ML20211K519
| ML20211K519 | |
| Person / Time | |
|---|---|
| Issue date: | 12/04/1986 |
| From: | Bernero R Office of Nuclear Reactor Regulation |
| To: | Beckjord E, Parler W, Taylor J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE), NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES), NRC OFFICE OF THE GENERAL COUNSEL (OGC) |
| Shared Package | |
| ML20209E138 | List: |
| References | |
| FOIA-87-10 NUDOCS 8612100292 | |
| Download: ML20211K519 (52) | |
Text
,
4
.$ v' '
- M
% UNITED STATES f( 4 NUCLEAR REGULATORY COMMISSION g WASHINGTON, D. C,20555 g ,
~s...../ .
DEC 4 1986 MEMORANDUM FOR: Eric S. Beckjord, Director Office of Nuclear Regulatory Research James M. Taylor, Director l Office of Inspection & Enforcement William C. Parler, Director 1 Office of General Counsel FROM: Robert M. Bernero, Director Division of BWR Licensing
SUBJECT:
GENERIC REQUIREMENTS FOR BWR CONTAINMENT RESPONSE TO SEVERE ACCIDENTS Enclosed for your comments is a Draft Generic Letter for Proposed BWR Containment Requirements which will enhance containment performance in severe accidents. The requirements specified in the Draft Generic Letter have been derived by regulatory analysis comparing the behavior of BWR containments in severe accident environments, and the benefits derived from proposed containment enhancements. The regulatory analysis is also enclosed for your comments.
Your comments and concurrence are requested by December 12, 1986. The material is scheduled to be presented to ACRS on December 9 and 12, 1986.
A CRGR review is planned for late December, 1986. The Generic Letter is scheduled to be issued for industry and public comments by the end of January 1987 after review by the CRGR and with the Commission and as final by May, 1987. Please provide your comments in time to support the schedule. -
m .
l Robert M. Bernero, Director i
Division of BWR Licensing i CONTACT:
M. C. Thadani i
X28649
Enclosures:
- 1. Draft Generic Letter on fo.T/J -g'7-o / o Proposed BWR Severe Accident 2.
Containment Requirements Regulatory Analysis Oho cc: See Next Page ,
f {[
.- ,, ,m, e --
,-y , -- _ .-.,v.- .
,r.-_
- . o-960 d cc: w/ enclosures H. Denton R. Vollmer W. Russell T. Speis F. Miraglia T. Novak H. Thompson D. Ross, RES J. Partlow, IE E. Jordan, IE J. Scinto, OGC J. Funches
/W Houston G. Lainas G. Holahan DBL PD's DBL BC's d' *4 *r *d'i a *8" * / %. e * ,4 y h . y'+' 3 ia -( #,',T_' 8 e"' a h - - , - . , _
UNITED STATES
- y 't, NUCLEAR REGULATORY COMMISSION
$, .$ WASHINGTON, D. C. 20666 5.,....../ 09.Jb v ,EI TO ALL BOILING WATER REACTOR (BWR LICENSEES AND APPLICANTS FOR BOILING WATER REACTOR LICENSES Gentlemen:
SUBJECT:
PROPOSED BWR SEVERE ACCIDENT CONTAINMENT REQUIREMENTS (GENERIC LETTER 87- )
Severe accidents dominate the risk to the public associated with nuclear power plant accidents. A fundamental objective of the Commission's Severe Accident Policy is to take all reasonable steps to reduce the chances of occurrence of a severe accident and to mitigate the consequences of such an accident should one occur. The Reactor Safety Study report issued in 1975 found that for BWRs the probabilities of accidents resulting in core melt were low, but the containment performance following a severe accident was poor and tended to offset the benefits of low BWR core melt probabilities. Subsequent actions resulting from the TMI Action Plan have led to several plant modifications and required improvements in plant procedures to further reduce the likelihood of severe accidents. In December 1980, an industry initiative on severe accidents resulted in the formation of the Industry Degraded Core Rulemaking (IDCOR) group to address the concerns related to core damaging accidents. The IDCOR effort has led to industry methodology for Individual Plant Evaluations (IPEs) to search for the risk outliers and to address system reliability and -
containment performance on a plant specifi4M basis. The staff has concluded, "
however, that for BWR containments, a set of generic requirements has been identified that moots the need to await plant specific analyses of containment performance and will 1 ad to speedier implementation than would be possible via the IPEs. Severe accident analyses have indicated several areas for improvement in BWR containments which should be promptly pursued as follows:
- 1. Hydrogen Control [
0 Present requirements imposed by 10 CFR Part 50.44 and the Technical 5 Specifications shall be adhered to, no additional requirements are proposed.
t
~
- 2. Con'tainment Spray All BWRs with Mark I containment shall provide at least two backup water l
supply systems for the containment drywell spray, one of which shall be functional during station blackout. Water to the spray system from these
- backup supplies shall be available by remote manual operation or by simple procedures for connection and startup which can be implemented during a severe accident scenario.
, In addition, the spray nozzles shall be adjusted so that an evenly
! distributed spray pattern will be developed in the drywell whether
o .
. - - -- - 2) water is supplied by the primary source or either of the~ backup '
sources. A flow rate on the order of 1/10 of the present flow rate is considered typical, the licensee shall select the flow based on an analysis of plant specific parameters. ;
- 3. Pressure Relief The licensee shall select a pressure between design pressure and 1 times design pressure at which to open an exhaust path from the wetwell vapor space to the highest vent point (stack or pipe) available. This line should be capable of handling water vapor flow equivalent to 1% decay heat at the vent pressure selected without significant chance of rupture before the desired release point. The line shall be equipped with isolation valves which can be opened and reclosed by remote manual operation or by simple procedures which can be implemented during severe accident scenarios including station blackout
- 4. Core Debris Control The licensee shall ensure that the water in the suppression pool in the event of torus failure is held within the confines of the torus room and the corner rooms and cannot flow out to other parts of the plant.
- 5. Procedures and Training The licensee shall implement emergency operating procedures and other procedures based on all significant elements appropriate to its plant of Emergency Procedure Guidelines, Revision 4.
Since these requirements are intended to be an optimized use of existing equipment it is expected that added equipment, of itself, need not meet the quality or design standards of safety related equipment. Nevertheless, modifications to or near equipment or systems which are already safety related shall not compromise the quality of such equipment or systems.
The equipment changes required herein shall be installed during the first -
-frefueling outage which begins nine (9) months after the effective date of this 1 letter. The procedures and training required shall be implemented on a i 1( schedule reviewed and approved by the NRC. Given the implementation of the generic improvements of Mark I containments there is no need for an Individual Plant Evaluation (IPE) for containment performance. This does not remove the need for an IPE which covers the system reliability or core melt frequency portion of the severe accident question.
l l
6 . .
1 3-This proposed generic letter has identified areas for improvement in BWR containments which the staff believes to be effective in reducing risk and which can be implemented at a reasonable cost. We welcome comments on the proposed actions and other suggestions on the subject matter. The goal is to significantly reduce the; likelihood of containment failure given a core melt.
Sincerely, Robert M. Bernero, Director Division of BWR Licensing Office of Nuclear Reactor Regulation
Enclosure:
BWR Mark I Containment Performance During Severe Accidents
=.. . . . - . , ,
_m,____-_ , _ _ _ .._ _ _ , _ . _ . .
,L BWR MARK I CONTAINMENT PERFORMANCE DURING SEVERE ACCIDENTS
1.0 BACKGROUND
A fundamental objective of the Commission's Severe Accident Policy of August 6,1985 is to take all reasonable steps to reduce the chances of occurrence of a severe accident and to assure substantial capability to mitigate the consequences of such an accident should one occur. The Commission also called for a balancing of accident prevention and mitigation, and special consideration of containmen't performance in searches for risk outliers.
Enhancements to the performance of containments in severe accidents should increase assurance of mitigation of severe accident consequences. The Reactor Safety Study report issued in 1975 found that for BWRs the probabilities of accidents resulting in core melt were low, but the containment following a severe accident could be severely challenged and tended to offset the benefits of low BWR core melt probabilities. Subsequent actions resulting from the TMI i
Action Plan have led to several plant modifications and required improvements in plant procedures to further reduce the likelihood of severe accidents.
