AECM-85-0341, Forwards Addl Info Re ECCS Performance Analysis Which Supports Increase in Peak Cladding Temp of 51 F Compared to Current FSAR Analysis,Per Request.Fsar Changes Will Be Made If Design Change Implemented

Forwards Addl Info Re ECCS Performance Analysis Which Supports Increase in Peak Cladding Temp of 51 F Compared to Current FSAR Analysis,Per Request.Fsar Changes Will Be Made If Design Change Implemented

ML20133M625
Person / Time
Site:	Grand Gulf
Issue date:	10/22/1985
From:	Dale L MISSISSIPPI POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
AECM-85-0341, AECM-85-341, TAC-59440, NUDOCS 8510280158
Download: ML20133M625 (13)

Text

1 1

I MISSISSIPPI POWER & LIGHT COMPANY  !

=

Helping Build Mississippi P. O. B O X 16 4 0 J A C K S O N. MISSISSIPPl 39215-1640 October 22 1985 NllCLE AR UCEN5hG & SAFETV Cf PARTMENT U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D. C. 20555 Attention: Mr. Harold R. Denton, Director

SUBJECT:

Grand Gulf Nuclear Station Unit 1 Docket No. 50-416 License No. NPF-29 File: 0260/0840/L-860.0 Response to Request for Additional Information on PCOL-85/ll AECM-85/0341 NRC staff reviewers have recently requested additional information on Mississippi Power and Light's (MP&L's) request for amendment of the CGNS Unit 1 Operating License NPF-29 submitted on August 12, 1985 by letter from Mr. O. D. Kingsley, Jr. to Mr. Harold R. Denton, MP&L correspondence number AECM-85/0228. NRC staff reviewers requested more information on the ECCS performance analysis performed in support of MP&L's proposed change to the technical specifications resulting from a design change to add high/ low pressure interlocks to the injection valves on the low pressure ECCS systems.

Attached to this letter is a description of the ECCS performance analysis which supports an increase in peak cladding temperature of Sl'F compared to the current FSAR analysis. Provided the design change is implemented, the necessary FSAR changes will be incorporated into the appropriate annual update.

Yours truly,

,,f M /h M:

L. F. Dale

['

/

Director WJH/SHH:vog Attachment cet (See Next Page) 0510200150 051022 FOR ADOCK 05000416 s P PDR

,\

J13AECM85101601 - 1 Member Middle South Utilitius System

• Ie s -- J

. AECM-85/0341 Page 2 cc: Mr. J. B. Richard (w/a)

Mr. O. D. Kingsley, Jr. (w/a)

Mr. R. B. McGehee (w/a)

Mr. N. S. Reynolds (w/a)

Mr. H. L. Thomas (w/o)

Mr. R. C. Butcher (w/a)

Mr. James M. Taylor, Director (w/a)

Office of Inspection 6 Enforcement U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Dr. J. Nelson Grace, Regional Administrator (w/a)

U. S. Nuclear Regulatory Commission Region II 101 Marietta St., N. W., Suite 2900 Atlanta, Georgia 30323 9

8 I

J13AECM85101601 - 2

_ Ab

. Attachment to AECM-85/0341 Page 1 LOW PRESSURE INTERLOCK ECCS PERFORMANCE ANALYSIS An ECCS performance analysis was performed for Grand Gulf to determine the effect of adding a pressure permissive to the LPCI and LPCS injection valve opening logic. This additional logic delays the low pressure ECCS injection for large breaks which rapidly depressurize the vessel. The results of this study show that for a valve opening time of 30 seconds, the delay in ECCS injection causes a delay of approximately 15 seconds in core reflooding and an increase in peak cladding temperature of 51*F compared to the current FSAR analysis. However, there is sufficient margin in the Grand Gulf FSAR analysis to absorb this increase and remain within the 2200*F PCT limit.

The design basis accident (DBA) recirculation suction line break with failure of the LPCI diesel generator was analyzed to determine the effect of the logic change. This break and failure combination was identified in the FSAR as the limiting LOCA event and is also the most sensitive to this change. Smaller breaks depressurize more slowly and are less affected by the pressure permissive. The analysis was identical to that in the FSAR with the exception of the low pressure ECCS injection. For the previous case presented in the FSAR, the LPCI and LPCS were assumed to inject at 45.9 seconds

• after the break occurs (i.e., 40 seconds after the water level fell below Level 1 at 5.9 seconds *). For the case with the pressure interlock, the injection valves were assumed to begin opening only after the reactor pressure dropped below 450 psia at 27.8 seconds. Low pressure ECCS injection was then assumed only after the valves were fully open at 57.8 seconds (i.e., assuming a 30 second valve opening time).

