ML19350B019

Forwards Response to 810112 Request for Info Re Emergency Feedwater Sys Design.Tech Specs Modified Nov 1977 to Conform W/Std Wording for Valve Testing for Eccs.Emergency Procedure 1202.26 Identifies Supply Requirements

ML19350B019
Person / Time
Site:	Arkansas Nuclear
Issue date:	03/12/1981
From:	Trimble D ARKANSAS POWER & LIGHT CO.
To:	Reid R Office of Nuclear Reactor Regulation
References
IR-0381-04, IR-381-4, NUDOCS 8103190121
Download: ML19350B019 (18)

Text

,

d ARKANSAS POWER & LIGHT COMPA hI__ / f POST OFRCE BOX 551 UTTLE ROCK. ARKANSAS 72203 ( 'p$1

~ '/

March 12, 1981 C

f@

k 9

,h d

IR-0381-04 s

5-g,.l([

Director of Nuclear Reactor Regulation "Q

ATTN: Mr. Robert W. Reid, Chief Operating Reactors Branch #4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C.

20555

Subject:

ArkansasNuclearObe-Unit 1 Docket No. 50-313 License No. DPR-313 Submittal of EW Design Information (File:

1510.3)

Gentlemen:

Your letter of January 12, 1981 requested AP&L to provide considerable information on the ANO-1 Emergency Feedwater (EFW) system. The items of interest were listed as Enclosures 1 and 2, as AP&L had already responded to Enclosure 3 on December 3, 1980 and has provided a schedule for the responses per our February 13, 1981 letter. Attached are our responses to Enclosures 1 and 2 of your January 12, 1981 letter in the order in which they were requested.

Very truly yours, d

Uk f/

David C. Trimble Manager, Licensing DCT:nak 8103190 h i

MEMBER MOOLE SOUTH UTtWTIES SYSTEM

i

{.

REQUEST FOR INFORMATION ARKANSAS NUr! EAR ONE, UNIT 1 AUXILIARY FEEDWAn.R SE TEM (AFWS) REQUIREMENTS

• iJCKET No. 50-313 Comment A.I.

Recommendation (GS-2) - The licensee should lock open single valves or multiple valves in series in the AFW system pump suction piping and lock open other single valves or multiple valves in series that could inter-rupt all AFW flow. Monthly inspections should be performed to verify that these valves are locked and in the open position. These inspections

. should be proposed for incorporation into the surveillance requirements of the plant Technical Specifications. See Recommendation GL-2 for the longer term resolution of this concern.

Response

Manual valves in the EFW system pump suction and others that could f

interrupt EFW flow are locked in their correct positions for EFW supply to the Steam Generators. A current status of all category 'E' valves is maintained in the control room. All changes from the required position must be approved by the shif t supervisor. Following completion of the -

task which changed the position from the required posirion, the valve is returted to the correct position and is verified locked in its required position ~by two individuals. Valves which are not locked in position, such as the EFW flow control valves, crossover valves and service water suction valves which may require operation during EFW supply or emer-gency actions are verified correctly positioned at least monthly.

Further, t e function and stroke time of valves which are or may be h

required to operate during EFW supply operations are verified operable quarterly per ASME Section XI requirements. A Technical Specification change submitted June 8, 1979 and modified in November 1979 required that:

Each EFW train sha7. De demonstrated operable:

at least once per 31 days on a staggered test basis by verifying that each valve (manual, power operated or automatic) in earb EFW flowpath that is'not locked, sealed, or otherwise secured i

>si-

+

tion, is in its correct position.

• Also referred to as Emergency Feedwater. System (EFW) 9

prior to exceeding 280F Reactor Coolant temperature and af ter any EFW system alignment alterations by verifying that each manual valve in each EFW flowpath which, if mispositioned may degrade EFW operation, is locked in its correct position.

At least once per 92 days on a staggered test basis by cycling each motor-operated valve in each flowpath through at least one complete cycle.

at least once per 18 months by functionally testing each EFW train and:

1.

Verifying that each automatic valve in each flowpath actuates auto =atically to its correct position on receipt of an actua-tion signal.

2.

Verifying that the automatic stean supply valves associated with the steam turbine driven EW pu=p actuate to their cor-rect positions upon receipt of an actuation signal.

This specification is consistent with the standardized Technical Speci-fication wording for valve testing for emergency core cooling systems.

Comment A.2.

