ML18094A310

% Tube Plugging Large LOCA Bash Analysis

ML18094A310
Person / Time
Site:	Salem
Issue date:	03/31/1989
From:	Akers J, Ivey J, Petzold J WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
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ML18094A309	List: ML18094A308 ML18094A310
References
WCAP-12192, NUDOCS 8904110159
Download: ML18094A310 (130)

Text

WCAP-12192 Westinghouse Class 3 SALEM UNITS 1 AND 2 10% TUBE PLUGGING LARGE BREAK LOCA BASH ANALYSIS Author: J. S. Petzold Transient and Safeguards Analysis Additional Contributors:

J. S. Ivey J. J. Akers S. E. Saunders

• March 1989 Prepared by Westinghouse Electric Corporation for use by Public Service Electric and Gas Corporation, New Jersey. Work performed under Shop Order SEDP-610A.

.<:~. --

WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division.

P.O. Box 355 Pittsburgh, Pennsylvania 15230

• *99(;4*1--10159 090381*

• PDR AIIOCK 05000272 P

PDC

. I

ABSTRACT Presented in this report are results of a SALEM Large Break Loss of Coolant Accident analysis performed using the 1981 Westinghouse Emergency Core Cooling System with BASH evaluation model.

The analysis was performed at enhanced plant conditions mutually agreed upon by Westinghouse and the customer, Public Service Electric and Gas Corporation.

ii

TABLE OF CONTENTS Section Title Page I.

INTRODUCTION 1

I I.

ANALYSIS CONDITIONS 1

I I I.

METHOD OF ANALYSIS 3

IV.

RESULTS 6

v.

CONCLUSIONS 7

VI.

REFERENCES 11 APPENDIX A FSAR MARKUPS A-1

ii

TABLE NO.

LIST OF TABLES TITLE 1

2 3

TIME SEQUENCE OF EVENTS FOR LARGE BREAK LOCA LARGE BREAK RESULTS PLANT CONDITIONS IN THE LBLOCA ANALYSIS iv 8

9 10

FIGURE NO~

la lb le 2a 2b 2c 3a 3b 3c 4a 4b 4c 4d Sa Sb Sc Sd 6a 6b 6c 6d LIST OF FIGURES TITLE Core Pressure OECLG (CD=0.8)

Core Pressure DECLG (CD=0.6)

Core Pressure DECLG (CD=0.4 and CD=0.4 Max SI)

Core Flowrate DECLG (CD=0.8)

Core Flowrate DECLG (CD*0.6)

Core Flowrate DECLG (CD=0.4 and CD=0.4 Max SI)

Accumulator Flow DECLG (CD*0.8)

Accumulator Flow DECLG (CD*0.6)

Accumulator Flow DECLG (CD*0.4 and CD*0.4 Max SI)

Pumped ECCS Flow (Reflood) DECLG (CD=0.8)

Pumped ECCS Flow (Reflood) DECLG (CD*0.6)

Pumped ECCS Flow (Reflood) DECLG (CD=0.4)

Pumped ECCS Flow {Reflood) DECLG (CD=0.4 Max SI)

Flood Rate {In/Sec) DECLG (CD*0.8)

Flood Rate {In/Sec) DECLG (CD*0.6)

Flood Rate {In/Sec) DECLG (CD=0.4)

Flood Rate {In/Sec) DECLG (CD*0.4 Max SI)

Clad Average Temperature Hot Rod DECLG (CD=0.8)

Clad Average Temperature Hot Rod DECLG (CD=0.6)

Clad Average Temperature Hot Rod DECLG {CD=0.4)

PAGE 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 Clad Average Temperature Hot Rod DECLG {CD*0.4 Max SI) 33 y

I. Introduction This document reports the results of an analysis that was performed to demonstrate that Salem Units I and II meet the requirements of Appendix K and 10CFRS0.46 for Large Break Loss-of-Coolant-Accidents (LOCA).

The analysis, performed at uprated power and with increased peaking factors, was primarily initiated by an increase in Steam Generator Tube Plugging levels for Unit 2.

Additional bounding plant conditions considered and analyzed are presented in Section II.

The Computer codes and Evaluation Model used are described in Section III.

The analysis resulted in a maximum Peak Clad Temperature (PCT) of 2091°F which meets the 10CFRS0.46 criteria of 2200°F. A discussion of the effect of transitioning from 17 x 17 Standard (STD) to VANTAGE 5 Hybrid (VSH) without Intermediate Flow Mixers (IFMs) and additional results are presented in Section IV.

Recommended Large Break LOCA FSAR changes resulting from this analysis are presented as FSAR markups in Attachment A.

II. Analysis Conditions Pertinent analysis assumptions include the following plant conditions:

1)

An enveloping 10% uniform steam generator tube plugging level was assumed.

This assumption bounds an actual plant condition of up to 10% tube plugging in any or all steam generators.

2) In order to support future fueJ upgrades and upratings, the total P.eak i ng factor was increased to 2. 40.

I

3) Both Units were bounded. This includes:

A. Analyzing the more conservative (highest) Tavg of the two plants, which is 573.4°F. This value includes documented cycle to cycle variations observed,in the past, and anticipated future variations.

B. Analysis of the more limiting fuel type in either unit.

The 17 x 17 STD fuel (fresh for Unit 2 cycle 5) was evaluated as more limiting than the VSH fuel w/o IFMs which is being introduced in upcoming cycles.

The largest contributor to this conclusion was the benefit associated with the enhanced rewetting provided by the thicker V5H grids.

The fuel analyzed has a backfill pressure of 275 psig.

4) In order to support future fuel upgrades and upratings, the hot channel enthalpy rise factor was increased to 1.60.

5) Thermal Design Flow (TDF) was assumed to be 1% degraded.

An evaluation determined that TDF would continue to be maintained at the 10% Tube Plugging level. The initial flow assumed was conservatively degraded an additional 13 to ~llow for future degradation, or increased measurement uncertainties.

6) Safety Injection (SI) performance. *some SI pump degradations relative to the previous analysis were incorporated to reflect the current.plant configuration. This includes an increase in the centrifugal charging pump miniflow to 120 gpm and a decrease in the intermediate head safety injection flow of approximately 10 gpm.

7) Diesel Generator/ SI delay.

The delay was increased from 25 seconds to 30 seconds to support future Tech Spec relaxation.

8) Low Pressurizer Pressure Reactor Trip & Low Low Pressurizer Pre.ssure SI setpoints. The analysis i_nputs were conservatively lowered to 1715 psia each to support future Tech Spec relaxation.

2

9) Bound Thimble Plug deletion.

An evaluation was performed to determine

. the more limiting of 'with' and 'without' thimble plugs.

The condition conservatively analyzed was 'with' thimble plugs which will envelope the

'without' configuration.

10) Uprated Power.

The Salem Large Break LOCA analysis was performed using 102 percent of uprated core power of 3588 MWt.

This supports an NSSS power of 3600 MWt (12 MWt pump heat).

A summary of the assumed plant conditions used in the analysis is given by Table 3.

III. Method of Analysis The analysis was performed using the Westinghouse 1981 Evaluation Model (EM) with BASH.

A full spectrum of Moody discharge coefficients were analyzed.

The Westinghouse 1981 ECCS Large Break Evaluation Model with BASH was developed to determine the RCS response to design basis large break LOCAs (see References 6-12).

The 1981 Evaluation Model with BASH is comprised of the SATAN-VI, HREFLOOD, COCO, LOCTA-IV, BART and BASH computer codes (References 3, 4, 5, 1, 15 and 16 respectively, see also Reference 17).

The 1981 EM includes changes required to incorporate the NUREG-0630 burst and blockage phenomena as identified in Reference 6.

When the BASH code (Reference 16) is used along with the 1981 EM, a slightly different version of HREFLOOD (INTERIM HREFLOOD) is required.

The purposes and structures of this code do not differ significantly from the standard 1981 EM Model HREFLOOD, and are described in Reference 15.

The Large Break accident is characterized by three distinct phases: Slowdown, Refill and Reflood.

Slowdown is the initial highly dynamic portion of the overall transient, and is terminated when a number of Reactor Coolant System (RCS) conditions are met.* The Refill transient picks up from the end of blowdown to the point in time when the lower plenum has refilled to the bottom of the active fuel elevation. Reflood then consists of the slow flooding of the active fuel region until the termination of the transient.

3

The SATAN-VI code was used to predict the system hydraulics of the blowdown phase of the transient. The ~REFLOOD CODE was used to predict the refill phase system hydraulics. The COCO code was used to evaluate the containment response for all phases of_ the transient. The ~REFLOOD code also predicts system performance during the reflood portion of the transient, but this prediction serves only to provide releases to containment that are input to COCO.

COCO and WREFLOOD run interactively to obtain the most realistic containment conditions within the conservative Appendix K modelling requirements.

The BASH code predicts system hydraulic performance for the reflood phase of the transient.

BASH initializes at system conditions taken from the results of INTERIM ~REFLOOD, COCO and LOCTA-IV at Bottom-Of-Core recovery (BOC) time.

BASH provides a more realistic thermal-hydraulic simulation of the reactor core and RCS during the reflood phase of a LOCA.

The containment backpressure used by BASH is supplied from COCO in tabular input form.

Cladding the~mal analyses were performed with the LOCBART code (a merger of LOCTA and BART), which for a BASH analysis has the FLECHT heat transfer correlation in LOCTA-IV replaced by the BART code.

LOCBART is run for the duration of the transient, using the RCS pressure, fuel rod power history, steam flow past the uncovered part of the core, and mixture height history from the SATAN-VI, ~REFLOOD, and BASH codes as input; The fuel parameters used as input for the LOCA analysis were generated using the Revised PAD Thermal Safety Model (References 13-14).

Consistent with the standard single failure criteria position of reference 11,

'Minimum Safeguards' (Min SI) assumptions are modelled for the majority of the cases co~~idered. The Min SI assumption described in reference 11 is the failure of one Low head injection pump.

As an added conservatism, the standard Westinghouse analysis additionally assumes the failure of the remaining pumps in that SI train. This is not a large conservatism since the low head SI pump is typically the most significant in providing flow to the RCS for a large break LOCA~ In particular, this analysis assumed, for the Min SI cases, the failure of one of each of the three SI pumps comprising one train. Inherent in the Min SI modelling are various other assumptions made to minimize the delivered SI flow, such a~ conservativ~ly high system resistances in the SI lines.

4

For the Min SI assumption, analysis of three Moody break discharge coefficients

{Co) were performed: Co=0.4, Co=0.6, and c0=0.8. This break spectrum is consistent with reference 9, Appendix 0.5 which establishes a c0=0.4 as the minimum value based on physical limitations.

Following the analysis of those three cases, a fourth case was analyzed called

'Maximum Safeguards' (Max SI).

Max SI assumptions include no failures of any SI pumps and other related assumptions such as conservatively low system resistances in the SI lines. The analysis of the Max SI case is required for all Westinghouse 4-loop non-Upper Head Injection (UHI), non-burst node limited plants. The Max SI case is a rerun of the limiting PCT break determined from the results of the three Min SI cases.

In this analysis, the Max SI case was a rerun of all aspects of the evaluation model described above, except for SATAN.

