ML14121A424

Westinghouse Responses to NRC Request for Additional Information on Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions

ML14121A424
Person / Time
Site:	Cook
Issue date:	04/29/2014
From:	Westinghouse
To:	Office of Nuclear Reactor Regulation
References
AEP-NRC-2014-27, LTR-PL-14-17, TAC MF2916
Download: ML14121A424 (69)

Text

ENCLOSURE 5 TO AEP-NRC-2014-27 DONALD C. COOK NUCLEAR PLANT UNIT 1 Westinghouse Letter LTR-PL-14-17 NP-Attachment Westinghouse Responses to NRC, "Donald C. Cook Nuclear Plant Unit I -

Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions (TAC No. MF2916)"

Request for Additional Information (RAI) Reactor Systems Branch (SRXB) RAI-1, SRXB RAI-2, SRXB RAI-3, SRXB RAI-4, SRXB RAI-5, SRXB RAI-6, SRXB RAI-7.a, and SRXB RAI-7.b NP-Attachment (Non-Proprietary)

Westinghouse Non-Proprietary Class 3 LTR-PL- 14-17 NP-Attachment Westinghouse Responses to NRC, "Donald C. Cook Nuclear Plant Unit 1 -

Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions (TAC No. MF2916)"

RAIs SRXB RAI-1, SRXB RAI-2, SRXB RAI-3, SRXB RAI-4, SRXB RAI-5, SRXB RAI-6, SRXB RAI-7.a, and SRXB RAI-7.b NP-Attachment (Non-Proprietary)

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066

©2014Westinghouse Electric Company LLC All Rights Reserved NP-1

LTR-PL-14-17 NP-Attachment SRXB RAI-1) Section 5.1.1, "Best-EstimateLarge-Break LOCA [loss of coolant accident],"of WCAP- 17762-NP indicates that the proposednormal operatingpressure (NOP)/normaloperatingtemperature (NOT) restorationwas evaluated using the analysis of record (AOR), approved in 2008, as a baseline. The WCAP is clear that the AOR included the 571 °F value within its range of reactorcoolant system (RCS) average temperature (Tave); however, the hot full power RCS pressure is presented, from the AOR, at both 2100 and 2250 psia. The 2008 ASTRUM implementation LAR (ADAMS Accession No. ML080090268) also includes an allowance for both pressurebands (see Table I of Enclosure 2 to ASTRUM LAR), but it is not clearhow the analysis accounts for these pressure bands.

1.a) Explain how the AOR accounts for the two pressure bands.

1.b) Explain whether the AOR peak clad temperature (PCT)case reflects the higher RCS pressure.

1.c) Explain why the AOR value provided in WCAP-17762-NP-A 3 (2128 OF) differs from that contained in the ASTRUM implementation LAR (2106 °F).

1.d) Explain whether the thermalconductivity degradation(TCD) estimate (ADAMS Accession No. ML12088A104) treated the RCS Pressureconsistently with the AOR and/orthe WCAP

Response

1 a) The analysis of record (AOR) bounds pressurizer pressures of 2100 psia +/-67 psia and 2250 psia +/-67 psia (Table 1 of Reference 1).

2'c Since modeling the pressurizer pressure at 2100 psia was found to conservatively bound modeling the pressurizer pressure at 2250 psia, the uncertainty analysis runs described in Reference 1 were performed considering a pressurizer pressure range of 2100 psia

+/-67 psia.

[

Ia'c lb) As discussed in the response to RAI SRXB la), the AOR uncertainty analysis was executed considering a pressurizer pressure range of 2100 +/-67 psia. Therefore, the AOR peak clad temperature (PCT) case reflects the lower nominal reactor coolant system (RCS) pressure of 2100 psia.

NP-2

LTR-PL-14-17 NP-Attachment ic) The final analysis PCT is shown to be 2128°F in Table 1 of Reference 3. The original LAR submitted (Reference 1), was supplemented by Reference 4, which evaluated a reduction in emergency core cooling system flow.

1d) The thermal conductivity degradation (TCD) estimate (Reference 5) treated the RCS pressure consistently with the AOR. Since no changes were made to the nominal pressurizer pressure in the TCD evaluation, the cases executed are within the AOR uncertainty analysis range of 2100 +/-67 psia, and the pressure assumed in the TCD evaluation is identical to the pressure in the AOR for each case.

a,c Figure 1: [ I a,c NP-3

LTR-PL-14-17 NP-Attachment SRXB RAI-2) Sections 5. 1.1, "Best-EstimateLarge-BreakLOCA, " of WCAP- 17762-NP, contains the following passage:

"Due to the non-lineareffects of the design input changes (which were updated relative to the assessment reportedin Reference 3 [ADAMS Accession No. ML12088A104 - estimated effects of thermal conductivity degradation{TCDJ]),

the return to NOP/NOT evaluation is being assessed againstthe Cook Unit I BE

[best estimate] LB [large break] LOCA analysis of record (AOR), which was submitted in Reference 7 [ADAMS Accession No. ML080090268 - ASTRUM LAR] and approved by the USNRC ... Additionally, due to different cases becoming limiting at NOP/NO T conditions, the priorPCT assessment reported in Reference 10 [August 30, 2013, 30-day reportof significant emergency core cooling system (ECCS) Evaluation Model error/change]is also re-considered..."

The method of selecting limiting cases to determine the effect of a model change on the PCTprediction has been previously reviewed and accepted by the NRC staff; however, the method of identifying and analyzing the case sub-set is a topic of plant-specific review (see, for example, ADAMS Accession No. ML12173A025

- D.C. Cook Response to Request for Additional Information related to TCD estimate). Pleaseprovide information to enable NRC staff review of the case subset selection and validation process.

2.a) Provide a matrix of the significant sampled input parametersfrom the AOR and the various cases executed to estimate the effects of TCD, model changes and error corrections, and the restorationof NOPINOT conditions.

2.b) Provide a summary of the case sub-set selection process: explain how the limiting cases were identified, and what attributeswere identified for the newly limiting cases.

2.c) Explain how the case sub-set selection method was validated,and how the results were verified to be limiting.

Response

2a) Tables 1 through 3 provide data that includes information on the Return to RCS NOP/NOT Evaluation, the original TCD evaluation and the AOR. The key results and attributes are consistent with the information previously provided in Reference 8. Model changes and error corrections were evaluated as follows.

Description of the Evaluation for the Revised Heat Transfer Multiplier Distributions Errors were identified in the development of the original heat transfer multiplier distributions, and revised blowdown heatup, blowdown cooling, refill and reflood heat transfer multiplier distributions were determined. This error was assessed for the D.C.

Cook Unit 1 AOR, and the impact of this error was reported to the NRC in Reference 6 as a 30-day report pursuant to 10 CFR 50.46. This error was also assessed on the Return to RCS NOP/NOT evaluation, following the same evaluation technique described in the Reference 6 enclosure, resulting in a PCT decrease of 91 *F.

NP-4

LTR-PL-14-17 NP-Attachment Similar to the AOR, the Return to RCS NOP/NOT evaluation is considered a "Late Reflood Limited Plant" for the purposes of assessing this error. For "Late Reflood Limited Plants" licensed with the ASTRUM EM, limiting runs tend to sample low reflood heat transfer multipliers. As a result, late reflood plants tended to receive large benefits from the change to the heat transfer multiplier distributions.

Description of the Evaluation for the Error in Burst Strain Application An error in the application of the burst strain was discovered in HOTSPOT. The equation for the application of the burst strain is given as Equation 7-69 in WCAP-16009-P-A and in WCAP-12945-P-A. The outer radius of the cladding after burst occurs should be calculated based on the burst strain, and the inner radius of the cladding should be calculated based on the outer radius. In HOTSPOT, the burst strain is applied to the calculation of the cladding inner radius. The cladding outer radius is then calculated based on the inner radius. As such, the burst strain is incorrectly applied to the inner radius rather than the outer radius, which impacts the resulting cladding geometry at the burst elevation after burst occurs. Correction of the erroneous calculation results in thinner cladding at the burst node and more fuel relocating into the burst node, leading to an increase in the Peak Cladding Temperature (PCT) at the burst node. This issue has been evaluated to estimate the impact on existing Best-Estimate (BE) Large-Break Loss-of-Coolant Accident (LBLOCA) analysis results.

