ML19326C382

Concurs W/Encl Tech Specs Change & Comments on Safety Evaluation

ML19326C382
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/30/1975
From:	Culp R NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD)
To:	Anderson F NRC COMMISSION (OCM)
Shared Package
ML19326C376	List: ML19326C375 ML19326C379 ML19326C381 ML19326C382 ML19326C384
References
NUDOCS 8004220921
Download: ML19326C382 (1)

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April 30, 1975 Note to Fred Anderson AMENDMENT 2 TO TECH SPECS. AN0-1 We are returning, with our concurrence, the ANO-1 Tech Spec change package with the following comments on the SER:

olC 1.

There are two page 3's.

0 $ 2.

On page 3 under Evaluation, after "added by the Staff" insert "and agreed to by the Licensee."

6 $ 3.

Page3,(3). Please explain further the basis for your conclusion that " Deletion of the CFT pressure and level instrumentation

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from the maintenance aspect of Specifications 3.3.5 does not affect the safety of the system or reactor operations and therefore is acceptable. "

oK 4.

Page 3, (6).

Identify the document where steam generator tube rupture and loss of load incident were previously analyzed.

t/q 5.

is there any discrepancy between your statement on page 4 (6) that the steam line break accident outside containment is more probable than a loss of coolant accident and the statement in the basis of Tech Spec 3.10 referring to the "less probable accident of a steam line break."

p.Page4,(8),Line5.

Change "eveery" to "every".

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ATTAQMENT TO LICENSE AMENDMENT NO. 2 CH.*NGE NO. 2 'IT) 'IllE TECilNICAL SPECIFICATIONS

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FACILITY OPERATING LICENSE NO. DPR-51 E

i' DOCKET NO. 50-313 i-Delete pages 13, 14, 15, 16, 19, 20, 23, 24, 37, 38, 39, 48, 48e, 48f, 60, 66, 67a, 68, 71, 72, 73, 73a, 74, 75, 76, 83, 84, 100a from the Appendix A Technical Specifications and insert the attached replacement pages 13,14,15,16,19, 20, 23, 24, (27aN 37, 38, 39, 42a, 48, 48e, j..

48f, 60, 66, 67a, 68, 71, 72, 73, 73a, 74, 75, 76, 83, 84, 100s. The

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changed areas on the revised pages are shown by a marginal line.

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Othsr ch:nn21s tre subject cnly to " drift" crrors induced within the.

Instrumentation itself and, consequently, can tolerate longer intervals

'between calibrations.

Process system instrumentation errors induced by drift can be expected. to remain within acceptable tolerances if re-

. calibration is-performed at the intervals of each refueling period.

Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures.

Rus, minimum calibration frequencies for the nuclear flux (power range) channels, and once each refueling period for the process system channels is considered acceptable.

Testiing On-line testing of reactor protective channels is required once every 4 weeks on a rotational or staggered basis.

The rotation scheme is designed to reduce the probability of an undetected failure existing within the system and to minimize the likelihood of the same systematic test errors being introduced into each redundant channel.

He rotation schedule for the reactor protective channels is as follows:

Channels A,.B, C, D Before Startup if shutdown greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

' Channel A One ifcek After Startup Channel B Two ifceks After Startup

. Channel C Three lieeks After Startup 7

Channel D Four liceks After Startup ne reactor protective system instrumentation test cycle is continued with one channel's instrumentation tested cach week.

Upon detcetion of a failure that prevents trip action, all instrumentation associated with the protective channels.will be tested after which the rotational test cycle is started again.

If actuation of a safety channel occurs, assurance will be required that actuation was within the limiting safety system setting.

%c protective channels coincidence logic and control rod drive trip breakers are-trip tested every four weeks.

% e trip test checks all logic

-combinations and is to be performed on a rotational basis.

The logic and breakers of the four protective channels.shall be trip tested prior to startup and'their~ individual channels trip tested on a cyclic basis.

Discovery /of a failure requires the testing of all channel logic and breakers, after which the trip test cycle is started again, yu Relequipment. testing and ' system sampling frequencies specified in Table 4.1-2 and Table 4.1-3 'Ts considered-adequate to maintain the status of

-l2 the equipment and systems to assure safe operation.

REFERENCE ~

FSAR Sectionc7.1.2.3.4 68

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clarify the intent of sanpling requirenents end censurements.

Changes 6 and 7 vould increase the surveillance require: rents by changing the

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acceptance test Ing f or the personnel batch and e=erpency hatch door seals end bettery chercers.