SATAN was conservatively not rerun because the Max SI modelling would be a small benefit for the blowdown portion of the transient.

The NRC Safety Evaluation Report (SER) on the 1981 EM with BASH (reference 16, prior to addenda) specificalJy imposes three restrictions on its subsequent plant specific use.

The conformance to these three restrictions is discussed next.

Restriction #1. The BASH EM may not be applied to the analysis of Upper Plenum Injection (UPI) or UHI plants. The Salem plants are neither UPI or UHi so conformance to restriction #1 is satisfied.

Restriction #2.

The analysis must continue to include both the Minimum and Maximum Safeguards cases in order to confirm the limiting break case.

Conformance to this restriction has been met by the specific analysis of the four cases described previously.

Restriction #3.

Sensitivity studiesJ confirming the limiting nature of the peaked-to-the-center chopped cosine power shape, must be performed for the first application to~ard each of the. 2-loop, 3-loop or 4-loop plant type.

For 4-loop plants, like Salem, this confirmation is provided by Addendum 1 of reference 16, therefore conformance to this restriction has been satisfied.

s

IV.

Results Of the three Min SI break cases and one Max SI case performed, the Min SI Co=0.4 break case proved to be the limiting (highest PCT) case with a PCT of 2091°F, compared with PCTs of 2071°F, 1911°F and 2011°F for the Co=0.6, c0=0.8 and Co=0.4 Max SI cases, respectively.

The highest Local metal/water reaction was 7.033, for the Co=0.6 case.

These transients were considered to be terminated when the hot rod clad average temperature "turned around" (i.e. - hot rod clad average temperature began to decline) indicating that the peak clad temperature had been reached.

Table 1 shows the time sequence of events for the Large Break LOCA transients considered. Table 2 provides a brief summary of the important results of the LOCA analysis for ~ach case.

Figures la-le and 2a-2c show important core characteristics during the blowdown phase of the*transient {Core Pressure and Core Flow versus Time, respectively). Figures 3a-3c and 4a-4d illustrate the flow of ECCS water into the RCS (Accumulator Flow during blowdown, and Pumped SI Flow during Reflood, versus time, respectively). The flooding rates during the reflood portion of the transient are given in Figures Sa-Sd. These figures show the smoothed flooding rate required by the BASH SER and described in the BASH WCAP, reference 17.

Clad Average Temperatures as a function. of time, for the peak clad temperature and clad burst locations, are given in Figures 6a-6d.

When a core.contains more than one type of fuel, it must be determined if the transition core can have a greater calculated PCT than a complete core of either fuel design.

For a given peaking factor, the only mechanism available to cause a transition core to have a higher PCT than a full core of either fuel is the possibility of flow redistribution due to fuel assembly hydraulic mismatch.

The current EM and codes that comprise it do not specifically model such geometry. Typically one fuel assembly type will be flow starved and result in hi~her PCTs than the analysis of that fuel type predicts. However, test results for the VSH assembly and grids have shown that hydraulically 17 x 17 VSH w/o IFMs and 17 x 17 STD are identical. Therefore there is no cross-flow penalty to consider.

6

For the VSH transition cores that will be forthcoming for both units, an NRC SER and subsequent SER clarification (references 18 and 19) to th~ VSH report (reference 20) have been obtained from the NRC.

The summary conclusion of references 18, 19 and 20 is that, for a 1981 EM with BASH, the 17 x 17 STD to VSH w/o IFMs transition core is assessed no Large Break LOCA PCT penalty.

Therefore, the quantitative results of this analysis will continue to apply for all transition core instances.

V.

Conclusions For breaks up to and including the double ended severance of a reactor coolant pipe,* the Emergency Core Cooling System will meet the Acceptance Criteria as presented in IOCFRS0.46.

That is:

I. The calculated peak clad temperature does not exceed 2200°F based on a large break LOCA total peaking factor Qf 2.40 and a hot channel enthalpy rise factor of 1.60, and up to 103 uniform Steam Generator tube plugging level.

2.

The amount of fuel element cladding that reacts chemically with water or steam does.not exceed I percent of the total amount of Zircaloy in the reactor.

3. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling.

4.

The local cladding oxidation limits of 17% are not exceeded during or after quenching.

5.

The core temperature is reduced and the decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core.

There will be no transition core penalty for cycles with mixed STD and VSH w/o IFMs fuel.

7

TABLE 1 TIME SEQUENCE OF EVENTS FOR LARGE BREAK LOCA Time

~

1.

(Co=.4)

Start 0.0 Reactor trip signal

.95 Safety Injection signal 1.81 Accumulator injection 19.4 Pump injection 31.81 End of blowdown 39.3 End of bypass 39.3 Bottom of core recovery 55.5 Accumulators empty 63.2

2.

(Co=.4 Max SI)

Start 0.0 Reactor trip signal

.95 Safety Injection signal 1.81 Accumulator injection 19.4 Pump injection 31.81 End of bypass 39.3 End of blowdown 39.3 Bottom of core recovery 54.5 Accumulators empty 70.3'

3.

(Co=.6)

Start 0.0 Reactor trip signal

.93 Safety Injection signal 1.49 Accumulator injection 13.9 End of blowdown 30.2 End of bypass 30.2 Pump injection 31.49 Bottom of core recovery 44.7 Accumulators empty 59.5

4.

(C0=.8)

Start 0.0 Reactor trip signal

.91 Safety Injection signal 1.32 Accumulator injection 11.8 End of blowdown 26.4 End of bypass 26.4 Pump injection 31.32 Bottom of core recovery 40.3 Accumulators empty 54.3 8

TABLE 2 LARGE BREAK RESULTS DECL DECL DECL DECL Co=0.4 CR=0.4 M X SI Co=0.6 Co=0.8 Results Peak Clad Temperature! (°F) 2091 2017 2071 1911 Peak Clad Temp. Location (ft) 8.5 8.5 8.5 8.5 Local Zr/H2o Reaction (max) %

6.96 5.34 7.03 4.63 Local Zr/H2o Location (ft) 8.5 8.5 8.5 7.25 Total Zr/H2o Reaction (%)

<0.3

<0.3

<0.3

<0.3 Hot Rod Burst Time (sec) 57~44 57.60 41.48 86.24*

Hot Rod Burst Location (ft) 6.25 6.25 6.00 7.25 1 Results continue to apply for STD to VSH transition cores 9

TABLE 3 PLANT CONDITIONS IN THE LBLOCA ANALYSIS Reactor Power 102% of Peak Linear Power (Hottest Rod) 102 % of Peaking factor (at above power)

Hot Channel Enthalpy Rise Factor Accumulator Water Volume per accumulator Accumulator Initial Pressure SG Tube plugging levell Thermal Design Flow (13 degraded @ 10% plugging)

SI delay time Lo Pressurizer Pressure Reactor Trip Lo Lo Pressurizer Pressure SI Tavg 3588 MWt 13.75 Kw/ft 2.40 1.60 850 ft3 610 psia 103 86400 gpm per loop 30 sec 1715 psia 1715 psi a 573.4°F

• Fuel Analyzed Fuels Evaluated 17 X 17 Standard Bounding 17 x 17 Standard & VSH w/o IFMs 1 Up to 10% Steam Generator Tube Plugging in any or all steam generators.

10 I

VI.

REFERENCES

1.

Bordelon, F. M., et al., "LOCTA-IV Program: Loss-of-Coolant Transient Analysis," WCAP-8301 (Proprietary Version), WCAP-8305 (Non-Proprietary Version), June 1974.

2.

"Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors: 10CFR 50.46 and Appendix K of 10CFR 50.46," Federal Register, Vol. 39,No. 3, January 4, 1974.

3.

Bordelon, F. M., et al., SATAN-VI Program: "Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant," WCAP-8302 (Proprietary Version), WCAP-8306 (Non-Proprietary Version), June 1974.

4.

Kelly, R. D., et al., "Calculational Model for Core Reflooding after a Loss-of-Coolant Accident (Wreflood Code)," WCAP-8170 (Proprietary Version), WCAP-8171 (Non-Proprietary Version), June 1974.

5.

Bordelon, F. M., and E. T. Murphy, "Containment Pressure Analysis Code (COCO)," WCAP-8327 (Proprietary Version),

WCAP-8326 (Non-Proprietary Version), June 1974.

6.

Eicheldinger, C.,

"Westin~house ECCS Evaluation Model, 1981 Version," WCAP-9220-P-A (Proprietary Version), WCAP-9221-A (Non-Proprietary Version), Rev. 1, 198L

7.

Bordelon, F. M., H. W. Massie, and T. A. Zordan, "Westinghouse ECCS Evaluation Model-Summary," WCAP-8339, July 1974.

8~

Bordelon, F. M., et al., "The Westinghouse ECCS Evaluation Model: Supplementary Information," WCAP-8471 (Proprietary Version, WCAP-8472 (Non-Proprietary Version), January 1975.

9.

Salvatori, R., "Westinghouse ECCS - Plant Sensitivity Studies,~

WCAP-8340 (Proprietary Version), WCAP-8356 (Non-Proprietary Version), July 1974.

• 10.

"Westinghouse ECCS Evaluation Model Sensitivity Studies,"

WCAP-8341 (Proprietary Version), WCAP-8342 (Non-Proprietary Version), 1974.

11.

Kelly, R. D., C. M. Thompson, et al4, "Westinghouse Emergency Core Cooling System Evaluation Model for Analyzing Large LOCAs During Operation With One Loop Out of Service for Plants Without Loop Isolation Valves," WCAP-9166, February 1978.

11

12..

Eicheldinger, C., "Westinghouse ECCS Evaluation Model, February 1978 Version, WCAP-9220 (Proprietary Version), WCAP-9221 (Non-Proprietary Version; February 1978.

13.

Letter from Cecil 0. Thomas (NRC) to E.P. Rahe, Jr.

(Westinghouse), "Acceptance for Referencing of Licensing of Topical Report WCAP-8720, Addendum 2, 'Revised PAD Code Thermal Safety Model'," Dated December 9, 1983.

14.

"Westinghouse Revised PAD Code Thermal Safety Model," WCAP-8720, Addendum 2 (Proprietary), and WCAP-8785 (Non-Proprietary).

15.

Young, M., et al., "BART-IA: A Computer Code for the Best

. Estimate Analyzed Reflood Transients," WCAP-9561-P-A, 1984 (Proprietary).

16.

Kabadi, J. N., et al., "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code", WCAP-10266 Rev. 2 with Addenda, 1986 Westinghouse Proprietary).

17.

Chiou, J. S., et al.,"Models fo~ PWR Reflood Calculations Using

.the BART Code," WCAP-10062

18.

Thadani, A. C. (USNRC) TOR. A. Wiesemann, (Westinghouse),

Letter dated Nov. I, 1988,

Subject:

'Acceptance for Referencing of Topical Report WCAP-10444-P-A Addendum 2'

19.

Thadani, A. C. (USNRC) to Wiesemann, R. A. (Westinghouse) letter dated 1/5/89,

Subject:

Clarifications on the Safety Evaluation of the Topical Report WCAP-10444-P-A Addendum 2

20.

"Vantage SH Fuel Assembly," WCAP-10444-P-A Addendum 2 (Proprietary), April, 1988; including Supplemental Information, Wiesemann, R. A. & Johnson, W. J.to NRC dated July 29, 1988.

