ML20003H625

Forwards Round 2 Requests for Addl Info to Complete Review of Application for Ol.Periodic Inservice Testing of Second Safety Valves Should Be Specified in Tech Specs

ML20003H625
Person / Time
Site:	Waterford
Issue date:	04/29/1981
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Aswell D LOUISIANA POWER & LIGHT CO.
References
NUDOCS 8105060524
Download: ML20003H625 (27)

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NUCLEAR REGULATORY COMMISSION r;

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APR 2 9 1981 Docket No. : 50-382 Mr. D. L. Aswell Vice President, Power Production Louisiana Power and Light Company 142 Delaronde Street New Orleans, Louisiana 70174

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION - WATERFORD 3 We have deteimined that-Sertain additional infonnation is required in order to permit us to complete our review of your application for an operating license fo'- Waterford Steam Electric Station, Unit 3.

The enclosed round two requests for additional infonnation were prepared by several branches.

Please advise us of the date you expect to provide responses to the;.pe,,,:,,

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enclosed request.

If you require any clarificatica, please contact the staff's assigned project manager.

Sincerely, hu Robert L. Tedesco,- Assistant Director for Licensing Division of Licensing i

Enclosures:

Request for Additional Information i

211.92 - 211.114 l

010.41 & 010.42 640.1 - 640.16 cc w/ enclosure:

See next page.

810 5060 fd y

Mr. D. L. Aswell Vice President, Power Production Louisiana Power & Light Company 142 Delaronde Street Nj'w Orleans, Louisiana 70174 cc:

W. Malcolm Stevenson, Esq.

Monroe & Lemann 1424 Whitney Building New Orleans, Louisiana 70130 Mr. E. Blake Shaw, Pittman, Potts and Trowbridge 1800 M Street, N. W.

Washington, D. C.

20036 Mr. D. B. Lester.

Production Engineer Louisiana Power & Light Company i

142 Delaronde Street New Orleans, Louisiana 70174 ~

Lyman L. Jones, Jr., Esq.

Gillespie &' Jones

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. P. 0. -Box 9216 ' * ^ * -

_Metairie,_Lquisiana 70005 l

Luke Fontana, Esq.

l Gillespie & Jones 824 Esplanade Avenue i

New Orleans, Louisiana 70116 Stephen M. Irving, Esq.

One American Place, Suite 1601 l

Baton Rouge, Louisiana 70825 Resident Inspector /Waterford NPS P. O. Box 822 Killona, Louisiana 70066 l

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APR 2 a lg8l DISTRIBUTION:

Docket File 50-382 LPDR PDR NSIC l

eTERAS

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UATECTOT.3 STE*C ELECTRIC STATIO:;, U:ll7 3 SEC0f!D COU:!D r.EOUEST TOT: ADDITI0liAL Il! TOR;;*sTIO!

21).92 Your. response to our Question 211.20 is not adequate.

The follou-(5.2A,5.2B ing information shculd be provided:

cod 9.3 0)

_ (1)

SRP 5.2.2 ' states that the high pressure reactor trip or second safety grade scra'm signal, whichever occurs later, should be used for sizing the primary system safety valves. The informa-tion provided in your response to our question 211.80 does not satisfy this requirement. Your FSAR indicates that during loss of load transient, the low steam generator level trip and the high pressurizer pressure trip will be reach about the scroc tiu::.

Confirm that the Waterford safety valves are sized assuming that the reactor is scrammed on the second safety grade scram signal.

(2)

Section 9.3.6.2 states that for overpressure protection of the RCS during low terperature conditions, safety relief valves (SI-485 and SI-487) are provided in the shutdown cooling system suction lines.

Each valve has a, design relief capacity of 3004 GPihat.the 415 psig setpoint. Since these safety relief valves are not included in the safety and relief valve testing program by EPRI, we require that you. provide additional informa- -

tion for those valves.. Specifically, state what versiun of the ASME cLC,:as.used for the., design of these valves, and,4uantify the' margin:available (i;e.,' maximum ~ expected'rel.ief f' low vs. design

. relief flow) in 'relleing'capacityifor1:he= worst overpressure transient conditions *forwhicifth'ese valves:are-designed to protect-.

against.

1 (3)

Periodic in-service testing of the second safety valves should be

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specified in the technical specifications.

To ensure that Waterford will have a reliable ove'rpressure protection 211.93 (5.2B) system for the.RCS during low temperature. condition's, it is our -

position that -

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1 (1)

The technical specifications should. include-req'uirements:to ensure the RCS is on-shutdown. cooling. systems wit _h all suction line valves open wherever the RCS ter3perature~is below-280V (or a temperature that is appropriate:for Waterford);

(2)

The technical specifications should prohibit.a'ctuat. ion' of. a reactor coolant pump if the associated steam generator to reactor coolant 0

systems AT is greater than 100 F; (3)

The set point for the automatic isolation of the SDCS should be raised to 700 psig (or a pressure which is appropriate for Waterfor l

(4)

The technical specificationshould _ include testing of valves SI-486 and SI-487, the SDCS safety relief valves, at intervals not to aceed thirty months in order to provide increased assurance of

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l 4'alve operability.

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211.H k Your shutdown cooling syste. :.hould meet the requirements icentifiol (5.4.7. a rd in BTP RSB 5-1.

