Text

Proposed Tech Specs Modifying RHR Sys to Allow for More Flexability Concerning Air Entrapment & Vortexing
	ML18094A701
Person / Time
Site:	Salem
Issue date:	09/11/1989
From:	; Public Service Enterprise Group
To:	;
Shared Package
ML18094A700	List: ML18094A699; ML18094A701;
References
	NUDOCS 8909210159
	Download: ML18094A701 (85)
	v • d • e

3.1 ~ t. 2 OIV J3Acl:::'

Dc:LETEO 3.1.l.3 The f w rate of reactor coolant through the reactor coo ant system shall be 3000 gpm whenever a reduction in Reactor Cool t

System boron conce tration is being made.

APPLICABILITY:

ACTION:

\\ilith the flow rate of reacto coolant through the actor coolant system

< 3000 gp"1,. irilnediately suspe all operations

valving a* reduction in boron concentration of the React SURVEILLANCE RE UIREMENTS 4:1.l.3 The flow rate of rea or coolant through he reactor* coolant

.system shall be detennined be > 3000 gpm within e hour prior to the start of and at least one per hour during a reductio in the Reactor Coolant System boron co entration by either:

a.

Verifying or least one reactor coolant pump

b.

Verif ing that at least one RHR pump is in operation a d SUR ying ~ 3000 gpm through the reactor coolant system.

~

SALEM - UNIT 1 3/4 1-4

(

8909210159 890911

~

PF.'DR ADOCK 05000272 I

PDC i

REFUELI~IG OPERATIONS COOLANT C.IRCULATIOM LI~ITING CONDITION FOR OPERATION 3.~.8. t 3.9.B At least one residual heat removal loop shall be in operation.

~PPLICABILITY: MODE 6.

ACTION:

a.

Wfth less than one residual heat removal loop in operation, except as provided in b. below, suspend all operations involvinp an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all containment penetrations providin9 direct access from the conta irment atmosphere to the outside.

atmosphere within ~ hours.

b.

The residual heat removal loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period durina'the perfonnance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot 119s.

c.

The provisf6ns of Specification 3.0.3 are not applicable

SURVEILLANCE REQUIREMENTS t/.9.8. I -

4.9.8 A residual heat removal loop shall be determined to be in operation and cfrculatina reactor coolant at a flow rate of !9 3000 ~pm at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SAL.EM-UNIT 1 3/4 9-8 Amendment No. 34

REFUELING OPERATIONS 3/4.9.8 RESIDUAL HEAT REMOVAL COOLANT CIRCULATION ALL WATER LEVELS LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal loop shall be in operation.

APPLICABILITY:

MODE 6.

ACTION:

a.

With less than one residual heat removal loop in operation, except as provided in b. below, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System.

Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

The residual heat removal loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.

c.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.8.1 A residual heat removal loop shall be determined to be in operation and circulating reactor coolant at a flow rate of greater than or eqYal to 3000 gpm at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SALEM - UNIT 2 3/4 9-8

ATTACHMENT 3

THIS PAGE LEFT INTENTIONALLY BLANK SALEM - UN IT 1 3/4 1-4

REFUELING OPERATIONS

COOLANT CIRCULATION LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal loop shall be in operation.

APPLICABILITY:

MODE 6.

ACTION:

a.

With less than one residual heat removal loop in operation, except as provided in b. below, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System.

Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

The residual heat removal loop may be removed from operation for up to one hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.

c.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.8.1 A residual heat removal loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SALEM - UNIT 1 3/4 9-8

REFUELING OPERATIONS

3/4.9.8 RESIDUAL HEAT REMOVAL COOLANT CIRCULATION ALL WATER LEVELS LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal loop shall be in operation.

APPLICABILITY:

MODE 6.

ACTION:

a.

With less than one residual heat removal loop in operation, except as provided in b. below, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System.

Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

The residual heat removal loop may be removed from operation for up to one hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.

c.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.8.l A residual heat removal loop shall be determined to be in operation and circulating reactor coolant least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SALEM - UNIT 2 3/4 9-8

ATTACHMENT 4

3/4.l REACTIVITY CONTROL SYST:MS BASES 3/4.1.1 30RATION cm;T~Ci..

3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDC~~ MA~G:N ensures that 1) the reactor can e

made subcritical froo a11 aperatin9 conciticr.s, 2.)

~he reac:ivit/

ra:is~e~:3 associated with ~ostulated accident conditions are ccntrollab1e wi n1n acc Qota~*Q **~~t-and

~i ~~Q rc*~-~r '*11',*1

~e ~-,*R-**R-~ -**==,*-~-M-,V

"'I*

111;11

i, I

J

l*'-'-'

n I

<i*C:

ilw::li**=-"

>u11

\\..1::.. _i_,

subcritical to preclude inadvertent criticality in the shutdown ccndi:icn.

SHUTDOWN MARGIN require~ents vary throughout core life as a funct~on cf fu.el de~letion, RCS boron :oncentration, and RCS T~.. ~*

The ~est 1 restricttve conaition cccLlrs at :c:.., 'Hi:ii 7~"'"' a: nc *1o:i: apera:~1*;

temperature, and is associated wi:h a pcstuT!~ed s:ea~ line trea~ a::'~e~:

and resu1t~ng uncontrolled RCS c:c~jo~n.

In the !na~ysis of t~is ac:ice~:,

a minimum sr.uT00 1

ri~i MP~~G~~-~ of 1.5~~.:k/k is ini:ia11y ~~~~ireC :: *::r.:"'":i the reactivity trar:sient..

..cc::r::!"n*;iy, the S~~7GC',.;:; ~1;.,;\\'.;Iii rer:::..;ir'="'"::-:

is based uoon tnis 1imitinc :cna1t1cn ana is consis:en: witn FS;;\\ sa~2:1 analysis assumptions.

With Tave:_ 200°F, the reactivity transients

~

resulting from a postulated ste~~ line break cooldown are minimal and a 1~ ~k/k shutdown margin prcvi~es adeqJate ~rote::i:~.

0 ! LUT: C~~

3/4.l.1.4 MODERATOR TEMPERATURE COEFFICI~NT (M7C)

The limitations on MTC are provided to ensure that the value of this *coefficient remains within the limiting ccnCition assu~2d in

~~=

accident and transient analyses.

='

SALEM - UN IT 1 B 3/4 1-1 Amendment No. 15

A\\)D£D

. REFUELING OPERATIONS BASES 3/4.9.6 MANIPULATOR CRANE The OPERABILITY requirements for the manipulator cranes ensure that:

1) manipulator cranes will be used for movement of control rods and fuel assemblies, 2) each crane has sufficient load capacity to lift a control rod or fuel assembly, and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING The restriction on movement of loads in excess of the nominal weight of a. fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array.

This assumption is consistent with the activity release assumed in the accident analyses.

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirements that at least one residual heat removal loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140°F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the -----

effects of a boron dilution incident and prevent boron stratification. f Flow limitations are specified in plant procedures, with the design basis documented in the Salem UFSAR.

These flow limitations address the concerns related to vortexing and air entrapment in the Residual Heat Removal system, and provide operational flexibility by adiusting the flow limitations based on time after shutdown. I The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability.

The provisions of Sections 3.4.1.4 and 3.9.8.2 [paragraph (b) of footnote (*)] which permit one service water header to be out of service, are based on the following:

1.

The period of time during which plant operations rely upon the provisions of this footnote shall be limited to a cumulative 45 days for any single outage, and

2.

The Gas Turbine shall be operable, as a backup to the diesel generators, in the event of a loss of offsite power, to supply the applicable loads.

The basis for OPERABILITY is one successful startup of the Gas Turbine no more than 14 days prior to the beginning of the Unit outage.

With the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling.

Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

SALEM - UNIT 1 B 3/4 9-2 Amendment No. 72

'w ADDED REFUELING OPERATIONS BASES 3/4,9.6 MANIPULATOR CRANE The OPERABILITY requirements for the manipulator cranes ensure that:

1) manipulator cranes will be used for movement of control rods and fuel assemblies, 2) each crane has sufficient load capacity to lift a control rod or fuel assembly, and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the accident analyses.

