ML18082A920
| ML18082A920 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 08/07/1980 |
| From: | Librizzi F Public Service Enterprise Group |
| To: | Harold Denton, Schwencer A Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18082A922 | List: |
| References | |
| 80-09, 80-9, NUDOCS 8008190519 | |
| Download: ML18082A920 (49) | |
Text
Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 Phone 201/430-7000 Ref. 80-09 August 7, 1980 Mr. Harold R. Denton Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Attention:
Mr. Albert Schwencer, Acting Ch,i,ef Licensing Branch #3 Division of Licensing
Dear Mr. Denton:
REQUEST FOR AMENDMENT FACILITY OPERATING LICENSE DPR-75 UNIT NO. 2 SALEM GENERATING STATION DOCKET NO. 50-311 In accordance with the Atomic Energy ~ct of 1954, as amended and the regulations thereunder, we hereby transmit copies of our request for amendment and our analysis of the changes to Facility Operating L.i;cense DPR-75 for Salem Generating Station, Unit No. 2.
This request consists of proposed changes to the Safety Technical Specifications (Appendix A) involving ch~nges to temporary exempt ceitain Technical Specification re~u,i,rements during the conduct of the Special Low Power Tes:t ;I?rogra;m.
The 'Technical $J?ecifi-cations requiring exemption are listed in 'I'a,ble 3-.1 of the Sa,fety Evaluation which is included as an attachment to th,:i,s request.
Westinghouse has reviewed our program and procedures and we ha,ve incorpora,ted their comments.
This change involves an issue which has acceptab.i,lity for the issue clearly identified by an NRC position, and is deemed not to involve a significant hazards consideration.
Therefore, it is determined to be a Class III a.mendmerit as def.i,ned by lOC;E'R 170.22 and a check for the amount of $4,000.00 iis enclosed.
This submittal includes three (3) originals and forty (401 copies.
CENTENNIA~.F*.,
LIGH.-TJ I
, /\\
LJ S0081 9 0 9
~/'l p
Very truly yours, l~
!'4tfs13 Frank P. Librizzi General Manager...
Electric Production 95-2001 {300M) 1-79
Ref. LCR 80-09 U.S. NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-311 PUBLIC SERVICE ELECTRIC AND GAS COMPANY FACILITY OPERATING LICENSE NO. DPR-75 NO. 2 UNIT SALEM GENERATING STATION Public Service Electric and Gas Company hereby submits proposed changes to Facility Operating License No.
DPR~75 for Salem Gen-erating Station, Unit No. 2.
This change request relates to Safety Technical Specifications (Appendix A) of the Operating License, and pertains to exemption from certain Technical Specification requirements during the conduct of the Special Low Power Test Program.
Respectfully submitted, PUBLIC SERVICE ELECTRIC AND GAS COMPANY VICE PRESIDENT
Ref. LCR 80-09 STATE OF NEW JERSEY)
)
SS:
COUNTY OF ESSEX
}
FREDERICK w. $CHNEJ;DER being d.uly sworn accord,in,g to law deposes and says:
I am a Vice President of Public Service Electric and, Gas Company, and as such, I signed, the request for change to FACILITY OPERATING LICENSE NO. DPR-75.
The matters set forth in said change request are true to the best of my knowledge, information, and belief.
~t.)~~
Subscribed and sworn to before me this Z'r/.._,day of ~
1980.
(Jt_a/YM'UJ Ii. ~J
~tary Public of New Jersey FREDERICK W. SCHNEIDER
- .*My corn.mission expires on (/,d. !, /9f..,3
EG460A Publ.ic Service Electric and Gas Company 80 Park Place Newark, New Jersey 07101 Attached Is our check In full payment of Items described below.
Check No. --------
U NUCLEAft REQULATOftY COMMISSION 0ate
~* 01941 a Date
- AUG Statement of Remittance 1
REQUEST FOR AlEN:>MENT LICENSE DPR-75 092213 Amount of Invoice, IJ 0
- Deductions J) 0 fat Amount J!J 0 8 Detach end retain this statement. Please refer to ebove date end check nwnber In your correspondence to the Vice President and ComptroDer.
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- *. **To Fidelity Unioii Tr~ Colllpariy, Newark, til.J.
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Cents
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Authorized Signature
6110A
,*.-c--*-* -*.......
SAL!H t7HIT 2 SP!CL\\L LOW-POWD. T!STS SA!"!TY !VALUATION JULY 1980
1.0 I!ITRODUcnoN Alm
SUMMARY
In mi actempc to *ove the licensing process for Sales, PSE&G propoaed to the *im.c the *ate special teat* to be performed at reactor power level*
'?heae teats vould deaoaatrate the plant'* capability in several simu-lated degraded *odes of operation and vould provide ~pportunities for operator training.
The basic mode of operation to be demonstrated is natural circulation with various portions of the plant equipment not operating, e.,., pre**urizer beaters, loas of offsite power C*imulated),
loH of ousite AC power {ailllula-ted), and JlCPs for plant cooldovn.
Weatinghouae baa reviewed the propoaed teatl and has with the exception of TVA proposed teats 8 and 9 {atartup from stagnant.,c:onditiona and boron mixing and cooldovn) has determined that vi.th cloae operator sur-veillance of parameters and suitable operator action points in the event of significant deviation from test conditions, the teats as outlined in the *Salem Special Test procedure's are acceptable and can be performed with minimal risk. It is recognized that in order to perform these tests some~automatic safety functions, reactor trips and safety injec- *
- tion, will be defeated. Westinghouse has determined a set of operator action points which should replace these autoaatic actuations. It is also recognized that several technical specification requireaents will not be met while either preparing for or performing these tests.
A.gain Westinghouse baa determined that the lov power levels and operator action will suffice during these tillle periods.
Westinghouse has reviewed the effect of the proposed test conditions on the incidents and faults which were di1cu11ed in the Accident Analysis section of the Salem Final Safety Analysis Report.
In moat cases, the 1 FSAR diacuaaioo vaa found to bound the consequences of such events occurring under testing conditions.
Conaequences of an ejected RCCA have not been aialyzed because of the low probabilities.
1or aome inci-denta, becauae of the far-off-normal conditions, the analysis sethods available have not shown that, vith reliance on automatic protection system action alone, the ?SAR analyses are bounding.
In those cases reliance is placed on expeditious operator action.
'nle operator action points as defined will provide protection for such events.
1-l'
2.0 DISCl.IPTIOR 07 T!STS 2.1 NATURAL Cil.CULATIOM T!ST (TEST 1)
Objective.- To demOll*trate the capability to remove decay beat by
.natural circulation.
Method - '?be reactor ia at approxi.. tely 3%' power and all 1eactor Cool-ant Pumps
(~CP'a) are operatin1. All 1CP'* are tripped *imultaneoualy with the eetabliahment of natural circu.lation indicated by the core exit theraocouplea and the vi.de range JI.TD' a.
2.2 MATUliL CIB.CULATION WITH SIMULATED LOSS 01 Ol'l'SIT!
AC POWER (TEST 2)
Objective - To dmaoaetrate that folloving a 1011 of offaite AC power, natural circulation can be eatabliahed and maintained 1lb ile being powered from the emergency diesel generators.
Method - 'l'he reactor ia at approximately 1% power and all l.CP'* are operating *. All RCP' 1 are tripped and a ~tation blackout ii aimulated *.
AC poveT ie returned by the diesel generators and natural circulation is verified.
2.3 MATUJlAI. CIRCULATION WITH LOSS OF Pl!SSURIZ!R HXAT!RS (TEST 3)
Objective - To demonstrate the ability to maintain natural circulation and saturation margin with the loss of pressurizer heaters.
Method - Eatabliah natural circulation aa in Test 1 and turn off the pressurizer heater* at the.ain control board.
Monitor the 1y1tea pres-1ure1 to determine the effect on 1aturation.. rgin and the depre1aur-ization rate.
Demonstrate the effects of charging/letdown flow and steam generator preaaure on the saturation margin
- 2-1 61101'.