As part of our review to deterr.ine your ccipliance 9.3.E) with.BTP, we will require the follouing information:

-(l)

A cooldown. analysis and curve should be provided which shows how much time is required to bring the plant from power operation j

to the RHR entry point. The analysis should be done assuming only 1

safety grade systems are available, only onsite power available, and the most limiting single active failure.

Your response to i

questions 211.14 which states that both steam generators can be blown down from a single atmospheric dump valve, is unacceptable, since the cross tie is located dounstream of the MSIY in non-seismic piping. The analysis must be conducted taking no credit for the cross tie. Also, provide a detailed description of all actions necessary to reach cold shutdown.

(2)

Detennine how much auxiliary feedwater would be required to maintain the plant at hot standby for four hours and 'then cool-doven in the manner spcified in your response to Part (I) above.

Confinn that your facility has sufficient seismically qualified water storage to meet this need.

(3)

Your planned natural circulation test does not_ include provisions for demonstrating adequate boron mixing when forced circulation ir n6t 'present.

Expand the scogef. the.tsst 'trinclude:this t < v v

n demonstration! Also,; discuss your actions.to_(1). prevent voiding in the-upper head. during. natural circul.ation' conditions and-(2)-

train the operators to recognize voiding should it occur. '-

(4)

Document ~ that the unavailability of the letdown system will not

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hamper efforts to cool and depressurize the primary system.

Demonstrate the inventory shrinkage due to cooldown will allow for sufficient pressurizer spray to meet the objectives of de-pressuri.ng the plant to shutdown cooling system 1:imits within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Also, demonstrate that any realign =ent of the charging pump suction, due to the unavailability of th'e letdown system,

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can be accomplished from the control room.

(5)

Discussion with the applicant has indicated that a number.of. valves-in the shutdown cooling system are being modified to include -

motor operators,- Provide-details-of these-modifications---Demon-strate that with these modifications, all actions necessary to initiate shutdown cooling can be taken from the control room, utilizing only safety grade systems, and assuming only onsite power is available. Also, indicate whether there are any systems or components needed for shutdown cooling which are de-energized or have power locked out during plant ' operation.

If so, indicate what actions have to be taken to restore operability to the components or systems.

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cooling system instrumentation de ig icd to c1tss 1E remire-nients is the inlet temperature m:asurcm:r.ts 01 the fi..tD.s and the low prc::urc header tcrper..ture indic: tor:,. The staff believes that this limited instru::.cntr. tion previd.. it.-

sufficient infor7r.ation to aid the operators in assetsir LPSI pump performance.

Instrumentation which is nore.lly relied upon to place the picnt into :hutdoun :coling, tuch as LPI heade,r pressure and shutdown cooling flow, has been previously found acceptable by the staff and should be available for the shutdown scenerios addressed in BIP RSS 5-1.

We request that you provide sufficient instrumentation to monitor SDC performanc'e, in particular SDC pump flow and pressure, or justify how the present design will provide the necessary information to the operators.

Any additional instrumentation required should be shown to be available in the absence of offsite power.

(7)

Additional protection must be provided to preclude damage to the LPSI pumps due to overheating, cavitation, or loss of adequate pump suction.

The staff will require that alarms be provided to alcrt the operating staff to conditions where low LTSI flow may be symptomatic of difficulties with the pump suction on: dis. charge which could cause pump damage.

Y.

In section 15.441.4 of the FSAR dealing with Boron Dilution transients, 211.95

(,1,5.4.1.4)

-_J.. it states mode _5_(s ble :tfriifor dets,the, situation which-results in.

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l support for this;.statementi.2 Also-indicater for all. six modes--

what alarms would. identify-torthe op~erjatersTthat,a Soron di.lutioh cevent-- '

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. =..-.. s.o ccur.ri..ng.

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~' ~ Show that the plant is protected for all postulated baron dilution events.

assuming the worst single active failure.

In particular, consider the~'.

L failure of the first alarm.

If a second alarm is not provided, show that:

the consequences of the most limiting unmitigated boren dilution' event mee-I te staff criteria and are acceptable.,,

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l, 211.96 k

. In regard to Question 211.49 concerning baron dilution events, severaL..

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(15.4.1.4) recent LERs indicate there ha's been a deficiency in the inadvertent ~~ ~

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born dilution analysis at some plants. Provide an analysis of the -

l dilution event when the RCS is drained to the. hot leg.

211.97 g In response to Question 211.84, it was stated that followimg~ a moderate (6.3) energy line break when in shutdown cooling, the operators would take ~

corrective action 10 minutes after the first alam. For any possible break locations in the moderate energy ' piping, indicate all actions which must be taken to identify the break location, isolats'the break and restor core cooling. A'ssuming the most limiting single ~ failure (such as nonfunc-tion ~ of first alarm indication) demonstrate that all actions described above can be completed within the 10 minute interval claimed in the l

question response. Justify time estimates for each step discussed above, include appropriate transit times for operators to reach locations outside

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of the control room. When answering the above question, consider that the piant m:y be in refueling mode, and consequently the pressurizer low level alacm may not be available.

EBCLOSUF.E

If your t.r.aly:.is ir.dici:1(s that 30 minutes is nst.. n.

