ML20010B556

Requests Approval to Operate Facility at 100% Power by 780915.Draft Facility Investigation Fluctuation Status Encl

ML20010B556
Person / Time
Site:	Fort Saint Vrain
Issue date:	08/11/1978
From:	Justin Fuller PUBLIC SERVICE CO. OF COLORADO
To:	Gammill W Office of Nuclear Reactor Regulation
Shared Package
ML19263D739	List: ML19263D738 ML20010B552 ML20010B554 ML20010B556 ML20010B559 ML20010B561 ML20010B562
References
FOIA-81-127 P-78137, NUDOCS 8108170232
Download: ML20010B556 (77)

Text

%..

g t,

e v

o (n----..--

I pubHe Service Company @ Odmede k's i, P. O. Box 361, Platteville, CO 80651

. t.

s k--

NJ August 11, 1978 Fort St. Vrain Unit No. 1 P-78137 Nr. William Gammill Assistant Director for Advanced Reactors Divicion of Project Management U. S. Nuclear Regulatory Commission Washington, D.C. 20555

~

Docket No. 50-267

Subject:

Safety Evaluation - Reactor Outlet Temperature Fluctuations Ref:

NRC Correspondence, July 6, 1978 Gentlemen:

In the referenced correspondence, the Public Service Company of Colorado was directed to provide a safety evaluation addres:ing the safety significar.ce of the Fort St. Vrain reactor region helium out:let temperature fluctuations that have been observed on occasion.

In response to your request, a draft safety evaluation was forwarded on July 14, 1978.

Comments on this first draft were received from members of the Staff at meetings at the plant site and at the PSC Engineerink Division offices on July 17 and 18, 1978.

PSC and GAC met with members of the Scaff in Bethesda on August 1,1978, and reviewed the second draft of the SER. At this meeting a discussion concerning the effects of broken fuel alignment dowel pins and the effects of fuel coolant

. hole blockage ensued. PSC and GAC vere informed by the Staff that they should expect a call on Friday, August 4, 1978, to further discuss the effects of fuel coolant hole blockage.

In a telephone call received on Monday, August 7, 1978, the following inform-ation was verbally given to the Staff:

1.

If thr'; )

dowel pins associated with a single fuel element were o, ~ed to be brokett and the element was assumed to exp rience exactly 0.625" of lateral displacement, complete blockage of all coolant holes in that column could result.

(

810F,170232 810529 P DF. FOIA MULLEN 81-127 PDR

.e J

t r

August 11, 1978

(

Mr. William Gammill Page two 2.

If it were assumed that only one (1) dowel pin remained intact, engaged to both the top and bottom of the element and the element experienced 2.75' rotation about these two dowel pins, coolant hole blockage vould range from 20% to 100%, depending on the location of the coolant holes relative to the rotation axis.

NOTE: Rotation of the fuel element would be possible if only one (1) dowel pin or two (2) dowel pins along the same vertical axis remained intact.

Any other combination of intact dowel pins would preclude fuel element rotation.

3.

Complete short time blockage ( 20 minutes) would result in a substantial temperature rise and fuel failure.

It should be noted that the design of the fuel columns in the Fort St. Vrain reactor results in their acting as a single tall fuel element and not as individual fuel elements. Any physical movement of the core support blocks

~

'results in coordinated motion of all those fuel ele =ents supported by it.

Lateral motion of a fuel element column due to lateral forces imposed by dif ferential pressures between refueling regions of the reactor results in fuel column tilting, not individual fuel element movement.

To etjerience dowel pin damage that would permit J neral motion of one fuel element with

. respect to another within the same coa =n, it must be assumed that forces exist that will act on a single fuel element in a column.

This is not

. generally the case, and therefore, lateral motion or rotation of one fuel element with respect to another is highly unlikely.

"In addition, the reactor core is designed so tha't the central fuel column in a region of seven columns is vertically displaced from the others so that no continuous horizontal plane exists in the core except at the fuel column 4

bottom reflector / core support blc :k interface.

As previously discussed with the Staff, fuel damage likely will result in an increase in the primary coolant system activity inver. tory. The activity is monitored and recorded on a continuous basis and samples are analyzed on a weekly basis to determine specific activity and to identify the various isotopes present, per Technical Specification requirements.

In the event the activity should increase by a factor of 25%, daily samples must be analyzed until a new equilibrium inventory is established.

These monitoring requirements are established in Technical Specification SR 5.2.11 - Primary Reactor Coolant Radioactivity Surveillance.

Primary coolant system activity inventory limits are established by Tech-nical Specification, LCO 4.2.8 - Primary Coolant Activity, Limitinst Conditions for Operation.

Should these limits be reached, the reactor would be shut down.

(

l

a:

c*

~

q August 11, 1978 EMr. Wil lam Gammill Page three It is again stressed that the reactor iji not operated continuous 1v in the fluctuating mode.

As previously discussed with the Staff or. numerous occasions, the facility operators have explicit instruction; that upon observing a fluctuation during routine operation, they are to take imme-diate corrective action to arrest it.

This action consists of reducing reactor power to whatever level is required until the fluctuation is damped out.

It should also be.noted that the fluctuations are not divergent and that the temperature amplitude and period are independent of power level.

Testing to date has established a " threshold line" of reactor core pressure drop vs. reactor power below which fluctuations have never been observed or have been successfully induced.

The plant is routinely operated below this

" threshold line" where no fluctuations are observed.

If data is to be collected with the reactor core outlet helium temp. sture fluctuating, operating conditions are established to raise the reactor core pressure drop and the reactor power increased to' induce the fluctuations.

Fluctuations can easily be detected by the normal plant instrumentation, i.e.,

steam generator main and reheater steam outlet temperatures, core linear power channels, and core region outlet thermocouples.

The steam generator main and reheater steam outlet temperature recordings are probably the best indication available.

When the reactor is operated with the helium outlet temperatures fluctuating for test purposes, specific limits are placed upon the allowable variation of various measured parameters.

Data frc= individual test runs have been compared for both amplitude and periods of fluctuation.

Any future data will also be so comparad.

Insofar as the reactor is not operated on a continuous routine basis with the core helium outlet temperature fluctuating, no specific limits have been established for fluctuations on a long term basis. As previously mentioned, if fluctuations are noted during routine operation, reactor power is to be immediately reduced to terminate the fluctuation.

If enough data is avail-able from such an observed fluctuation, its amplitude and period will be' compared to those of previously observed fluctuations.

Public Service Company is most anxious to take additional base line data below and above 70% reactor power, utilizing additional diagnostic instrumentation recently installed. This data would provide additional information to be used in resolving the fluctuation problem.

This data would also provide a basis for comparison of reactor operating characteristics following application of appropriate corrective measures to eliminate the observed fluctuations.

The testing contemplated is described in tests RT-500 and 502 that are attached to the accompanying SER.

(

\\

J

2 2

August 11, 1978 Mr. William Gammill g

Page four Public Service Company requests an immediate release to continue fluctuation testing below 70% reactor power on the: basis of the attached SER.

PSC indicated in its correspondence R-78131, dated August 8, 1978, its intent to operate the Fort St. Vrain facilidy at as high a power level as is per-mitted. PSC's system demands have ex'ceeded all previous records throughout this summer period and system load projections indicate continuing record demands. This week PSC requested some of its larger customers to curtail their-usage due to the unanticipated loss of a generating unit at the time Fort St.

Vrain was down. The maximum achievable output of Fort St..Vrain is needed to meet these system demands, both now and in the near forseeable future.

PSC therefore requests thas'iNe NRC approve operation of the Fort St. Vrain facility to 100% reactor power no later than September 15, 1978.

Very truly yours, PUBLIC SERVICE COMPANY OF COLORADO

/

[

. K. Fuller Vice President Electric Engineering and Planning JKF;il r

Attachment

^

^

. <. ~.

9 i

STATE OF COLORADO

)

)

ss.

CITY AND COUNTY OF DENVER)

J. K. Fuller, being first duly sworn, deposes and says; That he is Vice President, Electrical Engineering & Planning of Public Service Company of Colorado, the Licensee herein; that he has read the attache.d submittals and knows the contents thereof, and that the statements and mat ers set forth therein are true and correct to the best of his knowledge, information and belief.

~

p. Fuller Subscribed and sworn to before me this lith day of August, 1978.-

Witness my hand and official seal.

My commission expires:

4-S-71 2 M Gctbat?

Notary Public

)

8 e

e g

l s.

l 1'd D'S DEFT 7-28-78

. -.... w FORT ST. VRAIN CORE FLUCTUATION INVESTIGATION STATUS 1.

INTRODUCTION /

SUMMARY

The investigation of the nature and causes of the fluctuations in the Fort St. Vrait, reactor continues to be studied vigorously.

The program includes both experimental work at the site,and extensive analysis of the data in San Diego.

The fluctuations, which affect the nuclear channels, the region exit temperatures, and the steam generator module temperatures, are irregular and complex.

The period of the fluctuations ranges from 5 to 20 minutes with the largest amplitudes observed in the northwest sector of the core.

The changes _

i on the nuclear channels are very rapid and do not shov-.he normal temperature feed-back_ effects,(see Figure 4-1).

The temperature changes are much faster than could be achieved with region power changes (see Figures 4-1, 4-2, and 4-3).

The total core coolant flow and core thermal power, however, remain essentially constant

'during fluctuations.

t A postulation of small movement of core components (fuel elements or reflector elements or core _ support floor) which changes the distribution of inter-region gaps and coolant flow and pressure drop through the gaps may

{

explain the observations. These movements may be induced by thermal and pressbre gradients. __ The observed abrupt changes in the steam generator module

'I helium inlet temperatures may' be~ ~bxplained by sudden changes in the fraction of.

cool gap flow entering the steam generator module.

The larger abrupt changes in nuclear detector readings appear to result from increased neutron streaming from the core to the detectors through narrow permanent sit'3 reflector gaps which are opened or widened by block movement during fluctuations.

Analysis has indicated that the fluctuations observed on the~ region exit gas thermocouples are consistent.with a small intennittant cool bypass flow which enters the thermocouole probe at the side of the core support floor.

. ~-. m u-.

-.~.

I' this flow is brought into intimate contact'with the thermocouple, it would have a large influence on the reading.

The coo 1 bypass could be interrupted by opening and closing of the gaps between core support blocks and thus give the

T,

o e-l 2

impression Lof large region power or flow fluctuation when, in fact, no significant changes exist.

The observed small drop in core flow resistance (less than 5% at the enset of fluctuations}is consistent with the concept of core component move-g.

me The theory that fluctuations are due to core component motion is supported by data obtained from recently installed instrumentation.

Displacement probes attached to the PCRV indicate that the amplitude at its normal resonance frequency -

is larger during fluctuations and is correlatable with the other fluctuating signals (see Figures 4-5 and 4-6).

F_ission _ chambers hay.g bg placed adjacent g

,to th,egV liner, but no.t p line wi._th permanent sp. reflector gay.

The fission chambef.s do not show the abruRjympM21erved in the PPS nuclear channels, Nang s ag not true reactivity changg.s.

However,

-t h indicating the abyugt u

0 fl small reactivity changes are occasionally observed which can be explained by parts of the core moving together to close up intra-region gaps.

Four thermo-l couples, inserted through the core region exit penetrations into flow gaps in the north-west sector of the side reflector, show constant readings at normal operating conditions. When fluctuations begin, they exhibit peak-to-peak varia-tions of up to 40*F.

These fluctuations can La correlated with other events st '

1 as rapid changes in nuclear flux channels.

- Fluctuations have been observed a number of times under a variety of core -

I conditions and at power levels'between 40% and 70%. The observed temperature swings during fluctuations-have s.tayed within plant operations and technical-

-... ~ ~

specification limits.. (During one test a few minutes of operaticn in a range i

limited to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> was experienced.)

However, since the cause of the fluctua-tions is not precisely known and so.me core motion is indicated, it is important l

that the fluctuations be characterized and a mode of stable operation be completely identified.

The test program conducted thus far has identified the core pressure 1

dgop (core resistance) as one important operating parameter which may be 'used to 3

vary the threshold power level of the fluctuations ~.

It has been demonstrated that a reduction in the flow resistance of the core (obtained by opening the variable orifices) does raise the threshold of fluctuations to at least 70 I

([

thermal power.

The core has been subsequently operated at about 70' power for an extended period of time without fluctuations.

Extrapolation of avail-able data suggests stab 1e operation can be obtained at power levels above 70%.

~.

c,

. (

2.

HISTORICAL

SUMMARY

OF FLliCTUATION EVENTS A total of 20 occurrences of core fluctuations have been identified ~at Fort St. Vrain.

The first fluctuation occurred on October 31,177.~at 58%

power during a aradual rise in power from 50% to 60%.

SubseqJently, a series of tests were performed to investigate the characteristics and source of the fluctuation and to determine the parameters which affect them.

Following the initial fluctuation event, plant parameters affected by oscillations were specifically. monitored throughout plant operations to enable imediate detection of fluctuation.

Additionally, limits were placed on the pemissible amplitude of temperature fluctuations to the individual steam generator modules.

In all cases it was demonstrated that reducing reactor power is a reliable means to stop the fluctuations.

The occurrence of fluctuations has been observed to be sensitive to

(

core pressure drop (see Table 2-1). This core pressure drop (a P) is a f

~

function of (1) helium flow rate and (2) the combination of the resistance cf flow through the orifice valves, the core coolant channels, and the gaps within the core and reflector.

