ML20151S307

Forwards Revs to Tech Spec Upgrade Program Draft Specs, Providing Clarifications & Corrections,Per

ML20151S307
Person / Time
Site:	Fort Saint Vrain
Issue date:	08/05/1988
From:	Brey H PUBLIC SERVICE CO. OF COLORADO
To:	Calvo J NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation
References
P-88287, NUDOCS 8808150145
Download: ML20151S307 (34)

Text

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• CN 2420 W. 26th Avenue, Suite 1000, Denver, Colorado 80211 August 5, 1988 Fort St. Vrain Unit No. 1 P-88287 U. S. Nuclear Regulatory Comission ATTN: Document Control Desk Washington, D.C. 20555 Attention: Mr. Jose A. Calvo Director, Project Directorate IV Docket No. 50-267

SUBJECT:

Technical Specification Upgrade Program (TSUP)

Revisions to Final Draft

REFERENCES:

1) PSC letter, Brey to Calvo, dated 5/27/88 (F-88184)

2) PSC letter, Brey to Calvo, dated 6/14/88(P-88205)

3) NRC nemorandum, Heitret to Calvo, dated 7/26/88 (G-88292)

Dear Mr. Calvo:

Attached are revisions to the Fort St. Vrain Technical Specification Upgrade Program (TSUP) draft specifications that were previously submitted in References 1 and 2. These revisions provide further clarifications and corrections as discussed with the NRC on July 13, 1988 (Reference 3). Included with the attachment is a tabulation of each affected specification and a brief description of the change.

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, P-88287 t Page 2 August 5, 1988 i i

If you have any questions regarding this information, please contact j Mr. M. H. Holmes at (303) 480-6960. 1 Very truly yours, k N ew H. L. Brey, Manager Nuclear Licensing and Resource Management HLB /SWC/ lmb Attachment cc: Regional Administrator, Region IV ATTN: Mr. T. F. Westerman, Chief Projects Section 8 Mr. R. E. Ferrell Senior Resident Inspector Fort St. Vrain l

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's ATTACHMENT TO P-88287

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1 s

As a result of a telephone conference with the'NRC on July 13, 1988, the following changes have been made to the FSV TSUP draft submittals dated May 25, 1988 and June 13, 1988: (the asterisked margin marks distinguish the most recent revisions).

Specification -Descriotion of Chance SL 2.1.1 BASIS Editorial clarification'; revised subparagraph #3 to indicate that actual time periods are determined for each P/F RATIO interval during which the limits are exceeded.

LCO 3.1.1, Action i Added Action to declare control rod inoperable if slack cable alarm cannot be fixed within 24 nours, to avoid entry into Specification 3.0.3 and shutdown.

LC0 3.1.6 BASIS Revised reserve shutdown system discussion to reflect maximum temperature defect.

LCO 3.2.6, Action c Editorial clarification; revised to indicate that the fraction of allowable operating time is determined for each P/F RATIO interval  !

experienced during the transient.

Figure 3.2.6-2 Resised to show 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> allowable operating time with P/F of 1.05 to 1.17, consistent with the requirements of current SL 3.1.3.4.

LC0 3.2.6 BASIS Editorial clarification; revised _ consistent with the LCO.

Table 3.3.1-1 Corrected Table 2.2.1-1 reference.

LCO 3.3.1 BASIS Deleted discussion on SLRDIS valves, as this is contained in TSUP specification 3.7.8.

LCO 3.3.2.1 Editorial clarification. rsvised "alterna'.e" to "alternately" -

LCO 3.3,2.2 BASIS Editorial clarification in neader: changed LCO 3.3.2 to LCO 3.3.2.2.

SR 4.3.2.5.1 Editorial correction: changed Tab'e 3.3.2.4 to Table 3.3.2-4 i

SR 4.4.2.2 Editorial clarification; revised Sr-90 surveillance to determine Done dose ecuivalent activit', consistent with the LCO.

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's Specification Description of Chance

-t LC0'3.4.2 BASIS Editorial clarification; revised Basis to discuss bone dose equivalent Sr-90 activity.

~LCO 3.4.3 BASIS Corrected discussions about carbon transport and weight loss allowances from "per year" to "per cycle". This reflects the original intent of the requirements and takes into account the fact that fuel cycles nave historically not been' equivalent to an annual cycle.

LCO 3./.1.5. Added action to avoid Specification 3.0.3 shutdown reouirements in event of inoperability of 2 EES safety valves.

LCO 3.7.1.5 BASIS Revision to clarify boiler feed pump capacity.

SR 4.8.1.1.2.e.3 Clarified that load rejection test involves rejection of single largest load in lieu of 202 kw, in the event that the largest load is not actually 202 kw.

LC0 3.8.1 BASIS Identified that single largest load is circulating water pump.

Table 4.8.4-1 Editorial clarification; abbreviations GTE and LTE were replaced with easier to read synbels for "greater than or_ecual to" and "less than or equal to", resnectively. -

LC0 3.9.6 Revisec applicability to when CORE ALTERATIONS are conducted from the_ refueling floor, to clarify that communications are not required between the control room and the FHM contrcl room when control rods are oaing withdrawn from tne control room, as during normal operations.

. Also, deleted footnote reference from the surveil'ance, for consistency.

LCO 3.9.6 BASIS Revised for consistency witn appl 4cability change.

DF 5.3.4 Added temoerature coefficient concerns as reload reauirements.

AC 6.5.1.6 PSC agreed to consider reinstating the recuirement for PORC to review unplanned radiological releases to the environs, with some reasonable threshold value. Subseauently, PSC nas found examples of several recent plant Technical Specifications that do not include tnis reauirement (River Bend 1, Grand Gulf 1, Nine Mile Point 2. and Palo Verde 1). Based on these examples, PSC oroposes that AC 6.5.1.6 remain as submitted in the May 25, 1988 draft.

Amendment No.

r Page 2 -.

DRAFT AUG 5 888 BASIS FOR SPECIFICATION SL 2.1.1 -

SAFETY LIMIT 2.1.1 limits tne P/F Integral  :

.raction cf Allowable Ocerating Time of tne sammation of a numoer of individual transients. Tne indivicual transients are limited

.by Specification 2.2.6. The EASIS for Specification 3.2.6 is also applicaole to SAFETY LIMIT 2.1.1. Furtner discussion on the reactor core SAFETY LIMIT is provided in FSAR se : ion 3.6.3.

To ensure fuel carticle integrity ss a fission crocu::

barrier, it is necessary to orevent tne failure of_significant cuantities of fuel carticle coatings. Failure of fuel carticle coatings can *esult f rom the migration of tne fuel kernels tnr:0gn :neir coatings. During cower coeration, nere is a temoerature gracient across eacn fuel rod, witn :ne nigner temoerature ceing at tne center of the fuel roc and tne lower temoerature at tne outer ecge of tne fuel rod. In an overtemoarature conci:1on, fuei kerneis can move througn neir coatings in. :nis temoerature gradient, in tne direction of tne nigner temperature.

The reactor core SAFETY LIMIT nas ceer established to ensure tnat a fuel (ernel migrating at the hignest rate in tne core will penetrate a cistance less tnan :ne c mcined thickness of tne buffer coating, olus tne ir.ner is:trocic coating on :ne carticle.

Tne fraction of failoc carticle coatings 11 une core at all times is cetermina0;e oy measureme,*. cf gaseous f i s s i c.,

procu:: a:tivity in tne primary coolant loop.

r A: stateo in LCO 3.2.6, tne Integral Fraction of Allowable Ocerating Times is ceterminec as follcws:

1. The range of oossiele 00WER-TO : LOW RATIOS acove tre limit of Figure 3.2.6-1 is ctvicac i rito intervals, for ease of :alculation.