This error was assessed for the D.C. Cook Unit 1 AOR, and the impact of this error was reported to the NRC in Reference 7 as a 30-day report pursuant to 10 CFR 50.46. From prior evaluations (Reference 5), D.C. Cook Unit 1 AOR has plant-specific HOTSPOT calculations that include the impact of fuel pellet TCD and peaking factor burndown.

[

]a~c The large-break LOCA transient response for D.C. Cook Unit 1 AOR TCD evaluation and the Return to RCS NOP/NOT evaluation is similar since the transient response is highly dependent on plant design.

a,c NP-5

LTR-PL-14-17 NP-Attachment I

ja,c NP-6

Table 1: D.C. Cook Unit I (AEP) Integrated Return to RCS NOPINOT Evaluation Runset Data(1 )

IF - a,c

--4 r-B,C-C0 NP-7

____ __________ I_________ - __________ _________ Ii ________ - ___________ = - __________ ____________ - ___________ - ac I-

= - - = - = - - - - = - - - - -D I-F z 0

2 CD

=

NP-8

Table 2: D.C. Cook Unit I (AEP) Original Integrated TCD Evaluation Runset Data(')

=1; a,c

= = U .- = 11= I Y = I I -

- = -. - p -

4 II .4. 4- II- 4. + 4- 4 4 4 II + II- 4- + + 4 4 .4-4 .11 4- .4- II- 4- -4. 4- 4 4 II + II- 4- + 4- 4- 4 4-4 41 + + II- 4 + + 4- 4 T T IF T II r T r T Ii 4 4 II I 4- 4 +

4 Ii II- I- I- 4 +/-

- - - - - - = == I = I I r-

-I I 3,=

I z

"O NP-9

1 Table 3: D.C. Cook Unit I (AEP) AOR Runset Data( )

___IIa,_

I'-

,-o NP-1 0

___________________________ ________________ _______________ JL ____________________ r a,c I"-

-- I z

z'

