3F0796-19, Provides Boron Precipitation Info Requested by NRC in Ltr

Provides Boron Precipitation Info Requested by NRC in Ltr

ML20116B635
Person / Time
Site:	Crystal River
Issue date:	07/26/1996
From:	Boldt G FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
3F0796-19, 3F796-19, NUDOCS 9607300153
Download: ML20116B635 (12)

Text

4 Florida Power CORPORATION oo.wUN i

July 26, 1996 I

3F0796-19 1 U.S. Nuclear Regulatory Commission ATTN

Document Control Desk Washington D.C. 20555-0001

Subject:

Response to Boron Precipitation Issue

Reference:

NRC to FPC Letter dated June 26, 1996 [3N0696-17]

Dear Sir:

The purpose of this letter is to provide the boron precipitation information requested by the Nuclear Regulatory Commission (NRC) in the referenced letter dated June 26, 1996. Florida Power Corporation (FPC) is a member of the B&W Owners Group (B&WOG) which is currently addressing the baron precipitation control issue for small break loss-of-coolant accidents (SBLOCA) generically.

On July 9,1996, a meeting was held between the B&WOG Regulatory Response Group (RRG) and Nuclear Reactor Regulation (NRR) to discuss SBLOCA boron precipitation control issue (s), safety significance and overall resolution. The NRC Staff concurred with the B&WOG determination that this issue poses no immediate safety  ;

concern. FPC is working with the B&WOG to bring about closure on issues related I to post LOCA boron precipitation control.

Sincerely, l G. L. oldt Vice President Nuclear Production Attachment

cc

Regional Administrator, Region 11 NRR Project Manager ND Senior Resident Inspector i\

\.

9607300153 DR 960726 ADOCK 05000302 PDR \

g~

CRYSTAL RIVER ENERGY COMPLEX: 15760 W Power Line St . Cryctal River, Florida 34428-6708 . (352) 795-6486 '

A Florids Progress Company l

I

4 U.S. Nuclear Regulatory Commission 3F0796-19 Page'2 of 12 STATE OF FLORIDA COUNTY OF CITRUS G. L. Boldt states that he is the Vice President, Nuclear Production for Florida Power Corporation; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information, and belief.

G. L. Bol t Vice President Nuclear Production G. L. Boldt, is personally known to me. Subscribed and sworn to before me, a Notary Public in and for the State and County above named, this ay of July, 1996.

Y)lW$ bNI$

Notary Public (print) No ry Public (signature)

(:--

MY COMMIS8 ION f CC 514300 I ONES December 18,1900 l

. Bened Thru Neeuy Puhen Unemtus '

l i

U.S. Nuclear Regulatory Commission 3F0796-19 l Page'3 of 12 l l

Request:  ;

1. For both large and small-break LOCAs, provide or reference analyses to justify that the concentration of boron in the core will not exceed the solubility limit at any time during a LOCA scenario. ,

I

Response

The B&WOG Report 51-1206351-00 is the analysis that provides the basis and justification that boron in the core will not exceed solubility limits during LOCAs. This report takes credit for internal flow through the gaps that exist between the reactor vessel outlet nozzles and the core support assembly interface. The report was primarily developed to address Large Break LOCAs; however, a qualitative evaluation was performed of the entire break spectrum to obtain the bounding assumptions used for the internal reactor vessel recirculation flows. Prior to this report, the overflow of liquid through the Reactor Vessel Vent Valves (RVVV) was a credited mechanism that aided in the prevention of boron precipitation. Work done in 1991 indicated that the i collapsed saturated liquid level would be below the RVVVs, and therefore the l overflow previously credited would not occur. l The B&WOG report, referring to the RV internal gap flow, stated that "This mechanism does not require operator action to preclude baron precipitation in the core." Nonetheless, active systems have been the historically preferred mechanism to assure adequate post-LOCA boron precipitation control.

Request:

2. In the analysis, if credit is taken for a gap in the reactor internals, please address the following considerations:

a. Was the as-manufactured gap ever measured and compared to the nominal specification? What are the nominal and actual tolerances?

b. What is the gap size verses temperature?

c. Has the potential for clogging of the gap from corrosion, boron precipitation, plateout or other means been considered in the analysis?

If not, why not? If so, quantify the flow area and impact of the clogging on the flow.

Response

2a. Yes, the gaps were measured during fabrication and the fabrication records document the as-built gap measurement of 0.092457 inches at 70*F, with a nominal cold gap dimension of 0.077 +/- 0.020 inches.