This total core resistance can be varied by changes in the orifice setting.

More closed (high resistance, R), intermeetiate (normal R and low R) and most open (low-low R) orificing conditions have been employed.

No. fluctuations.have been observed during extensive plant operation

'+W.

up to 70% power with the orificing in the low-low R configuration.

The only 2

times that fluctuations have been observed below 50% power is when the core was purposely orificed for high resistance.

Fluctuations have occurred in other than the low-low R or orificing con;

~

iguration as a result of ev.ents 'ausing rapid power and flow changes, in c

\\

....,a...

fact, a rapid load increase of 5% at the rate of 3 to 5% per minute has been used to start fluctuations for the purpose of testing.

A grid frequency dip and i

a plant feedwater upset, which through the contro1 system cause rapid power-flow f

~

r changes, have initiated fluctuations.

Other fluctuation events have started

(

spontancously following slow power increases, changes in one or more orifice 4

l valve positions, swapping of regulating rod and shim bank positions, and in l

I some cases even withou.t identifiable preceding event.

Spontaneous initiation at given conditions has nat_bestregeatable.

o

(

Table 2-1 contains a list of all the fl.uctuation events, the date and times of the fluctuations, the core resistance resulting from the orificing configuration, and the events that took place at the onset of the oscillations.

8 T

.e*

e 6

e f

4 s

9 e

(

0

, e q

n

Tabls 2-1 FSV Core Temperatu,_'fluctutt$on? Events Clock Time

^

Hrs.

% Power Core AP Date (ORIFICING)

Start Stop psi Events :at' Start * *

• • '

Comment 10/31/77 2230-2315 58 53

3.6 Spontaneous

Gradual Power (NORM - R) Rise from 50% 11/23/77 1250-1440 59 53

3.6 Upset

Incomplete Transfer (NORM - R) of ' A' BFP Steam St'pply 11/23/77 1730-2020 53 53

3.1 Spontaneous

Swapping of Fluctuations Stopped After (NORM - R) Shim Rods & Reg Rod Shim / Reg Rod and Orifices Adjustment 11/24/77 0044-0510 55 51

3.2 Spontaneous

Gradual (NORM - R) Power Rise frram 53% 11/26/77 1550-1930 68 53

4.4 Spontaneous

40 Min. After* (NORM - R) Power Rise to 68% 11/28/77 2040-2260 53 52

3.2 Spontaneous

Following ('iORM - R) Manipulation of Orifices Regions 20, 32 and 35 4/22/78 0210-0540 61 52 3.6 Rapid Power Rise from 57% SG Trim Valve (NORM - R) Into Auto Ori-Did Not fice Changes Stop Reg Rod Int Manual at 0454 Stopped 4/22/78 2304-0028 60 54 4.0 Rapid Power Rise from 55% Power Was Lowered from ' (NORM - R) in 1 Min. 55% at 0017 4/23/78 0038-0048 54 54

3.1 Spontaneous

No Event 1 Cycle of Fluctuations. (NORM - R) Stopped after Rod Swap (2 More Cycles) c4/23/78 1152-1500 60 55 3.6 Rapid Rise From 55% In Reg Rod in Manual at 58%, (NORM - R) 2 Min. 1250. 1350, Matched Main, Steam Temps and Orifice Adjustments Did Hot Stop.

~ ~ TableIs.antinued) Clock Time Hrs. % Power Core AP Date (0RIFICIllGI Start Stop psi Events at Start Coment c4/25/7a 6-14 65 56 3.5 Upset:. Grid Frequency Dip 4/26/78 1107-1300 51 40

3.7 Spontaneous

Reg Rod Change (HIGH - R) After High Resistance Orificing 4/26/78 1356-1416 40 40

2.5 Spontaneous

No Event 2 Cycles Spontaneous (HIGH - R) Fluctuations, Self-Stopp,ing

• A4/26/78 1515-1705 50 40 3.4 Power Rise from 45% in (HIGH - R) 1 Min.

'4/26/78 gl85g) 45 40 2.9 Power Rise from 40% in Fluctuations Started $3,Mia 10 Min. After Reaching 45%. 3 5/8/78 0726-0801 69 61 3.3 Power Rise from 65% in Fluctuations Stopped After (LOW - R) 6 Min. Power Reduction. 5/19/78 2049-0032 67 60

3.3 Spontaneous

Orifice (LOW - R) Adjustments Regions 8, 10 & 16 5/20/78 0655-0706 67 60 3.2 None, Spontaneous Spontaneous Fluctuations (LOW - R) Stopped by Core Power and AP Reductions 6/3/78 1000-1100 '50 .39

3.2 Spontaneous

Orifice Orifices Adjusted to Estab-(LOW - R) Adjustments lish a Uniform Core Region Flow Configuration A6/4/78 1020-1400 45 39 3.2 Rapid Power Increase Core Region Flows Were * (HIGH - R) from 40% Near a " Uni form Flow" Configuration. ~ FM Data Set Taken, Non-Fluctuating Base Data Also Recorded, a FM Data Set Taken.

. ( 3.

SUMMARY

OF TESTING ^ Power operation up to 70% of rated was authorized on October 28, 1977. On the initial rise to power levels above the previous authorized 40% level, fluctuations were experienced et 58% power ori0ctober 31,1977 (Table 2-1). Subsequently, in the latter part of November 1977, a series of tests were run to investigate their cause and nature. These tests consisted of making known changes to coolant flow and core reactivity by slowly varying circulator speeds, reactor power, orifice valve positions, and control rod positions. Five fluctuation events occurred during this test period. For November 1977 tests scries, data were obtained and recorded using _, stand:rd :1:nt instrumentation a.M d_ata acquisition systems. The instrumentation of major interest in these initial investigations were the core outlet' thermo-1 couples at each of the 37 regions, the thermocouples at the inlet of each of the 12 steam generator modules, and the 6 PPS nuclear channel detectofs. i Figure 3-1 is a schematic plan view of the core and principle plant components. The standard plant data acquisition systems used for recording of data in this test series were the Data 1.ogger, the Model Verification Computer, and the Steam Generator Performance Monitoring Computer. The capability of these data acquisi-tion systems is shown on Table 3-1. As can be seen, sampling rates are relatively slow, with the exception of time data analyzer. For.,the period December 1977 through, March 1978_, the plant was' down; k for maintenance except for a brief operating period in January. No fluctuation t events occurred during plant operation in January 1978 although attempts were made to induce them. Following the November 19771fliicStation events, an in-core inspectid~n of the upper plenum area with a TV 'camer'a~was ~rrade in December 1977'- In addition,-- . a..> Nr Region 34 control rod drive and orifice assembly was removed and inspected in; k j the hot service facility. Region 34 was of interest because the maximum amplitude ,Q . of fluctuations occurred in that sector (north-west) of the core, and Region 34 l b< y control rods were fully inserted during the fluctuations. The inspections l q. j revealed that the upper plenum was in the as-built condition witt. no evidence !j'} of any damage er deformations. The Region 34 contrcl rod drive and orifice

( assembly and the control rods vere in excellent condition with no noticeable unanticipated effects imm t5.e normal ' operation or the November fluctuation events. The same control rod drive and orifice assembly was put back in Region 34. It was recognized early that additional instrumentation and improved data acquisition sy!.tems would be required to understand and identify the cause of fluctuations. The additional instrumentation and data acquisitions were defined, purchased, installed, and made operational by the time of resumption of testing in April 1978. The new instrumentation, plus some of the existing plant instrumentation, were monitored on broad-band frequency measurement (FM) recorders. For logistic purposes, four FM recording stations were established. The types and number of instruments monitored by FM recorders are listed in Table 3-2. The location of the instrumentation relative to the PCRV is shown in Figure 3-2. The location of the traversing thermocouple at gaps in the core support floor blocks is shown ( in Figure 3-1. The two displacement probes are located on the PCRV at the upper l elevation in the NW and SE sectors (see Figures 3-1 and 3-2). FM data were recorded for three fluctuation events in April 1978. Two of these events included FM recordings of nonoscillation period as a data base (Table 2-1). During the April 23, 1978 fluctuation, the regulatiing rod was placed in manual, but this did not stop the oscillations. Results from the testing in April with the new instrumentation revealed that the accelerometers did not record a signal that was above the electronic background noise. The accelerometers have been removed since no useful data was being obtained. The electronic package for displacement probe #1 was not operational and is being replaced. Displacement probe #2 was operational and provided useful. and significant data. The balance of the new instrumentation and FM recording systems were and remain operational and provide mear:ingful ( data for the fluctuation investigation.

. ( Other tests have been' performed since the resumption of operation in ~ April to understand better the oscillation phenomena. These include determina-tion of region outlet thermocouple time constant [s, influence of control rods on nuclear detectors, and the results of flow and regulatory rod cycling on t' individual regions. These tests were run in a 30 to 50% power range. The l i f region outlet thermoccuple's time response was determined to be inversely pro- ' pertional to flow rate. Time constants for.both region and thermocouple response / varied from 3 - 5 minutes for high-flow regions and 6 - 10 minutcs for low-flow [ k regions. Control rod movements were clearly detected as signals in the same j j direction in all six PPS flux. detector channels. Region exit temperature response M to region or core flow variation followed analytical predictions. Further upgrading of plant instrumentation and data acquisitions system have been incorporated or are being implemented. All the data acquisi-tion systems have been placed on a common clock for improved data analysis. In the last part of July 1978, up to 26 traversing core outlet thennoccuples and b g \\ 5 PCRV displacement probes will be operational. These displacement probes will I be capable of measuring displacements at four axes, tilt and PCRV torsional movements. The data from these additional instruments will t'e monitored on FM recorders. Two instrumented control rod drives (ICRD) have been prepared for future-insertion into the reac, tor. The instrumentation incorporated in these assem-blies is shown in Table 3-3. The physical location along the core axis is illustrated in Figure 3-3. The instrumentation on-these control y drives was defined ~shortli j after the first fluctuation had" occurred to' provide in-core monitoring < \\ I capability for neutro[ flux ~,(tEmpeFatiire',[and/or flow. Based on today's kriow~ lege, it is expected that after; insertion the ICRDs will confirm the conc'iusions previously reached about various theories involving inter-region neutronic and ( i k flow stabilities. 4 ( ' U,a cc yn & c4

m TABLE 3-1 COMPUTER DATA SYSTEMS 6 9 NAME DATA RECORDED /NO. CHANNELS SAMPLING RATE 1 DATA LOGGER PLANT PARAMETERS /s500 1 MINUTE MODEL VERIFICATION PLANT PARAMETERS /80 5 SECONDS STEAM GENERATOR STEAM GENERATOR PARAMETERS /s450 7 SECONDS TEMPERATURE SCANNER TIME DATA ANALYZER CONTROL PARAMETERS /32 PATCHED CHANNELS 0:1 SECONDS e 5

TABLE 3-2' FM RECORDER SIGNALS ~ SIGNAL NUMBER NUCLEAR CHANNELS 7 FISSION CHAMBERS 3 4 CORE AP 1 STEAM GENERATOR MODULE INLET TEMPERATURE 12 GAP TEMPERATURES 4 DISPLACEMENT PROBES 2 ACCELEROMETERS l ENDEVC0 13 WILCOXON 3 B&K 1 6 6 e

• g.

4 e

V r TABLE 3-3 / f INSThuMENTED CONTROL R0D DRIVES INSTRUMENTATION 1 2 FISSION CHAMBERS 2 FISSION COUPLES / >ZEROSUPPRESSEDI 2 NEUTRON DETECTORS 6 THERM 0CCUPLES BEFORE IN-CORE 2 DELTA-P (ORIFICE) TRANSMITTERS LIFICATION INSTRUMENTATION 1 LVDT (DISPLACEMENT TRANSDUCER) 2 MICROPHONES j 4 e 6 'l

• FF F

r

i FIGURE 3-1 t_. _ DP-1 O 8 LOOP 1 LOOP 2 1 I B-2-6 FC-1 g CIRC D NUCLEAR CllANNEL B-1-6 g' s DETECTO VI O B 'T-4 C T-1 B-2-5 3 . 36 37 20 T-2 B-2'-4 34 18 19 8 21 .\\..' j B-1-4 ! 33 17 7 2 9 22 i fF )o IV I 32 16 6 1 3 10 23

• FC-2

] r 31 15 5 4 11 24 /~ j I ~ B-1-3 30 14 13 12 25 B-2-3 29 28 27 26 ~ h' Cone REGION - ~ -ss,/ 9 B-1-2 III l p III .- 2 ~ - STEAM CENERA10R EODUI.E FC-3 B-1-1 B-2-1 CIRC A CIRC C O Dri2.. ( s 'J . ~ - j

FIGURE 3-2 ' ~ FORT ST. VRAIN WIDE BAND INSTRUMENTATION '/ .oc h NUCLEAR FLUX \\ y .Tc .' M.-~ 4 CHANNELS (7) __ W-: e 5 := := ,s Ny

• \\f '.

REGION 34 / !.e 3 3 AXIS ACCELEROMETERS y s (( 1 >>. ~~ fl 6 DISPLACEMENT Cmdg'_l[U;C'?;T1 . ~ N ELIUM PERMEABILITY N r.;r-T,,'m PROBES (2) h g l

• l

' P' i T.~. jj ;; p i i i H '[E!!.i;X..]d j PE?iETRATIO:iS & KEY'!AYS I i l /dE HE!y;Wiif O i Wp (11 ACCELERCMETERS l AND 3 FISSION CHAM 3ERS) M;lijf,'.! U P i I . JI j]!ly'ja.W.;

.m..