12

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Q z Q u LI.I "' -

LI.I

E -

0 0 -

(l3S/NI) 31Yll 0001~

Figure Sa Flood Rate (In/Sec) DECLG (Co=0.8) 26 I

C)

ID

&I)

Q

~

Q y

l.M

&I) l.M

~

C)

~

C)

(~3S/NI) 3!YH 0001~

Figure Sb Flood Rate (In/Sec) DECLG (Co=0.6) 27

.-~-.~~.-~-r-~~~~--r~~oor-~-,.~~~~--.~~~~

~

t--~-+-~~t-~-r~~+-~-+~~+-~--+~~-+-~~-+~~8

~

C>

ll)

C>

C>....

L_~-L~~L-~-L~~L-~--L.~~.L...i==::::+/-====:t:::=---1~~~C>

C>

(~3S/NI) 31Vlt 0001~

Figure Sc Flood Rate (In/Sec) DECLG (C0~0.4}

28 C>

Cl)

~

z:

C>

~

~

Cl)

~

£....

0 -

I I

I I

~*

I I i co N

(~3S/NI) 3!Yll 0001~

Figure Sd Flood Rate (In/Sec) DECLG (Co=0.4 Max SI) 29 I

I i

0 in N

0 0

N 0 in -

0 0 -

0 in 0

0

&I)

Q z 0

c..>

Lr.I

&I) -

Lr.I s::: -

~

.-~-.~~r-~-.~~T'"""""""~~~~r-~-.-....;_~.....-~--~----8 C"'>

0 0

It)

N

.. I ><

0

z.

~)

i...J

...J 0

0 0 N

I 11 0

0 It)

/

I 0

0 0....

(Jo) 001 lOH 31n1Yl3dW3l 39Yl3AY OY1~

Figure 6a*

Clad Average Temperature Hot Rod DECLG {Co=0.8) 30 0

V)

Cl z

0 u V) -

z: -

I-

.--~~..--~~......-~~......-~~..-~~.....-~~..-~~..-~~..-~~...-~~ g C'f)

V) c:::......

> (,.)

al Q,.

c c

.. IE Cl z:

L&J C..!J L&J

...J L...~--1~~_.l.~~...L~_::::::::=:::=~~!!!:::=::l:~~~~~..L~~..L_~--10 c g N

c c.,, -

c c c -

C.:10 > 001.lOH 3Hn.lYll3dN3.l 39Yll3AY m:>

Figure 6b c c.,,

Clad Average Temperature Hot Rod DECLG (Co=0.6) 31 c

Cl)

Cl z

Q u LI.I Cl) -

LI.I E -

r-~---:,-~-,,-~-,,-~--.~~---.~~--.,--~~.--~~....-~~....-~--8

.. I ><

c z

. LI.I c.:s LI.I

....I

(")

L__J~_J~_j~_j:~~~=:l:L-__L~__l~_J~_Jo 0

0 "'

N 0

0 0

N 0

0 "' -

g C) -

(~o) 001 lOH 31fl.1Y113dN3l 39Vl3AY OY1~

Figure 6c C) 0 "'

Clad Average Temperature Hot Rod DECLG (Co=0.4) 32 0

V) c z 0

L&J V) -

L&J E -

I-V) a:: I-

=>U cc Q..

.. IE

/

c z u.J c.,:,

u.J

...J

(.:10 ) 001 lOH 31rtLYl3dW31 39Yl31\\Y OY1:>

Figure 6d Clad Average Temperature Hot Rod DECLG (Co=0.4 Max SI) 33 Cl)

Q z Cl

(,,)

Cl) -

E: -

APPENDIX A FSAR MARKUPS A-1

15.4 CONDITION IV - LIMITING FAULTS Condition IV occurrences are faults which are not expected to take place, but are postulated because their consequences would include the potential or the release of significant amounts of radioactive material.

They are the most drastic occurrences which must be designed against and represent limiting design cases.

Condition IV faults are n6t to cause a fission product release to the environment resulting in an undue risk to public health and safety in excess of guideline values

• of 10CFRlOO.

A single Condition IV Fault is not to cause a consequential loss of required functions of systems needed to cope with the fault including those of the Emergency Core Cooling System (ECCS) and the containment.

The following. faults have been classified in this category:

1.

Major rupture of pipes containing reactor coolant up to and including double-ended rupture of the largest pipe in the Reactor Coolant System (RCS) loss-of-coolant accident (LOCA)

2.

Major se.condary system pipe ruptures

3.

Steam ~enerator tube rupture

4.

Single reactor coolant pump locked rotor

5.

Fuel handling accident

6.

Rupture of a control rod drive mechanism (CRDM) housing (rod cluster control assembly ejection).

The ti~e sequence of events during applicable Condition IV events is shown in Table 15.4-1.

15.4-1 SGS-UFSAR A-2 Revision 6 February 15, 1987

15.4.1 Major Reactor Coolant System Pipe Ruptures (Loss-of-Coolant Accident)

The analysis specified by 10CFRS0.46 (1) "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Power Reactors," is presented in this section.

The results of the LOCA analyses are shown in Table 15. 4-2 and show compliance with the Acceptance Criteria.

The analytical techniques used are in complian~

Appendix K of 10CFRSO, and are described in Reference~

results for the smail break LOCA are presented in Section 15.3.1'------------~

and are in conformance with 10CFR50.46 and Appendix K of 10CFR50.

The boundary considered for LOCAs as related to connecting piping is defined in Section 3.6.

Should a major break occur, depressurization of the RCS results in a pressure decrease in the pressurizer.

Reactor trip signal occurs when the pressurizer low pressure trip setpoint is reached.

A Safety Injection System (SIS) signal is actuated when the appropriate setpoint. is reached.

These counter measures will limit the consequences of the accident in two ways:

1.

Reactor trip and borated water injection complement void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat.

2.

Injection of borated water provides heat transfer from the core and prevents excessive clad temperatures.

At the beginning of the blowdown phase, the entire RCS contains subcooled liquid which transfers heat from the core by forced convection with some fully developed ~ucleate boiling. After the break develops, the time to departure from nucleate boiling (DNB) is calculated, consistent with Appendix K of 10CFR50.

Thereafter the core heat transfer is based on local conditions with transition boiling and forced convection to steam as the major 15.4-2 SGS-UFSAR A-3 Revision 6

• February 15, 1987

beat transfer mechanisms.

During the refill period rod-to-rod radiation is the only beat transfer mechanism.

~/O When the RCS pressure falls below~sia, the accumulators begin to inject borated wa~er. The conservative assumption is made that i

accumulator water bypasses the core and goes out through the break until the termination of bypass.

This conservatism is again consistent with Appendix K of 10CFRSO.

15.4.1.1 Thermal Analysis

• 15.4.1.1.1 Westinghouse Performance Criteria for Emergency Core Cooling System The reactor is designed to withstand thermal effects caused by a LOCA including the double-ended severance of the largest RCS pipe.

The reactor core and internals together with the ECCS are designed so that the reactor can be safely shut down and the essential heat transfer geometry of *the core preserved following the accident.

The ECCS, even when operating during the injection mode with the most severe single active failure, is designed to meet the Acceptance Criteria (1).

15. 4.1. 1. 2 Method of Thermal Analysis the The through 6. The SGS-UFSAR code (6) in Table various aspec This document des among the compute the cceptance Criter1.

15.4-3 A-4 deta1 used ECCS e containment ressure are Revision 6 February 15, 1987 I

• presented here was performed with Evaluation Hodel is 10, 11, 12, 13.

The analysis was performed using the following assumptions for the col~ leg accumulator and core:

1.

Cold Leg Accumulator:

Water volume - nominal Temperature - minimum r,.,.,.facf loop Line resistance -

nominal calculated (average of the three highest line resistances) crDr)

2.

Initial Core - Thermal Design FlowhConditions Flowrate -*nrinimtn.. '1<f°lo 4T})f af /c% ft.~<! pfw~~;,l\\J't Temperature <:isl. f I d

• um hi;j~u-r Mer::i4qN!'.Y piQ,vf 74vcn Pressure - maximum The net effect of the above parameters is conservative so as to maximize the stored energy in the RCS

' 'AP' The analysis was performed using the conservative assumption that the fluid temperature in the upper head of the reactor vessel is equal to the reactor vessel outlet temperature.

The effect of upper head temperature on ECCS performance is discussed in References 14 and 15.

The containment back pressure is calculated using the methods and assumptions described in Reference 2, Appendix A.

Input parameters used in the analysis are provided in Table 15. 4-3.

15.4-4 SGS-UFSAR A-5 Revision 6 February 15, 1987 l

\\

\\

vpra+eL The containment initial conditions of 90°F and 14. 7 psia are represent;:tively low values anticipated during normal full power operation.

The initial relative humidity was conservatively assessed to be 98.9 percent.

The condensing heat transfer coefficient used for heat transfer to the ~l containment...., a..\\I structures for the limiting break is given on Figure 15. 4-1.

The containment sump temperature does not affect the analysis because the maximum peak cladding temperature occurs prior to initiation of the recirculation mode for Containment Spray System.

The mass and energy releases used in the containment back pressure calculation for the limiting break are presented on Figures 15.4-2 and 15.4-3 and in Table 15.4-4.

15.4.1.1.3 Results Table 15.4-2 presents the peak clad temperature, hot spot m~tals'

-tile ca.se~

reaction and other lt~y results for )' range of break ~* The ca.s~

range of break~ was determined to include the limiting case for peak clad temperature from sensitivity studies reported in References 16 and 17.

The SATAN VI analysi5r of the LOCA is performed at 102 percent of o~ 3.5"98 MWt.

core li&eas~power.~ The peak linear power and core power used in the analyses are given in Table 15.4-2.

The equivalent eere

_.pnameter at Che license application power level ts also-ttown-in Ti_ble 15 4-2 Siace tber&.--ia.margin..bet.ween the value *Of the peak

~iaear power densit1 used jn thjs analysis and the value expected -

in epeHtien, a looer peak dad te11pe!'att1re weuld be o~tained by __

. usjng tbe peak linear p 0 we;: deaaity--4HGpec:.ted.. du.ri-Ag-~Htion.

For the results discussed below, the hot spot is defined to be the location of maximum peak clad temperature.

This location is given in Table 15.4-2 for each break size analyzed.

15.4-5 SGS-UFSAR A-6 Revision 6 February 15, 1987 r

1-

Figure 15.4-1 through 15.4-47 present the transients for the principal parameters for the break sizes analyzed.

The following items are noted:

~igure 15.4-1 f, J v"< I r-y - Z..

Figure 15.4-3 Figures 15.4-4 through 15.4-12 Figures 15.4-13 through 15. 4-2/0.

SGS-UFSAR This figure provides the containment wall condensing heat transfer coefficient)); m, 0

-A11J~ br ~q;_,

T'hi.s

..fi~w,..e s-\\..ow.i

~l..e hti!d. l"\\"4d.f/uv-1 teleaA-e.!h

+he fo.v-f.G;NMe'r.i't d<At";rJ}

b/ow """""'*

This figure shows the break energy released to the containment during blowdown.