2b. The gaps are calculated from as-built measurements considering the effects of temperature and pressure on expansion. Gap size as a function of RCS temperature and the corresponding flowrate are presented in the following table:

U.S. Nuclear Regulatory Commission 3F0796-19 Page"4 of 12 RCS Gap Size Total Flow Temperature (in) (lbm/sec) 550*F 0.025296 4.855609 1 500*F 0.028112 5.765656 450*F 0.032655 7.234234 400*F 0.039044 9.298879 350*F 0.046031 11.55707 300*F 0.053486 13.96681 250*F 0.061608 16.59177 212*F 0.067965 18.64640 200*F 0.070101 19.33675 70 'F 0.092457 25.56211 2c. Variations in the calculated gap flow, including reductions in flow of 50 and 75 percent (due to clogging or some other unknown phenomena), showed only marginal increases in the maximum calculated core boron concentrations, which remains significantly below the solubility limit using the smallest gap sizes for the most limiting break. Based on preliminary calculations by FTI, a constant gap flow of between 0.6 and 1 lbm/sec is sufficient to hold the core concentration at a constant value, which is a function of core decay heat. The higher flow value is applicable under high decay heat loads with the lower value applicable after 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the decay heat contribution is lower.

l Request:

l 3. Identify and justify other important assumptions, behaviors, and actions in the analyses, including boron solubility verses time and temperature, reactor coolant pressure and temperature verses time, method of depressurization, procedures to implement the active method and its timeliness and effect of DH drop line throttling on depressurization, cooldown, boron disposition, and other event behaviors.

Response

Preliminary SBLOCA analyses and evaluations of boron concentration were presented at the July 9, 1996 meeting between NRR Staff and the B&WOG RRG. The analysis used many conservative assumptions including:

. Small core mixing volume.

. Bounding high RCS, CFT, and BWST initial boron concentrations.

. Saturated core boiling (ie. no inlet subcooling.)

. Minimum RCS inflow from one HPI pump.

. One LPI pump available.

U.S. Nuclear Regulatory Commission l 3F0796-19  !

Page 5 of 12 l

Figure 1 provides RCS pressure verses time for the most limiting break location, at the bottom of the CLPD, for various size breaks.

Figure 2 illustrates boron solubility as a function of temperature, at I atmospheric pressure, which includes a reduction factor in actual boric acid solubility versus temperature to conservatively define boron control measures.

Figure 3 provides the ratio of the boron concentration in steam to the boron concentration in water. At higher pressures and temperatures, solubility increases as does the volatility, which aids in transferri.ng boron out of the core region. RVVV steam flow can carry significant quantities of boron as a function of temperature. 3 1

Figure 4 illustrates the effect of passive flows obtained through the hot leg I nozzle gaps which halt the boron concentration increase and decreases the core concentration long term. The results of the evaluation show a very limited size of breaks, 0.015 to 0.05 ft,, located on the bottom of the CLPD piping that i represent the areas of greatest concern regarding elevated core boron l concentrations. At the time of maximum core boron concentration, RCS pressure  !

and temperature are high, such that the solubility limit is not approached. j Operator assisted RCS cooldown using steam generators will shorten the time period in which core concentrations are increasing and reduce the maximum concentration reached due to the development of the gaps and resulting flow. Gap I flow will maintain boron concentration below the solubility limit until an active i means for prevention of boron precipitation can be established. ~

Emergency Operating Procedure (E0P)-03, Inadequate Subcooling Margin, provides l guidance for establishing emergency core cooling. Three key initial actions oper'ators take are: 1) trip reactor coolant pumps, 2) initiate full high pressure l injection (HPI) fl ow, and 3) raise steam generator level (using emergency I feedwater) to the inadequate subcooling margin setpoint. An additional step l maintains the steam generators as a heat sink. E0P-08, LOCA Cooldown, provides ,

guidance to maintair steam generator induced cooling. If the rate of steam  ;

generator cooling decreases specific actions are directed to establish heat '

transfer which includes lowering steam generator pressure, bumping RCPs and opening the pilot operated relief valve (PORV). Once low pressure injection (LPI) is supplying adequate flow, HPI flow and 0TSG cooling are terminated.