A f~ ) tT (3 ACCELEROMETERS & igg,y\\ T ,d \\ iTRAVERSINGTHERMOCOUPLES) p j' - \\ D SI STEAM GENERATOR HELIUM INLET -1 TEMPERATURE (12) l r ? ' W f*.NAY _. w kO HELIUM PERMEABILITY / SURVEILLANCE TUBE (1 ACCELEROMETER)

/'Q. 3 y 0 1 -;n?

.u 99 4 x-e

• Relative axial locations of in-core instrumentation i

.i .l. t FIGURE 3-3 l l 1. e r-. _ _.D y, -r l l.! I d i: I I. t 'i* 3 j Thermocouple 8 .,I. I i No. 1 'I 3 j i Fission Chamber t e e e ) 6 i i L,j{ Gulton Miciophone: i i - (on CGO assembly l f-jj S/N 43 only) i .4 I

o.

4 4. i i / SPND s ,i. 7 .c 4 N l N. 'N. I N Thermocouple i No. 2 j p i i I Kaman Microphone I I (on CEO assembly i 4 S/N 20 only) ~~ e i n -*-V-1 'l i .l 1 l ( i ! } Thermocouple 4 i M' No. 3 e e u g *.s Tt,: Fission-couple m_. /i.. i j ...fij:l-Ji i :i :p: i [ !l-n i 1 !j -c.4 /, :..J T i \\.d,.

g. ai/ iy'r;..at.;/.r.

nl: - i ;..l - b s. ,.1 I ^'t / . 1 T- ~ - ~ ii i ig i b .. h(.....__1 ( e 5-l. i g... 3...... ; _ _ g,) g, p ; t Il fi 1F I i n

6 = 4. FLUCTUATIONDATAA5DINTERPRETdIION 4.1 DATA OBSERVATIONS More detailed observations are possible for events occurring in April and after, due to added instrumentation and data acquisition systems. However, the main characteristics ~of fluctuations _have.not. changed since the { . initial event in October _1977.' Following are key observations from the recorded data.

1) Fluctuationsare, irregular,compleEandcorewidi.

However, the fluctuations first occur in the northwest sector of the core and have the greatest amplitude in that sector.

2) The fluctuations affect core region outlet gas temperature,

}I steam generator inlet temperature and flux readings on neutron detectors, In spite of the above fluctuations 15 individual readings, the average system parameters are not significantly affected. 3) Fluctuations have occurred at power levels between 40 and 70% power. The threshold power level and fluctuation character are sensitive to core pressure drop which can be varied by orifice valve adjustments or by changes in the total helium flow rate.

4) The~ pErio'd~f6F'f1GEfrati6ns^rITge"Wom s to'20 ininutes.

~

5) At the onset _of a. fluctuation, changes in various instrumentation 3'

signals occur simultaneouslyt including nuclear channel VI in the northwest / quadrant, reflector gap temperatures, change in core flow resistance, and.. displacement probe #2 ' movement. _ ~ 6) Once fluctuations are initiated, there are rapid changes in steam generator module inlet temperature with some adjacent modules being out-of-phase. The out-of-phase characteristics are also exhibited by the core 7 outlet thermocouples readings in some adjacent refueling regions.

= .g. \\ 7) Steam temperatures lag inlet helium temperatures to the steam generator modules. 8) Placing the regulating rod in inainIiiWIfo'es"n'ot ' slop the fluciuations.

9) The characteristics of the temperature fluctuations (such as amplitude, period, and general shape of the temperature cycles) are observed only during fluctuation periods.

That is, the characteristics are not observed (at a much lower amplitude) during normal steady-state operation. Figures 4-1 through 4-8 re typical examples of the previously listed observations. Figure 4-1 shows the relationship of nuclear channel VI and steam generator m6aules B-1-5 and B-1-6 helium inlet temperatures. Note the square wave configuration of the flux signal on nuclear channel VI. Also note the outs of-phase fluctuations in adjacent steam generator modules B-1-5 and B-1-6. Note the changes in channel VI readings coincide with the rapid changes in steam generator temperatures. Figure 4-2 illustrates the mismatch of region outlet temperature; between four regions. Regions 34 and 35 are fluctuating

together but out of phase with Regions 36 and 37.

Figure 4-3 illustrates the ~ gas inlet temperature fluctuations to the six steam generator modules in loop 1. The greatest amplitudes are experienced in B-1-5'and B-1-6 which are out of phase H th each other. Figure 4-4 shows the gas inlet temperature and the j steam outlet temperature of steam generator module B-1-6. The helium temperature f, fluctuations' lead and have' grsater amplitudes than that of. the steiin. Figures \\' 4-5.and 4-6 show a typical relationship between the six nuclear channels, core pressure drop and PCRV displacement probe Op-2. In both figures a sharp shift in nuclear channel VI is accompanied by a sudden increase in movement of the DP-2 and a change in core pressure drop. Figure 4-7 shows the fluctuations of core support block gap temperatures in the north-west core quadrant as measured r with the four traversing thermocouples. Note the uniform temperature measure-ments,xter to the onset of the fluctuations. Figure 4-8 shows PPS nuclear detector '.hannel VI of the PPS during the one exp'eriment when the regulating rod was pp, in manual during a fluctuation event. As evidenced by the figure, ( I -a

this action did not stop the fluctuations. Figure 4-9 compares steam generator helium inlet temperatures for fluctuation events of 26 November 1977 and 23 April 1978. Although the cycles are typically irregular, the average basic period, amplitude and shape of the temperature fluctuations are essentially the same for these two events. It can be concluded that the fluctuation characteristics have not changed significantly since the initial occurrences in November 1977. Figures 4-10 through 4-14 show comparisons of plant parameters before, at the onset and during the 4/23/78 fluctuation event. At about hour 1120 a rapid rise from 55 to 60% power was made in two minutes. The plant parameters remain stable and typical of normal plant operation until the onset of fluctua-tions at about the 1152 hour. Figures 4-10 and 4-11 show the six nuclear detector readings. All six channels record the power increase followed by the expected gradual reduction due to the negative temperature feedback. Channel VI flux readings exhibit the largest amplitudes during the time of fluctuations Figures 4-12 and 4-13 show core outlet temperature readings for Regions 31 through 37. The temperature readings generally exhibit a gradual increase as a result of the power increase but maintain stable characteristics until the onse. of fluctuations. Figure 4-14 shows the helium inlet temperature to the loop 1 steam generator modules. The temperature increase due to the power change is clearly evident on all modules followed by stable operation again until the onset of the fluctuations. Once the fluctuations are started, the temperature readings become irregular with the out-of-phase relationship between modules being clearly evident. As previously stated in Section 2, the initiation of fluctuations has exhibited a threshold with core pressure drop which can be varied at a given flow by orifice valve adjustments. Figure 4-15 shows the threshold line for the onset of fluctuations and also shows the extensive time of power operation without the occurrence of fluctuations when the core pressure drop is maintained below the threshold line. 4.2 INTERPRETATI0NS'- ( A number of theories have been investigated as the cause of the observed fluctuations. The folicwing;is a discussion of theories and the c,onclusions; drawn from, the avail _able data.

t. 1) The secondary system was considered as a possible cause of the fluctuations since feedwater upsets had been observed to initiate fluctuations'.- It should be noted, however, that feedwater uosets through the plant control; system will initiate cha'nges in ' primary flow and power almost immsdiately. The secondary system was ruled out as the cause of the fluctuations since steam temperature lags helium temperature Further, the helium temperi~thri'fluctua a tion amplitude decreases through the steam generator.' ' 2) Coupling or interface with the control system was evaluated. ~ ~ It was cortcluded this was not viable due to localized character of the fluctuations, the low frequency of'paramet'er Shinges, and the small change in theioverall or~ average value of the system parameter's. 3) Flow instabilities were considered. It was concluded that while not impossible, no viable candidate could be identified for the mechanisms evaluated, flow induced temperature and neutron flux changes would be slower than the rapid changes observed. y

4) Mechanical movement of system components was considered and dis-carded because most system components would move at higher frequencies than that observed in the fluctuation data.

5) Nucl ea r i ns ta b i l i ti es weF~e' css i de red's ne e ' th'e'idicliiF ~ds t' ecto r's p exhibit sharp fluctuationi:. o l\\ -- ~ ~ This.Nas. discounted becau'se.' f.theLnucibir'dstector 3 / response does not include the .-.i= temperature feedbacE. 0n numerous occasions n.. - ~. -,. - norma (. t ( the detector flux readings move in opposite directions ' simultaneously, and in \\ S some cases the flux change shows' on'361y'one of ^th~e~iWdet'ectorsr 6) Thermal and pressure gradient induced motion inside~ the core i 'Cl cavity was invest l gated and best correlates with the available msasured data " during fluctuations. The rapid large square wave' changes in indicated flu 7 on ) r t'he nuclear channels,can be explai,ned_by_ne.utron_ streaming.due to opening and M closing of gaps.inithe side Veflector. Gap temperature fluctuations in the core l n ( support floor can also be explained by opening and closing of these bypass flow paths. The rapid changes of inlet gas ter.perature to a steam generator module e

e. ^ l FIGURE 4-2 REGIONS 34-37 EXIT GAS TEMPERATURES. LEGEHD REG 34 1400 ...i REG 35 REG 36 - : n;. 1300-t. REG 37 O u-, . ~, ' '\\ l' \\ E ^' a. /" N ^- ~~~~~~ G /% ~ s, y,j '\\/ \\/ ~ p v g 1200[ E QA ,I gfjs - _<qb 4A ' N E s ,- 3 p v,. /., g / / %s/ O s/ \\/ g / g / / \\,J d \\/ %. / / f ~ ', .,, / r , ~., l l\\ ~.,. ,,jjjg'lllb $ Y lll. . c. 1000-i, i ii ,,,i i .ii iIiii iiii i. 0 10 20 30 '.40 50 60 70 TIME: 11-23-77./.1240-1350-4' .g 9 [,

r-F,IGURE 4-3 LO0P1STEAllGENERATORTEllPERiiTUREHISilATCH ~ 250 f / W 'i f,p j [y # 4 f f ' k B-1-6(250F) 3 .79 t #d.3 200 j N /- .g ez "O-n-1-5 (200oF) yk ,9 H gi q es f n-1ti(15ffF) _ _/,\\,..,f,y,.f . y t u',- w / s,.,/e L 3 ,,,s. ,.~ s 'H 150 A T h w s.,,[N- ,4.-%.,#.,,#^(.,, a-1-3 (100 F) ' '/ ~._,,, ^:A .__-,_.s / 100 - n .0 D U 50' 1 h

--- --,~

,_,_;_......_ _.A/'i,. kj. #%k y.e. f ',T ij k t a-1-2 (50 F) O y F J M[ W..#b' ['\\..I Sf .!br: B-1-1 (00 ) 0-F s v o ~ ~ B-1-5 ~ igiisigsi: i iis ij. O -50= is is sisi iisi s.i i g_j_4 0 10 20 30. 40 50 60 70 ~- o B-1-3 TIME: '.'11-235]2110-1350 ~ o a-i-2 o .c - .y a_i_ t.

m. FIGt,...:. 4-4 STEAN GEHERATOR:;TEMPERATl!RE HISMATCH - 6, 12 260 ,s ,.*s. .i.

\\,

I /~'. A I l i '.,,:S b,-' M / L A b ? I 240 - ~ \\'(' 1 s n t z /- f, /:. .) I, ' 1' ) I [f \\' k. C ) f i-[.!.; k jl t M M LE B-1-6 k -, D U _k 7 J. \\.1 !!AIN STEMI p

0. ;

g / 220-I d J h (*r + 250) M I !,.J l

• f Y;'

L',-9 I,'jf 'I ~ h i. i i s U 1' w. p

covLE a-i-6

[ ,r 1 i .l IIELIUM INLET E 200 - g ( v + 250) ,q y . ~.. l 180 - i.'- e p' s iil iis l iil i i ia i iiii 0 10 ' 20 39 40 50 60 s . TINE's 11-23-77/1300 l g g h p /

'~~ c SUPE 4-5 .( e AP.RIL 0i, 1978 ~ TIME ~ ,12:(U:36 1 ? 12:30:14 ( 10 SECONDS - ^., NUCLEARblW0iEL8 ~ NUCLEAR GIANNEL 7

• r e--

NUCLEAR CIWCiEL 3 n.-- s r NUCLEAR CifANNEL 4 ',. r c. ~ ), .~ , NUCLEAR CilANNEL S ~ r .~ ' NUCL5AR Cl{ANNEL 6 ^ e g NWA CORE PRESSURE DROP P I y \\ $h N h E'8"" " "" "" ( g_2 e n .? i e, _2-J

m FIGURE 4-6 Al'RIL 2i, 1978 4. 10 SECONDS 13f t 1:36'. 13:01:05 TIME = - - -, - NUCLEAR CHANNEL 3 NUCLEAR CilANNEL 7 =- NUCLEAR Cl{AhNEL 3 t ~ / NUCLEAR CllANNEL,4 ? NUCLEAR Cl{ANNEL.5 e e ^

wMnn,

^ ~w iM\\ CORE PRESSURE DROP L i e O g g O ta g -~'~ e 4

rg,- I.., f. :., o.__ .i...., E L/E3/.h l J ..-[ ] i -Ad o /., {] ' lI. i i,.. f; e, . P: i t., 1 ,.r } 97&d iu ..,u.., t, l 1 9 ...s /.9/S 3902/c/ ?. !,-.i..i.El. ' ~l" i l s !... Y:.}.LN3W:g/JV7c!,E'I'1 5 %72.f/7).%,'f3dW3] J/.C 'WISO.{M.fEl.L\\. ' s' ., '. ;f .t i, .....,..( .l...g. .)..,,1... _ p.. s. ]. _.,.,-..1 l 3 i; i.{- 4., p } r, t

1.'.' :/ l '. f. 1 t.' C.'E

~: :E-dC, i s . E.1

i.. LLL!.

