ML20054J597

Responds to NRC 820430 Request for Addl Info Re Questions CS760.24,29,41,45,77,79,98,134,137 & 138.Responses Will Be Incorporated Into PSAR Amend 69,scheduled for Submittal in Jul 1982

ML20054J597
Person / Time
Site:	Clinch River
Issue date:	06/25/1982
From:	Longenecker J ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT
To:	Check P Office of Nuclear Reactor Regulation
References
HQ:S:82:055, HQ:S:82:55, NUDOCS 8206290320
Download: ML20054J597 (33)

Text

-

Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:82:055 JUN 25 G2 Mr. Paul S. Check, Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comission Washington, D.C. 20555

Dear Mr. Check:

RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION

Reference:

Letter, P. S. Check to J. R. Longenecker, "CRBRP Request for Additional Information," dated April 30, 1982 This letter formally responds to your request for additional information contained in the reference letter.

Enclosed are responses to Questions CS760.24, 29, 41, 45, 77, 79, 98, 134, 137, and 138; which will also be incorporated into the PSAR Amendment 69; scheduled for submittal later in July.

incerely, J n R. Longene er Acting Director, Office of the Clinch River Breeder Reactor Plant Project Office of Nuclear Energy Enclosures cc: Service List Standard Distribution Licensing Distribution 90 0I E206290320 820625 PDR ADOCK 05000537 A PDR

P:ga - 14 (82-0358) [8,22] #89 Ouestion CS760.24 i Please provide decay heat data as a function of core region and lifetime with their respective uncertaintles.

Resoonse It should be noted that S'ection 15.1.4 (Ef f ect on Design Changes on Analyses of Accident Events) bss been added to the original PSAR to reflect the ef fect of design changes. Section 15.3 has not been changed in technical content since It was originally written in 1974. The worst case Section 15.3 undercooling trenstent has been updated and the analyses incerporated into Secti on 15.1.4.1. The current decay heat rates used in this update are included in the following information. If the 1974 decay heat data is of Interest, it too can be supplied upon request.

The decay heat data used in the undercooling design event analysis presented in PSAR Section 15.1.4 are provided in Tables QCS760.24-1 through -5. Data are presented as a f unction of time af ter shutdown including uncertainty with the associated uncertainty value provided f or each time point. Data are provided for the specific assemblies selected for hot channel analysis (see PS AR Figure 4.3-3) . Decay times ranging from shutdown out to 500 seconds af ter shutdown were considered. The decay heat values are based on the heterogeneous core design in which the f uel and Inner blanket assemblies have reached 2-year burnup and the radial blanket assemblies have a 3.2-year burnup.

QCS760.24-1 ,

. /-

Table QCS760.24-1 CLI NCH RIVER SINGLE AS SEM8LY DECAY POWER V ALUE S h/ UNCE RT AI NTIE S -

CR BRP HETE10GENEOUS CORE FdEL ASSEMBLY NO. 52 .

TIME AFTER DECA Y UNCERTAINTY S HUT D 0hN PO WE R (PERCENT) *-

.}

(SECONOSI ( KI LOWA TT S) , *l

0. 2.682E+02 3.242E+01

'[

2.0000 E+00 2.3 79E +02 2. T 42 E+ 01 *

4. 0 000 E+00 2. 201E +02 2 507E+01 . ,

• 3l

6. 000 0 E + C0 2. 090E +02 2.370E+01 , '

2.2I4E+01 8.0000E+00 2 009E +0 2 -

i!

1.0000E+01 1 945E +02 2 200 E+01 '

1.5000E+01 1 826E +02 2 065E+01

• i 2.000 0 E +01 1 739E +02 1 96TE+01 -

3. 0030 E+01 1 615E+02 1 829E+01 . ,,

i 4.0000E*01 1 52 fE +02 1 729 E+01 ~

6. 0000 E+ 01 1. 40 4E +0 2 1.592E+01 -

a 1.496 E+ 01 t 8.0000 E +01 1. 319E +02 .

O 1. 25 5E +02 1 426E+01 M 1.0000E+02 -

i' 1.2000E+02 1 205E +02 1.3 73 E+ 01 P 1 334E+01

~ -

. I 1 4000E+02 1 165E +0 2

% 1. 6000 E+ 02 1 132E +02 1 305E+01

~ '

4 1 8000E+02 1 104E +02 1.2 83 E+ 01  !

3 2.0000 E+02 1 080E+02 1.2 66E+01 1 254E+01 $

2 2000E+02 1 059E +02 5 2 4000E+02 E. 041E +0 2 1 244E+01 -

2.6000E+02 1 024E+02 1.237E+01 -

l' 2.8000E+02 1 009E +02 1.232E+01 ' '

3.0000E+02 9. 956E +01 1.229E+01 '

~~

4 0030 E+02 9. 409E *01 1.218E+01 .

8.993E 601 1.210 E+01 . I,

5. 0000 E+02 *

' 1 4

> l C

GIE '* .

' ~

,j I

- 3 Table QCS 760.24-2 ,

CLI NC H RI V ER S IN GL E AS SE MB LY D EC AY POWE R ,

VALUES W/ UNCERTAINTIES CRBRP HET EtOGENEOUS CORE FUGL ASSEMBLY NO.101 '

TIME AFTER DECA Y UNCE RT AINT Y ,;

SHUT 0 0WN POWE R (PERCENT) ,

(SECON05) ( KI LOWA TT S) '

O. 3. 393E +02 3.231E+01 ';

2.0000E+00 3. 011E +02 2. 732 E+ 01 4.0000E+00 2. T86E +02 2 497E+01 .

i 6 0000E+00 2. 64 7E +0 2 2.36LEt01 , [.