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'. u.e io respond to t. :..:.A rz.t e energy line brer.k tihen i:

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e nl-ing, discu.5 that r.:t.:.ures you will take to incrt ::'

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we.. ;.r action tir: to r. crept::ble vi: lues.

21'l.98 Your response to our Question 211.7 indicated that you assume (6.3) a naximum leakage from e passive failure of ECCS to be LC3 cc/ min.

-This is not a conservative assumption. It is our position.

that you assume a maximum passive failure flow rate of 50 GPM in each ECCS pump room and discuss the effects of the passive failure to each ECCS pump operation and demonstrate that adequate protection is provided for ECCS pumps from possible flooding..

211.99 Your response to our question 211.61 is not acceptable.

A single

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(6.3) r.t.: ual valve (3SI-Y109 A/B) in the corson ECCS pump minimum recirculation line does not meet single failure criterion. A single operator error to close this valve or a mechanical failure It is of the valve will cause possible damage.of all ECCS pumps.

our position that you modify the system to prevent system damage due to the single failure. The following alternatives are considered acceptable to the staff.

(1)

Remove valve 3SI-V109 A/B from the co= mon ECCS pump recirculation line.

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(2)

Provide parallel valves in the common ECCS pump recirculation line. The valves should be normally locked operi by administra -

tive controls....Also. provide limit-switches on thWalves 'ar.d g, ',,,, '

the valve posit 4ons-sho.uld:be. indicated and alarmed ih the control room (using computiriCRT. display.and ~ alarms are__ --,.~...

acceptable). '-

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It is our position that limit' switches which enable valve posiden 211.100 to be indicated in the control room and monitored by computer should (6.3) be installed on all manually operated and normally locked open ECCS

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valves listed in Tsble 211.85 1.

Confirm that the proposed containment sump tests will meet-the test 211.101 (6.2.2.

requirements specified in Question 211.64 and provide a schedule for and 6.3) submittal of thh test report.

Your response to our Questio 211.66 is not ' sufficient.

Detailed.

211.102 (6.2.2)

. __. design drawings _of the SIS sump should be provided for-staff review as a part of ECCS review.

'The method that you have used for calculating the amount of, failed 211.103 fuel after an accident has n~ot been reviewed and approved.

It is our (15.0) po,sition that, fuel' failures be recalculated using the criteria that Radio-any fuel rod which has a CE-1 DNBR less than 1.19 fails.

logical' consequences should be calculated accordingly.

Se,ction 15.2.3.1 stated that the maximum RCS pressure during the 211.104 (15.2.3.1) postDiated feedwater system pipe break accident is 2832 psia (approximately 113% of design pressure). Confirm that this maximum RCS pressure will not result in the stress of the RCS components to exceed the allowable stress limit for emergency conditions (120% of design value) specified in ASME Boiler and Pressure Vessel Code P00RORM3AL Section III, Division 1.

211.105 Your rc:;ponse to Question 211J5 is not cdct;u:tb.

Provida re rei t <.

(15.3) of an analysis of the reactor coolant pump shaf t break as rcquinJ by Section 15.3.4 of the Standard Review Plan for staff reviec.

The amount of failed fuel should be calculated c:ir.; th: critui:

_. stated in Question 211.103 and assess the radiological consequences of this event accordingly.

Trip functions assuced to protect against this event should be safety grade.and should be acccmplished assuming the worst single failure,of a safety system active component.

1 21'l.106 The analysis of the reactor ~ coolant pump rotor seizure event should (15.3.3) be presented in a manner to show that the a'cceptance criteria of Section 15.3.3, Revision 1, of the Standi:rd Revien P1tn are tet Eac.h of these criteria should be specifically addressed, and any deviations from the criteria should be justified. The event should be analyzed assuming turbine trip and coincident loss of offsite power and coastdown of undamaged pumps.

Only safety-grade equip :nt should be used to mitigate the consequences of the accident. Safety functions should be accc=plished assuming the..orst single failure of a safety-system active cc=ponent. The amount of fuel failure should be calculated using the criteria specified in Question 211.103.

In regard to your response to Question 211.46, the assumption that release of activity to the atmosphere is terminated when shutdown cooling is initiated should be justified or leakage (assu=ing.a

" steam generatorfsirstfMlve stuck _opeh) sitouldC#esG:td teyond-the initiation' of: shutdown' cooling." Also, the technical specification limit. :r steam generator-tube-leakage of-l--gpm.should be. assumed....

for the analysisW- -

1 211.107

~ Vou~r response ~to question 211.60 is not-adequate. - Provide.an analysis. '

t (15.6)

~ of single failures in the CVCS as these relate to use of the1CVCS for a small break. The following specific points should be a~ddre~ssed: J i

(1)

Provide a discussion of flow splitting.if the.small brealIis ~

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a-DEG break of a charging line downstream of:the last isulation f.

val ves. -

(2)

Address ths consequencds of one.of the' chargingiine' injection

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valves (e.g., D{-518) to fail closed when the. break is.in the other charging line or the loop to which it is attached.

t (3)

Discuss the effect.of the failure of the VCT isolation valve

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CH-501 to close on SIS actuation.

(4)

For items.2) and 3),- consider each as.an independent single failure, as well as bo'th actions resulting from a comnon initiator l

such as failure of a diesel generator to start.

