ML18018B070

Responses to Questions Relating to Changes to the Technical Specifications

ML18018B070
Person / Time
Site:	Nine Mile Point
Issue date:	11/23/1973
From:	Niagara Mohawk Power Corp
To:	US Atomic Energy Commission (AEC)
References
Download: ML18018B070 (248)

Text

CONTENTS PAGE NO.

Question 1 1 Question 2 5 Question 3 ~ ~ 7 Question 4 9 Regulatory Guide 1.22 .10 Regulatory Guide 1.23 12 Regulatory Guide 1.24 13 Regulatory Guide 1.25 14 Regulatory Guide 1.26 15 Regulatory Guide 1.27 ~ ~ 16 Regulatory Guide 1.28 17 Regulatory Guide 1.29 18 Regulatory Guide 1.30 20 Regulatory Guide 1.31 21 Regulatory Guide 1.32 22 Regulatory Guide 1.33 23 Regulatory Guide 1.34 24 Regulatory Guide 1.35 ~ ~ ~ 25

...,Regulatory-.Guide -l. 36 "26 Regulatory Guide 1.37 27 Regulatory Guide 1.38 28 Regulatory Guide 1.39 29 Regulatory Guide 1.40 30 Regulatory Guide 1.41 31 Regulatory Guide 1 ~ 42 32 Regulatory Guide 1.43 33 Regulatory Guide 1.44 34 Regulatory Guide 1.45 35 Regulatory Guide 1.46 36 Regulatory Guide 1.47 37 Regulatory Guide 1.48 38 Regulatory Guide 1.49 39 Regulatory Guide 1.50 40 Regulatory Guide 1.51 41 Regulatory Guide 1.52 42 Regulatory Guide 1.53 44 Regulatory Guide 1.54 45 Regulatory Guide 1.55 46 Regulatory Guide 1.56 47 Regulatory Guide 1.57 49 Regulatory Guide 1.58 51 Regulatory Guide 1.59 53 Question 5 54 Question 6 57 Question 7 60 Question 8 . 62

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C QUESTION Since the Technica1 Specifications appended to Provisional Operating License No. DPR-l7 will be reissued. at the time of conversion, the proposed. changes to technical specifications should be submitted as soon as possible for our review. Therefore, please provide a listing of a11 changes to the Technical Specifications for Nine Mile Point Unit 1 (MMP-1) other than those identified herein that you wiU.

consider in connection with the license conversion. Include adequate supporting information for changes 'being proposed..

RESPONSE

The proposed modifications to the Technical- Specifications fall into two categories (a) those requested in letters already submitted and (b) those proposed to be made at the time of license conversion.

Since Technical Specification Change No. 9 there have been other proposed changes submitted to the AEC which have not yet been acted upon. They are as follows:

a. Proposed changes to Specificatjoy 3.3.6, "Vacuum Relief"; have been submitted in two letters.

b.--" Proposed 'environmental techriical specifications.

c. Reload fuel app)ication concerning rod block and peaking factor for 8 x 8 fuel.

d. Proposed changes to specifications 3.1.1 and 3.6.2 concerning the reanalysis of the rod drop accident and the APRM rod block sys-tem.

Other areas where changes to the Technical Specifications. are appropriate are discussed below:

a. Specification 3.2.5 change primary method of leak detection to the rate of rise monitoring in sump level.

As described in response to Regulatory Guide 1.45 a more sensitive method of monitoring primary system leak detection has been installed. This system will detect a 0.2 gpm change in flow in the range of 0-1 gpm flow and a 0.5 gpm change for inflows of 1-5 gpm.

1 Letter dated March 26, 1973 from R. R. Schneider to D. J. Skovholt.

2 Letter dated July 20, 1973 from R. R. Schneider to D. J. Skovholt.

3 Letter dated October 4, 1973 from P. D. Raymond to D. R. Muller 4 Letter dated October 16, 1973 from P. D. Raymond to A. Giambusso.

5 Letter dated November 15, 1973 from R. R; Schneider to Mr. A. Giambusso.

b. Specification 4.2.2 change withdrawl schedule to meet the following:

First capsule - one fourth service life.

Second capsule - three fourth service life.

Third capsule standby In the event the surveillance specimens at one quarter of the vessels service life indicate shift of reference temperature greater than predicted the schedule shall be revised as follows:

Second capsule one half service life.

.This:d r.capsule ."s tandby .

The above is a modification to ASTM Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels. It will assure that there is adequate margins of safety against fracture throught the vessel service life.

c. Specification 3.6.2, Tables 3.6.2b and 4.6.2b The set points for the high area temperature for the clean-up and shutdown cooling system isolation should be changed to read 190F and 170F respectively.

Due to the fact that the monitors are at ceiling level and good circulation of air through these rooms does not exist, normal

""ambient-temperatures have ranged -from 140 F to 160 F.

d. Proposed technical specification for the high pressure coolant injection system is presented here. ' new specification (3.1.8) will be added as follows:

3.1.8 HIGH PRESSURE COOLANT INJECTION Applies to the operational status of the high pressure coolant in j ection system.

~Ob 'ctive:

To assure the capability of the high pressure coolant injection system to cool reactor fuel in the event of a loss-of-coolant. accident.

ae During the power operating condition whenever the reactor coolant pressure is greater than 110 psig and the reactor coolant temperature greater than saturation temperature, the high pressure coolant injection system shall be operable excepted as specified in Specification "b" below.

b. If a redundant component of the high pressure coolant injection system becomes inoperable the high pressure

coolant injection shall be considered operable provided that the component is returned to an operable condition with 15 days and the additional surveillance required is performed.

c ~ If specification "a", and "b" are not met, a normal orderly shutdown shall be initiated within one hour and reactor coolant pressure and temperature shall be reduced to less than 110 psig and saturation temperature within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.1.8 HIGH PRESSURE COOLANT INJECTION Applies to the periodic testing requirements for the high pressure coolant injection system.

~OB 'ective:

To verify the operability of the high pressure coolant injection system.

The high pressure coolant injection surveillance shall be

.-performed'"as "indicate'd"bel'ow:

a. At least once per operating cycle automatic start-up of the high pressure coolant injection system shall be demonstrated.

b. At least once per quarter pump operability shall be determined.

c. Surveillance, with Ino erable Com onent When a component becomes inoperable its redundant component shall be demonstrated to be operable immed-iatelyi and daily thereafter.

Bases:

The High Pressure Coolant Injection System (HPCI) is provided to ensure adequate core cooling in the unlikely event of a small reactor coolant line break. The HPCI System is required for line breaks which exceed the capability of the Control Rod Drive pumps and which are not large enough to allow fast enough de-pressurization for core spray to be effective.

One set of high pressure coolant injection pumps consits of a condensate pump, a feedwater booster pump and a motor driven feedwater pump. One set of pumps is capable of delivering 3,800 gpm to the reactor vessel at reactor pressure. The performance capability of HPCI alone and in conjunction with other systems to provide adequate core cooling for a spectrum of line-breaks is discussed in the Fifth Supplement of the FSAR. 3

In determining the operability of the HPCI System the required performance capability of various components shall be considered.

a0 The HPCI System shall be capable of delivery rate of 3,800 gpm.

b. The motor driven feedwater pump shall be capable of automatic initiation upon receipt of either an automatic turbine trip signal or reactor low-, water-level signal.

c The Condenser hot. well level shall not be less than 48 inches (75,000 gallons) .

d. The Condensate storage tank inventory shall not be less than 105,000 gallons.

During reactor start-up, operation and shutdown the condensate and feedwater booster pumps are in operation. At reactor pressures up to 450 psig, these pumps are capable of supplying the required 3,800 gpm. Above 450 psig a motor-driven-feedwater pumps, is necessary to provide the required flow rate.

The capability of the condensate, feedwater booster and motor driven feed-water water pumps will be demonstrated by their operation as part of the feed-water supply during normal station operation. Stand-by pumps will be placed in service at. least quarterly to supply feedwatex during station operation.

An automatic system initiation test will be performed at least once per oper-ating cycle. This will involve automatic starting of the motor driven feed-water pumps and flow to .the reactorvessel..

Specification 3.6.2k will be added as follows:

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k. Hi h ressure Coolant In'ection Initiation - The high pressure coolant injection system shall be considered inoperable and specification 3.1.8c shall be applied.

Other technical specification changes are discussed in answer to questions 2, 3, 7 of this submittal.

Tables 3.6.2k and 4.6.2k vill be added as follovss AB E 6 2k a d 4 6 2k PS EC h E 0 3.6.2k LinitinS Condition for Operation 4i6.2k Surveillance Requireuent Mininun No. of Reactor Mode Svitch Mininun No. oi Operable Instrunent position in Which Instrunent Instrunent Tripped or Operable Channels per Set Function Must Se Sensor Channel Charms psraueter ri S scene 0 arable 1 S st ~n e abl Check Test ~Cllb 1

ti (1) Lou Reactor 1 ft. Shutdovn Refuel Starts Run On<<e per Once per Once per 3

'Mater Level belou vater nonth conths X X X X day level at El.

302'-5u X X X X (2) Autonatic hone Each Oper-Turbine ~ tinS Cycle Trip

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT BASES I Cont'd.l S.2.S REACTOR COOIAVT SYSTEM IEAKACE RATE C.2.S REACZOR COOIAVT SYSTESI IEALACE RATE

~ll~'U Applies to the I(nits on reactor coolant systen leak- Applies to thc nonitoring of reactor coolant systcn age rate lcakagc.

Ob ective:

To assure that the nakeup capability provided by the To deteznlne thc reactor coolant systcn leakage rate control rod drive Funp ls not exceeded. and assure that thc lea'kagc llnits are not exceeded.

cification: ~Sill I Any tine Irradiated fuel ls ln the reactor vessel and A check of reactor coolant systcn leakage shall be Allowable leakage rates of coolant frere the reactor reactor coolant tnapcrature ls above 2IIF ~ reactor nadc at least once pcr day. coolant systcn have been based on the predicted and coolant leakage into the prlnazy contalnaent frow cxperlnentally observed behavior of cracks ln pipes unidentified sources shall not exceed S gpn. In and on the ability to nakeup coolant systcn leakage ln addition, the total reactor coolant systcn leakage thc event of loss of offslte a-c power. The noznally into the prlnazy coats(anent shall not exceed 25 gpn. expected baclground leakage due to equipaent design and If these conditions cannot be net, the reactor vill be thc detection capability for dctezulning coolant systen placed ln the cold shutdown condition within ten hours. lea'kage were also considered ln establishing the llnlts.

The behavior of cracks ln piping systens has been ex perlaentally and analytically investigated as part of the IISAEC sponsored Reactor prinary Coolant Systen Rupture Study (the pipe Rupture Study) . work utllltlng the data obtained in this study indicates that leakage fran a crack can be detected before the crack grows to a dangerous or critical size by nechanlcally or ther-nally induced cyclic loading, or stress corrosion cracking or sons other nechanisa characterized by gradual crack grcwth . This evidence suggests that for lea'kage scuewhat greater than the llnlt specified for unideatlfied leakage, the probability ls swell that inperfectlons or cracks associated with such leakage would grow rapidly. However, the establlshtent of allowable unidentified leakage greater than that given In $ .2.$ on the basis of the data pzesently available would be prenature because of uncertainties associated with thc data. For leakage of the order of S gpa as specified in $ .2.S, the experinental and analytical data suggest a reasonable nargin of safety that such Icakagc nsgnitude would not result frow a crack approaching the critical size for rapid propagation.

leakage of the nagnitude specified can be detected reasonably in a natter of a few hours utilizing the available leakage detection schenes, and lf the origin cannot be deteznined ln a reasonably short tine the plant should be shut down to allow further investiga-tion and <<orrectlve action.

A total leakage of 2S gpn ls well within the capacity of the control rod drive systen nakeup capability (page Ill-7 of the First Suppleuent).e As discussed In S.I.O above, for leakages within this nakeup capablllty the core will rcaain covered and autonatlc pressure blow down will not bc actuated.

eFSAR.

4T

LIMITING CONDITION FOR OPERATION . SURVEILLANCE REQUIREMENT BASES (Cont'd.)

she prxaaxy scans of detexaining the reactor coolant leakage rate is by aonitoring the rate of rise in the lewis of the dxywll floor anx eguipeent drain lines. Daily checks vill he sade that no alaxae have been actuated due to high leak-age. ror cusp inflove of ona gpa changes on the oxder of 0.2 gpa can he detected vithin 40 ainutes. At inflowe ba-tveen one and five gpa changes on the order of O.S gpa can be detected in eight ainutes.

leakage is detected by having all unidentified leakage routed to the drywall floor drala tank, and identlficd leakage routed directly to the drywall equlpscnt drain tanks. identified leakage includes such lteas as re clrculaticu puap seal leakage and recirculation puap suction and discharge valve packing leakoff.

Another aethcd vill aonitor the th>> re quired to fill the tanks between tvo accurately de-texaincd levels. When the level ln the tank reaches the lov-level svitch setting, a tiacr vill start and operate for a preset tine interval. if the tlaer resets before the high-level svltch setting ls reached in-dicating a leakage rate vithin allowable liaits, no action <<ill result, and the systca "resets for the next filling and tining cycle. lf the leakage ls high enough to cause the level to reach the high level switch setting before the tiacr resets autoaatlcally, an alaza ls actuated indicating a leak rate above the predetexalncd llalt (First and Fifth guppies>>nts).

Additional lnfozxatlon ls available to the operator which can be used for tho daily leakage check 1 f the drywall susps level el rats are out of service. lhe integrated flov puaped froa the suaps to the vesta disposal systea csn be checked.

Qualitative infozaatlon ls also available to tha oper-ator ln the foxa of indication of dryvell ataospheric

<<ondltions. Continuous leakage froa the priaary cool-ant systea vould cause an increase ln drywall tesper-ature. Any leakage in excess of lg gpa of steaa vould cause s continuing increase ln dxywell pressure vlth resulting scras (First Supplcaent).

Reactor coolant systea leak detection vill be further studied during the first year of operation of the facility. lf the results indicate that significant iaprovcaents ln aonitoring capability csn ba really achieved on a practical basis, such isprovcaents will be inpleaented.

Tables 3. 6. nd 4. 2. 6b INSTRUMENTATIONTHAT INITIATES

. PRIMARY COOLANT SYSTEM OR CONTAINMENTISOLATION

3. 6. 2b - Linutin Condition for eration 4.6.2b - Surveillance Re uirement Minirnurn No. of Operable Instrument Reactor Mode Switch Position Minimum No. of Channels per In Which Function Instrulnent Instrument Tripped or Operable Operable Must Be 0 erable Sensor Channel Channel Parameter Tri S stems Tri S stem Set Point Shutdown Refuel Startu Run Check Test Calibration PRIMARY COOLANT ISOLATION (Main Steam, Cleanup, and Shutdown)

(I) Low-Low Reactor S5 ft below minimum X X X X Once/day Once per Once per Water Level normal water level at month 3 months Elevation 302'-9" (2) Manual X X X Once duriag each major refueHag outage MAIN STEAM LINE ISOLATION (3) High Steam Flow S l05 psid X X Once/day Once per Once per Main-Steam Idnc month 3 months (4) High Ralliation <5 tilnes normal X X Once/ Once/week Once per Main-Stealn Line background shift 3 monthS (5) Low Reactor >850 psig X Once/day Once per Once per Pressure month 3 months (6) Low-Low- Low >7 in. mercury (a) (a) X None Once during Once during Condenser Vacuum vacUuln each major each major refueliag refueling CUtagc outage (7) High Temperature <200F X X X None Once dur lag Once during Main-Steam- Line each major lnajor Tunnel refueling refueHng outage CUtagc CLEANUP SYSTEM ISOLATION (8) High Area (S 190r X X X Once/week Once during Once during Temperature each major each major refueling refueling outage CUtagc l00

Tables 3. 6. 2b and .. 6b (cont.)

INSTRUMENTATIONTHAT INITIATES PRIMARY COOLANT SYSTEM OR CONTAINMENT ISOLATION

3. 6. 2b - Limitin Condition for 0 eration 4.6.2b - Surveillance Re uirement Minimuxn No. of Operable Instrument Reactor Mode Switch Position Minixnuxn No. of Channels per In Which Function Instrument Inst~eat Tripped or Operable Operable Must Be 0 erable Sensor Channel Channel Parameter Tri S stems Tri S stem Set Point Shutdown Refuel Startu Run Check Test Calibration SHUTDOWN COOLING SYSTEM ISOLATION (9) High Area < 170r X X X Once/week Once during Once during i

Texnperature each major each major refueling refueling outage outage CONTAINMENT ISOLATION (IO) Low-Low'eactor <5 ft below xninimuxn X X X Once/day Once per Once per Water Level normal water level at xnouth 3 months Elevation 302'-9" (11) High Drywell 53.5 psig (b) X (b) (b) Once/day Once per 0nce pex' month months Pressure (I 2) Manual X X Once during each operat-ing cycle Notes for Tables 3.6.2b and 4.2.6b (a) May be bypassed in the refuel and startup positions of the reactor mode switch when reactor pressure is less than 600 psi.

(b) May be bypassed when necessary for containment inerting.

h

2. UESTION Certain technical specifications applicable to containment will be required. to be changed.; these are discussed. below:

General Design Criteria Nos. 55, 56, and. 57 specify require-ments for particular classes of primary containment penetrations.

You present general information to show conformance to these requirements in your report, "Technical Suaplement to Petition for Conversion from Provisional Operating License to Full-Term Operating License" (Application). Please provide a tabulation .of isolation, valve arrangement andfunction-on all fluid. line 'penetrations of the R&-1 primary containment. This tabulation is to be incorporated. in Table 3.3.4 of the Technical Specifications.

Include in the above tabulation a designation of all valves which are defined to be primary containment testable penetrations for purposes of Section 4.3.3.e(l) of the Technical Specifications.

Ce In genera1, the primary containment leakage rate testing speci-fications are to be updated to be consistent with the requirements of Appendix J to 10 CFR Part 50. In particular, Section 4.3.3.d(3) is to be revised. to specify that a set of three integrated leak rate tests (Type A) shall be performed. at approximately equal

..,mnterva1sMuring. each .ten-.year,.service period.. The third. test-of each set shall be conducted when IMP-1 is shut down for the ten-year plant inservice inspection.

d. Similarly, Technical Specification 4.3.3.e(2) is to be revised.

to specify that the personnel air lock door seals shall be tested., after each opening when the reactor is in a power operating condition, at a pressure of 10 psig and the leak rate extrapolated to 35 psig. The leak rate shall not exceed 5$ of La. Air locks shall be leak rate tested, at a pressure of 35 psig at six-month intervals.

e. Section 4.3.3.e(4) is to be revised. to specify that testing C

of main steam line isolation valves is to be performed, at 35 psig in accordance with Appendix J of 10 CFR 50.

RESPONSE

a,b Proposed revisions to Technical Specification 3.3.4 is attached and noted by marginal markings.

c,d,e Primar Containment Leaka e Rate Testin Proposed changes to Technical Specification 4.3.3 to bring it into conformance with 10CFRSO Appendix J are attached.

Changes to pages 65, 66 and 67 of the Technical Specifications and Bases would be required.

As a result of the first primary containment leak rate test, the ratio of L (22)/L (35) was determined to be 0.735. This results in a test leak rate equation from Appendix J, which is the same as the current specification 4.3.3b (2).

The test frequency require by the current Technical Specif--

ications is not the same as that contained in Appendix J of 10 CFR50. Changes are proposed in the attached specification 4.3.3.d.

Three primary containment leak tests have been performed as follows:

1. Test at 35 and 22 psig (August 1969) .

2. Test at 22 psig (June 1970).

3. Test at 22 psig (May 1972).

Changes are also proposed for specification 4.3.3.e.(2) and e.(4) to allow testing of air locks and main steam isolation valves in conformance with Appendix J of 10CFR50.

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT BASES ICont'd.)

3.$ .4 PRIlQRT CCÃTAIhsKVT ISOIATICII VALVES 4.3.4 FRIMARY OavTAIbOKIT ISOIATIOX VALVES b~\blll ~bb bill Applies to thc operating status of the systca of Applies to the periodic testing requiresents of the isolation valves on lines open to the free space priaary containnent isolation valve systcn.

of the priaary containaent.

ebb To assure that potential leakage paths froa the prl- To assure the aural'Illty nr thc privacy containscnt nsry containsent Ln the evcat of a loss-of-coolant isolation wives to liait tbotential Iesiare oaths frou accident are alniaited. the contalnaent In the cwnt oe s lose-oe-cnn)ant accident..

a<<lbl I ~SI ll l Double isolation valves ara provided on lines penetrating tha prlaary Ithcnever the reactor coolant systce teaperature Ihe priaary contslnaent isola'tlon valves surveillance ocntalnaent and open to the tree space ot the containaent. Closure ls greater than 2ISF, all contaitsseat isolation ot one of the valves in each line vould ba suttlcieat to naintain tha shall be pcrforaed as indicated (see Table 3.3.4) . integrity ot che pressure suppression systea. Sxcept vhere check valves on lines opea to the free space of the priaaty containaent shall be operable except as a. At least once per operating cycle the operable valves are used as one or both of a set of double isolation valves, specltied in $ .$ .4b belov. isolation valves that are powr operated and the Isolation closure tiaes are presantai in rebel 3.3 4. Isolation valve arrangeaents on lines not opening to the free space of tha con-autoastlcally initiated shall be tested for auto- tainnent are also presented ln Table 3.3.4. Autoaatlc initiation is

b. In the event any isolation valve bccoaes inoperable astic initiation and closure tines.

the systea shall be considered operable provided required to ainlaixe the potential leakage paths tree the containsent at least one valve in each line having an lnoper b. At least once per quarter all nornaIIy open ln the awnt ot a losswf-coolant accident. bstails of the isolation awe valve ls la the node corresponding to the valwe ara discussed in Section VI%.~ ror allovable leakage rate isolated condition.

pover operated Isolation valves shall be fully speclticattonsb see Section 3.3.3 above.

