IR 05000387/1989003
| ML17156B092 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 03/24/1989 |
| From: | Durr J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17156B091 | List: |
| References | |
| 50-387-89-03, 50-387-89-3, 50-388-89-03, 50-388-89-3, NUDOCS 8904120146 | |
| Download: ML17156B092 (34) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-387/89-03 50-388/89-03 Docket No.
50-387/50-388 License No.
NPF-14 NPF-22 Licensee:
Penns lvania Power
& Li ht Com an 2 North Ninth Street Allentown Penns 1vania 18101 Category Facility Name:
Sus uehanna Units 1 and
Meeting At:
NRC Re ion I Kin of Prussia Penns lvania Meeting Date:
March
1989 Approved by:
J. Du, Engineering Br anch, DRS
<p1 date
~ei
'
i
"
"
i 12, 8,
5D.
report by the licensee regarding their evaluation of the Reactor Building post accident temperature profile and the environmental qualification of electrical equipment in the Reactor Building.
Preliminary calculation of Reactor Building temperatures performed in support of Appendix R analyses indicated temperatures in excess of temperature used to establish equipment qualification for electrical equipment in the Reactor Building.
Justifications for Interim Operation were developed to establish equipment operability pending final resolution of this issue.
8904120146 890824 PDR ADOCK 0 <000387 Q
.,
DETAILS 1.
Meetin Attende'es Penns lvania Power and Li ht Com an
=A. Derkacs, Engineer Eg Group J. Blakeslec, Assistant Plant Superintendent G. Miller, Engineer Mechanical Group, NPE J.
Agnew, Sr. Project Engineer, NPE F. Butler, Manager of Nuclear Design E.
Heckman, Licensing Group Supervisor R. Harris, Sr. Licensing Specialist P.
Brody, Sr. Project Engineer, NPE - Electical B. Palmer, Project Engineer, Nuclear Design - IKC D. Roth, Sr.
Compliance Engineer M. Mjaatvedt, Project Engineer - Nuclear Design Mechanical J. Retling, Sr. Project Engineer - Nuclear Systems Boston Edison G. Whitney, Licensing Engineer F. Mogolesko, Project Engineer Nuclear Re ulator Commission J. Durr, Chief, Engineering Branch, RI C. Anderson, Chief, Plant Systems Section, RI R. Architzel, Section Chief, DEST, NRR H. Walker, Sr.
Engineer, DEST, NRR F.
Young, Sr.
Resident Inspector 2.
Licensee Presentation The licensee presented their Reactor Building post accident temperature studies and their evaluation of the associated electrical equipment qualification that led to a 50.9 report regarding this issue on December 12, 1988.
The licensee presentation is outlined in their handout which is provided as Attachment A to this document.
During June 1987, the licensee performed Reactor Building temperature calculations in support of Appendix R analyses.
The Reactor Build>ng (RB)
HVAC is not a required system.
The revised calculated vemperaiures, without HVAC, were significantly in excess of previous RB temperature calculations used to establish electrical equipment qualification in a post accident harsh environment.
However, these preliminary calculations were based on a calculational model with numerous conservatisms.
As an interim measure, the lice'nsee developed procedures to restore RB HVAC cooling in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> and to remove non essential heat loads in the RB. It
was determined that the original design calculational model was non conser-vative and the calculational basis did not include the most limiting conditions for establishing the maximum long term temperature profile in the RB.
Additional analyses were performed by the licensee in 1988 with a computer program that was developed to simulate the post accident conditions in the RB.
This model yielded a significant reduction in the RB temperatures.
Never the less, temperatures in excess of the equipment qualification temperatures were calculated for selected areas of the RB.
The increased temperatures impacted 124 items of electrical equipment.
However, only 80 of these items reflected a significant temperature increase.
The maximum increase in temperature was 20'F above the original qualified temperature.
A preliminary review by the licensee of the impact of the increased RB temperature determined that all of the impacted equipment is operable.
Procedures will still require that either the RB HVAC be restored or that non essential heat loads be removed to maintain RB temperatures at an acceptable level.
