ML18093A378

Rev 7 to, Westinghouse Engineering Svcs Rept for Salem Nuclear Generating Station Units 1 & 2 Concerning RHR Sys Mid-Loop Operation Re NRC Generic Ltr 87-12

ML18093A378
Person / Time
Site:	Salem
Issue date:	09/16/1987
From:	WESTINGHOUSE ENGINEERING SERVICES
To:
Shared Package
ML18093A376	List: ML18093A373 ML18093A378
References
GL-87-12, NUDOCS 8709280080
Download: ML18093A378 (71)

Text

REVISION:

SEPI'EMBER 16

'REV 7 ATTACHMENT 2 WESTINGHCXJSE ENG:mEERING SERVICES REroRI' FOR SAUM NUCI.FAR GENERATING STATION UNl'IS

• 1 AND 2 CX>NCERNING RHRS MID-IOOP OPERATION m REFERENCE '10.

NRC GENERIC IEITER 87-12 SEPI'EMBER 1987.

'lHIS REVISION IS '!HE RESUUI' OF CXJ.1MENl'S MADE. IXJRING A SEPI'EMBER

~.s, 1987 MEEi'ING m PITl'SWRGH PA.

ATrENDED BY:

C. ~

(PSE&G)

G. :roGGIO (PSE&G)

R. CMAC (WESTlNGHOOSE)

P. MC HAIE. (WESl'mGfIOOSE)

. E. FRANTz:. (WESTINGHCXJSE)

R. OFS'IUN* (WESTINGHOOSE)

A. LIN (WE.STINGEIOOSE)

REFERENCE:

Westinghouse Letter NISD/SIM-87-396

' 97092sooeo a~g6a~72

\\

PDR ADOCK 0 PDR l p

TABIE OF CDNTENTS Section Title 1.0 INmOIXJcrION 1

2.0 RERJRI' SCDPE 1

3. 0

'IHERMAL HYmAULIC ANALYSIS 2

3.1 Description of Task 1 2

• 3. 2 Description* of Analysis Methods am 4

Assumptions 3.2-1 Operational Concems 4

3. 2-2 Description of Analysis Model 5

3. 2-3 Description of Ii1p.lt arxi Assumptions 6

3.3 Detailed Analysis 10 '

3.3-1 case 1 -

Base case Analysis 10 3.3-2 case 2 - Sensitivity to SG Con::iensation 16 3.3-3 case 3 - Analysis with 3/4 Inch Vent 20 FlCM Path 3.3-4 case 4 - Analysis With 3/4 Inch Liquid 26 FlCM Path 3.3-5 case 5 - Analysis With Iarge Vent 32 FlCM Path 3.4 ~of Resaj.ts 38 3.4-1 RCS H~tup am.!l'ime to saturation 38 3.4-2 Core Uncxwery 38 3.4-3 RCS. P.ressuriZation Rate 38 3.5 Q:Jnclusions 42 3.6 References 43 i

TABIE OF <DNI'ENIS.

section Title Page 4.0 TASK 2 - RADIOI.DGICAL CDN~CES 43 4.1 Description Of Task 2 43 4.2 Task 2 Assmnptions Ani Bases 44 4.3 Task 2 Activity Concentrations 45 In 'Ihe Contairnnent Atioosphere 4.4 Task 2 Off Site Doses 46 5.0 TASK 3 - ASSESSMENl' OF VORIEXING AND 48 AIR ENrRAINMENT 5.1 Description of Task 3 48 5.2 Task 3 Conclusions 48 6.0 TASK 4 - DEI'ERMINAT.ION OF VORI'EX IBVEL 53 6.1 Description of Task 4 53 6.2 Task 4 Conclusions 54 6.3 Boron *Dilution Concerns.ASsociated 60 With Reduced RHRS Flow 6.4 Technical Specification Cllarges 61 6.5 FSAR~es 64 ii

TABI.E OF CDNI'ENI'S Section

.APPENDICES A

B c

D USS OF RESIIXJAL HFAT REMJVAL (RHR)

WHIIE 'IHE REACTOR CODI.AN!' SYSTEM (RCS)

IS PARl'IAILY FILIED (GENERIC IEITER 87-12)

SAilM RHRS MID-IDOP NRC RESK>NSES WESTINGHCXJSE IE.ITER NISD/SIM-87-396 MARKED UP TECHNICAL SPECIFICATIONS AND ACXXlMPANYING 10CFRSO. 59 E.'VAIIJATIONS AND

• SIGNIFICANI' HAZZARDS CONSIDERATION ANALYSIS MARKED UP FSAR CHANGES iii A-1

. B-1 C-1 D-1

Lisr OF TABIES TABIE Title Page 3.3-1 TIME TABIE OF EVENIS, ~

CASE 1 -

12 mmcr RCS, NO SG CDNDENSATION 3.3-3 TIME TABLE OF EVENIS, CASE 3 - 3/ 4 rnaI 21 VEN!' rn VAroR RmION, NO SG CONDENSATION 3.3-4 TIME TABLE OF EVENrS, CASE 4 - 3/ 4 rnaI 27 VEN!' rn LI~ RmION, NO SG CONDENSATION 3.3-5 TIME TABIE OF EVEN'IS, CASE 5 - 16 rnaI 33 SG MANWAY VEN1', NO SG a:>NDENSATION 4.3-1 CDNI'AINMENI' ACTIVJ!I!'i 46 4.4-1 SITE :IDJNDARY InSES 45 6.2-1

'IUl'AL RHR FI!:M VERSUS MID-I.OOP 53 INITIATION TIME iv

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

RFACIOR SHUTIXMN 3.3-1-1 RCS 'IOl'AL AND CXlMPONENT PRESSURES 13

- CASE 1 3.3-1-2 MIX'IURE AND VAPOR ROOION TEMPERA'IURES 14

- CASE 1 3.3-1-3 I.CMER ROOION VOilJME AND CDLIAPSED VOIIJME 15

- CASE 1 3.3-2-1 RCS PRESSURE cx:H>ARISON 18 3.3-2-2 VAroR/AIR ROOION TEm'ERA'ltJRE cxm>ARISON 19 3.3-3-1 RCS 'IOl'AL AND CXlMPONENT PRESSURES 22

- CASE 3 3.3-3-2 MIX'IURE AND VAroR ROOION TEMPERA'IURES 23

- CASE 3 3.3-3-3 I.CMER ROOION VOllJME AND CDLIAPSED VOIIJME 24

- CASE 3 3.3-3-4 UPPER ROOION VENT PA'IH FI.CMRATE - CASE 3 25 3.3-4-1 RCS 'IOl'AL AND CXlMPONENT PRESSURES - CASE 4 28 3.3-4-2 MIX'IURE AND VAroR RIDION TEMPERA'IURES 29 CASE 4 3.3-4-3 I.CMER ROOION VOWME AND CX>LIAPSED VOIIJME 30

- CASE 4 3.3-4-4 I.CMER ROOICN VENT PA'IH FUMRATE - CASE 4 31 3.3-5-1 RCS 'IOl'AL AND CXlMPONENT PRESSURES - CASE 5 34 3.3-5-2 MIX'IURE AND VAroR ROOICN TEMPERA'IURES 35

- CASE 5 3.3-5-3 I.CMER RmION VOIIJME AND CDLIAPSED VOIIJME 36 CASE 5 v

LIST OF FIGURES.

FIGURE 3.3-5-4 UPPER mx;:too VEN!' PA'IH. FI.CMRA'IE -

CASE 5 37 3.4-1-1 HFA'IUP RATE FOR LQSS OF RHR OJOLING IXJRING 40 MID-IOOP OPERATIOO 3.4-1-2 TIME 'ro SAnJRATIOO FOR I.OOS OF RHR OJOLING 41 IlJRING MIIrIOOP OPERATIOO 5.2-1 SAUM RHRS 51 6.2-1

'IOI'AL RHR FI.!M VERSUS MIIrIOOP 56 INITIATIOO TIME 6.2-2 RCS WATER ll.VEL VERSUS 'IOrAL 57 RHR SUCI'IOO FI.!M vi

'WESTINGHCXJSE ENGINEERING SERVICES REroRI' FOR SAllM NUCLEAR GENERATING STATION UNITS 1 AND 2 CONCERNING RHRS MII:rI.OOP OPERATION

1. 0 INI'ROilJcrION Operation of the Residual Heat Re.rro.ral. (RHR) system with the :Reactor Coolant System partially drained has been an in:iustcy concern for many years since this mxie of operation places the plant in a corrlition highly susceptible to the loss of RHR function.

'Ihe U. S. Nuclear Regulatory Ccnunission (NRC) Generic Letter 87-12 was issued on this subject pursuant to 10CFR50.54(f) arrl was prcmpted by the April 10, 1987 Diablo canyon loss-of-RHRS event. 'lhe principal concems of the NRC focus on whether

• the RHR system design meets the lioensi.rg basis of the plant for this mxie of operation arrl whether this mxie of operation can lead to resultant unanalyzed events that cc:W.d impact safety. Generic Letter 87-12 is provided as an attachment to this report in Appendix A.