- closet snd reopened.

c. If Specifications $ .3.4 a and b are not aet,the As Illustrated in Figure E-$ 4 of Appendix E'uel rod perforation

c. At least once per operating <<ycle. each Instrument- does not occur until about 1$ 0 seconds folloving the loss-of-coolant reactor coolant systea tcaperature shall be re- accident. A required closing tine of 60 seconds for all priaary duced to a value less than 21SF vlthln ten hours. line flow check valve will be tested for operability; containaent Isolation valves vill be adequate to prevent fissloa product release through lines connecting to the privacy contaiteent.

For reactor coolant systea tcaperatures less than $ 12F, the Contain-seat could not beccae pressurlted due to a loss-of-coolant accident.

The 2)SF liait Is based on preventing pressuritation of the reactor building and rupture of the blnvout panels.

The test interval of once per operating cycle for ptcaatic lnltla-tion results ln a failure probability of I.I x 10 that a line vill not isolate (Fifth Suppleaent,. p. IIS). Here frequent testing for valve operability results in a nore reliable systcn.

In addition to routine surveillance ss outlined ln First Addenbhss to Technical Supplesent to Petition to Increase Pover Level, each last-ruaent-line flov check valve vill be tested for operability. All instruaents on s given line vill be isolated at each Instrusent. The line vill be purged by isolating the flee check valve, opening the bypass valves, and opening the drain valve to the equlpaent draia tank. shen purgiag ls sufficient to clear the line of non-conden-sibles and crud the flcv-check valve vill be cut Into service snd tho bypass valve closed. The aain valve vill again be opened and the flov-check valve a)loved to close. the flov-check valve vill be reset by closing the drain valve and.opening the bypass valve depressurizing part of the systes. Instruaents vill be cut into service after closing of the bypass valve. Repressurlting of the lndlvldual Instrusents assures that floe-check valves have reset to the open posltlon.

'FSAR Issued &CO-TI 68

LLfITI:1't CotZ)III(CX PCR OFIRAII(t Table 3 3 6 PRL6'dtY CHAI: X'T ISOL TIC: VALVES t~atiot Rclatf c VJ s fnu

~ll AcCfon oa No. ot Valves to Prfnsry Nor J Oper. Ifac Inftiatfng f In 1 c at 1 ng Signal Potftfo; vJtfve Fever c All Valves Have Rcaote Manual gac'Eu D ell Vcnc d Pur e Coancct 1 on Outside Closed (a) 60 Close One Line Outsldc Closed (a) 60 Close Reactor vates level lov lov or dryvell high pressure Afr Conncctfott Outside Closed (a) P.o. 60 Close (One Line) Outside Closed (a) F.o 60 Close Sv cession Chanber Vent 6 Fvr e N Connection Outside Closed (a) F.o. 60 Close One ioe Outside Closed (a) F.o. 60 Close Rcaccor vater level lov-lov or drywll high pressure Air Connection Outside Closed (4) F.o. 60 Close (One Line) Outside Closed (a) F.o. 60 Close Drywll NI M~akeu Reactor vacer level lov-lov or dryvcll

'(ttw T~ae Ovicide Closed (b) P.o. 60 Close high pressure NI Makeup Reactor vacer level lov-lov or drywff (Oac Linc) Ov teide Closed (b) F.o. Close high pressure ll uf nt Drain Line Inside Opctl F.o. eo Close (Onc Line) Outside Open P.o 60 Close Reactor vatcr level lov-lov or drywll high pressure Inside Open F.o. eo Close (One Line) Outside Open F.o. 60 Close u rection hanber ttstcr Ecu Outside Closed (b) F.O. 60 Reaoce nanw1 (ons Lfae) Outside Sell Act. Ck.

Ataosphcre to Pressure Suppression Sysc<<a Outside Closed F.o. Open Negative prcssure relative co atnosphere (Three Lines) O Csfde Sell Acc. Ck, Reactor Cleaaup Syscea Relict Valve Dfscha e (One Line to Supprcssioa Cbaaber Oucsfde Selt Act Ck.

Saapliag DrZyell (Ihree Liaes) Oucsfde Closed (b) F.o. 60 Close Rcaccor vaccr level Ittc Iov or high u 60  !

dryvell pressure (one Line) Outside Closed (b) F.o. Close Notcst (a) These valves aay be open for coatafnacnc till vfth nitrorcn.

(b) Ihcse valves vill periodically be opened for sanplfng or nftrogea aaltcup.

(c) P.o. - Povcr Operated.

69 1m'CO 71

I LIMITINC CONDITIO~vOR OPERATION Table 3.3.4 ntinued PRIM'RT CONIAItDKhT ISOIATION VALVES RF.E PA F. NTA K Location Relative Maxfnun Action on No. of Valves to Prfuary Homal Motive Oper. Tine Initiattng Initiating Signal

~ne ~it K ~Es ct+~c a{we A sl ave . e s s ku

~Co ~Srs (d)

~Pn S~ut (Fout Lines froa Supp. Chan.) Outside AC Hotor 90 Rcnotc annual

~PG. b (I'o Test Lfnes to Supp. Chan,) Outside Closed AC Motor Msx, 90 Close Reactor vster level lov-lov D veil Vent 6 N Connection Outside Closed Air/DC Sol. Msx. 60 Close (Oac Line) Outstde Closed AC Hotor Hax, 60 Close Reactor vater level lov-lov, or dryvell high pressure

~al C cf Outside Closed Air/DC Sol, Hsx. 60 Close (Oac Line) Outside Closed AC Hotor Hsx. 60 Close Rectr. Pu=p Cooltng d

Vi eQ Su Supply Line Outside Open Self hct. Ck.

Return Lfnc Outside Open DC Motor Max+ 30. Recete nsnua1

.ve11 Cnole Mat Su 1 '(d)

Supply Lfae Outside Open Self hct, Cki Return Ltne Outside Open DC Motor Hex'0 Rcnote nanua1 INE W F. E 0 A 1' uue 11 sn Air/DC Sol. Reactor level lov-lov and

~51 (P Lt ~ )

Outside Open Msx, 60 Open high dryvell pressure u II II h Outside Self hct. Ck, (Four Ltacs) cist Cbecb r Branch Outsfde Self hct. Ck, (One Branch for Each Systcn) u~ S c n n Su e sto Outside Open AC Ptor 60 Resote nsaual

~Chaebc (Four Lines) valve in each separate line and oae valve in each cocoon line.

(d) Ihese are classified ss not-testable valves and penatratfonsi

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT BASES ICont'd.I 3.3.3 LEAKAGE RATE 4. 3. 3 LEAKAGE RATE Al i Applies to the allowable leakage rate of the primary con Applies to the primary contalnmcnt system leakage talnment system. rate, Ob ectlvet Ob ective:

To assure the capability of thc containment in limiting To verify that the leakage from the primary containment radiation exposure to the public from exceeding values ~ ystem ls maintained <<ithin specified values.

specified in IOCFRIOO In the event of a loss-of-coolant accident accompanied by significant fuel cladding failure and hydrogen genrratlon from a metal <<ster reaction.

Whenever the reactor coolant system temperature ls above a. Intc rated Primer Containmcnt Leaks Rate Test 215F thc primary containment leakage rate shall be <<ithln e The primary contalnmcnt preoperational test pressures are based upon the calculated primary containment pressure the limits of 4. S. 3. b. (I) Integrated leak rate tests shall be performed response In the event of a loss-of-coolant accident. The peak prior to Itdt(al Station operation at the test dry<<ell pressure <<ould be 35 psig which <<ould rapidly redoce pressure of 35 pslg (Pp) and the test pressure to 22 pslg wlthln 100 seconds following the pipe brea'k. The (Ptl of 22 psig to obtain the respective measured total time the drywall pressure would be above 22 pslg ls cal-leak rates Lm (35) and Lm (22). culated to be about 10 seconds. Following the pipe break. the suppression chamber pressure rises to 22 pslg wltldn 10 (2) Subsequent leakage rate tests shall be performed seconds, cquallscs <<ithdrywell pressure and thereafter rapidly

<<ithout preliminary leak detection surveys or decays with the dry<<ell pressure decay. (I) leak repairs immediately prior to or during the test. at an initial prcssure of approximately 22 The design pressures of the drywall and absorption chamber are ps ig. 62 pslg and S5 pslg. respectively. (2) The design leak rate ls

0. 5~A/day at a pressure of 35 pslg. As pointed out above. the (3) Leak repairs. If necessary to permit Integrated pressure response of the dry<<ett and suppression chamber leakage rate testing, shall be preceded by local following an accident <<outd be the same after about 10 seconds.

leakage measurements. The leakage rate differ- Based on the calculated primary containment pressure response ence. prior to and aRer repair when corrected discussed above and the'suppression chatnber design pressure, to Pt shall be added to the final integrated lca'k primary containment preoperational test pressures were chosen.

age rate result. Also. based on the primary containment pressure response and the fact that thc drywall and suppression charnbcr function as a (4) Closure of the containment isolation valves for unit. the primary containment will be tested as a unit rather the purpose of the test shall be accomplished by than testing the Individual components separately.

the means provided for normal operation of the valves. ~ The design basis loss-of-coolant accident was evaluated at the primary containment maximum allowable accident leak rate of (5) The test duration shall not be less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ).9yA/day at 35 pslg. The analysis showed that with this leak for integrated leak rate measurements. but shall rate and a standby gas treatment system fitter.efficiency of be extended to a sufficient period of time to veri- 40 percent for halogens. 95 percent for particulates. and fy. by measuring the quantity of air required to assuming the fission product release fractions stated In return to the starting point (or other methods of TID-14844. the maximum total whole body passing cloud dose equivalent scnsltivhy). the validity and accuracy ls about 6. 0 rem and the maximum total thyroid dose ls about of the lea'kage rate results. 150 rem at the she boundary considering fumigation conditions (I) Appendix E. FSAR (2) Volume I.Section VI. FSAR Isauod O~-71

4 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIRBhENT BASES (Con)'d )

b. Aces tance Criteria over an exposure duration of two hours. The resultant doses would occur for the duration of the accMent at the low population (I) The maximum allowable tea4xe rate Lp shall distance of c mlles are lower than those stated due to the vari-not exceed I.5 <<eight percent of thc contained ability of meteorological conditions that would be expec*d to air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the test pressure of 35 occur over a 30-day period. Thus. the doses reported are the psig (Pp). maximum that <<ould be expected ln the unlikely event of ~ design basis loss. of-coolant accMent. These doses are also based on (2) The allo<<able test leak rate Lt (22) shall not the assumption of no holdup in the secondary containment re-exceed the value established as follower sulting in a direct release of fission product ~ from the primary containment through thc filters and stack to the environs. There-i,t (22) ~ 1.5 i,m(22)/I m (35) fore, the speci((ed prltnary containment leak rate and filter efficiency (Specification c.c. <) are conservative and provide margin (3) The allo<<able operational leak rate. Lto (22) bet<<ecn expected offslte doses and IOCFR)00 guideline Ifm)ts.

<<hich shall be met prior to resumption o( po<<er operation folio<<lng a test (either as rncasured The maximum allowable test leak rate as specified ln 4.3 ~ 3 b or folio<<ing repairs and retestl shall not exceed is 1.5%/dayat a pressure of 35 pslg. This value for the test

0. 75Lt (22). condition <<as derived from the maximum allowable accident leak rate of about 1.9%/day when corrected for the effects of

c. Corrective Action containment environment under accident and test conditions.

In the accident case. the containment atmosphere lnltlally If leak repairs are necessary to meet the allo<<able wouM be composed of steam and hot air depleted of oxygen operational leak rate. the integrated leak rate test <<hereas under test conditions the test medium would be air need not bc repeated provided local lca4ge measure- or nitrogen at ambient conditions. ConsMerlng the differences ments arc conducted. and the leak rate differences in mixture composition and temperatures, the appropriate prior to and after repairs. <<hen corrected to Pt correction factor applied <<as 0.8 and determined from the and deducted from thc integrated leak rate measure- guide on containment testing. (3) rnent. yield a lea4ge rate value not in excess of the allo<<able operational leak rate Lt (22). Although the dose calculations suggest that the allowable test to leak rate could be allo<<ed to increase to about 3.0%/day before three integrated leak rate tests shall he perfor<<ed the guideline thyroid dose limit given ln IOCFRIOO would be at approxinately equai intervals during each 10-year exceeded. establishing the t(m(t at 1.5%/day provides an service period with the third test in each ten~ear adequate margin of safety to assure the health and safety of the interval corresponding with the ten~sr scheduled general public. It is further considered that the allowable leak in-service Inspection shutdovn. rate should not deviate significantly from the contalnmcnt design value to take advantage of the design leak-tightness capability of the structure over its service lifetime. Addft)ona(margin to maintain the containment in the "as built" condition ls achieved by establishing the allo<<able operational leak rate.

The operational limit is derived by rnultlplying the allowable test leak rate by 0. 75 thereby providing a 25% margin to allow for leakage deterioration <<hich may occur during the pc rfed bet<<ceo teak rate tests ~

The primary containment leak rate test frequency I ~ based on maintaining adequate assurance that the leak rate remains within the specification. The leak rate test frequency ls based on the AEC guide for developing leak rate testing and surveillance of reactor containment vessels.( )

(3) TID-20583. Leakage Characteristics of Steel Contalntnent Vessels and the Analysis of Leakage Rat ~ Determinations.

(c) locttt50 appendix J, Reactor contain<<ant Leakage Testfng for I water cooled rover tteactors .

Issued

tl LIMITING CONOITION fOR OPERATION SURVEILI.ANCE REQUIREMENT BASES I Cont'd. I

e. Local Leak Rate Tests (I) primary containment testable penetrations and isolation valves shall be tested at a pressure of 35 pslg each major refueling outage except bolted double gaskeced seals shall be tested <<henever the seal ls closed aHer being opened. and at least at each refueling outage. The penetratioa and air purge piping leakage test frequency, along <<,ith the containment leak rate tests. Is adequate to 4llow (I) Personnel air lock door scale shall be tested detection of leakage trends. Whencvcr a double-gasketed pane after each opening <<hen the reactor is in a tratlon (primary containment head equipmcnt hatches and the poser operating oonditioa, at a pressure of 10 pslg and the leak rate extrapolated to 35 Psig. suppression chamber access hatch) is broken and remade. the air locks shall also be leak rats tested at a ~ pace bet<<een the gaskets is pressurized to determine that the pressure of 3$ psig at six aonth intervals. In seals are performing properly. The test pressure of 35 pslg is 44eh test. Chs Iesk r4te Cnrreeted to 35 psig consisteat with the accident analyses and the maxlcnum preop-shall not exceed 5 percent of L . cratlonal leak rate test pressure. Ic ls expected that the major)ty (5) Containment components not included in (I) and of the leakage from valves. penetrations and seals <<ould be Into

-(2) <<hich required teak repairs following any the reactor building. However. it is possible that leakage Into integrated leakage rates in order to rnect the other parts of the facility could occur. Such leakage paths that allowable leakage rate unit Lt shall be subjected may affect significantly the consequences of accidents are to be to local leak tests at 4 pressure of 35 psig at minimised. If the leakage rates of the double-gasketed seal each refueling outage. penetrations. testable penetration isolation valves. contalnmeat air purge inlet ~ and outlets and the vacuum relief valves are at (S) ghe aaln steaa line isolation valves are to the maximum specified. they <<ill total 90 percent of the allowed be Seated at 4 pressure of 35 psiil dtiriag 44th leak rate. (5) Hence. 10 percent margin is left lor leakage refueling outage.

through <<elis and untested components.

f. Corrective Action Monitoring the nitrogen makeup requirements of the lnertlng (I) lf the total leakage rates listed below as adjusted system provides a method of observing leak rate trends and to a test pressure of 22 pslg, are exceeded. re- would detect gross lea'ks in a very short time. This equip-pairs and retcsts shall be performed to correct ment must be periodically removed from service for test the condition. aad maintenance. but this out-of-service time will be kept to a practical minimum.

(a) double-gasketed seals l0 Lt (22l (b) (I) testable penetrations 30>> Lco [22) and isolation valves (t) any one penetration or 5>>v Lto (22)

I s ola t ion va I vs (c) primi ry contalament air 50" I o (22) purge penet rations and reactor building to torus vacuum relief valves

g. Continuous Leak Rate Monitor (I) When che primary containment is inerted the conCalnment shall be continuously monitored for gross leakage by review of the inerting system makeup requirements.

(t) This monitoring system may be taken out of service for the purpose of maintenance or tcstlng but shall be returned to service as soon as practical.

The accessible interior surfaces of the drywall shall be visually Inipccted each operating cycle (5) Volume i.Section VI. FSAR Issued for evidence of deterioration. 67

I

,L I

I

~UESTZON Certain other technica1 specifications will be required to be changed in addition to including requirements not presently specified. These are discussed.'below:

Concurrent maintenance involving removal of ccntrol rod. drives and. draining of the torus has been required previously on two occasions. The Technical Specifications should 'be revised. to permit this type of concurrent maintenance and to specify the applicable controls.

Your "an~sis 'of a'nticipated transients without scram indicates that the recirculation pumps would 'be tripped whenever reactor vessel pressure is equal to or greater than 1150 psia. You are requested to propose appropriate technical specifications to reflect this requirement if the tri~ing of the recirculation pumps is, in fact, what you are proposing. In this regard, you state that the installation of these trips would. be done by the end of the first refueling outage.

c ~ In Section 3.2.3.c of the Technical Specifications, the basis for allowing nigher chloride ion concentration and conductivity at high steaming rates is the lower oxygen levels anticipated at high steaming rates, and the data shown in Figure 3.2.3.

The referenced "Corrosion and lear Handbook" from which this figure is taken states that data were taken, on U-bend specimens of severa1 stainless steels, primarily type 347, at 500 F, in pH-10; 6 water containing 50 cpm POg, exoosed. in the steam phase with intermittent wetting. The original figure title states: "Curve is based. on observations made under specific conditions, (and) therefore is not intended. for general use."

Therefore, we believe Figure 3.2.3 is an unacceptable basis for technical soecifications of chloride levels in neutral, phosphate-free water at 600 F in PrrRs made of type 304 stainless steel. Tnerefore, Section 3.2.3.c should be deleted from the Technica1 Specifications or revised to conform with Table 2 of Regulatory Guide 1.56, "Maintenance of Mater Purity in Boiling Mater Reactors", which gives 5 umho/cm ynd 0.5 cpm chloride as acceptable limits at steaming rates P 10~ lb/hr for a B!R with a freshwater-cooled condenser.

The Surveillance Requirement specified in Section 4.2.3 of the Technical Specifications should be revised to define a "short term spike" and to indicate actions to be taken when the continuous conductivity monitor is inoperable. A maximum time for .operation of the plant without the continuous conductivity monitors being operable should be specified.

RESPONSE

a. Proposed revisions to Technical Specifications 3.1.4 and 3.3.7 is attached and noted by marginal markings.

b. Since the time of the Technical Supplement to Petition for Conver-sion from Provisional Operating License to Full Term Operating Li-cense a number of changes have taken place with, regard to Anticipated Transients Without Scram.. One important change has been the effect on transients of a revised scram reactivity curve. The second change relates to the staff review and conclusison, included in "Technical Report. on Transients Without Scram for Water Cooled Reactors".

"We are currently reviewing our analysi's an'd conclusion against the staff position described in that report. Information concerning this revised analysis and proposed changes to the Technical Specifications will be discussed in our response to the AEC request of October 19, 1973.

c. Proposed revisions to Technical Specification 3.2.3 is attached and noted by marginal markings.

1 Letter dated October 19, 1973 from Mr. A. Giambusso to Mr. P. D. Raymond

BASES (COIIt'd.I LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.1.4 CORE SPRAY SYSTtzt 4.1.4 CORK SPIIAY SYSTOt

~lr bl a Appliaa to the operating status of tbe core spray Applies to the periodic testing zequirencats for systeas the coze spray systcas.