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EVA!.UATI01:

I Cur eva}urtice p'oUsi M Mf the changes proposed by the licensee and edded by the 7

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(1)

Table 2.3 Cur review of the reactor protection syster (T:FS)

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r.csi'f G~n't'Gn as given in the safety evaluaticn appetided to our letter 5

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deted February 12, 1975, for the shutdown bypass fircuitry rodification

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l coceluded thct the rodification did not affect iny other satety relcted systec, setisfied the requirerents of IEEE Std' 275-1971 end enhanced 3

scfety by replacin;; an cdninistrative cont rol function with cn auto-r.atic cont ret function.

This change to the' technical specificction reflects cc.pletion of this approved ETS riodification and is acceptable.

H (2) rect ion 3.1.4, '7.eactor coolent Systert $ctiv,ity" - Ue performed a reenalysis of the postulated double-ended rupture of a stesa generator tube using current analytical models and ceteorolerical parcreters as discussed in the bases to the rev specifications.

This analysis tes perforced to deternine the :cceptable specific a :t ivity li:-its j

for ratioiodine in both the reacter coolant systen eta secondary w

coolent system.

The specific activity lit,its for the recctor coolent 3

have l'een defined in ter s of mass (grans) rather than volume

. frilliliter) er previously used to elinincte possible error in defining

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te percture and pressure associated with the ecnple volere.

The half-l life lir-itt.t ion valve vs.s de leted free the speciiication since this Mr i

percreter does not chenre the possible exposure fro.1 cloud passage of a niven radicisotopic nivture.

rovever, the r iniruc tire for decay l

enroute frc::' the source to the nearast site-bcundary for the assumed eteorolorical cenaiticos sheuld be consicered during the scnrle anclysis and is discussed Icter.

A requirc ent hcs been cdded to the g

i specificationwhic[specifiesthecetionstobetchen if the cpecific w

activity li.its ste e::ceeded.

Tech recuire: ents were net previously l

included in thefspecificaticn.

'Ibe specific cctivity limit for redio-g~

iodice ucs not/previously defined for the reactor coolant.

These i

. licit t. ore d.(fined for ct eady st ate reacter conditions end do not 5~

j reflect po[sible spikinc conditions :ssccicted with trcnsient reacter E

conditions, fuch conditions are considereo 1cter for surveillence receirerents..The minirun rctio determined between the rcdiciodine specific activity for the reacter coolar.t end the seccndcry coolant i

.tes conservctively ascessed on the basis of the naxieun allovsble

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the half-life of I-131 f or ecuilibriuri conditions.

The actual g;

ratio of radioiodine specific' activity in the reactor coolant to the recondcry coolant would be expected to be significantly greater than the calculated value of 20 to 1.

I'pecifications 3.3.5, 3.3.6 cnd 3.3.7 - The changes delete the (3)

S lT pressure and level ' Instrumentation from the list of systees 4

for which provisions have been made for riaintenance.

The restrie-tiens on this systec cre delineated in specification 3.3.3(D).

Exceptions to Specification 3.3.6 conditions given in Specification 3.3.7 provide adequate relief for perferning necesscryameintenance functions on both the CFT instrument chcnnels and on,.the NiST level instrurent channels.$TAlie' tion of the CET pressure and level

=E-inst rumentction from the reintencece aspect of Epecification 3.3.5 does not effect the scfety of the systen or reccter crerations and therefore is acceptable.

Continued reactor' operation for seven devs with incperable instru,ert channels in the CFT cnd L' ST sys' tens as given in Specification 3.3.7honcistent trith exceptions permitted for instrument channels in siviilar systers and therefore is acepetable.

(1)

Erecification 3.5.1.7 - This added specificction delicestes the N6rcRiaEE$ifiElation velve closure setroints or the suction line to escure proper orcration of the D1 K5 when required and the the 14:F5 relief valve setting necesscry to protect the systes arninst overpressure.

Proper settin::s for these valves would be verified durine the testing and calibrction required by Table 4.1-1.

(5) Ficure Fo, 3.5.2 Ihe chrege in the pervirsive operating region-for power i-balance reduces the allorable operation to be conyctible with the. protective systex rexit u:: allovtble setpoints.

The chance to thir, figure does not chente the ella cble reactor operation since the reccter had to be operated t:itoir the core restrictive licits es'tablished for the resetcr prctective system.

The enan;e is cceep/t/ct ie.

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the stecr cer.orrtor. tube rup tu re,, rs, encl yce. to cre t e rv,ne recet or i

ced reconJcry coolent ccJ ivi t y lir it, a lors M IdfE"ir'ic'ent' c

ce nrevicesly entlyzec.. n e - t-to ceterntue

.6 schondcry system ac t iv it y th

'li-it cne a stenn lire,brech accident cutride certcirrent.

- N ice the secondcry ecclent netivity lic it deterrined f rcy, the r t.an eene rr t or tube rertere as tre::ented ir. t' e 1.w e ter the reeeter ecclstt cycteu cctivity, Fect ice

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.resulting from the other two postulated accidents were deternined.

