. We also note that the pressure in this case reaches, and may even exceed, the 110% design criterion.
It is therefore the staff position that a complete pressurizer and steam generator safety valve sizing analysis must be. submitted specificall;.
for Davis-Besse. Also, modifications to the secondary system design (overpressureprotection)appearwarranted.
5.
Compare the margins (minimum psi below ASME limit) calculated
,to occur in the Davis-Besse pressurizer and steam generator safety valve sizing analyses with Oconee, Rancho Seco, North Anna, and o-,i e-
, REQUEST NO.
5.5.7 Recently, an unusual event occurred on Oconee Unit 3 in which the anti-rotational device on'one of the reactor coolant pumps failed D*
to function, and the pump rotated in the reverse direction. With re-gard to this occurrence, discuss the consequences that such a failure of the anti-rotational device could have on normal operation, transients, and accidents.
5.5.8 With regard to the response.to question 9.2.5, provide E.
a plot of reactor coolant temperature versus time.(from full power operation) using one and two RHR trains.
Describe and justify the procedure for attaining a cold shutdown condition with a malfunction (fail closed) of isolation valve DH-11 or DH-12 (inside the containment vessel).
, ' 6.0 ENGINEERED SAFETY FEATURES 6.3.5 With regard to question 6.3.5, include on the same diagram all other protection sequences recuired to A-mitigate the consequences of this event. That is, in addition to CORE COOLING, the diagram should also depict the systems reouired to produce
.other safety actions, such as REACTOR TRIP, CONTAINMENT ISOLA-TION, AND PRESSURE RELIEF.
With regard to Figure 6.3.5-1 (Revision 12);
1.
Why are the core flood tanks shown to be required for small 2
breaks of approximately 0.04 ft 7 2
2 2.
Show the required sequence for breaks betgeen 0.04 ft - 0.1 ft 0.3 ft*- 0.5 ft2 and 0.75 inch - 0.0 4 ft.
3.
The figure shows a requirement for low pressure injection during the short term of a CFT line break.
Since your analyses show that low pressure injection is not required, this sequence should be corrected.
4.
T'he SFAS channels should be included.
5.
Why was the AFS deleted during the long term?
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i KEQUEST NO.
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6.3.6 With regard to question 6.3.6, the answer is insufficient to 8.
allow an adequate evaluation.
Identify the specific systems or ccmponents which exist to provide the listed services.
Include any support sub-systems essential to the operation of each auxiliary system or components during a LOCA.
6.3.10 With regard to question 6.3.10, the resconse is insufficient C*
to allow an adequate evaluation.
Provide chronological lists of all manual actions that are required by the operators following a LOCJ, until stable long-term cooling is achieved.
These lists should include both large and small LOCA, and should indicate:
1.
the actual physical action taken by the operator (i.e.,
switch thrown, gauge checked, button pushed, etc., and the operator's physical location necessary to perform the sequired action).
2.
effect on the reactor systems of theaction (i.e., the system or items-of equipment turned on, turned off,. or whose operating state is changed--power source changed, watar source changed, water destination changed, etc.).
3.
the information required by the opentor to know when or if he should perfom the operation, that is, wht parameter must reach what level before the operation is required,and through what in-strument does the operator obtain that inforation, where is that instrument's readout physically bcated, andhow is the'information conveyed to the operator (meter or graph postion, audible or visible alarm,etc.)?
4.
the time delay in each case during which his failure to act properly will have no unsafe consequences, and the consequences if the action is not performed at all.
Also, include all of the information requested above for automatic operations which are to be verified by the operator, indicating what manual actions he is required to take if the automatic action is not properly executed.
In regard to the specific manual actions cited in the response to question 6.3.10, it is not clear why such an operational procedure is required. All safety analyses assume a single failure of a complete LPI chain, thereby ensuring that such loss of LPI flow as is referred
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- % REQUEST NO.
to in the answer to question 5.3.10 does not adversely affect the health and safety of the public. Justify the need to require these two operators to perform these potential high-radiation duties in all cases that the ocerator notes less than 1500 com in one of the DH injection lines. Also, the following points.are noted:
1
't 1.
