ML20141A515

Informs That Proposed Change 8 to License DR-15 for SEFOR Acceptable

ML20141A515
Person / Time
Site:	05000231
Issue date:	11/12/1971
From:	Skovholt D US ATOMIC ENERGY COMMISSION (AEC)
To:	Wolfe B GENERAL ELECTRIC CO.
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FOIA-97-34 NUDOCS 9705140341
Download: ML20141A515 (20)

Text

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. . DISTRIBUTION I

! .) W. Dobly, DR Q R. Engelken, CO (2)

H. Shapar, OGC 4 N. Dube, DRL (5) o J. Buchanan, ORNL l ,

NOV 4A~ 1971 T. Laughlin, DTIE PDR j ocket File Docket No. 50-231 DRL Reading Branch Reading R. Boyd R. DeYoung l

General Electric Company D. Skovholt ATTN: Dr. Bertram Wolfe R. Vollmer General Manager R. Schemel

, Breeder Reactor Development Operation R. Woodruff l 310 DeGuigne Drive S. Teets ,

Sunnyvale, California 94086 bCb(N Change No. 9 Gentlemen: License No. DR-15 l

By letter dated October 16, 1971, you submitted Proposed Change No. 8 to the Technical Specifications appended to Operating License No. DR-15 for the SEFOR reactor. the proposed change would reduce the number of tests required before l initiating transient testing with Core II.

In evaluating your request, we have considered the effect of the proposed change on margins of safety to fuel clad damage.

As discussed with your staff, certain modifications to the proposed change are necessary to meet our licensing requirements. We conclude that the proposed I

change, as modified, will not..present significant hasards considerations not described or implicit in the Safety Analysis Report and that there is reason-able assurance that the health and safety of the public will not be endangered.

We have redesignated the proposed change as Change No. 9; and pursuant to 10 CFR 50.59, the Technical Specifications appended to Operating License No. DR-15 are changed as shown on the replacement pages in Attachment A. New requirements are identified by underlining and deleted requirements are identified by lines through the material.

l l Sincerely,

.;.:d FDtd D

..'1.C M Donald J. Skovholt Assistant Director for Reactor Operations i Division of Reactor Licensing Enclosure ,

Attachment A - Change to '

Technical Specifications -

i cc: Paul B. Van Buren, Attorney ~fIf/b/O f (, f $,

l Genatal Electric Company

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9705140341 970505 PDR FOIA VARADY97-34 PDR i, _

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ls Q **,, UNITED STATES .

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, g./ ] ATOMIC ENERGY COMMISSION g g '03 in regard to descriptiodMPle7fERi8n 'fdMi requirements, operation, and l ce requirements of auxiliary structures stN#ssnhenssilssii@llnes.'

l Docket.No.50-23L Regulatory requirements specific for particular land types (see Table 4.1.2-1).

l . Regulatory nositions and specific criteria necessary to meet the regulations identified above are Genehfl Electric Company ATTN: Dr. Bertram Wolfe a

, Regulatorpfamidst4WaRga with respect to evaluation of alternative systems designs.2 Breeder Reactor Development Operation hh%h"p'gongn, "Altemative Electrical Transmission Change No. 9 9%tt3ttquittheir EnvironmentalImpact," NUREG-0316, August 1977. License No. DR-15 l

By letter dated October 16, 1971, you submitted Proposed Change No. 8 to the IrchsucaW2 min @cifications appended to Operating License No. DR-15 for the SEFOR reactor. The proposed change would reduce the number of test e required before i

The ld$ni8dl fMidtEfl8'f6f B6pfiUlfibHVtfi@ ESPE 6ptIIIite criteria is discussed in the following l ParaHPRhaluating your request, we have considered the effect of the proposed change on margins of safety to fuel clad damage.

nsideration of alternatives is the essence of the NEP rocess. The. review conducted c p lp @dwith ggtpfour a:aff gpnpfgf;gthresin.hodificaf

