ML20195H528
| ML20195H528 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 06/14/1988 |
| From: | Craig Harbuck Office of Nuclear Reactor Regulation |
| To: | Tison Campbell ARKANSAS POWER & LIGHT CO. |
| References | |
| REF-GTECI-024, REF-GTECI-NI, TASK-124, TASK-OR NUDOCS 8806280316 | |
| Download: ML20195H528 (72) | |
Text
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o UNITED STATES l
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NUCLEAR REGULATORY COMMISSION n
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WASHINGTON, D. C 20555 g
June 14,1988 l
l Occket No.:
50-368 Mr. T. Gene Campbell Vice President, Nuclear Operations Arkansas Power & Light Company Post Office Box 551 Little Rock, Arkansas 72203
Dear Vr,
Campbell:
SUBJECT:
ARTR SAS NUCLEAR ONE, UNIT E - P.ESOLUTION OF GI-124, AUXILIARY FEECWATER SYSTEM REllABILITY An Auxiliery Feedwater System (AFWS) review group was fortred on Septerber ES, 19M to prepare an overall reliability assessment for each of seven plants with a two train auxiliary feedwater system, GI-124, Auxiliary Feedwater System l
Peliability. This effort included a plant-specific review and an on-site audit Of the AFW system, and included calculated estimates of the reliability of the AFPS given various initiating events. The staff selected this approach for resching Gl-124 rather than a strictly analytical approach because the staff believed that a first-har.d aedit of the AFWS design and cperation more directly addressed the rect causes cf AFW syster, unavailability ano unreliability.
l The resolution apprcach adopted by the AFWS review team relied on an audit of several parar,eters that directly or indirectly affect the availability and reliability of the AFW syst2m. These parameters include design configurations; maintenance, surveillance, and testing procedures and practices; operating procedures; personnel training; system layout; operating experience; instrumentation ard control; and environment and accessibility for operator recovery actions follcwing potential malfuncticus. The Standard Pevi (SRP) Section 10.4.9. AFV system numerical reliability criterion (10'gw Planto 10'5 per demand) served as the basis for concluding that the AFV system in the seven plants of concern was acceptably reliable.
Because the SRP criterien specifies consideration of compensating factors such as other reliable decay heat removal methods to justify a larger AFW system unavailability, the AFWS review team evaluated compensatory features as part of its effert.
By letter dated July 31, lee 7, a draft report of the results of the GI-l?4 team's findings for Arkansas Nuclear One, Unit 2 (ANO-2) was transmitted to the licensee. By letter dated November 17, 1987, the staff requested licensee coments on the concerns identified in the draft report. The licensee's responses were provided by letter dated January 29, 1988. The enclosed report documents the results of the staff review of the licensee's responses and conclusion concerning AFWS reliability at ANO-2.
Based on that review, the staff concludes that' improvement in the reliability of the AFWS and secondary side decay heat removal capability for ANO-2 is warranted. This conclusion is based on the staff's evaluation of the AFWS 8806200316 800614 PDR ADOCK 05000360 p
4 T. Gene Campbell reliability, which indicates a reliability below the 10-4 to 10-5 per demand acceptance criterion, and the staff's determination thLt the uncertainties and disadvantages associated with the use of the feed-and-bleed decay heat removal method cannot justify it as a suitable ccrrpensatory feature.
Feed-and-bleed should therefore be considered only as a last resort for decay heat retroval in the event of a loss of all feodwater.
You are requested to provide a response to findings of the inclosed report within 60 days of the receipt of the report.
In your response you should specifically address how you propcse to upgrace ycur AFWS reliability to include a tentative schedule for 'mplementation.
Please contact us if you have any questions concerning the enclosed report.
If necessary we are available tc meet with you to discuss the recomendations contair,ed therein.
1 The reporting and/or recordkeeping requirements contained in this irtter affect fewer than ten respondents, therefore, OPB clearance is not required i
under P.L.96-511.
Sincerely, C. Craig Harbuck, Project Panager Project Directorate - IV Division cf Reactor Projects - III, IV, Y and Special Projects
Enclosure:
As stated j
cc w/ enclosure: See next page DISTRIBUTION Docket File i
NRC t. Local PLRs PDIV R/F DCrutchfield LRubenstein JCalvo PNoonan CHarbuck OGC-Rockville RJordan JPartlow ACRS (10)
Plant Service List 1h PD4:PM PD4:D g 4 PD4:L
/
PNoon n l CHarbue : c/kb JCalvo 06/jd/8 06/M/88 06//y/88
r-4 T. Gene Campbell relist,ility, which indicates a reliability below the 10~4 to 10-5 per demand acceptance criterion, ar.d the staff's determinttion that the uncertainties art disadvantaccs associated with the use of the feed-anc-bleed decey heat removal methcd cannot justify it as a suitable compensatory feature.
Feed-and-bleed should therefore be considered only as a last resort for decay heat retroval in the event of a loss of all feedwater.
You are requested to provide a response to findings of the enclosed report within 60 days cf the receipt of the report.
In your response you should specificelly address how you propose to upgrade your AFWS reliability to include a tentative schec'ule for implerentation.
Please contact us if you have any cuestions ccrcerning the enclosed report.
If recessary we are available te freet with yct to ciscuss the recontrendations Cd'w II.t.6 T.h61 e i T.
The reporting and/or recordkeeping re.quirer,ents centained in this letter affect fewer tl.ar; ten respcndents, therefore, OPP clearance is not reo,uired under P.L.96-511.
Sincerely, C. Craig Harbuck, Project Manager Project Directorate - IV Divisior cf Reactor Projects - III, IV, V and Special Prcjects Enclost:re:
As stated ce w/erclosure:
See rext page
s
$r.T.GeneCampbell Arkansas Power & Light Company Arkansas Nuclear One, Unit 2 cc:
Mr. Can R. Howard, Manager Mr. Charles B. Brinkman, Manager Licensing Washington Nuclear Operations Arkansas Nuclear One C-E Power Systetr.s P. O. Box 600 7910 k'oodront Avenue Russellville, Arkansas 72801 Suite 1310 Bethesda, Maryland 20814 Mr. James M. Levine, Executive Directer Site Nuclear Operations Mr. Frank Uilson, Director Arkansas Nuclear One Division of Environmental Pealth P. O. Box 606 Protection Russellville, Arkansas 7280)
Department of Health Arkansas Departrient cf Pealth Mr. Nicholas S. Reynolds 4815 West Markhan Street Bishop, Cook, Percell & Reynolds Little Rock, Arkansas ???01 1400 L Street, N.W.
Pasbington, D.C.
20005-3E0?
Honorable William Aternathy County Judce of Pope County Regional Administrator, Region IV Pope County Courthouse U.S. Nuclear Regulatory Cermission Russellville, Arkansas 72801 Office of Executive Director for Operations 611 Ryan Plaza Crive, Suite 1000 Arlir.gter, Texas 76011 Senior Pesidert Inspector U.S. Nuclear Regulatory Comission I huclear Plant Poad Russellville, Arkansas 7?P01 Ms. Greta Dicus, Director Division of Environtrertal Health Protection Arkansas Cepartrent cf Health 4815 West Markan Street Little Pcck, Arkansas 72201 Mr. Robert B. Borsum Babcock & Vilcox Nuclear Power Generation Division 1700 Rockville Pike, Suite 525 Rockville, Maryland 20852 a
l 1
I Arkansas Nuclear One, Unit Auxiliary Feedwater System Reliability Assessment A.
Executive Summary and Conclusions This report contains the staff's assessment of the overall reliability of the auxiliary feedwater system (AFWS) for Arkansas Nuclear One, Unit 2 (ANO-2).
This review was performed in connection with the resolution of Generic Issue (GI-124), "Auxiliary Feedwater System Reliability," which addresses AFWS reliability in certain plants.
AFWS reliability analyses indicated that many plants fell in the high reliabil-ity range; however, several plants fell in the lower reliability range.
While these plants met applicable licensing requirements for the AFWS, their system reliability was still in question.
Some licensees for this latter group of plants implemented modifications to increase AFWS reliability to an acceptable range.
However, AFWS reliability for seven plants remained questionable.
The plants in this category are ANO-1 and 2, Crystal River, Ft. Calhoun, Prairie Island Units 1 and 2, and Rancho Seco.
The objective of the review under GI-124 is to evaluate the AFWS reliability for these seven plants and to document any recommendations for further licensee action.
The resolution approach adopted by the staff in its review of ANO-2 relied on an audit of several plant features that directly or indirectly affect the availability and reliability of the AFW system in addition to an assessment of numerical unavailability.
These variables included design configurations; main-tenance, surveillance and testing procedures and practices; operating procedures; personnel training; operating experience; instrumentation and control; and environment and accessibility for operator recovery actions following potential malfunctions.
The AFWS numerical reliability criterion (10 4 to 10 5 per demand) 1 1
4
'given in Section 10.4.9 of the Standard Review Plan (SRP) served as the basis for evaluating the AFWS in the seven plants of concern.
Because the SRP criterion specfies consideration of compensating factors such as the availability of other reliable decay heat removal methods to justify a larger AFWS unavailability, an evaluation of compensatory features was also conducted.
When determining whether or not to give credit for compensatory decay heat removal features, the staff position has been and continues to be that only features which relate to secondary side decay heat removal capability (e.g. a startup feedwater pump, AFW pump discharge crossconnections between units, or a third AFW pump) can be considered acceptable for satisfying the SRP criterion.
While the staff recognizes the capability to remove decay heal. in the "feed-and
-bleed" mode utilizing the primary system safety / relief valves and high pressure injection pumps, such a method involves large uncertainties in operator response.
Therefore, it is considered to be a suitable backup to the AFWS in emergency procedures as a last resort for decay heat removal, but is not sufficiently reliable to justify it as a compensatory feature in order to j
meet the SRP goal for AFWS reliability.
A five person review team reviewed documents and interviewed maintenance, operations, ermineering, and training personnel and management.
The review included a three day plant site visit.
The licensee's statistical analysis conducted for the last six year period for the ANO-2 AFWS-related failures shows a decreasing failure rate trend with no significant failures since 1983.
This indicates an improvement in equipment performance.
As discussed subsequently, the majority of equipment failures over a five year period from 1981 through 1985 was determined to be readily recover-able at the equipment location.
Therefore, the staff belicves that ease of access to various AFWS equipment is important to overall system reliability and availability.
Ease of access includes factors such as normal and emergency sighting, adequate communications, clear and legible equipment ident1fication.
These issues are discussed in detail in Section D.5.
2
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' Although significant improvements in valve performance has resulted over the last few years, the licensee's adoption of the M0 VATS
- methodology for valve setpoint setting (see Section D.2.2) is expected to further. improve valve performance.
The licensee's emergency operating procedures (EOPs) explicitly and clearly instruct operators, if the AFW flow cannot be established, to attempt to reestablish the MFW flow or, if that is unsuccessful, to rely on the condensate pump flow.
The E0Ps provide guidance and precautions about degraded modes of equipment operations.
If all MFW and AFW are not readily recovable, the E0Ps instruct the operator to initiate the "feed-and-bleed" mode of decay heat removal.
The staff has certain concerns with the "feed-and-bleed" capability including operability requirements, and emergency operation (see Section D.3.).
The ANO-2 "feed-and-bleed" system does however provide redundant flow paths with a dedicated 3-inch diameter line with two motor operated valves in series and a parallel path with two low temperature overpressure protection valves.
This size flow path is sufficient to depressurize the reactor coolant system to allow the high head safety injection pumps flow in the reactor vessel for core cooling.
Thus, "feed-and-bleed" can serve as an acceptable backup to a reliable AFWS as a last resort in emergency procedures for decay heat removal.
should all feedwater be lost.
l As discussed in Section D.7, the plant's rate of unanticipated automatic reactor scrams is high (5 per year in 1986).
However, it has been decreasing.
- Also, the failure rate of its two emergency diesel generators has been relatively low.
- H0 VATS = Motor Operated Valve Testing System 3
e 4
'On the basis of its review, the staff finds that the ANO-2 AFW system is generally well designed and instrumented.
The staff also finds that the AFW system design and operation adequately consider other staff generic concerns raised within GI-124 (i.e., GI-68 with respect to environmental qualifications of the motor driven AFW pump, GI-93 with to steam binding of the AFW pumps, GI-122.1.a, b, and c with respect to isolation valve failure, and interruption and recovery of AFW flow, GI-122.2 with respect to initiation of "feed-and-bleed,"
and GI-125.II.1.b with respect to single failure protection of existing AFW systems).
However, the staf f initially found that the licensee should take certain steps before the AN0-2 AFW system could be found accetably reliable.
The steps were as follows:
1.
The licensee should justify, using a reliability analysis based on plant-specific operating experience, whether the closed position of the first AFWS discharge valve in each discharge line is optimum for safety (See Section D.1.2).
2.
The licensee should propose appropriate Technical Specifications for the pressurizer vent valves and low temperature overpressure protection valves (Section D.3.2).