1. F.4

-i 1

1

• i

.?, j $ t lo t f -r i. ,,. = e l,s. ...T. 1,. .b I A 1 912'm,, ,' e-- d e / 2 ~-A l s

r.. -

r r /2 4-- .t p _V y.-a-~ .439/ 9 ~*-- 3

t-w

\\. t s . r. 1I ' l. n i. I ve P 3. j -e<l n,I . c...;- a.,l ; 3. i I j' -{ 8 ..l 1.. .js.; < 1, ij W,....w,i..:,.. ' u. .:3... l. .- \\: m-s ~ ,. I. ,.{

• 1

.s. g. ..{ l. i... 1 e, e p g.[i. t.,, l t;,S. :. i_j 1 n,1 j 4 1,.. 1 I,. y g

l

.-? .a i.. } .f.f. 8; ' j.,. pi. ! d. if.r. .:i-g.j j, :; g. l1 1.. .i ;.( i: i a i ...;y%.,.g, -l , l- [.- ,'..... I, '} .i.c ..f. .~ l x. .g ..e. je.... t ..t.

• g: *f

..I ..)

l... f.1 i..

i .i..

t..

i C

• , g g4

.i e y.

\\

r r 't-l'g l3*. g. i {_4.,y s I EI 1 ! ::i .'i....N.it, .. r. i. tY: P t-I4# T-3v

n....... :

t. r -

8 -

I-. -Q' - 't! e- .t ,is. h' - : ' r -. 4. f rs t 1. A, x.,. ,ji ,.v, .., 7,, . 2,W. ......,.y. .s} ..,.g. t ,.,n. ,2 .. I. i 1 g... .- e j. 1 'i.- t. fr.

l. l

...,j g.. ,..'. 1 [,. 8 ! l'r, i 'g. (- i-e' s L ., 'i j.

.; g:

} '.. f-g 1.. g ;, ;,', l

.)

p n. I e j. - ..i i., 1 ..i.-...,, r. t .\\ f. , -c.,..l.y I' c p y. p [ b, i :e

. ; 3...:r '

4 4.....,.c

i.. i.,

p, ;: 4;.3. 1 t, .,.. I j t. . a.. ) ; . )f 1. - ,iti 3: e I., t j ii ..g =. 82.

J

..---{P;.'.) li. . 1 (.r ,j ' ;. j ;; 1. { />.. c.p 'g.. c.~ i.: 3

.si;#

4..., .i,.

4... '

,. r !

t t , 1. J. p.. s r. 1 ,. g. I i,g i e.. . l.7..!,.... g ..,.3,* . 'l s l.} ;- :., i i i-y g 1 ..8 g.,,. i g., .?

I

.. w g,.. ' s ..t i ..l. 6, s -i'; lr. p.; .k2 4 (h .. g ..,k'iljo m;. -- j.., l i I.,I I4 %., }

C*

l. .( ,.4.g. ..,{ i .r i,... 1.4 : ..i.. t ;.:. .:... 3 w qy

i.

., i.e: ;. r r.. ..,. :o. a # .q '. 3. 6!y L

i ;--:,e m

... ;.,' w t.. wt,. p e. g.

u. _.....,. i...c..

u...

L.. .?.;....- :..:,._ ,... :....;_;.y :5.... c. 1. E. I... ;fa 2.._*j...l

,..}.

t,' _.

f,.j i.s i..,.

.x p g . f_..., j.... .i k$ ',; ;' d d.-- N g

[- f j.

85 f I d c. d.[.l.;.jk.i,. J.* - l. [* 5$ .,.'e g g y s s.., g.. g - ..} Ijh....q

e. l.. -

s.p i i. t. -; t-h.4,,F. pp.3 " ...., g ;,.. i e

i." } I,

3. - ;

i. ;-- 8.. b--. ;,,..

r,- e. .-( l.l...: ..

• l. { ' l'..

i.'

t : --! }-- 1. .I t l L.: U. i, e. g ;.;;,J -( ..,... f. :l't..:p;r. l. 3,., !. 1 as. 4

i

. q f-r i

I,- ;N y .'.,., : 5 ' g i,. f, 1 .y *,p. 19.M

},...b t/

.l:. ,,,4 4 .. i i i. 5, i; t. j t .i ,6, i f,. i J.- s' s : F ( r,) ' r ?. I. e .ji. {i.g,-:...g ..'., q., {sp'i :l',c 1. ',p.- ..'l ,1.g : g.

0. 4. <

,.s 'l l t. . r !-- - ,i .,;.. j...- .a. .. i g, .41 ,' : s i 4...:i , e:, h.. ,.f; i r.'.v.. -.. j .c e L;

-l I _

4,,,e,.... ;

y ;d

} s,. 1.; ,- 3 '-.y s. u. g 1 ,.a. . e,.s... 7,--{ 1 .on. or > r. e i p , p. I, [ ;,,, .. ) : - },.,,.. 3 1, e - } s nas .. ]

• , r y,.;.;3...i.1,. j,.

,a u.. g. y q,. .I. i I .h, .- 1,. rj j 1.

8. l O. r.

{4 l7 .,,,.*. [ g;,,

-.e v.v.

i f '- ji: f^ l-f,; {! . : : -l' I I ,.,.,.........y.

,i ;

3 : .. j [,

1.....

\\, l.. 1 .t .,. p,.

• f sI.

.,.,s ... i t. ; . 'g' t. .4 y,... ;r ll t 4 g s. J l 6. l.6' 8.' i < U.)

t

!. l . J, s......t 4.. _... r /.;,. a. ,.f. 3.. j.- t, .I p.. t, ;& g i. l. ) ,} i. 3.- 3, .t .t ;.% 6(Ql

• s

.a. at I c.F..- ll...... 8 - 4..- g, i. c. j' f . [ :i .L. I,. 7J e .6

1...

j ;., *in... 2.g. .,..,g e .- ? - g. 8 -., j !.

n.. ; i. g,. t1,p r,. 4 g

u . g e, q... t r... y. t t. . i m,. .l. f)3 j.'. .g 1.,. I.* .e j.y 1 !....y

• 1l

,? { t. p ... t.,' 3 i.; a .,q ( _.- (. - i. a he -.I,i f gd;..,g.}}.!; r,. l..a.;j :t. ,t .i.

. :.: t.

l.. .. i..

. 1:-

i _ i .1 I I 3,2 u. ) , 'j '.,.,,.i e m

7. c,.

.g. .8 I 'pfg,. ,I.t 1-,. {. [. o l4 ff c e, g -}-- } ~

9.,.:;% e } = c {g. - \\ %p YyN ) 1 E - 7,st -i :%n. . v' t o e V 1 a f {... f .g i 'e e i . i..... 5* I ( . = :....a' t

• = *.[. * *i

. i) . *..... t. I

j...

t S. ..~? 1.

i..

ig t 4 i f g.:. ;... .l. .j io l s g. 8 2 ... '

• Y.

a p 8 ] r I i $........, t,P g [ N l l I. W t.

• ' ' * [. g i

= 1 [ {,

• 2*

( {- *, l 1 [. Ed }...:..... i. ce d'""""'" g j. L.... l,. (

--I z

t-f==

• * =

i, h# I.. 9 ; .a 9..;.* .I f a

• b
• E.

1 p +j 4 2 jj I-l. / .. i .g / i.. L... ..i. I .i 8. .l .l

7..t.

{. .. t. 8 9 ,5.. g, e 3 s l l 1,. ^ g-a.. j e .t -l k

• l tt 1

j.. 8 a h>. >* i l I i +.e l1 i.

2...

,-( I: {. p. l* h. .i g .e l .i. \\ 7, s r .4 \\I l. l t Q ^

n s. 9-1 6.rciaar/or man.c7Easr/c ccMmarcs f/.23/76 //30 ~ /3/.2 NMS 8-/-9 //E. /A//E7-7*Etff 8-/-3~ k t ~. F 'Q( -( .. ~ g

== e e e B-/-f g - V / l ]n q fu e X T ~ " - ~ srq ( ( (

$a 1 l f. ~P_

• ~n

<g

e. -

f ,a

1. f

~ - n-

c. n, O-P.,%

+* + n, + + + + + + + ~ P* M p f4 r 5 _rj

0. 0 O-

+ + + + + + + + + -92 On u E = ,, o e-OO M D=.>=>= + *4 } O-e 7 3 .Z, w ee4 c:: + +.c 4+ + + + + + ouo n-o.

=

www On h & LL ed wj .r ^d / E o .a s Ow c a: = e,w 5== r, + + + + + + '+ -Q k mg LL. a 5aa K uu 0 ,, y 2 s2 n, bAA Ow w MM O w aae C-e, www .=

+

-MM M E EE + + .+ -+ ~+ + .a a u >- ;e e -,o. w e-.a m. 94 m' r @ P% f o-. w UFE I 1 i: e- + + -8: S + + + + + ~+ ,;A % i -,m tJ 00O QOO g) W;i / e, g; ne4 L-e- s

0.,0 e,

O- + + + ~+ - + ~+ + - + tJ M l J tJ o PJ s

0. 0 O-g

' + + + + + -g + + + g py h q , td h r., p Ft Ow + -. ( O= ,p _ L. 't._ _1,_ t._ 1 L. f._ .1 1 .s e -e

== ,n O O O O O O O O O n O g4 + + + + + + + + + +

• O O

O O O O O O O O O O O O O O O O O O O O .t.t O. m. O. o. O. e. O. e. Oe E e to a m w w n et to t p'.'-- l = ! cia,. ;. of t w o. - - 4,;,. ,,.7..,

.,.7.oco20jl[ ~ ~ t .') f 'l 1 - \\ J 6.500 01-i. .{. .}. .t. + + -t- -l- ) 6 003 01- ). k

d

.+. + + + +

5. son.01

+ + 1 f, 5 000 01 7 +- + + + + .t. + + .1 i f 4 a 4 500 01 r + + + + + ,+ + .t. 1 a t 4.000 01 + + .}. .f. .t-1- -t- -t-1- 1 1 1 ~ ' ~ h ~ PERCENT. 1- .t. .g. .g. g 006 NIN1136 n.INEAR PWR CHRW 6 e 007 NIM 1137 1.INEAR PWR CHRN 7 PERCENT

A 003 WIN!!'30 LINEAR PWR CHRM S PERCENT t

3 030'08- [ + + + + 1 2.503+01 .g. .g. .g. .g. .g. ' 4 .g. 4 S 1g* = 2. . h -2.163 03 I.44C+03 2.830+03 3.600+03 4.300+03 5.040+03 5.760+03 5.450+S3 7 203 03 71239} 23/11:14s2 23/11:28:2 23/13:39:2 23/11:L0:2 23/12:02:2 23/12:14 2 23/12:26:2 23/12:30:2 23/12stGi2 23 r 13.C2 e! i FIGURE 4 STEADY STATE VS. FLUCTUATION COMPARISON

e. OfJ

• t3 n;;

~, y.c y 7 -_. r,, j g. ~. -1 n'.t

o..c, n

2 -U + + - P" +* + + + ._ r 4r f I } n. C.m O -h + + + + + + + + .n u f n-c.as e swwww n-d 9 66.,666 + + + + + + w wmm ~.- QQOOD U] g a eAALL EL L L I. 6 e w.a a'. en-fJ p t= r= s= 8-g R = ooooa C C C C= H OOO w CJ L + + + + + + +

o. s

= m- -ee 23 m3 = mn L:- ooOoo u w,ww auau. v, C3 E9 DD N v a u u .C,' eg n. ( ., m. o o~ h 39 - een 55555 + + + + + + ' + -A" v u 7 7.. e7 7. 7 @ E g t. ee.eee F u e. nce &~2 + + + + + + + + - 8. s 4

f. f E S. 3 an el c

we L-a t r-uogx+ o t A - 6 n 3 om s h c'3C + + + + + + j; + + + - E.: s Nf"o 5 t i s ~ i n_ q OQ s a" n + + + + + '+ + + e of NM s CJ '4 t 4 ja e ed.

a.,

C- . +.... f. _ t. t._ f._ L.

f. a

' ^ - t f-M e M m m e m m m m e G G G. O. O. O. O. O. O g4 n e h O O O O O O O O O

1. 4 2

A. n. n. 8 5

8. -~

Q.m -__ .I. .6 .u. p. .u. .uu. '.."c,n"--

.7c5 - 1 0 ,e m t 1 c-re. to 00, J (3 - +. + + + + +' + + - P,2 r~ 'd m.a G.m O u- + + + + + + + + - C.U ~ en fJ g 8 94 n-ww O. ~te c: O SS + + + + + ,+ - E.U E OO s c mg v. =0 c a H m= ww ,, si m e GG % rL. =1 oo O t;; + + + + + + + - q" si Oo .nn u.. to ww

== my 9( oo L) LI ,4 m-wr t4 W( OO gr. n 4 c:-h z2 O~ s' 22 + + + + +

+

-42 m s ,n tJ ON" e I s se r-w c5 FS H m. ,J n-wW I -i O. O, O-m + + + + + + + -8.s: 7 o-e nn e. w ~ e i c: ao 8 E FJ M n-u 4) O C.m O- + + + + + + '+ - E.:s + NM / tJ 9 94 ~- C. 0 N 0- + + + + + d- + +- + -) .s ~m + a N .,J.. e m. O *t e em O= L .L 1 _L _L 1 1 _L 9 + _ i ** 4 a 4 a e a a a a m m m m m. m m m m m n m O O O O O. O. O. O O O O. 94 e e + + + .O O O te O O O O O O O O tc eJ ets

• e O

to tri w O N, N O e4 = O. O. m. m. eJ FJ.