_ ___ ___ ___ li ____ ____ ___ _____ ____ ____ ____

0 3

CD NP-11

a,c I-

"-I I--

zS..5 3

3 e'D NP-12

axc

~~~~~~.11 __________ __________ _________ A { I z

0 2r NP-13

LTR-PL-14-17 NP-Attachment 2b) The cases from the ASTRUM run matrices that were chosen to assess the effects of TCD for the Return to RCS NOP/NOT evaluation are identified (Table 1) in the response to SRXB RAI-2.a).

Since the effects of fuel TCD are known to increase with burnup, explicit cases were executed in a burnup range that can generally be considered to be more limiting based on the competing effects of initial stored energy and peaking factor burndown. [

]a,c A total of 45 WC/T executions were performed for the D.C. Cook Unit 1 Return to RCS NOP/NOT evaluation,

]ac. The most limiting PCT result was selected as the Integrated TCD PCT value. This is the same process followed in the run selection process for the original TCD evaluation.

a]C For the Return to RCS NOP/NOT evaluation, all runs selected for the integrated run set were also selected for the margin recovery runs. This included the top 31 AOR cases (top 25% of all cases) to be selected for the margin recovery runs. The most limiting PCT result was selected as the Margin PCT to allow estimation of the effect of the margins taken on PCT (consistent with the original TCD evaluation).

The estimate of effect for the compensating model changes was provided by the difference between the AOR PCT and the Margin PCT. The estimate of effect of TCD was provided by the difference between the Integrated TCD PCT and the Margin PCT (consistent with the original TCD evaluation).

]ac NP-14

LTR-PL-14-17 NP-Attachment a,c Based on the results presented for the D.C. Cook Unit I Return to RCS NOP/NOT evaluation, [

Ia,c NP-15

LTR-PL-14-17 NP-Attachment Table 4: Run Selection for D.C. Cook Unit 1 Return to RCS NOP/NOT Evaluation a,c 4 4 4 + + +

4 .4 4 -4 4 4 4 .4 +/- 4 4 4 .4 .4 + 4 4 4 4 4 4- 4 4 4 .4 +/- 4 4 4 4 4 + 4 4 4 4 4 + 4 4 4 I .4 + 4 4

.4 + + 4 -4

• NP-16

LTR-PL-14-17 NP-Attachment 2c)

]a.c Based on the above information it was concluded with a high degree of certainty that the limiting PCT run with the integrated effects of TCD and the margin sources was captured in the runset selection.

In this evaluation, engineering judgment was applied to select runsets of limiting cases for the purpose of evaluating the effects of the design input margins and TCD on the D.C. Cook Unit 1 large break LOCA PCT. The remaining cases from the ASTRUM AOR which were not explicitly evaluated are expected to remain non-limiting and therefore would not be expected to influence the PCT estimate. The evaluations of TCD and peaking factor burndown support the full life of the fuel operation.

NP-17

LTR-PL-14-17 NP-Attachment

References:

1. Letter from J. N. Jensen (I&M), to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1, Docket No. 50-315, License Amendment Request Regarding Large Break Loss-of-Coolant Accident Analysis Methodology," December 27, 2007 (Agencywide Documents Access and Management System (ADAMS), Accession No. ML080090268).

2. WCAP-16009-P-A, Revision 0, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment Of Uncertainty Method (ASTRUM)," January 2005.

3. Letter from T. A. Beltz (NRC) to M. W. Rencheck (I&M), "Donald C. Cook Nuclear Plant, Unit 1 - Issuance of Amendment to Renewed Facility Operating License Regarding Use of the Westinghouse ASTRUM Large Break Loss-of-Coolant Accident Analysis Methodology (TAC NO. MD7556)," October 17, 2008 (Agencywide Documents Access and Management System (ADAMS), Accession No. ML082670351).

4. Letter from L. J. Weber (I&M) to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1, Response to Requests for Additional Information Regarding Reanalysis of Large Break Loss-of-Coolant Accident (TAC No. MD7556)," July 14, 2008 (Agencywide Documents Access and Management System (ADAMS), Accession No. ML082040584).

5. Letter from J. P. Gebbie (l&M) to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2, Response to Information Request Pursuant to 10 CFR 50.54(f) Related to the Estimated Effect on Peak Cladding Temperature Resulting From Thermal Conductivity Degradation in the Westinghouse-Furnished Realistic Emergency Core Cooling System Evaluation (TAC NO. M99899)," March 19, 2012 (Agencywide Documents Access and Management System (ADAMS), Accession No. ML12088A104).

6. Letter from J. P. Gebbie (I&M) to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit I 30-Day Report of Changes to or Errors in an Evaluation Model," August 30, 3013 (Agencywide Documents Access and Management System (ADAMS), Accession No. ML13247A174).

7. Letter from J. P. Gebbie (I&M), to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1, 30-Day Report of Changes to or Errors in an Evaluation Model," February 27, 2014. (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14063A043).

8. Letter from M. H. Carlson (I&M) to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2 Response to Request for Information 10 CFR 50.46 Report for Emergency Core Cooling System Model Change or Error Associated with Thermal Conductivity Degradation," June 11, 2012 (Agencywide Documents Access and Management System (ADAMS), Accession No. ML12173A025).

NP-18

LTR-PL-14-17 NP-Attachment Introduction Responses are provided for all parts of the following request for additional information (RAI) from the Nuclear Regulatory Commission (NRC) regarding the small break loss-of-coolant accident (SBLOCA) analysis described in [1].

SRXB RAI-3) The licensee presents its evaluation of the NOP/NOT restoration with respect to the small break (SB) LOCA analysis in Section 5.1.2 of WCAP-17762-NP. The evaluation is based on an SBLOCA analysis that was provided to the commission by letter dated August 31, 2012 (ADAMS Accession No. MLL12256A685). The succinct evaluation provided in WCAP-17762-NP concludes that the revised SBLOCA analysis explicitly accounts for the restored NOP/NOT conditions.

Noting that the August, 2012, SBLOCA analysis was provided to the Commission, rather than submitted for review and approval, the NRC staff is reviewing the SBLOCA analysis as part of the NOP/NOT review effort to verify that it satisfies applicable regulatory requirements and confirm that it accounts for the proposed NOP/NOT operating conditions.

NP-19

LTR-PL-14-17 NP-Attachment 3.a) Figure 6 provides the core mixture level for the 3.25-inch (limiting) break. The figure shows that the core mixture level remains below 20 feet for a significant period of time (i.e., about 2500 seconds), despite that the PCT node is located at 11.75 feet (NRC staff infers that this elevation corresponds to approximately 21.8 feet on Figure 6). At the time of PCT, 1483 seconds, the hot node does not appear to be covered. The mixture level appears closer to 14.5 feet. Additionally, the rod film heat transfer coefficient depicted in Figure 15 shows that the coefficient is reasonably stable below approximately 50 BTU/hr/ft2 /°F, from 1000 through 3000 seconds of the transient.

Furthermore, the accumulators begin to empty 200 seconds prior to time of PCT.

Please explain how the PCT temperature excursion is being terminated and provide supporting tables and plots with additional data from the NOTRUMP and LOCTA runs.

Response The limiting 3.25-inch break described in [I] depressurizes to the accumulator injection setpoint of 600 psia at approximately 1264 seconds (Figure 3.a-1).. With accumulator injection, the mixture level in the core begins to increase, which decreases the vapor region at the top of the core. With a decreasing core region exposed to vapor above the quench front, the integrated heat up rate experienced by the top elevations of the core decreases. Correspondingly, the top of core vapor temperature begins to decrease (Figure 3.a-2). Clad heat removal depends on both the heat transfer coefficient and the temperature difference between the clad and the surrounding fluid. Since the fluid temperature is decreasing due to accumulator injection, more heat is transferred from the clad to the surrounding fluid, even while the rod filn heat transfer coefficient remains relatively stable; this results in the termination of the cladding temperature excursion (Figure 3.a-3). In addition, the total safety injection to all four loops equilibrates with the total flow out the break (Figure 3.a-4) near the PCT time. The combined effect of accumulator injection (until they empty) and continued safety injection prevent subsequent cladding heat up. This is illustrated by Figure 3.a-5 which shows the core mixture level on the left axis and the total break flow versus the combined accumulator injected flow and safety injection flow on the right axis.

NP-20

LTR-PL-14-17 NP-Attachment 3.25-Inch Break Accumulator Flow and RCS Pressure Moss Flow Rote (Ibm/s)

Total Accumulator Flow Pressure (psio)

- Reactor Coolant System Pressure 2

(I)

C,)

0 C,)

U~)

0

- 400 1000 1200 1400 1600 1800 2000 Time (s)

Figure 3.a- 1: Accumulator Flow and RCS Pressure NP-21

LTR-PL-14-17 NP-Attachment 3.25-Inch Break Core Mixture Level and Vapor Temperature Mixture Level (ft)

Core Mixture Level Temperature (F)