,Sgjg tgi,o ns to theEMWdithyMretHH8pdVo osed ch nge are i ndthuguigans a6ptivena;mstrtissienticpdewsuminetyimthecentrehissurekvimudpapstadysnot t@W8Elihfi4P6ahWds9i6h ble assura ce that the idM ihlfthY d%4MifdalintrihirWiL4setit ddWifiidEosttason- e w" hen compare"d to the propose system.ealth and safetv of the pub'lic will not be endangered, We have redesignated the proposed change as Change No. 9; and pursuant to 10 III. MRyp@gggggtgp3 pal A Specifications appended to Operating License No. DR-15 are clanged as shown on the replacement pages in Attachment A. New requirements are identified by underlining and deleted requirements are identified by lines The gtrinnight objeohmsrafadiis analysis procedure are 1) te pmvide assistance to those ESRP Chapters 4 and 5 reviewers concerned with identifyingjnd verifying

• means to mitigate adverse l impacts associated with the proposed transmission sysddi aEiN)'to identify and analyze reason-l able alternatives to the applicant's proposed system to the ent ne em rpm t ra j an environmental standpoint, as preferable, equivalent, o ,

te y ,ed

's h " Donald J. ov olt Assistant Director for Reactor Operations Division of Reactor Licensing The depth of the analysis should be governed by the nature and magnitude of proposed transmis-sion gtsprggeteg At n bv,the y ESRP Chapters 4 and 5 reviewers. When adverse impacts are predigrfi t&q rgpeagrthqldd;poperate with these reviewers in identifying and analyzing means to mitigate these impacts. The proposed system with any verified mitigation schemes (i.e.cfrieasMdaHd cVHr81H8beUTEM impacts) should be the baseline system against General Elec.tric Compan7 which alternative transmission systems will be compared. The nature and adversity of the DRAFT Rev. 0 - May 1997 9.3.4-6 l _ _ . . -___.._. _ .-. _ _. _ _ -_ . _ . - - . _ . _ . _ . . - -.

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ATTACHMENT A CHANGE NO. 9 TO THE TECHNICAL SPECIFICATIONS PROVISIONAL OPERATING LICENSE NO. DR-15 GENERAL ELECTRIC COMPANY  ;

1 DOCKET NO. 50-231 ,

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3.3 Reactor Core -  ;

Applicability

, Applies to reactor c-ore loading configurations.

~

- Objective To assure that core phycies parameters remain within the expected range and that fuel rod cladding integrity is maintained.

l Specification I l A. The reactor shutdown margin at 350 F shall be equal to or greater 1

{

than 1$ with one operable reflector segment raised to its most l reactive position, and extrapolation of data obtained at or above j 350 F shall demonstrate that the reactor would be suberitical at l-300 F with one operable reflector segment raised to its most reactive position.

B. The excess reactivity available at rated power (20 MWt) shall be  !

j equal to or less than 0.5$ when the core inlet temperature is at l j 700 F and neither the oscillator poison slug nor the FRED poison slug is inserted into the core. The core reactivity shall not be increased by adding fuel rods to compensate for an inoperable reflector segment, C.

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.-The reactor power coefficient of reactivity at constant inlet temperature l and constant coolant flow rate shall be negative.

D. The isothermal temperature coefficient of reactivity at "zero" power shall be negative.

E. Following initial operations at a power level of 10 MWt, the reactor j shall not be operated unless operating data from SEFOR demonstrate that the net.non-fuel coefficient is negative and that the Doppler coefficient (T ) is negative with a magnitude equal to or greater than 0.005.

F. The reactor shall have a phase margin of at least 30 at the point where the Nyquist plot crosses the unit circle.

l G.

1 The reactor shall have at least 600 fuel rods in the core if the scram trip point is set at a power level greater than 1 MWt.

H. Cuinea pig fuel rods of 25.07. plutonium enrichment shall only be located below the six refueling ports, d

Change No. 9 l

3.3-1 November 12, 1971

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No' fuel' rods shall be placed in the center drywell.

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

reactivi,ty worth of less than 60c and containing a total of not more than 1 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

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material may be irradiated at reactor levels above 100 KWt.

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i Fuel rods which have defects as defined below shall not be reinserted in ,

the core: *

1. Cladding rupture, cladding perforation, or other observable defects  !

which may cast reasonable doubt on the integrity of the rods.  !

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2. Local swelling of the cladding in excess of 10 mils or bowing of t

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l the rod sufficient to prevent reinsertion of the rod into the core.

! 3.

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

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

M. If.the gross gamma monitor becomes' inoperable, the reactor shal) be shut

]

l down, except under the following circumstances: l 1,.., 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 indicative 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.