2. 0

REEURl' SCOPE

'Ibis report is prepared in response to Public service Electric arrl Gas Ccmpany's response to Westinghouse Letter NISD/SIM-87-390 which is provided as an attac.hment to this report in Appendix B. * '!he report provides Westin:Jbouse analyses specific to the Salem NUclear GeneratirxJ station Units 1 arrl 2 to _support response to Generic Letter 87-12. '!he work scope as defined in the technical description contained in Appendix B lists the followi.rg five tasks to be carpleted urxler this engineeri.rg services authorization: 1) 'lhermal HydraW.ic Analysis, 2) Radiological Con.sequences Evaluation, 3) Assessment of Vortexin;J arrl Air Entrainment,

4) Determination of Vortex Level arrl 5) Report Doo.mentation. 7his report, which was coordinated by the Plant & Systems Evaluation Licensing 1

(PSEL) group of the Westinghouse Nuclear Safety Deparbnent, cxmstitutes conpletion of task 5. Tasks 1 through 4 are di sa..issed in the follCMin3 sections a.rrl are the result of inp..rt provided by several functional groups of the Westinghouse Nuclear Safety am Systems Engineerin:J Departments.

3. 0 TASK 1 - ~

IMEAIJLIC ANAINSIS 3.1 Description of Task 1 Followi.n:J the April 10, 1987 loss of RHR event at Diablo canyon, the NRC issued Generic letter 87-12, "loss of Residual Heat Removal (RHR) While the Reactor Cbolant system (RCS) is Partially Filled." Item 5 of this letter requested :R4lR plant licensees to provide a summacy deseription of plant procedures for RCS draim.own am operation in the partially filled ex>n::tition.

'!he :response is to include:

- 'lhe analytic basis used for the procedures develcpnent

- Treatment of the drairxiown of the RCS

- Treament of air entrainrrent arrl de-entrainment

- Treatment of boilirg in the

• ex>re with am without RCS pressure bourrlary integrity.

- calall.ations of awroxililate tine fran loss of RHR to ex>re damage

- level differences in the RCS am the effect up:m

inst.runvantation Wications

- Treatment of air in the RCS/RHR

- Treatment of vortexin3 at the connection of the RHR suction line (s) to the RCS of Generic letter 87-12 specifically m:mtione:i several tq>ics that need t6 he addressed am Ul'rlerstood.

'1hese include:

2

- Unexpected RCS pressurization due to air in the RCS

- Water loss thralgh openin;J in the cold le:J

- RCS water level instrumentation uncertainties

- Vortexing ard air in;estion from the RCS into the RHR suction With the loss of RHR, the only m:rle of energy rerrova.l is coniensation in the steam generators. In the absence of air or other non-condensible gases, the prmacy system will pressurize mrt:il sufficient te:npllature difference exists between prilllary an:l secorXiacy to drive the reflux condensation process. Nonnal.ly the teuperature difference required is small, so the primacy pressure will essentially equal the secorrlacy pressure.

Recoveey from this condition should be relatively s.i.nq:>le, ard ex>re uncovery and damage is unlikely.

'!he presence of non-condensible gases such as air in the loops cxirrplicates the picture cx:msiderably. '!his air Im.JSt be displaced by the steam before reflux condensation can take place.

Deperx:ling on the extent of this displacement, RCS pressurization may be significant:.. It is likely that fairly ccmplex 2 or 3-dllrensional flow patterns will exist which substantially reduce the pressurization.

~ite the canplex nature of this problem, it is possible to determine*

nost of the parameters of interest usin:J a s.i.nq:>le but conservative boiloff m:del with non-condensibles.

'!his mc:Xlel will provide a macroscopic view of the RCS ard, with appropriate m:rleli.rg a.ssurrptions, provide a conservative estinate of RCS pressure, teuperature, ard mass invento:ry duri.rg tne boiloff ard ventin; processes.

Sane minim.mt annmt of RCS to SG heat transfer can also be included to realistically :t:nJrxi the RCS pressure rise.

'!he cinal.yses presented in this section of the report describe various loss of RHR CXX>lin; scenarios for mid-loop q>eration for Salem Units 1 ard 2

• 3

'Ihe :resW.ts of these analyses can thf>.n be used to validate the plant procedures for RCS drain:lown an:l operation in the partially filled corrlition.

3. 2 Description of Analvsis Methods an:l Assumptions

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

3.2-1 Operational Concerns

'!be analysis presented in this report is interrled to provide inprt to an:l SUQ?Ort the operation procedures for drain:lown an:l mid-loop operation for the Salem Generatin; station. Several analysis concerns related to the loss of RHR cool~ event dur~ _mid-loop operation have been identified arrl are smmna.rized below.

First, dur~ mid-loop operation, the core exit thenoocxxJples nay be disconnected in preparation for upper head rem:wal. If the core exit thennocouples are not functional, a direct in:tication of RCS temperature would not be readily available if forced RHR flow is lost. A COnsel:Vative estimate of the RCS heatup rate arrl time to saturation is therefore important to know to detennine when core boil~ will ~in arrl for-the ti:min:;J of subsequent -reo:Nery actions.

Seccnily, it is also i.Irp:>rtant to know the RCS pressurization transient that folldNs the initial heatup to boil~. If pressure remains lc:YN for an extended period of time (e.g., less than 25 psig), it would be possible to increase RCS inventory by gravity feed fran the ~

(this m:xie of r:ecavery was used at Diablo canyon)

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

one dlarg~ p.mp c:nlld be used to increase RCS inventory. Additional 4

. high-pressure injection 'WOUld be required for increased makeup requirements. If the heatup am pressurization transient is allowed to continue further, the RHR cut-in cpn::litions (350 F, 375 psig) and design corxlitions (400 F, 600 psig) could be exceeded. It 'WOUld then be necessary to use an al teJ:nate m:xle of coolirg for interhn or lorg tenn recovery.

Finally, the RCS boiloff rate. am tine to core uncovery are important to knaN in order that core damage be prevented. Additional radiological concerns are introduoed if oo:re uncovery is allowed to ocx::ur.

3.2-2 Description of Analysis Model

'llle cx:mp.iter nrdel used to analyse the Reactor Coolant System response durirg the mid-loop event is a sirgle node non-equilibrium l!Ddel.

'1he node is divided into an upper region and a lower region. 'llle lJR?er region contains a mixture of steam am a non-comensible gas (air or nitrogen).

'!he lower region contains only water in a liquid or a pool boilirg state.

Water mass is transferred from the lower region to the upper usirg a Constant bubble rise noiel. Water mass is transferred from the tJF.Per region ~

the lower region through a droplet fall or a o:m:lensation nrdel.

A sinple orifice flow nx:xiel. allows one flow pa.th into or out of the ~

region. It either tranports the non-con:iensible/steam mixture out or draws my oontaimnent air back into the upper region.

A silililar orifice nrdel allows a drain flow pa.th to be defined for the lower region. It only allows transport out of the lower region whenever the vessel pressure is.greater than the cutside :reference pressure.

5

A time deperrlent liquid flCMIOOdel is included to simulate flooding of the reactor vessel with cold water by startirg a p.mp or openirg a valve to the ms'!'.

'Ibis nx:del feeds only the lower region.

Decay heat is m:xleled as a tima varyirg heat i.np..It added directly to the l0i.1er region water.

structural heat capacity is noieled in both the uwer am the l0i.1er regions. In each region, m.tl.tiple heat sinks may be IOOdeled with different heat transfer ex>efficients am heat capacitances. 'lliese are also used in the ui;per region as ex>rrlensirg surfaces.

'!he fluid node calculations use a 6 equation IOOdel.

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

upper region water vapor; upper region non-comensible; am lower region water.

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

• A set of auxiliacy calculations are then perfonned to solve for the in1ividual carponent enthalpies am for the camponent partial pressures _in the uwer region. -

'!he water properties (specific volumes am 'te!rperatures) are obtained usirg pieoe-wi.se linear steam tables, exten:ied to lc:M pressures.

Non-oorrlen.sibles are assunei to be ideal gases.

3.2-3 Description of Irpit arrl Assunptions Q:)nservative ~

assunptions have been made which maximize the core heatup rate am pressurization an:l minilnize the tima to lx>ilirg an:i core urx:overy *. A-number of these assunptions are explained belc:M.

1. At the time of mid-loop operation, the RCS 'ten1;>erature is 140 F arrl the 6

time after reactor shutdown is 72 llours* '!his is the shortest decay time am highest ~ture

~licable to the mid-loop ~tion conlitions.

2. At 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown, the decay heat p::Mer is 0.41% of full pc:1.Yer (102% of 3411 MWt).

'!his corresponjs to a decay energy of 14.265 MWt

( 48. 67 MBIUjhr) *

'lhe decay heat power was detennined usi.rg the ANSI/ANS-5.1-1979 decay heat starx1ard am includes 2-sigma uncertainty plus conservative est:ilnates for heavy element decay an:1 fission product absozption.