2.4 l!'Y!CT OF STEAM G!H!llA'l'OR S!CORDARY' SID! !SOI.ATON 01'*1'ATOl.AL CillC'OLATION (T!ST 4)
Objective - To determine the effect* of *temn generator *econdary *ide iaolatiOD 0111 natural circulation.
Method - latabliah natural circulation condition* H in THC l but at 1%
paver. bolate the feedvater and *tum line far one ace.. generator and eatablish equilibrium. lepeat thia for ooe more *tea1.generator 10 that two are iaolated and eatabliah equilibrium. leturn t~e *eeaa seuerators to 1ervice in reverse order.
2.S RATURAI.. CIRCULATION AT l!DOCED PRESSURE (TEST 5)
Objective - To demonstrate the ability to aaintain natural circulation at reduced pressure and 1aturation margin.
The accuracy of the aatura-tion meter will also be verified.
Method - 'l'he test method ie the aame as for Teet 3, with the exception that the pressure decrease cc:i be accelerated with the use of auxiliary pre1surizer sprays.
The saturation margin will be decreased to approxi-o
.. cely 20 F.*
2.6. COOLDOWN CAPABILITY OP THE CHARGING AND ~
SYST!M (TEST 6)
Objective -
To determine the capability of the charging and letdown ayetem to cooldovn the RCS with the atesm generators iaolated and one RCP operating.
Method - With the reactor 1hutdovn, trip three of the RCP'* and iaolate all four of the *team 1enerator1.
Vary the charging and letdown flow*
abd.anitor tbe primary-1y1tem temperatures to detennine the heat removal capability.
2-2
I 2..7 SIMULATXD LOSS or ALL OlfSIT! AND On'SIT! AC POWn. (TEST 7)
Objective - To desoa*trate that following a lo** of all ou*ite and offsite AC power, includinc the emergency die*el generator*, the decay beat can be r..oved by uain& the auxiliary feedvater *Y*tem in the aanu.al aode.
Method - '!'he reactor i* 1hut down and all 1CP'1 are running. A total atation blackout i1 11.aulated.
Instrument and lighting paver i1 provided by the backup batteries 1ince the dieael1 are ahutdovn.
2.8 ESTABLISHMENT OF NATUR.AL CIR.ctn.ATION FROM STAGNANT CONDITIONS Westinghouae does not believe that it i1 advisable to perform thia test as noted in a letter from T. M. Anderson, Weatinghouae, to R. Denton, NR.C, RS-'l'MA-2242, April 29, 1980.
- 2. 9 FORCED C.IRCULATION COOLOOWN This test is performed as preparation for the Boron Mixing and Cooldovn Test.
Since Westinghouse does not believe it is advisable to perform the Bo?'on Mixing Test as defined uaing core heat,* it iis not necessary to per-form the Forced Circulation Cooldown Test.
2.10 BORON MIXING AND COOLDOWN Westinghouae does not believe that it is advisable to perform this test utilizing core heat ae noted in NS-TMA-2242, T. M. Anderson, Westinghouse, to B. Denton, NRC.
2-3
3.0 IMPACT OR PLANT TEClDIICAI. SP!:CIF!CATIONS In the evaluation of the propo1ed teat* Weatin&bou*e baa determined that approxiaately thirteen technical 1pecification1 will be violated, and thu1 require exception*, during the performance of the teata. Table 3-1 liata the technical 1pecification1 that vi.11 require ezception* and the te1ta fOT llbic::h they vill not be iaet.
'l'he folloviJ:lg notes the rea1on1 tbeae apecification* mu1t be e:zcepted and the basis for continued opera-tion during the teats.
3.1 3.1.1 T.S. 2.1.1 IEACTOll CORE SAFETY LIMITS
'nle core limits reatrict RCS T as a function of ftftVer, RCS pre1sure avg r--.
- (pres1urizer preaaure) and loops operable.
'l'heae li.sita provide protec-ticm by insuring that the plant is not operated at higher temperatures or lover pres1ures than those previously analyzed.
'nle core limits in the Salem tech *pees are for four loop operation.
Obviously 1ilhen in natural circulation vi.th no RCP'a running the1e limits would not be met.
However, it lhould be noted that the tests will be performed with limits on core exit temperature (< 610°F), T
(< 578°F) and avg -
Loop 6T (< 65°F) such that no boiling will be eXl>erienced in the core and the limits of 1pecification 2.1.1 for temperature vill be met.
'nle limits will not be met aimply becau1e leas than four RCP'* would be running.
I 3.1.2 T.S. 2.2.1 IEACTOR TRIP SYSTEM I~STRUMENTATION SETPOINTS
'nle Reactor Trip Sy1tem provides protection from variou1 transients and faulted conditions by tripping the plant llhen variou1 proceas parameters e:zceed their analyzed values.
When in natural circulation two trip
- function* will be. rendered inoperable, Overtemperature 6T and Over-power 6T.
'n'iere ia a temperature input to these functions llhich originates from the XTD bypass loops.
Due to the low flaw conditiona, 5% or less, the teml>erature indications from theae loops will be highly 3-1 6110A
suspect.
To prevent the.inadvertent tripping of the plant when in the natural circulation mode these functions will be bypassed.
Their pro-tection functions will be performed by the operator verifying that Pres-surizer Pressur.e and Level, Steam Generator Level, and subcooling margin (Tsat) are above the operator action points for Reactor Trip and Safety Injection.
Steam Generator Level (Low and Low-Low) is the third trip function that can be affected.
When at low power levels it is not uncommon for this function to be difficult to maintain above the trip*setpoint.
This function assures that there is some volume of water in the steam generators above the tops of the u-tubes to maintain a secondary side heat sink.
The amount of water is based on the decay heat present in the core and to prevent dryout of the steam generators.
With the plant limited to 5%
RTP or less and being at BOL on Cycle l there will be little or no decay heat present.
The heat source will be.the core operating at the limited power level.
Tripping*the reactor on any of the different operable trip functions or the operator action points will assure that this require-ment will b~ met.
Thus, Westinghouse finds that it is a~ceptable to lower the trip setpoints from 17% span and 25% span respective to 5%
span for all of the special tests.
3.1.3 T.S. 3.1.1.4 MODERATOR TEMPERATURE COEFFICIENT The Moderator Temperature Coefficient is limited to 0 pcm/°F or more negative.
When performing tests with the plant critical below 541°F this coefficient may be slightly positive.
However, it is expected that the Isothermal Temperature Coefficient will remain negative or approxi-mately zero.
The tests will be performed such that this is the case and thus minimizing any impact from rapid heatups or cooldowns.
In addi-tion, the effect of a small positive Moderator Temperature Coefficient has been considered in the accident analyses performed for the test conditions.
3-2
3.1.4 T.s. 3.1.1.s KilfIMDM 'l"ZMPXJLAT'CJJl! 10a CRITICALITY the Kini.mum Temperature for Criticality i* liaited to 541°1 by *pee.
3.1.1.S and S31°7 by 1pec. 3.10.3.
To perform te*t 4 it i1 expected fhat the B.CS avera1e temperature will drop below 531°1. Ve1tinchou1e ha* determined that operation with T.,,1 a1 low a1 485°? i1 accept-able a11uming that:
- 1. Control lank D ia iuaerted to no deeper than 100 1teps vithdr.avu, and
- 2.
Power B.ange".lleutron Flwc Low Setpoint and Intermediate llange Neutron Flux reactor trip aetpointa are reduced from 25% B.TP to 7% B.TP.
This will considerably reduce the con1equence1 of possible tran1ient1 by
- 1) reducing individual control rod worth* (lank D) on unplanned with-drawal, 2) reducing bank worth {Bank D) on unplanned withdrawal, 3) maxim~zing reactivity insertion capability consi*tent with operational requi~ements, 4) limiting maximum power to a very low value on an unplanned power excuraion, and 5) allowing the uae of the "at power" reactor trips a1 back-up trip* rather than a* primary trips.
3.1.5 T.S. 3.3.1 1.EAC'I'OR TRIP SYSTEM INSTRUMENTATION The reactor trips noted in Section 3.1.2 will not meet the operability requirements of *pee. 3.3.1. Specification 3.3.1 can be excepted for the reasons noted in Section 3.1.2 of thi1 evaluation.