2. When the reactor head is not in place, the reactor may be operated, as permitted by Section 3.9, with the gross gamma monitor not operating.

N. When guines pig rods are located under the innermost refueling ports, the B4 C polson rods in the coro shall be distributed such that the number of L

poison roda in any quadrant of the core (determined by N-S, E-W center-

! lines) does not exceed the number of poison rods in any other quadrnnt by more

!. than two. This specification shall not be applicable when the high flux trip icvel is set more than 10% below the LSSS.

Change No. 9

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3.3-2 November 12, 1971

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(1) observe. the continuous' monitoring system under operating conditions to diagnose the cause of. failure or maloperation of the system;

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(2) permit an orderly completion of a test series, so that tests.

. complet'ed prior to the failure do not have to be repeated;

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

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, During periods of reactor operation when the continuous fission gas monitor is inoperable, batch samples will be taken at intervals of approximately four j hours;; however samples need not be taken when operations are conducted with the rgpetor head removed. This sampling frequency will assure that any trends that l

.might develop will be' identified.  !

Specificatioti 3.3.N requires a reasonably uniform arrangement of B 4C poison rods l in the core to provide assurance that the power density in guinea pig rods under j the inuormost refueling ports is not significantly greater than the value used j to deterinine the LSSS. A uniform distribution of poison rods is desirable for

! most of the planned tests. However, some non-uniform arrangements at' low or l f intermediate power may be required for special tests such as determination of material worths at zero power or determination of available excess reactivity i

during the approach to power. Such arrangements are permissible when'the high j

flux trip level is reduced more than 10% below the LSSS, since'the maximum f') effect of a single poison rod on guinco pig rod power density la only 1/4%. }

The intent of limiting the applicability of this specification is not to permit .

I grossly non-un'iform arrangements of poison rods, but rather to permit the flexf bil-icy of arrangement 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 I

(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) SEr0R 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.

(7) SEFUR FDSAR, Volume I, Para. 4.2.2.2.2 , p 4-2 Change No 9

, 3.3-7 November 12 , 1971

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(8) SEFOR FDSAR, Volume II, Para. 12.2.1, pp 12-3, (9) SEFOR FDSAR, Volume II, Para. 16.2.4.3.

(10) SEFOR FDSAR, Supplement 21 pp 2, 3.

., (11) .SEFOR FDSAR, Supplement 3.

(12) SEFOR FDSAR, Supplement 21, pp 1-17.

• I

! (13) Addendum No. 2 to Proposed Change No. 3 to the Tech'nical Specifications, j January 27, 1971.

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- ~Change No. 9 ,

,3.3-8 November 12, 1971

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i 3.10 Approach to' Power

, Applicability l

Applies to reactor power limits during the initial approach to full power for Cvu I m.J else fer Core II.

Objective To procid T.cthod of :::uri .; a :sf d-eedecly appreach-tMull

-pewer, assure that reactor power is stable, i

Specification

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L A. Conventional ocillator tests shall be nerformed at a nnmfnnl oower level of 10 MWt. If at any power level, the analysis of the con-ventional oscillator tests indicates that the . stability criterion of specification 3.3 F will not be met at some higher level of power, 1

r the reactor power may be raised only as high as the halfway point j

! I between the level at which the test is made. and that at which the failure to meet the specification is indicated.

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one111 ster taats which were nerformed on Core I demonstrated the stability of the rameter. The channes made for conversion ,to Core II are not expected to have a sinnificant effect on the phase u.arninw. Conventional osculator tests will be performed at 10 MWt on Core II to veri'fy that the reactor operates in a stable manner and that the phase marain is arester than the value specified in paraarsah 3.3.F.

Change No. 9 3.10-2 November 12, 1971 b

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• -9 . Propored Change No. 8 f

for the Southwest Experimental Past Oxide Reactor, October 16, 1971, paragraph IV.C.

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Change No. 9 '

3.10-3 November 12, 1971 I l

3 p-- . .w.*e, ...%, .e.. .-%. . - - r- = s m- +4==+= =e= == **~*-ee s=.