Top of Core Vapor Temperature 1-ý Ij-75 a)

=3 XJ 1000 1200 1400 1600 1800 2000 Time (s)

Figure 3.a- 2: Core Mixture Level and Vapor Temperature NP-22

LTR-PL-14-17 NP-Attachment 5.25-Inch Break Clad Temperature at PCT Elevation (11.75) w

'I)

I-0

~1) 0~

1000 1200 1400 1600 1800 2000 Time (s)

Figure 3.a- 3: Cladding Temperature at PCT Elevation NP-23

LTR-PL-14-17 NP-Attachment 3.25-Inch BrE ak Break and Safety Inje ction Flows Total Break Flow Total Pumped Safety Injection E

Cl)

M, 1000 1200 1400 1600 1800 2000 Time (s)

Figure 3.a- 4: Total Break Flow and Safety Injection Flows NP-24

LTR-PL-14-17 NP-Attachment 3.25-Inch Break Mixture Level and Break vs. Safety Injection and Accumulator Flows Mixture Level (ft)

Core Mixture Level Top of Core = 22.0762 ft Moss Flow Rote (lbm/s)

Totol Break Flow Totol Pumped Safety Injection + Total Accumulator Flow (I)

.4-

'4- E Q) 0 ci~

C,,

W 1000 1200 1400 1600 1800 2000 Time (s)

Figure 3.a- 5: Core Mixture Level and Total Break Flow vs. SI and Accumulator Flows NP-25

LTR-PL-14-17 NP-Attachment 3.b) Describe the loop seal clearing behavior depicted in Table 6 in greater detail. For all the breaks, provide the thermal-hydraulic conditions present in the intact loop seals. For the limiting break in particular, describe the reactor coolant conditions immediately prior to and following the loop seal clearing, especially with regard to the effect that the loop seal clearing has on the mixture level transient and system pressure.

Response The loop seal clearing behavior depicted in Table 6 of [1] reflects the timing of the broken loop, loop seal clearing, which allows vapor to vent through the reactor coolant pump suction cross-over leg (i.e., loop seal) and out the break. For break sizes with equivalent diameters less than [ ]a,, (including the limiting 3.25-inch break),

the intact loop seals are restricted from clearing. This is illustrated by Figure 3.b-l, which shows the vapor mass flow rate in each of the loops for the 3.25-inch break.

Figure 3.b-2 scales the vapor flow to the period surrounding the loop seal clearing time.

Figure 3.b-3 provides the mixture level in the pump suction crossover legs (loop seals) for the broken loop and one intact loop (the remaining intact loops show the same behavior but are not provided here). Figure 3.b-3 shows that the broken loop, loop seal clears while the intact loop, loop seal does not. Similar behavior is observed for all break sizes less than [ Ic.

Prior to the broken loop, loop seal clearing for the 3.25-inch limiting break, the RCS pressure experiences an initial depressurization to a pressure near the minimum lift pressure of the main steam safety valves (approximately 1150 psia) (Figure 3.b-4).

During this early part of the transient, the pump suction pipe U-bend remains filled with liquid (i.e. the loop seal is not clear), sealing off steam flow to the break (See Figures 3.b-1 and 3.b-2). Therefore, the break flow is entirely liquid (Figure 3.b-5),

which in conjunction with the low safety injection flow rates associated with the high system pressure, results in a net reduction in primary system mass (Figure 3.b-6).

Figure 3.b-7 illustrates how the core mixture level has also decreased during this time.

Once the loop seal clears, the steam generated by the core decay heat can vent through the break and increase the depressurization rate (See Figure 3.b-4). The brief mixture level depression observed at approximately 500 seconds in Figure 3.b-7 corresponds to the reduced broken loop, loop seal vapor flow seen in Figure 3.b-2. However, the loop seal does not replug, and the core mixture level recovers when the broken loop, loop seal vapor flow increases again. Continued loss of system mass and core level reduction due to the boil-off leads to core uncovery at approximately 780 seconds for the 3.25-inch break (See Figure 3.b-7).

For break sizes equal to or greater than [ ]aC equivalent diameter, the loop seal restriction is removed, allowing the loop seals in each loop to clear. A check is made to confirm that the broken loop, loop seal clears first. Figure 3.b-8 shows the vapor mass flow rate in each of the loops for the 8.75-inch break and Figure 3.b-9 scales the vapor flow to the period surrounding the broken loop, loop seal clearing time (30 seconds). Figure 3.b-10 provides the mixture level in the pump suction crossover legs (loop seals) for the broken loop and one intact loop. It is observed from Figures 3.b-9 and 3.b-I 0 that the broken loop, loop seal clears first for the 8.75-inch break.

NP-26

LTR-PL-14-17 NP-Attachment 5.25-Inch Break Loop Seal Vapor Flow Loop 1 (Broken Loop) Loop Seal Vapor Flow Loop 2 Loop Seal Vapor Flow

-Loop 3 Loop Seal Vapor Flow Loop 4 Loop Seal Vapor Flow 300 250-200.

E 150 0) 100..

Cl)

M i) 0 1000 2000 3000 4000 5000 Time (s)

Figure 3.b- 1: Loop Seal Vapor Mass Flow Rate for All Loops NP-27

LTR-PL-14-17 NP-Attachment 3.25-Inch Break Loop Seal Vapor Flow Loop 1 (Broken Loop) Loop Seal Vapor Flow Loop 2 Loop Seal Vapor Flow Loop 3 Loop Seal Vapor Flow Loop 4 Loop Seal Vapor Flow 300 250

~200-E a:: 150" 0 C/)

C/)

0 200 400 600 800 1000 Time (s)

Figure 3.b- 2: Loop Seal Vapor Mass Flow Rate for All Loops Scaled to Loop Seal Clearing Time NP-28

LTR-PL-14-17 NP-Attachment 3.25-Inch Break Pump Suction Cross-Over Leg Mixture Level Broken Loop Mixture Level Intact Loop Mixture Level Bottom of Cross-Over Leg Top of Cross-Over Leg 30 25-~ Z:I . . . . . . . .- -.- - -. - -. - . -

-- . - - ---- - - - - L - - -

20"

-D X3 i ii I I I ' /

15" IA-i IV 0 200 400 600 800 1000 Time (s)

Figure 3.b- 3: Mixture Level in the Pump Suction Crossover Leg (Loop Seal)

NP-29

LTR-PL-14-17 NP-Attachment 3.25-Inch Break Reactor Coolant System Pressure E1500-W, (n~

r1) 0 200 400 Time 1000 (s) 600 800 Figure 3.b- 4: Reactor Coolant System Pressure NP-30

LTR-PL-14-17 NP-Attachment 3.25-Inch Break Break Flow Mass Flow Rote (Ibm/s)

Break Liquid Flow Moss Flow Rote (Ibm/s)

Break Vapor Flow 140 120 100 11- U)

E E 41) 0 M3 0'

0 200 400 600 800 1000 Time (s)

Figure 3.b- 5: Break Vapor and Liquid Flows NP-31

LTR-PL-14-17 NP-Attachment 3.25- Inch Break Primary System Mass 0

800 1000 0 200 400 lime (s) 600 Figure 3.b- 6: Primary System Mass NP-32

LTR-PL-14-17 NP-Attachment 3.25-Inch Break Core Mixture Level Core Mixture Level Top of Core = 22.0762 ft X3 0 200 400 600 800 1000 Time (s)

Figure 3.b- 7: Core Mixture Level NP-33

LTR-PL-14-17 NP-Attachment 8.75-Inch Break Loop Seal Vapor Flow Loop 1 (Broken Loop) Loop Seal Vapor Flow Loop 2 Loop Seal Vapor Flow Loop 3 Loop Seal Vapor Flow Loop 4 Loop Seal Vapor Flow

KAn .,

""UU 400-

-~300' 0

to Co 200" 0J Ii I IJ .1 100, 01 0 1000 2000 3000 4000 5000 Time (s)

Figure 3.b- 8: Loop Seal Vapor Mass Flow Rate for All Loops NP-34

LTR-PL-14-17 NP-Attachment 8.75-Inch Break Loop Seal Vapor Flow Loop 1 (Broken Loop) Loop Seal Vapor Flow Loop 2 Loop Seal Vapor Flow Loop 3 Loop Seal Vapor Flow Loop 4 Loop Seal Vapor Flow C,)

E 0 20 40 60 80 100 Time (s)

Figure 3.b- 9: Loop Seal Vapor Mass Flow Rate for All Loops Scaled to Loop Seal Clearing Time NP-35

LTR-PL-14-17 NP-Attachment 8.75-Inch Break Pump Suction Cross-Over Leg Mixture Level Broken Loop Mixture Level Intact Loop Mixture Level Bottom of Cross-Over Leg Top of Cross-Over Leg 30 25-20-L.)

-J 20 . -.- - ---

0 20 40 60 80 100 Time (s)

Figure 3.b- 10: Mixture Level in the Pump Suction Crossover Leg (Loop Seal)

NP-36

LTR-PL-14-17 NP-Attachment 3.c) The greatest fraction of pumped safety injection flows into the broken loop. Due to the large variation in liquid flow out the break throughout the duration of the transient, it is difficult to evaluate the broken loop flow behavior. Please provide detailed plots of liquid and vapor flow rates at the junctions or links connecting the broken loop to pumped safety injection sources, the break, the reactor coolant pump, and the vessel, for the first 2000 seconds of the limiting break. Include scaling appropriate prior to and following the loop seal clearing.

Response The figures described below are provided for the limiting 3.25-inch break. Note that the broken loop, loop seal clearing time is 445 seconds for the limiting break.

The broken loop pumped safety injection flow rate is shown in Figure 3.c-l for the first 2000 seconds of the transient and in Figure 3.c-2 for the period prior to and following the broken loop, loop seal clearing time.

The liquid and vapor break flow rates are shown in Figure 3.c-3 for the first 2000 seconds of the transient and in Figure 3.c-4 for the period prior to and following the broken loop, loop seal clearing time.

The total (liquid plus vapor) break flow rate is shown in Figure 3.c-5 for the first 2000 seconds of the transient and in Figure 3.c-6 for the period prior to and following the broken loop, loop seal clearing time.

The liquid and vapor flow rates from the reactor coolant pump (RCP) to the broken loop cold leg (CL) are shown in Figure 3.c-7 for the first 2000 seconds of the transient and in Figure 3.c-8 for the period prior to and following the broken loop, loop seal clearing time.