~b To assure the capability of the core spray systeas To verify the operability of the core spray systeas.

to cool reactor fuel ia the event of a loss of-coolant accident.

a> ~5<<! fl 1

a. Whenever irradiated fuel is in the reactor vessel, The core spray systea surveillance shall be pcrforaed The core spray systea consists ol tvo autonatlcally each of the two core spray systems shall be oper- as indicated belov. ~ ctuated, independent, douMe-capacity systaas capaMe able except as specified in Specifications b, c and of cooliag reactor fuel for ~ range of loss-ol-coolant d, below. a. At each as]or refueling outage Autoatatic startup accidents. For the vorst liae brcak, a loss-of-coolant of oae set of Ixzaps in each core spray systea accldcnt ~ a core spray of at least $ 400 gpn is required shall be dcaonstrated. vithin SI second>> to provide fuel stability sufficiently

b. If a redundant coraponeat of a core spray system to assure effective core cooling.

becomes inoperable, that system shall be consid b. At least once per quarter puap operhbillty shall ared operabla provided that the component is be decked. Each core spray systea has 100 percent cooling capacity returned to an operable condition within 15 days free eath spray header aad each supply puap aet. Thus>

and the additional surveillance required is perfozmed. c. At least once per quarter the operability of specifying both systeas to be fully operational vill pover operated valves required for proper sys- ~ asurc to ~ high degree core cooling capability if the

c. Il a redundant component in each oi the cote sptay tca operation shall be checked. core spray systea ls required.

systems becomes inoperable, both systems shall d. Core sprav header p I nstruacntat ion be considezed operable pzovided that the component Allovable ourages are specified to account for ccraponcnts that becoae Inoperable in both systeas and for sere than is rctuzned to an operable condition within T days check Once/day one coaponeat in s systea.

aad the additional surveillance required is per- calibrate Once/3 aonths lormcd. test Once/I senths goth core spray cyst>as contain redundant supply pvrap

d. Il a e. Survel I lance el ah Inc rgb le Co nants sets and blocking valves. Operation of oae puap set core spray system becomes inoperable and ~ nd blockiag valve ls sufficient to establish required all the components are operable in the other system, shen a cceprment or systca bccoaes inoperable its the reactor may remain in operation for a period delivery rate and flov path. Therefore, even vith the redundant coaponcnt or systca shall bc deaonstratcd loss ol one of the reduadsat coupoacats ~ a system is not to exceed I days. to be opera! lc lracdiarcly and daily thereafter. ~ till capable of perforaiag its intended function. If

~ rcdbadant couponeat is fouad to have failed, corrective

e. If Specifications a, b, c and d are not met, a normal aaiotensnce vill begin proeptly. Nearly sll nainteaaace orderly shutdowa shall be initiated vrithin one hour can be coupleted vlthin ~ fev days. Infrequently> hov-and the reactor shall be in the cold shutdown condi- ever> aa]or aalatenaace night be required. Replace>>eat tion within ten hours. of principal systea couponcnts could necessitate outagcs in excess of those specified. In spite of the best efforts of the operator to zetura equlpaent to service, soae aain If both core spray systeae becca>a inoperable the tensnce could require up to 6 aonths.

reactor shall be in the cold sbutdcvn condition vithin tea hours and no vork (except as specified ln dctcruiaiag the operability of a core spray systea ln f belov) shall be pezforacd on the zeactoz the required perfomaace capability of its various or its connected systaas vhich cauld rea>lt in cor>ponents shall be considered. Eoz exaapler loveriny the reactor vater level to sere than seven feet eleven Inches bolov alai>usa noraal l. The delivery rate frets one core spray puap and level. topping puap shall not be less chan 3400 gpa

~ t ~ coubined puap developed head of 691 feet of vster. At this delivery rate adequate core cooling is provided to prevent fuel welting. (Section VII)

'TSAR 32

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT BASES ICont'd.)

f. Work aay be pextoxaed on control xod drives at The puwp shall be capable of autcaatic initiation times when vater fs not fn the euppressfcn cham- from u fow-fow water level signal in the reactor ber and the core spray eyatea shall be considered vessel or a high containment pressure signal. The operable provided that the following are aetx blocking valves shall be <<apable of autoaatically opening free either s low-low vater signal or high

1. wo aoxe than one control xod drive housing or containment pressure signal simultaneous with low trna penetration vill 1>> opened at any tim>>. reactor prcssure pczaissivc signal. (Section Vil)'

2. A blind tfance vill be installed on the ocn The core spray delivery rate of 3400 gpa shall be trol zo4 drive housfnc vhenever a contxol xod available at the core spray nox les inside the re-drivo has been reaoccd tox maintenance. actor vessel vlthin 33 seconds.

3~ work vill not bo peztozae4 xn the reactor vessel

6. backup diesel generator pover shall be avaflabl ~ to vhlle ~ control zod drive housfno fs open. ~ ll motor operated components.

4. A contxol rod drive vill not be removed it the gefore the first major refueling outage, instrumenta-backseat seal 4oee not function. tion vill be installed to monitor the fntegrlty of thc

5. A afniaua confensate storore volume of 300,000 qallons core spray piping vithin the reactor pressure vessel.

and a afniaua hot well storage volume of CO,OOO Valfons Following installation of this fnstriacntatfcn, the vill bo aaintaine4 dvzfor the pczfod that the torus requirements stated in specification C.l.s 4 shall bo water levol fa belov that correspcodfxvf to ainfxsas argy folfowd.

x equi rcaent.

The testing specified for each aajor refueling outage

6. The control zod drive removal an4 frxxc replaces>>nt shall vill demonstrate component response upon automatic not be concurrent items. mystes initiation. For exaxple, puap sot starting (lov-low level or high dzyvcll pressure) and valve opening (low low level or high dryvell pressure and low reactor pressure) mast function, under simulated conditions in the sane aanner as the systeas are required to operate under actual conditions. The only differences vill be that dcaincrallxcd vatcr rather than suppression chamber water will be puaped to the reactor vessel and the reactor vill be at atmospheric prcssure. The core spray systeas are designed such that dcainerallxed water ls availablc to the suction of one set of pumps ln each systea (Section Vlf Figure Vlf-l).

The systca test interval betvecn operating cycles results in a system failure probability of 1.1 x 10 6 (Fifth Supplcaent, page 113) and is consistent vith practical considerations. The nore frequent coapon ent testing results in a acre reliable system.w At quarterly intervals,'tartup of core spray pumps will demonstrate pump starting and operability. Ão flow will take place to the reactor vessel due to the lack of a fow-prcssure permissive signai required for opening of the blocking valves. Flow, instead will be re-cycled to the suppression chaaber via a test loop. An orifice has been provided in the test loop which vill simulate the pressure loss during emergency operation of the systca. In addition, the noxaslly closed power operated blocking valves will be aanually opened and re-closed to demonstrate operability.

The intent of speciticatfon 3.1 ct fm to allov contxol xod dxfve aalntenance and trxcc replacement at the tfs>> that tho suppression chamber fs unwatered and to pertoxa normal fuel aovcaent actlv-itics fn the refuol aude vith an unwatered suppression cf>>aber.

Xiascd on the liaitcd tfae fnvolved in pertoxacnce of the con-current aaintenance tasks, procedural controls to ainiaize the potential anf duration ot feakaoe trna the control ro4 drive housfno or fpkw penetration an4 available coolant aakrxp pzc vfdes adequate protection acafnst draf naqe of the vessel while the suppression cheater fe drain<<d.

33

l LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT BASES (Cont'd.)

3.3.7 coztalzmbt srzar srgtgx g. 3.7 CONTAINMENTSPRAY SYSTEM Applies to the operating statva Of the contalnnent Applies to the testing of the containment spra'y sprs) svsten system.

Ob ective:

to assure the capability of the containoent spray To verify the operabilhy of the containment spray systez to it>>it containzent pressure and teoperature system.

In tbe event of a loss-of-coolant accident.

During all reactor operating conditions whenever The containment spray system surveillance shall For react<<r coolant tcnpcratvres less than 215 F not reactor coolant temperature i ~ greater than 215 P be per!ormed as Indicated halo>>u <<not gh stean Is generated during a loss.of.coolest and fuel is in the reactor vessel: each of the two accident to prcssurltc the contalnnent. In fact. for containment spray systems and the associated a. Containment Spray Pumps coolant teuperatures up to 312 F the resultant loss-

~

raw water cooling systems shall be operable of-coolant accident pressure vould not exceed the design except as specified in 3.3.7.b. pressers of 35 pslg.

(I) At least once Pvr operating cyclo automatic startup of the containment op<<ratios of only one coatainoent spray poop ls sufficient

b. If a redundant component of a containment spray spray pump shall he demonstrated. co provM<< the required cont<<inn<<ac spray flow. tbe specified system becomes inoperable. Specification 3 3.7.a.

~ flow of 3000 gpn (approxinetcly 9S percent to the drywall and shall be considered fulfilled, provided that the (2) At least once er uar ter. pump tbe balance to tbe suppression chaober) is sufflcfent to

<<omponent is returned to an operable condition operability shall be checked. rewove post sc<<ident core coergy releases including a sub-within 15 days and that the additional surveillance stantial cbcuical reaction Involving hydrogen generation required is performed. and vill also lint<< preset.re and teopersture rises ln tbe

b. Nozzles pressure suppression systcu to belov design values

c. If a redundant component in each of the containment (Appendix E-II 2.2.3 p.g-)d and the Fifth At 1<<ast once per operating cycle. an Supple>>eat).'ach contain>>ant spray systcu is eonsldcrcd opcraMe vhen spray systems or their associated raw water air test shall be performed on the spray boch p4npz are Capable of deltvering at less<< 3000 gpn at ~

systems become Inoperable, both systems shall headers and nozzles. ~ ptztp developed head of 3)S fact of vatcr at 50 F.

be considered operable provided that the compo Reqidrln boch Funps ln boch systcoo op<

>qthin 7 c. Raw Water Cooling Pump>> 4ndancy) vill assure tbe availability of cbe contatnzent days and that the additional surveillance required sprs't'rst<<it ~ is performed. At least once per quarter ntanual startup '11Iovablc outogtc Or4 Ape<<if lcd to account for conponents that and operabllRy of the ra>> water cooling I <<conc inopcraitlc ln I oth s)>>ters znd for nore than one ccepon-

d. If a containment spray system or its assoclmed clt'I In a systctt ~

pumps shall be detttnnstrattvl. raw water system becomes inoperable and ail the The corresponding rav eater cooling systce ls dcslgncd to components are operable in the other systems. d. Surveillance <<Ith inoperable I',orv~ncnttt stztntaln contalnoent spray vater cusp<<recure no greater the reactor may remain in operation for a period tbsn IAO I' tndcr tbc sesc liniting op<<racing conditions. not to exceed 7 days. zhcn a coaponcnt or systcn)<<coors inoperable its the contalnoent sprat rav vat<<r cooling systeu is con-redundant cooooncnt or zyztce shall Ie Jcnonstratctl OMered operable vben the floe rate ls not less than 3000

e. If Specifications a or b are not met, shutdown to be operable iswcdiztcl>> and ~tally t!<<rczftcr. zpn at a p ep developed head of 540 feet of vater. Thts shall begin within one hour and the reactor cool- pvnp developed bead util nalntaln a higher pressure on the rau voter side st the hest cxchangers than on the ant shall be below 215F within ten hours. contatnoenc side so that any leakage vill be into the contalnnent spray systez.

lf both contafnoonc spray systcns becnne inoper- Electrical pover for all s stew conponcnts is nomsliy able the reactor shall be in the cold shutdown ~ val lable frost cbc reserve transfer>>sr Epoti loss of condition within tcn hours and no llork (except this service tbe puuping rcqutreoent vill be supplied as specified in f belov) shall be perforsed froa the diesel rcnerator. At least one diesel generator on the reactor vblch could result in leertrv) shall always be avallsMe to provide bac'kup ~ lectrlcal the reactor vater levol to notre than seven feet peer for one contalnnent spray systcn. corresponding eleven inches belov nozual level, (Elevation rav water coollnx systen and associated ~ lectronic cquip-302' ). eent required for ancona<<le s!stcu initiation. 73 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT BASES IG)nt'd.) Work nay be perfumed on contxol xod drives at tines Autonatic initiation of the containvent spray systen assures vhen vatcr is not in the suppression cfvxsber. 'the that the containnent vill cot b>> overpressurited dw to containeent spray staten shall be consfdemd oper hydrogen generation. This autonatic feature vould only be able pmvfded the folloving are nett required if all core spray systcvs naltunctfoncd and eigni natal-vatcr reaction occurred. yor the norasl opxr 'fcant

1. No nore than one contxol zo4 driw housing or sting condition of 90y suppression chanbcr vater and tvo lplot penetration vill be opened at any cine. psig contafrxscnt pressure, coctainnent spray actwtion vovld not be necessary for about 1$ ninutcs. Rav vater

2. A blind flange vill be installed on the con- cooling affects the tcnperature of the spray vaccr and the trol zo4 drive housing vbcnever a control zcd suppression chanber pool. Taking into account the reduced drive has been zeaovcd foz naintenance. stean condensation capability snd increased suppression chaster vapor pressure. Che rav vatcr cooliog vould not be

3. Nork vill not be perfozxed ln the reactor vessel required for nore than 20 ninutes for initial suppression vhile a control rod drive housing ls open. chaeber tcvperatures up to 110y. This assuues that all core spray systces fail. Therefore, nanwl initiation of A contxol xod drive vill not be rescued if the the rav vater systcn is acceptable.

backseat seal does not function. $~ A nfnfnun condensate storage veloce of 300,000 gallons Nearly all naintenance can be ccopleted vithin ~ fev days. and a nfninun hot vali storage volune ot 40,000 gallons Infrequently, hovever, na)or nalntenance nfght be required. Rcplaccnent of principal systen coepooents could necessitate vill be nalntaine4 dazing the perfod that the torus oucaccs of nore than 1$ days. In spite of the best efforts vater lewl is belov that oorxesponding to nininun NRSH Of the OperaCOr CO CCCurn equi pnent CO Seluic ~ s Sane nafn requf reeent. tenance could require up to 6 nonths. 6~ The contxol rc4 drive rcnoval and fpRN replaceaent shall not be conccxzwt fteea. In con)unction with containment spray pump operation during each operating cycle. the raw water pumps and associated cooling system performance will be observed. The containment spray system shafl be capable of auto-matic initiation from simultaneous low-low reactor water level and high containment pressure. The associated raw water cooling system shaB be capable of manual actuation. Operation of the containment spray system involves spraying water into the atmosphcrc of the contalnmcnt. Therefore, periodic system tests are not practical. Instead separate testing of automatic containment spray pump startup wfll bc performed during each operating cycle. During pump operation water will be recycled to the suppression chamber. Air tests to determine flows to spray headers will also be performed at this time and compared to initial pre-operational air testing, verifying that piping andior noscle conditions have not changed significantly. De sign features are discussed in Volume I,Section VII B.Z.0 ipage VII-19e). The valves in the containment spray system are normally open and are not required to operate when the system is cafled upon to operate. The tesC lntcrval between operating cycles results ln a system faflure probabflity of 1. I x 10 6 Ipffth Supplement. page 11$ e) and ls consistent with practical considerations. Pump operabflity wifl be demonstrated on a morc frequent basis and wfll provide a more reliable system. The intent of specification 3.3.2f is to allov control rcd drive nafntenance ad fpRN replacesent at the tine that the suppression chasber is unvatere4 an4 to parfum noxnal fwl novenent activ-ities in the refwl node vith an unentered suppression chsaber. based on the linited tine involved in pertoxnance ot the con-current naintenance tasks, pxocedural oontrols to ninhaixe the potential ud duration ot leakage fron the control rod drive housing or cpRN penetration an4 available coolant nakeup pm-vidcs adequate protection against drainage ot the vessel vhile the supprecsion chauber is drained. 74 I LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT BASES (Cont'd.) 3.2.3 COOIAuT OIDIISTRY 4.2.3 COOLAIT s3lfuISTRY aafffbbflbfff f ufff bill App)ics. to the reactor coolant systen chsaicsl require- Applies to the periodic testing requirencnts of the ments. reactor coolant chenistry. Ob ective: To assure the chcaical purity of the reactor coolant To dcteraine the chcnical purity of the reactor vatcr. coolant vater. f~lf\

• l

a. The reactor coolant vater shall not exceed the Sasnlcs shall he taken at least evczv 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> and anaivzsd following linits vith stcaaing rates less than for conductivity and chloride ion content. In addition, If Ãaterials in the priaary systea are priaarily 304 stain-l00,000 pounds per hour except as specified in the condwtivity I scones abnosvai (other than short trna less steel and the zircaloy fuel claMinu. Tbe reactor 3.2.3bs snlkcs) ss indicated hv the continuous conducitivity aoni- vater cheaistry Iiaits are establlshe4 to prevent daa-tor, sannlca slsll Ie tHen and analvzcd. aqe to these aaterials. tdaits are place4 on chloride Conductivity Safnho/ca concentration and coxluctivity. The asst Iafpoztant I Chloride ion O.S ISaa llait is that placed on chloride concentration to pre-vent stress oorzosion crackkwz of the stainless steel.

b. For reactor start ups the maximum value for con- when the continuous conductivity aonitor ie inoperable, when the steafaln9 rate is less than 100,000 pounds per ductivity shall not exceed IOxmho/cm and the maxi- a reactor ooolant sfffaple shall be taken at least daily boursf a aors restrictive lialt of 0.1 ppa has been es-mom value for chloride ion concentration shall not  ! and analyxsd for ccfxluctivity and chloride ion content. tablished. At steaainu rates of at least 100,000 pounds exceed 0. I ppm. for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after placing per hour, boikinu occurs causinu deaeration of the re-the reactor In the pofffer operating condition. ~ ctor vater, thus aaintaininu oxygen concentration at lov levels.

c. The reactor coolant water shall not exceed the following I inits with steauing rates greater than A short tera spike is defined as a rise in conductivity or equal to 100,000 pounds pcr hour. such as that vhich could arise froa Infection of addit-ional feedvater flow for a duration of approxiaately 30 Conductivity IOxnho/ca ainutes in tine.

Chloride ion 1.0 ppa

d. If Speclflcatlons 3.2.3.a, b, and c are not net, nozasl orderly shutdofna shall be initiated vithin one hour snd the reactor shall be in the <<old shut-doffn condition vithin ten hours.

e. If tbe continuous oonductlvlty aonitor is inoperable Ihen conductivity is in its proper nozaal range, pN for nore than 7 days the reactor should be placed ln the cold s.'futdovn ccsbdition vithin 2S bours. and chloride and other lapurities affecting conductiv-

! ity aust also be vithin their nozaal range. When and if conductivity becones abnozasl, then chloride aeasure-sents are asde to detezaine vhether or not they are also cut of their nomal operating valws. This vould zxft necessarily be the case. Conductivity coul4 be high due to the presence of a neutral salt, e.g., Ica2304, vhich would not have an affect on pN or chloride. In such a case, high conductivity alone is not a cause for shutdcwn. In soae types of water-cooie4 reactors, con-ductlvities are in fact high due to purposeful addition of additives. In the case of Svg's, hovever, where no additives are used and where neutzsl pN is aaintained, conductivity provides a very good aeasure of the quality 43 1 SURVEILLANCE REQUIREMENT BASES (Cont'd.l Llt>IITING CONDITION FOR OPERATION of the reactor vater. Significant changes therein pro-vide the operator with a varning aechanisa so he can investigate and rcaedy the conditton causing the change before liaiting conditions, with respect to variables affecting the boundaries of the reactor coolant, are exceeded. scethods available to the operator for correct-ing the off-standard condition include, operation of the reactor clean-up systen, reducing the input of ta-purtties and placing the reactor in the cold shutdown condition. The astor benefit of cold shutdovn ls to reduce the tenperature dependent corrosion rates and provide tine for the clean-up systea to re.establish the purity of thc reactor coolant. Durtng start-up periods, which are tn the category of less than l00.000 pounds per hour, conductivity aay exceed 2 uaho/ca because of the initial evolution of gases and the ini-tial addition of dissolved actals. Ouring this period of tine, vhen the conductivity exceeds 2 unho tother than short tera spikes). saaples vill bo taken to assure that the chloride concentration is less than O.l ppa. The conducttvity at the reactor coolant ls conttnu ously aonttored. The saaples of the coolant vhlch are taken every o6 hours vill serve as a reference for calibration of these aonttors and ts considered adequate to assure accurate readings of the aonltors. If conductivity is vithtn tts noraal range, chlorides and other lapurlties vill also be within their noraal ranges. The reactor coolant sanptes vill also be used to dcteraine the chlorides. Therefore, the sawpllng frequeacy ls considered adequate to detect long-tata changes ln the chloride loa content. Hovever, lf the conductivity changes significantly, chloride acasure-nents vill be aade to assure that the chloride liaits of Specification 3.2.3 are not exceeded.

4. ~UES1lON Provide an assessment of the conformance of NMP-1 design and.

operation with the current Regulatory Guides of Division 1 extending the description of the Application from Guide No. 3..21 through No. 3..59. In this regard, those Regulatory Guides addressed in response to particular inquiries herein need. not be repeated..

RESPONSE

The conformance of Nine Nile Point Unit 1 with regulatory guides 1.22 through 1.59 follow.

REGULATORY GUIDE 1.22 PERIODIC TESTING OF PROTECTION SYSTEM ACTUATION FUNCTIONS The Nine Mile Point Unit 1 protection system is designed to permit periodic test-ing to extend to and include the actuation devices and actuated equipment.

The reactor protection system automatically initiates a reactor scram to prevent exceeding established limits. In addition, other protective instrumentation is provided to initiate action which mitigates the consequences of accidents or terminates operator error.

1 A detailed description of the system is available in the FSAR and a description of the capability of testing sensors, channels and channel calibrations has also been previously presented.

As. presently,.designed, the, protection system periodic tests duplicate, as closely as practicable, the performance that is required of the actuation devices in the event of an accident.

The protection system and the systems whose operation it initiates are designed to permit testing of the actuation devices during reactor operation.

