ML20216G340
| ML20216G340 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 06/24/1987 |
| From: | Gagliardo J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Campbell G ARKANSAS POWER & LIGHT CO. |
| References | |
| EA-86-151, NUDOCS 8707010071 | |
| Download: ML20216G340 (2) | |
Text
.
t JUN 341987 s
.In Reply Refer To:
Docket: 50-313/86-01 EA 86-151 Arkansas Power & Light Company ATTN:' Mr.. Gene Campbell Vice President, Nuclear Operations P. O. Box 551 Little Rock, Arkansas 72203 Gentlemen:
Thank you for your letter of December 26, 1986, in response..to Part II of the Notice of Violation and Proposed Imposition of Civil Penalty dated November 12, 1986. As a result of our review, we find that additional information, as discussed with you and other members of your staff on April 1 and May 1,1987, is needed. The specific areas are identified by the paragraph numbers uss~' in l
the Notice of Violation.
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, 1
including a valve located approximately 4 feet closer to a main steam line than l
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 j
L can be achieved, including the above considerations.
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 events.. Please submit a summary of.your EFW pump room cooling analysis.
Paragraph II.B.I.a: Your response did not address weaknesses noted in the post-modification test for battery D07, including:
(1)lackofconsideration for the minimum temperature (60 F) permitted by your procedures for battery operation; (2)-lack of monitoring of the acceptable 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.
e.j I
RIV:P C:RPB C: RPB MEMurphy/dp DRHunter
/J JEGagliardo W G ///87 Ifb/pf/87 h
///6/87 8707010071 870624 DR ADOCK 05000313" 3
r Arkansas Power & Light Company Paragraph II.D.1:
Your response allows the depiction of erroneous information on P&lDs 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.
After careful review of your response and other documents and information provided by your staff following the inspection, the following items are j
deleted as examples:
II.A.2.d, II.A.2.e, II.A.3, II.B.2, II.B.4, II.C.1, and 4
II.C.7.
The violations and severity levels remain as set forth in the original i
Notice of Violation and Proposed Imposition of Civil Penalty.
Your responses were acceptable and followup inspection and closeout has been performed (reference NRC Inspection Report 50-313/87-13) for the following items:
II.A.2.a, II.A.2.b, II.A.2.c, and II.B.1.b.
Your responses to Items II.B.3 and II.B.5 appear to address the NRC concerns for these items.
Corrective actions will be reviewed during a future inspection.
Please provide the supplemental information within 30 days of the date of this letter.
Sincerely, acds nol trane,1 ma r c. nrtw
\\u f J. E. Gagliardo, Chief Reactor Projects Branch cc w/ enclosure:
J. M. Levine, Director Site Nuclear Operations Arkansas Nuclear One P. O. Box 608 Russellville, Arkansas 72801 Arkansas Radiation Control Program Director bec to DMB (IE01) bec distrib. by RIV:
RPB RRI R. D. Martin, RA RPSB SectionChief(RPB/B)
D. Weiss, RM/ALF RIV DRSP RSB MIS System RSTS Operator Project Inspector, RPB R. Hall NRR Project Manager D. Powers
y a
~ ARKANSAS POWER & LIGHT COMPANY POST OFFICE BOX 551 LITTLE ROCK, ARKANSAS 72203 (5Q1)371-4000 December 26, 1986 l
@ b LT UMM %
. f
~2C07 d
'b Mr. James M. Taylor, Director Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission I
Washington, DC 20555
SUBJECT:
Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 i
Response to Notice of Violation and Proposed Imposition of Civil Penalty Part II and Request for Additional Information (NRC Inspection Report No. 50-313/86-01)
Dear Mr. Taylor:
The Nuclear Regulatory Commission issued its Notice of Violation and Proposed Imposition of Civil Penalty (NRC Inspection Report No.
50-313/86-01) relating to the Safety System Functional Inspection conducted on January 6-31, 1986 at Arkansas Nuclear One, Unit 1 on November 12, 1986 (1CNA118603).
Arkansas Power.& Light responded to Part I of the Notice of Violation in its letter of December 12, 1986 (1CAN128601).
Part II of the Notice of Violation is addressed in this letter.
The Nuclear Regulatory Commission also issued a request for additional information (ICNA078602).
A response is attached.
Very truly yours,
. Ted Enos, Manager Nuclear Engineering and Licensing JTE/sg Attachment
.;,{'
>J H "
g 2* 97 00 %
MEMBE A MIDOLE SOUTH UTILIT*4 av4TFAA i
w
m
.r w_
.n~.~~~.~,-,~~.n-,~~~~~
- ~. ~ - ~ ~ ~ ~
. December 26, 1986 cc:
Mr. Robert D. Martin Regional Administrator U. S. Nuclear Regulatory Commi Region IV ssion Arlington, TX611 Ryan Plaza Drive, Suite 100 76011 0
Division of Inspection PrograMr. James G. Partlow, Office of Inspection and Enf ms Washington, D.C.U.S. Nuclear Regulatory Commis i orcement 20S55 s on
\\
\\
1
\\
\\
i
\\
\\
\\
\\
\\
\\
\\
/'
/
_.._ _, J -._... a.._...
--~'----...]
s y
~ ARKANSAS POWER & LIGHT COMPANY POST OFFICE BOX 551 UTTLE ROCK. ARKANSAS 72203 (50.1) 3714000
(
December 26, 1986 i
Ml@ $ UVM h
4 h
ICAN128612
~2C07 bh
' '4 I
Mr. James M. Taylor, Director Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission Washington, DC 20555 l
i
SUBJECT:
Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 1
Response to Notice of Violation and Proposed Imposition of Civil Penalty Part II and Request for Additional Information (NRC Inspection Report No. 50-313/86-01)
Dear Mr. Taylor:
The Nuclear Regulatory Commission issued its Notice of Violation and Proposed Imposition of Civil Penalty (NRC Inspection Report No.
50-313/86-01) relating to the Safety System Functional Inspection conducted 1
on January 6-31, 1986 at Arkansas Nuclear One, Unit 1 on November 12, 1986 (1CNA118603).
Arkansas Power & Light responded to Part I of the Notice of Violation in its letter of December 12, 1986 (1CAN128601).
Part II of the Notice of Violation is addressed in this letter.