3.1.6 T.S. 3.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM DISTRUMENTATION jo prevent inadvertent Safety lnjection and to allow perforaance of the 0 1pecial te*ta, all automatic Safety Injection function* will be blocked.
Indication of partial Safety Injection logic trip* and manual initiation will be operable, bovever, the* automatic Safety Injection actuation functions will be made inoperable by forcing the logic to see that the reactor trip breakers are open. Westinghouse believes that this mode of operation is acceptable for the short period of time these tests will be carried out based on the following:
6110A 3-3
- 1. Clo** obeervatiOG of the partial trip indication by the operator,
- 2. ai1id adherace to th* operator action pointe u defined by WHting-houae, *** Section 3.2.
- 3. Little or no decay heat i1 pre1ent in th* 1y1t.., thu. Safety Injec-tiCJD 1ervH. primarily H a preuurizatiCJD function (1butdawn *argin capability ii cou11derably *ore than 1.60% b:l../'r. for control rode at or aboT* th* in1ertioa liait1).
Blockin1 th*** lunction1 will allow the performance of tbe1e te1ta at low paver, pre11ure, or te11perature and cloae operator 1urveillance will H8ure initiation of Safety Injection, if required, within a 1bort time period.
3.1.7 T.S. 3.4.4 PR!SSURIZER The PreHurizer provides the meana of maintaining preHure c011trol for the plant.
Kormally thia ia accompliahed through the uae of prea1urizer heaters and spray.
In 1everal te1t* the preaaurizer heater* will be either turned off or rendered inoperable by 101a of power.
'rhia mode of operation ia acceptable in that prea1ure control will be.. intained through the uae of prea1urizer level and char1in1/letdown flow.
3.1.8 T.S. 3.7.1.2 ADXILI.AllY ~DWAT!I. SYSTEM The auxiliary feedvater aystem will be rendered partially inoperable for two te1t1.
'nie two teats aimulate acme fona of lou of AC power, Le.,
motor driven aUlCiliary feedvater pumps inoperable. Weatinghouae has deterwined that thia ia acceptable for theu two teeta becauee of the 1~ ttle or no decay heat pre1ent allovi.nc 1uf ficient ti..e (.r 3o Ki.nutea) for operating per1oanel to rack in the ~
paver aupplie1 and re1ain 1tea:m 0 1enerator level.
3-4
3.l.9 T.s.-3.8.1.1, 3.a.2.1, 3.8.2.3 POWEi. SOUIC!S
'l'heae 1pecificatioaa are outlide We1tinahouae control,. however it ia acceptable to alter power source availability ** lOQ& ** 9a11ual Safety Injection i* operable and 1afety related equip:aent will function when
~*quired.
3.1.10 T.S. 3.10.3 SP!CLU. T!ST IJ:CEPTIORS - PHYSICS T!STS
'l'hie 1pecificaticm allows the minimum teaperature for criticality to be
- lov a* 531°1.
Since it i1 expected that I.CST will be taken avg as low aa 485°1*. tbi1 1pecificati011 will be excepted.
See Section 3.1.4 for ba1i1. of acceptability.
3.1.11 TECHNICAL SP!CI1ICATIOKS MOT EXC!PT!D While not applicable at power levels below 5% RTP the following tech-nical 1pecification limite can be expected to be exceeded:
- 2.
3.2.2 11!.AT FLUX BOT CBAKH!L FACTOR - FQ(Z)
At lov temperatures and flows FQ(Z) can be expected to be above normal for 5% l.TP with llCPa running. B.ovever at 1uc:b a low power level no significant deviation* in burnup or Xe peak* are ezpected.
- 3. 2. 3
- I.CS FLOWl.AT! AND R -
(F tui)
At low temperature* and flow FAR can be expected to be bi&her than if pumps are running.
However, no 1iguificant conaequence1 for full power operation are expected.
- 3.
3.2.4 QUADRANT POWER TILT RATIO With no, one, tvo, or three pump* running and critical, core power di1tributio111 re1ulting in quadrant paver tilt.., form.
At lov power level* and for lhort period* of times the1e tilts will not significautly influence 0 core burn-up.
3-5 61101.
1
- 4. 3.2.S DD PMAM!!TD.S In the perforsance of,..... ral te8t1 the plUlt will be cleprea1urised below 2230 peia.
At low operatin& power levell tbi9 depreHurisa-tiOG ia Dot aipificant u 10111 H 1ubcoc:tlin1.. r1in b.. intained.
3.1.12 SPZCIAL TZST 1%CXPTIORS
- l. Special T**t lzceptioca Specificatiou J.10.3 allov1 liaited exce}r tiou for the follawin1:
J.l.l.4;Hoderator T.. perature Coefficient 3.1.l.5 Hinimma Temperature for Criticality J.l.3.1 Movable Control A.11emblie1 3.1.3.5 Shutdown ~od In1ertion Liait1 3.1.3.6 Control tod In1ertion Lilli.ti
- 2.
Special Te1t Exception Specification 3.10.4 allOVI limited exception I
for 3.4.1.1 Reactor Coolant Loops - Norm.al Operation.
3-6
3.2 OPllATIORAL SAnT! CJ.I!XllA l>uriq the perfonaance of theu ta*tl the operator muat -et the follow-ina 1et of criteria for operation:
- 1.
- Maintain ?or ~11 Te*t*
a) Priaary Sy*tma Sub-coolin1 (T1at Margin) b) Steaa Generator Water Level c) Pre11uriser Water Level (l) With 1tc_p1 running (2) Hatural: Circulation d) Loop AT e) Tavg f) Core !xit Temperature (bi&he1t) g) Power ~ange Heutron Flux Lov Setpoint and Intermediate l.ange Neutron Flux Reactor Trip Setpoints b) Control !auk D
> 20°1
> 30% Barrow ~ange Span
> 22% Span
~Value when JlCPa tripped
! 65°1
! 578°F
! 6100p
< 7% l.TP 100 1teps withdrawn or higher
- 2. Reactor Trip and Teat Termination muet occur if any of the following condi-tion* are met:
a) Prim.ary System Sub-cooling (T1at Margin) b) Steua Generator Water Level c) MIS Paver Range, 2 channels d) Pressurizer Water Level e) Any Loop t:,,T f) T avg.
- 1) Core !xit Temperature (hignest) i) Uncontrolled rod motion
~ lS°F
< 5% Harrow ~cige Span or lquivaleut Wide i.ange Level
> 10% RTP
< 17% Span or an unexplained decrea1e of aore than 5% not concurrent vi th a T avg change
> 65°1
> 578°1
> 6100F j) Control !ank D leas than 100 atepe vithdrawn 3-7 6110A
(
\\
- 3. Safety Injection aua-t be aanually initiated if any of th* follovin1 condi-tion* are aet:
a) Primary Sy*teta Sub-coolinc CT t Marcin) 1a b) Steam Generator Water Level c) Contaimaent Pre11ure d) Pre11urizer Water Level e) Prea*urizer Prea1ure
~ 10°!'
< 0% Xarrow *a.mae Span or Equivalent Wide Range Level
.! 4.7 p*ia
< 10% Span or an unexplained decreaae of aore than 10% not concurrent with a T "
avg change.
Decrea1es by 200 pai or more in an unplanned or unexplained Safety Injection must not be terminated until the Westinghouse criteria aa def~ned in EOI:E-2, Loss of Secondary Coolant are met.
These operating and function initiating conditions are 1elected to assure that the base conditions for 1afe operation are met, i.e.,
1..
Sufficient margin to 1aturation. temperature at 1y1tem pressure to assure adequate core cooling (no boiling in the bot channel),
- 2. 1ufficient ateaa generator level to assure an adequate 1econdary aide heat 1ink,
- 3.
aufficient level in the pressurizer to assure coverage of the heaters* to maintain pressure control
~. 1uf ficient control rod worth to ensure adequate shutdown margin and minimize impact of uncontrolled bank withdrawal,* and
- 5.
limit maximum po11ible pover level in the event of an uncontrolled power increase.