8.0000E+00 2. 545E +02 2.2 64 E+ 01 * '

1.003 0 E+01 2. 46 4E +0 2 2.191E+01 1.5000E+0L 2 313E +02 2 056E+01 . *;

2 0000 E+01 2. 203E +02 ' 1.9 58E+01 " *8 2 047E +02 1.820E+01 3.003 0 E+01 ,

1.936E+02 1.7 21 E+01  ?

l @ 4.0000 E +01 '

6 0000E+01 1. 7d 1E +0 2 1.5 83 E+ 01 .  :

O 1.488E+01 -

8 8.0000E+01 1. 6 73E +02 , ,

1. 0000 E* 02 1 593E +02 1.41TE+01

• m 1.3 65 E+ 01

,I

? 1 2000E+02 1 53 0E +02 *

" 1. 4000 E+02 1 4 79E +02 1.326E*01 }:

1.297E+01 'l

1. 6000 E+02 1. 43 7E +02 .

1.2 75 E+ 01  ;'

1 8000E+02 1 402E +02 2 0000 E+02 1.3 T2E +02 1.258E+0L ,.

8 2 2000E+02 1.345E*02 1.246E+01 1.237E+01  !

2 4000E+02 1. 322E +02 1.2 29 E+ 01 l.

2 6000E+02 1. 301E +0 2 2 8000E+02 1 282E+02 1.2 24 E+ 01 f 3.0000E+02 1. 265E +02 1.221E*01

• 4.0000E+02 '1.196E +0 2 1 209E+01 .

5.0000E+02 1 143E +02 1.201E+01 f L

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c s { =^ WuW W mWd E I W WWWWWWW NemmediO 8WWWW NmeON mmN W $ W W WNM $ W:o s E W W $

M Z O W O O N N 1lP M m M N *= to m m 9 :D 4 0 & M W N P eD to ? N N OOAJP5mmtgmNMMo@mNONetmMO@@mN

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- Z 3 w O e.e m O f @W Wg m OOOOMMMMMMMNNNNNNNNNNNNN OOOOOOCOCCOOOOOOOCC00000 gs W4 D ZWZe WM20 0WOZ

++++++++++++++++++++++++

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>> 0400 000000000000000000000000 ame g WW$O per wrzo W O8O8OOO8nOOOOOOOOOOO8OO8 0 0 000 c00000000000 00 e

z WMaw e Oe Oe O. O. e O m. e O. eO O. eO e O. e oOe N o 4e de e. O o N < 4to O. O. O u 2 >= O N 4 & @ M M N M 4 etD M M *4 M =4 N N N N N m 4 m

<Z M n d E g

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QCS760.24-4 Amend. 69 July 1982 f _ _ __ _ _ _ _ ___

'. Table OCS760.24-4 .

. ,j

CLIN;H RIVER SINGLE ASSEMBLY DECAY POWER . .

V ALUES N/UNCERT AINT IE S . ,

• l

. ~

CR8RP HETEROGENEDUS CORE RADI AL BLANKET POSITION NO. 201 '

TIME AFTER DECA Y UNCERTAINTY

( P ERCEt* T )

SHOTD0hN PO WE R

( SECO NDS) ( KI LOWA TT S) .

'l '

1.196E +02 2.936E*01 -

0.

2. 000 0 E +00 1.050E+02 2 421E+01 *'. -

l

4. 0 00 0 E+00 9. 684E +01 2.177E+01. '

• i '

6.0000E+00 9. 200E +01 2.035E+01 .

- l 8.0000 E+00 8. 854E +01 1 9 34E+01 -

1.857E+01 1 0000E+01 8.585E+01 ~

• 1.5000E+01 8. 090E +01 1.715E+01 .

2.0000E+01 T. 738E+01 1.615E+01 -

/

a 3 0000E+01 7.244E +01 1 4 73 E+01 ' '

6.8995+01 1.3 75E+01 - '? I O 4.0000E+01 3 l d 6.0030E+01 6. 422E +01 1 239E+01

'. I i P! 8.0000E+01 6. 093E +01 5 84 BE +01 1.147E+01 1.0 79E+ 01

~! l

$ 1.0000E+02 ij 0 1.2030E+02 5. 656E +01 1.029E+01

• l 1.4000E+02 5 500E *01 9.912 E +00 ,

9.626E+00 l l 1.6000E+02 5. 372E +01 l 9.409E+00

5. 262E +01 1.8000E+02 *

l 2 0030E+02 5.168E +01 9.2 45 E+00

• 2.2000E+02 5. 085E +01 9.117 E+ 00 ' .

2.4000 E +02 5. 011E +01 9.014 E+ 00 .

8.9 33 E+00 -

2 6000E+02 . 4.945E +01 .

2.8000E+02 4.684E+01 8.8 71 E+ 00 3.0030E+02 4. 829E +01 8 825E+00 j'

4.0000E+02 4.601E +01 8.6 61 E+ 00 i 8.549E+00 I

5.0000E+02 4. 423E +01 .

a

. l bf. k Mg -

66 03to 1

N g

'.j

. f

. Table QCS760.24-5 .

.I CLI NCH RIVER SINGLE ASSEM8LY DECAY POWER *:

V AL UE S w/ UNCE RT AI NT IE S ,

CRBRP HET E40GENEUUS CORE M ADI AL ULANKET POSI TION NO. Zo3 TIME AFTER DECA Y UNCERTAINTY S HUT D 0hN P0 HE R (PERCENT) -

'[

( S ECO ND S) ( KI LUWA TT S) .

1 0.

