ML19241B320

Forwards Advisory Rept,Reactor Sys Cooldown,TMI-2

ML19241B320
Person / Time
Site:	Crane
Issue date:	04/10/1979
From:	Desiree Davis Office of Nuclear Reactor Regulation
To:	Stello V Office of Nuclear Reactor Regulation
References
NUDOCS 7907160078
Download: ML19241B320 (29)

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/,..,n'b. "~" '" UfJIT E D ST ATES -[.. TJUCLEAR REGULATORY COMMisslON f.,m.-g..;. m, f VJASHitJGTOfJ. D. C. 20555 g).) = April 10,1979 ,. g . ~... MEMORANDUM FOR: Victor Stello, Jr., Director, TMI Operations ~""4 FROM: Don Davis. Duty Officer Shift B

SUBJECT:

RESPONSE TO REQUEST FOR HEADQUARTERS ADVICE ON C00LDOWN Attached is the Headquarters Advisory Report on Cooldown from 280 F to 1500F at 1000 psia. Ot.r review of secondary systems has not =I~ been completed and has been hampered to some degree by a lack of g. understanding of the planned operational modes and any plant $ -IEl modifications that will be made, s.x e:. Don Davis, Duty Officer Shift B cc: R. Mettson L p 555nej e s p, D \\x_/ 7 9 0716 007s F

~ ADVISORY REPORT REACTOR SYSTEM C00LDOWN . ::=.. TMI-2 .;.T = .Ud.'.... J.:.g.e [ The purpose of this report is to summarize the recomendations of 3 the HQ Technical Support Teams with respect to the cooldown of the

== reactor system. In particular, that phase ;f cooling down the reactor 0 system from a temperature condition of about 280 F, to a temoerature of 1500F at a coolant pressure of about 10CJ psi (see attached figure). . ~.. ~"" The following are those subjects or areas that were considered: 1.0 Systems 2.0 Radiolysis 3.0 Potential Criticality 4.0 Engineering (mechanical capability of components) 5.0 Containment I.

6. 0 Radiological Consequences The following discusses each subject in detail:

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BAS'E CASE SUFJMRY Q 9/P Revision 1 h.. e &= h -7 (I ;q &c &A . Q[g 1 cce y 8 h-2 (INM s fue n A v p f..._. e (,g.g ga54 C= co aio =. sccoaou y j,=..__ d er 6-zdq.1) O A, k-g HE phgg Paunnr.Y

12. C. TEMP ;, *=

=~ g g, g 2W ..~.;-. _ 14o-/w v2O 7.ower Pressure (Aa A') while degassing, . ~ ~ " (1) Dcgas at A; then return to A. ~[ Continue Design / Installation of stat.ic and active fo primary makeup /pressuro control and (2) systems secondary cooling syst.em for "B" S/G. "rn.. (AAB) by steaming on "A" S/G (3) Redirec temperature to minimum (B"Cj. (4) Take "A" S/G colid - drop primary temp. - Entablish natural convect ion - Trip FC Pump "A" (5) Establish cooling to "B" S/G if available.. Drop primary prensure to selected value ( C - D,) (6) Tako primary. system solid - Control pressure (7) makeup with static or new active system

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[... . ~ ~. ~ END POINT g. Primary Natural Circ, solid liquid, Long-term P/V Contralf Secondary - Solid water, Long-term Heat Dump Syst,cm _ 5} Approved for Issue-P00RBREE L ?! sse!m _Ar u'b} W R. FA:c1b ~~ "M& W-w, m y

1.0 SYSTEitS FAILURE !iODE. AllD EFFE IS a Going From Point A to Point B in Base' Case Summary (0C/06/79) - holding primary pressure at 1000 psig and reducing prirrary temperature to ap-proximately 150*F. .. :.=-- ..= I~nstrumentation Required: ~ f 1. Pressurizer level 2. Primary pressure ~ ~~ - 3. Incore outlet T/C ~~~ E E--- 4. Loose parts monitoring system h.... {. 5. Relief block valve position je 6. 'Pressdrizer spray valve position p.s ,-= i 7. Steam generator level p:----- -- Areas to be Considered: _ l.1 Loss of Secondary Cooling - Increases primary pressure (a help for NPSH for RCP and absorption of gases). Some corrective acti.ons are P for the operator to restore secondary cooling by turning on either f another condensate pump, an auxiliary feedwater pump or use steam h I. generator atmosphere dump valve as required. Relief valves in the secondary system may also be employed to dump steam to the atmosphere.- If the steam becomes contaminated by primary system water due to a leaking steam generator tube, dumping to the atmosphere 'must be 32-terminated and an alternate method of cooling such as HPI or the Decay _ Heat Removal System must be used. f1onitor pressurizer level and pressure and stabilize conditions in crimary system.

