ML20140H842

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Proposed Tech Specs Pages 2.1 Re Safety Limits
ML20140H842
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Site: 05000231
Issue date: 01/28/1971
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GENERAL ELECTRIC CO.
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References
FOIA-97-34 NUDOCS 9705130296
Download: ML20140H842 (17)
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Text

. . . . - -

. 2.1 Safety Limits Applicability

, Applies, to process variables which affect the integrity of the primary system.

Objective To assure the protection of the primary process system barrier's against uncontrolled release of radioactivity.

Specification

/ The r2xir = pereircible r:2 ter cere flux ch211 be 110% ef r2tak flur, except th 2 t the limit deer net npply durin;; 2n excurrien tect initiated by FFID.

A. The maximum permissible reactor core flux shall be determined as follows, except that this limit does not apply during an excursion test initiated by the FRED:

1. If guinea pig rods are not present under the innermost refuel-ing ports, the limit shall be 110% of rated flux.
2. If guinea pig rods are present under the innermost refueling ports, the limit shall be the lower of Ly or L , where:

2 Ly =

1.05 (110N/600X) % of rated flux.

L = 110% of rated flux.

2 N = total number of fuel rods in the core.

X = ratio of guinea pig rod power to the power of a j standard rod located nearest to the center of the l l

core, as given in Reference 11. _'

. B. The maxim e permissible reactor core flux during an excursion test 3

after the FRED is fired shall be 6.25 x 10 times rated flux.

C. The maximum permissible integrated energy deposited in the core

, during an excursion test shall not exceed the limit shown in Figure 2.1-1. The limit is exceeded when reactor conditions

! result in a point above the limit line.

l l D. 'Ihe reactor vessel outlet coolant temperature shall not exceed i l

1050*F. j E. The maximum permissible reactor vessel cover gas pressure shall be 45.5 psig.

Change No. 3 l

2.1-1 9705130296 970505 PDR FOIA l VARADY97-34_ PDR

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$ Limit without Guinea Pig Rods located under  :

$ 800 the innemost refueling ports. -

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Q ts

  • Limit with Guinea Pig Rods located  !

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g under the innermost refueling ports j .

$ 400 -

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5 95

$b o l l l l 2 0 3 6 9 12 15 i . -

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p MAXIMUM ALLOWABLE ENERGY ADDITION FOR PLANNED TRANSIENT TESTS T

w Figure 2.1-1 i

i e

4 P Bases  ;

  • 1 l The ::fety lisit er reactor core flux of 110% of rated flux (defined in j l Section 1.6). corresponds to a reactor power level of 22 MWt. The steady -)

, state reactor power level is equal to the heat removed from the reactor l

8 vessel by the main primary and auxiliary primary coolant systems and is determined by measuring the coolant flow rate and coolant temperature rise from reactor vessel inlet to reactor vessel outlet for both of these cool-ant systems.

The reactor will normally operate at rated flux corresponding to a reactor power (defined in Section 1.5) of 20 MWt. The maximum linaar power density in the standard hot fuel rod with the reactor operating at 20 MWt is 21.8 KW/ft using the minimum core loading of 600 fuel rods specified in Section 3.3, the core power peaking factors given in Reference 1, and assum-ing the full 20 MWt is being generated in the fuel meat. Oper-:tica at a reacter pruer cf 22 "!t :: cult in : maximum linen: peror d:ncity of 24 ""/f t in the rtenderd het fuel red. Operation at the safety limit given in para-graph 2.1.A.1 results in a maximum linear power density of 24 KW/ft in the

- standard hot fuel rod. Operation at the safety limit given by paragraph j 2.1. A.2 results in a maximum linear power density of 25.2 KW/ft in' a guinea pig rod located under the innermost. refueling port. The values for X given in Reference 11 for calculating the limit of paragraph 2.1.A.2 will remain constant for the range of fuel rod loadings allowed by the Technical Speci-fications because the overall neutron flux profile will be constant. All 648 rod positions in the fuel channels (the central channel is a drywell for test devices) contain either fuel rods, B C rods, or stainless steel 4

rods. The radial flux profile is constant with changes in core loading between the 600 and 648 fuel rods allowed by the Technical Specifications because loading changes _ are accomplished by interchanging fuel rods and B or stainless steel rods rather than changing the size of the core.