Two operator's are required to effect the safety action (LPI flow equalization).
2.
The performance of the safety action depends on the amount of radiatica in the area.
No radiation levels are discussed and i
it is clear that this decision'(no safety action versus high radiation exposure) should not be imposed on the operators.
3.
No times are given.
4.
No consequences are discussed.
5.
No statement is made as to whet'her or not the need for this procedure was sensitive to break size.
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6.
No rationale is discussed as to why there would be less than 1500 gpm in the low pressure injection line.
7.
It is not obvious that sufficient personnel are left in the control roce after these two operators leave to respond to the procedures required during an accident (conitoring plant parameters, handling communications, etc.).
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6.3.13 With regard to question 6.3.13, the response is insufficient to all w an adequate evaluation. The statement is made that the maximum cal-D*
culated control rod centerline temcerature is 1295'F at 43 seconds and that the melting temperature is 1470 F.
Our concern is that temperatures
, may be high enough within the control rods to conpromise rod integrity.
The following additional information is required:
1.
Confinn that the 1295'F.was the maximum predicted temperature throughout the control rod.
2.
What cold leg break was assumed for the analysis (split or guillotine)?
_ REQUEST NO.
3.
FSAR Figure 6-38 shows the peak fuel clad temperature occur' ring at about 33 seconds. Account for the 10 second difference between the peak centerline fuel temperature and the peak centerline control rod temperature.
4.
Overlay. a plot of maximum control rod temperature onto a plot of peak fuel cladding temperature.
Identify the location of these peaks in the core.
Shcw the plot through the reflood peaks.
5.
Provide the data base for the poison material melting point.
6.
Confirm that the analysis was conducted using 102% of 2772 MUT.
~7.
Discuss analytical methods and describe all calculations.
8.
Discuss the uncertainties associated with this calculation.
9'.
If calculated temperature are high enough,' discuss all as-pects of the consequences of molten poison material within the stainless steel clad. Address such items as phase changes and degradation of the control aspect of poison i
material, pressure buildup within control rods; flow blockage potential due to bowed, ballooned or collapsed control rods, and all local and gross core effects.
10.
For the preceding types of control rods, assess the potential for eutectic formation between dissimilar metals.
6.3.14 With regard to question 6.3.14, the response is incomplete.
What is the melting point of the three combinations of B C-Al 023 4
E.
With regard to the resconse to, ouestion,6.3.18, explain how the apparent analytical error resulted in a completely different worst break l 6.3.18 F.
size and worst break location.
6.3.19 With regard to question 6.3.19 (and in light of your clarifi-cation of the FSAR statement that a change of position' of valved l
H*
is c
considered incredible applied only to the core flood line isolation valves), confirm that the consequences of such a single failure as a valve change-of-state (simultaneous with an ace _ident) was corisiliered plant-wide.
- REQUEST NO.
Provide the analytical basis for all pressure setpoints associated 6.3.23 with the Core Flooding Tanks (600 psi for CFT actuation, 700 psi -for valve g-position alarms, 800 psi for valve interlocks).
l Describe the sequence of events which take place during the 6.3.24 automatic switch to the long-tenn recirculation cooling mode of operation. Provide a system description and state the rationale for K*
Describe the relative order in which these automatic actions occur.
any operator action required to shift into and maintain this mode.
6.3.25 A recent occurrence at Oconee allowed several feet o.f water to build up in an ECCS pump room and consequently jeopardized the avail-M.
ability of ECCS pumps. Evaluate and provide means for decreasing the potential for a similar occurrence in Davis-Besse.
Oces Davis-Besse have sump pump monitor alarms? Are the ECCS pump rooms watertight?
In a recent occurrence at Oconce, cavitation damage to ECCS 6.3.26 valves was reported due.to decay heat removal operation at certain flow
.N.
rates. Eval.uate and provide means for eliminating the cavitation problems associated with flow rates through these valves in Davis-Besse.