3.
The licensee should confirm whether one low temperature overpressure protection valve is sufficient for decay heat removal via "feed-and-bleed" (Section D.3.2).
4.
What cor.stitutes "abnormal" in the E0P should be adequately discussed during training (Section 0.3.2).
5.
Clear instructions as to whether or not to use certain key equipment should be included at the appropriate steps in the Inadequate Core Cooling section of the E0P (Section D.3.d2).
6.
The licensee should specify in the operator training length of time, after loss of all feedwater conditions are reached, in which "feed-and-bleed" cooling will be effective (Section D.3.2).
4
4 7.
The licensee should verify the basis for the environmental qualification of the AFW valves, and remove discrepancies in referencing Q-listed valves (Section D.5.2).
8.
The licensee should maintain all available communication means, including the public announcement (PA) system in a ready and operable condiition (Section 0.5.2).
9.
The licensee should improve normal and emergency ac lighting, and emergency dc lighting, in the vicinity of the AFW pump rooms (Section D.5.2).
10.
The licensee should maintain the cleanliness of the AFWS equipment and its surroundings (Section 0.5.2).
11.
The licensee should relocate and orient (as appropriate) the AFW discharge piping temperature, pressure, and turbine speed gauges to be readily visible (Section D.5.2).
l 12.
The licensee should ensure the integrity of all equipment supports in the 1
AFW system (Section 0.5.2).
13.
The licensee should complete the retagging program for the AFW system (Section 0.5.2).
14.
The licensee should develop a "maximum time to access" from the control
-room, for each piece of AFWS equipment that may potentially be manually exercised during emergency decay heat removal, and for the atomspheric dump valves (Section 0.5.3).
15.
The licensee should pursue its trip reduction program to decrease the unscheduled automatic reactor trip rate (Section D.7.2).
16.
The licensee should complete its AFWS reliability analysis and implement changes that provide significant improvements to the system reliabili.y based on that analysis (Section D.T.2).
5
O 4
4 17.
The licensee should provide an alternate method of providing water to the steam generators such as a connection to the auxiliary feedwater systems at Unit 1.
The operability of this source of water should be required by the plant's Technical Specifications (Section 0.7.3).
18.
The licensee should emphasize in its operator training program, the grave-ness of not initiating the "feed-and-bleed" where a genuine total loss of feedwater is suspected.
Training on recognition and mitigation of such an event should also be emphasized (Section 0.7.2).
By letter dated July 31, 1987, a draft report of the results of the staff review team's findings for AN0-2 was transmitted to the licensee.
By letter dated November 17, 1987, the staff requested licensee comments on the concerns identified in the draft report.
The licensee's responses were provided by letter dated January 29, 1988.
The following listing is the staff's evaluation of those responses.
The discussion below corresponds to the item numbers above.
1.
In response to the staff's concern regarding the normally closed position for the first AFWS discharge valve and its affect on system reliability, the licensee stated that it had evaluated the advantages and disadvantages of a normally closed versus open valve.
It was their determination that the benefits in prevention of hot water backleakage for reduction in potential AFW pump steam binding, and improved isolation capability for a depres-surized steam generator (such as following a mainsteam line break) with a closed valve outweigh the minor benefits resulting from a normally open valve.
This issue is resolved.
2.
In response to the staff request that technical specifications be proposed for the pressurizer vent and low temperature overpressure protection valves, the licensee stated that because use of these valves in the "feed-and-bleed" mode is beyond the design basis for AN0-2, the Commission Policy Statement on Technical Specification Improvement indicates that incorporating these valves in the technical specifications is inappropriate.
Since the staff is proposing that the reliability of the AFWS be furth'er improved '(see 6
4 Item 17. below) so that the reliance on "feed-and-bleed" as a compensatory decay heat removal means for supplementing AFWS reliability is not assumed, such technical specifications are unnecessary.
This issue is resolved.
3.
In response to the staff request to confirm the adequacy of a single low temperature overpressure protection (LTOP) valve for "feed-and-bleed", the licensee stated that analysis has indicated that sufficient oepressuriza-tion capability is provided by one LTOP valve because of its relatively large size compared to a standard PORV.
This issue is resolved.
4.
In response to the staff concern that the meaning of "abnormal" with regard to E0Ps be adequately discussed during training, the licensee indicated that a review of operator training coverage of what constitutes "abnormal" in the E0Ps has been initiated.
The licensee further stated its belief that current training and control board indication of abnormal conditions is adequate.
This issue is resolved.
5.
In response to the staff concern that clear instructions on the use of certain equipment be included in the inadequate core cooling (ICC) secton of the E0P, the licensee agreed that the current wording in the E0P on the use of the turbine driven AFW pump was not adequate.
The licensee committed to revise the ICC section of the E0P to be indicated to the operator that an attempt should be made to establish flow from the turbine driven AFW pump before using the condensate pumps.
This change will be made during the scheduled E0P revision in April 1988.
This issue is resolved.
6.
In response to the staff request that the licensee provide operator training guidance on the time frane in which "feed-and-bleed" cooling will be effective following a loss of all feedwater, the licensee stated that current procedures call for initiation of "feed-and-bleed" anytime feedwater flow is not established and RCS (Tc) temperature is greater than 560 F and 7
increasing.
The licensee stated that operator training will continue to emphasize the importance of following this guidance without delay in order to ensure the effectiveness of "feed-and-bleed" for decay heat removal.
This issue is resolved.
7.
In response to the staff regt.est that the licensee verify environmental qualification and Q-listing of AFW valves, the licensee stated that it is verifying that the maintenance procedures for the AFW valves are in agree-ment with the Q-list in order to confirm compliance with environmental qualification requirements.
Any discrepancies will be corrected.
This issue arose from an incorrect application of a 1977 ANO-2 pipe break analysis.
The licensee indicated that the AFW pump room doors are fire doors and thus remain normally closed.
Therefore, the valves in the room ere not subject to a harsh environment from postulated high energy pipe breaks.
Thus, the valves are not required to be qualified to a harsh environment under the environmental qualification requirements.
This issue is resolved.
8.
In response to the staff concern that the public announcement (PA) system be maintained and operable, the licensee stated that the broken PA system head set noted by the AFW review team has been repaired.
The licensee further stated that in order to ensure the operability of communications critical to remote shutdown capability, certain telephones are inspected for oper-ability on a quarterly basis.
This issue is resolved.
9.
In response to the staff concern that emergency lighting in the AFW pump rooms be improved, the licensee stated that it has conducted a review of lighting in the pump rooms.
Based on that review, the licensee stated that they have verified that adequate emergency lighting is available for I
s performance of manual actions in the area.
The licensee further noted 1
that a review of the existing emergency lighting as part of the fire 8
i
protection evaluation conducted for compliance with 10CFR50 Appendix R found the emergency lignting arrangement acceptable.
This issue is resolved.
10.
In response to the staff concern regarding the cleanliness of the AFWS j
equipment and its surroundings, the licensee indicated that it will pro-vide for regular inspection of the housekeeping conditions for the AFWS.
The licensee stated that it recognizes the importance of equipment and structural cleanliness to ensure proper operation and maintenance of plant systems.
This issue is resolved.
11.
In response to the staff concern regarding access and visibility of certain AFW system instrumentation, the licensee stated that it is evaluating modifications necessary to relocate the AFW pumps discharge temperature and pressure gauges and the AFW pump turbine speed gauge to be more readily visible.
The licensee is also evaluating possible rotation of the AFW turbine steam supply valve in order to improve access to it for manual operation.
This issue is resolved.
12.
In response to the staff concern that the integrity of equipment supports j
in the AFWS be maintained, the licensee stated that it has established a short and long term program to ensure equipment support integrity.
In the short term, a plant isometric drawing update program has been implemented which includes a piping system walkdown/inspecton to verify drawing accuracy.
For the long term, a continuous rotating schedule of safety related system walkdowns has been implemented for detection of equipment deficiencies during normal operation.
The licensee stated that the above program should identify the types of equipment sppport problems identified by the staff AF45 review team.
This issue is resolved.
1 9
l
j 13.
In response to the staff concern that the AFWS component retagging program be completed, the licensee stated that retagging was nearly complete.
This issue is resolved.
14.
In response to the staff request that the licensee identify access times for various AFWS equipment and the atmospheric dump valves which may require local manual manipulation during an emergency, tne licenste stated that time-lines developed for Appendix R (post-fire alternative shutdown) accom-plish this purpose.
Maximum access times have been incorporated into the operator training program.
This issue is resolved.
15.
In response to the staff request that the licensee pursue its unscheduled automatic reactor trip reduction program, the licensee stated that their current Transient Reduction Program provides a systematic method for pro-cessing improvement recommendations intended to decrease plant trips.
It
- e. compasses information from the programs underway for ANO-1 including the B&W Owners Group Safety and Performance Improvement Program, and those o' the CE Owners Groups Operations Subcommittee.
Several modifications l
have been identified for future incorporation in the plant to reduce trips.
l Recent operating history for ANO-2 shows a reduction in the number of trips.
This issue is resolved.
16.
In response to the staff request that the licensee complete its AFWS reliability analysis and implement changes which improve system relia-bility, the licensee stated the analysis has been completed with the ANO-2 AFWS shown to be highly reliable.
The licensee stated its position that using a plant specific data base, it has demonstrated that the system
-4 meets the SRP Section 10.4.9 numerical reliability criterion of 10 per demand.
The need for further AFWS modifications to enhance reliability is currently under evaluation by the licensee on a cost / benefit basis.
The licensee also stated its intent to develop full plant PRAs for ANO 1 and 2 in the near future which will provide insights into overall plant risk as well as AFW reliability.
The licensee concluded that the cbrrent 10
s e
a i
level of AFWS reliability was adequate.
The staff cannot concur with the licensee's conclusion as discussed further in item 17 below.
j 17.
In response to the staff position that the licensee provide an alternate method of providing water to the steam generator such as connection to the-ANO-1 AFWS, the licensee indicated that it.'isa;,reeo with the staff.
The licensee stated its position that the existing decay heat removel capability.
consisting of the main feedwater system, AFWS, "high-head" condensate pumps and "feed-and-bleed" system is adequate to :,atisfy the SRP reliability 1
goal.
The licensee also noted that several AFWS design and plant program-
)
matic changes have also contributed to improved AFWS reliability.
These include 1) the additton of a solenoid valve to redun potential AFW pump turbine oserspeed trips, 2) new (motor driven) AFW flow control valves, 3) new AFW discharge isolation valvcs, 4) a lower setpoint for low steam gene-l rator level initiation of AFW5 to reduce AFWS challenges, and 5) implemen-tation of a trip reduction program to reduce AFWS demands, fhe licensee further stated its belief that the staff's position is incon,istent with 1
other activities related to decay heat removal adaquacy such as USI A-45 and "severe accident" policy where the contribution of AFWS to core melt is being considered, and where other new %.uirements to aooreis this area may be imposed in the future.
The staff has considered thL licensee's position and as.sertions.
The staf f does not agree and concludes that impreyer, ant in the seccndary side decay heat removal capability is necessary for ANO-2.
Tha staff I
l recognizes that the improvements made by the licensee should in, prove AFWS reliability. However, AFWS reliability is still considered to be below
~4 the SRP criterion (10 per demand).
Uncertainties associated with the operator's response for initiation of "feed-and-bleed" and tN intentional release of reactor coolant into containment as a means for xitigation of a loss of all feedvater makes "feed-and-bleed" an unsatisfat ory compenuting feature.
The staff recognizes that ANP-2 has the capability to rap Mly depressurize the primary system and utilize the safaty infection pumps in 1
a feed-and-bleed decay heat removal made, but recogniling the encertainty associated with operator response in the avaiiable time frame, the staff believes it should only be considered a last resort action for decay j
l 12
s heat removal.
The staff also concludes that while use of the sndensate pumps and "feed-and-bleed" for decay heat removal is appropriate in I
emergency procedure guidance for coping with loss of all feedwater, it is not an adequate compensatory feature and does not justify AFWS reliability below the SRP value.
The staff notes that use of a condensate pump is i
not unique to AND-2. Further, even though these pumps may have a slightly higher discharge pressure, secondary side depressurization is required in order for the pumps to deliver water to the steam ge1erators as is the case in all pressurized water reactors.
With regard to the impact of GI-124 on other staff ai..tions, the staff believes that any action implemented to increase the level of AFWS reliability to achieve the goals defined in GI-124 should not conflict with USI A-45 or the "severe accident" policy.
The staff does not believe that furthes uelay in improving AFWS reliability is justified.
Therefore, the licensee should make secondary side decav heat r;moval capability improvements to increase AFWS reliability.
s issue is not l
resolved.
18.
In response to the staff request that the licensee emphasize the importance of initiating "feed-and-bleed" when required in its training program, the licensee stated that explicit guidance for recognizing and initiating "feed-and-bleed" is provided in the emergency operating procedures and j
use of the procedures in the operator training program is detailed and exten,ive.