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirements that at least one residual heat removal loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140°F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the

~~~---

effects of a boron dilution incident and prevent boron stratification. I Flow limitations are specified in plant procedures, with the design basis documented in the Salem UFSAR.

These flow limitations address the concerns related to vortexing and air entrapment in the Residual Heat Removal system, and provide operational flexibility by adjusting the flow limitations based on time after shutdown. I The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability.

The provisions of Sections 3.4.1.4 and 3.9.8.2 [paragraph (b) of footnote (*)] which permit one service water header to be out of service, are based on the following:

1.

The period of time during which plant operations rely upon the provisions of this footnote shall be limited to a cumulative 45 days for any single outage, and

2.

The Gas Turbine shall be operable, as a backup to the diesel generators, in the event of a loss of offsite power, to supply the applicable loads.

The basis for OPERABILITY is one successful startup of the Gas Turbine no more than 14 days prior to the beginning of the Unit outage.

With the reactor vessel head removed and 23 feet of water. above the reactor pressure vessel flange, a large heat sink is available for core cooling.

Thus, in the.event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

SALEM - UNIT 2 B 3/4 9-2 Amendment No. 46

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.l BORATION CONTROL 3/4.1.1.l and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS Tavg.

The most restrictive condition occurs at EOL, with Tavg at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown.

In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.6% _k/k is initially required to control the reactivity transient.

Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions.

With Tavg_ 200°F, the reactivity transients resulting from a postulated steam line break cooldown are minimal and a 1%

_k/k shutdown margin provides adequate protection.

3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC)

The limitations on MTC are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the accident and transient analyses.

SALEM - UNIT 1 B 3/4 1-1 Amendment No. 16

REFUELING OPERATIONS BASES 3/4.9.6 MANIPULATOR CRANE The OPERABILITY requirements for the manipulator cranes ensure that:

1) manipulator cranes will be used for movement of control rods and fuel assemblies, 2) each crane has sufficient load capacity to lift a control rod or fuel assembly, and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array.

This assumption is consistent with the activity release assumed in the accident analyses.

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirements that at least one residual heat removal loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140°F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification..

Flow limitations are specified in plant procedures, with the design basis documented in the Salem UFSAR.

These flow limitations address the concerns related to vortexing and air entrapment in the Residual Heat Removal system, and provide operational flexibility by adjusting the flow limitations based on time after shutdown.

The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability.

The provisions of Sections 3.4.1.4 and 3.9.8.2 [paragraph (b) of footnote (*)] which permit one service water header to be out of service, are based on the following:

1.

The period of time during which plant operations rely upon the provisions of this footnote shall be limited to a cumulative 45 days for any single outage, and

2.

The Gas Turbine shall be operable, as a backup to the diesel generators, in the event of a loss of offsite power, to supply the applicable loads.

The basis for OPERABILITY is one successful startup of the Gas Turbine no more than 14 days prior to the beginning of the Unit outage.

With the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling.

Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

SALEM - UNIT 1 B 3/4 9-2 Amendment No. 72

REFUELING OPERATIONS BASES 3/4.9.6 MANIPULATOR CRANE The OPERABILITY requirements for the manipulator cranes ensure that:

1) manipulator cranes will be used for movement of control rods and fuel assemblies, 2) each crane has sufficient load capacity to lift a control rod or fuel assembly, and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the accident analyses.

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirements that at least one residual heat removal loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140°F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification.

Flow limitations are specified in plant procedures, with the design basis documented in the Salem UFSAR.

These flow limitations address the concerns related to vortexing and air entrapment in the Residual Heat Removal system, and provide operational flexibility by adjusting the flow limitations based on time after shutdown.

The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability.

The provisions of Sections 3.4.1.4 and 3.9.8.2 [paragraph (b) of footnote (*)] which permit one service water header to be out of service, are based on the following:

1.

The period of time during.which plant operations rely upon the provisions of this footnote shall be limited to a cumulative 45 days for any single outage, and

2.

The Gas Turbine shall be operable, as a backup to the diesel generators, in the event of a loss of offsite power, to supply the applicable loads.

The basis for OPERABILITY is one successful startup of the Gas Turbine no more than 14 days prior to the beginning of the Unit outage.

With the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling.

Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

SALEM - UNIT 2 B 3/4 9-2 Amendment No. 46

ATTACHMENT 5 REVISICN:

SEPl'EMBER 16 REV 7 SAUM NUCU'AR GEm:RATm:; STATICN UNITS 1 AND 2

~

RHPS MID-IDOP OPERATICN IN REFERENCE 'IO NRC GmERIC IE'lTER 87-12 SEPl'EMBER 1987

'!HIS REVISICN IS '!HE RmJir1' OF

~

z.moE roRD<<; A SEPl'EMBER 15, 1987 ~

IN PITl'SElllGi PA.

ATl'ENDED BY:

c. IASl<ARI (PSE&G)

G. RCGGIO (PSE&G)

R.CMAC~)

P. MC HAIE (WE.S'l'lNGEIJJ)

E. FRANl'Z (WES'l'I?GIXJSE)

R. OFS'IUN (WESTI?GDJSE)

A. LIN (WESTINQDJSE)

REFERENCE:

Westin3hcuse Ietter NISO/SIM-87-396

TABIE OF CDNI'ENI'S

Section Title Page 1.0 INmOIXJCI'ION 1

2.0 RERJRr SCDPE 1

3.0

'lHERMAL HYDRAULIC.ANALYSIS 2

3.1 Description of Task 1 2

3.2 Description of Analysis Methods arXl 4

Assunptions 3.2-1 Operational Concems 4

3.2-2 Description of Analysis Model 5

3.2-3 Description of Input arXl Assunptions 6

3.3 Detailed Analysis

10.

3.3-1 case 1 - Base case Analysis 10 3.3-2 case 2 - Sensitivity t.o SG Condensation 16 3.3-3 case 3 - Analysis with 3/4 Indl vent 20 FlCM Path 3.3-4 case 4..; Analysis With 3/4 Indl Liquid 26 FlCM Path 3.3-5 case 5 - Analysis With Iarge Vent 32 FlCM Path 3.4 SUmmal:y of Results 38 3.4-1 RCS Heatup arXl Time t.o Saturation 38 3.4-2 Core uncxwery

38 3.4-3 RCS Pressurization Rate 38 3.5 conclusions 42 3.6 Referel"DeS 43 i

TABIE OF CDNI'ENI'S section Title

4. 0 TASK 2 -

RADIOI.DGICAL CDNSEXJ)ENCES 4.1 4.2 4.3 4.4 Description Of Task 2 Task 2 Assunptions Arrl Bases Task 2 Activity Concentrations_

In '!he Containment A1::rOC>Sphere Task 2 Off Site DJses

5. 0 TASK 3 - ASSESSMENI' OF VORI'EXING AND 5.1 5.2 AIR ENI'RAINMENI' Description of Task 3 Task 3 Conclusions

6. 0. TASK 4 -

DEI'ERMINATION OF VORl'EX LEVEL 6.1 6.2 6.3 6.4 6.5 Description of Task 4 Task 4 Conclusions Boron Dilution Concems.ASsociated With Reduced RHRS Flow Technical Specification CllanJes FSAR Cllanges ii 43 43 44 45 46 48 48 48 53 53 54 60 61 64

Section APPENDICES A

B c

D I.CSS OF RESmJAL HFAT REMJVAL (RHR)

WHII.E 'IEE RFACIOR CXX>I.ANI' SYSTEM (RCS)

IS PARrIALLY FilIBD (GENERIC LEITER 87-12)

SAUM RHRS MID-LOOP NRC RESFONSES WESTINGHOUSE LEITER NISD/SIM-87-396 MARKED UP TECliNICAL SPECIFICATIONS AND ACCXl-fi>ANYlNG 10CFRSO. 59 E.VAIIJATIONS AND

SIGNIFICANI' HAZZAROO OJNSIDERATION ANALYSIS iii A-1 B-1 C-1 D-1

LISI' OF TABI.ES TABIE

Title Page 3.3-1 TlJw1E TABIE OF E.VENTS, ~

CASE 1 -

12 mrACI' RCS, NO SG CXlNDENSATION 3.3-3

TlJw1E TABIE OF E.vm1S, CASE 3 - 3/ 4 maI 21 VEN!' m VAroR.RmIOO, NO. SG CX>NDENSATION 3.3-4 TrnE TABIE OF E.VENTS, CASE 4 - 3/ 4 maI 27 VENT m LIQUID RmIOO, NO SG CDNDENSA.TION 3.3-5 TIME TABIE OF E.VENTS, CASE 5 -

16 maI 33

SG :HANWAY VENT, NO SG CDNDENSATION 4.3-1 CDNrAINMENI' A<:nVI!I!'f 46 4.4-1 SITE EDJNDARY D:6ES 45 6.2-1.

'lUrAL RHR FI.CM VERSUS. MID-I.OOP 53 INITIATION TIME iv

LIST OF FIGURES FIGURE Title Page 3.2-3-1 DECAY HEAT FC.WER VERSUS TIME AFTER 9

RFACIDR SHtJI'Ix:MN 3.3-1-1 RCS 'IOrAL AND a:MroNENT PRESSURES 13

- CASE 1 3.3-1-2

~

AND VAroR :ROOIOO TEMPERA'IURES 14

- CASE 1 3.3-1-3 I.OOER :ROOIOO VOIIJME AND CDLI.APSED VOIIJME 15

- CASE 1 3.3-2-1 RCS PRESSURE ClliPARISON 18 3.3-2-2 VAEOR/AIR :ROOIOO TEMPERA'lURE cx:MPARISON 19 3.3-3-1 RCS 'IorAL AND a:MroNENT PRESSURES 22

- CASE 3 3.3-3-2

~

AND VAroR :ROOIOO TEMPERA'IURES 23

- CASE 3 3.3-3-3 I.OOER :ROOIOO VOIIJME AND CDLI.APSED VOIIJME 24

- CASE 3 3.3-3-4 UPPER :ROOIOO VEN!' P.MH FLCMRATE - CASE 3 25 3.3-4-1 RCS 'IOrAL AND a:MroNENT PRESSURES - CASE 4 28 3.3-4-2

~

AND VAroR :ROOIOO 'l'EMP.ERAni"RE.S 29

- CASE 4 3.3-4-3 I.OOER :ROOIOO VOIIJME AND CDLI.APSED VOIIJME 30

- CASE 4 3.3-4-4 I.avER :ROOIOO VEN!' P.MH FI.a-mATE -

CASE 4 31 3.3-5-1 RCS 'lOI'AL AND a:MroNENT PRESSURES - CASE 5 34 3.3-5-2 MIX'IURE AND VAroR mx;roo* TEMPERAWRES

.35

- CASE 5 3.3-5-3 I.OOER :ROOIOO VOIIJME. AND CDLI.APSED VOIIJME 36

- CASE 5 v

I.J:ST OF FIGURES FIGURE 3.3-5-4 UPPER :REXiIOO' VF.NI' PA'IH FI.CMRATE - CASE 5 37 3.4-1-1 HFA'IUP RATE FOR I.OSS OF RHR CXlOLING IXJRING 40 MIIrIOOP OPERATIOO' 3.4-1-2 TIME 'IO SAWRATIOO' FOR I.OSS OF RHR CXlOL!NG 41 IlJRING MII>-IOOP OPERATION 5.2-1 SAUM RHRS 51 6.2-1

'lUI'AL RHR FI.CM VERSUS MII>-IOOP 56 INITIATION TIME 6.2-2 Ra; WATER LEVEL VERSUS 'IOrAI..

57 RHR SUCI'IOO' FI.CM vi

WES'l'INGHaJSE ENGINEERING SERVICES REroRI' FUR SAllM NUCIEAR GENERATING STATION UNI'IS 1 AND 2 CDNCERNING RHRS MID-I.OOP OPERATION 1.0 mrnorucrrON Operation of the Residual Heat *Rerroval. (RHR) System with the Reactor Coolant System partially drained has been an imustI.y concern for nany years since this mxle of operation places the plant in a corrlition highly susceptible to the loss of RHR ftmction.

'Ihe U. S. Nuclear Regulatory Commission (NRC) Generic I.etter 87-12 was issued on this subject pursuant to 10CFRS0.54 (f) an:l was prompted by the April 10, 1987 Diablo Canyon loss-of-RHRS event. '!he principal concerns of the NRC focus on whether the RHR System design mee~ the licensirg basis of the plant for this noie of operation am whether this noie of operation can lead to resultant unanalyzed events that cc:W.d inpact safety. Generic I.etter 87-12 is provided.as an attachment to this report in Apperxilx A.

2. 0 REroRI' SCDPE

'Ibis report is prepared in response to Public SeJ:Vice Electric am Gas C.ompany's responSe to Westinghouse I.etter NISD/SIM-87-390 which is provided as an attachment to this report in Apperxilx B. * 'Ihe report provides West~ analyses specific to the Salem Nuclear Generating station Units 1 arrl 2 to _support response to Generic I.etter 87-12. '!he.

work scope as defined in the technical description contained in Apperxilx B lists the following five tasks to be ccmpleted unier this engineering services authorization: 1) 'Ihermal Hydraulic Analysis, 2) Radiological Con.sequences Evaluation, 3) Assessment of Vortexing an:l Air Entrai.riment,

4) Detennination of Vortex Level am 5) Report Dx:umentation. 'Ibis report, which was coordinated by the Plant & Systems Evaluation Licensing 1

1, (PSEL) group of the Westinjlouse Nuclear Safety Deparbtent, constitutes completion of task 5.

