ML20024G674

Proposed Tech Specs Re Reporting Requirements

ML20024G674
Person / Time
Site:	Monticello
Issue date:	12/16/1974
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20024G458	List: ML20024G457 ML20024G460 ML20024G674
References
NUDOCS 9104230499
Download: ML20024G674 (18)

Text

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AMD;D':E';T REQl' EFT DATED - DECD BER 16, 1974 EX111 BIT B Ihis exhibit consists of the following pages revised to incorporate the proposed Technical Specification changes:

Page v Page 150 Page 1 Page 164 Page 1A Page 167 Page IB (new page) Page 170A Page 81 Page 201 Page 131 Page 211 Page 134 Page 212 Page 136 Page 213

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i 9104230499 741216 PDR ADOCK 05000263 p PDR

i 4

i j 3. Standby Diesel Generators 182 r i

4. Station Battery Systems 183 3.9 Bases 185 4.9 Bases 186 3.10 and 4.10 Refueling 187 r

j A. Refueling Interlocks 187 I

B. Core Monitoring 188 ,

i C. Fuel Storage Pool Water Level 188 '

i.

D. Movement of Fuel 188

E. Extended Core and Control Rod Drive M1Lntenance 188A i 3.10 and 4.10 Bases 189

.i 5.0 DESIGN FEATLTES 190 .

l l 6.0 ADMINISTRATIVE CONTROLS 192 6.1 Organization 192 i

6.2 Review and Audit 195 i 6.3 Actions to be taken in the Event of an Abnormal Occurrence 201 '

! in Plant Operation i l I l 6.4 Action to be taken if a Safety Limit is Exceeded 201 l J 6.5 Plant Operating Procedures 202 6.6 Plant Operating Records 209 l l 6.7 Peporting Fequirements 211 v

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INTFnDUCTION These Technical Specifications are prepared in accordance with the requirements of 10 CFR 50.36 and

1. The bases for these Specifications app!v to the Monticello Nuclear Generating plant, Unit No.

are included f or information and understandability purreses.

1.0 DEFINITIorg dM ined so t hat a uniform interpretation of The succeeding frequently used terms are e xplicit1-tbo Speciffeations may be achicved.

l A. Abnormal Occurrence 1

.l . Prompt Notification with Written Followup

,. Failure of the reactor protection system, or other systems subject to limiting safety system settings, to initiate the required protective function by the time a monitored parameter reaches the value specified as the limiting safety system setting in the Appendix A Technical Specifications, or failure to complete the required protective function.

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b. Operation of the unit or af fected syst.ms when any parameter or operation subject a limiting condition for operat ion is less conservative than the least conservative aspect of the limiting condition for operat ion established in the Appendix A Technical Specification =.

c. Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary or primary containment.

d. Reactivity anoualles involving disagreement with predicted value of reactivity balance under steady state conditions greater than or equal to $1.00T a calculated reactivity conservative than specified in the Appendix A balance indicating shutdown margin lescterm reactivity increases that correspond to a reactor Technical Specificat!ons; short period of less than 5 seconds, or if suberitical, an unplanned reactivity insertion of l

I more than 50c; er any unplanned criticalitv.

I 1.0 REV

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c. Failure or malfunction of one or more components which prevents or could prevent, by

! itself, the fulfillment of the functic"al requirements of systems required to cope 1

with accidents analyzed in the SAF.

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f. personnel error or procedural inadequacy which prevents or could prevent, by itself, j i the fulfillment of the functional requirements of systens required to cope with accidents

, analyzed in the SAR.

c. Conditions arising from natural or man-made events, that, as a direct result of the event, require plant shutdown, operation of safety systems, or other protective  !

measures required by Appendix A Technical Specifications.  ;

, h. Errors discovered in the transient or accident analyses or in the methods used for such ,

analyses as described in the SAR or 'n the bases for the Appendix A Technical Specifications that have or could have permitted react or operation in a manner less conservative than i assumed in the analyses. i j 1. Performance of structures, systems or components that requires remedial action or corrective' l measures to prevent operation in a nanner less conservative than assumed in the accident analyses in the SAR or Appendix A Technical Specifications bases or discovery during plant life of conditions not specifically considered in the SAR or Appendix A Technical

! Specifications that require remedial action or correcti"e measures to prevent the

existence or development of an unsafe rondition.  ;

j 2. Thirty Day Written Reports t

a. Reactor protection system or engineered safety feature instrument settings which are  !