A high average rumup was also assumed. in these calculations (core average il:radiation time of 800 days or 30,000 MWD/MIU).

As a cross-check on the results, the ANSI/ANS-5.1-1979 values were then cxxrpared to a Westin#leuse 3-region core calculation an:1 fo..II'rl to be within 2% over the rarge of decay times of interest.

(For the 3-region core calculations, the irradiation times were 333, 667, am 1000 days.)

'lhe higher of the two decay heat powers was then used in the analysis.

Figure 3.2-3-1 shows the decay heat power as a function of tine after reactor shutdown *

3. 'lhe water volume assumed for the single node core boiloff m::xiel is c:::anprised of the water in the core am upper plenum (to the middle of the hot legs) plus one third of the water in the bottom half of the hot legs.

'!he water volmne assumed for the analysis is 1260 cubic feet. Water in the cold legs, downcamer, lower plenum, am remain::ier of the hot legs is neglected in the initial heatup calculation even tha.lgh sare of this water would eventually heat. For the Salem Plant during mid-loop operation, the water level is nomal.ly one foot above centerline am there is an alann if the water level drops to 6" above centerline. '!he additional hot leg arxl ua;:er plern.nn vohnne to the alann setpoint Which was oonse:rvatively neglected in the present analysis is roughly 200 cubic-feet, i.e., m:>re than 15% of the liquid volume assummed in the calculations.

4. '!he heat capacities arxl overall heat transfer coefficients for the fuel, metal structure in the~ internals, plus water am structure in the barrel-baffle region has been in::luded in the heatup of the lower 7

(mixture) region. Similar in?it for the upper internals, ~support plate, vessel metal, hot le:;J am nozzles, am steam generator inlet has been specified for the ~

(vapor) region.

s. '!he initial steany'gas voll.Jil)a used in the si..rqle rxx:le boiloff nxxlel was assured to consist of the remaimer of the hot le:;J am uwer plenum, the upper head, the SG inlet plernnn, an:i only one half of the SG tube volumes.

'llle bottan of the surge line is initially cx:wered with water, so initially the pressurizer voll.Jil)a is not in direct contact with the rest of the RCS (except possibly to the upper head through the PRl' if the upper head is venteci to the mr instead of air). F\\Jrthenrore, the initial RCS pressurization will cause.water to expan::l into the surge line without significant ~e in the steam volume. It is conservative to neglect the pressurizer volume c::arpletely for p..irposes of det:ermininJ a high RCS.

pressure. 'lhe down side of the SG tubes has also been neglected. If it is p:::stlll.ated that the non-con:iensible gasses collect at the top of the SG tubes ani restrict the flCM of steam, then the SG volume considered should be somewhat less than the whole voll.Jil)a.

One half the total SG volume will.

certainly be consei:vative.

6. In m:>st of the ana.Iysis cases, heat transfer to the~

is not inclu:ied to simulate the chy layup situation.

steam con:iensation is minllnized to consei:vatively maxllnize the pressurization. An SG corQ.ensation sensitivity study is also corducted to derron.strate-the effects of havin.;J m:::>re realistic secomacy heat transfer.

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

Figure 3.2-3-1 Decay Heat Power Versus Time After Reactor Shutdown

_J

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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown (.41% of full power) am a conservative (1260 o.Jbic-feet) lato1er region mixture volume (s.illlilar to NUR00-1269 methc:dolcgy) *

'lhe namenclatUre for the various plots presented is as follows:

P-total, P-steam, P-air : 'Ibtal, steam, an:i air (or N2) partial pressures (psia)

Upper arrl lower Temp Temperatures in the lc:Mer (mixture) arrl upper (vapor) regions (F)

I.ewer arrl Collapse Vol :

Lower mixture region volmre arrl

. equivalent volume if all voids

.were collapsed (cubic-feet)

Ir:M Vent FlOW' :

Vent FlOW' :

Lower region (liquid) break flOW'

( lJ:xtVsec) uwer region (vapor) break flOW' (lJ:xtVsec)

J.J-1 ease i - Base case Analysis In this analysis, :oo SG comensa.tion is m:deled arrl the Ra; is a.ssune:l to remain intact (no vent path) * 'lllis case can be used to estimate the maximum pressurization for loss of RHR cooling at mid-loop conditions, 10

with no water in the steam generators (di:y layup).

'!he time table of events for this case is given in Table 3. 3-1. '!be pressure, te:npera:ture, an:1 lower region volmne transients are presented in Figures 3. 3-1-1 through 3.3-1-3.

After ai:\\)rox:iinately 9 minutes, the RCS lower region volmne (core an:1 upper plernnn) reaches 212 F an::l the RCS starts to pressurize.

By 18 minutes, the RCS pressure exceeds 25 psig. Pressure. continues to increase to 100 psig at approx:iinately 30 minutes am exceeds 400 psig by one hour (Figure 3.3-1-1).

'!he lower region tenperature increases all!ost linearly with time at the rate of about 8 F/min tmtil saturation is readied (212 F at 9 minutes).

After this time, some of the decay energy produces steam but nuch of it still heats up the lower region saturated mixture. By 35 minutes, the RCS tenperature readies the RHR cut-in te:n;;>erature of 350 F (Figure 3.3-i-2).

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

• 11

Table 3.3-1 Time Table of Events Salem Loss of RHR Coolin;J at Mid-loop Operation Base case 1 - Intact RCS, No SG Corrlensation Loss of RHR Coolin; at Mid-loop Corrlitions Core am Upper Plenum Terrpera:tures = 140 F RCS Pressure = 0 psig (14. 7 psia)

Core am UJ;:per Plenum Terrperatures Reach 212 F RCS Pressurizes to 25 psig (39.7 psia)

Corrlitions at 2000 secx>ms (33.3 minutes):

Core Exit Tenpera:ture = 338.8 F RCS Pressure = 101.6 psig (116.3 psia)

RCS Intact with Collapsed Ievel AR;>roxinately J Inches Above Mid-Ii=q>

Core am t]Wer Plenum Terrperatures Reach 350 F RCS Pressure :Reaches 375 :psig (390 psia)

ErXl of Transient M:xieled 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)

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- - - P - AIR 600.0 500.0 400.0 300.0 200.0 100.0 0.0 o.o 500.0 1000.0 1500.0 2.000.0 2500.0 3000.0 3500.0 4000.0 TIME (SECONDS)

Salem Mid-Loop No Vent, w/o Condensation Figure 3.3-1-1 RCS Total *and Component Pressures - case 1

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

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Sale* Mid-Loop No Vent, w/o Condensation Figure 3.3-1-3 Lover Region Volume and Collapsed Volume - case 1 i

3.3-2 case 2 - Sensitivity to SG Corxlensation

'!he comensation heat transfer coefficient can vary between 10000 arxi 50 BIUjhr-ft2-F, depenling upon the amount of non-corrlensible gas present.

'!his section presents the results of two analyses which were perfo:aned to determine the effects of corrlensation on the RCS heatup arxi pressurization rates follc:Min;J the loss of RHR during mid-loop operation.

Both the overall heat transfer coefficient arxi heat transfer area are ilt'lportant in det.enninin; the annmt of con:iensation.

As noted above, the heat transfer coefficient is sensitive to the annmt of non-corrlesible gas present.

As steam corx:lenses on the corrlensing surface, a layer of non-corrlensible gas is left behirrl.

As this layer builds up, the corrlensation rate degrades since steam must first diffuse through this layer of gas. 'Ihus, even a srrall anomt of non-corrlensible gas in the system will cause the corrlensation rate to decrease significantly

• At !CM pressures, steam is less dense than air arxi will not :penetrate into the downhill section of.the SG tubes under natural circulation con:titions. '!his leaves approximately half of the SG tube area for corxiensation.

'!he effect of corrlensation in reducing the pressurization of the RCS after -

the loss of RHR cooli.ng was studied by perfor:m:iig two analyses in which both the comensa.tion heat transfer coefficient an::l area for corrlensation

-were varied. For case 2a, a corx:lensation heat transfer coefficient of 1000 BIUjhr-ft2-°F an::l 20% of the uphill SG area were used. '!his

-resulted in an overall UA of 900 mU/sec-°F 'Which-is believed to be realistic if the steam generators are filled with subcooled water. For case 2b, a corrlensation heat transfer coefficient of 100 BIU/hr-ft2-°F ani only 5% of the uphill SG area were used~ '!his resulted in a nv:::>re conservative estimate for the overall UA of 100 BIU/sec-°F.

Note, in both cases, all 4 steam generators were assuned to be filled with subc:xx:>led water to the nonnal operating level. 'Ibis assurrption influences 16

". the heat up rate calculation after saturation is reached am steady state con::iensation is taJd.rq place. If less water were asstmed, the heat up rate (am con.sequently the pressurization rate) durin;1 the steady state corrlensation phase would be higher.

Figures 3.3-2-1 am J.2-2-2 present the pressure am temperature canparisons. '!he effect of even the miruJmJm ancunt of cOn::iensation heat transfer is to reduce the RCS pressurization significantly.

