Text

Forwards Final Rept on 780423 Reactor Trip Incident
	ML19312C982
Person / Time
Site:	Crane
Issue date:	05/04/1978
From:	Seelinger J; METROPOLITAN EDISON CO.
To:	Geoffrey Miller; METROPOLITAN EDISON CO.
References
	TASK-TF, TASK-TMR NUDOCS 8001200075
	Download: ML19312C982 (46)
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Reference 33

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METROPOl.lTAN EDISON COMP A NY a% o, c.

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TMI Nuclear Station sahy REACTOR TRIP /ES INCIDENT OF 4/23/78 Locadon Middletown, Pa.

To MR. G. P. MILLER

Enclosure:

Subject Report 1.

The finalized report is forwarded.

!w 0

J. L. Seelinger Unit Superintendent /

Technical Support JLS:pls cc:

J. J. Barton G.

P,. Hitz Tidb~d'@.j-@;M H. L. Beers R. W. Keaton D. ti. Shovlin R. W. Bensel R. M. Klingaman

8. G. Smith H. L. Benson G. A. Kunder R. J. Toole T. G. Broughton L. C. Lanese G. J. Troffer s

J. A. Snminer L. L. Lawyer A.

Tsaggaris J. J. Chwastyk S.

Levin E. G. Wallace

.R. C. Cutler T. A. Mackey R. W. Zechman W. J. Fels B. A. Mehler W. H. Zewe J. R. Floyd J. P. C'Hanlon J. G. Fritzen W. T. Gunn W. E. Potts J. P. Moore J. G. Herbein L. C. Rogers R. W. Heward M. J..

Ross J. F. Hilbish W. F. Schmauss hafd 3

cms zV ;gg INTER-OFFICE MEMORANDUM D L ---~s m ~~- m w

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' Table of Contents ~

i l

Pace

I Synopsis of Event 2

/

II Sequence of Events.'

4

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III Conclusions ~

10 i

' IV

~Recormendations/ Action. Items 14 LAooendices

'f A..

. Graphs 18 B i-Chemistry Parneters after Event 26 C.

Pressurizer Level Calculations 29 0.

Supporting B & W Analyses 31 E.

Rough-Data 40

-(1) ' Post Metrory. Trip Review 41

.(2) Computer Alarm Printcut 46

. (3) - Reactimeter Delog Data 67 i

(4)- ICS Transient. Response Traces 84 u

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. Synopsis of Event :

1 At 1651:23 on 4/23/78;THI Unit-2 experienced a reactor trip while at t

30% reactor power.due;to a noise. spike on HI8-power range detector. The

. reactor tripped because' RpS channel C was already in the tripped state c_

- as required by Technical Specification 3.3.1.1 due to the inoperability of NI7.

l

. [1dhen (the : reactor tripped the turbine tripped ' causing a very rapid ' pressure i

increase in the B OTSG and a slightly slower pressure increase in the A OTSG.

Four of six main steam relief valves lifted on the B OTSG.and very rapidly T

blew pressure down in the B-steam generator.

One main steam relief valve

-lifted on the-A~0TSG, and it'also caused a rapid pressure blowdown but about w

S 40 seconds delayed from the B OTSG. The B Turbine Bypass valve received a signal:to go full open but-almost imediately received a signal to' go -

full closed due,to the rapid pressure decrease.in the B OTSG. The A Turbine Bypass

- valve received:a signal to open at the proper pressure but the signal to oper the bypass. valve was lower in magnitude than-it should have been.

The operators took the proper iEediate action in nianually cutting back feedwater demand, shutting MUV37.5B (the letdown isolation valve),

starting a second makeup pump,.and opening'the high pressure injection valves on the side of the operating. makeup pumps. The operator failed to initially recognize that the feed pump was in manual and did not run the feed pu:np speed back until approximately 1 minute and 20 seconds had elapsed.

Earlier action here would have mitigated the causualty but due

to low initial feedwater flow' and the effect on feedwater after the

' operator took this action, the mitigating effect should probably not have prevented going dry or the SFAS actuation.

.f The 4 B side main steam safety valves and 1 A side main ; team safety valve failed to properly reseat. The _ safety valves on the B side started to reseat just prior to 2 minutes into the event with the remainder of the B safety valves and A safety valve reseating almost 4 minutes into

- the event.,.The-steam generator pressures when the safety valves reseated

-were between 550 - 600 psig.

The ICS control 'of = the'feedwater valves had not yet been tuned at the

. tice of the event.

Integral vice proportional control was the dominating

, signal.on the feedwater valves and although the valves responded in the

~

prcper direction, they responded much slower than the traditionally expected response.

Thus.with feedwater valves slowly going shut, rapidly decreasing steam generator pressure, Land a constant feed pump speed too much water cwas fed into the -steam generators.

LThe safety valves failing to reseat coupled with overfeeding caused a rapid depressurization and cooldown of the reactor coolant system.. The RCS' sent from 5B3F to 464F in 3 minutes.- The RCS shrinkage from the cooldown caused the pressurizer to drop.below the indicated level range approximately l' minute after the' reactor trip. Due to the-rapid depressurization of the RCS,?SFAS safety. injection occurred lapproximately one minute after the reactor

. trip..

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.. Pressurizer level was restored 2 minuties into the. event as a result of

.p' safety injection, the A Turbine Bypass valve going shut, and some of the B yj main' steam reliefs goingishut. : Feedwater latch occurred 21/2: minutes into the V

g event and terminated feedwater flow. Feedwater latch was the key event in. success-fully teminating the transient..

. 1The event caused viol'atiions and Technical Specification action statement Lentries in that:

-_(l) RCSLcidewn limit 'of4100F in any 1; hour was exceeded (actual 134F)

- T.S.3.~4.9.1.

I

-(2): Pressurizer cooldown liinit of 100F in any 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months was exceeded (actual l ut 136F) -T.S. - 3.4. 9.2.

E (3) Pressurizer was emptied and consequintly was below the Technical

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Specification limits - T.*S. '3.4.4.

-(4) Transient chloride limit of the RCS was exceeded in that safet/

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7 injection pumps sodium hydroxide into the RCS 'and there is ~ inherent chloride

- contamination in the sodium hydroxide.

Highest level seen was 3 ppm -

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T. S. 3. 4. 7.

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- Tne following limits and precautions were violated:

(1) B&W. fuel pin compression cooldown curve.

Tne: net positive suction head for the reactor' coolant pumps was not

violated, s

..7 Control' red drives limitations were not violated.