2. Tne Allowaele Ocerating Tire aco<e tne limit of Figure 3.2.5-1 is ce: ermine: for eacn P/F RAit0 i ntervai from lk Figure . . 2.6-2.

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Amendment No.

• Page 2-3 DRAFT AUG 5 1988 BASIS FOR SPECIFICATICN SL 2.1.1 (Continued)

, 3. For any P/F RATIO transient, the actual transient time period for each P/F RATIO interval during which the limit of Figure 3.2.6-1 is exceeded, is divided by tne

• Allowable Operating Time for that interval.

4. The individual fractions determined in Step 3 above are summed for eacn fuel segment, over its lifetime in tne core. This is the Integral Fraction of Allowable Operating Time wnicn may not exceed 1.0.

APPLICABILITY is limited to cower levels above 15% RATED THERMAL POWER,, in tnat Figure 3.2.6-1 covers only the range of 15% to 100*. oower. Soecification 3.2.4, Core Inlet Orifice Valves / Minimum Helium Flow and Maximum Core Region Temoerature Rise, includes oower levels below 15'. wnere core temoeratures are lower, and also overlaps tne power levels aadressed by this SAFETY LIMIT.

BASIS for Orderly Snutdown Following determination (Soecification 3.2.6 ACTION c.1) that SAFETY LIMIT 2.1.1 has oeen exceeded, shutdown is allowed to l be performed in an orderly manner (20 nours to be in at least SHUTOOWN), thus minimizing unnecessary transient effects on othet plant conoonents. Any 5evere transient that significantly exceeos the limits of S:ecification 3.2.6 would reouire a much faster plant snutcun (Specification 3.2.'6 ACTION b), if it did not result in a scram by automatic response of tne PLANT PROTECTIVE SYSTf.M.

Amendment No.

1 Page 3/4 1-3 SPECf FfCATf 0N LCO 3.1.1 (Continued) )F /4F1' AUG 5 E68

h. With tne r. nock-out cot for tne CRD purge flow lines flooded:

1. Se in at least SHUTOCWN witnin the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, arc

2. Perform surveillance SR 4.1.6.2.d.4

i. With a ' slack caole alarm, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> determine wnether a slack caole condition exists (i.e., a carted cable,  : eta:ne: caole, or failed instrumentation that is inaccessicle for recaia curing operation). If an a :ual slack caole concition exists, ce in at least SHUTDOWN witnin :ne next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the alarm is due to some other Condition, restore the alarm to OPERABLE status witnir. One next 24 nours or ce lare the affectec control rod cair inoceraole and comoly witn tne requirements of )r Action o.

J. The orovisions of See:ification 3.0.4 are not aoolicaole for enanges oe: ween STARTUP, LOW FCWER, and POWER. Prior to entry into STARTUP from SHUTCOWN, all recuirements of this LCO must ce met, without reliance on provisions contained in the ACTION statement,.

SURVEILLANCE REOUIREMENTS 4.1.1 Each control roc cair shall ce cemonstrated OPERABLE:

a. Prior to witocrawal of control rod cairs te acniese ,

criticality (if not oerformec in ene orev'ous 7 cays) cy cerforming a cartial scram test of at least 10 in:nes on all con:rol roc cairs ceing witnerawn. anc verifying trat the extracolated scram time is less tnan or ecual to 152 seconds.

b. At least once oer 24 nours cy: *

1. Verifying :nat all CRD motor temoeratures are less tnan or e;ual to 250 cegrees F.

a) aitn one or more CRC motor tercerature(s) e<:se:ing 215 cegrees ::

) 7ne te cerature of any CRC motor exceecing 215

e;rees : snail ce ecorce:,

T-

Amendment No.

3 Page 3/4 1-44 DRAFT BASIS FOR SPECIFICATION LCO 3.1.6/SR 4.1.6 ___

AUb b@

The reserve shutdown (RSD) system must be capable of achieving reactor shutdown in the event that the control rod pairs fail to insert.

After extended power coeration, the RSO system must add sufficient negative reactivity to overcome the temperature cefect cetween 1500 and 220 cegrees F, the decay of Xe-135, and some decay of Pa-233 to U-233. The ouildup of Sm-149 also adds negative reactivity and is taken into account in reactivity evaluations.

The calculated worth for the RSD system as noted in FSAR Section 3.5.3 is at least 0.14 delta k in tne initial core, and 0.13 delta k in tne eovilibrium cort. Based on calculated excess reactivity data in Table 3.5-4 and Section 3.5.3 of the FSAR, tne maximum allowaole temoerature defect is 0.065 delta k, oer LCO 3.1.5. Full Xenon cecay is wortn 0.032 delta k, per FSAR Tacle 3.5-4 Sm-149 ouilc-uo and 2 weeks of Pa-233 decay are worth aoout 0.007 delta k, and this value increases to aoout 0.024 delta k over several montns, including full Pa decay. Based on 5 the above, tne total reactivity increase for 2 weeks after snutdown is 0.104 delta K. Tnerefore, reactor shutdown is assured for at least 2 weeks using only the reserve shutdown system, in the unlikely event that all control rods failed to insert.

Fu r t.he rmo re , per F3AR Se: tion 3.5.3, tre acrth of the RSD system with the maxinun worth RSD unit inocart.cie is at Icast 0.12 delta j k i r. the initial core and 0.11 delta L it tne equilibrium co e.

This is sufficient to ensure snutdown curing the first 2 weeks of Da-233 cecay.

Generally, incoeraole RSD units are capable of ceing restored to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. However, in the unlikely event that an in:perable RSO unit ca,not ce restered to OPERABLE witnin tnis time, tnere is accouat9 We (at letc; 1s days due to the slow Pa-233 cecay as ciscussec in :n1 SAS'.S for Soecification 3.1.3) following a snutcesn usinc tne 950 syste9, to allon- for corrective action of enangirg aut a CRD asse oly. A se ra RSD unit is consicered available if it is on site.

_ Amendment No.

Page 3/4 2-26 DRAFT

'AUG 5 1988 SPECIFICATION LCO 3.2.6 (Continued)

c. Determine tne P/F Integral Fraction of Allowable Operating Time.

As soon as practicable, but no more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after any indivicual transient with the maximum POWER-TO-FLOW RATIO (P/F RATIO) exceeding Figure 3.2.6-1, for each fuel segment witnin the core:

1. Determine tne fraction of the Allowable Operating Time specifiec in Figure 3.2.6-2, for each P/F RATIO experienced curing tne transient (or tne transient may ye be divicec into smaller P/F RATIO intervals), as follows:

a) P/F RATIOS Less Than or Equal to 2.5 and above the g limit of Figure 3.2.6-1:

For eacn P/F RATIO interval experienced during the transient, tne fraction of Allowable Operating Time snall be the transient time that the P/F RATIO is wi '.h i n tne bounds of the P/F RATIO g interval, divided by tne Allowable Operating Time per Figure 3.2.6-2 (for each interval) based on tne maximum P/: RATIO experienced during the interval, b) P/F RATIOS Greater than 2.5 and Less Than or Equal l to 15:

Tne fraction of Allowabie Operating Time for this P/F RATIO interval experienced during each transient shall.ce that time ceriod from tne point onere the P/F RATIO exceecs tne limit of Figure 3.2.6-1, until it crops below 2.5, not including tre first 100 seconcs, divicsd oy the Alloaaole Cecrating "fee for tris P/F RATIO interval (cer

Figure 3.2 6-2). Tne calculation of additional fractions for P/F RATIOS less than 2.5 are given in c.1.a acove.

l

2 Amendment No.

., Page 3/4 2-27 DRAFT AUG 5 1988 c) P/F RATIOS Greater Than 15: j )t.