Tasks 1 through 4 are di sa.JSSed in the follc:7Ning sections am are the result of inp.l't provided by several functional groups of the Westin3house Nuclear Safety an1 Systems Engineering Departments.

3. 0 TASK 1 - ~

lMEAIJLIC ANAll'SIS 3.1 Description of Task 1 Follc::Mirg the April 10, 1987 loss of RHR event at Diablo canyon, the NRC issued Generic letter 87-12, "Loss of Residual Heat Rem:lval (RHR) While the Reactor Coolant System (RCS) is Partially Filled." Item s of this letter requested PWR plant licensees to provide a summary description of plant procedures for RCS draWam an1 operation in the partially fllled corxlition.

'!he response* is to include:

- '!he analytic basis used for the procedures developnent

- T:reat:Jnent of the drain:iam of the RCS

- Treanent of air entrairment an1 de-entrainment

- T:reat:Jnent of boiling in the core with am without RCS pressure boun1aJ:y integrity

- calculations of awroxllnate time fran loss of RHR to core damage

- Level differences in the RCS an::l the effect upon instrumentation Wications

- T:reatJnent of air in the RCS/RHR

- Treatment of vortexing at the connection of the RHR suction line(s) to the RCS of Generic letter 87-12 specifically mentioned several topics that need to be addressed am urrlerstood.

'lhese irx::lude:

2

- Unexpected RCS pressurization due to air in the RCS

- Water loss through openin:J in the cold leg

- RCS water level instrumentation uncertainties

- VortexinJ an::1 air inJestion fran the RCS into the RHR suction With the loss of RHR, the only m::xle of energy :rerraval. is corxiensation in the steam generators. In the absence of air or other non-oorrlensible gases, the primacy system will pressurize until sufficient terrperature difference exists between primary an:l secol'Dal:y to drive the reflux comensation process. Nonnally the terrperature difference required is small, so the prilnary pressure will essentially equal the secorrlary pressure. Recove.cy from this corxlition shc:W.d be relatively sin'ple, an::1 core uncovecy an::1 damage is mll.ikely.

'!he presence of non-con:iensible gases such as air in the loops ~licates the picture considerably. 'Ihis air must be displaced by the steam before reflux corxiensation can take place.

Depending on the eXtent of this displacement, RCS pressurization may be significant. It is likely that fairly complex 2 or 3-di.nen.sional flow patterns will exist 'Which substantially reduce the pressurization.

Despite the carplex nature of this prci:>lem, it is i;:ossible to detennine m:st of the paraneters of interest usin;J a sin'ple but conservative boiloff ncdel with non-con:iensibles.

'Ibis m::xlel will provide a macroscopic view of the RCS arrl, with appropriate mxielin;J asstmptions, provide a conservative est.llnate of.~ pressure, tenperature, arxl mass inventoz:y dUrin;J tne boiloff arrl ventin;J processes.

Sane :minllmlm rnmt of RCS to SG heat transfer can also be included to realistically boun:i the RCS pressure rise.

'!he analyses presented in this section of the report describe various loss of RHR ooolin;J scenarios for mid-:-loop operation for Salem Units 1 arxl 2

3

'lhe results of these analyses-can then be used to validate the plant procedures for RCS drairx:lown arxloperation in the partially filled con:iltion.

3. 2 Description of Analvsis Methods arrl.Assumptions

'!his section describes the analysis concerns, m:xiel description, arxl input assurrptions used for the mid-loop operation analysis.

3.2-1 Operational a:mcerns

'!he analysis presented in this report is interrled to provide i.np.It to ai:rl support the operation procedures for drai.mown arxl mid-loop operation for the Salem Generatin;J station. Several analysis concernS related to the loss of RHR.coolirg event durirg mid-loop operation ha'Ve been identified am are summarized below.

First, during mid-loop operation, the core exit thenrocouples may be disconnected in preparation for upper head :rerocwal. If the core exit thernDoouples are not functional, a direct irx:lication of RCS te.nperature would not be readily available if forced RHR flow is lost. A conservative estimate of the RCS heatup rate am time to saturation is therefore important to know to detennine when core lx>iling will begin arxl for. the tinrln;J of subsequent recovery actions.

Secarrll.y, it is also important to know the RCS pressurization transient that follC7w'S the initial heatup to lx>iling. If pressure remains low for an exten:Ied period of time (e.g., less than 25 psig), it would be possible to increase RCS inventory by gravity feed fran the I&s.l' (this m:xie of recovery was used at Diablo canyon)

If the pressure is higher tut the makeup requirements are relatively low (on the order of 100 gpn or less),

one charging p1II!p coold be used to increase RCS inventory. Additional 4

high-pressure injection 'WOUld be required for increased makeup requirem:mts. If the heatup am pressurization transient is allowed to cxmtinue further, the RHR cut-in corrlitions (350 F, 375 psig) am design corrlitions ( 4~0 F, 600 psig) could be exceeded. It 'W'C:Uld then be rieoessai:y to use an alternate noie of exx>ling for interim or lorg tenn J:eCXNery.

Finally, the RCS boiloff rate am time to core uncovery are i.np::>rtant to krx:lw in o:rder that core damage be prevented. Additional radiological concerns are introduced if core tmCOVery is allowed to occur.

3. 2-2 Description of Analysis Model

'!he cx:mpiter m:xiel used to analyse the Reactor Coolant System response during the mid-loop event is a single node non-equilibrium m:xiel.

'!be node is divided into an~

region am a lc:Mer.region.

'!he~* region contains a mixtUre of steam am a non-cOrrlensible gas (air or nitrogen).

'!he larer region contains only water in a liquid or a ~l boiling state.

Water mass is transferred frcan the lc:Mer region to the ~

usirg a constant bubble rise m:xiel.

Water mass is transferred frcan the upper region into the lc:Mer region through a droplet fall or a oorrlensation m:xiel.

A sinple orifice flCM m:xiel allCJl.tlS one flCM path into or out of the upper region. It either tranports the. non-con:iensible/steam mixture rut or draws ch:y containment air back into the upper region.

A similar orifice m:xiel allCJl.tlS a drain flCM path to be defined for the lower region. It only alla.i.'S transport cut of the lower region whenever the vessel pressure is greater than the c::utside reference pressure.

5

A time depen:ient liquid flOW' ncdel is included to si.nUllate flooding of the reactor vessel with cold water by startin;J a i;:urnp or openirg a valve to the msr. * '!his n'Ddel feeds only the 1C'1Ne.r region.

Decay heat is nmeled as a time varying heat inpit added directly to the lower region water.

stzuctural heat capacity is m:x:Jeled in both the uwer arrl the lower regions. In eadl region, mlltiple heat sinks may be JOOdeled with different heat transfer coefficients arrl heat capacitances.

'1hese are also used in the uwer region as corrlensin;J surfaces.

'lhe fluid node calculations use a 6 equation JOOdel.

One mass. conservation equation is defined for each of the 3 components:

upper region water vapor; upper region non-comensi.ble; arrl l~

region water.

One energy equation is solved for the total enthalpy of each region. '!his is coupled to a differential equation to solve for the total pressure of the node.

A set of auxiliary calculations are then perfonned to solve for the irrlividual carponent enthalpies arrl for the component partial pressures in the upper region.

'lhe water properties (specific volumes am temperatures) are obtained using piece-wise linear steam tables, exten:ied to lOW' pressures.

Non-comensi.bles are assumed to be ideal gases.

3. 2-3 Description of :mp.rt: am Assunptions Conservative IDp.rt:. aSsunptions have been made whidl maximize the core heatup rate arrl presslirization arrl minilnize the time to boiling arrl core urxxwery..A rumbp..r of these assurrptions are explained belOW'.

1. At the time of mid-loop operation, the RCS temperature is 140 Fam the 6

time after reactor shut.dO'tm is 72 hclirs. 'lhis is the shortest decay time and highest temperature awlicable to the mid-loop cperation c::onntions.

2. At 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . after shutdO'tm, the decay heat power is 0.41% of full :paver (102% of 3411 MWt).

'llrls corresporrls to a decay energy of 14.265 MWt

( 48. 67 MBIUjhr) *

'!be decay heat power was detennined usirg the ANSI/ANS-5.1-1979 decay heat starrl.ard ard includes 2-sigma uncertainty plus conse:rvative estimates for heavy element decay ard fission product absorption. A high average bJmup was also assumed in these calo.llations (core average irradiation time of 800 days or 30,000 MWD/MIU).

As a cross-check on the results, ~

ANSI/ANS-5.1-1979 values were then cxxnpared to a Westin#louse 3-:region core calo.llation an::l fc:urn to be within 2% over the rarge of decay times of interest.

(For the 3-:region core calo.llations, the irradiation tilnes were 333, 667, an::l 1000 days.)

'!be higher of the two decay heat pc7.N'erS was then used in the analys~.

Figure 3.2-3-1 shCMS the decay heat pcYWer as a function of time after reactor shutdc:Mn *

3. '!he water volume assumed for the sin;le ncxle core boiloff m::rlel is oc:mprised of the water in the core ard upper plernnn (to the middle of the hot legs) plus one third of the water in the bottom half of the hot legs.

'!he water volume assumed for the analysis is 1260 cubic feet. Water in the cold legs, downcamer, lower plenum, ard rema.imer of the hot legs is neglected in the initial heatup calo.llation even tbaigh sane of this water

'WOul.d eventually heat. For the Salem Plant durirg mid-loop cperation, the water level is nonnally one foot above centerline an::l.there is an alann if the water level drops to 6" alxwe centerline. 'Ihe additional hot leg ard uwer plenum volume to the alann setpoint which was conse:rvatively neglected in the present analysis is raighly 200 cubic-feet, i.e., m:>re than 15% of the liquid volume assunm3d in the calo.llations.

4. '!be heat capacities an::l overall heat transfer coefficients for the fuel, metal structure in the uwer internals, plus water an::l structure in the barrel-baffle region has been i.rci.uded in the heatup of the lower 7

(mixture) region. Similar irp.rt for the upper internals, upper support plate, vessel metal, hot leg

an::l nozzles, arrl steam generator inlet has been specified for the ~

(vapor) region.

s. '!he initial steanv'gas volune used in the sirgle oode boiloff m::xiel was assumed to consist of the renairrler of the hot leg an::l upper plenum, the tg;>er head, the SG inlet plern.nn, an::l only one half of the SG tube volumes.

'!he bottan of the surge line is initially covered with water, so initially the pressurizer volune is not in direct eontact with the rest of the RCS (except possibly to the upper head through the PRl' if the upper head is vented to the PRl' instead of air). F\\Jrthenrore, the initial RCS pressurization will cause water to e>epan:i into the surge line without significant dlarge in the steam volune. It is co:n.serJative to neglect the pressurizer volume oc:mpletely for puzposes of determi.nin; a high RCS.

pressure.

'!he down side of the SG tubes has also been neglected. If it is postulated that the non-corxiensible gasses collect at the top of the SG tubes am restrict the flow of steam, then the SG volune considered should be sanewhat less than the whole volune.

One half the total SG volume will certainly be conservative.

6. In m:>St of the. analysis cases, heat transfer to the secx::u'X3acy is not included to simulate the dJ:Y layup situation. steam corxiensation is.

minimized to conservatively maximize the pressurization.

An SG corxiensation sensitivity study is also co:rductei to derronstrate the effects of havirg IIDre realistic secorxiary heat transfer

8

'° 1.2 r

I i

i I I

-*... **o-I i

! i I'! '::'I I..

I I

, 11

*r-~*-i-: -.~~-. I,

