1CAN078705, Forwards Addl Info Re Items Discussed in Util Response to Insp Rept Re Safety Sys Functional Insp Conducted at Plant During Jan 1986

Forwards Addl Info Re Items Discussed in Util Response to Insp Rept Re Safety Sys Functional Insp Conducted at Plant During Jan 1986

ML20236Q565
Person / Time
Site:	Arkansas Nuclear
Issue date:	07/30/1987
From:	James M. Levine ARKANSAS POWER & LIGHT CO.
To:	Gagliardo J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
Shared Package
ML20236Q558	List: 1CAN078705, Forwards Addl Info Re Items Discussed in Util Response to Insp Rept Re Safety Sys Functional Insp Conducted at Plant During Jan 1986 IR 05000313/1986001
References
TASK-2.E.1.2, TASK-TM 1CAN078705, 1CAN78705, NUDOCS 8711200131
Download: ML20236Q565 (13)

Text

_

\

% i ARKANSAS POWER & LIGHT COMPANY b Le ej ,[

July 30, 1987

1. G -1 1987 I l

1CAN078705 Mr. J. E. Gagliardo, Branch Chief Reactor Projects Branch U. S. Nuclear Regulatory Commission j Region IV 611 Ryan Plaza Drive, Suite 1000 l Arlington, TX 76011

SUBJECT:

Arkansas Nuclear One - Unit 1 I Docket No. 50-313 i License No. DPR-51 Additional Information Relating to SSFI Inspection Report

Dear Mr. Gagliardo:

By letter dated June 24, 1987 (ICNA068705), you requested additional information concerning several items discussed in our response to the inspection report related to the Safety System Functional Inspection (SSFI) conducted at ANO-1 during January,1986. Based on the dialogue between our staff and NRC inspectors, we trust that the subject matters have been thoroughly discussed and that AP&L's general positions on the items have been previously communicated. We understand, however, that several specific matters related to the remaining issues require more detailed information and documentation to assist in your evaluation and closure. The items identified in your request for information are thdrefore addressed individually in the attachment to this letter.

Sincerely,

. M. Lev ne '

$[k#$ob!k$Ecoo313 Executive' Director G PDR Nuclear Operations JML:DRH:djm attachment h~ usuaea uioote sours uTitmes sysreu

v. ,, Attachment to ICAN078795 July 30,-1987 Page 1 Paragraph II.A.1.a:

Your response did not address the impact of a main steam line break on the safety-related components added during.the EFW-upgrade, iecluding a valve located approximately^4 feet closer-to a main steam line.than was. assumed in the FSAR-analysis,.nor did your.. response ~ include consideration.for steam jet.

Impingement on the main steam isolation system instrument,

. lines. Please submit a. summary:of your analysis which demonstrates that. safe shutdown can be achieved, including the-above considerations.

Response

During the upgrade of'the EFW system, efforts were made to .

n maintain the system within the bounds of original HELB analysis whenever possible. ~Both the valve and main steam isolation-

~

system (MSIS) instrument lines addressed above were located with this. goal in mind.and are considered to be in optimum locations given the constraints imposed by the existing plant configuration. The results of an evaluation of the acceptability of the MSIS instrument line locations have been previously incorporated into the FSAR. 'During.an inspection conducted in April of this year, however, it was discovered thatthroughanerroronourpartthe-subjectvalvewasnot properly evaluated and thus 'not included 1 n. the FSAR ~ update.

This evaluation has been revised to address this valve and is a presented below. The FSAR will be revised accordingly at the j

.next annual update.

HELB ANALYSIS

SUMMARY

.EFW SYSTEM UPGRADE-Recent re-review efforts relative to the Emergency: ,

Feedwater System upgrades have raised several questions regarding the completeness of the design changeLpackages involved. In particular, several concerns regarding the  !

required High Energy Line Break (HELB) analyses ha've been raised. These in' turn have led to some questioning of.the original analyses as refl ANO-1FSAR(sectionA.7)/gctedinthechapter14ofthe

.In order.to provide a.more-completely documented bases for the analyses involved, the 1 following evaluation is provided.