2 0000E+00

9. 938E +01

8. 75 7E +01 2.8 54 E+ 01 2.351E+01

]  ;

4. 0000 E + 00 8.096E+01 2.111E+01 . ,

1 9(2E+01

6. 0 000 E + 00 7. 704E +01 .

8. 0000 E +00 7. 423E +01 1 8 73 E+01 -

1 0000 E+01 7. 205E +01 1.797E+01 .-

• 1.5000E+01 6. 803E +01 1.660E+01 2.0000E+01 6. 516E +01 1 561E*01 . ,l

3. 0000 E *01 6.114E+01 1.423E+01 ' i

4. 0000 E+ 01 5. 832E +01 1.326E+01 fR *

$ 6.0000E+01 5. 44'2E +01 1.194E+01 *'

j 5.173E +01 1.104E*01

$ 8.0000 E+01 .

1 0000 E+02 4. 9(2E +01 1 038E+01 ,

n3

4. 814E +01 9.897E+00

? 1 2000 E+02

• 1. 4000 E +02 4. 68 FE +01 9.531E+00 'l-4.5d1E +01 9.25dE+00 *{

1. 6000 E+ 02 '

1.8D00E+02 4. 491E +01 9.04TE+00 -

2 0030 E+02 4. 413E +01 8.883E+00 -

2 2000E+02 4. 344E +01 8.758E+00 -

2. 4 000 E + 02 4. 283E +01 8.654E+00 .

2.6000E+02 4. 228E *01 8.5TSE+00 '

- S'

2. 8000 E +02 4.117E+01 8.518E+00 '

fc 3 0000E*02 4.131E +01 8.4 65 E+ 00 j,

4.0000E+02 3. 941E +01 8.309E+00 ,

5 0000E+02 3.791E*01 8.193E*00 . ,  :

i-l l

EF '. f.

cra" .

i G .

i mm sno .

i

. F

Paga 1 (82-0374) [8,22] #97 Question CS760.29 The Intent of Chapter 15, Section 3 Is to demonstrate the adequacy of the main heat transport systems to remove reactor and plant generated heat during protected undercooling accident conditions. Our concerns include: ,

f Once a transient event is initiated there are several factors that could affect the outcome. In Section 15.3, these possibilities are touched upon, but not with any real consistency. Our assessment of important f actors, and how they may be expected to vary, is organized in Table 15.3-2. It is Implicitly assumed throughout the PSAR that one of the plant protection systems can f all to recognize a problem. Therefore, these transients must be analyzed with the more conservative signal, i.e., the one that leads to more severe conditions. The number of pony motors that come on line is certainly an Important variable, especially since a pony motor driven pump in one loop could adversely af fect the natural circulation in the other loops. Auxillary feedwater could be supplied by two diesel driven pumps, one (50%) diesel driven pump (If one is out for service and turbine driven one falls), one turbine driven pump (which draws steam from the system), or none at all. The protected alr-cooled condensers remove heat by natural draf t air circulation.

When they are needed, louvers are opened and f ans come on, supposedly. How these various modes of operation impact on the undercooling events must be addressed.

Provide the basis and analyses to support the position that all events listed in Table 15.3-2 have been addressed on a consistent basis. Note that for several of the events, only a limited number of cases may need to be considered. In general, these are transients that are limiting only in the short term, i.e., in the longer term other related transients are likely to be more severe. Theref ore, the events listed in Table 15.3-3 may need be analyzed only for the plant protection system and number of loops initially operating cases. However, at least six events should be analyzed for all reasonable cases. This is because all could lead to difficulties in long term heat removal, and bound some of the other events. These events are listed in Table 15.3-4.

QCS760.29-1 Amend. 69 July 1982

Page 2 (82-0374) [8,22] #97 TABLE 15.3-2 UNDERC00 LING EVENT CASES Plant Protection System Available (PPS)

Primary Shutdown System Secondary Shutdown System Pony Motor Pumps Available 0, 1, 2 or 3 In Primary Loops 0,1, 2 or 3 In Intermediate Loops Auxiliary Feedwater (AFW)

Both Diesel Driven (100%)

One Diesel Driven (50%)

One Turbine Driven (100%)

Protected Air Cooled Condensers (PACCs)

Natural Draf t, Louvers Closed (0,1, 2 or 3 loops)

Natural Draft, Louvers Open (0,1, 2 or 3 loops)

Fans On, Louvers Open (0,1, 2 or 3 loops)

Number of Loops initially Operating 3-Loop Operation 2-Loop Operation 1

r QCS760.29-2 Amend. 69 July 1982 i - ___ - - - . - - . _ _ , . _ . _ _ _ . _ . . _ . . __ _ _ . _ , _ _ . _ . . _ ___ ,_ _

Page 3 (82-0374) [8,22] #97 TABLE 15.3-3 LIMITED CASE ANALYSIS EVENTS e

1. Spurious Primary Pump Trip

2. Spurious Primary and Intermediate Pump Trip

3. Spurious Intermedia 1e Pump Trip

4. Inadvertent Closure of an isolation Valve Evaporator inlet Superheater inlet Superheater Outlet

5. Turbine Trip

6. Inadvertent Actuation of Na/H O 2 Reaction System ,

7. Single Primary Pump Seizure

8. Single intermediate Pump Seizure

9. Small Water-to-Sodium Leaks in Steam Generator Tubes

10. Primary Heat Transport System Leak

11. Intermediate Heat Transport System Pipe Leak

12. Loss of One Recirculation Pump TABLE 15.3-4 FULL CASE ANALYSIS EVENTS

1. Station Blackout (LOEP)

2. Loss of Normal Feedwater

3. Failure of the Steam Bypass System

4. Steam- or Feed- Line Break Main Steam Line Rupture Steam Line from Superheater to Steam Header Saturated Steam Line from Steam Drum to Superheater Feedwater Line Break .