The licensee plans to upgrade the Eg files for all but 13 of the affected items by June 1989 to document qualification at the increased RB temperatures.
For the remaining 13 items, the licensee maintains that qualification to NUREG-0588 Category 2 requirements could be established based on existing data'owever, the licensee plans to move 11 of the these items by December 31, 1989 as part of an Appendix R
modification.
The two remaining items will be moved to a mild environment by December 31, 1990.
Conclusion The licensee stated that the overall safety impact of the increased RB temperature was relatively minor.
The NRC noted that a review would be made by the NRC of the supporting qualification data for the items that are to be moved to a mild environment by December 31, 199 AGENDA NRC/PP&L MANAGEMENTMEETING REACTOR BUILDINGENVIRONMENT PO T-A IDENT INTRODUCTORY REMARKS REACTOR BUILDINGENVIRONMENT POST-ACCIDENT EQUIPMENT QUALIFICATIONIMPACTS MANAGEMENTOVERSIGHT CLOSING REMARKS F. G. BUTLER
~ J. E. AGNEW A. P. DERKACS J. A. BLAKESLEE F. G. BUTLER
REACTOR BUIUII% HVAC POST LOCA TPPERATURE ANALYSIS o
NM PAL DISG%3E)
EVENTS THAT LED I DISGNHF INTERIN IRRECTIVE ACTINI ASSURE CNTINKD SAFE OPERATION
%M REACIR BUILDING TEtPHNTURE ANALYSIS DEVELOPED SGUJTIN TO REDUCE X%ERATlIE
OVERVIEW SEQUENCE OF EVENTS APPENDIX R ANALYSIS'ARCH 1987 ECAP MODEL OF REACTOR BUILDING TEMPERATURE, JUNE 1987 REVIEW OF ORIGINAL SSES REACTOR BUILDING TEMPERATURE ANALYSISi JULY 1987 SOOR GENERATED TO DOCUMENT WEAKNESSES.
JULY 1987 DEVELOPMENT OF NEW REACTOR BUILDING TEMPERATURE MODEL.
OCTOBER 1988 COTTAP
APPENDIX R ACTIVITIES o
M-1002 PREPARED TO IDENTIFY SYSTENS REQUIRED FOR SAFE SHUTDOWN Of PLANT IN THE EVENT OF A FIRE o
SAFE SHUTDOWN LIST (N-1002) DID NOT IDENTIFY REACTOR BUILDING HVAC AS A REQUIRED SYSTEN o
NECHANICAL DESIGN ANALYZED THE RESULTING TENPERATURES IN REACTOR BUILDING WITH NO HVAC o
ANALYSIS PERFORNED UTILIZING ECAP
APPENDIX R ACTIVITIES o
ECAP ENVIRONMENTAL CONDITIONS ANALYSIS PROGRAM I
o STEADY-STATE, MULTI-ZONE ENERGY BALANCE o
PREDICTS BUILDING TEMPERATURES CONSIDERING:
AIR FLOWS HEAT GAINS/LOSSES FROM:
OUTSIDE WALLS ADJACENT ROOMS MECHANICAL EQUIPMENT ELECTRICAL EQUIPMENT o
ECAP MODES MODELED APPENDIX R NORMAL OPERATION LOCA
ECAP RESULTS POST-LOCA o
REACTOR BUILDING BULK AVERAGE AIR TENPERATURE 140 F
o REACTOR BUILDING TENPERATURE RANGE 107 F TO 170 F o
FSAR TABLE 3.11-6 INDICATED A TENPERATURE RANGE OF
.104 vo 130 F
o, ECAP RESULTS INDICATED FURTHER RESEARCH WAS NECESSARY
ORIGINAL SSES REACTOR BUILDING TEMPERATURE ANALYSIS o
ORIGINAL ANALYSIS FOR REACTOR BUILDING POST LOCA INDICATED A PEAK TEMPERATURE OF 104 o
CONSIDERED CONCURRENT LOCA/LOOP/COMPLETE FAILURE OF ALL DIVISION II COMPONENTS o
ONE-ZONE MODEL o
A SEPARATE ANALYSIS PREDICTED A PEAK AIR TEMPERATURE OF 150 F IN THE REACTOR BUILDING ECCS ROOMS
A/E REVIEW OF ORIGINAL REACTOR BUILDING TEMPERATURE ANAIYSIS
CALCULATION BASIS.