Testing of actuation devices and actuated equipment is performed on an individual basis or in selected groups depending upon the particular system design. For example: At least once per operating cycle, power-operated isolation valves (except feedwater and main steam .) are fully closed and reopened. And, at least twice per week, the feedwater and main steam line power-operated isolation valves are exercised by partial closure and subsequent reopening.

A detailed description..ofa3.l,testing .for the,.protection .system and the systems whose operation it initiates, during reactor operation, is available in Tables Specifications. When a sensor check is 3.6.2a-j and 4.6.2-j in the Technical shown, it indicates individual testing. A channel test indicates a group.

Where the ability of a system to respond to a bona fide accident signal is in-tentionally bypassed for the purpose of performing a test during reactor oper-ation:

a. Key-lock bypass switches are provided for high drywell pressure and high water level scram discharge volume. All others are ad-ministratively controlled. This is provided to prevent expansion of the bypass condition to redundant or diverse systems.

b. Each bypass condition is automatically and individually indicated to the reactor operator by means of annunciators located in the main control room.

Actuated equipment of the systems initiated by the protection system are tested periodically during reactor operation and in a manner that does not adversely affect the safety or operability of the plant. However, in some cases full in-itiation of the system is not practical. For example, the containment spray 1 Nine Mile Point Unit 1, FSAR,Section VIII 2 Nine Mile Point Unit 1, Technical Supplement to Petition for Conversion from Provisional Operating License to Full Term Operating License.

system uses an air test and the liquid poison is recirculated to a test tank.

Testing of sensors in the area of main steam line isolation valves can only be done during periods of station shutdown because of high radiation levels. Test-ing of scram, associated with shutdown position of the mode switch also can only be done in periods of shutdown since it always involves a scram.

The probability that the protection system will fail to initiate the operation of the actuate equipment is very low as indicated on page III-12 of the Technical Supplement to Petition for Conversion from Provisional Operating License to Full-Term Operating License. The failure probabilities are low because all protective instrumentation has the capability of being tested and calibrated and has the cap-ability of sensor checks.

REGULATORY GUIDE 1.23 ON SITE METEOROLOGICAL PROGRAMS A wind speed and direction sensor is presently in service atop the reactor

~

building. Read out for this sensor is provided on a strip chart recorder

~

~ in the control room.

A 200 foot high meteorological tower equipped with wind speed and direction and temperature sensors is presently being installed at a location approximately 3000 feet west of the Unit 1 Station and at a base elevation approximately the same as Unit 1 grade.

Temperature and wind sensors are located at 36,100, and 200 foot elevations and relative humidity sensors are located at the 36 and 200 foot elevations. As discussed in Volume II Appendix A of the FSAR,wind speed changes with height.

Therefore,to estimate what it would. be .at the 350.stack .heightia .power law .approximation per 'extrapolation has been and will be used to obtain wind speed at stack height.

Both analog and digital recording will be provided at the tower site for all sensors.

Temperature will be recorded as ambient at 36 feet and differential from 36 to 100 and 36 to 200 feet. In addition, strip chart recorders in the control room will provide analog records of the temperature and. differential temperature and the wind data from the 36 and 200 foot elevations.

All wind instruments will have a directional accuracy of at least +5 degrees and a wind speed accuracy of +0.5 mph with a starting speed of approximately 1 mph.

Zn addition, a low threshold wind measuring system will be installed at the 36 elevation which will provide an accuracy of +0.13 knots (0.15 mph) or 1 percent foot with a starting speed of 0.5 knots (0.6 mph) . Temperature and differential temperature sensors will have an accuracy,,of Qess,than,,0,.5C. -. Relative .humidity ~s accurate to +'3 percent. from 15 to 95 percent relative humidity.

Continuous surveillance of the remote recorders in the control room and weekly inspection of the recorders at the tower site will assure prompt service and high reliability of the meteorological instrumentation. Semiannual calibration checks will be made to assure accuracy.

Monthly and semiannual joint frequency distributions will be maintained of wind speed and direction by atmospheric stability class. A reporting format will be established on a basis acceptable by the AEC.

REGULATORY GUIDE 1.24 - ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A PRESSURIZED >1ATER REACTOR GAS STORAGE TANK FAILURE This guide is not applicable to Nine Mile Point Unit 1, which uses a boiling water reactor.

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REGULATORY GUIDE 1.25 - ASSKIPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT IN THE FUEL HANDLING AND STORAGE FACILITY FOR BOILING AND PRESSURIZED 3/ATER REACTORS The fuel handling accident as described. in the FSAR for Nine Mile Point Unit 1 is the dropping of a fuel bundle into the core from 30 feet, the maximum allowed by the refueling equipment. The expected. number of failed rods is less than 76.

For the FSAR analysis, the reactor was assumed. to be operating at 18/0 hire for 1000 days up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before the fuel assembly drop. A peaking factor of lO was assumed. in that calculation. The depth of water above the core is greater than the 23 feet specified in the guide. This guide assumes total release is over a, 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period whereas we have considered both 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and.

~$ 0'day'do'se"although most of'%he activi'ty 'is release'd immediately.

Table 1.25-1 below compares the doses obtained using the assumptions of this guide with those previously presented. In either case, the results are within the limits of 10CFR100.

Table 1.25-2 gives comparisons of fission products release fractions and trans-port Meteorlogical and dose conversion factors are given in Table 1.25-3. The effects that these factors have in the calculations are illustrated by Table.

As described. in Volume II of the FSAR the filter system for Nine Mile Point Unit 1 was sized for the design basis accident.

14

TABLE l 25-.1 FUEL HANDLING ACCIDENT DOSES (REM)

Safety Guide Technical Supplement (2hr @ 0.78 mi.) (2hr 8 0.'78 mi.)

Thyroid 1.1 x 10 0 g Whole Body 2 1 x 10

TABLE 1'.25"2 FISSION PRODUCT RELEASE ASSUMPTION'S Safety Guide Technical SQ olement lg'(3Q K -'85) 'Nobel Gases l.g 10$ Iodines (total) 0.5jo 99 75$ Inorganic 0.25$ Organic 85$ Filter Reduction None Reactor Building Mixing Yes 133 for inorganic Retention of Iodine 100 1 for organic in Fuel Pool effective decontamination factor 100

TABLE 1.25-3 ATMOSPHERIC DISPERSION AND DOSE CONVERSION FACTORS Safety Guide Technical Su lement Actual Stack Release Height Actual Stack 3."47xlo

,43 m /sec Breathing Rate 3.47xl0 4.

m 3

/sec Iodine Dose Conversion Factors Same (ICRP Committee II-l959)

Infinite Cloud Whole Body Dose Infinite Cloud.

(Centerline Conc.) (Centerline Conc.)

.x/Q,.Factors 0-2 hrs (Site Boundry) 0-30 days (9.5xlO 5funigation) 1.25 x 10

TABLE 1.25-4 EFFECT ON DOSE OF FACTORS USED IN THE CALCULATIONS

~Th aid Whole Bod

'(2 hr'O'O."78'mi.') (2 hr 6 0.78 mi.)

I 1.3 Airborne Activity 17 5 15 Filter efficiency 12.5. Reactor 'building holdup 12 5 75 Meteorology 75

REGULATORY GUIDE 1.26 QUALITY GROUP CLASSIFICATIONS AND STANDARDS A listing of the systems which fall into the three quality groups described by this guide is included in Table 1.26-1. 'The codes applicable to Nine Mile Point Unit 1 components at the time of construction are given in response to Regulatory guide 1.48. Any future modifications will be in accordance with standards appli-cable to the quality g'oup as classified here.

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TABLE 1.26-1 UALITY GROUP CLASSXFICATXONS Qualit Grouo B Quality Groun C Qualit Grou D Reactor recirculation Reactor Cleanup Fire protection (foam 6 water)

Main steam Reactor G Waste bldg. closed Drywell sump pump piping Low and High Pressure feedwater loop cooling Resin transfer 6 regeneration Condensate Service Water Sulphuric acid transfer Control rod. drive Radioactive waste disposal Hydro pump discharge Core spray Off-gas Turbine bldg. closed loop Liquid Poison Diesel generator fuel oil, starting cooling Shutdown cooling air, and cooling water Screen washing Head. spray Instrument and breathing air House service air-Emergency Condenser Drywell vent and purge Roof and floor drains Containment spray Control room ventilation City water Reactor bldg. emergency ventilation Drywell and tores vacuum relief Reactor instrumentation Drywell Instrumentation and leak Reactor vent and drain monitoring Fuel pool filtering and. cooling

REGULATORY GUIDE 1.27 - ULTIMATE HEAT SIlK The ultimate heat sink for Nine Mile Point Unit No. 1 is Lake Ontario. This includes the primary and secondary forebays as shown in Figures III - 16, 17 of Volume I of theFSAR. The following systems obtain water from this heat sink.

1. Main condenser 8c Circulating water

2. Service water 3: Fire pumps

4. Containment Spray raw water pumps

5. Diesel generator cooling water pumps In case of blockage of one entrance to the forebay another entrance exists

'to a11ow water in from the Lake. The volume of water available from the lake is virtually inexhaustable and is sufficient to last longer than the required.

30 days.

As discussed in response to regulatory guide 1.59, flooding is possible but this would not have any adverse effect on the ultimate heat sink. Other phenomena such as earthquakes of large enough proportions to empty the lake are not considered credible.

Should an abnorma1 occurrence associated with the lake arise, the following Technical Specifications would be applied:

'pecification Specification Specification 3.1.3 - Emergency Cooling System

,3.3.7 - Containment Spray 3.6.3 Emergency Power Sources These specifications give appropriate action to 'be taken.

16

REGULAEORE GUIDE I. 28 - QUALIEE ASSURANCE PROGRAM REQUIREMENTS DESIGN AND CONSTRUCTiON It is the intent of the Niagara Mohawk Power Corporation to comply with 10CFR50 Appendix 3, Quality Assurance Criteria for Nuclear Power Plants and. Fuel Reprocessing Plants. Niagara Mohawk Power Corporations Quality Assurance Ywnuals are currently being revised. to comply with ANSI-45.2-1971.

17

REGULATORY GUIDE 1.29 SEISMIC DESIGN CLASSIFICATIONS The structures, systems and components, including their foundations and supports designated as Class I in the FSAR Second Supplement are designed to remain func- ~

tional*following the Maximum Credible Earthquake outlined in the (PHSR) Prelimin-arey Hazards Summary Report, Volume II, "Engineering Seismology." These structures, systems and components are listed in the Final Safety Analysis Report Second Supple-ment. The criteria of this guide are met with the following exceptions:

(1) Structures:

The spent fuel storage pool fuel racks are not specifically designed to resist eqrthquake forces and axe not listed as Class I components.

The false ceiling and lights in the control room are not designated as Class I components and are not designed to resist earthquake forces.

The primary containment is designated a Class I system and stresses resulting from various loading conditions combined with earthquake forces meet code requirements. However, the suppression chamber columns were not allowed the 1/3 stress increase.

(2) ~sstems:

"'Three piping systems were not designed to Class I criteria as set forth in this regulatory guide. These are:

(a) Main stream - after isolation valves (b) Reactor head spray (c) Reactor clean-up (3) Electrical:

All contxol boards, equipment and devices mounted on the control boards were designed to remain operable during and after a seismic disturbance of the in-tensity described below. The design is such that no control device or relay shall malfunction causing any inadvertent operation because of such seismic disturbance.

The maximum ground motion acceleration for the design seismic distrubance is eleven percent of gravity and the maximum resulting response acceleration is forty-five percent of gravity for oscillators in the period range of 0.2 to 0.3 seconds.

The diesel generators as wel3. as the emergency service portion of the A-C power distribution system are designed. and. built to meet the same design as mentioned above. criteria Miscellaneous electrical components (limit switches, local starters, local controL panels, etc.) were not specially designed to withstand. the effects of an earthquake.

18

However, these components are either located. in Class I structures or are identical to Class IE components that are documented. to comply with the seismic design criteria stated. above.

The systems containing these components are:

(a) Feedwater System in the Turbine Building (b) Spent Fuel Pool Cooling Control Panel located in the Reactor Building (c) Portions of radioactive Waste System located in the Haste and Turbine Buildings.

"Ele'ctrical 'cab1es asso'ciated with the protection system and the emergency service portion of the A-C distribution system, including the on site electrical power sources, are routed. in cable trays located in the turbine building

REGULATORY GUIDE 1.30 - UALITv ASSURANCE REQUIREMENTS FOR THE INSTALLATION INSPECTION AND TESTING OF INSTRUMENTATION AND ELECTRIC E UIPIKNT Adequacy relevant to.IEEE-336-1973. has been discussed. previously in the "Technica1 Supplement to Petition for Conversion from Provisional Operating License to Full Term Operating License" on pages III-60, 61. To assure the quality of the instrumentation and electrical equipment the following were applied..

(1) Detailed specifications were issued.

(2) Selected. shop surveillance

,(.3) .Field.inspections,,and tests (4) Field qua1ity control checks (5) Installation in accordance with approved. drawings and instructions (6) Preoperational testing Since this time IEEE-336-1971 has been incorporated as ANSI N45.2.4 "Installation, Inspection, and Testing Requirements for Instrumentation and Electric Equipment During the Construction of Nuclear Power Generating Stations". This will be used as a guide in conforming with ANSI N45.2

".Quality Assurance Program Requirements for Nuclear Power Plants".

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REGULATORY GUIDE 1.31 - CONTROL OF STAINLESS STEEL WELDING Weld material met the requirements of Section III of the ASME Code. Ferrite content was controlled in the filler metal itself so that a small amount of ferrite was present in completed welds. Experience has shown that these controls are adequate for producing satisfactory welds in austenitic stainless steel without any problem of fissuring or hot.:cracking.

Except for repair welds in pump and valve castings, ferrite welding control was accomplished by either mainta'ining a minimum of 5 percent ferrite in the filler material or a chromium to nickel ratio of at least 1.9 to 1. The chromium to nickel ratio was also used to control ferrite content of the castings themselves.

Heat input during welding was specified in .termsof amperage, voltage and inter-pass temperature limits.

The above procedure was felt adequate to control stainless steel welding to pre-vent. fissuring or hot cracking.

21

0

'EGUIATORY GUIDE 1:32 - USE OF IEEE STD 308-1 71 CRITERIA FOR CIASS 1E ELECTRIC SYSTEMS FOR ItUCLWR POLY GZiP~BATDlG STATIONS Adeauacy relative to IEEE Std. 308-1971 has been addressed. on pages III-50 thru 56 of the "Technical Supplement to Petition for Conversion from Provisional Operating License to Full Term Operating License."

This regulatory guide identifies two potential conflicts between IEEE Std. 308-1973. and General Design Criterion 17 of Appendix A to 10CFR50.

The design of Nine Mile Point Unit 1 complies with the regulatory position.

Nine Mile Point Unit 1 has two independent circuits norae1ly available

.from.the, transmission network. Also, -the.'battery-charger supply capacity for Nine Mile Point Unit 1 is based. on the largest combined. demands of the various steady state loads. They are also capable of supplying sufficient capacity to restore the batteries from the design minimum charge to their fully-charged state while supplying normal steady state loads.

'22

REGULATORY GUIDE 1.33 UALITY ASSURANCE PROGRAM RE UIREMENTS (OPERATION)

A response to the regulatory position as stated in this guide is, included in the response to a July 26, 1973, letter from the AEC.

This is included as response number 8 in the attachment to the letter from Mr. P. D. Raymond to Mr. A. Giambusso dated November 16, 1973.

23

'I REGULATORY GUIDE 3..34 - CONTROL OF ELECTROSLAG klELD PROPERTIES The electroslag ~fielding process vas not used for components imoor ant to safety at Bine Nile Point Unit 1.

24

REGULATORY GUIDE 1.35 - INSERVECE SURVEILLANCE OF UNGROUTED TENDONS IN PRESTRESSED CONCRETE CONTADTbKNT STRUCTURES This guide is not applicable to Nine Mile Point Unit 1 which uses a steel primary containment.

25

REGULATORY GUIDE 1.36 NON-METALLIC THERMAL INSULATION FOR AUSTENETIC STAINLESS STEEL The non-metallic insulation used at Nine Mile Point Unit 1 was manufactured by Pittsburg Corning Corporation. The insulation conforms to Military Speci-fication MXL-I-278 ID as described under 1.2 Classification Grade IX. Class C and Grade III, Class F. In addition, "Unebestos" insulation is covered by the Department of the Navy in Pittsburgh, Pennsylvania under QPL-2781.

Certified chemical analysis tests were performed on 6 samples under MIL-1-24244.

The results of these tests showed that leachable chlorides ranged from 86-102 ppm, and that leachable sodium plus silicate ranged from 135,000 to 150,000 ppm.

'This analysis Cal'ls 'in the acceptable range as shown on figure 1 of the guide.

Xn addition, the tests results did not vary by 50 percent from one another.

26

REGULATORY GUIDE 1.37 UALITY ASSURANCE RE UIREMENTS FOR CLEANING OF FLUID SYSTEMS AND ASSOCIATED COMPONENTS OF HATER-COOLED NUCLEAR POWER PLANTS The cleaning process at Nine Mile POint Unit 1 for the reactor vessel and primary system consisted of using 0.5 percent trisodium phosphate heated to 180F. Chloride content was kept to less than 10 ppm. The-system was then flushed with demineralized water equal in quality to that used during opera-tion.

Procedures are presently being developed at, Nine Mile POint Unit.l using ANSI N45.2.1 as a guideline.

27

REGULATORY GUIDE 1.38 UALITY ASSURANCE RE UIREMENTS FOR PACKAGING, SHIPPING, RECEIVING, STORAGE, AND HANDLING OF ITEMS FOR WATER-COOLED NUCLEAR POWER PLANTS Procedures are being developed at Nine Mile Point Unit 1 to comply with this guide. ANSI N45.2.2 1972 is being used as a guide in their preparation.

28

REGULATORY GUIDE l-39 HOUSEKEEPING RE UIREMENTS FOR WATER COOLED NUCLEAR POWER PLANTS Procedures are presently under development to control work activities, l

conditions and environments at Nine Mile Point Unit No. using ANSI N45.2.3-l973 as a guide for their preparation.

29

REGVIATORY GVIDE 1.40 QUALIFICATION 1ESTS OF CONTINUOUS DU1Y ROTORS INSTALLED INSIDE THE CONTKIRKNT OF HATER-COOLED NUCLEAR PO)/ER PLAVZS Adequacy relative to IEEE-334 has been discussed on page III-59 of the Technical Supplement to Petition for Conversion From Provisional Operating License to Full-Term Operating License. Were are no continuous duty Class-I motors installed in the containment. Therefore, Regulatory guide 1.40 does not apply to Nine Mile Point Unit l.

30

REGULATORY GUIDE 1.41 - PREOPERATIONAL TESTING OF REDUNDANT ON'-SITE ELECTRIC PONrR SYSTEMS TO VERIFY PROPER- LOAD GROUP ASSIGNIKNTS A rigorous preoperational test was performed for the on-site electric power system. The isolation was effected by direct actuation of'he undervoltage sensing relays on Power Boards 102 and, 103.

Each on-site electric power system (D.G. 102 5 103, Power Boards 102 and 103, and all redundant engineered. safeguard load. groups fed from these power boards) was tested independently and each test included. the following:

(1) Injection of simulated accident signa1s.

'(2) 'Startup of the 'diesel generator (3) Startup of and sequencing of redundant load group under test (4) Functional performance of the loads The duration of each test permitted. the system to achieve stable operating conditions and no adverse conditions were detected.

During each test the D-C and on-site A-C buses and related loads not under test were not required to be disconnected.

31

REGULATORY GUIDE 1.42 - INTERIM LICENSING POLICY ON AS LOW AS PRACTICABLE FOR GASEOUS RADIOIODINE RELEASES FROM LIGHT-WATER-COOLED NUCLEAR POWER REACTORS A revised off-gas system is being installed at Nine Mile Point Unit 1. This upgraded system will reduce the off-site exposure to less than 0.025 mr/yr.

This system provides 20 days holdup for Xenon and 33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> for Kryptons. A flow chart for the system is presented on pg III-27 of the Technical Supplement to Petition for Conversion from Provisional Operating License to Full Term Operating License.

In addition, valve leakoff hoods have been placed on valves in, areas where there have been high radiation levels. These hoods vent directly to the venti3; lation,.system.

The tyroid dose to a child from radioiodine intake via the milk pathway has been calculated using the assumptions as outlined in Appendix A of this guide.

Assuming an annual release of 0.38 Ci of I-131, the dose conversion factors of Appendix C of this guide and taking into account that the nearest dairy farm which is 0.64 mi southwest of the station, the resultant dose is 3 mr/yr., This is based on a X/Q of 4 x 10-8 secs/m 3 and an offgas release rate of uci/rec after 30 minutes decay.

An environmental monitoring program as described in Section 4.3 of the Prepared Environmental Technical Specifications will limit the dose to a childs thyroid to less than 15mrec/yr for the entire site.

1. Letter'from Mr. P. D. Raymond to Mr. Daniel R. Muller dated October 4, 1973.

32

REGULATORY GUIDE L.II3 - COIIEROL OF STAINLESS SEEEL IIELD CLADDDlG OF LOS ALLOY STEEL COMFOli~TS Tliere are no stainless steel cladding welded. to low-a1loy steel components important to safety at Nine Mile Point Unit 1.