4, .+- l-E.< [.g-t= Increasing Pressurizer Level - due to loss of letdown, increased

l. 2.

flow or loss of secondary cooling. . - - -.. - ---... g --.-.. =.._. Any of the above actions could lead to having the primary system = = -- JE water-solid with the potential for system overpressure if fairly

== For loss of letdown flow, rapid corrective means are not employed. modified Emergency Procedure 5 mitigation steps (not rest of pro-If increased makeup flow is detected, cedur2) should be employed. it may be reduced t4y throttling valves MU-V17 or MU-Vl8 or by f temporarily turning off the makeup pump (s); if caused by inad iss ..=. start of another makeup pump, shutdown the second pump. Loss of secondary cooling is addressed in 2.1.1. 7: l Isolation of the seal return line on the shutdown RC pumps may also ~ i=.-.. . be used to help mitigate the potential overpressurization,of the [M = If the pressurizer level continues to increase and water-solid con-

.=-

ditions appear imminent in spite of all actions take'.i above, the pressurizer vent valve (RC-V137) should be used to prevent over-pressurization and if this is insufficient, the block valve (RC b5f should be opened (use of RC-V2 should be kept to a minimum since ~[ continued or intermittent' flow across the valve seat may effectively =. - destroy the seat after a time since this valve may not have been 555005

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4====:- !an or multiple closures with large dif-designed for flow modu? p.: t l ) ,ferential pressures a ross the va ve. .. =. h per-the above procedure has not recommended tripping t e o Should the RCP be allowed to contir.ue to operate whe Note: ating RCP. hould in a water solid condition (as we recommend) the operators s ~ ) to mitigate be prepared to open the pressurizer vent valve (RC-V137 f-any slow pressure transient due to tripping the RCP. is Should the RCP be' turned off and the MU pump starte Note: 7 be imperative that the makeup pump flow control valve MU-V e.-_. l shocking the h. closed initially to prevent overpressurizing or therma MU-Vl7 could then be carefully opened to control S primary system. If the makeup pump bypass line is n.ot available, it system pressure. h in a water-may not be desirable to operate the makeup pumps w en D solid condition. l" d Effect o'f opening safety valve - gives an uncontrolled blow o t. 3. which may reform vessel bubble. .. = - as If the valve recloses, the system must be checked for any l for pre-g..... bubbles; these must be removed for they i sve thu potentia 5.. Degassing must continue, in any case, r venting the cooldown process. RCS as a result [ to remove any small bubbles occurring throughout the If. the valve sticks open, maintain of inadvertent depressurization. 555p06 .i .w.-..

.s water levels, with the makeup system. This method of cooling can ] only be maintained as long as the wat'er inventory within containment [j.

==~ y is at a low enough level so as not to affect containment integrity [

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or continued operation py flooding systems and instrumentation and ' only as.long as there is a borated water supply. Eventually, con-Z r.EE : tinued cooling of the core tauld require the use of the decay heat ....lf~ removal system. ... = - - PORV Block Valve Open - Similar effects, but not as severe as 3. 1.4 =m-1" Vent Line Isolation Valve Open - Similar effects, but not as . = = ~1. 5 .:~:: ~ severe as 4. 1.6 Loss of RCP w/ Restart - tio problems anticipated. Loss of RCP w/flo Start o,f RCP - assume there is no natural circula-l.7 tion and the RC pump trips. Assume cannot restart any..RCPs. Go to EP-4 which involves makeup' injection. An alternate method,is use of the decay heat removal system. EP-4 controls pressure from the MU pumps via PORV block valve and 1" vent valve isolation valve. We recommend using valve MU-V17 to ~~ = control pressure as much as is practicable Use of PORV block valve is least acceptable since if the block valve stays open it will cause an uncontrolled blowdown. Should the PORV block valve fail closed, the system may overpressurize and cause the safety valves to open, resulting in an uncontrolled blowdown - addressed above in Section 3. ~ e .r yQ9 *) 0