(Itshouldbenoted safety rods were cocked after this event and based upon possible gas in the CRD's this~ should not have been done).

Calculations-performed imediately after the event and subsequent

~

chemistry analyses showed ;that the core remained covered at. _all times. Although u

'_I the bubble left the pressurizer and went into one or both hot legs, the f

hot leg with the. bubble, if only one had it,.was.at least still_ filled with 4~ ~ '

water half way up the height of the leg.

If;both hot legs had a bubble then m

the leis would have been filled 3/4 of the way up the legs.

1

~

Daring recovery and while beginning to cooldown 45 minutes after the event, a secor.dJSFAS~ safety injection occurred. This occurred due to a mixmatch 7

between' console pressure indication and-what was seen by the E.S. bistables.

The t

' operator w s reading 1720 psig'on wide range pressure instrumentation when the

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second SFASJsafety injection occurred at 1630 psig.- The SFAS signal.was

~ icmediataly bypassed and cooldown proceeded normally-from that point.

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a Initial Conditions

-Tim's Zed 116:51:23

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Reactoi Power 30%

- Reactor. coolant pumps 1A,18.- 2B running f:

~ Tav-- : 533F -

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Pressurizer heaters in' automatic ~

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- Feedwater' Flow: -

6 l

A-

.B. x 10 lbsMr.

i-6 8

1.7 x 10 lbs/hr.

Reactor doolant Pressure,2217 psig MW 235 g

[

ICS Stations:

01.amond in manual r

_.O Bailey Reactor, master in manual

~

.A Feedwater pump running and'in manual

~B Feedwater pump secured

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- A and.B main. and - startup feedwater valves in Auto -

ie;

. A and 61feedwater demand stations in Manual Sters generator / Reactor demand in Auto C

-Turbine in Auto

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. A and B turbine bypass valves ~in: Auto c a

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A.

S/G controlling.on _ low level limits and A feedwater demand manually set at O.

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Caused by noise spike on NIB:(C RPS cabinet was already tripped

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.due to inoperability'of NI7).;

p.:

T3 0:01 Turbiar generator. trip.

T-

0:06 B S/G BTU limit due'tol increasing S/G pressure (graph A pt 1).

. T.

0:10 B S/G increased momentarily to J1040 psig.' (graph B pt 1)

'j' MSRIB, 23, 48=and 5B lifted 1

8 FW flow decreased to O due.to rapidif; increasing B S/G i

1

-pressure (graph A pt 2)' -

x l

B-turbine byl pass -valve opened comentarily and then closed.

. Tc icil2' B ~S/G(BTU limit cleared and operator ran'FW demand station 8

~

)

to.0.

(A.FW demand? station was.already at 0 as the A OTSG j

,was on the low level limits).

Y B FW rapidly increated due.to rapidly decreasing S/G pressure.

I (graph-A pt'3)

r

B FW-valves started-to close*

Currently the FW valves have not been tuned and integral

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vice' proportional control is-dominating. - Consequently the FW valves _ responded very slowly.

-T.

0:13-A S/G pressure rose' to 1045 psig and remained there for 27

'I seconds (graph A pt 4)

'Due to high S/G pre.ssure A FW: flow dropped off scale low.

1 -

(graph.A pt 5)

- T 0:18' pressurizer level at 180" and rapidly decreasing (graph 0 pt 5)

T-

0:23 Shut MJV 376

=

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.0:28 A S/G below low, level limits and SU FW valve starts to open

'(graph A pt 7)

. T

'0:30 B S/G BTU limited which is dominating the FW demand signal..

however both signals are telling. the FW valves to_ go shut and -

the B FW' valves are driving shut (graph A pt 8) s

A turbine by pas,s' valve went shut (graph a pt 9)

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.0:43 Started MUPlA' 10pened MUV16A/B

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-T-fQ:47 Pressurizer. heaters tripped at' 80" in pressurizer (graph 0

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pt 10)-

4

-T? - 0:48' A FW flow back on scale and rapidly increasing due to

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~ decreasing S/G pressure and SU FW valve opening (graph A pt 11)

't-.

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. A S/G 1evel increasing due to A turbine by pass valve shutting.

-and decreasing steam pressure allowing flow to get into S/G r

-(graph A pt 12)

T' 1:08 pressurizer level off scale (graph C pt 13)

T-

- 1:11 Safety injection A and B side due,to low-pressure.

1630 psig E

(graph D pt 14)

Operator verified.MUP1A' and IC on, OHV SA and B opening, OhY 8A and 'B opening.

T' 1:12 A S/U FW valve going shut due to level being above the low level limits and the FW demand signal being at 0.

(graphA pt 15)

A FW flaw remaining constant even though SU FW valve going shut because of decreasing S/G pressure (graph A pt 16)

T' 1:17 Safety injection bypassed.

-r (g'raph.0'pt 17)

T.

1:20 Operatar ran FW pump speed to minimum.

Fw flow to B S/G started to decrease faster.

(graph-A pt 18)

T 1:40 Closed OHV8A and B (These valves were not yet full open)

- (graph D.pt 19)

HPI flows were approximately:

A 130gpm.

B.

160gpm C

300gpm t.

-240ggn l

Flows low primarily due to MUV 17 and 18-being open.

-T 1:48 8 S/G. level increasing.

. Sane of-:th.e safety valves on B S/G shut.

(graph A pt 20).

.B.5/G = pressure at 560 psig.

' [.

Based on-this pressure FW latch should have occurred sooner.

on.B S/G (graph A pt 21). -

t i

pressurizer level restored to the indicating range (graph C pt. 22) j

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SFAShighpressureinfaccion'due'to(low!.C.-pressure.

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. bypass valves. He should have had 'an alarm at-1820psig to-J-

tell him safety injection was?not bypassed. Per procedure

j:

. hef is to ' bypass' safety: injection -between 1820psig 'and ~1700psig, e

"At the time of the safety injection signal the operators indicated f

11=

pressure on the console was 1720psig.- Thus there was a discrepancy.

between console indication and _what the safety injection bistables

+

i.

~ were seeing. -

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'44:15' SFAS bypassed.

OHV8A ~and B closed. -

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Ti 2:15 A S/G pressure 580 psig (graph A-pt 23)

F FW 1atch on A S/G relative to the time this pressure was reached j

. appeared to be satisfactory.

3F

. T.:

'2:24 B : side FW 1atch (graph A pt 24)

~

.B S/G 1evel decreasing..

Some-B-5/G relief valves still open (graph A pt 25).