'!he RCS pressure 30 mirrutes after the loss of :RHR went from 80 psig in the base case to 25 psig in case 2b am a psig in case 2a

• 17.

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120.0 tOO.O 80.0 60.0

.. 0.0 20.0 0.0 0.0 No Condensation small Condensation

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

Reasonable condensation

.. oo.o 800.0 1200.0 1600.0 2000.0 200.0 600.0 1000.0 1400.0 1800.0 Time (seconds)

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

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small condensation

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~00.0 800.0 1200.0 1600.0 2000.0 200.0 600.0 1000.0 1400.0 1800.0 Time (seconds)

Salem Mid-Loop Condensation Sensitivity Figure 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 sec:tiori 3.3-1 (no SG oon:iensation, RCS intact) was repeated, this time with a 3/4 inch diameter break in the vapor space. '!his is representative of the size of the reactor vessel head vent. It is also cxxrparable to the size of a tygon tube (used for vessel level :in:lication) that could potentially rupture due to the RCS pressurization. 'lhe 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.

Ccmparinj Table 3.3-1 to 3.3-3 arx:l Figures 3.3-1-1 through 3.3-1-3 to Figures 3.3-3-1 through 3.3-3-3, th9 smail vapor space break has a slight inpact on the transient. As noted in Figure 3.3-3-4, the vent flow rate is on the o:rder of 1.0 lbm/sec. '!his is an *order of magnitude sma1ler than the steam production due to decay heat. 'lhus, one (or several)* vent paths *of this size will not have a significant inpact on the RCS heatup am pressurization transient.

20

Table 3.3-3 Tilne Table of Events Salem IDss of RHR Coolin;J at Mid-Ioop Operation case 3 - 3/ 4 Inch Vent in Vapor Region, No SG Con:iensa.tion Event IDss of RHR Cooling at Mid-loop Con:iition.s Core Exit Tenperature = 140 F RCS Pressure = o psig (14. 7 psia) 3/4 Inch Diameter Vent Path in Vapor Region Core an:i Upper Plemnn Temperatures ReaCh 212 F RCS Pressurizes to 25 psig (39. 7 psia)

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

RCS Has Small Vent Flow, Collapsed Level is Approximately 3 Inches Above Mid-loop (same as Case 1) 21 Tiree sec Cmin>

0 (0) 546 (9.1) 1124 (18.4) 2000 (33.3)

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-- P - *STEAM

-- - - P - AIR 120.0..------------------------------------------------------.

100.0 80.0 60.0 40.0 20.0 0.0

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---

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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 3.3-3-1 RCS Total and Component Pressures -

Case 3

a.

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-- UPPER TEMP 340.0 320.0 300.0 280.0 260.0 240.0 220.0

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200.0

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180.0

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160.0

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140.0 LOWER TEMP

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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 '------------------------------------

0 >

1200.0 UJ en a.. 1100.0

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0 u 1000.0 UJ

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700.0 Top of Core 600.0 500.0 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 3.3-3-3 Lower Region Volume and Collapsed Volume -

Case 3

N UI u

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-- UPR VENT FLO 1.8 1.6 1.4 1.2 1.0 0.80 0.60 0.40 0.20 0.0

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. 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 J

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

'!his scenario is the same as the previc::us case except the break is located in the liquid region. '!his could 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 lbnVsec (36 gpn) which represents the flow expected at a distance of 12 feet below mid-loop (this coincides with the bottan of the crossover leg). Inventocy loss was nmeled for the entire duration of the transient. It should be noted, however, that after the level drains to the elevation co:rresporrling to the top of the reactor coolant p.lil1p weir ( 5" above the bottan of the cold leg or about 911 below mid-loop), the RCS inventocy loss for the core arrl upper plenum water could be expected to stop (a depletion of only a one to two hundred cubic-feet). '!he analysis also conservatively assumes that r

only the water in the core arrl upper plenum is lost am totally ignores d.rairrlown of the cold leg am downcanmer (this over-estilllates the break flow by al::xJUt a factor of two). '!he time table of events for this case is given in Table 3.3-4 am parameters of interest are presented in Figures 3.3-4-1 through 3.3-4-4.

nie to the reduction in inventocy, the time. to boiling is reduced when c::cmpared to the* base case. Additional steam production. is also predicted which in turn 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 corrlensation cases, break flow would not increase as fast)

• However, level is still nuch higher than the active fuel at the errl of the transient (Figure 3.3-4-3). In view of the oonservatisms-n:>'ted above, the actual volume at the errl of the transient wruld be expected to stabilize at a nuch higher value. 'Ihus, the expected behavior for the 3/4 inch liquid break case wruld be si:milar to the corresponding case withcut the the break

• 26

_J

Table 3.3-4 Time Table of Events Salem Ios.s of RHR Coolirg at Mid-loop Operation case 4 - 3/4 Inch Vent in Liquid Region, No SG Corxlensation Event Loss of RHR Coolirg at Mid-Loop Conlitions Core Exit Ter!perature = 140 F RCS Pressure = 0 psig (14. 7 psia) 3/4 Inch Li.quid Flow Path (5 n:mvsec Break Flow)

Core am tJH?er Plernnn Ter!peratures Reach 212 F RCS Pressurizes to 25 psig (39. 7 psia)

Errl of Transient Modeled Core Exit Teirperature = 349.5 F RCS Pressure = 119.1 psig (133.8 psia)

RCS Break Flow= 23.7 n:mvsec, Collapsa:i Ievel is More 'lhan 3 Feet Above Top of Active FUel 27 Time sec Cminl 0 (0) 537 (9.0) 1079 (18.0) 2000 (33.3)

IC M c I

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TOTAL

--P -

STEAM

- - - P - AIR 140.0 120.0 100.0 80.0 60.0 40.0 20.0 0.0 0.0

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

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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 Fiqure 3.3-4-1 RCS Total and component Pressures - Case 4

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---

UPPER TEMP 350.0 300.0 250.0 200.0 150.0' 100.0

/

- - - LONER TEMP

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I 0.0 400.0 800.0 1200.0 1600.0 2000.0 200.0 600.0 1000.0 1"400.0 1800.0 TIME (SECONDS)

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

• case 4

0 >

l&J en D..

-c 0 u l&J x

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s cc l&J

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1500.0 1*00.0 1300.0 1200.0 1100.0 1000.0 900.0 800.0 700.0 600.0 500.0

-- LONER VOLUME

- - - COLLAPSE VOL*

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

Top of core 0.0 AOO.O 800.0 1200.0 1600.0 2000.0 200.0 600.0 1000.0 1.COO.O 1800.0 TIME (SECONDS)

Sale* Mid-Loop 3/A Inch* Liquid Vent Fiqure 3.3-4-3 Lower Reqion Volume and collapsed Volume - case 4

25.0 -- LOW VENT FLO 20.0 u

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it 15.0 m

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

3.3-5 case 5 - Analysis With I..arge vent Path F,'or this analysis, a large 16 inch diameter vent path is m:xleled in the vapor region. '!his 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 paraneters of interest 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 approximately one atm::>Sphere for the* duration of the transient (Figure 3.3-5-1). Note that after boiling occurs (again after 9 minutes), all the air is expel.led am the steam partial pressure becomes the same as the total pressure.

As expected, the RCS boils at :near atnospheric pressure am the temperature is maintained at an approxilna.tely contant value of 212 F (Figure 3.3-5-2).

'!he vent flow approaches the constant boiloff rate

~g = 13.94 l.bm/sec (Figure 3.3-5-4).

('!he equivalent makeup flow requirement is approxilna.tely 100 gpm.)

D.le to the nearly constant boiloff rate, the volurre also decreases at an approxilna.tely 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 downcomer 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 exterrled further, by nore 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 Ioss of RHR

• Ox>lirg at Mid-Loop operation case s - 16 Inch SG Manway Vent, No SG Con1ensa.tion Loss of RHR Ox:>lll'g at Mid-loop Corditions Core. and Upper Plenum Temperatures = 140 F RCS Pressure = O psig (14. 7 psia)

Core and Upper Plenum Temperatures Reach 212 F revei SWel1 Reaches a Maxini.nn and start:S to Decrease

'Vapor Region Temperature R0aches ?12 F, Air Partial PressUre I.ess '1han o.1 psia Collapsed tevel Reaches Top ol. Active FUel Errl of Trans.j.ent MJdeled RCS and Vapor Temperature = 212. 7 F RCS and Vapor Pressure ::::: 14. 9 psia Boiloff Rate ~roximately 13. 5 l.l::atVsec (13.9 lbnVsec I..orq Tenn Core Boil.off) 33 Time*

see Cminl 0 (0) 545 (9.1) 586 (9.8) 621 (10.35) 3500. (58.3) 4000 (66.7)*

z c II.I t-en I

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w..

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TOTAL

--P - STEAM

- - - P - AIR 14.0 12.0

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8. 0 6.0

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500.0 I

1000.0 I

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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 U1

n.

x UJ t-cc w

z 0

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a.