In performing the re-review of the EFW system up' grade, it-became apparent that ~although the effects of high energy line breaks were considered, the methods applied were in some cases inadequately documented. .It is-the intent of ,

this evaluation to systematically address each concern. '

The major changes reflected in this summary are'. listed below followed by an individual detailed assessment of' each.

i i

'f i

m 3 -.

,,' , Attachment to ICAN078705 i

' July 30, 1987 Page 2 1

. i' A. HELB' analyses.'of the 4-inch steam supply piping to the EFW turbine between the AC-operated branch line isolation valves and the DC-operated common line steam admission valves.

B. Re-consideration of the previously identified j high energy line breaks relative to the new 1 targets related to the EFW upgrade efforts.

C. Re-evaluation of the 8" line to the atmospheric- i dump valves.

A. HELB analysis of the 4-inch steam supply line to the ,

EFW turbine .i i

Section A.7.2 of the chapter 14 safety analyses previously analyzed the portion of this piping from j the steam lines to the AC-operated isolation valves in i each branch line. It is not intended to re perform l the previously approved analyses; however, the l emergency feedwater system upgrade resulted in a j change to the HELB boundary which required further i analysis, j Specifically the AC-operated valves were changed from normally closed to normally open, and normally closed i DC-operated steam admission valves were installed downstream forming the new HELB boundary. In response-to specific NRC questions regarding the required HELB considerations (reference 1), AP&L provided a summary of the conclusions drawn (reference 2). AP&L's approach was subsequently accepted by;the NRC in their safety evaluations of the EFW upgrade submittals.

(references 3 and 4). In re-reviewing those conclusions, it was considered appro)riate to re-examine the HELB requirements as applica)le in.this case.

_A re-assessment of the specific cr'!teria applicable to ANO-1 was made and piping analyses were performed to-identify the required: locations at which breaks must be postulated. Walkdowns were performed to' identify the potential targets from pipe whip or jet impingement.

i Next, careful considerations of safety functions were l

made in assessing the breaks and their consequences.

As a result,it was confirmed that the consequences I

were acceptable in each case based on the approach documented in reference 2 and' summarized below.

Several breaks in the 4-inch piping were identified '

i which would result in the loss of'one' train of EFW;

, however, the redundant train is available to provide EFW if needed. These 4-inch line breaks do not require protective system actuation, that is, they do

,o -- - - -

a

- - ~-

of , ,

' T 3-

+ &m Attachmentito ICAN878795f ,

' July 30, 1987 ,

.Page 3 not result' in aTreactor trip 'since the ' break is' within the capacity of the mainLfeedwater pumps. This'was? f' l confirmed by B&W (reference'5) and is the same position' >0 l taken by AP&L for the ANO-2 4-inch line breakeand approved by:NRC'(reference 6). .This1 approach is also consistent with the~ applicable HELB criteria for ANO-1 "

'(reference:7 item 11),; and is further substantiated by-ANSI /ANSJstandard 58.9l(reference'8). Finally, it.

should be noted that single failure consequences'(lossi

'of the' motor driven EFW pump) were considered and found acceptable based on the availability of-the .

Auxiliary Feedwater;and High Pressure! Injection.

systems to achieve' safe shutdown.(reference 2, 3,L4).:

Other breaks postulated in the 4-inch' piping could-

_. . conceivably result:in reactor trip.and initiation of e emergency feedwater. Specifically, certain breaks either tiy pipe whip or jet impingement could cause ' ' '

severence of two'MSIS: bundles (containing pressure; '

sensinglines). Failure of:two bundles could cause Main Steam Isolation. Valve closure on both' steam lines ,

y

.and initiation'of emergency feedwater to both steam.

generators'(by opening of steam; supply to'the turbine,.

opening of EFW supply valves, and startup of the motor driven purp. Loss of: safety- function'(i.e. supply of-EFW) does not occur; however, the capability of the-EFIC upgrade described b i" Feed Only Good Generator" (F0GG) would not'be assured. ;The.F0GG capability is a positive design. feature of.the'ANO-1 Emergency Feedwater. system as upgraded;;however, itmistnot-required by the. applicable > regulations (i.e. NUREG . ,

