ML20039D192
| ML20039D192 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 12/22/1981 |
| From: | Clayton F ALABAMA POWER CO. |
| To: | Varga S NRC |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-***, TASK-TM NUDOCS 8112310408 | |
| Download: ML20039D192 (77) | |
Text
._
M;iling Addrns Alabama Power Company 600 North 18th Street Post Office Box 264; Birmingham, Alabama 35291 Telephone 205 783-6081 F. L. Clayton, Jr.
L'ag;%y"'
Alabama Power n saven aune saiem December 22, 1981 Docket No. 50-348 p
50-364 nECgNED Director, Nuclear Reactor Regulation DEC3 01981 *f U. S. Nuclear Regulatory Comission E-Washington, D. C.
20555 sama p T, W4 1
Attention: Mr. S. A. Varga
/,
~.
Gentlemen:
Joseph M. Farley Nuclear Plant - Units 1 and 2 l
l NRC Commitment Compliance Status 1
In order to provide clarification status and required submittals as related to various Post-TMI issues; the Unit 2 license; and other l
NRC requirements, Alabama Power Company provides the following enclosures.
Enclosures 1 and 3 provide the status of NUREG-0737 requirements for Units 1 and 2, respectively.
Enclosures 2 and 4 provide the status
(
of the NUREG-0737 Technical Specification requirements for Units 1 and 2, respectively. provides the status of the Unit 2 license requirements. provides the status of other NRC requirements for January 1, 1982; the 3rd Refueling Outage for Unit 1; the 1st Refueling Outage for Unit 2. includes the basis for each extension indicated in Enclosures 1 through 6 and provides all submittals not previously provided.
If you have any questions, please advise.
Yours very truly,
\\
f le le0 F.1.Clayt'n, o
r.
6 i
FLCJr/ JAR:jc Enclosures cc:
Mr. R. A. Thomas w/enciosures)
Mr. G. F. Trowbridge w/ enclosures)
Mr. J. P. O'Reilly w/ enclosures)
Mr. E. A. Reeves (w/ enclosures)
Mr. W. H. Bradford (w/ enclosures)
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ENCLOSURE 1 l
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J. M. FARLEY NUCLEAR PLANT UNIT 1 l
STATUS OF NUREG-0737 REQUIREMENTS l
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Encl. 1 UNIT 1 NUREG-0737 STATUS l'
Clarifi-Implemen-Tech.
cation Shortened tation Spec.
NRC APCo Response APCo-Ext 7nsion Item Title Description Schedule Req.
Remarks Letter Date Remarks Requested I.A.1.1 Shift technical
- 2. Tech specs 12-15-80 Yes 12-15-80 Complete N/A advisor l
- 3. Training per 1-1-81 Yes 1-14-81 Complete N/A LL Cat B 2-5-81
- 4. Describe long-1-1-81 No 1-14-81 Complete N/A i
term program 2-5-81 i
I.A.1.3 Shift manning
- 1. Limit overtime 11-1-80 No 6-26-80 Complete N/A
- 2. Minimum shift 7-1-82 Yes Amend TS 6-26-80 Complete N/A crew on shift manning I.A.2.1 Immediate upgrade 1.
SR0 experience 5-1-80 No Completion 7-10-80 Complete N/A of R0 & SR0 to be training and verified
- 2. SR0s be R0s, 12-1-80 No Completion 1-14-81 Complete N/A 1 yr.
by OIE
- 3. 3 mo training 8-1-80 No Completion 1-14-81 Complete N/A on-shift by OIE
- 4. Modify training 8-1-80 No NRR staff 7-10-80 Complete N/A to review 1-14-81
- 5. Facility certi-5-1-80 No 01E veri-7-10-80 Complete N/A fication fication l
I.A.2.3 Administratior.
Instructors' com-8-1-80 No.
NRR to 1-14-81 Complete N/A of training plete SR0 exam verify programs confor-mance N/A - There are no further requirements for this item.
No - P.emaining requirements are expected to be completed on schedule Yes - An extension for this item was requested in the letter listed below.
l Encl. 1 UNIT 1 NUREG-0737 STATUS
- Clarifi-Implemen-Tech.
cation Shortened tation Spec.
NRC APCo Response APCo Extension Item Title Description Schedule Req.
Remarks Letter Date Remarks Requested I.A.3.1 Revise scope and
- 1. Increase scope 5-1-80 No 7-10-80 Complete N/A criteria for licensing exams
- 2. Increase pass-5-1-80 No 7-10-80 Complete.
N/A ing grade
- 3. Simulator exams 10-1-81 No Plants 1-14-81 Complete N/A
- b. All w/o 2-9-81 (Plant simu-specific lators simulator exams sche-duled to be given by 7-1-83)
I.C.1 Short-tem
- 2. Inadequate accf Jent and core cooling l
procedure review
- a. Reanalyze 1-1-81 No 1-14 Complete N/A and propose W.0.G letter
. guidelines80-179, dated 12-15-80
- b. Revise First No 1-14-21 Suspense:
No procedures refueling 9-9-81 (not 4th Refuel outage after applicable Outage 1/1/82 due b early.
See Encl. 7, outage) item 1
- 3. Transients &
1-1-81 No 1-14-81 Complete N/A accidents
- a. Reanalyze &
propose guidelines
- b. Revise First No 1-14-81 Suspense:
No procedures refueling 4th Refuel outags after.
Outage 1/1/82 See Encl. 7, item 1
Encl. 1 3'
UNIT 1 NUREG-0737 STATUS Clarifi-Impl emen-Tech.
. Extension ~
cation Shortened tation Spec.
NRC APCo Response APCo Item
- Ti tle Description Schedule Req.
Remarks -
Letter Date Remarks Requested-I.C.5 Feedback'of Licensee to imple-1-1-81 No 1-14-81 Complete-N/A operating ment procedures experience I.C.6 Verify correct Revise performance 1-1-81 No 2-5-81 Complete N/A 2-23-81 performance of procedures operating activities I.D.1.
Control-room Preliminary TBD.
-=
Final 2-23-81 Awaiting -
No design reviews assessment &
guidance-finaliza-schedule for will.be tion of correcting issued require-deficiencies 1981 as' ments by.
NUREG-0700 NRC i
See Encl. 7, i
' item 2 f
Guidance 1-14-81 Awaiting No I.D.2 Plant-safety-
- 1. Description.
TBD parameter display per NUREG-11-16-81 finaliza-console 0696.Rev. 2 tion of require-ments by.
NRC' Guidance 1-14-81 Awaiting No
- 2. Installed TBD
. per NUREG-11-16-81 finaliza-0696 Rev. 2 tion of-require-ments by NRC
- 3. Fully imple-TBD
= Guidance 1-14-81 Awaiting.
No mented per NUREG-11-16-81 finaliza-0696 Rev. 2-tion of.
require-sents by-NRC.
- - - ~
Encl. 1 UNIT 1 NUREG-0737 STATUS 4
Clarifi-Implemen-
.ech, caticn Shortened tation
' Spec.
NRC APCo Response APCo Extension Item Title Description Schedule Req.
Remarks Letter Date Remarks Requested II.B.1 Reactor-coolant-
- 2. Install vents 7-1-82 Yes 1-14-81 Complete N/A system vents (LL Cat B)
- 3. Procedures 1-1-82 Yes 1-14-81 Suspense:
No when vent system is operational and NRC ap-proved See Encl. 7, item 3 II.B.2 Plant shielding
- 2. Plant modi-1-1-82 No 2-23-81 Complete N/A fications except for (LL Cat B) the elec-i trical dis-connect devices which are scheduled to be in-stalled prior.to startup following the current refueling outage See Encl. 7, item 14
- 3. Equipment 6-30-82 No 2-23-81 Suspense:
No j
qualification 9-9-31 6-30-82 II.B.3 Postaccident
- 2. Plant modifi-1-1 Yes 1-14-81 Complete N/A l
sampling cations l
(LL Cat B)
Encl. 1' 4-UNIT 1 NUREG-0737 STATUS 5-Clarifi-Implemen-Tech.
cation Shorteced tation Spec.
'NRC APCo Response APCo Extension Item
' Title Description Schedule Req.
Remarks Letter Date Remarks Requested II.B.4-Training for
- 1. Develop train-1-1-81 No 2-9-81 Complete N/A mitigating core ing program damage
- 2. Toplement program
- a. Initial 4-1-81 No 2-9-81 Complete N/A
- b. Complete 10-1-81 No 2-9-81 Complete N/A II.D.1 Relief and
- a. Complate 7-1-81 No 6-25-81 Additional No-testing 9-30-81 information to be pro-vided upon completion of the EPRI program l
- b. Plant speci 1-81 TBD 6-25-81 Additional No i
fic report 9-30.information to be pro-vided upon completion of the EPRI program-
- 3. Block valve 7-1-82 TBD 2-23-81 Suspense:
No testing
'7-1-82 APCo is participa-ting in EPRI sub-mittal II.D.3.
Valve position
- 2. Tech Spec 12-15-80 Yes 12-15-81 Complete N/A
Encl. 1 5
UNIT 1 NUREG-0737 STATUS l
Clarifi-Implemen-Tech.
l cation Shortened tation Spec.
NRC APCo Response APCo Extension Item Title Description Schedule Req.
Remarks Letter Date Remarks Requested II.E.1.1 Auxiliary feed-
- 1. Short term 7-1-81 Item 1-14-81 Complete N/A water system specific evaluation
- 2. Long term 1-1-82 Item 1-14-81 Complete N/A specific II.E.1.2 Auxiliary
- 1. Initiation feedwater system (b) Safety 7-1-81 Yes 2-9-81 Suspense:
Yes initiation and grade 7-1-81 4th Refuel 7-1-81 flow Outage See Enci 7, item 4 2.-Flow indication (b) LL Cat A 12-15-80 Yes 12-15-80 Complete N/A tech specs (c) Safety 7-1-81 Yes 2-9-81 Suspense:
Yes grade 7-1-81 4th Refuel 7-1-81 Outage See Enci 7, item 4 II.E.3.1 Emergency power
- 2. Tech specs 12-15-80 Yes II.G.1 12-15-80 Complete N/A for pressurizer heaters II.E.4.1 Dedicated hydro-2.
Install 7-1-81 No 1-14-81 Not N/A gen penetrations required II.E.4.2 Containment
- 5. Containment 2-13-81 Complete N/A isolation press setpoint 9-9-81 dependability.
- a. Specify 1-1-61 No pressere
- b. Modifica-7-1-81 les tions
i Encl. 1-UNIT l'NUREG-0737 STATUS Cl ar:,fi-Implemen-Tech.
cation Shortened tation' Spec.
NRC APCo Response-APCo Extension ges Title Description Schedule Req.
Remarks Letter Date Remarks Requested
- 6. Containment 1-1-81 Yes 1-14-81 Complete N/A :
purge valves
- 7. Radiation 7-1-81 Yes 1-14-81 Complete N/A l-signal on l
purge valves
- 8. Tech Specs 12-15-80.
Yes -
12-15 Complete N/A-II.F.1 Accident-
- 1. Noble gas 1-1-82 Yes-2-13-81 Complete F/A-monitoring monitor See Enci 7,
-item 5
- 2. Iodine /
1-1-82 Yes 2-13-81
. Complete N/A.
particulate See Enci 7, j
sampling item 5 l
- 3. Containment 1-1-82
_Yes 2-13. Complete N/A i
high range monitor l
- 4. Containment 1-1-82 Yes-2-13-81 Complete See pressure Enc 1 2, item l
II.F.1.4
- 5. Containment 1-1-82 Yes 2-13-81' Complete N/A water level
- 6. Containment 1-1-82 Yes 2-13-81 Complete N/A hydrogen II.F.2 Instrumentation
- 2. Tech Spec 12-15-80 Yes 12-15-80 Complete See for detection of (LL Cat A) _
. item Encl 2, inadequate core-cooling II.F.2 l
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i Encl. 1 1
8 UNIT 1 NUREG-0737 STATUS l
Clariff-Impl emen-Tech.
l cation Shortened tation Spec.
NRC APCo Response APCo Extension Item Title Description Schedule Req.