Local neutron flux perturbations due to the presence of a B C poison rod 4

which could affect the value of X are small because the number of poison rods is limited by reactivity requirements and because the poison rods are distributed throughout the core, as required by 3.3.N, to minimize perturbations in the overall neutron flux profile. Burn-up effects on the flux profile are Change No. 3 -

! 2.1-3

m 1

. also negligible for the planned SEFOR program. They are in the conserva- l i

tive direction (reducing guine'a pig rod power density) because the flux l t

profile flattens as plutonium is preferentially burned out at the core center. Changes in the power profile at the guinea pig rod locations with variations in re'flector position are also negligible because allowable reflector patterns are restricted by the excess reactivity limit given by 1 3.3.B. At 20 MWt, an unbalanced pattern of nine reflector segments fully raised and one segment partially raised would have no significant effect on the peak power density of the guinea pig rods. .

The peak temperature in a fuel rod is calculated to reach the fuel solidus temperature ( at a linear power density of 26 KW/ft. The accuracy of these calculations, determined by the present state of technology, does not com-pletely preclude the possibility that a small amount of centerline fuel melting may occur at a linear power density of 26 KW/ft. Tests conducted under Task 3 of the SEFOR Preoperational Research and Development Program and PA-10 of the AEC LMFBR De'elopment v Program ( ' have demonstrated that oxide fuel can be operated for a large number of cycles at linear power densities greater than 26 KW/ft without damage to the fuel. For example, results given in Reference 3 demonstrate that fuel can be cycled more than 100 times between power densities of 23-29 KW/ft (cycle frequency %.01 cps) with no damage. Fuel tests conducted under PA-10 have shown the capability of this type fuel to sustain burnups of tens of thousands of MWD / Ton at linear power densities in excess of 24 KW/f t. In contrast, the SEFOR fuel will experience an estimated burnup of only 1500 MWD / Ton during the three-year experimental program. All of these tests were performed using sodium coolant at temperatures similar to those that will occur in SEFOR.

If the reactor outlet temperature were to increase from the nominal design value of 804*F to the safety limit of 1050*F at the same time the reactor power approached the safety 14 ** ^'"w* *'^-^-"'*'-~'--~^#- '"^'

t- perature in the etendard p"9 fuel red reu1&cor-respend te enly e 0.' /f*

iacreece in the lineer perer daarity, limits given in paragraphs 2.1.A.1 or 2.1.A.2, the resulting increase in fuel temperature in the standard peak fuel l

l rod and guinea pig rods would correspond to only a 0. 7 KW/f t or 0.8 KW/f t increase in the linear power density of these rods, respectively. ,

Change No. 3 2.1-3a l

- _ _ . .- .. -- _ _ - ~ - - - . . . -- -

t l

l I

, Additional margin at the safety limit cf 22 "'!t is provided by the fact that the 24 KU/f t maximum linear power density discussed above. I: the fuel et th

-,rm... se_4. _c',,""

l is based on the total ' power entir; 22 "Wt being generated

'inthe'fuelmea$t for a minimta core loading of 600 fuel rods. Physics calcu-

. i lations indicate that only 94% of the total energy generated inside the reactor vessel is generated directly in the fuel. (5) The remaining 6% is deposited in the coolant, structure and shielding inside the reactor vessel. This reduces the actual KW/ft generated in the fuel of the hottest standard rod at 22 MWt from 24 KW/ft to 22.6 KW/ft. Ir, addition, th: :::1: 21 1: ding of rFF0F is p*aA4-*ad ta be 630 f uel rede uhich vill furthcr r duce th pech K"/ft gcr. crated ir th e fuel at th cafety limit a f 2 2 "'?: :: 21.^ ""/ft. Ecced en there dat2, 1* 4=

^^"-1"A^A

  • bat sufficient r rgi  : ict: betu: r oper tier et the 22 "'d:

2 fety limit 2nd the print et h!:S cignif!:n.t fuel red d2r2ge zeuld accur.