0737, Item I.E.1.1). Furthermore, it should be noted' that a 4-inch line' break would notirequire F0GG capability for an extended period of time since steam generator pressures'would.not immediately reach 600 psig. In any case, the operator could take actions to-isolate the.4-inch.line break (if downstream of the AC-operated valves) or isolate feedwaterito the affected generator.if the break is' upstreas off the -

L valves on one'of the' branch lines. Since the failure-i mechanism postuluted~does,not' result-in-loss'of safety. .

function (i.e.~ emergency feedwater) and single. failure -

l consequences are acceptable as documented by NRC in.

references 3 and 4, the current configuration 'of the ,'

4-inch lines is" acceptable.- o B. Previously Identified HELB's~ Relative to New Targets Installed as Part of the EFW Upgrade. '

The ANO-1 FSAR original HELB analyses specifically addressed the following high energy lines

1. Main Steam (MS) .

2. MS to Emergency Feedwater Pump Turbine Driver .

3. Main Feedwater

g; m

'E i d!r . Attadhir.ent to ICAN078705 4' July 30,.1987-m 4

Page 4

4. Reactor Coolant Letd an Line

5. Steam Generator and Pressurizer Sample Lines

6. Decay Heat Removal Each of these lines.was reconsidetOf.with respect to the new targets installed as part of.the EFW Upgradu.

Items 3 through 6 are not af fected by the EFIC upgrac'es. The main feedwater line analysis involves  ;

only a critical crack analysis. Since the lines are located in room 77 (upper south piping penetration room) and no EFW vpgrade modifications were made in

'this area, no further efforts were required. The f

letdown line break and.its effects are limited to areas which do 'not ?af fect the EFW system or the modifications thereto. The Steam Generator and Pressurizer Sample liv.s are routed in the same area l

as the letdown lines and.do not affect the EFW upgrades.

Furthermore, EFW is not required to mitigate the consequences of the Steam Generator and Pressurizer sample line breaks or the letdown line break. The Decay Heat Removal line HELB was excluded from further consideration in the FSAR for reasons unrelated to the EFW system or the recent upgrades.

The previous HELB analyses for Case 2 were addressed in A above. The general conclusions drawn in the original FSAR analyses are not invalidated by the newer considerations; hower, significant additional l

information is new pertinent and specific details were inaccurate. Therefore, the 1SAR was revised via amendment number 4 submitted in July,1986 to l

reflect the necessary changes.

The final case involves the Main Steam line breaks.

As was the case for the 4-inch lines, the 36-inch main steam lines are located in the vicinity of safety-related targets related to the Emergency Feedwater System

_(i.e. the steam supply piping to the EFW turbine) and certain EFW upgrades.

as described in the.fSAR, In thespecific original HELB were ruptures analysis postulated, the dynamic effects evaluated, and protective .

measures taken in some cases. Concerning the EFW upgrades, specific additional tar ets have been considered. The targets pertaini..g to the new DC-operated steam inlet valves (including their cabling) are within the direct effects of main steam liaa critical cracks (i.e., less than 20 feet remond). Therefore, it is conceivable that a main -,

staam line critical crack could render the DC operated steam admission valves inoperable as a result of steam impingement upon the valves and/or their associated cabling. Therefore, the steam driven turbine would be considered unavailable per this

'i sc .: Attachment to.1CAN978795 '

. July 30, 1987 ,

Page.5-

. i scenario; however, this is. considered acceptable- .

based on the following factors. First, the mass and energy release from a main steam line critical crack; is insufficient.to cause a plant trip; therefore, main feedwater would still be available to reach safe shutdown' conditions for~the same reasons. described-for the 4 inch line breaks in section AL above.

It-is conceivable that a maint steam:line critical '

crack could fail the'DC valves.and the MSIS bundles mentioned previously with regard to the'4Linch lineL

' break. Though considered a' remote possibility,_this .