Recommendation GS Emergency procedures for trinsferring to alternate sources of AFW supply should be available to the plant operators. These' a

procedures should include criteria to inform the operator when, and in what order,- the transfer to alternate water sources should take place.

The following cases should be covered by the procedures:

-The case in which the primary water supply is not initially available. The procedures for this case should include any op-erator actions required to protect the AFW system pumps against self-damage before water flow is initiated; and,

-The case in which the primary watdr supply is being depleted. The procedure for this case should provide for transfer to the alter-nate water sources prior to draining of the primary water supply.

Response

Emergency Procedure 1202.26, " Loss of Steam Generator Feed", covers the following EW supply requirements:

I.

-Loss of Normal Feedwater.

II. - Loss of Feed to ore OTSG.

III. Loss of.All Feedvater Including EFW.

There is a concensate storage tank low level alarm 2d low EFW suction line. pressure alans annunciator 'in the control room. Coupled with the Tecnical Specification level requirement on the Condensate Storage Tank there is adequate assurance that the'EFW pumps would not be started with

- inadequate supply on'the suction line.

Ii:the feedwater suctica k

k m'--

m.

m y

a.

1 from.the primary source (Condensate Storage Tank) is lost or depleted after initiation of EFW the procedure mentioned above instructs the operator to open the service water supply to the EFW suction upon re-f seip of either low suction pressure alarm.

Comment A.3.

Recommendation GS The as-built plant should be capable of providing the.rcquired AFW flow for at least two hours from one AFW pump train 4

independent of any alternating current (AC) power source. Your sub-mittal1of the proposed system design upgrade indicated that the AFWS t

will have the capability for automatic start with loss of all AC power, but'it is not clear whether the revised system will have the capability of operating for two hours without AC power.

If manual AFV system initiation or flow control is required following a complete loss of AC l

. power prior to implementing your proposed design revisions, emergency

-procedures ahoL3d be established manually initiating and controlling the system under these' conditions.

If the cooling water for the turbine-driven and motor-driven pumps lube oil coolers and pump room coolers is dependent' on alternating current power, design or procedural changes Jehould be made to eliminate this -dependency as soon as practicable.

Until this' is done,. the emergency procedures should provide for an individaal to be stationed at the turbine-driven pump in the event of s

the loss lof all AC pcuer to monitor pump bearing and/or lube oil tem-

[

peratures. If necessary, this operator would operate the turbine-driven

[

pump in an on-off mode until alternating current is restored. Adequate lighting powered by di

:ect current power sources and communciations at E

local stations should also be p'rovided if manual initiation' and control of:the AFW system is'needed.-

s s

Response-

-Emergency-Procedure i

1202.02, " Blackout-Loss of All Auxiliary Power

< Except Batteries", specifies action'is required for manually initiating l

Land: controlling the'EFW-system in the event of loss of all AC power.

l' The cooling waterifor both the turbine-driven and -motor driven pump. lube

oil cooling.is supplied from the pump suction supply water and thus are not_ dependent upon any auxiliary system supplied-by AC power. There are L no ' safety-related room coolers associated with EFW pumps on ANO-1.

l-Comment'A.4.

i-

. Recommendation GS-6 -'The.licenseeLshould confirm flow. path. availability of-an AFW: system flow train that has been.out of service to perform periodic: testing or maintenance as follows:

o b

-Procedures should be implemented.to Tequire an operator to de-L termine.!that the: AFW ' system ' valves are properly: aligned and a

- second operator. to independently yerify'that the valves are pro-y L

cperlyTaligned!

w _

M, 3

e v4 Y-

-w...w

--w.

s-

-a.,--

--9 re-W.'

-r

- w -

rwwr w

1

\\

~

-The licensee should propose Technical Specifications to assure that prior to plant startup following an extended cold shutdown, a flow test would be performed to verify the normal flow path from the primary A?7 system water source to the steam generators.

The flow

)

test should be conducted with AFW system valves in their normal alignment.

Response

ANO procedures presently require flow path availability verification following any testing or maintenance that has been performed upon the system which has affected availability. A log of all changes to the flow path is maintained and all valves which are out of their safeguards alignment are addressed during shift turnover between operating crews.

This is documented on our shift turnover checklist.

The periodic sur-

.veillance flow path has been chosen such that monthly tests to verify operatibility per ASME Section XI do not require loss of safeguards flow path alignment.

Amendment 50 to the ANO-1 Technical Specifications added the requirement to demonstrate EFW operability at least once per 18 months by function-ally testing each-EFW train and verifying that feedwater is delivered to each steam generator _using the electric motor-driven EFW pump.