The following quantities are presented at the clad burst location and at the hot spot (location of maximum clad temperature), both on the hottest fuel rod (hot rod):

tf

1.

Fluid quality (Figures 15. 4*/ to

2.

Mass velocity (Figures 15.4-7 to 3~

Heat transfer coefficient (Figures 15.4-10 to 15.4-12~ ~ 15, i-t-1 ;>.A Tae heat tna&far eeeffieieat shewn- -is-calculate~ 8y t&e LQGTA IV ode.

The Jvf\\oJiblow{a~

system pressure {\\ (Figures 15. 4-13 to

15. 4-15) shown is the calculated pressure in the core.

The flow rate out the break (Figures 15.4-2, 15.4-16, 15.4-17) is plotted as the sum of both ends for the guillotine break cases.

This core pressure drop (Figures 15.4-18 to 15.4-20) shown is from the lower plenum, near the core, to the upper plenWil at the core outlet.

15.4-6 A-7 Revision 6 February 15, 1987

Figures 15.4-21 through 15.4-29 These figures show the hot spot clad temperature transient and the clad temperature transient at the burst locatio (Figures 15.4-21 to 15.4-2 The fluid..__--~ 1S-.'f-*

(Figures 15.4-24 to 15.4-26~

,temperature shown is also for the hot spot and burst location.

The core flow (top and bottom) is also shown A

(Figures 15.4-27 to 15.4-29).

I

"{- 'J'J..

~ r5, :J *

.J Jo1.<1tJC()M ~'("' M

• , X.fv.....r<

~~

Figures 15.4-30 These figures present the coreuefleed threttga 1~

l

~

c through 15. 4-,)5 ___ ~_!!.B!!.~~-- _ ~"-~ ~-~

~-- +~ ~ r_~ J}o~~ '!! ::__-" ~. ('_~ t Fi,..,,...,,1fy-1J fl.io... 3~1~i.;-J~

-rt,~C".

-A6~t°"'U) f~~-t..µ,..~

cor-e..+fc.,.,t;..,'J--

Figures 15. 4-36 These figures show the ECCS flow for all cases fctt~ i" hrcc,- 3 "-

( iS: 'f..-'1I A __

t;;,;;h;.;;r..;.o~ug~h;;;...;;~.;,,,_;~

t \\... ~ re.{:-1 c. c *-'

As described

earlier, the

+- -r"'1.-.J J i ~ N t.

analyzed.

accumulator delivery discarded until the during blowdown end of bypass is is calculated.

Accumulator flow,

however, is established in refill-reflood calculations.

The accumulator flow assumed is the sum of that injected in the intact cold legs.

Figures 15. 4-42 These figures show the containment pressure through 15.4-44i, transient.

L~1~-t-f'7'1\\

Figures 15. 4-45 These figures show the core power transient.

through 15.4-47 In addition to the above, Tables 15. 4-4 and 15. 4-5 present the

~

mass and energy,releases to the containment and the broken

/p.J,.,pe~ ~"-f~f.7 i""iec.ft*g" !>Pill loop accumulato71 mass and energy flowrate to the containment for the limiting break, respectively.

The clad temperature analysis is based on a total peaking factor

"). *~

""d--1i IV" (A"'

03 of r'* TheA hot spot metal water reaction reached is °'6'":T 7.

percent, which is well below the embrittlement limit of 17 percent, as required by 10CFRS0.46.

In addition, the total core 15.4-7 SGS-UFSAR A-8 Revision 6 February 15, 1987 l

metal water reaction is less than 0.3 percent for all breaks as compared with the 1 percent criterion of 10CFRS0.46 15.4.1.1.4 Conclusions - Thermal Analysis For breaks up to and including the double-ended severance of a reactor coolant pipe, the ECCS will meet the Acceptance Criteria as presented in 10CFRS0.46.

That is:

1.

The calculated peak fuel clad temperature provides margin to the requirement of 2200°F even with containment parameters as conservative as thos~

presented in Table 15.4-3.

2.

The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.

3.

The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling.

The cladding oxidation limits of 17 percent are not exceeded during or after quenching.

4.

The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core.

The time sequence of events for all breaks analyzed is shown in Table 15.4-1.

15.4-8 SGS-UFSAR A-9 Revision 6 February 15, 1987 I --

JN-~

15.4.1.2 Fuel Rod Model Discussion dated November 9, 1979 (18), to operators regarding fuel rod models used in LOCA This letter describes a meeting called by I

the NRC on November 1, 1979, to present draft report NUREG-0630, welling and Rupture Models for LOCA Analysis" (19).

At representatives of Nuclear Steam Supply System (NSSS) vendors and el suppliers were asked to show how plants licensed, using their L CA/ECCS evaluation model, continued to conform to 10CFRS0.46 fuel rod models presented in Reference 19.

We tinghouse representatives presented information on the fuel rod mo els used in analyses for plants licensed with the Westinghouse ECC evaluation model and discussed the potential impact of fuel rod m el changes on results of those analyses.

That information was rmally documented in Reference 20, and formed the basis he Westinghouse conclusion that the information presented in erence 20 did not constitute a safety problem for Westinghouse pla s and that all plants conformed with NRC regulations.

requested that operators of* light water reactors provide, within 60 days, information which would enable the staff to determine, *n light of the fuel rod model concerns, whether or not further ion was necessary.

This section provides information on Units 1 and 2 required to respond however, that a significant amount of exchange between Westinghouse and the Reference 20 was prepared and the compliance with 10CFRSO has been Salem such a request.

Note, outline of the significant events that has transpired since for demonstrating following is an occurred since November 2, 1979, and provides an update on As a

result of compiling information for Westinghouse recognized a potential discrepancy in of fuel rod burst for cases having clad heatup rate rupture) significantly lower than 25°F/s.

15.4-9 SGS-UFSAR A-10 Revision February 20, 987

\\ '

,}jy

~ to the NRC staff, by telephone, on November 9, 1979, and although independent of the NRC fuel rod model concern, the combined effect this issue and the effect of the NRC fuel rod models had to be Details of the work done on this issue were presented to on Novemb*er 13, 1979, and documented in Reference 21.

included development of a procedure to determine the clad

.rate prior to burst and re-evaluation of operating rod burst model.

As part of this re-evaluation, the Westinghouse osition on NUREG-0630 was reviewed and it was still constitute a

presented in Reference 1 did not problem for plants licensed with the Westinghouse ECCS On December 6, 1979, information thus discussion the NRC C and Westinghouse personnel discussed the At the conclusion of that quested Westinghouse to provide further detail on the potential fuel rod models used in of modifications to each of the LOCA analysis and to outline analytical model improvements :i:

the potential benefit associate additional information resulted and documented parts of the analysis and those improvements.

This various LOCA analysis Another meeting was held in Bethesda on December 20, 1979, where.

NRC and Westinghouse personnel establis ed:

1) The currently accepted procedure for assessing the tial impact on LOCA analysis results of using the fuel Reference 1, and 2) acceptable benefits model improvements that would for the interim until difference between concern are resolved.

g from analytical plant operation models of Part of the Westinghouse effort provided to resolution of these LOCA fuel rod model differences is in the in Reference 23, which contains Westinghouse comments o NUREG-0630.

As stated in that letter, Westinghouse believ 15.4-10 SGS-UFSAR A-11 Revision 6 February 15, I

\\

I

~ent Westinghouse models compliance with Appendix K.

to be conservative and to be in

1.

SGS-UFSAR Evaluation of the potential impact of using fuel rod models presented in draft NUREG-0630 on the LOCA analysis for Salem Units 1 and 2.

This evaluation is based on the limiting break LOCA analysis identified as follows:

Double-Ended Cold Leg Guillotine 0.8 Westing ouse ECCS Evaluation Model Version February 1978 2.32 Hot Rod Ma im~ Temperature Calculated For The Burst Region Of Th Clad:

2130°F = PCTB Elevation:

Hot Rod Maximum Te erature Calculated For A Non-ruptured Region The Clad - 2003°F= PCTNN Elevation: 7.5 feet Clad Strain During 3.3 Percent; Maximum 6.2 Percent Maximum temperature reflood rate is greater than reflood heat transfer is calculation.

Average Hot Assembly Rod 15.4-11 A-12 at This

_Elevation:

This Elevation:

ch per second and the FLECHT 6.0 feet

SGS-UFSAR Hot Assembly Blockage Calculated:

45 Percent

a.

Burst Node The maximum potential impact on the rupture clad node is expressed in Reference 22 in terms of the change in the peaking factor limit (FQ) required to maintain a peak clad temperature (PCT) of 2200°F and in terms of a change in PCT at a constant FQ.

the clad-water reaction rate increases at temperatures (such as ~CT due to changes in indicated here may not apply over large

ranges, but a

in FQ which causes the PCT to remain i the neighborhood of 2200°F justifies use of this For the Use of the NRC bur reduction of 0.015 The maximum estimated strain model Therefore, the maximum burst node.is:

model could require an F Q of using the NRC reduction of 0. 03.

for the hot rod

~CT 1 = (0.015 + 0.03) (150°F/

= 675°F Margin to the 2200°F limit is:

15.4-12 A-13 Revision 6 February 15,

b.

{

SGS-UFSAR The FQ reduction required to maintain the 2200°F clad temperature limit is

(

0.01 MO

= (M>CT1 - M>CT2)

)

= (675 - 70) (.Q.Jll.)

150 150°F

= 0.04 (but not less than zero).

The temperature calculated for a non-burst section of elevation above

. mid-plane during the core reflood phase of the LO transient.

The potential impact on that.maximum fuel

• rod models aspects of change in*

temperature of using the NRC e estimated by examining two The first aspect is the conductance resulting from a non-burst maximum clad temperature node clad strain all along the fuel burst occurs and model can change the time at calculated.

Three sets of LOCA Note that after clad were studied to establish an acceptable nsitivity to apply generically in this evaluati The possible PCT increase resulting from strain (in the hot rod) is +20°F decrease in strain at

• locations. Since the clad strain the RCS blowdown phase of the accident is changed by the use of HRC fuel rod models, maximum decrease in clad strain that must considered here is the difference between 15.4-13 A-14 Revision 6 March 15, 1987 be the I

\\

I

\\ I

\\

\\ \\

'1J

SGS-UFSAR "maximum clad strain" and the "clad strain during blowdown" indicated above.

Therefore:

20°F MCT3 = (

) *(Max Strain - Blowdown Strain) 0.01 strain 20

= (~) (0.062 - 0.33) 0.01 S8 The PCT greatest of the analysis that can increase blockage calculated.

Since the blockage indicated by the NRC blockage model 7S percent, increase can be est ated by current level of (indicated above) is the maximum PCT assuming that the -

in the analysis 7S percent and then applying an appropriate in Reference 22.

. itivity formula shown Therefore, MCT4 = 1.25°F (SO - Percent

+ 2.36°F (75-50)

= 1.2S (SO - 45) + 2.36 (75-50)

= 65°F If PCTN occurs when the core reflood rate greater than 1.0 in. per sec ~CT4 = 0.

The total 15.4-14 A-15 Revision 6 February 15, 1987

2.

potential PCT increase for the non-burst node is then Margin to the 2200°F limit is required to maintain this 2200°F limit is (from HS-TMA-2174) (22) 0.01 t:iF

(~CT5 - ~CT6

) (

Q )

10°F~CT 0.14 but not less than zero.