Recirculation Line Break

5. Dump of Evaporatory Water inventory with inlet Isolation Valve Failure (0 pen)

6. Large Water-to-Sodium Leaks in Steam Generator Tubes QCS760.29-3 Amend. 69 July 1982

P;ge 4 (82-0374) [8,22] #97 Resoonse The evaluation of the undercooling design events is discussed in Section 15.1.4.2 of the PS AR. As noted in that section, the set of events listed in Table 15.3-1 which were analyzed earlier were exanined to determine the limiting undercooling event which turned out to be the loss of of fsite power.

The question expresses a concern that all the f actors (cases) listed in Tables 15.3-2 were not considered In the evaluation of events given in Tables 15.3-3 and 15.3-4.

l Each of the events listed in Table 15.3-1 of the PSAR which involved a scram l did consider the ef fects of primary system shutdown only, as well as only the secondary system shutdown. In addtion, it was f urther assumed that the highest worth rod in each of the shutdown systems was stuck In the out posi tion. This provided a consistent basis for inclusion of a single f ailure in addition to the initiating event.

Some general observations about the thermal-hydraulic response of the CRBRP are in order:

1. Th.e peak temperatures in the core (the real basis f or the evaluation of udnercooling events) are in general seen immediately (within 20 seconds) af ter the onset of the event. The exact magnitude of these temperatures wil l be a f unction of the control rod worths, the delays in reactor scram and the reductions in primary flow prior to rod i nsertion. This is why the loss of of fsite power is the limiting event for that list given in Table 15.3-1 of the PSAR.

2. For the unique case of loss of all primary pony motors, (the natural circulation event) the peak core temperatures wil l be seen between 200 and 300 seconds af ter scram af ter which the power to flow ratio begins to decrease and the core temperatures likewise decrease.

3. One primary pony motor will furnish more core flow than the case of no primary pony motors (natural circulation). Two primary pony motors will furnish more flow than the case of only one operating primary pony motor even if the check valve f alls open in the loop with the Inoperable pony motor. Thus, the case of no primary pony motor represents the limiting case.

4. Operation of pony motors in the intermediate loops (with no operation of primary pump pony motors) will enhance the primary natural convection flows because of the f aster shif t in the prinary sodium temperatures In the IHX. Thus, this is not a limiting case.

5. Upsets in the steam generator system will not af fect the peak core temperatures because of the long transport delays in the primary and Intermediate piping. For example, the evaporator sodium outlet transient produced by a steam or feedline break in the shortest loop (loop 2) would require more than 200 seconds to be seen at the reactor vessel inlet even if the pony motor speeds in the primary and Intermediate loops were 10% of rated flow. The effect of the transient produced in the af fected loop would be f urther mitigated at QCS760.29-4 Amend. 69 July 1982

P gs 5 (82-0374) [8,22] #97 the core inlet due to the mixing of flows from the unaf fected loops in the large mixing volume In the reactor inlet. The total decay power at 200 seconds would be less than 3.3%. Consequently, events 4, 6, 9 and 12 of Table 15.3-3 of the question are not limiting in the short term.

6. Events which af fect the heat sinks f or all three heat transport loops (and associated steam generator systems), given in Tables 15.3-3 and 15.3-4 of the question ares a) Turbine trip - see Section 15.3.1.5 of the PSAR.

b) Station Blackout - Natural Circulation Analysis provided in CRB RP- ARD-0308.

c) Failure of the Steam Bypass System - see Section 15.3.2.4 of the PSAR.

d) Steam of Feedline Break - see Section 15.3.3.1 of the PSAR; Main Steam Line Rupture and Feedwater Line Break.

As noted in the appropriate sections of the PSAR, none of these events result in signficant peak core temperatures.

7. Events which af fect the heat sinks of individual loops (events 4, 6, 9,12 of Table 15.3-3 end events 4 (except Main Steam Line rupture), 5 and 6 of Table 15.3-4) will not af fect the peak core temperatures f or the reasons given in 5 above. In the long term, the plant is f ully capable of removing decay power through a single loop. Loss of a single loop due to the postulated events will not challenge the plants' decay heat removal capability; and in terms of peak core temperatures, would not represent a true undercooling event.

The f actors (cases) provided in Table 15.3-2 of the question were considered and are discussed below:

! " Plant Protection System Available (PPS)

Primary Shutdown System Secondary Shutdown System".

The limiting event in Section 15.3, Loss of of fsite power, was analyzed f or the secondary shutdown system only (PSAR pages 15.1-127 and 128) . In addition, the hot rod analysis of the. natural circulation event given in CRBRP-ARD-0308 also assumed a secondary shutdown system only. Results of analyses of the other events reported in Section 15.3 of the PSAR are given for both primary shutdown system only and secondary shutdown system only, where this aspect is important to the event being analyzed.

" Pony Motor Pumps Avail able 0,1, 2 or 3 In Primary Loops 0,1, 2 or 3 in intennediate Loops."

The limiting case is that in which it is assumed that there are no pony motors available, Event 1 of Table 15.3-4 of the question. This case has been reported in the natural circulation assessment (CRBRP-ARD-0308).

QCS760.29-5 l -

Amend. 69 i Julp 1982

Peg $6(82-0374)[8,22]#97 The combination of a primary pump seizure along wl-h f ailure of the primary pump pony motors in the other two loops would be beyond the design base and has not, theref ore, been considered.