Comment A.5.

Recommendation (Plant Specific) - The licensee's letter of December 30, 1979.. " Auxiliary Feedwater System Reliability Study" stated that the atmos?heric dump valves fail 50 percent open on loss of control signal.

The licensee should verify that control power will not be lost in the evrat of a loss of offsite power (LOOP) or in the event of a complete loss of AC power (LOAC).

If control power is not available for either of the above two events, then modifications should be made to prevent an uncontrolled cooldown in the event of a LOOP or a LOAC.

Response

Modifications have been made at ANO-1 to let ce the atmospheric dump isolation valves normally closed with safety grade (diesel backed) power to allow opening the isolation valves, if needed. This modification will allow operator action to enhance cooldown without the concern for uncontrolled cooldown upon loss of the control signal to the atmospheric dump valves.

Comment A.6.

Recommendation The licenseef should assure that there are no temporary

- strainers in place in the AFW flow path that may cause flow blockage if plugged. Operating experience at'several plants.has shown this to be a potential common cause failure mechanism which could fail the entire AFWS.

.The' suction _ strainers between the condensate storage tank and the pumps are an example.

Response

The ANO-1 EFW system is designed to contain a strainer on the suction of each EFW pump between the condensate storage tank or Service Water Header, whichever supplies suction. A single failure of one of these strainers thus could not cause failure of the entire EFW system. How-ever, these strainers have been removed several years ago, as they were installed for startup testing purposes.

Comment-B.I.

For the short term we require the licensee to provide redundant low-level indication and alarm indication and alarm in the control room for the condensate storage tank. The low level alarm setpoint should allow at east 20 minutes for operator action, assuming that the largest capa-i seity AFW pump is running. The level indications should be redundant all

.the way from the detectors to the readouts and alarms inside the control room.

Power supplies for the indication and alarm should also be re-dundant. Use of non-Class IE circuitry is acceptable provided one power train has a back-up battery source.

i

Response

The present annunciator on the Condensate Storage Tank (CST) level to

'EFW pump suction' includes a low suction pressure alarm and a low CST level alarm. The low suction pressure alarm would annunciate in the control room when the level in the CST at maximum EFW flow would allow greater than two hours before the water supply would run out.

The level alarm would alarm in the control room when, at maximum EFW flow, the operator would.also have approximately two. hours to react before the water supply would be depleted.

The EFW suction pressut-alarm is Class IE.

ANO-1 annunciators in the Control Room are backed-up by DC

power.

Comment B.2.a.

JANO-1 has performed endurance tests of'their AFW pumps. The following requests for additional information has resulted from our review of the

preliminary submittal of.their test results.

a.

Explain why the motor bearing temperatures 'were not measured during

testing of Unit No. P-7B.

Response

The pump bearing oil. temperature instead of the pump-bearing temperature

.was measured. This was selected due to (1) the.ulative ease of obtain-ting;an accurate measurement in an oil bath rather than a contact read-

.ing, (2) the fact that' the pump' manufacturer.specified limits.upon the

~

. oil temperature versus the bearing temperature and (3) the fact that pump performance testing criteria-in AP&L's ASME Section-XI-program is-based upon bearing oil ~ temperature.

• e-v

-e-

+,

,---m

.e

-v

-v 9

During testing upon other pumps which are nearly identical on ANO-2 (same manufacturer, function is HPSI vs. EFW) conducted in November, 1980 bearing race temperatures were measured concurrently with bearing oil temperature. The bearing race temperature was found to be approxi-mately 5*F higher than the oil temperature in all cases.

Comment B.2.b.

b.

Provide the maximum allowable value of bearing reservoir oil tem-perature recommended or specified by the pumps / drivers manufac-turers, with and without external cooling.

Response

The pump manufacturer (Ingersol-Rand) has conservatively set a recom-mended bearing oil temperature limit at 160*F.

This limit is based on the assumption that cooling water is being supplied to the bearing housings.

For bearings without cooling water, higher temperatures would be expected during operation and a higher limit than 160*F is reasonable in this case. Data from bearing manufacturers and industrial experience has shown that bearing temperatures in the range of 180*F - 200*F are not detrimental to pump operation. We have chosen to use 180*F as an upper bearing temperature limit.

~ Comment B.2.c.

c.

