ML19319D301
| ML19319D301 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 12/04/1975 |
| From: | Stello V Office of Nuclear Reactor Regulation |
| To: | Moore V Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8003130952 | |
| Download: ML19319D301 (5) | |
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Voss A.'lloore Assistant Director for LWR's, Group 2, RL
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REQUEST FOR ADDITIONAL ECCS I!GORMATIO:I We are currently evaluating the submitted ECCS analyses for Crystal River 3..In addition to Topical Report BAW-10103, "ECCS Arnlysis of B&W's 177-FA Lovered-loop NSS", we are reviewing the cor. tones of the latters from Flarida Power Corporation dated September 19, 1975 and October 15, 1975.
In addition to questions which were subnitted to B&W pertaining to EAU-10103. enclosed is a list of e&litienel infomation time is required for our reviev.
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gOf FOR ADDITIONAL INFORMATION
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_C_RYSTAL RIVER 3 1.
With regard to the potential for boron precipitation:
-a)' Page 2 of Florida Power Corporation's letter indicates that the vulnerability of dilution Modes I, II, III, and IV to single failures is analyzed in Section 10 of Topical Report BAW-10103.
Further information is needed from Florida Power to confirm the single failure position stated in BAW-10103. It is the staff's position that bbde I should not be atte=pted as the primary method to control hsran concentration in the core during long-term cooling.
The possibility of gas or steam entrainment in the decay heat suction nozzle can result in severe damage to the decay heat removal pump.
Long-term heat removal requirements can exist for bng durations (days or months) after the accident and continuous operation of one train of the decay heat removal system is required.
In the event of en equipment malfunction in this train, no =ethod is available to remove the decay heat if the other train'has been previously da= aged.
Therefore, i=ple=entation of Mode I should not be atte=pted since this action could result in the decrease of required safety equipment.
Since initiation of ibde I is not allowed, it must be established that Modes II, III, and IV in combination are single failure proof.
An alternate to consider is the feasibility of utilizing a single = ode, with modifications to take this = ode single failura proof and relatively
, free of pu=p cavitat!.on problems. We suggest you examine the possibility of maintaining containment sump suction for the LPI pumps, while -utilizing gravity drain to the containment su=p.
To achieve this drainage, s=all lines may have to be installed Jrom the DhR drop line to the containment. sump (with suitable motor-operated isdlation valves),
b) Provide the specific operating procedures required to implem,ent each of these dilution = odes.
Indicate each operator action necessary, and identify which valves cust
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be manually operated outside the control room.
Confirm the
~ apability for the operator.to close or open these valves at cthe time specified in the' September 19th letter-(to be instituted 15 minutes af ter a LOCA) and evaluate the radiation levels which could be present at the valve locations.
c) Florida Power Corporation indicates that a backup electrical supply,is present to run the pu=ps and actuate the motor-driven valves for all modes in case of electrical failure (see reference to BAW-10103).
List all pumps and valves included in'the dilution modes with their associated power supplies. Specify their power requirements and relate to the capacity of the backup power supply.
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. Indicate the location of this power supply in the plant.
Specify those motor-operated-valves in the dilutien modes which, if rendered
-inoperable due to a-common power supply failure, would fail all dilution modes. Discuss your means to connect power to these valves
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should: chey be inside containment.
n, 2-d)
Confirm that remote readouts of dilution flow rates for each mode are available in the control room.
e) -Discuss the capability to test the dilution modes.
f)
Explain the type of valves that DHV-39 and DHV-40 are, and identify their mode of actuation.
g) Provide the elevations of the piping and other components in the decay heat drop line from the hot leg nozzles through each'of the trains to the reactor building sump.
This infor-wation uill verify that gravity draining for boron dilution for Mode II is possible.
h) Mode III: Justify the 40 gpm predicted to occur through the auxiliary pressurizer spray line.
Since water is being pumped from the containment floor, discuss the ability of the spray to pass all material admitted through the sump screens.