The peaking tion required to maintain the 2200°F clad temperature.

the greater of MQB and MQH' or; MQPEN

= 0.04 The effect on LOCA analysis analytical and modeling techni ues approved for use in the upper h ad analyses) in

• the RCS computer code) has been quantified has recently been submitted to Recognizing that review of that complete and that the of using improved (which are currently injection plant LOCA calculation (SATAN an analysis which for review.

is not yet model improvements can change for other the HRC has established a credit that is this interim period to help offset penal ties resulting from application of the HRC fuel rod models. Tha

'. for two-, three-and four-loop plants is an incr SGS-UFSAR 15.4-15 A-16 Revision 6 February 15,

3.

the LOCA peaking factor limit of 0.12, 0. 15 and 0. 20 1 respectively.

eaking factor limit adjustment required to justify plant ope tion for this interim period is determined as the te

~Q credit identified in Section 2

above, minus the

~QPENALTY calculated in Section 1 above (but not greate than zero)

FQ Adjustment = 0.2 ~ 0.04

~-----------

.....__.---...__..--.,.- ~

15.4.1.3 Environmental Consequences of a Loss-of-Coolant Accident 15.4.1.3.1 Sources Initial Release Fractions Offsite doses are analyzed for two cases.

In the first case, the design basis accident, it is assumed that the entire inventory of volatile fission products contained in the pellet-c.ladding gap is released during the time the core is being flooded by the ECCS.

Of this gap inventory, 50 percent of the halogens and 100 percent of the noble gases are considered to be released to the containment atmosphere (a plate out and condensation factor of 0.5 is assumed).

It is also assumed that 10 percent of the airborne halogens available for leakage from the containment are in organic forms and are not subject to removal by sprays.

The remaining 90 percent of the airborne inventory is considered to be in the elemental form and subject to removal by containmeqt sprays.

The bas.ic inventories used in the analysis of this case are identified in Section 11.1.

In the second

case, the offsite doses resulting from a

hypothetical accident with larger activity releases are analyzed.

Activity releases of these magnitudes have a considerably lower probability of occurrence than those associated with the design basis accident.

SGS-UFSAR For the analysis of the hypothetical it is 15.4-16 A-17 Revision 6 February 15, 1987

. I

l This analysis is *based on the assumption that the containment pressure signal is the one signal which will actuate the safeguards.

In reality there are other signals which could cause the initiation of safeguards.

15.4.8.4.2 Loss of Normal Containment Cooling A loss of offsite power incident is not sufficient to cause a loss of normal containment cooling.* The normal containment cooling can only be interrupted by a coincident loss of offsite power and plant trip.

In the event of a loss of normal containment cooling due to the loss of offsite power and plant trip, a small temperature rise will occur in the containment until the high fan cooler inlet temperature signal setpoint is reacned.

When this signal, is received, the operator will actuate the fan coolers and stop the temperature rise.

The setpoint is sufficiently low that the co~ponents are not endangered.

15.4.9 References for Section 15.4 *

1.

"Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10CFRSO. 46 and Appendix IC of lOC~O. Federal Register, Volume 39, No. 3, January 4, 1974.

2.

Bordelon, F. K.;

Kassie, H. W.;

and

Zordan, T. A.,

"Westinghouse ECCS Evaluation Hodel - S0mmary, 11 WCAP-8339, July 1974.

3.

Bordelon, Space-Time June 1974 F. K.

et al, "SATAN~VI Program:

Comprehensive Dependent Analysis of Loss of Coolant," WCAP-8302, (Proprietary) and WCAP-8306, June 1974 (Nonproprietary).

15.4-123 SGS-UFSAR A-18 Revision 6 February 15, 1987

_ _J

4.

5.

~

(

i;\\l""(

et al, "LOCTA-IV Program:

Loss of Coolant I

~e..~ ~'-\\--\\

Transient Analysis," WC 74 (Proprietary) and

.JL~ *'1'X \\>'-

\\

A ~'i.

WCAP-8305, June 1974 (Nonproprietary).

~--~

• y Kelly R. D. et, al, "Calculational Model for Core Reflooding after a Loss-of-Coolant Accident (WREFLOOD Code)," WCAP-8170, June 1974 (Proprietary) and WCAP-8171, June 1974 (Nonproprietary).

6.

Bordelon, F. M.

and Murphy, E. T., "Containment Pressure Analysis Code (COCO)," WCAP-8327, June 1974 (Proprietary) and WCAP-8326, June 1974 (Nonproprietary).

7.

Bordelon, F. M. et al, "Westinghouse ECCS Evaluation Model -

Supplementary Information,"

WCAP-8471-P-A, April 1975 (Proprietary) and WCAP-8472-A, April 1975 (Nonproprietary).

8.

9.

10.

"Westinghouse (Nonproprieta C. Eicheldinger ssallo 1975 V~on:*,)

d WCAP-8~

of Westinghouse Electric of the Nuclear Regulatory Commission.

Letter No. NS-CE-924, January Thompson, C. M.; et al, "Westinghouse Emergency Core System -Evaluation Model for Analyzing Large LOCAs During

• h One Loop Out of Service for Plants Without WCAP-9166, February 1978.

11.

c.,

j:plitta'lt:10il"-1Mode 1,

) and WCAP-9221 February 1978 Version,"

(Nonpropri

, February 1978.

15.4-124 SGS-UFSAR A-19.

Revision 6 February 15, 1987

(

from*

T. H. Anderson of Westinghouse John Stolz of the Nucle

13.

Electric

14.

Letter from C. Eicheldinger of Westinghouse Electric Corporation to V. Stello of the Nuclear Regulatory Commission.

Letter No. NS-CE-1163, dated August 13, 1976.

15.

Beck, H. S. and Kemper, R. H., "Westinghouse ECCS Four-Loop Plant (17 x 17)

Sensitivity Studies,"

WCAP-8865, October 1976.

16.

Salvatori, R.,

"Westinghouse ECCS Plant Sensitivity Studies," WCAP-8340, July 197 4 (Proprietary) and WCAP-8356, July 1974 (Nonproprietary).

17.

Johnson, W. J.;

Hassie, H. W.;

and

Thompson, C.

H.,

"Westinghouse ECCS - Four-Loop Plant (17 x 17) Sensitivity Studies," WCAP-8565, July 1975 (Proprietary) and WCAP-8566, July 1975 (Nonproprietary).

18.

S. Nuclear Regulatory Coumission letter, D. G. Eisenhut to

./

Util1 *es With Operating Light Water Reactors, er 9, 1979.

19.

Powers, D. A. and Swelling and Rupture 1979.

20.

Letter SGS-UFSAR of D. G. Eisenhut of the Letter No. NS-THA-2147, 15.4-125 A-20 November.8, Electric Revision 6 February 15, 1987

21:

from T. M. Anderson of Westinghouse to D. G. Eisenhut of the Nuclear No. NS-TMA-2163, 1979.

22.

Letter from Electric Corporation Regulatory Commission.

23.

Letter Anderson of Electric to R. Denise of the Letter No. NS-TMA-2175, December

24.

Geets, J.M., "MARVEL - A Digital Computer Code for Transient Analysis of a Multiloop PWR System," WCAP-7909, June 1972.

25.

Moody, F. S., "Transactions of the ASME Journal of Heat Transfer," Figure 3, page 134, February 1965.

26.

Bordelon, F. M., "Calculation of Flow Coastdown After Loss of Reactor Coolant Pump (PHOENIX Code),"

WCAP_-7973, September 1972. *

27.

Burnett, T. W. T.; Mcintyre, C. J.; Buker, J. C., and Rose, R. P.,

"LOFTRAN Code Description," WCAP-7907, June 1972.

28.

Hunin, C.~ "FACTRAN, A Fortran IV Code for Thermal Transients in a uo2 Fuel Rod," WCAP-7908, June 1972.

29.

Burnett, T. W. T., "Reactor Protection System Diversity in Westinghouse April 1969.

Pressurized Water Reactors,"

WCAP-7306,

30.

Taxelius, T. G.

(Ed),

"Annual October 1968, September 1969,"

IN-1370, June 1970.

15.4-126 SGS-UFSAR

• A-21 Report - Spert

Project, Idaho Nuclear Corporation Revision 6 February 15, 1987 I

\\ I

INSERT A TO FSAR TEXT REFERENCES TO BE INSERTED 4; WCAP-10266-P-A, Rev 2, including Addendum 1 & 2, "The 1981 Version of the Westinghouse ECCS Evaluation Model using BASH."

9.

Rahe, E.P. (Westinghouse) letter to Tedesco, R.L. (USNRC),

NS-EPR-2538, December 1981.

10.

"Westin~house ECCS Evaluation Model - 1981 Version",

WCAP-9220-P-A & WCAP-9221-A; Revision 1, Februa~y, 1982.

INSERT C TO FSAR TEXT Calculations of cold leg double-ended guillotine pipe breaks are performed ov~r a range of Moody discharge coefficients (Co) to identify the case which produces the highest peak clad temperature (PCT).

For this analysis, calculations were performed for discharge coefficients of 0.4, 0.6, and 0.8. This spectrum of breaks was performed assuming the availability of only minimum safety injection flow capacity (Minimum Safeguards), in accordance with the single failure criteria of 10CFRSO, Appendix K.

A break discharge coefficient of 0.4 was found to result iri the highest PCT of 2091°F. This ts acceptably below the 2200°F limit of 10CFRSO.

Consistent with the methodology described in reference 9, an additional calculation is performed for the identified worst break case, assuming maximum safeguards availability. In this calculation, the delivered ECCS flow is maximized by the following assumptions: minimum injection line resistances, and no safet~ injection system pump failures. This case was resulted in a PCT of 2017 F, which is bounded by the Minimum Safeguards case.

INSERT D TO FSAR TEXT SATAN was conservatively not rerun for the 'Max SI' case because it would have been only a small blowdown benefit.

As a result, the following Co=0.4 blowdown figures apply to the Max SI case as well: 15.4-2, -3,

-15, -20, -29, -38, -47.

A-22

'~

INSERT B TO FSAR TEXT Descriptions of the various aspects of the LOCA analysis are provided in references 2, 4, 7, and 10.

These documents describe the major phenomena modeled, the interfaces among the computer codes and features of the codes which serve to maintain compliance with the acceptance criteria of IOCFRS0.46.

The analysis of a large-break LOCA transient is divided into three phases:

Slowdown, Refill and Reflood.

A series of computer codes has been developed to analyze the transient based on the specific phenomena which govern each phase.

During the blowdown portion, the SATAN-VI code (3) is used to calculate the RCS pressure, enthalpy, density and mass and energy flows in the primary system, as well as the heat transfer between the primary and secondary system.

At the end of the blowdown, information on the state of the system is transferred to the WREFLOOD code (5) which performs the calculation of the refill period to bottom of core (BOC) recovery time.

Once the vessel has refilled to the bottom of the core, the reflood portion of the transient begins.

The BASH code (4) is used to calculate the thermal-hydraulic simulation of the RCS for the reflood phase.

Information concerning the*core boundary condition is taken from all the above codes and input to the LOCBART code (4) for the purpose of calculating the core fuel rod thermal response for the entire transient.