Compare the data obtained from the subject endurance tests to similar data obtained during periodic, shorter-term testing of the units. Provide results of future tests (e.g., during the next refueling) to NRC when available.

Response

Data from short-term testing of the EFW pumps such as during the peri-odic ASME Secton XI tests has indicated that bearing temperatures ap-proach stability in these one_ hour tests and as such are adequate for trending pump performance. However, longer periods approaching six hours are required before observed bearing temperatures cease to in-crease.

Otherwise, no differences in test results between long-term endurance test and short-term tests have been observed.

No plans-exist for future long-term endurance tests for the ANO-1 EFW pumps. The only tests which are routinely performed are related to measuring pump performance curves at various head and flow conditions.

Comment B.2.d. & e.

d.

Provide the supplier's conclusions with regard to the tests, in particular with regard to bearing temperature.

State the source of pump-bearing external cooling and whether it is e.

normally in use.

. _ = -

.._ _ _ _. _ =. _ _ _

i Y

L:

Response

t No formal review of the subject test results has been conducted by 4

Ingersol-Rand. However, informal. discussions with their personnel l

- indicated that they believed it desirable to add forced cooling to the

- bearing oil housings to extend pump bearing lifetime. We have elected to' add a source of bearing cooling water. This water is common with the pump suction fluid and as such is continuously in use whenever the EFW

~

-pump _is in service.

~

. Comment B.3.

If the licensee's plant requires local realignment of val 'es to conduct

periodic tests on one AFW system train, Technical Specifications should l'

' be proposed that provide's dedicateJ individual who is in communication i

with the control room be stationed at the manual valves in the AFW I

p system train in the test mode to shift the AFW system train from the stest mode to its operational alignment if necessary.

Response Plant procedures for periodic te'sts on EW system trains do not require c

local realignment of valves, nor'do the-tests cause a loss of EFW flow-h t

path to the steam generators due to selection of test flow path. There-l fore, Technical Specification. changes are not necessary.

~

.:1 Comment-C.I.-

y,-.

~

iRecommendation GL-2:- Licensees with plant designs in which all (primary-and-alternate) water supplies t'o the A W systems pass through valves in I-na single flow path:should install-redundant parallel flow paths-(piping c

and valves).

p f

Licensees with plant designs in whichithe primary AW system water 1

. supply passes-through valve's in.a single flow path, but the alternate U

L AFW system water supplies connect to the AW system pump suction piping downstream of the above valve (s), should install redundant velves par-

~

~ allel to the above. valve (s) or provide automatic opening of the valve (s) w

{

cfrom.the alternate water; supply upon low pump suction pressure.

lThellicensee should propose Technical ~ Specifications to incorporate-

~

appropriate periodic inspections to-verify the valve positions'into the l

surveillance requirements.

l

. The prima'ry AW ' system water supply passes through single' series ' valves j

i in the suction flow path but-the alternate water supply :(Service Water).

?conne' cts to the pump: suction' piping doOnstream of these' valves. The y'

flicensee1should make~one'of_the following modifications:

a

.a)f. install redundant valves parallel.to the existing valves (note:f' cdisabling the_ suction valves byl removing the valve internals will-a

;be

considered an acceptable substitute),-

p j

g r

fJ

[_1 'z

.D. d

^

r

1. ~

~

, w j

~

l8 :.

s[

2.

.;~,_.

L :,.,. _

s_.

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

4 b) provide automatic opening of the valves form the alternate water l

2 supply upon low pump suction pressure (Note: if this modification is used, the licensee must demonstrate that the control system and valve response time is adequate to protect the pumps from the 4

- effects of suction flow termination),

[

c)_

provide redundant safety grade indication and. alarm for the exist-ing valves combined with either:

1) automatic trip on low suction pressure (Note: if this mod-ification' is used, the licensee must demonstrate that the response time of the trip is adequate to protect the pump), or l

2) operating procedures to place the pumps in manual start mode u

- if valve (s) is not open and Tech Specs to make the reactor suberitical within one~ hour and place the facility in a shut-down cooling mode which does not rely on steam generators for cooling within twelve hours, or at the maximum safe shutdown rate', if a. water supply is not available (

Reference:

NUREG-

[,;

0645~, Vol. II IE.Julletin 79-05A).

-The licensee should also. propose Tech' Specs to' incorporate appropriate

periodic inspections to verify valve positions into the surveillance requirements.

t

Response' I

f.