- t. 1 1 nA 3-8
TABLE 3-l TECHNICAL SPECIP'ICATION IMPACT Test Technical Specification 1
2 3
4 5
6 7
2.1.1 Core Safety Limits x
.. x x
x x
2.2.1 Various Reactor Tripe Overtemperature 6.T x
x x
x x
x Overpower 6.T x
x x
x x
x Steam Generator Level x
x x
x x
x 3.1.l.4 Moderator Temperature Coef-x ficient 3.1.1.5 Minimum Temperature for x
Criticality 3.3.l Various Reactor Trips Over temperature 6.T x
x x
x x
x Overpower 6.T x
x x
x x
x Steam Generator Level x
x x
x x
x 3.3.2 Safety Injection - All x
x x
x x
x automatic functions 3.4.4 Pressurizer x
x x
3.7.l.2 Auxiliary leedvater x
x 3.8.l.l AC Power Sources x
x 3.8.2.l AC Onsite Power Distribu-x x
tion System 3.8.2.3 DC Distribution System x
x 3.10.3 Special Test Exceptions -
x Physics Tests 3-9 6110A
4.0 SAFETY !VALUATION In thil 1ection the 1afety effect* of tho1e 1pecial te*t condition*
Wiich are out*ide the bounds of condition* aa1uaed in the FSAJl are evaluated.
'the interaction of these conditiona with the transient analyses in t:he FSAJl are discus1ed.
4.1 EVALUATIOM or TLUiSIENTS
'the effect of the unuaual operating condition* on the tranaients.
analyzed in the FSAR are evaluated.
4.1.1 CONDITION II - FAULTS OF K>D!BAT! PJl.!QUENCY 4.1.1.1 Uncontrolled lod Cluster Control Asaembly Bank Withdrawal from a Subcritical Condition R.estriction of control rod operation to 11&nual control, and constant operator monitoring of rod position, nuclear power and temperatures greatly reduces the likelihood of an uncontrolled R.CCA withdrawal.
Operation without reactor coolanf pumps, and in acme cases vi th a posi-tive moderator temperature reactivity coefficient, tend to 11&ke the consequences of R.CCA withdrawal vorae coapared to the operating condi-tions assumed in the 'FSAB..
For these reasons the operating ~rocedures require that following any reactor trip at lea*t one reactor coolant pum~ will be reatarted and the reactor boron concentration will be auch that it will not go critical with lesa than 100 steps withdrawal on D Bank.
An analysis of this event is pre1ented in Section 4.2.1.
For Test 7, this transient i1 bounded by the FSAR analysis, since all reactor coolant pumps are operating *
- *4.l.1.2 Uncontrolled lod Control Clu1ter Assembly Bank Withdrawal at Paver
'nle same considerations discussed in Paragraph 4.1.1.l apply here.
In addition, the low* operating power and tbe Power Range Neutron Flux Low and Intermediate Range Neutron Flux trip setpointa act to mitigate this 4-1 6ll0A
incident, while lack of the Overtemperature AT trip remove* *aae of the protection provided in the FSAI. caae.
AD analy*i* i* diacu**ed in Paral't'aph 4.2.2.
4.1.1.3 Kod Control Cluater A.asembly Miaaligmaent the lSAll di*cu*aioa concernin1 *tatic RCCA mi*aligament applie* to the teat condition*.
'11le conaequence* of a dropped RCCA vould be a decreaae in power.
Thu* n:) increaae in probability or, severity. of Chia incident is introduced by the te1t conditions.
4.1.1.4 Uncontrolled Boron Dilution
'11le con1equences of, and operator action time requirements for, an uncontrolled boron dilution under the te1t conditions are bounded by those discussed in the FSAR.
'11le fact that the control rods will never be inaerted to the insertion limits, aa well aa the Power Range Neutron Flux Low Setpoint and the constant operator monitoring of reactor power, temperature and charging 1ystem operation, provides added protection.
4.1.1.5 Partial Lois of Forced Reactor Coolant Flow Because of the low paver limit* the con1equencea of loa1 of reactor coolant pump power are trivial; indeed they are bounded by norm.al opera-ting condition* for theae teats.
4.1.1.6 Startup of an Inactive Reactor Coolant Loop When at least one reactor coolant pump is operating, the power limit for these teat* re1ulta in auch 1mall temperature difference* in the reactor coolant 1yatma that *tartup of another loop cannot introduce a signifi-cant reactivity disturbance.
In natural circulation operation, inadver-tent atartup of a pump would reduce the core water temperature and thu*
provide a change in reactivity and power.
Becauae of the am.all modera-tor reactivity coefficient at beginning of life the power increase in the worst condition would be small and gradual and the flow-to-power
ratio in the core would be increa*ing.
Tbe Power :a.nge Weutron lluz Low Setpoint reactor trip provide* an upper bound an paver. hcau*e of the increa*e in flow-to-power ratio and becauae of the low aetpoint on the reactor trip, DRB ia precluded in thi* tranaient.
4.1.1.7* Lo** of External Load and/or Turbine Trip lecau*e of the law power level, the diaturbance cau*ed by any loaa of load i*... 11.
'l'he PSAR ca*e i* bounding.
4.1.1.8 Lo** of Wormal Peedvater Because of the lov power level, the coneequencea of a 1011 of feedvater are bounded by the 1SAB. case.
In the case of losa of all feedvater 1ources, if the reactor i* not ahutdOV11 aanually, it would be tripped on Lov-Low Steam Generator Water Level.
Ample ti.me is available to re-institute auxiliary feedwater source.a.
4.1.l.9 Lose of Offaite Power to the Stati0t1 1 1 Auxiliaries (Station Blackout)
Because of the lov paver level, the consequences of a lo** of off-site paver are bounded by the 18.AR case.
4.1.1.10 Exce11ive Beat Removal Due to Feedwater Sy1tem Malfunctions The main feedwater control valve1 will not be uaed while the reactor is at power or near criticality on these teats.
Thus, the potential water flow i1 reatricted to the main feedvater
- bypua valve flow or auxiliary feedvater flow, about 15% of nor.al flow.
'11le tran1ient ia further mitigated by the low operating paver level, 1111all *oderator temperature
.*reactivity coefficient, the* lov 1etpoint1 on the Intermediate and Power
~ange Meutrou 11U% Low aetpoint t'ripa, and cloae operator 1urveillance of feed flow, RCS teaperaturea, RCS pre11ure, and nuclear power.
The case of exce11 heat removal due to f eedvater system malfunction* with very low reactor coolant flow ia mnoag the cooldown tranaienta discussed in more detail in Section 4.2.* J.
4-3
4.1.1.11 lxce**i*e Load Increa*e Incident l'he turbine Yill not be in uae durin1 the perfonaance of theae te*t*,
and load control will be li9i.tea to operation of a *in1le *taca dump or 1te.. relief valv*.
l'he... 11 aoderator temperature reactivity coeffi-cient al*o reduce* the con*equence* of thi.. tran*ient.
Clo1e operator
- urveillance of 1team pre*1ure, cold lea temperature, preaaurizer prea-aure, and reactor power, with 1pecific initiation criteria for manual reactor trip, protect against an unwanted reactor power increaae.
In addition, the lov 1etpoint1 for Power-Range and Intermediate-Range Neu-tron Flux reactor trip1 limit any pover tran1ient. In addition, modifi-cation of the High Steamline Flov setpoint allows a reactor trip on Low Steam Pres1ure only.
Analyse* are discu***ed in Section 4.2.3.
4.1.1.12 Accidental Depre1surization of the Reactor Coolant System Close operator 1urveillance of pressurizer pres1ure and of hot leg sub-coolin~, with specific initiation points for manual reactor trip, pro-vides protection against DHB in the event of an accidental depresauriza-tion of the RCS.
In addition, automatic reactor trip cau1ed by the Lov Pressurizer Prea*ure Safety Injection signal would occur when core out-let subcooling reached approximately lO°F as an automatic backup for manual trip. During teat 3 and 5, llhen thi* trip ia bypaHed to allov deliberate operation at low pre*aure, the pre11urizer PORV block valves vill be clo1ed to remove the aajor credible aource of rapid inadvertent depre11urization.