For a guinea pig rod, the actual KW/ft at the safety limit defined in paragraph 2.1.A.2 is reduced from 25.2 to 23.7 KW/ft. In addition, the nominal loading of SEFOR is predicted to be approximately 636 fuel rods which will further reduce the peak KW/f t generated in the fuel at the safety limit defined in paragraph 2.1.A.1 to 21.2 KW/ft. The loading will not affect the peak KW/ft for the guinea pig rods at the safety limit. The 23.7 KW/ft generated at the safety limit in a guinea pig rod located under the innermost refueling ports corresponds to a linear power density 5% greater than that of a standard rod )

at minimum core loading and 110% of rated flux. This increase is justified 1

by increased surveillance of the guinea pig rods as specified in paragraph 1 4.3.E. Based on these data, and the limited' number of guinea pig rods that can be loaded under the inne'rmost refueling ports, it is concluded that I

- sufficient margin exists between operation at the safety limits given in paragraphs 2.1.A.1 and 2.1.A.2 and the point at which significant fuel rod damage would occur.

The rapid expansion of the fuel due to heating during planned prompt power transients exerts a dynamic load on the fuel rod, grid plate, and the core support shroud inside the reactor vessel. This dynamic load can be treated as an impulse in these structures since it is applied during a time interval that is small relative to the natural frequency response of the structures.

l The impulse is proportional to the peak power reached in the transient.(6)

Analysis has shown that the 304 SS bolts (which attach the core support shroud to the reactor vessel) are the critical items (6,7); i.e. , they reach their working stress limits for this dynamic load before the other components do. The safety limit of a maximum core flux of 6.25 x 10 times rated flux Change No. 3 2.1-4

(125,000 MWt) limits the impulse load so that the peak stress in the bolts does not exceed the yield strength. Significant margin exists above this safety limit because the stainless steel bolts are capable of carrying

, dygamic , loads 'of several times the value corresponding to the yield strength

.without failing grossly'.

s.: cur > ch'r t:- er Figure 2.1-1 def4 :: th e te t:1 energy-*Meh-can-be-depes44ed in th: :::: durin; 2 pl::::d  ::::12 t aftF :t ::.:::lin; ; pech f t:1 red center-l lin: ' energy derrity ef 250 c21/g. The two curves shown on Figure 2.1-1 define i

~

the total energy which can be deposited in the core during a planned transient with and without guinea pig rods present under the innermost' refueling ports without exceeding a peak fuel rod centerline energy density of 250 cal /g. The _, j 250 cal /g value is the calculated energy density from mixed oxide fuel just '

below its solidus temperature as obtained using the latest available data for mixed oxide fuel heat capacity.( ) Results from transient tests ( on sodium-cooled oxide fuel have demonstrated that repeated transients up to fuel

~

energy densities of 250 cal /g will not damage the fuel. Initial' fuel tempera- '

tures are based on the data and methods given in Reference 10. The difference between the two curves shown on Figure 2.1-1 is the result of differences in power peaking factor f'r the standard hot rod and the guinea pig rods which l influence the' initial fuel temperature and the energy deposited in a particu-lar rod.

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} Change No. 3 2.1-4a i

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- - . . .- + ,, -n ..-,n. --- ~ , . - . , w , n e

y . - - - . . - - _ . _ . - - -- - - - . - - - - . - - - . ~ .

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. The safety limits of 1050*F vessel outlet temperature and 45.5 psig reactor vessel cover gas pressure are selected to limit the pressure and temperature t

.for all compone,nts in the primary system to 'their design conditions. Margin

'between these-safety limits and a damage limit is inherent in the codes used

?

l ,in their design (ASME Shetion III and the ASA Piping Code B 31.7' i The fuel will not approach h::: flux limit (buracut limit) b :: :: c f the-12:;  ::u-t :f cub:2011 ; (sp; ::ximately 600*F :: th: ca fee-y--14 mi4-ef 1050*F) : d th ; :d h::: :::::fer charcateristic: ef th: : dir :alant.