AP&L has modified the EFW suction line on ANO-1 to ensure redundant fparallel flow paths. This is true for both the primary (Condensate 1 Storage. Tank) and alternate (Service Water) supply paths to each EFW -

pumpfsuction. LAs stated in.the response to Comment A.I.,.there are

Technical Specification requirements-in place for verifying valve po-sitions.

- Per AP&L's ?long-tern EFW upgrade, program ~ we are evaluating. modification

of;the EFW suction flow to provide automatic opening of the valves from

' the ? Service Water supply upon low pump' suction pressure. The final

~

. design.and: Technical Specification submittals will be made according to

the schedule provided in our February 19, 1981' letter;on this topic.

f }-

~

I

(

t

'l F

f -

s L

6 y

_k^

{ d,'

[~[ -

i s

vu -i;.

~

" ;? -

Y e

s Ih'

[_ k. -

I

.g j;

+

~

w:

= a.. -.... -.

t

Comment C.2.

Recommendation GL At least one AFW system pump and its associated flow path and essential instrumentation should automatically initiate flow and be capable of being operated independently of any AC power source for at least two hours.

Conversion of DC power to AC power is acceptable.

(See item A.3)

Response

Reference to our response to Comment A.3 will adequately respond to this item concerning the independent operation of at least one EFW pump on loss of all AC power for at least two hours. AP&L's long term EFW up-grade will include DC operated valves such that manual operator action will not be required after implementation of that upgrade.

Comment C.3.

Recommendation GL Since the ANO-1 plant has an unprotected tormal AFW system water supply, the' licensee should evaluate the design of

the AFW system to determine if automatic protection of the pumps is necessary following a seismic event or a tornado. The time available before' pump damage, the alarms and indications available to the control

. room operator, and the time necessary for assessing the problem and taking action should be considered in determining whether operator action'can be relied on to prevent pump damage.

Consideration should' be given to providing pump protection by means such as automatic switch-over of the pump suctions to the alternate safety-grade source of water, automatic pump trips on low suction pressure or upgrading the normal source of water to meet seismic Category I and tornado protection re-quirements.

(For automatic switchover or pump trip, see the stipulations

' on response time in item C.I.)

~

Response

AP&L is evaluating a design to automatically switchover the EFW supply primary water source to the alternate source as discussed in our response l..

l to Comment C.I.

This response will adequately address this item as well.

l

' Comment C.4 Recommendation - Evaluate the ANO-1 AFWS design with regards to appli-cable high energy pipe break criteria in Branch Technical Positions ASB 10-1, ASB 3-l 'and MEB ~ 3-1 including assumption of a concurrent single l

active failure. Provide the effects of pipe whip and jet impingements, I

as well as the environmental effects of postulated pipe failures. The l

latter should include the resultant temperature, pressure and humidity.

LThe results of this analysis should be compared with the environmental-l design criteria of vital AFWS electrical components. The effects of a main steam or fer water line failure on the capability of the AFWS to s

[

provide safe shatdown should be included in this analysis.

r

The licensee should evaluate the postulated pipe breaks stated at;ve

-and (1) determine any AFW system design changes or procedures r ecessary to detect and isolate the breaks and direct the required feed,ater flow to -the intact steam generator (s) before they boil dry or (2) describe how the plant can be brought to a safe shutdown condition by the use of other system., which would be available following such postulated events.

Response

The steam supply piping for the turbine driven EFW pump up to the steam admission valves meets the criteria for high energy piping.

No vital-EFW system components are located in the room contain '

ing the proposed high energy piping except those related to steam supply for the turbine driven EFW pump. A postulated break in this piping can be isolated by closing the motor operated valves in the lines from each steam generator. Once the break is isolated,

-the plant can be brought to a safe shutdown condition, even assum-ing a concurrent failure of the motor driven EFW pump, by using the normal feedwater system (Auxiliary Feedwater) or the high pres-sure injection system.

t' F

l 1-T s

r

-4

REQUEST FOR INFORMATION BASIS FOR AUXILIARY FEEDWATER SYSTEM FLOW REQUIREMENTS ANO-1 Comment 1.a.