From the boundary conditions, LOCBART computes the fluid conditions and heat transfer coefficient for the full length of the fuel rod by employing mechanistic models appropriate to the actual flow and heat transfer regimes.

Conservative assumptions ensure that the fuel rods modeled in the calculation represent the hottest rods in the entire core.

The containment pressure analysis is performed with the COCO code (6),

which is run interactively with the WREFLOOD code (5). The transient pressure computed by the COCO code is then input to the BASH code (4) for the purpose of supplying a backpressure at the break plane while computing the reflood transient. The containment parameters used in the COCO code to determine the ECCS backpressure are presented in Table 15.4-3.

A-23

TABLE 15.4-1 TIME SEQUENCE OF EVENTS FOR CONDITION IV EVENTS Major eactor Coolant System ipe Ruptures Double-E ded Cold Leg Guillotin

1.

2.

(C =

D (CD = 0.8)

3.

(~ = 0.6)

SGS-UFSAR Event Start Reactor trip signal Safety injectio*n signal Accumulator injection End of blowdown Bottom of core recovery Accumulators empty Pump injection End of bypass ctor trip signal ty injection signal ulator injection o blowdown f core recovery Accumula ors empty Pump inj e ion End of bypa s Start Reactor al Safety injection *gnal Accumulator inject n End of blowdown Bottom of core recovery Accumulators empty Pump injection End of bypass 1 of 3 A-24 Time (sec) o.o 1.65 0.86 14.1 28.1 40.34 51.15 25.86 25.4 o.o 1.66 0.92 14.6 28.8 40.95 51.6 25.92 26.0 o.o 1.66 1.03 16.8 30.46 42.5 53.64 26.03.

27.51 Revision 6 February 15, 1987

TABLE 15.4-2 LARGE BREAK RESULTS DEC LG DECLG DECLG (Cn=l.O)

(CD=0.8)

(Cn=0.6)

Peak Clad 2108 2130 1968 Peak Clad Lo ca *on (ft) 6.0 6.0 7.5 Local (max)%

5.97 6.1 2.87 Local (ft) 6.0 6.0 7,5 Local Zr/H20 Reaction (%)

<o.3

<0.3

<0.3 Hot Rod 31.4 28.1 33.0 Hot Rod Burst Location (

6.0 6.0 6.25 Calculation NSSS Power (MWt) 102% of 3411 Peak Linear Power (kW/ft) 102% of 12.63 Peaking Factor (at license rating) 2.32 Accumulator Water Volume (ft3 per

-accumulator) 850 Fuel region + cycle analyzed Cycle 1

1 of 1 SGS-UFSAR A-25 Revision 6 February 15, 1987

TABLE 15.4-3 CONTAINMENT DATA Service Outside Spray System Number of (Of) r Temperature (°F)

Runout Flow Rate (eac ) (gpm)

Actuaton Time (sec)

Safeguards Fan Coolers Number of Fan Coolers Fastest Post Accident of Fan Coolers Structural Heat Sinks Thickness (in.)

0.0075 Paint, 0.375 Steel, 54 Concrete 2.5 Insulation, 0.375 Steel, 54 Concrete 0.0075 Paint, 0.5 Steel, 42 Concrete.

0.018 Paint, 42 Concrete 0.018 Paint, 12 Concrete 0.018 Paint, 20.5 Concrete 0.014 Paint, 18 Concrete 1 of 2 SGS-UFSAR A-26 2.62 x 106 ft3 14.7 90 40 32 0

2 3800 27 5

30 Area (ft2 )

10,416 35,000 Revision 6 February 15, 1987

\\

TABLE 15.4-3 (Cont) 0.187 Steel, 23 Concrete 0.0075 Paint, 0.1 Steel Paint, 0.25 Steel Paint, 0.5 Steel 0.75 Steel 1.0 Steel 0.0075 Steel 0.0075 Steel 0.0625 Steel 1.125 Steel 0.125 Steel 0.86 Steel 1.41 Steel 2 of 2 SGS-UFSAR A-27 17,536 73,870 90, 110 23,688 10,864 9,441 3,370 1,916 53,460 1,832 133,056 7,274 4,915 Revision 6 Febrµ~ry 15, 1987

TABLE 15.4-4

• REF OOD HASS AND ENERGY RELEASES FOR LIMITING CASE <<=n=0.8)

Time (sec) 42.0 47.S 55.9 68.3 83.5 100.4 118.8 159.7 SGS-UFSAR Total Hass Flowrate (lbm/sec) 1 of 1 A-28 Total Energy Flowrate (Btu/sec) 0.0 51,192.0 179, 111. 0 217,064.0 221,579.0 216,319.0 209,858.0 194,033.0 Revisir.m 6 Febr*.1a:i:*y 15, 1987

~------------

TABLE 15.4-1 TIME SEQUENCE OF EVENTS FOR CONDITION IV EVENTS Time Accident Event illli Major Reactor Coolant System Pipe Ruptures Double-Ended Cold Leg Guillotine

1.

(CD=.4)

Start 0.0 Reactor trip signal

.95 Safety Injection signal 1.81 Accumulator injection

• 19.4 Pump injection 31.81 End of blowdown

.39.3 End of bypass 39.3 Bottom of core recovery 55.5 Accumulators empty 63.2

2.

(CD=.4 Max SI)

Start 0.0 Reactor trip signal

.95 Safety Injection signal 1.81 Accumulator injection 19.4 Pump injection 31.81 End of bypass 39.3 End of blowdown 39.3 Bottom of core recovery 54.5 Accumulators empty 70.3

3.

(CD=.6)

Start 0.0 Reactor trip signal

.93 Safety Injection signal 1.49 Accumulator injection 13.9 End of blowdown 30.2 End of bypass 30.2 Pump injection 31.49 Bottom of core recovery 44.7 Accumulators empty 59.5

4.

(CD=.8)

Start 0.0 Reactor trip signal

.91 Safety Injection signal 1.32 Accumulator injection 11.8 End of blowdown 26.4 End of bypass 26.4.

Pump injection 31.32 Bottom of core recovery 40.3 Accumulators empty 54.3 A-29

TABLE 15.4-2 LARGE BREAK RESULTS DECL DECL CD=0.4 CD=0.4 MAX SI Results Peak Clad Temperature (OF) 2091 2017 Peak Clad Temp. Location (ft) 8.5 8.5 Local Zr/H2o Reaction (max) %

6.96 5.34 Local Zr/H2o Location (ft) 8.5 8.5 Total Zr/H20 Reaction (%)

<0.3

<0.3 Hot Rod Burst Time (sec) 57.44 57.60 Hot Rod Burst Location (ft) 6.25 6.25 Calculation Reactor Power (MWt} 102% of Peak Linear Power (Hottest Rod) (Kw/ft). 102 % of Peaking factor (at above power)

Accumulator Water Volume (ft3)

Fuel Analyzed 17 X 17 Standard Bounding A-30 DECL CD=0.6 2071 8.5 7.03 8.5

<0.3 41.48 6.00 3588 13.75 2.40 DECL CD=0.8 1911 8.5 4.63 7.25

<0.3 86.24 7.25 850 per accumulator

I TABLE 15.4-3 CONTAINMENT DATA Net Free Volume Initial Conditions Pressure (p~ia)

Temperature (°F)

RWST Temperature (°F)

Service Water Temperature (°F)

Outside Temperature (°F)

Spray System I Number of Pumps Operating Runout Flow Rate (GPM) (per pump)

Actuation Time (sec)

Safeguards Fan Coolers I Number of Fan Coolers Operating Fastest Post-Accident Initiation of Fan Coolers I

Assumes Loss Of Offsite Power I of 4 SGS-UFSAR A-31 2.62 x 106 ft3 14.7 90 35 32

-10 2

3800 27 5

30 l__

Structural Heat Sinks Thickness (in.)

TABLE 15.4-3 (Cont) 0.005 Surface Paint (for steel),

0.0025 Primer Paint (for steel), 0.375 Steel, 54 Concrete 2.5 Insulation, 0.005 Paint (for steel), 2 0.375 steel, 54 concrete 0.005 Surface Paint (for steel),

0.0025 Primer Paint (for steel), 0.5 Steel, 42 Concrete 0.008 Surface Paint (for concrete),

0.010 Primer Paint (for concrete), 42 Concrete 0.005 *Surface Paint (for steel),

0.0025 Primer Paint (for steel), 0.0625 Steel 0.005 Surface Paint (for steel),

0.0025 Primer Paint (for steel), 0.375 Steel 0.0625 Steel Area (ft2) 49,924 15,655 32,328 13,139 390,880 62,253 330,012 2

Both Surface Paint and Primer Paint combined.

Thermal properties assumed as the more conservative ~f the properties for either paint type.

2 of 4 SGS-UFSAR A-32

TABLE 15.4-3 (Cont)

Structural Heat Sinks Thickness (in.)

0.375 Steel 0.8333 Steel 0.8125 Steel, 1.000 Insulation, 0.216 Steel 0.008 Surface Paint (for concrete),

0.010 Primer Paint (for concrete), 18 Concrete 0.004 Surface Paint (for concrete),

0.010 Primer Paint (for concrete), 41 Concrete 0.25 Stainless Steel, 51 Concrete 0.09375 Stainless Steel 0.005 Surface Paint (for steel),

0.0025 Primer Paint (for steel), 1.125 Steel 0.005 Surface Paint (for steel),

0.0025 Primer Paint (for steel), 1.375 Steel 0.005 Surface Paint (for steel),

0.0025 Primer Paint (for steel), 2.5 Steel 1.25 Steel 1.875 Steel 3 of 4 SGS-UFSAR

• A-33 Area (ft2) 11,354 4,789 7,037 15,400 10, 911 46,490 8,538 24,333

" 3,679 4,053 17,087 2,259 6,809

Material Properties Material Steel (Carbon)

Stainless Steel Concrete Insulation Surface Paint (for steel)

Primer Paint (for steel)

TABLE 15.4-3 (Cont)

Thermal Conductivity (BTU/ft3-°F}

28.0 8.5 1.04 0.024 0.119 0.292 Surface Paint (for concrete) 0.292 Primer Paint (for concrete) 0.292 4 of 4 SGS-UFSAR A-34 Volumetric Heat Capacity (BTU/hr-ft-°F) 58.8 58.8 23.4 3.94 30.53 32.48 38.4 52.8

TABLE 15.4-4 RCS TO CONTAINMENT MASS AND ENERGY RELEASES. FOR LIMITING CASE (CD=0.4)

DOES NOT INCLUDE ACCUMULATOR/SI SPILL Total Mass Total Energy Time Flowrate Flowrate

.llitl Cl bm/sec)

(Btu/sec)

0.

0.

0.

0.25 532,371.

28,153,000.

13.0 14,314.

9,352,000.

32.

5,289.

2,194,000.

38.0 954.

258,600~

40.

0.

0.

54.

0.

0.

59.

20.

. 2,573.

74.

82.

101,100.

99.

104.

127,300.

139.

378.

208,300.

189.

444.

209,300.

239.

456.

202,300.