('l.'he Lov Pre11ure trip i1 automatically reinstated when pressure goe1 above 1925 p1ig and the pORV block valve1 vill be reopened at that time.)
- 4. l. l.13 Accidental Depreuurization of, the Kain Steam System
'Jhe FSAR analysi1 for accidental steam 1ymtem depressurization indicates that if the tra1i.ient atart* at hot lhutdovn condition* vith the worst RCCA 1tuck out of the core, the oegative reactivity introduced by Safety Injection prevent* the core from 1oing critical. !ecau1e of the 1mall moderator temperature reactivity coefficient which vill exist during the 4-4 6110A
teat period, the reactor would re.ain aubcritical even if it were cooled to room temperature without Safety Injectiou. 'l'hua the SAR analyais ia bounding.
4.1.1.14 Spurioua Operation of the Safety Injection Syate:m at Power In order to reduce the poaaibility of wineceaaary thermal f atiaue cycling of the reactor coolant ayatem component*, the actuation of high head char1ing in the 1afety injection mode, and of the aafety injection pumps, by any 1ource ezcept unual acti.on will be dilabled. 'lbua, the most likely aource~ of apurioua Safety Injection, i.e., 1purioua~or
"*pike" pressure: or pre1eure-dif;erence aignall from the primary or 1econdary 1yate1iae, have been eliAinated.
4.1.2 CONDITION Ill - IHFUQOEN'I PAULTS 4.1.2.1 Loee of Reactor Coolant fr01D Smail Ruptured Pipes. or from Cracks in Large Pipes aiich Actuates Emergency Core Cooling A review of the plant 101* of coolant.accident behavior during the low power teating sequence indicates that without automatic Safety Injection there ie sufficient cooling water readily available to prevent the fuel rod cladding from overheating on a 1hort term baais. 'l'he aystem inven-tory and normal charging flow provide the 1hort term cooling for the
.. all break transient.
A 1ample calculation for a 2 inch break shovs that the core remains covered for at least 6000 1econds.
".Chis ia 1uffi-cient time for the operator to aanually initiate SI and align the system for long term cooling.
It must be aoted that the.. gnitude of the reaulting clad heatup tran-sieat during a l.OCA event froaa these conditions ia si'°ificantly reduced from the PSil basis scenario. by the lov-decay heat and core atored energy resulting from the low power level and short operating h htory *
- 4-5 6110A
4.1.2.2 Minor Secondary System Pipe Breait.
'I'be c~equence* of minor *econdary sy*tem pipe breakll are within the bound* di*cu**ed in Paragraph 4.2.3.
4.1.2.3 Single Rod Clu*ter Control Aa*embly Withdrawal at Power The PSAi. analy*i* *how* that aa*uming limiting parameter* for normal operation a aaximta of 5 pe~cent of the fuel rods could ezperience a DHBB. of leu than 1.3 following a *ingle ICCA 'withdrawal. u the FSAR point* out, no s~ngle electrical or mechanical failure in the control
- Y*tem could cau1e such an event.
The probability of 1uch an event happening during the teat period ia further reduced by the abort dura-tion of this period, by the restriction to aa.anual control, and by the cloae operator surveillance of reactor paver, rod operation, and hot leg temperature.
4.1.2.4 Other Infrequent Faults The conaequencea of an inadvertent loading of a fuel auembly into an improper position, complete loss of forced reactor coolant flow, and waste gas decay tank rupture, as described in the PSAR, have been reviewed and foUlld to bound the conaequences of 1uch event* occurring during test operation.
4.1.3 CONDITION IV - LIMITING FAULTS 4.1.3.l Major Reactor Coolant Pipe Ruptures (Loss of Coolant Accident)
A review of the plant 1011 of coolant accident behavior during the low
~~er testing sequence indicate* that without automatic *afety injection there ia 1ufficient coolin1 water readily available to prevent the fuel rod cladding from over heating on a aho.rt term basis.
During the large break event the sy*tem inventory and cold leg accumulator* will have removed enough energy to have filled the reactor vessel to the bottom of the nozzles.
Following the sy*tem depressurization there is enough 4-6 J
vater in the reactor vessel below the nozzles to keep the core covered for over one hour uaing c0111ervative a1sumpti011s.
'?hi* is 1ufficient tiae for the operator to manually initiate SI and align the 1y1tem for long term cooling.
At no time during thi1 tran1ient vill the core be uncovered.
It..sit be noted that the magnitude of the resulting clad heatup tran-
- ient during a LOCA event frcm the1e condition* i1 1ignificantly reduced from the FSAR basis *scenario by the low decay heat and core stored energy re*ulting frcm the low power level and abort operating history.
4.1.3.2 Major Secondary Sy1tem Pipe Rupture The small moderator temperature reactivity coefficient, close operator surveillance of pres*urizer pressure, cold leg temperature, and reactor.
power, with specific initiation criteria for reactor trip; low trip setpoints 011 the Intermediate-Range and Pover-i4nge Neutron Flux tripe; Low Fl?" Mismatch setpoint for Reactor Trip and MSIV cloture on High Steamline Flow in coincidence with Low Steam Pressure; and Low Pressur-izer Pre*sure trip (S.I. initiation) assure a Reactor Trip vithout excessive reactor power following a cooldown transient caused by the secondary system.
Following reactor trip, a*suming the worst RCCA stuck out of the core, the reactor would remain subcritical even if it were cooled to room temperature.
Transient analyaes for a steam pipe rupture are provided in Section 4.2.3.
The con1equence* of a aain feedline rupture are bounded in the cooldown direction by the *team pipe rupture discussion.
Because of the low operating power, the heatup aspects of a feedline rupture are bounded by the PSAR discussion.
4.1.3.3 Steam Generator Tube Rupture
>he ateam generator tube rupture event may be categorized by tvo dis-tinct phases.
The initial phase of the event is analogou* to a small LOCA event. Prior to operator-controlled system depre*aurization, the steam generator tube rupture is a *pecial class of small break LOCA 4-7 6110A
tran*ient*, and the operator action* required to deal with thi* *itua-tion durina thi* pba*e are identical to tho*e required for aitigation of a 1aall LOCA.
Bence, evaluation of th* *tema aenerator tube rupture durin& thi* pha*e ia wholly covered by the 1af ety evaluatioa of the I
n~ll LOCA.
After the appropriate operator action* have taken place to deal with the initial LOCA pbaae of the event, the remainder of the 1teaa generator tube rupture accident aitiaation would con*iat of tboae operator actions required to i1olate the faulted 1team generator, cooldovn the RCS, and depreuurize the RCS to equilibrate pri11&ry RCS preuure with the faulted *team generator 1econdary pre**ure.
Theae actions require util-ization of the following *y1tema:
- 1. Auxiliary feedvater control to the faulted a team generator *
- 2.
Steam line isolation of the faulted steam generator.
- 3.
Steam relief capability of at lea*t one non-faulted steam generator.
- 4.
RCS depressurization capability.
Evaluation of the PSE&G 1pecial teat procedures has verified that all of the above systema are U..ediately available for operator control from the control room.
Therefore, it i1 concluded that the ability to miti-gate the steam generator tube rupture event ia not compromised by the modifications required for operation at 5% power during the proposed tests, and that the analyses perfoti11ed for the SAR regarding this event remain bounding.
4.:.3.4 Single Reactor Coolant Pump Locked Rotor Because of the lov paver level, the locking of a single reactor coolant pump rotor is incon1equential.
4-8
4.1.3.5 Fuel Bandling Accidents l'be FSIJl aa.&lyaia of fuel handlin& accident* ia boundin1.
4.1.3.6 aupture of a Control J.od Drive Mechanism Bouain1 (aod Cluster Control Aaaembly Ejection)
'Jhe control rod bank inaertion will be ao limited (i.e., only Bank D inserted, with at leaat 100 step* withdrawn) that the worth of an ejected rod rill be aubatantially leas than the delayed neutron frac-tion. thu*, the pover riae follovi:ng a control rod ejection would be relatively grad~al and terainated by the Power aange and Intermediate Range Neutron Flux reactor trips. lilile the core power tran1ient and power di1tribution following an RCCA ejection at thi1 time would be less 1evere th.m those shown in the FSAlt, the reault of combining these aae-liorating effect* vith tbe effect of the natural circulation flow rate on clad-to-water heat transfer and RCS pressure have not been analyzed.