For these specific items, as well as others which may be appropriate, show that the small break assumptions and analysis are conservative and that the results meet the acceptance criteria.

ENCLOSURE

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2il.108 Your response to Question 211.57 is not :,ufficimit.

Tiit cr.i. lysis (IE.f) of the inadverter t ope:iing of a pressurizer sefcty v;.1ve :,huuid Lt presented to show that the acceptance criteria cf secticn 15.0.1 of the Standard Review Plan are met. Each of thesc criteria sL::cid be specifically addressed ar.d any deviations fro the criteria should

_be justified.

A curve of DNBR vs. time for this transient should 4

be provided to confirm that there is no DNB during this event.

211.109 The HRC is currently considering what actions may be necessary to (15.8) reduce the probability and consequences of anticipated transients without scram (AWS). Until such time as the Cornission determines what plant modifications are necessary, we have generally concluded thtt pressurized water plants can continue to operate becausc the risk from anticipated transient without scram events in a limited time period is acceptably small.

However, in order to further reduce the risk from anticipated transient without scram events during the interim period before completing the plant modificaticns determined by the Commission to be necessary, we have requi, red that the following actions.be taken:

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O) Develop emergency procedures to train operators to recogni c anticipated transient without scram events including consideration of scrara indicators, rod position indicators, flux monitors, pressurizer leve1~ and pressure indicators, pressurizer relief" 4-and safety valve indicators, and any other alarms annunicated in#-'-

the control room with emphasis on alarms not processed through--

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the electrical portion of the reactor-sciam system. --- -

.(2) Train operators to take-actions.in'-the-event of an anticipated.

transient:without scrama-including corisi'deration of mtnually scramming the reactor by using the manual scram button, prompt.

l actuation of the,axuiliary feedwater system to assure delivery to the full capacity of this system, and initiation of turbine.

trip. The operator should also be trained tu initiate boration by actuation of the high pressure safe injection system to bring the facility to a safe shutdown condition.

r Describe how you will meet the above requirements, and provide a schedule for submittal of your AWS procedures for staff review.

l 211.110 Your response to item H.B.1 of. NUREG-0737-requirements-15 riot ~ ~

l (II.B.1)

-- sufficient. Section 5.4.15 of FSAR does not include adequate' informa-tion for Reactor Coolant Gas Venting System.

Provide the following:

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(1) Provide diagrams.and-a description of-the vent discharge., vicinity. -

Verify that adequate ventilation is provided and that equipment in this area 1s. capable of' withstanding discharge of gases and liquids from.the vents.

(2) What size are the flow limiting orifices and what are the calculated

. flow rates through the vent system for both gas mixtures and liquids at operating pressures?

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(3)

Provide drawings of the piping sysic. from the vessel head and pressuri cr through the disch rgt p ths.

In ptriicular.

show the location of the solenoid operated valves and consider potential missile hazards from them.

211.111 Your response to item II.K.1 indicates that.a review of all ESF (II.K.1 )

valve positions, controls and test and maintenance procedures is conducted during procedure preparation to ensure proper ESF function-ing.

Confirm that your review of all procedures concerning ECCS valve operations will be completed and documented prior to fuel loading in order to meet the schedular requirements of liUREG-0737.

211.112 Your response to items II.K.2.13 and II.K.2.17 is not complete.

(II.K.2.13 T, Provide a commitment that you will provide the results of your II.K.2.17) evaluation of the items by January 1,1982, in order to meet the schedular requirements of liUREG-0737.

211.113 Your respense to item II.K.3.17 is not complete.

Provide a comit-(II. K. 3.17) ment that you w.ill establish a program prior to fuel loading for data collection on information regarding ECCS outages.

The informa-tion will contain: 1) outage dates and duration of outagesy (2) cause of the outage; 3) ECCS systems or components involved in the outage; r,. and 4) corrective action taken.-- -

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211.114

- Your response 'to items.II.Kb.30 and II~.K.3.'31.is not 'c6n@lete. -

(II.K.3.30.. Provide"a cosnitm'ent to. submit-a revise'd-SBLOCA.model. nd/or r:-Z #

. A II.K.3.31P-Ydditio' al,iustificatibn.foiTthe piesentc.model_by:Januar-y-1 r-1982; --- _.-

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-and rev.ised SBLOCA ECCS analysesr'if necessary, within one year-after staff approval of the SBLOCA models, per schedular' requirements of NUREG-0737.

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l P00R ORS'Ml EliCLOSURE

010.00 A5XILIARYSYSTEMSB' RANCH 010.41 Identify sensitivities and response times for intersystem leakage -

(5.2.5) detection systems.

010.42 Provide a technical specification which states that should leakage be alarmed and confirmed in a flow path with no flow meters, a water iiiventory material balance must be begun within one hour to determine the extent of the leakage, t

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640.1 Regulatory Guide 1.68 (C.7) states that, to the extent (14. 2.1, 14.2.9) practical, plant operating and emergency procedures should be trial-tested and corrected during the initial test program prior to fuel loading to establish their adequacy. Modify Subsections 14.2.1 and 14.2.9 to state that procedures will be upgraded as necessary.

