05000346/LER-1980-028-01, /01X-1:on 800417,discrepancy Discovered Between Existing Plant Operating Practices & Safety Analysis.Assumed Availability of Pressurizer Spray Following Steam Generator Tube Rupture to Reduce RCS Pressure Questioned: Difference between revisions

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#REDIRECT [[05000346/LER-1980-028-01, /01T-0:on 800417,during Review of Station Procedures,Discrepancy Identified Between Plant Operating Practices & Safety Analysis.Caused by Safety Analysis Assumptions Being More Restrictive than Actual Conditions]]
| number = ML19345G883
| issue date = 04/10/1981
| title = /01X-1:on 800417,discrepancy Discovered Between Existing Plant Operating Practices & Safety Analysis.Assumed Availability of Pressurizer Spray Following Steam Generator Tube Rupture to Reduce RCS Pressure Questioned
| author name = Lingenfelter J
| author affiliation = TOLEDO EDISON CO.
| addressee name =
| addressee affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
| docket = 05000346
| license number =
| contact person =
| document report number = IEB-79-05C, IEB-79-5C, LER-80-028-01X, LER-80-28-1X, NUDOCS 8104220628
| package number = ML19345G882
| document type = LICENSEE EVENT REPORT (SEE ALSO AO RO), TEXT-SAFETY REPORT
| page count = 6
}}
{{LER
| Title = /01X-1:on 800417,discrepancy Discovered Between Existing Plant Operating Practices & Safety Analysis.Assumed Availability of Pressurizer Spray Following Steam Generator Tube Rupture to Reduce RCS Pressure Questioned
| Plant =
| Reporting criterion =
| Power level =
| Mode =
| Docket = 05000346
| LER year = 1980
| LER number = 28
| LER revision = 1
| Event date =
| Report date =
| ENS =
| abstract =
}}
 
=text=
{{#Wiki_filter:U. S. NMLLG BLMiLWt/LoOJV (sOMMoSS CN NRC FOE 87-7 7)
LICENSEE EVENT REPORT OON T ROL BLOCK: l l
l l
l l
l (PLEASE PRINT OR TYPE ALL REQUIRED INFt RMATION) 1 6
lo l t ] l 0 l 11l D l B l S l 1 l@l 0 l 0 l - l 0 l 0 ] 0 l 0 l 0 l - l 0 l 0 ]@l 4 l 1 l 1 l l j 1 ]@l l
lg 7
3 O LICENSEE CODE 14 15 LICENSE NUMBEH 25 26 LICENSE TYPE JO 57 CA T 68 l0l1l 3gORT lL l0 l5 0 l 0 l0 l3 l4 l6 Ql 0 l 4 l 1 l 7 l 8 l 0 l@l 0 l 4 l ll 0l 8 l 1l@
CO%'T R
7 0
60 61 OOCKET NUveER 6d b3 EVENT DATE 74 75 REPORT DATE 80 g
EVENT DESCRIPTION AND PROB ABLE CONSEQUENCES h 1012 l l (NP-32-80-04) During a review of station procedures, a discrepancy between existing l
plant operating practices and the safety analysis was identified. Of primary concern l o 3 l is the assumption that pressurizer spray would be available following a steam genera-l g o j,, ) l This was determined tor tube rupture to reduce reactor coolant system pressure.
lo Jsj 1 There l
immediately reportable as potential operation outside the safety analysis.
101s l l was no danger to the public or station personnel.
The assumptions used in the safety l lol7l l analysis are substantially more restrictive than actual conditions.
l j 3,3 y ;
SYSTEM
 
==CAUSE==
CAUSE COMP.
VALVE CODE CCOE SUBCODE COMPONENT CODE SUSCODE SUBCODE lC lB lh h
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O 9 7
8 9
10 ft 12 33 18 19 20 SEQUENTIAL OCCURRENCE REPORT REVISION LE R< RO EVENT YE AR R EPOR T NO.
CODE TYPE No.
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_ 21 22 2J 24 26 27 28 29 30 31 32 SUPPLIE MA ER l,,,X JglJ4H lg l3bZ jg gg l1 0l0l0l l
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33._
36 3
40 41 42 43 44 47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS required procedural changes ;
l l o l l NRC IE Bulletin 79-05C, in response to the TMI II event, j, ;i j l whereby reactor coolant pumps are tripped on receipt of Incident Level 2 Safety g
Features Actuation System 1650 psig trip. The steam generator tube rupture accident,
,, g has been reanalyzed. Procedural guidelines are being developed as part of the
{
i ATOG program by B&W a.d utilities owning B&W plants.
l 80 7
8 9 ST S
% POWE R oTHER STATUS IS O RY DISCOVERY DESCRIPTION NA l
lAlgl Procedure Review l
]@ l 0 l 0 l 0l@l i s ACTIVITY CONTENT RELEASED OF RELE ASE AMOUNT OF ACTIVITY LOCATION OF RELEASE NA l
l NA l
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PERSONNEL EXPOS ES OEstn PTiO~ @
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10;el01@lZI@l i 7
,ERSONsa',Na'iES oESCniPiiON@
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1010 l01@l NA i s 80 7
8 9 11 12 LOSS OF OR DAM AGE TO F ACILITV TYPE
 