/ controlled manner thrcugh Actuation of an HPI train should be in a the manipulation of any HPI pump bypass lines and/or return test g {~m:3 _3 lines in order to minimize thermal shock on the primary system. ?_.AN ] ~~ 1,8 ' Inadvertent Start of Reactor Coolant pump - additional diff i.T1. ift pressures across the core from a second running pump might l ~ ~ ~ ~ ~ parts of the core, break free particulate matter and relocate pie The additionai in other positions of the reactor coolant system. E =.. Immediate cor-F 5 MWt of pump heat would create a heat imbalance. g __. i F rective action would be to secure the RC pump and stabilize pr mary k Removing power on standby pumps would reduce the system parameters. lity

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potential for inadvertent startup; however, the imm.ediate av . 7=='..; ="E of the pumps is viewed as the more important consideration.

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h.., Inadvertent Start of Makeup Pump - this can pose a serious proble e is9 (1) the system is already water.-solid, (2) the pump starts up if: 6 = :- without having valve MU-V1'7 control the flow to maintain the lev in the pressurizer, and (3) there is no bypass operating so that The maximum full operational pump pressure is applied to the system. head attained by the pump would be applied to the reactor coolant When the sys, m is not water-system (RCS) under these conditions. solid, the makeup pump adds water to the RCS, compressing and Operator action ime .densing the steam bubble in the pressurizer. required to shutoff the pump after an inadvertent actuation

==- 5.=- insufficient for the operator to prevent RCS overpressurization =; 552 0s 9

Safety valve action ma,, .,'t mitigate the overpressurization transient,-... > articulai y under water-solid conditions; therefore, a bypass flow .1. l line or test line should be utilized foi each standby pump to pre- . vent maximum HPI dischaFge pressure being applied to the primary xx.=x-

. = -

system. 1.10. Inadvertent Opening of Pre:;urizer Spray Line . inadvertent pres-surizer spray would cause a pressure decrease in the RCS. If the decrease is sufficient, RCP cavitation may occur, together with .].7 increased gas release from the coolant throughout the RCS. =..... Corrective action would be to immediately isolate the spray line by means of the spray block valve and then to stabilize primary system conditions. If the line cannot be isolated, venting 5f the pres-surizer should be discontinued and additional pres arizer heaters ~ turned on, if available, while observing pressurizer level and primary system pressure. If this action i. insufficient t maintain the pressurizer pressure, it is recorrmended that the alternates RCP (the one without direct pressurizer spray) be used. If de-pressurization cannot be stopped, the RC pumps will have to be stooped.before they cavitate and operation maintained by cooling ~ in the natural circulation mode, by the method described in EP-4 ~.. or by means of the decay heat removal system. 1.11. toss of Pressurizer Heaters - if all pressurizer heaters are lost, the plant may be unable to maintain system pressure. Under these SSb 03 -3' 0 ~ ~- . 1..

t conditions (unable to control system pressure), the plant should be operated in the natural circulation mode, or, if not available, the mode described in EP-4. The decay heat removal system may be ~ employed if these other inethods prove to be unsatisfactory for plant co01 do'wn. 7-1.12. Baron Dilution - Potential return to criticality would exist by the addition of unborated water to the primary system. Periodic p-- = sampling of primary coolant boron concentration should verify shut-down margins. On-line SRMs/IRMs should be monitored to detect a i..f loss of shutdown margin. All sources of unborated water should be ~2.3.; ; carefully controlled administratively from the control room. ..z= b Immediate corrective action for any loss of shutdown margin would [.r i be to identify and secure the source of dilution and use borated makeup to restore margin. .._l_ _.1 3 floise Diacnost %

• Reduction of primary side temperature by steaming on OTSG "A" should not produce any sudden changes which can be monitored by the incore noise analyses system.

Increasing coolant viscosity from the c'oler temperatures.... would tend to produce larger differential pressures ac.ross restrictions,7 thereby increasing the potential for loosening parts. Virtually any anomalies heard with the noise analysis system would be cause for concern, .g -

=;;-.

The and cooldown should allow time for feedback from the noise analysts. .... = S55010

detection of. anomalies should be evaluated before proceeding in the h.. cooldown. To allow a more quantitative evaluation of the progress of the cooldown ~._7. ['- a baseline " background" noise map must be available upon initiation o Periodic noise survefs thereafter could be compared to g.;;. -- the cooldown. [2 'the baseline to provide a more positive trend indicator. f-:-) TE-E::'.. :'.. giy -. f.