Other p6st TMI actions-have also involved containment enhancements;
- particularly in the areas of isolation dependability and hydrogen control.
In concert with the Commission's policy to further reduce the chances of occurrence of severe accidents and to mitigate their consequences, an industry initiative is underway to develop a methodology for Individual Plant l
l Evaluation (IPE) directed to search for risk outliers. The resulting approach will be applied on a plant-specific basis. The initial IPE trials l
l
L ;. ; SiMFij 2-have been made by industry, and have encompassed internal accident initiators and systems reliabilities to the point of estimating core melt probabilities.
Source terms, containment performance and offsite risks have not been considered, but have been discussed as future extensions of initial IPEs.
With respect to BWRs with Mark I, II and III type containments, the staff has reviewed these initial IPEs, historical probabilistic risk assessments and the plans for completing the search for individual plant outliers. The review has indicated that sufficient bases exist to complete the search for outliers for all such plants with respect to accident mitigation by backfitting in five areas as discussed in the subsequent sections. That is, by requiring
- improvements in five areas, no further evaluations of accident mitigation for BWR reactors with Mark I, II and III type containments are considered necessary.
The staff has identified five potential containment enhancements which lend
~
~ ~
themselv'es to ge'neric implementation' and have the potential to~ s'ignificantly mitigate the consequences of several severe accident sequences including station blackout and ATWS sequences. In the Policy Statement the Commission stated that the rulemaking route would generally be a preferred route to implement future severe accident related actions. However, rulemaking is extremely time consuming. The Commission's statement regarding operating reactors recognized the time element and the continued severe accident risk to public health and safety, and provided other options to dispose of the issues l
l l
5 through the conventional practice of issuing Bulletins and Orders or Generic Letters. In the interest of time the staff has chosen the Generic Letter as a preferred approach to achieve closure of the BWR containment performance issues. Containments are required to protect the public from the consequences of accidents. The design and sizing of containment are required to assure that the containments are essentially leak-tight barriers against the uncontrolled release of radioactivity to the environments and to assure that containment design conditions important to safety are not exceeded for as long as i
postulated accident conditions require. The containments should accommodate with sufficient margins, the pressures and temperatures resulting from any loss-of-coolant-accident (LOCA).
Although a postulated design basis LOCA is not expected to produce more than a 4
few percent fuel failures, an accident radiological " source term" used in in calculating offsite dose consequences is representative of a substantial core melt accident (10 CFR 100). Even for this source term, containments are i _ designed such that calculated offsite doses are unlikely to result in an early or major latent health hazard if the containments were to maintain their low i leakage capability *. What is at issue is the capability of containments to perform a mitigating safety function as long as practicable during very low probability severe accidents, where the stress on containment may significantly exceed that of a design basis LOCA and the consequences of containment failure may be very significant. The structural integrity of BWR containments is seriously challenged for accidents with high energy release i
"rt 200 specifies "the expected demonstrable leak rate from the '
i containment", a value which is made part of each licensee's Technical Specifications.
i .
to the containment because in spite of the positive pressure suppression feature, they are either relatively small (Mark I & II); or have a very low design pressure (Mark III), and the likelihood of their failure in a severe accident is perceived to be higher than necessary.
Overall plant core melt probabilities for BWRs with Mark I, II and III contain-ments have been estimated to range from one in a thousand per reactor year to two in ten million per reactor year for BWR designs evaluated by the NRC and the industry. Many of these estimates have not fully included assessments of the benefits of post-TMI backfits, operator responses, or the increases in core melt probabilities arising from factors not considered in plant specific analyses such as earth-quakes, floods and fires. Contemporary analyses break down such probabilities into classes and subclasses of accidents. The sum of the core melt probabilities for all classes and subclasses of accidents is considered to be the overall core melt probability. For BWRs with Mark I containments, IDCOR1 ,2 has proposed the following five classes of events for core melt accidents:
o c. . .
e ,. n. , <...,.
, ., ,y .,-
1 Individual Plant Evaluation - Peach Bottom Atomic Station, May 1986.
10COR Technical Report 85-3-81, BWR Accident Sequence - Individual Plant Evaluation Methodology, April 1986.
l l
y loss of core cooling with containment at low pressure and failure after core melt; loss of core cooling with containment failure before core melt; loss of core cooling with containment failure soon after core melt due to high containment pressure at the time of core melt; loss of core cooling with containment failure before core melt due to failure to depressurize; and containment bypass.
Our review of the IDCOR core melt probability estimates to date generally indicates that they are low. The BWR core melt frequencies of past evaluations are summarized in Table I. Given a core melt, the estimates of likelihood of Mark I, II and III containment failures have been high relative to other containment types. Ii1 all of these past evaluations, little or no credit has i
been given to features which can be used with relatively modest upgrading to prevent or mitigate accidents.
M L: w w r ~ .t .we.o.# -
The Reactor Safety Study (WASH-1400, NUREG-75/14) indicates a conditional containment failure probability for the BWR Mark I containment reference plant (Peach Bottom) of about 90% (inferred from Table 5-3, page 81). That is, given a core melt in a BWR with a Mark I containment (Peach Bottom) there is l about 90% chance of containment failure. In the November 1984, IDCOR Technical Summary Report, Nuclear Power Plant Response to Severe Accidents, the estimate for Peach Bottom was about 20% (inferred from Table 10-1, page 10-6). More recently, the Vermont Yankee Containment Study provided an estimate that
9 TABLE 1 - U.S. BWR PLANT-SPECIFIC PRA STUDIES V .
REACTOR CORE-DAMAGE EVENTS MEDIAN. CONTAllOIENT PROGRAM REPORT CORE /
CONSIDERED MEAN OR CON 0lTIONAL MAME REPORT YEAR CONTAlleENT POWER (WT) FREQUENCY PRA PLANT POINT FAILURE ESTIMATE ESTIMATE PR08 ABILITY 3x10
-5 Internal / Median Not evaluated RSS WASH-1400 1975 BWR-4/MK I 3293 Peach External Botton ,
4x10
-5 Internal Maan 0.2 Peach IDCOR Tech Summary 1984 BWR-4/MK I 3293
~
Botton Task 21 -5 Not evaluated IPE IPE 1986 BWR-4/MK I - 3293 2x10 Internal Mean l Peach
- Bottom
~4 .nternal Median Hot evaluated 1983 BWR-3/MK I 1727 3x10 Millstone IREP NUREC/CR 3085 ~4 1727 5x10 Internal Mean Mot evaluated 1'
M111 stone NUSCO Millstone 1 1986 BWR-3/MK I PSS 2x10
-4 Internal Point Not evaluated Brown. Ferry IREP NUREG/CR 1982 BWR-4/MK I 3293 2801 Estimate 3x10
-5 Internal Mean 0.07 Vermont VYCSS VYCSS 1986 BWR-4/MK I 1593 Yankee 1x10
-3 Internal / Mean 0.25 Big Rock Consumers Big Rock 1981 BWR-1/ Dry 158 Point Point PRA External 3 0.25
, Big Rock EG&G/BNL EG&G-EA- 1982 BWR-1/ Dry 158 1x10 Internal / Mean Point 5533 Rev. 1 External
~
-5 Internal / Mean 1.0 Limerick PEPC0 Limerick PRA 1981 BWR-4/MK II 3293 7x10 External 1x10
-4 Internal / Mean 1.0 Limerick BNL NUREC/CR- 1983 BWR-4/MK II 3293 3028 External Shoreham LILCO Shoreham PRA* 1983 BWR-4/MK II 2436 5x10-5 Internal Point not evaluated
-4 Estinata i
Shoreham BNL NUREG/CR- 1985 BWR-4/MK II 2436 1x10 Internal Point Not evaluated 4050 -5 Estimate Shoreham IPE Shoreham IPE 1986 BWR-4/MK II 2436 8x10 Internal Mean Not evaluated
~7 Susquehanna IPE IPE 1986 BWR-4/MK II 3293 2x10 Internal Mean Not evaluated 3833 4x10
-5 Internal Median Not evaluated Grand Gulf RSSMAP NUREG/CR- 1981 BWR-6/MK III
, 1659 -6 l Grand Gulf IDCOR Tech Summary 1984 BWR-6/MK III 3833 8x10 Internal Mean Not evaluated Task 21 4x10
-6 Internal / Not evaluated f GESSAR GE GESSAR II PRA BWR-6/MK III 3579 Mean External l
l
m
],
m . ,
Vermont Yankee, a somewhat smaller BWR with Mark I containment, has a condi-tional containment failure probability of about 7%. In all these Mark I containment failure estimates the challenges come from a spectrum of accidents including ATWS, station blackout, and ordinary transients. The principal causes of failure are overpressure and direct attack of the drywell.