The Nuclear Regulatory i
Commission also issued a request for additional.information (1CNA078602).
A response is attached.
I Very truly yours,
. Ted Enos, Manager Nuclear Engineering and Licensing 1
JTE/sg Attachment
, -a - ] '
,y J
/[
$97 00>
i j
_ -me eav1-or,m,-,_
(~...~~,-~m.--m~~-.~,...-~~-,.m----m,a.,--x.,s-,,---~.-.
1 y
' December 26, 1986 i
cc:
Mr. Robert D. Martin Regional Administrator U. S. Nuclear Regulatory Commission i
Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 Mr. James G. Partlow, Director Division of Inspection Programs Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, D.C.
20555 i
)
t J
f -n..,.-,- u n ~ n n ne v na- ~nn a,-,=ensam - >, - ~ ~~-..~w-+. -,. s,-, ~
-m ~ ~
~:--s..-.
.we s.
ma nn +w
..s,,
<< ws,sen
(
_l0 b
ATTACHMENT RESPONSE TO NOTICE OF VIOLATION (NRC INSPECTION REPORT NO. 50-313/86-01)
SAFETY SYSTEM FUNCTIONAL INSPECTION i
PART II AND REQUEST FOR ADDITIONAL INFORMATION (ICAN128612) 1 l
l l
l 1
wmaysmaaviewwa,kenaMaww +wwwawamaemiarm.;w
}
)
\\
General Discussion j
l The Safety System Functional Inspection (SSFI) was first conducted by the Nuclear Regulatory Commission (NRC) in 1985.
The SSFI is a very comprehensive approach to systems inspections.
During the month of January, i
1986, a team of eleven inspectors conducted a detailed examination of the
{
Emergency Feedwater (EFW) System at Arkansas Nuclear One - Unit 1 (ANO-1).
The EFW Systems of Pressurized Water Reactors (PWRs) had been the subject of an NRC-mandated reanalysis and upgrade following the Three Mile Island accident.
This was particularly extensive regarding Babcock & Wilcox j
designed plants, such as ANO-1.
The NRC issued NUREG-0737 which required ANO-1 to provide automatic initiation and control for their EFW systems.
j NUREG-0737 also required a reliability analysis.
AP&L also took the j
initiative to perform related improvements to this system during the period between 1980 and 1985 in an integrated program of EFW Upgrade.
After its detailed inspection of the EFW System at ANO-1 in January,1986, the NRC's inspection team concluded, "In general, the inspection team found the design of the ANO-Unit 1 EFW system to be sound..."
As a result of this inspection, there were findings in the four general areas of design control, test control, maintenance procedure control and drawing control.
One of these findings has been proposed as a Severity Level III and has been addressed in AP&L's letter of December 12, 1986 (ICAN128601).
The remaining findings were of such a nature that it was appropriate to group them into four Severity Level IV violations.
AP&L has reviewed each of the individual findings which comprise the four proposed Severity IV violations. While, overall, the SSFI was a detailed and thorough technical review of the EFW system at ANO-1, we feel that it is necessary to consider the relevance of these findings with respect to the ANO-1 licensing basis.
We would make the following general observations:
We have not previously had the opportunity to make a formal response to i
a number of the individual findings.
The SSFI Inspection Report of March 31, 1986 (1CNA038611) listed numerous observations.
It was impracticable for AP&L to formally respond to all of the detailed inspection findings because of their number, although each was the subject of thorough internal reviews.
We did formally respond to all of the inspection findings which were deemed significant by the NRC, as was specifically requested in the NRC's letter of March 31, 1986 (1CNA038611).
A number of the inspection findings or open items which became sub-elements of the four level IV violations were not designated as significant by the SSFI report.
During the SSFI it was noted that in addition to Design Basis, findings would also be written based on deviation from current " good engineering practice" even though there was not a design basis requirement to adopt a more current practice.
In a number of the specific items listed in association with the proposed violations, we believe the deficiencies to be based on current " good engineering practice" as opposed to design basis.
Such deficiencies, we believe, are inappropriate to be used as a basis for a violation.
2 j
I e a
- n. -.- w w.,e m a m men w e s - =<.aana.w.awmaw.ksuswwmm 5
We request reconsideration of particular findings.
It is our understanding that the NRC Enforcement Policy would provide for fair application of NRC regulations among different licensees by not permitting violations to be issued relating to implementation of new-NRC requirements being implemented on a previously agreed upon
. schedule. Therefore, it is of concern to us that two proposed violations were based in part on findings relating to motor operated valves which are the subject of a formal NRC I.E. Bulletin program and findings related to a drawing contr01 upgrade ' program already undertaken at ANO.
We are certain that the concerns outlined above will be resolved'by a careful NRC licensing review of the individual proposed violations. We believe that the level of detail indicative of the thoroughness with which the NRC conducted this investigation should also be applied to the evaluation of the licensing applicability of the proposed violations.. We believe that the NRC will conclude, not only that the ANO-1 EFW system is sound, but that we have taken a serious and thorough approach in considering and responding to the concerns raised during the inspection.
In summary, we request that Violation A be reconsidered and that the Severity Levels of Violations B through D in Part II of the Notice dated November 12, 1986 (ICNA118603) be reduced from Severity Level IV to Severity Level V because the safety significance of the applicable discrepancies is minor.
A complete discussion follows.
t 4-Ea I
e sde g 9 i
, g" I
'
- Pked
'i b
I S
3-s e
mamuyawwansexoammmwawrmnuuwwaanmwmmww.unmunmame _-
)
{
NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTY During an NRC inspection conducted on January 6 - 31, 1986, violations of NRC requirements were identified.... The particular violations...
are set forth below:
g II. VIOLATIONS NOT ASSESSED A CIVIL PENALTY VIOLATION A:
10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.
The ANO Quality Assurance Manual - Operations, Section 3.0, Design Control, implements this requirement and commits the licensee to the provisions of Regulatory Guide 1.64 and ANSI N45.2.11-1974.
I Sections 4.0 and 6.0 of ANSI N45.2.11-1974 require, in part, that design activities are to be performed in a controlled, planned, and correct manner that is traceable to the design basis.
i Contrary to the above, as of January 6, 1986, the licensee's program for design control did not assure that design and design reviews were properly performed.