33

REGULATORY GUIDE 1.44 CONTROL OF THE USE OF SENSITIZED STAINLESS STEEL The unstabilized, austenitic stainless steel of the AISI Type 3xx series used in the systems mentioned in tne guide was all solution annealed per the ASTM standards at the time of order. Material was tested to ASTM A 262-70 to de-termine the degree of sensitization.

The material used at Nine Mile Point, Unit 1 is 304 stainless and is furnace sen-sitized. No "L" grade material was used at Nine Mile Point Unit 1. On material other than welded, testing was performed to ASTM A-262 or material was resolution heat treated. No demonstration as to susceptibility to intergrannular stress corrosion was provided.

..The cleaning .process involved the use of 0.5 percent TSP solution which was circulated and heated to 180F. Chloride content was kept to ~ 10ppm. As de-scribed in the Sixth Supplement to the FSAR systems cleaned with TSP and then flushed with demineralized water act to inhibit chloride attack.

Welding practices were such as to prevent excessive sensitization.

34

REGULATORY GUIDE 1.45 - REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTION SYSTEMS There are two systems which collect leakage in the drywell at Nine Mile Point Unit 1. One system, classified as identified leakage, collects leakage from

~

~

drywell equipment drains. This can detect a change of approximately 1 gpm in the 8-10 gpm flow range in 20 minutes.

Unidentified leakage is measured by the rate of rise of sump water level in the'rywell. For sump inflows of one gpm, changes in the order of 0.2 gpm can be detected within 40 minutes. At inflows between one and five gpm, changes in the order of 0.5 gpm can be detected in eight minutes.

A second type of unidentified leak detection system monitors airborne particulates.

This would indicate any leakage from the primary system by showing a rise in activity.

A third, method employed is the analysis of isotopic composition of effluents from the drywell air coolers, the clean up system, or water from condensed reactor steam. In addition, drywell temperature, humidity and dew point are monitored which would alert the operator of abnormal leakage conditions.

Alarms in the control room indicate abnormal leakage. The Technical Specifications define the limits for unidentified leakage as 5 gpm and total leakage as 25 gpm.

The leak detection systems although not designed as seismic class I are located in seismic class I structures whose integrity will remain during earthquake conditions.

REGULATORY GUIDE 1.46 - PROTECTION AGAINST PIPE IrHiIP INSIDE THE PRIMARY CONTAIiSKNT

~

As presented in the answer to question 10, there is adequate redundancy to maintain core cooling capability. An analysis made to determine the effect on the primary containment shows that a deformation can occur and in one case exceed 50 percent of the'ltimate strain. The only case in which this can occur is for one of the recirculation loops. This will be protected by appropriate pipe whip restraints.

36

REGULATORY GUIDE 1.47 BYPASSED AND INOPERABLE STATUS INDICATION FOR NUCLEAR POWER PLANT SAFETY SYSTEMS The reactor protection system consists of two independently powered logic channels. Each logic .channel consist of two independent trip logic channels.

This arrangement is commonly called one out of two taken twice logic.

The system as designed is normally" energized. It is designed to trip when de-energized and fail safe on loss of electric power. During normal operation of the station, operating bypasses which are automatically or deliberately induced are indicated in the control room. Reactor pressure below 600 psi automatically bypasses main steam line isolation valve closure scram and low condenser-vacuum scram. Control room indication is "Condenser Vac Bypass."

Reactor pressure less than 850 psi is bypassed to prevent main steam isolation in all reactor operating modes except "Run". Control room indication is "Reactor Press Low.".Reactor scram as a result of a turbine trip is bypassed below 40 percent power. Control room indication is "Turbine 1st State Press Low" Other bypasses are operated from the control room. Control room indication of "CRD Scram Discharge Volume Bypassed" is key operated but only operates in shutdown or refuel modes. Control room indication of "Dry Well Press Bypassed" is also key operated.

Emergency Ventilation System operation is changed during the refueling mode so that operation is from.radiation monitoring on the refueling platform,

""Emergency "Vent -'System'Bypassed, Refueling 'Mode". 'This provides more sensitive monitoring during refueling operations.

Bypass switches on neutron monitoring systems are not alarmed since the switches are located on the operating console under the visual control of the operator.

Visual surveillance of the console can determine which neutron systems are by-passed. Engineered safeguards such as core spray and containment spray have no bypasses in their control circuits. Breaker motor control'voltage is monitored and alarmed in control room.

During the surveillance testing of the reactor protection system, deliberate bypassing of some sensors are necessary to perform the required tests. The bypassing of required sensors can only be accomplished by isolation of such devices or of the attachments which block or prevent the actuation of signals.

This procedure is under the control of the control room operators. Surveillance test procedures describe the means for bypassing when required. Although a particular sensor in a.'.logic channel may be bypassed during testing other process sensors in the same logic channel are fully operable and capable of performing their intended function. A redundant sensor in the redundant logic channel is operable and capable of performing the intended function of the sensor being tested should the requirement arise during the test. During surveillance testing only one sensor is tested at a time.

37

REGULATORY GUIDE 1.48 DESIGN LIMITS AND LOADING COMBINATIONS FOR SEISMIC CATEGORY I FLUID SYSTEM COMPONENTS The following codes were used in the design of the Nine Mile Point Unit 1 seismic Class I systems important to safety.

a) ANSI B 31.1 1955 b) ANSI B 16.5 1955 In addition, to these code recpxirements various other items such as seismic and thermal analyses were performed. Ultrasonic and radiographic examinations were made. An independent third party review of calculations was performed by Teledyne Materials Research and were found acceptable.

38

REGUIATORY GUIDE 1.50 - CONTROL OF PREHEAT TEMPERATURE FOR WELDING OF LOW-ALLOY STEEL There are no structures or components important to safety which use low alloy steel.

40

REGULATORY GUIDE 1.51 INSERVICE INSPECTION OF ASME CODE CLASS 2 AND 3 NUCLEAR POWER PLANT COMPONENTS Response to the position of this guide is covered in answer to Question 9 of this submittal.

41

1 REGULATORY GUIDE 1 52 DESIGN TESTING p AND MAINTENACE CRITERIA FOR ATMOSPHERE CLEANUP SYSTEM AIR FILTRATION AND ABSORPTION UNITS OF LIGHT-WATER-COOLED NUCLEAR POWER PLANTS At Nine Mile Point Unit 1 the reactor building emergency ventilation system mini-mizes the radioactivity released from the reactor building.

The entire system is located within the turbine and reactor building. The system is designed to operate under the most severe overall building environment of 150F, 100 relative humidy and + 0.5 psig.

The emergency ventilation system is designed to handle source terms comparable to those of Regulatory Guides 1.3 and 1.25.

The system is redundant and consists of inlet duct work taking suction from the normal reactor building discharge. There are dual banks of filters for removal of particulates and halogens with a motor driven blower in each. Also, there are redundant flow controllers, indicators and exhaust ducting to the stack. The flow diagram for the system is contained in the Sixth Supplement to the FSAR. The system has been designed to seismic Class I standards and is housed in seismic Class I and Class II structures, the reactor and turbine buildings.

For each of the two loops position of inlet and outlet valves, discharge flow rate and low flow are indicated 'in the control room. High differential pressure across the filters is indicated in the control room also.

A 10 kw heater is placed before each filter train to lower the relative humidity from 100 percent to 70 percent. There is also provision for 60 cfm of turbine building air to cool filters due to fission product decay heat. This flow limits temperature to 200F for the design basis accident.

Each system is capable of 1600 cfm flow rate which can hold the reactor building at a negative pressure of 0.25 inch water gage to keep building exfiltration to a minimum.

All materials associated with the system including filters were designed to per-form their required function even under the possible radiation effects associated

...with the design basis accident.

The power to the syst: em is from redundant power boards, one supply to each filter train. These are in accordance with IEEE-308. Instrumentation is in accordance with IEEE-279.

There are no bypasses which allow unfiltered air from bypassing the filter train.

In addition, no outdoor air intake openings exist for this system.

The high efficiency particulate filters are Cambridge filters suitable for air temperatures up to 550F.. The filter media is glass asbestos with aluminum sep-arators. The filter elements are in zinc plated carbon steel and the filter housing is galvanized steel. The elements are of such a size as to be disposable in a 55 gallon drum.

The charcoal filter train consists of six trays, each of which are 20 inches wide, 31 inches long and 7.5 inches deep. Each bed is 2 inches deep and has a flow area of 4.36 square feet with a dwell time of 0.3 seconds. The filter casings are type 304 stainless steel and the unit housings are of galvanized steel.

42

The system was tested at design flow and each filter train tested individually.

Penetration of the particulate filters ranged from 0.002 to 0.022 percent penetra-tion. The charcoal filters tested at 99.9 percent efficiency. The charcoal filters are. MSA85851.

The system is designed for ease of maintainance and testability. Bolted covers close off the filter access openings. The housings are easily accessible from the side since the banks are stacked on top of each other. The filters themselves are easily disposable in 55 gallon drums. Special tools for filter removal, re-placement and testing are provided. At least 2 inches of clearance exists be-tween filter elements. The filters are in rigid housings eliminating any special alignment procedures. There are permanent test taps for performing the required testing. Electrical, water and compressed.air are available outside the filter housing.. There is no lighting inside the filter housings.

The particulate filters were also shop tested for minimum efficiency of 99.97 percent with dioclylphthalate before their installation.

Technical Specifications have been developed, limiting condition for operation 3.4.4 and Surveillance Requirements 4.4.4 cover system operability and testing requirements to assure that it performs its intended function when called upon.

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REGULATORY GUIDE 1.53 - APPLICATION OF THE SINGLE-FAILURE CRITERION TO NUCLEAR POWER PLANT PROTECTION SYSTEMS The prime objectives in the design of the Nine Mile Point Unit 1 reactor protection system were:

(1) Component or channel failure does not. interfere with the proper operation of its redundant counterpart.

(2) All potential single failures are detectable failures either by periodic tests, anamolous indications, or by alarms.

An orderly single failure analysis was performed on the Nine Nile Point Unit 1 reactor protection system prior to preoperational testing. Potential undetectable failures were identified and assumed to be in their failed mode in this analysis.

The specified surveillance requirements in the Technical Specifications defines the testing procedure to include the final relay action resulting from sensor and instrument channel testing.

The operating mode switch at Nine Mile Point Unit 1 does provide signals to redundant protective channels. The redundant circuits are supplied from inde-pendent switch sections which are separated by bakelite barriers on the switch itself.

The collective protection system and logic-actuator system has been analyzed for single-failure modes which could disable control power for one channel and for the redundant actuator circuit. This condition does not exist for the Nine Mile Point Nuclear Station Unit. gl.

The..single--failure analysis included a .systematic investigation of potential faults

~

and failures on a per channel basis.

~

~ It was determined that the system could

~

withstand the loss of a redundant channel and still perform its protective function.

Interconnections between channels were analyzed and it was determined that, no component failure> short circuit, open circuit or ground could cause the loss of a protective function.

The logic system was analyzed for single failure in a manner similar to the single failure analysis performed on the channels. No single failures in the system logic caused multiple= failures in the channels or actuator circuits that would violate the single-failure criterion.

A single failure analysis was also performed on the actuator circuits. The analysis assured that no single failure could cause a significant loss of function due to an improper connection of the actuators to a source of energy.

The location and arrangement of protection system equipment were analyzed and was determined that an acceptable degree of separation and independence exists it to prevent loss of a protective function due to missiles, fire, flooding, earthquake, temperature and chemicals. ~

It was determined that a stack failure in the right direction could cause loss of the emergency buses, the normal A-C supply to these buses and the emergency on site A-C supply to these buses.

In the overall system-failure analysis, interconnections were analyzed to show that a single failure of a channel, component or actuator would still result in the remainder of the protection system meeting the single failure criterion.

44

REGULATORY GUIDE 1.54 - UALZTY ASSURANCE REQUIREMENTS FOR PROTECTIVE COATINGS APPLIED TO WATER-COOLED NUCLEAR POWER PLANTS Procedures are presently being developed at Nine Mile Point Unit 1 to control the use of protective coatings. The procedures are being prepared using ANSI N 101.4-1972 as a guide.

0 REGULATORY GUIDE 1.55 - CONCRETE PLACEbZNT IN CATEGORY I STRUCTURES The design of Nine Mile Point Unit 1 was performed by the Niagara Mohawk System Project Engineering Department. Communication between the Niagara Mohawk design office and the field constructor was direct. The Niagara Mohawk design office had a'ield representative who was liaison between the office and the constructor.

This liaison assisted in the communication of ideas between Office and Constructor.

The Niagara Mohawk design office sent specifications, drawj ngs, revisions of these, plus any other instructions, to the constructor with a copy to the Niagara Mohawk field representative. Thus, the contractor had two sources for consultation or direction. The Niagara Mohawk design personnel visited the work site on a frequent schedule and had direct telephone communication via company tie-line among all parties concerned.

.Tne.System Project Engineering:Department-was responsible for all design and drawings of Class I reinforced concrete, structures. The design and drawings were prepared in accordance with the ACI codes, applicable, at that time. The primary code which applied was ACI 318-63 Building Code Reauirements for Reinforced Concrete. Control of design drawings consisted of a design progress system showing current progress and checkoff of the required approvals (Responsible Design Engineer, Chief Mechanical.and/or Chief Electrical Engineer, and the Chief Structural Engineer).

All structural designs were reviewed by the Chief Structural Engineer or his assistant.

Placement of reinforcing, watersteps, embedments, construction joints, congestion of reinforcing and other embedded items were shown in considerable detail on the drawings. Large scale sketches developed the more congested areas to assure proper placement. Manufacturers reinforcing detail drawings were completely checked for compliance with Niagara Mohawk design requirements. The type and. required strength of concrete was stated on the drawings.

Design personnel had considerable experience in the use of mass concrete in prior designs. The requirements and techniques necessary for placing large complicated concrete pours was continually reviewed.

Prior to the start of concreting in any segment of the work, the Constructor would discuss the work at meetings with the various contractors (piping, electrical etc.)

and. with his field engineer and inspectors. The contractor also had ~ediate access to design personnel, if design changes were required due to field conditions.

A check-out routine for each item as completed consisted of an inspection report card where signatures of both the foreman for the contractor and the assigned inspector were entered after completion and approval. Final sign-off was by the Chief Field Engineer. Check items included:

(1) Preparation of surfaces to receive concrete (2) Formwork - for line and grade (3) Formwork - construction (4) Installation of reinforcing steel, embedded items, box-outs, piping, electrical, etc.

(5) Final clean-up in preparation for concrete Before concreting commenced, the constructor's (inspector) reviewed details of the concrete such as:

(1) Class concrete specified - (strength, slump, etc.)

(2) Readiness of concrete plant (3) Suitable equipment for delivering and handling concrete (4) Placement technique (5) Pr'eparation for protecting concrete during placement (6') Preparation and material for curing concrete

0 REGULATORY GUIDE 1.56 - VAINTENANCE OF WATER PURITY IN BOILING WATER REACTORS The condensate demineralizers at Nine Mile Point Unit 1 have been deisgned. and operated to permit an order~ shutdown of the reactor in case of high coolant conductivity levels. Also sufficient instrumentation available to monitor the following:

1) Conductivity of condensate

2) Available capacity of the demineralizers

3) Purity of demineralizer effluent Maximum limits on the reactor coolant composition and conductivity have been set as outlined in Specification 3.2.3 of the Station's Technical

-Specifications. They-are as Tollows:

TABLE l.56-1 TECHNICAL SPECIFICATIONS Conductivit Chloride 10u mho/cm O.l ppm Steaming rate g 100,000 lb/hr 2u mho/cm 0.1 ppm 0 Seeeming'rene ) 100;000 lb/hr . 5u mho/cm 0.5 ppm These rates reflect the prepared changes made to the Technical Specifications in response to Question 3c.

REGULATORY GUIDE 1.56 Start-up 10u mho/cm 0.1 ppm Steaming rates(100,000 lb/hr 5u mho/cm 0.1 ppm Steaming rates +100,000 lb/hr 5u mho/cm 0.5 ppm A recording conductivity meter with a range of 0-10u mho's/cm is located between the hotwell oulet and. the inlet to the demineralizers. This meter is calibrated. using certified. conductivity solutions.. A recording flow meter is also used. to measure flow rate through each demineralizer. There are three conductivity meters which monitor feedwater conditions as follows:

(1) Outlet of each demineralizer (0-lu mho/cm)

(2) Outlet of each demineralizer (0-10u mho's/cm)

(3) Upstream of feedwater isolation valves (0-20u mho's/cm) 47

0 These recording" conductivity meters alarm in the control room.

Initial exchange capacities of resins are not measured and. total capacities are not determined after operation. However, the regeneration cycle is closely monitored. $ lhen a resin batch does not rinse down properly after regeneration, it is replaced.. This is done in place of reconditioning the resin. Resin replacement is at such a rate that, the average life of a resin batch is five years and as such precludes significant loss of exchange capacity.

Total flow and. operating hours are used. as the criteria for regenerating the demineralizers. The minimum residual demineralizer capacity is not determined. Due to the water purity required. for operation and. the use of high flow rate condensate demineralizers, quality of demineralizer influent wi11 have,a.large effect on the demineralizer.-effluent quality long before residual capacity is exhausted. Therefore, the demineralizers are normally regenerated.'before they reach a minimum exchange capacity which would. require regeneration.

Because of the water quality required and. the use of highflow rate demineralizers, a change in conductivity of the demineralizer effluent is a more significant factor than ion breakthrough. Therefore, there is no need to determine the quantity of principal ions.

Due to the normal operating range of condensate purity the demineralizers are regenerated 'based on throughput long before residual .capacity would effect quality.

,Chloride concentration is. determined by ASTM Standard Method D512-67, Reference Method C. To ensure that conductivity and chlorides are at their lowest practical level, concentrations at the demineralizers outlet are normality maintained. at the levels shown below; Conductivity O.lu mhoicm Chloride 10 ppb The reactor water is normally maintained at a and a chloride concentration of ( (

conductivity of 1.0~mho's 0.1 ppm (10 percent of the Technical Specification limit) for steaming rates ) 100,000 lb/hr. For rates less than 100,000 lb/hr the conductivity is maintained at ( 2p.mho's and chloride of ~

0.1 ppm (same as Technical Specification limit).

48

REGULATORY GUIDE 1.57 - DESIGN LIMITS AND LOADING COMBINATIONS FOR METAL PRIMARY REACTOR CONTAINYiENT SYSTEM COMPONENTS The Niagara Mohawk specification for the Nine Mile Point Unit 1 Containment dated September 10, 1964 with revision 1 on November 25, 1964 defines the requirements for the reactor containment system. The mandatory code addendum is the Minter 1973 Addendum to Section III - of the ASME code.

The l963 Code requires that stress levels during test cannot exceed 90 percent. of the tabulated yield strength at test temperature for a primary membrare stress intensity.

Furthermore, the primary membrane plus primary bending stress intensity shall not exceed 125 percent of the tabulated yield strength at test temperature. The membrane stresses in,the Nine Mile Point Unit 1 containment have been limited to 1.15 times the basic. Code allowables. This is substantially lower than those stress levels permitted by NE-6222 and NE-6322.

Loading conditions for which the containment function is required to sustain, in combination with specific seismic events, that is 1/2 S.S.E. and S.S.E. were not defined in the l963 code. The conservatism of that code and the effect of changes in the technology of seismic analysis, enhance the possibility that this vessel would meet the current code.

H For the purpose of determining the effect of jet impingement, the Chicago Bridge and IronCo.has generated a report entitled "Loads'n Spherical Shells" dated Aug. 1964.

,For the portion backed'y concr'ete with a 2 inch air space between the shell and concrete,. a.600,jet force was applied .resulting in a maximum displacement equal to3 inches without rupture.

In additi,on, General Electric Company has generated a topical report entitled, "Jet Loading in Primary Containment Vessel", by V. R. Netzel and J. P. Brackton, dated April, 1972. his report was intended to analyze the shell when subjected to jet impingement loads and -to determine that the shell would not rupture. The report demonstrates the capability of the containment vessel to withstand jet forces. The vessel shell, where not back-jap by concrete, is designed to resist

'these jet loads with maximum allowable stress values limited to 90 percent of the yield of the material at temperature in accordance with the allowable stresses.

For that part of the vessel that has a 2 inch air space, the allowable strain in the steel structure may be conservatively based on the fatigue criteria of Section III of ASME. The evaluation of allowable strain is based on Figure 1-9-1 of the 1971 issue of this Code. For carbon steel material based on 10 cycles of fatigue loading, the quarter cycle allowable strain is shown to be at least 3.9 percent.

The results of the report indicate that the maximum strain to be found is approximately 3.3 percent and therefore within the ASME criteria.

FATIGUE EVALUATION Paragraph NB 3222.4 (d) defines various criteria by which it can be shown that the vessel is exempt from a fatigue analysis., Although there are no calculations in the Stress Report to substantiate that this vessel qualified for the exemption several similar various BWR type containment vessels which have had fatigue calcu-lations indicate that the vessel qualifies for this exemption. The Summer 1972 Addendum to ASME III, Paragraph NE-3131 (d) indicates that the fatigue analysis does not have to include cyclic activity due to earthquake. Paragraph NE-3222.4(d) 49

of Section III of ASME defines significant cycles as those which will provide a stress level to 3 S . None of the earthquake stress cycles would provide a stress level greater than Sm and therefore, none of the earthquake cycles are normally

.considered. to be significant.