The accidents of interest span a spectrum of sequences and will have a probability distribution unique to each plant. Nevertheless, because of the uncertainties in calculating the dominant accident sequences, it is prudent to consider each principal type as the cause of large scale core melt and containment challenge.
For a BWR Mark II containment (Limerick), Brookhaven National Laboratory (BNL) estimated almost a 100% likelihood of containment failure give a core melt (BNL 33835; April 1984). The staff evaluation of the GESSAR II standard plant
- desig'n'(NUREG-0979, Supplement 4,: Tables 15.1, 15.'2, and 15'.12) also, indicates a conditional containment failure probability of close to one for a Mark III type containment. Only IDCOR and GESSAR II evaluations considered containment venting.
l i
a .
Q,),
3M IU For most accidents considered, the core is postulated to melt, interact with steam, water, and the structural features in the vessel and coolant system, melt through the vessel, and attack the concrete and structural features of the lower containment. Depending on the sequence of events, the' containment has the potential to fail either before or after vessel melt through. For the remainder of the accidents postulated, the containment would be bypassed, allowing radioactivity a direct path to portions of a plant not designed to contain the releases, but with some capability to attenuate radioactivity.
BWR containments respond to heatup of the fuel in the vessel directly or indirectly. The direct transfer of energy is through pipe breaks, through blowdown into the suppression pool or by the aerosols generated when the core melts through the vessel. Indirectly, radiant heat is transferred through the vessel and piping. Th'e blowdown or depressurization process, and the use of the relatively large quantity of suppression pool water as a heat sink and fission product scrubbing device, act in combination with the structural
. capability of the containment (including penetrations) to mitigate the high temperatures, pressures and radioactivity releases in a core melt. Core melt scenarios have been identified, however, which can produce conditions that could lead to containment failures, and release of fission products to the environment without the benefit of the suppression pool scrubbing. There is strong evidence that BWR containments are capable of withstanding substantially higher stresses than those for which they have been explicitly designed and this potential containment strength can be drawn upon to demonstrate additional
WJT E
i 1
protection to the public at low to modest cost. The long'er a containment can be expected to hold, the greater the likelihood that failure can be avoided.
If failure were to occur, however, reductions in the radioactivity released would be achieved. Actions that can be taken to prevent a catastrophic failure of containment before the fission products are adequately attenuated include such items as operator actions to vent the wetwell space above the suppression pool, and providing reliable spray capability.
In a core melt accident with temperatures in excess of 5000 degrees F, fission 1
products are released from the fuel in three general groups. The noble gases and the more volatile species of fission products are released from the fuel relatively early in a core melt accident. Later, the less volatile species are released as the fuel melts down into the vessel and combine with the in-vessel structural m'aterials. Finally, after melting through the vessel, refractory materials may be released during interactions of core debris with concrete on the floor of the containment.
i . .% m ,. - , -
,,,w s -- ,.1., . . ,
y4_, , -, .. .
! The amount of radioactivity that could be released to the environment in core i melt or degraded core accidents has been the subject of considerable analysis for a number of years. Present estimates (NUREG-0956) for MK I and III BWRs indicate that substantial quantities of important fission products can be j
released in a core melt accident and these analyses provide clues that suggest
, that releases can be reduced by a number of actions to enhance containment performance.
i 4
. - - . - . - . . -,-.-,. _.. . - . - ~- , - - . . . - - _ _ _ - _ - - _ . . . . . . . - . _ _. . _ _ _
- .. .. . - - -= -- - - - - - . . .
, j.
.g.
Within the core of a contemporary BWR with MK I, II, or III containment at full power there are over five billion curies of radioactivity. Severe accident releases to the environment for a MK I, II or III have been estimated to exceed 40% of such important fission products as iodine and cesium (releases of over 300 million curies of iodine and over seven million curies of cesium for a 3458 MWt reactor).
2.0 NEEDS AND STRATEGY FOR CONTAINMENT IMPROVEMENT ,
1 Consideration of the insights drawn from previous analyses suggests that no single simple feature can be added to a BWR pressure suppression containment
, to provide substantial assurance that it will successfully mitigate the consequences of a large scale core melt should one occur. Rather, one must I
conceive of some integral approach which deals with the principal concern,s.
! Consider now only the Mark I containments, 24 of which are now found in 1
licensed U.S. reactors. This analysis and development of requirements will deal first with Mark I containments because they constitute about 2/3 of the "to'. ,-
BWR population." Subsequent analyses willcdeal with Mark-II and III j containments.
1 Compared to many other U.S. reactor containments, the Mark I containment (Figure 1) has a small volume relative to the size of the reactor it contains. With a i
free volume of less than 300,000 cubic feet, the drywell wall is very close i
(
i to the reactor and to the lower head area where molten core material would l
l
(
l.
+ - , .. . . , - .,- - ,, . .- - - - - - - - - - - - - - - - - - - - - -- - ---
t .
1 PRIMAny gygygg ORYWELL I NMMfddjl g PEDESTAL CAVITY i -=
W Tit K' i
, j j 4 IGG %
',pF g ; b' '
eg .2 _
" -y7 c7 , 9 .
TYPICAL MARK I CONTAINMENT DESIGN:
FIGURE 1 l
- , - - - _ < - , - - - - - ,.__.,--,---,a- , - - . , - - - - - , - - - . - - . _ , , , . . - - , - - - -- - - - - -
O most likely fall from the reactor vessel. Even with a relatively high design pressure, typically about 50-60 psig, the small volume makes the Mark I containment more vulnerable to overpressure failure, given a comparable core melt event. Any strategy to enhance Mark I containment performance must certainly consider preventing hydrogen combustion, cooling the non-condensibles in containment, and as a last resort, venting serious overpressure through some available path, where the consequence of venting is known and is preferred to the potential of uncontrolled release due to containment loss.
i Should molten core material (corium) reach the drywell floor, the direct attack of the drywell becomes a serious concern. If the corium is sufficiently hot
! to flow with low viscosity it can easily reach the nearby wall of the drywell.
There it will attack t.he steel wall of the drywell between the vents or attack one of the 1,arge steel vent passages leading dcwn to the wetwell. The steel shell of the drywell is typically backed by a 1-2 inch construction gap filled it with a plastic spacer and then by a'very thick, reinforced concrete biological shield. Most analyses do not attempt to treat attack through the shell and shield mechanistically because of the complexity of the path, but it is apparent that this path to the reactor building and the ambient is not an open one, especially if some means are available to reduce the vigor of the attack by the hot corium.
i 1
1 1
,--,e,-.,, - - - - - -- - ,,,_-mm,--,-- , - - - . , - - . - , - , - , - - . . - = , -
)?(
The presence of hot corium on the floor of the drywell raises other challenges.
The corium can be expected to react with the concrete floor, thus generating large quantities of non-condensible gas (including hydrogen) as well as voluminous aerosols carrying non-volatile but health threatening radionuclides such as lanthanum and plutonium. In addition, radiative and convective heat transfer can directly attack the steel shell and its penetrations. Any strategy to enhance the performance of Mark I containments must seek some means to cool or quench the core debris. That strategy should also include means to cool the drywell wall to prevent overheating.