1.
The emergency feedwater (EFW) system design analysis was inadequate, in that:
The consequences of certain high energy line breaks in a.
the area referred to as the " penthouse" on EFW system components coincident with a single failure and the i
consequences of a EFW high energy steam line break on other safety-related equipment in this area were not addressed.
h j
Discussion - II. A.1.a g
Potential high energy line breaks (HELB) in the " penthouse" area were p
addressed in the ANO-1 Final Safety Analysis Report (FSAR) at the time of ANO-1 licensing. Although the EFW upgrade introduced a new section of high energy line, it did not introduce any safety-related equipment in this area, other than EFW turbine steam admission equipment (i.e. CV-2613, SV-2613, CV-2663, SV-2663). A break in this line would render the turbine-driven EFW pump inoperable due to loss of steam supply.
Therefore, subsequent damage due to the line break is of no additional consequence.
Further there is no other safety related equipment in the area, that could be adversely impacted by an HELB, the.t is necessary to address a design basis event.
This analysis was qualitative in nature and concluded that following any l
4
=
1 ma&.wwwkwum%. 4QuadixAwarMr.RMMShs& " :wadelblditdidMW66%Mn%%%4nSmNdW&wram 2 1
e
}
l I
~
postulated HELB in this area the plant could safely shut down using either the motor EFW pump, the motor auxiliary feedwater (AFW) pump, the main feedwater (MFW) pumps or high pressure injection (HPI) cooling.
1 As no significant equipment changes were introduced in this area, the EFW design change did not explicitly address HELBs, as the design was judged to be enveloped by the original Design Basis HELB analysis.
The EFW design change, as installed, did not violate the Design Basic HELB analysis or requirements.
AP&L acknowledges this evaluation could have been better documented in the design change package (DCP).
As previously committed, the HELB evaluation for the EFW system has been reviewed, updated, and included in the most recent FSAR amendment.
II.A.1.b. The determination as to whether safety-related room cooling was needed when both EFW pumps were operating was not performed.
Discussion - II.A.1.b:
The design basis of ANO-1 as discussed in the FSAR does not require safety grade room. cool,ing in the EFW pump room.
NUREG-0737 provides additional
\\
requirements for EFW beyond the FSAR, specifically, that a safety grade initiation and control system be installed.
EFW pump room cooling was not addressed by NUREG-0737 nor in the NRC's review of our compliance to NUREG-0737, nor are we aware of any other requirement applicable to ANO-1 for safety grade room cooling.
Original design basis calculations for the EFW pump room concluded that the existing room cooling system is adequate.
As the EFW upgrade modifications did not add any additional heat sources to the EFW pump room, it was not necessary to re-evaluate the room cooling system.
II.A.2.
Design calculations and verifications associated with certain electrical equipment were inadequate in that:
a.
The correction factors for operation of the station battery at minimum temperature, specified as 60 F in Procedure 1307.006, 007 Quality Surveillance, were not included (Calculation GE-83D-1032-01, January 25, 1984).
(Note:
Procedure 1307.006 is correctly entitled D06 4
Quality Surveillance).
sctssion - II.A.2.a:
r The design basis of the ANO-1 batteries is to provide continuous emergency DC and vital AC loads for a minimum period of two hours (as required by the FSAR), in addition to supplying power for the operation of momentary loads during the two hour period.
It should be noted that the ANO-1 design basis does not specify a minimum battery design temperature.
5
imuum am+m,
.*mensMwwtweMmmewwwwmm%wwnenmus.uc~~~
Calculation GE-83D-1032-01 demonstrated conservative battery sizing for the new 58 cell, 1500 ampere-hour, (which replaced the 60 cell 1350 ampere-hour battery) configuration since it actually assumed DC loads for a period of eight hours, rather than the two hour FSAR requirement.
The specified tempera hre of 60 in Procedure 1307.006, D06 Quarterly Surveillance, does not constitute a design basis requirement, and is therefore not technically a part of the ANO-1 licensing basis for the battery design.
Nevertheless, acting upon the concern expressed by the SSFI team, Calculation GE-830-1032-06 was produced.
This calculation demonstrated adequate battery capacity for the FSAR load duty cycle duration of two hours, while assuming correction factors for aging and 60 F electrolyte temperature.
II.A.2.b. The station battery sizing calculations did not consider meter inaccuracy or increased loads during a design 1g basis event (Calculation 83-1032-06, January 24, 1986).
4,/
Discussion - II.A.2.b:
With respect to meter inaccuracies, this comment is applicable only to inverter loads.
Subsequent investigation with calibrated instruments confirmed our position that these contributions are insignificant and have a negligible effect on establishing overall battery size.
Furthermore, conservatisms in the calculation (e.g., assumptions for worst case load sequence, 25% margin for DC panel continuous loads and 40% margin for DC panel momentary loads) provide more than adequate design margin considerations for the battery sizing calculation.
II.A.2.c. The valve actuator sizing analysis for motor starting torque did not adequately consider the design requirements (Calculation 80-1083A-02, January 14, 1986; 80-1083A-04, January 24, 1986; and 83D-1032-07, p-January 24, 1986).
)fDiscussion - II.A.2.c:
The valve actuator in question is the EFW test loop isolation valve, CV-2870.
This valve is required to operate only if EFW initiation occurs during a turbine-driven pump flow test.
The remainder of the time this valve is kept closed.
The question regarding the adequacy of the calculation to demonstrate adequate valve operating torque involves a postulation that CV-2870 will be required to operate under worst case conditions associated with battery sizing methodology.
Although not part of the original valve operator selection criteria, under this assumed worst case voltage drop, AP&L justified that the torque developed will be above that required by the operator to close the valve.
J 6
j
d.
_.;.N.
1 Whysiskwam,M.M:we _
4~'e r
Subsequent data from the valve manufacturer confirmed this and has since been incorporated into the calculation.
Using the valve manufacturer's minimum torque requirements and the battery worst case voltage dip, the calculation confirmed that CV-2870 will close.