Remarks Letter Date Remarks Requested l
- 3. Install level 1-1-82 Yes 3-14-80 Complete N/A instruments 2-9-81 (LL Cat B) 6-24-81 (Incore ther-6-29-81 mocouple 10-6-81 panel test l
12-31-79 to be per-1-14-81 formed l
prior to startup following current refuel outage)
II.G.1 Power supplies
- 2. Tech specs 12-15-80 Yes See 12-15-80 Complete N/A for pressurizer II.E.3.1 relief valves, block valves, &
level indicators II.K.1 IE Bulletins 79-05, 06, 08 Bulletin No NRR eva-6-22-79 Complete N/A Specific luating licensee responses II.K.2 Orders on B&W
- 13. Thermal-1-1-82 As 1-14-81 See Enci 7, No Plants mechanical required W.0.G.81-138 item 6 report dated 4-1-81 5-20-81
- 17. Voiding in RCS 1-1-82 No 1-14-81 Complete N/A See Enci 7, item 7
- 19. Benchmark 1-1-82 No 1-14-81 Not N/A analysis seg NRC letter required AFW flow 7-6-81
Encl. 1 UNIT 1 NUREG-0737 STATUS' 9
Clarifi.
Implemen-Tech.
cation Shortened tation Spec.
NRC APCo Response APCo Extension-Item Title Description Schedule Req.
Remarks Letter Date Remarks Requested
'II.K.3 Final recommenda-
- 1. Auto PORV 5-26-81 Not-N/A tions, B&O task isolation required force
- a. Design 7-1-81 Yes if required by II.K 3.2
- b. Test / install 1st refuel Yes 6 mo. after staff appro-val
- 2. Report on 1-1-81 No 1-14-81 Complete N/A PORY failures WCAP-9804 3-13-81
- 3. Reporting SV &
1-1-81 Yes Initiate 1-14-81 System in N/A RV failures &
data begin-9-9-81' place challenges ning 4-1-80
- 5. Auto trip of RCPs 1
a.
Propose mo-7-1-81 No 6-30-81 Suspense:
No
)
difications Within 3 months after NRC determina-tion of accepta-bility of SB LOCA Model See Encl. 7, item 15
Encl. l' UNIT 1 NUREG-0737 STATUS 10 Clarifi-Implemen-Tech.
cation Shortened tation Spec.
NRC APCo Response APCo Extension Item Title Description Schedule Req.
Remarks Letter Date Remarks Requested b.
Modify 3-1-82 Yes If required 6-30-81 Suspense:
No Within 11 months after NRC determina-tion of accepta-bility of SB LOCA model during an appro-priate out-age or 1st such outage thereafter 9.
PID controller 1-1-81 No Implemen-6-26-80 Complete N/A tation to be verified
- 10. Proposed Plant Yes 6-26-80 Complete N/A anticipatory specific trip modifi-cations
- 11. Justify use Plant No See NUREG-1-14-81 Complete N/A of certain specific
- 0611, PORVs Sect.
3.2.4.d.
- 12. Anticipatory trip on tur-bine trip
- a. Confirma-1-1-81 No 6-26-80 Complete N/A tion or propose modifica-tions
r Encl. 1 UNIT 1 NUREG-0737 STATUS l
l Clarifi-Implemen-Tech.
l cation Shortened tation Spec.
NRC APCo Response APCo Extension l
Item Title Description Schedule Req.
Remarks Letter Date Remarks Requested
- b. Modify 1st refuel Yes 6-26-80 Complete N/A i
or 6 mo.
after staff approval
- 17. ECCS outages 1-1-81 As 2-13-81 Suspense:
No required Every 5 year period after initial criticality
- 25. Power on pump seal s
- a. Propose 1-1-82 No 1-14-81 Complete N/A mods prior to SER
- b. Modifica-7-1-82 No 1-14-81 Complete N/A tions See Enci 7, item 8
- a. Schedule 11-15-80 No Westinghouse Complete N/A outline letter dated 11-25-81 NS-EPR-2524
- b. Model 1-1-82 No Westinghouse Complete N/A letter dated See Enci 7, 11-25-81 item 9 NS-EPR-2524
- c. New 1-1-83 or No Westinghouse Complete N/A analyses 1 yr after letter dated See Enci 7, staff appro-11-25-81 item 9 val.
Encl.1 12 UNIT 1 NUREG-0737 STATUS Clarifi-Implemen-Tech.
cation Shortened tation Spec.
NRC APCo Response APCo Extension Item Title Description Schedule Req.
Remarks Letter Date Remarks Requested
- 31. Compliance 1-1-83 or TBD 1-14-81 Suspense:
No with 10 CFR 1 yr after 1-1-83 50.46 staff appro-val III.A.1.2 Upgrade emer-
- 2. Design TBD TBD 2-13-81 Comple te N/A gency support 5-19-81 facilities
- 3. Modifications TBD TBD 2-13-81 Suspen!.e:
No 5-19-81 10-1-82 III.A.2 Emergency
- 1. Upgrade emer-3-1-81 Yes Procedures 11-7-80 Complete N/A preparedness gency plans submitted to App E, 3-1-81 10 CFR 50
- 2. Meteorological 6-1-83 Yes Staged 2-9-81 Suspense:
No data implemen-7-1-82 tation III.D.1 Primary coolant
- 2. Tech specs 12-15-80 Yes 12-15-80 Complete N/A outside con-tainment III.D.3.3 Inplant 12
- 2. Modifications 1-1-81 Yes 1-14-81 Complete N/A radiation to accurately 9-9-81 monitoring measure 12 III.D.3.4 Control-room
- 1. Review 1-1-81 No 6-26-80 Complete N/A habitability
- 2. Modifications TBD Yes 6-26-80 Not N/A required I
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l
y-M ENCLOSURE 2 e
J. M. FARLEY NUCLEAR PLANT UNIT 1 STATUS OF N.UREG-0737 TECHNICAL SPECIFICATION REQUIREMENTS
Encl. 2 J. M. FARLEY NUCLEAR PLANT UNIT 1 1
STATUS OF NUREG-0737 TECHNICAL SPECIFICATION REQUIREMENTS TECHNICAL 0737 SHORTENED SPECIFICATION ITEM TITLE SECTION COMMENT / EXTENSION REQUESTED I.A.1.1 Shift Technical Presently in FNP-1 Tech Spec. Section 6.2.3 and Advisor Tabl e 6.2.1.
Also, part of FNP-1 Tech. Spec. Up-grade Package Section 6.2.4, Figure 6.2-2 and Tabl e 6.2-1.
I.A.1.3 Shift Manning Presently in FNP-1 Tech. Spec. Section 6.2.2. and Table 6.2.1. Also, part of FNP-1 Tech. Spec. Up-grade Package Section 6.2.2 and Table 6.2-1.
II.B.1 Reactor Coolant Pressurizer Vent (PORV) is in FNP-1 Tech. Spec.
Reactor Vessel Head Vents are installed.
System Vents 3.4.4.a.
Reactor Vessel Head Vents are not A Tech. Spec. will be written upon NRC in FNP-1 Tech. Spec. Also, Pressurizer Vent approval of installed system.
(PORV) is part of FNP-1 Tech. Spec. Upgrade l
Package Section 4.5.
l II.B.3 Postaccident Not currently in FNP-1 Tech. Spec.
Sampling Part of FNP-1 Tech. Spec. Upgrade Package Section 6.8.3.d.
j II.D.3 Valve Position Presently in FNP-1 Tech. Spec. Tables 3.3-11 l
Indication and 4.3.7.
Also, part of FNP-1 Tech. Spec. Up-grade Package Tables 3.3.11 and 4.3.7.
l l
Encl. 2 2
J. M. FARLEY NUCLEAR PLANT UNIT 1 STATUS OF NUREG-0737 TECHNICAL SPECIFICATION REQUIREMENTS TECHNICAL 0737 SHORTENED SPECIFICATION ITEM TITLE SECTION' COPMENT/ EXTENSION PZQUESTED
.II.E.1.2 Auxilliary Feed-Initiation is part of current FNP-1 Tech.
Water System Spec. Tables 3.3-3, 3.3-4, 3.3-5 and 4.3-2.
Initiation &
Flow is part of current FNP-1 Tech. Spec.
Fl ow -
Tables 3.3-11 and 4.3-7.
These Tech. Specs.
are also part of the FNP-1 Tech. Spec.
Upgrade Package.
II.E.3.1 Emergency Presently in FNP-1 Tech. Spec. Section 3.4.4.
Power for-Also, part of FNP-1 Tech. Spec. U; grade Package Pressurizer Section 3/4.4.
Heaters l
II.E.4.2 IContainment Containment Isolatica Dependability is part l
Isolation of current FNP-1 Tecii. Spec. Section 3.3.2.1, l
Dependability 3.3.3.1 and 3.6.3.
These same Technical l
Specifications are also part of the FNP-1 Tech.
l l
Spec. Upgrade Package.
Containment purge l
valves are' part of FNP-1 Tech. Spec. Upgrade l
Package 3.6.1.7 L
II.F.1 Accident-1.
Noble Gas Monitor is part of FNP-1 Tech.
Monitoring Spec. Upgrade Package Table 3.3-6.
Not currently in FNP-1 Tech. Spec.
2.
Iodine / particulate sampling is contained in FNP-1 Tech.' Spec.-Section 6.8.3.b and is also part of FNP-1 Tech. Spec. Upgrade Package Section 6.8.3.d.
Encl. 2 3
J. M. FARLEY NUCLEAR PLANT UNIT 1 STATUS OF NUREG-0737' TECHNICAL SPECIFICATION REQUIREMENTS TECHNICAL 0737 SHORTENED SPECIFICATION ITEM TITLE-SECTION COMMENT / EXTENSION REQUESTED II.F.1 3.
Containment High-Range Monitor is part of (continued)
FNP-1 Tech.- Spec. Upgrade Package Table 3.3-6.
Not currently in FNP-1 Tech. Spec.
4.
Containment Pressure is not in FNP-1 4.
Containment pressure monitor has Tech. Spec.
been installed and is operational l
and is included in the current Unit 1 Upgrade to.the Tech.' Spec.
5.
Containment water level is part of FNP-1 l
Tech. Spec. Upgrade Package Table 3.3-11 and Table 4.3-7.
Not currently in FNP-1 Tech. Spec.
l 6.
Containment Hydrogen is part of current L
FNP-1 Tech. Spec. Section 3.6.4.
l l
II.F.2
' Instrumentation Subcooling meter is presently in FNP-1 Tech. Spec.
Reactor vessel water level is to be in-l for Detection of
. Table 3.3-11 and 4.3-7.
Subcooling meter and corporated in Tech. Specs. upon approval Inadequate Core incore thermocouples are part of FNP-1 Tech.
by NRC of the FNP "Special" level Cooling Spec. Upgrade Package Table 3.3-11 and 4.3-7.
system. However, the BF3 prototype t
Reactor vessel water level is not in FNP-1 system is currently under a testing pro-
~
Tech. Specs.-.
gram described in APCo's letter of 6-29-81.
Upon completion of this test program a schedule for the Tech. Spec.
II.G.1 Power Supplies Presently pressurizer level indication and block
.will be developed.
for Pressurizer valves.are part of current Fl!P-1 Tech. Spec.
Relief Valves, Sections 3.8.2.1 and 3.8.2.2, and PORVs.in Sections
Encl. 2 4
J. M. FARLEY NUCLEAR PLANT UNIT 1 STATUS OF NUREG-0737 TECHNICAL SPECIFICATION REQUIREMENTS TECHNICAL 0737 SHORTENED SPECIFICATION ITEM TITLE SECTION COMMENT / EXTENSION REQUESTED II.G.1 Block Valves, &
3.8.2.3 and 3.8.2.4.
These Tech. Specs.
(continued) Indicators are also part of the FNP-1 Tech. Spec.
Upgrade Package.
II.K.3 Final 3.
Reporting SV & RV Failures & Challenges Recommendations is part of FNP-1 Tech. Spec. Upgrade B & 0 Task Force Package Section 6.9.1.10.
Not in current FPN-1 Tech. Spec.
j 10 & 12. Anticipatory Trip on turbine trip is part of current FNP-1 Tech Spec. Table 3.3-1.
These Tech. Specs are also part of the l
FNP-1 Tech. Spec. Upgrade Package.
l l,
III.A.2 Emergency Presently is Part of current FNP-1 Tech. Spec. Section Preparedness 6.8.1.
These Tech. Spec. are also part of the FNP-1 Tech. Spec. Upgrade Package.
III.D.1.1 Primary Coolant Presently is part of current FNP-1 Tech. Spec. Section Outside Contain-6.8.3.a.
These Tech. Specs. are a'.so part of the ment FNP-1 Tech. Spec. Upgrade Package.
l III.D.3.3 Inplant Radiation Presently in part of current FNP-1 Tech. Spec. Upgrade l
Monitoring Package Section 6.8.3.b.
These Tech. Specs are also part of the FNP-1 Tech. Spec. Upgrade Package.
Encl. 2 5
J. M. FARLEY NUCLEAR PLANT UNIT 1 STATUS OF NUREG-0737 TECHNICAL SPECIFICATION REQUIREMENTS 0737 SHORTENED SPECIFICATION ITEM TITLE SECTION COMMENT / EXTENSION REQUESTED III.D.3.4 Control-Room Presently is part of current FNP-1 Tech. Spec.