Overpressure failure of the Mark I containment may be averted even in a large scale fuel melt if debris cooling and quenching limit the amount and the temperature of the non-condensible gases in containment. Nevertheless, it is possible that pressure and temperature can build up to levels which could cause containment penetration (seal) failures or catastrophic rupture of the containment. It is desirable to have available a procedure of last resort whereby the threatening overpressure can be relieved from the wetwell. vapor space so that all gases released from containment will have passed through the water in the suppression pool, and thus will have been scrubbed of most non gaseous fission products. The pressure at which such relief should be taken into account must account for the ultimate strength of containment, the reliability of the valves used for venting, and backpressure effects on SRV operation. In addition, consideration should also be given to the material vented from the containment. At a minimum it will contain water l
) \
vapor, nitrogen, unburned hydrogen, and (depending on the stage of the accident) fission product noble gases (principally Xe 133). If the path through the downcomers and the pool water is bypassed, perhaps through a vacuum breaker line, the effluent could contain other fission products as well.
The core debris of concern includes not only the corium which melts through the' pressure vessel but the large amount of aerosols which may be released and captured by the water in the suppression pool. Using Three Mile Island experience as a guide, the suppression pool water might absorb radioactive material on the order of 0.01 to 0.1 Ci/ml (0.4-4 gigaBq/ml). That water, almost 1 million gallons of it, would be so hot (radioactive) that it would be very desirable to see to it that it stays in the torus or at least in the torus room and immediately adjacent spaces should the torus fail. If some molten corium does pass down through one or more of the eight drywell-to-torus q vent ducts, then it would most likely cause torus rupture. In that event the h'
water, if retained, would be available to quench the corium.
Finally, it is evident from all previous studies that the Mark I containment should not be treated as a simple passive body. To be effective in mitigating core melt accidents, its features must be used by trained on-site personnel who are prepared to deal with these extreme events using the equipment at hand.
Thus, the containment improvement strategy should include procedures and training for such accident management.
t
3.0 MARK I MITIGATION FEATURES After considering the technical factors identified in the preceding, a 5 element strategy stands out as an effective solution to improve Mark I containment performance to the point that there would be reasonable assurance that Mark I containments can substantially mitigate the consequences of a large-scale core melt accident. The five elements are:
- 1. Hydrogen Control
- 3. Pressure Relief
- 4. Core Debris Control
- 5. Procedures and Training 3.1 Hydrogen Control Under the present requirements of 10 CFR Part 50.44, all Mark I containments are required to have their containments inerted (with nitrogen gas) during operation. Allowance is made for a period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the beginning
_ and at~the end of the power operations cycle to* operate with air in <the containment to enable operators to inspect equipment in the containment for leaks etc. From time to time small, unidentified system leaks will start inside the containment during power operation. . Although operators sometimes question whether the 10 CFR 50.44 based technical specifications permit deinerting and entry during a cycle to investigate such leakage, they have used the 24-hour deinerted periods permitted in the technical specifications to investigate and to the extent practicable repair such leaks. Data recently I
~' '
..t-
= -
,e .
9
'i I.
Ubik.f '
TABLE 2 -
s BOILING WATER REACTORS
. WITH MARK I CONTAINMENT CONFIGURATION PLANT NAME PROD LIC ': ECCS LPCI CONT DES DRYWELL WETWELL WETWELL LINE PWR N CONFIGURATION ~ RATED PRESSURE VOLUME AIR VOL WATER (MWT) FLOW (PSIG) (CU FT) (CU FT) (CU FT)
? (GPM)
I Browns Ferry 1 BWR 4 3293{ RCIC/HPCI/LPCI/LPCS 10000 56.0 159,000 129,300 123,000 Browns Ferry 2 BWR 4 3293j RCIC/HPCI/LPCI/LPCS 10000 56.0 159,000 129,300 123,000 Browns Ferry 3 BWR 4 32931 RCIC/HPCI/LPCI/LPCS 10000 56.0 159,000 129,300 123,000 Brunswick 1 BWR 4 2436j RCIC/HPCI/LPCI/LPCS 5775 62.0 164,100 124,000 87,600 Brunswick 2 BWR 4 2436' RCIC/HPCI/LPCI/LPCS 5775 62.0 164,100 124,000 87,600 Cooper Station BWR 4 2381 d RCIC/HPCI/LPCI/LPCS 7700 56.0 132,465. 106,050 87,660 Dresden 2 BWR 3 25278 IC/HPCI/LPCI/LPCS 3625 62.0 158,236 116,645 112,203 Cresden 3 BWR 3 2527] IC/HPCI/LPCI/LPCS 3625 62.0 158,236 116,645 112,203 Duane Arnold BWR 4 16583 RCIC/HPCI/LPCI/LPCS 3600 56.0 118,000 94,270 58,900 Fermi 2 BWR 4 32924 RCIC/HPCI/LPCI/LPCS 7500 56.0 163,730 130,900 117,450 FitzPatrick BWR 4 2436[ RCIC/HPCI/LPCI/LPCS 4375 56.0 154,500 113,089 105,600 Hatch 1 BWR 4 2436' RCIC/HPCI/LPCI/LPCS 5775 56.0 146,010 112,900 87,300 Hatch 2 BWR 4 24363 RCIC/HPCI/LPCI/LPCS 5775 56.0 146,266 109,800 87,300 Hope Creek BWR 4 3293: RCIC/HPCI/LPCI/LPCS, 10000 62.0 169,000 133,500 118,000 Millstone 1 BWR 3 20115 IC/FWCI/LPCI/LPCS 7500 62.0 146,900 109,900 98,000 Monticello BWR.3 1670} RCIC/HPCI/LPCI/LPCS 3000 56.0 134,200 103,510 68,000 Nine Mile Point 1 BWR 2 1850j IC/LPCS N/A 62.0 180,000 120,000 89,000 Oyster Creek 1 BWR 2 1930f IC/LPCS N/A 62.0 180,000 127,000 83,400 j
Peach Bottom 2 BWR 4 3293,d RCIC/HPCI/LPCI/LPCS 10900 56.0 175,800 127,700 122,900 Peach Bottom 3 BWR 4 3293; RCIC/HPCI/LPCI/LPCS 10900 56.0 175,800 127,700 122,900 Pilgrim 1 BWR 3 19987 RCIC/HPCI/LPCI/LPCS 3600 56.0 '
147,000 120,000 84,000 Quad Cities 1 BWR 3 2511k RCIC/HPCI/LPCI/LPCS 3625 62.0. 158,236 116,645 112,203 Quad Cities 2 BWR 3 2511I RCIC/HPCI/LPCI/LPCS 3625 62.0 158,236 116,645 112,203 Vermont Yankee BWR 4 1593[ RCIC/HPCI/LPCI/LPCS 7200 56.0 134,000 112,200 68,000 1
~ - -
- presented in the Vermont Yankee Containment Study indicate that non-inerted
- operation at that plant amounted to 1.1% of power operations in a period of 14 years. Such a low fraction indicates a very low risk from hydrogen even now as long as inerting system power sources operate as intended The impact of the station blackout sequences for inerting system operation must be considered in any evaluation of hydrogen control for inerted containments. Taking Vermont Yankee's experience as representative, the Mark I strategy here should reaffirm the existing controls for hydrogen.
3.2 Containment Spray All Mark I containments except Oyster Creek and Nine Mile Point 1, are equipped with a dual header drywell spray system. The two spray headers are rings located well up in the cylindrical part of the drywell with branches holding spray nozzles pointing down at an angle. The headers are fed through
! each division of the RHR system with spray operation as an alternate mode of RHR operation. Due to the characteristically large size of RHR pumps (3,000 -
10,000 gpm) the drywell spray has a very high flow rate.
Precautions are usually included in operating procedures to avoid excessive use of this powerful spray system. Oyster Creek and Nine Mile Point 1 have separate dedicated spray systems. See Table 2 for a summary of key features of the
.24 plants with Mark I containments.