! found to be less conservat.ive than those established by the Appendix A Technical j Specifications but which do not prevent the fulfillment of the functional requirements j of affected systems.

l b. Conditions leading to operation in a degraded mode permitted by a limiting condition

! for operation or plant shutdown required by a limiting condition of operation. .(

I j c. Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection

systems or engineered safety feature systems.

t i 1.0 1A

! REV l

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_ _ _ s - _ - - - . _ _ _ _- _ _ __ _____ ________._._ _ _ _ _ _

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d. Abnormal degradation of systems, other than those specified in T.S.1.0.A.1.c above, l designed to contain radioactive material resulting from the fission process.

B. Alteration of the Reactor Core l The act of moving any component in the region above the core support plate, below the upper j grid and within the shroud. (Normal operating fun.tions such as control rod movement usiig the normal drive mechanisn, tip scans, SR!-f and IR?t detector novements, etc. , are not to be 4

considered core alterationm) i C. Ilo t Standhv -

i l lin t Standby means opei nt iori with the reactor c- itical in the rt;.rtup inode at a power level

just sufficient to maintain reactor pressure a

I t<,perature.

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1 poses 3.6 And h.6 Continued:

"he initini UDT Tomporature of the vensel shell e,terin] "pposito the ranc'or core region is O F.

1" initi,1 UDT temper,ture in tho main done flenges, and tho chell and head r:ntori,1 connecting to these El won in In"F, and "1sevbere is 400F. % e design life ei th- rmetor vessel Le is ho years and the maximum UDT temperature linit curve rnst neut ron . xposur- nt h1 years is es1culnted to be S.h x Id 7 evt.

in Fic;ure 4.6.] uses the "vorst case" curvo cf the PSA3 to est9di: b the UDT t gerature 1 shift and is ,

l thore fore con 7ervativo. %e expected UDT ter:perature shif t for this vencel nt 5.4 x 19 7 nyt isthe E in fact (r is based upon expected require- l 0

to be 00 F. Figure h.6.1 n1so incorporatos n (0 F factor of safety.

considerotions which resulted in those raquiranents. %erefore, the l ments of the M7tE ced" and for "the worst case" data ns voll as GWF of en,-in to provide assurance that specifiention providos eporntion in t he nor.- luctile region vill not occur.

3e ran" tor vessel hond flange and the vessel finnge in embinntion with the double "O" ring type seal l

'rSen the vessel head is placed on the are designed to provide a leak-tight seni when bolted together.

re,etor vessel, only thnt portion of the head flange near the inside of the vessel rests on the vessel finnge.

As the hend Folts nro repinced and tensicned, the vessel honi Is rioxed slightly to bring together the entire

( contnet surfnoes ndj, cent t o the "O" rinns o.~ the hen 1 and ve~el flonge. Doth the head and the hendflanceAere-i have nn ITDT temperat ure of OIC F, and they nre not subjectfor to on;/ nrpreciable 1olting tho hendneutron f2 nngeradiation nnd vessel exposure.

flange fore, the mininum vessol hend and hend finnce tempernturo 0

is establishod ns IO"F + 6J F or 'iG F.

Numerous d,ta are available relating integrated flux nnd the chan6e in nil-ductility transition temperature (ND?f) in various steels. De most conservative thto h7s been used in Specification 3 3 he integrated flux nt the vessel vall is calculated from core physics data and vill be measured using flux monitors installed incide the vessel, he measurements of the neutron flux at the vessel voll vill be used to check and if necessary correct, the enlculnted datn to determine nn occurate NDTf.