EXPANSION BELLOWS The containment has 10 - 90 inch expansion bellows connecting the vent lines to the suppression chamber. These expansion bellows are positioned inside of the chamber and are designed for an external pressure of 35 psi. At the time of fabrication of the Nine Mile Point Unit 1 containment the Code did not include any design criteria for expansion bellows. Recent Addenda include criteria for establishing minimum burst ~ressure, meridional permanent strain, an analysis to determine instability, and a definite program by which each of these will be demon-strated. Paragraph NE-3810 also includes a requirement for a cyclic life determination.

This criteria did not exist at the time this vessel was designed.

50

REGULATORY GUIDE 1.58 UALIFICATION OF NUCLEAR POWER PLANT INSPECTIONS AND TESTING PERSONNEL The formal indoctrinational and training program for personnel performing quality control related activities consists of:

Familiarization with the content of:

a) Regulatory criteria such as Appendix B to 10 CFR 50; b) Niagara Mohawk Quality Assurance Manuals 6 Procedures; c) Niagara Mohawk Quality Control Procedures; d) Regulatory Guides;

-e') Safety 'Analysis Reports; f) Files of Quality Control Type Records; g) Engineering codes such as ANSI and ANSI Standards 2~ On-The-Job Trainin Personnel performing Quality Control related activities are assigned responsibilities within an active project. Their activities are reviewed by a supervisor or other experienced member of the Quality Assurance/

Quality Control organization. On-the-job training includes:

a) Auditing, i.e. planning, preparation, conduct, reporting, responding and follow-up; b) Reviewing of purchase documents for the adequacy of Quality Assurance content; c) Preparation or revision of detailed procedures implementing the Nine Mile Point, - James A. FitzPatrick Site Quality Control Procedures; d) 'reparation of responses to inquiries from Regulatory Agencies regard-ing Quality Control/Quality Assurance; e) Evaluation of vendor's Quality Control .programs and manuals; f) Observation of skilled Quality Control/Quality Assurance consultants and inspection personnel inside and outside Niaqara Mohawk; g) Use of files containing Quality Control type records/

3 ~ Provision for participation in college level courses in various special processes, metallurgy, non-destructive testing, etc.

4. The provision of texts and periodicals concerning Quality Control related activities.

In addition to the foregoing, the Quality Assurance will conduct meetings and seminars to include as a minimum the following:

a) The history of Quality Assurance; b) The need for Quality Assurance; c) The functioning of a Quality Assurance System; d) Detailed explanations of all Quality Assurance/Quality Control policies, procedures and instructions; e) Application of various regulations, standards, codes and guides; f) The role to be performed by all personnel performing activities subject to Quality Assurance Coverage, including engineering, plant operating and maintenance, purchasing and storeroom.

51

The purpose of audits, to insure that the system is functioning and to recommend improvements.

5. The formal -indoctrination and training of personnel involved verify conformance of work activities to inspection, examination and testing to various codes and standards; a) Training courses will be scheduled on a periodic basis; b) Non-destructive testing programmed instruction manuals produced by General Dynamics, will be used for instruction.

c) On the job participation shall also be included in training program.

d) Specific required capabilities for inspection, examination and test personnel are- presently maintained in the Quality Control Inspection file, such as: personnel vision requirements.

e) The Non-destructive testing program will meet the requirements of the American Society of Non-destructive testing recommended practice SNT TC 1A and supplements.

f) Proficiency testing of training shall be in accordance with SNT TEC lA and supplements.

6. In addition to the above, indoctrinational and training program proce-dures, are currently being developed at Nine Mile Point Unit 1, using ANSI N45.2 and this guide as guidelines.

52

REGULATOR GUIDE 1.5 - DESIGN BASIS FLOODS FOR NUCLEAR POWER PLANTS A discussion of the design basis flood. is given in Sections 2.3.3 and.

12.3.7 of the Nine Mile Point Unit 2 PSAR.

The worst site related. flood at Nine Mile Point would. result in a screen-well flood. level at elevation 2/2.g feet. This i"s based: on a maximum probable setup of Lake Ontario of 4.1 f'eet above mean lake level and a maximum probable rainfall of 0.35 feet. The maximum wave run up associated with that flood level is elevation 263 fee~. To prevent the uni<

from being affected a stone faced. dike with a top elevation of 263 feet will be constructed in Lake Ontario extending from the existing dike in front of Unit 1 on the west and. to the eastern part of the site where the ground. rises to elevation 263 feet.

The questions of maximum probable flood. level and. required. protection are currently being resolved, with the AEC Staff during the licensing review of Nine Mile Point Unit 2 (Docket 50-410). The fina1 resolution will also app+ to Unit l.

53

5. ~UESNION Provide a summary description of the current status of modifi-cations which you stated. ~could, be completed. at the time of the first major refueling outage that began in April 1973. include a discussion of the design changes that were made to KP-1 since submitta1 of the Application.

RESPONSE

The following is a summary description of the status modifications pre-

.'sented in the "Technical Supplement to Petition for Conversion from Pro-visonal Operating License to Full Term Operating License" which were to be completed by the end'f the'irst'efuel'ing outage.

a. Vibration Monitorin A committment was made for installing a rudimentary impact system at Nine Mile Point Unit Nl. Since this time, General Electric in connection with Empire State Electric Energy Research Corporation has been developing a plan for installation of a prototype vibration monitoring system. The present schedule is to 'install this system at the James A. FitzPatrick plant in 1975. Niagara Mohawk being a member of Empire State Electric Energy Research Corporation will have access to all results from this program.

b. Recirculaticu P~um Tri A trip of the recirculation pumps on high pressure was discussed in the Technical Supplement. The status of this change is included in the response to Request 3b.

c. Hi h Pressure Coolant Ins ection S stem The high pressure coolant injection system serves to cool the reactor core for small line breaks and backs up, the core spray and auto&epress-urization. This system has been operable since the first refueling outage. A proposed change to the Status Technical Specification is included in response to request 1.

d. Instrumentation for Monitorin Dr well Conditions Pressure and temperature instrumentation were installed before initial operation. Additional wide range drywell pressure instruments were installed prior to the first refueling outage. These have a range of 0 75 psig with indication in the control room.

Since July, 1973, several other design changes were made to station oper-ating equipment and systems to further enhance the safe and efficient op-eration of the plant. In summary, they are as follows:

1 Nine Mile Point Unit 1, Technical Supplement to Petition for Conversion from Provisional Operating License to Full Term License.

54

a. D ell to Su ression Chamber Vacuum Breakers Lever arm modifications were made to improve the closing torque on the valves. These modifications enhance the closing of the valve following tests or actual operation by providing a more constant tor-que throughtout the valve cycle. In addition, low hysterisis indi-cating switches were installed to detect'pen positions less than the maximum allowable bypass area on, each valve.

b. Su ression Chamber Baffles All suppression chamber baffling was removed to improve the discharge flow pattern from the electromatic relief valves during their oper-ate:on. It was possible for hydraulic forces and local hot spots to develop on individual baffles. Therefore, to improve the safe oper-ation of the system all baffling was removed and "rams head" discharge installed.

The brake timing relay was reduced from 1.0 seconds to 0.5 seconds and the control circuit was reconnected such that hoist speed will drop instantaneously to one half speed when limit switches open the control circuit. In addition, a second upper limit switch was installed to open contacts in the primary motor leads, and repositioning of the load cell was performed. This modification will accomplish smoother braking of the grapple.

d. Reactor Protection S stem S nchronization H

Panel mounted dial indicators were installed to provide sensitive instrumentation for synchronizing Reactor Protection System busses 162 and 172 and computer bus 167 to and from the maintenance bus.

This design change eliminates the possibility of tripping the unit during transfers to and from the maintenance bus.

e. Emer enc Diesel Generator Mode Switch The emergency diesel generator mode switch was a two position toggle switch located at the diesel local control panel. This allowed an operator to defeat the automatic starting logic of a diesel generator for maintenance purposes. There was no indication in the control room of the switch position. A mechanical lock was placed on this switch to ensure that it always remain in the automatic position. Control procedures were set up for maintenance on the diesels.

f. Condensate Pum Room Protection The condensate pumps were protected from flooding by adding bulkhead ..

doors to the pump rooms and sealing the opening around the lines. This was done to provide protection of the High Pressure Coolant Injection System from flooding due to a break in the circulating water lines.

This analysis was at the request of the AEC in a letter dated August 3, 1972 from Mr. D. J. Skovholt to Mr. T. J. Brosnan.

55

g. Snubbers on blain Steam Line Differential Pressure monitors Pulsation dampening snubbers have been installed on the steam flow differential pressure switch lines to enhance the feedwater control.

A valve has been added- to the discharge line of the liquid poison pump so that the system can be tested at design pressure.

i. Leak-off Lines from Recirculation Valve Packin

-Valves'have been pla'ced in the leak-off lines which go to the equip-ment drain tanks. This modification will allow the second and third sets of packings in series to see service and minimize leakage to the drain tanks.

j. Feedwater Control Power Su 1 This change involved the separation of power supply channels. Re-dundancy of power supplies is accomplished by using different buses for each feedwater train.

k. Antici atin Tri Test Circuitr With the initial design it was impossible to test the scram logic at

'ful'1 'design rating without scraming of the reactor . Originally, two stop valves had to be operated. A change was made to monitor the current in the circuit when stroking one valve. This allowed testing of one valve at a time so that no scram results.

1. Electromatic Relief Valve Test Circuitr Originally, yhen the valves were energized and subsequently de-ener-gized, the solenoids would not fully reset due to a small current flow through the contact monitoring. lights. A time delay. drop out relay was installed to eliminate current flow through the lights until the solenoid is de-energized and returned to the fully de-energized position.

56

6. ~UES1 ION The following additional information regarding the containment atmosphere dilution system is required:

a. Provide curves of the oxygen concentration versus time in the drywell and suppression chamber following a loss-of-collant accident (LOCA) assuming no dilution by steam.

b. include a curve of containment pressure as a function of time assuming zero containment leakage.

c. Also, provide a curve of nitrogen addition requirements for dilutions as a function of time using the assumptions given in

a. and. b. above.

d. The ACRS has recommended for other similar plants that the peak containment, repressurization level be limited to a value substantially below its design pressure. Define and justify the repressurization limit for IRK-1. identify'equired. purge rates, initiation time, and radiological doses at the site boundary due t 0 purging o

e. Provide discussion and analyses to support the adequacy of the design bases for the containment atmosohere dilution (CAD) system and discuss how the system will be operated. The discussion should include the following:

(1) The sampling equipment, principles, design, operating procedures, equipment qualification for LOCA service, time to sample or monitor, location of sampling points in containment, location of measurement readout, sampling errors and, stratification considerations.

(2) The preoperational checkout and evaluation of the sampling and. CAD systems and. the testing procedures and. frequency for these systems during the plant lifetime.

(3) Relate the maximum required rate of nitrogen makeup to the design flow rate of the CAD system. Specify the capacity of'he CAD system nitrogen tank and the provisions for monitoring the nitrogen level in the tank.

(4) The design pressure limitations of CAD system components and piping, delivery capability of the CAD system against pressure head. of the containment, and makeup limitations due to inadequacy of onsite nitrogen inventory or time to obtain offsite makeup (specify).

57

HYDRO%EN AND OXYGEN CONCENTRATIONS IN CONTAINMENT FOLLOWING LOSS OF COOLANT ACCIDENT 15 DRYWELL

~10 0-cL 5 SU PP RESS ION CHAMBER N2 INJECTION BEGINS 0.1 1.0 10 100 TIME AFTER ACCIDENT (DAYS)

SUPPRESSION CHAMBER DRYllELL C)

I 2

CL NJECTION BEGINS 0.1 1.0 10 100 TINE AFTIACCIDENT (DAYS)

Fi ure 6-1

CONTAINMENT PRESSURE WITH CONTAINMENT ATMOSPHERIC DILUTION OPERATION.

ZERO CONTAINMENT LEAKAGE 40 30 20 0.

20 40 60 80 100 TIME AFTER ACCIDENT (DAYS)

Fi ure 6-2

NITROGEN ADDED BY CONTAINMENT ATMOSPHERIC DILUTION,OPERATION FOLLOWING LOSS OF COOLANT ACCIDENT 500 400 CD CD g 300 200 C) 100 C) l 20 40 60 80 100 TINE AFTER ACCIDENT (DAYS)

I.igure 6-3

on a continuous basis to minimize stratification and sampling errors.

These monitors will be capable of meeting the most severe environ-mental conditions following an accident. The location of sampling points is shown on Figure 6-5. Readout will be in the control room.

Equipment will be designed to operate in the most severe environmental conditions.

The maximum required rate of nitrogen make-up comes in the first 20 days following an accident and amounts to 7 scfm. The design flow rate of the vent system is 0-100 scfm. The storage tank holds approxi-mately 1,100,000 cubic feet of nitrogen. Approximately 600,000 cubic

.feet,.are required to inert the,drywell,so that in the unlikely event that the accident occurred, there would be sufficient capacity for over 2 months of Containment Atmospheric Dilution system operation.

Delivery of nitrogen can be continued for containment pressures in excess of 40 psig. Design pressure limits will be 350 psig. Level monitoring of the nitrogen tank will be in the control room.

Piping will meet the requirements of ASME Section III Class 2.

Pre-operational and operating procedures will be developed commensurate with progress in system completion. In addition technical specifications which will include test frequencies and procedures will be developed on an appropriate time scale.'he:only sources which could potentially contribute oxygen to the con-tainment following a loss of coolant accident are oxygen entrained in the coolant and leakage from air supply systems.

The maximum amount of oxygen entrained in the reactor coolant is 0.0043 lb- moles. This is based on a maximum coolant concentration of 3 ppm as described in the bases to the Technical Specifications (3.2.3).

There are service air and breathing air connections which are valved closed during normal operation. However, leakage on the order of 0.1 scfh could be possible. The 0.0043 lb moles from the reactor coolant and the 0.001 lb moles from valve leakage is negligible in comparison to the 8 lb-moles of oxygen generated in the radiolytic decomposition.

first 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> due to Figures 6-4 and 6-5 show the containment atmosphere dilution system with back-up venting and the containment gas analysis system.respec-tively.

The dotted lines on Figure 6-4 show the lines which are used for con-tainment atmosphere dilution and venting. The system also utilizes the existing storage facilities (with additional storage added), vapor-izer, isolation valves and emergency ventilation filters and fans.

The containment atmospheric dilution system including nitrogen storage tanks, vaporizors, piping and valves is an engineered safeguards system and is being designed to meet seismic Class I requirements. The system.

will be designed in accordance with the following codes and standards:

(1) USAEC Regulatory Guides 1.7 and 1.26 (2) AS&IE Section III Class 2 (3) IEEE 279

STACK TURBINE BUILDING DUCT WORK PBV ENERGEttCY 8'EACTOR VENT ILATIOH FANS DRYWELL VENT BUILDING EXHAUST FANS

~IV

" ~iV AND PURGE FAH BV BV ABSOLUTE BV CHARCOAL W BV IV F

ABSOLUTE ILTERS l- ~ I l COttDEttSER I VAPORIZER I

DUCT g

HTR FE FCV W I N2 RE Cg w ~w rsm ssrace BV I l STORAG TANK TURBINE tt BUILDING REACTOR REACTOR BUILDING BUILDING VEttT DUCT NORMAL BV l I SYSTEI4 NEyRppENI m l MAKE-UP l I IV tt I AM'V I

IV l M IV IV EL I

I

~l 315.0'TMOSPHERE a

M IV 'I IV DRYWELL BV I V IV NORMAL l M-MANUALLYOPERATED NITROGEN TORUS TORUS

---CAD SYSTEM FLOW LINES MAKE-UP 1

CONTAINMENT ATMOS PHERE D ILUTION SYSTEM WITH BACK-UP VENTING Figure 6-4

CONTAINMENT GAS (H2 AND 02 ) ANALYZER SYSTEM RMC J H2 5ARPLK WLV

~ IIC"AlWIN

,P9 CtlCTCN Og

~INC

~ CIS

UESTION To assure that ferritic materials of pressure retaining components of the reactor coolant pressure boundary will have adequate fracture toughness during service hydrostatic and. leak tests, provide revised temperature and. pressure limitations established by using the requirements-.of Appendix G 2000 of the Summer 1972 Addenda to Section III of the ASME Boiler and. Pressure Vessel Code as a guide. Also, provide the temperature limitations for core operation specified by the recently revised and issued Appendix G of 10 CFR 50 (Published in the Federal He ister on July 3.7, 1973). Indicate the operating limitations on heatup and. cooldown plus the above information that wi.ll be included in proposed changes to the Technical Specifications.

RESPONSE

t A fracture toughness analysis has been completed for the Nine Mile Point Unit Nl pressure vessel. The analysis indicated that the present l00 F/hz limit on normal heat-up and cool-down are acceptable throughout the life of the plant.

Operating limits on reactor vessel pressure and temperature during normal heat-up and cool&own, and during inservice hydrostatic testing have been established using Appendix "G" of Section III of the ASME Boiler and Pressure Vessel Code, 1971 Edition as a guide.

These operating limits assure that a lazge postulated surface flaw, having a depth of one-quarter of the material thickness, can be safely accommodated in regions of the vessel shell remote from d'scontinuities.

For the purpose of setting these operating limits the reference temperature, RTNDT was determined from the impact test data taken in accordance with re-quirements of the ASME Code to which this vessel was designed and manufact-ured. If the dropweight Nil Ductility Transition Temperature is known, the reference temperature to be used would be the Nil Ductility Transition Tem-perature. If the dzopweight Nil Ductility Transition Temperature is not known, RTN would be taken as the lowest temperature at which the minimum Code allowRle Charpy V-notch, energy requirement would be expected to occur on the basis of reported Charpy V-notch test data.

The highest reference temperature of any part of the reactor pressure vessel, pressure boundary material is used as the reference temperature for calcu-lating one set of operating temperature and pressure limits for the shell remote from the core beltline region. A second set of temperature and pres-sure limits for the core beltline region have been calculated based on the core beltline,region material reference temperature appropriately adjusted for irradiation shift with vessel neutron exposure. The requirements of the Code to which the vessel was designed and manufactured results in a third set of vessel shell temperature pressure limits. These are Nil Ductility Transition Temperature +60F or Charpy V-notch +60F at pressure greater than 20 percent of preoperational system hydrostatic test pressure.

The most conservative of the above three limits was used to set pressure and temperature limits for the vessel shell.

60

Figure 7-1 gives the temperature and pressure limits for inservice hydro-static testing. The upper curve will be limiting until the predicted Nil Ductility Temperature shift reaches 30F. Figure 7-2 gives the changes in Charpy V-notch temperature shift as a function of neutron fluence. This curve will be used to determine the shift during vessel life.

A proposed change to the Technical Specifications is attached as a revision to Specification 3.2.2 and bases.

PRESSURE AND TEMPERATURE LIMITS FOR INSERVICE HYDROSTATIC TESTING 200 150

~ REACTOR PRESSURE VESSEL SHELL RENTE FROH CORE BELTLIHE REGIOH REACTOR PRESSURE VESSEL SHELL CORE BELTLIHE REGIOH 100 rr 50 0

600 800 1000 1200 1400 1600 INSERVICE HYDROSTATIC TEST PRESSURE (PSIG)

MEASURED IN TOP DOME Figure -1

THE EFFECT OF IRRADIATION ON VARIOUS HEATS OF A302B/A533B-CL'ASSTEEL

'600 500 400 300 I

200 100 0

016 1017 101 8 1019 1020 1021 INTEGRATED NEUTRON DOSAGE (>1 MeV) (<t), nvt Figure 7-2

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT BASES (Cont'd.)

$ .2.2 NINON REACTOR VESSEL TDO'ERAIURE MR PRESSURIIATICeI 4.2.2 NININNI REACIOR VESSEI. TENPERATURE FOR PRESSURI ATIOV f~lfflff II Applies to the ninisacs vessel tcnperature required Applies to the required vessel tcnperature for for vessel pressurization. pressurization.

~ff f To assure that no substantial prcssure is inposed cm To assure that the vessel is not subjected to any the reactor vessel unless its tenperature is consider- substantial pressure unless its tcnperature is greater ably above its Nil Ductility Transition Texperature than its NUTI.

(NUIT) .

~Slfl l ~Slfl \

a. the reactor vessel shall ba vented and shall not be a. Reactor vessel tesperature and pressure shall be Figures 3.2.2a ani 3.2.2b plot respectively the pressure vs.

in the paver operating oondition whenever the pressure nonitored and controlled to assure that the pres- tesperatuze Iinits ani tha change in 30 ft. - lb. Chaxpy vessel texperature is less than Nnxr plus 60r ae sure snd teeperature linits are net. v-notch capability vs. integrated neutzon dossage. Vba basic shovn in Pleura 3.2.2a. data foz figure 3.2.2b tor A302n/AS330 - class 1 steels is

b. Neutron flux nonitors installed in thc reactor based ce 30 ft. - lb. Charpy V-notch energy transition tem-

b. Iho reactor vessel head bolting studs shall not be vessel adjacent to the vessel vali st core ~ Id- peratures which have been correlated vith drop veight spec-under tension unless tho tcnperature of the vessel plane level shall be renovcd and tested at the lnen nil ductility transition for this steel. At the design head flange and the head are equal to or greater first refueling outage. exposure of s x 1012 not the change in NUIT is 6sp.

than IOOF.

c. Naterial sexples, installed in the stean, stean/water, and vater Ieuees inside the reactor pressure vassal, used to nonitor tha sensitized stainless steal shall be inspected on the following schedules First capsule - ona touzth service life second capsule - three fourth service lite Third capsule standby the reactor vessel head flange and tho vessel flange in In the event the suxveillance specinans at ccebinatlon with tha double 'xy'ing type seal are de-one quarter of the vessels cervioe life indicate signed to provide a leak tight seal when bolted together.