Most plants have other systems already connec'ted to the spray header feed lines outside of containment. They include such systems as RHR Service Water, Condensate, and in some cases bolted blind flanges which are removed to 1
r
%=3 twd FL!al install. lines for periodic containment integrated leak rate tests. Thus, it is easy for a plant to provide one or more backup supplies for the drywell spray, even in the event of a station blackout, because of the availability of fire main systems with independent pumping capability. But the available backups are all smaller systems, on the order of 10% the size of the RHR. If they were used they would probably not be able to develop sufficient header pressure for even spray flow distribution in the drywell. If there is a high assurance of drywell spray during severe accidents, even in station blackouts, a number of benefits accrue. First, the walls and penetrations of the drywell are cooled to reduce the threat of heat induced failure. Second, the drywell floor is kept flooded to provide a quenching pool for molten corium if it melts through the reactor vessel. Third, the continuing spray cools any corium which begins to travel over the open floor toward the wall of the l drywell or its vents. Fourth, the spray is available to begin washout of aerosol particles even before they pass to the suppression pool; this is j another filtering and condensing mechanism which will reinforce defense-in-depth l . - if some flow were to bypass the suppression pool. -
j Thus, the Mark I strategy is to replace all spray nozzles with smaller sizes (about a tenfold reduction) and to provide at least two backup water supply systems (including one for station blackout) which can be turned on by remote manual operation or by simple procedures for connection and startup.
l l
l 1 _ _ - . -_.- - .. - - _.- _ ._ - - .- . - _ - _ _ _
.- .- . . - . - _ - - ~ -_-_--_ _- - --_ _ ---_-
ger k hk
~
3.3. Pressure Relief Currently available structural analyses for Mark I containments show ultimate failure at about twice the design pressure, usually failing at the knuckle between the upper cylindrical and lower sections of the drywell. However, these analyses have not taken into account the mechanical backing which may be provided by the biological shield surrounding the drywell. The ultimate strength of the Mark I may be quite a bit higher.
Other factors may control the selection of a pressure limit. The vent valves already on containment are tested or qualified to levels about equal to design pressure and may not be reliable at pressure far above that. In addition, such high back pressures would reclose SRVs, possibly exacerbating the accident.
The size and the durability of the vent path involves questions of accident scenario. Assuming that pressure considerations lead one to select a relief
! - at a level on the order of design pressure, some alternatives become apparent.
l First, this vent need not be the large steam escape path desired for an ATWS scenario; for the ATWS the operator would use one or more main steam lines to the turbine bypass. With ATWS set aside, only a decay heat level vapor flow vent capacity is needed to blunt the pressure rise. Since the containments are already designed to absorb initial sensible heat and the high, early decay heat within the design pressure, a flow equivalent to 1% of rated power at the .
I
- . - - - - + - ._---,m . - - . -
+-g.y - yy +--- -a.n-y----r------w,py c.y --m.s r
, . . . . ... .... . .... - a ...... w .,.. ...- ...-.... ~ .. _ .. ... . . . ~ . . , , ,
- . ., l
~
i venting pressure is sufficient. One percent power is equivalent to the decay heat generation-rate after' ten minutes. For Mark I containment, this
' translates-to six to eight inches vent diameter, based on venting at about 60
~
psig. Figure (2) was obtained from the August.1986 IDCOR study submitted to BWROG, and gives the estimated vent diameter as a function of power level.
The figure can be used to determine venting capacity for ATWS. For a 2800 MWe reactor ATWS power levels can range from 30 to 40% of or 840 to 1120 MWt. A vent diameter of 40 inches or so may be needed to manage ATWS by venting the containment.
One has the choice of designing a special purpose vent for this purpose leading directly to the stack or to use existing vent valves and ducts. The staff knows of no plant which already has a high. pressure vent path in place.
Given the highly undesirable effects of the potential vent path rupturing inside the plant, the Mark I containment venting strategy is to provide a
~
- bur'st-resistant path with reliable' valves;-capable'ofYemote manual opening .
and reclosing even in station blackout, to vent steam equal to 1% of rated power to the plant stack or a high point vent. The use of stack or other high
, point release will assure a substantial reduction of radiation doses due to
! post-accident venting. Figure (3) shows expected whole body dose as a
, function of distance for ground level and elevated release for average and adverse meteorology. The figure shows a substantial reduction in whole body ,
doses for elevated releases.
1,_ _
F .
i .
1.0 ;
i -l l 3 g i g 40 -
- CONTAINMENT PRESSURE = 0.5 MPa (58 psig) 0.8 L = 1M m (328 h)
E f = 0.02 32
@ g ;
F O.6 - H ,
m -
24 o u ;-
2 .
< i-E k 0 N !.
0.4 [ y
- z.
16 Z -
j m -
i 0.2 - a ,.
' - . 8 .- t O I I I I I I I I f
0 100 200.2 300 400 500 600 700 800 900 i,
'~l VENTING CAPACITY, Mw !'
- I' l'
i -
. Figure 2. Vent Size Requirement as'a Function of Power d :
(This figure is reproduced from IDCOR Report, " Evaluation of BWR Accident Mitigation Capability Relative to proposed NRC Changes," August 1986.)
7N e :.
l
~' _ __
_ , , , n . . . - . - . . . . . . . . . . _ - . . .. - . _ . . . . . - - . - .
~
QQEIT
. L:m_
. = = == = = :u =m= m - - - -
= - -
=
m = r= w m- = = = =m
-y-r -
FIGURE 2 -' a===
- EXPECTED WHOLE BODY RADIATION DOSE (REM) i -: H_ IH : : :
EE ai ii i FROM RELEASE OF 100% NOBLE GASES
-~ -~
-- ~ ' -
(1 HOUR DECAYED AND 5 HOURS DURATION OF r RELEASE) FROM 3412 MWt LWR VS. DISTANCE NOTES:
- 1. Graphs assume one hour holdup and decay prior 10M to release. Greater delay in release can produce {000
. mn: : L : : :: lower doses (e.g., as much as a factor of about e --~ ~ ~ ~ ~ -
y 'RE ==1 r 30 at one mile for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of inreactor holdup -==
< a comparedtoonehour).
(t g 8
b m 2.' Dose estimates are based upon MCCS computer gi = 9 _ - -
- code calculations using revised (relative to u' ,i n ____
CRAC and CRAC2) meteorological sampling models. --- '
5; a
- N 3. The likelihood of exceeding the
E
estimated 95 percentile dose is less
- i ' than 5% given release of 100% of ,, i m
~
noble gases as's'pecified above. , ,5 :
m i 5 :
p ". I l" N if5 M Ed NHU M i l-" g '100
- sei ;; ::3 a ==-, N ESTIMATED 95 PERCENTILE - - -
,h 1 -"
- i
, GROUND LEVEL R; LEASE 10 METERS g ty c:: 4
.?
=
CD :
)"!u&
8 =
4 ~
_-__w ii
= ESTIMATED 95 PERCENTILE
!gg
.- = = = =
]!
ii' Ni @
l t si i_. iF !! f :~_ 2 g g ;s;; g.g si d
' ili E _- -= I is = 5 - ELEVATED RELEASE 100 METERSj ;
- 8xg -== ~ ~ ~ ~
=
E
v y 3=
4 --=---
o .
g;a z _=_ -
-; w i b
g ;; -- - -
3 w :
5:r wu'O*uE % "
o
-al 88
"( , i ' " 3l
- m. < <
3g =-
m m, .
1 3
1 L i h A. 5
! . I d 10 w ,:._ m 9 ,, ,
- sz m ez
_ ! 9
=. = e se _:2 u es s m w rEt 10 4=- r-c e
g
=
2 ovy si i
- = r :
= = =.
er s =
- . = -g y a gr ie = s s : s= wi g II 1 r = M gs+ := s g - - t :r o M E : : = n =
3-
= == ==- = -
== = = == 5 EdT=_
.e,og 1 2 =
~-
ai
- =
M_-
sa 5-eo-x . __ __.. __ _
-_ . -Ms_;
m . GROUND LEVEL RELEASE 10 METERS i --' -~
og ,
- m m m a :s i r ewnm
: n ==z
, s y' m w gi f =
hi f 53 i si !w!i 55 i
W 5 55
-4 .
is i it p : :: :
. -2 55 5 5 '5 1 il- Me 25[ iE -
a 'E i 55
~
st E2 M 1 M m =-:- 53 :5
= =~
- i:!