Operating Procedure (0P)-404, Decay Heat Removal System, provides guidance for establishing the active method for boron precipitation control. This is accomplished by throttling open (to the full open position) the first valve (DHV-

3) in the dropline to reactor building sump flowpath. Certain prerequisites must be met prior to establishing the "dropline-to-sump" alignment which include:

. adequate subcooling margin does not exist,

. LPI pumps are aligned to draw suction from the RB sump,

. incore temperature < 280*F,

. RCS pressure = RB pressure.

For LBLOCAs, the above conditions will develop in a short period of time.

Certain SBLOCAs will require a longer period of time to depressurize the RCS.

It was assumed that LBLOCA analyses bounded SBLOCAs regarding the establishment of an active method without reliance on gap flow. Gap flow, at CR-3, begins when RCS temperature reaches 550*F. Average RCS temperature during normal operation is 579'F. Therefore, essentially as soon as the reactor trips due to low RCS

U.S. Nuclear Regulatory Commission 3F0796-19 Page'6 of 12 pressure, gap flow begins which, in turn, maintains the boron concentration substantially below the solubility limit throughout the entire transient. Once the conditions above are met, the dropline-to-sump configuration is established providing the active method for boron precipitation control.

1 The action of establishing the dropline to RB sump alignment should not result in a significant RCS pressure or temperature change and is not expected to create any adverse behavior. This is due to approximate equilibrium fluid conditions between the RB sump and the RCS. Throttling DHV-3 to the full open position is l a conservative measure to reduce the potential for negative hydraulic impact on the screens due to instrument error or other unknown phenomena.

1 Request: 1

! 4. If the sump screen were redesigned or replaced to withstand reverse flow 1 pressure differences, would it still be necessary to take credit for the l hotleg nozzle gap for preventing boron precipitation?

Response. '

l Assuming the RB sump screens were modified, the dropline-to-sump alignment would be performed when RCS temperature is < 300*F, which is a piping analysis based limit. In a saturated system 300*F translates to a saturation pressure of 67 l

psia. The time to reach the boron solubility limit compared to the cooldown time to reach 300*F may require some limited reliance on RV internal gap flow.

Further analysis and' evaluations, being performed by FTI, will determine if that is the case. Notwithstanding, the dropline-to-sump method is not single failure proof as discussed in the response below.

Request:

i

5. Justify that your present method for preventing boron precipitation can perform its intended function with a passive or active failure upon initiation of the sump recirculation mode of recirculation.

Response

Figure 5 illustrates the dump-to-sump flowpath which is established by. opening 5 valves that connect the RCS to the RB sump. The LPI pump in the flowpath is secured prior to configuring the dump-to-sump alignment to prevent vapor (steam) binding the LPI pump. Only active failures are considered in the CR-3 mechanical system design basis. From that perspective a failure of any one of the 5 valves in the decay heat dropline flow path to open would prevent the establishment of l the active method to control post LOCA boron precipitation. Therefore, the RV internal gap flow would be relied upon to prevent boron precipitation until a i recovery plan and subsequent opening of the affected valve.

? Request: '

L

6. Describe how the 10 CFR 50.59 evaluation for FSAR changes reconciled the

, boron dilution methodology described in the FSAR with staff letter, (Ashok i Thadani to P. S. Walsh dated March 9,1993), regarding use of hotleg gap as a backup but not a primary method for recirculation flow path.

l 4

I. l l

U .S. Nuclear Regulatory Commission 3F0796-19 Page~7 of 12

Response

The 10 CFR 50.59 evaluation focussed on the B&WOG boron dilution report, taking credit for RV internal gap flow as the only necessary mechanism to prevent boron precipitation, as a basis for the FSAR change. The conclusions reached in the 50.59 evaluation did not appropriately consider the subject letter. This is recognized in the evaluation. However, the preference for establishing an active means of boron precipitation control has always been in effect. Although the (

FSAR identified the RV internal gap flow as preferred, procedural guidance for establishing dropline flow to the RB sump remained intact with enhancements to avoid damaging the RB sump screens. FPC maintains a Nuclear Operations l Commitment System (N0CS), which addresses the need for procedural guidance to '

establish the active method for post LOCA boron precipitation control. This i commitment and the related procedural guidance continued to require and maintain l an active method, t

l l

l

)

i l

l l

I l

SBLOCA BORON PRECIPITATION STUDY WITHOUT SG COOLDOWN ,me m

m PRIMARY PRESSURE FOR THE CLPD BREAK SPECTRUM S.