'nle extremely low probability of an RCCA ejection during ths brief period.in the te1t aequence does not warrant auch an analysis.
4.2 AHALYSIS OF TRANSIENI'S 4.2.l AKAI.YSIS OF RCCA !ANK WITHDRAWAL FROM SUBCRITICAL CONDITION An analysis was performed to bound the teat tranaients. 'Ihe*methoda and a1aumptions used in the FSAll, Section 15.2.l were uaed vith the follow-ing exceptions:
- 1.
- 2.
- 3.
ieactor coolant flow vaa 0.1% of D011linal.
Control rod incremental worth and total vortb were upper bound values for the D bank initially 100 atep* withdrawn.
Moderator temperature reactivity coefficient vas an upper bound (positive) for any core average* temperature at or above 485°F
- 4-9 6110A
- 4. the lower bound for total delayed neutron fraction for the be,innin' of life far Cycle l vaa uaed.
- 5.
~eactor trip vaa initiated at 10% of full power.
- 6.
DHB vaa aaaumed to occur apontaneouely at the hot apot, at the be1ianin' of the tranaient.
'Die reeulting nuclear paver peaked at 65% of full power, a* ia shown in Figure 4.2.l. the peak clad temperature reached was under 1300°1, a*
is ahovn in Fii\\lre 4.2.2.
Ho clad failure ia expected aa a reault of th is tr ans ient.
4.2.2 ANALYSIS OF RCCA BANK WITHDRAWAL A! POWER Analyses of RCCA bank withdrawal transients were performed for natural circulation conditions.
the tranaients were aaswaed to atart from steady-state operating condition.1 at either 1% or 5% of full power, and with either all ateamline isolation valves open or tvo of thoae valves closed.
A range of reactivity inaertion rates up to Che maximum for tvo banks moving was assumed for case1 with all steamlinea open, and up to the maximum for one bank moving for the cases with two stemlllines iso-lated.
Both maxilllWll and minimum bound1 on reactivity feedback coeffi-cients for beginning of life, Cycle l, were investigated.
In ~ll caaes, reactor trip was initiated at 10% nuclear power.
Reactor conditions at the time of m.aximum core heat flwc are 1hown in Figures 4.2.3 and 4.2.4 as function* of the reactivity inaertion rate for three four-loop active ca1e1.
P'or hiifl reactivity inaertion rates, the minimum reactivity coefficient caaes give the greatest heat flux after the trip 1etpoint ia reached, and have the loweat coolant flow rate at the time of peak heat flux.
For these caaes even the slowest in1ertion rate* atudied did not re1ult in any increase in core inlet temperature at the time of peak heat flux.
For maximum feedback cases, however, the transients for very lcv inaertion rates go on for ao long 4-10
Chat the core inlet temperature finally increa*e* before trip, i.e.,
after approximately one and one-half minute* of continuous vithdraval.
Thus, the ca*es 1hovn bound the worst cases.
4.2.3 A!IAI.YSIS OF COOLDOWN Tl.AHSIEHTS Cooldovn transients inciude feedvater 1y1tem malfunctiou1, exce1sive steam load increa*e, accidental depre**urization of the aain 1team sys-tem, and minor and major *econdary 1ystem pipe ruptures.
Attention has been focu1ed on the poasibility and magnitude,of core power transients resulting from 1uch cooldowns before reactor trip would occur.
(Follow-ing reactor trip, no cooldown event would return the reactor to a criti-cal condition.)
During natural circulation operation, approximately 01\\e to tvo minutes would elapse following a secondary aide event before cold water from the steam generator reached the core; thus, considering the close and con-stant =surveillance during these teats, time would be available for the operator ~o respond to 1uch an event.
Analyses were also performed to determine the extent of protection provided by automatic protection systems under trip conditions.
4.2.3.1 Load Increases A load increase or a naall pipe break, equivalent to the opening of a single power-operated steam pressure relief valve, a dump valve, or a safety valve, would cause an increase of less than four percent in reac-tor power, with a corresponding increase in core flow vith natural cir-culation, assuming the bounding negative moderator temperature coeffi-cient for the beginning of life, Cycle 1.
Thus no automatic protection is required, and ample time is available to the operator to trip the 1 reactor, isolate feedvater to the faulted steam generator, and isolate the break to the extent poasible.
Calculated results for the *udden opening of a single steam valve, a11uming the moat negative !OL Cycle one.moderator reactivity coefficient and 5% initial power are 1hown in Figures 4.2.5 and 4.2.6.
4-11 6110A
- 4.2.3.2 Bigh Flux Protection Reactor trip oa high nuclear fluz provide* backup protection for larger pipe bruu or load incru*e*. Analy*es were performed to detenci.ne the war*t core condition* that could prevail at the time of high-flu::ic trip, independent of the cau*e.
?he follovin1 ***uaption.t vere u*ed:
- 1. Upper-bound ne1ative llOderator i*other11&l temperature coefficient, T*. core aTerage temperature, *far beginning of life, Cycle l.
I
- 2.
Lover-bound fuel temperature - power reactivity coefficient.
- 3. Initial operation with core inlet temperature 555oF.
- 4. Initial powers of 0% and 5% of full power vere analyzed
- S.
Hot leg coolant at incipient boiling at the time of reactor trip.
Thia re*ults in some boiling in the reactor.
The negative reactiv-ity introduced by core boiling would effectively limit power; this negative reactivity vaa coa*ervatively neglected.
- 6.
Uniform core inlet temperature and flow.
- 7.
Reactor trip equivalent to 10% of full power at the initial inlet temperature.
The power.. mea*ured by the HIS ia aasumed to be diminished fro. the true power by 1% for each loF decrease in reactor inlet temperature, re*ulting in a true power of greater than 10% at the time of trip.
- 8.
Core flov rate a* a function of core power wa1 aa1umed equal to the predicted flaw under 1teady-state operating conditions
- 4-12 6110A j
ADalyaea of _core condition* baaed on the1e aaauaption* indicate that the DD criterion of the FSAll ia *et.
4.2.3.3 Secondary Preaaure Trip Protection
..i.arse 1teamline rupture*. 11hid1 affect all loop* unifonaly will actuate reactor trip and 1temaline i1olation on LoV Stemaline Praaaure aisuals in any tvo linu, becauae the required coincident 1li&b Ste-line Flow
- iaual ia aet to sero flow.
Low Pre11urizer Preaaure and Pover 1ange Beutroa Flux law 1etpoint trip* 1erve.. further backup*.
An example i*
the double ended rupture of a aain 1tealine_ dovuatreaa of the flow restrictor* with all 1teamline i1olatian valves i_nitially open. ricures 4.2.. 7 and 4.2..8 shov the response to 1uch an event, with an initial power of 5% and ti&tural circulation.
'Dle Lav Steaaline Pre11ure trip occurs almost mediately. In the example shown~ the main 1tecnline iaolation valve on loop one wa1 aHumed to fail to cloae.
Ro power excursion resulted, and_ the react.or remained subcritical after the trip.
4.3
- ADDITIONAL CONSIDERATIONS In the great majQrity of cases it was concluded, either by reanalysis or by comparison with previously analyzed FSA.R conditions, that fuel clad integrity would be maintained without need for operator mitigating action. ?or the LOCA or 1teambreak events, it vas concluded that the operator would have more than ample time (> l hour) to respond by aanual action, e.g., aanually initiate 1afety injection, to preclude fuel damage.
Finally, in certain other cases, primarily associated with certain inadvertent RCCA vithdraval events, the postulated accident conditions were neither amenable to direct analy1i1 nor credit for operator inter-vention.