l The maximum heat flux of a fuel rod is limited by boiling, not by a burnout -"

limit, because gross sodium bolling and associated expulsion of sodium from l the core will occur before a burnout heat flux is reached. However, gross sodium boiling will not occur because of the large amount of subcooling in

.a channel (approximately 600*F in the hot channel at the safety limit of 1050*F when guinea pig rods are not present under the innermost ports and approximately 400*F at the safety limit of 1050*F when guinea pig rods are present under the innermost ports) and the good heat transfer characteristics of the sodium coolant. During 44w> steady state operation at less than the _.

limiting safety system settings specified -in Section 2.2, the maximum com-bined stress in the fuel cladding in the standard hot rod and auinea pig rods is less than twice the at-temperature yield strength of 316 stainless steel (9) so that cyclic fatigue will not occur. During transient operation l (either planned transients or accidental transients which exceed the normal reactor power and coolant temperatures) the maximum combined stress in the cladding of the standard hot rod and guinea pig rods may exceed the twice-

.- yield design limit and the fatigue life of the cladding can become the limiting condition. The predicted fatigue life,for the cladding is 750 cycles ( 0) and 300 cycles for the hot standard and guinea pig rods, respec-tively, for strain cycles associated with the Maximum Planned Transients specified in Section 3.12.

Thde f ti;ur l!-'t "e= These fatigue limits were obtained using the methods outlined in Section III of the ASME Code and the fatigue curves given in l Code case 1331.1. The cumulative fatigue damage for the planned transient l

[ test program is,less than 0.02 for. standard rods and less than 0.05 for guinea pig rods. Additional fatigue damage due to unplanned power or Change No. 3 i

2.1-5 4

i 1

l temperature transients approaching the safety limits of 22 '"'t c.:J 2050*r

ccrel eutlet temperature t

paragraphs 2.1. A and 2.1.D will not contribute sig-nificantly to this cumulative total. For example, the minimum predicted

, fat $gue life for the cladding cycled from refueling conditions to er oper2t-

.ing p e ae r o f 2 2, '"!t 2nd 2 1050*F fecesi cut 1ct temperature ic 2000 cycica.

_the safety limits of 2.1.A and 2.1.D is 1000 cycles. Consequently, one cycle of this type (involving violation of two limiting safety system settings) would increase the cumulative usage factor less than 0.0005. 0.001.

e 9

l Change No. 3 2.1-Sa l

. ._ _.. . - _ . . . . _ . _ _ . . . _ _ . . . _ . . _____.__.._.._.__.__.m.-._._.....___..__ _ _ ....... ...- _ _ -.

l~.

References ,

t l.

'.(1)' SEFOR FDSN , Volume I, Table IV-7, p 4-33.  ;

^*

(2)' SEFOR FDSAR, Supplement 10, pp 1-110.

'(3) SEFOR FDSAR, Supplement 3.

(4) GEAP 5198. 19th Quarterly Progress Report, pp 5-22 and 23.

(5) GEAP 5576, . " Final Specification for the SEFOR Experimental Program" ,

January 1968, p II-2.

(6) SEFOR FDSAR, Supplement 18, p 35.

l (7) SEFOR FDSAR, Supplement 19, p 57. .

(8) SEFOR FDSAR, Supplement 10, pp 1-126.

, (9) SEFOR FDSAR, Supplement 10, pp 1-92.

l.

l (10) SEFOR FDSAR, Supplement 10, pp 1-109.

(11) SEFOR FDSAR, Supplement 10, p 1-113.

l .

l i

t r

i

. Change No. 3 i 2.1-6 l

l

. . ~ . . . - - . - . . . -. _ _ -.. . - - - -.-. . -~ --

, , , TABLE 2.2-1 SCRAM FUNCTION

' j l ,. FUNCTION SAFETY SYSTEM SETTINGS l -

  • j High Flux, -

=

105% of Rated Flux, if guinea.

i j

l Wide Range Monitor pig rods are not present under i l~

the innermost refueling ports If guinea pig rods are present  !

under the innermost ports the -  !

setting shall be equal to or less I than the lower of Tp or T , where:

2 110N Ty =

600X l T = 105 2

% of Rated Flux N = total number of' fuel rods in the core X = ratio of guinea pig rod power of standard rod located nearest the center of the core. (Reference 6)- ,_

Low Level,  ; 4 inches below lip of operating Reactor Sodium level overflow pipe High Temperature, = 900*F Core Outlet Low Flow, =

20% below the operating flow set Main Primary point

  • High Temperature = 350*F for thermocouples on the Reflector Region ** reflector guide structure inner diameter and radial web.