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220.0 210.0 200.0 190.0 180.0

. 170.0 160.0 150.0 140.0

-- UPPER TEMP

- - -- LOWER TEMP

---~-----'-*-* ---*----*-----****-**** *'*****-----. *-*** ----*-*- ---~

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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 Inch Vapor Vent Fiqure 3.3-5-2 Mixture and vapor Reqion Temperatures - Case 5 j

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- - - COLLAPSE VOL 2000.0

[-------~----~-***-*--***-**--*-

• 1eoo.o 1600.0 r 1400.0 I

~

1200.0

~

I

~

1000.0 ~

I aoo.o ~

'~

-~--

~

Top of core *

~---------. ---------~:::--

600.0

..-oo.o 0.0 500.0 1000.0 1500.0 2000~0 2500.0 3000.0 3500.0 4000.0 TIME (SECONDS)

Salem Mid-loop 16 Inch Vapor Vent Fiqure 3.3-5-3 Lower Region Volume and Collapsed Volume - case 5

u cu UI Iii

.a --

]t 0 -

w c m

-- UPR VENT FLO 14.0 12.0 -

10.0 8.0 I-.

6.0

• .o...

2.0 I-0, 0 F~--___,

-2.0 0.0 I

500.0 I

I I

I I

I 1000.0 1soo.o 2000.0 2500.0 3000.0 3500.0 *ooo.o TIME (SECONDS)

\\

Sale* Mid-Loop 16 Inc~ Vapor Vent Fiqure 3.3-5-4 Upper Region vent Path Plowrate - case s

3. 4 summary of Results

3. 4-1 RCS Heatup an:i Tirre to Saturation Based on the previous analysis, the RCS heatup to 212 F occurs at an approximately constant rate. 'Ihis rate is proportional to to the decay heat p::Mer an:i inversely proportional to the thermal capacities of the water, fuel, an:i structure in the core an::i upper plenum regions. 'Ihe RCS heatup rate an:i tine to reach saturation can thus be determi.ne1 at various decay heat rates corresponiing to different times after reactor shutdown.

'Ihese resultes are plotted in Figures 3.4-1-1 an:i 3.4-1-2. 'Ihe cases analysed in this report (startirg fran 140 F initial RCS temperature, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after trip) were chosen to conservatively calculate the heatup an:i pressurization rates following the loss of RHR cooling. 'Ihese cases reached saturation ra.ighly 9 minutes after trip as expected based on the initial con:titions given. If RHR cooling is lost later than 72 hoUrs after trip, or the initial RCS temperature is lower than 140 F, the heatup rate will be lower an:i tirre to reach saturation will be longer.

3.4-2 Core Uncovery

'Ihe minimum tine to core uncovery following the loss of RHR cooling while in mid-loop operation was conservatively estinated to be one hour. 'Ibis analysis assumed a loss of RHR coolirg 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after reactor shutdown, an open SG manway vent* path for steam flow, no steam generator con1ensation an:i no operator recovery actions. It shool.d be pointed cut that a loss of RHR cooling does not nea:ssarily lead to core uncxNery in all cases since the various operator recovery actions have not been analysed.

3.4-3 RCS Pressurization Rate

'Ihe maximJm RCS pressurization rate (fran 14. 7 to 39. 7 :psia) was conservatively estimated to be 2. 77 psi/min while the RCS is saturated.

38

'!his m:nservative estimate was based on a loss of RHR CX)Qlin;J 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after reactor trip, no con::lensa.tion heat transfer to the_ steam generators, am no operator r:ecovery actions.

'!he case without a:ey open vent paths was ccmpared with two cases with 3/4 inch vent paths open to detennine if steam or water ventin;J 'WOUld have any significant effect on the pressurization rate. No awreciable difference in pressurization _rate was fam:l for the vapor vent path case, l:ut the liquid vent path pressurization rate was slighly higher since less water was available for sensible heat addition. 'lherefo:re, the calculation of a cxmse.rvative pressurization rate is based on the results f:ran the 3/4 inch liquid vent path analysis

• 39

'2..

• ii: *

0... I.

r....

• I I.

I I

20.0 tl.O tl.O t4.0 t2.0 so.o 1.0 1.0 4.0 2.0 o.o o.o

\\'

\\

~

90.0 Flgurt 3.4-t-t.

Hlttup Ritt for Lo11 119' Caallnt During Mld-t.aap Ollerttlan

~

~

~

too.o t90.0 200.0 Tl* Afttr Altetar lllutdnn lhr*J 290.0 300.0

"2.....

I..... *

!s I

ii

• I 0..

Flll'l'I 3.... S-2 Tl* to S.tW'1Uan for La11 of "" Cooling Dartn1 Mld-1.aop 0p1r1tton

• .o--* --------.---------..-------------------...... -------------------

a.o TRCS

• 100°1' 20.0 S!l.O TRCS

• 1400p 10.0 0.0'--------------------"'----------'----------...... --------...... ---------

0.0 IO.O 100.0 SIO.O 200.0 250.0 300.0

/ Tl* After lhutdcnll 0..1)

As pointed out earlier, one of the i;x:ssi.ble recovery actions to restore RCS inventory and RHR ftmction is to gravity drain water from the :ms'l' to the RCS.

'!his recovery action is only effective if the RCS pressure is less than the head f:ran the RWST.

If a 25 psig RWST head is assumed, the RCS pressure would e.xoeed this within 20 minutes after the loss of RHR cooling.

'!he analyses presented in section 3. 3-2 shoil that the RCS heatup and p:reSsurization rates are very sensitive to the annmt of condensation assumed.

Even a small am::Jlll'lt of condensation can increase the time available for operator recovery actions.

3.5 Conclusions A conservative analysis was performed to detennine the mininn.nn time to core uncovery following a loss of RHR while in mid-loop operation.

An open manway vent path for steam was assumed and no credit for either reflux condensation or q;erator recovery actions was assumed in this conservative analysis.

Based on this analysis, it was determined that the mininn.nn time to core unccvery following a loss of RHR CX)Qling at the Salem Nuclear Plants while in mid-loop operation is approximately l hour.

Following the loss of RHR c0oling, operator recovery action is necessary to either restore a heat sink or provide* adequate makeup water into the RCS to prevent core urxxwery.

Makeup water can be added by either gravity feed fran the :ms'!', the charging system, or Mr:f other available high pressure injection system. A heat sink can be restored by either refilling the steam generators or restoring RHR flow.

Pressurization of the RCS may prevent gravity feed fran the :ms'!' as a nw:ans of recovery. A separate analysis to detennine the :maximum pressurization rate was also perfo:rmed.

No vent paths were assumed to be open and no credit for either reflux condensation or operator rea:Nery 42

~ctions was assumed in this analysis.

Based on this consez:vative ~ysis the RCS is predicted to pressurize to 25 psig in approximately 20 minutes f ollowin;; a loss of RHR cooling at the Salem Nuclear Plants while in mid-loop operation. When the effects of SG reflux corrlensation were mXlelled, it was dem:>nstrated that the RCS pressurized at a nruch slower rate than the consez:vative analysis predicted. 'lherefore, it would be beneficial to keep one or 11Dre steam generators partially full of water while in mid-loop operation.

3.6 References

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

2. us NRC Augmented *Inspection Team 'Report "loss of :Residual Heat Removal System - Diablo canyon, Un.it 2 - April 10, 1987," NUREX;-1269, May 1987.

4. 0 TASK 2 -

RADIOLOGICAL CDN~CES 4.1 Description of Task 2.

Task 2 was interned to provide an evaluation of the radiolcqical consequences of a loss of decay heat rerooval event during operation with the RCS partially filled. 'lhe task description as it was presented in the technical description is _contained in Apperd.ix B., 'lhe task consists of providing a plant specific detennination of the off-site thyroid ani whole body doses that 'WOUld result due to evaporative losses of reactor coolant ani the resultant releases of icxlines ani entrained gases as the c6olant boils in absence of RHR System forced flow.

'lhe intent of this task is to

• deroonstrate that the off-site doses due to the event would present no significant hazard to the plblic

• 43

It was requested that the off-site dose evaluation take into ac:x::ount the possibility that the cxmtairnnent bcurxiacy might be open for passage of activity to the environment.

A calculation of site bourxlacy doses was made assuming that full containment p.n:ge is operating an:l a separate calculation of the ail:OO:rne activity concentrations in the containment was made assuming that the containment is closed~

'Ibe Plant & Systems Evaluation Licensing (PSEL) group of the Westinftlouse Nuclear Safety Depart:Irent corrlucted the evaluation an:l has provided the following disaJSSion as presented in sections 4.2 through 4.4.

4. 2 Task 2 Assunptions an:l Bases For the p.rrposes of detenninin:J the activity in the reactor coolant, it was assumed that the mid-loop operation takes place in IDDE 5 am the RCS is vented with the reactor coolant degassed to a concentration of o. 5 micro Ci/g of Xe-133.

'!his is a factor of ten greater than the value recamrerxied by Westirghouse for the canpletion of reactor coolant degassing operations an:l is five times greater than the concentration present at the initiation of the biablo canyon event.