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l ~ Pressurizer Level Conditions 1.14 - ~.... Cooldown of the primary system will cause shrinkage of the ~.~E? moderator volume. Si.nce cooldown is expected to proceed slowly, pressur.izer level changes would also be expected to occur slowly, and within the capability of the normal makeup system.

--55 i

~"~ A~ of all levsl~ indication during the cooldown evolution wouM remove a key piece of information necessary to this evolution. The immediate corrective action should be as indicated in EP-21. "~~~~~ ."~~ The predicted moderator volume shrinkage in going from 280 F to 3 160 F' at a constant pressure of 1000 psia is approximately 540 ft, Since there is approximately 800 ft of water in the pressurizer at the initial condition, loss of level in the pressurizer ] E75-Pressurizer sprays should be monit.ored to predliid5 should not occur. activation which will cause a decrease in pressure.

== L e ~ S55012 .__l. ~ . [U555 ^ ~ s.

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/ I s /* '/ 1.15 INSTRUMENTATION OVERVIEW ~~ 1. Primary System Instrumentation

1. J b Primary coolant loop instrumentation appears to be powered from the 120 volt vital (safety-related) buses.

These buses are normally sup_ hEj~~ plied from a de bus through inverters. The dc bus is normally supplied T ~ ~~ flom the auxiliary ac system by means cf static rectifiers and is backed up by a battery floating on the de bus. The plant computer is also powered from the vital buses. Assuming the loss of offsite ~ ~ ~ ~ ~ power, the aforementioned instrumentation should continue to monitor the plant parameters without any intr, ruption. 2. Pressurizer Valves ee. VALVE POWER SUPPLY CONTROL Sw INDICATION RC-R2 ~ Safety-Rel ated Auto /Open/Close . Power Available Relief Valve Only Hand Indicating Controller RC-V1 Spray Valve Controller

~

Open/Close Open/Close RC-V2 Relief Block Valve RC-V3 Open/Close Open/Close Spray Isolation Valve RC-137 ? Open/Close Open/Close Vent Valve B&W says presently an effort to connect back-up power for these valves in case of loss of offsite power. Not done yet. ~ 55b(313

13 - l s t ,/ / 3. Incore Thermocouple Readout ' With regard. to the incore T/C being read by the plant computer, we have determined that there is no low temperature reading cutout built-in

g; the analoc. anversion program in the computer.

The value of the incore temperatures will therefore be computed and printed for the low" temper- [N(( ature range. The reactor ' coolant loop hot and ccid leg temperature ~ indicators (0*-800*F and 50*-650 F) could be used as substitutes. 4. Reactor Coolant Pumos Motors 'The close and trip circuits for the reactor coolant pump motor breabars include permissive and trip interlocks. The RCP motors will trip on a number of electrical signals and either of two cooling water signals. The cooling water signals include low seal injection flow or low cooling water flow to the seal heat, exchanger. Consideration should be given g to bypassing (jumped out) these trip signals. B&W informed us on April 10,1979 that these two tr.ips have been bypassed since April ~ 8, 1979. The electrical fault protection on the motor should be retained to protect the containment electrical penetration assemblies. Other electrical trips, such as under voltage, not associated wi'.h _ 11.. interrupting fault currents, could by bypassed. With respect to the permissives to start (in case of loss of the running ((~ pump), there are e number of permissives including: 555 14 .i 0

14 - I f ~ 4. Reactor Coolant Pumos Motors (Continued) ..2 1. RCP oil lift pressure li;;..