2== =- ==-
5 :
=
==
E -5 M
2 5 = =- =
gjyM
,yjQf'
'~ ~ --'
5 __ __@. -
S =- - # -' -
_~-N-
,y =- 'ELEVATEDRELEAS[100 METERS %- __
- y -__
- :::::: ::: =:::: =: ::: == ==::: ::: =: :=
E-:5EE =E EE =EEEE EEEE ?! fisij iiii a:::::::::: :::p: ::::: ::: ::::::::::::::: ::::: :: :::::: ::: ::::: :: ::: :-@mEii -
_ DISTANCE FROM RELEASE POINT (MILES) = ,
Ag i iii i i i iii i i i iii i i i iii i i i iiE i i i iii i i i 1 4
' i n
a'3 , 4 5 6 7 8 9 10 11 12, 15 14
w .. . - .:: ...:a. .= a.. . ;e.: w . a z .::. . .c .: .. = ._ . :. .;= .- . . ., . : . . :. ;. -
l 3.4 Core Debris Control The. strategy identified in sections 3.2 and 3.3 should provide cooling of molten corium should it come out of the reactor vessel. If molten corium does reach the drywell wall, the combination of a spray-cooled interior
'and a heavily backed exterior make drywell to reactor building failure unlikely.
The vents on the other hand are a remaining possible debris travel path and F the torus room is pneumatically open to the reactor building. Therefore, the Mark I strategy is to ensure that, if the torus fails, the water in the torus will be retained in the torus room and the corner rooms, and will quench any corium which might reach there and limit the spread of damage by intensely radioactive material.
3.5: Procedures and Training The Emergency Procedure Guidelines, Rev. 4, now under review, have the scope and content to satisfy the needs identified in Section 2.0. The Mark I strategy then is to require that all licensees adopt all principal elements of EPG'Rev. 4, and' revise or modify as necessary to reflect iihe changes ,
occasioned by 3.1 to 3.4 above.
i 4.0 FORMULATION OF REQUIREMENTS Based on the preceding analysis the following requirements should be met by
{ any BWR with Mark I contcinment.
l 4.1 Hydrogen Control Present requirements imposed by 10 CFR 50.44 and the Technical Specifications shall be adhered to, no additional requirements are proposed.
_ . ~ _ __ ._ _., _ _.____ -. . . .__ _ _ _ , _ _ _ . _ _ _ . . _ _ . _ _ _
i . .
4.2- Containment Spray
~
All BWRs with Mark I containment shall provide at least two backup water supply systems for the containment drywell spray, one of which shall be functional during station blackout. Water to the spray system from these backup supplies shall be available by remote manual operation or by simple procedures for connection and startup which can be irplemented during a severe accident scenario.
In addition, the spray nozzles shall be adjusted so that an evenly distributed spray pattern will be developed in the drywell whether water is supplied by the primary source or either of the backup sources. A flow rate on the order of 1/10 of the present flow rate is considered typical, the licensee'shall select the ficw based on an analysis of plant specific parameters.
4.3 Pressure Relief The licensee shall select a pressure between design pressure and 1 times design pressure at which to open an exhaust path from the wetwell vapor space
. to the highest' vent point ~(stack or pipe) available. This line should be capable of handling water vapor flow equivalent to 1% decay heat at the vent pressure selected without significant chance of rupture before the desired release point. The line shall be equipped with isolation valves which can be opened and reclosed by remote manual operation or by simple procedures which can be implemented during severe accident scenarios including station blackout.
4.4 Core Debris Control &
l The licensee shall en g that the water in the suppression pool in the event l
^
. . ~ .: ...- . ~ . a ... . . w . _ . . ... . a .... .:.. ..:. ,
- a. .. .. . . - ... .....
o . .
\l of torus failure is held within the confines of the torus room and the corner rooms and cannot flow out to other parts of the plant.
4.5 Procedures and Training !
'/-ll h The licensee shall implement emergency operating procedures and other procedures based on all significant elements appropriate to its plant of
,J}( -
j Emergency Procedure Guidelines, Revision 4.
4.6 Quality and Design Standards Since these requirements are intended to be an optimized use of existing equipment it is expected that added equipment, of itself, need not meet the quality or design standards of safety related equipment. Nevertheless, modifications to or near equipment or systems which are already safety related shall not compromise the quality of such equipment or systems.
4.7 Implementation T'he' equipment changes-required'herein shall be' installed during the first refueling outage which begins nine (9) months after the effective date of this letter. The procedures and training required shall be implemented on a schedule reviewed and approved by the NRC. Given the implementation of the generic improvements of Mark I containments there is no need for an Individual l
Plant Evaluation (IPE) for containment performance. This does not remove the i
i need for an IPE which covers the system reliability or core melt frequency ,
i portion of the severe accident question.
~~ . . .- ,.
. WB 5.0. JUSTIFICATION FOR REQUIREMENTS There are two possible bases for justification of Mark I containment improvements, one on the basis that they are needed for safety and the second on the basis that they are justified backfits by cost-benefit analysis'.
Examination of both bases shows that they support the containment improvements.
5.1 Needed For Safety The present General Design Criteria (GDC) set requirements for containment performance. GDC 16 - Contr.inment Design says, "an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to sa'fety are not exceeded for as lona as postulated accident conditions require."
(Emphasis added). It is clear from the long application of this GDC to many designs that " postulated accident conditions" are design basis accident conditions, not severe accident conditions. In a similar way GDC 50 -
- Containment Design Basis says, "[the containment can accommodate with
" sibffli:1~e nt' margin] the'iiressirre?and' temperature' resulting from any loss-of-coolant accident. This margin shall reflect consideration of (1) the effects of potential energy sources which have not been included in the determination of the peak conditions, such as the energy in the steam generators and as required by $50.44 energy from metal-water and other chemical reactions that may result from degradation but not total failure of emergency core cooling functioning, (2) the limited experience and .
experimental data available for defining accident phenomena and containment
'I responses, and (3) the conservatism of the calculational model and input
i
- a l
~
parameters." The words " degradation but not total failure of emergency core cooling" clearly limit the application of this GDC to design basis accidents.
Thus, consideration of both these GDC indicates that any mandate for change to protect against core melt accidents lies outside the requirements of the existing regulations.
The Commission spoke to the need in the Severe Accident Policy Statement of August 8, 1985:
o Operating nuclear power plants require no further regulatory action to deal with severe accident issues unless significant new safety information arises to question whether there is adequate assurance of no undue risk to public health and safety.
o In the latter event, a careful assessment shall be made of the severe accident vulnerability posed by the issue and whether this u-- vulnerability istplantz or site specific or of generic importance.
1 .
o The most cost-effective options for reducing this vulnerability shall be identified and a decision shall be reached consistent with the cost-effectiveness criteria of the Commission's backfit policy as to which option or set of options (if any) are justifiable and i
required to be implemented.
'I i
- c. --- ..y. . _ , - _ _ . - - _ . - - - _ , - - _ _ - , . ~ . . . _ _ - ,
,, . .. ~ _ _ - _ _ _ _ _
.o In those instances where the technical issue goes beyond current regulatory requirements, generic rulemaking will be the preferred solution. In other cases, the issue should be disposed of through the conventional practice of issuing Bulletins and Orders or Generic Letters where modifications are justified through backfit policy, or through plant-specific decisionmaking along the lines of the Integrated Safety Assessment Program (ISAP) conception.