I could. result in-a reactor trip. As indicated previously,;

the motor driven EFW pump would.still be available to.

provide emergency feedwater as well as the auxiliary feedwater pump. If a' loss of off site power:is also postulated, high' pressure injection is available as a1 further backup.. (references 2,3,4)

~

f The MSIS bundles previously mentioned'are also potential targets due.to other. critical crack' locations. As indicated in-the previous FSAR analysis,.and' supported l- by the above additional' considerations, an MSLB or

critical crack should not prevent steam supply to the.

EFW turbine. . Failure of two MSIS bundles could result in loss of the F0GG logic,. including failure to.

isolate'the' steam supply from the affected-steam generator;;however, this should not impair the steam supply to the turbine since the original-analysis demonstrated that the postulated breaks do not affect =l the other steam supply line. .It-should also be noted that the ANO-1 MSLB safety analysis. takes no credit for the previous SLBIC logic or the new MSIS/EFIC- ,

features. Rather,-it assumes Main Feedwater is i supplied to both generators with and without operator i

action. Therefore failure of the MSIS bundles does not invalidate the conclusions of the previous HELB analysis or the Chapter 14 (design basis accident)

-safety analysis. However, the additional evaluations relative to the new targets in the area is' appropriate for inclusion in the FSAR and will therefore be reflected in-the next annual-FSAR update to be submitted in July,1988.

Notwithstanding the above argument, it is AP&L's intention t'o provide additional assurance that the postulated breaks are incredible'and therefore require a no additional direct provisions ~for protection. -In-

)

accordance with previously accepted criteria protection L ,.

from the' effects of postulated pipe ruptures is not required if (a) the protective measures "are not practicable and (b) a supplemental inservice inspection  :

program is required "....which consists.of 100 percent- a j

a

. _ _ _ . _ _ _ _ _ . 2

s

, Attachment to ICAN078705 July 30, 1987 Page 6 inspection of the circumferential welds at these locations during each inspection period as specified in section XI of the ASME Code." (reference 9, 10).

Based on the present design of the steam lines, existing supports, etc. , additional protection against l the effects of postulated pipe ruptures is considered impracticable since the existing structural members are not capable of supporting significant additional loads such as would be required for. pipe whip / jet impingement protection. Designing and constructing additional members to support the necessary protective features would require an enormous and unwieldy structure which is not warranted due to the accompanying cost and other complications. Therefore, the specific welds in question are considered non-credible break locations based on AP&L's existing inservice inspection program which meets the intent of the NRC criteria.

C. Re-evaluation of the 8-inch Atmospheric Dump Valve (ADV)

Lines The original FSAR HELB analyses for ANO-1 did not reflect a specific evaluation of the 8-inch ADV lines.  !

Based on conversations with Bechtel Inc. personnel, this apparent omission was attributed to the following factors. First, a break in an 8-inch line is clearly enveloped by the Main Steam line break as far as total energy released, thermal hydraulic system response, and so on. Second, the routing of the ADV lines is relatively short and in close proximity to the steam lines, suggesting that consequences of pipe whip and jet impingement were enveloped by the steam line cases analyzed. Third, the steam lines are generally interposed between the ADV lines and the targets considered (i.e. AC-operated steam admission valves).

Therefore, it was not considered necessary to analyze

_the 8-inch lines in the same manner as the other high energy lines.

Notwithstanding the~Bechtel arguments, it is extremely difficult to analyze the consequences of the 8-inch line break for new targets (for example) in the area without the same level of analysis provided for the other casesr. Therefore, AP&L has evaluated existing pipe stress ana'yses and determined postulated pipe break locations in general conformance with reference 7.

Consistent with the approach described in the

-~

above cases, the potential targets in the area were identified. Without performing detailed pipe whip motion analyses, etc. , it is apparent that ruptures in the 8-inch lines could cause failures of the MSIS bundles in the area, thus resulting in loss of the

. , . . Attachment to ICAN078705 July 30, 1987 Page 7 F0GG design ~ feature as described in the case of the steam line break. The HELB consequences are considered-enveloped by the above evaluation for the main steam HELB.and the FSAR safety analysis (Chapter 14) for the MSLB. envelopes the smaller (8-inch) break as a design basis accident.