A-35

TABLE 15.4-5 BROKEN LOOP ACCUMULATOR/PUMPED SI CONTAINMENT MASS AND ENERGY SPILL FOR LIMITING CASE CD=0.4 Total Mass Total Energy Time Flowrate Flowrate ti.ill.

Cl bm/secl

{Btu/sec)

0.

2, 719.

162,200.

5.0 2,115.

126,200.

10.

1,796.

107,100.

15.

1,582.

94,356.

20.

1,425.

85,000.

25.

1,304.

77,800.

30.

1211.

72,200.

32.

150.

450.

55.

151.

453.

59.

159.

477.

94.

160.

480.

179.

162.

486.

301.

163.

489.

A-36

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I

~

I 0

0 P) 0 in N

0 0

N 0 in -

0 0 -

0 0

V)

Q z 0

c..>

L.i.I V) -

L.i.I

E: -

I-Heat Transfer Coeff;c;ent DECLG (Co=0.6)

Updated FSAR Figure 15. 4-11 A-50 J J

0 in I-(/)

c:: I-

>u co c..

.. I ><

Cl

z L.W

(.!J L.W

-I I

--:J>

~

0

('f)

\\/

\\

~

~

~

I --f,_,.

--t<~x

'\\

0 N

\\

\\.

/\\

}:-

~

~

( b-:-

',\\

/

l"r

-r~

I

( ~

~

,">., i I,

-t~

'/~

'v*

.::::~~

~

\\,... _

~]

~

j \\

0 -

}-.

0 0

0

('f) 0 LO N

0 0

N 0

LO -

0 0 -

0 vr c z 0 u LI.I V) -

LI.I

E -

PUBLIC SERVICE ELECTRIC AND GAS COMPANY Heat Transfer Coefficient DECLG (Co=0.4)

SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.4-12 A-51

I i

I I

I I I

i-0 U')

I-V>

c:: I-

>U cc c..

.. I ><

c z:

LiJ

~

LiJ

-I

--.........._ ~

I I

0 M

\\

I I

I<

~

I\\

.~,,

I

\\

0 N

l:t

\\.-

• ~

~

• -~

1\\ - /

/ '";

/

.)

~- ;

~

/

/

~

./

-i

\\

/-::

I l 1: ~*~

I

~

t~

I' l

\\ '

~ L

-2 lJ' 0....

(~0 -HH-i1:1/n1e) 1N3I~I~~30~ H3~SNVH1 1V3H l

0 0

0 M

0 LO N

0 0

N 0

LO....

0 0....

0 Cl')

c z 0

(,,)

LI.I Cl') -

LI.I

E I-Maximum Safeguards PUBLIC SERVICE ELECTRIC AND GAS COMPANY Heat Transfer Coefficient DECLG (Co=0.4)

SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.4-12A A-52

~

Cl

z UJ c.!:l UJ 0

0 Lt)

N

...J

E:

0 C..1-01-1-0 c::l

-IC I r

0 0 0 N

I I

J

/

I I

J I

v*

0 0

Lt) -

0 0

0 -

(VISd) 3HnSS3Hd 3HO~

0 0

Lt)

)

I J

I I

I I

0 0

N

~ -

N -

co 0

en Cl z

0 u LI.I en -

LI.I

E: -

I-PUBLIC SERVICE ELECTRIC AND GAS COMPANY Core Pressure DECLG (Co=0.8)

SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.4-13 A-53 I

I

~

I I

L I

I I

i I !

' I i

i I

I I

I i

0 0

ID N

E:

0 C..1-01-1-0 c:::I

.. I Cl

z:

lJ..J c.!:l lJ..J

.....J 0

0 0

N

/"

f

~I 0

0 ID I

r J

};*

0 0

0....

I

)

I I

/

I I

0 0

ID

~

0 0

M 0

N ID 0....

0 V) c z 0

c.J I.I.I V) -

I.I.I

E -

I-(VISd) 3HnSS3Hd 3HO~

PUBLIC SERVICE ELECTRIC AND GAS COMPANY Core Pressure DECLG (Co=0.6).

SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.4-14 A-54

0 ll;f'

E:

)

0 0..1-01-1-0 cc

-IC Cl z:

LL.I c.!:I LL.I

....I 0

(")

I I -

V) c I

z lo 0 u LI.I V)

N -

LI.I

E i

I-I I L-~~~-!...~--oc==-...1.-~~~~~~~~_;_~__:_~---'* 0 0

0 ID N

0 0

0 N

0 0

ID -

0 0

0 -

0 0

ID 0

(VISd) 3HnSS3Hd 3HO~

PUBLIC SERVICE ELECTRIC AND GAS COMPANY Core Pressure DECLG SALEM NUCLEAR GENERATING STATION Updated FSAR A-55 (Co=0.4)

Figure 15.4-15

)

\\

I

' I

/

J I J

___-" ~

E /

0 0

<")

LO N

0 N

LO....

Q....

Q Q z Q u LI.I.,, -

LI.I

E: -

PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Break Flow DECLG (Co=0.6)

Updated FSAR Figure 15.4-16 A-56

(

~

I/

I I

I/

I v

I j

..,,.----v

.! v 0

(~3S/81 ~Ot) MOl~ ~V3~8 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Break Flow DECLG SALEM NUCLEAR GENERATING STATION Updated FSAR A-57 a>

N 0

N IQ -

N -

0 Cl)

Cl

z:

0 u LaJ Cl) -

LaJ

E -

(Co=0.8)

Figure 15.4-17

0 co

\\

)

J

{

I

~

I I

2 5

I !

0 (ISd) dOHC 3HnSS3Hd 3HO~

0 o:t' I

0 co I

0 N

!'D....

N....

0 V')

c z 0 u LL.I V') -

LL.I

E -

PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Core Pressure Drop DECLG (Co=0.8)

Updated FSAR Figure 15.4-18 A-58

\\-

0 co

)

I

• ~

(

}

0 (ISd) dO~O 3~nss3~d 3~0~

0

~

I 0

C")

Lt)

N 0

N Lt) -

0 -

0 0 co I

Cl)

Q z 0

(,,)

L.iJ Cl) -

L.iJ

E: -

I-PUBLIC SERVICE ELECTRIC AND GAS COMPANY Core Pressure Drop DECLG (Co=0.6)

SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.4-19 A-59 I,

0 co I

\\

)

(

\\

~

0 (ISd) dOHO 3HOSS3Hd 3HO~

0 od' I

0 co I

0 N

0....

0 Cl)

Q z 0

(.)

I.LI Cl) -

I.LI

E -

PUBLIC SERVICE ELECTRIC AND GAS COMPANY Core Pressure Drop DECLG (Co=0.4)

SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.4-20 A-60

~----,-----------------------.-----

0 0

Lt)

N f

t-(/')

0::: t-

> u cc c..

.. I ><

c::::i

z UJ C!l UJ

....I 0

0 0

N I Ii i I

' I

-~

L.1 l

J' /I

\\

\\

. 1; t

5~

'. \\

, /

/.

~*

( l

~

"/

~ I

~*

! -:"'\\i.

"' ~~

~

.--J 0

0 Lt) -

I II i

I L.

I

~

I I

I.,*

0 0

0 -

0 0

Lt)

I 0

C~o> OOH !OH 3HnlVH3dW3l 39VH3AV OV1~

Clad Average Temperature PUBLIC SERVICE ELECTRIC AND GAS COMPANY Hot Rod DECLG (Co=0.8).

SALEM NUCLEAR GENERATING STATION 0

0

('t) 0 0

N 0

0 -

0 Updated FSAR Figure A-61 V)

Q z 0 u LI.I V) -

LI.I

E: -

I-15.4-21

0

,---..-----.--..--..--.---..--..-......--..---..-..--..-~-..--..-....--..--..-..---..-___,..-..--..--..--..------, 0 M

l--..--1-..--..-...L-..--..-1-~.:::::::::~::3:~==d:::::i"--~L_-..--1~-..-....L~__Jo 0

0 i.n N

0 0

0 N

0 0

i.n 0

0 c....

0 0

i.n 0

(~0 ) oo~ !OH 3~nlVll3dW3! 39Vll3AV ovi~

• PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Clad Average Temperature Hot Rod DECLG (Co=0.6)

Cl) c z 0 u LI.I Cl) -

LI.I

E -

Updated FSAR Figure 15.4-22 A-62

0

.--~-..~~-.-~~--.-..--~...-~---.~~---.-~~--.-~~-.-~-.....-~---. 0 1-Vl ex:: 1-

> u cc c..

('t) 0 l--~-l-~~--~~----~++-~---1--~-+-~~-r-~~--~~---~--10 0

0 Lt)

N 0

0 0

N 0

0 Lt) -

0 0

0 -

0 0

Lt) 0 C~o> oou !OH 3Un!VU3dW3! 39Vlt3AV ov1~

Clad Average Temperature (Max SI)

PUBLIC SERVICE ELECTRIC AND GAS COMPANY Hot Rod DECLG (Co=0.4)

SALEM NUCLEAR GENERATING STATION Cl) c z 0 u I.LI Cl) -

I.LI

E -

I-Updated FSAR Figure 15.4-23A A-63

1--

0 0 ll)

N I

I-V) 0::: I-

>u cc c..

.. I ><

c z

LL.I

(.!:I LL.I

-I 0

0 0

N J

~

.:r:

" ' I sµ.

' \\ I

~

(.' v t-('

~

{'

)

~

I

\\

l1

...-,:. J

~

~I

~'

r-.....

0 0 ll) -

.. \\

0 0

0 -

I 7

) ~

I.

-.;;;;;; ~\\.-

'~

. '\\

0 0

ll)

I I

I I I

I 0

0

('I')

0 0

N 0

0 -

0 0

Q z 0 u LI.I PUBLIC SERVICE ELECTRIC AND GAS COMPANY Flu;d Temperature DECLG {Co=0.8)

SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.4-24 A-64

0 0 Ill N

t-(/)

0:::: t-

>u cc 0..

.. I ><

Cl

z:

L.l.J c.!J L.l.J

.....J 0

0 0

N 1-

/

A

~

I 1

-I-

> 1j i

! i..-1

~

5

~ I(

..... ~ J

)(

{~

I, ~

~\\~

~-

\\~/

0 0 Ill

/\\

/

,/

.[_

/ /

I I

~

lf

"\\..

---*~

I 0

0 0....

_r :r. I 0

0 Ill

~.....

' ~

0 0

0

(")

0 0

N 0

0....

0 en Q z 0 u.

LI.I en -

LI.I

E -

I-PUBLIC SERVICE ELECTRIC AND GAS COMPANY Fluid Temperature DECLG (Co=0.6)

SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.4-25 A-65

0 0

i.n N

I-Vl c::: I-

>u cc c..

.. I ><

0

z 1.1.J

(.!:!

1.1.J

_J 0

0 0

N

~*

~

~l

' '/ I ti

\\

\\

+f -}

Sf.

i f

5 (I)

SR 't 2!',,

~' ->

t)

<~' <

~

~

~7

'~

~

I

,~

~~\\I 0

0 i.n -

1\\

PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION I

~

b

/

I/

I

~

~. I

-=~

I

• ~

• \\

~ I

,I~

0 0

0 -

.... \\

0 0

i.n

~~

{'

0 0

M 0

0 N

0 0 -

0 0

.en Cl z

0 u

~

V) -

Fluid Temperature DECLG (Co=0.4)

Updated FSAR Figure 15.4-26 A-66

0 0 in N

I I-Vl ex: I-

~u cc c..