LOCA/LOOP WITH THE MOST LIMITING SINGLE FAILURES ALTHOUGH CONSISTENT FROM A LICENSING STANDPOINT. MAY NOT BE THE MOST LIMITING IN ESTABLISHING A BULK LONG TERM TEMPERATURE PROFILE
WEAKNESSES IN ORIGINAL REACTOR BUILDING TEMPERATURE ANALYSIS o
LOOP NOT WORST CASE o
MOST LIMITING SINGLE FAILURE NOT CONSIDERED o
SINGLE ZONE MODEL NOT ADEQUATE o
ELECTRICAL LOADS UNREALISTIC
INTERIM CONCLUSION o
REACTOR BUILDING AIR TEMPERATURE POST LOCA IS INDETERMINATE o
EQUIPMENT QUALIFICATION NAY BE AFFECTED o
JUDGEMENT INDICATES THAT ACTUAL TENPERATURE MILL BE LESS THAN THAT PREDICTED BY ECAP, BUT GREATER THAN THAT PREDICTED BY THE ORIGINAL TEMPERATURE ANALYSIS
INTERIM SOLUTION o
REMOVE HEAT TO REDUCE TEMPERATURE o
DEVELOPED A CONSERVATIVE ESTIMATE RATE OF TEMPERATURE RISE IN REACTOR BUILDING TO BOUND THE EVENT o
RESTORE REACTOR BUILDING HVAC COOLING IN 10 HOURS
NEW REACTOR BUILDING TEMPERATURE ANALYSIS o
DEVELOPED NEW MODEL IN HOUSE TO FULLY UNDERSTAND PHENOMENON o
TRANSIENT TEMPERATURE ANALYSIS o
MULTI-ZONE CAN ANALYZE UP TO 200 ROOMS o
PREDICTS TEMPERATURES CONSIDERING:
AIR FLOWS HEAT GAINS/LOSSES FROM:
OUTSIDE NLLS
,ADJACENT ROOMS MECHANICAL EQUIPMENT ELECTRICAL EQUIPMENT o
RUNS ON PPRL MAINFRAME COMPUTER o
INDEPENDENTLY VERIF IED
. o COTTAP COMPARTMENT TRANSIENT TEMPERATURE ANALYSIS PROGRAM
NEW REACTOR BUILDING TEMPERATURE ANALYSIS o
ELECTRICAL DESIGN DETERMINED REALISTIC ELECTRICAL LOADS o
THREE DESIGN BASE ACCIDENTS CONSIDERED.
CASE
UNIT 1 LOCA. UNIT 2 NORMAL OPERATION NO LOOP
'AILUREOF ESW LOOP B CASE 2 UNIT 1 LOCAL UNIT 2 NORMAL OPERATION LOOP FAILURE OF ESW LOOP B CASE 3 UNIT 1 LOCA, UNIT 2 FALSE LOCA NO LOOP
COTTAP RESULTS o
POST LOCA REACTOR BUILDING TENPERATURE AVERAGE 131 F
o HIGH TPIPERATURE IS A CONCERN o
TO REDUCE TENPERATURE REDUCE HEAT LOAD
COTTAP RESULTS o
.