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Excursion Tests

. 3.12 Applicability t

These limits apply when excursion tests are conducted with the Fast

-~

Reactivity Excursion Device (FRED). The pr ;t critir,0-1 : st 7::;;r -

3

f th: :p:;iel ;;p;;;; dc;erited in Op;;ifica .isu 5.5." .5 and d;;;rsin:d

.e ;!.cz e s ...:. eJJi:1e.. 1 ayesifle .;.iv... ..e .e3. ired. For Core II.

the transient test program may be initiated following satisfactory completion of static and oscillator tests at 10 MWt. orovided the con-ditions listed in Section 3.3 and Paranraoh 3.12 B.1 have been met.  !

Objective To specify additional limits which are applicable only during the 1

excursion tests, i

i Specifications A. Experimental Program with FRED l

The experiments with the FRED shall be carried out in three phases i .

l as indicated below. Progress to the next phase shall be contingent upon adequate agreement with predicted results.

1. Familiarization Tests

a. The worth of the poison slug used in these tests shall be

( 0.5$ or less.

b. The initial reactor power level shall be equal to or less *

! than 15 MWt.

I f 2. Sub-Prompt Critical Tests

a. The worth of the poison slug used in the sub-prompt critical tests shall be 0.98$ or less. '

l

b. The initial reactor power level for sub-prompt critical tests with the FRED shall be equal to or. less than 15 MWt.

l l . . . ... Change No. 9 . ,

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* . November 12 3,})7L l

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3. Super-Prompt Critical Tasta i

a. The worth of the poison slug used in the prompt critical tests shall be equal to or.less than 1.3$ if the nominal-value of the magnitude of the sodium-in negative Doppler coeffi,cient ( ) is equal to or greater than 0.008. Wi

.:::th Of the pci;;n :,1ug ; hell t; ;@al to s lese ihm. 14 if the namin:1 ;;1u: cf the ec;;nitude cf the : diu in n;gntiOc l'eppic-i cemffimient 'Td)i i'*" h"" C*c

• f If--

the nominal magnitude of the sodium-in negative Doppler Co-efficient is less t_han 0.008 the worth of the poison slug shall be equal to or less than the value determined from the linear relationship in Figure 3.12-1.

b. The initial reactor power level for prompt critical tests shall not exceed 11 MWt.

(-0.008, 1.30) 1.30- -

- C c jl o

O eO v l

Edu '

0:E 1.20 --

% .gj e t.- (-0.005, 1.20)

2. in -

-0.005 -0. rjn8 SODIUM-IN DOPPLER COF.FFICIENT Figure 3.12-1 Limits for Poison Slug Worth e- Change'No.'9 3.12-1.1

. November 12 P1971 l

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1 - - - - - - - - - - -

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Bases The experimental program with FRED is graded so that small transients precede larger transients. The information from the small transients will be used (1) to evaluate the performance of the reactor, (2) to compare the performance with predicted behavior, and (3) to predict performance of the reactor for the' larger prompt critical transients.

The characterization of the tests into the categories of (1), (2), and (3) above I is scif-explanatory. The maximum power levels indicated in each case are to assure that the safety limits as given in Section 2.1 will not be violated and  !

l are in accordance with Figure 2.1-1 and the explanation in Section 2.1. The limits given in this specification also assure that the maximum energy addition to the core during a planned transient does not exc'ed that calculated for the Maximum Planned Transient.( } This in turn assures that the maximum consequence of inadvertently running a transient with a defective (sodium-logged) fuel rod in the core would be limited to deformations corresponding to about 0.6% strain of the cladding of the defective fuel rod.(11) This amount of strain is only 4%

of the minimum ductility of the SEFOR cladding at the end of the three-year experimental program.( )

The value of the Doppler coefficient for SEFOR Core I with sodium in the core is cutimated to be T = -0.0085. This value was verified experimentally by means of Doppler measurements on the SEFOR mockup in the ZPR-III Critical Facility (.l' '

From further measurements on this mockup, it was established that the Doppler co-dk efficient with sodium out is 17.5% lower than the value with sodium in, or T y

-0.0070 for SEFOR Core I with sodium out.

The value of the Doppler coefficient (T ) for SEFOR COPS II with sodium in the core is predicted to be -0.0069.

This value is based on calculardnos_which were normalized to the Core I value of the Doppler coefficient -0.0031.(14) ,

sodium out Doppler coefficient for Core Il is estimated to be -0.0057.