The liquid and vapor flow rates from the broken loop CL nozzle to the vessel are shown in Figure 3.c-9 for the first 2000 seconds of the transient and in Figure 3.c-10 for the period prior to and following the broken loop, loop seal clearing time.

NP-37

LTR-PL-14-17 NP-Attachment 5.25-Inch Break Broken Loop Pumped Safety Injection 40-30-a 20-.. . ..

11) 10) 0500 1000 100 2000 Time (s)

Figure 3.c- 1: Broken Loop Pumped Safety Injection Flow Rate (0 to 2000 seconds)

NP-38

LTR-PL-14-17 NP-Attachment 3.25-Inch Break Broken Loop Pumped Safety Injection Co-E 0n U/

MI 200 300 400 Time (s) 500 600 700 Figure 3.c- 2: Broken Loop Pumped Safety Injection Flow Rate (200 to 700 seconds)

NP-39

LTR-PL-14-17 NP-Attachment 3.25-Inch Break Break Flow Moss Flow Rote (Ibm/s)

Break Liquid Flow Moss Flow Rote (Ibm/s)

Break Vapor Flow 2000 140 120 I /'I \.

1500 i ,t I 100 E

Cn' 0.)\

U-r 1000.

0) 0 Cl)

Cnl I 500-50 . .. ... .. . .. . .. . . . ....... . .. ..

0 500 1000 1500 2000 Time (s)

Figure 3.c- 3: Liquid and Vapor Break Flow Rates (0 to 2000 seconds)

NP40

LTR-PL-14-17 NP-Attachment 3.25-Inch Break Break Flow Moss Flow Rote (Ibm/s)

Break Liquid Flow Moss Flow Rote (Ibm/s)

Break Vapor Flow 140 120 100 C,) C,)

E 2

-o -80

0) '3) 0 0 0 60~

C,) C,)

C,) C,)

0 AO 20 200 300 400 500 600 700 Time (s)

Figure 3.c- 4: Liquid and Vapor Break Flow Rates (200 to 700 seconds)

NP-41

LTR-PL-14-17 NP-Attachment 3.25-Inch Break Total Break Flow A-o Cl) 0 500 1000 1500 2000 Time (s)

Figure 3.c- 5: Total (Liquid + Vapor) Break Flow Rate (0 to 2000 seconds)

NP-42

LTR-PL-14-17 NP-Attachment 3.25-Inch Break Total Break Flow 11-1 E

0 M

200 300 400 500 600 700 Time (s)

Figure 3.c- 6: Total (Liquid + Vapor) Break Flow Rate (200 to 700 seconds)

NP-43

LTR-PL-14-17 NP-Attachment 5.25-Inch Break Broken Loop Reactor Coolant Pur np Moss Flow Rote (Ibm/s)

Reactor Coolant Pump to Cold Leg Liquid Flow Moss Flow Rote (Ibm/s)

- Reactor Coolant Pump to Cold Leg Vapor Flow 10000O 8000"

~6000-

.a* ~ ~ IIl ,

E Q.)

EIt, IIl 0

LjCI) 0 ..

I..

IrI

• CI) 0 500 1000 1500 2000 Time (s)

Figure 3.c- 7: Liquid and Vapor Flow Rates from the RCP to the Broken Loop Cold Leg (0 to 2000 seconds)

NP-44

LTR-PL-14-17 NP-Attachment 3.25-Inch Break Broken Loop Reactor Coolant Pump Moss Flow Rote (Ibm/s)

Reoctor Coolant Pump to Cold Leg Liquid Flow Moss Flow Rote (Ibm/s)

Reoctor Coolont Pump to Cold Leg Vopor Flow 1500-1000.

.1-1 E

E)

S500.

0_o 200 300 400 500 600 700 Time (s)

Figure 3.c- 8: Liquid and Vapor Flow Rates from the RCP to the Broken Loop Cold Leg (200 to 700 seconds)

NP-45

LTR-PL-14-17 NP-Attachment 3.25-Inch Break Broken Loop Vessel Inlet Mass Flow Rote (Ibm/s)

Cold Leg Nozzle Liquid Flow Mass Flow Rate (Ibm/s)

Cold Leg Nozzle Vopor Flow C1) 0 0 0A U,

C,)

M 0 500 1000 1500 2000 Time (s)

Figure 3.c- 9: Liquid and Vapor Flow Rates from the Broken Loop CL to the Vessel (0 to 2000 seconds)

NP-46

LTR-PL-14-17 NP-Attachment 3.25-Inch Break Broken Loop Vessel Inlet Moss Flow Rote (Ibm/s)

Cold Leg Nozzle Liquid Flow Moss Flow Rote (Ibm/s)

- Cold Leg Nozzle Vopor Flow C) 03 0)ý 0 (D) 0-03 M'

200 300 400 500 600 700 Time (s)

Figure 3.c- 10: Liquid and Vapor Flow Rates from the Broken Loop CL to the Vessel (200 to 700 seconds)

NP-47

LTR-PL-14-17 NP-Attachment 3.d) Describe the modeling of flow paths between the downcomer and the upper plenum and core barrel.

Response The NOTRUMP code models a flow link (FL) between pairs of nodes identified below.

Figure 3-14-1 of [2] illustrates the noding scheme used by NOTRUMP to model the broken loop and one intact loop. The FL between each pair of nodes is provided below. (Note that the analysis described in [)) modeled all four D. C. Cook Unit I loops explicitly.)

ac NP-48

LTR-PL-14-17 NP-Attachment 3.e) Provide plots of the hot assembly void fraction as a function of height for the limiting break at the time of minimum core level, and again at the time of PCT.

Response The NOTRUMP RCS model contains four equally spaced nodes that represent the core at four elevations: the bottom of the core, the lower-mid core, the upper-mid core, and the top of the core. The void fraction for these core average nodes is provided in Figure 3.e-1. A scaled figure for the period just prior to and following the time of minimum core mixture level (approximately 1340 seconds) is provided in Figure 3.e-2.

A scaled figure for the period just prior to and following the maximum clad temperature time (1483 seconds) is provided in Figure 3.e-3. Note that when the entire node is voided, the void fraction goes to zero. This is observed between approximately 1000 and 1800 seconds for the top core node in Figure 3.e-1, and for the entire duration shown in Figures 3.e-2 and 3.e-3.

NP-49

LTR-PL-14-17 NP-Attachment 3.25-Inch Break Core Void Fraction Bottom of Core Lower-Mid Core Upper-Mid Core Top of Core 0.7 0.6 0.5" 0 0.4-"

> 0.3 .

0.2" Ai oil 20 00 F 0 000 3000 4000 5000 Time (s)

Figure 3.e- 1: Core Average Void Fraction NP-50

LTR-PL-14-17 NP-Attachment 3.25 -Inch Break Core Void Fraction Bottom of Core Lower-Mid Core Upper-Mid Core Top of Core 0

U-4 0Q0 0

1000 1100 1200 1300 1400 1500 1600 Time (s) 3.e- 2: Core Average Void Fraction Scaled for Minimum Core Mixture Level Time NP-51

LTR-PL-14-17 NP-Attachment 3.25 -Inch Break Core Void Fraction Bottom of Core Lower-Mid Core Upper-Mid Core Top of Core 0.7 0.6 0.5-hut 0 0.4-LVat '

~0.3 1300 1400 1500 1600 1700 Time (s)

Figure 3.e- 3: Core Average Void Fraction Scaled for Maximum Clad Temperature Time NP-52

LTR-PL-14-17 NP-Attachment NP-53

LTR-PL-14-17 NP-Attachment 5.25-Inch Break Core Exit Mass Flow Rate Moss Flow Rote (Ibm/s)

Core Exit Liquid Flow Moss Flow Rote (rbm/s)

Core Exit Vapor Flow 40000 EI 20000<

1000 0 " ..... .

-210000"

-o L0

-100(0 0 1000 1000 2000 2000 3000 3000 4000 4000 Time (s)

Figure 3.f- 1: Core Exit Mass Flow Rate NP-54

LTR-PL-14-17 NP-Attachment

References:

I. Letter from Joel P. Gebbie, Indiana Michigan Power Company (I&M), to Nuclear Regulatory Commission (NRC) Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1, Revised Small Break Loss-of-Coolant Accident Analysis," AEP-NRC-201 2-71, dated August 31, 2012 (ADAMS Accession Number ML12256A685).

2. Lee, N. et al., "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code,"

WCAP-10054-P-A, August 1985.

NP-55

LTR-PL-14-17 NP-Attachment Nuclear Regulatory Commission (NRC) Request for Additional Information (RAI) SRXB RAI-4 The D.C. Cook post-LOCA long term cooling (LTC) analyses demonstrate that boric acid concentration control measures are adequate, and that the ECCS recirculationflows "dilute the core and replace core boil-off, thus keeping the core quenched." WCAP 17762-NP refers to an analysis (ADAMS Accession No. MLI1l95A025), which is performedfor D.C. Cook Unit 2.

4.a) Please explain how the calculationconcludes that the ECCS recirculationflow is adequate.

4.b) Please addressdifferences between the D.C. Cook Units.

4.c) The Unit 2 analysis states the following: "The current hot leg switchover time and plant operating procedures result in ECCS flows that temporarily drop below the injectedflow necessary to replace core boil-off (plus entrainmen't)during the HLSO process."

4.c (i) Please explain what consequence, if any, the hot leg swapover evolution could have on maintaininga stable core quench.

4.c (ii) Please explain how this calculationaccountsfor entrainment.

4.d) Section 5.1.3 of WCAP-17762 does not appear to indicate, as other sections of the WCAP do, that the boric acid precipitation analysis reflects the Unit 1 NOP/NOT values. The WCAP states, "The inputs used to perform post-LOCA LTC analyses include core power levels, fuel dimensions, and RCS and ECCS volumes, temperatures,pressures, and boron concentrations." Explain whether, and how, these inputs are affected by the NOP/NOT restoration,and whether, and how, the analysis accounts for the NOP/NOT restoration.

Response to NRC RAI SRXB RAI-4 4.a) Pleaseexplain how the calculationconcludes that the ECCS recirculationflow is adequate.

D.C. Cook's post-loss-of-coolant accident (post-LOCA) long-term cooling (LTC) calculations conclude that the emergency core cooling system (ECCS) recirculation flows are adequate by comparing the minimum hot leg recirculation charging and safety injection (SI) pump flow rates to the core boil-off rate at hot leg switchover (HLSO). In particular, the calculations performed for D.C. Cook show that:

• For a break in a reactor coolant system (RCS) hot leg, the charging pump flow rate to the RCS cold legs exceeds core boil-off and provides sufficient dilution flow during HLSO and hot leg recirculation.

• For a break in an RCS cold leg, the charging pump flow rate to the RCS cold legs and available liquid inventory above the active fuel are sufficient to keep the core covered during the 15-minute HLSO.

" For a break in an RCS cold leg, the concentration of boric acid in the core remains below the atmospheric solubility limit during the 15-minute HLSO.

• For a break in an RCS cold leg, the SI pump flow rate to the RCS hot legs exceeds core boil-off and provides sufficient dilution flow during hot leg recirculation.

NP-56

LTR-PL-14-17 NP-Attachment 4.b) Please addressdifferences between the D.C. Cook Units.

