ML18051B539
| ML18051B539 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 09/11/1985 |
| From: | Wambach T Office of Nuclear Reactor Regulation |
| To: | Zwolinski J Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8509120138 | |
| Download: ML18051B539 (15) | |
Text
September 11, 1985 Docket No. 50-255 MEMORANDUM FOR~.. John A. Zwolinski, Chief
- operating Reactors Branch #5, DL FROM: :
SUBJECT:
Thomas V. Wambach, Palisades Project Manager Operating Reactors Branch #5, DL MEETING
SUMMARY
- CONSUMERS POWER COMPANY PROBABILISTIC RISK ASSESSMENT (PRA) FOR MAIN STEAM LINE BREAK
,*'01') August.21, 1985 a meeting was he id between the NRC staff and representatives of Consu.mers Power Company, the licensee for Palisades Plant.
The licensee provided an overview of the PRA performed to support their position that modifications to preclude a.single failure of a main steam isolation valve resulting in the blowdown of. both steam generators were not necessary.
The PRA also justifies, in the licensee's view, not adding redundant isolation capability. for the main feedwater* lines. The licensee's presentation is given in.Attachment 1. The staff provided preliminary feed back and indicated that there would be additional information required... The staff stated that a high priority was established for completing the review of these issues and both the staff and the licensee must apply the necessary resources to expedite the review.
The attendees are listed on Attachment 2.
~~inal lUBned by:
Th6mas V. Wambach, Palisades Project Manager Operating Reactors Branch #5, DL cc: See next page
Enclosures:
As stated DISTRIBUTION Docket NRC PDR Local PDR ORB#5 Rdg JZwolinski
. ACRS (10)
, v?tl' DL%RB#5 TWambach:jg
'f //0/85 TWambach HThompson CJamerson DCrutchfield OELD GLainas EJordan TSpeis BGrimes RBernero NRC Participants DL:ORB#5~
JZwolinski
., ;,, /85 eP' s.
ov\\
<.f:l-
cc:
M. I. Miller, Esquire Isham, Lincoln & Beale Slst Floor Three First National Plaza Chica~0, Illinois 60602 Mr. Thomas A. McNish, Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Judd L. Bacon, Esquire Consumers Power Conpany 212 West Michigan Avenue Jackson, Michigan 49201 Regional Administrator, Region III U.S. Nuclear Regulator.v Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137 Jerry Sarno Township Supervisor Covert Township 36197 M-140 Highway Covert, Michigan 49043 Office of the Governor Room 1 - Capitol Building Lansing, Michigan 48913 Palisades Plant ATTN:
Mr. Joseph F. Firlit General Manager 27780 Blue Star Memorial Hwy.
Covert, Michigan 49043 Resident Inspector c/o U.S. NRC Palisades Plant 27782 Blue Star Memorial Hwy.
Covert, Michigan 49043 Nuclear Facilities. and Environmental Monitoring Section Office Division of Radiological Health P.O. Box 30035 Lansing, Michigan 48909 David J. VandeWalle Director, Nuclear Licensing Consumers Power Company 1945 W. Parnall Road Jackson, Michigan 49201
- 1.
NRC Concerns PALISADES PLANT PRA Evaluation of the Risks Associated with Main Steamline Break Events 8-21-85
- 2.
Scope and Assumptions of PRA Evaluation
- 3.
Conclusions
- 4.
Recommendations & Cost Benefit Evaluations RP0885-0001A-NL03 ATTACHMENT 1
Main Steamlipe Break Inside Containment With Coincident MSIV Failure Original NRC Concerns:
Main steamline break inside containment with coincident single failure of the MSIV on the intact steam generator results in the blowdown of both steam generators into the containment.
Reference -
SEP Topic XV-2 (NUREG 0820 10/82)
NRC Identified Safety Issues (Ref NUREG 0820)
- 1.
Following blowdown of both steam generators, the steam driven portion of the auxiliary f eedwater system would fail leaving only the motor driven auxiliary feedwater pump.
- 2.
There exists only a limited time frame following steam generator dryout in which to.establish heat removal due to possible steam generator failure caused by the addition of cold feedwater.
- 3.
A potential for operator interference with the feedwater system exists in attempting to control the cooldown rate.
- 4.
Difficulties may be encountered in controlling heat removal given the existence of open steam system piping.
- 5.
The ability to feed High Pressure Safety Injection (HPSI) water to the primary coolant system under these conditions has not been demonstrated.
Proposed Resolution Backup MSIVs; $2M (CPCo experience with similar backfits -
AFW upgrade, SIRWT support repair - indicates this may be low by factor of 2 or 3).
/
RP0885-0001A-NL03
Main Steamline Break Inside Containment With Coincident Continuous Feedwater Addition To The Failed Generator Original NRC Concerns:
Main steamline break with coincident single failure of the feedwater regulating valve or feedwater bypass valve in the open position results in the blowdown of the inventory of the failed steam generator plus the additional steam generated by continuous feedwater addition into the containment.