Habitability Section 3.3.3.6 and 3.7.7.1.
These Tech.
Specs, are also part of the FNP-1 Tech. Spec.
Upgrade Package.
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1 l
l l
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St 4
n ENCLOSURE 3 t _,_
4
{
i l
(
.i f
i J. M. FARLEY NUCLEAR PLANT UNIT 2 STATUS OF NUREG-0737 REQUIREMENTS l
--m m
__._m-m
F' i
Encl. 3 UNIT 2 NUREG-0737 STATUS 1
Clarifi -
Implemen-Tech.
1 c.
' catior.
Shortened tation Spec.s NRC' AFCo Response APCo Extension Item Title Description-Schedule _
Req.
- Remarks, Letter Date Remarks Requested <
I. A'.1.1 Shift technical
- 1. On shift Fuel Load Yes 1-14-81 Complete.
N/A-advisor
- 2. Training per Fuel Load No 1-14-81 Complete N/A LL Cat B
- 3. Describe long-Fuel Load No 1-14 Complete N/A term program I'. A.1. 2 Shift supervisor Delegate nonsafety Fuel Load No 6-20-80 Complete N/A-responsibilities duties I.A.1.3 Shift manning
- 1. Limit overtime Fuel Load No 2-23-81 Complete N/A
- 2. Minimum shift Fuel Load Yes Case by 1-14-81 Complete
!1/A crew case I.A.2.1
.Immediate upgrade
.1.
SR0 experience Fuel Load No 1-14-81 Complete N/A of RO & SR0
- training and qualifications 2..SR0s be R0s, Initial No 1-14-81 Complete N/A 1 -yr.
Criticality _
- 3. 3 mo training Fuel Load
. No 1-14-81 Complete N/A on-shift
- 4. Modify training Fuel Load No 1-14-81 Complete N/A
- 5. Facility certi-Fuel Load No 1-14-81 Complete N/A fication (I.A.2.3)
N/A - There are no further requirements-for this item.
.No - Remaining requirements are expected to be completed as scheduled.
Yes - An extension for this item was requested in the letter listed below.
r r
y r,
~ __
' Encl.'3.
~
UNIT 2 NUREG-0737 STATUS
~
2-Clarifi-Implemen.
Tech.
~
. cation Shortened tation Spec.
NRC APCo Response APCo Extension Item Title Description Schedule Req.
Remarks Letter Date Remarks Requested i
l 1.A.2.3 Administration Instructors com-2 mo prior No 1-14-81 Complete N/A of training plete SR0 exam to issuance programs of license I.A.3.1 Revise scope and
- 1. Increase scope 10-1-80 No 8-1-80 Complete N/A criteria for licensing exams l
- 2. Increase pass.
10-1-80 No 8-1-80 Complete N/A ing grade
- 3. Simulator exams 10-1-81 No 1-14-81 Complete N/A
- b. All plants 2-9-81 (Plant specific
-simulator exams to be given by 7-1-83)
I.B.1.2 Evaluation of Organization,-
Fuel Load Yes Draft 8-8-80 Complete N/A organization resources tng. &
guideline 1-14-81 and management qualifications available for operators'&
accidents I.C.1 Short-term
- 2. Inadequate Fuel Load No 10-24-79 Complete N/A core cooling 1-14-81
- a. Reanalyze and propose guidelines l
l
UNIT 2 NUREG-0737 STATUS Encl. 3 Clarifi-Implemen-Tech.
cation Shortened tation Spec.
NRC APCo Response APCo Extension l
Item Title Description Schedule Req.
Remarks Letter Date Remarks Requested t
j
- b. Revise First No 1-14-81 Suspense:
No l
procedures refueling end of 1st i
outage after refuel l
1/1/82 outage See Encl. 7, j
item 1
- 3. Transients &
Fuel Load No 10-24-79 Complete N/A accidents 1-14-81 l
- a. Reanalyze &
propose guidelines
- b. Revise Fi rst No 1-14-81 Suspense No procedures refueling end of 1st outage after refuel out-1/1/82 age See Enci 7, l
item 1 l
1.C.2 Shift &
Revise procedures Fuel Load No 6-20-80 Complete N/A l
relief turn-to assure plant l
over procedures status known by l
new shift i
I.C.3 Shift superviscr Corporate direc-Fuel Load Yes 6-20-80 Complete N/A l
responsibility tive to establish i
command duties and revise plant procedures l
I.C.4 Control-room Establish author-Fuel Load No 6-20-80 Complete N/A l
access ity and limit access l
1.C.5 Feedback of Review & revise Prior to No 1-14-81 Complete N/A operating procedures issuance of experience OL l
l
UNIT 2 NUREG-0737 STATUS Encl. 3 Clarifi-Implemen-Tech.
cation Shortened tation Spac.
NRC APCo Response APCo Extension Item Title Description Schedule Req.
Remarks Letter Date Remarks Requested I.C.6 Verify correct Revise performance Fuel Load No 2-23-81 Complete N/A performance of procedures operating activities I.C.7 NSSS vendor rev
- 1. Low-power test Fuel Load No 2-9-81 Complete N/A of proc program
- 2. Power ascension Full Power No 2-9-81 Complete N/A and emergency procedures I.C.8 Pilot mon of Correct procedure Full Power No 6-30-80 Complete N/A selected emer-based on NRC i
gency proc for sample audit NT0Ls I.D.1 Control-room Preliminary Prior to No Guidance 2-23-81 Awaiting No design reviews assessment &
issuance of and finalization schedule for OL schedule of require-correcting being ments by NRC deficiencies developed See Enci 7, item 2 I.D.2 Pl ant-safety-
- 1. Description TBD No Guidance 1-14-81 Awaiting No parameter display and sche-11-16-81 finaliza-console dule being tion of developed require-in NUREG-ments by 0696 NRC
- 2. Installed TBD No Guidance 1-14-81 Awaiting No and sche-11-16-81 finaliza-dule being tion of developed require-in NUREG-ments by 0696 NRC i
l l
UNIT 2 NUREG-0737 STATUS Encl. 3 Clari fi-Implemen-
- Tech, cation Shortened tation Spec.
NRC APCo Response APCo Extension Item Title Description Schedule Req.
Remarks
__ Letter Date Rengrks Requested
- 3. Fully imple-TBD No Guidance 1-14-81 Awaiting No mented and sche-11-16-81 finaliza-dule being tion of developed require-in NUREG-ments by 0696 NRC I.G.1 Training during
- 1. Propose tests Fuel Load No 6-20-80 Complete N/A low-power testing l
- 2. Submit analysis Fuel Load Yes 9-2-80 Complete N/A and procedures 9-11-80
- 3. Training and Full Power No 5-18-81 Complete N/A results II.B.1 Reactor-coolant-
- 1. Design and Full. Power No 1-14-81 Complete N/A system vents analysis J
l
- 2. Install 7-1-82 Yes 1-14-81 Complete N/A
- 3. Procedures 1-1-82 Yes 1-14-81 Suspense:
No when vent system is operational and NRC app-l roved See Encl. 7, item 3 II.B.2 Plant shielding
- 1. Radiation and Four months No 1-14-81 Complete N/A shielding before 0.L.
review
- 2. Correcting Full Power No 1-14-81 Complete N/A actions to assure access
UNIT 2 NUREG-0737 STATUS Encl. 3 6
Clarifi-Impicmen-Tech.
cation Shortened tation Spec.
NRC APCo Response APCo Extension Item Title Description Schedule Req.
Remarks Letter Date Remarks Requested
- 3. Complete mods 1-1-82 No 2-23-81 Suspense:
Yes 4-1-82 The Unit 2 Full Power License requires this item hv d 1-82.
See Enci 7, item 14
- 4. Equipment Four months No 2-23-81 Suspense:
No qualification before 0.L.
4-1-82 II.B.3 Postaccident
- 1. Design review Four months No 2-9-81 Complete N/A sampling before 0.L.
- 2. Corrective Full Power Yes 2-9-81 Complete N/A actions
- 3. Procedures Full Power Yes 2-9-81 Complete N/A
- 4. Complete 1-1-82 Yes 2-9-81 Complete N/A actions II.B.4 Training for
- 1. Develop train-Fuel Load No 2-9-81 Complete N/A mitigating core ing program 1
damage l
I l
I i
UNIT 2 NUREG-0737 STATUS Encl. 3 Clarifi-Implemen-
- Tech, 7
. cation Shortened tation Spec.
NRC APCo Response APCo Extension Item Title Description Schedule Req.
Remarks Letter Date Remarks Requested
- 2. Complete Full Power No 2-9-81 Complete N/A training II.D.1 Relief and
- 1. Describe pro-Fuel Load No 7-17-80 Complete N/A safety-valve gram & schedule test requirements
- 2. RV & SV tests Fuel Load TBD 2-9-81 Additional N/A 6-25-81 information 9-30-81 to be pro-vided upon completion of the EPRI program.
- 3. Block valve 7-1-82 TBD 2-9-81 Suspense: No tests 7-1-82 II.D.3 Valve position Install in control Four months Yes 1-14-81 Complete N/A indication room before 0.L.
II.E.1.1 Auxiliary feed-
- 1. Analysis Full Power No 6-20-81 Complete N/A water system evaluation
- 2. Modification Full Power No 1-14-81 Complete N/A II.E.1.2 Auxiliary
- 1. Initiation Four months Yes 2-9-81 Suspense:
No feedwater system (a) Control before 7-1-81 1st Refuel initiation and grade Fuel Load Outage flow (b) Safety grade
- 2. Flow indication Four months Yes 2-9-81 Suspense:
No (a) Control before 7-1-81 1st Refuel grade Fuel Load Outage (b) Safety grade
UNIT 2 NUREG-0737 STATUS Encl. 3 8
Clarifi-Impl emen-
- Tech, cation Shortened tation Spec.
NRC APCo Response APCo Extension Item Title Description Schedule Req.
Remarks Letter Date Remarks Requested II.E.3.1 Emergency power Installed ccpa-Four months Yes 1-14-81 Complete N/A for pressurizer bility prior to heaters issuance of SER II.E.4.1 Dedicated hydro-
- 1. Design Four months Yes 1-14-81 Not N/A gen penetrations before 0.L.
required
- 2. Review and Fuel Load No 1-14-81 Complete N/A revise H2 control proc
- 3. Install 7-1-81 No 1-14-81 Complete N/A l
II.E.4.2 Containment 1-4 Implement Prior to Yes 1-14-81 Complete N/A isolation diverse issuance of t
dependability isolation OL
- 5. Containment 7-1-81 Yes 2-13-81 Complete N/A press setpoint
- 6. Containment
. Prior to Yes 1-14-81 Complete N/A purge valves issuance of OL
- 7. Radiation 7-1-81 Yes 1-14-81 Complete N/A signal on purge valves II.F.1 Accident-
- 1. Procedures Fuel Load No 7-24-80 Complete N/A monitoring instrumentation
- 2. Install instru-mentation
- a. Noble gas 1-1-82 Yes 2-13-81 Complete N/A monitor See Enci 7, item 5
- b. Iodine /
1-1-82 Yes 2-13-81 Complete N/A l
See Enci 7, particulate item 5 sampling 1
~. -
l UNIT 2 NUREG-0737 STATUS Encl. 3 9
l-- LClarifi-Implemen-Tech.
I cation Shortened tation Spec.
NRC APCo Response AFCo Extension-Item' Title Description Schedule Req.
Remarks Letter Date Remarks Requested i
- c. Containment 1-1-82 Yes 2-13-81 Complete
'N/A high range l
monitor l
l
- d. Containment 6 mo prior Yes 2-13-81 Complete See pressure to. issuance Enci 4, of OL item-II.F.1.d i
- e. Containment 7-1-82 Yes 2-13-81 Complete N/A water level
- f. Containment 1-1-82 Yes 2-13-81 Complete N/A hydrogen II.F.2 Instrumantation
- 1. Procedures Fuel Load.
No 6-20-80 Complete N/A for detection of instruments inadeq0 ate core-cooling
- 2. Subcooling Fuel Load Yes 6-20-80 Complete N/A meter
- 3. Describe other Fuel Load No 6-20-80 Complete N/A instrumentation
- 4. Install addi-1-1-82 Yes 2-9-81 Installa-No tional instru-6-24-81 tion com-menation 6-29-81 pl ete.
10-6-81 Testing to 6-20-80 be com-7-17-80 pleted prior 7-24-80 to startup 8-1-80 from the
.8-6-80 1st Refuel 8-19-80 outage per 1-14-81 6-29-81 let-
.ter.