The generic problems, related to maintenance of manufacturer's actuator sizing analysis (torque requirements) are currently being addressed by the industry under I.E.Bulletin 85-03, which was issued just prior to the' SSFI inspection and allows a two year completion period.
r'[%)
II.A.2.d. The compatibility of the DC distribution system E
components with the new, larger batteries was not 8
Lp
/ f)j' determined (Calculation 83D-1032-02, February 17, 1984).
Discussion - II.A.2.d:
Compatibility of the DC distribution system components 'with the new, larger batteries wjLs determined consistent with the design basis requirements.
The inspector noted that one component, which could add approximately 2% to the short circuit current in the calculation, was not considered.
This was a conscious omission in the calculation as it was deemed insignificant and covered by other conservatisms.
However, the inspector also noted that several current limiters were equally not considered -for conservatism.
Our calculation concluded that short circuit current was acceptable and that compatibility was established.
Although the inspector disagreed with some of the conservative and simplifying assumptions made in the calculation, he did not challenge the adequacy of the calculation or compatibility of the system.
l II. A.2.e. The protective relay. study for breaker A311 tor the EFW
[h
/
pump motor failed to document the source of'the critical
,1 f
/
parameters (Calculation 84E-0083-12, November 19, 1985).
[l 4
Dis /cussion - II.A.2.e:
~
/
4 Safe stall time, starting time and locked rotor current are ' variable parameters, dependent on such items as bus voltage and temperature of motors.
A particular value of these variables is selected and used to select a relay setting.
The relay setting is, again, a variable dependent on both time and current.
The calculation selects a particular value.
The relay protection setting selected by this process is an engineering.
compromise between at' equate motor protection versus reasonable assurance of avoiding unwanted trips.
Extensive operating experience at ANO-1 supports the validity of the relay setting in question.
The issue, therefore, is not a question of the accuracy of the calculation but of the adequacy of references to any documents which the engineer may have relied upon when performing the calculation.
Documentation of the sources of all of these parameters was available and have been identified by AP&l..
Safe stall time and starting time were documented in the original l
i Bechtel Relay Protection Study, which is included as a reference 'in the current protection study.
Locked rotor current was documented in a conversation memo with Westinghouse dated October 22, 1984.
7
mamma.mam,mmamusuwsuaummm.mumemmmewammmann
.,w i
i Although the sources from which all these parameters were drawn were not explicitly referenced in the portion of the Relay Protection Study reviewed by the SSFI team, AP&L had appropriate documentation in its possession.
'll A.3.
Vendor-supplied motor operated valve design data were not
[5^/.
\\
adequately translated into controlling documents, in that:
,/
g' /
f Valve maintenance procedures did not include certain f
a.
h'f vendor information for torque switch settings, and as a 0
-,0 3'/
result the torque switch open setting for valve CV-2627 g 6 g/5 ' -
was set below the manufacturer's tested setpoint.
/
b.
There were not documented bases for the torque switch settings for valves CV-2870 and CV-3851.
Discussion - II.A,3.a. & II.A.3.b:
The torque switch setting in question for CV-2627 was found to be below the manufacturer's tested setpoint and could have possibly prevented the valve from going full open.
The valve is normally open.
Therefore, its design basis safety function would be to close for steam generator isolation.
While performing routine surveillance testing of the EFW system, the valve is closed.
If EFW was required while the test was being conducted, the valve would be required to open.
This test condition exists for only a few hours a month.
Concerns regarding inadvertant closure requiring reopening
(
are being addressed in accordance with our response to I.E.Bulletin 85-03.
Documentation of manufacturer's torque switch settings for CV-3851 was not available during the inspection.
CV-3851 was subsequently determined to be set within limits.
j No documentation was available to establish manufacturer's torque switch settings for CV-2870.
No operational problems had been identified on CV-2870 which would impact its capability to perform its safety function.
However, because of the lack of documentation relating to the the operator, the operator was replaced during our recent refueling outage as part of our response to I.E.Bulletin 85-03.
I.E.Bulletin 85-03 was issued by the NRC prior to the SSFI at ANO-1.
This Bulletin outlined potential generic problems with valve actuator performance, operation, maintenance, and testing and specifically noted problems with torque switch settings.
The Bulletin outlined actions to be taken by licensees to investigate and correct such problems.
A program and schedule was established for accomplishing the work and for providing information to the NRC.
We believe that it is inappropriate for these three valve issues to be cited as part of this subject violation as the potential problem had been identified and corrective actions were underway in accordance with the instructions and schedules specified by the NRC.
These three valves were included in our I.E.Bulletin 85-03 program and have been satisfactorily addressed during the recent 1R7 refueling outage.
G j
aan.mn+e.. s.>-aeu.warnavmnwxmasn-amamemm.,u,.mmmmmmmmvuwe,w.en zuwn.mu i
Admission or Denial of " Violation A" - Severity Level IV For the reasons discussed above, AP&L does not believe that a violation of applicable regulatory requirements and/or the ANO-1 design basis has occurred.
&E_such we deny the violation.
Many of the items discussed involve performance of activities in accordance with current-day requirements and current engineering practices which are beyond the ANO-1 design basis. We constantly change our programs in an effort to improve and to maintain abreast of improved practices.
Each of the items above have been evaluated and in many cases procedural changes have been made to improve the quality of our operation.
Although we do not believe that a violation of regulation has occurred, we have considered your observations and have taken action, where appropriate, to ensure station reliability and to improve our programs.
l 9
J
]
-~,.__-.- -~.- -.- - -.- - -.- -
x VIOLATION B:
10 CFR Part 50, Appendix B, Criterion XI, Test Control, requires, in part, that a test program shall be established to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.
The ANO Quality Assurance Manual - Operations, Section 11.0, Test Control, implements this requirement and commits the licensee to Regulatory Guide 1.33 and ANSI N18.7-1976.
ANSI N18.7-1976, Section 5.2.19, Test Control, requires, in part, j
~
that a test program be established, including pre-operational tests, surveillance tests, and maintenance / modification tests such that the safety of the plant is assured by demonstrating the satisfactory performance of a system or component following plant maintenance or modifications.
Contrery to the above, as of January 6, 1986, the licensee's program for test control did not demonstrate that components would perform satisfactorily in service.
1.
The modification acceptance tests for the station batteries did not demonstrate that the batteries would perform acceptably.
a.