.=

2. RCP low cooling water seal flow to heat exchanger j~lj ' 3. RCP' seal injection water flow })){-- 4. Upper and lower oil pot level 5. Neutron power level (start-up) 6. Core lift (start-up) ~ ~ ~ 7. Mator heat exchanger cooling water J;.:.;;-- Number 2 an'd 3 could be bypassed as it was done in the trip circuits. Numbers 5 and 6 involve start-up concerns. These could be bypassed with no loss of function.- To prevent inadvertent signals being present,1, 4 and 7 could be Fp bypassed. However, additional precautions should be.taken to. assure !l... that oil and cooling water pumps associated with the RCPs are [ ~~- started. B&W also ir. formed us that there will not be RCP clearance problems due to operati'on at the lower temperatures. ...me.ee. ~~' e -.s.. S5bo15

i

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z-1.15 INSTRUMENTATION OVERVIEW

}))~'.} (Additional secondary side failure mod'es & effects are being - - = = - 1.16 [ considered and will be forvarded when complete). '~ 2.0 RADIOI.YS IS Consideration was given to the potential effects of radiolysis of the reactor. coolant water, i.e., the effects of decomposition of the water to form hydrogen and oxygen. In camma and neutron fields,17 ccH /kg water is needed.to suppress 2 a radiolysis of the primary coolant (Ref: US Patent 2937981, 5/24/60). In operating plants the usual concentration of hydrogen is maintained between 25 cc/kg and 35 cc/kg. In the TMI plant the saturation coricentrations of hdyrogen are [.~ listed below:

f. :--

Operating Point "A" (2800F,1000 psia) 1670 cc/kg Operating Point "B" (220 F,1000 psia) 1540 cc/kg 0 b Operating Point "C" (140 F,1000 psia) 1430 cc/kg Operating Point "0" (140 F,100 psia) 140 cci' q. 0 i=) These values are considerably higher than the concentration of hyo., gen required for the suppression of radiolysis of water at the operating i.'.~ conditions of the plant. It can be concluded therefore, that no radiolysis will take place at the operating conditions defined by remainssaturabd ~ points A, B, C and 0 as long as the primary coolant 566016

with hydrogen. Some radiolysis ay be expected if the concentration of { hydrogen is reduced below the saturation limit.and it reaches the value lower than 17 ccH /kg but we cannot see how this would occur. 2 ...[ ~ Therefore, we conclude that the decomposition of the water into hydrogen and' oxygen during cooldown is not of concern. 3.0 CR_ITICALITY_ The following discusses the need to increase the boron concentration i in the reactor coolant system to prevent possible criticality. E_ =-.- F Assuming the initial concentration of boron in the primary coolant 7.=-- system of 3000 ppm, precipitation of boric acid will not occur until the temperature of the primary coolant is reduced to below 320F. g-Since there is no boil-off of coolant in the primary system the {"~ .... ~ " concentration of boric acid remains unchanged. It would be necessary to increase the concentration of boric acid by a factor lE of 24.5 at operating point "B" (220 f and 1000 psia) and by' a E factor of 6.5 at operating poir3t "C" (140 F and 1000 psia) before i It is concluded that any precipita' tion of bor.ic acid could occur. vecul d. occ ur no precipitation of boron in the primary syste. when the operating conditions of the plant is char.ged fre "A** to "S" ~~ and subsequently to "C". In the. letdown system the primary coolant is cooled from its initial .*=..e operating temperature to about 120 F (FSAR value). Then it is depressurized rrE:: to 14.7 psia. During the depressurization the liquid remains in a sub cooled condition and no boiling takes place. Thet CCG ?

1, concentration of boric acid in the letdown system stays the same as in the primary system and no precipitation is expected. ~ ...~;..- teved to contain 220C ppm The reactor coolant is presently bt T_..Z~ m-.--- The reactor core geometry if fully intact, at 150 F, 2200 ppmb 0 u boron. boron, all rods in and burnable poison intact, would have a K -= P5.~.' Thus if rods are 10hk; the burnable poison 47, ak. Rods are worth k ?. out, k goes to N.95 and if the burnable poison is also out the goes to N.99 (at 2200 ppm). If the clad is removed and borated water (2200 ppm) substitute g= =% (i.e., Zr to oxide and washed away) there would be little, if any, the .['. change in reactivity, since at this boron level and temperature moderator reactivity coefficient is near zero )probably slig t y hl Similarly the readivity state is not sensitive to the positive). temperature in this range (150 F to 280 F). 0 0 Thus, if fuel has not redistributed, a boron level of 2200 ppm keep the system subcritical even if everything but fuel is re (small effect from thimbles and grids). If fuel is redistributed (in addition to the above rem itical system could - under optimum conditions of moderation - go cr moderation is required since solid U02 spheres at 2200 ppm (note: For the worst case require enrichments over 5% to be critical). b t all fuel in a cylinder, or sphere - at optimum moderation - a ou a The calculations 3000 ppm would be required to stay subcritical. (KENO) are by the B&W Naval criticality group, using Monte Carlo 555Ms