• y po-su.

.s. .m. 5 .uk .=. es. ene p. %W T -. - - - -... - -., ~., y; - 79l.. ,75. ,ic,9; s~, a; C'.3 ',,.T. i '. :"7 f *.,- '3 C.,[- ~ ~. ' 7;. k 4 ?%

• ~

m.- -

l.. 3 9a-u o. e. ]* ',m 2c gr o c-no () o.2 + +( + + + + + -y s n M2 ng 9; a n., + d + ~+ + + + -n; T -- h a ,s 9a, e + + ~ + + + -F - 20 e 4 h79955 3 4^' 202000 ng E m 000000 4 E .s

---

~~

o, 3 + + + + + - 0!; E s 8 as p g d 3 ^ O s i N N COCCCC 9 P 4 h .s..s.s _s.s_.s. n8 :. y g og u C: CR + + +- t wwwwww + + - 20 t. p n 2. T7?TVT "A = u 3 p.. TTT TT e 3

==

s. I.' ' 5 82R228 n8 3 .2

==

a? - m m, - - o, m ) + >d + + K g "' M " ; + + -8; p d } E9 fy i; ee<xow u y,. g , l1-29 + DD an d 6 + + + 8 + + + + + -i~ @@9

Q

~ D 7 \\ c. D o e gg s a.T =a. d w-e + + + + + -F + - 8._2. e _s 2 = Q,4 -y,N~. at +' + + i + 22 + + a 2 2 2 2 2 2 2 2 2

JR A

a a .a -a a a a a a 2 M E. 0 a 4 2 S S 2 2 e

t 4 .5.0 -- / / 4.0 - / FLUCTUATIONS HAVE S OCCURRED IN THIS AY TOTE 3.0 - u-8. OPERATION WITH NO FLUCTUATIONS 2.0 -- k 57 DAYS 116 L DAYS 1.0 63 i DAYS 35 29 DAYS 17 DAYS DAY,S 0 i 4 ( 30 40 50 60 70 80 % POWER FIGURE 4-15

l S. SAFETY CONSIDERATIONS 5.1 Summary Since first being encountered on October 31, 1977, fluctua-- tions have been initiated several times in a continuing effort to under-stand their cause. The power levels at which fluctuations were initiated have ranged from 40% to 70%. A total of about 40 hours has been spent in a fluctuating mode, which is equivalent to 240 cycles given a period of about 10 minutes. From first observation, the potential safety consequences ~ of the core fluctuations were evaluated; and it was concluded that there was a small probability of damage to fuel and reflector elements in the core and that the risk of an incident endangering the safety of the public was not increased. This conclusion was reached based on the mag- ' nitudes of the fluctuations, both global and local, and on an evaluation of what is believed to be the cause of the fluctuations: motion of core ( components. These considerations are discussed in more detail below. Various tests have proved the character and magnitudes of fluctuation to be predictable and repeatable except for minor differences. The core operation conditions which initiate the fluctuations have been well defined, and it has been demonstrated that reducing power is a reliable means of stopping the fluctuations. An' inspection of the top plenum in December 1977 after fluctuation testing at power levels ranging between 53% and 68% revealed no indication of damage and nothing amiss. An inspection at that time of the control rods in region 34, which were inserted throughout the entire period of fluctuation testing, also showed no signs of excessive temperatures or impact. Finally, fluctuations have not resulted in an increase in circulating activity, which shows that no serious damage to the active core has occurred.- The secondary system is being driven by changes in the primary system and is affected only slightly by average parameter feedbacks through ( the control system. j e ~.

~ a 5.2 Fluctuation Detection -( A range of reactor operating conditions has been identified where fluctuations do not occur. These conditions involve operation with core pressure drop below the curve shown in Figure 4-15, and the reactor is normally operated in this mode. Fluctuations for diagnostic purposes have typically been induced by operator action and according to a test plan. Various instrumentation is available to the operator to diagnose fluctuation behavier. The onset of fluctuations is typically detected by changes in signals on a number of displays in the control room of continuously operating normal plant instruments. Instruments available in the control room that are capable of detecting fluctuations are shown in Table 5-1. Nuclear channels, flux controller, and reg rod motion are among the first ~ to register fluctuations. Main steam temperatures are also used. Usually measured steam temperatures exiting each bundle are within a few degrees of each other, but at the onset of fluctuations the differ-( ence between individual modules increases. Additional instrumentation listed in Table 5-2, not part of the normal plant instrumentation, display signals indicative of the onset of fluctuations useful to the operator. In the control room, the trend recorder on the data logger can display four data logger inputs on a strip chart: typically nuclear channel VI, reg rod position, and outlet temperatures from regions 34 and 36 are used. This instrument has proved to be quite useful in the past because fluctuations are obvious and the display is easily accessible to the operator. Reg rod position and reactivity both show increased activity at the onset of fluctuations and are displayed in the control room on the reactivity computer strip chart. I Outside the control room, not readily available to the operator, steam generator module helium inlet temperature can be seen on the steam l generator data acquisition system in the auxiliary control room and nuclear ! (, channels can be displayed on the CRT at FM station 1. l l

l ~ ~~ i, Table S-1 FLUCTUATION DETECTION ON NOIDIAL PLANT INSTRUMENTATION (DISPLAYS IN CONTROL ROOM) 1. Main steam and reheat steam strip charts e Onset of fluctuations eyident from increasing variations 2. Reg rod motion Fluctuations cause increase in frequency of reg rod motion e (noticeable to operator - definite click) 3. Main steam temperature alarm Alarms if main steam temperature deviation (module to loop ( e average) exceeds 20*F 4. I-04 board region temperatures e CRT display of three selected regions. Region temperatures show fluctuations. S. Flux controller l Fluctuations cause small indicated average nuclear power changes, e noticeable in flux controller (.

,( Table 5-2 - F111C111AT (N DETECION ON ADDITIONAL INDICA'"3RS A. Located in control room 1. Trend recorder - strip chart Display of any four data logger inputs. Typically have e. channel VI, regions 34 and 36, and reg rod position. These variables show fluctuations clearly. 2. Reactivity computer - strip chart Display of reg rod and raactivity, both of which show in-e creased activity with f1tetuations i B. Outside control room 1. Steam generator data acquisition system CRT display on demand, in auxiliary control room, of helium e and steam temperatures 2. FM station 1 CRT display of nuclear channels, in reactor building e (

Combining knowledge of operating conditions where fluctuations are expected with this diverse group of instrumentation, no difficulty has been experienced in promptly identifying the onset of fluctuations. 5.3 Safety Evaluation As discussed in preceding sections, the data obtained during core fluctuations can only be consistently explained by presuming move-ment of components within the core barrel with resultant opening and closing of eme bypaw flow paths. , " ~ i Nuclear channel signals indicative of neutron streami'ng and [,3 exponential shaped changes in reflector gap flow exit terhperatures both 3 (( strongly suggest the movement of the permanent ~ side reflector blocks ~ during fluctuation periods. While a numbeF of nuclear channels exhibit the abrupt changes in signals, the larger magnitude and more frequent occurrence of such jumps in channe1 VI strongly suggest the movement is ~ more prominent in the northwest portion of the core. 1 While it is most probable that core permanent side reflector movement is essential to the observed data, refueling region motion or fuel element column motion may also be occurring and are therefore con-sidered in this safety evaluation. Fuel element motion can best be characterized by motion of regions rather than as individual fuel elements 4 or columns of fuel elements. Fuel elements iti a region' are constrained ~ to move together because (1) their columns rest on a common large support. block, (2) they 'are~ keyed togeter 'at' the top by ke'ys'iiithe plenui elements, (3) the pressure outside a regica tends to be higher than that inside the region resulting in a clamping effect on the columns.' Minute r aement of PCRV (millionth,s_of_an' inch) is sensed by the displacement probes mounted on the surface of the vessel. It is possible that some energy is transmitted from moving parts inside the core through the core barrel to the PCRV liner, resulting in the obser':d <:. u -.. - ~.a _reas.e). Avail-f increase in normal PCRV movement ( w.0002 to .0004' inch inc ~ v l 7 {' able masses and gaps (movements) within the core cavity have been :tnaly:ed { and indicate that most probably permanent side reflector columns and t possibly refueling regions are moving. 1 ~. -~ .. - =. - -. -

o c Quantitative evaluation of fuel temperatures, reactivity (_ effects, possible velocities of gotion, potential impact loads on the y, core and core support structure, control rod insertability during fluc-tuations, and effects on the secondary system are discussed' bel'ow. 5.3.1 Fuel Temperatures 5.3.1.1 Changing Bypass Flow Only small changes in the total. flow bypassing the active core are required to explain the observed effects in the temperature fluctuations which can be accomplished by redistribution of the bypass gaps and bypass flow. The o'bserved abrupt changes in the steam generator [ inlet helium temperature cannot be the result of region coolant flow .l and are thus attributed to sudden changes in the fraction of cool gap flow entering the steam generator module. Therefore, no significant global temperature effects of the fluctuations are expected.

However, the varying bypass gap flow distribution in the core will cause the fuel and graphite temperature to change locally near the changing gaps.

5.3.1.2 Quantity of Fluctuating Bypass Flow The helium flow at the steam generator inlet is a mixture of helium flow streams coming from the various regions and bypass flow gaps around the regions and reficctors. The temperature of the combined / flow is a flow weighted average of all the temperatures of various streams entering into the steam generator inlet. Figure 5-1 illustrates graphically the combining of two streams, i.e., the region and bypass exit streams which mix together in the bottom core plenum before entering the steam generators. The steam generator helium inlet temperature is made to fluctuate as observed by changing the fraction of the cooler bypass flow' entering the steam generator. l The amplitude of the steam generator inlet temperature fluctuation depends on the fractional changes in the bypass flow and the differences between l ( the exit temperatures of flow streams from regions and bypass gaps. The magnitude of abrupt changes is of the order of 20*F in the observed

. c . N f, c. w . A l.- l .I i l p . y TitoCou6tl Hixin6 n.s. macte. t .c v).... ... O o 3 p.- 2 . k,. _ yy Nc T1Ht Vw,1 %s.d s. M} v~,... y. y j .... (. g _~. .n wy T. .... 9 r g.yti n,g> e ,g ,.. s}. _.. Ct).OFF. \\..... r,.,.,..,.<> _. s

s.........

w. ,..... f ..i .... o.. 00 ... U. tR u...m... =_.

t....

.a.. .... -...i \\.. 3 t -7.. >....I s. ....... 1........-. QU V v ......i_.................._.._... _.... _...,... ,.s. ...........i......, i. s.. _.._.... _.. ..i_...

g....

. s........ ...~ _ _........ co. .N..........N........... , (. ..._._._.__.._s..,...1_._. d .__.y_...._____.. _. _.. ... s,. .. s ,... ~.. Q s i.,.... %... __.gp ......N.. . J,.... _ :-_ _:_.... s 3o. t p, _. t w g4.,.r. s ,.w-2_o r ........i...._._... .....e....__. ...e..._....... .I_.._........... ......,i. ....to. r e \\ o to g_ o ..- So 4o . _... _.............. i..........., \\,e u ./7, T t o . A, p. c c ,,.,i, ..i.. _.. _.. _.. _ _.. _. _.. _...i._....................i..... ... i. i....._..... ....i...... .I. i ..,..t....... I. .._ _. _.s_._._.__ _.,.... __ _ _....,... . e. Ct t. .... _... _ _..... tru.. ssA.... t..M s r.3.=. it._e_.. ..r................ ._s, 4.. t . _. y.g. W. . +,.... w. 5 w e..I ).. ( y sr_s.h %.. b..e.. . s t = _ " - ... ~..... _....._ _ %..,., k. ........_.._ _ _.. T. R.g

• s e....._.._..__._.g a

ure.

• .... _.... y.. 4_ta,,, s om. car.e.

ss gy u b.. 3m.; lagkvaw.q

qt-

• (

n: th. %lo oJ, S*.. ei.h le *5.W*

• :. :. M s

9 =o hw g sw.%t oh- ~ <. ..........'T'. =4... E."ft...,.,.