The fraction of Allowable Operating Time for this P/F RATIO interval exoerienced during each transient shall be that time period from the point wnere the P/F RATIO exceeds the limit of Figure lk 3.2.6-1, until it drops below 2.5, not including the first 60 seconds, divided by the Allowable Goerating Time for tnis P/F RATIO interval (per Figure 3.2.6-2). The calculation of additional fractions for P/F RATIOS less tnan 2.5 are given l*

in c.1.a above.

2. Determine tne P/F Integral Fraction of Allowable Ooerating Time by summing tne fractions of Allowable

. Operating Time for each P/F RATIO interval determined acove. accumulated over tne lifetime of each fuel segment vitnin the core.

3. Verify tnat the P/F Integral Fraction of Allowable Onerating Time is lest tnan or equal to 1.0, consistent with tne Reactor Core SAFETY LIMIT cf Soecification 2.1.1.

SURVEli. LANCE REQUIREMENTS __

4. 7. 6 a The P/F RATIO shall ee determine: to be below the curve of N Cigure 3.2.6-1 at least once per 12 nours. .

b. Witnin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after any oceriting transient wnere the P/F RATIO exceeds the limit of Figure 3.7. 6-1. deternine .lx .

tne P/~ Integral Fraction of Allewable Ocorating Time per Soecification 3.2.6, ACTION c. ,

, c. At least once oer 7 cays, cetermine tne P/F Integral Fraction of A'lo-aole 0:erating Time oer Soecification 3.2.6, ACTION c.

s Amendment Mc.

Page 3/4 2-29 2 miat" DRAFT U AUG 5 1988 15 -

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Alloweble Operating Time (Hours)

ALLOWABLE OPERATING TIMES WITH POWER - TO - FLOW RATIOS EXCEEDING FIGURE 326-1 1

1 Figure 3 2.6-2

Amendment No.

a Page 3/4 2-31 DRAFT AUG 5 1988 EASIS FOR SPECIFICATION LCO 3.2.6/SR t. 2.6 (Continued)

The measured INDIVIOUAL REFUELING REGION OUTLET TEMPERATURE for :ne nine regions with tneir ' orifice valves most fully closed and- all regions with control rod pairs inserted more tnan 2 feet. was assumec to ce not more tnan 50 cegrees F greater snan the CORE AVERAGE CUTLET TEMDERATURE, consistent with Specification 3.2.2.a.1.a. Tne measured INDIVIOUAL REFUELING REGION OUTLET TEMPERATURE for tne remaining core regions was conservative?y assumed to ce uo to 200 degrees F creater  : nan tne CORE AVERAGE OUTLET TEMPERATURE, Isoecification 3.2.2.a.2 and Figure 3.2.2-1). A measurement uncertainty for tne core region outle; temoerature of plus or minus 50 cegrees F was asseed, and 5% uncertainty in _ flow rate measurement anc a 5% uncertainty in reactor THERMAL PCWER measurement were assumed in estaolisning :ne limit consisten:

witn FSAR See:fon 3.6.7. Tne 9 5'. confidence interval on exoerimental cata was used in tne most conservative manner ::

determine tne rate of migration of tne fuel. kernel as a function of. tne fuel kernel temoerature and tne average temoerature gradien; across the fuel kernel.

For the ::tal fuel lifetime in the core, cased on calculation incorocrating plant carameters and 'uncertairties aopropriate for longer time, migration of :ne fue' carticle kernel througn its coating would be less than 20 mic-ons for the fuel with tne most camaging temoerature nistory, and with the core coeratec constantly at any of tne ?^aER-TO-FLOW RATIOS and power comoinations snown on the curve of Figure 3.2.6-1. Out of a total inner coating tnickness of 70 microns, only 50 microns nave oeen used for :ne determination of fuel particle failure in estaolishing :ne limit curve in Figure 3.2.6-2.

Determination of Intecral Fraction in ACT 0N c.2 ine Integral Fraction of Alicaabie Coerating Times is ceterminec as folicws.

1. Tne range of cossible POWER-TO-FLCW RATIOS aoove tne limit of :icure 3.2.6-1 is diviced into intervals, for ease of calculation.

2. The Alio-aole Ocerating Time aoove tne limit of Figure 3.1.6 .' 4 5 ceterminec for eacn D, F RATIO interval from k Fi;ure 3.2.6 .

Amendment No.

,. Page'3/4 2-32 DRAFT AUG 5 1988 BASIS FOR SPECIFICATION LCO 3.2.6/SR 4.2.6 (Continued).

3. For any .P/F RATIO transient, the actual transient time period for each P/F RATIO interval during which the -limit K of Figure 3.2.6-1 is exceeced, is divided by the Allowable Operating Time for tnat interval.

4. The individual fractions determined in Step 3 above are

)

summed for each fuel segment, over its lifetime in. tne j core. Tnis is the Integral Fraction of Allowable Operating Time which may not exceed 1.0, per SL 2.1.1. i BASIS for DOWER-TO-FLOW RATICS Less Than or Equal to 1.17 For an individual transient witn a maximum POWER-TO-FLOW RATIO aoove the curve of Figure 3.2.6-1 and less than or equal to 1.17, a 30 minute limit has been established from an operating viewpoint as acequate for reactor ooerator action. This provides sufficient conservatism since Figure 3.2.6-2 allows a l total of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> for tne Integrated Operating Time of all l

l such transients. If tne transient is not reduced below Figure 3.2.6-1 within 30 minutes, an orderly reduction in power to at least STARTUP is aporopriate.

BASIS for POWER-TO-FLOW RATIOS Greate- Than 1.17 and cess Than f

or Eoual to 2.5 The minimum time to orevent exceeding the curve of Figure 3.2.6-2 is 2 minutes, which occurs at DOWER-TO-FLOW RATIOS of 2.5.

To reach a POWER-TO-FLOW RATIO of this magnitude through an increase in core power. significant eauipment malfunction or failure, and/or significant ceviations from operating procecures would have to occur.

Tr.e re f o re , a 2 minute limit on incividual transients is sufficiently conservativa. For example. as can be seen from Figure 3.2.6-2, sufficient time (at least 9 minutes) is l available for tne reactor coerator to take corrective action to prevent tne core SAFETY LIMIT from oeing exceeded for POWER-TO-FLOW RATICS less than or ecual to 2.0.

Amendment No.