The staff has noted that Davis-Besse Figure 6-17 shows TWO 6.3.27 High Pres' ure' Inj'ection'Emps (550 gpm each), while THREE (of varying s
1 0.
capacities from 250 gpm to 700 gpm) are shown on such plants as Rancho j
Seco, Oconee, WPPSS-1, Greenwood, and Sellefonte.
Explain this variance in HPI design concept on Davis-Besse.
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15.0 ACCIDENT ANALYSES REOUEST NO.
15.1.2 With regard to question 15.1.2 (partial loop operation), the re-sponse is partially acceptable.
Provide the following additional A*
information:
1.
A plot of peak clad temperature versus time for the worst case LOCA during two-pump operation.
2.
For two-pump operations, evaluate the applicability of each generic sensitivity study presented in topical report BAW-10091.
3.
For the Unccmpensated Operating Reactivity Changes event, why were the assumed maximum reactivity iates the same for 2-and 3-pump operation, yet quite different from the rates for 4-pump operation on page 15-66 of the FSAR? Why did thd rate of average RCS temperature change for xenon builduo beccme less negative in going from 2-pump operation to 3-pump cperation, yet more negative in going from 3-pump operation to 4-pump operation?
Similarly, why did the rate of average RCS temperature change for xenon burnout decrease in going from 2-pump operation to 3-pump operation, yet the rate increased in going from 3-pump operation to 4-pump oper-ation?
4.
For the worst-case main steam line break, provide a plot of Total Reactivity (% 4 k/k) versus Time After Break (seconds) for 2-pump operation.
5.
Provide a plot of DilBR versus time for the rotor seizure during 2-pump operation. Justify the initial Dt BR.
15.1.6 Provide the information listed below (1-5) for each of the following transients or accidents.
a.
The worst case feedwater line break (See cuestien D)
'b.
The worst case rain steam iine break (See question J)
'c._,, Loss of offsite ocwer (See cuestion F)
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Plots of the following parameters versus time:
Steam generator pressure (affected and unaffected)
--MinimumDfiBR(W-3 correlation)
.g-PEQUEST NO.
-- Break flow rate
-- Safety and relief valve flow rates (primary and secondary
ystem) s
-- Mass and energy transfer within the containment (for breaks insidecontainment)
' - Turbine bypass valve flow rates
-- ECCS flow rates
-- Reactor coolant pressure and temperature
-- Thermal and neutron power Carry the precedi.ng parameters out to such a time period as to ensure that reactor conditions have sufficiently stabilized or abated.
Identify all trip setpoints on the plots.
2.
Credit for operator action to mitigate the consequences of these events is not acceptable unless analyses show that sufficient time exists for the operator to recognize the initiating event, ascertain the proper response, and perform the appropriate manual action.
In this regard, identify the exact manual operations required by the operator to meet the acceptance criteria stated in the FSAR and bring the plant to a final, stabilized condition.
Speci fy:
(1) the infor-mation available to the operator, (2) the time delay during which his failure to act properly will have no unsafe conse-quences, and (3) the consequences if the action is not perfonned at all.
3.
Substantiate the assumed worst single failure by a sensitivity study.
Examples include the auxiliary feedwater steam admission valve or the turbine by-pass system (bypass system actuation logic not single failure proof-page 10-18).
_ _ REQUEST NO.
4.
a.
Provide a table of key initial assumptions employed in each analysis.
Include in the table such parameters as initial power level, core flow, coolant system pressure at the core
.. outlet, core inlet fluid temperature, volume average fuel temperature, and secondary -system conditions (such as feed-water flow, steam flow, and steam generator pressure and temperature).
b.
Indicate the appropriateness of these values by comparing to the expected operating rances for Davis-Besse (for example, operating pressure (psig) = 2185 + xx).
5.
To complement the FSAR discussions, provide for each of these three events a su=ary of -a functional analysis of systems required. The'sunmary should be shown in the
.-form of simple block diagrams beginning with the event and branching out to the various possible sequences.
When complete, the diagram should clearly identify each system required to function d'uring any plant operating state. See the response to question 6.3.5 and subsequent staff co=ents.