Identify the plant transient and accident conditions considered in establishing AFWS flow requirements, including the following events:

1) Loss of Main Feed (LMFW) 2)' LMFW w/ loss of offsite AC power

3) LMFW w/ loss of onsite and offsite AC power

4) Plant cooldown

5) Turbine trip with and without bypass

6) Main. steam isolation valve closure

7) Main feed line break

8) Main steam line break

9) Small break LOCA

10) Other transient or accident conditions not listed above

Response

The original design of the Emergency Feedwater System established a re-quirement for a minimum flow sufficient to remove heat load equal to 3 % full power operation. This is reported in Chapter 10 of the FSAR.

l This_ flow rate was not based on any specific transient. However, where appropriate, this flow rate is used as part of the transient analysis j

.for the accidents considered in the~FSAR. Table 1 below contains a i

list. of those transients considered in the FSAR along with the accep-tance criteria.

As part of the-EFW upgrade, the requirements for.the minimum flow were

~

re-examined.

A value of 500 gpm was established as a conservative min-imum flow for the EFW System taking into account a single failure and allowing for pump recirculation flow, seal leakage and pump wear..This flow was then used to analyze the accident with the maximum heat load.

This accident is a loss of main feedwater from 102% full power. Reactor coolant pump heat was included and 1.2 times ANS 5.1 decay heat was assumed. ' No credit was taken 'for the anticipatory reactor trip. The reactor was tripped on high RC ' pressure. The primary acceptance criteria

~

were met for this analysis. These criteria are:

RCS Pressure < 110% of design DNBR > Applicable correlation limit Doses ( 10 CFR 100 Even though these primary criteria were met, the pressurizer did go solid for a substantial length of time. For this reason, it was de-cided that an additional reactor trip function should be added to the system. The reactor trip function selected is the powerffeedwater flow trip. This function will also initiate EFW. The power /feedwater flow trip will detect both loss of feedwate,r events caused by a main feed pump trip and those caused by valve closure or feed pump run-back without trip. For this reason, the trip function is taken credit for in an accident analysis. With the addition of this earl-ier reactor trip, the pressurizer will not go solid even for the con-servative conditions used in the analysis.

Accidents 1, 2 and 3 of Table 1 also specifically require AFW for mit-igation. The impact of a 500 gpm EFW flow on the results of these an-alyses has also been reviewed.

It has been determined that the results reported in the FSAR would not change with a 500 gpm EFW flow. The other accidents listed in Table 1 do not require EFW for mitigation although the availability of the EFWS is assumed.

)

The other events listed in the question but not included in Table 1 are discussed below:

LMFW with Loss of Onsite and Offsite AC Power - This event was not a design basis for the plant and subsequently is not included in Chapter 14 of the FSAR. Although a specific analysis of the event is not in-cluded, the upgraded EFW System will be designed to supply at least 500 gpm even with the loss of both onsite and offsite AC power.

Plant Cooldow.- - Plant cooldown with EFW is a controlled event with decay heat lesels equal to or lower than the loss of feedwater event idratified as the design basis event. The design basis event bounds this case for EFW flow required.

Turbine Tr.p With and Without Bypass - This. event does not affect the EFWS unless MFW fails.

In which case, the loss of MFW event previously.

addressed would bound the EFWS design.

' Hain Steam Isolation Valve Closure - Again, this event does not directly affect the EWS unless MIN is lost as discussed above.

Main Feedline Break - This event was not a design basis for this plant and is not included in the FSAR Section 14. Main feedline break is a more abrupt case of LOFW and the results of an analysis are expected

' to yield approximately the same requirements for EFW flow.

e

Small Break LOCA - The EW criteria assumed for this event are described in Topical Repcrt BAW-10052 updated by letter report, J.H. Taylor (B&W) to S.A. Varga (NRC), 7/18/78 and the B&W report entitled " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 FA Plant", 5/07/79. These documents show that an EW flow of 500 gpa for-ANO-1 will not lead to the violation of the acceptance criteria.

w S

I s

't g

k t

(

E

TABLE 1 Accident Description _

FSAR Section Acceptance Criteria

1) Loss of Electric Power 14.1.2.8 A,B,D

2) Main Steamline Failure 14.2.2.1 D

3) Loss of Coolant Flow 14.1.2.6 A, B

4) Startup Accident 14.1.2.2 A, B

5) Rod Withdrawal Accident 14.1.2.3 A, B at rated power operation

6) Moderator Dilution Accident 14.1.2.4 A, B

-7) Cold Water Accident 14.1.2.5 A, B f

8) Stuck Out, Stuck In, or 14.1.2.7 A, B Dropped Control Rod p

Accident

9) ' Steam Generator -Tube 14.2.2.2 B, D Failure

10) Rod Ejection Accident 14.2.2.4 C, D

11) Loss of Coolant Accident 14.2.2.5 D, E Jggc Acceptance Criterion Technical Basis A