- 1.8 - =.. and tested cross sections. The, calculation used pellet nuclear g.,._.: parameters (because these maximize reactivity) and fuel-water ..[ (boron) ratios which have been optimized by sensitivity studies. ~N~ j_:=.. =- The configurations which require boron levels above 2200 appear to. -2m; be not very probable (unless it is probable that the rods are not there), but B&W believes (strongly)that there should be protection from redistribution criticality by going to 3000 ppm baron. The There are no problems with 3000 ppm from a physics viewpoint. problems which. have been expressed appear to'be only the potential for systems blockage from baron at this level - a concern which appears to have no theoretical basis (see Section 1). Detectability If the reactor were to get to a k of about 0.95 subsequent changes of the order of 1%ak should be reasonably detectable (25% increase - in count rate for.95 to.96) an an excore startup detector if conditions are reasonably normal. It may, in these circumstances. be partially masked by (relatively) weak. sources, disturbed geometri .l~.5 nonnucl ear changes (e.g., 'downcomer density changes). 555019 At 2200 ppm the worst configuration is estimated to be several percent It is very difficult (if not, impossible) supercritical (i.e., ksi.03). to estimate the power level such a system would maintain to compensate However, going to a water density of about 0.5 for this reactivity. It is, of course, not generally possible should (at least ) do it. to predict the (hypothetical) rate of reactivity addition to the case

J of redistribution (,'wh.en detecta.ble above 0.95} and thus provide the ::-:: needed boron insertion rate Conc 1usions and Recommeadations _At thir time we do not have sufficient concern either to require z _m_ , _ _._. 5".[$.

==r horation or to prevent it. A primary. coolant sample would likely '~ = provide information that would allow a more definitive recom a+= En (e.g., th,e presence or non-presence of control rod material and a y:.... better value on existing baron concentration), [.._ There i;s noth.ing inherent about th.e cooldown process which would in 5 ~ ~ 3.h"9. = = - If boron block. age is not a problem, itself affect th.e reactivity. _5 ' however, th.e boron leyel sh,ould be taken to 3000 psi to. cover all rgs If decision to borate is made, a 100 ppm (remote} possi.bilities. stepwise addition with monitoring of potential letdown blockage If blockage is suffic'iently strongly suspected ~~~~ sh.ould be consi-dered. to cause problems so th.at this is not carried out it would be high,ly qdvisable to be sure th.e startup range instrumentation i,s likely. to be in good order (would. be wise in any case) and it would lT. also be advisable to be sure that b~ oration at some " reasona is avai1.able. 4.0 ENGINEER ING We have studied the mechanical capability of the reactor system g---i. ..~. . =.. components for ;he conditions th.at would be experienced during The results of this study are discussed as follows: cooldown. 555020 .i

L'.'.

4.1 FRACTURE MECHANICS ~

.. =

Fracture mechanics calculation have been performed for several cases {_ that could be encountered in the planned cooldown of THI-2. In all --.g-- cases, the possible atypical. weld metal in the lower head is limiting. ] Nevertheless, assuming reasonable mixing of the water, our calculations show that there is no need'for concern about brittle fracture of the ,r.z== vessel unless extremely unlikely conditions would occur [-- y. We first performad Appendix G calculations using all of the conserva-tive Appendix G assumptions. These include a 1/4T flaw, the Appendix This G bound K g curve, and a factor of 2 on pressure stresses. i b gave a minimum temperature of 1600F for 1000 PSIG pressure and a cooldown ra te of 500F/hr. t Next, we calculated [t'herMa.1_itie.s.ses.ind stress int' ens'ityhactors [. _ [ ~ for the proposed cooling parameters. This g.ve a slightly highe~r cooldown rate, then slightly hi.gher thermal stresses and stress intensity factors. Again, using the Appendix G factor of 2 on pressure, the K g curve, and the 1/4T flaw,. the minimum temperature to comply with I Appendix G was 1700F. ~ [ If the pressure were reduced to abcut 900 PSIG, Appendix G requirements " - ~. and margins would be met at 150 F. ~ 555021 q. p =.. We also performed calculations asuuming a pressure increase to 2500 9= 5 PSIG. Using KIC instead of Kyg, with a, factor _of 2 on pressure stress, and a 1/4T flaw a temperature of 1850F would be required. With yg no factor of 2 margin on pQssure, a temperature of 140 F is 2-- 0 .g still tolerable. _j-Therefore, we conclude thak, there is a very low probability of vessel [ 5 e failure under conditions postulated to occur d iring the planned cooldown. 4.2 Solid Conditions in Steam Generator Secondary piping I e For water solid conditions on the steam generator secondary -ide, the piping systems affected,out to the first isolation valve, are the main steam line, main feedwater line, and the auxiliary feedwater line. The design of both feedwater lines is predicated upon being filled with water during operation and therefore, normal code allowable stresses will not be exceeded. Whil e the main steam line is not filled with water nomally, the additiorial dead weight contribution = to the piping is accommodated within normal code limits fcr that portion inside of containment. The spring hangers (one on one main steam line and three on the other) will tattom out and act as rigid restraints. For the main steam piping in the auxiliary building, ~ the spring hangers will be pinned so as to carry the additional dead weight load of the water in the piping within normal code limits. ~ll.. The information on the main steam lines is based upon verbal input E~~ from Burns and Roe, the architect engineer for TMI-2. At this .l.1 point in time the architect engineer is re-evaluating the seismic capability of these lines; the results of this reevaluation ars not 55Go22