~-

~

1--*---,

: I i11!i ;;,;.. " ::*: ij 1

1!i !Iii !

1jl::!:,::::/

1 1

f 't !j. :,.

Not~~

I I

. 111* ' 'I"" "

' I I

I I

j_

I I

I t

-r-1-,-*I- -n 1 : -- :-:- - - * !)\\---* -~* rn-r:- :mm: ~._ ~.. -,1 Tnl 1-

~ * -* * -

I I

I

' i' I

!" 111:: i I' I i 11' 11;*

I I I

11

'1

2.

I, 1 I I

I r I,*

1

, 11 I

i I I *,1 1

I I '

- l* 1:*

I I

i i

~

I

~

: I l

' I I

l ;

I I I 11 1 l ' :

! I '

I I

'"I: I\\': I ',1 1 I'" *11*

I

fr:* *j.I

'I: 1 1 *~!***

i I 1

r I

~1.o !,

!li':_1 l:l; 1 111:.,,IN.lt!:i,!!i::*;.ii..

111 :!

QI i I I I : "

11 1

1 f

: i I*: '1 ::111N.1 I I

1*

11 111

1 1

.: ! "1 i.

I 11' *1*.

I

):

I ! '

I 11"

I 11 J, '

jr 11

1'-1..

I i I ;I j'*

I j I 1

I o

i 1 I 1 1 rl 1 I ! *'ii* !i 1I Ii 1:x: 1.111:

I ! 1 i *

, 1!

I Curve based on maximum of Westinghouse 3-Region Core Calculation (plus Uncertainties)

ANS-5.1-1979 (plus 2-sigma Uncertainty) for High Average Burnup (approx. 30,000 MWD/MTU) l--

r-i r-i

I

~

0

.µ s:::

Q) 0 M

QI

~ -

M QI 31: 0

~

.µ RS QI

i::

. >t RS 0

QI 0

~

~-11:~i:-1--

J r1ll1':,;:::::r1:1 ii ~'F-::--:1:~*1J..

1--11 1111111nmimr11111rl'1i-11;11:11:1L

1 1**'1. 111imnr11:**-*:-

11*1-i I i 11 1*

1 il1 I 1i i !1. I I

i*1:1:K i :1... I I

!. ! 11 *1 l,i !:!I:::* I, i II i: :: :j11 ;::. :,1....

" *It

_, I:!

I i I I illi "' 'i' i

.1 K

'I' I Ii! ! i I 1! 'I :1!'

I iii: 1:.,: ::',,,

1 I ! :: !Ii:!,!

o.a. *1-1111 i

'1

!I I '1 l'1i ;t\\: i ; :11\\i II j

\\'1'1

111,: ii:\\ *1i.~N:

I 1 j !j \\

II I \\\\!i, :\\ :;

ii'~

fl '1 :: !\\\\; ;*;: ;,

j 1'1' iii:' \\iii.

- I** I -,. I I

' i i 't I i ;; : : : :

ii

ii! I I it 1*

'11 *1 i : ii! : I!'. ::

j

!I I II I ti ii i': i '. ! : ' ! I i

.,, ; I!! : l ! : i ! !

II 11 I t I i i ! '!

I. I' rlj I II r**.: "':

'.'"I'""'!"

1 ':"':"*.I', *,I,I'll ':i,*,

~------*

I j j, 1'1!

,.:,,1 I

I

j.

.1 T

I..

I I

I!*!

I I

I i 1 1 1 -1~1*

i... ~T "i::-r7.

1,: -~rn-1

'"'"1 i i-~ !IT! trl:

1,..,.;.

.--,.--1 ".

I 'I I 11 1 1 : : ~ ' ii *11 : 1 :. t -,1 r.,

1

.. 1.

1

'I 1 1 1

1 :

1

: i 1

I I

I l

I

' l

~

t I I i I

f

. : ~ I I l :..

I

.. I t

I l '

l I

I I I I

l t I : I I '

1' 'ti. i : I!! i ! I 1[ ; : I:

11:: :::I!

Ii 1' Ii,:: [ ;j!. ::I:

~: "

i 1*.

. m ' 'i Iii II li!ll 1;11:* r. I : :1*; 1'!:: :1;>1':.. *I_~*-~,-, *I. !

I I!,

!, i ',!',: :::

I '

. ' I

. ! I.

j '

I I'

~

x_1 _I

1 I

~

I I '

I 11:

': 11 I

,, 1 :

1 I r 1 1: '" 1: *

'; : 11. 11 * :*::

I

ix 11 11 1.!1 r ii 1 Iii!*;:*':

--

1** *** 1 1 '

lr' 11' 0.6~:.-1~1.-1~1'1-i~i h:~~1:-t--:-*!~l!;-,-,-~,~:;-.. ~1~i~lir.~il7!-1-,-I,.,+---+--+-.~. -+1.~i-1~. ~,-l-T-+l~I! H1-.-~;l~'"T.......;ii...........

i.~'1ml1r

.. h1~1,+11rh-i11+1~i. ~'.'1~,;~;:_!..........

!i:~-i-11*~1~1 "1-++-1+-+--

llM! 41Tl-,+-'-"4;,~.,~.~~

1**1 I ;

i '.1', '1,. Ir' 11! !

'1* I r

111, ll!'i,,.

I '1*:

! I' >,I' 11.,:!.'.'.

., : 11:: !Ii:";;,;

>----;-!~:

!!~1'~l-r-l!... ~~C0:11:1 :!1i1:1:

!!;:i::;,

i I:

',,I; II !!i! 111i!i~i""'11!(ill.'!i;1': I i: 11:; :i:',!',

1:1 iii:: ;i:ll 1i:'.ii::

11

'1\\i

!il!!li'::::_

i* *l*1*

i'1'1 1

1 1 1iliiil li~ll:ili 1 \\ii:!li<-=-=:..

--1~1 i 1

11ii!l!::/l':

1 1 ';1li!

11 :!

1 t:'iil:'

i 1!i::::i* *ii l1*ll

i 1ri:il!

!i:!*~...:_!11 1

ll!*i!l1i::

0.4~:11~l~i~!,hi~,,~;~111~,11*r.!~-11,:M1i,_,~:iii~lffi~*,il~i~,.!~1~:~::!+!':*:1~.~.,l~H11 +1++r+rn1*1811ti,1~:!*l:~lliHi~i~:~111H14i!~::i:~~+.~+~

__ + __ +1*M14-~11~;~; ~!::~:~ilii~i,~'1!~~~

" I I -.. : II 11 ' i. i!f1 i 111' ' I P:!,;I: "

I!

i ! I i;:1

! I' Ii ! I Ii 'I::, *t'I!

I I : ! f;. ! ! 1 I i lri I '1 I 'I

'I.

I

. I*

I t

' ' I

' I' I 1 *. I I : I '

I ' "

I ' I 'I I ;

' I I :

" I '

' i ' "

11 ! : i ! ! ! ; : ' !

~:

I I i i

! ! I I I ! I

I I ;. ! : ! I ' [ !.i '

I I.

I i :

I I :. : I ! :

I l I

i I 11 I

I!:' I'" I 11 I ' 11 I' I::'

fl

'I I

I I ' I

'I' I Ti':I '

[ 11 : : 11' "1* ~

I I I ~ ' '.

I I I '

- n :*i-

1*1*

Ii: 11:1;1ii;:11*1 11 1, ! i::;,:::1:-* *****.. :I*

I! !ii 11::: i i:: 1il1i 1'i',1: **

11

No.I..L.;!, :1:: :1:

If I!

I I 11 ii:! ::j 1111,1 I' *1 i' '!.*:*r I! ;il!i i:i! I I ii; ;II'.'!: ;'J'il,~~--

11* I 11'*1 i~.,_;.;j jll: :;i I

! I l i I

I 1 I

I ~ I ' ' 1 11 *

! ' I. :

~ ! '

1 !.. *....

i I ! : I : I : ! I. i I I ! I ' I

i. : 1' I.* ! ; **

' I i : i :

I. j I 0.2

1 i i ! Ml Ii ! i :111 i!< !i*:

'111 ! !i'1*i II Ii

,j,

1....

I i I 11:11 :l!I !Iii! i; i ::Ii:

'11*1 :: 'ii.

1-t I: i !

11*1: :11, I

' ! i :

I I i i ! I!,I *: : 11i; I 'I I

lj "i; 1il1 I !

II I,. ' 1li11,1* 'I : ! 1: i :.;,. 11.,i' i

.. *1 I., I lJrl I!

l!::il*:i.11 ll 1:! 1i 1* 1;,.,i:-

i',

i! li,.*;;,*' 1,i'"!ii :;:'"!

I I !i1!;*

i I:.

11 Ii! :I'll'!

I !ill! 11 I'. '11 :;'

11 ;'.,.'II ;m 1j::.J *:*"

I I I Ii:!

-:-n* -:-1.

1 1

1

,1111 1

1 11

111:1~,

1111 *11:ir1

11111-::

1.1 1:

1 1,

1 11T 1*:rr:

11:

1 1

1

,11 1

'!B 1

111-~1i::11 =--=--

1 11 1

11 1

1

rn: :*:~*--

. I i '

. "11 I 11 1;1 i,:. !

.J!' 1;: ill'

,ii I ! Iii; 1* '

I I I I! i111 11:1; i,

'Ii/..

11.*. :

1.. ":I*

1*

,**1 1*I

!l1'!'1!.!l 1 !.,

I

,11'::

0._

1 r

1

-~ ** :

1

, 1" ac,1 1 1

i 1

1 2

3 4

6 8

10 20 JO 40 60 80 100 200 300 Time After Reactor Shutdown (Hours)

Figure 3.2-3-1 Decay Heat Power Versu me After Reactor Shutdown

3.3 Detailed Analvsis Transient analysis for various cases of interest are presented in this section.* As explained previously, the analyses are based on decay heat at 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after shutdown (.41% of full pc:Mer) am a conservative (1260 cubic-feet) lower region mixture volune (similar to NUREXr1269 methcxiology) *

'Ihe :ncxrenclature for the various plots presented is as follows:

P-total, P-steam, P-air : Total, steam, am air (or N2) partial pressures (psia)

Upper am I..ower Temp :

Terrperatures in the lower (mixture) am upper (vapor) regions (F) lower am Collapse Vol :

I.ewer mixture region voltntie am equivalent volune if all voids were collapsed (cubic-feet)

II:M Vent Flow :

lower region (liquid) break flow Vent Flow :

. ( 11:xrV sec) uwer region (vapor) break flow

( ll::!tV sec) 3.3-1 Case 1 - Base Case Analysis In this analysis, no SG comensation is nmeled am the RCS is assumed to remain intact (no vent path) * '!his case can be used to est.llnate the maximum pressurization for

1oss of RHR cooling at mid-loop comitions, 10

with no water in the steam generators (dcy layup).

'!he time table of events for this case is given in Table 3.3-1. '!he pressure,* temperature, an:l lower region volume transients are presented in Figures 3.3-1-1 through 3.3-1-_3.

After cq:proxima:tely 9 minutes, the RCS lONer region volume ( ex>re an:l upper plenmn) reaches 212 F an:l the. RCS starts to pressurize. *By 18 minutes, the RCS pres.sure ~

25 psig.

Pressure. cxmtinues to increase to 100 psig at cq:proxbnately 30 minutes arrl exceeds 400 psig by one hour (Figure 3.3-1-1).

'!he lower region temperature increases alI!DSt linearly with ti.me at the rate of about 8 F/min tmtil saturation is reached (212 Fat 9 minutes).*

After this ti.me, same of the decay energy produces steam rut much of it still heats up the lower region saturata::i mixture. By 35 minutes, the RCS temperature reaches the RHR cut-in temperature of 350 F (Figure 3.3-1-2).

Since the RCS is assumed to be intact in this analysis, the lower region volumes increases due to the heatup swell (Figure 3.3-1-3).

11

Table 3.3-1 Time Table of Events

salent Loss of RHR Cooli.rg at Mid-I.oop Operation Base case 1 - Intact RCS, No SG Corrlensation Event Loss of RHR Cooli.rg at Mid-Loop Corrlitions Core am Upper Plemnn Tenperatures = 140 F RCS Pressure = 0 psig (14. 7 psia)

Core am~

Plemnn Tenperatures Reach 212 F RCS Pressurizes to 25 psig {39.7 psia)

Con:litions at 2000 secorrls (33. 3 minutes) :

Core Exit Tenperature = 338.8 F RCS Pressure = 101.6 psig {116.3 psia)

RCS Intact With Collapsed level AR;>rox:imately 3 Inches Above Mid-I.oop Core am ~

Plemnn Tenperatures Reach 350 F RCS Pressure Readies 375 psig {390 psia)

Erxi of Transient M::xleled for Base case 12 Time sec Cminl 0 (0) 547 (9.1) 1104 {18.4) 2000 {33.3) 2140 (35. 7) 3480 {58.0) 4000 (66.7)

~

t-4 c

I Q.

z c UJ UJ I

t:

Q.

l -

0....

I Q.

--P-TOTAL

--P - STEAM

- - - P - AIR 600.0 500.0 400.0 300.0 200.0 100.0 0.0 0.0

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

500.0 1000.0 1500.0 2000.0 2500.0 3000.0 3500.0 4000.0 TIME * (SECONDS)

Salem Mid-Loop No Vent. w/o Condensation Fiqure 3.3-1-1 RCS Total and Component Pressures - case 1

~

UJ I-a:

UJ

s 0

..J

~

a.

x UJ I-a:

UJ.

a.

-- UPPER TEMP

- - -* LOWER TEMP 500.0 450.0 400.0 3!50.0 300.0 250.0 200.0 150.0 100.0 0.0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 3500.0 4000.0 TIME (SECONDS)

Sale* Mid-Loop No Vent. w/o Condensation Fiqure 3.3-1-2 Mixture and Vapor Reqion Temperatures - Case 1

5 UJ UJ IL 0 u w 0 >

cc UJ

s 0 _.

1500.0 -- LONER VOLUME

COLLAPSE VOL 1400.0 1300.0 1200.0 1100.0 1000.0 900.0 800.0 700.0 600.0 500.0 0.0 Top of Core 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 3500.0 4000.0 TIME (SECONDS)

Sale*.Mid-Loop No Vent, w/o Condensation Pigure 3.3-1-3 Lower Region Vol\\ime and Collapsed Volume -,case 1

3.3-2 case 2 - Sensitivity to SG COn:iensation

'!he con:iensation heat transfer coefficient can vary between 10000 arrl 50 BIUjhr-ft2-F, depen:ling upon the amount of non-con:iensible gas present.

'!his section presents the results of two analyses which were perfonned to determine the effects of con:iensation on the RCS heatup arrl pressurization rates foll~ the loss of RHR dur~ mid-loop operation.

Both the overall heat transfer coefficient arrl heat transfer area are i:rrportant in dete.nninirg the allDl.D'lt of CX>n:iensa.tion.

As noted above, the heat transfer coefficient is sensitive to the am::Jlmt of non-con:iesible gas present.

As steam con:ienses on the ex>n:iensirg surface, a layer of non-Condensible gas is left behin:i.

As this layer builds up, the con:iensation rate degrades since steam must first diffuse through this layer of gas. 'Ihus, even a small am:x.mt of non-con:iensible gas in the system will cause the con:iensation rate to decrease significantly.

At !CM pressures, steam is less dense than air arrl will not penetrate into the downhill section of the SG tubes un:ler natural circulation con:iitions. '!his leaves approximately half of the SG tube area for con:iensation.

'!he effect of ex>ndensation in*reduc~ the pressurization of the RCS after the loss of RHR coolinJ was studied by performirg two analyses in which both the con:iensation heat transfer coefficient an1 area for ex>n:iensation were varied. For case 2a, a condensation heat transfer coefficient of 1000 BIUjhr-ft2-°F arrl 20% of the uphill SG area -were used. '!his

.resulted in an overall UA of 900 BIU/~°F which is believed to be

realistic if the steam generators are filled with subcooled.water.

For case 2b, a condensation heat transfer coefficient of 100 BIUjhr-ft2-°F an1 only 5% of the uphill SG area -were used~* 'Ihis resulted in a nore ex>nservative estimate for the overall UA of 100 BIU/~°F. Note, in both cases, all 4 steam generators -were assumed to be filled with subcooled water to the nonnal cperat~ level. '!his asst.nnption influences 16

the heat up rate calculation after saturation is reached and steady state oorrlensation is takin:J place. If less water were assumed, the heat up -

rate (and consequently the pressurization ;rate) dur.i.n; the steady state con:Iensation ~

would be higher.

Figures 3.3-2-1 and 3.2-2-2 present the p~*

and temperature cxxnparisons. '!he effect of even the rirln:inu.nn aircunt of oorrlensation heat transfer is to requce the RCS pressurization significantly. '!he RCS pressure 30 minutes after the loss of RHR went f:ran 80 psig in the base ca:se to 25 psig in case 2b and 8. psig in case 2a. *

'17

a. -*

L

J.,.,

L Q.

120.0 100.0 80.0 60.0

.-o.o 20.0 0.0 0.0 Ho Condensation small Condensation