15.1.7 Provide a list of all the transients and accidents conside. red in establishing the auxiliary feedwater system flow require-H.
ments and response. times. For each of these transients and accidents, state the maximum delay in the initiation of aux-iliary feedwater flow that can be tolerated, whether the initiation is automatic or requires operator action, and the minimum required auxiliary feedwater flow rate required to mitigate the consequences of the transient or accident.
I 15.1.8 For each accident and transieot analysed in Chapter 15.0 of the SAR which results in a reactor trip, the control rod worth curve 0.
(reactivity versus time) used in the analysis must be provided.
Show that the control rod worth curve is based on an axial power profile and peaking factors selected to produce the worst conse-quences from the events analyzed.
Relate the axial power limits and control rod limits specified in the Technical Specifications.
Reference the Technical Specifications which require the operator to maintain this flux shape over the life of the core and the maxi-mum allowed variation permitted in this flux shape.
t" i REQUEST NO.
Justify the quality of two-phase mixture exiting the steam e.
generator via the break and discuss why the blowdown could not be initially single phase (feedwater).
Relate this same discussion to breaks at other elevations on the steam generator (e.g., auxiliary feedwater line).
states that the reactor trips on high reactor Page 15.2.12-2 f.
system pressure (2355 psig setpoint from page 2.2-1 of Chapter Since the pressurizer electromagnetic relief valve 16.0).
setpoint is 2255 psig (response to question 5.2.3), analyze the consequences and extent of delay of the pressure trip due to the relieving capacity of this valve (and therefore more time for energy to build up in the core before reactor trip). State the capacity of this relief valve and provide the tolerances in,+ psig.
Include the uncertainty in + psig of the reactor trip pressure setpoint.
Also, include a discussion of the potential delay in reactor tripduetotheprobabjeactuationofthepressurizersprays.
g.
The explanation on cage 15.2.12-2 which compares the described break situation with the event analyzed in the FSAR is not detailed en'ough to be clear.
For example, the statement is made that for the described accident, the reducticn in the secondary system heat removal capability..." is not extreme since only the unaffected steam generator heat removal capa-city is reduced." Ocesn't the ccmplete loss of the affected steam generator reduce heat removal capacity?
~
This discussion also states that,'"Since the results of both accidents are similar, Figure 15.2.8-1 is sufficient to show the eventual effect of a feedwater line rupture uostream of the first feedwater line upstream check valve with offsite power available." This statement makes little sense because it conflicts with page 15.2.12-1 of the response which placed the break downstream of the check valve (in accordance with the staff question). Also,theFSAR(page15-53) states:
"The loss of normal feedwater due to a feedwater line break between the first feedwater line up-stream check valve and tne steam generator produces results no worse than the steam line break accident presented in Section 15.4.4."
This above conclusion from the submitted information is not obvious, especially after noting the following:
i
- The reactor trips on low pressure for the main steam line break, but on hiah pressure for the feedwater line break.
l
. REQUEST NO.
- The critical time period in the main steam line break shows a reactor coolant temoerature and pressure decrease, but the critical period in the feedwater break shows a coolant temp-erature and pressure increase.
With regard to the feedwater line break with a loss of off-2.
site power concurrent with the feedwater line ruoture (Case II):
The statement that the event is less severe than, but a.
similar to, the results of the station blackout (in lieu of the requested analyses).is not acceptable.
b.
Amend the FSAR to verify that the plant auxiliary boilers are not used to supply steam to drive the auxiliary feed-water system.
Di.scuss the term "high level control" which is stated to c.
be initiated by reactor trip on the secondary side of the steam generators to enhance natural circulation.
Provide a description and available data to substantiate and quan-tify this operating capability.
Subsection 5.5-2, " Steam Generators," does not appear to address this safety feature.
Is it required to mitigate the consequences of the feedwater line break?
d.
Page 15.2.12-3 of the resconse indicates that the unaffected steam generator pressure increases to the turbine bypass valve setpoint and the steam l'ine safety valve setpcint to remove decay heat.