The valve logic change results in a delay in both injection and core reflooding, and an increase in the calculated peak cladding temperature.

Since the PCT for the FSAR case in 2098'F, there is sufficient margin to the 2200*F limit to absorb the increase for the case with the pressure interlock, and therefore it is acceptable from an ECCS viewpoint.

To document the effect on the LOCA analysis of adding the low pressure ECCS pressure permissive, attached are some recommended changes for FSAR Tables 6.3-1, 6.3-2, 6.3-3 and 6.3-6. Revised copies of FSAR Figures 6.3-14, 6.3-15 and 6.3-16 are also attached which show the water level, pressure and peak cladding temperature response with the addition of the injection valve pressure permissive. These recommended changes to the FSAR will be i incorporated, if applicable, at the appropriate annual update.  !

1

• Table 6.3-1 in FSAR amendment 57 gives actual expected response as opposed to response assumed in the analysis. Level 1 indication at 7 seconds was a typographical error. Revised Table 6.3-1 attached to this letter depicts response assumed in current analysis.

J13ATTC85101701 - 1

/

Attachment to AECM-85/0341 Page 2 GRAND GULF FSAR RECOMMENDED CHANCES TO INCORPORATE A PRESSURE PERMISSIVE ON THE LOW PRESSURE ECCS INJECTION VALVES

1. Table 6.3-1 would be updated and modified as shown on the next page.

2. In Table 6.3-2, insert the following information immediately following the " maximum allowable time delay" entry, and modify the " injection valve fully opened" entry as below for both LPCI and LPCS systems.

o Injection valve fully open - seconds af ter pressure permissive si 30.0.

o Pressure at which injection valve may open - psia $1 450,

3. In Table 6.3-3, add footnote "(3)" after "3.1 ft2 (DBA)" and add to the bottom of the page "(3)". The impact of the LPCI and LPCS injection valve pressure interlock is estimated to be a 51*F increase in PCT and a 0.004 increase in oxidation fraction. Consistent with the estimated increase in PCT, the CWMWR% would be 0.16."

4. In Table 6.3-6, under "high enrichment IC fuel" exposure 200.0, add "*"

after (PCT) 2098 and (0XID FRAC) 0.0201, and the footnote to the bottem of the page "*" the impact of the LPCI and LPCS injection valve pressure interlock is estimated to be a Sl*F increase in PCT and a 0.004 increase in oxidation fraction."

1 J13ATTC85101701 - 2

. Attachment to AECM-85/0341 Page 3 GGNS FSAR TABLE 6.3-1 OPERATIONAL SEQUENCE OF EMERGENCY CORE COOLING SYSTEMS FOR DESIGN BASIS ACCIDENT Time (sec) Events 0 Design basis loss-of-coolant accident assumed to start; normal auxiliary power assumed to be lost.

ese 0 High drywell pressure and reactor low water level (level 3) are reached. All diesel generators HPCS, LPCS, LPCI signaled to start on high drywell pressure *. Scram initiated on level J.

1

<%/ 3 Reactor low-low water level (level 2) reached. HPCS receives second signal to start.

As 6 Reactor low-low-low water level (level 1) reached. Second signal to start LPCI and LPCS; main steam isolation valves close. Auto-depressurization sequence begins.

<se 13 All diesel generators ready to load; erergize HPCS pump motor; open HPCS injection valve: begin energizing LPCI and LPCS pump motors.

em/28 Pressure permissive for LPCI and LPCS injection valve reached.

N/30 HPCS injection valve open and pump at design flow, which completes HPCS startup.

ev 33 LPCI and LPCS pumps at rated speed, ev 58 LPCI and LPCS pumps at rated flow, LPCI and LPCS injection valves open, which completes the LPCI and LPCS startups.

See Figure Core effectively reflooded assuming worst single failure; heatup 6.3-14 terminated.

> 10 min Operator shifts to containment cooling.

NOTE: For the purpose of all but the next to last entry on this table all ECCS equipment is assumed to function as designed.

Performance analysis calculations consider the effects of single equipment failures (see subsections 6.3.2.5 and 6.3.3.3).

• No credit taken in LOCA analyses for ECC System start on high drywell pressure signal.