Quali-fication of equipment I
and detailed operating.
procedures
UNIT 2 NUREG-0737 STATUS Encl. 3 10 Clarif f-Implemen-Tech.
cation Shortened tation' Spec.
NRC APCo Response APCo
.. Extension Item Title Description Schedule Req.
Remarks Letter Date Remarks Requested
'to be deve-loped after system acceptance.
II.G.1 Power supplies Power supply from Fuel Load Yes 1-14-81 Complete N/A
.for pressurizer emergency buses relief valves, block valves, &
level' indicators II.K.1 IE Bulletins
- 5. Review ESF Fuel Load Yes 6-20-80 Complete ~
N/A valves
- 10. Operabili'j Fuel Load No 6-20-80 Complete N/A status
- 17. Trip per Fuel Load Yes 6-20-80 Complete N/A low-level B/S II.K.2 Orders on B&W
- 13. Thermal 1-1-82 As 1-14-81 See Encl No Plants Mechanical Required-W.0.G.81-138 7, Item 6 Report dated 4-1 5-20-81~
i
- 17. Voiding in RCS 1-1-82 No 1-14-81 Complete N/A See Enci 7, Item 7
- 19. Benchmark 1-1-82 No 1-14-81 Not-N/A-analysis seg NRC letter required AFW flow 7-6-81 II.K.3 Final recommenda-
- 1. Auto PORY-1st refuel Yes 5-26-81 Complete N/A tions, B&O task isolation 6 mo. after force staff approval
- 2. Report on
-Four months No 1-14-81 Complete N/A PORY failures prior to
'WCAP-9804 1-1-81 3-13-81 a
UNIT 2 NUREG-0737 STATUS Encl. 3 11 Clariff-Implemen-Tech.
cation Shortened tation Spec.
NRC APCo Response APCo Extension Item Title Description Schedule Req.
Remarks Letter Date Remarks Requested
- 3. Reporting SV &
Four months Yes 1-14-81 System N/A RV failures &
before 0.L.
in place challenges
- 5. Auto trip of RCPs
- a. Propose mods Prior to 0.L.
No 1-14-81 Suspense:
No Within 3 months after NRC deter-mination of accepta-bility of SB LOCA Model See Encl. 7, item 15
- b. Modi fy Full Power Yes 1-14-81 Suspense:
No Within 11 months after NRC deter-mination of accepta-bility of SB LOCA model during an appropriate outage or 1st such outage thereafter
- 9. PID controller Four months No 6-20-80 Complete N/A before 0.L.
- 10. Applicant's Four months Yes 9-16-80 Complete N/A propose anti-before 0.L.
cipatory trip at high power
UNIT 2 NUREG-0737 STATUS Encl. 3 Clarifi-Implemen-Tech.
cation Shortened tation Spec.
NRC APCo Response APCo Extension
_ Item Title Description Schedule Req.
Remarks Letter Date Remarks Requested
- 11. Justification Fuel Load No See NUREG-1-14-81 Complete N/A use of certain
- 0611, PORVs Sect.
3.2.4.d.
- 12. Confi nn anti-cipatory trip l
- a. Propose Four months No 6-20-80 Complete N/A l
modifica-before 0.L.
l tions l
l
- b. Modify Four months Yes 6-20-80 Complete N/A before 0.L.
- 17. ECCS outages In accord-As 2-13-81 Suspense:
No l
ance with required Every 5 review year period schedule after ini-for licens-tial criti-ing cality
- 25. Power on pump seal s
- a. Propose 6 months No 1-14-81 Complete N/A mods prior to SER
- b. Modifica-Full Power Yes 1-14-81 Complete N/A tions See Encl 7, item 8
- a. Schedule In accordance No Westinghouse Complete N/A outline with review letter dated schedule 11-25-81 NS-EPR-2524
- b. Model In accordance No Westinghouse Complete N/A with review letter dated See Encl 7, I
schedule 11-25-81 item 9 l
UNIT 2 NUREG-0737 STATUS Encl. 3 13 Clarifi--
Impl emen-Tech.
-cation Shortened tation.
Spec.-
NRC APCo Response APCo Extension Item Title Description' Schedule Req.
Remarks Letter Date
~ Remarks Requested-l
- c. New In accordance No Westinghouse Complete N/A-analyses with review letter dated See. Encl 7, schedule
'11-25-81 item 9
- 31. Plant-1-1-83 No 1-14-81 Suspense:
No specific 1-1-83 analysis III.A.1.1 Emergency Short-term Fuel Load No Use NUREG-11-7-80 Complete N/A preparedness, improvements 0654 until short term Rev. 1.is issued (due out
-10/80)
III.A.1.2 Upgrade emer-
- 2. Design TBD TBD 2-13-81 Complete N/A 5-19-81
- 3. Modifications TBD TBD 2-13-81 Suspense:
No 5-19-81 10-1-82 III.A.2 Emergenqy
- 1. Upgrade emer-
. Fuel. Load No 11-7-80 Complete N/A preparedness
.gency plans
.to App E, 10 CFR 50
- 2. Meteorological Fuel Load No 2-9-81 Suspense:
No data 7-1-82 III.D.1 Primary coolant Measure leak Full Power Yes 2-9-81 Complete N/A outside con-rates & establish 9-8-81 tainment program to keep leakage ALARA l
l
UNIT 2 NUREG-0737 STATUS Encl. 3 14 Clarift-Implemen-'
Tech.
cation Shortened tation Spec.
NRC
~APCo Response
'APCo.
Extension
-Item Title Description Schedule Req.
Remarks Letter Date Remarks ~ ? Requested III.D.3.3 Inplant I2
- 1. Provide means Fuel Load Yes 1-14-81 Complete N/A radiation to determine monitoring
-presence of radiofodine
- 2. Modifications Prior to Yes' 1-14-81 Complete N/A to accurately licensing measure radio-iodine III.D.3.4 Control-room
- 1. Identify and Full Power No 7-17-80 Complete
- N/A' habitability evaluate
. potential haz-ards
- 2. Schedule for Fall Power No 7-17-80 Complete N/A modifications
- 3. Modifications Full Power
.Yes 7-17-80.
Complete N/A' I
ENCLOSURE 4 J. M. FARLEY NUCLEAR PLANT UNIT 2 STATUS OF NUREG-0737 TECHNICAL SPECIFICATION REQUIREMENTS
1 Encl. 4 1
J. M. FARLEY NUCLEAR PLANT UNIT 2 STATUS OF NUREG-0737 TECHNICAL SPECIFICATION REQUIREMENTS TECHNICAL 0737 SHORTENED SPECIFICATION ITEM TITLE SECTION COMMENT / EXTENSION REQUESTED I.A.1.1 Shift Technical Part of current FNP-2 Tech. Spec. Section Advisor 6.2.4, Figure 6.2-2, and Table 6.2-1.
I.A.1.3 Shift Manning Part of current FNP-2 Tech. Spec. Section 6.2.2 and Table 6.2-1.
I.B.1.2 Evaluation of Part of current FNP-2 Tech. Spec. Section Organization &
6.2.3.
Management I.C.3 Shift Super-Part of current FNP-2 Tech. Spec. Section visor Respon-6.1.2, 6.2.2 and Table 6.2-1.
sibility I.G.1 Training During Not part of current FNP-2 Tech. Specs.,
Low-Power however, training has been completed.
Testing II.B.1 Reactor-Reactor Vessel Head Vents are not in FNP-2 Reactor Vessel Head Vents a e installed.
Coolant-Tech. Specs.
A Tech. Spec. will be written upon NRC System Vents Pressurizer Vent (PORV) is part of current approval of installed system.
FNP-2 Tech. Spec. Section 3.4.5.
Encl. 4 J. M. FARLEY NUCLEAR PLANT UNIT 2 2
STATUS OF NUREG-0737 TECHNICAL SPECIFICATION REQUIREMENTS TECHNICAL 0737 SHORTENED SPECIFICATION ITEM TITLE SECTION COMMENT / EXTENSION REQUESTED II.B.3 Postaccident Part of current FNP-2 Tech. Spec. Section Sampling 6.8.3.e.
II.D.3 Valve Position Part of current FNP-2 Tech. Spec. Table Indication 3.3-11 and 4.3-7.
II.E.1.2 Auxiliary Part of current FNP-2 Tech. Spec. Section Feedwater 3.3.3.8 and Tabics 3.3-3, 3.3-4, 3.3-5, 4.3-2, System 3.3-11 and 4.3-7.
Initiation and Flow II.E.3.1 Emergency Part of current FNP-2 Tech. Spec. Section Power for 3.4.4.
Pressurizer Heaters II.E.4.2 Containment Part of current FNP-2 Tech. Spec. Sections Isolation 3.3.2, 3.3.3.1, 3.6.3 and 3.6.1.7.
Dependability i
l l
i Encl. 4 3-J. M. FARLEY NUCLEAR PLANT UNIT 2
-STATUS OF NUREG-0737 TECHNICAL SPECIFICATION REQUIREMENTS TECHNICAL l
0737 SHORTENED SPECIFICATION ITEM TITLE SECTION COMMENT / EXTENSION REQUESTED l
II.F.1 Accident a.
Noble Gas Monitor is part of current Monitoring FNP-2 Tech. Spec. Table 3.3-6 and Instrumentation 4.3.3.
b.
Iodine / Particulate Sampling is part of current FNP-2 Tech. Spec. -Section 6.8.3.e.
l c.
Cuntainment High Range Monitor is part of current FNP-2 Tech. Spec. Table 3.3-6 and 4.3-3.
j d.
Containment Pressure is not part of,FNP-2 d.
Containment pressure monitor has Tech. Spec.
been installed and is operational.
and will be included in the Unit 2 l
e.
Containment Water Level is part of current Upgrade of the Tech. Spec. upon l
FNP-2 Tech. Spec. Table 3.3-11 and 4.3-7.
receipt of_the Unit 1 Tech. Spec.
j
- Upgrade, f.
Containment Hydrogen is part of current i
j FNP-2 Tech. Spec. Section 3.6.4.
l II.F.2 Instrumentation Subcooling Meter and Incore Thermocouples are Reactor Vessel Water Level is to be for Detection of part of current FNP-2 Tech. Spec. Table 3.3-11 incorporated in Tech. Spec. upon appro-Inadequate Core and 4.3-7.
Reactor Yessel Water Level is not val by NRC of the FNP "special" level in FNP-2 Tech. Specs.
system. However,' the BF3 prototype -
system is currently under a testing program described in.APCo's letter of l
6-29-81.
Upon completion of this test program a schedule for the Tech. Spec.
will be developed.
l
Encl. 4 J. M. FARLEY NUCLEAR PLANT UNIT 2 STATUS 0F NUREG-0737 TECHNICAL SPECIFICATION REQUIREMENTS TECHNICAL 0737 SHORTENED SPECIFICATION ITEM TITLE SECTION COMMENT / EXTENSION REQUESTED II.G.1 Power Supplies Part of current FNP-2 Tech. Spec. Sections for Pressurizer 3.8.2.1, 3.8.2.2, 3.8.2.3, and 3.8.2.4.
Relief Valves, Block Valves,
& Level Indicators II.K.3 Final Recom-3.
Reporting SV & RV Failures and Challenges mendations, is part of current FNP-2 Tech. Spec. Section B&O Task Force 6.9.1.10.
110 & 12. Anticipatory Trip on Turbine Trip is part of current FNP-2 Tech. Spec.
Table 3.3-1.
III.D.1.1 Primary Coolant Part of current FNP-2 Tech.-Spec. Section Outside Contain-6.8.3.a.
ment III.D.3.3 Inplant 12 Part of current FNP-2 Tech. Spec. Section Radiation 6.8.3.b.
Monitoring I
III.D.3.4 Control-Room Part of current FNP-2 Tech. Spec. Sections Habitability
-3.3.3.6 and 3.7.7.
r
ENCLOSURE 5 J. M. FARLEY NUCLEAR PLANT UNIT 2 STATUS OF FULL POWER LICENSE REQUIREMENTS I
i l
I.