Differences between D.C. Cook Units 1 and 2 include core power levels, fuel dimensions, and RCS volumes, temperatures, and pressures. The post-LOCA LTC analysis of record (AOR) for D.C. Cook Unit 1 is the same as the post-LOCA LTC AOR for D.C. Cook Unit 2 and considers these differences. For example, the HLSO time determined for D.C. Cook Units I and 2 is based on the Unit 2 licensed core power level which is larger, and therefore more limiting, than the Unit 1 licensed core power level.

Similarly, the HLSO time determined for D.C. Cook Units 1 and 2 uses an initial RCS liquid mass that is based on the Unit 1 RCS volume; for D.C. Cook, the RCS liquid is a post-LOCA dilution source and as such, the smaller Unit 1 RCS volume is more limiting than the larger Unit 2 RCS volume. A summary of D.C.

Cook's boric acid precipitation control (BAPC) AOR was provided to the United States Nuclear Regulatory Commission (USNRC) in fulfillment of an American Electric Power commitment to perform an updated analysis of the potential for boric acid precipitation to occur during the recirculation phase of a postulated LOCA (Reference 1).

4.c) The Unit 2 analysis states the following: "The current hot leg switchover time and plant operating procedures result in ECCS flows that temporarily drop below the injectedflow necessary to replace core boil-off (plus entrainment)during the HLSO process."

4.c (i) Please explain what consequence, if any, the hot leg swapover evolution could have on maintaining a stable core quench.

4.c (ii) Pleaseexplain how this calculationaccounts for entrainment.

The post-LOCA LTC calculations performed for D.C. Cook demonstrate that a stable core quench will be maintained during and after HLSO. In particular, calculations performed for D.C. Cook show that:

• For a break in an RCS hot leg, the charging pump flow rate to the RCS cold legs exceeds core boil-off and provides sufficient dilution flow at an early HLSO time of 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

• For a break in an RCS cold leg, the charging pump flow rate to the RCS cold legs and available liquid inventory above the active fuel are sufficient to keep the core covered for more than 30 minutes at an early HLSO time of 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />; these calculations assume that the downcomer is initially filled with liquid to the bottom of the cold legs, that the core/upper plenum mixture level is initially at the bottom of the hot legs, that the core and downcomer are in hydrostatic balance, and that the relative pressures in the upper plenum and downcomer remain unchanged during HLSO. For D.C. Cook, the switchover from cold leg recirculation to hot leg recirculation is required to be performed in 15 minutes or less.

• For a break in an RCS cold leg, the SI pump flow rate to the RCS hot legs exceeds core boil-off and provides sufficient dilution flow during hot leg recirculation.

The post-LOCA LTC calculations performed for D.C. Cook account for entrainment in the following manner:

NP-57

LTR-PL-14-17 NP-Attachment

• The impact of core mixing volume decreases on the boric acid concentration during HLSO is determined using a core boil-off multiplier of 1.2. The core boil-off and entrainment are assumed to be boron-free.

• The time at which the steam flow in the hot legs drops below the entrainment threshold is determined using the Ishii-Grolmes (Reference 2) and Wallis-Steen (Reference 3) liquid entrainment onset criteria. Calculations performed for D.C. Cook show that steam flow in the hot legs should drop below the entrainment threshold at approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 3 minutes based upon the Appendix K decay heat function. The early HLSO time of 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> evaluated for D.C. Cook is well after the calculated time at which the decay heat steaming rate is able to entrain SI pump flow around the loops.

4.d) Section 5.1.3 of WCAP-1 7762 does not appear to indicate, as other sections of the WCAP do, that the boric acid precipitation analysis reflects the Unit 1 NOP/NOT values. The WCAP states, "The inputs used to perform post-LOCA LTC analyses include core power levels, fuel dimensions, and RCS and ECCS volumes, temperatures,pressures, and boron concentrations."Explain whether, and how, these inputs are affected by the NOP/NOT restoration,and whether, and how, the analysis accounts for the NOP/NOT restoration.

The post-LOCA LTC inputs reviewed in support of the D.C. Cook Unit 1 restoration of NOP/NOT are summarized in Section 5.1.3.2.5 of WCAP-17762-NP, Revision 1 (Reference 4) and consist of the initial RCS liquid mass, ice melt rates, total ice melt mass, and ECCS recirculation flows. The review performed concluded that:

• The initial RCS liquid mass used in D.C. Cook's post-LOCA BAPC AOR is based on an RCS pressure equal to the proposed Unit 1 NOP, but an RCS average temperature that is less than the proposed NOT; for D.C. Cook, the RCS liquid is a post-LOCA dilution source and as such, a smaller RCS liquid mass is more limiting than a larger RCS liquid mass. As a result, a sensitivity calculation was performed using the proposed NOP of 2250 psia and an RCS average temperature that bounds the NOT for D.C. Cook Units 1 and 2. The HLSO time determined in this sensitivity was virtually the same as the HLSO time determined in D.C. Cook's BAPC AOR.

• The ice melt rates used in D.C. Cook's post-LOCA BAPC AOR bound the change in initial RCS stored energy resulting from the restoration of NOP/NOT.

• The total ice melt mass used in D.C. Cook's post-LOCA BAPC AOR is based on the Technical Specifications Surveillance Requirement 3.6.11.2 minimum-allowable ice bed mass which is not changing as a result of the restoration of NOP/NOT.

• There are no changes to the ECCS design or method of operation associated with the restoration of NOP/NOT that would invalidate the ECCS flow rates or flow rate reductions (e.g., at HLSO) evaluated in D.C. Cook's post-LOCA LTC calculations.

NP-58

LTR-PL-14-17 NP-Attachment References

1. M.H. Carlson (AEP) to USNRC Document Control Desk, "Donald C. Cook Nuclear Power Plant Unit 2 Docket No. 50-316 Updated Boric Acid Precipitation Analysis for Recirculation Phase of a Postulated Large-Break Loss-Of-Coolant Accident (TAC No. ME1017)," June 30, 2011. (ADAMS Accession No. ML11193A047)

2. Ishii, M.; Grolmes, M. A., "Inception Criteria for Droplet Entrainment in Two-Phase Concurrent Film Flow," AIChE Journal, Vol.21, No. 2, pp. 308-319, 1975.

3. Wallis, G. B., "One-Dimensional Two-Phase Flow," pp. 390-393, 1969.

4. WCAP-17762-NP, Revision 1, "D. C.Cook Unit 1 Return to Reactor Coolant System Normal Operating Pressure/Normal Operating Temperature Program - Licensing Report," September 2013.

NP-59

LTR-PL-14-17 NP-Attachment Request for Additional Information SRXB RAI-5:

Subsection 5.2.1, "Introduction and Background," to Section 5.2, "Non-LOCA Transients," discusses evaluation for events that take credit for the lower temperature/pressure,stating, "In particularwere the overtemperatureAT (OTA T) and overpower AT (OPA4T) setpoints, which utilized T' and T" values that were restricted below the full power Tavg primarily to provide overpower protection while maintainingthe same AT selpoints."

Subsection 5.2.3.2 discusses the UncontrolledRod Withdrawal at Power, and states, "Additionally,it was confirmed as part of the Return to RCS NOP/NOT Program that the OTAT setpoints modeled in the current analysis remain valid at NOP/NOTconditions. " Please explain how this confirmation was performed and provide additionaldetail regarding the results of the confirmation. In particular,explain whether the T' and T" values assumed in the Rod Withdrawalat Power analyses reflect the more restrictive values, and if so, how the setpoints remain validfor the proposed operatingconditions.

Response to RAI - SRXB RAI-5:

Applying the method described in WCAP-8745-P-A (Reference 1), the current OTAT and OPAT setpoints were found to protect the core thermal limits without restricting T' and T" below the full power Tavg. This is consistent with previous calculations, which also showed that the core thermal limits would be protected without any restrictions on T' and T"; however, restrictions were previously placed on T' and T" for operational flexibility and to maintain consistency with design bases. These restrictions are no longer required with the implementation of the Return to Reactor Coolant System (RCS) Normal Operating Pressure/Normal Operating Temperature (NOP/NOT) Program.

A restriction on T' was previously instituted to demonstrate that the negative side of the f(AI) penalty function provided adequate protection for reduced primary operating temperature and pressure conditions. With the return to a higher RCS temperature and pressure, the conditions for which the T' modification was required no longer exists; therefore, the restriction is no longer needed.