In addition, vessel mterial samples vill be located within the vessel to monitor the effect of neutron exposure on these materin1s. We samples mese include specimens of bnse metal, veld zone metal, heat affected sn=ples will receive neutron exposur" more rapidly than the vessel vnll rene metal, and stanlard specimens. Rese samples vill provide material and therefore vill 1 cad the vessel in integrated neutron flux exposure. An nnolysis and report further nasurnnee thnt the shift in ND'I'r used in the cpecifiention is conservnt ive.renetor vessels in nccordance will be sulnitted to the AEC on all such surveillnnce specinens re-oved from theinformation specified in ASTM E-185-66, 10 CFR Part 50, Appendix H. %ese reports shall include the lvith Tests on Structural ISter191s in Nuclenr Peactor," and information "f ece : mended P nctices rer Surveillanco routron irrn listion roceimd by the specimens and actual vessel obtnined on tha level of integrntad fast material.

131 3.6/h.6 REV 4

e Bases Continued 3.6 and 4.6:

D. Coolent 1.eaka ge lhe former 15 gpn limit for leaks fron unidentified sources was established assuming such leakage was coming f rem the primary sys t em. Tests have been conducted which deme, strate that a relationship exists between the size cf a crack and the probability that the crack will propagate. Fron the crack size a leakage rate can be determined.

for a crack w ize uhich gives .1 leakage of 5 gpm, the probability o r rapid propagation is less tha. 10-5 Th u s , on unidentified leak of 5 gpn when assumed to be from the primary systm had less than one chance in 100,000 of prop'-

Fating, which provides adequate margin. A leakage of 5 gpm is detectable and ceasurable. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period allowed for detemin ttien of leakage is also based on the low probability of the crack propagating.

The capacity of the du,fwell sump pumps is 100 gpm and the capacity of the drywell equipment drain tank pumps is also 100 rpm. Remove.1 of 25 gpm f ron either of these straps can Se acccmplished with considera51e cargin.

1 E. Safetv/ Relief Valves Testing of all safety / relief valves each refueling outage ensures that any valve deterioration is detected. A tolerance value of 1% for safety / relief valve setpoints is specified in Section III of the l

ASME Poller ud Presnure Vessel Code. Analyses have been performed with all valves assumed set 17. higher (1080 psig + 17.) than the nominal setpoint; the 1375 psig code- linit is not exceeded in any case.

The safety / relief valves are used to limit reactor vessel overpressure and fuel thermal duty.

The required safety / relief valve steam flow capacity is determined by analyzing the transient accompanying the mainsteam flow stoppage resulting from a postulated MSIV Closure from a power of 1670 N t . The analysis assumes a multiple-failure wherein direct scram (valve position) is neglected. Scram is assumed to be fr indirect means (high flux). Tn this event, the safety / relief valve capacity is assumed to be 717. of the full power steam generation rate.

1.6/4.6 BASES 134 REV

pnses Centinv'd 3.6 and h.6:

Donicn -enfimntion utel construction adequ,cy will bo d' :enstrated durinr the plant stnrtup and pm or accon : ion test pro. rne. As part of this progrnm, cold nnd hot vibration tacts on certain r"neter vosu'l internals vill be parfonwd. The tests, described in a letter to Dr. P. A. Forris, Intel March 5 F)W. are dm irned to ol tnin confimatory dat on ! he daign fmtures of Mont iceJ10 o3 a q , rod *o Lresder " nit ' design. Tler , tim basis for the "enti':cIlo vibration test progrnm is predicated on obtnining sntisfeetory datn which confims ermon d^oicn features from earlier FG p! ants mch ,s Drcr 1"n Unit 2. In the vent that data from theco earlier plants are not avuilobin before i

l routine pcvar oper' tion of Manticello, the rmtter vill b" revieve i by the AEC.

m The progrnn ou tlined in Table h.6.1 is limited to inspections of the prir"wr cool.ent acu. It is "nticiputed thnt t he data collected during the first fiv" years of operation vill provide o suitable 1 , sis to evnlunto the need for inspecting other pertions of the facility (such as the mein stens lines

!ownstroom of the min steanline isolation valves). Dione data nJong with the overall operating oxperiences vill be revioved to detemine the inspection prcr. ram to be inplorented for the lifetire of the facility. Die results of this study tocother with the prcpos"1 Ilfetime inspection program vill be

'ulcitted to th< !P in necordance with Cpecifiention 6.7. B.