It should be noted that the pony motors do not "come on line". They operate conti nuously. The load is picked up by the pony motors when the shaf t speed reduces to the point where the over-running clutch engages. In addition, the two pumps in the same loop (one primary pump and one intermediate pump) have their pony motors f urnished with powe- from the same buss. Thus, while it may be postulated that there may be many combinations of operable primary and intermediate pony motors, there are no common cause f ailure which could provide a mechani sm f or th i s. Nevertheless, a case has been analyzed which assumed the following: following a plant trip, the primary pump pony motors in loops 1 and 3 f all (loop 2 primary pump pony motor is available) and intermediate pump pony motor is not operating). The peak power to flow ratio seen was <0.9 (at 120 seconds into the transient). This event would be considered beyond the design basis because it induces three Independent f ail ures.

" Auxillary Feedwater (AFW)

Both Diesel Driven (100%)

One Diesel Driven (50%)

One Turbine Driven (100% ) . "

The particular combination of AFW pumps that may be assumed will have no impact on the short term undercooling of the core. While the mote Prtven AFW pumps are designated "hal f capacity", this is with respect to the worst e.ase conditions of:

o a pipe break on loop #1 with flow limited by the control valves, o steam drum venting on loop #2, and o superheater venting on loop #3.

If there are no leaks In the SG systems and normal venting takes place, either of the motor driven AFW pumps will furnish suf ficient water to maintain steam drum l evel s. Thus, multiple f ailures are required to result in loss of the SG system as a heat sink f rom a f eedwater standpoint.

" Protected Air Cooled Condensers (PACCs)

Natural Draf t, Louvers Closed (0,1, 2 or 3 loops)

Natural Draf t, Louvers Open (0,1, 2 or 3 loops)

Fans On, Louvers Open (0, 1, 2 or 3 l oops) ."

The Protected Air Cooled Condensers f or each of the loops consists of two units rated at 7.5 MWt , each with its own f an and set of louvers. The fans f or the two units on each loop are f urnished power from the same division of 1-E power.

If it is assumed that the louvers remain closed and the f ans do not operate, the heat loss is negligible (0.45 MW, p,7 goop),

QCS760.29-6 Amend. 69 l July 1982 l

k$ge'7(82-0374)[8,22]#97 If the louvers are assumed to open but the f ans do not f unction, the heat removal is approximately 305 (4.5 MW t ) of rated.

i The PACCs are Intended f or long term decay heat removal. In the event that one or more units are not f unctioning, more venting (and thus more feedwater) would be required. However, so long as the drums do not dry out, decay heat and sensible heat will be removed independent of PACC operation.

" Number of Loops initially Operating 3-Loop Operation 2-Loop Operation."

NRC review of GBRP on two-loop operation is not being requested by the aoolicant at this time, n l

l l

l QCS760.29-7 Amend. 69 July 1982

P go - 2 [82,0357] 8,22 #92 Question CS760.41 What are the bases (or the plans f or determining the bases) for setting the conditions at which the DND signal will alert the reactor operator? What will be the operator responses to the DND signal in conjunction with other plant parameters? If these responses have not been determined, what are the plans for formulating them?

Resoonse Plans for determining the bases f or setting the conditions at which the DND signal will alert the reactor operator are dependent upon results to be obtalnod from the on-going Run Beyond Cladding Breach (RBCB) program. Fuel assemblies having Indications of fuel exposure beyond a defined limit, are to be removed f rom the core. Development of this limit Is dependent on the development of appropriate technology through the RBCB Program which will assure the benign nature of operation with limited f uel sodium contact.

The applicant is committed to removing all failed f uel at each scheduled refueling outage. If a f ailed f uel assembly is characterized by the DND during reactor operation as having f uel exposure beyond the defined limit, then an orderly reactor shutdown will be initiated f or the purpose of removal of the f uel exposure to sodium at that shutdown, the other f ailed f uel in the reactor may also be replaced.

Procedures f or operator action will be f ormulated based upon the results from the on-going RBCB program and will be finalized during the OL review. A generalized approach is discussed in QCS760.39.

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l QCS760.41-1

! Amend. 69 l

July 1982

4 Page - 3 [82,0357] 8,22 #92 1

Question CS760.45 What change wilI the increased plutonium content of the f uel have on the of fsite doses due to f ailure in the EVTN?

Response

The radioactivity source term for analysis of fuel failure in the EVTM is based on the maximum design basis conditions during plant life, including the use of LWR recycle plutonium (see PSAR Section 12.1.3) which contains the largest plutonium content.

l i QCS760.45-1 Amend. 69 July 1982

Prge 8 W82-0320 [8,22] 98 Question CS760.77 Section 4.4.2.7 of the CRBR PSAR presents a discussion of pressure drop calculations and experimental results at f ull flow conditions. Is there a similar analytical and experimental base for low flow conditions? If not, how are pressure drops for low flow conditions determined? If there is a similar data base for low flow conditions, please provide detailed Information in the form of calculated and experimental pressure drop data including uncertainty factors?

Resoonse The experimental results and pressure drop correlations and uncertainties presented in Section 4.4.2.7 of the PSAR were based on data which generally cover a range of Reynolds numbers f rom-20 to 30% of the design Reynolds numbers to 100% or greater. In addition, the fuel, inner blanket and radial blanket rod bundle f riction f actors were based on preliminary data which range f rom less than 1% to greater than 100% of the design Reynolds numbers. The f uel assembly friction f actor used in the PSAR is shown in Figure QCS760.77-1.

It is significant that the f uel assembly rod bundle f riction f actor, the largest single primary system hydraulic resistance component was characterized over the f ull range of reactor flow rates. The remaining high resistance components are primarily orifice form losses which are not as Reynolds-number-sensitive as the f riction losses. Consequently, low flow rate calculations performed with the PSAR correlations are valid down to approxl-mately 20 to 30% flow and are expected to be quite close down to 1% flow.