'lllree I-131 concentrations -were considered:

case 1

'lhe I-131 concentration as required by the technical specification limit of 1. O micro Ci/g.

44

case 2

'Ihe I-131 concentration of 0.1 mien:> Ci/g which is ten times the value specified in typical operati.rg instructions to pennit head lift.

case 3

'1he I-131 concentration of 0.02 micro Ci/g whidl is twice the.

value specified in typical cperati.rg instructions to pennit head lift.

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

In the determination of the activity released to the containment it is assi..nned that all of the Xe-133 in the total RCS volume is released to the c::Ontainrnent 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 containment (recx:we:ry of the event is assumed to occur before the core is unc:xJVered).

In the determination of the site bourrlacy doses it is assi..nned that the contairnnent is open ard that all of the activity released due to the event is released durixg the first two boors.

'lhe FSAR X/Q value of s.o x 10-4 sec;m3 was used in the evalutation.

4. 3 Task 1 Activity Coramtrations in the o:>ntainment Atm:sphere

'!he results of the evaluation for the containment activity concentrations assumin:J that the contaiment is not open to the environment are presented below

• 45

TABIE 4.3-1

. a:NrAlNMENl' Acr!IVrr'f ISOIOPE ACI'IVITY

~133 6.3 x lo-4 micro Ci/ml I-131 (1.0 micro Ci/g in coolant) 4.05 x 10-4 micro Ci/ml I-131 (0.1 miC:ro Ci/g in coolant) 4.05 x io-5 micro Ci/ml I-131 (O. 02 micro Ci/g in coolant) 8.1 x lo-6 micro Ci/ml 4.4 Task 2 Off-Site Doses

• '!he results of the evaluation for site bourrlary doses assumin:J that the contaimnent is q:en to the. environment are presented belCM:

I-131 at 1.0 micro Ci/g I-131 at 0.1 micro Ci/g I-131 at 0.02 micro Ci/g

'mBIE 4.4-1 srm~zxes Doses, rem

. Whole Body 1.53 x 10-3 3.04 x 10~

1.95 x 10-4 46

'lhyroid 7.378 0.737 0.147

In summacy, if the loss of RHRS during mid-lc:x:p operation is considered an accident, oc:mparison with the 10 CFR 100 limits shows that the calculated doses are 'Well urxier the "small fraction of 10 CFR 100 limits" (30 REM thyroid arrl 2. 5 REM whole body) which the NRC assigns for accidents which have the greatest probablility of oc:x:m:ence.

Alternatively, the calculated doses can be c:::aipared with the dose limit of 0.5 rem whole body (or its equivalent to aey part of the body) which is identified in Regulato:cy Guide 1.26 as the limit for specify:i.rq if equipnent is to be categorized as Quality Group C or D (equivalent to Safety Class 3 arrl NNS).

In other 'WOrds, if a canponent failure walid result in a dose of less than o. 5 rem whole body, the i.Jrpact :i,.s considered not to be of significance to safety arrl the cxmiponent can be classitied as a Non-Nuclear Safety Ccrnponent.

Based an the weight.:i,ng factors provided in ICRP Publication 26, 0.5 rem whole body is equivalent to 16. 7 rem thyroid. In all of the cases the calculated doses are less than 0.5 rem whole body arrl 16. 7 rem thyroid. It is note:i that ICRP Publication 26 recamnen::is an~

dose lllnit of 0.5 rem whole body arrl also recc:.mmenjs a limit of 5. o rem to any *organ.

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

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

It sh~d be noted that durin;J the event at Diablo canyon an increase was ci:Jsel:ved in the co_ncentration of xe-133 in the contairnnent atltos];tiere while little c::harge was observed in the corx::entration of I-131.

'lbe assmrption that 100 percent of the iodine contained in the coolant assumed to have evaporated beca:DeS aimo:rne is considered to be an extremely conservative assurrptian

• 47

5. 0 TASK 3 -

ASSESSMENI' OF VOR1'EXING AND AIR ENmAINMENI' 5.1 Description of Task 3 Task 3 was inten:Ied to provide an un::ierstarxlin;J of the ~

of vort:exirg an:i air entrainment as it relates to the RHR System.

Additionally, a review of Salem design features was to be made an:i oc:mments provided.

'!he task description is contained in Appe.rrlix B.

In addition, the assessnent was to include reoc:mnerx3ations for operatirg procedures to limit the potential for loss of RHR system durin;J mid-loop

~tion an:i to respon:i to the loss of the RHR system should that occur.

5. 2 Task 3 Corx:lusions

'!be Safeguards Systems (SS) grrup of the Westinjiouse Systems Engineerirg Department has provided the following di sa.ission of the P"ienanenon of vort:exirg an:i 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 beoames a concern as there is the potential for drawirg air into the p.mp suction due to Res loop level fluctuations an:1/or development of a vortex. If air bi.n:iirg of the the RHR p.mps occurs, RHR system capability may be lost.

However, the RHR system i.nst.Iumentation provides a number of in:iications that would provide rapid detection of air bin::lin:;J of the pmps such as fluctuations in the in:iications of RHR pmp m::>tor current, suction pressure (PI-631 & PI-632~, an:i discharge flCM (FE-64lA & 641B).

Additional RHR system perfonnanoe ncnitorirg capability an:i.inproved RCS level m::>nitorin;J capability as 'Well as i.nprovements to operatin;J procedures to limit the likelihood of air entrainment in the RHR p.mps an:i to provide guidance for a rapid an::i effective response to loss of RHR system operability waild enhance the safety an:i reliability of the 48

Salem Units while operatirg with RCS level at mid-loop.

'Ihe foll~

paragraphs. identify sate operating procedure guidelines to help preclude air bi.rxilng of* the :mm. pmp and to properly respom to loss of the :mm.

pump due to air binling should it occur. Also identified are additional capabilities for ioonitoring the :mm. system performance and RCS level which, if added, 'WOUld improve the ability to avoid entering operating corxlitions in which air entrairnnent 'WOUld be likely to occur and to pratpt.ly identify its oocurrenoe.

While the

• developnent of detailed operatirg procedures is the responsibility of the utility, Westinghoose Operating Instnlction M-1, "Draining the Reactor Coolant System" recammends guidelines tO follow* when.

draining the RCS for mallrt:enance or refueling. 'lhis guideline recammends that a tygon hose be connected to the drain line of the reactor *vessel flange leakoff: connection.. 'Ihe water level in the reactor vessel can be

• clOsel.y IIDnitored by means of this tygon hose.

However, the reactor coolant loop water level cannot be IIDnitored by this azrangement.

To provide a. llEal'lS of IIDnitoring 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, instJ:ument connection) of one of the four reactor coolant loops and be extemed at least two feet above the top of the pressurizer, where it 'WOUld be vented to the contairnnent aboosphere.

'!he water level in the RCS loops can then be ioonitored by means of this secom tygon hose.

'!he following are precautions that should.be follaNed..to prevent the loss of :mm. capability when the RCS has been lCMered (to the hot leg

. centerline) to drain the steam generator tubes for maintenance

• 49

1.

'!he RHR. flowrate should be reduced to 1800 gpm or less to preclude vortex fonnation at the RHR hot leg connection (Developrent of this flow limit is djscusse.d in Section 6.2).

2.

ruring steam generator tube draining,* the water drains in a slugging fashion, there!fo:re RCS water level indication may be erratic. For this reason, the draining operation should be stopped periodically to allow the water level in the system to stabilize.

3.

4.

'!he water level should be m::>riitored continuously while pertuming the RCS inventory to assure that the RHR System inlet line does not become uncovered and gas birrl the RHR pumps.

When RCS draining is from the loop which is provided with the tygon hose level indicater, the draining operation should be stopped when it is desired to obtain an accurate reading. If drainirg is fram. a reactor coolant loop not provided with the tygon hose, proper CXJl'l'IIIl1ll'cation should be established to coordinate t:he draining operation with the level m::>nitoring.

'!he Salem RHRS (Figure 5.2-1) consists of a single drop line connecting two parallel and identical trains, each consisting of an RHR pump, RHR heat exchanger am the associated piping, valves and i.nst:runentation required* for operational control. '!he system is sized to pennit RCS cooldown from 350°F to 140°F in 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> with both RHR trains in operation.

However, after cooldown is conplete, only one RHR train is required to provide adequate :residual heat reroval capability; thus one of the two RHR trains may be out of senrice during this period. Following RCS drainiown, if the RCS water level should, fluctuate or be inadvertently reduced to the point at which air is drawn into the suction of the operati.rq RHR pump, resulting in air bin:linJ, the I."lll1l'rinJ RHR pump must be stopped (do not 50

ts

~ ;;q

~ I j~

I 44

~

ii n ~z-~li I I I I I' I I I I I 41 I II I

• I I

I 11 I

~*

~*

(J1.

N I....

' It

~

i iii 1*1 II

~g It 111

start the mWnpaired RHR p.m!p), the water level should be restored an:l the

~

RHR pump should be then started.