= 275*F for thermocouples on the reflector guide structure, outer diameter.

= 450*F for thermocouples in the lower end of the reflector seg-l ments.

l l *The operating flow set points shall be specified

'.l[

.in written procedures.

l **At least ten thermocouples shall be connected in l the safety system, including at least one in each reflector bay.

Change No. 3 2.2-2

_- - - __ -- - - -~. - - -- - - . .-. - . -

Bases Th; limi t iag c afe ty cy a t;; : t ting (LCC C) of 105*' o f ret-ed-41e p re videc 2-5%.

. , m4m x m i m_, e n'm c.eroro 14m4r mr 31nt rn.ic .,111 acc ure _ pretce tig._.t.he sftylimit fer normal reenter operatic; '"he cetual safety cyctem ce'tt4ng  ;

5 yill gencr:11y'be lecc than-105%-cf rcied flux cince c large percentage of

  • b e plant opernting ti-= eill be cpent at pcuer lave-le-below-204Mt rhere the trip c etting-would-noren11y be reduced-a-eeercepending amount. (Only c limited

-nu ' ~ ^ # ^xperimentc zill be conducted at r2ted flun.)(1)

The limiting safety system setting (LSSS) of 105% of rated flux provides a 5%

margin below the safety limit of 110% defined in paragraph 2.1. A.1. The limiting safety system setting of 110N/(600X) % or 105% of rated flux,which-ever is less, provides a 5% margin below the safety limit of paragraph 2.1. A.2.

This will assure protection of the safety limit for normal reactor operation.

The LSSS for operation with guinea pig rods under the innermost refueling ports will limit the reactor power so that the maximum linear power density of a guinea pig fuel rod does not exceed the maximum power density for standard fuel rods at the safety limit of paragraph 2.1. A.1 with the minimum core loading. (Only a limited number of experiments will be conducted at rated flux.)( ) -

The LSSS for reactor vessel sodium level provides assurance of reactor scram in the event that reactor cooling capability should be jeopardized because of a leak in the coolant system and consequent loss of sodium from the reactor vessel. Normal operation of the pump-around loop and overflow nozzle in the

.vassel will maintain the sodium at a constant level in the vessel. A loss of about 15 gallons of sodium from the reactor vessel will cause the level to fall below the level trip probe and scram the reactor. The level trip probes are two inches below the overflow nozzle, providing margin with respect to the LESS of four inches.

The core outlet sodium high temperature trip at 900*F provides a 150*F margin to prevent the sodium temperature from reaching the safety limit. Analyses presented in the FDSAR( } show that the coolant temperature will not approach

( the safety limit for accident conditions except for extreme assumptions involv-ing failure to scram or failure of both main primary pump flywheels.

Change No. 3 2.2-3 l

.i The low -flow trip for the main primary coolant system provides assurance that the coolant, temperature will not approach the safety limit due to loss

- of coolant flow I'f the main primary coolant flow rate decreased to 80%

. of the set point value, the temperature rise'across the vessel would increase  ;

less than 25% (to a vessel outlet temperature of 830*F) before the safety system would receive the scram signal and shut down the reactor. Thus, the low flow trip provides the earliest trip in the event of sudden reduction in coolant flow. .

Adequate cooling of the reflector guide structure, segments, and neutron i flux monitors, is required to assure operability of the reflectors and the neutron monitors. Thermocouples installed in the reficctor guide structure and segments are monitored by the safety system to provide this assurance.