In order to provide additional assurance regarding the consequences of breaks in the 8-inch line, the welds in question (as identified by the piping stress analyses) have been added to the existing Inservice Inspection program and examined as su applicable NRC guidance (reference 9)ggested by the

-These welds were inspected during the most recent refueling outage in late 1986 This is considered acceptable.

^'_ based on the above evaluations and the impracticability of providing the protection features which would otherwise be involved. The impracticability is due to the same reasons previously enumerated for the main steam high energy line break.

AP&L recognizes the need to formally incor'porate the evaluations discussed above into engineering design documentation. This is scheduled to be completed by September 30, 1987.

REFERENCES

1. NRC letter 1N-81-003 (1CNA018103) dated January 12, 1981

2. AP&L letter 1R-0381-04 (ICAN038104) dated March 12, 1981

3. NRC letter 1CNA068202 dated June 18, 1982

4. NRC letter ICNA108301 dated October 6, 1983

5. AP&L calculation 86D-1005-20 revision 0 (B&W calculation 32-1159704-00)

6. ANO-2 FSAR section 3.6.4.1.5;2

7. ANO-1 Safety Evaluation Report dated June 6, 1973 Appendix C

8. ANSI /ANS standard 58.9 " Sing"le Failure Criteria for Light Water Reactor Safety-Related Fluid Systems , section 4.6

9. ANO-1 Safety Evaluation Report section 3.6 (1NSE010000 through 1NSE ..

220000 and 1NSE0A0000 through 1NSE000000)

10. NRC letter dated December 26, 1973 (ICTA127306)

t Attachment to ICAN078705 July 30,1987 Page 8 Paragraph II.A.1.b:

Because of the upgrade of the EFW system to safety-related, and because of the addition of safety-related EFW initiation and control and flow indication per NUREG-0737, Item II.E.1.2, you must confirm that this equipment will remain functional following a loss of the existing nonsafety-related room cooling to ensure the operability of the equipment during design basis event. Please submit a summary of your EFW pump room cooling analysis.

Response

In response to concerns expressed regarding room cooling in the 1 EFW pump room, an analysis of room temperature response to a

~

design basis accident was performed. This analysis assumed no l active room cooling from the nonsafety-related room cooler and

~

j heat load from both EFW pumps operating at design ratings. No credit was taken for reduction of heat load due to throttling of EFW necessary to match decay heat removal later in the accident scenario.

The room was modeled as an enclosed space bounded by a ceiling, walls and floors. Heat transfer is from the equipment to air and then through the walls to constant temperature adjacent spaces. A 2' X 8' wall opening (wire mesh) provides additional  ;

room cooling by convective stack effect as the temperature I increases. The heat transfer to piping within the room was not modeled as a conservative simplification.

For the purpose of this analysis, assuming operation of both  !

EFW pumps, a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without cooling conservatively bounds all credible scenarios including natural circulation.

The analysis indicates a room temperature of 143 F at the end of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

i The components added per NUREG-0737, Item II.E.1.2 which are I required to function during this period include EFW flow transmitters, DC solenoid operated control valves with associated position contro11ers', and various AC and DC valve motor operators. A review of associated testing and analysis indicates these components are capable of functioning at temperatures greater than 200 F for the required time interval.

These components are generally suitable for much more severe environments; some art in fact qualified for post LOCA containment conditions. (Motor operated valves which perform their safety function upon EFW initiation and are not required further were not included in the above review.) _.

As noted in our previous response (1CAN128612), NRC requirements did not require upgrade of other EFW components to safety grade (i.e., EFW motor driven and turbine driven pumps). Although such

a I

~ '- t Attachment-to ICAN878795 -)

Ju'ly 30, 1987' I Page 9 upgrade was not required and the system 1 1s. felt to be highly j reliable in its presert configuration,.in order to provide the additional information requested, an evaluation of the capability of these components.to function under postulated loss

- of normal room cooling.was .also performed.' It should be noted that AP&L's typical design specification for such equipment is ',

140'F. The calculated room-temperature of 143'F is only- .

slightly in excess 'of.this value, and'could likely be reduced by_.

removing analytica1' conservatism. In this case, however, review of specific testing and analysis applicable to these components demonstrates their capability to function for.-.the-required time: interval at the-calculated room temperature.