The post-modification service test for battery 007 (ANO l
JO 05396) performed on March 24, 1984, in accordance with design change package DCP 83-1032 and calculation p,f GE-83D-1032-01 did not include certain design requirements, including electrolyte temperatures, actual test current, battery minimum voltage limits, and discharge voltage profile calculations.
fh Discussion - II.B.1.a:
The design basis of the ANO-1 batteries is to provide continuous emergency DC and vital AC load for a minimum period of two hours in addition to supplying power for the operation of momentary loads during the 2-hour i
period.
Adequacy of the batteries to perform this function (and determination of OPERABILITY) is established by conducting a performance discharge test in accordance with Technical Specification 4.6.2.2.
This specification requires a performance discharge test to be conducted in accordance with manufacturer's recommendations.
A performance discharge test was performed by the manufacturer on the batteries.
This test proved the capacity of the batteries and initial compliance with the Technical Specifications was established by this test.
The item noted in this finding pertains to the post-modification service test performed by AP&L under a Special Work Plan as part of the design 10 j
gundLawaredd:amaantn#esawmee*Na*MWsm$******ENm*S****
y o
change package.
The service test further demonstrated that the battery was properly installed and capable of meeting its intended design function.
In spite of perceived test inadequacies by the SSFI team, the post-modification service test fully demonstrated satisfactory system performance per ANSI N18.7-1976 and was in accordance with the ANO-1 battery design basis.
II.B.1.b. The post-modification testing performed approximately 1983-1984 in accordance with Special Work Plan 1409.29, after the removal of two cells from each station battery (DCP-83-119), used a battery temperature for which its
[A) basis could not be determined and the temperature correction factor was incorrectly applied.
/
Discussion - II.B.1.b:
AP&L agrees that certain apparent anomalies with regard to temperature correction existed in Special Work Plan 1409.29.
However, in spite of those anomalies, the test results clearly demonstrated that the battery capacity was in excess of its 1500 ampere-hour rating.
Therefore, the battery c'pacity was determined to be adequate.
a 7
II.B.2.
The EFW system discharge piping valves (CV-2620, CV-2626, CV-2627, CV-2670, CV-2869, and CV-2870), installed during the 1984 refueling outage as part of the EFW upgrade modification f
(DCP 80-1083), had not been tested or evaluated to
/p jf g demonstrate the capability to operate under system flow j
p conditions.
q D scussion - II.B.2:
/
ANSI N187-1976, Section 5.2.19.3, Tests Associated With Plant Maintenance, Modifications, or Procedure Changes, delineates test requirements as follows:
e.
" Tests shall be performed following plant modifications or significant changes in operating procedures to confirm that the modifications or changes reasonably produce expected results and that the change does not reduce safety of. operations."
o,s -
. o.;
.u w.
x-AP&L performed an Emergency Feedwater Initiation and Control (EFIC) System test to verify system performance as designed.
Further, a hot standby EFIC test demonstrated the integrated Emergency Feedwater (EFW) System operation following an EFIC initiation'to obtain full flow data for the EFW pumps, and to obtain system response time from EFIC initiation to delivery of water to the steam generators.
During this test the NSSS was at normal operating temperature and pressure.
This testing was performed to the extent possible without adversely affecting critical plant parameters.
It is AP&L's opinion that this met the intent of the referenced ANSI standard.
11
Masa@//&ineWeseWesamedstifes@eug@ise4MhtkmshwAah,m
Additionally, more stringent testing requirements were part of I.E.Bulletin 85-03, issued shortly prior to the SSFI.
The EFW discharge valves and flow testing of those valves were specifically required in I.E.Bulletin 85-03.
For reasons discussed in our discussion of Violation II.A.3.a and II.A.3.b above, we believe that it is inappropriate to include this item as part of the violation.
II.B.3.
Post-maintenance testing had not been performed in accordance with Procedure 1402.09, Emergency Feedwater Pump Maintenance, following three separate maintenance activities performed during the 1984 refueling outage for EFW pump P7A.
Procedure i
1402.09 provides detailed guidance for taking post-maintenance vibration readings in the horizontal, vertical, and axial directions and requires that they be 1
compared to a set of pre-maintenance vibration results.
a.
The testing documented for the December 23, 1984, maintenance on EFW Pump P7A under JO 76916 was conducted on January 17, 1985; however, the test records did not document that additional modifications to the pump had been made on January 7-11 and therefore, the testing did l
not reflect the work done under JO 76916.
b.
The testing documented on JO 81212 for replacement of a thrust bearing was missing some axial measurements and there was no pre-maintenance data for comparison.
c.
There was no post maintenance test data or reference to test data recorded on J0-75648 for maintenance conducted on January 11, 1985.
Discussion - II.B.3.a - II.B.3.b - II.B.3.c:
j/
1
[
The job orders were conducted in an integrated sequence.
Job Order 76916
'/
was performed to reassemble Pump P7A after outage maintenance and testing, j
on December 23, 1984.
At this time the plant was shutdown for a refueling I
outage and no steam was available to run the pump for conduct of the vibration tests.
The vibration test section of the procedure was left open and carried over until the test could be performed.
When steam was available some surveillance testing was done and, in the course of this testing a thrust bearing overheating problem was identified.
The vibration testing could not be performed at this time because of the problems with the bearing.
The bearing was disassembled and a shim was replaced under Job Order 81212 on January 7-8.
On January 8, vibration readings were taken, based on test data taken under plant operations Procedure 1106.06.
Because the operations surveillance procedure does not require axial readings for pump inboard and outboard, no axial data was taken.
Job Order 75648, performed on January 11, was issued to reverify and replace the shims.
j Following this task, the operations surveillance procedure was again i
utilized to verify pump operability.
Although the full data initially requested by the post maintenance test job order was not taken, we believe that the operations surveillance procedure was sufficient to show operability after these maintenance activities.
12
uw9awc. caw.msweesemwamWWEmewsmwaawmemmewwwm%=,.- e I
We acknowledge a failure to specifically comply with all of the procedurally specified data requirerrents. However, we believe that the intent of the testing to verify pump operability following maintenance was met.
Maintenance Procedure 1402.09 will be revised to clarify post-maintenance test requirements following delays in testing due to plant conditions, f~
' JI.B.4.