However, page 15-57 of the FSAR states that for a loss of off-site power..." turbine bypass valve steam relief is lost due to the loss of power to the condenser circulation pumps." Please explain this apparent conflict and, if credit for turbine bypass was improperly utilized for the feedwater line break, re-assess the consequences of this event.
3.
With regard to the break analyzed with a loss of offsite oower at reactor trio- (Case III), the brief discussion provided is inadequate to quantify any consequences and leaves unanswered the concern as to whether or not this is a worst-case situation rslative to the status of offsite pcwer.
\\
, REOUEST NO.
15.2.13 The response to question 15.2.13 is partially acceptable.
With regard to the overpressure protection of the RHR system, provide a discussion of the design features which protect the RER system against overpres-E.
surization during shutdown (while the shutdown cooling system is functioning).
Include detailed analyses, with calculations, of needed Justify all worst case events RHR relief valve capacity, if applicable.
considered in the relief valve sizing analysis (see Chapter 15.0 events).
Include consideration of starting up with a " solid" pressurizer at the time of the pressure transients. Describe, with diagrams, the relief valve design and operation.
With regard to question 15.2.14 (loss of offsite power), the 15.2.14 Regulatory staff review of the computer code POWER TRAIN used F.
for this event (BAW-10070) is not complete. Should modifi-cations to this code be required, the impact of such modifi-cations. upon the consequences of this event will have to be assessed.
1.
The top of page 15-58 states:
" Excess steam is relieved until the reactor coolant system pressure is below the pressure corresponding to the lowest setpoint of the steam safety valves."
With regard to this statement, state the setpoints, and capa-city at these setpoints, of the steam relief and safety valves.
How is the reactor coolant pressure incorporated in the coerating logic of the secondary system safety valves (as indicated by
' the statement)? Describe the basis for this arrangement.
i 2.
Page 15-58 of the FSAR also states:
"The turbine-driven auxiliary feed pumps provide feedwater to the steam generator by taking suction frcm the condensate storage tanks and are driven by steam from either steam cen-erator."
With regard to this statement, it appears to conflict with the response to question 15.2.12 which indicates that for a feedwater line break with a loss of offsite cower the operator would have to rely upon the plant auxiliary boilers to supply steam to drive the auxilairy feedwater pumps.
Please clarify.
- s 15-l
. REQUEST NO.
t 3.
Comment on the validity of the statement in the introduction to Chapter 15.0 (page 15-2) that all systems _ utilized in these Chapter 15.0 analyses have been designed in a manner such that a single failure of an active compenent will not prevent then from meeting their performance require-ments.
15.2.15 With regard to question 15.2.15 (feedwater system malfunctions),
the response is insufficient to allow an adequate evaluation.
G.
Address the following comments:
1.
Page 15-59 states, "ftonnally operator or ICS action would correct feedwater system malfunctions, however, such actions were not considered in the analysis of this accident."
This statement conflicts with the response which indicates
~
that although credit for ICS was not assumed, the operator was required to control stean generator water level.
Correct this apparent inconsistency.
If operator action is not recuired, provide the analytical consequences of this event using the protective systems referred to in the following s+atement (page 15-59 of the FSAR):
"Only the low reactor coolant pressure and high neutron flux trip were used in the analys!s to assure reactor protection."
If the existing FSAR plot: are intented to represent this situation, then the description on page 15-60 should be clarified; specifically, the following statements are confusing if no credit for ICS or operator action is assumed:
"Without temperature compensation of the feedwater flow the steam generator level rises to the high level limits where feedwater flow is reduced to prevent flooding of the steam generator."
And for the feedwater flow malfunction:
- The steam generator level rises to the high level limit where feedwater flow is reduced."
REQUEST NO.
To repeat for clarification, how is feedwater flow reduced for the analysis which states than no ICS or operator action was assumed?
2.
Provide the follo4ing additional parameters versus time for the feedwater malfunctions.
The analyses should be performed at 102% power (2% to account for power uncer-tainties):
- Minimum DNBR
.- Pressurizer water level
-- Main feedwater flow rates
- Turbine bypass flow rate
- Steam generator water levels
- Safety and relief valve flow rates (primary and secondary. system).