The licensee provided a summary cf the procedural guidance for coping Hth loss of feedwater and core cooling events.
This issue is resolved.
In conclusion, the staff finds that the above issues have been resol ed with the exception of the need for additional features to improve secondary side decay heat removal reliability.
The licensee should propose appropriate compensatory features as necessary to improve the AEWS reliability of ANO-2.
12
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j l
Arl.ansas Nuclear One, Unit 2 - Auxiliary Feedwater l
System Reliability Assessment i
B.
Introduction i
This report contains the staff's assessment of the Auxiliary Feedwater System
)
(AFWS) reliability of Arkansas Nuclear One, Unit 2 (ANO-2).
This review is being done in connection with the resolution of Generic Issue (GI) 124.
GI-124, "Auxiliary Feedwater System Reliability," addresses the reliability of AFWSs.
Reliability analyses
- for AFWSs indicated that many plants fell in the high reliability range.
However, several plants fell in the lower reliability ranges.
Licensees for some of these plants implemented sufficient modifi-cations to increase their AFWS reliability to an acceptable range.
- However, the reliability of the AFWSs for seven plants, including ANO-2, remained questionable.
The six other plants are ANO-1, Crystal River, Ft. Calhoun, Prairie Island, Units 1 and 2, and Rancho Seco.
The objective of this task is to determine whether the AFWS of each of the subject seven plants is sufficiently reliable and to document any recommenda-tions for further licensee or staff actions.
This report pi
=nts the resolution approach and evaluation philosophy in Section C, and detailed evaluations in Section D.
The summary and conclusions are presented in Section A of this report.
Appendix A contains the references.
Appendix B lists the names of the NRC and licensee personnel who participated in this task.
- NUREG-0611, and NUREG-0635, Generic Evaluation of Feedwater Transients and Small Break LOCA in Westinghouse and CE Designed Plants, respectively, and NRC memoranda from A. Thadani to 0. Parr dated October 17, 1983, Octob'er 23, 1983, and November 9, 1984.
l l
20 i
o
' C.
Resolution Approach The staff believes that a high degree of availability and reliability for the AFWS can or.ly be achieved if such a system is adequately designed, properly maintained and well operated, Proper maintenance ard operating practices help reduce component failures.
These practices are enhanced by good training gronrams for the maintenance and operations personnel.
Good training programs also help the operations personnel understand the system's capabilities and its importance to safety.
System understanding reduces failure due to maloperation of equipment and improves the likelihood of recovery in case of unanticipated component failures.
C The staff believes that assessment of the above variables should provide a significant indication of the degree of reliability of the AFWS.
Therefore, the resolution approach adopted by the staff relied on an audit of several plant variables that directly or indirectly affect the availability and reliability of the AFW system.
The main items of this task include the following:
1.
Consideration of relevant information pertaining to the AFWS and support systems capability and reliability (e.g., Systems Desc"iptions, Piping and Instrumentation Diagrams, Logic Diagrams, Safety Analyses Reports, AFWS Reliability Analyses, INP0 Reports, AE00 and IE Reports, and staff Safety Evaluation Reports).
2.
Evaluation of plant operating experience with emphasis on the degree of failure repetitiveness and the rate of unanticipated automatic scrams.
3.
Evaluation of the AFWS maintenance, surveillance and testing, backlog of maintenance work requests, and ability for failure root-cause identification.
4.
Evaluation of the clarity and accuracy of the AFWS-related Emergency Operating Proredures, with emphasis on ease of recovery from fault'ed conditions, accessibility of equipment and adequacy of instrumentation and controls.
21
' 5.
Evaluation of the licensee's training programs for maintenance, operations and engineering personnel.
6.
Appraisal of the system layout, accessibility, indication and control, environment during an accident, and cleanliness, etc.
7.
Consideration of alternate plant features to maintain adequate core cooling if main and auxiliary feedwater systems became inoperable.
8.
Review of the plant post-THI modifications.
The AFWS review team audit of the above plant variables is not intended to replace other systematic staff reviews'of any of these variables but merely assesses their effects on the overall AFWS reliability and availability.
The licensee is developing an AFWS reliability analysis which was provided to the staff in draft form.
The staff has made a preliminary review of the analysis, which is discussed in Section 0.7.
The seismic qualification review
- in response to Generic Letter 81-14, Seismic Resistance of AFW Systems, was previously performed by the staff.
That review determined "that the emergency feedwater system has sufficient seismic capabil-ity to withstand a safe shutdown earthquake and accomplish its safety function."
The Appendix R fire protection staff review of ANO-2 is still in progress.
In performing this AFW reliability evaluation the staff conducted reviews of licensee-supplied documents at the NRC Headquarters, reviews of additional documents at the plant site, and numerous interviews with the maintenance, operations, engineering, and training personnel and management.
Memorandum f rom L
- 5. Rubenstein, to G. C. Laints, "Arkansas Nuclear One, Unit 2:
Seismic Qualification of the Emergency reedwater System", MPA C-14, dated June 18, 1984.
22
'The AFWS review team consisted of four multidiscipline team members and a team leader.
The team effort included three day plant site visit' and pre-visit and post-visit reviews.
The names and organizations of the NRC and licensee participants are listed in Appendix B.
D. 1 Design and Configuration 0.1.1 Approach The staff conducted a review of the design and configuration of the Arkansas Nuclear One - Unit 2 (ANO-2) auxiliary feedwater (AFW) system (ir. this report the phrases "auxiliary feedwater" and "emergency feedwater" are used interchangeably).
The staff reviewed the system descriptions in the updated FSAR, together with pertinent drawings and figures (Reference 1), and the Technical Specifications (Reference 2).
The staff also reviewed the following for additional informa-tion:
NUREG-0635 (Reference 3); and Status and Safety Analysis Reports for TMI Action Plan NUREG-0737 (Reference 4).
The staff also conducted a comparison of the ANO-2 AFWS design with the Criteria of Standard Review Plan Section 10.4.9.
A walkdown of the AFW system was con-ducted by the staff to determine the degree of ANO-2 compliance with applicable criteria and drawings.
D.1. 2 Evaluation ANO-2 is a CE designed 2815 L'(t) reactor, with two U-tube steam generators, two MFW trains, each with a turbine-driven pump, and with three motor-driven condensate pumps.
The reactor is also equipped with three high head safety injection pumps with a shut-off head of 1450 psia.
The pressurizer has two low-setpoint safety valves each with two motor operated isolation valves (low temperature overpressure protection system) and a 3-inch vent valve arrangement for use in the "feed-and-bleed" mode of decay heat removal.' The reactor is located in a large ary reinforced concrete containment.
The plant is provided with two 100% capacity diesel generators for shutdown cooling if offsite power is lost.
23
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'The ANO-2 AFW system is a two-train system.
One train contains a centrifugal pump driven by an electric motor (2P78) and the other train contains a steam turbine driven pump (2P7A).
Diversity in pump drivers eliminates common mode failures in the AFWS motive power.
The AFW system configuration is shown in Figure 1-1.
The pumps, 2P7A and 2P78, are identical.
At rated flow, each pump is capable of providing 575 gpm of water, which is sufficient for removing decay heat loads of 3 percent of rated thermal power.
The plant is also equipped with a steam bypass system.
Each of the plant's two steam lines is equipped with an atmospheric steam dump valve (ADV) upstream of the main steam isolation valve (MSIV).
This arrangement makes the dump valves operable even if the MSIVs are closed.
Each ADV is capable of removing about 11% of rated steam flow.
The AFW pumps are horizontal, multi-stage pumps with horizontally split cases, using double volutes at each stage and impellers arranged in opposed groups for hydraulic balancing.
The steam turbine driver (for 2P7A) is a single stage,.olid wheel, non-condensing, horizontal, split case Terry turbine unit.
It is designed for variable speed operation and is equipped with an electrohydraulic actuator for speed control, an overspeed trip mechanism, and an integral trip throttle valve.
It is also designed for rapid starting and will operate with steam generator pressures ranging from 1,100 psia to 60 psia.
Steam can be supplied to the tur-bine driver from either or both steam headers.
The steam supply lines are insu-lated to minimize condensation.
Also steam traps are installed on the turbine steam supply lines to continuously remove any condensate.
The turbine exhausts to the atmosphere.
Cooling water for the turbine oil cooler is piped from the pump suction line rather than from the pump recirculation line.
This is because the pressure available at the suction provides adequate cooling water flow through the cooler.
Pump and motor bearings do not require auxiliary cooling.
The electric motor driver is capable of accelerating the pump to rated speed within 4.5 seconds with 80 percent voltage and within 2.4 seconds at full voltage.
The motor can be powered from normal, preferreo,'or emergency' power sources.
25
'The AFWS is used for plant startup.
Below five percent full power, normal suction for the AFWS pumps is the startup and blowdown demineralizer effluent, with provisions for transfer to the in-use condensate storage tank or, if off-site power and the condensate storage tank are unavailable, to the Service Water System (SWS).
Above five percent full power, AFWS suction is shifted to the condensate storage tanks (CSTs), and the startup and blowdown demineralizer source is isolated by closing valve 2CV-0706.
The plant is provided with a 200,000 gallon capacity CST with a Technical Specification minimum value of 160,000 gallons.
There is also 100,000 gallons of water available from a swing CST, which is shared by both AN0 units, and can be manually lined up to either unit.
The two redundant and independent 6-inch SWS lines provide an assured source of water in the case of failure of the Seismic Category 2 condensate supply line.
The condensate supply line is furnished with two motor operated isolation valves, 2CV-0795 and 2CV-0789, and two pressure switches 2PIS-0795 and 2PIS-0789.
The service water lines are each provided with a non return check valve and motor operated isolation valves:
The normal pressure in the condensate line is about 10 psig.
If the pressure drops to 7 (+1,-0) psig, local and control room alarms will be actuated.
If the pres-sure drops to 5 (+1,-0) psig during an emergency operation, the pressure switches will automatically close the condensate line isolation valves, 2CV-0795 and 2CV-0789, and will simultaneously open the SWS valves 2CV-0711 and 2CV-0716.
The AFW pump discharge lines are provided with cross-connected flushing lines with normally closed block valves (see Figure 1-1).
These flushing lines are
{
provided in order to meet feedwater quality requirements.
An air operated valve on each flush line is automatically closed if emergency operation of the AFWS is initiated.
A pump minimum flow recirculation line is provided from each pump discharge bypassing the air operated flush valves.
Each minimum flow line is provided with an orifice and one common globe valve for limiting and regulating the flow while providing a normally open flow path to ensure that each pump has a minimum flow of 50 gpm.
The AFWS discharge piping and valving arrangement, shown in Figure 1-1, is designed to allow either pump to supply cooling water to either or both' steam generators.
Each supply line to each steam generator is provided with j
26
o I
' redundant control valves, in accordance with the single failure criteria, to ensure isolation of any faulted steam gen'rator, and feeding of the intact steam generator as required during emergenev operation of the auxiliary feed-water system following a postulated main steam or feedwater line break.
Each discharge line has a closed MOV, an open MOV, and two check valves.
In addi-tion to the check valves the closed MOVs provide an additional barrier to hot water backleakage and potential AFW pump steam binding.
The closed MOVs, thus, help avoid the steam binding problem.
However, these valves must open success-fully in order-for the AFWS to serve its safety function.
The likelihood of these valves' failure to open is directly related to the adequacy of their main-tenance.
Given two operable check valves in series and a discharge temperature check every shift (see Section D 5, Walkdown) the AFWS review team initially requested that the licensee review the risk associated with the normally closed MOVs failing to open on demand.
The licensee responded to this issue in their January 29, 1988 letter.
The staff considers this concern resolved as discussed in S.~: tion A of this report.
Feeding of the steam generators during non-emergency cooldown is by means of the electric motor driven pump and power operated control valves PCV-1025 and 2CV-1075.
It is also possible to use the steam turbine driven pump by adjust-ing the variable speed control.
The steam supply line feeding the steam tur-bine driver provides an assured source of steam to the turbine, even when the main steam isolation valves are closed.
The steam admission valves 2CV-1000 and 2CV-1050 are reormally open.
When an AFWS actuation signal is received, bypass valve 2SV-0205-2 ano the turbine cooling water solenoid valve 2SV-0317-2 will open, and the turbine will reach idle speed, allowing the turbine governor hydraulic system to pressurize.
Fifteen seconds later, the main steam valve 2CV-0340-2 will open, 2SV-0205-2 will close, and the turbine speed comes under the governor control.
Upon receipt of the automatic actuation signal the four closed discharge MOVs receive signals to open if their respective steam generators are intact.
27
4
'During plant startup, the AFWS is placed in operation until sufficient steam is available to run the main feedwater pump turbine drivers and the condensate and feedwater system is placed in operation.
This, in effect, is a full flow test of the AFWS before it is put in the automatic standby mode.