Addltlonal data and correlations such as those shown in Figures QCS760.77-2 and -3 for the blanket and control assembly rod bundle f riction f actors are under development. Also, a more recent f uel assembly friction f actor correlation than the preliminary correlation shown in Figure QCS760.77-1 was developed based on 266 additional data points (Reference QCS760.77-1) but which dif fers f rom the Figure QCS760.77-1 correlation by only 0.2% at high flow rates and a maximum of ~2% in the transition region.

Data are becoming available on the overall fuel, blanket and control assembly pressure drops similar to those shown in Figure QCS760.77-4. Correlations to those data are under development. Flow tests to characterize orifices down to

~2% flow have been completed for the f uel and control assemblies. Simil ar testing of the radial blanket orifices located in the LIM is in progress.

Orlfice testing is under development for the Inner and radial blanket assembly orifices down to low flow rates. The remaining hydraulic resistance components in the reactor inlet and outlet regions are alI low pressure drop components which have been tested at high flow rates and will be extrapolated to Iow fIow rates wIth an appropriate Increase in uncertalnty.

QCS760.77-1 Amend. 69 July 1982

Page 9 162-0320 [8,22] 98 Rigorous data-based hydraulic correlations and uncertaintles are being developed for all major reactor pressure drop components over the full range of operating flow rates. Final results will be presented in the FSAR.

Reference I

QCS760.77-1 D. R. Spencer, R. A. MarkIey, "FrIct!on Factor Correlatton for 217-Pin Wire Wrap Spaced LMFBR Fuel Assemblies", AliS Transactions. 39, pp.1014-1015, November, 1981.

t QCS760.77-2 Amend. 69 July 1982

PERCENT OF DE84GN Flour (ZONE 1) s.5 1.0 2.5 5 10 25 as see .

8 10 8 5 l 11111 11 I l llll ll l 1111l ll l 8 2

[ f = f, [1.000 + 5.0027 (1000/Re ) + A.13b4 (1998/Re")) FOR Re > 1000 4

- = es/Re tw Re <18es 2 -

10-1

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• * = .

4 4 - mneos l O g - 11.s WlRE LEAD 3.068 WIRE DIARIETER 2 2 -

g 4.32s DUCT ACROSS FLATS E 0.230 ROD DIAAGETER (INCHES)

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,,_3 I I I IIlli l I I IIlil I I I IIlli l Ig2 2 4 8 8103 2 4 6 8104 2 4 6 810 5 2 REYNOLDS NUMBER, Re Figure QCS 760.77-1. Friction Factor Data and Correlation for 217 Pin Wire Wrap Spaced Fuel Assembly 7209-1 QCS760.77- 3 l

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! PERCENT OF DESIGN FLOW (ZONE 9) 0.1% 1% 10% 10e%

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V CNeu & TOOREAS l

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Figure QCS 760.77 2. Friction Factor Test Data for Tight Pitch to Diameter Rod Bundles With 4 Inch Wire Wrap Spacer Lead  !

7209-4 j

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Figure QCS 760.77 3. Primary Control Assembly Rod Bundle Friction Factor i

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k Figure QCS 760.77 4. Overall Fuel Assembly Loss Coefficient as a Function of Reynold Number from CRBRP Fuel Assembly

Flow and Vibration Test i

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Page 11 WB2-0320 [8,22] 98 Question CS760.79 Do the uncertainty factors presented in Tables 4.4-18A through 4.4-31 apply to low flow conditions? If not, have uncertainty f actors been determined f or low flow conditions? If the af orenentioned uncertainty f actors do not apply to low flow conditions and none have been determined f or low flow conditions, are there plans to perf orm experiments and calculations to detennine these uncertainty f actors? If the above ref erenced uncertainty factors do apply to low flow conditions, please provide data in the f orm of experimental and calculated results to support their use for low flow conditions.

Resoonse A) The uncertainty f actors presented in Tables 4.4-18A through 4.4-31 were developed using data for high flow rates. Hydraulic uncertainties at low flows are discussed in the response to Question CS760.77.

B) Maximum (highest) core hot rod temperatures are calculated f or the low flow conditions of natural circulation. These are used f or design and saf ety evaluations. Application of f ull power and flow uncertainties is conservative since:

o At the low power flows and conditions of natural circulation, the ef fect of uncertainties which locally af fect rod temperatures is insigni ficant since heat fluxes are small, i.e., local temperature dif ferences between f uel, clad and coolant are minor for a particular axial position.

o Maximum cladding / coolant temperatures f or natural circulation occur near the top of the heated core axial position and are thus the integrated ef fect of coolant over channel length. Uncertainties which af f ect Integrated coolant temperature (enthalpy rise) are for full power and f low. At l ow f l ow, these uncertainties would be smaller for maximum temperature channels because of significant increse in Inter-and intra-assembly flow and heat redistribution which results in a smaller temperature consequence due to an uncertainty in either power or flow.

QCS760.79-1 Amend. 69 July 1982

kgel2WB2-0320[8,22]98 C) for natural circulation analyses the f ull power uncertainties are increased multiplicatively for 3r decay heat uncertainties. Af ter the Initial cooldown phase and during the period when maximum temperatures are l reached the hot rod power is essentially all from decay heat, thus PHC = P x (1 + 8 P') x (1 + 5 P")

where PHC = hot rod power generation f rom decay heat; P = nominal decay power calculated f or hot rod (includes radial and axial peaking);

JP' = total f ul l power uncertainty;

&P" = 3e decay heat uncertai nty from D4DF-B4 data f il es D) Numerous other uncertaintles/ conservative assumptions in addition to the ones previously mentioned are incorporated into all transient evaluaisons.

