ML20064G357
| ML20064G357 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 01/05/1983 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8301110444 | |
| Download: ML20064G357 (14) | |
Text
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,f t TENNESSEE VALLEY AUTHORITY CHATTANOOGA. TENNESSEE 37401 400 Chestnut Street Tower II January 5, 1983 Director of Licensing Attention:
Mr. Domenio B. Vassallo, Chief Operating Reactors Branch No. 2 U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Vassallo:
In the Matter of the
)
Docket Nos. 50-259 Tennessee Valley Authority
)
50-260 50-296 In response to your letter to H. G. Parris dated October 26, 1982, subject "10 CFR SO Appendix R,Section III.G," we have prepared the enclosed information.
This information addresses the seven questions contained in your letter and incorporates supplemental information as a result of subsequent conference calls with your staff.
If you have any questions, please call Jim Domer of my staff at FTS 858-2725.
Very truly yours, TENNESSEE VALLEY AUTHORITY o
L. M. Mills,. Manager Nuclear Licensing Sworn }ognd.subso ibed before me thi (.).sa day of
. 1983 a4A PER Notary Public d
My Commission Expires 7'"
Enclosure 00: See page 2 8301110444 830105 PDR ADOCK 05000259 F
PDR I
An Equal Opportunity Employer J
. Mr. Domenio B. Vassallo January 5, 1983 oo (Enclosure):
U.S. Nuclear Regulatory Commission Region II ATTN: James P. O'Reilly, Regional Administrator 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 Mr. R. J. Clark Browns Ferry Project Manager U.S. Nuclear Rogulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20814 l
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e ENCLOSUHE Hequest for Additional Information Post-Fire Safe Shutdown Capability firowns Ferry Nuclene Plant Unita 1, 2, and 3 Dockot, Nos. 50-2S9, S0-260, and 50-296 Quention 1 In your proposed method of alternate safe shutdown, A.
How much of the core will be uncovered and for how long? How does this comparo with the extent of core uncovery using normal shutdown (noilities?
11. What is the predicted punk totus water temperature?
Hosponnn 1 A.
In a telecon betwoon NHC nnd TVA on November 15, 1982, TVA agreed'to provido additional information for the case or manuni blowdown after ten minuten using four SHVs from ther backup control panel with low pressure makoup consinting or one LPCI pump only. The results presented in attachment 1, which woro obtained from NEDO-24708 entitled, " Additional Information Hequired for NHC Starr Cenoric Hoport on !! oiling Water Henctors," are indicative of this caso. The water level inside the shroud in shown to approach the TAF with recovery provided with low pressure makeup using ono LPCI and two core spray pumps. The only dirrerence between the subject cano and t.he caso presented in attachmont 1 is the rato nt which the water level in recovered. Due to the comparatively large contribution to level recovery by the 1.PCI pump, the water level inside the nhroud is expected to stay near or above the TAF. 1r any core uncovery in experienced, it would be minimal, and surricient rerlood capability is easily provided by one LPCI pump to recover and maintain the water level nhovo the core bororo any coro damage occura. Theno results are nimilar t.o neveral events which require emergency deprennurization using the relint valves in manuni or automatic modo.
11. What is the predicted peak torun wat,er temperaturo? This question rorers l
to the caso analyzed for timo greater than one hour.
The ponk calculated bulk torun water temperature in 197.5 F which occurs at approximately 10.S hourn into the transient. The nasociated subcooling (difference between local pool tempernt,ure and saturation temperature at tho T-quenchor submergence) will be in excess or 30 F.
Comparison with HUHEG-07113 indinnten that smooth and stablo nteam condensat,lon would be expected to extnt, under thone conditions. Thun, we bellovn that, wo are in complianco with NkC requiremontn.
i l
. Question 2 Will the proposed addition of manual safety relief valve controls to the new backup control panel preclude spurious SRV operation as a result of a fire in the A1, C1, and E1 shutdown board rooms?
Response 2 No.
The fire scenario for board rooms A1, C1, and E1 is given on sheet 2-4, paragraph 2 of the TVA appendix R submittal dated June 30, 1982.
It says that, ".
. it is possible that some S/R valves can spuriously operate."
Question 3 Have the following systems been considered in the alternate safe shutdown analysis? If not, provide justification for not including them.
A.
Response 3 A.