During hot standby and hot shutdown, the AFWS is placed in operation tc maintain steam generator level.
The average feedwater flow requirement in this case is about 320 gpm, which is within the rated capacity of a single AFW pump (575 gpm).
During normal cooldown, the AFWS is placed in operation to maintain steam genera-tor level during decay heat removal operation until the RCS temperature is brought down to about 350 F, at which time decay heat removal is switched over to the Shutdown Cooling System (SCS).
The operation of the AFWS in this mode lasts about 3.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at an average flow of 510 gpm.
The licensee noted that it is their intent to share the new, partially tornado-protected, CST (originally installed for ANO-1 use) with ANO-2 and to eliminate the automatic switchover to the service water system.
While the new CST will provide additional water inventory, this arrangement will also minimize inadver-tent admission of low quality service water into the steam generators because of spurious signals as has occurred in the past.
The AFW pumps start automatically on a steam generator low level signal.
However, if the low level signal occurs because of a break in the steam line or the feedwater line as would be evident by low pressure in one steam genera-tor, the closed MOVs do not receive a signal to open and the open MOVs receive a signal to close for the steam generator with the lower pressure.
The essential valves in the turbine driven pump system are all powered from a de source since this system is required to operate in the absence of all ac power.
This includes all motor and solenoid operated valves at the suction and disciange sides of the pump, with the exception of steam supply valves 2CV-1000, and 2CV 1050, and condensate supply valve 2CV-0707.
These valves are ac operated, but are normally open and do not change positions upon AFW actuation.
28
In the event of a complete loss of the AFW system, other methods are available to remove the decay heat, including (1) use of the condensate pumps, and.(2)
"feed-and-bleed." These decay heat removal methods are discussed in detail in Section D.3, Emergency Operating Procedures, of this report.
D.1. 3 Conclusion On the basis of this evaluation, the staff concludes that the AFW system in the ANO-2 nuclear power plant complies with the applicable criteria of Section 10.4.9 j
of the Standard Review Plan, including single failure considerations (this also 1
addresses staff concerns raised in GI-125.II.1.b, with respect to single fail-ure protection for existing AFW systems), and including the guidelines of NUREG-0737, Item II.E.1.1.
The staff notes that use of the condensate pumps and the "feed-and-bleed" capability are effective means of decay heat removal and enhance the plant's overall capability of decay heat removal.
However, as indicated previously, "feed-and-bleed" is appropriate only as a last resort for decay heat removal in emergency procedures in the event of a loss of all feeduster.
D2 Maintenance, Surveillance and Testing D.2.1 Approach The reliability of a system depends to a large extent on the maintenance pro-grams applied to such a system.
The adequacy of system maintenance will be reflected by component failure rates during operation, surveillance and testing.
The AFWS evaluation team performed a detailed audit of the AFWS-related mainte-nance and surveillance procedures and surveillance results.
Members of the team interviewed maintenance and operations personnel and management on the practices and organization of the maintenance programs.
Training of the maintenance and operations personnel was also explored, and is further discussed in Section D.6, Training.
29
i l
ANO-2 is one of several plants for which the NRC staff evaluated the maintenance program practices.
The staff completed its evaluation report, "Technical Letter Report, Site Survey of Arkansas Nuclear One Unit-2 Maintenance Program and Practices," and issued it in June 1986.
This technical letter is provided in NUREG-1212 (Reference 5).
In this reference the staff presents its evaluation of five areas at ANO-2.
These are:
Organization and Administration, Facilities and Equipment, Technical Procedures, Personnel, and Work Control.
NUREG-1212 lists the ANO-2 maintenance program as "Good and Improving." The NUREG lists several significant observations.
The ones most relevant to the subject of this report are listed below:
(1) Maintenance Department Work Control - This included a description of the newly implemented Work Control Center (WCC), and the Station Information Management System (SMIS), both of which are improvements over previous practice.
The WCC and SIMS are discussed below.
(2) The use of procedures and setting of formal management goals were found to be good and highly developed.
(3) Maintenance backlog due to material shortages was rare.
(4) Component labelling for ANO-2 was not started at the time of staff review.
The potential for mislabelling of equipment was expressad as a staff
)
concern.
(5) ANO-2 was reasonably staffed to minimize maintenance backlog.
(6) There was a formal training program in place.
This is discussed in detail in Section D.6, Training.
This section and Section 0.6, Training, of this report supplement NUREG-1212 as it pertains to the ANO-2 AFWS.
The AFWS review team endorses the NUREG-1212 findings as supplemented by this report.
30
e
'O.2.2 Evaluation The ANO-Unit 2 maintenance program is basically composed of two parts.
The first part is the periodic maintenance (or, preventive maintenance), and sur-veillance and testing.
These are identified in the plant operating procedures, maintenance procedures, and Technical Specifications.
These assure that the AFW system and its support systems are kept operable and capable of reliably i
performing their intended function.
The second part of the maintenance program is the corrective maintenance.
The responsibility si any plant worker is to spot potential system or component failures and report them.
The maintenance program revolves around the work control center (WCC).
This system processes, monitors, and controls the maintenance job until it is com-pleted.
The process works as follows (refer to Figure 2-1 for job order flow path):
When a problem is suspected a job request is v,dtiated.
The job request identi-fies the component by number, the problem (how the component failed), the loca-tion, and any other pertinent information.
The jcb request then goes to the "schedulers" in the WCC.
The "schedulers" check whether the job has already been identified.
If the job has not been previously identified, the job request is then forwarded to the "planner." The "planner" screens the problem, at the component location if necessary, evaluates any previous events involving the component, and writes a job order.
The job order fully describes the component, the work to be done, and other pertinent information.
The "planner" alto assigns the job priority to one of the following categories:
A.
Emergency B.
Priority 1 - to be done quickly C.
Priority 2 - to be done within 2 days D.
Priority 3 - to be done within 3 weeks Priority 2 and 3 may be done during refueling outage depending on the job.
31
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'The "planner" then returns the job order back to the "schedulers" to schedule the job and coordinate the work with "health physics," the parts expediter, and the outage schedule for the plant.
The package is then sent to the craft super,-
intendent for assignment to a craftsman to perform the work.
The craftsman per-forms the work in accordance with the job order and reports the status of the work back to the "schedulers" at the end of each shift.
When the job is com-pleted the paperwork is returned to the "planner" for review and evaluation of the job order and the work.
Finally it is sent to the "work closing" group which is responsible for updating the maintenance history.
This group evaluates the job order for correctness, completes a failure report which may be input in the nuclear plant reliability data system (NPRDS), and updates the component maintenance data base.
This data base is an important part of the station information management system (SIMS).
The "work closing" group then forwards the paperwork to "records."
A root cause analysis is performed throughout this process.
The job requester does a preliminary root cause analysis when the job is requested.
The planner continues the root cause analysis by evaluating the job and the maintenance and f ailure history of the component in the preparation of the job order.
The craft superintendent contributes to the root cause analysis when he receives the job order package which includes a repetitive failure printout on the com-l ponent.
The root cause analysis program also includes evaluation of unplanned trips and transients by the significant events review committee (SERC), and evaluation of reportable abnormal conditions (RACs) and licensee event reports (LERs).
Since the group of people evaluating a component failure may change from failure to failure, the staff finds that no one particular person is responsible for the AFWS operation or failure evaluations.
This means that the root cause analysis for AFWS failures may rely on the general knowledge and expertise of the person or persons who happen to review the failure.
However, the newly operational SIMS data base provides a promising tool for failure trending analysis which is a significant factor in root cause determination and analysis.
The SIMS data base, however, has been in operation for less than a year and equipment' 33
)
' histories are not being entered into the system retroactively.
Therefore, the benefits gained from such a system in root cause analysis may be more apparent as equipment data accumulates.
The usefulness of the SIMS data base for this purpose will depend to a large degree on the accuracy and amount of detail of the equipment history that is entered in the system and subsequently retrieved for use.
The review team believes that as the SIMS data base increases, and if the system is properly used for failure trending analysis, a significant increase.
in AFWS reliability and availability will result.
The evaluation team has the following additional observations on the maintenance, j
surveillance and testing program:
1.
All maintenance personnel hired by the plant start at the entry level and must go through the Arkansas Power and Light Company (AP&L-the owner of Arkansas Nuclear One plants) training program.
This assures uniformity in training of maintenance personnel and their knowledge
)
of the systems and procedures, l
2.
If a job is unfamiliar or new, a pre-job meeting is held by the main-tenance crew to discuss the procedures required and any special instructions.
In addition, for a design change, the work package would specify any additional or specific training necessary for the job.
These are positive features in the maintenance practices.
3.
The licensee stated that the rate of maintenance staff turnover is low (between 3 - 6% per year).
In previous years this rate has been high.
The change is a6tributed to an aggressive program to retain plant workers which includes hiring more local people, training, com-petitive salaries and job security.
The staff believes that a reduc-tion in staff turnover rate leads to an expertise enrichment, which in turn leads to enhancing systems availability.
4.
Plant officials take pride in their high rate of early reporting of incipient failures.
The plant has a mandatory pr'eventive mai'ntenance program for safety related equipment which goes beyond the normal technical specification requirements.
34
- 5.
The periodic maintenanc: procedures, surveillance procedures and system operating procedures for the AFWS and its components were of sufficient detail to properly complete the task.
However, the staff found some discrepancies in procedures for similar components.
Examples of these discrepancies include:
cautionary notes, equipment and parts needed for certain tasks, incomplete procedure statements, and undefined setpoints or tolerances existed in one procedure but not in another.
These items were brought to the attention of the licensee who explained that he was in the process of revising all of the procedures and that this revision would be completed by the end of 1987.
The licensee also explained that some discrepancies may be due to the fact that different people revise different sections for similar equipment.
The procedures revision includes:
(a) making the procedural steps shorter and simpler (e.g., one action per step);
(b) revising the format and content; (c) checking the procedure against the component technical manual to assure accuracy; and (d) including in the maintenance procedure a requirement that the first line supervisor go over the procedure with the craftsman prior to commencing work.
6.
In response to IE Bulletin 85-03, with respect to motor-operated valve (MOV) limit, torque, and bypass switch setting, the licensee is under-taking an aggressive program in testing and calibrating its valves (many with full operating AP across the valves) using the motor operated valve analysis and test system (MOVATS) methodology.
In this methodology, mea-surement of changes ir, the electric current required to operate the valve motors during the valve travel are interpreted to indicate the various stresses on the valve motor, valve stem, and valve seats or gates.
These electric current measurements can also be used for early detection of degrading valve conditions (e.g., insufficient valve lubrication.
broken stem threads, etc.).
The AFWS review team believes that the licensee's approach and commitment to valve testing and calibration adequately address the concerns raised by GI-122.1.a. and c with respect tu MOV failures and AFW flow interruption.
This approach by the licensee enhances the reliability of the MOVs when called upon to operate, which i
)
j 35 1
in turn enhances the AFWS reliability.
The licensee's response to IE Bulletin 85-03 with respect to motor operated valve settings is still ender staff review at the time of this report writing.
7.
The licensee stated that the findings in the ANO, Unit 1 safety system functional inspection (SSFI) report (Reference 6) relevant to ANO, Unit 2 have been implemented.
Examples of the SSFI report findings included:
inadequate torque switch, and torque switch bypass settings; valves that were required to be periodically tested were not tested; controlled design documents and drawings contained errors and omis-sions.
The AFWS review team notes, however, that the ANO-2 FSAR continues to have errors and discrepancies.
0.2.3 Conclusions The staff review of the ANO-2 maintenance, surveillance and testing was somewhat limited due to the major procedure revision effort underway.
The staff reviewed the latest revisions of selected AFWS procedures and relied mainly on the licensee's statements in describing the procedure revision effort.
Based on our review of the maintenance, surveillance and testing, and based on the licensee's statements, and the status of the procedures revision effort, the staff notes that the licensee appreciates the significance of important maintenance related elements like work control, failure trending, root cause determination, and record keeping.
The AFWS review team, therefore, expects that upon revision completion, the maintenance and surveillance program will enhance the AFWS reliability.
D.3 Emergency Operating Procedures D.3,1 Approach The staff reviewed the emergency operating procedures which included the overall emergency procedures and the abnormal operating procedures relating to.the AFW system (References 7, 8).
The AFWS review team's objective was to determine whether these procedures can easily be followed and understood and whether they 36
o o
enhance the plant's overall capability to remove decay heat under emergency conditions.
The licensee conducted a limited simulator demonstration of a loss of feedwater event.
D.3.2 Evaluation The ANO-2 Nuclear Power Plant emergency operating procedure (E0P), E0P 2202.01, deals with various accident conditions.
The conditions which initiate use of E0P 2202.01 include automatic or manual reactor trip or indications of a steam generator tube rupture.
The reactor may be tripped manually when an operator senses a condition that may damage the plant or pose injury to personnel.
The operator may also trip the plant because of a failure of the reactor protection system (RPS) to function when reaching any of its setpoints.