• • 1 x Cl x

z x

L..r..J

~

L..r..J

_J 0

0 0

N A,/

,7\\

it+\\

I I. lt

~

~

~--

I

~

.:j '

~

~

/\\

-r}

,P--!;

'~

~* ~,

• O 0 in -

\\ /\\ --f v

j,/

/

I If f

)

Li

~

(

.~

f

.r,.

---..-n-..

I LL ll' I

I'-

I

. I

~

• I\\~

0 0

0 -

0 0 in 0

0 0

('f')

0 0

N 0

0 -

0

(.:10 ) 3ltn!Vll3dW3! orn1.:1 Maximum Safeguards PUBLIC SERVICE ELECTRIC AND GAS COMPANY Fluid Temperature DECLG (Co=0.4)

SALEM NUCLEAR GENERATING STATION V) c z 0

(,,)

LIJ V) -

LIJ

E -

Updated FSAR Figure 15.4-26A A-67

~

~

0 0

0 co I

0 z:

LLJ t!J LLJ

...J

E:

0 0..1-01-1-0 a::i

-IC I 0

0 0

'l:t'

~.

I I

\\*

~

I

'\\

~

I I

~

• I i I\\

~

l*

I r l(

j)

-)r

\\.

~

~

.~<:'.'.'

0

(~3S/81) 3!VllM01~ 3HO~

I 0

0 0

'l:t' I

I I

I I i

co N

0 N

'° -

N -

0 0

0 0 co I

Cl) c z 0 u LI.I Cl) -

LI.I

E -...

PUBLIC SERVICE ELECTRIC AND GAS COMPANY Core Flowrate DECLG (Co=0.8)

SALEM NUCLEAR GENERATING STATION Updated FSAR.

Figure 15.4-27 A-68

E:

0 r

c.. I-01-1-0

~

co

-IC

---.. I Cl z:

LLJ

~

LLJ

-l

~

/'

1t J

J I

(

)

)

~

........... ~

0 0

0 co 0

0 0

Olll:t' 0

0 0

0 Olll:t' I

(~3S/81) 31VllM01~ 3HO~

PUBLIC SERVICE ELECTRIC AND GAS COMPANY Core Flowrate DECLG SALEM NUCLEAR GENERATING STATION Updated FSAR A-69 I

I 0

0 0 co I

0

('I')

0 N

in -

0 -

0 c z 0

(.)

1.1.i 1.1.i

E -

(Co=0.6)

Figure 15.4-28

0 0

0 ClO Cl z

L.l.J

(,!'

L.l.J

.....J

E:

0 0.. I-01-1-0 CJ

-IC I I

I I

i i I l

)

r 1

~

!~

/

t

< I:

0 0

0

~

_:)

~

0 I

I I

1 I

I I

I I

I

• I I

I

(~3S/81) 3!VHM01~ 3~0~

I I

I 0

0 0

~

I 0

N 0....

0 0

0 0

ClO I

V) c z 0 u LI.I V) -

LI.I

c -

I-PUBLIC SERVICE ELECTRIC AND GAS COMPANY Core Flowrate DECLG (Co=0.4)

SALEM NUCLEAR GENERATING STATION

• Updated FSAR Figure 15.4-29 A-70

0 0

L.U ('I)

L.U a:

a:

c c

(,,)

(,,)

LI..

LI..

c c

E c..

L.U c

c a:

c z

(,,)

c 0

cc a:

Q LI..

LI.I a:

Cl)

L.U

c c..

E

(,,)

cC 0

z

...I

(,,)

LI.I

...I z

c

a:

a

(,,)

0 LI.I Q

a:

0

(,,)

0 0 N -

Cl)

Q z 0

(,,)

LI.I Cl) -

(,,)

0 cc a:

L.U LI..

cC LI.I

E -

0 0....

L-~--1.~~L-~-L~____:L_~_L~~L--.~....I:::::::;~;:::::~o 0

N N....

(l.:I) 13A31 H3!VM PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION co Water Level Updated FSAR A-71 (Ft) DECLG (Co=0.8)

Figure 15.4-30

,j

0 LLI 0

LLI rt>

a::

a::

0 0

u u

L&..

L&..

0 0

a.

LLI

E:

0 a::

0 0

z u

0 0

a::

Q ca LL.I L&..

V)

c

a.

a::

u c:C LL.I z

...I

E:

LL.I

...I 0

0 u

a u

z 3

LL.I 0

a::

Q 0 u 0

0 N -

V)

Q :z:

0 u LL.I V) -

u 0 ca a::

LL.I L&..

c:C LL.I

E: -

I-0 0....

{LJ) 13A3'1 H3lVM PUBLIC SERVICE ELECTRIC AND GAS COMPANY Water Level {Ft) DEC LG {Co=0.6)

SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.4-31 A-72

0 L.U L.U 0

a::

a::

M 0

0

(.)

(.)

LI..

LI..

0 0

Q.

E:

0 L.U 0

a::

0 z

(.)

0 0

ca a::

LI..

a::

c L.U

E:

(.)

z 0

L.U

(.)

~

z 3

a 0

L.U c

a::

0

(.)

0 0 N* -

V) c z 0 (.)

L.U V) -

(.)

0 ca a::

L.U LI..

cc L.U

E: -

0 0 -

{J.J) 13A31 H31VM PUBLIC SERVICE ELECTRIC AND GAS COMPANY Water Level {Ft) DECLG {Cp=0.4)

SALEM NUCLEAR GENERATING STATION Updated -FSAR Figure 15.4-32 A-73

LI.I a::

0 u LI..

0 z

Cl.

0 a::

0 a::

L&J LI..

E:

0

c u

u z

z 3:

L&J 0

c a

(!:I) 13A31 H3!VM PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION.

0 LI.I 0 a::M 0 u LI..

L&J 0

a::

0

c u

0.....

c I.I.I 0

V)

CQ Cl.

...J

...J 0 u lJ 0

0 C\\I -

V) c z 0 u LI.I V) -

u.

0 CQ a::

LI.I LI..

LI.I

E: -

0 0 -

Maximum Safeguards Water Level (Ft) DECLG (Co=0.4)

Updated FSAR Figure 15.4-32A A-74

0 Ill - -

Cl) c z 0 u LI.I Cl) -

LI.I

E -

I-0 0 -

(~3S/NI) 3!Vll 0001~

PUBLIC SERVICE ELECTRIC AND GAS COMPANY Flood Rate (In/Sec) DECLG {Co=0.8)

SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.4-33.

A-75

0 i.n - -

V) c z 0

~

LI.I V) -

LI.I

E: -

0 t-0 -

.__~-'-~-'-~~...._~_,__~__.~~--~_...~~---...__._~~o 0 -

c:o

(~3S/NI) 3l'lll 0001~

PUBLIC SERVICE ELECTRIC AND GAS COMPANY Flood Rate {In/Sec)

SALEM NUCLEAR GENERATING STATION Updated FSAR A-76 0

DECLG {Co=0.6)

Figure 15.4-34 L

Cl in V'l Cl

z Cl c..>

LI.I V'l LI.I

~

Cl Cl

(~3S/NI) 3!Vll 0001~

PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Flood Rate (In/Sec) DECLG (Co=0.4)

Updated FSAR Figure 15~4-35 A-77

0 -

i I

(~3S/NI) 3!Vll 0001~

I I

I I I i

I I

I i

0 l.t'l N

0 0

N 0

Lt) -

0 0 -

0 Lt) 0 0

V) c z 0-(..)

LU V) -

LU

E -

I-Maximum Safeguards PUBLIC SERVICE ELECTRIC ANO GAS COMPANY SALEM NUCLEAR GENERATING STATION Flood Rate (In/Sec) DECLG (Co=0.4)

Updated FSAR Figure 15.4-35A A-78

0 0

0 IO I

j

\\

\\

'2

~---

0 0

0 o:f' 0

0 0

N

(:>3s/e1> Mo1.:1 110.unnwn:>:>v

---r-.......

'\\ <

co N

0 N

IO -

N -

co 0

0 Cl)

Cl z

0 u LI.I Cl) -

LI.I

E -

I-PUBLIC SERVICE ELECTRIC AND GAS COMPANY Accumulator Flow DECLG (Co=0.8)

SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.4-36 A-79

0 0

0

'°

. I

\\

\\

~:___

0 0

0 "It' 0

0 0

N

(~3s/e1> Mo1~ uo1v1nwn~~v

~

)

0 0

('I')

LO N

0 N

LO....

0....

0 V)

Q z:

0 c..>

L&.I V) -

L&.I

E -

I-PUBLIC SERVICE ELECTRIC AND GAS COMPANY Accumulator Flow DECLG (Co=0.6)

SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.4-37 A-80

0 0

0

'°

~

\\

}

\\

\\ \\

)

0 0

0 o:t' 0

0 0

N

(~3s/a1> MOl~ Ho1v1nwn~~v

{

0 0

N 0 -

0 II) c z 0 u LI.I II) -

LI.I

E -

I-PUBLIC SERVICE ELECTRIC AND GAS COMPANY Accumulator Flow DECLG (Co=0.4).

SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.4-38 A-81 j

r*

550

(,,)

LU Cl) m 500 3

0.....

~

450 400._,_ ____________________________________________ __,

0 40 80 120 1&0 200 240 TIME AFTER.BOC (SECONDS)

PUBLIC SERVICE ELECTRIC AND GAS COMPANY Pumped ECCS Flow (Reflood) DECLG (Co=0.8)

SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.4-39 A-82 L __

(.)

LI.I en ca D 3 C)

~

0

~

120 160 2~

TIME AFTER BOC (SECONDS)

PUBLIC SERVICE ELECTRIC AND GAS COMPANY Pumped ECCS Flow (Reflood) DECLG (Co=0.6)

SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.4-40 A-83

u LU V) ca 500

JC 0 _,

LI..

0 40 80 120 160 200 240 TIME AFTER BOC (SECONDS)

PUBLIC SERVICE ELECTRIC AND GAS COMPANY Pumped ECCS Flow (Reflood) DECLG (Co=0.4)

SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.4-41 A-84

11 105

(,.)

LI.I Cl)

CCI 100 3

0.....

LI..

95 gn+------....---..---...---.-----.---------~.....------.---..--4 0

80 120 160 200 240 TIME AFTER BOC (SECONDS)

Maximum Safeguards PUBLIC SERVICE ELECTRIC AND GAS COMPANY Pumped ECCS Flow (Reflood) DECLG (Co=0.4)

SALEM NUCLEAR GENERATING STATION Updated FSAR Figtire 15.4-41A A-85

/

/

I/

/

~

~

/

/

0 N

~

/

,/

~

v 0 -

~

(91Sd) 3~nSS3~d !N3WNIV!NO~

' " " \\.

\\

\\

\\

0 C"l 0....

N 0....

0....

0 0....

en Cl z

0 u I.I.I en -

I.I.I

E -

PUBLIC SERVICE ELECTRIC AND GAS COMPANY Conta;nment Pressure (PSIG) DECLG (Co=0.8)

SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.4-42 A-86

..)

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