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

Reasonable condensation

.-oo.o 800.0 1200.0 1600.0 2000.0 200. 0 600. 0 1000. 0 1 *00. 0 1800. 0 Ti me (seconds)

Salem Mid-Loop Condensation Sensitivity Figure 3.3-2-1 RCS Pressure comparison

L :l

~

ca L

cu

~-

cu...

L.....

c L

0 a.

m 350.0 300.0 250.0 200.0 150.0 100.0 0.0.

/

'/

/

small condensation Reasonable condensation

oo.o eoo.o 1200.0

1600.0 2000.0 200.0 600.0 1000.0 1*00.0 1800.0 Time (seconds)

Salem Mid-Loop Condensation Sensitivity Fiqure 3.3-2-2 vapor/Air Reqion Temperature comparison

3. 3-3 case 3 - Analysis With 3/ 4 Inch Vent Flow Path

'!he base case analysis of Section 3. 3-1. (no SG con:lensation, RCS intact) was repeated, this tllne with a 3/4 inch diameter break in the vapor space. 'Ihis is representative of the size of the reactor vessel head vent. It is also carparable to the size of a tygon tube (used for vessel level in:lication) that could potentially rupture due to the RCS pressurization.

'!he time table of events for this case is given in Table

3. 3-3. Parameters of interest are illustrated in Figures 3. 3-3-1 through 3.3-3-4.

Comparin;J Table 3.3-1 to 3.3-3 an:1 Figures 3.3-1-1 through 3.3-1-3 to

~igures 3.3-3-1 through 3.3-3-3, the small vapor space break has a slight impact on the transient.

As noted in Figure 3.3-3-4, the vent flow rate is on the order of 1.0 11:Jm/sec.

'!his is an order of magnitude smaller

. than the steam production due to decay heat. 'lhus, one (or several) vent paths of this size will not have a significant impact on the RCS heatup an:1 pressurization transient

20

Table 3.3-3 Tine Table of Events Salem loss of RHR Cool~ at Mid-I.oop ~tion case 3 - 3/ 4 Inch Vent in Vapor Region, No SG Comensation Event Loss of RHR Cooling at Mid-IDop Con:litions Core Exit Temperature = 140 F RCS Pressure = O psig (14. 7 psia) 3/4 Inch Diameter Vent Pa.th in Vapor Region Core arrl Upper Plenum Temperatures Reach 212 F RCS Pressurizes to 25 psig (39. 7 psia)

End of Transient Modeled Core Exit Temperature = 333.4 F RCS Pressure = 93.6 psig (108.3 psia)

RCS Has Small Vent FlOlirl, Collapsed Level is Approximately 3 Inches Above Mid-IDop (sane as case 1) 21 Tine sec Cminl 0 (0) 546 (9.1) 1124 (18.4) 2000 (33.3)

llJ N

a:

H <

a.

E ct UJ I-en I

a.

l-o I-I

a.

120.0 100.0 80.0 60.0 40.0 20.0 0.0

-- P -

TOTAL

--P -

STEAM

/

-- - - P - AIR

/

/"

/.

/*

/

~~

/

0. 0 400. 0 BOO. 0 1200. 0 1600. 0 2000. 0 200.0 600.0 1000.0 1400.0 1800.0 TIME (SECONDS)

Salem Mid-Loop 3/4 Inch* Vapor Vent Figure 3.3-3-1 RCS Total and_ Component Pressures - Case 3

n.

E

. UJ t-cc UJ 3:

0 _J N w

n.

E UJ t-cc UJ

n.

J

-- UPPER TEMP 340.0 320.0 300.0 280.0 260.0 240.0 220.0

/

200.0

/

180.0

/

160.0

/

140.0

-- - - LOWER TEMP

/

//

/

0.0 400.0 800.0

.1200.0 1600.0 2000.0 200.0 600.0 1000.0 1400.0 1800.0 TIME (SECONDS)

Salem Mid-Loop 3/4 Inch.Vapor Vent Figure 3.3-3-2 Mixture and Vapor Region Temperatures -

Case 3

-- LOWER VOLUME COLLAPSE VOL 1500.0 1400.0 1300.0

...J

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

L-----

0 >

1200.0 w

UJ Q..

1100.0 ct

...J

...J 0 u 1000.0 w

~

IV

l

...J 900.0 0 >

cc UJ 800.0 3:

0

...J 700.0 Top of Core 600.0 500.0

0. 0 400. 0 BOO. 0 1200. 0 1600. 0 2000. 0 200.0 600.0 1000.0 1400.0 1800.0 TIME (SECONDS)

Salem Mid-Loop 3/4 Inch.vapor Vent Figure 3.3-3-3 Lower Region Volume and Collapsed Volume - Case 3

-- UPR VENT FLO 1.B 1.6 1.4 1.2 u

QJ UJ E

1.0

.c

s 0.80 0.....

LL

.µ c

0.60 QJ 0.40 0.20 0.0

0. 0 400. 0 800. 0 1200. 0 1600. 0 2000. 0 200.0 600.0 1000.0 1400.0 1800.0 TIME (SECONDS)

Salem Mid-Loop 3/4 Inch.Vapor Vent Figure 3.3-3-4 Upper Region Vent Path Flowrate -

Case 3

3.3...:4 case 4 - Analysis With 3/4 Inch Liquid Flow Path

'!his scenario is the same as the previous case except the break is lcx::ated in the liquid region. '!his cx:W.d represent an RCS inventory loss through a drain valve or failure of the tygon tube at the low point connection.*

'!he initial break flow for this case is 5 n:mvsec (36 gpn) Wich represents the flow expected at a distance of 12 feet below mid-loop (this coincides with the bottan of the crossover leg). Irwentory loss was mxieled for the entire duration of the transient. It should be noted, however, that affu the level drains to the elevation corresporxiln::J to the top of the reactor coolant :p.mp 'Wir (5" above the bottan of the cold leg or about 9" below mid-loop),

the RCS inventozy loss for the core arrl upper plern.nn water could be expected to stop (a depletion of only a one to two hurrlred cubic-feet). '!he analysis also conservatively assumes that only the water in the core arrl upper plenum is lost arrl totally ignores drairxiown of the cold leg an:i downcammer (this over-estimates the break flow by aboUt a factor of two) *

'!he time table of events for this case is given in Table 3.3-4 arrl paraneters of interest are presented in Figures 3.3-4-1 through 3.3-4-4.

rue to the re::luction in inventory, the time_ to bolling is re::luced when compared to the* base case~ Additional. steam production is also predicted Wich in tum causes slighly higher pressurization than the base case.

Referring to Figure 3. 3-4-4, the inventory loss speeds up after saturation is reached due to the pressurization (for the SG corxiensation *cases, break flow wotild not increase as fast)

However, level is still nuch higher than the active fuel at the erxi of the transient (Figure 3.3-4-3). In view of the conservatisms noted above, the actual volurre at the erxi of the transient wculd be expected to stabilize at a nuch higher value. 'Ihus, the expected behavior for the 3/ 4 irK::h liquid break case wa.lld be silni.lar to the correspon:lirg case withrut the the break.

26

Table 3.3-4 Time Table of Events Salem I.oss of :RHR Coolin; at Mid-IDop Operation case 4 - 3/4 Irrh Vent in Liquid Region, No SG Corxiensation Event I.oss of :RHR Coolin:J at Mid-IDop Con1itions Core Erit Temperature = 140 F RCS Pressure = 0 psig (14. 7 psia) 3/4 Inch Li.quid FlCM Path (5 lJ::mVsec Break FlCM)

Core am~

Plern.nn Temperatures Reach 212 F RCS Pressurizes to 25 psig (39. 7 psia)

End of Transient Modeled Core Erit Temperature = 349.5 F RCS Pressure = 119.l psig (133.8 psia)

RCS Break FlCM = 23.7 ll::mVsec, Collapsed Ievel is More 'Ihan 3 Feet Above Top of Active F\\lel.

27 Tine sec Cminl 0 (0) 537 (9.0) 1079 (18.0) 2000 (33.3)

--P - TOTAL

.140.0

--P - STEAM

- - - P - AIR a:

120.0 M c I

CL 100.0 z c 80.0 t-en I

CL 60.0 N

Q)

..J c....

40.0 a....

I CL 20.0 0.0 0.0 400.0 800.0 1200.0 1600.0 2000.0 200.0 600.0 1000.0 1400.0 1800.0 TIME (SECONDS)

Sale* Mid-Loop 3/4 Inch Liquid Vent Figure 3.3-4-1 RCS Total and component Pressures - case 4

a.

x IA.I....

a:

IA.I

a 0.....

a.

x

'° IA.I....

a:

IA.I Q.

Q.

350.0 -- UPPER TEMP

- - - LONER TEMP 300.0 250.0

. 200.0 150.0

~-.J 100.0 0.0

'400.0 800.0 1200.0 1600.0 2000.0 200.0 600.0 1000.0 1400.0 1800.0 TIME (SECONDS)

Sale* Mid-Loop 3/4 Inch Liquid Vent Figure 3.3-4-2 Mixture and Vapor Region Temperatures - Case*4

-- LONER VOLUME

- - - COLLAPSE VOL 1500.0 t*oo.o 1300.0 0 >

1200.0 en a..

c( 1100.0

..J

..J r------... ~*~~

0 u 1000.0 x

w 0

900.0 0 >

a:

UJ 800.0

s 0 -'

700.0 Top of Core

~--

600.0.

500.0 o.o

oo.o eoo.o 1200.0 1600.0 2000.0 200.0 600.0 1000.0 1*00.0 1eoo.o TIME (SECONDS)

Sale* Mid-Loop 3/* Inch Liquid Vent Figure 3. 3-4-3 Lower Region vo.lume and Collapsed Volume -

Case 4

25.0 -- LOW VENT FLO 20.0 u

UI UJ

it 15.0 m _, -

z 0 _,

II.

~

10.0 w

c UI cc m

C3

_/~

5.0.

0.0 0.0 400.0 800.0 1200.0

1600.0 2000.0 200.0 600.0 1000.0 1400.0 1800.0 TIME (SECONDS)

Sale* Mid-Loop 3/4 Inch Liquid Vent Figure 3.3-4-4 Lower Reqion Vent Path Flowrate - case 4

I I, tl 3.3-5 case 5 - Analysis With I.arge vent Path F,'or this analysis, a large 16 inch diameter vent path is m:xieled in the vapor region.

'lb.is represents the opening of a SG manway prior to installation of the SG nozzle dam.

'!he time table of events for this case is presented in Table 3. 3-5 am paraimters of i.ntereSt are illustrated in

Figures 3.3-5-1 through 3.3-5-4.

Since the vent opening is very large, the total pressure is maintained at approxilnately one atmosphere for the duration of the transient (Figure 3.3-5-1). Note that after boiling occurs (again after 9 minutes), all the air is expelled am the steam partial pressure becc:me; the same as the total pressure.

As expected, the RCS boils at near a'btx:>spheric pressure am the temperature is maintained at an approxilnately contant value of 212 F (Figure 3. 3-5-2) *

'lhe vent flow approaches the constant boiloff rate

~g = 13.94 lbnVsec (Figure 3.3-5-4).

('lhe equivalent makeup flow requirement is approxilllately 100 gpm.)

rue to the nearly constant boiloff rate, the vol'l.IltVa also decreases at an approxilnately constant rate (Figure 3.3-5-3).

By one hour, the collapsed mixture level (for this conservative calculation) reaches the top of the active fuel. If cold leg am do;vncomer water above the top of the fuel is assumed to be available to replace some of the water being boiled away, the time to core uncovery would be extended further, by mre than 20 additional minutes. 'lhus, the expected time to core uncovery for this limiting case is expected to exceed one hour.

32

Table 3.3-5 Time Table of Events Salem loss of RHR o:>oli.n:; at Mid-I..oop Operation case 5 - 16 Inch SG Manway Vent, No SG Condensation Event loss of RHR o:>olirg at Mid-Loop Corxlitions Core ani Upper Plenum Temperatures = 140 F Ra3 Pressure = O psig ( 14. 7 psia)

Core ani Upper Plenum Temperatures Reach 212 F level SWell Reaches a Maxim.mt ani starts to Decrease Vapor Region Tenperature Reaches 212 F, Air Partial Pressure I.ess 'Ihan o.1 psia Collapsed level Reaches Tc:;? of Active F\\lel Erd of Transient M.xlel.ed

RCS arrl Vapor Temperature = 212. 7 F

RCS ani Vapor Pressure = 14. 9 psia Boiloff Rate AR;>roxilnately 13.5 ll:ra/sec (13.9 l..brrVsec I.on;J Tenn Core Boiloff) 33 Time sec Cmin>

0 (0) 545 (9.1) 586 (9.8) 621 (10.35) 3500 (58.3) 4000 (66.7)

JC c

la.I I-en I

a.

w illo

..J 4(

I-0 I-I

a.

-- P -

TOTAL

--P -

STEAM

- - - P - AIR 16.0 14.0 12.0 10.0...

h 8.0...

6.0...

.O...

I

~-J' 2.0...

0.0 0.0 I

I I

I

\\

500.0 I

1000.0 I

I I

1500.0 2000.0 2500.0 3000.0 3500.0 4000.0 TIME (SECONDS)

Sale* Mid~Loop 16 Inc~ Vapor Vent.

Figure 3.3-5-1 RCS Total and Component Pressures - case s

w UI Q.

x w

a:'

w

s 0 _,

Q.

% w a:

w Q.

Q.

-- UPPER TEMP

- - -- LOWER TEMP 220.0.

210. 0.

200.0 190.0 180.0 r

I I

I

- - - - ----**-*** ---***------~

I

. I I

j I

170.0 I

I I

160.0 I

I I

150.0 1 I I

140.0 130.0 0.0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 3500.0 4000.0 TIME (SECONDS)

Sale* Mid-Loop 16 Inc~ Vapor Vent.

Fiqure 3.3-5-2 Mixture and vapor Reqion Temperatures -

Case 5

2000.0

1800.0

~ 1600.0 r IA.I i

~ 1400.0

~

-c _, _,

0 u IA.I

E

J _,

0 >

a:

l&J

s 0 _,

1200.0 I

~

1000.0 800.0

~

Top of Core

600.0 ~

400.0 0.0 500.0 1000.0 1500.0 2000.0 2500.0 30~0.0 3500.0 4000.0 TIME (SECONDS)

Salem Mid-Loop 16 Inch Vapor Vent Fi~re 3.3-5-3 Lower Reqion Volume and Collapsed Volume - case 5

u cu.,,

Ii

~ -->>

0 -

w

~

c cu

-- UPR VENT FLO

.1'4.0 12.0...

10.0

~

8.0...

6.0 t-4.0 -

2.0...

0. 0 -------J

-2.0 0.0 I

500.0 I

I I

I 1000.0 1500.0 2000.0 2500.0 3000.0 3500.0. ~000.0 TIME (SECONDS)

Sale* Mid-Loop 16 Inc~ Vapor Vent Figure 3.3-5-4 Upper Reqion Vent Path Flowrate - Case 5

l l

l'

3. 4 Sumrnazy of Results 3.4-1 RCS Heatup an::l Time to Saturation Based on the previous analysis, the RCS heatup to 212 F occurs at an approximately constant rate. '!his rate is proportional to to the decay heat power an::l inversely proportional to the thennal capacities of the water, fuel, an::l stnicture in the ex>re an::l upper plenum :regions. '!he RCS heatup rate an::l ti.me to reac:::h saturation can thus be determined at various decay heat rates corresporrling to different tillles after reactor shutdown.