Identify all trip setpoints on the plots.
15.2.16 The response to cuestion 15.2.16 is partially acceptable.
Provide the specific analysis, and properly classify in H.
the FSAR, the inadvertant opening of a cressurizer safety or relief valve (highest caoacity).
State the acceptance criteria, the computer model utilized, and all initial conditions and assumptions.
Provide the parameters versus indicated in part 1 of question K.
15.2.17 In the response to question 15.2.17, the statement is made that:
I.
...it is a routine safety analysis assumotion that unless an action is guaranteed by the orotection system it does not occur. Therefore, no credit is taken for runbacks, interlocks, etc."
T
,.- _ REQUEST NO.
of less than 1.3 was assumed to be released to the reactor coolant. Provide the percent fuel rods involved in this release.
3.
Discuss the conservatism in the initial DNBR shown in Figure 15.2.5-5 (abcut 1.9) and compare the initial value assumed for the 4-pump coastdown event. Justify any difference in assumed values.
15.4.1 With regard to question 15.4.1 (the steam line break), the re-sponse is insufficient to allow an adequate evaluation:
1.
The analysis assumes credit for the shutdown baron addition of the HPI System. Credit for this additional shutdown margin from the HPI System is not acceptable unless it can be shown that the assumed portion of shutdown reactivity contributed by the HPI System would be injected at the times assumed.
Along these lines:
Justify the conservatism of the 25 second ECCS delay, a.
noting that the ECCS delay assumed for the recent generic LOCA analyses (BAW-10091) was 35 seconds.
Discuss also the breakdown of this time from the beginning of the event (time zero) to actual injection of the HP boron into the core (setpoint delay, pump start delay,, transport delay, etc.).
b.
Show that the HPI actuation setpoint, high containment pressure (4 psig) or low reactor coolant pressure (1600 psig), is reached for breaks inside or outside the con-tainment and specify the time to reach these setpoints.
Choose conservative assumptions, such as maximum heat sinks to delay the containment pressure rise to 4 psig (for a break inside containment).
For breaks outside containment, the time to the low reactor coolant pressure HPI initiation time shculd also be conservatively accounted for in the analysis.
2.
Provide the minimum reactivity margin for the following five main steam line break situations.
Specify the worst single active component failure for each case:
Case I - 102% power.
Breakisinsidecontainment(36"line).
No offsite power.
Case II - 102% power. Break is inside containment (36" line).
Offsite power is available.
s
.- _ REQUEST NO.
Case III - 102% power.
Break is outside containment and uostream of isolation valves.
No offsite power.
. Case IV - 102% power. Break is outside containment and downstream of isolation valves.
Offsite power is av.atlaol Case V - Hot standby or low power operation.
Choose other assumpt. ions based on worst-case above.
3.
Page 15.4.1-2 of the response indicates that, for the steam break situation analyzed, the reactor trips on low pressure at 1.13 seconds after the rupture. This appears to be a significantly faster time-to-trip than other more recent B&l reactors.
Provide a breakdown of this time, including.all delays.
4.
The analyses indicate that, deperiding on break size, the reactor will trip on either high-flux or low-reactor pressure.
Provide a graph of time to reach high flux trip point versus break size, and. time to reach low pressure trip point versus break size (both on the same plot).
5.- We note that additional protection system trips have recently been added to mitigate the consequences of the steam line break (Main Steam Line Rupture Control System).
The statement that the original FSAR analysis would be more severe (in lieu of a new analysis) is not acceptable. A new worst-case analysis should reflect the current design of Davis-Besse.
6.
With regard to part "e" of question 15.4.1 the requested de-finition in terms of line size should be provided (e.g., at what point does a " minor" steam line break become a " major" steamlinebreak?).
7.
The response to question 15.4.1 states that periodic full-closure testing of the tuibine stop valves will disclose any sticking conditions, so that a shut down could be made to make the necessary correction.
Pruidecrreference this surveillance requirement in Chapter 16.0.
8.
Clarify the need to include the feedwater control valves and stop valves in the Main Steam Line Rupture Control System.