'lhe gas-bouni pump may be restored to operational status by ventin;J the air from the pump casing an:i cormecting piping. '!his can be accomplished by refloc::x:lin3' the pump suction piping. am by opening the manual vent valves on the pump suction an:i discharge piping.

Once this is done the pump may then be re-started as req.rlred.

'!he preceding steps should be completed quickly such that minimal interruption in RCS cooling will be experienced.

In addition to closely oonitoring RCS loop water level the follc:Ming in:lications could be incorporated into a RHRS punp protection system design to prevent am;or provide rapid iniication of air bin:lirg of the RHRS ~. '!his RHRS pl.IlTp protection system would consist of a microprocessor which utilized RHRS iniications, such as RHR flCM, suction am discharge pressures, RCS suction water level, RHRS pump current an:l other available irrlication to calculate the margin to loss of suction corxiltions.

An alann would be provided to alert the control roam operator to take corrective action before the pumps air bi.rxi.

In:lications of abno:nnal RHR purrp operation are curreritly provided by readout of the purrp discharge pressure an:i punp current on the Main eontrol Board an:l by remote readout *of the p.mp suction pressure.

In sunmmy, to prevent air birrlin;J of the RHR pimps, it is.important to keep the RCS loop water level above the hot leg centerline.

Because of this requirement, it is ilnportant that close oonitorin;J of the RCS loop water level be. maintained during aey operations involving re:i_ucing the RCS water level belCM the top of the RCS hot leg. If air birxiing of the RHR pumps occurs, reprimirg (filling am ventirg) the p.nnp should be canpleted as quickly as possible to preclude the loss of RCS coolin;J for an extenled period of tilre.

In tl'ie event the RHR purrps cannot be restarted, the followin:J should be.

considered as options available to restore cooling of the RCS:

52

1.

Use the dlarg.ing system to refill the RCS and establish a natural circulation cooling path, :rerroving decay heat via the steam generators and the Auxiliazy Feedwater System.

2.

Use the dlarg.ing system to refill the RCS and establish circulation by starting one or ioore reactor coolant pumps * (the starting criteria am precautions for RCP start should be met);

as in Option (1), decay heat :rerroval would be provided by the secorrlacy system.

3.

Allow the core to boil-off coolant and use the charging system to supply makeup to the RCS; this may be a viable option only if the RCS is open to the contairnnent atioosphere *

6. 0 TASK 4 -

DEI'ERMINATION OF VORI'EX LEVEL 6.1 Description of Task 4 Task 4 was inten:ied to provide an evaluation to detennine the sensitivity o~ RCS loop water elevation to critical submergence depth as a function of RHR System flow..Additionally, a cala.ll.ation is to be made to detennine acceptable reduced RHR System flow dur.ing mid-loop operation. Appropriate FSAR and Technical Specification C'hanges along with accarnpany.ing 10CFRSO. 59 and Significant Hazards Evaluations are to be provided to reflect the throttled RHR System flow rates during mid-loop operation.

'!he task description as it was presented in the technical description is contained in Apperdix B

• 53

6.2 Task 4 Conclusions

'Ihe Safeguards Systems (SS) group of the Westin;house Systems En3'ineering Departmant has c::onjucted a review of RHRS operation an:l design for the Salem nuclear units. 'Ihe review considered minimum RHR flow :necessai:y to:

1) rem:::we decay heat to maintain RCS terrpera.ture, 2) preclude boron stratification an:l 3) provide adequate fla,.r rate for boron dilution accident concerns.

'Ihe effect of RHR flCM an:l RCS level was *evaluated to determine susceptibility for air entra.iranent. It should be noted that, in all cases We.re RHR flow is identified, it refers to RCS delivered flc:Mra.te. If the miniflow path is opened this should be acca.mted for by operating the :pmp at the flCM requirement plus minifla,.r.

HFAT ROOvAL

'Ihe primacy. function of the RHRS is to rerrove decay heat after a

. reactor shutdown.

'Ihe anount of decay heat rem:::wal necessacy to.

maintain the RCS terrpera.ture constant or decreasing is a function of time after initial reactor shutdown.

Table 6.2-1 presented below lists the required RHR flc:Mra.te requirements as a function of t:iine after shutdown.

TAB!E 6.2-1

'lOrAL RHR FLCM VERSUS TIME AFTER SHtJI'Inm TIME AFTER SHtJI'In2N

'lOrAL RHR FI.CM m InJRS mc;m 18 3000 42 1800 54 1500 78 1250 114 1000 54

Figure 6.2-1 depicts the data presented in Table 6.2-1 in graphic fonn.. It should be noted that the Salem Tedm.ical Specification Limit of 3000 gpm (SUrveillance Requirement 4.9.8 for Unit 1 arrl 4.9.8.1 for Unit 2) is sufficient to maintain RCS te.ITpera:ture as early as 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> after plant shutdown.

As the time after shutdown increases the decay heat rerroval :requirenents 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 renoval requirements to reduce the RHR flowrate.

roMP NPSH Punp NPSH was evaluated at the 1000 to 3000 gpril flowrates (assuming RCS at 200°F) arrl. it was determined that sufficient p.mp suction head is available in this range.

OORON ST.RATIFICATION Since no intentional changes in boron concentration are made during refueling, the only nechanism which could i.muce a gradient in boron concentration is local mass evaporation~ However, the fluid temperature during m::de 5 arrl 6 operations is below 200°F, arrl. the evaporation effects are minil1lal.; also the fluid temperature is well above the precipitation value for the expected range of concentration.

'Ihe potential for boron stratification has been evaluated for a total RHR flow of greater than 1000 gpn arrl. determined to be acceptable.

'Ihe bases for preventirg boron stratification in the RCS is to minhni.ze the potential for a boron dilution accident. A flow of greater than 1000 gpn 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 prbnary circuit.

55

~v 0

0 0

M

-/

J

~'

)

I I

7 v

/

0 0

0 N

TOTAL RHR FLOW (G~M) 56 l

I 0

0 0 "

0 0 "

0 O'I 0 co 0

~

0

'° 0

in 0

~

0 M

0 N

~

es I

~.

~ -

0

~

~

i c

f'4

~

tll

~

. p:

~

8

~

~

~

~

~

E:

H 8

I N.

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~

8 12 M

~

OPERATIONAL BAND M

t4 9

to 0 <

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c 0 8 BORON

~

~

6 C>

n M z 8

DIWTION

/

LIMIT VORTEX M

~

3 H

~

LIMIT

---,i z

M H z n

i: -

0 1000 1800 2000 3000 TOTAL RHR SUCTION FLOW (GPM)

FIGURE 6.2-2:

RCS WATER IEVEL VERSUS IDI'AL RHR SUCTION FLCM

In view of the prece:ting, it is concluded that there is no concern for boron stratification at the RHR minimum flowrate of 1000 gprn.

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 1800 gprn but not greater than 3000 gpn, the suction water level shall be maintained at least 6 inches above the hot leg centerline. '!his irx:rease in RCS water level has been determined to be conservative for the flow ran;;re of 1800 to 3000 gprn.

It should also be noted that during mid-loop operation it is recommerxied that only one train of the RHR system be in operation.. At the reduc00 RHR flow rate recarnnetXied (1000 - 1800 gpm), it would be diffio.llt to read the flow rate irxilcation if both :P,mips were in use because the flow element is calibrated to the nonnal RHR flow rate of approximately 3000 gpm an:i the flow rate in each train would be in the rarge of 500 - 900 gpm.

Also, if air entrainnentdid occur while both ?JI11PS are in operation, it is likely that both would be lost fran service.

~potential air entrainment, current Westinjlouse guidelines (Salem Reference Operating Instruction M-1, Rev. 1, dated September, 1975) recommerxi reducing the RHR flow to 1500 gpn per* RHR :p.1lTp ~

the RCS water level is lowered in the loops. 'Il1is original recanmen1a.tion was based on engineering judgement an:i q>era:ting experience.

westinghouse has re-evaluated this limit based on previous research, available literature, scaled test results, an:i additional cperatirq histo:cy an:i has determined that the 1500 gpn limit, should be revised to

• 1800 gpn (total} with no correspomin;J iocrease in RCS water level. It is noted that the recamrnen:led flowrate of 1800 gpm is only a qualitative limit based on the criteria to maximize available core cooling 58

-I I

while providing ad.equate assurances against unacceptable oonsequences of.

vortex fonna.tion *.

Although the analytical techniques derived from test results do not entirely preclud.e same degree of vortex fonna.tion an::l air entrairnnent at 1800 gpm, operating experience supports the Position that, at this reduced flowrate, pump perfonnance an::l operability is not adversely inpa~.

However, given that a partial vortex fonna.tion at 1800 gpm cannot be precluded, Westinghouse does suggest that strict administrative/procedural assurances be inco:cporated to ensure that maximum punp flCMate arxl minimum RCS level limits are not exceeded.. In. addition, since any increase in RCS level or decrease in pump flowrate provides significant benefits in suppressing vorteX fonna.tion the follCJWi.n;J suggestionsi are provided:

1)

Mid-loop operation should be limited to only thoSe times required. Whenever :r;x>ssible RCS level should be mamtained at

'the highest :r;x>ssi.ble eleVa.tion an::l preferably with the RCS hot legs water solid.