22 - s yet available to the staff. ' 4.3 Steam Generator Tube Integrity Steam Generator tubes are required to maintain their integrity during .2==- postulas design basis acc.16ents including a LOCA or a steam line brdak in combination with an SSE. The design basis LOCA corresponds to a 925 psia secondary to primary pressure diffarcatial and the ~ design basis MSLB corresponds to a 2200 psia primary to secondary j;; - G: 0 differential pressure at approximately 600 F. The required margins = " " of safety against tube failure during postulated accidents are consistent with the margins of safety determined by the stress limits specified in NM-3225 of Section III of the ASME. Further:no re, a ] ~ "= factor of safety of three against tube rupture is required during normal operating conditions which correspond to a 1250 psia primary ~ "- to secondary pressure differential at approximately 6000F. Babcock and Wilcox has provided results of laboratory tube burst and collapse tests. The burst tests conduc,ted on specimens with defects up to 70% thrcugh Hall resulted in no tube failures at pressures less than 3900 psi and the collapse tests on similarly defected tubes resulted in no tube failures below 3500 psi. Three Mile Island Unit 2 conducted a baseline inspec tion of 100% of the tubes in both steam generators in November 1977 following the hot functional tests. Tubes with imperfections 40% or greater were ~{ plugged which is consistent with the basis delineated in Regulatory 555023 .i

.s Guide 1.121,. to maintain the factors of safety described above and proside an addi?.ional cargin for possible operational degradation. Based on the above de. sign bases and the steam generator inspection and tube plugging which was conducted,the steam generator tubes will '"5.'.5: have conservative margins [f safety against failure under the ~ _;; ~ proposed condition of 1000 psi primary to ' secondary pressure differ-0 ential at temperatures up to 600 F. = = - - . = =.. Condenser Flooding Potential safety concerns associated with flooding of the condenser were Since condenser integrity i' not normally included in our considered. safety reviews, little information is available in HQ to determine the If operation safety margins for static cr dynamic flooding forces. in a partially flooded condition is anticipated additional information = as to the expected operating conditions, condenser design parameters Our contacts with Burns & and test results (e.g., hydro) is needed. Roe have not been successful in obtaining this information and further effort has been stopped pending feedback as to the potential operating modes in a flooded condition. 555024 g m t ....e ,e .eme. .m*w*

i g, 5.0 CONTAINMENT = The containment internal pressure has been slight'y lower than ambient e=- pressure for most of the time since the accident. At the present time, the containment is at approximately a 0.9 psi reverse pressure differential. = Since the. design pres'sdre"is"2.5 'pE, the current pressure is not~ of .i 3M-immediate concern. Current ope' rating proce'dures indicate that the i ater 3 flow to the fan coolers should be terminated if the reverse pressus differential reaches 2.0 psi. This action would effectively tenninate I... p... ~ further cooldown of the containment atmosphere thereby terminating the l=--