A restriction on T" was previously instituted to limit the potential overpower condition to 118% power, which simplified the process of confirming that overpower protection was provided for all of the applicable non-LOCA events, particularly the steamline break (SLB) event initiated from full power conditions. For the Return to RCS NOP/NOT Program, an explicit analysis of the SLB event from full power conditions was performed. The full power SLB analysis, in conjunction with the AT setpoint confirmation and evaluation of the continued applicability of the current RWAP analysis, demonstrates protection against overpower conditions without relying on a restriction on T".

In terms of impact on the Rod Withdrawal at Power analysis, no credit was taken in the analysis for the benefit of restricted T' or T" values; the values were set consistent with the Tavg modeled in the analyses. Thus, removing the restrictions on T' and T" had no impact on the Rod Withdrawal at Power analysis.

Reference:

1. Westinghouse Report WCAP-8745-P-A, "Design Bases for the Thermal Overpower Delta T and Thermal Overtemperature Delta T Trip Functions," September 1986.

NP-60

LTR-PL-14-17 NP-Attachment SRXB RAI-6) A copy of NSAL-07-11 is provided in Attachment 2.

SRXB RAI-7.a) Table I provides a comparison of the analytical assumptions used in WCAP- 10698-P-A (Reference 1) to those used in the D.C. Cook Unit I NOP/NOT SGTR MTO analysis.

Table 1: Comparison of WCAP-10698-P-A Modeling to the D.C. Cook Unit 1 NOP/NOT SGTR MTO Analysis Assumptions SMo n D.C. Cook Unit 1 SGTR MTO Parameter WCAP-10698-P-A Modeling Analysis I Direction of Conservatism NOP/NOT Initial Conditions______________ _________________

Power Full-power Full-power (nominal)(')

Power_ _(nominal + uncertainty)

RCS Pressure Minimum Nominal(')

Pressurizer Water Level Maximum Nominal("

Steam Generator Secondary Mass Maximum Nominal("

Break Location Cold-leg Cold-leg Offsite Power Availability Offsite Power Loss of Offsite Power (LOOP) LOOP Protection Setpoints and Errors Reactor Trip Delay Minimum Minimum Turbine Trip Delay Minimum Minimum SG Relief or Safety Valve Pressure Minimum (PORV) Minimum (PORV)

Setpoint Minimum_(PORV)_Minimum_(PORV)

Pressurizer Pressure Trip Setpoint Maximum Nominal(")

Pressurizer Pressure SI Setpoint Maximum Nominal(')

Safeguards Capacity SI Flow Rate Maximum Maximum AFW Flow Rate (isolation on Maximum Maximum operator action time)

AFW System Delay Minimum Minimum AFW Temperature Maximum Maximum Control Systems CVS Operation, PZR Heater Control Not operating Not operating Turbine Runback Mass Penalty Included Not included')

RCP Running Not operating Not operating Decay Heat Decay Heat Maximum ANS 1979-2a(2)

Single Failure Single Failure Included Not included")

(1) These deviations are consistent with the current approved D.C. Cook Unit 1 SGTR MTO analysis (Reference 2).

(2) For this revised analysis, the 1979 American Nuclear Society (ANS) decay heat model minus 2y uncertainty is used. Plant specific sensitivities performed to address the NSAL-07-11 (Reference 3) determined that for D.C. Cook Unit 1 the use of 1979-2y decay heat is conservative compared to the 1971+20% ANS decay heat model specified by the methodology of WCAP-10698-P-A.

NP-61

LTR-PL-14-17 NP-Attachment SRXB RAI-7.b) The D.C. Cook Unit 1 NOP/NOT SGTR MTO analysis modeled the replacement steam generators, consistent with the current approved SGTR MTO analysis (Reference 2).

References:

1. Westinghouse Report WCAP- 10698-P-A, "SGTR Analysis Methodology to Detennine the Margin to Steam Generator Overfill," August 1987.

2. Letter from John F. Stang (U.S. NRC) to Robert P. Powers (AEP), "Donald C. Cook Nuclear Plant, Units 1 and 2 - Issuance of Amendments (TAC Nos. MB0739 and MB0740)," dated October 21, 2001. (Available in NRC ADAMS under Accession Number ML012690136)

3. NSAL-07-1 1, "Decay Heat Assumption in Steam Generator Tube Rupture Margin-to-Overfill Analysis Methodology," dated November 15, 2007.

NP-62

LTR-PL-14-17 NP-Attachment : NSAL-07-11 NP-63

LTR-PL-14-17 NP-Attachment Nuclear Safety wW8singhouse

,Adv Letter This is a notification of a recently identified potential safety issue pertaining to basic components supplied by Westinghouse.

This information is being provided so that you can conduct a review of this issue to determine ifany action is required.

P.O. Box 355, Pittsburgh, PA 15230

Subject:

Decay Heat Assumption in Steam Generator Tube Rupture Number: NSAL-07-11 Margin-to-Overfill Analysis Methodology Basic Component: Steam Generator Tube Rupture Analysis Date: 11/15/2007 Affected Plants: See page 4 Substantial Safety Hazard or Failure to Comply Pursuant to 10 CFR 21.21(a) Yes El No Z N/A El Transfer of Information Pursuant to 10 CFR 21.21(b) Yes El Advisory Information Pursuant to 10 CFR 21.21(d)(2) Yes E]

References:

See page 5

SUMMARY

A generic methodology for the steam generator tube rupture (SGTR) margin-to-ruptured SG overfill analysis was developed by a subgroup of the Westinghouse Owners Group (WOG) and is documented in WCAP-10698-P-A (Reference 1). The methodology included the conclusion that higher decay heat was conservative with respect to margin to overfill, based on sensitivities performed for a reference plant.

Recently, a customer identified an issue with the Reference 1 decay heat conclusion. The customer determined that lower decay heat is conservative for their plants' SGTR margin-to-overfill analyses.

Westinghouse confirmed the customer's results and determined that lower decay heat could be conservative for some plants analyzed with the WCAP- 10698 method.

In the development of the Reference I methodology, Westinghouse evaluated the dose consequences of SG overfill following a SGTR using best-estimate assumptions. This evaluation was documented in WCAP- 11002 (Reference 2). The results were acceptable on a best-estimate basis and accepted by the Nuclear Regulatory Commission (NRC) although they noted the work could not be used in a plant's licensing basis since it does not utilize conservative (FSAR Chapter 15) methodology. Although the issue described in this NSAL may result in the design basis SGTR analysis predicting SG overfill, dose consequences are bounded by WCAP- 11002. Since the dose consequences resulting from a SGTR event that includes SG overfill are acceptable on a best-estimate basis, this issue is not a substantial safety hazard pursuant to 10 CFR 21.21(a).

Additional information, if required, may be obtained from Sean Kinnas, (412) 374-4640 Originator:(s) Approved:

J. T. Crane J. A. Gresham, Manager Regulatory Compliance and Plant Licensing Regulatory Compliance and Plant Licensing S. T. Kinnas Containment and Radiological Analysis Electronically approved records are authenticated in the Electronic Document Management System NP-64

LTR-PL-14-17 NP-Attachment NSAL-07-11 Page 2 of 5 ISSUE DESCRIPTION As a consequence of the January 1982 Ginna SGTR event, the NRC questioned assumptions about the analyses presented in licensees' safety analysis. The NRC required some plants to address the issues raised by the Ginna event. In response, a WOG subgroup developed a generic SGTR methodology, WCAP-10698-P-A, to serve as the framework for plant specific analyses. Participants in the WOG subgroup could then use the WCAP's generic methodology to perform their licensing basis SGTR analysis, or contract a vendor to perforn the analysis. The WOG subgroup consisted of Shearon Harris, Byron and Braidwood, Catawba, Beaver Valley Unit 2, South Texas, Millstone Unit 3, Diablo Canyon, Ginna, Vogtle, Watts Bar, Comanche Peak and Seabrook.

WCAP-1 0698-P-A examined the competing effects of decay heat on the margin-to-SG overfill following a SGTR. Showing margin-to-ruptured SG overfill demonstrates that potential consequences of water releases do not have to be considered. For the margin-to-overfill analysis, higher decay heat yields a benefit by increasing steam releases from the ruptured SG, but results in a penalty from a longer cooldown and a conservatively delayed break flow termination. Conversely, lower decay heat yields a penalty by reducing steam releases from the ruptured SG, but results in a benefit from a shorter cooldown and earlier break flow termination. WCAP-10698-P-A concluded that higher decay heat was conservative for the SGTR margin-to-overfill analysis, and therefore 120% of the 1971 ANS decay heat curve should be used in future SGTR margin-to-overfill analyses.

Recently, it was determined by a licensee that lower decay heat was conservative for their plants' margin to overfill analyses, resulting in a loss of steam generator margin to overfill. Westinghouse confirmed the lower decay heat could be conservative for some plants' margin to overfill analyses, while others will not be adversely affected.

This issue is only known to be applicable to those plants utilizing the WCAP-10698 method for their SGTR licensing basis. Combustion Engineering (CE) plants and Westinghouse plants that do not use the WCAP- 10698 method are not impacted by this issue.

TECHNICAL EVALUATION Although SG overfill could result in increased radiological consequences, there is a reasonable basis to conclude that the applicable offsite dose guidelines would continue to be met. This is based on best-estimate dose evaluations performed as part of the generic SGTR methodology and presented to the NRC at that time as detailed below.