640.2 (RSP) The qualifications requirements for testing personnel (14.2.2.8) as stated in 44.2.2.8 are not acceptable.

It is the staff's position that personnel qualifications be in e3

..,, conformance with ANS 3.1 Draft ofc.0ctc5en 23,1980.

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Modify 14.2.2.8 to comply with this position..

640.3 Subsection 14.2.3.b refers to Regulatory Guide 1.33, Appendix (14.2.3.b)

A, Section 6 for low power and power ascension test procedures to be reviewed by Combustion Engineering. Delete this reference.

The current Regulatory Guide does not address low pcwer and i

I power ascension tests. Your reference should be consistent with 1.9.14 of your FSAR.

640.4 Certain regulatory guides listed in Subsection 14.2.7 do not (14.2.7) reference the correct revision applicable to the Waterford #3

__ plant. Modify Subsection 14.2.7 as listed below:

1.32 (Rev 2, 2/77)

(1) 14.2.7.5 1.52 (Rev 2, 3/78)

(2) 14.2.7.9 1.63 (Rev 2, 7/78)

(3) 14.2.7.11 (4) 14.2.7.18 1.95 (Rev1,1/77) 1.129 (Rev1,2/78)

(5) 14.2.7.22 640.5 (1)

Regulatory Guide 1.68 states that the pro' er operation (14.2.7.13) i of decelerating devices used to prevent mechanical damage to control rods should be demonstrated during rod drop test N o.regraph 14.2.7..13 1,dpdicates that no-flow,.,,..

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rod drop tests to be performed will be hot, group drop tests during which the computer will be used to measure drop times.

Explain how the oroper operation of all decelerating devices will be tested.

640.5 (2) The Regulatory Guide 1.68 criterion for minimum l

(14.2.7.13) acceptable neutron count rate on the startup channels, before startup begins, is 1/2 count per second with a signal-to-noise ratio greater than two.

14.2.7.13.2 indicates that Waterford 3 may see significantly less at the initiation of initial approach to criticality but when the ARO condition is attained the prescribed count rate will be attained and multiplication will ' eo 0.98.

Provide expanded technical justification for this exception to Regulatory Guide 1.68.

i 640.5 (3) Subsections 14.2.7.13.3 and 14.2.7.13.6 state that the (14.2.7.13)

Xenon oscillation control, pseudo-rod-ejection, and dropped CEA tests will not be performed since they will have been conducted on a core essentially identical to the Waterford #3 unit (San Onofre #2).

Provide the acceptance criteria that will be used to establish that an adequate similarity exists between the two plants.

These criteria will be for CEA* symmetry checks, CEA group worths, critical' boron concentrations and nuclear instrumentation data. The acceptance criteria should be quantitative agreement tolerances between measured values for San Onofre Unit 2 and the measured valves for Waterford Unit 3.

They must be wtightes than the quantitative c;,rehment'tb10anEeTthat San ' ' # *U Onofre Unit 2 used when comparing. predicted and measured values.

If these criteria are not met, commit to perfn'rming -

l-the tests which were deleted (because core identica1 ness has not been demonstrated).

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640.5 (4) Provide acceptance criteria that will be used to compare (14.2.7.13) the power reactivity coefficients determined at 50% and 100% power at San Onofre Unit 2 and Waterford Unit 3.

These criteria should be quantitative, and tighter than those used by San Onofre Unit 2 when comparing predicted and measured valuves.

Commit to a complete set of power reactivit)coefficientmeasurementsifacceptancecriteria are not met (14.2.7.13.5).

- 4 640.5 (5) Subsection 14.2.7.13.7 states that flux instrumentation (14.2.7) will not be tested for the ability to detect CEA misalignment because CEA position is determined by redundant reed switches.

Explain how the flux detectors will be tested for sensitivity.

640.5 (6) Subsection 14.2.7.13.9 takes exception to the Regulatory (14.2.7.13)

Guide 1.68 position that testing should be sufficiently.

comprehensive to establish that the facility can operate in all operat,ing modes for which th'e facility has been designed to operate stating that step and ramp changes of full dys,ign,;va;1ue wi,11 no,tje,oerformed., Step and 3

ramp changes of full design value need not be performed if changes of lower value can accurately establish the ability to properly respond to changes of design value.

Explain how the data will be used to determine ability to properly respond to design value step and ramp changes.

640.5 (7)

In subsection 14.2.7.13.10 provide technical justification (14. 2.7) that a loss of Reactor Coolant System (RCS) flow at 80%

power provides the same plant verification that performance at 100% power provides, or delete this exception and modify the appropriate test description.

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5-640.5 (8) Your statement in subsection 14.2.7.13.11 provides (14.2.7.13)-

inadequate technical justification for substituting the 100% turbine trip test for the 100% main steam line isolation valves closure. Provide technical justification for the test substitution, provide l

technical justification for performing the test at i

a lower power level in accordance with Regulatory Guide 1.68 (App. A, 5.mm), or delete this exception l

and include the appropriate ' test description.

Explain your statement in 10.3.3 that operation of MSIV's during normal operation would cause severe L w,

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transients.

640.5 (9) Subsection 6.3.4 states that calculations will be

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(14.2.7.15) l performed to verify vortex and NPSH conditions.

14.2.7.15 states that model tests will be used.