The initial calibration and functional checkout of the
/
e' Condensate Storage Tank (CST) level indication transmitter
' p hp #
/
(LIT-4203) were not adequately performed and documented, and M-
/
routine surveillance procedures were not developed and
,eT performed following installation of LIT-4203 during the 1984 refueling outage as part of the EFW upgrade modification (DCP 80-1083 and DCP 84-1045).
Discussion - II.B.4 :
! At the time of the inspection, AP&L was unable to retrieve calibration data j
on the initial post-installation calibrations of the condensate storage tank (CST) level indication transmitter.
However, when the calibration of the j
transmitter was checked, it was clear that the transmitter war in specification.
This indicates that the calibration was performed but that calibration data was not retrievable.
AP&L acknowledges that we were unable to locate documentation showing the date and method of calibration.
However, it is evident that the calibration i
had been done, since followup testing demonstrated that the transmitter was in calibration.
This is believed to be an isolated case of inadequate documentation, and this conclusion is further supported by the SSFI findings that, for the several components reviewed by the team, post-installational functional testing was performed before restart; which constituted sufficient initial surveillance testing for all but the subject transmitter.
An additional concern of the SSFI team was raised related to the fact that, at the time of the inspection, routine calibration procedures had not been formalized for this instrument.
Routine calibration procedures are not normally prepared until a time closer to when they are needed, in this case the subsequent refueling outage.
Routine calibration procedures have been prepared and used during the recent 1986 ANO-1 refueling outage (1R7).
AP&L has a formal mechanism for identifying the need for such procedures which has generally been quite effective.
The mechanisms in place would have, in the normal course of events, provided a calibration procedure for this instrument at an appropriate time.
II.B.S.
The required routine inservice testing of certain valves 1
associated with the EFW system was not adequately performed
/p fl and documented.
/(
The operability and the full stroke response of valves CS-98, CS-99, CS-261, CS-262, FW-55A, FW-558, FW-56A, FW-568, FW-10A/FW-61, and FW-10B/FW-62 had not been demonstrated and documented, sithough acceptable flow had been demonstrated through the flowpaths in routine pump flow tests.
13 j
rm-+
w
~
~
o ce _.t n gggg g.
ww
- n.Arg DISCUSSION - II.B.5:
Because of their configuration, it is impractical to individually test check valves CS-98, CS-99,.CS-261 and CS-262.
For this reason an exemption request from the testing requirements of Section XI had previously been submitted to the NRC in AP&L's letter of February 20, 1985 at page 21 (ICAN028507).
Further, with the addition of the seismic condensate storage tank these valves are no longer in the primary flowpath for EFW.
Check valves FW-55A, FW-558, FW-56A, FW-56B, FW-10A and FW-10B are specifically identified in AP&L's ISI program.
This program references the monthly surveillance testing requriemetns of Procedure 1106.06 " Emergency Feedwater Pump Operation" to fulfilling the exerciring requirements of ASME Section XI.
As you indicated, the testing of these valves was not specifically identified in the surveillance procedure.
However, it is intuitive from'the system design that the check valves are in fact tested by the monthly surveillance requirements.
To avoid future misunderstandings, Procedure 1106.06 has been revised to indicate flow verified stroking of these check valves.
Currently there is no available indication to measure flow through check valves FW-61 and FW-62 in the mini-recirculation line.
AP&L is evaluating the need for this indication.
Appropriate instrumentation, will be installed during thenext refueling outage, if our evaluation concludes that such instrumentation is required, or an exemption request will be made.
Admission or Denial of " Violation B" - Severity Level IV AP&L admits that relatively minor violations of procedures and testing requirements occurred as identified in three of the items above.
We have or will be taking steps as outlined above to address each deficiency.
Many of the other items cited above involve performance of activities in accordance with the latest standards or regulatory guidance which is beyond the ANO-1 licensing basis. We constantly change our programs in an effort to improve and to stay abreast of improved practices.
Each of the items above have been evaluated and in many cases procedural changes have been made to improve the quality of our operation.
Although we do not believe that a violation of regulation has occurred in several of the instances, we, nevertheless, will be using your input to consider changes that will improve our overall performance.
We admit the violation based on the identified discrepancies and request reconsideration of the Severity Level because of the minor safety significance of these procedural discrepancies.
14
- -mmwwpwesvwwwtew:@{uumkusamantwmntunnenkwwkksmshacW%d41ukM;eric.cdtab.,tsai,4,ar- -
- "- 1 VIOLATION C:
Technical Specification 6.8.1 requires, in part, that written procedures shall be established, implemented, and maintained
(
covering the applicable procedures recommended in Appendix A
^
of Regulatory Guide 1.33, November 1972.
Section 1 of Appendix A to Regulatory Guide 1.33, November 1972, states, in part, that maintenance affecting safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances.
Contrary to the above, several discrepancies were identified regarding EFW motor operated valve (MOV) maintenance Procedures 1402.160, Limitorque Motor Operated Valve _SMB-000 Maintenance, Revision 3; 1402.161, Limitorque Motor Operated Valve SMB-00 Maintenance, Revision 1; and 1402.71, EIM Motor Operated Valve Maintenance Revision 2.
(NOTE: 'The actual
~ ~ ~ ~
procedure number prefix is 1403, not 1402.)
s y
T
/
1.
All three procedures referenced drawing E-195 for the description of the motor operated valve limit switch operation.
However, this drawing did not show the Limitorque f
limit switch contact scheme and the EIM limit switch scheme y/f' incorrectly showed contact "LSO/G" as being closed continuously throughout valve travel.
/ Discussion-IIC1
/
This inspection finding does not identify a deficiency which is related to maintenance activities at ANO-1.
Drawing E-195 and the three MOV maintenance procedures do not cross-reference each other.
The drawing is a schematic and would not, therefore, be used to set. limit switches.
The i
drawing does, in fact, show limitorque limit switch contact schemes-(Reference' schemes A, B, C, and M).
The EIM limit switch scheme is incorrect, and a drawing change has been initiated to correct it.
- However, this discrepancy would not affect torque switch or. limit switch settings as these are done under specific steps in the procedure.
2.
Valves CV-2663, CV-2620, CV-2870, CV-3850, CV-3851, CV-2627, CV-2626, and CV-2869 were not identified as de powered MOVs or as Q-listed valves.