, .Page 3/4 3-3 s

l 4

DRAFT 1 TABLE 3.3.1-1 (Part 1) M6 5 W

. 1 INSTRUMENT OPERATING REQUIREME.iTS FOR PLANT PROTECTIVE SYSTEM. SC;AM i TRIP ALLCWABLE NO. FUNCTICNAL UNIT SETDOINT VALUE la. ."anual Scram Not Applicaole Not.Applicaole (Control Room)

16. Manual Scram Not Aoplicaole Not Acolicaole (Outside Control Room)

2. -Startup Cnannel-Hign 5 S.3E+04 ps 5 9.3E+04 ces Coun Rate 3a. Linear Cnannel-Hign ---------See Table 2.2.1-1-------

Channels 3,4,5 (Neutron Flux) 3b. Linear Cnannei-High ---------See Table 2.2.1-1-------

-Channels 6,7,3 (Neutron Flux) 4 Primary Coolant Moisture

-Hign Level Monitor 5 60.5 degrees F 5 62.2 cegrees -

cewooint cewooint

-Looo Monitor 5 20.4 cegrees F 5 22.1 degrees F cewooint cewpoint

5. Reheat Steam Te :erature $ 1055 cegrees F ; 1067 degrees F

-Hign 6 ~. Primary Coolant Pressure ---------See Table 2.2.1-1------- 9E

-Programmec Loa'

7. Primary C:o: ant Pressure ---------See Tacle 2.2.1-1-----+-

-Programmed Hign lt

3. Hot Reheat heacer Pressure 2 40 osig  ; 43 psig

-Low

9. Main Steam ;ressure-Low  ; 1529 osig 31517 osig

10. Plan Ele::-i:ai System-Loss > 275V 266V

[31.5secencs }35seconcs

11. Two L0co Trow::+ NO: Acolicaole Not Applicable

12. High Reactor E. 'a-  ; 161 cegrees F < 166 cegrees ?

Temperature (;i ce 'avity) -

Amendment #

Page 3/4 3-36 i

DRAFT BASIS FOR SPECIFICATION LCO 3.3.1/SR 4.3.1 (Continued)

The ACTION statements for inoperable SLRDIS detection and information crocessing eouipment allow one cnannel in eacn building to be inoperable for up to 7 cays; a seconc inoceraole channel in eitner ouilding requires that cower be reduced to below 2*; witnin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The 7 day ACTION time for a single detector cnannel is acceptable based on preservation of a 2 out of 3 coincidence cetection system still in operation. ACTION 3 is applicaole to otner _ functions within the SLRDIS instrumentation canel sucn as loss of power from instrument buses, or otner failures in the logic trains and associated electronics. A 12-nour time ceriod in ACTION 3 for inoperability of those associatec SLRDIS functions minimi:ed the time trat SLRDIS may coerate witn limited fanctional capability, Steam Leak Cetection in :ne Turbine Building is reauired for l*

eouipment oualification of SAFE SHUTDOWN COOLING Systems. Tnus, tne limits anc 2 ASIS are tne same as discussed in the BASIS for steam leak cetection in tne reactor building.

Rod Withdrawal Prohibit Incuts The termination of control rod withdrawal to prevent further reactivity addition will occur with the following conditions:

Startuo Channel - Low Count Rate Start-uo Channel - Low Count Rate is proviced to orevent control roc cair withcrawal anc reactor startup ithout adeouate neutror flux indication. The TRIP level is se'ected to be aoove the backgrounc noise level.

Linear Cnannel - Low Power RWP Linear Channel ( 5*. RATEC THERMAL PCWER) cirects tne reactor operator's attention to eitrer a cownscale failure of a cower range cnannel or improcer oositionir; af the Interlock Seouence Switen. (FSAR Sections 7.1.2.2 and 7.'. 2.3)

Linear Channel - High ?ower RWD Linear Channel (3C'. RATED THERMAL PCWER} is provided to prevent control roc cair withcrawal if reactor cower exceeds tne Interlock Seouence Switen limit for LCW PCWER. (FSAR Sections 7.1.2.2 arc ' . 2.3)

The specifiec survei.ance cneck anc test minimum frecuencies are based on estaolisrec 4 custry cractice and ocerating experience at conventional anc r : ' ear cower olants. The testing is in accorcance with tne IEEE Criteria for Nuclear caer Plant Protection Systems, and in accorcance aitr accectec incustry stancarcs.

Amendmont 0 Page 3/4 3 'l7 DRAFT BASIS FOR SP:CIFICATICN LCO 3.3.1/SR 4.3.1 (Continued)- 5E Calibration frequency of tne instrument channels listed in Tables 4.3.1-1 tnrough 4.3.1-4 are civiced into tnree categories: 1) passive tyoe indi:ating cevices tnat can ce comoared witn like units l on a continuous easis; 2) semiconductor cevices and cetectors that may crift or lose sensitivity; anc 3) on-off sensors wnien must ce tripoec cy an external source to cetermine tneir setooint. Drift tests ey G4 on transcucers similar to :ne reactor pressure transcuters (FSAR Section 7.3.3.2) incicate insignificant long term drift. Inerefore, a ence oer REFUELING CYCLE calioration was selected for cassive cevices (:nermoccuoles, cressure transducers,  :

etc ). Devices in:orocrating semiconcu: tors, carticularly  !

amplifiers, will be also :aitoratec on a once oer REFUELING CYCLE '

basis, anc any cri': in resconse or Distacle setpoint will ce discovered from the'tes: crogram Orift of electronic apparatus is not tne only consiceration in cetermining a calibration frequency; for example, tne Onange in cower cistricution and loss of detector cnamcer sensitivity recuire tnat tre nuclear cower range system oe calibratec every montn. On-off sensors are caliorated and tested on a once per REFUELING CYCLE oasis.

Tne surveillance reavirements for the Steam Line Rupture Detection / Isolation System instrumentation in Table 4.3.1-3 incluce orovisions for CHANNEL CHECK, CHANNEL CALIBRATION, CHANNEL FUNCTIONAL TEST and an ACTUATION LCGIC TEST. The frequency of CHANNEL CALIBRATION, at least once oer REFUELIN3 CYCLE, not to exceed 18 months, is consisten; witn tne interval for testing and calibrating similar dete::o-s (neat sensitive cabling usec for fire detection).

Tne manufacturer of :ne instrumentation recommencs an 13 month interval for test /calioration of the electronics portion of the Steam Line Rupture Detection / Isolation System, tnus, ne CHANNEL FUNCTIONAL TEST is soecifiec for tnat interval. Tne ACTUATION LOGIC TEST verifies oroper operation of :ne SLRDIS Cetection and Logi: Racks from a simulatec rate-of-rise inout signal tnrougn and ln:luding actuation of ne cutout logic relays. Time resconse of :ne SLRDIS Detection and Logi: Racks is ve-ified to ce ecual to or less tnan 7.1 se:0ncs as assumec in :ne nign energy line creaA analysis. Tne potential for an inacverten; a::vation curing testing suggests tna logic testing De performec only onen :ne clant is in SHUTOCWN. Thus, the surveillance reovirements are soecified for REFUELING out not to exceed 18 montns. Ine SLROIS control unit incluces a suo2rvision system nat con-inuous'y enc automatically monitors critical circuitry anc internal 00moonents, anc alarms SLRDIS troucle conditions to : e ::erators.

Tests anc caliora: -s o# instrument :nannels in Tables 4.3.1-1 tnrougn 4.3.1-4 a: :+ :er#orme witn eitner internal or external test signais. .3e  : :ne internal tes signal is oreferrec, ahile ecuivalent exterra' te n s';ca's are e:ually a::sotaole.

Amendment No, r Page 3/4 3-38 4

DRAFT INSTRUMENTATION AUG 5 1988 3/4 3.2 MONITORING INSTRUMENTATION ANALYTICAL MOISTORE MONITORS LIMITING CONDITION FOR OPERATICN 3.3,2.1 ihe following analytical moisture monitors shall be OPERABLE:

a. Upon entry into anc operation in STARTUP from SHUT 00WN, two anaiytical moisture monitors (or alternately, PPS llM-cewpoint moisture monitor (s) placed in the "Indicate" mode),and l

b. Upon entry into and operation in STARTUP from LOW POWER, one analytical moisture monitor (or alternately, a PPS ll t dewpoint moisture moni?.or placed in the "Indicate" mode). l APPLICABLILITY: STARTUP ACTION:

a. Upon entry into anc operation in STARTUP from SHUT 00WN:

1. With only cne moisture monitor" OPERA 9LE, restore a seconc monitor to OPERABLE status or De in SHUTOCWN or LCW POWER within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. l

2. With no moisture monitors" OPERABLE:

a) Restore one monitor to CPERABLE status or be in SHdT00WN or LOW POWER witnin the next 90 l minutes, anc b) Restore a seconc monitor to OPERABLE status or be in SHUT 00WN or LOW POWER within 12 nours of l tne first monitor being mace OPERABLE, -

c. Ucon entry into anc operation in STARTUP from LCW POWER, witn ao moisture monitors

• OPERABLE, restore one monitor to 00ERAELE status or be in SHUT 00WN or LCW POWER within 90 minutes.

l A PPS cewooint moi sture monitor ciaced in tne "Indicate" moce can be utili:ec to meet tre intert of Soecification 3.3.2.1 for an CPERABLE analyti:al moisture monitor.

eg .