Clarify the discrepancy and provide references l

for the model testing to be performed.

l 640.5 (10)

Indicate if a control room gross leakage test will (14.2.7) be performed in accordance with Regulatory Guide 1.95(CS).

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640.6 Modify Subsection 14.2.8 to establish that your operating (14.2.8) and testing experience program will at a minimum consist

- of an initial review to be administered prior to conducting preoperational tests and an ongoing review during the remainder of the test program.

640.7 Subsection 14.2.5 states that Phase I and II testing will not (14.2.

12.2) necessarily be completed prior to comencing Phase III testing.

If portions of any preoperational tests are intended to be conducted, or their results approved, after fuel loading:

(1) List each: test; (2) State what portions of each test will be delayed until after fue,1,rdoad.inr;-

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(3) Provide technical justification for delaying these portions; (4) State when each test will be completed (key to test condi-tions defined in Chapter 14).

(5) Document that completed portions of each test will be sub-ject to conditional review and approval prior to fuel load.

Note that any test not begun prior to fuel loading should be included in Phase III instead of in Phase II test descriptions...

640.8 Provide a table or figure that indicates the power levels at (14.2.12.3) which each of the Phase III tests will be performed.

Provide assurance that all tests to be performed at a given power level are performed prior to exceeding that

' power level for the first time, or indicate and justify exceptions.

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

7-640.9 Identify any of the initial startup tests described in Subsection (14.2.12.3) 14.2.12.3 which are not essential towards the demonstration J

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of conformance with design requirements for structures, systems, components, and design features that:

(a) will be relied upon for safe shutdown and cooldown of the reactor under normal plant conditions and for main-taining the reactor in a safe condition for an extended shutdown period; or (b) will be relied upon for safe shutdown and cooldown-of-the reactor under transient (infrequent or moderately frequent.pven,ts) conditions and pos'tulated accident conditions, and for maintaining the reactor in a safe condition for an extended shutdown period following y:.., w c,..

such conditions; or (c) will be relied upon for establishing conformance with safety limits or limiting conditions for operation that will be included in the facility technical specifications; or l

(d) are classified as engineered safety features or will be relied upon to support or ensure the operations of engineered safety features within design limits; or (e) are assumed to function or for which credit is taken I

in the accident analysis for the facility (as described in the Final Safety Analysis Report); or-(f) will be utilized to process, store, control, or limit the release of radioactive materials.

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. 640.10 Our review of your test program description disclosed that thG (14.2.12) operability of s,everal of the systems and components listed in Regulatory Guide 1.68 (Rev. 2), Appendix A, may not be demonstrated by your initial test program.

Expand your FSAR to include appro--

priate test descriptions (or identify existing descriptions) to address the following items from Appendix A of the guide:

(A) Preoperational Testing:

(1) residual heat removal system (as described in 9.3.6.4)

(2) tests of protective devices such as leak-tight covers, structures, or housing,s provided to protect engineer'ed safety features from flooding.

(3) containment design overpressure structural tests (ChaptcE 14 should acknowledge the test described in 3.8.2.7).

(4) containment penetration pressurization system l

(5) containment penetration cooling system (6) leak detection systems used to detect failures in ECCS and containment recirculating spray systems located outside containment (7) pressure control systems designed to prevent i

leakage across boundaries (feedwater leakage control system)

(8) auxiliary startup instruuents (neutron response checks)

(9) instrumentation used to detect external and internal flooding conditions that could result-from such sources as fluid system piping failures such as the system described in 3.6A.6.4.22

-g.

640 10 (10) isolation features for liquid radwaste effluent (14.2.12)

Continued.

systems (11) operability and leak tests of sectionalizing devices and drains in the refueling canal and fuel storage pool (12) vent and drain systems for contaiminated and potentially contaminated systems and areas, and drain and pumping systems serving essential areas (B) Fuel Load and Pre-critical Tests (1) final cal:bration of source-range neutron flux measuring instrumentation and proper operation of alarms and protective functions of source-and intermediate-range monitors (C) Low Power Tests (1) demonstrate adequate overlap of source-and intermediate-range neutron instrumentation

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(D) Power-Ascension Tests (1) response times of main stream line isolation (MSIV)

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and branch steam line isolation valves; measure the full travel of the valves or provide technical justi-

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fication for extrapolating the full closure time.

(2) demonstration of adequate beginning-of-life performance margins for shielding and penetration cooling systems or demonstration that concrete temperatues do not exce.ed design limits when coolers are not used.

(3) process and effluent radiation monitoring systems using independent analysis of samples to verify system operation.

640.10 (4) demonstration that the dynamic response of the plant (14.2.12)

Continued is in accordance with design for the loss of or bypass of the. feedwater heaters from a credible single failure or operator error that would result in the most severe case of feedwater temperature reduction 640.11 We could not conclude from our review of your Phase II and Phase III 1

(14.2.12) test descriptions that comprehensive testing is scheduled for several systems and components. Therefore, clarify or expand the appropriate test descriptions to address the following:

(1) 14.2.12.2.1 - Verify that each emergency load can start and operate 2t the minimum voltage level at which it can be postulated to operate, i.e., after the design discharge s,

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(2) 14.2.12.2.13.4 - Provide a reference or actual criterion that stipulates the required design performance for the Nitrogen -

l System.