Reference -
IE Bulletin 80-04 (2-8-80)
NRC Identified Safety Issues (Ref IE Bulletin 80-04)
If a main steamline break inside containment coincident with the addition of feedwater continuously through the.feedwater regulating or feedwater bypass lines occurs, then the following conditions can result:
Containment Overpressure Continuous feedwater steaming to the containment in addition to the initial inventory of the steam generator can result in pressurization of the containment beyond the design basis.
Return to power Continuous f eedwater addition to a failed steam generator coincident with (a) the most reactive control rod stuck in the fully withdrawn position, (b) end of life reactivity conditions, and~(c) the most restrictive active single
. failure in the safety injection system can cause a
- reactivity increase during the transient beyond previously analyzed design bases.
Proposed Resolutions Automatic closure of the feedwater stop.valves on steam generator low pressure; $30K.
Additional valve in series with each feedwater bypass valve; $200K.
RP0885-0001A-NL03
. e PRA Evaluation of Main Steamline Break Issues Scope The purpose of the PRA was to investigate and identify the underlying safety issues associated with main steamline break events, and propose resolutions to these issues keeping in mind the integrated plant design.
Events Considered All initiating events which could lead to a blowdown of one or both steam generators inside or outside containment including:
passive piping failures transients leading to stuck open atmosphere dump valves or secondary safety valves Systems Modeled Twenty front line and support systems including:
for example, AFW, HPSI, MSIVs, SWS, AC and DC power Data Sources Plant Specific Ten years of control room log books, maintenance orders, surveillance tests Generic WASH-1400, NREP, IEEE-500, NRC and EG&G LER Summaries Transient/Consequence Analysis In plant response -
MAAP Dose Consequences -
CRAC-2 RP0885-0001A-NL03
Results Determine core melt frequency and public does risk associated with steamline breaks given existing plant design.
Propose backfits to address identified safety issues and determine net change in core damage frequency and public dose associated with each alternative.
Produce cost benefit analyses for proposed backfits.
RP0885-0001A-NL03
PRA Evaluation of Main Steamline Break Issues Important Insights And Assumptions
- 1.
Steam generator blowdown to the containment can occur as a result of the passive failure of main steam, main feed or auxiliary feedwater piping attached to the secondary side of the steam generator.
Steam generator blowdown can also occur outside the containment.
Potential initiators include main steam lines, main feed lines, auxiliary feedwater lines, steam supply lines to the auxfeed pump turbine and steamlines to the atmospheric dump valves up to the first normally closed isolation valve outside containment.
Initiators also include stuck open ADV's or secondary SRV's which are actuated as a part of a routine plant trip from power.
Only failure of the main steam piping inside containment is considered to be important as an initiator during continuous f eedwater events'.
Failure of feedwater or auxiliary feedwater piping coincident with continuous feedwater will result in water (rather than steam) addition to the containment.
- 2.
EEQ.
Two steam generator blowdown to the containment through a large break (main steam or main feed line rupture) will result in exceeding the existing EEQ pressure and temperature profiles.
Continuous feedwater to the failure steam generator through either the stop valve or bypass valve will also result in exceeding EEQ limits.
In that regard it is assumed that automatic instrumentation and equipment which operates early in the incident are not degraded by the environment prior to fulfilling their
-safety function.
On the other hand, equipment required to function in the long-term is assumed to be unavailable.
Small steamline breaks (auxiliary feedwater piping) inside containment are not assumed to lead to EEQ related failures as the rate for energy addition to the containment is substantially less than the DBA.
- 3.
As the primary system reheats during sequences when there is no-secondary cooling in service, the ADV's will reopen when Tav returns to the ADV setpoiny.
If the break is inside the containment, this results in an opening of the containment to the outside atmosphere.
RP0885-0001A-NL03
CONCLUSION$
Two Steam Generator Blowdown Issue The original NRC concerns have been addressed by analysis or backfit:
The original NRC evaluation of this issue assumed only a single motor driven auxiliary f eedwater pump would remain should both steam generators depressurize.
The MSIV backfit was to have provided additional redundancy by keeping tqe steam driven pump in service.
Since that evaluation, an additional motor driven pump has been installed, providing the additional redundancy even if both steam generators are depressurized.
If auxiliary feedwater were to be disabled following a large
'steamline break event, approximately 15 minutes would be available to the operator to re-enable the system prior to primary safety valve actuation, and core uncovery would not be threatened for more than an hour.
Re-establishment of feedwater during this time would prevent core damage and would not threaten steam generator integrity.
The steam generators have been designed to withstand eight cycles of adding 70°F water to a dry steam generator at 600°F.