==DESCRIPTION==
l W @l NA i 9 80 7
8 9 to l*;SUt @ DESCRIPTION l
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l NA 2 0
" 5 rg'3 0 4 o o 0(c0 6 7
a 2 DVR 80-066 N AME OrIRIPAREFt Jacoue Lineenfelter PHONE:
 
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TOLGO EDISON COMPANY l
DAVIS-BESSE NUCLEAR POWER STATION UNIT ONE SUPPLEMEKfAL INFORMATION FOR LER NP-32-80-04 DATE OF EVENT: April 17, 1980-FACILITY:
Davis-Besse Unit 1 t
IDENTIFICATION OF OCCURRENCE: Determination of Existing Conditions Outside Safety Analysis-Conditiont Prior to Occurrence:.The unit was in Mode 5, with Power (MWT) = 0, and-Load (Gross MWE) = 0.
Description of Occurrence: During a review of station procedures, a discrepancy between existing plant conditions and the Davis-Bosse Unit 1 safety analysis was j
identified.
Chapter 15 of the FSAR discusses the steam generator tube rupture acci-
]
dent and makes certain basic assumptions.
Of. primary concern is the assumption that pressurizer spray will be available following the tube rupture to reduce Reactor-i Coolant System (RCS) pressure below the main steam safety valve 'setpoint pressure.
~
The accident analysis also identified the escape paths by which primary activity.
I could be released to the atmosphere.
i This analysis shows that during the steam generator. tube rupture accident, the RCS pressure will drop and the Safety Features Actuation System (SEAS) Incident Level _2 j
1650 psig trip will be actuated. As a result of emergency procedural changes which l
require the tripping of all Reactor Coolant Pumps (RCPs) on receipt of an SEAS 1650 l
psig Incident Level 2 alarm, the driving force for pressurizer spray will be lost un-til the RCPs are restarted.
Standard operator action time assumption for. accident' analysis at Davis-Besse is ten minutes.
When all four RCPs are tripped, the Steam and Feedwater Rupture Control System (SFRCS)-
will initiate auxiliary feedwater to both steam generators with each auxiliary feed j
The exhaust pump turbine being supplied steam by its associated steam generator.
i for the auxiliary feed pump turbine being supplied by the leaking steam generator constitutes a new path for primary system activity to the atmosphere. The loss of~
pressurizer spray.and the new leakage path are not in. agreement with the existing accident analysis.
4 These issues -appear to be generic in nature and have.been reviewed by the NRC in NUREG 0651.
[
f Designation of Apparent Cause of Occurrence: NRC IE Bulletin 79-05C, in response to-the Three Mile Island Unit II event, required the procedural changes by which RCPs l
are tripped on receipt of Incident Level 2 SEAS 1650 psig trip.
LER #80-028 1
s..
.-~#.,._
.u. -.
 
TOLEDO EDISON COMPANY DAVIS-BESSE NUCLEAR POWER STATION UNIT ONE SUPPLEMENTAL INFORMATION FOR LER NP-32-80-04 PAGE 2 Analysis of Occurrence: There was no danger to the health and safety of the public or to station personnel. The steam generator tube rupture accident analysis has been re-evaluated specifically addressing the loss of pressurizer spray associated with the trip of all reactor coolant pumps (RCPs) and the additional releases associated with the exhaust of steam from the affected OTSG in the auxiliary feed-pump turbine (AFPI). This analysis considers'the effect of extended cooldown and resultant delay in isolation of the affected OTSG associated with the loss of pres-surizer spray. Consistent with the FSAR analysis, the analysis assumes that off-site power is available throughout the transient. However, several assumptions have been altered to eliminate extra conservatism from the simplified FSAR analysis.
The FSAR analysis assumes that all of the reactor coolant leaking into the secondary side of the OTSG (88,000 lbs) is released directly to atmosphere. The revised analy-sis identifies that most of the flow goes to the condenser (where a large partitioning coefficient results in negligible off-site releases) and only a fraction of the reactor coolant released to the affected OTSG is subsequently released to the atmos-phere (16,000 lbs.) via the Main Steam Safety Valves (MSSV). This is consistent with the Abnormal Transients Operating Guidelines (ATOG) analyses performed by B&W for ANO-1 and is considered equally applicable to Davis-Besse Unit 1 for the purposes of this analysis. In addition, pending similar analysis for Davis-Besse, ANO-1 ATOG parameters have been used to demonstrate Davis-Besse Unit 1 compliance with 10CFR100 limits. See Tables 1 and 2 for these parameters. Assuming 1% failed fuel (5 micro-curies / gram I-131 dose equivalent) and using the Davis-Besse meteorology a site boundary dose of 4.9 rem has been estimated excluding the release from the AFFT ex-haust. The dose-reported in Davis-Besse Unit 1 FSAR is 27.1 rem for the steam gen-erator tube rupture accident.
This estimated dose of 4.9 rem nas been readjusted to take into account the noncon-servatisms in the analysis based on the particular features of Davis-Besse design as outlined below:
1.
Because of trip of the four RCPs on low RCS pressure, AFPs are automatically started. This provides an additional offsite release path through the AFFT exhausts.
2.
The primary to secondary leak rate assumed in the Davis-Besse FSAR is 435 gpm, 20 gpm higher than that assumed in ANO-1 ATOG analysis (see Table 1).
In addition, the relief setpoint of the lowest set Main Steam Safety Valve at Davis-Besse is 1065 psia as compared to 1150 psia assu=ed in the ANO-1 ATOG analysis. This results in an extended ti=e required to isolate the affected OTSG since the system should be depressurized to 1065 psia (before isolation of affected OTSG) as compared to 1150 psia assumed in the ANO-1 ATOG analysis.
Based on the time interval (from AFP start to manual isolation of AFPTs from affected OTSG) in Table 1, the steam released from AFPT exhaust has been conservatively cal-culated to be 6,620 lbs and has been assumed totally to be primary coolant. Adjust-ing for this additional dose, the resultant dose is calculated to be 7 rem.
Considering the higher primary to secondary leak rate and additional time required to isolate the affected OTSG (approximately 4b minutes), the total amount of reactor coolant released to the atmosphere (including the auxiliary feed pump turbine exhaust LER #80-028
 