Sie spoci,1 inspection of the min foed and steam lines is to provide nilad protection against pire whip. Die Group I velds are selected on the basis of an onolysis that shows these velds are the highest stress velds and thet due to their physical location, a break would result in the least interference nnl maximum energy upon impnet with the dryvell. Dioso velds are the only ones which offer nny signific,nt risk and vill be included in future inspectiens as detemined by the study described nbove.

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Grorp II velds are selected because without regard for the cperating stress levels and inter-l fering equil-:ent, they have sufficient theoretical enerry to penetrate and would propel the pipe toward the contaiment. They are therefore included in the first inspection. Upon consideration of i~ pact anC l e, interfering equipment and distance pipe traveln, no substantini risk is involved onl no extrn inspecticn is needed.

.2 critical systems.

In addition, extensive visual incroction for lenPs will le m,de periodically The inep"ct irn prerrnm sTncified enecqnsces the ma,inr areas of the vessel and piping systems within the dryw]l. The inspection period is l aced on the observed rate of grovth of defects from fatigue studios crensore11; the AEC. 21ese stulies show that it renuires thousands of stress cycles at 136

".'i/h.6 REV

l 4.0 SURVEILTANCE REQUIREENTS 3.0 LI?flTING CONDITIONS FOR OPERATION C. Secondary Containment C. Secondary Containment i Secondary containment integrity, shall be 1. Secondary containment surveillance shall

1. be performed as indicated below:

maintained during all modes of plant operation except whsn all of the following conditions are met.

The reactor is suberitical and Specifi- a. Secondary containment capability to a.

cation 3.3.A is met.

maintain at least a 1/4 inch of water vacuum under calm vind (2 <u <5 mph) conditions with a filter train flow rate of64,000 scfm, shall be dem-onstrated at each refueling outage prior to refueling.

l l b. The reactor water temperature is below 2120 and the reactor coolant system is vented.

c. Ne activity is being performed which can reduce the shutdown margin below that specified in specification 3.3.A.

150 REV 3.7/4.7

Bases Continued:

The acceptd le values for local leak rate teste hwa been rpecifie1 in terms of s .andsr3 cubic fer' per hcur (scf/hr) far purposes cf clarit'/. IcIlowing is thr. list of equivalent values gi ten in terms of an 31lowable percentre of the allowable operational leak rate (Lto)-

17.2 scf/hr = 5% Leo

@ 41 psig 3k.h scf/hr = ILT, I g i  ? 41 psig 103.2 scf/hr = 301 L g3

+ ~1 psig whera Ltg = .75 L t(the maximum allowable leak rate) and L., = 1.2 wei. ht percent of the contained air at the test pressure of kl psig.

Results cf loss of coolant accident analyses indicate that fission produ..ts would not be released directly to the environs because of leakage through the main line isolation valves due to holdup in tk.e steam syr' m cocplex. Although this effect shows that an adequate margin exists with 1

regard to relcaer of fission products, the results of leak tests on the main steam line isolation

( valves will te ciosely followed in order to detarmine the adequacy of these valves to perform

! their intended fbra tion.

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Monitoring the nitrogen makeup requirements of the inerting system provi. des a method of ctserting leak rate trends and would detect gross leaks in a very short ti e. This equipment must te l periodically removed frem service for test and .aintenance. vut this out-of-service time will be kept to a practical minimum.