In particular, the po1tulated accident conditions were outside the bounds of accepted analy1ia techniques so that fuel damage vaa not precluded either by analy1i1 or identified operator action. ror these caaee, the basis for acceptability was primarily auociated vi th the low 4-13 6110A
probability of an inadvertent rod withdrawal event durin& the limited duration of the 1pecial te1t1.
thi* 1ectioa provide* an additicmal a11e1ameat relative to the potential f ~ and conaequencu of fuel failure for these "uaanalyzed" accident conditiona usociated vi th certain rod withdrawal event*. * 'J:hia u*eH-aent ia.partially ba*ed upon an attempt to bound certain effects which aay esi1t for condition* removed from the range of direct model appli-cability. Additional information (attached) i1 pro~ided for four area*:
- l. thermal lll&I'gin aaaociated with normal teat condition*.
- 2. the potential for DD during accident conditiona.
- 3.
'the clad temperature reaponae a11uming that DHB occurs.
- 4. Radiological conaequ~nces associated with presumed groaa fuel
- failure.
the concluaiona of this aaaessment are aa follows:
- 1. DD is not expected for the limiting thermal condition associ-ated with any l.CCA withdrawal event.
- 2. lven a11uming DMB, there should be adequate heat tranafer to prevent clad ovetheating.
- 3.
Fuel clad failure ia not expected.
- 4. !ven aaauming 100% clad failure and other extreme conservatisms, the resulting offsite doae would be llllall.
4.3.1 DESIGN CONSIDERATIONS Margin to hot channel boiling ha1 been incorporated with all normal test conditiona by establishing a lower bound requirement on the degree of 4-14 6110A
reactor coolant 1ubcooli11&.
'lhi1 ta1t requirement a1*ure1 that poatu-latad accident* are initiated fram a condition of esce11 thermal aargin.
4.3.2 DHB CONSI.DEBA'IIONS
, ror c:ert.ain cooldovn tranlient1, th* conclu1ion that DD ii precluded wa1 dr&VD ba1ed oa. uae of the W-3 critical beat flus correlation.
Although the analy1e1 for the cooldova event* dilcu11ed in Mction 4.2.3.2 re1ult in UH velocity below Che rmce of direct applicability of the correlation, the reactor beat flus was 10 low relative to the predicted critical beat fluz that even a factor of 2 would not reault in 1eriou1 concern** for I>NB for th i1 event.
Por the n0n-coold~ transients the limiting condition*, with respect.to DNB, are farther Bay from the W-3 ru1e of applicability becau1e the coolant temperature ii higher and the paver-to-flow ratio i1 larger.
Compa;iaou of the W-3 DNB correlation to low flow DHB te1t data and correlatiou (references 1 and 2) indicate that it will conservatively predict critical heat flux at lov pre11ure c~ 1000 pai) conditions with low coolant flow.
Pool boiling critical heat flux values (refer-ence 3) at these pressures are higher than tho1e predicted by the low flow correlations. Further review of the data in reference 1 indicates that the critical heat flux at -higher preuure i1 lignificantly lover than the above data at 1000 p1i. 'l'he ainimum cri.tical heat flux of the data 1et ia.16 x 106 BTU/hr-ft2 for a data point at 2200 p1ia at a mass velocity of.2 x 106 lbm/hr-ft2.
Since the exit quality for this data point vaa 64%, it i1 unlikely that the reactor would be able to maintain a heat flux of that level due to the nuclear feedback fr'c:a voiding.
l'he power di1tributio11 vould tend to
- peak. tovarda the bottom thua further reducing the local quality at the peak flux locations
- 4~15 6110A
Al10 the pool boilin1 correlation* in refereuce 3 1hov 1ome decrea1e in critical heat flux above 1000 p1ia to tbe aazimua pre11ure of appli-cability of 2000 p1ia.
However extrapolation of the correlation* to a value of sero critical heat flux at the critical pre11ure (3206.2 p1ia) wuld aot re1ul t in lower critical heat fluxe1 than 1bOV11 in the data
- t fTa. reference 1.
Since the core average beat flux at 10% of nomi-nal power (highut expected power for heatup eTentiil ii only on the order of.02 z 106 BTtJ/hr-ft2 a large pe~ng factor would be required to put the reactor heat flux a1 high ** the critical heat flu.z.
For the tran1ien~1 con1idered, the only one1 that lead to 1ignificant off normal peaking factor* are rod motiOD tran1ients.
The rod with-drawal from 1ubcritical i1 a power burat concern.
A.I auch, it i*
expected that even if DRB occurred, the rod 1urface would revet.
Por the rod bank withdrawal, the combination of auimum power and peaking factor would result in a peak power lover than the data referenced above.
Given the lack of data, it i1 difficult to completely preclude DNB, although a prudent judgement indicate* that it i1 indeed remote.
4.3.3 CLAD T!KPERATUIX CONSIDERATIONS Should DNB occur, the peak clad temperature reached would depend prima-rily on the local nuclear tran1ient following DRB and on the behavior of the po1t-DNB heat tran1fer coefficient.
ror a rapid power tran1ient, a* i* illu1trated by the SER analy1is for RCCA bank withdrawal from a 1ubcritical condition, the fuel temperature reactivity feedback and reactor trip on a nuclear flu~ signal would abut dovn the reactor before aufficient energy could be generated to cause a dcaaging ri1e in clad temperature.
In that case, the maximum clad tem-perature calculated va1 under 1300oP even a1suming an eztre.ely lov beat tran1fer coefficient (~ 2 !TU/hr-ft2-oF).
A po11ibly more limiting condition for ICCA vithdraval would be the caae in which a-power increaae cau1e1 DNB but would either not re1ult in reactor trip on high nuclear flux or the trip is delayed.
In the foriner 4-16
caae, a atead' atate condition with bot apot Dim could be poatulated.
In thia atate the clad temperature could be calculated 1iven only the total core power, local beat flux channel factor, heat tranafer coeffi-cient md aaturation temperature.
the core power i* poatulated to be ***entially at the power which would cau.e a reactor trip an high Power lmge-Weutron Pluz law aetpoint.
'?he trip 1etpoint ia at 7% for the1e teats. To allow for caloriaetric error1 and nor.al 1y1te11 error*, trip i1 aaa'9ed to occur at 10% of rated theraal p'ower (l.TP), unle11 a large ducreaae in dOVllCOller coolant temperature occur* durin1 the teat. In teata\\,3 and 5, depre11urization
~
to leu than approziutely 1450 p1ia could require temperature reduc-tion, a1 ia indicated in Figure 4.3.1; however, 1uch low pres1ure. are not ezpected.
Figure 4.3.2 1hovs the relatiou1hip of peak clad temperature, local heat tran1fer coefficient, and the product of heat fluz bot channel factor (FQ) ti.mes core paver (fraction of JlTP).
For the event of an uncon-trolled JlCCA bank or 1ingle l.CCA the upper bound of thi1 heat flux product is approzi.aately 0.34. U1ing this value, the heat tran1fer coefficient required to keep the peak clad temperature below 1800oF, the threshold of 1iguificant heat flux increase1 due to sirconiU11-vater reaction, can be found from Figure 4.3.2.
Various film boiling heat tran1f er correlations have been reviewed to evaluate the-heat tran1fer coefficient for po1t-DRB condition1.
Although no correlations were found which cover the complete range of condition* being tested, 1ome data ezi1t vbich can be extrapolated to obtain repre1entative heat tran1fer coefficient1.
The We1tinghou1e UBI fil11 boiling correlation (reference 4), va1 developed at low flow condi-ticm9 similar to tho1e po1tulated for incidents occurring during the
-PSE&G te1ta.
'rhi1 correlation va1 extrapolated to the hi1her pre1sure conditions of the teat1 to obtain representative film boilin& coeffi-cient1.
Thi* r~1ulted in a heat transfer coefficient in ezce11 of (100 !TU/hr-ft2-or)a,c at 2200 p1ia and 5% flow with quality 4-17 6110A
between 10-50%.
Other film boilins heat tran1fer correlation*, devel-oped at hi&her pre11ure1, were al10 examined.
'l'heae correlation. were H:trapolated dovu 1:0 t1le lower flov conditiona of t1le PSE&G te*t* aa another approach to obtain repre1entative film boilina coefficienta.