Paradraph II.B.1.a:

- Your.. response did not address weaknesses noted in the- .. .

post-modification' test for battery D07, includingi (1) lack of-consideration for the minimum temperature (60'F) permitted by your procedures for battery operation;1(2) lack of monitoring. 1 of the acceptab1' e minimum voltage during-the critical period

-(one minute into test) identified in your calculations; and (3) lack of documentation for the actual test ' currents used.

Please discuss the effect of these weaknesses on the-acceptability of your post-modification test.

Response

IEEE-450 defines two types of. battery tests.- A " service test",

similar to the post modification test conducted on D07, subjects a battery to series of discharge currents and durations intended to. simulate the design load profile for the battery. . This test is normally used periodically during a battery's service life to confirm its ability to perform its intended function. 1 The second test defined by IEEE is a " performance test."1 This test subjects a battery to a constant discharge current for an extended period of time.and is intended to verify successful delivery q.f the entire rated capacity of the battery. Such deep

. discharges are normally performed much less frequently than service discharge test.and are'inore demanding _in terms of11ong" term effect on'the battery.

It is important to note that ANO-1 Technical Specification-4.6.2.3 requires a performance test.be conducted once every 18 months. (There is norrequirement for periodic service tests.) 3 This is much more severe than. Standard Technical Specifications _ R (and ANO-2 Technical Specifications) which require a service 1 test at 18 month intervals and a performance test-only once per  %

60 months. As a result of the ANO-1 requirements, performance-

• tests have served as the requirement for establishing operability- l of the station batteries.

-l j

?i

. l Attachment to ICAN078705 l July 30, 1987 l Page 10 1

Prior to the replacement of D07, AP&L required a performance test on the new battery be conducted by the manufacturer. Since I the deep discharge required by such test causes long' term .I permanent degradation it was not desirable to repeat the test following installation. Rather it was determined that the 1 manufacturers performance test, in combination with the sizing  !

calculations, adequately established operability of the battery i provided that additional testing be conducted to verify proper installation. It is the latter testing.which was the subject of 1

Item II.B.1.a. We concur that this test did not satisfy all the provisions of IEEE-450 for a service test and the specific deviations are discussed below. The primary purpose of this  ;

testing, however, was not to confirm battery capability but rather to verify proper installation. The specific items addressed in your recent request (1CNA068705, dated June 24,

_ 1987) do not compromise the validity of the testing to accomplish

- this purpose, but rather are concerned with the verification of battery capacity.

l Item 1. " Lack of consideration for the minimum temperature (60 F) permitted by your procedures for-battery operations."

The overall effect of including a correction factor 4 for operation at 60 F is to increase the test discharge  !

current by 11% (i.e., 11% more energy is removed from the battery). Therefore, the battery must be larger to supply identical duty cycles at 60 F versus the 77 F rating. As noted above, the intent of the post modification testing was to verify proper installation and integration of the new battery into the DC system, not to verify battery capacity. The battery capacity was verified by the factory performance test, vendor calculations, and AP&L calculations. During the SSFI inspection in January 1986, AP&L prepared additional calculations specifically addressing the 60'F factor and a 20% aging allowance. This calculation also concluded that the battery capacity is adequate to Tceet the emergency duty cycle when it is 20 years old and at 60 F. It is ,our position that the use of the factory performance' test, and the analysis to verify battery sizing, and the post modification testing to verify proper installation is acceptable in meeting the ANSI N18.7-1976 and QA Manual Section 11.0 guidelines.

Item 2. " Lack of mo'nitoring of the acceptable minimum voltage during the critical (one minute into test) identified in our calculations."

When sizing a battery the emergency duty cycle is broken down into segments and analyzed both individually and collectively. The results indicate the battery capacity required at any point in the duty cycle.