1)
Indicate the feasibility of monitoring boron concentration levels during the long term.
2.
With regard to the single failure analysis; a)
Confirm that a single failure or operator error that causes any motor norated valve to inadvertantly actuate could not adversely affect the ECCS (i.e., Service Water System Valves, Building Spray System Valves, Boron Dilution Valves, CFT Vent Valves, etc.).
b) Provide a list of each valve considered under item a) and indicate the consequences of the spurious actuation.
c)
FSAR Figure 9-6 shows LPI valves DH-V4A and DH-V4B to be normally closed.
To allow low pressure injection subsequent to a CFT line break and a single. active component failure, these valves must be required by Station Technical Specifications to be open, power removed, and breakers locked o'en.
Station Technical Specifications p
must also require a periodic test be performed to warn of abnormal leakage of the check valves in the LPI injection lines inside containment and that this test be performed at least annually.
These changes provide assurance that abundant core cooling is available for a CFT line break and further minimize the potential for a LOCA outside containment.
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3.
It is noted that motor-operated valves DH-5A and DH-5B from the BWST are shown normally closed.
It appears that, assuming sufficient static head were available, the potential for a water hammer when ECC is injected into a drv line would be reduced considerably if these valves were normally left open.
Please discuss.
Also, confirm that pariodic venting of ECCS lines and ECCS pump casings are perforced and reference this surveillance requirement in your station technical specifications.
With regard to the potential for submerged valves, the September 19th'
-4.
letter indicates that there are no valve motors which would be flooded that are required to operate following a LOCA. Provide the level of water assumed after the LOCA. Provide the elevation of the rotors for DHR suction valves DHV1, DHV2, and DHV3.
Confirm that any other equip-ment essential to ECCS perfor=ance which could be submerged would not be adversely affected.
5.
With regard to the partial loop analysis (Reference to BAW-10103):
a) No discussion was offered as to the consequences of a break in an active cold leg of the fully active loop.
b) Technical Specifications will prohibit 2-pump cperation unless an analys:s is provided to support this mode of operation.
Compare a break in the inactive cold leg to a break in the active cold icg.
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Indicate and justify the worst-case pump status assumed at the time of the LOCA (tripped vs. povered).
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's d) Provide assurance that the PCT versus Break Size curve in BAW-10103 would not be significantly altered by either mode of partial loop operation.
e) Submit the LOCA parameters of interest identified in the
" Minimum Requirements for ECCS Break Spectrum Submittals,"
dated April 25, 1975.
f) The statement is made that the load ratio between the steam generators is changed to 2.27:1 by control in the feedwater flow.
Discuss the sensitivity of peak clad temperature to the load ratio.
g) The analysis states that the ruptured node cladding temperature decreases rapidly after rupture because of the reduced gap heat transfer from the fuel to the cladding and the increase in surface area for cooling.
It further states that the reflooding heat transfer coefficients are high enough to prevent a rise in the ruptured cladding temperature after rupture.
'These statements do not appear to agree with the temperature transients in Figures 1 and 2.
Explain why the rapid decrease in cladding temperature after rupture (shown to be over 600*F within 5 seconds for 4-punp operation) did not occur for 3-pump operation. Also, explain why it is stated that the reflood heat transfer coefficients prevent a rise in the ruptured cladding temperature after rupture even though over a 100*F rise occurs in Figures 1 and 2 at the time of rupture.
h) The analysis states that the peak linear heat rate for the hot bundle is the maxicum kw/ft LOCA limit as shown in Figure 2-2 of BAW-10103 at the six-foot elevation for this code of operation.
Clarify the effect of such an assumption on Station Technical l
Specifications.
If this assumption would allow 3-pump operation at this kw/ft, discuss the consequences of the abnormal events in Chapter 15.0 should they occur during three (or two) pump operation.
6.
Provide your schedule for submitting revisions to the proposed Technical Specifications affected by the LOCA ana'.ysis.
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