== transient. In any case, this would be a rather slow transient allowing sufficient time for proper action. We believe, however, a more severe transi;nt should also be censidered. This transient is the inadvertent operation of the containment sprays. Initiation of the sprays would result in rapid cooling of the containment atmosphere causing a corresponding rapid ~ decrease in containment pressure. The magnitude of the pressure decrease will _3 e. depend upon the inlet spray water temperature (BWST water temperature). ~ i) ~"~ assure that the containment does not exceed the design reverse differential pressure of 2.5 psi, the containment parameters should be maintained above minimun values as shown in the enclosed figure. The figure indicates that for a given inlet spray water temperature, the containment temperature as well as containment pressure should be maintained above minimim values. The pressure could be controlled by the addition of a noncondensible gas such as nitrogen or dry air _ This procedure has apparently been followed previously to decrease T.. I."U the reverse differential pressure to below 1 psi. Control of containment m;; 5E5025 ' * ~ * * " ' ~==** W Y

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temperature could also be achieved by. thrminating the water to the fan coolers.

L.lt l. Since the fan operation would continue, proper mixing of the atmosphere would be maintained while eliminating the heat removal mechanism. Since the con-d sequences of exceeding the reverse design pressure differential is unknown) we believe it prudent to maintain containment conditions as indicated above .. ;zz. Ei..ze.- to allow for inadvert'ent spray opeIation -{- 6.0 RADIOLOGICAL CONSIDERATIONS The potential radiological consequences of loss of let down flow use of the RHR system, and steam generator leakage have been identified for consideration in this section. 6.1 Purification Demineralizer Heatup/ Degradation ~ .;g; Substantial radioactivity may ha,e built up on the purir tion ife- ._-[ mineralizer such that if the flow is stopped, the bed will heat up due to decay heat. Rough calculations indicate that the re' lief yalve will lift and discharge small amounts of water and possibly traces of steam to the Reactor Coolant Holdup Bleed Tanks. (RCHBT) if the system is isolated. As long as some flow is~ maintained, there should i not be any steam. If water and traces of steam are relie.ed to the RCHST, the offsite consequences should be nil because these tanks vent to the waste gas vent' header which can be placed at a negative pressure i by venting back to containment. Procedures should exist for venting -.== the waste gas. vent header back to containment should this become a ~~t_. problem. _.g- __.g.. SEG 27 9 l. ~

Heat in combination with radiation dalnage could result.in degradation 'of the demineralizer resin. Radiation degradation which would lead to physical property changes should not occur within the next few weeks. If there has been more fuel degradation than the 0700 3/30 primary j-

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coolant sample indicated, it is possible that the resins could physically ..'.~5:. break down. This 'could 1"ead to pl'ugging of the demineralizer lower -h ~ retention screens, thus blocking flow. It is our understanding that the [. g[ valve operator for the inlet to the purification demineralizer has failed thus making easy realignment or letdown flow difficult. We recomnend that procedures be considered for flow blockage in the ".. =. purification system. J] The radiation exposure for the demineralizer resins will also decrease g a their ability to ion exhange. It is expected that decreased ion exchange is now taking place and that radioactivity could leach off of the resins in the future. This should not be a significant-concern because downstream componects are: heavily shielded; however, radiation levels could increase. ~ = 6.2 RHR System Contingency Plan }:]; [ If it is necessary to use the RHR system, leakage and resultant iodine releases could be a problem. A method to minin. :e radiciodine re- ~ leasn would be to install a skid mounted charcoal filter system in the RHR room. Such units already exist and could' fairly easily be [M lowered through the RHR pump room equipment hatch. This should be C.[.. considered for installation prior to reactor systems operation which could load to a likelihood of RHR syster. operation. 555 28 0 .i y

The design f1'ow rate of air from the RHR pump.ooms is only 350 SCFM. This is a small flow and a small charcoal filter system could be l t installed in the exhaust ducting if room exists. This would supplement "~"~~" the large Auxiliary Building Filter Units which may become degraded with ~ time. A small fresh charcoal filter would reduce iodice releases by at least a factor of 100 if RHR had to be used. ...::;.x 6.3 Coatincency Plan For A St.am Generator Leak =- Consideration should be given for methods of detecting "A" steam ug_ generator. leakage with a flooded secondary side condition. Procedures should exist for minimizing releases should leakage occur--e.g., ~~ use of condensate. polishers on recirculation to the hotwell and main-taining the condenser at a pressure negative to the condenser ' "] circulating cooling water. = -{.

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