'Ihese resultes are plotted in Figures 3.4-1-1 an::l 3.4-1-2. '!he cases analysed in this report (starting fran 140 F initial RCS temperature, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after trip) were c:::hosen to conse:rvatively calculate the heatup an::l pressurization rates follawing the loss of RHR cooling. 'Ihese* cases.

reached saturation ra.ighly 9 minutes after trip as expected based on the initial oon:titions given. If RHR eooling is lost later than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after trip, or the initial RCS temperature is lower than 140 F, the heatup rate will be lower_an::l time to reac:::h saturation will be lo:ncJer.

3.4-2 Core Uncovecy

'!he mi.niJnum time. to core uncovecy following the loss of RHR c.Ooling while in mid-loop operation was conservatively estimated to be one hour. '!his analysis assumed a loss of RHR cooling 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after reactor shutdown, an open SG manway vent path for steam flaw, no steam generator corx:lensation an::l no operator recovecy actions. It should be pointed out that a loss of RHR cooling does not ~ily lead to oore uncovecy in all cases since the "VariOJS cp3rator -recovecy actions have not been analysed.

3.4-3 RCS Pressurization Rate

'!he maximum RCS p:resSurization rate (fran 14. 7 to 39. 7 psia) was

consenratively estimated to be 2. 77 psijmin while the RCS is saturated

38

'lhis conservative estimate was based on a loss of RHR cxx:>lirg 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after reactor trip, no oorrlensation heat transfer to the steam* generators, am no operator r:ecovery actions.

'!he case without any open vent paths was compared with two cases with 3/4 inch vent paths open to detennine if steam or. water ventirg would have any significant effect on the pressurization rate. No awreciable difference in pressurization rate* was fwrxl for the vapor vent path case, blt the liquid vent path pressurization rate was slighly higher sin::le less water was available for sensible heat addition. 'lherefore, the calculation of a CC>nSaVative pressurization rate is based on the results fran the 3/4 inch liquid vent path analysis

39

'2

~

I&,;

at !

0 I

r I

20.0 11.0 11.0 14.0 12.0 10.0 1.0 1.0 4.0 2.0 0.0 o.o

\\'

\\ '\\

' ~

!IO.O Figur1 3.4-1-1 H11tup Alt1 for Lo11 "" Cooling During Mid-Loop Op1r1tlon

~

r----___ r---_

100.0 l!IO.O 200.0 Tl* Aft1r Altctor Shutdown (hr1)

I 290.0 300.0

'2.....

I...

~ *

!J

'61 I fi

I 0

'61 t-Flgur1 3 *.f-1-2 Tl* ta Sttur1Uan for La11 af "" Coalln1 Dlrln1 Mld-laop 0p1r1t1an l!l.0-----------..--------------------------------------------------

IO.O 1-------1--------J------4.....;__----~-----1-----~

211.0 TRCS

lOOoP' 20.0 19.0 TRcs

140°r 10.0 O.O'----------i~--------'---------.&.---------"'-------------------

0.0 90.0 100.0 190.0 200.0 250.0 300.0 Tl* Afttr lhutllnn 1hr1J

As pointed out earlier, one of the possible recovery actions to restore RCS inventory and RHR function is to gravity drain water from the ms'!' to the RCS.

'!his -recavery action is only.effective if the RCS pressure is less than the head fran the 001'. If a 25 :psig ~head is asslll'lled., the RCS pressure would exceed this within 20 minutes after the loss of RHR coolilg.

'!he arialyses presented in section 3. 3-2 shCM that the RCS heatup and pressurization rates*are very sensitive to the annmt of condensation asslll'lled..

Even a small annmt of Condensation can increase the tline available for operator recovery actions.

3.5 Conclusions A conservative analysis was perforined to detennine the minimum tine to core uncovery following a loss of RHR while in mid-loop operation.

An open manwa.y vent path for steam was assumed and no credit for either reflux corrlensation or operator recovery actions was assumed in this conservative analysis.

Based on this analysis, it was determined that the

mininu.nn tine to core uncovery followilg a loss of RHR cooling at the Salem Nuclear Plants while in mid-loop CJPeFtion is approxbnately 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months *.

Followin;J the loss of RHR cooling, operator recovery action is necessary to either restore a heat sink or provide. adequate makeup water into the RCS to prevent core uncovery.

Makeup water can be added by either gravity feed fran the ~, the dlargilg system, or aey other available high pressure injection system_.

A heat sink can be restored by either refillin;J the steam generators or restorin; RHR flow.

Pressurization of the RCS nay prevent gravity *feed fran the RWST as a ma.ans of recovery. A separate analysis to detennine the maximum

, pressurization' rate was also performed. No vent paths were asslll'lled. to be open and no credit for either reflux comensation or operator recovery 42

actions was assumed in this analysis.

Based on this consei:vative analysis the RCS is predicted to pressurize to 25 psig in approximately 20 minutes following a loss of RHR cooling at the Salem Nuclear Plants while in mid-loop operation. When the effects of SG reflux corrlensation were IOOdelled, it was dem:>nstrated that the RCS pressurized at a much slower rate than the consezvative analysis predicted. 'Iherefore, it would be beneficial to keep one or nore steam generators partially. full of water while in mid-loop operation.

3. 6 References

1. American National stamard "Decay Heat Power in Light Water Reactors, 11 ANSI/ANS-5.1-1979.

2. US NRC Augrrented Inspection Team Report "Loss of Residual Heat RemJval system - Diablo canyon, Unit 2 - April 10, 1987," NUREX;-1269, May 1987.

4. 0 TASK 2 - RADIOIJ:X;ICAL CDN~CES 4.1 Description of Task 2 Task 2 was intemed to provide an evaluation of the radiological consequences of a loss of decay heat renoval. event dur~ operation with the RCS partially filled. '!he task description a$ it was presented in the technical description is contained in Apperrlix B. *'!he task consists of.

providing a plant specific determination of the off-site thyroid arrl whole 1xxiy doses that would result due to evaporative losses of reactor coolant arrl the resultant releases of iodines am entrained gases as* the coolant boils in absence of RHR System forced flow.

'!he intent. of this task is to dennnstrate that the off-site doses due to the event would present no significant hazard to the public

43

It was request¢ that the off-site dose evaluation take into account the possibility that the cxmtainrnent bouOOary might be open for passage of activity to the environment.

A calculation of site bouOOary doses was made assuming that full cxmtairnnent pn:ge is operating arrl a separate calculation of the airlx>:rne activity conoentrations in the contai.rment was made assuming that the containment is closed.

'!he Plant & System; Evaluation Licensing (PSEL) group of the Westinghouse Nuclear Safety Deparbtent corrlucted the evaluation arrl has provided the following djsa.JSSion as presented in sections 4.2 through 4.4.

4. 2 Task 2 Assunptions am Bases For the p.n:poses of detennining the activity in the reactor ccx::>lant, it was assumed that the mid-loop operation takes place in M:>DE s arrl the RCS is vented with the reactor ccx::>lant degassed to a concentration of 0.5 micro Ci/g of Xe-133.

'!his is a factor of ten greater than the value reccmnended by Westinghouse for the carpletion of reactor ccx::>lant degassing cperations arrl is five times greater than the cona:mtration present at the initiation of the Diablo Canyon event *. 'lhree I-131 concentrations wre considered:

case 1

'!he I-131 conoentration as required by the tedmical specification limit of 1. o micro Ci/g

44

case 2

'lhe I-131 Concentration of 0.1 micro Ci/g \\trhl.c::h is ten times the value specified in typical c:peratirg.instruction.s to permit head lift.

case 3

'lhe I-131 concentration of 0.02 miero Ci/g \\P.tlic:h is twice the value specified in typical q>e.ratirg instructions to permit head lift.

No other isotopes of iodine or of the noble gases are assumed to be present in quantities. that would have any consequence.

In the detennination of the activity released to the c::ontairinent it is.

assumed that all of the Xe-133.in the total RCS volume is released to the contai.rmertt am that 100 percent of the I-131 in the RCS volume of the coolant above the top of.the active fuel is released to the contairnnent (reoovei:y of the event is assuire:l to occur before the core is unccv~).

In the detennination of the site bcJurdal:y doses it is assumed that the containinent.is q:>9n an:l \\:hat all of the activity :released due to the event is released durirg the first two boors. 'lhe FSAR X/Q value of s.o x 10-4 secjm3 was used in the evalutation.

4.3 Task 1 Activity Coocentrations in the Cl:>ntainment Atm::>sphere

'lhe resW.ts of.the evaluation for the.containment activity concentrations assum:llg that the containment is not open to the environinent -are presented belOW'

45

TABI.E 4.3-1 a::NrAINMENT ACTIVIT:'i ISOIOPE ACI'IVITY Xe-133 6.3 x 10-4 micro Ci/ml I-131 (1.0 micro Ci/g in coolant) 4.05 x lo-4 micro Ci/ml I-131 (O.l micro Ci/g in coolant) 4.05 x 10-S micro Ci/ml I-131 (0.02 micro Ci/g in coolant) 8.1 x 10-6 micro Ci/ml 4.4 Task 2 Off-Site Doses

'!he results of the evaluation for site bourrlary doses assumi.rq that the contai.rnnent is open to the environment are presented below:

I-131 at 1.0 micro Ci/g I-131 at 0.1 micro Ci/g I-131 at 0.02 micro Ci/g TABIE 4.4-1 SI'l'E ~

InSES n:ses, rem

'Whole Body

. 'Ihyroid 1.53 x 10-3 7.378 3.04 x 10-4 0.737 1.95 x lo-4 0.147 46

In summary, if the loss of RHRS durin;J mid-lc:x:ip operation is considered an accident, cc:anparison with the 10 CFR 100 limits shows that the calo..ilated doses are well urrler the "small fraction of 10 CFR 100 limits" (30 mM thyroid am. 2. 5 REM whole body) whidl the NRC assigns for accidents which have the greatest prabablility of occurence.

Alternatively, the calo..ilated doses can be extrpared with the dose limit of 0.5 rem whole body (or its equivalent to aey part of the body) which is identified in Regulatory Guide 1.26 as the limit for specifying if equipnent is to be* categorized as Quality Group c or D (equivalent to Safety Class 3 an1 NNS).

In other 'WOrds, if a carponent failure wruld result in a dose of less than O

~ 5 rem whole. body, the ilrpact is considered not to be of signif ican6e to safety an1 the component can be classified as a Non-Nuclear Safety Ccilponent.

Based on the weighting factors provided in ICRP PUblication 26, 0.5 rem whole body is equivalent to 16. 7 rem thyroid. In all of the cases the calo..ilated doses are less than 0.5 rem whole body an1 16. 7 rem thyroid. It is noted that ICRP Publication 26 rec:x::mrrerrls an annual dose limit of 0.5 rem whole body an1 also reccmnen:3s a limit of 5.0 rem to any organ.

'!he results show that if the coolant concentration is at the Tedm.ioal Specification value of 1. O micro Ci/g the thyroid dose would be greater than 5.0 rem.

'!he coolant I-131 concentration associated with a 5.0 rem thyroid dose is 0.68 micro Ci/g.

It shc:uld be roted. that during the event at Diablo canyon an increase was d:>served in the concentration of xe-133 in the contairnnent atJtosii1ere while little dlan;Je was cbseJ:Ved in the concentration of I-131.

'Ibe assmrption that 100 percent of the icxline contained in the coolant assumed to have evaporated becanes. ail:ixlrne is considered to be an extremely consei:vative assunption~

47

5. 0 TASK 3. -.ASSESSMENI' OF VORI'EXING AND AIR ENI'RAINMENI' 5.1 Description of Task 3 Task 3 was interrled to provide an urrlerstan:li.rg of the i;:tienanenon of vort:exllg arrl air entrairnnent as it relates to the RHR System *

. Additionally, a review of Salem design features was to be made arrl camerits provided.

'lhe task description is contained in AJ;pm:lix B.

In addition, the assessment was to include :recamnerrlations for q>eratin3' procedures to limit the potential for loss of RHR System durin3' mid-loop q>eration arrl to resporxi to the loss of the RHR System should that occur.