Since we consider the probability of occurrence for either the

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steam f.enerator tube rupture or the loss of load incident to be comparable, the cecertable thyroid dose linit for either incident was ~ t aken as 1.5 ren.

Le consider the occurrence of a.stean line break accident outside containgi1[Yc0NvI$iisidOF'tlii secdt-to be ecre likely.than,a

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loss of coolant accident [^ hor N iI rea N'

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able thyroid dose limit to be 30 Fen or significantly less then, the ruideline doses of 10 CFR Part 100 As stated in the bases for this section of the technical specificcticns, the resulting thyroid doses using the specified secondary systen activity 'iimit of 0 17 n i/ge of I-131 dose ecuivalent are c

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crpror.imately 1.5 Fem for the steam reneretor tute rupture (as stated in the Iases to Speciticction 3.1.4), 0.6 for the loss of leod gg incident end 22 re-for the steam line bresh eccidert outside con t a inr en t. All of these doses are less thcn the cbove stated F=4 dose guidelines for these accidents and indicate thct the contrcllin"

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accident for deternining the secondcry coc1cnt rcdiciodine linit is the stece generttor tube repture.

1,n increase 'in the ratio of radiciodine specific ectivity for the reactoi coolent to the secondcry coolant vould directly reduce the fliculated close for tbc two accidents involving only secondaryfeoolant re lea se s.

Eovever, no increase in this ratio vould 6ot significertly reduce i

j the eciculated dose for the stern generator tube rut ture thich releases both reactor coolent and se dndary coolent rcdioactivity.

(7)

T.chl.e. 4 1.1.( It.en. 30) cnd Table 4.(1..2 (.I.t.e.n.11..), - rot e 3 ha s Esen

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etanced. to reflect the newly est'ablished setpotnts on the isolction velves of DITS given in Specification 3.5.1.7 and givec the pressure range within vhich the test nurt be perforred.

The test y

vill verify the correct setpointo for the isolction vcives.

The sore chen;'e vas raade 'to the t est f requency colurn in Table 4.1-7 for consistencv.

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9 (F) "Ichte 4.1 The r.ibinun sannling and enclysis frec;oency and tests beve been chenced for the reector coolant savples.

The Cross Activity teterrioction (previously designated es Crosc lete end Cecro Activity)/frecuency hcs been reduced fre-5 tires per veek to 3 tines perrweek and et tenst eve /ry third day.

~his frecuency also has been desirosted for yensuring the Cbe iistry cod Eoron Concentration in the t:ecctor Coolant.

f.r.pe r ienc e has shown that such frec:vencies are adecuate to detect chances in coolent chenistry

u en a tir.elv bacis cud pernits reactor operction over.a veeh-end or bolidev vitheut the need for a reactor coolent sFrle enclyais.

0c~ a Irotopic J.nalysis frec,uency has been increased from conthly

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to bi-5 eekly (once every two weeks) anddhe liadiochectical Analysis

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Q Ecth of these changes in frequency is,c ta detect on a tinely basis

"'"E any chance in quality of the gross radiocctivity contained in

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the reactor coolant.

Experience has shown that such f requencies

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will detect changes in quality of radioactivity due to additional a:s:[

f ailed fuel or chante in reactor operations.

The Cross Radiciodine 99 Deternination has been added to detect radiciodine activity levels in the reactor coolant for compliance with Section 3.1.4 requirements.

The specified frequency for the analysis is veekly but shall be r ore frequent if the gross activity increases by a given *arount as specified by rote 3.

Experience has shown that a t.eekly frequency with this condition for nore frequent analysis is adequate to detect

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on a tieely basis any chances in reactor coelent radiciodine levels.

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A determination of dissolved pares concentration in the reactor is coolant is required by Specification 3.1.9.1 chich pieces a limit R

of 100 sto cc per liter of water of dissolved reses in the reector coolcnt for control rod operation.

The buildup of dissolved Tcses in the reactor coolant is a slow process and therefore, deekly detereinction of this parat eter is considered adequats for tirely detection of any unusual increases.

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The existinr recuire-ents f or determining t ritiu.~ concentrzt ic n, l

Sr-S9 and Sr-90 concentrction end gross alpha activity in the t

recctor coolant are not recuired because t'here are no liaits

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necescry for these rcdioisotopes for operating of the facility.

The existing reouirerent for detereininc gross beta-pcma' activity in the secondcry coolant N d not receired because such nctivity vould not be present and no limite on operation for gross activity are required.

The re fore, these recuirerent s hcve been deleted f rom the trble. Analyses for' radioactivity levels and dissolved coses tre net required nen the plcnt is in the cold or ref uelina ehet 6 condition because thece perc-eters do net af fect the setety of the plant vhen in these shutdorn conditions.