These are summarized i n Tabl e QCS760.79-1. Conservatively all the af orementioned assumptions / uncertainties are assumed to occur simultaneously.

E) A prime example of the conservatism in this approach of maintaining f ull power / flow uncertainties f or low flow plus the other assumptions of item D is the comparison of calculated predictions using this method as compared to prototypic FFTF natural circulation experimental data shown by Figure QCS490.38-3. The second curve f rom the top rhows 3r hot channel calculations inf orporating Inter- and intra-assembly heat and flow redistribution ef fects; the lowest curve shows the measured temperature f or the highest temperature core rod. The large dif ference between the top and bottom curves demonstrates the extreme conservatism of the overall uncertainty approach.

QCS760.79-2 .

Amend. 69 July 1982 '

Paga 22 W82-0320 L8,22] 59 TABLE QCS760.79-1 MAJOR ASSU@TIONS USED IN TRANSIENT HOT ROD ANALYSIS o Conservative plant THDV Initial conditions (e.g., 750 reactor Inlet) o Worst case Doppler coef ficient including uncertaintles o Minimum control rod shutdown worth (one stuck rod) o Highest power and temperature hot rods at worst time in lIfe o ) and Worst endgap f uel/ clad of uncertainty conductancerange for both used for properties power (e.g., fuel and temperature C@lons calcula o Maximum decay heat loads including time in lif e ef fects o No credit taken for inter- and Intra-assembly flow and heat redi stri bution o Negative reactivity feedbacks neglected (e.g., core radial expansion, bowing, axial expansion of f uel and cladding) o Conservative 0.2 second delay used for PPS logic, scram breaker and f the control rod unt atch time delays QCS760.7 9-3 Amend. 69 July 1982 M _ - _ . - _ - _ _ _ _ _ _ _ - - _ _ _

Prge - 7 (82-0358) [8,22] #91 l

Question CS760.98 in Section 5.5.2.3.3 It is stated that the pipes were sized to limit mass flow rates to 20 feet per second for water through 175 fps f or superheated steam.

What are the maximum anticipated mass flow rates for steady-state operation and also during the various postulated events?

Resnonse Section 5.5.2.3.3 states that the steam and water piping is sized to limit fluid velocities at 100% power to the values Indicated in the question. The fluid velocities at 100% power are given in the following table:

Pi pe S iz e, Vel ocity, Piping Section Fluid in, fps Feedwater inlet to Steam Drum Water 10 17 Drum to Recirculation Pump Water 18 11 Pump to Tee Water 12 24 Tee to Evaporator inlet Water 10 17 Evaporator Outlet to Drum Water / Steam 16 32 Drum to Superheater inlet Saturated Steam 12 104 Superheater Outlet to isolatior. Superheated Steam 16 164 Valve Drum urain Line Water 6 5 The fluid velocity design limits are based primarily on acceptable pressure losses in the system. Thus, the 24 fps velocity in the 12-inch pump-to-tee line was judged to be acceptable, since the pressure drop calculated for this section of line is only 5 psi, :rJiuding the tee and 12 x 10 reducer. In addition, the standard outlet nozzle size for the selected recirculation pump was 12 inches.

The various postulated events (Table 5.7-1 of the PSAR) begin f rom initial conditions of 115% power. At this power level, the fluid velocity in the feedwater inlet, evaporator outlet, superheater inlet, and superheater outlet lines, will increase by 15% from the above values. The fluid velocities in the other lines will remain essentially constant.

The postulated events where line rupture is involved would have maximum anticipated fluid velocities dependent upon the nat e of the rupture, but plant safety is maintained by isolation of the ruptured line section while other system components f unction to a saf e shutdown situation.

PSAR Section 5.5.2.3.3 has been updated to reflect average water velocities net to exceed 25 fps.

1 QCS760.98-1 Amend. 69 July 1982

p:ge - 1 (8,22) #96

4. Piping shall be designed with suitable access to permit in-service testing and Inspection.

5. All " horizontal" piping shall be sloped. Steam traps and drain valves shall be located at the low points to permit complete draining of thm piping.

6. Piping sizes shall be chosen such that average fluid velocities at +he 100% plant power condition will not exceed the following values:

a. water 25 fps

b. water-steen mixture 50 fps

c. saturated steam 125 fps

d. superheated steam 175 fps System Descriotion All Steam Generation System piping is shown in Figure 5.1-4. The design characteristics and ASME Code classifications are presented in Table 5.5-7.

The only field run piping planned for the steam generator system is non-saf ety class piping. The internal diameter of the piping wilI be 2 inches or less and is used for drain lines f rom steam traps. The design pressure would not exceed 100 psia and the design temperature would be less than 3000F.

The Seismic Category I design requirements are placed on the Steam Generation System's steam-water piping. Superheater and evaporator modules and the steam drum are provided with quick acting isolation valves. Design pressures of all piping are nominally 110% of the operating pressure at rated power.

The use and location of rigid-type supports, variable or constant spring-type supports, and anchors or guides will be determined by flexibility and stress analysis. Piping support elements wilI be as recommended by the manuf acturers and will meet applicable code requirements. Direct weldment to thin wall piping will be avoided where possible.

Attachment and penetrations shalI be designed and f abricated according to the ASME Code requirements.

5.5-8 Amend. 69 July 1982

P;ge - 2 (8,22) #95 Question CS760.134 The design basis assumes five (5) natural circulation occurrences per lifetime but the fuel damage analysis assumes only one.

Does this Imply that a new fuel loading will be required af ter any natural circulation event?

Resoonse The CRBRP plant duty cycle is given in Appendix B, Table B-1, of the PSAR.