This technique is described in IEEE 484.

E-  !

, . 1

. " . .  ? Attachment to'ICAN878795 ' 1 3- July 30, 1987!

Page 11

]

8 It should be noted that calculations performed during  !

the SSFI inspection indicated that bank voltage used-in the calculated duty cycle would:not fall below the. -]

~

'105 volt acceptance criteria at any time during the

. cycle. .-These calculations, in combination with.the-performance. discharge test verifying battery capacity, provide assurance that the 1 minute voltage is acceptable.

'This item did not have a significant' adverse impact on the post modification test results.

Item 3. Lack of documentation for the actual test current's.-

used.

.. . . . u As noted in the SSFI report the test data' sheets did.

not contain recorded data of the actual; discharge-currents at various times during the test but rather contained only cell' voltage data. -The DCP did,.

however, specify the' load profile to'be~used and  ;

contained a verification signature denoting that the .i battery bank had " satisfied the duty. cycle without dropping terminal. voltage below 105 VDC.." While-recorded discharge current data would have provided  ;

better documentation of,the. test, we are confident the 1 test was performed using the specified load cycle.

Incorporation of the comments noted in the SSFI report would have resulted in a more rigorous test meeting the definition . 1 of a service test in IEEE-450, however, the post modification' l test is considered adequate to verify proper battery installation.  ;

This, with the factory performance test and calculations verifying  !

battery size, meets the intent of the post modification testing guidelines. Currently, AP&L is making several' changes to our-battery surveillance program. Specifically a Technical:

S) edification change request is being prepared which will make tie ANO-1 Technical Specification more consistent with current Standard Specifications and IEEE-450. -This change will propose .

replacing the requirement for 18 month performance discharge tests and, instead, requiring the less severe service discharge .

test to be conducted in full compliance with IEEE-450._ l Paragraph II.D.1: i Your response allows the depiction of erroneous information on P& ids and is considered unacceptable even if a drawing. note indicates that the information is "for information only".

Please describe your plans to remove or correct any erroneous valve position or locked status.information on P& ids. i Response: '

^

The status of locked valves is controlled at ANO by various .

operating procedures. This is considered more appropriate than '

depicting such information on design drawings. Therefore, the j

, . ,- -Attachment to ICAN078795 i

' July 30, 1987

)

Page 12 j 1

, .l

" locked open" and " locked closed" designations are being deleted ]

from P& ids. This effort, being done in conjunction with other a drawing revisions, began earlier this year and should be.

completed by July of 1988.

With regard to depiction of valve positions on P& ids, our previous response indicated that valve position is'also controlled'  ;

via procedures and noted that, due to valve alignment changes 1 resulted from plant mode changes, equipment outages, etc., the P& ids can at best only depict typical valve positions. The  ;

valve positions indicated on P&ID's remain generally l consistant with.the conventions utilized by the architect- )

engineer and since, this.information'is not relied upon j

. for plant operations or design, discrepancies would not  ;

l adversely affect-safe operation. However, methods to further '

enhance the P&ID's-in this area are currently under discussion-

/ with the Region IV staff. The -final resolution' of this item will be forwarded under separate cover by August 24, 1987.

l l

I l.

k 4

4 e

1 i

l l

L

. .h .' .g <

7 h

,/~x .

t.

x- .g.

ARKANSAS POWER & LIGHT COMPANY ,

October 1, 1987

.M O * ~ -

Ai ,

ICAN108702 o ,,

o 2/7 '9 6; Mr. J. E. Gagliardo, Chief c.'

co Reactor Projects Branch , W":

U. S. Nuclear Regulatory Commission *'

Region IV

. ,,, t/Giff(;n-611 Ryan Plazs Drive, Suite 1000 '"'0 34 09~ ^

Y Arlington, TX 76011 M

C

SUBJECT:

Arkansas Nuclear One - Unit 1 C Docket No. 50-313 License No. OPR-51 Additional Information Concerning Safety System Functional Inspection (SSFI) k s

Dear Mr. Gagliardo:

By letter dated July 30, 1987 (1CAN078705) AP&L provided certain requested information concerning the January 1986 Safety System Functional Inspection (SSFI). The response to one item concerning the indication of valve positions on P& ids was deferred to allow for further discussions among our respective staffs. This item was discussed with Mr. Dorwin Hunter during the week of September 16, 1987. Attached is our response to this item.

t m,

i Sincerely -

^ ,1 y

lJ.M. evine

/ Executive Director

.AN0 Site Operations -

JML: DRH: djm 1l attachment cc: U. S. Nuclear Regulatory Commission I Document Control Desk -

Washington, DC 20555 -

, O E 4:

(G) v' Regional Administrator Region IV .C '

U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 7.

Arlington, TX 76011 *T h

c:;

a '

l Attachment to ICAN108702 , .

• i' s a

Page 2 J

n

/ i Requested Information (II.D.1)

U i

"Your response allows the depiction of erroneous information on P& ids and is considered unacceptable even if a drawing note indicates' that the information is for "Information Only " Please describe your plans to remove or correct any erroneous valve i position or locked status information on P& ids."

v

Response

a The response to the portion of this request dealing with locked status of valves shown on P& ids was previously provided in AP&L's c letter dated July 30, 1987 (1CAN078705). A portion of this letter ,

dealing with valve positions is repeated below: 1 00 l

"~

"With regard to depiction of valve positions on P& ids, j our previous response indicated that valve position l 7 is also controlled via procedures and noted that, due j to valve alignment changes resulted from plant mode j M changes, equipment outages, etc., the P& ids can.at {

best only depict typical valve positions. The valve j O positions indicated on P& ids remain generally consistent j

~

with the conventions utilized by the architect-engineer  ;

and since, this information is not relied upon for l

plant operations or design, discrepancies would not adversely affect safe operation. However, methods to {j (A)

V further enhance the P& ids in this area are currently under discussion with the Region IV staff."

The basis for this position has been discussed in detail with your staff and AP&L continues to believe this issue does not adversely affect the safe operation of our facility. We agree, however, that valve position information consistent with an established convention is desirable and would be an enhancement to the P& ids. Therefore, the following actions are planned. A specific convention for the valve positions to be shown on P& ids will be established. This convention (s) will be established considering that used by the i <

architect-engineer during original P&ID development and input from the principal enct users of the P& ids.

Following establishment, this convention will be incorporated into appropriate procedures and/or drawings to assure that valve ,

positions are properly reflected during future design work. l l

l In addition, various groups who routinely utilize P& ids (e.g., '

operations and training) will be made aware of the convention and I requested to identify discrepancies to our engineering organization. Such discrepancies will be collected and corrected on ~

{

the P& ids during the normal P&ID revision process. Due to the i extensive usage of P& ids, frequent revisions required by design j O changes, and changes resulting from ongoing efforts such as component '

/ labeling reviews, it is anticipated that valve positions differing

(- from the established convention would be corrected within approximately two years. The success of this effort will be verified via a sample review. ];

A l

t a 1

' '* Attachment to ICAN108702 . -

y

• - Page 3 i

1

[~'}

\x_,/

Since this issue is not considered by AP&L to impact safe operation, we believe the plan outlined above is responsive to your request and I will allow this issue to be addressed in an orderly fashion. As discussed with your staff a separato effort on a shorter schedule  !

1 would adv'ersely impact several other efforts which involve P&ID '

changes of much greater significance. These include.DCP closeouts to tn reflect as built conditions and plant labeling programs.

The first phase of this effort, to establish appropriate conventions and modify appropriate procedures and/or drawings, will be complete c) by December 31, 1987. Verification of P&IO valve position consistency j with this convention is planned to be complete by December 31, 1989.  ;

c3 We will apprise you of any anticipated change should this schedule be j c3 impacted by other unforeseen efforts of greater safety significance. '

T 9

C O

l 1

l f r~'s k

v

' d.5

-c..

~

l l

l -

/'~'N .

) .

\s_ / ~

1

',y

+ -s , -

Q___--_-_______.--._. _ - - - _ _ - - - -