~
Discussion - II.C.2
, This inspection finding does not identify a deficiency in that the list in
[_i
/. question does not attempt to identify all-of the characteristics of every
/
valve nor does it control maintenance activities.
The list is Attachment 3 to the Limitorque MOV Maintenance Procedures (1402.'160 and 1402.161).
The-
,/
purpose of the list is to provide information to address specific topics and-it is not intended to be an exclusive or all inclusive reference.
For i
example, the -list may identify specific concerns'such as testing l
1 15:
I
_j
_ yo--
mc -
ew
--*w****=+m n=ra"*********"
s requirements, torque settings, special wiring schemes, DC power, EQ, Q and LLRT specifics but none of these are required to be acted upon by maintenance personnel as a result of being listed there.
The fact that DC power or Q designation is not a specific item on the list does not affect maintenance done in accordance with these procedures, since maintenance activities are governed by other documents in addition to these procedures.
3.
The open and closed adjustments.for torque switches were shown as p
reversed in Procedures 1402.160 and 1402.161.
iscussion - II.C.3
/
The Limitorque MOV Maintenance procedures contained a drawing which was
[
intended to represent the torque switches.
The procedural language provided proper directions for setup of the torque switches.
The' drawing was reversed and has been revised.
Maintenance action was not directed by the drawing, however, and the procedural language correctly directed maintenance activities relating 'to adjustment of torque switches.
4.
Procedure 1402.161 incorrectly listed CV-3851 and CV-2620 as model SMB-00 operators, rather than as SMB-000 operators as g
installed.
l j
Discussion - II.C.4
/
CV-2620 was correctly listed as an SMB-00 operator but the NRC has correctly 4
identified that CV-3851 had an SMB-000 operator which was incorrectly listed as an SMB-00 operator.
The procedures have been revised with respect to CV-3851.
The procedural misidentification would be expected to be rapidly identified and corrected during setup.
Further the setup' procedures for the two sizes of Limitorque operators would not be significantly different under these procedures.
Therefore, the safety significance of this error is minor in that it would not be expected to result in improper setup of CV-3851.
5.
The operation of the closed limit switches was incorrectly described in all three procedures.
Discussion - II.C.5 The EIM MOV Maintenance procedure was correct but the Limitorque MOV Maintenance procedures incorrectly described the coast-down operation in making closed limit switch adjustments.
We have reviewed the description and determined that it is of minor significance and would not have been expected to result in switch settings which were contrary to the intent.of' the procedure.
It should be noted that the the subject procedure has been revised to change the manner in which limit switches are set.
I.
Therefore, this concern has been addressed along with the' accomplishment ~of other recent improvements wh.ich AP&L has made in its MOV program.
i 16 i
e
m m-peyanamweswmmermmesomewsmaamwmamawauerme nemmuwww 6.
Procedure 1402.71 incorrectly identified CV-3850 as a
" modulating" valve rather than a " seal-in" type design.
(
iscussion - II.C.6 The procedure noted that opening and closing torque switches for CV-3850 must be set to a high value to allow the valve to be " modulated" while subject to a high differential pressure.
We feel it is clear that it was not the. intent of this sentence to describe the design function of the valve.
Because of the fact that the language actually triggered this misunderstanding, we have changed the word to " operated."
ph 7.
Procedure 1402.71 did not provide for independent
[
verification of the removal of a test jumper which could
/
p interfere with normal valve operations.
^>*1Di cussion - II.C.7 The issue raised by this finding is whether there was adequate independent verification of the removal of the dead man switch and replacement of the external power lead.
If the wiring configuration was not returned.to normal status the valve would not perform.
Post-maintenance testing verifies valve performance.
Therefore there was no need for independent verification, and AP&L contends that there is no discrepancy in this case.
This is a Severity Level IV violation (Supplement I).
Admission or Denial of " Violation C" - Severity Level IV The discussion of the individual findings under this violation shows that of seven identified inspection concerns only three-involved valid discrepancies; and we believe these discrepancies ~to be minor.
We have considered these findings, and made corrections to the procedures as needed to bring us into full compliance.
However, we have determined that the safety significance of the individual discrepancies is minor and that because they were few, and would not have.resulted in improper setup of the MOVs, that they do not indicate the programmatic concerns indicated by the Severity Level of this violation.
In addition to actions taken for each individual concern, the three MOV procedures reviewed by the SSFI team have subsequently been rewritten and reformatted according to AP&L's recently developed guidelines for-improved maintenance procedures.
The guidelines for maintenance procedures, discussed above, should ensure that consistent
{
quality of these procedures-is achieved and maintained.
We admit the violation based on the identified discrepancies and request reconsideration of the Severity Level because of the minor safety significance of these procedural discrepancies.
J 17 o
o j
1 VIOLATION D:
10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings appropriate to the circumstances.
The ANO Quality Assurance Manual - Operations, Section 5.0, Instructions, Procedures, and Drawings, implements the above requirement and requires, in part, that instructions, procedures,
)
and drawings are provided for the control of activities which i
affect quality and safety at the nuclear plant, and include, as a minimum, administrative, general plant operation, modification, l
maintenance and repair, and control of activities related to the ASME Code,Section III, Class 1, 2, or 3 components required to be operable.
Contrary to the above, as of January 6,1986, the licensee's program failed to assure that instructions and drawings were appropriate to the circumstances.
6
/
1.
The following piping and instrumentation diagrams associated I
[)/ gl f with systems important to safety were found to have incorrect valve positions, incomplete locked valve positions indicated,
,g-or missing instrumentation bubbles:
4}r a.
Drawing M-204, Sheet 3 of 4, Piping & Instrument Diagram, Emergency Feedwater, Revision 2 (two examples).
b.
Drawing M-206, Sheet 1 of 2, Piping & Instrument Diagram, Steam Generator Secondary System, Revision 45 (three examples).
Drawing M-202, Piping & Instrument Diagram, Main Steam, c.
Revision 33 (two examples).
3..
Drawing M-212, Sheet 1 of 2, Piping & Instrument Diagram, Plant Makeup Domestic Water Systems, Revision 29 (two examples).
2.
The piping design specification for ANO-Unit 1, M-84, Piping Class Drawing, Revision 22, contained errors and was not adequately controlled.