A FW 1atch (graph A pt 26) ~

' A S/G-level decreasing due to MSRIA being ~open (graph A pt 27).

- t-T,

~ 2:30 Operator took manual control of A and B FW valves and attempted to i

. open 'for 'short ' period. Then operator attempted to close for.short period.,(This did not-have any ef.fect due to FW 1atch haying taken place i' ?

RCS teenperature at 475F.

Rate of-decrease slowed considerably -(graph E pt 28)

Throttled MUV 16C to 260gpm

.MUY l60 to 250gpm

. (graph C pt 29)

T 3:00 RCS pressure reached its lowest value of 750 psig before turning and increasing (graph.D pt 30)-

. Stopped MUPlc

. Oper.ed MJY36 and 37 e

'.T

' 3:48-B S/G level stepped decreasing dte-to remaining B S/G relief ~ valves shut (graph A pt 31).

MSR1A shu't (graph A pt 32)'

T-4:00 At approximately this time MSV 4B and 7B went shut from the previous B side FW 1atch signal.

.T.-

'4:11 MSV 4A and 7A closed f' em the previous A side FW latch signal.

x-r T'

.4:14~ 0perator started EFP2A A S/S level started to rise:

-B S/G 1evel stayed constant (Operator has noL indication of EFVilA/B valves in the control ' room and he did not open EFV118 far enough to get water into the. 8 S/G (graph A pt. :33).

lT 8:43 EFP2A secured (graph pt. 34).

ST-22:16' Secufed RCP 28.

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. Operator. response.to reactoritrip was prompt and correct'with one exception--the.FW pump.was not run back'until 1:20 into the event.

2i fWhile.the.~ operators. responded correctly.to the reactor trip, they

~

~did not' realize the casualty ~ they were really. dealing with was a -

~

-major-steam leak (through'the relief valves).

53.- 'The operators 'during the transient never fully grasped. the damaging L.~

?effect of feedwater on his: situation.. This is evidenced by the d

action taken with the energency feed pump 4 minutes into the event.

4.

. Operator action with feedwater shows-that the operator watched the needles on the. Bailey stations instead wof the actual measured parameters..In retrospect', Jfurther instrumentation of such

critical things as FW block valve position is badly needed. Also 7.1 after FW latch,-signal indication and measured parameters Will

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l disagree..

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Operators, having been trained on the ICS on the B&W simulator, ith the untuned.ICS and slowly operating feedwater valves. The w

slow response of the feed valves while.in auto significantly hindered the ability of-the unit to cope with the casualty.

6.

When the second E.S. occurred the operator was following his cooldown procedure and E.S. should not have had to have yet been y

bypassed.

7.

On the E.S. when the operator saw inadequate flow through the

,4 MUV16's, Lhe did not realize flow was also going in through MUV17 c

and M V18.

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The operators' action in response to the E.S. in both cases was very good.

The E.S. was quickly bypassed to regain control and and the DHV8's were promptly shut to keep from contaminating the' R.C.~systen.

Ecuiorent performance 1.

' Excessive blowdown by the S/G relief valves is totally unsatisfactory.

12.~-

S/G~ safety valve pipe liners ~ must be repaired as some did not remain intact.

3.

_ The A-turbire bypass valve did not appear to open.far enough.

While-the B turbine bypass valve appeared to perform properly, alll bypass valves. opening and closing characteristics should be.

examined.'-

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TThe main a'nd startup fe::dwater valvesin'eed to be turned as-1 Fk

. soon as possible to;significantly ' decrease their. response times.-

j 5.

jThe actuation pressure of..feedwater latch needs'.t'o zbe checked.

4 l

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6.

}Qpeningithe:DHV8'ionasafety1injectionsignalmustbechanged Jto; avoid inadvertent Na and C1 contamination of the 'RCS.

f 97.fThe response of theQEPMB must be checked in; that-the operator 1

triedito feed the'~ B OTSG by " bumping" this valve, but.no water 4'

- went in.s The valve lineup alsoineeds to be confirmed.

i.

8.

The' console wide range; pressure indicator's.must be checked 4

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against what the safety injection bistables.ar_eiseeing.

1

. L1~ Why the. sequence of events did not-print out' on the computer.

9 f.

must be resolved.

cl0.. The time lag'in selected parameters such as flux in the ccmputer

. memory trip review must be resolved.

p

-;11. Nuclear. instrumentation spikes must be found and corrected. NI 7

- must be restored to operability.

' Procedures 1.

More specific. guidance in some areas could have keyed the

. operator into such things as running feed water pump speed down--

see recocmendations.

Philosochy

-r

1 '.

We must understand each specific evolution and.the most likely consequences of equipment failure er malfunction.

We must' approach evolutions with the question, "how would I respond if-happened."

~ 2.'

We must stop and regroup when something is marginally working or

,1 marginally understood. Adverse affects can be additive.

3.

. On' Wednesday, April 19 the unit tripped from 15% power due to a loss

- of' feedwater accident ~ caused by operator-error when.b1_owing-down a condensate puap suction : strainer. The trip was analyzed and parameters

' thought' at - the' time. to be adequately understood.. Reanalyzing the-trip with hindsight from the April 23 trip shows the'same excessive bicwdown Ecccurred in both steam generators.

Had loss of feed-not been the initiating

- event,'E.S. initiation could 'have occurred on the earlier trip.... Thu s when investigating aJtrip or transient each parameter must not be analyzed

- with " tunnel vision,"L but rather must be looked at independently and robjectively fort proper performance.

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- Training;

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E1.: 'Itfis'not likely that operator training on the simulator would' "E

have; helped the operator to better cope with the! slow. response

of the FW valves.. however it.should have'. helped him recognize

g..

r the ' steam' 1eak. -

2A-talk though with "what ifs" might have 'significantly helped

- 2.

. ith recognition of the. steam leak / overfeed problem.

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- See reconnendations.

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. 0ther.

11 More indicatiori of. critical operator items in the. control-room is' needed: -

FW block valve position. indication

'EFvalve position indication..

'i Turbine bypass valve position indication n

acnospheric relief valve position indication Main steam relief valve posttion indication

- 2.

The r. umber of inoperable alarms in untrol-room needs to be

. reduced to a manageable : level.

3.

' More alar:n acknowledgement-capability in~ control room is needed.

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' EQUIPMENT CHANGES '

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1 isolve-relief ' valve discharge piping expansion joint problems - GPUSCz 3.

- Toole/ Levin ~ prior to startup.

.2..