5.2 Task 3 OJnclusions

'!be Safeguards systems (SS) group of the Westinghouse Systems Engineering Department has provided the following discussion of the P"ienanenon of vortexirg arrl air entrainment as it relates to the RHR System.

When the RCS water level is drained to the RCS hot leg centerline (e.g.,

for steam generator maintenance), air bin:ting of RHR p.mps becomes a conceJ:Tl as there is the potential for drawi.RJ air into the p.nrp suction due to RCS loop level fluctuations am;or develc:pnent of a vortex. If air bi.rxlin;J of the the RHR p.mps ocx:urs, RHR system capability may be lost.

However, the RHR system instiumentation provides a number of in::tications that would provide rapid detection of air bin:ting of the p.Illp; such as fluctuations in the.in:lications of RHR p.m1p JJDtor current, suction pressure (PI-631 & PI-632-), arrl disdlarge flCM (FE-64lA & 641B).

Additional RHR system perfonnance m:mitoring capability arrl inproved RCS level nonitor.in;J capability as lt.1ell as improvements to q>erat.in;J procedures to limit the likelihcxxl of air entrainment in the RHR p.mp; arxi to provide guidance for a rapid arrl effective response to loss of RHR system operability watld enhance the safety arxi reliability of the 48

Salem Units while operatin:J with RCS level at mid-loop.

'Ihe following paragraphs identify some ope.rating procedure guidelines to help preclude air birrling of* the RHR p.nnp an:i to properly :resporrl to loss of the RHR pump due to air birrling should it occur. Also identified are additional capabilities for m:>nitoring the RHR system perfonnance arrl RCS level which, if added, would i.nprove the ability to avoid entering operating conditions in which air entra.irment would be likely to occur arrl to promptly identify its ocx:urrence.

While the development of detailed operating procedures is the responsibility of the utility, Westinghouse Operating Instruction M-1, "Draining the Reactor Coolant System" recamnends guidelines to follow when draining the RCS for maintenance or refueling. 'Ihis guideline recamnends that a tygon hose be connected to the drain line of the reactor vessel flange leakoff connectiQn.

'Ihe water level in the reactor vessel can be clOsely _monitored by neans of this tygon hose.

However, the reactor coolant loop water level cannot be monitored by this arrangement.

To provide a neans of monitoring the water level in the RCS loops, a secom tygon hose should be connected to the drain line (or another available connection: e.g., vent path connection, instnnnent connection) of one of the four reactor coolant loops an:i be exten::ied at least two feet above the top of the pressurizer, where it would be vented to the contaimnent atmosphere.

'Ihe water level in the RCS loops can then be m:>nitored by

~of this second tygon hose.

'Ihe following are precautions.that should be followed to prevent the loss of RHR capability when the RCS has been lc:Mered (to the hot leg centerline) to drain the steam generator tubes for maintenance

49

1.

'!he RHR flCMrate should be reduced to 1800 gpm or less to preclude vortex fonnation at the RHR hot leg connection (DevelCl[:l1Yal1t of this flow lllnit is discussed in Section 6.2).

2.

D.lring steam genera~r tube draining, the water drains in a slugging fashion, therefore RCS water level in:lication may be erratic. For this reason, the draining operation should be stopped periodically to allow the water level. in the system to stabilize.

3. -

'lhe water level should be m:mitored continuously while perturt>ing the RCS inventory to assure that the RHR System inlet line does not become uncovered an:l gas bin:i the RHR pumps.

4.

When RCS draining is from the loop which is provided with the tygon hose level irrlicater, the draining operation should be stopped when it is desired to obtain an accurate reading. If draining is from a reactor coolant loop not provided with the tygon hose, proper carmmmication should be established to coordinate t:he draining operation with the level monitoring.

'lhe Salem RHRS (Figure 5.2-1) consists of a single drop line connecting two parallel an:i identical trains, each consisting of an RHR pump, RHR heat exchanger an:i the associated piping, valves and.instrumentation required for operational control. 'lhe system is sized to pennit RCS coolda;.m from *35o°F to 140°F in 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months with both RHR trains in operation.

However, after coolda;.m is c:arplete, only one RHR train is required to provide adequate residual heat renoval capability; thus one of the two RHR trains may be out of sei:vice during this period.

Following RCS draindown, if the RCS water level should fluctuate or be inadvertently reduced to the point at which air is drawn into the suction of the operating RHR pump, resu1 ting in air bin:lin;J,. the rtmning RHR pump must be stopped (do not 50

l><J

r:i:

- -N M'1MA"WIUM..

-- '1:1

--..J "iiiimi

-=D

.. _G-

start the unilrpaired RHR pump), the water level should be restored arrl the uniJrpaired RHR p.mip should be then started. '!he gas-boun:i pump may be restored to operational status by ventin] the air from the pump casing arrl connecting piping. '!his can be ac:carrpli.shei by reflooclin:} the p.mip suction piping. arrl by opening the manual vent valves an the pump suction arrl discharge piping.

Once this is done the pump may then be re-started*

as required. 'Ihe preceding steps should be campleted quickly such that minimal interruption in RCS cooling will be experienced.

In addition to closely m:mitoring RCS loop water level the following imications could be incorporated into a RHRS purrp protection system design to prevent an::vor provide rapid imication of air bi.rxtin] of the RHRS pmps. '!his RHRS puirp protection system would consist of a microprocessor. which utilized RHRS iniications, such as RHR flow, suction arrl discharge pressures, RCS suction water level, RHRS pump current an::i other available irxlication to calculate the margin to loss of suction corrlitions.

An alann would be provided to alert the control roam operator to take corrective action before the pumps air bim.. Irxlications of abnormal RiiR puiri:>- operation are currently provided by readout of the purrp discharge pressure arrl puirp current on the Main Control Board arrl by renote readout of the pump suction preSsure.

In summazy, to prevent air bin::lin] of the RHR pumps, it is illlportant to keep the RCS loop water level above the hot leq centerline.

Because of this requirement, it is illlportant that close n:mitoring of the RCS loop water level be maintained during any operations involving reducing the RCS water level below the top of the RCS hot leq. If air bi.rxtin] of the RHR pmps occurs, reprim:in;J (filling* arx:l venting) the pump should be completed as quickly as possible to preclude the loss of RCS cooling for an extended period of time.

In the event the RHR p1MpS cannot be restarted, -the followirg should be,

considered as option$ available to restore cooling of the RCS:

52

1.

Use the c::h.arging system to refill the RCS arrl establish a natural circulation cooling path, :rerroving decay heat via the steam generators am the Auxiliacy Feedwater System.

2.

Use the c::h.arging system to refill the RCS arx:i establish circulation by starting one or more reactor coolant pumps (the starting criteria arrl precautions for RCP start should be met);

as in Option (1), decay heat rem::wal would be provided by the sec:orxmy system.

3.

Allow the core to boil-off coolant arx:l use the charging system to supply makeup to the RCS; this may be a viable option only if the RCS is open to the contairment aboosphere.

6. 0 TASK 4 -

DETERMINATION OF VORI'EX I.E.VEL 6.1 Description of Task 4 Task 4 was interrled to provide an evaluation to detennine the sensitivity

~f RCS loop water elevation to critical submergence depth as a function of RHR System flow. Additionally, a calculation is to be made to detennine acceptable reduced RHR System flow during mid-loop operation. Appropriate FSAR arx:i Tec::hnical Specification changes along with accompanying lOCF.R.50. 59 arrl Significant Hazards EValuations are to be provided to reflect the throttled RHR System flow rates during mid-loop operation.

'!he task descript:ion as it Was presented in the technical description is contained in Apperrlix B

53

6. 2 Task 4 Conclusions

'Ihe Safeguari)s Systems (SS) group of the Westin3house Systems Engineering Department has oomucted a review of RHRS operation an:i design for the Salem nuclear units.

'Ihe review considered mininnlII1 RHR flaw necessaz:y to:

1) renx:we decay heat. to maintain RCS. tenperature, 2) preclude boron stratification an:i.3) provide adequate flaw rate for boron dilution accident c:xn:eJ:TIS.

'Ihe effect of RHR flaw an:i RCS level* was evaluated to determine susoeptibil_ity for air entrairmlent. It shalld be noted that, in all eases WeJ:'e RHR flaw is identified, it refers to RCS delivered flawrate.. If the miniflaw path is opened this shool.d be acxumted for by operati.rx.J the pump at the flaw requirement* plus miniflc:JW.

HFAT REM:>VAI..

'lhe primacy function of the RHRs is to rem:we decay heat after a reactor shutdown.

'Ihe am::iunt of decay heat renrwa1 necessa.z:y to _

maintain the RCS teirpera.ture constant or decreasinif is a function of time attar initial reactor shutdown. Table 6.2-1 presented below lists the required RHR flawrate requirements as a function of tiine after shutdown.

18 42 54 78 114 TABIE 6.2-1 54

. 3000 1800 1500 1250*

1000

Figure 6.2-1 depicts the data presented in Table 6.2-1 in graphic fo:rm. It should be noted that the Salem Ted'U1ical Specification Limit of 3000 gpm (SW:Veillance Requirement 4.9.8 for Unit 1 an1 4.9.8.1 for Unit 2) is sufficient to maintain RCS temperature as early as 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months after plant shutdown.

As the time after shutdown increases the decay heat renoval. requ:irerents for RHR flow are reduced. Since there is a relationship between RHR flowrate arrl required RCS water level to prevent vortexing in mid-loop operation, the plant can take advantage of the reduced decay heat renova1 requirements to reduce the RHR flowrate.

ruMP NPSH Pl.mp NPSH was evaluated at the 1000 to 3000 gpm flowrates (assuming RCS at 200°F) an:1 it was detennined that sufficient purrp suction head is available in this range.

OORON STRATIFICATION Since no intentional Changes in boron concentration are made during refueling, the only nechanism which could induce a gradient in boron concentration is local mass evaporation.

However, the fluid temperature during nme 5 arrl 6 operations is below 200°F, arrl the evaporation effects are min:irnal; also the fluid temperature is well above the precipitation value for the expected range of concentration.

. '!he potential for boron stratification has been evaluated for a total RHR flow of_ greater than 1000 gpm am detennined to be acceptable.

'!he bases for preventin:J boron stratification in the RCS is to minimize the potential for a boron dilution accident. A flow of greater than 1000 gpm ensures that adequate mixing within the RCS takes place such that no significant accumulation of coolant, with a boron content different to that in the core, can occur elsewhere in the primary circuit.

55

3000 -\\

~

0 1-i

)ii t1

~

2000 b

~ -

Cl i -

1000 20

/

~

~ ~

~

r---- r-._ -

30 40 50 60 70 80 90 100 TIME AFTER SHUTDOWN (HR)

FIGURE 6.2-1:

'lOl'AL RHR FUM VERSUS MID-I.OOP* INITIATIOO TIME 110

~

8 12 tlJ

~

OPERATIONAL BAND tlJ I:"'

9 to

o tlJ

i:

0 8

BORON

~

6 Cl n

tlJ z 8

DI WT ION

'LIMIT

/

VORTEX tlJ

~

3 H

LIMIT z

tlJ H z n

z: -

0 I.

1000 1800 2000 3000 TOTAL RHR SUCTION FLOW (GPM)

FIGURE 6. 2-2:

RCS WATER IBVEL VERSUS* 'IUI'AL RHR SUCTION FI.CM.

In view of the preoedinq, it is concluded that there is no cxmcem for boron stratification at the RHR min.ilm.nn flowrate. of 1000 gpm..