RCS Pressure- <110% of Design ASME code i

B DNBR > Applicable Correlation Limit SRP 4.4 L

C-280 Cal./ Gram Fuel Limit RG 1.77 D

Doses (10 CFR 100 10 CFR 100-E-

Fuel Cladding < 22000F 10 CFR 50.46

i TABLE 1 Accident Description FSAR Section Acceptance Criteria

1). Loss of Electric Power 14.1.2.8 A,B,D

2) Main Steamline Failure 14.2.2.1 D

3) Loss of Coolant Flow 14.1.2.6 A, B

4) Startup Accident 14.1.2.2 A, B

5) Rod Withdrawal Accident 14.1.2.3 A, B at rated power operation

-6) Moderator Dilution Accident 14.1.2.4 A, B

7) Cold Water Accident

-14.1.2.5 A, B

8) Stuck Out, Stuck In, or 14.1.2.7 A, B Dropped Control Rod Accident-

'9) Steam Generator Tube 14.2.2.2 B, D

-Failure

10) Rod Ejection' Accident' 14.2.2.4 C, D

-11) Loss of Coolant Accident 14.2.2.5 D, E Kjgt -

Acceptance Criterion Technical Basis A

RCS Pressure. 110% of Design ASME code B

DNBR Applicable Correlation Limit SRP 4.4 C

280 Cal./ Gram Fuel Limit RG 1.77

'D Doses 10 CFR 100 10 CFR 100 E

Fuel Cladding 22000 F 10 CFR 50.46 g

Comment 1.b Describe the plant protection acceptance criteria and corresponding technical bases used for each initiating event identified above.

The acceptance criteria should address plant limits such as:

- Maximum RCS pressure (PORV or safety valve actuation)

- Fuel temperature or damage limits (DNB, PCT, maximum fuel central temperature)

- RCS cooling rate limit te avoid excessive coolant shrinkage

- Minimum steam generator level to assure sufficient steam generator heat transfer surface to remove decay heat and/or cool down the primary system.

Response

The design basis event for sizing the EFWS is the Loss of Feedwater Event discussed in the response to Comment 1.a.

The acceptance cri-teria for the other transients which assume the availability of EFW are also given in the response to Comment 1.a. (Table 1).

The acceptance criteria for these accidents include RCS pressure limits, fuel limits and offsite dose limits. The RCS cooling rate is not an acceptance criterion for accident analyses.

An overcooling event that drains the pressurizer is not desirable, however, it does not violate any of the accident analysis acceptance criteria.

Maintaining a minimum steam generator level is also not an acceptance criterion for accidert analyses.

It is desirable that the reactor is tripped and EFW initiated prior to steam generator dryout, but this is not required in order to obtain acceptable results. After EFW has been initiated, the high injection point in the steam generator reduces system dependence on a specific level for adequate heat transfer. The steam generator level control is set 16w for decay heat removal with forced circulation and high for natural circulation.

The level is also set high for small break LOCA events.

Comment 2.

Describe the analyses and assumptions and corresponding technical justification used with plant condition considered in 1.a. above.

Response-

.As discussed in response to Comment 1.a., the design basis event which verifies the EFWS design flowrate is loss of main feedwater.

The analysis assumptions for this event are listed below. Corres-ponding technical justification where not specifically listed, is

based on licensing requirements and prudent engineering judgement at the time of the analysis.

The information is not provided for the other events identified in Comments 1.a. and Table 1 because the LOFW event is the most limiting.

a)

Maximum reactor power (including instrument error allowance) at the time of the initiating transient or accident.

- 102% full power; this includes an allowance for 2% power level measurement uncertainity.

b)

Time delay from the initiating event tc reactor trip.

- The reactor will trip on the power /feedwater flow trip approx-imately 4 to 5 seconds after the LOFW event.

c)

Plant parameter (s) which initiates EFWS flow and time delay between initiating event and introduction of EFWS flow into steam generator.

- The EFWS will be initiated by the power /feedwater flow trip.

It is assumed that the time delay between receiving the initiate signal and full EFW flow to the steam generators is 25 seconds for a case without the loss of offsite power. This is a total delay of approximately 30 seconds from the LOFW event.

d)

Minimum steam generator water level when initiating event occurs.

- Steam generator inventory rather than water level is used as an input to this analysis, e)

Initial steam generator water inventory and depletion rate before and after EFWS flow commences - identify reactor decay heat rate used.