When responding to an emergency condition the operator proceeds to the propee section in E0P 2202.01, either "Reactor Trip" or "Steam Generator Tube Rupture Within Charging Pump Capacity" wherein further directions are provided for recovery from these conditions.
The reactor trip section may direct the opera-i tor to other sections as appropriate.
The other E0P sections are contained in the same document and include:
l 1.
Degraded Power, 2.
Blackout.
3.
Overcooling, 4.
Main Steam Isolation, 5.
Safety Injection Actuation, 6.
Steam Generator Tube Rupture Within Charging Pump Capacity, 7.
Steam Genorator Tube Rupture Greater than Charging Pump Capacity, 8.
Inadequate Core Cooling, and 9.
Emergency Reactivity Control.
Any section iay then refer to et.her sections or to other actions required in
)
the recovery process.
All of these sections refer to the operation of the AFW l
system, either directly in the section, or indirectly by referring the operator to another section.
j 37 s
4 0
4 The licensee uses a "Procedures Writers Guide" for reference when writing a procedure.
The procedure is then independently reviewed for completeness and accuracy.
Procedure validation is then performed using one or more of the following tools:
(a) round-table discussion by experienced operators; (b) procedure walk-through., and (c) actually conducting the procedure.
Before the procedure is implemented it must be reviewed by the Plant Safety Committee.
1 The emergency procedures for ANO Unit 2 also include abnormal operating proce-dures which deal with events of lesser severity than does E0P 2202.01.
Those which deal with the emergency feedwater system include 2203.120 (Annunciator 2K04), 2203.12F (Annunciator 2K06), 2203.12G (Annunciator 2K07),
and 2203.13 (Natural Circulation Cooldown), Reference 8.
When specific windows in the annunciator are activated, the operator reverts to the appropriate abnormal operating procedure to find the instructions to clear the fault.
The review team has the following observations:
1.
At the beginning of the procedure there is a procedure entry condition (PEC) list that clearly defines when that procedure is entered and lists the RPS setpoints.
The operator is instructed to manually trip the reactor if any of the RPS setpoints is exceeded, and automatic trip had not occur-red.
The PEC list is followed by an immediate actions list.
This list is a reminder to the operator of the expected ranges of plant parameters.
The team finds that providing thes) lists at the front of the E0P is a I
very useful practice, l
2.
The E0P section dealing with degraded power provides useful estimates of the condensate water requirements at different timas following a loss of offsite power.
Also, useful instructions are provided to monitor the emergency diesel generator fuel oil inventory, and economize the station batteries loading.
l 3.
Useful notes, cautions, and direc+. ions are frequently,used thro 1ghout the 4
l i
E0P.
These alert the operator that under certain degraded condit ons l
control room indications may be inaccurate, explain certain anticipated l
38 i
1
equipment performance, and instruct the operator on how to reset and operate equipment under degraded conditions, if necessary.
4.
The E0P clea.ly indicates that if the AFW flow cannot be established the operator should attempt to restore T.he main feedwater (MFW) flow.
These instructions are provided in several locations of the E0P.
Additional guidance and precautions are provided to the operator so that he may start the MFW pumps under degraded conditions, if necessary.
Reminders are also provided that the NFW pumps may be run as long as sufficient steam exists in the steam generators.
If starting the MFW pumps is unsuccessful the operator is instructed to attempt establishing the condensate pump flow.
If that, in turn, fails the operator is :learly instructed to "feed-and-bleed." This is accomplished by manually initiating the safety injection, verifying maximum safety injection and charging pump flow, then opening the emergency core cooling system (ECCS) vent valves on top of the pres-surizer.
If the vent valves fail to opan the operator is instracted to open the low temperature overpressure protection (LTOP) valves and the pressurizer and reactor vessel high point vent valves.
The vent valves will reduce the reactor pressure sufficienti,' to allow the high head a fety injection (HHSI) flow into the reactor vessel.
These valves are schematically shown in Figure 3-1.
5.
The licensee stated that the ECCS vent valves have teen full op tasted using the M0 VATS methodology (See Section D.2).
The licensee ststed that the motor torque capability and the torque switch bypass settings for the LTOP valves are such that these valves car opn against the full system differential pressure.
However, the staff notes that neither the vent valves nor the LTOP valves are in the plant's Techni.al Specifications j
(TS).
Therefore, the licensea should prepose for staff review ppropriate
{
fechnical Specifications for these vrb es.
The licensee responded to this issue in their January 29, 1988 letter.
The staff considers this concern resolved as discussed in Section A of thit report.
1 6.
Each of two HHSI pumps wil' deliver 3TC gpm at about l'250 psig.
A' third, similar capacity HH5I pump is on st3ndby and can be started manually it one of the operable pumps bc.comes uncvailable.
In addition, each of three, safety grade, charging punps :an deliver 44 gpm at pressures up to i
39 1
O a
2300 psia.
Two of the changing pumps are required to be operable by the olant's TS when the reactor is operating.
Only one safety injection pump is sufficient to remove decay heat.
Since a single failure of tne vent valve 4740-2 can disable the veat path and an LTOP valve (see Figure 3-1) the licensee is requested to verify whether one LTOP valve is capable of full decay heat removal via "feed-and-bleed".
The ANO-2 "feed-and-bleed" system is unique in that it provides a dedicated 3-inch diameter flow path with two series motor operated valves.
The review team believes that, with verification of the LTOP valve capability, as discussed in Sectiti A, Item 3, the ANO-2 "feed-and-bleed" system constitutes an alternate decay heat removal means, and therefore supplements the AFWS in decay heat removal.
Howeser, as indicated previously, because
]
cf uncertainties in operator action associated with it: ini'.iation, it serves only as a suitable last resort fcr decay heat removai but is not an appropriately reliable compensatory feature.
It does serve as a supplement to the AFWS for decay heat removal as a last resort in emergency procedures.
This is discussed further in Section A of this report.
7.
At the end of the E0P there are useful figures and appendices th;t may be used for quick reference during an emergency.
However, the AFW review team also has the following additional observations:
8.
We noted some minor apparent inconsistencies between similar steps in the E0P.
9.
In several locations of the E0P the operator is di~ected to do certain actions if a parameter was "abnormal." Further guidance, in the E0P, to define what is "abnormal" may help avoid costly mistakes under emergency circumstances.
What constitutes "abnormal" should be adequately discussed during training.
10.
The inadequate core cooling section of the E0P (Step 2.C MFW), ins'trt. cts the operator to attempt to start the motor driven AFW pump (2P78).
No instructio, addresses the turbine driven AFW pump (2P7A).
Clear instruction as.to whether or not to use 2P7A should be included.
40
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'uring our review the licensee calculated an ANO-2 specific steam generator dryout time of about 19 minutes, and hot leg saturation time of about 52 minutes using CEPAC, a PC-based CESEC (a CE transient code)..The licensee also stated that the "feed-and-bleed" mode of decay heat removal will be effective if it is started anytime up to hot leg saturation time.
- However, since there may be a natural tendency for the operator to resist initiating "feed-and-bleed," which is, in effect, a small break LOCA, the licensee should specify in the opvator training material a time limit window after the "feed-and-bleed" Londition is reached (the E0P specifies this as T 560 F and increasing).
The operator must initiate "feed-and-bleed" before c
this time limit has expired or if T rises excessively, whichever occurs first, c
or the risk for core uncovery and possible fucl failure will be too great.
Guidance for what constitutes excessive rise of Tc should be provided in the training program.
The review team believes that the above measures will also resolve staff concerns raised by GI-122.2, with respect to initiation of "feed-and-bleed."
D.3.3 Conclusion j
l The staff found that, with the exception of the above observations, the ANO-2 emergency operating procedure is generally well organized and provides clear instructions, notes, cautions, and directions.
As indicated in Section A of this report, the licensee hu provided satisfactory responses to these concerns, and therefore, the staff concludes that, the AN0-2 emergency operating procedure enhances the AFWS and alternate systems reliability and the overall plant capability for decay heat removal.
D. 4 Instrumentation and Control D.4.1 Approach fhe review team audited the instrumantation and control circuits for the Auxiliary Feedwater System (AFWS).
As part of that audit we conducted a walkdown of the as built system and compared it to the design in the following areas:
1.
Review electrical instrumentation and control system drawings O r confor-mance to IEEE Standard 279(i.e., Independence, Single Failure, Autoatic and Manual Initiation and NUREG-0737, Item 11.E.1.2.
42
~. _
l
'2.
Verity from the control room and drawings that the operator has indication for normal and abnormai conditions (LOMF, LOOP) in accordance with the i
Emergency Procedures and Regulatory Guide 1.97, "Instrumentation for Light, Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident."
D.4.2 Evaluation AN0-2 has a Class 1E Auxiliary Feedwater (AFW) system consisting of one turbine driven pump (2P7A), one motor driven pump (2P78), and two independent trains, each capable of supplying water to either of the two steam generators as shown in Figure 1-1.
Each pump train has two parallel flow paths to allow a given pump to feed one or both steam generators.
Each flow path in turn has two elec-trically operated valves in series.
The first valves in the electric motor driven pump trains are normally closed and are used for feed flow control.
The second valves are normally open.
The first valves and the motor driven pump are powered from Class 1E, diesel generator backed power supplies (power train I-RED).
The second valves are powered from the redundant Class 1E, diesel generator backed power supplies (power train II-GREEN).
The turbine driven train also has two parallel paths from the pump discharge and each path has two valves in series.
The first set of valves (normally closed) are powered from the Class 1E DC power system (power train II-GREEN)
The second set of valves are normally open and are powered from the redundant DC power system (power train I-RED).
The turbine steam admission and control valves are powered from a Class 1E battery backed DC system.
The turbine driven train is independent of AC power.
Therefore, upon loss of all AC and a demand for AFW system, the turbine driven train can supply water to the steam generators.
43
e 1
'The normal water source for the AFW system is the 200,000 gallon (with a Technical Specification minimum value of 160,000 gallons), non-seismic condensate storage tant (CST).
There is another 100,000 gallon swing CST that is shared between the two AN0 units.
The secondary source of water is the Seismic Category-: service water system (SWS) whose supply is either the emer-gency cooling pond or the Dardanelle Reservoir.
The AFW system is designed i
so that the supply is automatically shifted from normal to secondary on low AFW pump suction pressure (5 psig).
The instrumentation and control power supplies to s mport the ANO-2 configuration consists of a two train system each associated with a separate diesel generator.
Each train provides power to a 480V motor control center (MCC) vital bus.
The redundant MCC provides the AC power for valve control in the motor driven pump train.
The DC power for the turbine driven train comes from redundant DC buses each having its own battery charger.
There is a thire standby battery charger which can be switched by operator action to replace either normal battery charger.
The initistion signals and circuits are part of the Engineered Safety Features Actuation System (ESFAS).
The ESFAS consists of sensors, logic and actuation circuits which monitor steam generator water level and pressure.
If these para-meters reach preselected setpoints, an emergency feedwater actuation signal (EFAS) is initiated.
The EFAS is initiated to Steam Generator 1 either by a
{
low steam generator water level coincicient with no low pressure trip present on Steam Generator 1, or by a low steam generator water level coincident with a differential pressure between the two steam generators with the higher pressure in Steam Generator 1.
The EFAS for Steam Generator 2 is identical to that of Steam Generator 1.
The actuation signals are derived from 2-out of-4 coinci-dence logic measurement channels.
Each of the four channels have a 120-volt uninterruptable AC power supply system which consists of four inverters and four distribution panels.
Each inverter has three sources of power:
- 1) the normal Class 1E MCC source, 2) the battery bank as the backup source, and 3) a second Class 1E MCC as the emergency standby sources.
44
o o
When an actuation signal is received from the EFAS, both pumps will start and simultaneously all valves in the discharge lines will open, unless there is a steam'or feedwater line break, then the EFAS logic will open only the valves leading to the intact steam generator.
There is a time delay of 90 seconds arter the EFAS signal and before the motor driven pump can start.
This start-ing delay ensures that the turbine has accelerated to full speed and under governor control to preclude drawing suction pressure excessively low (cavita-tion point) which could cause the turbine pump to oversped and trip Based on our review of the AFWS, we conclude that the system is designed with due consideration of safe failure modes if conditions such as system disconnec-tion or loss of power are experienced.
Therefore, we find that the AFWS satis-fies the requirements of GDC-23, "Protection System Failure Modes." We also conclude that the design provides the necessary instrumentati a to sense acci-dent conditions and anticipated operational occurrences.
This instrumentation as designed will actuate the AFWS.
Therefore, we find that the AFW system satisfies the requirements of GDC 20, "Protection System Function."
A..,-2 instrumentation provides information for the operator's use during all modes of normal operation, including operational transients, and enables the operator to verify safety system performance following an accident and manually perform required safety functions.
The information includes indications, records (for level, pressure and flow), status lights for pumps, valve position indi-cation, annunciators and alarms.