J13ATTC85101701 - 3

.]

l Attachment to AECM-85/0341 Page 4 GGNS FSAR TABLE 6.3-2 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS Plant Parameters o Core Thernal Power MWt 3993 o Vessel Steam Output LB /hr 17.3 x 10 6 o Corresponding percent of rated percent 105 steam flow o Vessel Steam Dome Pressure psia 1060 o Maximum Recirculation Line ft 3.1 Break Area Emergency Core Cooling System Parameters Low-Pressure Coolant Injection System o Vessel Pressure at which flow psid (vessel to 225 may commence drywell) o Minimum Rated Flow at Vessel GPM 22000 Pressure paid (vessel to 20 drywell o Initiating signals low-low-low water level ft above top je 1.0 or of action fuel high drywell pressure psig 2; 2.0 o Maximum allowable time delay sec 27.0 from initiating signal to pumps at rated speed o Injection valve fully open sec after pressure 30.00 pe rmissive o Pressure at which injection psia JG450 valve may open J13ATTC85101701 - 4 J

Attachmeat to AECM-85/0341 Page 5 GCNS FSAR TABLE 6.3-2 (Cont.)

Low-Pressure Core Spray System o Vessel pressure at which flow psid (vessel to 289 may commence drywell) o Minimum rated flow at Vessel GPM 7000 Pressure psid (vessel to 122 drywell) o Initiating signals los-low-low water level ft. above top ;h 1.0 or of action fuel high drywell pressure psig 3E 2.0 o Maximum allowed (runout) flow GPM 9100 o Maximum allowed delay time sec 27.0 from initiating signal to pump at rated speed o Injection valve fully open sec after pressure 30.00 permissive o Pressure of which Injection paia JE 450 valve may open High-Pressure Core Spray o Vessel pressure at which flow psid 1177 may commence o Minimum flow available at See vessel to pump suction head Figure 6.3-3 o Initiating signals low-low water level ft. above top of jt 10.5 or active fuel high drywell pressure psig ;E 2.0 o Maximum allowed (runout) flow GPM 9100 o Maximum allowed delay time sec 27.0 from initiating signal to rated flow available and injection valve wide open J13ATTC85101701 - 5 L a

_. ;_..=--.-~__-.--- . . _ - _ - . . . . - . .

Attachment to i

AECM-85/0341 l Page 6 4

i GCNS FSAR TABLE 6.3-2 (Cont.)

Automatic Depressurization System o Total number of valves installed 8 o Number of valves used in analysis 8( }

analysis 6

o Minimum Flow Capacity of Ib/hr 6.4 x 10 8 valves at vessel psid (vessel to 1125 pressure suppression pool)

( } Additional LOCA analyses in Section 6.3.3.7.8 with seven ADS valves justify "7-one ADS valve out of service for an extended period of time.

I i

1 i

i i

I i

J13ATTC85101701 - 6 i _ . . .__. _ .-_ ___.,.

Attachment to AECM-85/0341 Page 7 GGNS FSAR TABLE 6.3-3

SUMMARY

OF RESULTS OF LOCA ANALYSIS Peak Local Oxidation

% of Initial Cladding Break Spectrum Analysis PCT (F) Thickness Break Size Location Single Failure 3.1 ft2 (DBA) (3) 2098 (1) 2.01 Recirc. Suction LPCI D/G 2.5 Ft (80% DBA) 1973 (1) 1.23 Recire. Suction LPCI D/G 1.9 ft2 (60% DBA) 1791 (1) < I.0 Recire. Suction LPCI D/G 1.0 ft Large 1990 (1) 1.71 Recirc. Suction Break LPCI D/G Method Small 1718 (2) < l.0 Break Methods

.09 ft 2 1404 (2) < l.0 Recirc. Suction HPCS i

l The corevide metal-water reaction for the subject plant has been calculated using method I described in Reference 2. The value is as follows:

Corewide Metal-Water Reaction % = .13 NOTES:

! (1) CHASTE - Large break method (2) Non-DBA reflood (3) The impact of the LPCI and LPCS injection valve pressure interlock is i estimated to be a 51*F increase in PCT and a 0.004 increase in oxidation fraction. Consistent with the estimated increase in PCT, the CWMWR would be 0.16.