,s Amendmers No, y i Page 3/4.3-46

@' DRAFT AUG 5 1988 BASIS FOR SPECIFICATION LC0 3.3.2.2/SR 4.3.2.2 (Continued) g

2. Radioactive' gaseous effluent monitoring includes tne following, for snicn recuirements are given ir Seecift:ation 8.1.1. Tnis includes control reem ventilation system recirculation control on. nign radia:ien (FSAR Section 7.3.5.2).

a) ventilation exhaust monitors - RT-7324-1,-2, RT-7325-1,-2, RT-73437-1,-2, RT-4801, RT-4502, RT-4303 b) Gas saste neacer er.naust RT-6314-1,-2

) Secencary coolant air ejector- RT-31193

3. Racica:tive licuic effluent monitoring incluces tne following, for wnien :ne recuirements are given in Soecification 5.1.2 anc Soecification 8.1.3. -

a) Racioactive liauid waste -

RT-6212, RT-6213 l disenarge b) Gas waste comoressor coc'ing - RT-46211,-

a:tivity RT-46212 l

A. Tne seconcary coolant reneat steam ciping monitors (RT-93250-10,-11; RT-93251-10,-11: and RT-93252-10,-11) are inclucec as cart of :ne PPS loop snutcown (Specification 3.3.1).

5. Tne rer. eater / steam generator interscace process monitors (RT-2263 anc RT-2264) nave recuirements as soecifiec in 5:eci fication 3.6.1.5.

c. Tne ac:icent monitoring instru ents incluced in Taole 3.3.2-1 involve :ne nign range reactor evilcing radiation enitor (RT-93250-14). ne reacter clant exnaus filter n Onitor (RT-93251-1), anc tne criticality alarm for tne new fuel st rage cuilcing.

Tne ACTICN statements are consistent fo- comoarable instrumentaticn i tne LWR 5:ancarc Tecnnical 5:ecifications.

The SURVEILL ',CE INTERVA.. s:ecifie: f0- CHa% E. CHECK, CHANNE.

FUNCTICNAL TE5 . ar: C"ANNE. CALIERATION conform to incustry cra:: ice an: :e SL;VE!LLAN^E INTERVALS given in tre Stancarc Technical 5:e: :3:':rs f:- LW;s anc are tnerefore consicere: '

acecuate to +ns.-e : e :-::e ::eration of tnese cete:: ors.

Amendment No,

,.g Page 3/4 3-55 INSTRUMENTATION DRAFT AUG 5 Kh39 3/4.3.2 MONITORING INSTRUMENTATION FIRE DETECTION AND ALARY SYSTEM LIMITING CONDITION FOR C?ERATION 3.3.2.5 The fire cete:: ion instrumentation for each fire dete: tion area snown in Taoie 3.3.2-4 snsll ce OPERABLE.

APPLICABILITY: At all times 1

ACTION: With tne num er of CPERABLE fire cetection instrument (s) 1 l

for a fire cetectice area less tnan tne minimum aumoer  !

OPERABLE recuirement of Table 3.3.2-4:

a. Within 1 nour establish a fire watch patrol to insce t ne area (s) with the inoperable instrument (s) at least once per hour, i b. Restore the inocerable instrument (s) to OPERABLE status witnin la days, or in lieu of any other report recuirec ey Scecification 6.9.1, prepare and sucmit a Scecial Recor to tne C: mission pursuant to Scecification 6.9.2 witnin tne next 30 days outlining l :ne action taken, tne cause of tne inocerablility and I the plans and senecule for restoring tne instrument (s) to CPERAELE status, and l  :. Tne orovisions of Specifications 3.0.3 and 3.0.4 are j no: a;olicacle.

1 SURVEILLANCE REQUIREvENT3 l

4.3.2.5.1 Eacn of :ne recuirec fire cetection instruments listec in Tacle 3.3.3-4 anier are a: essiole curing :lant oceratior l yr snal' Oe :e ons; rate: 0 ERAELE at least on:e ter 6 mon ns by :e-f:r ar:e of a CHANNEL FUNCTIONAL TEST. Fire cete::ces a -4:n are n : a::essicle curing clant oceration snall :+ :e :as:-a:e: 0 ERABLE cy :ne cerformance of a CHANNE. :/, '::NA. EST curing each SHUTOCWN exceecing 24 nours uniess erfer.e: la :ne crevious 6 months.

~..

Amendment No.

, Page 3/4 4-10

'/

DRAFT SURVEILLANCE REQUIREMENTS AUG 51988 4.4.2.1 The primary coolant gross gaseous activity level shall be examined at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by:

a. Use of the gross activity monitor (RT-9301), or

b. If the primary coolant gross activity monitor is j inocerable, by collecting and analyzing a primary coolant samole.

4.4.2.2 The primary coolant gaseous and plateout activity levels shall ce determined to be within the . limits of Specification 3.4.2 as follows:

a. At least once oer 7 days, by collecting and analyzing a grab samole of crimary coolant. This grab sample analysis shall be used to cetermine the following: l

1. E-EAR (See Note l',

2. Curies - MeV/lb,

3. Plateout curies of DOSE EQUIVALENT I-131, 4 An estimate of the circulating DOSE EQUIVALENT I-131, and

5. An estimate of tne Sr-90 cone dose equivalent total [h plateout activity level.

b. If the crimary coolant activity level reaches 25% of the limits of Specification 3.4.2.a, b, or c aoove, at least once oer 24 nours a grao samole of primary coolant shall ce taken ano analyzed oer Scecification 4.4.2.2.a acove. l Normal samole freauency (i.e., at least once per 7 days) may be resumed wnen tne activity level is reducea to celow 25' of tne limits of 5cecification 3.4.2.a. c. or c, or wnen tne activity level reacnes a new equilibrium level, as defined by four consecutive daily samoles wnose results agree within 10'. of tne average of the four samples.

Amendment No.

,_ Page 3/4 4-11 DRAFT SPECIFICATION 4.4.2 (Continued) AUG 5 1988

c. One plateout probe shall be . removed for evaluation coincicent with the second, fourth, and sixth refueling, and at intervals not to exceed 5 REFUELING CYCLES thereafter. If, during the fifth REFUELING CYCLE, or any REFUELING CYCLE following the sixth REFUELING CYCLE, the primary coolant circulating gas activity is greater tnan 7,725 Ci, the olateout probe shall be removed at the end of that REFUELING CYCLE. The probes.shall be analyced for Sr-90 and :-131 inventory in the primary circuit. The results SDall be used to determine the total plateout activity level of Sr-90 bone cose eouivalent and tne circulating activity of I-131 in the crimary circuit. lNe 4.4.2.3 The gross activity monitor (RT-9301) shall be demonstrated OPERABLE:

a. At least once per 31 days, by determining its sensitivity from the grab sample analysis from SR 4.4.2.2.a.

b. At least once per 18 months, by performance of a CHANNEL CALIBRATION.