Table 4-1 lists the information available at the control room.
Table 4-2 lists information available at the equipment locations.
The scope of our review audit included tables of system variables and component states to be indicated, functional diagrams, electrical drawings, emergency procedures and submittals on conformance to R.G. 1.97 (Reference 9).
Since the errangement, clarity, and layout of control room instrumentation is adequately addressed by the staff in the Detailed Control Room Design Review (DCRDR) as required by NUREG-0737, the review team did not actively pursue that aspect of the control room instrumentation.
However, the team had concerns about the 45
j
' locations, orientation, and clarity, especially under emergency lighting conditions of the instrumentation at the equipment locations.
These concerns are discussed in Section D.5, System Walkdown.
0.4.3 Conclusion j
Based on document reviews, licensee statements, and direct observations the staff determined that adequate independence between trains was provided such that any single failure of components in a train will not prevent the other train from completing its safety function.
In addition it was demonstrated that the automatic and manual initiation signals and circuits for the AFWS also comply with the single failure criterion.
Based on the audit, the staff concludes that the ANO-2 AFWS can perform its intended function, and that it conforms to the design basis requirements for IEEE Star.dard 279, 1971.
Tne staff also concludes that the information available to the operator includes appropriate variables and that their range is consistent with the guidelines identified in R.G. 1.97.
Therefore, the staff finds that the information systems satisfy the requirements of GOC-13, "Instrumentation and Control" for monitoring variables and systems over their anticipated ranges for normal operation, anticipated operational occurrences, and accident conditions.
With the exception of staff concerns identified in Section D.5 of this report, the staff concludes that the AND-2 Instrumentation and control provisions are adequate and enhance the AFW system reliability.
46
Table 4-1 AFWS Information Available to the Operator in the Control Room 2P7A Steam Driven Pump Indication Lube Oil Temp 50 -250 F Speed (2K03) 0-5000 RPM Discharge Press 0-2000 PSIG 2P7A+S/G B Flow 0-750 GPM 2P7A+S/G A Flow 0-750 GPM Main Steam Pressure to 2P7A 0-2000 PSIG 2P7A Controls (With Indication)
Steam Bypass Valve Position Open/ Closed 2K03 Seal Water Open/ Closed Steam Supply Valve Open/ Closed Main Steam to Turb, Open/ Closed Main Steam to Turb.
Open/ Closed 2K03 Speed Controller "Dailed" Speed Output Signal-0-100%'
2P7A45/G A Valve-1026 Open/ Closed YP7A+S/G A Valve-1037 Open/ Closed 2P7A+S/G B Valve-1076 Open/ Closed 2P7A+S/G B Valve-1039 Open/ Closed Condensate to 2P7A Open/ Closed Service Water to 2P7A Open/ Closed Room Cooler Service Water Open/ Closed 2VUC 6A Rooni Cooler On/0ff 2P7A Annunciators Turbine Overspeed Trip Suction Pressure Low Discharge Pressure High/ Low I
A S/G Flow High/ Low B S/G Flow High/ Low Service Water Pressure Low Lube Oil Temp. High/ Low EFAS Overriden 2P7B Electric Driven Pump 2P7B Discharge Pressure 0-2000 PSIG 2P7B-SG A Flow 0-750 GPM 2P78-SG B Flow 0-750 GPM 47
j Table 4-1 (Continued) 2P7B Controls With Indication Indication 2P7+SG A Valve-1038 Open/ Closed 2P78+SG B Valve-1036 Open/ Closed 2P7845G A Modulative Valve-1025 Open/ Closed and 0-100%
Open 2P7845G B Modulative Valve-1075 Open/ Closed and 0-10C%
Open Condensate to 2P78 Open/ Closed Service Water to 2P78 Open/ Closed 2P78 Start Switch On/0ff/ Springs Charged EFW Flush Valve A Open/ Closed EFW Flush Valve B Open/ Closed SG A/SG B Wide Range Level 20-460" Room Cooler Service Water Open/ Closed 2VUC-6B Room Cooler On/0ff 2P7B Annunciators 2P7B Overload Breaker Trip Suction Pressure Low Discharge Pressure High/ Low A S/G Flow High/ Low B S/G Flow High/ Low Service Water Pressure High/ Low Additional Instrumentation (Water Sources)
Service water Header #2 Pressure 0-200 PSIG Service Water Header #1 Pressure 0-200 PSIG CST A/B Level 0-100%
Q-CST Level G-100%
Additional Instrumentation (Steam Generators)
S/G Narrow Range Level (4 Channels Per Generator) 0-100%
S/G Narrow Range Level (2 Channels Per Generator) 0-100%
S/G Wide Range Level (2 Channels Per Generator) 20"-460" S/G Pressure (4 Channels Per Generator) 0-1200 PSIG S/G Narrow Range Level 0-100%
S/G Pressure 0-120.0 PSIG 48
a e -
l Table 4-1(Continued 1 Additional Annunciators Service Water pump (A, 8, C) Discharge Pressure High Service Water Header (1,_II) Discharge Pressure Low CST (A, B) Level Low-Low CST (A, B) Temperature Low e
4
[
49 1
Table 4-2 AFws Information Outside Control Room
\\
2P7A Turbine Pump Room Indication S.W. to 2P7A Pressure 0-160 PSIG 2P7A Suction Pressure 0-100 PSIG 2P7A Discharge Pressure 0-2000 PSIG 2P7A Disch. Header Temp.
0-240 F 2P7A Speed 0-5000 RPM 2P7A Governor Oil Pressure 0-60 PSIG 2P7B Motor Pump Rooni S.W. to 2P78 Pressure 0-160 PSIG 2P7B Suction Pressare 0-60 PSIG 2P78 Discharge Pressure 0-2000 PSIG 2P7B Discharge Header Temp.
0-240 F Penetration Rooms S/G A Level 0-100%
S/G B Level 0-100%
50
o o
0.5 System Walkdown D.5.1 Approach A key part of the staff's site visit to ANO, Unit-2 was an AFWS walkdown.
The walkdown afforded the staff the opportunity to examine the as-built system-configuration, specific components, and potential undesirable interactions.
The system walkdown had two main objectives, one was to confirm that the installed system conformed to the staff's understanding of the system design basis as identified in previous evaluations, and to determine if the system may be sub-ject to common mode failure mechanisms or hazards (e g, flooding, fire, mis-siles suction strainers, etc.).
The other main objective was to examine the ease of operator access to vital equipment for performing necessary ecovery actions.
This includes assessment of emergency local lighting, communications and other factors (e.g., cleanliness, equipment labeling, use of locking devises, posting of simple instructions at equipment locations, etc.).
The walkdown covered the entire pipinn and component layout from the condensate storage tanks, through the pumps to the containment penetration and included the turbine driven AFW pump steam supply lines, switchgear, and the instrumen-tation and control.
D.5.2 Evaluation The system walkdown, in addition to confirming the "as built" configuration of the system, was also used to determine the location of the AFW system motor operated valves (MOVs).
Determining the AFW system valve locations was impor-tant in verifying whether the valves had been properly qualified, that is environmentally qualified and Q-listed.
Prior to the site visit the staff reviewed the limitorque motor ope"ated valve maintenance procedures (procedure numbers 1403.160 for model SMB-000 and 1403.161 for model SMB-00, Reference 10), which identified whether a valve was environmentally qualified or Q-listed Based on this review and the walkdown the staff determined that the AFWS motor operated valves listed in Table 5-1 were either not environmentally qualified, not Q-listed, or both.
The status of these v.alves was discussed with the licensee.
51
~
1
'For the Q-listing of valves on Table 5-1, the licensee stated that all AFW system MOVs were qualified.
The licensee also stated that the component Q-list but not the maintenance procedures (1403.160 and 1403.161), is the controlling factor in determining whether a valve is qualified (Q-Listed).
The staff inspected the Q-List and verified that valves 2CV-0,: 1-2 and 2CV-1025-1-were on the list.
In order to avoid discrepancies and confusion, the licensee should update procedures 1403 160 and 1403.161, and other similar procedures, to conform to the Q-List.
For the environmental qualification of valves on Table 5-1, the licensee stated that an analysis was performed in 1977 to determine temperature conditions in the AFW pump rooms following a unit cooler failure.
For the worst case condi-tion in the turbine diiven AFW pump room, the temperature after th ee hours operation would approach 115 F.
Based on this analysis it was determined that the valves did not need to be environmentally qualified.
The analysis was not available for the AFWS team review.
Thus, the staff could not verify the assumptions used, initial conditions or other pertinent parameters, which could have an effect on the room temperature.
The licensee stated that this analysis was performed prior to plant fire protection modifications (fire doors have subsequently been installed on pump room entrances' These modifications could have an impact on the analysis.
The licensee was aquested to reverify his analysis and the adequacy of the environmental qualification of these valves.
Part of the system walkdown included an evaluation of the plant communication and lighting systems in the AFWS rooms.
The communication system for this area consists of a telephone system, the plant public announcement (PA) system, and hand-held radios.
The AFWS review team noticed a broken handset on the PA system located in the hallway outside the AFW pump rooms.
Although the telephone system and hand-held radios act as redundant communication means, the staff urges the licensee to maintain all available communication means, including the PA system, in a ready and operable condition.
The lighting system for this area consists of the normal and emergency AC fluorescent fixtures, ar.d bat:e'y pack lighting.
The lighting was found to be inadequate particularly during the postulated station blackout scenario when 52
e Table 5-1 "NOT QUALIFIED" AFW MOTOR OPERATED VALVES Not Not Valve Identification Description EQed Q-Listed 2CV-0789-1 Pump 2P-78 Condensate Suction MOV X
2CV-0795-2 Pump 2P-7A Condensate Suction MOV X
2CV-0711-2 Pump 2P-7A Service Water Suction MOV X
X 2CV-0716-1 Pump 2P-78 Service Water Suction MOV X
L 2CV-1529-1 Service Water to AFW Room 2P7A unit cooler MOV X
2CV-1532-1 Service Water to AFW Room 2P78 unit cooler MOV X
2CV-1025-1 Pump 2P-78 Discharge to SG-A control valve X
C 53 l
' proper operation of the turbine driven AFW pump is imperative.
The emergency DC lighting for the turbine driven AFW pump room is powered from the battery pack located in the motor driven AFW pump room.
A battery failure due to an accident could result in no lighting in either room.
The review team noted that the hallway leading to the two AFW pump rooms was dark with potentially hazardous stairway steps especially under stressful emergency conditions.
The review team did not observe any normal or emergency AC lighting in the hallway.
As a part of the system walkdown, the staff reviewed the emergency lighting dia-gram for the AFW pump rooms area (see Figure 5-1) and the surveillance procedure that is performed quarterly on the emergency lighting (Ref. 11).
A review of the surveillance procedure and the system walkdown showed that the aiming of the emergency lighting may not be correct to provide adequate illumination to perform manual actions.
This is particularly true for the turbine driven pump Since this room is crowded and access to most valves is difficult, even room.
under normal conditions, adequate illumination is necessary.
According to Table 1 of Reference 11 the lighting in this room is aimed toward the west wall the opposite wall from where the lights are mounted.
This would not provide adequate illumination for the trip and throttle valve, the local control panel, or the condensate water supply valve, 2CV-0795-2.
The trip and throttle valve is located north of the lights, and behind some pipes when seen from the location of the lights.
The local control panel faces the west wall and may receive an inadequate amount of indirect lighting by reflection.
The condensate water sup-ply valve 2CV-0795-2 is located in the southwest area of the room and may receive an inadequate amount of indirect lighting.
It is the AFWS review team's assess-ment that the lighting especially emergency DC lighting is inadequate for the turbine driven AFW pump room as well as the turbine and motor pump control panels, 1
and valves.
The review team requests that the licensee ensures the adequacy of normal and emergency AC lighting, and the adequacy and aiming of emergency DC lighting.
The review team did not observe or check the emergency light'ng intensity but notes that the illumination level should be a minimum of 10 foot-candles at the work station to conform to the NRC's criteria for station lighting (NVREG-0700).
54
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A Ficure 5-1 Emergency DC Lighting Diagram For The AFh' Pumps Vicinity 55
The evaluation team has the following c Jitional observations:
1.
During the AFW system walkdown the staff noted that the degree of equip-ment and structural cleanliness was less than desirable That is possibly amplified by the dark colors of the walls and floors.
In one case, in order to read a nameplate on a valve limitorque a thick coating of dust and dirt had to be removec from the label.
Dust and dirt may contribute to malfunctions of equipment with low tolerance moving parts, or electrical relays and contacts.
The malfunctions of the latter may also be compounded by moisture.
2 The review team noted that operator access to AFWS equipment for normal surveillance, or for emergency remedial actions is generally poor.
This problem would be more pronounced during stressful conditions, or poor lighting (in case of station blackout).
Examples of this are shown in Figures 5-2, 5-3, and 5-4.