WCAP-1 1002 (Reference 2) presents an analysis of the consequences of SG overfill resulting from a SGTR. The analysis includes consideration of the failure of a safety valve on the ruptured SG following overfill and the resulting continued releases from the ruptured SG. The analysis modeled the recovery actions to cool down and depressurize the ruptured SG to cold shutdown. WCAP-1 1002 presents a dose calculation for the SGTR with overfill using realistic or best-estimate assumptions and concludes that the resulting doses are within the guideline values of 10 CFR 100.

The NRC reviewed WCAP- 11002 and presented the conclusions of the review together with those of the review of WCAP-10698-P-A in their evaluation dated March 30, 1987. This evaluation is included in WCAP-10698-P-A. The conclusion of the evaluation of WCAP-1 1002 includes the following statements, essentially accepting the conclusion that the dose guidelines would be met with best-estimate assumptions but not allowing its use for licensing basis analyses:

"The report concludes that the consequences of SGTR overfill are acceptable based on 'best estimate' offsite dose calculations and the waterhammer evaluation as discussed above. The NP-65

LTR-PL-14-17 NP-Attachment NSAL-07-11 Page 3 of 5 staff concludes that the analytical approach utilized in the report is tectmically sound with respect to providing "best estimate" information regarding the effects of SGTR overfill.

However, this report should not be used as a licensing basis document since it does not utilize conservative (FSAR Chapter 15) methodology."

The above demonstrates that the radiological consequences of SG overfill were evaluated during the development of the SGTR methodology. The consequences were determined to be within 10 CFR 100 guidelines using best estimate assumptions which were found acceptable to the NRC for evaluation purposes. It is reasonable to extend this discussion for plants licensed to the dose guidelines of 10 CFR 50.67 (Alternate Source Term), since the transition from thyroid to total effective dose equivalent (TEDE) doses generally results in increased margin to the limits.

Both WCAP-10698 and WCAP-1 1002 analyzed a single reference plant that was judged to apply generically. However, the recent discovery concerning the decay heat assumption in the generic methodology (WCAP-10698) raises the concern of whether the best estimate WCAP-1 1002 remains generically applicable. It is tile Westinghouse engineering judgment that WCAP- 11002 continues to be applicable and may be used for operability determinations and other non-licensing basis applications that allow for the use of best-estimate assumptions. The basis for this judgment is presented below.

The method employed in the development of WCAP-I 1002 was to force the reference plant from WCAP-10698 to overfill. The assumed operator action times were extended until SG overfill occurred and water was released through a safety valve on the ruptured SG. Once water went through the safety valve, a consequential failure of that safety valve was assumed. Two scenarios with this consequential failure were examined: the safety valve was assumed to either fail full-open or to fail partially open. Both scenarios result in an uncontrolled depressurization of the ruptured steam generator with continued break flow and atmospheric releases until cold shutdown conditions are reached.

With respect to decay heat, once overfill occurs, a higher decay heat is more limiting. After overfill occurs, higher decay heat results in a longer time to reach cold shutdown conditions and a later break flow termination, leading to more limiting dose consequences. Thus, the decay heat assumption used for the WCAP-1 1002 reference plant does not affect its continued applicability to demonstrate that dose limits are not challenged by a SGTR on a best-estimate basis.

For the SGTR thermal and hydraulic analysis for input to the radiological consequences analysis discussed in Reference 3 (modeling a failed-open PORV on the ruptured SG), there are no competing effects with respect to decay heat. Higher decay results in increased steam releases from the ruptured SG and a longer cooldown, leading to a later break flow termination. These effects are conservative for the SGTR radiological consequences calculation, and thus, lower decay heat does not need to be considered for the SGTR thermal and hydraulic analysis for input to the radiological consequences analysis performed following Reference 3.

AFFECTED PLANTS This issue is only applicable to plants utilizing the WCAP-10698 method for their SGTR margin-to-overfill analysis.

The WOG subgroup consisted of Shearon Harris, Byron and Braidwood, Catawba, Beaver Valley Unit 2, South Texas, Millstone Unit 3, Diablo Canyon, Ginna, Vogtle, Watts Bar, Comanche Peak and Seabrook.

Since these plants sponsored WCAP-10698-P-A, it is likely their plant-specific tube rupture analysis is based on the generic methodology, and they should review their assumptions with respect to decay heat.

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LTR-PL-14-17 NP-Attachment NSAL-07-11 Page 4 of 5 All other utilities who have a SGTR margin to overfill analysis, whether licensing basis or supplemental, based on the WCAP-10698 method, should review their assumptions with respect to decay heat.

Therefore, the potentially affected plants will include the WOG subgroup participants and those plants that have utilized the WCAP-10698 methodology. Based on a review of Westinghouse engineering records, the table below lists those plants where Westinghouse has performed a SGTR analysis per WCAP- 10698, and those plants in the WOG subgroup.

PLANTS THAT POTENTIALLY USE WCAP-10698 METHODOLGY WOG SGTR Subgroup Other Plants Using WCAP-10698 Beaver Valley 2 Beaver Valley I Braidwood 1&2 D. C. Cook l&2 Byron 1&2 Farley 1&2 Catawba 1&2 Indian Point 2&3 Comanche Peak 1&2 Kewaunee Diablo Canyon 1&2 Point Beach l&2 Ginna Turkey Point 3&4 Millstone 3 V.C. Summer Seabrook 1 Angra Shearon Harris 1 Beznau 1 South Texas 1&2 Ringhals 2&3 Vogtle 1&2 Watts Bar 1 CE plants and Westinghouse plants that do not use the WCAP-10698 methodology are not impacted by this issue.

SAFETY SIGNIFICANCE The issue may result in the design basis SGTR analysis predicting SG overfill for a given plant, but the dose consequences are bounded by the WOG work discussed above (Reference 2) which demonstrates there is no safety concern based on best estimate analyses.

NRC AWARENESS Westinghouse has not formally notified the NRC.

RECOMMENDED ACTIONS

1. It is the Westinghouse engineering judgment that WCAP-1 1002 continues to be applicable and may be used for operability determinations and other non-licensing basis applications that allow for the use of best-estimate assumptions

2. Westinghouse will present a project authorization to the Analysis Subcommittee of the Pressurized Water Reactor Owners Group (PWROG) in December 2007 to develop a resolution plan for this issue. In addition to decay heat, other input assumptions with known competing effects will be examined. Once that program is complete, plants should examine the effects of the revised assumptions on their SGTR margin-to-overfill analyses.

3. The proposed PWROG program will investigate input parameters with competing effects, including decay heat. If a plant does not participate in the proposed PWROG program, it is NP-67

LTR-PL-14-17 NP-Attachment NSAL-07-11 Page 5 of 5 Westinghouse's recommendation that the SGTR analysis use decay heat based on full power operation. Early into an operating cycle, the decay power rapidly (on the order of hours) approaches an equilibrium value. It is not anticipated that a SGTR would occur during this time period since the likelihood of a SGTR occurring during this short period is low. A similar justification is contained within WCAP-10698 and has been accepted by the NRC. Hence, while lower decay heat has been shown to be conservative for some plants' SGTR margin-to-overfill analyses, only decay heat based on full power operation needs to be considered.

4. As noted above, the SGTR thermal and hydraulic analysis that provides input to the radiological consequences analysis (as discussed in Reference 3) is not impacted by this issue.

REFERENCES I. WCAP-10698-P-A, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill," August 1987.

2. WCAP- 11002, "Evaluation of Steam Generator Overfill Due to a Steam Generator Tube Rupture Accident," February 1986.

3. Supplement I to WCAP-10698-P-A, "Evaluation of Offsite Radiation Doses for Steam Generator Tube Rupture Accident," March 1986.

This document is available via the Internet at rle.westinghousenuclear.com. This web site is a free service for Westinghouse Electric Company LLC (Westinghouse) customers and other electric power industry-related organizations. Access will be provided based on Westinghouse's judgment of appropriate business affiliation.

Westinghouse reserves the right, at its sole discretion, to grant or deny access to this web site. Requests for access should be made to giampora@westinghouse.com.

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