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t (3) 14.2.12.2.17 - Modify this test description as necessary to conform to Regulatory Positions 2a(3), (5), (6)(a),

(6)(b), and (9) and 2.b in Regulatory Guide 1.108.

(4) 14.2.12.2.18.3.E - Revise the test method to state that l

full operational testing will be accomplished at 100% of l

rated load.

(5) 14.2.12.2.19.3.E - Describe how you will verify that containment l

recirculation fan motor currents will be within design value un er post-accident conditions. Provide a description of testing to be performed during the containment pressure l

test and address such issues as air density, temperature, humidity, fan speed and blade angle.

. 640.11 (6) 14.2.12.2.21.4 - Include Subsection 9.4.5.5 along with (14. 2.12 ) '

Continued 9.4.5.8 as the references in acceptance criteria.

(7) 14.2.12.2.31.4 - Supply missing heading for acceptance criteria section.

(8) 14.2.12.2.36.4 - Modify the acceptance criteria to reference 5.4.13.4 or expand the criteria to more clearly verify adequate pressurizer safety valve perfonnance.

(9) 14.2.12.2.44.3 - Modify the test descript'on to verify

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correct flow paths and sampling procedures and to assure,

that sample holdup times are within allowable limits, orifices are correct, and relief valves tested.

l (10) 14.2.12.2.49; 50; and.51 - Describe tests that show

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conformance to positions 2.b.(2), (3); 2.c.(1), (4), (5);

and 2.e of Regulatory Guide 1.79.

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(11) 14.2.12.2.52.4 - Provide dynamic and static load testin'g of all lifting and rigging equipment in accordance with NUREG-0554.

(12) 14.2.12.2.69.4 - Expand the reference so that it includ'es ~

10.4.10.

(13) 14.2.12.2.70.4 - Add Table 6.2-32 to references (to include MSIV closure time tests).

(14) 14.2.12.2.73.5 - The information listed in Section 10.2 is not sufficient to adequately verify the proper operation of the MSR and feedwater drain systems.

Provide a reference or modify this subsection to clarify the acceptance criteria necessary to validate this system.

12 -

640.11 (15) 14.2.12.2.75.4 - Provide a reference or actual criteria (14.2.12)

Continued that stipulates the required design perfonnance for the

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Feedpump Lube Oil System.

(16) 14.2.12.2.77.4 - Include Subsection 7.7.1.4.2 as a reference for the acceptance criteria.

(17) 14.2.12.2.90.4 - List or reference the applicable code requirements (FSAR Subsection 5.4.2.4.1).

(18) 14.2.12.2.91.4 - List or reference the applicable code requirements that must be met.

(19) 14.2.12.2.94.4 - Expand the acceptance criteria to list which applicable sections of the FSAR, and what exact design criteria will be used to ensure that the objectives of the

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Hot Functional Testing have been achieved.

(20) 14.2.12.3.1.4. A - List or reference the design criteria to which CEDM operation will be compared.

(21) 14.2.12.3.3.4. A - List or reference the design tolerances to which the leakage resistance of the incore detectors will be compared.

(22) 14.2.12.3.4.4 - Provide bases for the acceptance criteria that will be used to determine the degree of agreement between the appropriate outputs and the plant conditions.

(23) 14.2.12.3.5.4 - List or reference the desired amount of continuous spray and pressurizer pressure reduction design

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requirements referred to in Subsections A and B.

(24) 14.2.12.3.6.4 - Provide bases for the criteria used to determine satisfactory agreement between measured and actual heat loss.

. 640.11 (25' 14.2.12.3.10.4 - Provide bases for the acceptance criteria (14.2.12)-

Continued to be used to determine satisfactory agreement (a) between the measured and predicted ITCs, and (b) between the derived moderator temperature coefficients and the Technical Specifi-cations as well as the safety analysis.

(26) 14.2.12.3.11.4 - Provide bases for the acceptance criteria to be used to determine satisfactory agreement with the measured values.

(27) 14.2.12.3.12.4 - Provide the bases for the acceptance criteria to be used to determine satisfactory agreement with measured CEA group worths, and specify how the verification of adequate shutdown margin will be conducted

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v.,z in SubsE;tLYiWO (28) 14.2.12.3.13.4 - Provide bases for the ac,ceptance criteria to be used to determine that the measured boron worths are in satisfactory agreement with predicted values.

(29) 14.2.12.3.15.4 - List or reference the design criteria used to verify natural circulation cooling in Subsection B.

Expand objectives to include those stated in our letter of November 14, 1980.

(30) 14.2.12.3.18.4 - List or reference the criteria that will be used to determine that the recorded radiation levels are acceptable for full-power operation.

l (31) 14.2.12.3.19.4 - List or reference the chemistry specifica-tions citid in Subsection A, the criteria that will be used to demonstrate the appropriateness of sampling frequencies in Subsecticn B, and provide the bases for the acceptance criteria to be used to determine satisfactory agreement between laboratory analyses and process radiation monitors l

in Subsection F.

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. 640.11 (32) 14.2.12.3.20.4.A - List or reference the criteria or provide (14.2.12)

Continued bases for the criteria that will be used to determine that sufficient allowance is made to allow for thermal expansion, and that the shock and vibration measurements satisfactorily agree with the stress analysis.