Emergency remedial actions may be greatly hampered by poor access to equipment, especially if compounded by poor emergency lighting.
Therefore, ease of access is of vital importance (see D.3.2 for available time discussion).
3.
To guard against AFW pump steam binding due to hot water back-leakage through the pump discharge piping the licensee checks the discharge piping temperature locally once every shift.
In order to achieve that, an operator has to climb over equipment or under piping to get to the far corner of the room and possibly lean over other equipment to see the temperature gauge.
This is shown in Figures 5-5, 5-6. and 5-7.
Although this once-a-shift task addresses staff concerns raised in GI-93 with respect to pump steam binding due to hot water backleakage, the current location of the tempera-ture gauge increases the potential of equipment malfunction due to an operator inadvertently stepping on, or bumping into such equipment.
In addition, this may increase the possibility of personnel injury.
The licensee is requested to move the AFW discharge piping temperature display closer to the room entrance so that the operator may avoid entering the room every shif t if the only objective is to take the
- temperature reading.
The licensee stated that ANO-2 has not tjad any backleakage steam binding events.
56
e 4.
The review team noted that at least one pipe support was loose such that, j
if sufficient force was exerted on that pipe, a pipe crack may develop.
Figure 5-7 shows a loose pipe support in the turbine driven AFW pump room.
This pipe is located in a tight passage way where it may be leaned against by an operator attempting to carry out the surveillance procedure described in item 3 above.
The licensee is requested to ensure the integrity of all equipment supports in the AFW system, and improve its surveillance program to sustain such integrity.
1 5.
Under certain emergency conditions the only means of decay heat removal is l
through injecting water in the steam generators by manually and slowly starting the turbine driven AFW pump using the trip and throttle valve (TTV).
Ir that process the operator needs to monitor the discharge pres-sure and possibly the turbine tachometer in order to ensure proper pump operation.
However, the pressure gauge and tachometer are not directly visible from the TTV location.
This may necessitate pump starting in several stages if one operator is assigned this task, since such operator will have to manipulate the TTV then move to check the readings'of the gauges.
Alternately, the task may require two operators.
The review team feels that this is an important (last resort in case of station blackout) remedial recovery action that can be made much easier by properly locating the necessary gauges so that they can be clearly seen from the TTV location.
The gauges should be clearly visible under the postulated station blackout lighting conditions.
Therefore, the staff suggests that the licensee move and reorient the above gauges, considering the proper emergency DC lighting, so that recovery actions by the operator may be easily carried out.
l 6.
The review team noted that equipment identification (e.g., labelling) is
{
generally not adequate.
While some equipment was clearly marked, most labels or nameplates were not easily legible.
The licensee stated that a reta;ging program is underway to improve equipment identificatior.
The licensee also stated that the ANO-2 inside-of-containment retagging has been completed, and that the AFW system will soon be retagged.
%e AFWS team notes that after the system is retagged, verifying that the correct tag is attached to the intended equipment is an equally important task.
57
(
e 7.
The review team noted that the licensee used adequate locking devices on manual valves in their standby mode (see Figure 5-8)
The licensee stated that all manual valves that, if not properly aligned, can divert or block the AFW flow are secured in their standby position by chains and locks.
The keys to these locks are readily available to the control room operator.
The licensee has a procedure for a pre-heatup valve position alignment (Reference 12) which includes physically-checking the locking chains to ensure that they are not cut.
By procedure the licensee also performs a pre-heatup valve position verification which specifically instructs that the verification be done by an individual different from that who did the alignment check.
The review team finds this independent verification useful in assuring system readiness.
The licensee also stated that by procedures the steam trap valves are ensured to be unisolated before plant startup, and the steam lines purged by steam via steam trap bypass valves every shift to further ensure no condensate in the lines.
8.
The review team noted that the licensee posts some system piping and instrumentation diagrams (P& ids) locally at the equipment location.
While the team did not check the accuracy of the P& ids or whether they are of current revision, it is the team's opinion that this is a very useful practice that tends to reduce confusion and improve equipment identifica-tion.
The review team also noted that the licensee is currently revising the turbine driven AFW pump overspeed trip reset and manual control pro-cedures (Reference 13).
The licensee stated that he intends to post exhibits A, B, and C of that procedure at the turbine driven pump room for quick reference.
With proper consideration of equipment location and lighting, the review team finds that posting of these simple exhibits can be very helpful during any potential recovery actions.
i 9.
The AFW pumps are located in separate isolated rooms such that a postulated steam line break in the 2P7A pump room will not adversely affect the motor
[
operator of pump 2P7B in the next room.
This pump arrangement addresses staff concerns raised in GI-68 with respect to AFW motor operator environmental qualifications following a steam line break in a pump room.
l l
i 58 r
r.
s e
- 0. 5. 3 Conclusion Based upon its review, the AFWS review team found tot the "as built" configuration of the ANO-2 AFWS is consistent with the design documentation.
The staff, however, had several concerns as discussed in the evaluation section above.
Based on discussions with the staff the licensee indicated that it is already taking certain steps to address some of the staff concerns.
Of the several staff concerns raised above, one is of particular significance.
This concern is over the ease of access to equipment to carry out manual reme-dial actions under emergency conditions.
Since the plants capability to safely mitigate transients and accidents depends to a large degree on the operators ability to perform various recovery actions, soine of which may be outside the control room, it is the staff's opinion that ease of access to the AFW system equipment is of paramount importance.
Therefore, the staff requests that the licensee develop a "maximum time to access," (MTA) from the control room to each piece of the AFWS equipment that may potentially be exercised during the course of an event.
Also, since the atmospheric dump valves are key equipment that may be used in combination with the AFWS for decay heat removal, an MTA should be developed for these valves.
The MTA should be introduced in the operator training program, and actual practice be performed by all operators as a part of that training.
MTA training should be introduced in the same context as the "time window" concept discussed in Section 0.3, Emergency Operating Procedures.
In developing the MTA, the licensee should consider potentially adverse environ-ments (e.g., elevated temperatures due to lack of ventilation from station blackout) and emergency lighting conditions.
The licensee should also ensure the availability of necessary tools (e g., ladders or wrenches) that may be used to operate various pieces of AFWS equipment.
The MTA should be determined as the maximum time for an operator to reach a particular piece of equipment and be able to start operating it.
After operating the equipment the AFW flow should be resumed to the steam generators before their water levels reach the zero level mark.
This will help the plant avoid temporary loss of the secondary heat sink.
The staff concludes that, provided the licensee adequately addresses staff concerns discussed in this section and in Section D.S.2 above, the AFW system as installed, instrumented, and operat.ed addresses staff concerns raised 59
4 in GI-68 with respect to AFW motor operator environmental qualifications follow-ing a steam line break in a pump room, in GI-1221.6, and c with respect to AFW
' interruption and recovery, and in GI-124 in that it is conducive to safe and reliable operation during accident or transient conditions.
As indicated in Section A, the licensee provided satisfactory responses to these concerns, and therefore, the staff considers the issues identified during the walkdown to be resolved.
1 D.6 Training 0.6.1 Approach The AFW team reviewed the training program to establish an understanding of the licensees commitment to maintain and enhance the proficiency level of its main-tenance, operations and engineering staffs with respect to normal, and emergency operation the AFW system.
The team review consisted mainly of interviews with training instructors and staff members.
The team also inspected some training documentation.
This review does not address the completeness of the training program and should not be considered a3 an acceptance review.
D.6.2 Evaluation Prior to being hired by Arkansas Power and Light Company (AP&L) the prospective maintenance employee is screened for aptitude and trainability.
In addition the prospective employee's educational background must include one or more of the following:
Vocational-Technical School, Technical School, or college.
The utility assigns new hires as entry level trainees.
A person would have to be exceptionally qualified to be hired above an entry level position.
Newly hired maintenance personnel must take the AP&L training program.
This program consists of five years of both schooling and on-the-job (0JT) training.
The program also includes continuing training and refresher courses.
The training program, however, is not strictly geared to the AFWS but in general to all plant systems and equipment.
i 60
(
s.
The entry level personnel undergoes four r.onths of training after he is hired.
Dur'ing these four months, he attends a company-owned offsite training school.
'This school introduces the treinec to the basic courses (mathematics, mechanics, thermodynamics, pumps, controls, etc.).
The entry level personnel has ti,e option of "testing out" or qualifying for the individual courses.
To "test out," the trainee takes a test in the course subject matter, if he passes the test, he gets credit for that course and does not have to take the course.
Some of these courses have been transferred to the onsite training facility.
Upon completion of the entry-level training course, the trainee becomes an apprentice trainee and selects an area of specialization:
rigger, steam fitter / welder, electrician, mechanic, etc.
The apprentice trainee undergoes four years of training.
During these four years, he attends a company-owned on-site training school, which provides the trainee with basic-skills training for the specific craf t he has cLsen.
This training provides some hands-on experience in disassembling and reassembling equipment similar to that found at the plant.
The balance of the four years is spent at the plant as on-the-job training.
The on-the-job training consists of a check-off list of typical jobs and duties which the apprentice trainee must complete prior to advancing to tha next level.
The check-off list starts out with general items and becomes mora specialized as one progresses.
Upon completion of the apprentice trainee program, the trainee becomes a journeyman.
The journeyman course includes required schooling and on-the-job training over a minimum one year period.
Upon completion of the program the journeyman becomes a mechanic, electrician, etc.
Both the apprentice trainee and the journeyman courses are INP0 certified, and take full advantage of manufacturer material.
In addition to the formal training courses, the ANO Unit 2 Maintenance Training Department has a continuing / refresher training program.
This program allots approximately 25% of the total yearly man hours for continuing education during a refueling outage year, and 15% to 20% during a non-refueling outage year.
The program makes use of plant training facilities, company training facilities 61
s e
and manufacturer traini.1g programs.
Its purpose is to maintain and enhance the skills of the maintenance personnel.
The licensee informed the review team
'that part of the new parts and component procurement process will be to negot-iate the inclusion of vendor training (maintenance and troubleshooting) as part of the purchase contract.
By inspection of training documentation, interviews, and a partial walkdown of the training facility, the review team found the maintenance training program to be thorough and deta. led.
It makes extensive use of manufacturer materials, technical bulletins, sales brochures, technical manuals and demonstration models, as well as on-the-job training.
Since operator licensing is a regulated activity and has been extensively considered by the NRC, the evaluation team spent a minimum amount of time in the operator training area.
ANO Unit 2 management emphasized the licensee's commitment to training, and stated that it encouraged the plant engineers and plant superintendents to maintain their reactor operator (RO), and ser'or reactor operator (SR0) licenses current.
A few select individuals in this group were required by company policy to maintain current licenses.
However, due to the amount of required training, testing, and serving on control room shifts (approximately 25%-30% of the yearly man hours), it was stated that it was difficult for many ex-R0s and ex-SR0s to maintain the license and perform there assigned regular duties.
Thus, many of these individuals allow their licenses to lapse.
The staff does not encourage this approach since it lessens the level of awareness of plant operation and systems interrelations.
0.6.3 Conclusions The staff finds that the licenses's commitment and imp.ementation of the training programs as discussed above promotes good understanding of plant operation in general, and in particular may enhance the licensee's ability to minimize system malfunctions and improve the likelihood of recovery once a malfunction takes place.
The staff, therefore, concludes that with resolution j
62
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l
+
a of,the concerns raised in Section 0.5, and other sections, the ANO-2 training program contributes to and er5ances the plants capability for decay heat removal.
D.7 Operating Experience and Reliability Analysis D.7.1 Approach The staff reviewed several documents including the Licensee Event Reports LERs) and Nuclear Plant Reliability Data System (NPROS) files pertinent to ANO-2 (References 14 and 15).
The staff also discussed with the licensee the licensees statistical analysis relative to equipment failures.
i D.7.2 Evaluation The staff searched the LER and NPRDS files for the period of 1981 through 1985 for data pertinent to the ANO-2 AFW System (References 14 and 15).
The search revealed a substantial number of discharge valve failures and several failures in suction valves, steam admission valves to the turbine drives for the AFW pump, turbine driven pump overspeed, and pump packing.
The licensee replaced some of the discharge valves, and stated that the rate of valve failures has significantly decreased.
Some valve failures continued to occur as of late 1985, the period covered by this data search.
In nine instances the failures disabled or degraded an AFW train.
However, all of these were readily recover-able by an operator at the equipment location.
Discharge valve failures par-tially disabled an AFW train.
These failures were all readily recoverable at the equipment location.
Only one instance was found that is a precursor to a nonrecoverable AFW train failure.
That was an excessive packing leakage which resulted in the pump bearing oil being contaminated by the leaking water.
The licensee conducted a statistical analysis for ANO-2 AFW train failures and component failures for the period of 1980 through 1986.
This analysis shows a rapidly decreasing failure rate trend (e.g., from 16 single train failures in 1980, to 8 failures in 1981, to zero in 1982 and subsequent years).