40 =

cm e N IJGEND

• EE

.o,, e n W

^

0.50 ft2 CDD C $

" O 0.10 ft2 CLPD

0.06 ft2 CLPD g' x 0.05 ft2 CLPD a 2000- -------------j

+-- '- -----+ ---------t------ -

W 0.04 ft2 C&D " b

, l

_ 0.03 ft2 CGD ,

. p

c. 0.02 ft2 CLPD @

C 0.015 ft2 CUD x l

0.01 ft2 CUD p

! i I i 1600- [ + 4 ~~4 + f- T p W ,

m i r i I v2 '

De  !

a j i i .

g' -* - + - - ~ - - - - - - +

1200-  ;-

--t-- --

I

~ '

i r t .

5.~... ..e-.~ s

. .. . . .. p ~ 6 .

i i

L3 l 9 i

! , >4

g. - _ . . _ ,

g

.b

~

i f O,~.- _ __

h

^ ^l:

s _ _ T - 4 .- -

. _ _  : I'_ m

+

g _ _ . . _ _

w  : . i 0

0 4 8 12 16 20 24 28 32 Time (hours) i

4

' ~uc -

a 2-

• 2?

'*Tz GE Boron Solubility Versus Temperature.  ;

E-50000  %

a 8 aooo , x s' 9 40000 - D g a n x

a. 35000 -

I 1

=

i k # 0 a

o 8 30000 5

m 5 < - -* - solubility Umit c 25000 e o e

c o y _

," + Imposed Solubility Umit u

5

- 15000 - #

g ,

W 10000 -

5000- -

O  : _

0 50 100 150 200 250 =rt Phsid Temperature (F) g C

llO m

1%3

? =. zhm" A9rreoE OH g

• MS.T MMg$M W

?% 5 N 0

5 6

. 0

' i 0 6

0

' i 5 5

t n

i e

_. c 0 _

i f ' e 0 _

f 5

_ e o

C _

n i

o -

t u 0 b ' a 5 F 4 , _

i t

r e r

i s t u

D a r

e _

d i

p _

c m A

c

' i 0 e 0

4 T

i

_ r o

B 0

' a 5 3

0

_ ' e 0 3

0

' i5 2

/

/

W

- - - - 0 0

2 2 1 8 6 4 2 O 1

O 0 0 0 0 0 0 O 0 0 5 O J$ oe1oo

mwc tb *2 m@ oJ C/3

, @ =

W@

W 5 2".

W c O CO m W e

W lu 0.03 FT2 SBLOCA BORON CONCENTRATIONS VS TIME WITH CR-3 GAPS.

m e

M c

4000 == - - - .__. _ . _ _ _ _ . _ . _ _ . _ _ . - . . _ _ . _ . . ._ . _ _ _ _ _ . . .._.._..._ g

% O

,/ s n 3500 - =

s 2 ~ , --'-- - --- . . - . . - . ._ . . . . _ - - _ _ _ _ . - . . . .__ - . . . . _ . _ . . _ . _ _ _ _ _ _ ._. .___

r a

/ b'

. . _ ~ ~ ~ . -~~ O r ~ - - - - - - - - .- - - - - - - ..- - -- - -- -- - - - - - .. - -- -

3000 - r f -- - - _ - _ _ _ - _ . . _ _ _ _ _ _ . _ _ . _ . _ _ _ _ __ .___ . . _ _ _ _ . . _ . _ _ _ . _._

k

_ f ,,,........ ,

s E .

s '. g m W

2- 2500 = t '. -r -- .-

g o

c o \j r't.......

i2:2000 = = -_ . - - _ .___ ___ . _ . _ _ _ _ _ _ .

E .__ . _ _ _

y1500= =- 3 g

o 1000 = * * .-

CORE /10 500 = DOWNCOMER r

O m m m

m m

" m m m m ,

m O 5 10 15 20 25 30 35 40 45 50 Time (hours) c

[

c M

m b

_ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ __ _m. _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ __- _ _ _ _ . -_ _ _ _ _ . - _ _ _ _ _ . _ _ _ _ _ _ _ _ _ __

U. S. Nuclear Regulatory Commission Figure 5 s '3F0796-19

- Page 12 of 12 l

j (fi} "B" Hotleg i

I DHV-3 i

s BWST BWST i

l -

DHV-4

! \/ \/

i

- 1 DHV-41 -

DHV-34 - - -

DHV-35 l

/\ /\

LPI LPI DHV-39 DHV-40 DH F'-1 A ' I L DHf'-1B DHV-42 - -

DHV-43 '

l RB SUMP -

i .