NOTE 1: Calculations recuired to determire E-BAR shall consist of the following:

a. Quantitative measurement of the radionuclides making up at least 95?. of tne noble gas ceta olus gamma cecay energy in :ne primary coolant in units of C1/lb of helium corrected to 15 minutes after samoling,

c. A cetermination of tne average ceta plus gamma energy cer disintegration of eacn nuclice ceterminec in NOTE 1.a aoove, by aoplying known cecay energies and schemes, anc

c. A calculation of E-EAR by accrocriate weighting of eacn nuclide's ceta anc gamma energy with its concentration as cetermined in NOTE 1.a aoove.

Amendment No.

r Page 3/4 4-12 t .; -

DRAFT BASIS FOR SPECIFICATION LCO 3.4.2/SR-4.4.2 AUG 5 1988 BASIS for Noble Gas Beta olus Gamma Activity Limit The whole body cose is a direct function of the gross gamma activity in the pri.Lary coolant. The cose to tne skin of the whole body is a direct function of tne gaseous beta activity in the primary coolant.

The primary coolant noble gas ceta plus gamma concentration limit, (Soecification 3.4.2.a) is cased on tne Maximum Credible Accicent (MCA) (FSAR Section 14.3), wherein the entire "design" primary coolant circulating gaseous racioauive inventory is carried out of the PCRV and is released to the atmospnere through the reactor building exhaust system.

Correcting the noble gas beta plus gamma activity to 15 minutes after sampling would conservatively indicate the activity that would reach the Exclusion Area Bouncary (EAB), following tne postulated accident, taking into account tne cecay of snort half-life radionuclides during atmospheric transoort to ne EAB.

The U.S. Atomic nergy Commission Staff (Table 4.1 of Ref.1) used a number of conservative assumptions to calculate the MCA doses at the EAB. Tnese conservatisms included a short-term atmospheric dilution factor of 2.6 E-3 sec/m3 resulting from an assumed downdraft of the exhaust clume at a wind speeo of only 0.3 m/sec during Pasouill atmospneric condition F. This produced a wnole body dose for the MCA of 8.6 rem at tne EAB, which is well celow the 10CFR Part 100 guidelines.

BASIS for SR-90 and I-131 Activity Limits The Sr-90 cone cose eouivalent and 00SE EQUIVALENT I-131 limits are llW based on tne AEC's evaluation (Ref.1) of Design Basis Accident No. 2 (PCRV racid ceoressuri:ation-FSAR Section 14.11), wnerein tne entire primary coolant circulating inventory anc fractions of the platecut iodines and s trontiurn are carriec out of tne PCRV and out of tne reactor cuilcing tnrougn tne louvers.

The U.S. Atomic Energy Commission Staff (Tacle 4.2 of Ref.1) used a number of conservative assumptions to calculate the accicent consequences. However, tnese assumotions result in calculated EAB doses which are well celow 10 CFR 100 guicelines. The maximum l equivalent activity levels (e.g.. Sr-90 and I-131 limits) cetermined by the Commission staff from the Design Basis Accicent No. 2 (CEA-2) are summari:ec in :ne following taole:

f .

Amendment No, ]

i Page 3/4 4-20 1,'

DRAFT BASIS FOR SPECIFICATICN LCO 3.4.3/SR 4.4.3 For plant coerati0n in the range of aoout 25'. to 10C'. RATED THERMAL POWER, maximum imourity levels nave oeen establisnec to restrict graonite oxication anc caroon transport from tne reacter core to cooler portions of tne p rima ry coolant system to about 330 lo/ cycle. Limiting the cuantity of graphite oxici:ed or caroon transoortec from tne reactor l*

core ensures the structural integrity J the fuel elements and tne core succort structure, and limits the caroon cecosition effect on tre steam generator heat transfer properties. Tne caroon corrosion will ce fairly uniformly districuted tnrougneut tne outlet third of the core, resulting in a rate of weign loss from tnis portion of ne core cf aoout 0 . 3'. cer -cycle whicn is within allowances r

- assumed in tne cesign. (FSAR 4.2,1).

Primary cociant is monitored and alarmed by the PPS Dewooint Moisture Monitoring System. Tne Primary Coolant Pressure-Hign instrumentation would also indicate tne presence of imouritie; in tne Primary Coolant System.

DGY graonite soecimens have oeen olaced in modified coolant inannels in five transition reflector elements in the hottest columns of regions 22, 24, 25, 27, and 30. Tne surveillance test soecimens are sucjected to the same primary coolant conditions, as well as other reactor parameters, as seen by the PG) core support blocks.

Examination anc tests of tne surve lance test soecimens at regular intervals can readily te utili:ed to assess oxidation rates, oxication profiles, as well as general cegracation of tne PGX core supoort blocks to preciet aceouately the structural integrity of the core supoort blocks over the opera;.ing life of tne reactor, (FS AR 5ection 3.3.2.2 and Accencix A.12.5.5).

Visual examination of tne core succort blocks in : nose regions cncsen for insertion of ?GX graonite specimens orovices accitionai assurance tnat integrity of tne core support clocks coes not cegrace cue to piant coerating conditions, since tnose regions were selected because of their higner Detential for'PGX graonite ournoff. Analysis shows that tne hignest tensile stresses Occur on the top surface of ne core succort clocks, at tre keyways, and at the weo cetween reactor coolant cnannels. Consecuently, any crack.ing ac;ic ce exoectec to originate at tnese locations, anc snouic ce c15coverec curing insoection.

Amendment No.

,u. _ . ., Page 3/4 7-12

'. em DRAFT PLANT AND SAFE SHUT 00WN COOLING SUPPORT SYSTEMS AUG 5' t@98 3/4.7.1 TURBINE CYCLE SAFETY VALVES - OPERATING LIMITING CONDITION OR OPERAT!ON 3.7.1.5' a. At least one steam generator economizer-evaporator-suoerneater (EES) safety valve per operating loop (V- lW 2214 V-2215, V-2216. V-2245. V-2246, or V-2247) shall be OPERAELE for eacn boiler feed pump in oceration supplying feecwater to tne EES sections. OPERABLE valve setooints snall be in accorcance with Table 4.7.1-1 *

o. Botn reneater safety valves (V-2225 and V-2262) shall os OPERABLE witn setooints in accorcance witn Taole 4.7.1-1.*

APPLICABILITY: POWER, LOW POWER and STARTUP l ACTION: a. With one or more of the above required EES safety valves inocerable in any one loop or with one reheater safety g valve inocerable, restore tre required valve (s) to OPERAELE status within 72 nours or restrict plant operation as follows:

1. With an EES safety valve (s) inoperable, restrict plant operation so tnat tne numoet of boiler feed cumos in operation corresponds to the numoer of OPERABLE safety valves as required above.

2. With a reneater safety valve inocerable, be in at least SHUTCO'nN witnin :ne next 24 nours.

c. With ore or more of :ne acove recuired EES safety valves inoceraole in cotn locos, restore tre reouirec valve (s) to CPERABLE status ithin 12 nours or restrict plant ik operation so tnat tne numoer of coiler feed pumos in oceration corresconcs to tne numoer of CPERABLE safety val,es as reouirec above.

. Tne e-ovisions of See:ifi:ation 3.0.4 are rot a:cli:10's.

Setooint verifd:at':- is not -e:utrec unti; 7 days after acnieving steacy state ciant coerating c:ncitions at a power level acove 50'. R'TE: THE; MAL 00aER.