For other component failures, the licensee shows a similar trend, from 20 in 1980, to zero in 1983 and later.) The staff believes that this trend in reduced number of failures is significant, and is indicative of a substantial licensee effort 63 1
o to, improve equipment performance.
Most of the equipment failures are recoverable given an easy access route and suf ficient time before reactor coolant system degradation (See Sections 0.3 and 0.5 for discussion of time available for action, and of equipment accessibility).
Since reactor scrams place the plant in a condition where decay heat must be removed, such scrams constitute potential challenges to the AFW system.
The licensee's operating experience relative to mscheduled automatic reactor scrams and associated challenges to the AFVS, shows a promising trend.
Although AND-2 had a scram rate of 24* scrams / year in 1981 the rate has generally declined since then to 5 scrams / year in 1986.
The plant had 8 unanticipated scrams in 1985.
This is about double the national average for that year (4.3 scrams / plant year, Reference 16).
However, the licensee appears to be committed to reducing the plant scram rate through its trip reduction program.
The trip reduction program is directed at minimizing the causes for reactor trips.
e.g., reactor protection (including the core protection calculator) anomalies, main feedwater upsets, and turbine generator control malfunctions.
If the above improvement rate continues, much reduction in the challenge rate of the AFWS will result.
A survey of the failure rate history of the emergency diesel generators (DGs) on all operating nuclear power plants in the period of 1983 through 1985 (Reference 17) shows that ANO-2 emergency diesel generators have an average failure rate of 0.016 and 0.02 failures per demand, and meet the reliability criterion identified in the proposed Regulatory Guide, "Station Blackout."
This Regulatory Guide has been issued for comment and identifies a DG relia-bility acceptance criterion of no more than O'.025 failures / demand for the ANO-2 diesel generator configuration.
- The licensee attributed most of the scrams to the CE core protection calculator (CPC), which is currently being improved.
64
e e
The licensee has performed an AFWS reliability analysis.
Below is a summary of'the licensee's analysis.
After performing a benchmark analysis against the
'NUREG-0635 (Reference 3) results, the licensee conducted two additional analyses us%g the same model but with different failure rate data.
These analyses are raferred to as "Bayesian Update" and "Best Estimate" analyses.
The results are provided below for the loss of feedwater (LOFW) and the loss of offsite power (LOOP) events.
LOFW LOOP Benchmark Analysis 1.1E-3 2.2E-3 Bayesian Update 9.9E-4 2.1E-3 Best Estimate 6.3E-5 3.3E-4 The failure rate data used for the benchmark analysis is the same as that used for the NUREG-0635 analysis.
The failure rate data used for the Bayesian Update analysis was obtained by combining plant-specific data with industry generic data.
The failure rate data used for the Best Estimate analysis was obtained from the generic data base and additional modeling assumptions.
The above analyses do not account for:
(a) operator action to recover faulted conditions or to use alter-nate equipment; (b) support system failures (other than ac power); (c) AFWS failure due to control system faults; or (d) common cause contributors to system unreliability.
Consideration of these factors could affect the reliability esti-mate in different ways.
The AFWS review team places significant weight on each of the above factors and suggests that the licensee complete its reliability analyses.
The licensee should consider implementing those measures identified in the analyses that will enhance the AFWS reliability.
The above estimated AFWS reliabilities may be combined with the frequencies of the LOFW or LOOP to arrive at an estimate of the expected frequency of AFWS failure when it is needed for decay heat removal.
ANO-2 has experienced 12 LOFW events in the period of 1981 through 1986.
While this is a high rate of occurrence the licensee stated that all events were readily recoverable.
- Also, the rate of LOFW decreased steadily from 5 events in 1981 to no events in 1986.
The plant experienced no LOOP events during the same period.
In order to 9
65
e est.imate the frequency of failure of AFWS when needed the staff assumed a
, frequency of 0.646 non-recoverable LOFW events per reactor-year.
This was estimated by Duke Power Company for Oconee, Unit 3 (Reference 18) and is believed to be a reasonable estimate for other plants.
In the following ColCulations the plant-specific bayesian update results were used.
Number of AFWS failures when needed=
0.646 (LOFW events / year) x 25.4 (years-assumed remsining operation life)^
x 9.9E-4 (failure / demand)= 0.016 Therefore, there is a 1.6% chance that during the remaining life of the plant it will experience a total loss of main and auxiliary feedwater.
In a case of total loss of feedwater the cperator must ruccessfully initiate the "feed-and-bleed" mode of decay heet removal or risk fuel failure and radieactive releases.
The staff estimates the risk to the public as a result of total loss of main feedwater and auxiliary feedwater systems to be 1480 person-rems.
If the AFWS and back-up systems reliability is increased to IE-4 failures / demand as per the Standard Review Plan Section 10.4.9, Auxiliary Feedwater System, risk to the public would be decreased by 1330 person-rems.
In light of this risk reduction the staff believes that the licensee should provide an alternate source of water to the steam generators such that the reliability of decay heat removal means may be enhanced.
The staff believes that the addition of the alternate means of water delivery to the steam generators will enhance the AFWS reliability to the IE-4 failures / demand criterion.
Furthermore, the licensee should emphasize, in its operator training program, the graveness of not initiating "feed-and-bleed" if a total loss of feedwater is suspected.
Training on recognition and mitigation of such an event should also be emphasized.
80% of the remaining calendar plant life of 31.7 year 66
o w
D.7 3 Conclusion Based on staff review and observations, ano based on t h licensee's statements, the staff concludes that the licensee is actively pur;uing improvements in tht:
AFW system reliability and availability.
The licensee'4 statistical an3 lysis conducted for the last six ye:rs fo.- the AN0-2 AFWS-related failures shows a rapidly decreasing fuilure rate trend with no significant failures since 1983 This indicates a suostantial improvement in equipmer.t performance.
As discussed above the majority of equipmtat failures, over a five year perivi from 1981 through 1985, were dei.errained to be readily recoverable at the equip-ment location.
Therefore, the staff believes that ease of access to various AFWS equipment is of paramount importance to the overall system reliability ard availability.
Ease of ac,ess requires adequate normal and emergency lighting, adequate communications, clear and legible equipment identification, and avail-ability of tools that may be necessary to operate such equipment.
These issues are discussed in detail in Section D.5 above.
Although significant improvement in va've performance has resulted over the last fe., years, the licensee's ado;. tion of the MOVATs methodology for valve setpoint setting (see Section D.2.2) is expected to further improve these i
valves performance.
The licensee's E0Ps explicitly and c!adrly instruct the operators, i' the AFW flow can not be established, to attempt to reestablish the MFW flow or, if that is unsuccessful, to rely on the condensate puhr flow.
The E0P sections provide guidance and precautions about degraded mod 9s of equipment operations, if neces-sary.
If all MFW and AFW are not readfiy recov v able, the E0Ps instruct the operator to initiate the "feed-and-bleed" mode of decay heat removel.
.ilthough the staff has certain concerns regarding operator initiation of "ferb and-bleed" as discussed previously, the staff believes that ANO-2 "feed-and-teleed" system with its 3-inch flow path with two motor operated valves in series, enhances the decay heat removal capability.
This flow path is sufficiently large to depressurize the reactor coolant system to allow flow from the high head safety injection pumps.
However, while this provides additional capability fter 67
+
~
1 decay heat removal the overall decay heat removal system reliability remains less than adequate.
This is due to the large uncertainties in the operators
'use of "feed-and-bleed" as mentioned previously which limit its consideration as a last resort for decay heat removal rather than an adequately reliable I
compensatory decay heat removal feature to justify a larger AFWS unavailability, Therefore, the stsff believes that an alternate method of providing water to the steam generators is needed so that the overall decay heat removal system reliability may be enhanced.
The alternate water source for ANO-2 could use existing equipment such as an interconnection to the ANO-1 AFWS, or could incorporate a separate startup feedwater pump.
As discut ed above, the plant's rate of unanticipated automatic reactor scrams is high, however, it is steadily decreasing.
Also, the licensee has had a relatively low failure rate of its two emergency diesel generators.
Based on the above, the staft concludes that, upon resolution of the staff concern identified in Section A of this report for an additional method of secondary
{
side decay heat removal capability, the ANO-2 AFW system is properly designed, instrumented, maintained, and operated, and that the licensee has an adequate E0P, and training program for the AFW system The staff, therefore, concludes that, upon resclving the above conce n, the ANO-2 AFWS will be sufficiently reliable.
1 68 1
e s
Appendix A
'Referencas
-l l
1.
ANO-2 FSAR Section 9.2.1, Service Water System., 10.J, Main Steam Supply System, 10.4.9, Emergency Feedwater System j
2.
Plant Technical Specifications Section 3.7.1.2, 3.7.1.3, 2.8 1.1 3.
NUREG-0635, "Generic Evaluation of Feedwater Transients aad Smell Breaks Loss of Coolant Accidents in CE Designed Plants"Section X.1, pages X-2 through X-15, ANO-2 Emergency Feedwater System.
4.
ANO-2, II-E.1.1 Safety Analysis Reports dated NcVamber 10, 1980, May 12, 1981, February 3, 1982, March 25, 1982, and November 18, 1983.
5.
NUREG-1212, "Status of Maintenance in the U.S. Nuclear Power Industry 1985," Volume 2, June 1986 6.
Inspection Report No. 50-313/86-01, Safety System Functional Inspection of ANO, Unit 1 AFWS, dated March 31, 1986 7.
Emergency 0parating Procedure, E0P 2202.01, Rev. 2 8.
Anncoicators 2K04, 2K05, 2K06, and 2K07 Corrective Actions 9.
Drawings audited for conforman:e to IEEE Standard 279 and NUREG-0737 Item II.E.1.2.
Title OWG. No Station Single Line Diagram E-2001 Single Line Relay Diagram E-2002 Single Line Relay Diagram (6900 Volt)
E-2003 Single Lina Relay Diagram (4160 Volt)
E-2004 Single Line Relay Diagram ESF E-2005 (4160 Volt)
Low Yoltage Safety System E-2006 Single Line Relay Diagram (480 Volt)
E-2008 Schematic Diagram (125 Volt DC)
E-2085 Functional Logic Diagram EFWS E-2403
{
Functional Logic Diagram Aux Cooling E-2406 Water e
69
w Title DWG No.
M2202 Condensate & Feedwater (P&ID)
M2204 Steam Generator Secondary System (P&ID) M2206 EFW Turbine MOVs 2CV-1000-1 &
E229 2CV-1050-2 EFW Pump Suction MOVs E2296 EFW Discharge Valves E2300 EFW Condensate Suction MOVs E2301 Emergency Power Supply for Control E2420 Room dC System i
EFW Turbine Driver Steam Isolation E2443 Valves 10.
Maintenance Procedurec 1403.160 for model SMB-000, and 1403 163 for model SMB-00 11.
System Operating Procedure 2107.04, "0C Electrical System Operation, Supplement VI" 12.
Plant Pre-heatup and Pre-critical Checklist 2102.01, Rev. 23.
13.
Emergency Feedwater System Operating Procedure 0P 2106.06, Exhbits A, 2P7A Turbine Overspeed Trip Reset., B, Drawing., and C, Manual Control of 2P7A.
14.
F. Hebdon to T. Speis, Preliminary Review of-AFW Related LERs, April 13, 1986.
15.
F. Hebdon tu T. Speis, Review of AFW Related LER and NPRDs Data, July 25, 1986.
16.
INP0 86-012, "Unplanned Automatic Scrams in U.S. Electric Generating Units in 1985," April 1986.
17.
NSAC-108, "The Reliability of E'.lergency Diesel Generators at U.S Nuclear Power Plants in 1983 tb*ough 1985," September 1986.
18.
NSAC-60, "A Probabilistic Risk Assessment of Oconee Unit 3," June 1984.
i 1
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Appendin B.
' Individuals involved in the,AFW System Review NRC Personnel i
AFWS Evaluation Team Paul Norian, Team Leader NRR Sammy Diab, Task Manager, NRR Norman Wagner, NRR Robert Giardina, NRR Joseph Joyce, NRR
. Robert Lee, ANO-2 Licensing Project Manager Charles Harbuck, Resioent Inspector, Region-IV Willim Johnson, Senior Resident Inspector, Region IV Arkansas Power and Light (AP&L) Personnel l
Ted Enos, Manager of Nuclear Engineering and Licensing, AP&L Charles Turk, Supervisor of Nuclear Engineering, AP&L Bill Converse, Supervisor of Plant Performance, AP&L Kurt Taylcr, Supervisor of Operational Technical Support, AP&L Jason Remer, Maintenance History Supervisor, AP&L 3
Kenny Coates, Mechanical Maintenance Supervisor, AP&L 1
Larry Taylor, Plant Licensing Engiraering, AP&L Jerry Peter, Operational Technical Support, AP&L David McKenny, Reactor Operator, AP&L Randal Golden, Senior Reactor Operator, AP&L Bill Cradacek, Reliability Analysis Group, AP&L 71
..