.t i q.T' e,m.; M* is(

• set.e y ; J-}q.*J *l E g

. g .a.... s ,.w, on ...i..u M. r4 3 %.(. if,,,. . 3 ....g. ..t .g s .i .I .. i... '1

steam generator inlet temperatures during fluctuations. The temperature of bypass gap flow near the exit was expected to be lower than the region , exit helium temperature by 200*F, as shown in Table 3.6-1 of the FSV Safety Analysis Report. Traversing thermocouple measurements have been made through the side reflector at 30% and 50% power which indicate that the helium exiting the side reflector is only slightly hotter than core inlet temperature, resulting in a 500*F difference. The required fractional change in the bypass flow, therefore, is less than 5% of the steam generator-module flow as given by Figure 5-1. The required 5% change in gap flow is consistent with the 5% steady-state value shown in Table 3.6-1 of the FSAR. 5.3.1.3 Bypass Flow Changes during Fluctuations The core pressure drop data taken during the fluctustions show that the core ficw resistance always decreases at the start of the core fluctuations by an amount varying from 1% to 5% of the ini-tial value. This reduction in core resistance may be explained by rearrangement of bypass gaps in the core cavity where larger gaps open up in other parts of the core. The bypass gap flow could increase by 2.5% (of total core flow) when the core pressure drop decreases by 5% ~ due to rearrangement of bypass gaps. The steady-state effect of increasing the bypass flow'by~ 2.5% of _ total cor,e flow on the coolant and fuel temperatures is shcE~ ~ n in Table 5-3, where changes in"si>me of'the import'aitt parameterTMin' ~ .n :. - -..-- __..,n Table 3.6-1 of the FSAR are shown. The steady ~ state increase in the ximum fuel temperature is estimated to be 25'F. If it is assumed tnat the total time of fluctuation testing is 100 hours, then increasing the fuel temperature by 25'F over and above the maximum fuel temperature i of 2300*F results in no fuel damage and no additional fission product ( release but enly a negligible additional kernel migration distance and therefore does not have any safety implications. If the transient effects are considered, the increase in fuel temperature will be smaller than 25'F due to the thermal inertia of the core. l ~.. ~ ~ i. w

c ~ ~ s 5.3.1.4 Local Temperature Effects of Bypass Flow Redistribution I Although changes in the total core bypass flow brought about by fluctuations.are small, local gap flow changes along the outside surface of fuel elements can be relatively large. The influence that gap flow fluctuations will have on the local peak fuel and graphite tem-perature in a fuel column depends mainly on the power profile across the fuel column. Steepest power profiles across a fuel column occur in a rodded region and in a buffer fuel column at the active core peri-phery. Figures 5-2 and 5-3 show fuel temperature distribution across a buffer fuel column and a standard column in a rodded region. Plots are made for the axial mid-plane of the core. The upper lir.it effect of replacing gap flow by an adiabatic boundary is seen in these figures. 'Mie peak fuel temperature, however, is influenced by the external gap flow only in cases where the power profile is higher near the surface cooled by the external gap flow' f The peak fuel temperature in a buffer block occurs near the buffer zone interface and hence is essentially insensitive to gap flow. However, in a standard block of a rodded region, the power peaks near the periphery of a region which is cooled partly by the external gap flow. In this case, the peak fuel centerline temperature could be affected by as much as 100*F. 5.3.1.5 'Velociti'es] Velocities of movement of core components are required to establish the consequences of motion on structural integrity. Two approaches have been used to determine the magnitude of core motion during core fluctuations: a single-impact acdel where measured PCRV displacement is reproduced and an approach involving lateral pressure forces to move columns. Results from these methods are summarized in Table 5-4 and indicate that the. fuel elements are experiencing low ( velocity during core fluctuations. e n

c +e. si + !l / f i e I .'.l. I t l y 5 ,I o. t yt 5 I 1 G:1 r s.s p. g I 6 p. = p .Q. . d eg f. ,1 [ .l f f. .j..... R ?.. G..

• ....e e..

.....s-p. ).. .6 i/ ,e g

Q

... =.... c< [ la u f U.* .......,......t p. 9...,... j ,,..,,.,...,,.. l... -..... e w . f* l 1 2 3 i.,. 6 .c t I l! /- f'!. *....

• <c.

.L .... s. l . al.. .,g., =- l. ,y ...j.. ...=..s. W

• +

vs '. [- 3.. e.t l

• • a O

v .g z A- . g ..i U 's r-. s .6-s: c t' g. ,p

==* f.%+ = g

t... =......

--...'.....t.. i ,.g ,.s... 1 .',s s q ( i P ~ .g s ......r . c >l j \\_ .m l s 3 u. .:: r-t a /n / L n y a o / l lll'. O * / 39 ~ >- n) / .z,. ) . O.. S. n o-o O o - Q* {. r O Q o o ,.s N o. gp 4 t ( g .), e v.'y 'I fII 1,4 3 !.N. y ~):1n J

• 1-

.o y 5 g

O-e 9 ] t o g *; 6 ~ ) g < '.e* '** /c) U v,J

• A 3

A a 4...

v.,

• • 4-

/' s t

• ~

J .J,.. \\ / / / t N. .a_s g .e .s, C 'i-f 1 a t 13 [ <* *Z ts ., o. 3 = a. .,..s m A. Y L <a s e / <E d y / 4.8 g t-- ,a. 4 8 . l........ .....".. a. _.,... <;~. / g .s t c . v J. Q t .1 -4 ,..;.~ ~ ,g,,.*,, / q, 4 2 [ g .a [.. ' :'...e..'....... - .a p. y [* y

s*

.9 =..

• ', ~.

...,... ~,

g.. "

A !.s 4 g Ad ,-'-r-* l 'l.*.. .g 2 .%g ..{ ( .s /.,.. a q F = .i. *' ~*- f l. 3 p %a J .n (. C h. Y, *, *.

r *. ****:

A .I ...."t*****. g. 3 .3 G s O' ,a g<. q. n u I ~ s. .. ~.. d d

r...

s* c-r- E w. .e t-

a., -

.~. " * = = -+- <. a Vy .s / m. s t Q O A u o O W l I - ] ( o o O a o o n. O o O O o 4 O gp WO eJ . { g 1 g f. 3, c./ w q yn nu rr.r) 19 0..}

~ m Table 5-3 REACTOR COOLANT PARAMETERS 9 100% POWER Reduced Region Flow FSV FSAR Core Avg. Flow at Core Avg. Flow at Exit Temp. Core Exit Exit Temp. Corc Exit- '(*F) (1b/hr) (*F) (1b/hr) 6 Coolant holes 1516 2.88x106 1496 2.96x10 Side reflector 886 0.20x106 970 0.12x106 Control rods 930 0.14x106 930 0.14x106 Space between sides of fuel 1260 0.17x106 1260 0.17x106 elements Total active core and bypasses 1444 3.39x106 1444 3.39x106 Maximum Fuel Temperature at' 100t, power level 2325*FA O O e f

^ 4 Table 5-4 VELOCITIES OF COLLISION EXPERIENCED BY CORE REGIONS Method Maximum Velocity (in/sec) Single-collision-momentum conservation 5 Lateral pressure gradients dynamic model <3 ' l (

C

n I

During core' fluctuations, displacement probe 2 was obscrved T .. building-up in magnitude from a nominal-background level-of '200' micro- ~ inches -(.0002 inches) to approximately 600 microinches. Since'the PCRV weighs 31 x 106 pounds, any displacement seemed to imply a large' amount. ~ m ,of energy being transferred to the PCRV'. However, the amount of energy s '. transfer required to increase the amplitude of the PCRV movement in the above magnitude is an incredibly small 1.5 in-lb. This simply means that 'we are using.very. sensitive measurements on the,outside of the vessel

which' measure a low energy ph'enomenon to perceive motion inside.the vessel.

.5.3.1.5.1 PCRV Excitation b^y'a Single Impact One of the most straightforward ways to estimate the core motion necessary to cause the indicated PCRV motion is to picture the . energy transfer through a single elastic collision. From such a cal-culation with conservation of-momentum, it was determined that a mass on the order of one fuel region'or one side reflector column moving at a modest velocity of about 5 inch /second would be required. Several re-- gions moving together would require even lower velocities. x It should be noted that in.the displacement probe trace the amplitude builds up to the peak value and then decays away. This implies .that there must be-more than one impact between a region and the core barrel. When multiple hits are involved, the velocity of the hits would be lower than that calculated for a single hit. Therefore, the velocity of 5 inch /second represents an upper limit value. 5.3.1.5.2 Lateral ~ Pressure Gradient in the Core Model l i One mechanism for moving core components are from radial pres-f sure' differences around regions or side reflector blocks which result in l a net force on the component.. Pressure gradients across core regions may exist'due to differences in orifice. valve. settings and due to non- !_( uniform' inter-region gap widths.. Calculations have shown that flow communication between inter-region gaps and intra-region gaps is poor 'so that;different orifice settings create only very small pressure differences in the inter-region gaps.

T, The model used assumes one dimensional motion of a plane through the reactor core. Figure 5-4 shows the basic elements which we*.e included in the model. Figure 5-5 shows the mathematical ideali-zation constructed to simulate the behavior. The model includes the PCRV mass and support skirt stiffness. The weight of the unsprung fuel, reflectors, core barrel, and core support floor were lumped into mass point. The core barrel / floor was sprung from the PCRV mass by a member with the stiffness of the core barrel keys and the core support columns. Several fuel regions and two reflector columns, representing a diametral slice through the core cavity,,are supported by graphite support columns which are designed in such a manner as to develop a spring stiffness of approximately 1500 lbs/in. in the lateral direction. Gaps of 0.2 inches are included between the core barre.1 and reflector columns. All other gaps are nominally 0.28 inch. The flow is calculated down each of the gap channels assuming a constant core pressure drop. The flow model includes exit losses, bend losses, friction losses, and the effect of heating of the coolant as it travels through the gaps. The model represents the spaces between regions as separate and unconnected. This ignores the continuous connec-tions throughout the core gap structure and tends to overestimate gap pressure differences. To approximately account for this crossflow between regions, the width of the channel was limited to one foot when calculating region forces calculated pressure differences. It is felt that this assumption will overestimate the impact velocity between regions. The PCRV motion calculated by the code for a cere pressure drop of 3 psi, which approximates the displacement probe 2 trace with respect to amplitude, is shown in Figure 5-6. Impact velocities of individual regions corresponding to the indicated PCRV motion of 0.0004 inch were calculated by this model to be less than 3 inches /second. 5.3.1.5.3 Nuclear Detector Signal During fluctuations, rapid changes in the measured nuclear I. power have been observed. Since these changes were detected in most cases on only one or two of the six detectors and because no temperature s

s A REGION GAPS. / FUEL / REGION / / COREN 7-+ N BARREL NJA f N P C.R V _ o 3 ~29,3 x 10' LBS O CORE O BARREL 6 M - RE FLECTORS

~l 6

4 ~ I "~ CORE SUPPORT FLOOR . CORE 2.. x.1 o' LBS..._. ,. CORE SUPPORT i . SUPPORT /' . COLUMN S ~ POST / .ai. s 3, i SUP PORT-~ - f RING 6 K= 2.0.2 x 10 ggj, y ELEMENTS OF CORE MOTION MODEL i

l, 8 t, c 3 g PTOP y 3' O Q O t t d a y v y y v v v v v / / / / / / / lh ih lb ib lb db d l-db db db / l R3 R9 R5 RG R7 R8 R7 Rio Rll Ri?. s u a / O F 2 Z o v o O 1-v uJ .J - -I W a wG w a u. ,w a u. W CM aN W / U K / 'l H3-H} H} H} H} H} W} Hl-H]- 's yw en en en n n Hw es-Fw / . P bot h CORE SUPPORT l . FLOOR ( CORE. SARREL n u R2. g -{h GAP h/ITH STIFF .5 P RING -i-DASH POT ~' l PCRV SPRING O -se n / M. R = SPRlWG NUMSER D FSV CORE MOTlON MODEL mA, n .n

^

I f..