(Correct nuclev instrwgentation problems NI7finoperability, NIS noise

~

7

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. prior; to startup. - -

spikes, NI5 possible crimped cable, NI4 zeroing problem - 8rumer -

a 3.-

'Get OHV8 open. signal: injection - GPUSC Broughton/Toole '- prior to startup.

4.

Solve relief valve blowdown ' problems - GPUSC Toole - prior to criticality.

^5.

- Actuation pressure of feedwater latch needs to be checked for. proper

. response -: Brrmmer -- ' prior to'startup.

.6.

_ Console-pressure 1 indicators must be checked.against SFAS safety injection j

bistable pressures '- Brumer - prior to startup.

~

7.

The response' of the EFV118 valve must be checked.. The valve -lineup for emergency feed must also-be confirmed -to be proper - Brummer - prior to

^

s tartup.-

8.

Turbine bypass valves proper opening and tuning must be performed - GPUSC Toole/ '- during heatup and power escalation testing.

9.

The main and.startup feedwater regulating valve control needs to be tuned GPUSC - Toole~ ~during power escalation testing.

e 10.

Computer problems with the sequence of events and' memory trip review need.

to be resolved - Fels - prior to startup.

11.

Control rocm position indication. is needed for_ the following valves:

a)

Turbine bypass. valves b

' Main and startup~ feedwater block valves c

Emergency feed control valve d

JAt:nospheric relief. valves GPUSC - Toole. 'detennine GPUSC position on making the plant changes-required to accomplish these changes.

.May 5.

J

12.

Provide some indication for main steam relief valve' position.

A_ temporary microphone from the vicinity-of the valves-to the control room has been suggested. -In.this way the operator could at least detect indep'endent-of analyzi.ng console indication that the relief valves are open. GpVSC -

Toole - prior to startup.

13. Change main feedwater block val _ve signal to open from 90% to 80% open on the startup.feedwater regulating valve.1 This will be consistent with Unit 1 and should reduce the power level at which block valve induced 1 flow oscillations occur.

GPUSC - Toole - prior to startup.

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J jj4 Priorito againfachic.ving -30% power.detennine' optimum mirii[num feedwater?

. s pumpTspeediso unit can go through the period of flow _ oscillation with

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! feedwater; pumpfin auto.:!GPUSC -qToole - durin'g: power.. escalation testing._.

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^ /PROCEhURECHANGES:

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^ ilh !.Re91se-proceduEes to insure:

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More specific guidance islincluded with respect to ICS stations:

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ibeing'in -hand on' a~ reactor trip and other mafor casualties. Flag.

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Lto the operator which stations to..specifically look for to run

~

back' -- 8rumer, May:10.

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b)

'Includellimitiand.precautioniin cooldown.and CR0 operating procedure

'to insure; safety rods are not'" cocked" until a depressurization -

f situation-hasLbeen ana19 zed. Bensel May 10..

Lc)

'. Flag to the operator. on safety injection. to not only monitor high -

.oressure injection flows, but also flow through MUV17 ar.d MUV18.

Fla~g hcw to properly throttle-flow'in this situation - Mackey - May 10.

..~_

d). L Revis.e startup1 procedure to provide the. operator with specific

'~

' guidance as, to how _to drive the plant through the 17". - 22%- feedwater

+'

' block valve oscillation problem.

Use the collective methcdology. of the shift supervisors and change procedure so-that this is done in

'aluniform fashion-from shift to shif t - Floyd - May 10.

~

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Include. in limits 'in precautions - for theICS procedure a sta:ement m

-to' minimize the use of ICScstations in hand.

Brumer - May 10.r

? PHILOS 0pHY-

~

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1. -

We must slow'down and proceed forward deliberately and correctly. Senior j

station management must' convey this philosophy to shif t personnel - G.P.

</J

Miller /R.J. Toole prior to startup.

+

2.

We must; analyze each unknown objectively and thoroughly and respond to the conclusions reached with' good engineering judgement.

G.P. Miller /

f _

rJ.t.. Seelinger

onigoing.

TRAINING L

~ 11.

Reviewithis transient and all other trips and transients to date with

~

all ~ operating shifts; ;Such. reviews should be conducted when feasibie in front of-the1 controls.to;get maximum, benefit from the situation.-

m v

2 Floyd -' Mayf10.

ia;

'. ' Write l supervisor of: operations memo.which discusses; pre-evolution briefings,~.

a2..

thefneed forithese briefings to" review the most probable expected.

abnormalities, theineed for.: these briefings to be conducted, whenever

feasiblei'inifront of the" controls. Floyd - May 1_0..

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3, TTRAININGr(Cont'd)

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43. - ?Ditvelop standa'rd formatifor how a: trip / major transient gets reviewed inxhoth units.1 and '2 with 'each operating shif t.

Floyd/Ross - May 10.

R 14.; iCRO training must. stress and teach' the. operator to watch' and monitor.

yf

, imeasured?indicationsivice ~8ailey ICS, signals.

Zechman - long; term' Q

i

-- 5.

Simulator' training;is.a definite asset in-helping the operator recognize.

4

.* l _

and respond' to';this type of; incident.

f ja)

-Reeval_uate-sending the single unit licensed ' operators to the simulator each year vice every other year. Zechman/Tsaggaris - long -

term.-.

... t..

b)! } Consider purchase of the small Omnid' ta $300K simulator for in-house a

-' operator training'.

Sendisomeone to Omnidata-to evaluate its performance and capability.

Zechman/Tsaggaris - long' term.

OTHER:

ilb Precalculate thc' exact volumes.of-critical piping. runs and tanks from -

v,arious ' taps and capture in a' procedure.: Such a procedure would have helped in' quickly analyzing the' pressurizer situation and in determining the core remained covered.1Mackey - May 12..

.2.

Escalate the alann window correction program-in priority. 'This will #

elimin' ate an excessive number of lighted panel alarms at the baseline

. condition and give the operator a' better chance to focus on what-to

! respond to.

Shovlin -_on going 3.c ; Add additional alann. acknowledgement capability-in the control rocm.

Currently the operator, at the controis has 'to go across-the control to read an alarm and then back to the center console to acknowledge it.