' TOLEDO EDISON COMPANY DAVIS-BESSE NUCLEAR POWER STATION UNIT ONE PAGE 3 SUPPLEMENTAL INFORMATION FOR LER NP-32-80-04 is 25,810 pounds. This results in a total " adjusted" dose of 8 rem. Since the dose reported in Davis-Besse Unit 1 FSAR is 27.1 rem for this accident, the new dose is less than 1/3 of the original FSAR value and, only a small fraction of 10CFR100 limits.
The analysis presented here includes a conservative assumption that no mixing and 1
transport occurs between the primary and secondary fluids.
It has since been cal-culated that only 31% of volume vented through MSSV would be the reactor coolant, however, this additional margin was not incorporated into the revised analysis.
Based on the above, it is concluded that the steam generator tube rupture accident does not result in dose consequences which exceed NRC guidelines, and further the Davis-Besse Unit 1 FSAR analysis envelopes the doses calculated using assumptions that are consistent with those reported in the NRC Safety Evaluation Report.
Therefore, this does not constitute an unreviewed safety question.
 
==Corrective Action==
Procedural guidelines which cope with this accident are being developed as part of the ATOG program being conducted by Babcock and Wilcox Company and the utilities owning Babcock and Wilcox plants.
Failure Data: There have been no previous similar reportable occurrences.
LER #80-028
 
TOLEDO EDISON COMPANY DAVIS-BESSE EUCLEAR POWER STATION UNIT ONE PAGE 4 SUPPLEMENTAL INFORMATION FOR LER NP-32-80-04 TABLE 1 STEAM UENERATOR TUBE RUPTURE COOLDOWN SCENARIO Time (Min. /Sec.)
Event Initiation of Rupture (415 gpm, Makeup 55-160 gpm, O
Letdown 55 gpm)
Pressure - Temperature Reactor Trip 11:00 (MSSVs Lif t)
Letdown Secured, 2nd Makeup Pump Started 11:14 (320 gpm) 11:17 Low RC Pressure SFAS Trip 11:30 HPI Flow Initiated 11:36 MSSVs Rescat 1
11:40 Operator Trips RCPs, AFPs Started 11:45 Operator Initiates Manual Cooldown 23:00 50 F Subcan'ing Achieved, RCPs Restarted from Manual Isolation of Steam Supply to AFPTr 23:00 Affected OTSG (T
= 540 F, P = 1470 PSIA)
HOT Pressurizer Spray Initiated (RCS Depressurization),
25:00 Periodic Use 40:00 Af fected OTSC Steamed (Level > 95% on Operate Range)
Affected OTSG Isolated (T
= 500 F, HOT 43:00 P = 1150 PSIA)
% 180:00 End of Transient (DRRS Conditions)
LER #80-028 l
t
--~.
 
TOLELO EDISON COMPANY DAVI5-BESSE EUCLEAR POWER STATION UNIT ONE
' SUPPLEMENTAL INFORMATION FOR LER NP-32-80-04 PAGE 5 TABLE 2 Value Parameter 24,750 lb.
RCS Leakage, Prior to Trip RCS Leakage, Post Trip (to condenser) 60,500 lb.
Until Affected OTSG Isolation 171,400 lb.
Until DHRS Conditions 16,060 lb.
RCS Leakage During Venting to Atmosphere 43 min.
Time from Rupture to Affected OTSG Isolation 32 min.
Time from Reactor Trip to Af fected OTSG Isolation 36 sec.
Venting Time to Atmosphere Time for Operator Initiated Cooldown to Affected 31 min.
OTSG Isolation Rate of RCS Cooling from Trip to Affected OTSG 180 F/hr.
Isolation Rate of RCS Cooling from Operator Cooldown 180 F/hr.
to Af fected OTSG Isolation LER #80-028
}}
 
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