COREMODEL3PSIPR5SSUREDROP5%FUELDAMPINGNOMINALGAPS 5!" i. l U i f!L g 2 9 q, g I M }. lit 1[ ji b i ] ]~I[j l I' l, lt I ,' I,, Y - , 't 'i '*i 'l. e g lgl. l'l i.f' I;;il fr gi..' .g 'l-1 f I I l -i i. ?.g ; j,i = .'I' i o J'L f.',) t '1 .~, l f~ ff{}~ { I!' d I ~[~~ ~ !;,1 I m j; Ier M -2 l N [ -3 ~4 ,y 0-0 16 24 32 40 48 56 64 72 88

• • • 1C-I SECONDS ms1 south 29 JUN 73 e7:53803 Fig. 5-6 3

k a

feedback was seen following jumps in channel VI and other channels, it was concluded that these changes were the result of movements of the I permanent side reficctors creating neutron streaming paths. Neutron streaming calculations show that a 1/4" gap between permanent side reflectors is sufficient to cause a measured flux change of 15%, which is about the maximum change detected in channel VI. From the expanded-time traces of nuclear channel data recorded in the FM systems, the time for this change to occur is shown to be between 1/4 second and 1 second. Based on these values, the velocity at which the streaming path crack opens is calculated to be a maximum of 1 inch /second. This tends to confirm the above derived values of 3 to 5 inches /second. 5.3.1.6 Fuel Element Impact Loadings Use of spring mass relations allows conversion of these impact velocities to the impact forces necessary to show the ability of the fuel elements to remain undamaged during core fluctuations. Top edge and side edge loads may be determined gongidering the rotational moments of inertia of the fuel element and stiffness in the respective impact orientations. Fuel element stiffness were previously determined for LRTGR fuel elements in Referenct 5-1, and factors for determining non-flut impact loads were developed in Reference 5-2. Using these factors and an impact velocity of 3 inches /second, loads on individual elements are determined to be: Flatface{ 2,100 lbf' Top and bottom edge 1,100 lbf Side edge 400 lbf' Dowel loads ~165 lbf' 5.3.2 Evaluation of Effects of Fluctuation In this[section, quantitative information developen in the previous' sections 6n danseq6cnces ofl fluctuations will be" examined for' ~ safety significance. ~

a ~ 5.3.2.1-Accident Analyses-( FSV accident analyses- (Chapter 14) were reviewed to deter-mine if any reevaluation was required if the plant is operated under conditions of primary system temperature fluctuations. It was deter-I mined that the localized initial conditions created by the fluctuations would not effect the accident consequences since the accidents are ini-tiated at Technical Specification limits and operation with fluctuations ( k is within those limits. Additionally, the fluctuations cease immediately on a large power reduction such as occurs on most FSV postulated accident scenarios. Safety systems such as.the plant protective systems, the primary cociant system, the secondary coolant system, the reserve shut-down system, and the liner cooling system which may be required to effect a safe shutdown are neither associated with nor impaired by the core outlet temperature fluctuation phenomena. ( It was concluded that no reevaluation of the FSAR accident analyses are required as a result of the observed temperature fluctua-2 tions. 5.3.2.2 Fuel Temperature I [ Expected fuel temperatures during fluctuations were discussed l in Sections 5.3.1.3 and 5.3.1.4. The effect of incie'a~ sing the fuel temperature by 100*F for ~ ~ l 100 hours over and above~the maximum' fuel temperature of 2300*F "r'esults i~' ~ in an additional kernel migration distance of only 0.8 microns. This' additional kernel migration distance is small compared-to the allowable distance of 22 microns $' ~ Accident consequences depend principally on the initial inventory of fission products,. initial fuel temperature, and initial l(_ . condition of fuel. Initial fission product inventory is controlled by .a specific Technical Specification limit which is not at this time-

being approached. The small increase in fuel temperatures, both global and local, resulting from the fluctuations will not have a major impact on the accidents. 5.3.2.3 Reactivity At operating temperatures, the fuel element colunns are separated by small clearances. The gaps in the core in the hot condi-tion total about 1.5 inches. The most extreme changes in reactivity which could result from the core motion is.to compress the core so as to close all these gaps. The resultant reduction in neutron leakage amounts to only J.00015 Ak.. In FSAR Section 14.2.1.3, a reactivity change of 0.000 ak was evaluated and no damaging effects were found. Therefore, no damaging effects from these small reactivity changes due to core motion are possible. 5.3.2.4 Structural Integrity of Core { 5.3.2.4.1 Fuel Elements Loads on fuel elements and dowels during fluctuations have been calculated and are presented in Section 5.3.1.6. Large margins exist for flat face, top and side edge, and dowel single impacts. Fuel element and dowel / socket strengths have been determined experimentally. Single FSV dowel socket systems have been shown to have a failure strength of 950 lbs. (vs 165 lbs. load). FSV fuel elements have been tested to failure 12 aoth static and dynamic impact loading. Mean static flat face failure loads were determined to be 103,900 lbs. (vs 2,100 lbs load). Single impact failure loads are slightly higher. Although top and side edge failure loads were not experimentally determined for the FSV fuel element geometry, they were determined for the LHTGR fuel elements (Ref. 5-1). The close similarity between these elements allows use of the LHTGR data to estimate k FSV fuel element side and top edge failure loads. These loads were determined to be:

Side edge 49,500 lb (vs 400 lb load) _ (1 Top and bottom edge 22,600 lb (vs 1100 lb load) Thus it can be seen that the fuel element and the dowel socket system easily cceepts single impact loads. The fatigue c.aracteristics of graphite ensure the integrity S of the fuel elements and dowels for more than 10,000 cycles (vs the 240 cycles experienced). The impact fatigue behavior of the FSV fuel element and dowel / socket structure can be determined from the results of full scale fatigue testing. Full scale testing has been done for FSV _ fuel elements in flat face impact fatigue and for LHTGR H-327 elements for top edge . fatigue and dowel and socket fatigue (Refs. 5-1, 5-3). l The large design margin of the core structures may be seen 4 by observing that even at the extreme value of 10 cycles the strength ! ( of the FSV element and dowels have only decreased to the following values: 4 j Configuration Strength 9 10 Cycles Max. Load Ext. during Fluctuation Flat face impact 40,000 lbf 2,100 lbf Topandboktomedge 16,700 lbf 1,100 lbf Side edge 36,600 lbf 400 lbf Dowel / sockets 590 lbf 165 lbf Even further conservatism is achieved for irradiated graphite (Ref. 5-4). l 5.3.2.4.2 Core Support Blocks and Posts Movement of core support blocks resulting in sliding contact on the block keys, rolling contact on the ends of core support posts, l and impact loads on support block to barrel keys. No significant wear is expected based on a 1970 simulated (dry helium ambient temperature) l3 normal operation test on FCV core support blocks and-posts. 40,000 cycles

~ 4 of motion (~.4 inche) representing 100 to 25% power operation resulted in s:me brightening of the contact areas in the support posts, but there were no dimensional changes or indications of structural damage. Impact of the 2000 lb. core support blocks at ~3 inches per second results in an axial load on the core support block to barrel key of roughly 14,000 lbs., which is within the combined design load of 16,000 lbs. axial and 8,000 lbs. shear. Impacts of the top reflector block on the reflector to barrel key result in loads of 860 lbs., com-pared to 21,000 lbs. design capacity. 5.3.2.5 Contioi Rod Insertability The ability to insert control rods under misaligned rod channel conditions is discussed in Sections 3.2.2.6, 3.2.3.4.2, 3.3.1.2, 3.8.1.2, and 3.8.2 of the FSV FSAR. As described in the FSAR, and recently confirmed, the maximum misalignment of a control rod channel available if all gaps across the core are stacked together is 1.5 inches. Rod insertion tests were conducted using over 1.6 inches misalignment in the insertion loca-t tion and 2.5 inches misalignment in the rod channel and no appreciable increase in scram times were noted when compared to similar tests with the core aligned. The conclusion was made that no predictable misalign-ment of the core will interfere with the ability of the control rods to be inserted or withdrawn. That conclusion is still valid, even if we assume that core regions are moved in such an unlikely manner as to close all gaps in the core. Rod drop tests at the plant February 10, 1978, and March 17, 1978, from 26 inches to 0 inches showed insertion speed of 1.3 inches /second, indicating no significant misalignme.it in the rod channel. 5.3.2.6 Effects on Secondary System Core outlet temperature fluctuations result in varying steam temperatures. Variations of roughly 20 degrees are representative, and i variations of up to 40 degrees have been observed. Components of the secondary system were reviewed for effects of this variation, and it was found to be well within,the endurcace fatigue stress limit of the system.

Endurance limits at full load for the relatively short duration of fluc- .tuation testing range frca a low of 30*F for the module steam outlet pipe to 150*F for steam generator tubes (assuming design temperatures are not exceeded). As shown in Appendix 1, operating limits for fluctua-tions are set well within these values; and pretest conditions are set to stay within design conditions. Small changes in feedwater flow and cir-culator speed result from the control system responding to the averaged nuclear flux channels and steam temperature 11uctuations, but parameters are not significantly perturbed. The plant responds as expected to core outlet temperature fluctuations, and the effect on the secondary system is minimal. 5.4 PORC and NFSC Actions The Public Service Company's internal safety committees have maintained cognizance of the fluctuation investigations and particularly of testing performed at the site. Special tests related to the investi-gation of the phenomena have been reviewed and accepted by PORC as not ~~ safety significant. Each of the approved test proposals contained limits on the magnitude of allowable fluctuation amplitudes and the duration of the fluctuations so as to assure that neither plant damage nor Techni-cal Specifications violation would occur. (See the attached proposed test plan for over 70% power testing for an example.) The NFSC has reviewed these PORC actions and findings related to temperature fluctuations and indicated their approval of them. Apart from its review of PORC actions and findings, the NFSC has been presented summaries of temperature fluctuation incidents and reports of the status of the related investigation. It has also reviewed the preliminary RO 77-43 report. On January 5, 1978, following its first review of RO 77-43 and the fluctuation incident details, the NFSC established an open item on its agenda, which is reviewed at each regularly scheduled meeting. O

s The item will not bg[onsidered for closing until the Final Supplement ~ to RO 77-43 has been submitted to the NRC. At its 38th meeting, held June 22, 1978, the NFSC discussed the ability to continue plant operation without additional risk to the public health and safety. A consensus judgement was reached that con-tinued operation should not be limited for safety reasons because: o Fluctuations are readily apparent by observation of normal plant instru-ments. o Ability to terminate fluctuations is always available through power reduction. o Ability to shutdown or scram reactor is not affected by fluctuations. Except for the initial fluctuation, none of the applicable Technical Specification limits have been exceeded during a fluctuation event. The duration of the single violation was for a total of about 40 minutes out of the 24 hours allowed. There is no indication that onset of a fluctuation could cause a loss of ability to maintain acceptable process control of the primary and secondary coolant systems. REFERENCES S-1. Sevier, L., "HTGR Graphite Fuel Element Seismic Strength," General Atomic Report GA-A13920, April 30, 1976. S-2. Shatoff, H. D., " Approximation of Corner and Edge Loads f.am HTGR Core Seismic Analysis Codes," General Atomic Report CA-A14247, April 1977. S-3. Chiang, D. D., " Fatigue Tests of Dowel-Socket Systems," General Atomic Report GA-A13861, June IS, 1976. 3-4. Price, R. J., " Cyclic Fatigue of Near-Isotropic Graphite: Influence of Stress Cycle and Neutron Irradiation," General Atomi-3eport ) GA-A14588, November 1977. \\

~6.1 TESTI!!G Future testing will concentrate on further characterizing the fluctuations and defining limits for operation in a nsn-fluctuation mode beyond 70% power. An understanding of.the fluctuatior. phenomenon and iden-tification of remedies will be pursued. To this end a number of activities are planned. These activities will be started after upgrading of the diag-nostic instrumentation as discussed in Section 3.

1) Threshold Testi~ng Below 70% Power Tests will determine fluctuation sensitivity to core resistance threshold and to rate of power change between 30% and 70% power.

This test wil! provide data from the updated diagnostic instrumentati9n and help characterize the fluctuations. Estimated completion data is August 31.

2) Threshold Testing Above 70% Pcwer Based upon the results of threshold testing below 70% power, extrapolations will be made to non-fluctuation operation above 70%.

Thres-hold testing will then be conducted similar to that performed for below 70%. Current data extrapolation indicates that the A P threshold will increase with power, that indicated neutron flux changes due to neutron streaming will increase, but that steam generator inlet temperature changes will not increase significantly. Current data also indicate that stable non-fluctuation operation can be achieved at rwer levels up to at least 80% power. The plan-ned additionel tests above 707, power will define the effect of core thermal p0wer on fluctuations, the effect of higher power to flow' ratios, the effect of power on changes in affected parameters, and operational constraints to avoid fluctuations at the higher power levels. The testing will be conducted using the same limits employed in current fluctuation testing that limit the mavimum amplitude of steam temperature changes. M Tests ilith The Instrumented Control Rod Drives This subject will be covered in a separate submittal.

e

4) Other Tests and Instrumentation Other in-pile tests are being evaluated to help identify possible specific fluctuation cr.sses or test potential remedies.

At the first refu'eling outage, an in-core inspection will be perfomed. Refueling region 35 is scheduled to be reloaded. This region is a perpherial region in the north-west quadrant of the core and, as such, has experienced large temperature changes during periods of fluctuation. After fuel block removal the.s.ide reflector will be exposed and will be inspected using the TV camera. Selected fuel blocks will be inspected to document there condition. Out of pile tests involving use of scale models to determine causes or phenomena affecting the fluctuation are being studied. Th se out-of-pile tests involve measurement of pressure and flow distribution in a core region, studying flow asymmetrics in the primary system (emphasizing effects of the thermocouple penetrations entering the core in the north-west), and studying distortions in side reflector columns. Finally, the feasibility of providing in-core motion measuring instrumentation is being avaluated. Motion transducers for the 700 F high radiation environment at the core inlet are within the state of the art but are mechanically complicated to allow for varying height between regions. Specific designs to allow measurement of. motion between regions are being prepared. Measurement of motion in lower portions of the core adds high temperature and conductor / connector / mounting problems to the already difficult environment and may not be possible. Feasibility of making such measurements is being investigated. 6.2 ANALYSIS Analytical models and correlating analyses of current data are being pursued in an attempt to simulate and understand the fluctuation phenomenon. These include the following:

i

1) Analyses of the effects of thermally-induced column bowing and thermally-induced dimensional change in the core support posts which could be triggering fluctuations. These processes are being modeled to determine their feasibility as causes of fluctuations and their associated characteric time constants.

2) Modeling of the core flow and pressure distribution is underway to detemine the magnitude of forces available to move blocks and the core gap structure favoring large lateral pressure differences.