!GPUSC, 'Tcole.- determine. GPU positionL 4

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- Power Generation Grouc P.O. Box 1250. Lynch::urg Va. 24505 -

May 5,19781 Telephone: (804) 384 5111 SOM-II.-lh 3 -

}

~

Mr. : G E P. -Miller :

LStation Superintendent -

Metropolitan Edison'Cenpany

-Post. Office Box h80' DMiddletown, PA. 17057

~

Subject:

Reactor Trip /E.S./Cooldo.n Incident of 23 April 1978

~De~ar Gary:

<Information relative to the ER evaluations of the subject transient is continuing to be developed and accu:ulated at -both the Nuclear Service

_ section' in Lynchburg ar.1 the. TMI Site Office.

The following;infor:E ion

.during the subject plant transient.contains; conclusions regarding the -lev

._ Conclusions Cor.curninr Voiding of Pressurizer

.p-

, There are four, observations or calculations which can be drawn to ascertai

. Whether or not the pressurizer-surge lins e=ptied during the transier.t.

Each,of these suggests that, in fact, no ' steam bubble was drexn into the

.~during the depressurics. tion.LReactor Coolant System proper frc:a th

. Calculations'vereicade cased on the nessured Reactor. Coolant System

, parameters to determine the rate of and enount of Reactor Coolant System coolant shrinkage during the first two minutes of the. transient, which

, includes rost' of. the'pericd when pressurizer levelivas low-off-scale.

~

These; calculations considered the effects'of the Makeup System and the l operator. initiation of: EPI,.estinations or: the effects of temperatu measurc:entL errors.:and flashing r.nd-subsequent cooling of liquid in the re pressurizerias the pressure rapidly. dropped._ Using conservative esti= ate's.

, for.these effects,. the eniculation :showed that the pressuritec steam

~

,did emptyi. but the' bubbic did.not reach the Lhot leg. -b

'the pressurize r-x T

i

. The B5bcock & Wacos Company / Citablehed 10G7.'.

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'1G[PeHiller 5/5/78 The second itcs of note is that the Reactor Coolant System pressure was a cmoothly ' varying parameter throughcut the transient. In-hou'se codes (e.g., TEAP2) have shovn that folleving the pressurizer-surge line emptying, the Reactor Coolant System pressure promptly flattens as hot

' legs. flash to steam, thus mainta.ining the Reactor Coolant System at saturation pressure for the highest temperature in the Reactor Coolant System. This. sort of phenemenon did not occur at E!I-II as can be seen

. from the loop Reactor Coolant Systes pressu e traces.

The third indication that no bubbic was drawn in the Reactor Coolant System is that the botter of the two Reactor Coolant System hot legs (loop "A")

never got within 20 F of saturation temperature as the Reactor Coolant System pressure. fell. At the 3 minute mark af ter reactor trip, the pressure turned sharply upward c.nd this aspect of the problem ceased to be a concern.

The fourth ites is a qualitative agreement base'd on the reactimeter data from the site. For pu poses of explanation, the plot of pressuriner level, Reactor Coolant System pressure and pressurizer temperature are attached. According to the site scurces, h3 seconds after reactor trip, the ITI-isolation valves were opened and the second makeup pu=p was started.

This point is =arked with an "A" on the pressuriser level plot. Design stroke time for the 'iPI valves is 10 seconds; design startup time for the makeup pumps is 6 seconds. Therefcre, it seems reasonable to e.ssume that full IGI flow vas achieved well before the 1 minute mark.

Since the makeup punps are centrifugal pumps, their flow rate increases as Reactor ' ',

Coolant System pressure decreases. !*otice on the pressuriner level trace

' - that when it comes back on scale (t = 2.2 minutes ), its slope is relatively.

linear for about half a minute. At t = 2.5 minutes, the operators throttled down JOI flow.

Notice that, at that time,the slope decreased as one vould expect.

Nov look at the dashed curve which covers that range in time when 1cvel was off scale. - For the level to have behaved in this way, makeup

. flow rate vould have had to decrease sometime before level came back on scale. Since Reactor Coolant Systes pressure was continuously decreasing throughout this time, and the EPI valves were at a fixed position, this conclusion does not seem physically realizable. The only curve shape which neets the physical conditions is a concave curve, or straight line.

This alternative is shown as a dotted line. According to the FPI flov e

versus Reactor Coolant System pressure curves put out by Component Endineer-ing,. the IFI flow rate should be continuously: increasing as the Reactor

. Coolant' System pressure falls, thus a concave curve shape seems most logical.

The conclusion fron this line of reasoning is.that pressurizer level could not have dropped very far below the lower level tap when IFI turned the level decrease around; The curves suggest that the level stayed within

,two feet' of the lover level tap.

9

s -

9)?-}:g.- W M b i-s-E 15/s/78

.l

^

-l Based 1 on' the -arSuments 'and.; observations :. outlined - above, B/,N'si conclusion a

.;is;;that the~ pressurizer vast neveriemptied. 'It : appears that 'only the

^

~

Loperator's ti=ely initiation of=HPI prevented this from occurring, but

.the. data Lseems jo supporticur contention that the pressurizer, was never

entirely drained.

- a.

5 As ' additional Tinfornatien becomes. available,'I vili forvard it to you

~

for; your reports and files.

,:If you haveL any further questions,- please do. not. hesitatc -to contact me.

y 9

--Very truly yours;-

.Lc.

~

.L. C.-Rogers-

. Site. Operations Manaser

.LCR/ bay ~

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G, K,- Vandling '

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,.1: PRESSURIZER I.EVEl. cal.CULATIONS 1

' " APPENDIX C:

w:

ic'~-

ji( s;uiW' (A;roughiessiniate to-determine-the extent of steam bubble formationLin the RCS was-t

~

@y$ < +,

iDATA & ASSUMPTIONS" "made initheifallowing manner:'

i,'.

~

9

_ D 1. + OTime perfod PZRiindicated 0" J= L1.0 ~niinutg.

M, -

A M 2.1 1RCS~Tave'at initiall 0"f PZR level ' 537.5 F p -

f3..

Equivalent volume of RCSi0. 537.5 F =. 7f,000 ' gals -

d:"?

A. E4.:- RCS Tave when PZR: level restorgd. =' 495 F-M5U ^ Equivalent volume ofs RCS 9 495 F :=: 74i400 gals b"

6.x l Approximately 6 Lft. ~of; PZR volum'e below level trans tap 17i c Assumei20' gal /in: PZR level; s..

' 8. -

Total 2HPILflow during 0"' PZR _ level :=: 63S 'GPM -

191;Normalt makeup ficw during 0" PZR level ~= 160 GPM

~.w.