~

Amendment No.

i Page 3/4 7-13

'i' DRAFT AUG 5 HN39 SURVEILLANCE RE0VIREMENTS 4.7.1.5 The superheater and reneater safety valves sna11 be cemonstrated OPERASLE by_ testing in accordance with tne applicable ASME Code requirements to verify setpoints. Tne test frequency is specifiec in the ASME Code, and tne lift setting 2 are scecifiec in Tacle 4.7.1-1.

TABLE 4.7.1-1 STEAM GENERATOR SAFETY VALVES VALVE NUMEER LIFT SETTINGS LOOP I V-2214 Less than or eaual to 2917 psig V-2215 Less tnan or eaual to 2346 osig V-2216 Less tnan or eaual to 2774 osig V-2225 Less enan or equal to 1133 psig LOOP 11 V-2245 Less than or eaual to 2917 psig V-2246 Less tnan or ecual to 2346 osig V-2247 Less snan or e ual to 2774 osig V-2262 Less tnan or e:ual to 1133 osig ,

t 4

.= _ -

Amendment No.

Page 3/4 7-15

's DRAFT AUG 5 1988 BASIS FOR SPECIFICATIONS LCO 3.7.1.5/SR 4.7.1.5 AND LC0 3.7.1.6/SR 4.7.1.6 The economi:er-evacorator-superheater (EES) secti:n of e :n steam generator loop is crotectec ey nree spring-loaced safety valves, ea:n with one-tnird nominal relieving caca ity of eacn 1000. Ine reneater section of eacn steam generator loop is protectec from overpressure transients oy a single safety valve. Tnese steam generator safety valves are cescrioec in tne FSAR, Section 10.2.5.3.

These steam generator safety valves are cesignec to relieve steam anc can ce camagec by racic :y:lic actuations tnat occur wnen tney relieve water. To crotect :nese valves, only one EES safety valve and tne reneater safety valve are maintainec in service in eacn loop, througn startuo evolutions with only one coiler feed pump sucolying feecwater to the EE3 sections. Each ooiler feec oump is capable of supolying acoreximately one-tnirc of the full ?ower feedwater  %

reouirements (FSAR Se: tion 10.2.3.1). As additional boiler feec oumos are olaced in service, accisional safety valves are also cla:ec in service. The use of one safety valve per steam generator section during SHUTCCWN.anc REFUELING is acceptacle, as it is capable of relieving the availaole flow. Also, c;ner power actuated valves that are capable of relieving cressure from the main steam and reneat piping are included in the FSV cesign.

The above valves are reouired to be teste: in accordance with ASME Section XI, IGV reouirements every 5 years (or less, cepending on failures) or after mai :enance. To satisfy :ne testing criteria, t .e valves must be testec witn steam. Since tnese valves are permanently i installed in steam oioing, tne approcriate means for testing recuires the olant to ce ocerating at steacy state concitions, and close to tne steam concisions exoectec at :ne setooint. Power levels aoove 5084 RATED THERMAL POWER are sufficient to acnieve this. Also, 7 cays ensures setooint verification witnin a reasonaole time, noting tna; the test scnecules are such tnat all valves are not tes:10 a tne same time and tnus, some valves will normally oe OPERABLE.

During all MODES. witn one EES sa f ety valve incoeraole, plant operation is restricted to a concition for wni:a. :ne remaining safety valves have suffic4 ent relieving cacacility to orevent j overpressuri:atior of s.v s:can generator se : ion.

Conversely, witn any reheater safety va ve inoceracle, olant oceration is restricted to a more restrictive 901.:.

A 72-nour at fon ti e for repair or SHUTCCWN cue to inoceraole safety valves ensures tra: *ese valves are returne to service in a relatively snor: :e-':: c# time, curing onfer an overoressure transient is un'.i.e!,. Coerat!0n at cower f0r 72 nours does not result in a si;*i# : ant 1055 of safety *e:: ion for any extence:

pericc.

Tne setooints d fo : e sa e::, ,a'ves <centifiec in Tacle 4.7.1-1 are tnose values icentie: tr : e :SAR aitn tolerances accliec sucn tra:

the Tecnnical Sce:'# t:aticas incorocrate an uceer counc setooint.

's -

Amendment No.

3 --

i Page 3/4-8-6 DRAFT SPECIFICATION 4.8.1.1.2 (Continued) AU6 59

d. At ' least once per 31 days oy ' performing a CHANNEL FUNCTIONAL TEST of tne SOG engine exnaust temoerature "shutcown" and "declutch" function.

e. At least once per 15 montns, curing SHUT 00WN oy:

1. Subjecting ne SOG diesel engines to an inscection in accercance with ne ' procedures prepared in-conjunction witn tne manufacturer's recommencations.

2. Performing a CHANNEL CALIBRATION of the 50G "shutdown" and "declutch" engine protective functions.

3 Ve ri fying tne SOG cacao 111ty to reject the single g largest loac wnile maintaining voltage at 480 olus or ninus 48 volts and frecuency at 60-plus or minus 1.2 H:.

4 Verifying the 50G caoacility to reject a lead of 1150 KW olus or minus 50 KW without tripping the 50G; tne SOG voltage shall not exceed 552 volts curing anc following the load rejection.

5. Simulating an undervoltage relay actuation signal:

a) Verifying ce-energi:stion of the essential 450 VAC cuses and l o t.d snecoing from the essential 430 VAC buses, b) Verifying tne SOG ciesel engines start on the aut0-start signal, energi:e the essential 480 VAC buses within 60 seconds, start the auto-secuencea loacs tnrougn the load secuencer, and ODERATE for greater tnan or ecual to 5 minutes wnfle tne associated SOG is ioacec with the programmed loacs; after energi:ation, tne steacy state voltage anc frecuency snali ce maintained at 480 clus or minus 48 volts and 60 clus or minus *. 2 H: curing this test, and c) Verifying tne overload anc antimotoring 50G trio

1. unctions are cycassec when tne 50Gs are in :ne a to-staat oce.

d) 'v e r ' ' ',

trat tne loac secuence timer is OPERABLE it- tre co?olete secuence loacec witnin plus or minas .'~, / o# its cesign time.

Amendment No.

Page 3/4 8-15 e

DRAFT 8 ASIS FOR SPECIFICATION LCO 3.8.1/4.3.1 (Continued) A G 5W The ACTION reovirements for various allowable levels of degradation of tne electrical power sources orovice restrictions uoon continued facility operation commensurate with the level of cegracation. The OPERASILITY of tne oower sources is consistent witn :ne initial conditions / assumptions of the FSAR, and is cased uoon maintaining at least one of tne recundant sets of on-site AC and DC electrical oower sources

-anc associated cistrioution systems operable curing accicent conditions anien costulate tne loss of all off-site cower, compounded by a single failure of :ne otner recuncant on-site sources.

The term "ve ri fy" as used in tne ACTION statements means to acministratively One:k oy examining logs or other information to determine if certain components are out-of-service for maintenance or c ner reasons. Tne term "ensure" as usec in ACTION statement 3.3.1.1.c allows 2 nours to verify OPERABLE or to restore to CPERABLE status af fected ecutoment, with any additional ACTION not recuired, if in compliance.

The surveillance recuirements are aceouate to cemonstrate the OPERABILITY of ne off-site and on-site AC electrical cower sources, sucn :na their intended safety functions uncer postulatec aonormal anc ac:icen conditions can ce performed.