The guide structure temperatures at the positions monitored are predicted to range in value from 200*F to 250*F with all reflector segments raised and a.

i

' reactor power level of 20 MWt. The variations depend on whether the thermo-couples are located in the inner or outer h

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i Change No. 3 I

  • L 2.2-3a l-g_ , . _ . . .m-, ._. , , , . . - - , _,, . ,,,_e,y. . , , , , ,

. . _. . _m.,_.. . _ _ _ . _ - _ . _ . _ . _ _ _ _ _ _ _ . ___._.

'Ihe ten reflector region thennocouples used by the safety system can be l chosen from any,of the applicable thermocouples listed in Table 2.2-1, j

since a different trip level can be set for each thermocouple. The ,

, chodce of' the ten thermocouples to be used for the safety system will be

.made so as to monitor temperatures in each of the ten reflector bays.  !

A se ety system trip at 50% to 60% of the normal coolant flow rate also i provides assurance that the temperature of the neutron monitors will remain .

below the manufacturer's certified operating temperature of 300*F.

t i

References GEAP 5576, " Final Specification for the SEFOR Experimental Program",

(1)

January, 1968.

( '

(2) SEFOR FDSAR, Volume II, Section 16.3, pp 16-10, ff.

-(3) SEFOR FDSAR, Supplement 11, Appendix A and B.  ;

(4) SEFOR FDSAR, Supplement 12, p 7-20.

(5) SEFOR FDSAR, Supplement 17, p G-5.  ;

i i

(6) SEFOR FDSAR, Supplement 10, p 1-113. j 9

f Change No. 3 2.2-5 1

i i

. I.

No fuel rods shall be placed in the center drywell.

J. Fission chambers, experimental foils or oxide fuel samples having a 4

total reactivity worth of less than 60c and containing a total of

, , not more than 0.5 Kg fissile material may be placed in the center channel,(or in a'drywell in the center channel) for irradiation at power levels equal to or less than 100 KWt. Experimental foils containing less than 10 mg of fissile material may be irradiated at reactor levels above 100 KWt. l K.

Fuel rods which have defects as defined below shall not be reinserted in the core:

1. C1' adding rupture, cladding perforation, or other observable defects which may cast reasonable doubt on the integrity of the rods,
2. Local swelling of the cladding in excess of 10 mils or bowing of the rod sufficient to prevent reinsertion of the rod into l i

the core.

3. An increase of more than 1/2 inch in the column height of either fuel segment.

L. The gross gamma cover gas monitor shall be demonstrated to be capable of detecting a fission gas release equivalent to about 1% of the 20 MWt equilibrium inventory in a fuel rod before the reactor is operated above 10 MWt. If such sensitivity is not demonstrated, a more sensitive monitor shall be installed.

M. If the gross gamma monitor becomes inoperable, the reactor shall be shut down, except under the following circumstances.

If a reactor test is in progress, (other than FRED transient test program) and the monitor should fail, reactor operation may continue for 24-hours, if no unexpected changes in cover gas activity indica-

'tive of changing fuel condition have been observed just preceding the failure, and if cover gas samples are taken for spectral analysis at intervals of approximately four hours.

,N. When guinea pig rods are located under the innermost refueling ports, the B4 C poison rods in the core shall be distributed such that the number of poison rods in any quadrant of the core (determined by N-S, E-W centerlines) does not exceed the number of poison rods in any other quadrant by more than two. This specification shall not be applicable when the high flux trip level is set more than 10% below the LSSS.

Change No. 3 3.3-2

(1) observe he continuous monitoring syste inder operating conditions to diagnose the cause of failure or maloperation of the system; (2) pe,rmit an orderly completion of a test series, so that tests completed prior to the failure do not have to be repeated;

, , .(3) plan for an orderly shutdown of'the reactor.

1 1

During periods of reactor operation when the continuous fission ' gas moni- l tor is inoperable, batch samples will be taken at intervals of approximately four hours. This sampling frequency will assure that any trends that might develop will be identified.