SHED NON-ESSENTIAL LOADS IN REACTOR BUILDING POST LOCA o
RESULTING TBlPERATURES AVERAGE 117 F
o HIGH TENPERATURES OCCUR ONLY IN SELECTED AREAS
POST LOCA LOAD SHED o
SHED NON-ESSENTIAL LOADS IN AFFECTED REACTOR BUILDING IF REQUIRED o
PARTIAL LOOP DRIVES THE PLANT TO THE LICENSED, ANALYZED, DESIGN BASIS ACCIDENT o
LOAD SHED CREATED BY EXERCISING 2 BREAKERS IN THE TURBINE BUILDING o
LOAD SHED INITIATED 2Q HOURS POST LOCA o
PROCEDURE PREPARED TO DICTATE LOAD SHED JEA/DEL J EA /EXC002C (20)
SUMMARY o DISCOVERED ANONLY IN ORIGINAL DESIGN ANALYSIS o INTERIN CORRECTIVE ACTION TAKEN o
NEM DESIGN ANALYSIS PREPARED o TEMPERATURE NNAGPtENT RESTART REACTOR BUILDING HVAC OR
.SHED NON-ESSENTIAL LOADS JEA/TRc003c(22)
ECAP ORIGINAL ANALYSIS COTTAP.
% SIME FAILURE SIBSY STAK NJLTI Z(NE ELEClRICAL LOAD HI6H 107'
170'F 1%'F AVG, FAIUIRE OF DIVISION II TRANSIENT SI%LE Z(WE ELECfRICAL LOAD HI6H 1%'F PEAK 6ENERAL 1%'F PEAK EKS LIMITINGFAILISE TRANSIENT l%LTI Z(WE REALISTIC ELECTRICAL LOAD 117'F AV6.
JEA/mc
.
Jv/zxc003c(20)
POTENTIAL DEFICIENCY IDENTIFIED DETERMINE IF REAL I
UNDERSTAND NATURE AND SCOPE DETERMINE OPERABIZITY (6 Reportability)
ESTABZISH WRITTEN JCO ( repeat last 4 steps as required)
ESTABZISH PZAN TO CORRECT DEFICIENCY EXECUTE PZAN F
CZOSE DEFICIENCY TRACKING PAPER (NCR)
EQ'MPACT LIMITED o CHANGES POST-LOCA TEMPERATURES IN THE REACTOR BUILDING o ORIGINAL TEMPERATURES 104 to 1.30 degrees F
o NEW TEMPERATURES ARE 92 to 140 degrees F
(In areas with EQ'd Equipment)
IMPACT ON EQ'd.EQUIPMENT IN REACTOR BUILDING.
o MANY TYPES ALREADY QUALIFIED FOR 144 to 410 degrees F
o IMPACT 9 EQ BINDERS ABOUT 20 TYPES OF EQUIPMENT 124 ITEMS OF 2129 IN EQ MASTER LIST
EQ ZONES
ROOMS IN PLANT
DEGREE F INCREASE IN TEMPERATURE TO QUALIFY TO IN WORST CASE
COTTAP EQ B 1 dg ROOM ZONE Room Room Description NUMBER OF I TEMS ORIG NEW TEMP QUAL'D MAX DIFF TEMP TEMP
15
17
88
14
13
16
25
71
92
R 1 k I-401 R lm I-100 R lm I-200 R lm I-201 R1IK I I-401 R 1m I I-200 R
1 m I I-201 R
1 p I-109 R
1 m I I-102 R lm I-102 R lm I-105 R lm I-203 R
1 m I-500 R lm I-513 R
1 m I I-500 R lm I I-513 R lm I I-203 R 1 p I I-109 R lm I I-105 Cont Access Area.....
RR Access Shatt......
Equipment Removal Are RHR PP/Hx Hatch Area.
Cont Access Area.....
RHR PP/Hx Hatch Area.
Equipment Removal Are Remote Shutdown Room.
Equip Access Area....
Equip Access Area....
Equip Access Area....
Hx/Pump Room.........
Circ Space...........
SLC Areao
~
~
~
~
~
~
~
~
~
~
~
~
Circ Space...........
SLC Area.............
Hx/Pump Room.........
Remote Shutdown Room.
Equip Access Area....
~
~
~
. 12
..2
..4
..1
. 12
..6
..