Encl.- 5 UNIT 2 LICENSE STATUS 1
Licznse NUREG-0737 License APCo Response APCo Extension Raftrence Reference Subject Due Date Letter Date Remarks Requested 2.C.(6)
Appendix R, 10 CFR 50.48 Per 3-19-81 3-19-81 See Enclosure 6, Yes APCo letter item 9 l
l 2.C.(7) 90 day report 6-29-81 6-25-81 Complete N/A III.A.1.2 Upgrade Emergency Support 5-17-81 5-19-81 Conceptual design N/A Facilities Submitted l
10-1-82 5-19-81 Suspense:
10-1-82 No III.A.2 Long Term Emergency Pre-3-1-82 1-14-81 Suspense: 3-1-82 No paredness 7-1-82 2-9-81 Suspense: 7-1-82 No l
9-1-82 Suspense:
9-1-82 No l
6-1-83 Suspense:
6-1-83 No Appendix B Emergency l
Preparedness Evaluation Report '
l Section B Onsite Emergency Organiza-9-1-81 Complete N/A l.
tion Section E Notification Methods and 6-1-81 5-21-81 Complete
.N/A Procedures 8-31-81 7-23-81, 10-20-81 11-1-81 6-29-81 5-21-81, 2-23-81 Complete N/A 10-1-81 9-4-81 11-1-81 10-20-81 l
12-15-81 2.C.(8)
Complete modifications to 6-29-81 6-29-81 Complete N/A both service water loops as one-time only exemption of T.S. 3.7.4 2.C.(9)(a)
. Provide results of seven Prior to 5-18-81 Complete N/A augmented low power tests exceeding 5% Power N/A - There are no further requirements for this item.
No - Remaining requirements are expected to be completed on schedule.
Yes - An extension for this item was requested in the letter listed below.
Encl.'S UNIT 2 LICENSE STATUS 2
License NUREG-0737 License APCo Response APCo Extension R:ference Reference Subject Due Date Letter Date Remarks Requested 2.C.(9)(b)
Provide results of test of Within 60 9-16-81 Now part of N/A-natural circulation with days after NRC letter dated 2.C.(12)(c) test boron mixing operation 10-20-81 to be run, if for 25,000 required, at the MW(e) days 1st Refuel Outage 2.C.(10)
II.B.3 Make fully operational post-Prior to 1-14-81 Complete N/A accident sampling system exceeding 5% Power 2.C.(11)
II.B.4 Complete training to miti-Prior to 1-14-81 Complete N/A gate core damage exceeding 5% Power 2.C.(12)(a)
Perform tests to demonstrate Prior to Complete N/A manual operation of. an at-exceeding mospheric steam dump valve 5% Power 2.C.(12)(b)
II.B.1 Make provisions or modifi-Prior to See Enclosure 7, No cations to assure the Startup item 3 l
I safety grade backup means following of the reactor coolant 1st Refuel-system depressurization is in accordance with require-ments 2.C.(12)(c)
Provide natural circulation Prior to 9-16-81 Suspense: ' Prior No cooldown procedures Startup to Startup follow-following ing ist Refuel, if ist Refuel required.
t l
'2.C.(13)
Revise procedures, modify Prior to 9-8-80 Complete N/A-l and test reset circuits for exceeding 9-12-80 l
containment air mixing fans, 5% Power 9-29-80
' containment purge. isolation 4-27-81 valves and auxiliary feed-water. pump discharge valve
Encl. 5 UNIT 2 LICENSE STATUS 3
License NUREG-0737 License APCo Response APCo Extension Reference Reference Subject Due Date Letter Date Remarks Requested 2.C.(14)
Demonstrate the operability Prior to Complete N/A of TS Table 3.3-12 fire exceeding detectors and TS Table 5% Power 3.4-1 reactor coolant system isolation valves 2.C.(15)
II.F.2 Complete modifications to 5-31-81 Complete N/A subcooling monitor system 2.C.(16)
Provide response to I.E.
5-17-81 5-12-81 Complete no later N/A l
Bulletin 80-11, including than 1st Refueling l
reevaluation report l
2.C.(17)
II.E.4.2 Provide design of modified 10-1-81 9-30-81 Complete N/A containment vent and purge 10-30-81 system to reduce use of 18" purge valves l
Install modified containment Prior to 9-30-81 Suspense: Prior No l
vent and purge system Startup 10-30-81 to Startup follow-following ing ist Refuel 1st Refuel 2.C.(18)(a)
Complete and have auditable 6-30-82 Suspense: 6-30-82 No records available at a central location to docu-ment compliance with envi-ronmental qualification requirements for Class IE equipment 1
2.C.(18)(b)(i)
Correct or commit to correct 7-14-81 7-1-81 See APCo resporise N/A noncompliance with 7-1-81 NUREG-0588 equipment quali-fication safety evaluation l
Encl. 5 UNIT 2 LICENSE STATUS 4
License NUREG-0737 License APCo Response APCo Extension Reference Reference Subject Due Date.
Letter Date Remarks Requested Commit to corrective actions 6-30-82 Additional informa-No 2.C.(18)(b)(ii) which will result in documen-tion will be pro-tation of compliance of vided in the applicable equipment with 12-28-81 submittal NUREG-0588.
2.C.(18)(c)
Qualify all safety-related 6-30-82 Suspense: 6-30-82 No electrical equipment in the facility Suspense: Prior No 2.C.(19)(a)
Prov ide additional evalua-Prior to tions of the Westinghouse Startup to Startup follow-fuel performance code.
following
.ing ist Refuel'
.lst Refuel Suspense:
No Complete modifications of Prior to 2.C.(19)(b)
. primary and backup circuit Startup Prior to Startup protection devices in con-following following ist tainment electrical pene-1st Refuel 1st Refuel (Seis--
tration circuits mic testing required prior to manufacturer's i
shipment 9/82 for 10/82 outage) 2.C.(19)(c)
Submit system final design 9-30-81 9-21-81 Complete N/A and implementation schedule of the modification of the
. lubrication system of=the two Fairbanks-Morse opposed l
piston diesel generators Modify lubrication system of Prior to 9-21-81 Suspense:
No the two Fairbanks-Morse
.Startup Prior to Startup opposed-piston diesel following following 1st generators 1st Refuel Refuel
Encl. 5 UNIT 2 LICENSE STATUS 5
License NUREG-0737 License APCo Response APCo Extension Reft.rence Reference Subject Due Date Letter Date Remarks Requested 2.C.(19)(d)
Inspect main steam turbine Prior to 2-1-82 See APCo letter Yes for low pressure rotor disc Startup letter 2-1-82 cracking or replace rotors following 1st Refuel 2.C.(20)
Provide a schedule for 4-30-81 3-30-81 See 11-16-81 N/A bringing the facility into 11-16-81 APCo letter compliance with Rev. 2 to R. G. 1.97 2.C.(21)(a)
I.C.1 Complete upgrading of emer-Prior to See Enclosure 7, No gency procedures and asso-Startup item 1 ciated operator training following ist Refuel f
after 1-1-82 i
2.C.(21)(b)
II.B.1 Submit design description 7-1-81 6-25-81 Complete N/A l
and operating procedures WOG letter for reactor coolant system 11-30-81 l
vents Complete installation of 7-1-82 Complete N/A reactor coolant system i
l vents 2.C.(21)(c)
II.B.2 Complete modifications to 4-1-82 Suspense:
4-1-82 Yes I
assure access to vital areas See and protection of safety Encl. 7, equipment following a item 14 l
degraded core accident 2.C.(21)(d)(1)
II.D.1 Report qualification of 10-1-81 6-25-81 APCo is a part N/A relief valves and piping 9-30-81 of EPRI program and will provide additional infor-mation as it be-comes available
Encl. 5
. UNIT 2 LICENSE STATUS 6
Lic:nse NUREG-0737
. License APCo Response-APCo Extension R;ference'
. Reference
. Subject
-Due Date Letter Date Remarks Requested-2.C.(21)(d)(2)
II.D.1 Report qualification of 7-1-82
. Suspense:
'7-1-82' No block valves APCo is partici-pating in EPRI submittal l
i 2.C.(21)(e)
II.E.1.2 Submit design of modifica-7-1-81 7-1-81 Complete N/A.
l tions to the control and protection circuits for the auxiliary feedwater systems Suspense: Prior No Modify control and protec-Prior to tion circuits for auxiliary Startup to Startup follow-feedwater system following ing ist Refuel 1st Refuel 2.C.(21)(f)(1)
II.F.1 Install noble gas effluent 1-1-82 Complete-N/A monitors See Enclosure 7, item 5 j
l 2.C.(21)(f)(2)
II.F.1 Install capability for con-Prior to Complete N/A l
tinuous sampling of plant
-exceeding See Enslosure 7, l
gas effluents 5% Power item 5 l
Complete N/A L
2.C.(21)(f)(3)
II.F.1 Install high-range radio-1-1-82 activity monitors in con-l tainment 2.C.(21)(f)(4)
II.F.1 Provide a description of 6-1-81 6-1-81 Complete N/A containment pressure in-struments
' Install containment pressure 1-1-82 Complete N/A instruments' s
u_____
" ' ~ - ' ' ' ' ' ' '
Encl. 5 UNIT 2 LICENSE STATUS 7
License NUREG-0737 Reference
_ Reference Subject License APCo Response APCo Extension Due Date Letter Date
_ Remarks Requested 2.C(21)(f)(5)
II.F.1 Provide a description of the 6-1-81 6-1-81 containment water level Complete N/A measurement system Install containment water 1-1-82 level system Complete N/A 2.C.(21)(f)(6)
II.F.1 Provide a description of the 6-1-81 6-1-81 installed hydrogen indica-Complete N/A tion monitors Make modifications to the 1-1-82 Not required N/A installed hydrogen indica-tion monitors, if required 2.C.(21)(g)(1)
II.F.2 Provide detailed design in-7-1-81 6-29-81 formation for proposed Complete N/A reactor vessel water level instrument 2.C.(21)(g)(2)
II.F.2 Provide results of FNP-1 7-1-81 6-24-81 tests of proposed reactor Complete N/A vessel water level instru-ment 2.C.(21)(g)(3)
II.F.2 Provide planned program to 1-1-82 6-29-81 complete reactor vessel Complete N/A water level instrument development, and feasibility data thereto 2.C.(21)(h)(1)
II.K.2.13 Submit detailed analysis of 1-1-82 5-20-81 thermal mechanical conditions See Enclosure 7, No in reactor vessel during item 6 recovery, from small LOCA with an extended loss of all feedwater
Erel. 5 UNIT 2 LICENSE STATUS 8
License NUREG-0737 License APCo Response APCo Extension R ference Reference Subject Jue Date Letter Date Remarks Requested See Enclosure 7,
'N/A 2.C.(21)(h)(2)
II.K.2.17 Provide analysis of poten-1-1-82 tial for voiding in reactor item 7 coolant system 2.C.(21)(h)(3)
II.K.2.19 Provide benchmark analysis 1-1-82 NRC letter Not required N/A-of sequential feedwater flow 7-6-81 following loss of main feed-water 2.C.(21)(1)(1)
II.K.3.1 Provide information required 7-1-81 5-26-81 Complete N/A (i) & (ii)
II.K.3.2 by NUREG-0737 if automatic PORV isolation. system is required as the result of safety examination q
2.C.(21)(1)(1)
II.K.3.1 Complete installation an'd Prior to 5-26-81 Not required N/A (iii) testing Eof. modified automa-Startup tic isolation system following 1st Refuel j
l 6 months after NRC design l
2.C.(21)(1)(2)
II.K.3.5 Provide evaluation or design Within 3 Suspense:
Wi thin No (i) modification for tripping months 3 months after NRC of reactor coolant pumps-after NRC determination of in the. event of a small determination acceptability of break LOCA of accepta-small break LOCA bility of
_ model. See small break Encl osure - 7, LOCA model-item 15' g
'k f
./;
_.y
,,7 t
_ Encl.15 '
~
.f '; ' _..
a-
=
o UNIT 2 LICENSE STATUS
' 9 '.
^
t i~
a
^
-License'. ~
NUREG-0737 License ~
APCo Response APCo Exthnstonic, R2ference
. Reference Subject Due Date.,
Letter Date Remarks ~
Requested)',"
5 2.C.(21)(1)(2)'
II.K;3.5 Complete plant modifications yWithin 11:
Suspense: Within
.No '
(ii) for small break LOCA trip ~
months after 11 months'after ing of reactor coolant pumps NRC determi-NRC determination.
)
nation'of of 'acceptabil ity -
l-acceptability of small break 4
of small brea LOCA model during L
LOCA model an appropriateL
,$^','
~,
during an outage or first.
~-
' appropriate such: outage there-i
-l C
. outage or.
after.
s
.firstisuch-outage"there"<
after-l-
2.C.(21)(1)(3)
' fI.it.3.25 Submit. results. of analysis 1-1-82 See Enclosure 7, N/A (1)'
-to determine consequences item 8 s
of loss of cooling water l_
~
.to the reactor.. coolant,
- l; pump seal ~ coolers and describe any modifications necessary j
2.C.(21)(1)(3)
II.K.3.25 Complete any.necessary modi-' 7-1-82 See Enclosure 7, N/A L
(ii) fications to the reactor-item 8 ~
coolant pump seal coolers i
2.C.(21)(1)(4)
II.K.3.30 Submit revised small break 1-1-82 Complete N/A' (i)
LOCA model to account for.