~

Specification 3.3.N requires a reasonably uniform arrangement of B C 4 poison rods in the core to provide assurance that the power density in guinea pig '

rods under the innermost refueling ports is not significantly greater than the value used to determine the LSSS. A uniform distribution of poison rods is desirable for most of the planned tests. However, some non-uniform arrangements at low or intermediate power may be required for special tests such as determination of material worths at zero power or determinatic n of available excess reactivity during the approach to power. Such arrange-ments are permissible when the high flux trip level is reduced more than 10% below the LSSS, since the maximum effect of a single poison rod on guinea pig rod power density is only 1/4%.( } The intent of limiting the applicability of this specification is not to permit grossly non-uniform arrangements of poison rods, but rather to permit the flexibility of arrange-ment which may be required during portions of the test program when the reactor is operated below the power levels at which special protection for guinea pig rods is required. _

References (1) SEFOR FDSAR, Volume I, Para. 4.5.3.1, pp 4-50 and 4-51.

(2) SEFOR FDSAR, Volume II, Para, 12.3.6, pp 12-15 and 12-16.

(3)' SEFOR FDSAR, Para. 16.4.2.5 and 16.4.2.6.1.1, pp 16-26 and 16-28.

(4) SEFOR FDSAR, Para. 16.2.1, p 16-4.

(5) SEFOR FDSAR, Appendix B, Para. B.5, p B-3.

(6) SEFOR FDSAR, Supplement 10, p 3-10.

l (7) SEEDR FDSAR, Volume .I, Para. 4.2.2.2.2, p 4-2.

(8) SEFOR FDSAR, Volume II, Para. 12.2.1, p 12-3. l (9) SEFOR FDSAR, Volume II, Para. 16.2.4.3 (10) SEFOR FDSAR, Supplement 21, pp 2, 3.

Change No. 3 (11) SEFOR FDSAR, Supplement 3.

l (12) iSEFOR FDSAR, Supplement 21, pp 1-17. 3.3-7 j l

l (13) Addendum No. 2 to Proposed Change No. 3 to the Techn:

l J anuary 27. 1971 l

\

l

I 4.3 Reactor Fuel Rods -

Applicability

' Applies to fuel rod examination made in the refueling cell.

?

l ,

Objective l To assure maintenance of fuel rod cladding integrity during reactor operation.

l Specification

! A. Two or more guinea pig fuel rods which have operated at power l densities higher than the power density of standard fuel rods l

nearest the center of the core shall be removed from the reactor aftar operation at reactor power levels of 15, 17.5, and 20 MWt, and shall be examined.in the refueling cell by visual observation, dimensional checks, and gamma scans. After reaching a power level of 15 MWt and before reaching 17.5 MWt, the interval between fuel rod examinations shall not exceed six months.

B. Before the start of the sub prompt critical excursion tests and before the start of the prompt critical excursion tests, a minimum e

of one guinea pig fuel rod and one standard. fuel rod shall be examined by the methods described in "A" above.

C. After each prompt critical excursion test, at least one guinea pig rod and one standard rod shall be examined by the methods described in "A" above.

- D. If the examination of a fuel rod should indicate a defect as described in Section 3.3K, additional fuel rods shall be examined to determine the extent of additional defects if any.

' E. Following the examination after operation at 20 MWt as specified in 4.3.A, two or more (if available) guinea pig fuel rods which have operated at power densities higher than the power density j of standard fuel rods nearest the center of the care shall be j examined by the methods described in "A" above at intervals such that the rod exposure during the first interval does not exceed Change No. 3 4.3-1

m . .. - _ _ _ . . . _, - - _ _ - . _ . . . . _ . . _ . _ _ . . - _ _ . . . . . . _ . _ _ _ . . _ . -

t a .. .-

a .

  • a cpre integrated power of ~300 MWt-days with the reactor operat-ing at power levels greater than 17.5 MWt, and during successive

- p . intervals,_does not exceed a core integrated power of 600 W t

. ' days with the reactor operating at-power levels. greater than 17.5 MWt. The maximum time interval between examinations'of guinea. pig fuel rods- located under the innermost ports shall

  • not exceed six months'. Extended outages.of more than one month shall not be included in the determination of this surveillance i interval. _

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Change No. 3 i 3
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