2)

RHR flowrate should be maintained between 1000 an::l 1800 gpm, at the :mininrum flowrate required tO keep up with the decay heat load. 'Ibis requirement should be inco:cporated into plant operating pi"ocedures.

In sununacy, Westinghouse ooncludes that 1800 gpm RHR flow is an

. appropriate limit to prevent unacceptable vortex fonna.tion.

However, it is recommerrled that system perfonnance be oontinuously ioonitored to detect the onsiet of vortexing an::l air entraimnent.

59

6. 3 Boron Dilution Concerns Associated With Reduced RHRS Flow

'Ihe Transient.Analysis II (TAI!) group of Westinghouse Nuclear Safety Department has* con:lucted a review of the imPact of throttling the RHR flCM during mid-loop operation arxi has provided the follc:Min;J discuSsions concentlng possible boron dilution concerns.

'Ihe Salem Units have adopte:i the guidelines of NS-'IMA-2273,issued on July 8, 1980, {Interim Operating Procedure For Boron Dilution In IDDFs 4.Arrl 5) to assure adequate operator action time for a boron dilution event in IDDFs 4 arrl 5. For each reload a curve of boron concentration verses core burnup is gerierated.

'Ihe curve establishes the boron concentration required in MJDFs 4 arrl 5 to ensure adequate operator acti6n ti.Ire is

• available should a boron dilution event occur in these m:xies of plant operation.

'Ihe anal:Ysis is sensitive to RHR flow rate. A wide range of RHR flCM rates in IDDE 5 would result in the need for IOC>re restrictive boron concentration curves than currently in existence.

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

'Ibis data will be applicable for IDDFs 4 arxi 5 for the range of RHR flow rate from 1000 gpm to 4500 gpm.

'!his range will be sufficient to CXNer the flow rates for mid-loop operation as discussed in section 6.2 above arrl also will be applicable for full-loop operation. 'lhe revised boron concentration curves will be transmitted urrler a separate CXNer letter.

'lhese requirements should be incm:porated into plant operating procedures along with the vortexing arxi boron stratification requirements of section 6.2 above.

It should be noted that the IDDE 6 boron dilution analysis is not impacted by the RHR flow reduction dismsse.d in section 6.2 of thiS report.

'Ihe Interim Operating Procedure is. not used for the boron dilution event in.

IDDE 6

• 60

6.4 Technical Specification Olan;Jes

'Ihe Technical Specification Services (TSS) group of westin;house Nuclear Safety Departnent has c:orxlucted a review of the ilrpact of throttling the RHR flow durirg mid-loop operation an:l has provided the followin;J di SCllSSions conc:e.miig possible technical Specification ~.

• 'Ihe Salem Technical Specifications contain the follc::Mirg requirements:

A.

Mode 5

1.

Operability of two RHR loops arxi operation of one (four

• filled RCS loops an:l two steam generators with level ~ 5% of narrc::M ran:Je 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B.

Mode 6

1.

Operation of at least one RHR loop.

2.

Verification that at least 3000 gpm of reactor coolant is beirg circulated via the RHR system alee per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

3.

Operability of two RHR loops when the level in the refueling cavity is less than 23 feet above the vessel flange.

c.

All Modes (Salem-Unit 1 Only)

1.

A flc:Mrate of~ 3000 gpn through the RCS.

2..

Verification that one reactor coolant p.mp is in operation or that one RHR punp is operatirg and supplyin:;J ~ 3000 gpm through the RCS

• 61

'!he Salem Technical Specifications do not (and in reality cannot) disallow operation of the RHR system with the loops drained to mid-loop.

'!he tedmical specifications do place limitations on the RHR system when operati,m with the RCS drained to mid-loop in the form of l.iJnitations on the number of loops which must be operable.

'!he technical specifications do not contain restrictions :based on min:iJDizi,m air entrainment in the RHR system as a resu1 t of vortexin; which may occur durirg mid-loop operation urrler certain corrlitions. '!he minilTlum flow requirement is for the p.n:pose

• of decay heat J:'el1KJVa1 and is considered conservative. 'Ibis flow rate has subsequently been used as an inp.It to the boron dilution analysis.

Reqli.red RHR flow is a :function of many factors in:::luding tllne after shutda;.m., boron dilution and stratification considerations, level in the refuelirg cavity, RCS pressure and tenperature, and level of the reactor coolant in the loops 'When the RCS is partially drained.

No one flCM requirement applies to all the possible RHR configurations. In view of the multiplicity of RHR flow requirements, lililitations on RHR flCM are more appropriate for operatllq procedures where prescriptive configurations and the correspon:lin;J flCM requirements can be described in detail.

Such detail is inapprot>riate for technical specifications. It is recammerrled that RHR flow requirements presently included in the technicai specifications be deleted and that operatirg procedures be revised as necessaey to include such requirements.

Per the above recammemation Salem Unit 1 and 2 Technical Specifications were reviewed and marked up as follows:

Specification 3/4.9.8 (Unit 1) and 3/4.9.8.1 (Unit 2) - :References to a flow rate of 3000 gpm in sw:veillance requirements 4.9.8 and 4.9.8.1 were deleted. Specification 3/4.1.1.3 for Salem Unit 1 was deleted.

It is proposed that specific RHR flow requirements be located in plant procedures.instead of technical specifications. '!he flow rate of 3000 gpm represents the design flow rate and is conseJ:Vative for decay heat 62

remcva1. It does not address all RHR flow considerations am therefore may not be ai;::plicable in all ci.rcumstarx::es. Relocation of flCM requirements to procedures 'Wa.lld allCM all RHR flCM requirements to be addressed in an appropriate manner.

For RHR System operation the required flCMiate of Specification 3/4.1.1.3 for Salem Unit 1 'W'OUld be located in plant procedures.

For Modes 1,2 an:i 3 flCM greater than 3000 gpn required by Specification 3/4.1.1.3 is assured by RCS specifications requirin;J cperatian of at least one RCS loop incl~ the reactor coolant pmp.

Salem specificatin 3/4.1.1.3 is not*

in stan:lard tedmical specifications (NUmx;-0452).

Bases 3/ 4. 4.1 an:i 3/ 4. 9. 8 - '!he followi.rq was added:

Adjusbnents to RHR flCM may be required durin;J operation of the RHR system.

For example, RHR flCM may need to be adjusted to control RCS temperature, to prevent RHR p..mp overheatin;J an:i to preserve RHR suction requirements for the existin;J RCS an:i RHR fluid con::li.tions

• Bases 3/4.1.1.3 for Salem Unit 1 was deleted.

RHR flow must be adjusted for proper oontrol of RCS temperature an:i other factors, possibly over a significant rarqe, as con::li.tions in the RCS an:i RHR systems chan:]e. Various minimum amjor maximum flCM restraints may need to be iirposed depen;:lin:J an the existing comitions.

FlCM rates which are sufficient for one con::li.tion may not be sufficient for another con1ition. '!he proposed bases stateirent has been added in recognition of this.

Possible altematives to, the recamren:led solution could be:

a. Inclusion of RHR flow requirements in Modes 5 an:i 6 for operation of RHR with the RCS drained to mid-loop. 'Ihis may :require specifyin;J both a mininn.Im an:i a naxinnJm flow if it is necessai:y to address decay 63

'f"-..'

.~*

heat :reiroval, boron dilution an::l vortexin;J in the tedmi.cal specifications. It is also possible that a requ.irem3nt on reactor coolant level in the loop would be required for completeness.

b.. Revision o; the RHR flow requ.irem3nt in M:xie 6 to a value consistent with mid-loop operation but which does not prevent operation with flows greater than that for which *a vortexin;J p?:Oblem could OCClJr.

'lhis would be consistent wi~ the existin;J specifications 'Which address analytical concerns but not vortexirq concems.

c. Reduction of the flc:M requirement in Specification 3/4.1.1.3 for Salem Unit 1 to the minilmim required for boron dilution.

6. 5 FSAR 0-Jarges

'Ihe Plant & systems Evaluation Licensin;J * (l?SEL) group of West.ii:~

Nuclear safety Depart:rrent has corrlucted a :review of the irapact of throttlirg the RHR flow durirg mid-loop operation ani has provided the following discusnion9 ~possible Salem Plant FSAR c.harges.

'Ihe mar.keel up FSAR chan]es a:re provide:\\ as an attad'lment to Appem.ix D of this report. 'lhe markups consis'c of providirg an insert for section

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

'llle insert describes the throttl.:in;J of the the RHRS flow durirg mid-loop operation am includes a discus.csior1 on factors which minimize the effects of air entrairil.uent on pump perfonnance.

A section is included which a; scusses procedure cormnitment an::l development w.ich *the plant may wish to inco:rporate

• 64