~ shitt of refezence tcnpsrature greater than then the vessel head is placed on the reactor vessel, predicted tha schedule shall be xevised as follovsx only that portion of the head flange near the inside of thc vessel rests on the vessel flange. As the head geconi capsule - one balt service life bolts are replaced and tensioned, the vessel head is Third capsule standby flexed slightly to bring together the entire contact surfaces adjacent to the 'xy'ings of the head and vessel flange. goth the head and vessel flange have a AUT tesperature of 40F and they are not subject to any appreciable neutron radiation exposure. Therefore, the nininun vessel head and head flange tcepersture for bolting the head flange and vessel flange Is established s>> 40 e 60F or lOOF.

The integrated neutron flux at the vessel vali is cal enlaced free core physics data and will be neasured using flux nonitors Installed inside thc vessel. This neasured flux vill be used to check and if necessary correct the calculated data to detexuine an accurate flux. Fran this data a ccesexvative NUIT tenperature can be detexuined. Since no shift vill occur until sn integrated flux of IOIP nvt is reached the confir-natice csn be sade vali in advance of any shift.

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT BASES (Cont'd.)

Vessel saterial surveillance saaples are located vlthin the core region to pcrsit periodic aonitoring of ex-posure and saterial properties relative to control saaples (Vol. IV, Section I-D. p. I-24). ~

In addition. sasples vill also be installed to aonlr tor the sensitited stainless steel cosponents.

Sasplcs consisting of sensltlted stainless steel forgings and strips and annealed aaterlal vill be located in the stean, a{xture, and vater phases inside the re~ctor vessel. Detailed laboratory exaaination of these sasples vould be required if inspections and/or analyses of other con ditlons, e.g. ~ substantial deviations In pri sary coolant chesistry, indicate that stress corrosion cracking of the sensitited stainless steel occurred.

'FSAR.

41

PRESSURE AND TEMPERATURE LIMITS FOR INSERVICE HYDROSTATIC TESTING 200 REACTOR PRESSURE VESSEl. SHELL REHOTE FROH CORE BELTLINE REGION 150 LLJ I ~ REACTOR PRESSURE VESSEL SHELL CORE BELTLINE REGION 100 50 0

600 .800 1000 ,

1200 1400 1600 INSERYICE HYDROSTATIC TEST PRESSURE (PSIG)

MEASURED IN TOP DOMlE Fig 3.2.2.a

THE EFFECT OF IRRADIATION ON VARIOUS HEATS OF A302B/A533B-CLASS 1 STEEL

"'600 500 400 lD 300 I

200 100 0

lpl6 1017 lpl8 1019 '020 INTEGRATED NEUTRON DOSAGE (>1 MeY) (<t), nvt Figure 3.2.2.b

8. QUESTION Provide the proposed surveillance capsule withdrawal schedule and, indicate the degree that KP-1 can comply with the recently revised.

Appendix H to 10 CFR $ 0 (published in the Federe1 ~Re ister on JnIF

17. 1973) ~

RESPONSE

.The surveillance..capsule .withdrawal schedule for Nine Mile Point Unit 1 will comply with Section II C.3a of revised. Appendix H to lOCPR$ 0 dated July 17, 1973. Reports will be submitted. pursuant to this Appendix. The present Technical Specifications will be revised to reflect this change.

62

9. ~UES1ION Provide sufficient information about your inservice inspection program for engineered safety features to indicate that the program provides a degree of assurance of system integrity comparable to the program recoaunended in Regu1atory Guide 1.51, "lnservice Inspection of ASIDE Code Class 2 and 3 Nuclear Power Plant Components",

May 1973, within the limits of accessibility designed into RP-l.

Znc3ude the corresponding revisions, as appropriate, in the proposed.

changes to the Technical Specifications.

RESPONSE

Nine Mile Point Unit 1 components were designed to the ANSI B31.1 and B16.5 codes. Those systems which would be included as ASME Code Class 2 and 3 components if being designed presently are:

ASME CODE CLASS Class 2 S stems Class 3 S stems

• Reactor recirculation Reactor Cleanup

• Main steam Reactor and Haste bldg. closed

• Low and High Pressure feedwater loop cooling

• Condensate Service Water Control rod drive Radioactive waste disposal

• Core spray Off-gas Liquid Poison Diesel generator fuel oil, Shutdown cooling starting air, and cooling Head spray water.

• Emergency Condenser Instrument and breathing air Containment spray Drywell vent and purge Reactor bldg. emergency ventilation Control room ventilation Reactor instrumentation Dxywell and torus vacuum relief Reactor vent and drain Drywell and Instrumentation and Fuel pool filtering and cooling leak monitoring

• PRIMARY COOLANT SYSTEM Presently, the inservice inspection program covers the primary coolant sys-tem, those engineered safeguards concerned with emergency core cooling and the main coolant piping. The guide calls for additional areas for inspection to include all those systems as classified as ASME Code Class 2 and 3 components.

Procedures are currently under development at Nine Mile Point Unit 1 which will update the present inservice inspection program. Regualtory Guide 1.51 will be used as a guide in their preparation. This revised program will include all areas which would be AS'ode Class 2 and 3 components designed today.

if being Appropriate Technical Specification changes will also be submitted.

64

10. UESTION The response on page III-2 of the Application to show conformance to Criterion 4 by stating that adequate bracing is provided to prevent pipe whip is not consistent with the FSAR information. It is our understanding that protection against pipe whip was not included in the design of NMP-l.

Therefore, to determine feasible means to protect the containment, the primary system and engineered safety features against the adverse effects of pipe whip in the unlikely event of a pipe rupture inside containment, provide the following information regarding high energy lines inside the NMP-1 containment:

a. An evaluation describing those systems that are ad'equately protected against the effects of pipe whip due to either of the following:

'(1) The use of two independent redundant systems where only one need be protected by augmented inservice inspection or pipe restraints, or (2) Sufficient spatial separation or structural separation exists to protect all trains or a redundant system from being affected by a single pipe rupture.

b. Where an augmented inservice inspection program is indicated, follow the guidance of Enclosure 1 attached hereto.

it will

c. Where structural design criteria are indicated to provide the protective structures, the criteria of Enclosure 2 attached hereto will be followed.

RESPONSE

a.

SUMMARY

All high energy lines inside the primary containment of Nine Mile Point Unit No. 1 have been analyzed for the effects of pipe whip. In all cases, the capability of core cooling is maintained. Table 10-1 lists all the high energy systems inside the primary containment. It has been assumed that anyone of these lines can break anywhere inside the primary containment.

If a line break occurred all the systems listed in table 10-1 could be affected.

However, because of redundancy and separation, the engineered safeguard systems would still perform their intended functions.

An investigation into their impacting the walls of the primary containment was also performed. This analysis showed that Pg postulated breaks and resulting impact, the containment would not lose,its integrity and would suffer no loss of function.

b. CONTAINMENT INTEGRITY ANALYSIS A review of the piping systems inside the drywell was made to determine which systems could, if postulated to fail, impact the containment with sufficient energy to cause concern. The three systems considered as containing sufficient 65

TABLE 10 1 High Energy Systems (1) main steam (2) feedwater (3) reactor reci rcul ation (4) core spray (5) containment spray (6) emergency condenser supply and return (7) control rod drive hydraulic (8) liquid poison (9) relief valve discharge (XO) shutdown cooling (ll) head spray (12) clean-up

energy upon impact to represent the worst cases are:

(1) Reactor Recirculation (2) Main Steam (3) Feedwater The fluid forces generated by the break have been calculated and are dis-cussed. below. These loads are then applied to a model of the pipe and the im-pact velocity at, the containment is obtained. The stresses in the containment are calculated by modeling the containment, air gap, concrete, and impacting missle with the appropriate .mass and .velocity.

FLUID FORCES Recirculation Loo The force at the break rises from a value equal to the product of pressure times area to about 1.125 times that value in a short time, then decays slowly. However, due to the proximity of the break to the reactor pressure vessel nozzle, some conservatism was added to cover possible impingement effects of the jet escaping from the nozzle. The actual load applied to the piping system was 1.5 times the product of pressure and area.

Main Steam The force at the break rises from an initial value equal to pressure times pipe cross sectional are to a maximum of 1.26 that value in 0.052 seconds.

Feedwater The force at the break rises from an initial value e'qual to a maximum of 1.125 that value in 0.068 seconds.

Jet Im in ement Force As a result of break in the Recirculation System at the vessel nozzle a jet is generated which can impinge on the containment vessel. The jet pressure is 40 psi over an area of 26,300 square inches.

Im act Velocities and Effects The velocities of the pipes and beams which have been analyzed for impact on the containment vessel are given below:

~Sstem Im act Velocit (ft/sec)

Recirculation 115 Main Steam 100 Feedwater 89 Structural Beam 115 (lOWF33) 66

The break in the main steam system represents a break at the vessel nozzle.

For a break near the main steam penetration, the steam line cannot reach the containment vessel without striking the feedwater system. This would reduce the velocity of the steam line significantly. Assuming that the steam line did not strike the feedwater line, an impact velocity of 270 ft/sec would re-,

sult.

The allowable strain in the containment vessel demonstrates the adequacy of the structure against impacts resulting from pipe break and associated whip.

The ultimate strain of the containment vessel is 10 percent. The calculated accumulative strains for the conditions analyzed are given below:

Condition Vessel Thickness (in) Strain Recirculation Loop Elbow 1.5 6.1 percent Recirculation Loop Elbow '768 9.2 percent Structural Beam(12 WF40) .768 4.0 percent Structural Beam(12 WF40) 1.5 1.1 percent With respect to the structural'beams, the highest velocity (115ft/sec) occurs for the 10WF33 beam. For conservatism, the analyses used a heavier 12WF40 beam I with a velocity of 115 ft/sec rather than the actual velocity of 60 ft/sec.

c. S stem Descri tions (1) Main Steam The main steam lines discharge from the vessel at Elevation 310 feet and at the 90 and 270 redial locations and descend. In this area there are two 10 inch emergency condenser lines at the 67.5 and 292 ' radial locations.

These lines are separated at great enough distances so as not to be effected by any break of the main steam lines.

The main steam lines proceed downward through Elevation 295 feet where they pass by some 1 inch instrument piging at the 90 radial location and a control rod drive exhaust line at the 270 radial location.

Continuing on down from Elevation 295 feet to Elevation 264 feet, the steam lines pass by containment spray sparagers, feedwater lines on each side of the steam lines, relief valve discharge lines, shutdown cooling, reactor recircu-

,lation, core spray and rod drive exhaust. All could be ruptured except for the recirculation lines because of their larger size. From this point, the steam lines heal toward 180 0 and exist the drywell adjacent to the two 18 inch in-coming feedwater lines.

The required systems for core cooling and safe shutdown in the case of a main steam line break are the containment spray, core spray, and feedwater. All of these systems have the required redundancy or backup as discussed in paragraph

d. below.

67

(2) Feedwater The two 18 inch feedwater lines enter the primary containment at Elevation 263 feet adjacent to the two 24 inch main steam lines. At this point the lines curve around to 90 and 270 radial directions. At + 45 degrees on each side of these lines, two 10 inch lines proceed inward and then ascend vertically.

In this run the feedwater lines pass by some 12 inch core spray lines, 14 inch shutdown cooling lines, 6 inch containment spray lines and 6 inch clean up system lines.

Proceeding upward to Elevation 295,feet, the feedwater lines enter the reactor vessel at the 45 , 135 , 225 and 315 radial locations. During this ascension they pass by the 12 inch core spray lines and containment spray sparagers, and

'a 'lb inch 'liqui'd 'poison 'line.

The only damage which could occur is to one of the redundant containment sprays and to the liquid poison lines. because of their size. As discussed in para-graph d. below, this system has adequate redundancy even in the event that one system is incapacitated.

The liquid poison system is a backup system only. The primary means of shutting down the reactor is the control rod drive system.

(3) Reactor Recirculation System There are five recirculation pumps each of which has a suction and discharge line. These lines are at the 0 , 42 , 73o, 114 144 , 186 , 216 , 258 , 288 an'd '330 radial locations between Elevations 225 feet to 275 feet. These are 28 inch and 26 inch diameter lines for the suction and discharge respectively.

Other lines in the area of these recirculation lines are:

. (a) 10 and 12 inch Emergency Condensers (330, 0 )

(b) 4 and 6 inch Containment spray (0 to 360o,)

(c) 10 and 18 inch feedwater (288 , 258 216 , 144 , 114 )

(d) 12 inch Core Spray (258 , 115 )

(e) 14 inch Relief valve discharge (288 , 216o)

(f) 6 inch clean-up (42 )

(g) 24 inch Main steam (258, 216, 186 144 , 114 )

(h) 14 inch Shutdown Cooling (330 )

The first four systems (a, b, c, and d) may be required following a break in the recirculation system. All have the required redundancy or backup.

(4) Containment Spray Due to the small size of the containment spray lines and the fact that they are not pressurized during normal operation, rupture would not cause damage to any other of the lines because of their larger size.

(5) Liquid Poison The liquid poison line due to its small size would not impart damage on any other system.

(6) Emergency Condensers The 10 inch emergency condenser supply lines are located in the area between P

I 68

270 and 90 radially at Elevation 306 feet. In this area there are only some small'nstrument lines, a 2 inch head spray line and some 14 inch containment spray headers. The supply "lines leave the drywell and the 10 inch return lines enter at Elevation 269 feet. In this there are four lines:

'a) 6 inch clean-up (b) 6 inch, 8 inch, and 12 inch containment spray The only required lines in the event of a break of an emergency condenser line are the containment spray. These are supply lines to sparagers and damaging one would not render the system unoperable since, there are four sparagers.

Only one sparager could be damaged by the break of any one emergency condenser line.

(7) Control Rod Drive Discharge Due to the small size of the discharge piping in relation to the other lines no other damage would be imparted due to a rupture of this line.

(8) Core Spray There are two 12 inch core spray lines which enter the containment at 240 feet at about, 45 0 on each 0 side of the 180 0 direction. The one line in the quadrant

~

0 from 180 to 270 passes by the following lines and then rises vertically:

(a) 10 and 18 inch feedwater (b) 24 inch main steam (c)'4 inch relief valve discharge (d) 6 inch containment spray The otHer core spray line in the quadrant from 90 and then runs north to the 0 -90 quadrant where line lines:

ittorises 180 enters the drywell vertically. This passes by the following

~

(a) 10 and 18 inch feedwater (b) 14 inch relief valve discharge (c) 6 and 8 inch containment spray Both lines rise to Elevation 295 feet where they enter the reactor vessel 180 0 apart. In their rise they pass by 10 inch feedwater liens and some small con-tainment spray headers. The only lines which the core spray could damage are the four smaller feed water lines and the four containment spray lines to sparagers.

(9) Relief Valve Discharge In each 180 radial segment of the drywell, that is from 0 -180 and 180 -360 there are three 14 inch discharge lines.

In the 0 -180o sector there are the following lines.

(a) 6 inch clean-up (b) 10 and 18 inch feedwater (c) 12 inch core spray (d) 6 ,- 8 and 12 inch containment spray (e) 24 inch main steam In the 180 -360 segment there are the following lines:

69

(a) 14 inch shutdown cooling (b) 10 and 18 inch feedwater (c) 12 inch core spray (d) 6, 8 and 12 inch containment spray (e) 24 inch main steam The lines which could be damaged by the relief valve discharge that are required for core cooling are the core spray, containment spray and feedwater lines. No single relief valve discharge line failure could eliminate redundancy to where the safeguards function is inadequate.

the'oint (10) Shutdown Cooling In the '270 'to 0 'radial location there are 14 inch supply and return lines to the shutdown cooling system at Elevation 270 feet. In this area there are the following systems:

(a) 10 inch feedwater line (b) 6 and 12 inch containment spray (c) 14 inch relief valve discharge (d) 3 inch exhaust from the control rod drive The first three systems may be required following a shutdown cooling system line break. However, no line break in this system could result in loss of redundancy in those required systems to the point where the safeguards function is inadequate.

(11) Clean-Up System In the 0 to 90 0 radial location a 6 inch line comes out of one of the recircu-lation lines at Elevation 263 feet cleaves the drywell, then re-enters at Eleva-tion 263'eet and goes back into the recirculation line again. The only lines in that area are:

(a) 10 inch feedwater (b) 12 inch core spray (c) 10 inch emergency condenser return line (d) 14 inch safety valve discharge (e) 6 and 12 inch containment spray Due.to the large sizes of these lines there would be no effect on them because of a rupture of the clean-up system.

(12) Head Spray Due to its location on the vessel head, there are no lines important to safety in the area of the line nor could it be damaged by any of the other lines.

d. ENGINEERED SAPEGUARDS PROTECTION The preceeding analysis show that engineered safeguard systems could be damaged as a result of pipe whip. However, in no case is the damage, extensive enough to result in loss of core cooling,a safe shutdown capability.

70

As descirbed in paragraph c above, the feedwater system (high pressure coolant injection) could be damaged as a result of a rupture in the main steam, recircu-lating, core spray, relief valve discharge or the shutdown cooling system lines.

In any event since there are two feedwater lines which are physically separated in the areas of concern only one could be damaged. In addition, core spray and auto depressurization are the prime sources of core cooling. High pressure coolant rejection is only a backup system. There is no single pipe rupture which could result in loss of feedwater and both core spray systems.

There are two independent core spray lines, 180 degrees apart. These could be damaged by rupture of either the recirculation or relief valve discharge lines.

However, because 'of redundancy and physical 'separation only one line could be damaged. High pressure coolant injection serves as a backup.

The containment spray system could be damaged as a result of a rupture in the following systems:

(1) reactor recirculation (2) feedwater (3) main steam (4) emergency condensers (5) core spray (6) relief valve discharge (7) shutdown cooling There are two containment spray systems each one consisting of a supply and set of sparagers inside the containment. Both sets of containment spray systems could be damaged as a result of a single line break due to close proximity of the sparagers. This would not result in a loss of containment cooling since the suppression chamber water would still be circulated through the containment spray Beat exchangers.

Degradation of spray efficiency could occur and would depend on the extent of sparager damage': In any event, some spray efficiency would remain.

The emergency condensers supply and return lines on both systems could be damaged by a rupture of the main steam or reactor recirculation system lines.

However, this system is not required to maintain core cooling. Feedwater,

'core spray, and auto depressurization provide the core cooling function in the event of a line rupture within the drywell.

The control rod drive hydraulic system could be damaged by a rupture in the main steam, relief valve discharge or reactor recirculation system. However, sould these lines be damaged the rods would scram on reactor pressure. The liquid poison system, which serves as a back up to the control rod system is not subject to damage by the same lines.

The only line whose rupture could damage the liquid poison system is a line in the feedwater system.

However, the liquid poison system is not normally used. It is only a backup to the control rod drive system which is not subject to damage by a ruptured feed-water line.

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11. UE STION The following additional information regarding electrical and instrumentation. systems is required.:

a. Paragraph seven of Section III.C.3..a of the Application does not clearly identify the systems (reactor trip, emergency core cooling, etc.) that satisfy the requirements of these criteria.

include this information in this paragraph.

b. Your system description in Section III.C.l.b of the Application

.doesnot .clear~1 .describe how your design satisfies the requirements of Section 4.4, 4.5, 4.6, 4.15, 4.17, and the 4.21 of IEEE-279. Provide a discussion in more detail of these IEEE requirements in Section III.C.l.b.

c. The description of how your design satisfies the requirement of Section 5.2.3(5) of IEEE-308 is not included. in Section III.C.2.b of the A~lication. Include this information in this paragraph.

d. A recent incident occurring in a BUR resulted. in temperature inside the primary containment exceeding those temperatures specified in the design of equipment required. for safety and located inside the containment. Provide the results of environ-mental qualification type tests for a11 electrical Class IE equipment, including electrical penetration assemblies'and.

connections, located. inside the primary containment that are required. for safety. Include this information in Section III.C.3 and. 8 of the Application.

RESPONSE

The systems that satisfy the requirements described in paragraph 7 of Section III C.l.a. of the Technical Supplement to Petition for Conversion from Provisional Operating License to Full Term Operating License are listed below:

(1) Class 1E electrical systems including:

(a) Station battery systems (b) Emergency diesel generator systems (c) Emergency service portion of the plant service a-c power distribution system.

(2) Reactor Protection System including power supplies.

(3) Engineered Safeguards including:

( 1) Core Spray System (2) Containment Spray System (3) Liquid Poison System

4) Containment Inerting

5) Emergency Ventilation 72

(6) Automatic Depressurization System (7) Contro3. Rod. Drive System (8) Emergency Condensers (9) .Containment and. Primary System Isolation Valves (3.0) Containment Vacuum Relief (11) High Pressure Coolant Injection System

b. For protection system components, vendor's certified, design data sheets are available. These verify that system equip-ment is'adequate for achieving system performance requirements.

Verification of adequacy on a continuing basis is available in preoperational test .results .and the results of protection system surveillance. The testing requirements are contained in the Technical Specifications.