Thus', tot e 7 provides for these exclusions.

To deternine the level end duration of possible rcdiocctivity spihing for both prost, snd ictice retivity, ndditional an ' lysis dependin; upon level et radiocetivity during the previous stei.dy st cte operat ion at the tine et resctur startt:r

- cnd duriny reacter startup is recui. red by rote 6.

This note cpr. lies only' to tt;e r;reca determinctient of totcl activity end

- radiciodin..

rote 1 basically renains the snre or previously steted for t*iir table excen thet the increased freceercy of ennlysis is receired ocly until stecdy state ectivity level is establiebed.

As discussed in f.ete I rnd f ote 2, the rress activity deter-incticy orricr>

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vill be based on the activity present 15 rinutes ef ter sanpling.

This tire period is equivalent to the minircum expected decey time

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fro:r the release point to the nearest :ite boundcry in case of a steen Fenerator tube rupture.

!;ote 2, which is associat2d with

. EE the deterraination of f, specifies the t'ethod to be used for the 52

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detern~netton of E, the freeuency of deterrination for E and radio-iodine, and the reference (or equivalent) source to be used to deternine the individual per:-a and beta energies per disintegration for the r'adioisotopes present in the reactor coolant.

!:ote 4,

to this sa:e se=ple determinction indicates that all r diciodine activit ies (I-131,132,133,134, and 135) are to be reighted to detere:ine the I-131 dose equivalent activity actually present in the reactor coolant for comparison with the linit established in

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Specification 3.1.4.1.b.

Ucte 8 states that the 0, cualysis is

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oct recuired when the plant is in the cold or refuUint shutdovn condition for. the serre reason ss gi ren for rote 7 acceptebility.

Note 9 states that a detereincticn of boron concentration in

.gc the Spent Fuel Pool is recuired only if fuel is presert in the pool and prior to fuel beia; transferred to the pool.

Since the boron in pool voter is to ascure that the fuel renains sub-criticci, it is not recuired when there is no fuel in the pool; therefore. Pote 9 is acceptable.

Notes 5 cnd 10 apply to the secondary coolant sampling and analysis program.

1:ote 5 recuires additienci senpling cnd analysis if the priricry to secondary leakage increcces significently.

Uote 10 elicinates the require-rent for sanpling and enalysis of the secondary systen coolant when steam generation is not occurring.

If steam is not being renerated, the postulated eccidenth, rhich established the secondary coolant cetivity licit, could not occur or would not result in cny sipnificant release of radiciodine to the environs frem the ryecondary coolant.

Fote 4 airo rppli s to the cecondcry ceolcet rcdioiodine activity deternination for tssesnin; the I-131 dose equivalent ectivity for conparison with the lirit established in Specification 3.10.

rote 7 elso applies to the secondcry coolcot scrpling and analysis when the pier.t is in the cold cr refueling shutdovn condition.

(9)

_Spe. c._i fi_c_c_t io_.n. 4. 4. 1 _2. 5( b_) - To rahe the test ine requirerent s 1-en the rerscnnel hatch cod eterrency hatch door sects consistent with /prendix J of 10 CiE Fort Sc, the first sentence of this E

specificctien ucs rodified.

The cddition of the phrase ' when reacter E='

buildine interrity is recuired satisfies I.ppendix J. 10 CEE 50,

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receire-ente provined the existing phrase, "but no ecre frecuently than daily duriorm nortc1 operct ier" is deleted.

The existine recuirerent for s eehly testic; durinr refueling or colo shutdowns cmcc>...........;........

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is not necessary for' the-proposed specification which limits tecting requirerents to tims when building integrity is required.

These changes cre acceptable and were made.

.=z (10) ' Sygei fica tion' 4.6. 2.l4 - The existing testing requirerent relating to the third battery charger hat been expanded to apply.,to all

three battery chargers to assure 'that adequate surveillance of all battery chargers is provided.

The additional, surveillance 7

requirerent-is considered appropriate and accepteble to the staff and should increase the reliability of the station battery system.

(11) Appropriate changes in the Bases were made f.lariitcation, but they do not affect the specifications goserning operation of the facility.

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Le have concluded, basec on the considerations discuseed above, that:

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~(1) because the change does not involve a significant increase in the probability or consecuences.of cecidente previously considered and does not involve a significant decrease in a safety carrin, the chante does not _ involve a signifier.nt hazards consideration, (2) there is reasonable assurcuce that the health and safety of the public will not be endancere d by operetion in the proposed ranner, cnd (3) such activities will 1.e conducted in conplicnce with the Comission's regulat ions and the issuance of this arendtent vill not be ini:uical to the.co=on -defense and security or to the health end safety of the public.

Date:

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