[.

from the dove, = thef amount ofitotal R_CS volume shrinkage'during the period of-x J

'0"Dp2R level indicationiis: equal 1to

~

L74,400 galsj-~ 71,h00 gals =!3400 gals LAmount of makeup /H?I 635 t 160

795 gals

Amount of'PZR water volume remaining at:0"' level indication =

,m l

l20 gals /in X.721n = 1440 gals e

Total 7 volume shrinkage not accounted for by-makeup or remaining PIR level =

~

t t3400 -J795 i1440 V ll65 gals.

!The': approximate volume of the surge line -(vertical runs only) isiequal:to.

2

.$m 17548 gals /ft3 3.M 5 :in J 3

b 1728 --in /f t 7 1n.) 1

=.49.95 gals. = (50) 3 Thus', the volume of the-RCS to fiash to steam is 1165 '- 50 = 1115 gals.

1Since,

~ 1)f :The: pres'surizerc su'rge ~-line: enters the l vertical hot leg at an

. l J elevation '41/2cfeet above the - topLof;the RV outlet nozzle, and-

}2). L The RCSLpressure e'.i' sting'at the--top of the. hot leg mu'st.be 'less.

t

-th~an: the RCS pres' ure ' existing atithe RV, outlet nozzle?by' at least 1

~

x

. the dit ference Tin head!'at these elevations, Tit p"s;as'sumed water! began reaching; saturation.p5 essure first, at the t p= f the.

,e,

'180 candycane".jAssuming: the watertin' the 180 head. has already flashed ~to r"

~

steam, the < level..of water in the1 vertical Lhotcleg would continue"to decrease-' s a q

isteam bubble'was be~tng drawnn a

~

' ;Conrnenc.

lM f the" volu.ingsith a waterflevelj at an elevation-~of 357?ft.u(Top off 0TSG) and as w

y ' o1..

me of RCSJtof flashetoisteamkdidtso:all?in the oneihot tleg. final ~~ water lleveliinj the hot leg.may;beidetermined. -

E

. ~

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. 1

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~

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4?$ llssl ' l e 3

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~

RCS Volune to flash =

~

11{/4 = -(3.14)(3)g ~X L = 7 06L = 149 ft y^

- : Hot leg Volume /ft = lxf 0 3

4-

~

4 L = 21.1 ft

'?.357 ft:(Top'of.0TSG) 4 21.1 ft

~

.335.9 f.t. El ey..

JSince this elevsti'chiis approximately 6 ft. above the elevation of the top of

~

the' R.V. 'and '.since the reactor vessel contains o' er 9,000 gals above the core, v

"it was concluded that-. the core remain covered throughout this transient.

,5 e

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APPEN0!X 0 I

SUPPORTING B&W ANAL.YSES

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x, Babcock &Wilcox

,,,,,, c,n,,c,,, c,eo, P.O. Sox 1260. Lyncheurg. Va. 24505 s

Tefephoner (804) 334 5111

.a

.May 2, 1978 u

SCM-II-lho

- Mr. G. ' P. Miller Station Superintendent" Hetropolitan Edison Company

. Post 10ffice Sox 4801 Mftiletovn, ?A 17057~

Sbbject: Reacter Trip /E.S./Cooldown Incident of 23 April 1978

- Dear Gary '

~

In cecordance with your request fer the evaluation of specific areas of concern fo'.lov'ing the subj ect plant transient, 3W has completed a pre-lbhry review of the effects of the supplied data. The findings con-

- cerning the reactor :colant pumps, the control rod drive mechanis=s, fuel co=ponsats, the ?,cactor Coolant Syste: vater che=istry,' OTSG transient

~

results, res: tor vessel transient results, and reactor coolatt piping, including p.rsscuri:er, are as follows:

- i

- I.

Reactor Coolant Pu=ps Du$ing the first 22 minutes of this transient,

^

three pt::sps vere ope:atiq, two in "3" loop and one in "A" loop. 'The vorst condition, frem an :.?SH

~ standpoint,.tuld have affected the single pump in the "A" loop.- rnis

.F

. pu=p vas floving apprc:cisately 117,000 spa at the time of the incident and vould ha.e required a minimum systes pressure of 'about 560 psig to' prevent cavitation-in the i=peller at a cold legitecperature of

'k6k'?.. As the indication ~is that the syste= pressure only dropped to

. 752 ;~sig, ve do not believe the pu=ps. operated under cavi:ating conditions.

We. note : hat after 22 minutes, single pump operation in each icop vas-

~ initiated, however, by this. tine the Reactor Coolant System pressure 7vas back up to 2140/ psia, which.vould provide adequate UPSI: available--

to the : p.=. :ps.

Frem verbal ' observations,.ve ' understand that injection vater was usin-J-tainedj and there vas no,noted. change in. shaft vibration. We vould-like.

y E

' to. get. additional information, 'such as seal cavity pressuro response L and:sesl' leal:nge during: this transient to further our ~ knowledge ss1to U hov these seals. respond: to such plant transients.

~

l 9

32 -

p3 u

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1 P

ITbcDabroc. Win:ot Co no.wy} EstMdis>d 18fJ 1m 4

{ "

.,._q_

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( v ko p

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e-a G. P. Miller - - 5/2/T8

<g 11,

' L JAs *;his was a very quick te=;erature and pressure ramp,.125'T in 3 minutes,-lh'uS psi in 3 minutes, ve cannot nake any statenent as ;

to the effect of long-ters fatigue life on the pump casing and cover.

These transients vould tare to be evaluated by the Mechanics Research Institute, the consultant that perforned the stress analysis on -these pu=;s. Ve'recently obtained a quotetion for $8,000 for a similar analysis on the SCD pu=ps.

If ve are advised to proceed with this ana!ysis,;ve vill pursue obtaining a new quotation for the TMI-2 Pumps.

Reeee=e:dat ioes

. Bf=W recenser.ds. continued operation of the reactor coolant pu=ps.

Startup' and pover. escalation data pertaining to the reactor ccolant pump seals and pump vibration data should be obtained and conpared

~ vith baseline. eference data ' This data should be forwarded to 34V for final reco:nendations and confirnation of our assessment. Ve vould recon =end pursuing the analysis described above, hovever, this vould not delay ~ the. present operation of these pu=ps.

II. CRDM's-

- Confirmation is requestei that the safety rods which were.vithdrawn during the transient were driven, not' tripped, back into place.

7 Reec =e::iatiocs Basei on the above, and the similarities of this transient to the recently analytei SWD transient (March 20,.1978), ve.do not feel there are any significant concerns regarding long-terr. damage to

the CEDM's cr their ability to. continue to perfors as designed.