U1ina both t1le Matt1on et al (reference 5) and the Tong (reference 6) film boiling correlationa re1ulted in po1t-DMB heat transfer coeffi-cient* in exce11 of 150 !TU/hr-ft2-0f at the condition* siven above.
'l'he1e reaulta indicate that a clad temperature excursion reaultin& in fuel daaage ia not likely to occur even if DHB, ia uawaed.
~
4.3.4 DOSE AHALYSIS COHSID!B.ATIONS
'l'he dose analyse* wer~ performed for a hypothetical accident senario using con1ervative a11umptioa.a 10 a1 to determine an extreme upper bound on poatulated accident consequences.
'l'he analy1i1 a11uaed a reactor accident involvina ao pipe-break with a coincident loss of condenaer vacuum. ~ 'l'his accident scenario is representative of tne Condition II type events analyzed in the FSAi..
'l'he bounding were asaUlllptiona made in the analysis mic:h include:
170 Mvt (5% power) 1.0 do1e-equivalent I-131 acs activity (tech apec limit) 500 gpd steam generator leak in each SG (tech spec limit) 100% clad dam&&e and 1ap activity releaae 10% iodine/noble gas in gap space 100 OF in ateam generators 500 iodine 1pike factor over steady atate 509,000 lb. atmo1pheric 1team dump over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.7 x lo-3 aec/-3 x/Q percentile value 1'he results of the analysia ahow that the tvo hour site boundary do1es would be 5 r~ thyroid, 0.9 rem total body and 0.4 rem to the akin
- 4-18 6110A
'?he analyaia of the accident* baa incorporated aome very c011aervative
&H\\mption.. which 1oea beyond tm ncmnal degree of canaervatin uaed in 7SAJ. anal ya es.
'?he mo*t prominent of these auumptiou* and a brief deacriptioa of the eztreme coaaervatiam include*:
- 1) Equilibrium radionuclide inventoriea eatabliahed at 5% paver.
For iodines, thi* require* ~ 1 aoath of ateady atate operation at 5%
uninterrupted.
- 2) fuel clad gap inventorie* at 10% of core inventory, thia ia a tiae dependeut, temperature dependent phenomon,.
At 5% power, very little diffusion to gap apace ia expected for the short teat period.
- 3) 100% fuel rod clad damage. *
- 4)
Primary to secondary leak.age to tech spec values.
Since Salem is a nev plant, no pr~ary to aecoudary leakage ia expected.
- If leakage were present, it would most likely alovly increue in atepa up to tech spec levels.
- 5)
Percentile meteorology, there is 95% probability of better diffusion characteristics and thus lover offsite doses.
For these reasons, in the unlikely event of a potential accident during the tests, the reaulting dose ia small, even a11uming 100% clad dmnage and other extreme conservatisms.
- 4. 3. 5 OTHER CONCERNS
'?be LOCA analyaes preaented indicate that there are over 6,000 seconds for the operator to take action.
Thia i1 more thmi aufficient time for
- the operator to take corrective action.
Some tranaienta were not analyzed or diaeuaaed in this aupplement due to the combination of the lov probability of the transient occurring and the very abort time 4-19 6110/..
period of the apecial tHta. thia i9 true for the rod ej.ection acci-dent.
?he cOllbinatioa of the low probability of occurriq and the boundin1 doae evaluatiou for a condition II tranaient 1iven here indi-cate that theae event* do not need to be analyzed.
Similar do*e calcu-lation* have been done for the ateamline break accident* Wiich reaulta in *011e1'bat hiFer doses the tha condition II analysia. IheH dose result* indicate that the fact that th* HIS channel* are uot coapletely qualified does not alter the concluaioa that the reault* are bounded
- 4-20 6ll0A
UFER.!NC!S
- 1. J. s. Geller1tedt, I.. A. Lee, w. J. Oberjohn, I.. B. Wil1on, L. J.
Stanek, "Correlation of Critical Beat Fluz in a Bundle Cooled by Pre11urized Water," Sympo1iWR on Tvo-PhA*e Flow in i.od Bundle1, Code 1127, ASKE Winter Annuai Meeting, Movember, 1969.
- 2.
Bao, I* ll.., Zielke, L. A., Parker, M. B., "Low 1lov Critical Beat lluz," ABS 22, 1975.
- 3. Lahey, i.. T., Moody, !'. J., "The Thenaal-Hydraulics of Boiling Water Nuclear B.eactor," American Nuclear Society, 1977.
- 4. WCAP-8582-P, Vol. II, "Blovdovn Experiments With Upper Bead Injec-tion in G2 17x.17 l.od Array," Mcintyre, B. A., Augu1t, 1976.** (Weat-
- inghouse Proprietary)
S.
Mattson, R. J., Condie, K. c., Bengston, s. J. and Obenchain, c. F.,
"Regression Analysi1 of Poat-cBF Flow Boiling Data," paper B3.8, Vol. 4, Proc. of 5th Int. Beat Transfer Conference, Tokyo, September (1974).
- 6.
Tong, L. s., "Beat Transfer in Water-Cooled Nuclear Reactors," Nuc:~
Engng. and Design§._, 301 (1967)
- 4-21
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10.a 20.0 30.0 40.0 Time (sec.)
Figure 4.2.l Uncontrolled Rod Bank Withdrawal from a Subcritical Condition, Neutron Flux vs. Time 4.22
Q
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u 1300.0 1200.0 1100. 0 1000.00
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Figure 4.2.2 Uncontro11ed Rod Bank Withdrawal from a Subcritica1 Condition, Hot Spot Clad Temperature vs Time, Assuming DNB at Time = 0.
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10 I-k IJ 11 REACTIVITY INSERTION RATE (PCM/SEC)
F1gure 4-2.3 Uncontrolled Rod Bank Withdrawal at.Power.
Peak Heat Flux, Core Coolant Flow. and Increase in Core Inlet Temperatu~e I
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Figure 4.2.4 Uncontrolled Rod Bank Withdrawal at Power.
Time of Reactor Trip vs Reactivity Insertion Rate 100
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TRANSIENTS IN THE RE.ACTOR CORE AND COOLANT LOOPS FOLLOWING THE OPENING OF A STE.AM DUMP VAJ..VE FROM 5% POWER, ALL LOOPS ACTIVE
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FlGUlU: 4.2.6 TRANSIENTS IN THE PRESSURIZER AND STEAM GENERATOR FOLLOWING THE OPENING OF A STEAM DUMP VALVE FROX 5% POWER, ALL LOOPS ACTIVE
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0 FIGURE 4.2.7 TRANSIENTS IN THE REACTOR CORE AND COOLANT LOOPS FOtLO'W'ING A DOUBLE ENDED RUPTURE OF A MAIN STE.AM-LIN! DOWNSTREAM OF THE STEAMI.INE ISOLATION l,.T Tt"\\l'\\nt-
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FIGUIU: 4.2.8 TRANSIENTS IN THE PRESSURIZER AND Sn.AM GENERATOR FOLLOWING A DOUBLE ENDED RUPTURE OF A MAIN STtA.).fLINE 4.29 DOWNSTREAM OF THE STEAMLINE ISOLATION
- Pressure, Psig 22 20 16 15 14 460 Test No.
4 Power Level (11)
[
- 1. 2 *
. (1 & 31)
I I
I L>l. 5
- 1. (l + 1 1/21)
I I --
1 J
4u111ted by Hot Leg
/~
Subcoo11ng
- 20°F at 1.51 480 500 520 540 560 580 Active Cold Leg Temperature. °F FIGURE 4.3.l NATURAL CIRCULATION TEST CONDITIONS
- e
120.
100 80 Heat Transfer Coefficient Dtu/Hr 0 f Ft.2
.I 60
~
40 20 0
0 Peak Clad Temperature = 1800°F C:::'I"
./.
- o. 1 0.2 0.3 0.4 0.5 FQ x Power Fraction FIGURE 4.3.2 HEAT TRANSFER COEFFICIENT VS. HEAT FLUX FOR CLAD TFMPFRATllRE OF 1A00°F *
-a 0.6 0.7
- ~J