See Enclosure 7,.
recent experimental data item 9 2.C.(21)(1)(4)
II.K.3.31 Submit results of plant 1-1-83 Suspense:
No l
(ii)
. specific. calculations using 1-1-83 NRC approved revised.small break LOC 4 model l
~~
Encl. 5 s
i i
UNIT 2. LICENSE STATUS 10.
l License NUREG-0737 License
-APCo Response
.APCo Extension Reference Reference' Subject Due Date
. Letter Date Remarks Requested ~
2.(D)
Implement commitments iden-8-1-81 7-31-81 Complete N/A tified in Chapter 15, items 15.1.d, f and g of Physical Security Plan f
l.
l t
l
-.e.-
r ENCLOSURE 6 J. M. FARLEY NUCLEAR PLANT STATUS OF UNITS 1 & 2 OTHER REQllIREMENTS
Encl. 6 0THER LICENSING REQUIREMENTS *
-1 NRC Due APCo Response Extension Item No.
Subject Date Date Remarks Requested 1.
Provide 90 day response on USI A-44 1-1-82 12-7-81 Complete N/A (Station-Blackout) (Unit 1 only) 2.
Complete an evaluation to extend the 1-1-82 To be submitted No qualified life of equipment (Environ-under separate mental Qualification) letter on 12-28-81 3.
Revise plant procedures (Control of 10-1-81 See Encl. 7, No Heavy Loads, NUREG-0612) item 10 4.
Correct remaining fire damper defi-Prior to Suspense: Prior No ciencies Startup to Startup fol-following lowing ist refuel 1st Refuel Unit 2 Prior to Suspense: Prior No Startup to Startup fol-following lowing 3rd Refuel 3rd Refuel See Encl. 7, item Unit i 11; I&E Audit Report dated 4-6-81 5.
Conduct UT (Flow splitters) inspection Prior to Complete N/A (Unit 1 only) _
Startup following 3rd Refuel 6.
Complete Hot Shutdown panel modifica-Prior to 9-30-80 Design complete No tions (Unit 2 only)
Startup Suspense: Prior following to Startup fol-1st Refuel lowi.g ist Refuel
- Unless otherwise noted, all items refer to both Units 1 and 2.
Encl. 6 OTHER LICENSING REQUIREMENTS
- 2 NRC Due APCo Response Extension Item No.
Subject Date Date Remarks Requested 7.
Oil Collection for Reactor Coolant Pump Prior to 3-19-81 See Encl. 7, See Startup item 13 Encl. 7, following item 13 3rd Refuel for Unit 1 Prior to 3-19-81 Suspense: Prior No Startup to Startup following following 1st Refuel 1st Refuel for Unit 2 See Encl. 7, item 13 1
8.
Radiological Emergency Response 1-1-82 12-29-79 See Encl. 7, N/A l
(Class A Model) item 12 9.
Fire Protection Upgrade (10 CFR 50, 3-1-82 3-19-81 Plan and schedule Yes Appendix R) for Unit 2 provided in 3-19-81 referenced letter 10-1-82 3-19-81 Plan and schedule Yes i
for Unit 1 provided in 3-19-81 referenced letter 10.
RETS (10 CFR 50, Appendix I) Upgrade Unit 2 T.S.
30 days within To be included as N/A and Snubber Upgrade Unit 2 Upgrade receipt of issuance part of the Unit 2 Package of Unit 1 Upgrade Upgrade Package package currently scheduled for December,1981.
- ll.
Unit 2 T.S. Upgrade 30 days within This package will N/A receipt of issuance be identical.
of Unit 1 Upgrade (less specific currently scheduled unit differences for December,1981.
from the Unit 1 Upgrade) l j
Encl. 6 OTHER LICENSING REQUIREMENTS *-
'3
'NRC Due APCo Response Extension.
~ Item No.
Subject Date Date Remarks Requested l
12.
Station Blackout Training W.0.G.
See Enclosure 7, This training is N/A 11-30-81 item 1 a part of the prc,cedure upgrade required by NUREG-0737, Item I.C.1 13.
D.G. T.S. Upgrade (Unit 2) 10-28-81 Same as section N/A-11-6-81 in Unit 2 T.S.
Upgrade except for 26 second delay in startup of D/G 2C. Scheduled to be issued by NRC in January, 1982.
14.
FQ Upgrade 11-16-81 To be handled N/A 11-23-81 as part of the 12-14-81 Unit 2 T.S.
Upgrade package.
15.
Masonry Walls (IEB 80-11)
N/A Unit 1 5-22-81 None N/A Unit 2 5-12-81 1
16.
M5L Break With Continued AFW None 5-8-80 None N/A Addition (IEB 80-04) 17.
HELB (IEN 79-22)
None Unit 1 10-5-79 None N/A Unit 2 6-5-81 NRC 9-1-81 letter accepting position 18.
Seismic Qualification of AFW System None 10-9 Complete N/A
__m
__.__-._m-_
Encl. 6-OTHER LICENSING REQUIREMENTS
- 4-NRC Due APCo Response Extension Item No.'
Subject Date-Date Remarks' Requested 19.
Natural Circulation Cooldown (Generic None 11-13-81 Procedures to be N/A Letter 81-21 St. Lucie Event) developed 90 days t
after receipt of W.0.G. guidelines 20.
NRC Approval of IST Program For Unit 2 None Scheduled for APCo is develop-N/A' l
3-15 ing a-revised program scheduled for submitted to the NRC by 3-15-82 21.
Approval of the Unit 1 and 2 Technical NRC letter 9-23-81 Tech. Spec. will' Specifications For Containment Purge 8-5-81 be' revised just-Valves prior to -instal-lation'of-the 8 inch purge
(
val ves.
I i
i l
I i
ElCLOSURE 7 i
J. M. FARLEY NUCLEAR PLANT INITIAL SUBMITTALS AND EXTENSION BASIS DOCUMENTATION
INDEX Item Enclosure Item /
No Subject Number Paragraph 1.
I.C.1, Guidance for the evaluation 1
I.C.1 and development of procedures for 3
I.C.1' transients and accidents 5-2.C.(21)(a) 2.
I.D.1, Control room design reviews 1
1.D.1 3
I.D.1 3.
II.B.1, Reactor Coolant System vents 1
II.B.1 3
II.B.1 5
2.C.(12)(b) 4.
II.E.1.2, Auxiliary Feedwater 1
II.E.1.2 Initiative and Indication 5.
II.F.1, Accident-Monitoring Instru-1 II.F.1 ments 3
II.F.1 5
2.C.(21)(f)(1) 6.
II.K.2.13, Thermal Mechanical Report 1
II.K.2.13 3
II.K.2.13 5
2.C.(21)(h)(1) 7.
II.K.2.17, Voiding in RCS 1
II.K.2.17 3
II.K.2.17
-5 2.C.(21)(h)(2) 8.
II.K.3.25, Loss of AC on Pump Seals 1
II.K.3.25 3
II.K.3.25 5
2.C.(21)(1)(3)(1) 9.
II.K.3.30, Revise SBLOCA Methods 1
II.K.3.30 3
II.K.3.30 5
2.C.(21)(1)(4)(1) 10.
3 11.
Fire Damper Correction 6
4 12.
Radiological Emergency Response 6
8 13.
Reactor Coolant Pump, 011 Collection f
7-System Upgrade per 10 CFR 50, Appendix R 14.
II.B.2, Plant Shielding 1
II.B.2 3
II.B.2 5
2.C.(21)(c) 15.
II.K.3.5, Auto Trip of RCP during LOCA 1
II.K.3.5 3
II.K.3.5 5
2.C(21)(1)(2)(1)
Item 1.
I.C.1 GUIDANCE FOR THE EVALUATION AND DEVELOPMENT OF PROCEDURES FOR TRANSIENTS AND ACCIDENTS [ Enclosures 1 and 3, item I.C.1;, para. 2.C.(21)(a)]
Requirement 1.
By letter to the NRC dated January 14, 1981, Alabama Power Company made the following commitment on this subject related to the Farley Nuclear Plant.
The Westinghouse Owners Group submitted a detailed description of its program to comply with the requirements of Item I.C.1 on December 15,1980 (WOG Letter 80-179). The program identified previous owners group submittals to the NRC, which we believe will comprise the bulk cf the response. The additional effort required to obtain full compliance with this item, as discussed with the NRC on November 12, 1980, together with a schedule for completion, was also identified in the December 15, 1980, submittal.
If additional guidelines are developed, the Farley Nuclear Plant will utilize these guidelines to further upgrade appropriate procedures and provide the associated operator training by the first refueling outage for each unit after January 1,1982.
2.
Unit 2 License Item 2.C.(21)(a) states that prior to startup following the first refueling after January 1,1982, Alabama Power Company should complete the upgrading of emergency procedures and associated operator training.
Response
1.
Alabama Power Company, as a member of the Westinghouse Owners Group, is participating in a significant upgrade of emergency procedures with accompanying appropriate operator training as a result of this requirement. The Vestinghouse Owners Group submitted a detailed description cf the revised Emergency Response Guideline Program (ERG) to Item I.C.1 of NUREG-0737 in 0G-61 dated July 7, 1981. The Westinghouse Owners Group transmitted all completed portions of the ERG set to the NRC for their review in letter 0G-64 dated November 30, 1981. Alabama Power Company is confident that the materials in the ERG set which has been transmitted to the.NRC are sufficient to satisfy the requirements oi NUREG-0737, I.C.1.
The upgrade of plant specific procedures must be coordinated with other NRC criteria. These criteria include the main control room review (NUREG-0700 and 801), emergency response upgrade (NUREG 835, 696, 814, R. G. 1.97), and operating procedures (NUREG-0799).
Some of these NUREG are in draft form and have been issued for comment. To enable Alabama Power Company to implement an integrated system which addresses all criteria, such new criteria must be finalized by the NRC and included in the integrated efforts associated with the main control room review and emergency response upgrade (SPDS).
Item 1 I.C.1 (Cont'd)
Alabama Power Company received the approved Westinghouse Owners Group guidelines in December,1981. These guidelines will be utilized to upgrade plant specific procedures and provide associated operator training by the fourth refueling outage for Unit 1 (scheduled for the spring of 1983), and the first refueling outage for Unit 2 (scheduled for November,1982).
The above proposed schedule is contingent upon timely finalization of new NRC criteria associated with the main control room review, emergency response upgrade and operating procedures.
Item 2.
I.D.1 CONTROL ROOM DESIGN REVIEWS [ Enclosures 1 and 3, item I.D.1]
Requirement By letter of January 14, 1981, Alabama Power Company made the following commitment on this item.
Alabama Power Company intends to implement similar commitments described in the July 17, 1980 letter for Unit 1 prior to return to power after the current refueling outage (second) except for denoting normal, alert, and alarm ranges on other significant main control room meters (other than those described in emergency procedures) which will be completed by the end of the third refueling outage. Alabama Power Company will address the long term control room design review after issuance of NUREG 0700. Alabama Power Company intends to implement the commitments described in the July 17, 1980 letter for Unit 2 prior to exceeding 5% power except for denoting normal, alert, and alarm ranges on other significant main control room meters (other than those described in emergency procedures) which will be completed by the end of the first refueling outage.
Response
Alabama Power Company has completed the Unit 1 and 2 modifications associated with the commitments described in the July 17, 1980 letter.
Denotation of the Unit I normal, alert, and alarm ranges has been completed on the other significant main control room meters and will be completed for Unit 2 by the end of the first refueling outage scheduled to begin in November,1982.
NUREG 0700 was issued in September,1981 and Alabama Power Company is in the process of reviewing this NUREG and developing an implementation plan. Subsequently, additional draft NUREGs which affect the main control room review have been issued for comment.
These draft NUREGs include NUREG-0801, " Evaluation Criteria for Detailed Colitrol Room Design Reviaw"; NUREG-0793, " Draft Criteria for Preparation of Emergency Operating Procedures"; NUREG-0814,
" Methodology for Evaluation of Eiergency Response Facilities"; and NUREG-0835, " Human Factors Accep'ance C-iteria for the Safety Parameter Display System".
To enable Alabama Power Company to perform the main control room review in a manner that will be consistent with criteria currently in review (draft NUREGs), such new criteria must be finalized by the NRC and included in the Farley main control room review plan.
Alabama Power Company will provide an implementation plan and schedule for the main control room review upon finalization of the NRC criteria.
Item 3.
II.B.1 REACTOR COOLANT SYSTEM VENTS [ Enclosures 1 and 3, item II.B.1; Enclosure 5, para 2.C.(12)(b)]
Requirement 1.