Final calculations to' support these. findings are 'anticipatel' by June 1, 1978.

In addition, the.nor-*1 ' drive 'ven';ing proc dure cust be, followed e

prier to returning - the CRDM's to ' service.

IIII. Fuel

It appecrs.that' the cooldovn limit for BOL clad conpression found in Limit / Precaution Curve 1.0, C5.2 vas violated by as such as 250 psig-

= duri::g the accidental depressurication.. However, this limit represents a vers case' enyelope and does.not realistically reflect the conditicas

encountered in the T:tI-2 transient.;' A specific analysis of the TMI-2

conditions _using the supplied reactor coolant temperature /pressee Linfern:ation and the TACO code. (Version 18) indicates that. in actuality e

s 1

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5/2/78 the fuelIod cledding never experienced a tensile force. This.

conclusion is based on a conservative analysis and all infor=ation available to date. Turther analysis details can be' found in cal-Eculational file 32-90,72-00.

r Reco= endatices Although the licits/ precaution curve vas violated, a transient specific analysis indicates the fuel design criteria :vas not; therefore,' EV does not feel there is acy concern regarding the ability of the fuel to continue t;o perform as designed.

Al'1 os the above findings are based on the information provided fres the site to Engineering through Nuclear Services.

IV. RCS Vater Chemistry-A reviev has been na:ie of the operating and chemistry data asscciated vith the chloride contamination of the Reactor Coolant Systes ca 23 April, and it is our opinion that the high chlorides ' vill have no deleterious 'effect on the structural integrity of the Reactor Ccolant Syste= or associated systens and equipnent. Therefore, the Resctor Coolant Systen remains aceeptable for continued-operatica,(heatup and startup).

This letter' constitutes the require:1 engineering evaluation.in accordance with Plant Technical Specification- (3/4 h-lT).

- This eva' uatica 'is hr. sed on the folloving conditions:

l Chlorides 3.0 pps (cax. )-

Oxygen 0.0 ppa as indicated by chemistry analysis Sodiusi S130 pp=

and the Reactor Coolant Systec at Hot Standb'y Conditions.

~ This1 evaluation is specifically based on the basic pH associated with the'sedium hyd cxide contamination and the presence of lov oxygen

, : levels..

Recc==en:iat io ns In agreement; -ith the directions ~ outlined for a recovery pr gram following Ja similar checistr/-contamination of ~ the plant, BIN vould expect a

' concentrated effdrt to' eliminate ;the currenc Reactor Coolant Systes, A

6 v=

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.,'B*437,]and auxiliary systemsland l piping contamination.,The Reactor P

Coolantf SystenL s:diu= levels must be reduced belov.2 ppn prior to

. achieving. criticality in the reactor with chlorides as lov;as pos -

, c sible,' and because of:the current: plant maintenance work, all control' N'

~

c rod drive mechanis=s villirequire fulliventing before return to SV c

' operations. >iIt =ust!be stressed;L novever, that Met-Li should ensure

, :thatisodium levels are belov.10 ppa in the control rod drive mechaniss f

~

J components during these venting operations, with. chlorides similarly, 4.. :

J teduced. ;

t L ' ' 51 -

i The'BiM.should have.all of /the contaminants reduced to the lovest 7<

achievable-cleanny conditions 1 _.Significan't fldshing operations on t all of thelister:onnecting piping should be accomplished to elisinate n

contarication in those areas. -

m-

- V. 2 Once Throurh' Steam Generators -

Bh*4 Mt. Vernon-section has reviewed the data sent concerning. the rapid

~

cooldevn which occurrei at 04I-2 on 23 April.. Based

on the infonation J

received by; telephone' and telecopy, it appears the hot leg reacter

~ coolant.te=perat:re dropped: from 5929 to h60'F ~ in 3-1/2 minutes; and

'. the sini=us tecpers.ture of k5k*F vas reached 6-1/2 minutes af ter the itrip.' ' The reacter ecolant te=perature remained relatively constant

^

aty this.te=perature,)for at least 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months , before a! gradual cocidovn -

s r

ve.s initiated.:

t is conservative to assuse that the maxi =us tube e

J tofshell te=;erature difference'is equal to the =aximus temperature drop _of 1389.1 3 J /.i t

' This is less than the 170*F tube to:shell temperature difference that Evas' analyzed; for the rapid cooldown at:SMU*D in March and based on that s

"ianalysistis acce; table.

k

.The; rapid _tenperature decrease vill?cause local ther= alt stresses which

vill haveito be evaluated:in order 'to ' determine the effect of this w, '

v k

i transient on the" fatigue, life of. the vessel.

Even though this transient-vas scre rapid than-chess!41Diccoldown'in l'. arch, the-total temperature

'N 2,.. e f ' drop vas; signifi:actlyiless;:and R&W feels the net effect of the C4I-2 ~

pn g

.coldeenl*Mb'e less severe than the SIEJD cooldown currently being Tanalyz ed. i h

'r

' skirt: duringithe critical. transient times considered 11n the' 0TSG Stress:

4 ttachedjis a graph 1sh6ving the tinperature distribution in the support O

r

~

JK 1Repertz l Sin:e the ree.ctor, coolant temperature did not. drop Jcelov; 45h*F t

r yd -.'

fduringitheiinitizi ecoldown,ethis transient should not' produce a-;te=per-.

C, "fature gradient anyj=cre ; severe' than theicondition analyced in.ithe: stress

" J%#' ; 1 report.h Thereferek the effect 'of :this : transient ~ on the. support skirt 1

< f1N

+.c iveuldstetth'e sa=e at s'100*F/hr' cooldown.'

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- G ' P. Miller l

^

5/2/78 l[

m.

3ased on the aferecentioned itecs, PM feels that the structural integrity of the EC-2 generators is. acceptable, for resusic6 normal

. plant. o;eration.-

VI.; Rea:ter Vessel

.A pre' 4-inary: review of the rapid. cooldovn transient telecopied to B?W Mt. Vernon section hasibeen reviewed and found to be acceptable.

J

. The telecopied'eurves of. this transient' vere difficult to read;' there-

-fore, the transient analyzed from a fracture mechanics standpoint is

~ described belov:

Ti=e Inlet Tescerature Time Pressure (Minunes).

( ?)

(Minutes)

(PSI)

0. 0 -

57 5

~ 0.0 22m c.2 :

575-1.0 1750

.2.6.

455' 2.5 820 25.c h58_

3.0 760 115.o.