! Are both these valves safety grade? State their required closure times a'nd the bases for these times.
s A REQUEST NO.
9.
Page 7-52a describes the Main Steam Line Rupture Control System.
Why is the trip of the turbine stop valves not included in this discussion? Where is the need and specifications for this trip addressed in the FSAR?
- 10. Page 15.4.1-4 of the answer to question 15.4.1 discusses and quotes pressure drops across a main steam line isolation valve.
However, the question was directed at the non-return check valves during accident conditions.
Provide the design capability of the non-return check valves, as requested.
15.4.2 With regard to question 15.4.2, the response to part "b" is insufficient to allow an adequate evaluation. The question Q*
refers to the statement on page 15-11-8 of the FSAR that, "An additional scurce of fission product leakage during the maximum hypothetical accident can occur from leakage of the engineered safety features external to the containment l
vessel during the recirculation phase for long-term core cooling."
The response refers to Tables 6-15 and 6-16 which do not appear applicable.
Provide an interpretation of these Tables with regard to leakage. locations, flow rates, and leakage detection instrucentation.
Include a description of any operator actions that are required witn the time needed for the action.
15.4.7 What initiates the auxiliary feedwater system for a main steam line break and a single failure of the auxiliary feedwater pump on the unaffected steam generator (e.g., what autcmatic signal N.
J would open the correct crossover line between the two auxiliary feedwater system)
m
. _ REQUEST NO.
15.4.8 To assess' the potential severity of a steam line break inside of containment that results in both steam generators blowing down, P.
provide the results of an analysis that assumes the single active failure that results in the most severe consequences regarding core thermal limits.
For example, a single failure that results
~
.in the opening of one or more of the steam dump valves in the steam generator not supplying the broken steam line.
Clearly state all assumptions used in the analyses, including the time in core life and baron concentration in the reactor coolant.
Justify the selection of the single failure assumed in your analyses.
Present plots of the following parameters as a function of time:
- 1. Neutron power level
- 2. ~ Minimum D 3R(N-3 correlation)
- 3. Total Reactivity
- 4. Averarte core moderator temperature
- 5. Reactor coolant system pressure
- 6. Water level in the pressurizer
- 7. For each steam generator:
a.
reactor coolant outlet temperature b.
steam pressure c.
feedwater flow rate 8.
Heat flux (average and maximum)
~
9..
Fuel temperature (average and maximum)
w q
l i
REQUEST NO.
15.4.9 What instrumentation would be relied on to single out a steam generator tube failure as the cause of an event 0*
so that the reactor operator would know that the required action at 20 minutes must be accomplished? Our concern is that other possible events, e.g. a small pipe break LOCA for which no operator action is required, would be incorrectly diagnosed by the operator. The operator could then fail to achieve the proper manual action at 20 minutes. Why does the initial discharge out the break (Table 15.4.2-1; 435 gpm) differ from WPPSS, even though steam generator tube diameters appear identical?
It is noted that the number of rods expected to experience DNB 15.4.10 for the control rod ejection accident increased from less than T.
6% quoted in the PSAR (Figure 14-28) to 45% in the FSAR (Figure 15.4.3-9). Why?
l Also, please explain the convex nature of the FSAR curve when comparing to the concave trend of the PSAR curve for Davis-Besse, i
Bellefonte, and WPPSS.
16.0 TECHNICAL SPECIFICATIONS 1.
Specification 2.1 - Must consider the effects of fuel densification.
The allowed power levels for partial loop operation in Figure 2.1-2 C.
(3'-pump and 2-pump) do not agree with FSAR Figures 5-10 and 5-11.
The relieving capacity of each pressurizer code safety valve (page D.
3.1-2) does not agree with FSAR page 5-14a (300,000 lb/hr versus 336,000lb/hr).
states that, "Two core flooding tanks erth Specification 3.3.1.2(3)of borated water at 600 +25 psig shall be E.
containing 1040 +30 ft 3
available." ConTirm that values of 1010 ft and 575 psig were assumed in LOCA analyses to justify the operating ranges.
..