By letter dated January 14, 1981, Alabama Power Company committed to the actions described below for this item.
The Westinghouse Owners Group, of which Alabama Power Company is a member, is developing generic procedure guidelines for the use of the reactor vessel head vent system which will be incorporated into the Farley Plant procedures when the Reactor Coolant System Yent System is operational and approved by the NRC.
The additional displays and controls added to the control room as a result of this requirement will be considered as part of the long-term human-factors analysis and will include:
a) the use of this information by an operator during both normal and abnormal plant conditions b) integration into emergency procedures c) integration into operator training, and d) other alarms during emergency and need for prioritization of alarms.
2.
Unit 2 License Item (2.C.(12)(b)) states that prior to startup following the first refueling, Alabama Power Company shall make provisions (or modifications) as necessary to assure that the safety grade backup means of-reactor coolant system depressurization is in accordance with the requirements of Table 1 in Branch Technical Position RSB 5-1, Rev. 1.
R_esponse 1.
This system has been installed and tested for Units 1 and 2.
This system will not be placed in an operational status on either unit until accepted by the NRC.
The Westinghouse Owners Group has developed generic procedures guidelines for the use of the reactor vessel head vent system.
Alabama Power Company will develop plant specific procedures based on the generic guidelines as described in response to item I. Col.
The additional displays and controls added to the control room as a result of this requirement will be included in the main control room review as described in response to item I.D.1.
l 2.
Alabama Power Company is reviewing, with Westinghouse Electric Corporation, the available options that address BTP-RSB 5-1, Rev.
1.
If conditions arise that prevent implementation of an acceptable safety grade backup depressurization system, Alabama j.
Power Company will notify the NRC Staff.
l
Item 4.
II.E.1.2 AUXILIARY FEE 0 WATER INITIATION AND INDICATION
[ Enclosure 1. item II.E.1.2]
Requirement By letter dated July 1,1981, from Alabama Power to the NRC:
In accordance with my letter dated December 4,1980 to Mr. R. L.
Tedesco and as required by(Farley Nuclear Plant Unit 2 Operating License NPF-8 Section 2.C. 21)(e), Alabama Power Company submits for NRC review-the attached description of proposed modifications to the Farley Nuclear Plant Auxiliary Feedwater control and protection design.
In order for Alabama Power Company to implement these changes on Farley Unit I during the third refueling outage, your concurrence is requested no later than August 1, 1981. These changes have been agreed to in principle as a part of the Farley Unit 2 full power licensing process.
Response
At the time that Alabama Power Company committed to implementing the proposed modifications on Farley Unit i during the third refueling outage, that outage was scheduled to begin on or about March 12, 1982.
Main generator problems forced Farley Unit 1 into an unexpected outage in early September,1981. Because of the extensive repairs required to restore the generator, a decision was made to refuel the reactor early thereby avoiding a refueling during the summer of 1982.
Design and procurement were criginally scheduled to support the outage beginning in March,1982. All NRC required modifications to the Auxiliary Feedwater have been completed except those identified in Alabama Power Company's December 4,1980 letter. These requirements will provide the operator with more positive control of the flow control valves as described in this letter. Design to implement requirements in the December 4,1980 letter was approved by the NRC in August, 1981. Procarement was delayed pending NRC approval to preclude unnecessary purchase of equipment. At the time of the unexpected outage a review was made to determine if icng lead time items could be procurred in time to support the revised outage schedule. Procurement lead time exceeded the original outage duration schedule. Subsequent to this review the outage duration expanded in a fragmented fashion as additional problems were discovered on the turbine generator.
In addition, the proposed design change is a major modification affecting a safety related system and would require significant review time by the plant staff to plan the implementation and verify that other safety related systems would not be impacted.
Outage manpower has already been allocated and any additions impact already scheduled work.
Alabama Power Company is continuing to review the impact of implementing these modifications and will make every effort to complete this item during the current outage but in any event all modifications will be completed by the end.of the fourth refueling outage currently scheduled for February of 1983. However, if any scheduled outage occurs of sufficient duration, Alabama Power Company will implement the auxiliary feedwater modification prior to startup following the fourth refueling outage.
Item 5.
II.F.1 ADDITIONAL ACCIDENT MONITORING INSTRUMENTATION
[ Enclosures 1 and 3, item II.F.1; Enclosure 5, para 2.C.(21)(f)(1)]
1.. Noble Gas Effluent Monitors Requirement By letter dated January 14, 1981 Alabama Power Company committed to develop procedures for operating, calitration, and dissemination of release rate information asociated with noble gas effluent monitors.
Response
The above procedures have been written for the Sping-4 uonitor and are being written for the condenser air removal exhaast and steam generator safety relief valves and atmospheric relief discharge monitors. The Unit 1 and 2 procedures currently being written are scheduled to be approved by January 1,1982. Procedures to disseminate release rate information have been approved for Units 1 and 2.
2.
Sampling and Analysis of Plant Ef fluents Requirement By letter dated January 14, 1981 Alabama Power Company committed that for vent stack effluents there will be a Victoreen vacuum pump with charco&l filters what will allow the Chemistry and
. Health Physics Group to draw 15 minute iodine and particulate samples to be analyzed in the. laboratory.
Response
A vacuum pump has been installed for Units 1 and 2 that will allow 15 minute samples to be drawn and analyzed in the laboratory.
i
Item 6. - II.K.2.13, THERMAL MECHANICAL' REPORT [ Enclosures 1 and 3, item II.K.2.13; Enclosure 5, para 2.C.(21)(h)(1)]
Requirement 1.
By letter to the NRC dated January 14, 1981, Alabama Power Company made the following commitment on this subject related to the Farley Nuclear Plant.
To completely address the NRC requiraments for a detailed analysis of the thermal-mechanical conditions existing in the reactor vessel during recovery from small breaks with an extended loss of all feedwater, Alabama Power Company,.as a member of the Westinghouse Owners Group, is participating in a program consisting of analysis for generic Westinghouse PWR plant groupings. The program will be completed and documented to the NRC by January 1,1982.
Following completion of this generic program, additional plant specific analysis, if required, will be provided. A schedule for the plant specific analysis will be determined based on the results of the generic analysis.
2.
Unit 2 License item 2.C.(21)(h)(1) states that prior to January 1, 1982, Alabama Power Company should submit a detailed analysis of-thermal-mechanical conditions in the reactor vessel during recovery from small LOCA with an extended loss of all feedwater.
3.
By letter to the NRC dated May 20, 1981, Alabama Power Company made the following commitment on this subject related to the Farley Nuclear Plant.
Verify by January 1,1982, completion of W.0.G. program and determine if future additional plant specific analysis add /or remedial actions are required.
Response
Westinghouse (in support of the Westinghouse Owners Group) is.
performing an analysis for generic Westinghouse plant groupings to address this issue which will be submitted to the NRC by the end of 1981. This generic study will be applicable to Farley Nuclear Plant.
Item 7.
II.K.2.17, VOIDING IN RCS [ Enclosures 1 and 3, item II.K.2.17,, para 2.C.(21)(h)(2)]
Requirement 1.
By letter to the NRC dated January 14, 1981, Alabama pner Company made the following commitment on this subject related to the Farley Nuclear Plant.
The Westinghouse Owners Group, of which Alabama Power Company is a member, is currently addressing the potential for void formation in the Reactor Coolant System (RCS) during natural circulation cooldown conditions, as described in Westinghouse letter NS-TMA-2298 iT. M. Anderson of Westinghouse to P. S. Check of the NRC). We be.lieve the results of this effort will fully address the NRC reqJirement for analysis to detennine the potential for voiding in the RCS during anticipated transients. A report describing the results of this effort will be provided to the 14RC by January 1,1982.
2.
Unit 2 License item 2.C.(21)(h)(2) states that prior to January 1, 1982, Alabama Power Company should submit an analysis of the potential for voiding in the reactor coolant system during anticipated transients.
Response
Westinghouse (in support of the Westinghouse Owners Group) has performed a study which addresses the potential for void fonnation in Westinghouse designed nuclear steam supply systems during natural circulation cooldown/depressurization transients. This study has been submitted to the NRC by the Westinghouse Owners Group letter 0G-57, dated 4-20-81, Jurgenson to Check and is applicable to the Farley Nuclear Plant.
In addition, the Westinghouse Owners Group has developed a natural circulation cooldown guideline that takes the results of the study into account so as to preclude void formation in the upper head region during natural circulation cooldown/depressurization transients, and specifies those conditions under which upper head voiding may occur. These Westinghouse Owners Group generic guidelines have been submitted to the NRC (letter 0G-64, dated 11-30-81, Jurgensen to Eisenhut). The generic guidance developed by the Westinghouse Owners Group will be utilized in the implementation of the Farley Nuclear Plant specific operating procedures as part of Alabama Power Company's NUREG-0737, item I.C.1 effort.
Item 8.
II.K.3.25, LOSS OF AC ON PUMP SEALS [ Enclosures 1 and 3, item II.K.3.25; Enclosure 5, para 2.C.(21)(1)(3)(1)]
Requirement 1.
NUREG-0737 require the submission of the evaluation and proposed modifications for the above item by January 1,1982.
2.
Unit 2 License item 2.C.(21)(1)(3)(1) states that-prior to January 1,1982, Alabama Power Company should submit results of analyses or experiments to determine consequences of a loss of cooling water to the reactor coolant pump seal coolers and decribe any modifications found necessary.
Response
This item requires that the consequences of a loss of RCP seal cooling due to a loss of AC power (defined as loss of offsite power) for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> be demonstrated.
During normal operation, seal injection flow from the chemical and volume control system is provided to cool the RCP seals and the component cooling water system provides flow to the thermal barrier heat exchanger to limit the heat transfer from the reactor coolant to the RCP internals.
In the event of loss of offsite power the RCP motor is deenergized and both of these cooling supplies are terminated; however, the diesel generators are automatically started and either seal injection flow or component cooling water to the thermal barrier heat exchanger is automatically restored within seconds. LEither of these cooling supplies is adequate to provide seal cooling and prevent seal failure due to loss of seal cooling during a loss of offsite power for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Item 9.
II.K.3.30, REVISE SBLOCA METHODS [ Enclosures 1 and 3, item II.K.3.30; Enclosure 5, para 2.C.(21)(1)(4)(i)]
Requirement 1.
NUREG-0737 requires the submission of additional information on the subject item by January 1,1982.
2.
Unit 2 License item 2.C.(21)(1)(4)(ii) states that prior to January 1,1982, Alabama Power Company should submit to the NRC a revised model to account for recent experimental data.
Response
This item requires that the analysis methods used by NSSS vendors and/or fuel suppliers for small-break LOCA analysis for compliance with Appendix K to 10 CFR Part 50 be revised, documented, and submitted for NRC approval.
Westinghouse feels very strongly and Alabama Power Company agrees that the small-break LOCA analysis model currently approved by the NRC for use on Farley Nuclear Plant is conservative and in conformance with Appendix K to 10 CFR Part 50. However, Westinghouse believes that improvement in the realism of small-break calculations is a worthwhile effort and has committed to revise its small-break LOCA analysis model to address NRC concerns (e.g., NUREG-0611, NUREG-0623, etc.). This revised Westinghouse model is currently scheduled for submittal to the NRC by April 1,1982 as documented in Westinghouse letter NS-EPR-2524, dated 11-25-81, Rahe to Eisenhut.
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Item 10. CONTROL ~ OF HEAVY LOADS [ Enclosure 6, Item 3]
Requirement In Mr. F. L. Clayton's letters to Mr. D. G. Eisenhut dated May 15, 1981, and June 24, 1981, concerning the above subject, Alabama Power Company comnitted to the following:
1.
Safe load paths inside containment will be defined in plant procedures.
2.
Alabama Power Company will revise plant procedures to minimize the time that the demineralizer hatch cover is positioned over the demineralizer during lifting operations.
3.
Alabama Power Company will develop procedures to cover the load handling operations for those loads listed in Table 3-1 by October 1, 1981. These procedures will meet the requirements of NUREG-0612, Section 5.1.1(2).
4.
Operator training will be revised to include the basic requirements of ANSI B30.2-1976 by October 1,1981.
5.
Crane inspection, testing and maintenance requirements will be revised to incorporate the t'esic requirements of Chapter 2-2 of ANSI B 30.2-1976 by October 1,1981.
6.
a.
Plant procedures will be reviewed with respect to the following:
(1) review of procedures for installation of rigging or lifting devices and movement of load to assure that sufficient detail is provided and that instructions are clear and concise; (2) visual inspections of load bearing components of cranes, slings, and special lifting devices to identify flaws or deficiencies that could lead to failure of the component; (3) appropriate repair and replacement of defective components; and (4) verify that the crane operators have been properly trained and are familiar with specific prcedures used in handling these loads, e.g., hand signals, conduct of operations, and content of procedures, b.