From these passages it is clear that the Commission intends to deal with severe accident issues if there is a question whether there is adequate assurance of no undue risk to public health and safety. As noted in'Section 1.0, the Reactor Safety Study estimate of BWR Mark I containment performance in the face of core melt was that it had a 90% chance of failure. Many .
expected that more refined' analyses of risk available now would show a much lower level of severe accident risks. In many ways that expectation has been satisfied but with the Mark I containment the later results have not been so
- encouraging. Again as noted in Section 1.0, the IDCOR Technical Summary Report presents an inferred containment failure rate of about 20%. The analysis d'one for the smaller Vermont Yankee plant yielded a 7% estimate.
Considering the continuing debate on uncertainties in these estimates, it is fair to say that the early failure rate for Mark I containment lies in the range of 90% to 10%, perhaps in the lower rather than the upper end of that range.
' ~ -
In discussing a plant with Mark I contairment in a Congressional hearing on July 16, 1986, the Commission responded as follows to the question:
Question Is a 90 percent chance of failure in the event of a core meltdown an acceptable failure rate?
Answer The NRC holds the position that the likelihood of core melt accidents in any plant should be very low and, in addition, that there should be
~
substantial assurance that the containment will mitigate the consequences of a core melt should one occur in order to ensure low risk to the public. It is not merely a question of having low risk but of having as well the defense-in-depth assurance of combined protection by prevention and mitigation... ,
l If Ee ar'edebating1'n the ranne of 90% to 10% failure protiability, even with ,
l the likelihood that it is closer to the lower figure, that is hardly
" substantial assurance that the containment will mitigate the consequences of a core melt should one occur." There is no quantitative synonym for substantial assurance but it is a defensible proposition if the range of debate can be shifted down to something more like 10% to 1%. The Mark I l
l strategy developed in the preceding sections is not quantified but it does ,
provide significant changes for the better in each of the areas of greatest uncertainty and significance for Mark I performance in core melt.
25-And.so it can be argued that these containment enhancements are needed for safety, to ensure low risk to the public by establishing substantial assurance that the containment will mitigate the consequences of a core melt should one occur.
5.2 Costs and Benefits The estimated costs of proposed action would vary substantially depending upon specific designs of plants and ease with which performance enhancements could be incorporated. IDCOR* has presented approximate ranges of costs. The estimated costs do not reflect any unique engineering difficulties or time available for modifications to be incorporated in plant maintenance butages, i
The cost considerations included in the IDCOR study include the following:
o Hardware o Installation o Test and Maintenance Plant Unavailability
~
- o o ALARA (Exposure Costs) o Costs of Procedure Changes and Training and Impact of Proposed I Backfits The cost of drywell sprays using fire pumps available at all plants was w - -.-r - + - - - - - - - ------------------------------------ - - -
+--w v--*m, ,-- . y - --- --- =--wwyw--- - - * - ---
,, ..-.......s.....&..... . ._...- _.. .. . .. - _ _ __.. _ _ --__.4 - ..... . . . - . . . - .
estimated to range from 0.6 million dollars to 1.1 million dollars. The cost of venting using vents of 6 to 18 inch ducts ranged from 0.1 to 1.1 million
-dollars. The cost of installing a short debris barrier was estimated to be 0.4 million dollars. No additional cost is expected for Mark I containment hydrogen control. No significant new costs are expected for implementing the emergency procedures guidelines that-the industry is pursuing as a result of TMI actions.- Based on the above IDCOR estimates it would appear that the cost of the proposed initiative should range from 0.7-2.2 million dollars per reactor.
To estimate the benefits of containment improvements one must estima'te the averted loss. The terms needed to estimate it are:
o FCM, Frequency of Core Melt
- . o CCFP, Conditional Containment Failure Probability
! o Loss, the monetized cost of a large release.
i
, . .c . n . . . ,. ,
,Tgn..e..,g Frequency of Core Melt first, the containeerifimprovements include ,
i operator training and procedures for handling the containment. Because of the J
close interaction of systems in a BWR improved procedures will undoubtedly have an effect of reducing FCM. For reference, in the Reactor Safety Study, l
about two-thirds of all core melts were caused by the failure of containment !
due to overheating which failure then caused the loss of core cooling. For
" Evaluation of BWR Accident Mitigation Capability Relative to Proposed NRC Changes, August 1986 l
simplicity in the calculation here the reduction of FCM will not be included, '
l thus apparently underestimating the value of the containment features. It is ,
I reasonable to do this because the reduction in FCM will come principally from the training and procedures whose costs are not included in the preceding section since the BWR owners are already committed to adopting most of them.
The proper choice of a typical FCM is difficult to make. Current IDCOR and NRC analyses of Peach Bottom suggest FCM on the order of 1x10 -5 /yr although the NRC results are dominated by station blackout sequences while the IOCOR results are dominated by ATWS. Examination of the results presented in Table 1 indicates a number of plant specific cases where FCM ranges up to and above 1x10~4/yr. Considering the diversity of systems and the flexibility of operation in a BWR, a FCM of 1x10 -5 /yr may well be attainable.
However, the modelling differences between IDCOR and NRC results at that level and the results from other plants at the higher level suggest the choice of 1x10 -4 /yr
_ *-as the typical value for cost-benefit analysis purposes.
The cost-benefit equation calls for a quantitative estimate of the CCFP before and after the containment changes. Considering the range of debate on the present state of containments, given previously as 90% to 10%, it is reasonable to use 50% as the "before" figure. If substantial improvement is
- achieved the exact value of the "after" figure is not important, but 5% will be used here.
l l
l l
. . - - ~ - __ .~.- _ -...... _ . _ _ . _ .. - - ... .. . _ _ _ .
~
~
The monetized consequences of a large release coming from early containment failure can be large. Consideration should be given to counting health effects above or counting offsite economic consequences as well. Previous work
- indicates that large early releases can cause on the order of 107 person rem offsite exposure. Monetized at $1000 per person-rem, this gives consequences of $10 10 per event. Studies also show** that offsite economic consequences of large early releases can be up in the tens of billions of dollars or even more. On the other hand, source term studies continue and some argue that these high consequences are derived from WASH-1400 vintage source terms. For this cost-benefit calculation the health effects only ,
consequence of 107 person-rem or $10 10 will be used. Another question arises in converting annual averted health effects into a present worth value. Since these are human health effects some argue that they should not be discounted in a present worth calculation. If that approach is taken then the averted loss per year is multiplied by the remaining years of plant operation. The
, m .ther
, o approach.7;is .to .use a , discount a factor puch ,as,.woul,d. be associated with averted economic losses. Table 2 has been prepared to illustrate the central
& lb & O^*f'A & ~.
Q. # A l
^
'YA NUREG/CR-2723, " Estimates of the Financial Consequences of Nuclear Power Reactor Accidents," Appendix A, September 1982
- NUREG/CR-3673, " Economic Risks of Nuclear Power Reactor Accidents"
- l p.2-13, April, 1984 l
I
]
^
k
~
Referring to Table 2, the cost is listed as $0.7-2.2 million, or in a rounder number, less than $3 million per reactor. The base calculation gives a benefit of $3 million of $12 million, indicating a balance of cost and benefit or a clear justification. The other calculations in Table 2 are a sensitivity analysis to explore the range'of outcomes with different assumptions. The lower FCM case uses the frequency of core melt currently being calculated by IDCOR and NRC for Peach Bottom. The next case illustrates less improvement in containment performance, only a factor of five. The next case assumes that containment performance now is better, the 20% CCFP inferred in the IDCOR report. The " optimistic" case uses the IDCOR values for FCM and present containment performance while the " pessimistic" case assumes a relat'ively high FCM and CCFP.
Comparison of these estimated benefits to the range of costs indicates that these proposed changes are justified backfits.
l ._
mos .. , .
l I
I
' - ~ - ~ - - . . .
TABLE 3 COST-BENEFIT ANALYSIS COST: $0.7-2.2M BENEFIT:(1) FM CCFP CCFP AVERTED AVERTED BEFORE AFTER LOSS /YR LOSS PRES. VALUE BASE
-4 CALCULATION- 1x10 /yr 0.5 0.05 '$4x105 /yr $3M/$12M
-5 LOWER FCM 1x10 /yr 0.5 ,
0.05 $4x104 /yr $0.3M/$1.2M LESS CHANGE
~4 IN CONTAINMENT 1x10 /yr 0.5 0.1 $4x105 /yr $3M/$12M BETTER CONTAINMENT
~4 ' '
. "~' TO' START 1x10 .