In particular, ne surveillance recuirements for the SDGs are consistent with tne intent of Regulatory Guide 1.103 "Periodic Testing of Diesel Generator Units Use: as Onsite Electric Power Systems at Nuclear Power Plants", ;evision 1, August 1977 and Generic Let:er 34-15 "Prooosec Staff A:tions to Imorove and Maintain Diesel Generator Reliacility". In SR 4.8.1.1.2.e.3, :ne single largest load to De rejected is a U circulating wate cump, wnien is ratec at 202 KW.

Tne 50Gs are recuirec to rea:n ratec soeed, voltage and frecuency on cemanc. If an 503 coes not reach :nese parameters or i' the SDG fails to start cue to ceoletion of ne starting air receivers, :ne 503 start is consicered a failure.

The SDG fuel oil samoling reovirements are sufficient ::

assess fuel oi' ouality at For: St. Vrain. ditn over 10 years of diesel generator coera ional exoerience, tnere nave een no fuel oil relatec failures of :ne 50Gs. ;uel oil is cistributec ce:weea a ciesel fuel oil storage tank for tne 503s anc a srare: .ac. arrangement witn :ne Auxiliary Boiler.

Tne turnover :# c'ese' 'uel in tne undergrounc storage tanks curing SHUT 00 a . S'iRT.; an: LCW POWER; tme cerformance of Surveillan:e Re:a' e ents 4.3.*.c; anc :ne cerforman:e of Surveillance Ren "e ents 4.3.'.1.2.D

. anc 4.3.1.1.2.0 cemonstrate ne Lal';y " c'esel fuel oil in uncergrounc storage. Figure 3.3.'.-1. Diesei Fuel Oil Systems, snows ne

Amendannt No.

Page 3/4 8 DRAFT  ;

AU8 5 1988 y TABLE ~4.3.4-1 ACM O!ESEL GENERATOR TEST SCHEDULE-Number of Failures in i Last 20-Valid Tests Test Frecuency ,

51 At least once per'31 days ly l i

i 22 At least once per 7 days' lp

^

i i

i 1

i

• This test frequency snall ce maintainec until 7 consecutive failure-free cemands nave ceen cerformed anc tne numcer of failures in the last 20 cemands nas oeen recuted to 1 or less. i 1

?

Amendment No.

A Page 3/4 9-15

'f DRAFT AUG 5 1988 FUEL HANDLING AND STORAGE SYSTEMS 3/4.9.6 COMMUNICATIONS CURING CORE ALTERATIONS LIMITING CONDITION FOR OPERATICN 3.9.6 Direct two way communications shall be maintained cetween operations control room cersonnel and personnel at tne Fuel Handling Machine (FhM) control room.

APPLICABILITY: During CORE ALTERATIONS conducted from the F Refueling floor ACTION: With no direct communications between tne operatioits controi room personnel and oersonnel at the FhM control room, suspend all CORE ALTERATIONS.

SURVEILLANCE REOUIREVENTS 4.9.6 Direct communications cetween tre operations control room anc personnel at tne FhM control ro: shall be comonstratec witnin one hour orier to the start of and at least once per 12 nours curing CORE ALTERATICNS. k 4c

Areedment No.

.*- Page 3/4 9-16 DRAFT BASIS FOR SPECIFICATION LCO 3.9.6/4.9.6 8 The requirement for communications ensures that refueling cersonnel can be promotly inf0rmed of significant changes in tre i facility status or core reactivity conditions curing CORE ALTERATIONS, and operations control room personnel can ce informed oy refueling cersontal onenever CORE ALTERATIONS are being perfor.mec so snat core conditions can be monitoreo.

The FHM centrol room anc operations control room cersonnel must coordinate control rod movemer.ts to ensure the required SHUTDOWN MARGIN is maintained curing CORE ALTERATIONS. Maintaining direct c0mmunicati0n also cermits the coerations control room to immeciately r,eouest a stop of any movements causing excessive count rate cnanges.

The surveillance times soecified give acequate assurance that communications will ce availacle as needed.

)V I

t s

t t

Amendment No,

. .j' "'S'"'

DRAFT  ;

DESIGN FEATURES AllG 5W In addition to the reference fuel elements, up to eight test fuel elements nave resicec in the reactor core, cepencing on whien fuel segments are inclucec in,a given fuel cycle. Tnese eignt test elements (FTEl-3) contain small cuantities of test fuel- particles that are in various ways different from tne reference fuel, _ The j cescription of tne test fuel elements is contained in Taole 5.3-1.

The coated fue; carticles are conced together with a caroonaceous material to form fuel rocs. Tne fuel rods are completely surrouncec and containec ey graphite whten forms tne structural cart of ne fuel element, and in accition to tne caroon containe within the fuel rods, also serves as tne sole mocerator. Tne reference fuel elements are facricatec from H-327 needle coke (anisotrooic) graonite, as cescrioed in the Fort St.

Vrain FSAR, Section 3.0. Tne test fuel elements are fabricated from H-451 near-isotropic graphite in anticipation of avalifying this material for future use in all reload fuel for the reactor.

Beginning with core Segment 9 (Reload 3), H-451 near-isotopic graonite is used in the f ac-i ca tion of reload fuel elements in addition to or in p' ace of the previous reference H-327 neecle coke (anisotro:ic) graphite, 5.3.4 Reload Seement Cesien Eacn reload segment comorises about one-sixth of the reactor core. Consecuently, tne reactor core after a refueling consists of six segments witn cifferent cegrees of core curnuo districutec tnrougneut tne core. In addition, tne ournacle poison ceing acce: for reactivity control is only present witnin tne rew fuel elements. As

& consecuence, eacn of tne 37 core regions nas a different effective multiplication constant, k (eff).

The new fuel arc curnable coison loacina for each reload cycle, in conjunction witn tne remaining fuel in tne  %

core, shall satisfy tne following recuirements:

a. Provice acecuate core reactivity fer ournuo curing eacn cycle.

5. Ensure ar ac:ectacle SHUT 0;WN M H GIN :nrougnout the cycle u t acy one cont ol cc cat- witnerawn witn tne co e at an average te cerature of 30 cegrees F, and

Amendment No.

j ' Nge 5-9 Dgp DESIGN FEATURES AUG 5 1988

c. Ensure an acceptable SHUTDOWN MARGIN :nroughout the cycle with any two control rod pairs withdrawn with tne core at an average temoerature of 220 degrees F for at least two weeks.

c. Ensure temoerature coefficients at least as negative as those used in tne FSAR analysis, tnroughout the N cycle.

To satisfy tne criteria for reactor oower distribution and maximum control rod worth, eacn REFUELING CYCLE has a control rod withdrawal sequence that is specified for use during operation.

Tne following criteria shall te used as the basis to establish any control rod withdrawal sequence:

a. The maximum calculated reactivity worth of any rod pair in any normal coerating rod configuration with the reactor critical shall not exceed 0.047 delta k.

b. The maximum allowable calculated single control rod pair wortn. at any core condition, during power operation snali depend on tne available core temperature coefficient. The accicental removal of the maximum worth single rod-pair shall result in a transient with consequence no more severe than those cescrioed for tne worst case red-pair withdrawal accident in :ne AEC Safety Evaluation of Fort St.

Vrain cated January 20, 1972.

c. Calculatec power ceaking factors in any normal coerating rod configuration shall ce within tne following specifiec range:

Region Peaking : actor  : Average Region D/ Average Core P CORE AVERAGE CUTLET TEMD. Core Region Peakino Factor Greater than er eaual to Between 0.4 and 1.83 1250 cegrees :

Between 950 arc '250 cegrees F Eetween 0.t. and 2,15 Less than 95; cegrees F Eetween 0.23 and 3.00

,