3) Assymetric core barrel heating due to primary helium themocouple penetration induced flow perturbation in the inlet annulus being investigated as a possible contributor to the northwest predominance.

4) Analysis of the signals from PPS nuclear channels and special fission chambers is underway to attempt to understand the causes of the large dislocations and the smaller reactivity induced detector changes.

5) Comparisons are undenvay relating the six nuclear channels, PCRV movement, gap temperature changes, and steam gt.ierator inlet tempera-tures in both the time and frequency domains in.an attempt to definitize relationships and identify a mechanistic scenario of movements that corre-late with the data.

6) Design solution efforts are underway, assuming postulated causes.

Feasibility of specific designs for restricting motion of the core is being inves'tigated. 6.3 OUTSIDE CONSULTANTS, A number of nationally renowned consultants have been assembled ( to review the effort and data to date for possible additional effort and/or different emphasis based on their experience and areas of expertise.

L981nt 2/-y-7/8 OL3 ~' s = ) RT-502: POWER INCREASE TO 100'e P0hTR SUhMARY From an initial steady-state condition at 70% power, the core power will be increased slowly (1/25,/ minute) to 75% and stabilized. If no oscil - lations occur, power will be reduced to 70%, stable operation achieved, and a 3%/ minute load increase to 75% power will be effected to attempt to trigger. oscillations. This process o'f slow and then rapid power increases of 55, .will be continued until o' sci 11ations are encountered or until 100% power or a plant limit is encountered. If oscillations occur, data will be recorded for a short period of time and the step which initiated the oscillation will be repeated to establish reproducibility of the onset of oscillations. If oscillations have not occurred by 80% power, SUT B-0 sequence 71 tests (steady-state) will be performed prior to exceeding 80% power for oscillation testing. This is to confirm that other plant systems are operating properly prior to exceeding 80% power. Initially, the core orifices will be adjusted to positions more open than normal and the main steam temperature setpoint will be reduced to lower core pressure drop. If osti11ations are not encountered or are encountered only at a high core power (>90%), an additional series of power increases will be performed with the core orifices in a more closed configuration and the main steam temperature setpoint increased. 1 I OBJECTIVE This test has two objectives: the first objective is to define the conditions for stable operation above 70% power and the second objective is to further our understanding of the oscillation phenomenon.

] w Since oscillations were first encountered, several tests have been ^ f conducted under various core conditions. In large part, these tests were designed to gather specific informatica on what key parameter or combination of parameters' leads to the oscillations,.since this knowledge could be instru-mented in understanding their cause.. These tests have shown fairly conclu-sively that' power level is not by itself a parameter of primary importance to the oscillation threshold, and they have established core pressure drop as a key parameter, probably closely related to the cause of the oscillations, j Another result from these tests is that it appears that the core pressure drop at which oscillations are produced is higher at higher core power levels. Figure 1 shows this trend. % ese tests have also served to help define conditions at which stable core operation can be maintained up to 70% power. Considerable operating time at power levels between 60% and 70*. have demonstrated that the reactor can reliably be operated in a stable mode. Stable operation at power levels up to at least 80% power is highly probable. With reheat steam attemperation the core power to flow ratio is increased, so that a reactor power of ~80% will be achieved before the core [ flow rate and pressure drop will equal those already achieved at 70% power without reheat attemperation. The limit may also be increased by the indi-cated increase in the core pressure drop threshold for stable operation shown on Figure 1. The proposed tests above 70% power are part of the continuing test pro-gram to identify the parameter or combination of parameters which lead to oscillations. For the same core flow and pressure drop a higher core power will be achieved which ui11 serve to confirm core pressure drop as a key parameter. l f SAFETY' CONSIDERATIONS l l ( Since first being encountered on October 31, 1977, oscillations-have l been initiated several times in a continuing effort to understand their i. l .~ ,~ -._

('+ cause. The power levels at which oscillations were initiated have ranged from 40%.to 68%. A total of about 40 hours has been spent in an oscillating mode, which is equivalent to 240 cycles for a perwod of 10 minutes. Although the cause of the oscillations is not known, there are several reasons for concluding that continued testing 1.s safe. The oscillation-12e primarily local in nature, with the core power, flow, and average-terperatures b'eing relatively stable. An. inspection of the top plenum in Decembel 1977 after oscillation testing at power levels between 53% and 68% showed it was in good condition. An inspection at that time of the control rods in region 34 (which were inserted throughout the entire period of oscillation testing) also showed no signs of excessive temperatures or impact. Although differences in oscillation characteristics have been observed, safety of the plant for operation at greater than 70% power is not believed to be a concern. The only parameter that has increased with power level during oscillation it the magnitude of the power fluctuation on nuclear channel 6. Other parameters, such as the steam generator temperature flue-tuations, appear to be constant. Also, there is no indication of differences in oscillation characteristics or magnitudes between recent oscillations and those experienced during the first oscillation tests of November 1977. Oscillations may be initiated two or possibly four times during tais test. If encountered, the oscillations will be allowed to continue far a short' period of time before power is reduced, which has been demonstrated to be a reliable means of stopping the oscillations. The total time in oscil-lation would be about one hour, or 6 cycles for a 10 minute peripd. 4 j b c( .l i n

1 OPERATING LIMITS In addition to the normal plant operating procedures av.d limitations, the following limits should be observed: 1. The steady-state steam generator module helium inlet temperature shall be limited to 40*F about the mean. 2. The hot reheat temperature imbalances between individual modules and loop averages should not exceed 30*F. 3. The main steam temperature imbalances between individual modules and loop averaged should not exceed +10*F, -30*F. 4 Prior to a power increase, the core region exit gas temperatures shall be limited both as shown in Figure 2 and as given in SOP 12-04. The temperature mismatches given in Figure 2 are based on evaluations of previous oscillation data and on knowledge of region temperature changes during a 5% power transient, such that if oscillations are encountered during a power increase or decrease the region exit gas temperature mismatch limits of LCO 4.1.7 of the Technical Specifications will not be exceeded. Since the +50 F exit gas temperature mismatch limit on all regions of SOP 12-04 will come into effect during this test (Figure 6 of S0P 12-04), all region exit gas temperature mismatches are to be within 50 F prF'r to a power increase. 5. The following limits apply to the steam generator modules when oscilla-tions are present: A. A temperature oscillation of module MS temperature (about its mean) up to 110*F is acceptable with no specific time limit. B. Temperature oscillation of module MS temperature (about its mean) greater than 10*F but less than 30.*F is tolerabic for a maximum of 1 hour duration per event. C. A temperature oscillation of module h5 temperature (about its mean) which reached 30*F amplitude is cause to take immediate corrective action. The brush recorders should be used to monitor module FG outlet tempera-l ture.during oscillations. l l 1

6-INSTRtDfENTATION/ DATA SYSTEMS J.r tugh the duration of this test, the following data systems will be operacing and personnel will be present: 1. Brush recorders with selected steam generator module outlet tempera-tures. A steam generator performance engineer will be present to moni-tor steam generator performance. 2. Data logger. A core performance engineer will be present to monitor the core temperature limiting conditions for operation. If either of these systems becomes inoperable, testing will be halted until the system is reinstat'el. If oscillations are encountered when either of these systems is inoperabic, core power should be reduced as discussed in item 14 of the test procedure, until the oscillatians cease. During power increases and for a period of in hour following a power increase, the following data systems and dath taking frequencies are de-i sired: 1. Data logger on a fa t sample rate (15 seconds or less, with the following variables displayed on the trend recorders: o wide range linear channel 6 recorder 1 regulating rod position o region 34 outlet gas temperature 'i i region 36 outlet gas temperature ( recorder 2 or other desired variable 2. Steam generator fox II computer on a fast sample rate (~5 seconds) 3. Model verification computer 4. FM recorders s 5. Time data analyzer At periods during the test other than those specified above, the following data systems and data taking frequencies are desired: (. 1. Data logger on a sample rate of 2 minutes 2. Steam generator fox II computer on a sample rate of 15 seconds 3. ' Model verification computer

s. PROCEDURE I Notes 1. The approach is to perform slow and then rapid power increases of 5%. -During the power increases, it is desired to minimize core perturba-tions, so that unless required to stay within the limits of this test or SOP 12-04 an exchange of the regulating control rods with the shim bank or orifice valve adjustments should be done no sooner than 1/2 hour after a load change. 2. If oscillations are encountered, proceed to step 13. Steps 1. Establish reactor therma ~1 and xenon equilibrium at 70% power, with cold ~ reheat attemperation in service and with the following operating condi-tions to achieve a low core pressure drop: A. Attemperation schedule for 72,000 lbm/hr flow rate at 70% power, i increasing linearly to 144,000 lbm/hr at 100S. power. B. Hot reheat temperature controller setpoint at 980*F and loop main } steam temperature controller setpoints at 940*F. C. Core orifices adjusted to achieve a region exit temperature mis-match of +40*F to +45*F in the maximum, power region with its orifice valve fully open. D. The regulating rod set at 110 inches to 115 inches withdrawn. 'The reheat temperature controller setpoint has been reduced from 1000*F to 980*F provide margin steam generator limits if oscillations are encountered. The maximum region exit temperature is given as +40*F to +45'F because the +50*F region exit gas temperature mismatch limit will come into effect during this test (Figure 6 of SOP 12-04). In addition to the core orifices, the main steam and attemperation trim valves may be used to aid in reducing steam generator module temperature mismatches. The turbine should be kept on load limit except during load changes. The expected core pressure drop at 70% power is 2.6 psi. 2. Start data recording systems and station personnel to monitor core and k steam generator performance. Slowly increase power (1/2%/ min) to '75% power and stabili:e for S hours. e J

(, 'o='*'*- 3. Reduce power to 70% and stabilize for 1 hour. 4. Increase power to 75% at 3%/ min, and stabilize for 2 hours. 5. Slowly increase power (1/2%/m!.n) to 80% and stabilize for 2 hours. 6. Reduce power to 75% and stabilize for 1 hour. 7. Increase power to 80% at 3%/ min and stabilize for 8 hours. 8. Complete SUT B-0, sequence 71, to verify proper overall plant performance at 80%. 9. Slowly increase power (1/2's/ min) to 85% and stabilize for 8 hours. 10. Reduce power to 80% and stabilize for 1 hours. 11. Increase power to 85% at 3%/ min and stabilize for 2 hours. 12. Repeat steps 9 through 11 for power levels of 90%, 95%, and 100%. 13. If indications of core, oscillations are observed, stop any power in-creases in progress and observe amplitude of oscillations to ensure that the limits of this test are not being exceeded. If any limit is approached, immediately reduce power and reduce oscillation amplitude. If the am-plitude of oscillations is not about to exceed limits, observe and record the plant response for 20 minutes of the oscillation. Reduce power in 'S MWe steps at 20-minute intervals until the oscillations cease. 14. If the observed oscillation magnitudes were acceptabic, attempt to retrigger the oscillations by repeating the road increase which previously appeared to initiate the oscillations. Once retriggered oscillations are confirmed, reduce power to 5 MWe at 20-minute intervals until oscillations cease. 15. If oscillations were not encountered or were encountered at a power level E90%, increase main steam controller setpoint to 980*F and adjust ' core orifices for optimum core region and steam generator module tempera-tures per S0P 12-04. (Optimum region temperatures are when the exit gas temperature mismatch on the highest power region is 0*F to +10*F and its orifice valve is fully open. Regions may have to be at somewhat higher temperatures chan this to achieve acceptable steam generator tempera-ture mismatches.) Then repeat the rise-in-power per steps 2 through 7 and 9 through 12 until oscillations are encountered or until 100% power or a plant limit is encountered. E

-m z 4. ANTICIPATED RESULTS I Figure 1 shows core pressure drop versus reactor flow rate for several operating points to date, including those points where oscillations were encountered. A cross-sectioned band has been drawn through these points, representing the area in which oscillations may-occur. Below this area, oscillations have not been observed and above this area the reactor has not been operated without oscillations. Also shown on Figure 1 are the projected operation lines for the two test conditions, one with the core orifices and main steam temperature control set for low core pre.ssure drop and one with nominal core pressure drop. Based on these operating lines, for the minimum core pressure drop configuration oscillations may be encountered at as low a flow rate as 82% or may not occur up to 955. power. The reactor power levels at which these' flows will be attained for the I programed reheat steam attemperation levels are also given on Figure 1. Note that with reheat uttemperation the core power to flow ratio is increased significantly, which explains the lower reactor flows for given power levels than those operated at to date. N l l e e 9

t .....~ ^- . _.:.~ ..;j 0 OSCIL LnTsu.4 O s rw.c. t.... ./*/ (ft & .(,. o. ..'.....E!AO oe '% M ,.s... t. . i s w s ..r t ....u - .u. .;.,,, payeciso Ny.hw (as sts c

• f D hnN4 un6

.N ~ n po ...s .t. g s. gy .g r g . s. r., y s g gs A..O g g gg. .s ., g ..... s. s. g,. s. r. y \\ t,, g.,, i .s s, g g g g., 3 s. -w y. w \\ <.M go_ g hc/o eg.O, c) e ~ w 43 -. A srcru 100 6 2 s 2 O ..O . t.) .. l.:. .. & E4 7~m(, u d ~ ,0_ .. s. .=.. 1.0 .~ ) o i 30 =40 SD 40 '70 80 40 ro: sYa 14 6. F~ L O G 8 l j </u l rr .}}