The operator is not considered any more likely to disable auxiliary feedwater during a two-steam generator blowdown than during a single steam generator blowdown.
This is because indications available to the operator for control of primary system cooldown (i.e., PCS temperature) are very similar for both single and double steam generator blowdowns.
It is acknowledged that feed and bleed cooling with high pressure safety injection has not been demonstrated for Palisades Plant.
No credit has been takeri for feed and bleed in this evaluation.
CPCo is evaluating possible system configurations which would allow establishment of the mode of core cooling.
Potential backfits were evaluated:
The main steam isolation valve backfit is two orders of magnitude below the cost beneficial threshold.
This is because its major benefit is the enabling of the third steam driven auxfeed pump (the two motor driven pumps must fail before the steam driven pump is required).
RP0885-0001A-NL03
The backfit has no known beneficial effect on transient sequences other than main steamline break events.
There is a larger risk associated with steam generator blowdown outside containment than inside containment - the concern identified in the
- Systematic Evaluation Program.
(The core damage probability associated blowdown inside containment is less than 1 x 10 -6/yr.)
This is simply because the initiator is more frequent by a factor of 1000.
A condensate pump alignment procedure is at least as effective in reducing risk as the backup MSIV.
This is because the condensate pumps address most of the same sequences addressed by the third auxiliary feedwater pump.
The proposed EEQ backfit is not cost beneficial.
This is because it addresses only steamline break events inside containment and has no beneficial affect on blowdowns outside containment or other accidents such as LOCA's.
Plant design and various programs (including EEQ and containment leak testing) have been developed and implemented without consideration of the two-steam generator blowdown event.
The cost of retrofitting the plant and these programs at this time would far outweigh the.safety benefits.
RP0885-0001A-NL03
Conclusions Continuous Feedwater Issue The plant as-built is designed to preclude the potential for continuous feedwater addition in that f eedwater reg~lating valves and f eedwater bypass valves receive a signal to isolate on low steam generator pressure.
Feedwater Isolation Backfits The backup feedwater bypass valves are several orders of magnitude above the cost benefit threshold.
This is because normal operation takes place with the existing valves closed which effectively precludes feedwater addition through these lines.
Installation of circuitry to close the main feedwater stop valves on steam generator low pressure is not cost-effective and may be contrary to plant safety.
This is because the safety issue being addressed is of very narrow scope (i.e., the effect on overall plant safety has not been stu~ied). The modification will in fact reduce the overall reliability of the f eedwater system and the resultant negative impact on a wide range of events may be greater than positive impact on steamline breaks.
RP0885-0001A-NL03
RECOMMENDATIONS Two-Steam Generator Blowdown Issue Develop procedures which permit the alignment of the condensate system as a backup to AFW.
Insure that the operator understands not to disable auxiliary feedwater in an attempt to control the cooldown rate.
Insure that the operator understands that the ADV's may come open automatically during events in which inadequate core cooling is approached.
For accident initiators such as steamline breaks inside containment it will be necessary for him to take manual control of this system to prevent a loss of containment integrity.
(These procedural modifications are planned as a part of the emergency procedure upgrade required by TMI Action Plan Item I.C.l.)
Continued Feedwater Addition Issue Continue operating the f eedwater.system as it is currently configured for normal power operation.
That is, maintain the risks of continuous feedwater through the bypass line at a minimum by keeping the feedwater bypass valves closed, limiting operation with them in the open position to periods of startup and shutdown.
RP0885-0001A-NL03
Cost Benefit Evaluations Cost CB
=
(Change in dose/yr) * (Yr)
Where Backfit Backup MSIV's Condensate pump EEQ backfits CB = Cost benefit Cost = Cost of backfit ($)
Change in dose/yr = Net annual reduction in dose received by the public as a result of the backfit Yr = Years remaining in license CBR
$140,000/Man:...rem l
V"'
procedure
$63/Man-rem "Z_<)O/<..
$37,000/Man-rem
' ('A FW bypass valves
$9,000/Man-rem
'"2.oO )<.
FW stop valves
$90,000/Man-rem
~()\\(
RP0885-0001A-NL03
Figur~I Pali sades SEP NO
~
C'J45l0 G~
Fr:'" *
11.1pply t
s J..
N c_..:.- SV:6C8 t
N:-~Sllc-sc:2 I
~sv~*
I Palisades main system and feed~ater system 19
I
- NAME T. V. Wambach H. L. Thompson D. M. Crutchfield J. A. Zwolinski S. Israel C. Grimes E. McKenna J. Shapaker D. Blanchard D. Vandewalle AFFILIATION NRR/DL/ORB#5 NRR/DL NP.R/DL/ADSA NRR/DL/ORB#5 NRR/DST/RRAB NRR/DL/SEPB NRR/DL/SEPB NRR/DSI CPCo CPCo
. ATTACHMENT 2