The protection system design incorporates dual independent tripping channels with'each channel containing two independent instrument channels. The tripping logic of the total system is referred to as a one out of two taken twice. This system will accommodate any single failure and still perform its in-tended. function and in addition, provide protection against spurious scrams. For further design details refer to Volume I,Section VIII A of the FSAR.

All protection system channels are designed to maintain necessary Junctional capaoility under extremes of conditions relating to environment, energy supply, malfunctions a'nd accidents.

Channels that provide signals for the same protective function are functionally and physically separated.. This results in decoupling of any adverse effects of accidents, electric tran-sients and. resulting environment.

The protection system design does provide positive means of assuring that a more restrictive set point is used where applicable. Positive means of assurance are obtained by the it is use of operation procedures and.,checklists during plant startup, power operation, shutdown and refueling. An example of this would be to switch Intermediate Range Monitor in service before Source Range Monitor's reached full scale during station start-up.

It is inherent in the reactor protection system design that an action once initiated at the system level will go to completion.

The entire system is designed to be fail safe. Initiation of a protective action in any one of the four sub-channels, will cause a half-trip. A coincident protective action initiation in the proper redundant channel will cause a full trip. De-liberate operator action is required to acknowledge the protective action and return the protection system to operation.

73

The protection system includes means for manual initiation of protective action at the system level. The manual initiation capability is described in the Reference Volume I,Section VIII A of the FSAR. It is inherent in the reactor protection system design that no single failure within the manual, automatic or common portions of the protection system, prevents initiation of a protective action.

Manual initiation is provided for:

(l) Reactor trip (2) Main Steam isolation (3) Reactor Cleanup isolation (4) Reactor Shutdown cooling isolation (5) Containment Isolation (6) Start Core Spray Pumps (7) Open Core Spray Discharge Valves (8) Containment Spray Pumps (9) Containment Spray Discharge Valves (10) Liquid Poison System (11) Containment Inerting (12) Emergency Ventilation (13) Control Rod Drive System (14) Containment Vacuum Relief'he protection system is designed to acilitate the recognition, location, replacement, repair or adjustment of malfunctioning components or modules.

Manual bypasses are incorporated for use during maintenance.

System protective actions are alarmed both on the station annunciator system and the station computer.

The preferred power supply portion of the station distribution, systems is monitored in the control room as follows:

C (1) 115 KV Bus Voltage (2) 115 KV Breaker Status Indication (3) Breaker status for 4160 V supply breakers R1012 and. R1013.

(4) Power board. 102 and. 103 bus voltage (5) Power board. 102 and. 103 frequency Environmental Qualifications The following environmental type tests were performed. on electrical Class'E equipment at Nine Mile Point Unit No. 1.

(1) Electrical penetrations-These penetrations were tested for leak tightness under the following accident environmental conditions:

74

Temperature Containment 310F Reactor building 50-150F Pressure 62 psig Relative Humidity 100 percent (2) Limitorgue valve operators These operators were exposed. to saturated steam at pressures up to 90 psig and temperatures ranging from 250F to 335F. The operators performed satisfactorily.

(3) Electromatic relief valves These valves were tested. for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> at 62 psig and.

300F. The valve was capable of performing its required function during this time.

(4) Control Cable High temperature cross-linked, polyethylene (Vulkene) cable was used., which is capable of withstanding 340F for a period. of one month.

75

12. ~UEST10N The additional information regarding the upgraded radwaste system for NMP-1 as identified belo~ is required:

As required. by General Design Criterion 64 in 10 CFR Part 50, Appendix A, indicate provisions made to monitor a11 normal and potential pathways for release to the environment of radio-active material in liquid and gaseous effluents.

b. For the proposed system modification to return equipment drain liquid. waste to the condenser hotwell without treatment in the waste collector subsystem provide:

(1) Description of the equipment and. piping modifications which wild. be required.

(2) Criteria and. means to be used. to determine whether the equipment drain liquid waste will be returned directly to the condenser hotwell.

c ~ For the upgraded off-gas radwaste treatment system, analyze the consequences of a ma1function or failure of essential components and-estimate%he resulting doses to plant personnel.

de For the solid waste data points listed in the Environmental Report Tab3.e 3.6-2 as numbers 25 and. 35, identify and. quantify the major radioisotopes expected from these sources.

e. Provide the design codes and. standards for the new waste concentrator, the new waste concentrator storage tank, and the new eauipment and processing lines for rerouting of liquid wastes ~

RESPONSE

a0 As'described on page III-22 of the Technical Supplement to Pe-tition for Conversion from Provisional Operating License to Full Term Operation License, provisions have been made to monitor the following releases:

(l) Gaseous releases from the stack (2) Liquid discharges to the circulating water tunnel (3) Reactor building ventilation (4) Waste building ventilation In addition, both on-site and off-site monitors associated with the environmental monitoring program are used to trace the effects of radioactive releases.

76

RESPONSE (cont'd)

b. The equipment drains which have been modified to return directly to the condenser hotwell contain only condensed steam from three sources in the off-gas system. They include (l) condensed steam from the off-gas preheater (2) condensate from the off-gas con-denser and (3) condensate from the off-gas vent cooler. All of these sources originate from main steam and do not contain sig-nificant. amounts of contaminants. This'llows the condensate to be returned directly to the main condenser hotwell without any

,processing. The criterion used to determine whether the drains are returned to the hotwell is that the conductivity be less than lp mho/cm. This conductivity is monitored and if the level reaches lg mho/cm, an alarm sounds in the control room and the condensate can then be routed to the turbine building equipment drain tank.

c. l. Introduction The pressure boundary of the entire upgraded off-gas system 's designed to withstand a hydrogen detonation. In addition, it is also designed to withstand a complete vacuum.

Redundancy exists in all major system components except the mixing nozzle, preheater, and charcoal adsorbers. All these latter three components have no moving parts. The preheater is only needed during startup of the system to ensure that the temperature of the inlet gas and steam mixture at the recombiner is at 350 P. Under these conditions, the steam is superheated to prevent saturated steam from entering the re-combiner. Normally, the mixing nozzle steam will provide sufficient superheat. Wetting of the catalyst is not a serious problem but can cause physical damage to the catalyst by crack-ing the pellets. Electric heaters are provided in each recom-biner to prevent wetting of the catalyst even in the event of a malfunction or danger to the preheater during the system startup.

The chillers or freeze-out heat exchangers are provided in the system to lower the dew point of the gas entering the preadsorbers and the charcoal adsorbers. The lower dewpoint enhances the per-formance of the charcoal adsorbers. The chillers operate by cooling the gas as it passes through the heat exchanger. Cooling is provided by a freon system. Any leakage at this location would be from the freon side to the off-gas side of the system since pressure is higher on the freon side (about 45 psia) than on the off-gas side (about l2 psia) . Leakage is detected and alarmed in the system. Freon does not hinder the performance of the charcoal.

The three chillers in the system perform on a timed cycle of about six hours. Normally, one unit is in service, another is on standby, while the other unit is deicing. Deicing requires about one hour per unit. The six hour in-service cycle is interrupted if the design dewpoint of--4 F is reached before the prescribed 77

time cycle terminates. This situation would put the standby unit immediately into operation automatically.

The preadsorber, which has a redundant, counterpart acts as an expendable charcoal bed. It collects particulate daughter products that result from radioactive decay. Provisions have been made to replace the spent preadsorber charcoal. This can be done with the system fully operated since the redundant pro- .

cedures are segregated by shield walls.

The charcoal adsorbers do not have redundancy but the charcoal adsorbers (6) can be valved such that they can be deivided into two 'banks 'of 'three each. In the event of a problem that is con-fined to the first adsorber, the first three adsorbers can be valved out of service leaving the remaining three operational.

Temperature is monitored at three different locations on each preadsorber and charcoal'dsorber. These monitored points will give an accurate measurement of the charcoal operating temper-ature for performance calculation. The temperature points also give indication and, alarm in the event of rising temperature in the adsorbers.

Either of two redundent vacuum pumps draws a slight vacuum on the entire system back to the mixing nozzle if the pump tripped off, the pressure in the entire system would slowly begin to rise.

h An alarm on the suction side of the vacuum pump would give an indication that the vacuum pump was not functioning. The stand-by unit could be put into service to prevent a further pressure rise.

Electrical power supplies for the system are designed such that all redundant components, except the chillers requiring power, are supplied from different power boards. ,The chillers are all powered through the same board, but there is a backup power source in the unlikely event that the main power source fails.

Malfunctions Table 12-1 lists the major components in the off-gas system.

The table also includes modes of damage or malfunction, action required and consequences of the indicated damage or malfunction.

For example, if there is a loss of dilution steam to the mixing nozzle, decreases in temperature and in flow downstream of the component occur. These result in an alarm and the effect is recorded. If the malfunction cannot be corrected by opening the valve to allow steam or more dilution steam to pass, then the system would be shutdown and repaired. This malfunction results in no release to the plant even if no action is init-iated. The recombiner dilution steam available.

still functions even if there is no 78

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It'IIIAZ55)AIA)CCS X (I) tcraertet lr ttcVllet la ~ Crtc< (generator end)- 2.8 Turbine operating floor (feed. pump end) 3.3 Feed rump area 0.2'.3 Electrical switchgear area Condensate demineralizer valve area 40 Regeneration area 120 Makeup demineralizer area 0.3 Reactor Building Reactor building radiation levels in generally accessible areas are less than 5 mr/hr except as noted., and. posted, as follows:

.(1) Elevation 237 feet: entire level is. considered. a radiation area due to rod. drive pumps, rod drive module area, rod drive decontamination sinks (20 ~/hr at edge of sinks),

rebuilt rod drive storage, and' ventilation duct in the east passageway (overhead reading 450 mr/hr at 1 inch and 15-20 mx'/hr 4 feet off the floor).

(2) Elevation 261 feet: last passage from airlock to precoat tanks, (5-15 mr/hr).

(3) Elevation 261 feet: reactor water sample sink; (25-70 mr/hr standing at sink; 150-200 mr/hr in sink; 100-150 mr/hr 3 inches from turbidity column).

82

t (4) Elevation 281 feet: area by clean-up surge tank; (12 mr/hr at, 3 inches).

(5) Elevation 281 feet: fuel pool filter precoat tank, (20 mx/hr at 3 inches from tank, 5 mr/hr at rope).

(6) Elevation 298 feet: south passageway, west of liquid poison tank, and. west passageway, due to moisture separator and dryer storage pit drain valve (300 mr/hr at 1 inch from valve in overhead.) and drain line.

(7) .Elevation '318 feet; condensate 'line near containment spray heat exchangers, (48 mr/hr at 3 inches).

(8) Elevation 318 feet,:. south. passageway. near, line to spent fuel pool, (12 mr/hr at 3 inches from drain valve) .

(9) Elevation 340 feet: entire level is considered a radia-tion area due to fuel pool (10-20 mr/hr at edge of pool),

tool storage in northeast corner (100-200 mr/hr at 3 inches from tools; 35 mr/hr at rope barrier), fuel rod. stored.

north of reactor top plugs (40 mr/hr at 3 inches), neutron source stored on south side (25 mp /hr at 3 inches; 5 mr i/hr at rope) .

Representative Reactor Building Area Radiation Monitor readings at 1725 K<T operation are as follows:

Area m~r. ar Fuel pool bridge(low-range) 22 Reactor operating floor (equipment hatch area) 0.4 Reactor building equipment drain tank area 2 Closed loop cooling area . 2 Cleanup system area (near pumps) 3 Elevation 281 feet - near fuel pool filters 1 5 Control rod drive module area 5 Spent fuel pool (east end) 2 Instrumentation room (elevation 237 feet) 9 Cont. spray heat exchange area 0.4 Haste Building (1) Control room: the waste building control room radiation level remains below 5 mr/hr under all operating or processing conditions.

(2) Elevation 261 feet - entire level is considered. a High Radiation Area due to the floor drain sample .tanks, which read. 100 mr/hr up to 5 feet away. A fence is under con-struction to limit access to these tanks so that the waste building may be returned.to Radiation Area status.

83

0 (3) Waste concentrator room: Periodic access is necessary for sampling, (1-2 R/hr field in area of sample point).

(4) Flat bed filter room: periodic access is necessary for operation and maintenance, (200 to 1000 mi/hr field de-pendent on process conditions).

(5) Elevation 225 feet,. drum filling and storage:

periodic access to barrel conveyor system is necessary for operational adjustment and maintenance. Radiation levels are dependent on contents of barrels in locale, (spent res'in 10-30 R/hr; concentrated. waste 500-1000 .

/hr).

Area radiation monitor readings are dependent upon process conditions rather than power level. Representative waste building area radiation monitor readings were taken at 1725 MNT operation.

Area m~r4r Radwaste convey Aisle 0.9 Radwaste - pump room 15 Radwaste - control room 0.5 Radwaste - storage and shipping 3

b. Radiation levels fuel handling During uel handling and incore instrumentation replacement operation radiation levels have been measured. The survey ~

made on April 24, 1973 showed the following results:

(1). General level 6 inches above surface of water in fuel pool, reactor cavity and storage pit, (13 mi/hr).

(2) Top rail of hand rails, (3-6 mr/hr).

(3) Bridge over fuel pool, (7mr./hr at 1 inch).

(4) Bridge over storage pit, (10 mr/hr at 1 inch).

(5) Sipping head control center, (3-5 mr/hr at 1 inch).

(6) Under head, southwest corner of Elevation 340 feet, (5 mr/hr).

Several operators, technicians and. supervisory personnel indicated over 10 percent of MPBB uptake of 60 Co during whole body counts from May 21 to May 25, 1973. Assuming a one-time uptake at the beginning of the outage, lung dose for the ca1endar quarter for a technician with 15.6 percent MPBB for 60 Co and 0.8 percent MPBB for 58 Co was 589 mrem. The equivalent whole body dose is 196 mrem, assuming a 3:1 ratio between lung dose and whole body dose.

This calculation assumes that 236 n Ci were absorbed in the lungs at. the start of th<<<<age. A typical CAM reading of 3000 counts 84

per minute (6 x 10" Ci/cc) would result in an uptake of 115 Ci during four 40 hr. weeks. The estimate of 236 Ci may therefore be high, but gives a reasonable upper limit to expected, uptake.

No distinction can be made between uptake due to fuel handling activities and. that associated with Local Power Range Monitor replacement.

No increase in ambient radiation levels were noted during transfer of spent fuel to the storage pit.

The cask for shipment of spent fuel will be built to the applicable standards. No dose estimates are currently avail-able.

The proposed access platform provides for convenient washdown of the reactor cavity and. storage pit areas after the water level has been lowered. Time saved in cleaning, and. lower airborne activity resulting from more efficient cleaning will result in lower doses and. uptake for maintenance personnel.

The resultant decrease will be small on a man-rem basis. The precise figures cannot be estimated from currently available data.

-.c. 'Uentilation Systems By design, building ventilation is provided. to promote air motion from areas of low potential contamination to areas of higher potential contamination. All station ventilation is exhausted by means of the main stack. In addition, the waste building atmosphere is passed. through high efficiency filters prior to entering the main stream at the stack.

Airborne activity levels in station building are monitored by means of low volume samples located in representative areas.

Monthly average readings for March, August and. September 1973 were as follows:

March Auguac September (1) 'urbine building, elevation 261 feet gear .Steam Jet 4.74 x 10 -12 u Ci/cc 2.81 x 10-12u Ci%c 1.98 x 10-12u Ci%c ir Ejector)

(2) Turbine building, elevation 300 feet (operating floor) 1.88 x 10" u Ci/cc 6.15 x 10 u Ci%c 1.64 x 10 u Ci/cc (3) Reactor building, elevation 237 feet -llu Ci/cc (near Control.Rod Drive) 5.61 x 10 3.41 . x 10 12 u Ci/cc 2.93 x 10 u Ci/cc (4) Waste building elevation 247 fe'et 3.30 x 10 -llu Ci/cc 8.10 x 10 -11. u Ci/cc 6.10 x 10 u Ci/cc (5) Large equipment decontamination room 8.43 x 10" u Ci/cc 5.20 x 10" u Ci/cc 4.64 x 10" 3u Ci/cc 85

Inhalation dose to individuals is verified to be low 'by periodic whole body counting. Results of whole body counting for statio personnel during operation would be expected to be highest for operators or mechanics spending significant portions of their time in the waste building. Nuclides detected have remained at low fractions of the maximum permissable body burden (MPBB) limit. This is reflected in the following data for individuals, with body burdens in excess of one percent:

\

work group isotope percent, MPBB counting date operator 6O Co x.6 3/19/73 mechanic 60 Co 1.3 3/2o/V3 operator 60 Co 1.5 3/19/73 operator 60 Co 1.2 3/21/73 mechanic 60 Co 1.9 3/19/73 operator 60 Co 1.7 3/19/73 mechanic 60 Co 1.4 3/19/73 supervisor 60 Co 1.4 3/19/73 mechanic 60 Co 2.8 3/20/73 operator 60 Co 2.1 3/21/V3 mechanic 60 Co 1.5 3/19/73 operator 60 Co 1.1 3/19/73 technician 60 Co 1'. 3/21/73 operator, 60 Co 3'-.2 3/19/73

~all 5.0 3/19/73 operator 60 Co 2.3 3/19/73 supervisor 60 Co 1.2 3/2o/V3 mechynic 60 Co 1.0 3/2o/73 mechanic 60 Co 2.9 3/20/73-mechanic 60 Co 1.4 3/20/73 technician 60 Co 3.'.0' 3/21/v3 mechanic 60 Co 1.1 3/2o/V3 mechanic 60 Co 2.8 3/2o/73 These 'percentages may be considered as chronic levels for the calendar quarter. For the individual with the highest uptake (4.2 percent of MPBB of 60 Co) this amounted to an uptake of 0.046 Ci with resulting dose to the lungs of 154 mrem/quarter.

If the effective whole body dose is considered to be one third of the lung dose, this chronic whole body exposure amounts to 51 mrem/ quarter. (MPBB for 60 Co is considered to be 1100 Ci)

Activity released by way of the main stack is continuously monitored for particulate and iodine by a filter and charcoal cartridge samples. The samples are withdrawn from the gas stream by isokinetic probe located high in the stack. Samples are changed at least weekly. Release levels of iodines and. par-ticulates from"the stack, including off-gas, averaged 6.6 percent of the technical specification release rate limit for the period January to June lg73.

86

Airborne activity is monitored by continuous air monitors located in .the Reactor Building and Turbine Building ventilation exhaust ducts. These monitors alarm in the control room. Also a portable Constant Air Monitor is normally located'n Elevation 261 feet of the Turbine Building. The three Constant Air Monitors cover a range of 50 to 500,000 counts per minute, or approximately 10 11 to 1$ 7 u Ci/cc. Alarm points are set at the 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> MPC for I and Co, ( 9 x 10 9 u Ci/cc) . Constant Air Monitors are calibrated semi-annually. Associated counting equipment is calibrated quarterly Maintenance of Personnel Radiation Exposure as Low as Practicable Records of personnel radiation exposures are maintained. by film badge, charged. every two weeks. Exposure is kept as low as practical by means of the administrative practices discussed below.

Pocket dosimeters are supplied. to all personnel working in radiation areas in the plant. Dosimeter records are kept on cards by the individual. Dosimeter logs are kept by the super-visors of personnel routinely subject to the Region radiation exposures in the plant. In particular, the maintenance and opera ions personnel keep such a log at all times. Logs are also. kept for instrument and radiation protection technicians during maintenance outages.

First - line supervisory personnel are responsible for initiating requests for Authorization to Exceed Radiation'xposure Guides for their personnel. This authorization consists of a review of the individual's exposure record. and authorization up to a specific number of mrem for the week for any individual exceeding 100 mrem/week.

The review is accomplished. by Radiochemistry and Radiation Protection supervisory personnel taking into account exposure history as recorded by'ilm badge, and dosimeter records for the intervening period,.

Radiochemistry and Radiation Protection supervisory personnel routinely compare dosimeter records with film badge results to detect dosimeter malfunctions which could lead. to inadvertent overexposure. As a Radiation Work Permit is required under the following conditions:

(1) Radiation exposure rates greater than 100 mrad. per hour.

(2) Neutron exposure.

(3) Contamination levels greater than 10,000 antres per minute per square foot.

(~) Airborne activity requiring use of respiratory equipment.

87

(5) Maintenance of equipment, controls or instrumentation in Radiation Areas or High Radiation Areas.

(6) Entry into an area of unknown condition.

Radiation Work Permits are made out by senior Radiation Pro-tection technicians with supervisory personnel on ca11 for consu1tation. Necessary work provisions are specified on the permit for control of radiation exoosure.

A system utilizing thermoluminescent dosimeters (TLD's) is being initiated. The TLD reader is now being calibrated.

'Personnel training and procedure writing is in progress.

Initial planning is underway with the Engineering Department for computer assistance to facilitate calculation of best "dose to date", information utilizing Film Badge, TLD and.

pocket dosimeter data.

88

J

4. QUESTION Your description of compliance with Regulatory Guide 1.17 as given on page III-44 of the Application is not adequate. Please provide the details of your Industrial Security Plan in accordance with Regulatory Guide 1.17 dated June 1973, as prop-rietary information in conformance with Section 2.790 (d) of 10 CFR Part 2.

RESPONSE

The reply to this question is being submitted under separate cover as proprietary information.

89

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