Figure 6.2-2 shows hot leg RCS water level vs. RHR flOW' with the limitations on mid-loop operation shown. It should be noted that the RCS water level shall be maintained no lower than the hot leg centerline for RHR flOW' up to 1800 gpn. For an RHR flow higher than 1000 gpm but not greater than 3000 gpn, the. suction water level shall be maintained at

.least 6 indles above the hot leg centerline *. '!his increase in RCS. water level has been detennined to be conseIVative for the flow ~e of 1800 to 3000 gpm.

It should also be noted that duri.rg mid-ioop operation.it is reccmrerrled that only one train of the RHR system be in operation._ At the reduced RHR flow rate :recommemed (1000 - 1800 gpn), it l<<IUl.d be difficult to read the flow rate i.rx:lication if both punps were in -use because the flOW' element is calibrated to the nonnal RHR flow rate of approxiJnately 3000 gpm arrl the flOW' rate in each train l<<JUl.g be *in the ~e of 500 - 900 gpm.

Also, if air entrainment did ocx::ur 'While both punps are in operation, it is likely that both l<<IUl.d be lost fran sezvice.

Regarding potential air entrainment, current Westin:Jbouse guidelines (Salem Reference Operating Instruction M-1, Rev. 1, dated Septelilber, 1975) recamnerxl reducing the RHR flCM to 1500 gpn per RHR p..mp Wen the RCS water level_ is lowered in the loops. 'lhis original recamrrerxiation was

. based on eilgineering judgement am q>era~*experi~.

West:injlcuse has re-evaluated this limit based on previoos research, available literatw:e, scaled test results,-an::l additional Operating histo:cy an:1 has detennined that the 1500 gpn limit, shculd be revised to 1800 gpn (total) with no mn:espon:ling ioorease in RCS water level. It. is noted that the recamren:ied flawrate of 1800 gpni is only a qualitative limit based on the criteria to maximize available ex>re cooling 58

~.

while providing adequate assurances against unacceptable consequences of vortex fonnation.

Although the analytical techniques derived from test results do not entirely preclude same degree of vortex fonnation arrl air entraimrent at 1800 gpm, operating experience supports the position that, at this reduced flowrate, pump perfonnanc::e arrl operability is not adversely inpacted.

However,* given that a partial vortex fonnation at 1800 gpn cannot be precluded, Westinghouse does suggest that strict administrative/prcx:edural assurances be inco:rporated to ensure that maxinn.Im pump flavate arrl mininn.mt RCS levei limits are not exceeded. In addition, since any increase in RCS level or decrease in pump flowrate provides significant benefits in suppressing vortex fonnation the following suggestions are provided:

1)

Mid-loop operation should be limited to only those tines required.

Whenever possible RCS level should be maintained at the highest possible elevation an:l preferably with the RCS hot legs water solid.

2)

RHR flowrate should be maintained between 1000 arrl 1800 gpm, at the minimum flowrate required to keep up with the decay heat load. 'Ibis requirement should be inco:rporated into plant operating procedures.

In summacy, WestinJhouse concludes that 1800 gpm RHR flow is an appropriate limit to prevent unacceptable vortex fonnation.

However, it

. is reconunerrled that system perfonnanc::e be continuously m:mitored to detect the onset of vortexing arrl air entraimnent

59

6. 3 Boron Dilution Concerns.Associated With Reduced RHRS Flow

'!he Transient Analysis II (TAII) group of Westinghouse Nuclear Safety Department has* c:x>rrlucted a review of the inPact of throttling the RHR flCM during mid-16op operation am has provided the follCMing discussions c:x>ncerning possible boron dilution c:x>ncerns.

'!he Salem Units have adopted the guidelines of NS~-2273,issued on July 8, 1980, (Interim Operating Procedure For Boron Dilution In IDDF.s 4 And 5) to assure adequate operator action til1le for a boron dilution event in IDDFs 4 am 5. For each reload a curve of boron c:x>ncentration verses ~re burn.up is generated.

'lhe curve establishes the boron c:x>ncentration required in IDDF.s 4 am 5 to ensure adequate operator action til1le is available should a boron dilution event occur in these modes of plant operation.

'!he analysis is sensitive to RHR flCM rate.

A wide range of RHR flow rates in IDDE 5 would result in the need for more :restrictive boron c:x>ncentration curves than currently in existence.

'lb address the potential RHR flow rates in IDDE 5, new data will have to be generated.

'!his data will be applicable for IDDF.s 4 am 5 for the range of RHR flow rate from 1000 gpm to 4500 gpm.

'!his rarge will be sufficient to cover the flow rates for mid-loop_operation as discussed in section 6.2 above am also will be applicable for full-loop operation *. '!he revised boron c:x>ncentration curves will be transmitted urxler a separate cover letter.

'Ihese requirements should be incorporated into plant operating procedures along with the vortexirg am boron stratification requirements of section 6.2 above.

It should be noted that the IDDE 6 boron dilution analysis is not inpacted by the RHR flow reduction discussed in section 6.2 of this report. '!he Interim Operating Procedure is not usa:I for the boron dilution event in IDDE 6.

60

L_

6.4 Tedmical Specification C11an;Jes

'!he Technical Specification Sel:vices {TSS) group of Westinjlouse Nuclear safety Department has corxiucted a review of the inpact of throttling t.he RHR flow durirg mid-loop operation arx:l has provided the followirq a; scussians conc::emin:1 possible technical specification ~-

'1he salem Technical Specifications contain the foll~ requirements:

A.

Mode 5

1.

Operability of two RHR loops arxi operation of one (four filled RCS loops arxi two steam generators with level~ 5% of narrow rarge may be substituted for one operable RHR loop).

2.

Verification that :reactor coolant is bein;J circulated via the RHR system once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

B.

Mode 6

1.

Operation of at least one :RHR loop.

2.

Verification that at least 3000 gpm of :reactor coolant is being circulated via the :RHR system once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3.

Operability of two :RHR loops when the level in the refueling cavity is less t.han 23 feet above the vessel flarge.

c.

All Modes (salem-Unit 1 only)

1.

A ~owrate of~ 3000 gpm through the RCS.

2.

Verification that one reactor coolant ?lllp is in operation or that one RHR p..nnp is operating arxi supplyin;J ~ 3000 gpm through the RCS.

61

'lhe Salem Technical Specifications do not (an:l in reality cannot) disallow operation of the RHR system with the loops drained to mid-loop.

'Ihe technical specifications do place limitations on the :RHR system when operatirg with the RCS drained to mid-loop in the fonn of limitations on the number of loops which must be operable. 'lhe technical specifications do not contain restrictions based on minimizirg air entrainment in the :RHR system as a result of vortexirg which may occur durirg mid-loop operation urrler certain oorrlitions. 'lhe miniJnum flCM requi.reirent is for the p.u:pose of decay heat ren¥JVal ard is considered conservative. '!his flew rate has subsequently been used as an inp..rt to the boron dilution analysis.

Required :RHR flCM is a function of many factors inclulirxJ time after shutdaNn, boron dilution ard stratification considerations, lev0I. in the refuelirg cavity, RCS pressure ard temperature, ard level of the reactor coolant in the loops when the RCS is partially drained.

No one flew requirement applies to all the possible RHR configurations. In view of the multiplicity of RHR flew requirements, limitations on :RHR flew are more awropriate for operatirg procedures where prescriptive configurations an:l the cori:'esporxlin:J flCM requirements can be described in detail.

SUch detail is inappropriate for technical specifications. It is reconunended that RHR flew requ:ire:ments presently included in the technical specifications be deleted ard that operatirg procedures be revised as necessaey. to include such requirements.

Per the above recammemation Salem Unit. 1 arxi 2 Technical Specifications were reviewed an:l marked up as follaws:

Specification 3/4.9.8 (Unit 1) arxi 3/4.9.8.1 (Unit 2) - References to a flew rate of 3000 gpm in surveillance requirements 4.9.8 ard 4.9.8.1 were deleted. Specification 3/4.1.1.3 for Salem Unit 1 was deleted.

It is proposed that specific :RHR flCM requirements be located in plant procedures instead of technical specifications. 'lhe flew rate of 3000 gpm represents the design flew rate arxi is conservative for decay heat*

62

rem::wal. It does not address all RHR flow considerations and therefore may not be applicable in all circumstances. Relocation of flow requirements to procedures wa.U.d allow all RHR flow requirements to be addressed in an appropriate manner.

For RHR System operation the required flowrate of Specification 3/ 4.1.1. 3 for Salem Unit 1 would be located in plant procedures. For Modes 1,2 and 3 flow greater than 3000 gpn required by Specification 3/4.1.1.3 is assured by RCS specifications requirin:J operation of at least one RCS loop incltxiirg the reactor coolant p.mtp.

salem specificatin 3/4.1.1.3 is not in standard technical specifications (NUREXH)452)

Bases 3/4.4.1 and 3/4.9.8 - '!he follc:Jf.tlin; was added:

Adjusbnents to RHR flow may be required duri.n; operation of the RHR system.

For example, RHR flow may need to be adjusted to control RCS temperature, to prevent RHR pmp averlleatin;J and to preserve RHR suction requirements for the exi.stin;J RCS and RHR fluid caniltions.

Bases 3/4.1.1.3 for Salem unit l was deieted.

RHR flow 1ID.lSt be adjusted for proper control of RCS temperature and other factors, possibly aver a significant ran::re, as carxli.tions in the RCS and RHR systems charXJe.

Various mininn.nn am;or maxinn.nn flow restraints may l1eed to be iJrp:>sed deperxiirq on the existin;r corxiitions. Flow rates which are sufficient for one cxmlition may not be sufficient for another corxiition.

'!he proposed bases statement has been added in recognition of this.

Possible *alternatives to. the recc:amnen::1ed solution could be:

a. Inclusion of RHR flCM requirements in Modes 5 and 6 for operation of RHR with the RCS drained to mid-loop. 'Ibis may require specifyin:J both a :m:in:iim.nn and a max:inn.nn flow if it is necessacy to address ~y 63

/

I heat ren¥JVa1, boron dilution am vortex:in;J in the tedmical specifications. It is also possible that a :requirement on reactor coolant level in the loop 'WOUld be required for completeness.

b. Revision o~ the RHR fla.r requirement in Mode 6 to a value consistent

with mid-loop operation but which does not prevent operation with.

flc:MS greater than that for which a vortex:in;J problem could occur.

'lllis would be consistent with the existirg specifications which address analytical concerns* but not vortex:in;J oancerns.

c. ~an of the fla.r requirement in Specification 3/4.1.1.3 for Salem Unit 1 to the min:llmJm required for boron dilution.

6.5 ~

Qlan;Jes

'!he Plant & Systems Evaluation Licensirg (PSEL) group of Westllghouse_

Nuclear Safety Deparbnent has coIXiucted a review of. the i.npact of throttlin;;J the RHR flCM durirg mid-loop operation am has provided the follc:iwixg disoJSsions ooncemi.rg possible Salem Plant FSAR chan]es *

'!he marked up FSAR dlan;Jes are provided as an attachment to Apperrl.ix D of this report.

'!he markups consist of providin;J an insert for section

5. 5. 7. 2, System Description of the Salem - FSAR.

'!he insert describes the throttlin;;J of the the RHRS fla.r durirg mid-loop operation am includes a diSCJ.JSSian an factors which minimize the effects of air entrainment an

pump petfonnance.

A section is included which a; scusses procedure conunit:rcent ard development which the plant may wish to inco:rporate.

64 i

------1

ML18094A701

Text

REFERENCE:

Navigation menu

Search