The safety analysis of the MIA for SEFOR was based on a sodium-out Doppler dk coefficient of T g(=)-0.004, which corresponds to a sodium-in Doppler coefficient (T ) of -0.005. The demonstration of a negative Doppler coefficient with a magnitude equal to or greater than 0.005 during the approach to maximum power will verify predicitions of this coefficient based on the ZPR-III measurements and will provide the basis for safe performance of prompt critical tests in SEFOR.

. . Change No. 9

. , . . . - ;3.12-3 November .12 , 1971,s

.--.,.e - - a-- . . - -m=..'.

l.- 1 .

i=

l Tho total reactivity worth of each poison slug used.in the FRED will be known-

!' and the value will be checked before each transient test. The maximum re-activity inserpion rate will be limited to less than 20$ per second by limiting

, the reactivity worth of the slug to 1.3S and by limiting the~ minimum allowable enee for the slug to travel the first 20 inches to 0.097'se'cond.(3) This time

. will be measured by means of the lift-off switch and a proximity switch which marks 20 inches of travel by the poison slug. The safety of the plant has been assessed for a maximum rate of 50$ per second with a sodium-out Doppler coefficient (4)

(Tf) of -0.004.  ;

l l

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l I

l

Change No. 9 3.12-3.1 November 12, 1971-

- - - - - - - + - -

I I

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

i starting from initial power levels' as low as 0.1 Wt is still well below the safety limit.

l-Initial citeckout tests of the FRED af ter it is installed' on the reactor head will be performed with the reactor either sub-critical or at low .

l ,

power level (less' than 0.1 Wt) . The FRED will have a negligible effect on reactivity when it is in a position more than 20 inches i above the core midplana.

The minimum limit of 700'F on the core coolant inlet temperature is-to assure that the total reactivity of the core is maintained at the equivalent of 50c excess at 20 W conditions.I9) At lower temperatures, the excess core reactivity would be higher. The 50c excess limit assures that the Maximum Planned Transient will not be initiated f rom a power ,,

l level in excess of 11 W e, and also limits the final reactor power if the reactor does not scram and the FRED slug remains out of the core following any excursion test.

The requirement for redundant flux monitoring equipment during the transient tests will provide a cross check on data and will also l

assure monitoring of reactor flux throughout the transients while pro-l - viding the flexibility to use wide range monitors, U-238 fission chambers or gamma chambers as required by the magnitude of each test. (13)

Obtaining overlapping data from both the U-238 fission chambers and the gamma chambers prior to switching-from one instrument to the other for those transients whose magnitude require the use of both instruments will provide a cross check between instrument readings during switchover.

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I Change No. 9 3.12-5<

l November 12 ,.1971' t- . - , ,, r , , , ,

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eye. e == -% ---a=.mee. - - pe e. mear, geessodene as -w een.em e w= w eem . - . w.m~me -ema e-e.- pa ge g.,,r e,

_ _. . _ _ _ _ _ _ . - _ _ . . . ..._ _ . ______-_.m _ _ _ . . - . _ _ _ _ _ _ _ . . _ _ . _ _ _ . - , _ _ .__.

l -

t R*fercnces l (1) SEFOR FDSAR, Appendix B Section B.5, p. B-3.

(2) SEFOR[ FDSAR, Section 16.4.2.6.1.1, p. 16-28.

l

. (3) Proposed Change No. 6 for the Southwest Experimental Fast 0xide .

Reactor August 5, 1971.

l (4) SEFOR FDSAR, Volume 11, Section 16.2.7.

(5) SEFOR FDSAR, Volume II, Section 16.2.7, p.16-10.

(6) SEFOR FDSAR, Supplement 17, p. G-1.

l (7) SEFOR FDSAR, Supplement 3, Section 5.1.3.

(8) SEFOR FDSAR, Supplement 19, p. 57.

l (9) SEFOR FDSAR, Volume II, Section 12.3.6, pp.12-15. -16.  !

l l

(10) SEFOR FDSAR, Supplement 10, p. 1-48.

(11) Additional Information Regarding Sodium Logging of SEFOR Fuel Rods, l l

February 1, 1971. )

l (12) SEFOR FDSAR, Supplement 21, p. 4.  !

i l (13) "Results of the Familiarization and Sub-Prompt Critical Transients

! for Core 1," pages 4 and 18, submitted to the AEC-DRL on July 16, 1971, j by the Breeder Reactor Department of the General Electric Company. l l (14) Op. Cit., Reference 13, p. 3.