- The initial steam generator inventory is dependent on power level. For this case, a liquid inventory of 33,639 lbm per steam generator was used. The depletion rate of the inventory before initiation of EFW averages about 720 lbm/sec in each steam generator. After initiation of EFW, the depletion rate averages about 145 lbm/see until the liquid inventory is essen-l tially depleted. From this point, the entire EFW flow is vapor-l ized until_ decay heat plus RC pump heat drops below the capablil-ity of the EFW System. At.that time, steam generator inventory begins increasing again. The decay heat used in this calculation was 1.2 times ANS 5.1 decay heat.

f)

Maximum pressure at which steam is released from the steam generator (s) and against which the EFW pump must develop sufficient head.

- The peak steam pressure occurs shortly after the ie:siating event and is about 1100 psig. Soon after the EFW is initieted, however, the steam pressure is controlled by the first bank of steam safety valves to a pressure of_about 1050 psig.

e

g)

Minimum number of steam generators that must receive EFW flow.

- This analysis was run assuming both steam generators were available.

The heat load can be removed with one OTSG and it is expected that the results would be approximately the same.

h)

RC flow condition - continued use of RC pumps or natural circula-tion.

- Continued operation of RC pumps was used for this analysis.

i)

Maximum EFW inlet temperature.

- An inlet temperature of 120 F was used.

j)

Following a postulated steam or feedline break, time delay assumed to isolate break and direct EFW flow to intact steam generator (s).

EFW pump flow capacity allowance to accommodate the time delay and maintain minimum steam generator water level. Also, identify credit for primary system heat removal due to blowdown.

- FSAR Section 14.2.2.1 contains the details of the assumptions used in the main steamline break analysis. Main feedwater line breaks have not been analysed for ANO-1.

k)

Volume and maximum temperature of water in main feedlines between steam generator (s) and EFWS connection to main feedline.

- There are no piping connections between the EFWS and the MFW System.

1)

Operating condition of steam generator normal blowdown following initiating event.

- The OTSG's do not have a blowdown system.

l m)

Primary and secondary system water and metal sensible heat used for cooldown and EFW flow sizing.

- As stated in our response to Comment 1.a for Enclosure 2 the EFW flow sizing was based on a decay heat load of 3\\% full power operation.

No cooldown rate restrictions were used as stated in i

our response to Comment 1.b. for Enclosure 2.

These two responses contain more detail that will adequately respond to the above.

l_

n)

Time at hot standby and time to cooldown RCS to RHR system cut-in temperature to size AFW water source inventory.

- The analysis did not consider a time at hot standby as outlined

'in our response to Comment 2.c.

Per the ANO-1 Technical Specif-l ication basis there is adequate storage in the CST for 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of EFW operation at 390 gpm. The decay heat removal system is

s.

expected to be in operation within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

If the CST were unavailable the Service Water can supply almost un-limited amounts of EFW to the OTSG's.

Comment 3.

Verify that the AFW pumps in your plant will supply the necessary flow to the steam generator (s) as determined by items 1 and 2 above considering a single failure.

Identify the margin in sizing the pump flow to allow for pump recirculation flow, seal leakage and pump wear.

The response should include the following:

(1) List of all events needing AFW to mitigate the consequences.

(2) Justification that the bounding non-LOCA calculation will serve as a conservative basis for sizing the AFW system for non-LOCA core cooling considerations.

In other words, show that the cal-culation will bound all of the non-LOCA events requiring AFW.

(3) The non-LOCA analysis should include a loss of feedwater event using FSAR type assumptions to maximize heat removal requirements (1.2 ANS decay heat, 2% power level measurement uncertainty, RCP heat input).

The calculation should not take credit for "anticipa-tory reactor trip" since it will not occur under all conditions.

Lifting of the PORV is not precluded; however, credit for pressure relief through the valve should not be assumed.

(4) For the small LOCA events, reference may be made to the B&W Report,

" Evaluation of Transient Behavior and Small Reactor Coolant Syst em Breaks in the 177 Fuel Assembly Plants dated May 7,1979.

The acceptance criteria for the event will be:

1.

Reactor Coolant System pressure remains less than 110% of design pressure (2750 psig).

2.

No fuel failure (DNBR > 1.30).

Response

(1) The response to Comment 1.a resolves this item.

(2) The response to Comment 2 resolves this item.

(3) The response to Comment 1.a. resolves this item.

(4) The response to Comment 1.a resolves this item.