'h80 5.0 980 8.0 1520 11.0 2220 13.0 2160 15.0-2200 g-115.0 2200

Cc:;onent T.ngineering 'nas received the computer printout of the fracture mecha:ics analysis for the core region and the outlet not:le-no::le belti r egion.. The stress intensity factors for this transient vere less than those' experienced by SMUD during a rapid cooldova.

Preliminary evaluationLet the analysis indicates. that this ar,ea is of no concern based on-the transient data.-

- A 'detailei ;ri-* y'. plus n secondaryiand subsequent - fatigue analysis vas not. performed.:. She short tine duration for the transient results 'in

.peah skin stresses on the inside ' surfaces and'small discontinuity-

- i I

stresses -(priscry *_ni secondary).'.Using an uppe.r bound peak skin

)

stress equatien vhich assu::es an' infinite fib coefficient and a i

ste;> deva' in. fluid. temperature, the resulting skin stress for this L transient tis-approximately 26,000, psi'. : One cycle of this 26,000 psi the-~*; stress and pressure stress lumped with ' the maximu:s stress.

-l resnitii.; ~ fro: nor=al operating transients vill! have. an. insigniticant i

effect ca.the et=ulativel fatigue usage factors of. the reactor vessel-i

'. con;ccentc ;

'l 1 g..g

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+

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- ~

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- A further: fatigue Justification:of this transient is. that it. is

-:less seversithan the SMUD rapid cooldown! The(SMUD rapid cooldovn d[ ' -

s Etransient 1siin the~ process of being analyzed by utilizing sinplified,.

7 y

' ;. conservative ^ analysis methods. This' analysis.is,in the final; stages.

- l-and;the incres.se in' cumulative factors at the :more critical locations is s s:::all. r kr uYII. 2 Reacter Cools =t' Pising and Pressu-izer W

-a

T 1The trassient data shovn in the table ~below has been revieved for preli=1: ary -icpact.on;the ^ operating ' lite of the' TMI-2 reactor cool' ant I piping and pressurizer.; The results or,this reviev shov there vould j

the. ::o 'significant inerease fin-the cucule.tive usage factor. for.any

portien of the reacter coolant piping or' pressurizer.. ' Eevever, a

~ ; detailed' analysis vould;:eed to be perforned at a later date to doc-1

- unest-this transient's actual effects for purposes of; the plant's L life histe'ry CL7.

x iMost ofithe reacter coolant piping has, at present a lov CUT as does

, : the pressuri:er..Therefore,. the increase in CUT due. to this one

- time occurring transient vould ;not. cause any portion of the reactor 7

. coolant pipingic pressurizer to have a CUT. greater than allovable

~

for. the life of: .the plant.

~

Thus,: there are olk cvn' consequences of this abnorcal transient'

~

on: the reactor ecolant piping and pressurizer.vhich vould prevent N,

u startup of the plant at this tise.

~

M' TETdATUP.I.

t ' DIE -

zCCIT:

-DLZr PRESSURE

0.0

'592.

57 5 2200

. C '. 2 -

592

,575

'[, [

O.8 *

.1. 0 :-

. 1520 455 1750-

2. 5 -

- - ~ -

820.

m

- 2.T;

. k55

~.

L..

v..

3.0-160 5

L5 0!

980 1

8.0-

.. ~

r

-1570 311.0; 2200 713.02

-1 21601 i G16.0L 12200:

126.0 4 58)

L458?

^

(115.0i h80l A80l

~4

. f 2200 s

~

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y o

y l#

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~

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+.,

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+

t5c et We

A further fat!6ue justification of this: transient.'is that.it is "j

iless severe than the SMUD rapid cooldown. The SMUD rapid cooldovn 18h transient isI in the' process _ of beir.3 analyzed by utili:ing'si=plifie1, h

! conservative analysis methods...Th'is analysis'is-in the: final. stages f

- a.nd the11screase in cu=ulative factors 'at the more critica.1 locations 1.s s=all.-

VII. P.eacter Coolant Pining and Pressurizer

~

n.The' transient' data-shovn in the table belov has been reviewed for-preliminary impact on the opere. ting life of the TMI-2 reactor 'c'colant

- pipin.g and pressuri:er. ::The results of' this review show there vould

'te :o..significant increase in the cu=ulative' usage factor-for any-pcrtion of the reactor coolant piping er pressuri:er. F.ovever, a 1

idetai2.ed analysis vould need to be performed at a later date to doc usent this transient's actual effects for purposes of the plant's

~

life history CU?.

.Mos of. the reacter coolant piping has, at present, a lov CUF as' does the pressurizer. Therefore, the increase in CUF. due to this one

.ti=e occurring transient vould not cause.any portion.of the reactor coolant piping er pressurizer to have a CUF e,reater than allovable

^

T for the life of the plant.

' Thus, there are no known consequences of this abner =al transient on -he reactor coolant piping and pressuri:er which would prevent

~s artup of; the' ;1 ant' at this time, t - i

- TIMPERATURE

! T ME

-OUTLIT INLI7 FRESSURE'

[0.0 -

.592 575-

~2200 o.2 592 575 o.8

520 455 J '.. o 1750-1-

2.5-820 27, kS5-3.o' 760 5.0 980'

~

T8.0 1570;

'11.0 2200

'13.0' 2160 16.o 22001 125.0-h58:

458' 11.15 0

-:h80 l 480.-

2200 4

.b' s.

37:--

x s-

.d 5

4

( p-i e

s

.n.

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rr....

- l; T-5/2/T8

.G. P. liller

a...:

t-4ll.

1, il_

c All of:the;above preli=' -y evaluation' results are transmitted to

you in order lto; support cyour presentation of. an ' initial report.

r

s 1i-3&'4 assunes that _a full-detailed report requiring a~ longer time

. span to prepare. vill be presented.

In. support of.that presise,~.

3&*J is ' ecatin.:ing to assenble the necessary infor=ation to provide i Met 'ui vith the detailed evaluations of the transient on'the NSS 4

supplied equi;=ent.. She Site Office vill be able to give you an estinate of the: detailed report delivery date.vithin a~ short period.

ff

?If you'have any further questions, please do not hesitate to contact me.

. Very truly yours,.

J

'3L,-

/

L. C. Bogers Site Operations Manager ICR/ bay'

=

cc:

L'. IR..' ?1ethe W. II.~Spangler 7

G. K. Vandling

"a J.'G. Herbein

R. M. Klinga=an L. L. La.fer J. 3.~ Logan-J.'L. Seeli=ser 5

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