Crane operators will be trained per basic requirements of ANSI B30.2-1976 and plant procedures by October 1,1981.
7.
The lifting devices identified in 2.1.3c will comply with the inspection criteria and operator qdificaiton requirments of ANSI N14-6-1978 or ANSI 330.9-1971 as appropriate by October 1, 1981.
Response
1.
Safeload paths inside containment have been defined in plant mainteaance procedures.
Item 10. CONTROL OF HEAVY LOADS (Cont'd) 2.
By April 1,1982 a precaution will be added to plant procedures to minimize the time in which a cation bed demineralizer hatch cover is suspended over its respective demineralizer cubicle.
3.
Alabama Power Company has completed development of load handling procedures meeting the requirements of NUREG-0612 Section 5.1.1(2) for those loads listed in Table 3-1.
4.
Crane operators involved in the lifting of heavy loads at Farley Nuclear Plant have been trained in accordance with ANSI B30.2-1976.
Licensed reactor operators involved in fuel manipulations and personnel involved in radwaste handling and decontamination activities are exempted from this training since they do not lift heavy loads.-
l 5.
Crane inspection, testing and maintenance requirements have been revised to incorporate the basic requirements of Chapter 2-2 of ANSI B30.2-1976.
6Property "ANSI code" (as page type) with input value "ANSI B30.2-1976.</br></br>6" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..
a.
Plant procedures have been reviewed and found to comply with the appropriate criteria, b.
Crane operators involved in the lifting of heavy loads at Farley Nuclear Plant have been trained in accordance with ANSI B30.2-1976 and appropriate plant procedures.
7.
The lifting devices identified in 2.1.3c comply with the inspection criteria and operator qual.fication requirements of ANSI N14.6-1978 or ANSI 330.9-1971 as appropriate.
Item 11. FIRE DAMPER DEFICIENCY CORRECTION [ Enclosure 6, Item 43 Requirement
References:
1.
NRC inspection of March 10 - 13, 1981 Report No. 50-348/81-06.
2.
Letter from Mr. F. L. Clayton, Jr. to Mr. J. P. O'Reilly, dated May 11, 1981 The reference (1) inspection reports states:
"Section 4.4.4.1.10 of the Farley Nuclear Plant Fire Protection Program Reevaluation (FPPR) dated September 15, 1977 with Amendments 1 through 4 (Amendment 4 dated January 3,1979) states that modifications will be made to the ventilation duct penetrations of fire barriers to '.chieve compliance with the NRC guidelines of Appendix A to Branch Technical Position 9.5.1, Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1,1976.
Appendix A Section D.1(j) requires ventilation duct systems to I
be provided with " Fire door dampers" installed in accordance with l
National Fire Protection Association Standard No. 80 (NFPA-80), Fire Doors and Windows, at all locations where the ducts penetrate fire barriers. NFPA-80 Section 2-8,9 requires fire door assemblies to M installed in accordance with the manufacturer's installation instructions.
Contrary to the above the fire damper was not installed in accordance g
with the manufacturer s installation instructions".
The reference (2) letter states:
"All other Un' c 1 fire dampere will be inspected and any deficiencies corrected prior to the return to power at the next refueling outage."
l i
Response
l l
All Unit i fire dampers have been inspected. As of the present time l
not all of the deficiencies have been corrected. It is expected that these deficiencies will be corrected prior to return to power based upon the nature of the deficiencies, corrective responses received from engineering to date, and the time it has taken to perform the corrective action on those dampers already repaired.
At the time that Alabama Power Company committed to correcting the remaining fire damper deficiencies on Farley Unit 1 during the third refueling outage, that outage was scheduled to begin on or about March 12, 1982. Main generator problems forced Farley Unit 1 into an unexpected outage in early September,1981. Because of the extensive repairs required to restore the generator, a decision was made to refuel the reactor during this outage in lieu of a refueling outage in the summer of 1982. Alabama Power Company is making every attempt to correct these deficiencies during the current outage.
Item 12. RADIOLOGICAL EMERGENCY RESPONSE [ Enclosure 6, Item 8]
Requirement By letter dated December 29, 1979, Alabama Power Company connitted to provide the following by January 1,1982:
a.
Installation date for the emergency response facility (TSC, EOF, OSC) hardware and software.
b.
Primary and backup meteorological equipment, Class A DCM and remote interrogation capability of Appendix 2 to NUREG 0654.
Response
a.
The emergency response facility hardware and sof tware will be installed in the Technical Support Center by July 1,1982, and is scheduled to be completed for the Emergency Operations Facility by July 1982.
b.
The primary and backup meteorological equipment has been installed and is operational.
The Farley Nuclear Plant Class A DCM System software programs have been loaded into the plant computer and are in the final testing phase. Upon completion of the final testing phase, scheduled for December 31, 1981, the Farley Nuclear Plant will have an operational Class A DCM. The Class A model's conceptual design (EDCM) is complete and has been reviewed by the NRC. The conceptual design was transmitted to the NRC by Alabama Power Letter dated February 9,1981.
The system has been designed on a modular basis. The major software components are:
1.
Puff Processing Module 2.
EDCM Reporting Module 3.
System Utilities The puff module simulates the emission of a gaseous puff from the plant vent containing a homogeneous isotopic concentration determined from standard tables in the computer files. This puff updated every 15 minutes, comprises both an elevated and a ground level component.
Height of the elevated component is determined by meteorological and stream release data located in computer file tables. The puffs are tracked as they traverse a ten mile circular boundary surrounding the plant. Puffs are purged from the data base after they cross the ten mile boundary or after twelve hours from origin, whichever event occurs first.
Data gathered from meteorological conditions and effluent grab samples are filed in a computer storage table.
.m
Item 12. RADIOLOGICAL EMERGENCY RESPONSE (Cont'd)
The EDCM reporting module utilizes the existing puff location file to simulate radiation dosare emissions based on current location of puffs, meteorological conditions, and the isotope table.
Meteorological conditions and the isotope table are located in-a computer storage table. The EDCM process will be initiated by the puff process module every fifteen minutes. This will produce report parameters which include current locations, arrival times, and organ dose calculations of all puffs emitted since the start of the incident. All report parameters generated during the incident are kept in the report parameter computer file and can be viewed on demand. After an incident is terminated, a utility is also available to archive the parameters to tape. A password secured routine is also avaliable to terminate the incident. This is the only manner in which an incident is terminated and is executed from the system console.
The system has several utilities designed to Insure data integrity, availability, and to edit the master file and tables. These ut'lities also can be utilized to generate initial files during system startup.
Alabama Power Company will provide data to the offsite groups via voice communications and/or telecopy rather than a computer data link between onsite and offsite groups. This position was described in Alabama Power Comp 69y's letter to the NRC 4ted May 19, 1981.
1 Item 13. REACTOR COOLANT PUMP,'0IL COLLECTION SYSTEM UPGRADE PER 10 CFR 50, APPENDIX R [ Enclosure 6, Item 73 Requirement
Reference:
Letter from Mr. F. L. Clayton, Jr. to Messrs' S. A. Yarga and A. T. Schwencer dated March 19,'1981.' to the reference letter shows " anticipated outage schedule for FNP fcr Appendix R Planning." For FNP Unit 1, the third refueling outage is shown as 2-22-82 to 4-1-82.
During that time interval, it was anticipated, as shown in attachment 2 to the reference letter, " Plan and Schedule for Fire Prote: tion Program for Operating Power Plants Part 50.48 and 10CFR50 Appendix R" that the installation would be completed for the seismic reactor coolant pump oil collection system. Such implementation in Unit 2 is to be ccepleted by the 1st Refueling.
Response
At the present time, part of the engineering is released and at the j
site, it is anticipated that the balance of the engineering will be released shortly.
Based upon the engineering released to date and information from engineering planning, it is anticipated that the installation of changes necessary to seismically qualify the existing PCP oil collection system will not be completed by the.end of the t
third refueling outage. Alabama Dower Company is making every effort to improve upon the current schedule for design and installation.
Every effort will be made to seismically qualify the oil collection system for Unit i during the current refueling outage but in no case will such qualification be completed. The implementation of the Unit 2 system is still scheduled to be completed during the first refueling outage.
4
Item 14.
II.B.2, PLAi4T SHIELDING [ Enclosures 1 and 3, item II.B.2;, para 2.C.(21)(c)]
Requirement In accordance with the Unit 2 Operating License, Alabama Pcwer Company must, prior to April 1,1982, complete all modifications to assure access to vital areas and protection of safety equipment following an accident resulting in a degraded core.
Response
UNIT 1 Currently work is progressing to install disconnect devices (scheduled for delivery in " vary,1982) during the current outage.
If procurement problems occur, the installation of such devices will be completed when the unit is in cold shutdown for sufficient duration but no later than the return to power following the fourth refueling.
The modification of the penetration door shielding is coplete.
UNIT 2 All but two modifications to meet the above requirement have been compl eted. The remaining modifications entail installation of a shield door on the electrical penetration room door and installation of electrical disconnects to allow operation of eight motor operated valves (during post-acciaantal conditions). Current installation and procurement schedules show completion of the door modification prior to April 1,1982. However, an outage is required to install the electrical disconnect devices that are to be located outside +he electrical penttration room. When Alabama Power Company originally discussed this with the NRC during Unit 2 licensing, the intent was to have a two (2) week generator outage during February to meet the turbine generator manufacturer's requirements. This requirement will be obviated by a planned shorter seven day inspection on the turbine generatne after return to power of Unit 1.
This shorter outage will be for the purpose of installing Turbine Generator monitoring equipment ind to perform a limited generator inspection to provide confidence of operation until the first refueling. The seven day period will be insufficient time to accomplish this change.
Specifically this design change would require extension of the outage for approximately two weeks and would necessitate bringing the unit to cold shutdown which is not currently planned. No outage of sufficient duration for installation of this change is planned until the first refueling outage. Alabama Power commits to having these devices installed prior to returning to power operation following the first refueling outage or other outage of sufficient duration.
Item 15.
II.K.3.5, AUTOMATIC TRIP 0F REACTOR COOLANT PUMPS DURING LOSS OF-COOLANT ACCIDENT [ Enclosures 1 and 3, item II.K.3.5; Enslosure 5, para 2.C.(21)(1)(2)(1)
Requirement 1.
By letter to the NRC dated June 30, 1981, Alabama Power makes the following commitment on this subject related to the Farley Nuclear Plant:
Within three (3) months of formal NRC notification that automatic RCP trip design modifications are required, it is Alabama Power Company s intention supply design infor7atica for Units 1 and 2.
2.
The Unit 2 License item 2.C.(21)(1)(2)(1) states that with respect to tripping of RCPs, Alabama Power Company must submit to the NRC for approval either (1) an evaluation which shows that sufficient time is
.ailable to the operator to manually trip the RCPs in the event or a small break LOCA, or (2) a description of design modifications required to provide for an automatic pump trip.
This suomittal is required within three months after NRC determination of acceptability of the sma?1 break LOCA model based on comparisons with LOFT test L3-6.
Response
Westinghouse (in support of the Westinghouse Owners Group) has performed an analysis of delayed reactor coolant pump trip during small-break LOCAs. This analysis is documented in WCAP 9584 and 9585, August 1979.
In addition, Westinghouse (again in supp1rt of the Westinghouse Owners Group) has performed test predtetisns of LOFT Experimentt L3-1 and L3-6.
The results of these pradietions are documented in letters OG-49(3-3-81), 0G-50(3-23-81) and 0G-60 (6-15-81).
Based on:
- 1) the Westinghouse analysis, 2) the excellent prediction of the LOFT Experiment L3-6 results using the Westinghouse analytical model, and 3) Westinghouse simulator data related to operator response time, the Westinghouse and Alabama Power Company position is that automatic reactor coolant pump trip is not necessary since sufficient time is available for manual tripping of the pumps.
Our understanding of the schedule for final resolution of this issue is:
A) Once the NRC formally approves the Westinghouse model, a 3-month stucy period will ensue during which the Westinghouse Owners Group will attempt to demonstrate compliance with some NRC acceptance criteria for manual RCP trip. The NRC acceptance criteria will accompany their formal approval of the Westinghouse models.
n;
~
Item 15.- ~ II.K.3.5, AUTOMATIC TRIP OF REACTOR COOLANT PUl@S DURING t.0SS OF-
, /[c COOLANT ACCIDENT (Cont' d) a B) If, 'at.the end of the 3-month period, the Westinghouse Owners s
Group cannot show compliance with the acceptance cirteria, the NRC y
will formally notify utilities that they must submit an automatic RCP trip design.
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