I h.7 IMCFS i

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a.0 SURVE1LLANCE REQUIREMENU__

3.0 LIMITING CONDITIONS OF OPERATION lavestieite to identify the 5. A determination shall be made of the

a. total I-131 releasad weekly. An analysis causes for such release rates. shall be performed on a saeple at leant De fine and initiate a progran nonthly for I-133 ead '.- 13 5 .

b.

to reduce such release rates 6. A determination shall be made of the to the as low as practical levels.

total radioactive part i culates with hal f-Provide a report describing lives greater than 8 days released we"kly.

c.

these actions within 30 days.

The particulate filters shall be recoved and analyzed for gross beta particulata radioactivity P.-ith half-lives greate r than 8 daya. Monthly, a composite of those filters used during the month

10. At least one of the two stack monitors, shall be prepared and analyzed for tha including the charconi cartridge and principal gamma emitting radionuclides, particulate filter, shall be operab13 at all t.2es that the stack is releasing Analysis for Sr-89 and Sr-90 shall be i

7.

effluents to,the environs. made quarterly. Gross alpha radioactivity shall be determined quarterly.

11. If both stack monitor s are made or found in-operable, the reactor shall be olaced in the hot standby conditioa within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

12. Except as specified in 3.8.A.13, the off-gas stack and reactor building vent monitors shall have automatic isolation set points consistent with Specification 3.8.A.1 and alarm set points consistent with Specifi-cation 3.8.A.2.

13. If operation is necessary with the Of f-gas lloldup System recombiners bypassed, the off gas stack monitors shall serve only an alarm function.

170A 3.8/4.8 REV

c. Mechanism for schedu11ng meetings

d. Meeting agenda

c. Use of subcommittees

f. Review a...! approval, by members, of OC actions

n. Distribution of minutes 6.3 Abnermal occurrence Action In the event of an abnormal occurrence as defined in the Appendix A Technical Specifications.

the Commission shall be notified and/or a report submitted pursuant to the requirements of

( r.s.6.7.A.

Each abnormal occurrence shall be reported to the Operations Committee, either by copy of the report previously submitted to the Commission or by a separate investigation report. The Operations Committee shall review the report and reconmend further action if necessary.

Copies cf the report and Operations Committee minutes documenting their review shall be l

I submitted to the Safety Audit Committee and the General Superintendent of Nuclear Power Plant Operation.

l 6.4 -Safety Limit Violation l

If a safety limit is exceeded, the reactor shall be. shut down and the Commission shall be notified l

immediately. It shall also be promptly reported to the General Superintendent of Nuclear Power Plant Operativa and the Chairmin of the Safety Audit Committee, or their designated alternates.

A safety limit violation report shall be prepared. This report shall describe (1) applicable circumsttnces precedirg the vfolation, (2) effects of the violation upon facility components, systems or structures, and (35 the basis for corrective action takca to preclude recurrence.

The report shall be reviewed by the Operations Committee. The safety limit violation report shall be submitted to the Commission, the General Superintendent of Nuclear Power Plant Operation, and the Safety Audit Committee with two weeks of the event.

Operation shall not be resumed until authorized by the Commission.

201 6.3 REV

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i 10. Feactor coolant system in-service inspections

11. Minutes of meetings of the Safety Audit Committee 6.7 Reportion Fenuirements f

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A. Routiae Reports and Event Reports j

! The operating information to be reported to the USAEC in addition to the reports required by i Title 10. Chapter 1, Code of Federal Regulations, shall be in accordance with the Regulatory i posit ion of Regulatory Guide 1.16. Revision 2, " Reporting of Operating Information - Appendix A l' Technical Specifications " with the following exceptions:

1. In case of conflict between definitions or requirements in Fegulatory Guide 1.16 and these j Appendix A Technical Specifications, the Intter will take precedent. In those cases where  !

the Regulatory Guide definitions of abnormal occurrences centain interpretive notes on what j

, is reportable, those interpretations will apply to the definitions listed in T. S. 1.0.A. i 1

2. Regulatory Guide 1.16, Revision 2 (R.G. 1.16-2) , para. C.1.b(2) : replace "five percent"

, with "20%."