..5
..1
..1
~
..5
~
~ 2
..2
..6
..2
..4
..9
..1
11
11
12
14
15
16
16
16
19
104 104 104 104 104 104 104 104 104 104 104 104 104 104 104 104 104 104 104 113 115 115 115 115 116 116 118 118 119 119 120 120 120 120 120 121 123 124 Sub-Total....80
.R lm I-513 SLC Area...................
71 R lm I I-513 SLC Area..................
71 R1m I I-500 Ci rc Space...............32 TOTAL...124
1 0.05 119 119 120 120 120 120
EQ COTTAP 81dg ZONE ROOM Room Room Oescription T2 4,0 PERCE,hJT rE.VP.
RiSE.
So
$ 0
R lm Rim Rlp R lm R lm R lm R lm R1k R lm R lm R lm R lm R 1k R lm Rim R lm R lm R lm Rip
13
15
17
19
25
71
87
88
91
I-105 I-102 I-109 I- '100 I-203 I-200 I-201 I-401 I-500 I-513 II-500 II-513 I I-401 II-203 II-200 II-201 II-102 I I-105 I I-109 Equip Access Area Equip Access Area Remote Shutdown Room RR Access Shaft.
Hx/Pump Room RHR PP/Hx Hatch Area Equipment Removal Area Cont Access Area Circ Space SLC Area Circ Space SLC Area Cont Access Area Hx/Pump Room RHR PP/Hx Hatch Ages Equipment Removal Area Equip Access Area Equip Access Area Remote Shutdown Room 100 100 100 i04 100 100 100 100 100
- 100 100 100 100 100 100 100
'I 00 100 104 1!9 119 118 115 120
'I 15 115 113 1'10 110 120 120 115 121 116,
16'I
124 123
/0
~ I o
30
50
70
90 l00 Tj: NE DfEVS
/
IMPACTED EQUIPMENT AND BASIS FOR OPERABILITY I
I BINDER I
EQDF-22B I
EQDF-29 I
EQDF-33 EQDF-34 EQDF-04 EQDF-09 EQEL-36 EQEL-38 EQPL Jol OPER BASIS ANALYSIS ANALYSIS TEST ANALYSIS ANALYSIS ANALYSIS ANALYSIS ANALYSIS TEST EQUIPMENT TYPE Comsip Control Panels Asco Temp Sws/FCI FSs Unit
ESQG Refrig Bailey MV Converter ITE 480 VAC MCCs
"GE 230VDC MCCs L LCs S8K/Amatek Flow Eleme S8K/Amatek Flow Xmtr Dwyer dp Press Switch WORST ROOM I I 109 I I-513 II-500 I I-109 I I-109 I I-109 I I-104 II-401 I-100 WORST TEMP CHANGE
1
20
20
11
ORIG I
QUAL'D TEMP 104 119 104 104 104 104 104 104 104 NEW I
WORST I
TEMP 124 120 120 124 124 124 115 115 115
BASIS FOR OPERABIZITY AND JCO ANAKYSIS Comparison of vendor's temperature rating of device to new postulated temperatures Compa'rison of materials temperature ratings to new postulated temperatures Re-evaluation of qualified life considering accelerated aging for new postulated temperatures for 100 days TEST Comparison of temperature to which equipment was tested to postulated new temperature Re-evaluation of qualified life considering accelerated aging for new postulated temperatures for 100 days Engineering calculations document conclusion PLANS FOR CORRECTION OF DEFICIENCIES o EQEL-38'&K/AMATEKFLO'6 TRANSMITTERS
ITEMS AZ READY COMPT ETE o EQDF-22B EQDF-29 EQDF-04 EQDF-09 EQDF-36 EQPL-J01 COMSIP CONTROL PANELS
ITEMS ASCO TEMP SWITCHES 6 FCI FZOW SWITCHES l2 ITEMS ITE 480 VAC MCCs
ITEMS SGK/AMATEK FLOW EZEMENTS 8 ITEMS DWYER dp PRESS SWITCHES 2 ITEMS Review analysis in calculations against final version=of COTTAP results (M-RAF-024), revise as needed, and place in EQ Binder.