For an emergency event like the natural circulation event, five (5) such events are considered per plant lifetime (one every 6 years) plus two consecutive occurrences of the most severe type (giving potentially seven total emergency events). For replaceable components such as the fuel assembly, one event of this type during the assembly lifetime is conservatively assumed f or fuel design damage analysis.* However, af ter such an event, an assessment of core damage (including analytical methods) of course must be made bef ore resumption of normal operation. If it were determined that there was little f uel rod damage and that acceptable f uel rod lif etime renained such that another event of this type could be taken, a new fuel loading would not be required.

• Assuming a linear distribution of emergency events over the 30 year plant lif e would result in a 7/15 probability of occurrence over the 2 year fuel l i f etime. This f requency fraction has been rounded-of f to a single occurrence.

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l QCS760.134-1 Amend. 69 July 1982

Page - 5 (8,22) #95 Question CS760.137 Please provide a list of the Instrumentation and their functional requirements which are necessary to monitor and control the heat transport systems.

Emphasize those Instruments that the operator will use for decay heat removal under both normal and emergency conditions.

Resoonse A. Tables QCS760.137-1 through 3 provide the Instrument list for the Heat Transport System (HTS) which may be used to monitor and control the sy stems.

B. The instruments the operator may use for decay heat removal under both normal and emergency conditions are identified below:

o Primary Heat Transport System (PHTS) Temperature Hot Leg o intermediate Heat Exchanger (IHX) Outlet Temperature (PHTS Cold Leg) o PHTS Pony Motor Speed o PHTS Flow o Intermediate Heat Transport System (lHTS) Temperature Hot Leg o Evaporator Outlet Temperature (IHTS Cold Leg) o IHTS Pony Motor Speed o IHTS Flow All the Instruments identified as being used by the operator for monitoring decay heat removal are provided on the Main Control Panel (MCP).

QCS760.137-1 Amend. 69 July 1982

Page - 6 (8,22) #95 l

TABLE QCS760,137-1 PHTS INSTRUENT CHANNELS FUNCTION SENSOR TYPE EASUREMENT NUE ER RANGE USE OF SENSORS Pump Outlet Nak Capillary Surveil lance 2/ Loop 0-20 0 Pressure Performance psi Evaluation PHTS/lHTS4P Reactor inlet Nak Capillary Survei l lance 2/ Loop 0-200 Press Performance psi Evaluation Pump Outlet RTD Survei l lance 2 dual / 300-Temp Performance Loop 12000F Eval uation Calorimetric Reactor inlet RTD Survei l lance 2 dual / 300-Temp Perf ormance Loop 12000F Evaluation Calorimetric Loop to Loop AT Main Loop PM Survei l lance 1/ Loop 0-Flow Fl owmeter Perf ormance 40000 gpm Eval uation llH X Outlet CA-T/C Control, PPS 3/ Loop 300-Tenp 10000F QCS760.137-2 Amend. 69 July 1982

Page - 7 (8,22) #95 TABLE QCS760.137-2 IHTS INSTRUENT CHANNELS FUNCTION SENSOR TYPE E ASUREMENT NUE ER RANGE USE OF SENSORS lHX Inlet RTD Survei l lance 1 Dual / 300-Temp Calorimetric Loop 8000F IHX Outlet RTD Survei l lance 1 Dual / 300-Temp Calorimetric Loop 10000F IHX Outlet Nak Capillary Survei l lance 1/ Loop 0-27 5 Press psig Pump Inlet Nak Capillary Survei l lance 1/ Loop 0-150 Press psig Pump Outlet Nak Capillary Survei l lance 1/ Loop 0-350 Press psig Main Loop PM Flowmeter PPS, Control 1/ Loop (1)4000-40000 gpm Flow Evap. Outlet CA-T/C PPS 3/ Loop 300-8000F Temp.

(1) PPS operating range. Functional range and Indication are from zero.

QCS760.137-3 Amend. 69 July 1982

Paga - 8 (8,22) #95 TABLE QCS760.137-3 PHTS MAIN SODIUM PUMP INSTRUENTATION FUNCTION SENSOR TYPE EASUREMENT NUE ER RANGE USE OF SENSORS Pump Shaf t Magnetic Perf ormance 2/ pump 0-1200 rpm Speed Survei l lance Pony Motor Speed Survei l lance 1/ pump 0-120 rpm Speed Indicator IHTS MAIN SODIUM PUMP INSTRUENTATION Pump Speed Magnetic PPS, Control 5/ Pump (1)120-120 rpm Pony Motor Speed Survei l lance 1/ Pump 0-120 rpm Speed Indicator (1) PPS operating range. Functional range and indication are f rom zero.

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l QCS760.137-4 Amend. 69 July 1982

P;ge - 9 (8,22) #95 Questfon CS760.138 Please include the core thermal response for item h (Section 5.7.3 - loss of one primary pony motor with check valve f ailure).

Resnonse This particular transient was analyzed using minimum decay heat along with various other conditions which were included to make the transient conservative f rom a plant piping and component design thermal transient standpoint. As such, the core thermal response from the analysis which produced the pump Inlet transient for item h shows very modest temperatures.

The core flows are, however, suf ficiently high to keep the power to flow ratio wel l below 1.0. For example, at the time the flow reverses (100 seconds into the transient) In the loop with the f ailed pony motor (and a stuck open check val ve), the core f low is 8.6%. At 600 seconds into the event, the reverse flow in the loop with the f ailed pony motor has steadled out at 200 lb/sec (5.2% of Initial steady state loop flow) and the core flow is 6.9% of its initial flow. Thus the core temperatures will be less than their Initial steady state temperatures.

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QCS760.138-1 Amend. 69 l July 1982

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