Yes. HPCI and RCIC were considered as an alternste means of safe shutdown as shown on the shutdown logic diagram (figure I.a);
however, because a low pressure system (RHR) is needed for torus cooling, it was decided to also use low pressure systems in conjunction with ADS for core cooling to minimize equipment required.
Therefore, detailed studies regarding HPCI or RCIC availability were 3
not performed and no credit is taken for these systems in the analysis.
B.
Yes. The emergency diesel generators (DGs) were considered, and it was shown that a minimum of four DGs would be available regardless of the fire area. The physical separation of the eight DGs (i.e., four at one side of the reactor building and four at the other side) precludes any power and control cables for one set of DGs, being involved in a common fire with the cables of the other four DGs. The automatic logic and support systems were evaluated in attachment 2, "BFN Diesel Generator Automatic Start and Diesel Fuel Supply System Evaluation for Appendix R."
Question 4 Are the following process instrumentation parameters available at the backup control station?
A.
Torus temperature and level, B.
Condensate storage tank level.
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.. Response 4 A.
Yes. These instrumentation parameters are available at the backup control panel. The torus water level was not initially required since the connections to the torus were analyzed to ensure there was no possibility of inadvertently draining the torus. Ilowever, consideration was not given to the possibility of overfilling the torus. After examination, there is only one major source of water that could possibly increase the torus water level. The condensate storage tank (CST) could drain into the torus if the HPCI and RCIC miniflow valves were inadvertently opened. This would be a very small leak and would take a considerable amount of time to affect the torus water level. Therefore, the following instrumentation will be available in the main control room (MCR) and backup control panel (BCP) for the given fire areas.
Fire Area Instrumentation Location RB/DGB CST Water Level MCR SBR CST Water Level MCR BBR Torus Water Level MCR TB/IPS Torus Water Level MCR CB Torus Water Level BCP B.
No.
CST level is not required at the BCP because it is not a source of makeup water to the vessel for the appendix R analysis.
Question S For the mechanical shutdown systems, have you considered the possibility of one pump in any system being out of service for maintenance at the outset of the fire? If not, provide justification for your basis.
Ronponso S No. The possibility of one pump in any system being out of service for maintenance is governed by the limiting conditions for operation specified in the technical specifications. Therefore, no considorations were given to equipment being out of service at the outset of the fire. Note that this is consistent with the way other events are handled. A fire occurring while one train of mitigating equipment is out of service is unlikely, and a fire in a particular area that takes out all mitigative equipment except one train in conjunction with the remaining train being out for maintenanco is extremely unlikely.
Question 6 In your submittal, sheet 4-84 is not consistent with sheets 4-85 and 4-86.
It appears that information is missing.
.. Response 6 There is a sheet that was inadvertently omitted following sheet 4-86.
. Attachment 3 is the missing correctit action sheet.
Question 7 On pages 4-111 and 4-112, you stated that new alternate power supply breakers may be required for the 480-volt reactor MOV board B and I&C power for shutdown and battery boards. What is the status of this potential change?
Response 7 The pages referenced in your request were 4-111 and 4-112.
These page numbers were revised during our telecon on November 9, 1982 to 4-108 and 4-109. You requested the status of the new alternate power supplies to 480-V RMOV board B and I&C power. The corrective actions are referring to manual transfers to alternate supplies which were already present. No modification was required.
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e T!rl e sei.IIIIr:at ion involves wrappine,the division i condicits (1isted below) containing,cabics shown on RIH-IT-TB-RilRSW-0, sheet I, for NilRSW pungm Al, A2, C1. and C2 with a one hoiar
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fire r.1tc<! harrier in tie intake pi:epInc, st at ion, viere 20-Iont ncp.1:ation r,r n three-isour IIrc-rated barrier does not eri=t 1.etween redundant divinfon'TT cable trays for RilRSW
-L a
pt ps R I. Pl2, til and n2.
I reno.3 i t s : ES75-T ESfiS-1 lS100-I ESil3-T ev n.
4 I
or l'
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m t
a i
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+
APPENDIX R
I C O R ii E C T I V E-AC II O N:
1' ' ' 1. I '!
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t
. _ff% _hf:s/o'f-...
R.virwr.n:
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i 1
B F N - 1 T - T B - R'll R S W - 0 '- M 1
.