(33) 14.2.12.3.26.4 - Provide bases for the acceptance criteria to be used in determining that the measured power and i

isothermal coefficients are in satisfactory agreement l

with the predicted values.

(34) 14.2.12.3.27.4 - List, reference, or provide bases for the l

criteria and/or tolerances that wilf te used to establish that:

(a) In Subsection A, the results of the measurements at N '

.each power level indicate that acqqptable..corform3nce.,

at the power level can be expected.

(b)

In Subsection B, the agreement between predicted and measured power distributions is satisfactory.

(c)

In Subsection C, the measured peaking factors are in satisfactory agreement with the predicted values.

(d)

In Subsection D, calculations are in agreement with predicted values.

(35) 14.2.12.3.28.4 - Indicate the source of predicted values and acceptance criteria to be used to. establish that satisfactory agreement exists between measured and predicted values.

l (36) 14.2.12.5.'29.4 - List or reference the minimum acceptable valve capacities, and the bases for those values.

(37) 14.2.12.3.30.4 - List, reference or provide bases for the criteria to be used to demonstrate that calculated and actual plant conditions are in satisfactory agreement.

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. 640.11 (38) 14.2.12.3.31 - State what transients and trips will be (14.2.12)

Continued ^

Performed and analyzed to determine proper operation of control systems and reference or provide bases for the

. acceptance ranges ~ referred to in the Acceptance Criteria.

Indicate if Reactor Power Cutback System is tested.

(39) 14.2.12.3.32.4 - List or reference the source of design l

temperature conditions that will be compared with measurements made during the ventilation capability l

test.

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(40) 14.2.12.3.34.4 - List or reference the design criteria that will be used,when calculating whether the response of the t

plant to a total loss of coolant flow transient was safe.

(41) 14.2.12g.,37jnd 14.2.12.3.38 -

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l (a)

Identify the parameters to be monitoreJ for each test.

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(b) Provide bases for acceptance criteria for each test based on realistic beginning-of-life predictions.

(c) Provide the basis for and state the required degree of convergence of actual test results with predicted results for the monitored parameters for each trip.

(42) 14.2.12.3.39 - State what transients will be performed and analyzed as a part of this test, state the power level at which each of the transients will be conducted, and provide bases for quantitative acceptance criteria.

(43) 14.2.12.3.40.4.C - Expand the criteria explanation or reference-other material that will specify what constitutes an acceptable determination of the loose parts alert setpoint.

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.. 4 640.12 7.2.1.2.1 states that protective device delay times will be (14.2.12) obtained from calculations and tests.

Expand applicable Phase II o

or Phase III test descriptions to describe these tests.

640.13 We have noted on other plant startups that the capacities of (14.2.12) pressurizer or main steam power-operated relief valves (PORV's) are sometimes in excess of the values assumed in the accident analyses for inadvertent opening or failure of these valves.

Provide a description of the testing that demonstrates that the capacity of these valves is consistent with your accidenh analysis assumptions.

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l 640.14 The first prerequisite for Phase II and III tests requires

v.em n(,..14.2.12) clarification. Most test descriptions state. that "all" construction activities are complete, while a.much smaller -

number more specifically require that all construction

_etivities "on the systems to be tested" are complete.

An additional problem is that certain tests given in 14.2.12 include two systems (ex. RB Polar Crane and FHB Bridge Crane).

Tables 14.2-5 and 14.2-6 indicate that the Fuel Handling Bldg. Crane Test will start 8 months before the polar crane l

1s released).

Prerequisites seem to imply that you cannot test one system until all systems to be tested are ready.

l (1) Modify the prerequisite requirement concerning the status of construction activities for all Phase II and Phase III_

tests to clearly indicate which systems are referred to when determining that all construction activities are complete.

640.14 (2) Modify the descriptions for those tests involving more (14:2.12) -

Continued than one systems, where all systems are not available on the date the test is scheduled to commence, to clarify the prerequisite on completed construction activity.

(3) The Phase II test (14.2.12.2.3) on safety-related and non safety-related inverters states tha', diesel generator testing will be performed concurrently with this test. Modify Tables 14.2-5 and 14.2-6 to reflect this.

(Note:

the diesel generator is not scheduled to be released until three months after this test staFts, and the present test schedules are not concurrent).

.a.

640.15 Our review of ifcensee event reports has also disclosed thst many '-

(14.2) events have occurred becane of dirt, condense.d moisture, or other foreign objects inside instruments and electrical components (e.g., relays, switches, breakers).

Modify Table 14.2-3 and 14.2-4 to ensure that formal inspections will be performed during each test to p(avent component failuras such as these on your l

facility.

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640.16 Tables 14.2-5 and 14.2-6 are incomplete and not consistent with (14.2)

Section 14.2.

Provide the following:

(a) Modify Table 14.2-6 so that the test listings. correspond to those given in Tables 14.2-1 and 14.2-2.

(b) Modify the early start date for the Supplementary Chilled Water l

System to reflect that system 46F1 (Table 14.2-5, Supplementary Chilled Water-Chiller Bldg) will not be released by that date.

(c) Modify Table 14.2-5 to indicate scheduled release date for system #45-1.

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. _ _. _ _