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Change No. 9 l

3.12-6 l l November 12 , 1971 '

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I l 4.3 Reactor Fuel Rods t

, Applicability l l  ; Applies to fuel rod examination made in the refueling cell.

I ,

Objective  ;

To assure maintenance of fuel rod cladding integrity during reactor

! operation.

Specification l A. At the intervals specified below, tre er = r: guin : pig fuel rods

^ich 52 :: :p: ret:d et ;:::: dca:itic; 'igher than the p _:r danzity l cf :::nd:rd f :1 : d  :::::t th ::nter of t?: = re, ch:11 b: r:= .cd frem th e reseter efter operatier : rec:ter pcuer 1:; 1 cf 15, 17.5, l

i nd 20 MUt, ...d shall be examined in the refueling cell by visual observation, deminsional checks, and gamma scans. '

..fter ::::hing ; ,

1

r 1;V;l of 15 MUt ;;d befer; ;;;;hing 17.5 MUt, the interval j h a t;;;.- fuel rad cx;;inction; chall not ex

cd sin

nths.

B. Before the start of the sub-prompt critical excursion tests and before the start of the prompt critical excursion tests, a mintmum 'of one i

j guinea pig fuel rod and one standard fuel rod shall be examined by the '

methods described in "A" above. If guinea pig rods are located under l

inner refueling ports, one of them should be the one examined.

C. Af ter each every other prompt critical excursion test, at least one guinea pig rod and one standard rod shall be examined by the methods described in "A" above. If guinea pig rods are located under inner refueling ports, one of them should be the one examined.

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. Fell ring th cx;;inction after operation at 20 MUt : specified in Wkr&dk, two cr amre (if ;ccilab1 ) guinea pig fuel rods, which have operated at power densities higher than the power density of standard fuel rods nearest the center of the core, shall be examined by the methods described in "A" above at intervals such that the rod exposure during the first interval does not exceed a core integrated power Change No. 9 4.3-1 November 12, 1971

Q. Samples of the primary coolant shall be taken for analysis, at intervals l not to exceed three months and following eeek every other prompt-critical l FRED transient.

. R. Three re' actor vessel outer head bolts shall be removed annually for ,

visual inspection, dimensional checks and dye penetrant testing. l S.

Five redctor vessel outer head bolts shall be inspected annually using ultrasonic testing with straight beem longitudinal scan from the top of the bolt. This inspection may be made with the bolts in place or removed from the, reactor. Bolts to be examined shall be selected so that all of the outer head bolts are examined within a 10 year period.

i 4

Change No. 9 -

4.4-2.1

'" ". November 12 , 1971 '  ;

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• '~~~

b5 6.6 Plant Operation Report's l ,

Applicability Applies to the preparatioh of a Quarterly Plant Operation Report and

. reports of experimental r'esults.

l- ,

Objective j To provide a timely summary of all aspects of facility operation and projected reactor experiments.

Specifications

/

! A. A Quarterly Plant Operation Report shall be prepared and submitt.ed l

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to the Director, Division of Reactor Licensing. The report sp'all l generally contain the following sections in the order listed:

1. Summary of plant operations.

2. Listing and classification of all plant shutdowns.

3. List of all " abnormal occurrences".

4. A detailed summary of changes in cover gas activity and any

,r g investigations involving suspected f ailed fuel.

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5. Tabulation of all major items of plant maintenance,

6. Tabulation of major items of instrumentation and control work.

/

j 7. Sum: nary of required surveillance tests and results.

l 8. Summary of health and safety operations including a summary l of all radioactive releases and shipments.

-9_ Discussion and justification of significant changes in plant l design.

10. Tabulation of significant changes in plant operating procedures related to transient tests.

11. Current schedule for planned transient experiments.

l B. The following special reports shall be submitted in a timely manner.

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Nov.e.mbe r 12,,1971

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g _,. , -- _ , . .m,.sc#_J J_ r.__ 1__ *3 1S i r------- - - - - - - - - - -~a 1, . -6 . A report which identifies changes made to the reactivity equation as required by Specification 3.13.A.3. This report shall describe i

the reasons for the changes and the bases upon which continued

] reactor operation is deemed to,be safe, a

Change No. 9 6.6-2 November 12 , 1971 4

_---