DCP 80-1083B identified that a revision to the piping a.
design specification was required; however, evidence of this change was not apparent.
b.
Errors were identified between a controlled copy of specification M-84 and the one in Document Control.
18 L
..~
J 3.
There were 18 examples of draf ting errors or incorrect information noted on controlled drawings.
This is a Severity Level IV violation (Supplement I).
Discussion - II.D The specific discrepancies referenced in the SSFI report have been corrected.
Discrepancies between piping and instrumentation drawings, operating procedures, and the as-built plant have previously been identified to AP&L as unresolved items in Inspection Reports 50-313/84-29 and 50-368/84-29, At that time a review was underway to determine a method for verification of the accuracy of operations procedures and associated Piping and Instrument Drawing (P& ids) with respect to the as-built status of the plant.
A formal description of the program AP&L has implemented to addrest these discrepancies was provided in response to Inspection Report 50-313/85-01 and 50-368/85-01, where this same item was followed up on by the Resident Inspectors.
In that response (0CAN038510) we indicated that in conjunction witu the equipment database collection effort for the ANO Maintenance Management System, walkdowns of plant components (e.g., pumps, valves, valve operators, pump drivers, instruments, breakers, etc.) are underway.
The P& ids are being used as a primary input to walkdown packages and discrepancies such as tagging, location or description of components are being documented by the walkdown teams.
These discrepancies are to be reviewed by ANO engineering and/or operations personnel to validate the finding and to determine corrective action.
Valid discrepancies are then corrected by changes to P& ids, procedures or tagging where applicable.
If discrepancies are determined not to be in accordance with design, then the as found condition will be evaluated for impact on function and reportability and the plant will be modified if necessary.
This effort was originally anticipated to be completed by first part of 1986; however, substantially more manpower than our original projection has been necessary to complete this task.
We currently anticipate completion of this project by mid-1987.
l i
Additionally, AP&L has initiated a program to close-out back-logged Design
)
Change Package (DCP) documentation.
Many of the discrepancies on the
_j drawings are a result of incomplete documentation updates following DCP j
installation.
The goal of this program is to update drawings within six months of design implementation, in the future.
As was indicated in our May 23, 1986 response (1CAN058608) to the SSFI report, it is impractical to' indicate " normal" valve position on the P& ids.
The normal position of valves is frequently a function of plant conditions and as such cannot be accurately shown on P& ids.
Such-indication of valve position is not used as design imput nor is it used by the operations staff to control valve position..P&ID drawing M-200, " Piping Symbols and Drawing Index" has been revised to include the following note:
i l
19 l
l
w -%admaussunemwmewmentwausesssnaamm:taxanwawumstaaraisa$e~
" Valve positions are not_ indicative of design requirements or operational conditions or requirements and are shown for information purposes only.
Valve positions will be controlled by
{
operational procedures."
Admission or Denial of " Violation D" - Severity Level IV AP&L admits that in general the findings associated with the violation are accurate.
This item had been previously identified by the NRC and action is currently underway to address this issue as stated above. We request reconsideration of the Severity Level because of the minor safety significance of these procedural discrepancies.
l 20
v o v n ' s d e c h 4 4. w m x M ::
i, c
a'
~
ewswamb
-/ z * :-
,x,,
e%%*
, j,, - ' i +.d
_f @ _ __ J
+
ADDITIONAL INFORMATION REQUEST LETTER (1CNA078602)
Item 1.
In regard to your response to Item II.A.I.9, your statement that the identified design inadequacy represented a minimal potential for degradation of the level of protection for public health and safety seems unsupported.
Complete consideration on your part of ef fects of not having check valves in the EFW turbine steam supply lines, other than the potential for a loss of EFW flow, was not apparent.
Your response was also considered weak concerning long-term corrective actions.
The response focused only on 10 CFR 50.59 reviews.
Improvements to the design review and verification process were apparently not considered.
Item 1 Response Please see our response of December 12, 1986 (ICAN128601) to NRC's Notice of Violation.
Item 2:
In your response to Item II.A.1.b, the statement was made that the high-energy line break calculations were not required because their were no safety-related components near the subject piping.
However, this does not appear to be consistent with the fact that the motor operators for CV-2617 and CV-2667 are installed on the subject piping in the area of concern.
There are other safety-related components such as CV-2613, SV-2613, CV-2663, and SV-2663, that apparently have not been considered and are located in the area of concern.
Item 2 Response Please see the discussion on II.A.1.a above.
Item 3 Your letter dated April 11, 1986, to Mr. John Stolz. in regard to ANO-1 seismic design, stated on page-4, "In addition, design procedures instruct the engineers to apply the latest criteria for 2 over 1 wherever practicable".
Please identify in your response the applicable procedure references referred to that pertain to ANO-1.
Item 3 Response Requirements relative to seismic II over I design are addressed in several Ap&L procedures applicable to ANO-1.
Specifically, ESP-202 Form 202F5 contains, among a series of questions which must be answered for each design change, the following:
"17. How was seismic category 2/1 considered in the design?"
This provision is also conta'ined in Attachment 3 to ANO procedure 1032.01--Design Control.
Civil Engineering Procedure CEQN-0002 which details the acceptable II/I methadology for ANO-2, provides for the 21
T q
m
.-m
, s..
2.ammm&u. v:3 Am...
'jp i c0. --
- 2: :;.L a,g. L x Mme..
_4 application of this criteria to ANO-1 designs at the discretion of the lead engineer.
In addition, subsequent to AP&L's letter of April 11, 1986 (1CAN048605), an Engineering Services Directive (ESD T-218) was issued to reinforce the intent of the procedures.
Item 4.
Your request to Item II.B.3 regarding overall actions being taken to address configuration control problems lacked sufficient detail.
Of the 30 individual configuration control items identified in the SSFI report, many identifying multiple problems, 8 items were in regard to the identification of normal valve positions.
Your statement that normal valve positions cannot be shown accurately on P& ids is not consistent with industry practice.
Additionally, your P&ID legend, drawing M-200, Revision 4, identifies normally open, normally closed, locked open, and locked closed valve symbols that are used uniformly throughout your P& ids.
Item 4 Response See our response item II.D above.
i 22 m