J13ATTC85101701 - 7 L _]

Attachment to AECM-85/0341 Page 8 GGNS FSAR TABLE 6.3-6 MAPLHGR, MAXIMUM LOCAL OXIDATION, AND PEAK CLAD TEMPERATURE VERSUS EXPOSURE High Enrichment Fuel Exposure MAPLHGR P.C.T. OXID MWD /T KW/FT DEG = F FRAC 200.0 12.0 2098* 0.0201*

1,000.0 12.0 2087 0.0193 5,000.0 12.4 2069 0.0178 10,000.0 12.6 2071 0.0177 15,000.0 12.6 2083 0.0184 20,000,0 12.6 2085 0.0186 25,000.0 12.1 2014 0.0147 30,000.0 11.1 1885 0.0093 35,000.0 10.2 1764 0.0060 40,000.0 9.6 1692 0.0045 Medium Enrichment Fuel Exposure MAPLHGR P.C.T. OXID MWD /T KW/FT DEG = F FRAC 200.0 11.7 2016 0.0152 1,000.0 11.8 2019 0.0152 5,000.0 12.4 2027 0.0154 10,000.0 12.4 2018 0.0150 15,000.0 12.4 2026 0.0154 20,000.0 12.1 1986 0.0135 25,000.0 11.2 1869 0.0090 30,000.0 10.4 1760 0.0060 35,000.0 9.6 1672 0.0042 40,000.0 9.0 1609 0.0031 i Low Enrichment Fuel l

Exposure MAPLHGR P.C.T. OXID MWD /T KW/FT DEG = F FRAC 200.0 11.5 1960 0.0125 1,000.0 11.4 1929 0.0112 j 5,000.0 11.3 1886 0.0095 10,000.0 11.5 1881 0.0092 l 15,000.0 11.5 1878 0.0091 20,000.0 11.0 1818 0.0073 25,000.0 10.4 1743 0.0055 30,000.0 9.7 1666 0.0040 35,000.0 9.0 1596 0.0029 l

!

• The impact of the LPCI and LPCS injection valve pressure interlock is

{ estimated to be a 51'F increase in PCT and a 0.004 increase in oxidation

fraction.

J13ATTC85101701 - 8 l

1

^

- . - - -

• r w ~e .,y4--%--m-as--sw--yym=w----wa y--w re -e Ftsw-c-t-r--p eet- www--*e,----,ree-geg-r,w--ee-em--_-e- e---- m--wy ,as v ,-++ g m- rw e,

, RC9 R n m9aUBC o  ?%e F F A A E T B '

lR 0

' t L "- 0 1

I t A

F R

O T

A R

E N

E C

L E

S '

E I

D .

K A

E I 0

' C 0 R P B L 3 R ,

E K T A F E A R B

E M N I O T I S

T )

I C .

I U C S S R E 4 E N S V (

1 O

- I l F. .

K 3

l t

i t

T A

/ / 0A 6 o C E R

L U , 0 R E l i D C 2B R S N R l

l A I R C E R C E I l i C E T F T R F A

E , .

D T E I N M S E I N D T I I C

L C E A V

E S L I S .

R A E B 0 l T 0 A N 1 W G I

• S '

E D '

I l '

_~ --__~ .

0 0 'O 0

6 4 2 0 OU* 8O$ sya Mu5 dB4 Uya

_ k

Ckn eO

~ 9keC

%E

• E .

R 0 U_ E 0

L 1 I t A

F R,

O T

A R

E N

E C

L '

E S

E T

D ,

K I 0 A C 0 E P R L 3 B ,

R K E A T E F R A B

)

E N .

M O C I I E T T S C

f

(

5 S i 1

- U S F. S I

K A

3 R U N E 6 E C OI R E

V,D, T 0

.B E N A -

R R

U R A L U R U 0E G S C CR 2 TF I S A F E I R C E P E M R .

• I L , T

. E .

S T S N

- E E V D I

R C

_ O C T A C

A S E I R S .

A B 0 N 0 1

G I

S E

D o 0 o 0 s 0

- 4 50-

1 gE~d wad $

e

3000 FTCURE 6.3-16 PEAK CLADDING TEMPERATURE VERSUS TIHE AFTER BREAK l GRAND GULF l

DESIGN BASIS ACCIDENT, RECIRCULATION SUCTION BREAK, LPCI DIESEL CENERATOR FAII.URE IIIGil POWER AXIAL PLANE REFLOODED itTCllEST POWERED AXTAL PLANE l

{

C - -

HIGHEST POWERED AXIAL PLANE L 2000 TO EXPERIENCE CPR=1.0 PRIOR T0-- -

JET PIR1P UNC0VERY /

[t- -

g BECIN!ilNC OF w /

SPRAY COOLING

/

/

E  ; /

5 /,

h f

'\ r /

f m

1000

__ )- ~~

[ lilCll POWER AXIAL N>>

Pi.ANE uNcovEREo S$"

o~

/ fg J

t ((

~ ag ONSET OF BOILING TRANSTTION CnO 6

0.

0.1 1 10 100 1000 TIME AFTER BREAK (SEC.) -

6 -.