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3. R.G. 1.16-2, para. C.1.b(3) : Exposure reporting will be for individuals receiving exposures ,

i greater than 100 mren in the reporting period and assignment of exposures to various duty functions will be based on best estientes from normally used personnel monitoring devices. ,

4. R .C. 1.16-2, pa ra. C. I .b(4) : The reporting recocecnded under sub-paragraphs (a)-(d) will not ,

be required. I t

5. R.G. 1.16-2, para. C.2.a: The prompt notification shall be as expeditiously as possible, but

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vithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed by telegraph, mailgram or facsimile transmission  !

i to the Director of the appropriate Regulatory Operations Regional Office, or his designate no later than the first working day following the event, with a written follow-up report ,

within two vecks.

l i 6. R.G. 1.16-2, para. C.2.a(1) : change to read " Failure of the reactor protection system,or other systems subject to limiting safety 5: t settings, to initiate the required protective function by the time a monitored parameter reaches the value specified as the limiting safety i

system setting in the Appendix A Technicai 4rcr ifications, or failure to complete the requir< d

! , protective function."

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l 6.7 211 REV

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7. R.C. 1.16-2, para. C.2.a (3)

change to read "Ahnormal degradation discovered in fuel cladding,

reactor coolant pressure boundary or primary containment."

8. R.C. 1.16-2, para. C.2.a(4): Delete this definition and reporting requirenent.

, 9. R.G. 1.16-2, para. C.2.b: Change the reporting time requirement to 30 days following the i event.

10._ E.C. 1.16-2, para. C.2.b(2): change to read " Conditions leading to operation in a degraded rede permitted by a limiting condition for orcration or plant shutdown required by a limition condition of operation.

11. R.C. 1.16-2, para. C.2.b(3): Delete this definition and reporting requirement and re-number C.2.b(4) es C.2.b(3) . Add new paragraph C.2.b( **) which reads " Abnormal degradation of system

. other than those specified in C.2.a(3) abe c, designed to centain radioactive material resul ting from the fission process."

12. R.G. 1.16-2, para. C.2.c: The events listed under this paragraph are not abnormal occurrencen

! and are not required to be reported.

]

j B. Special Reports i The following special reports shall be submitted in writing to the Director of the Regulatory j Operations Regional Office within the tine period specified for each report:

e j Area Reference Submittal Date

1. prtnary Containment Leak Rate Tests 4.7A 90 days after each integrated leak rate test 1

2. In-Ser* ice Inspection Evaluation & Developnent 4.6F 6 4.6F Bases October 1, 1976 I'

3. Failed Tuct Detection 3.2 Bases July 1, 1976 6.7 212 REV l

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C. Environmental Reports 1

The following reports relating to environmental activities shall be submitted to the Director of l the Regulatory Operations Regional Office and are included in this Appendix A Technical Specification section until an Appendix B Technical Specification has been issued for the Monticello Nuclear

! Generating Plant:

?

l i 1.

A Semiannual Radioactive Ef fluents Report shall be submitted within 60 days af ter '

I January 1 and July 1 of each year. The report will meet the intent of Regulatory

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Guide 1.21, Revision 1, and will include a surnary of the quantitles of radioactive liquid and gaseous ef fluents and solid wastes released from the plant during the previous six months of operation, 2.

An Annual Radiological Environmental Monitoring Report shall be submitted by April 1 of the subsequent year. The report will meet the intent of Regulatory Guide 4.1 (1/IS/73) l and will include summaries, interpretations, and statistical evaluation of the results j of the radiological environmental surveillance activities. In the event that some results j

+

are not available within the 90 day period, the report will be submitted noting and explaining the reasons for the missing results which will be submitted as soon as possible

! in a supplementary report.

3. An Annual Environmental Monitoring and Ecological Studies Program Report shall be submitted by August 1 of the subsequent year. The report will include summaries, interpretations, and statistical evaluation of the result's of the non-radiological environmental surveillance

, activities.

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