Plan to do by 6/30/89.
(77 items)
o EQDF-34 BAIZEY MIZZ IVOZT CONVERTER CARDS 3 ITEMS 1 Being moved to mild environment by Appendix R Mod by end of U14RIO (6/89)
RFP to move the other two to mild environment issued not scheduled.
o EQDF-33 Unit
ESGW REFRIG UNITS 36 ITEMS Electronic Controls to be moved to mild environment by Appendix R Mod.
Plan to do by 12/31/89.
(10 of 36)
Revise EQ analysis (based on test reports) for Compressor Motor and ITT NH90 Actuator.
Plan to do by 6/30/89.
(4 of 36)
.
Remaining items already qualified to 120 degrees F by ex'isiting binder.
-(22 of 36)
SUMMARY o
EQ IMPACT LIMITED o
DOCUMENTED ANALYSIS SHOWING OPERABIZ ITY IN PLACE o
EQ BINDER REVISIONS PZANNED FOR 6/30/89 FOR"-
AZZ EXCEPT 13 ITEMS o REMAINING 13 ITEMS WILE BE MOVED TO A MIZD ENVIRONMENT ll by 12/31/89 2 by 12/31/90
PROBLEM RESOLUTION PROCESS EVENTS REVIEWS/
ASSESSMENTS ENGINEERING EVALUATION/
ANALYSIS POTENTIAL PROBLEM IDENTIFICATION DOCUMENTATION VIA SOOR/NCR REPORTABILITY OPERABILITY EXTERNAL NOTIFICATIfNi t
MANAGEMENT EVALUATION PORC.
SRC ACTIONS INTERIM LONG TERM COMMUNICATION TRAINING PROGRAMS
REACTOR BUILDING HEATUP CHRONOLOGY OF KEY EVENTS SHORT TEfN SOOR INITIATED PORC REVIEWED INITIALJIO INITIALOPERABILITY/REPORTABILITYREVIEW SRC REVIEWED INITIALJ10 NRC RESIDENT INSPECTORS BRIEFED OPERATIONS HOT BOX TRAINING'NITIATED PORC REVIEWED STATUS OF ACTIONS-PORC REVIEWED STATUS OF ACTIONS PORC REVIEWED PROCEDURE CHANGES ADDITIONAL OPERATIONS HOT.BOX TRAINING PROCEDURE CHANGES ISSUED JULY 17, 1987 JULY 28, 1987 JULY 28, 1987 JULY 29. 1987
~l
~
~
JULY 29, 1987 JULY 30, 1987 AUGUST 17'987 AUGUST 31, 1987 SEPTENBER 8, 1987 SEPTBSER 8, 1987 SEPTBSER 8, 1987
REACTOR BUILDING HEATUP CHRONOLOGY OF KEY EVENTS -
LONG TERN-SRC UPDATE PORC UPDATE NCR'S 88-0679 AND 88-0680 ARRIVE ON SITE PORC REVIEWED NEW JIO AND EP-IP-055 JULY 27, 1988 SEPTEtrmER 8, 198'CTOBER 27, 1988 OCTOBER 27. 1988 EP-IP-055 ISSUED SRC REVIEWED NEW JIO
~ 10CFR 50.9 REPORTABILITY DETENNINED REGION 1 TELECON HELD PORC REVIEWED CLOSURE OF ORIGINAL SOOR 10CFR50.9 FOLLOWUP LETTER ISSUED AT NRC REQUEST NRC NANAGPIENT MEETING NOVBSER 3, 1988 NOVPSER 30,.198 DECBSER 12,.198 DECPSER 13, 198
DECBSER 15, 198 JANUARY 17. 1%5 NARCH 3.
1989
SUMMARY
COMPREHENSIVE PROCESS FOR PROBLEM RESOLUTION
SAFETY REVIEW PROCESS WAS EFFECTIVE
ENGINEERING ANALYSIS WAS TRANSLATED INTO USEFUL CONSERVATIVE OPERATOR ACTIONS