0.2 0.05 $2x105 /yr $2M/$6M "0PTIMISTIC"
-5 CALCULATION 1x10 0.2 0.05 $2x104 /yr $0.2M/$0.6M
" PESSIMISTIC"
~4 CALCULATION 3x10 0.9 0.1 $2x105 /yr $16M/$60M (1) FCM = Frequency of Core Melt CCFP = Conditional Containment Failure Probability AVERTED LOSS PRESENT VALUE expressed as A/B where A is the averted loss per year times 8 (roughly equivalent to discount at 12%/yr rate) and B is the averted loss per year times 30 (no discount).
t
_ - - . . ~ . . . - - . _ - . .,
.i
) BOILING WATER REACTOR i 1
PLANT INFO.~RMATION SYSTEM L r
1 t
1 DIVISION OF BWR LICENSING I t
l OFFICE OF NUCLEAR REACTOR REGULATION L WINTER 1986 l 1
i l
I i
l
1 ;_ : . w .. . ... . ._u..w:n.. . .a .: .w.- a u z .w.u_. . -.. . a . -
,. '. + ' .o .. .
Pcg3 Ns. 1
. 12/D2/C6
, BWR INFORMATION SYSTEM DATA IDENTIFICATION ITEM NAME PAGE IDENTIFICATION .
ADS ENGINEERED SAFETY FEATURES ARCHITECT / ENGINEER (A/E) SITE INFORMATION BACKUP PROJECT MANAGER PROJECT DIVISION INFORMATION BOTTOM OF CORE (IN) REACTOR DATA COND STORAGE VOLUME (CU FT) ECCS DATA CONDENSER COOLING PLANT SYSTEMS DATA
., CONSTRUCTOR SITE-INFORMATION
, CONTAINMENT CONFIGURATION PLANT SYSTEMS DATA CONTAINMENT DESIGN PRESSURE (PSIG) CONTAINMENT DATA CONTAINMENT FREE VOLUME (CU FT) CONTAINMENT DATA CONTAINMENT ULTIMATE PRESSURE (PSIG) CONTAINMENT DATA
~ ' CONTROL RODS NSSS DATA COOLANT ACTIVITY (TECH SPEC) NSSS DATA i CORE FLOW e WTD PWR (MLB/HR) REACTOR DATA
- CP DATE LICENSING ACTIVITY
! DIRECTOR PROJECT DIVISION INFORMATION
[ DISTANCE TO NPC (MILES) SITE INFORMATION i DOCKET NUMBER LICENSING ACTIVITY i
DRYWELL VOLUME (CU FT) CONTAINMENT DATA
! ECCS CONFIGURATION ENGINEERED SAFETY FEATURES EXCLUSION RADIUS (METER) SITE INFORMATION FEED PUMP TYPE NSSS DATA
! FEEDWATER TEMP (F) NSSS DATA .
l FUEL BUNDLES NSSS DATA FUEL CHANNEL THICKNESS (MILS) NSSS DATA HEAT GEN RATE, AVG (KW/FT) REACTOR DATA HPCI/HPCS FLOW (GPM) ECCS DATA
! JET PUMPS,. NUMBER REACTOR DATA LEVEL 1 TRIP SETPOINT (IN) REACTOR DATA LEVEL 8 TRIP SETPOINT (IN) REACTOR DATA LICENSE NUMBER LICENSING ACTIVITY
- , ,, $ LICENSED POWER-(MWT)< , . NSSS DATA LICENSEE / APPLICANT LICENSING ACTIVITY ;
i LICENSEE / APPLICANT . LICENSING ACTIVITY
. LOW POPULATION ZONE (METERS) SITE INFORMATION -
LPCI LOOP SELECTION LOGIC ENGINEERED SAFETY FEATURES LPCI PUMPS (NUMBER) ENGINEERED SAFETY FEATURES l LPCI RATED FLOW (GPM) ECCS DATA LPCI RATED PRESSURE (PSID) ECCS DATA LPCS RATED FLOW (GPM) ECCS DATA '
LPCS RATED PRESSURE (PSID) ECCS DATA NEAREST POPULATION CENTER (NPC) SITE INFORMATION e
l I
, ,n.,-- ..,,,,,...,c.-,..,.,_,,--.---,--n - - - _ . , _ _
- ~ . - Pcg3 No. 2 12/02/C6
. CWR INFORMATION SYSTEM
, DATA' IDENTIFICATION ITEM NAME PAGE IDENTIFICATION NON-IODINE COOLANT ACTIVITY (MCI /GR) NSSS DATA .
OFFGAS PROCESSING SYSTEM PLANT SYSTEMS DATA OFFGAS RELEASE HEIGHT (FT) SITE INFORMATION OL DATE LICENSING ACTIVITY POWER DENSITY, AVG (KW/L) REACTOR DATA PRODUCT LINE LICENSING ACTIVITY PROJECT DIRECTORATE PROJECT DIVISION INFORMATION PROJECT MANAGER PROJECT DIVISION INFORMATION RCIC FLOW (GPM) ECCS DATA RECIRCULATION CONTROL NSSS DATA RPV ID (IN) REACTOR DATA RV NORMAL WATER LEVEL (IN) REACTOR DATA SAFETY / RELIEF VALVES ENGINEERED SAFETY FEATURES SECONDARY CONTAINMENT VOLUME (CU FT) CONTAINMENT DATA SGTS FLOW RATE (CFM) PLANT SYSTEMS DATA SHROUD OD (IN) ECCS DATA SITE BOUNDARY (METERS) SITE INFORMATION SRV CAPACITY (LB/HR) PLANT SYSTEMS DATA SRV REFERENCE PRESSURE (PSIG) PLANT SYSTEMS DATA STEAM FLOW (MLB/HR) NSSS DATA T-G DESIGN (MWT) NSSS DATA '
TURBINE BYPASS CAPACITY (%) NSSS DATA ULTIMATE HEAT SINK (UHS) PLANT SYSTEMS DATA WETWELL DESIGN PRESSURE (PSIG) CONTAINMENT DATA WETWELL FREE VOLUMC (CU FT) CONTAINMENT DATA WETWELL WATER VOLUME (CU FT) CONTAINMENT DATA S
- 4?Cil tN(t0* '
i.- ?* ac
- t
, [
e 1
i i
p
$ e .
1
- ,., ._~.-...w
- .,...w....w~.~..~..>-cm..-...~ .
}
Page No. 1 12/02/96 BOILING MTER REACTOR INFORMATION SYSTEM LICENS!NS ACTIVITY PL'.~T MME DOCKET NO LICENSE PROD CP DATE OL DATE LICENSEE / APPLICANT
, NUMBER LINE lig Rock Point 1 05000155 DPR-6 BWR 1 MAY 31,1960 AUS 30,1962 CONSUMERS POWER Brouns Ferry 1 05000259 DPR-33 SWR 4 MY 10,1967 DEC 20,1973 TV4 BrownsFerry2 05000260 DPR-52 BWR 4 MY 10,1967 AUG 02,1974 TVA trownsFerry3 05000296 DPR-68 IWR 4 JUL 31,1968 606 18,1976 TV4 transuick 1 05000325 DPR-71 SWR 4 FEB 07,1970 NOV 12,1976 CAROLINA POWER & LIGHT Brunswick 2 05000324 DPR-62 SWR 4 FEB 07,1970 SEP 08,1974 CAROLINAPOWER& LIGHT Clinton 05000461 NPF-55 BWR 6 FEB 24,1976 SEP 29,1986 ILLINDISPOWER Cooper Station 05000298 DPR-46 IWR 4 JUN 04,1968 JAN 18,1974 NElRASKA PUBLit POWER DISTRICT Dresden 2 05000237 DPR-19 IWR 3 JAN 10,1966 DEC 22,1969 COM0 HEALTH EDISON Dresden 3 05000249 DPR-25 BWR 3 00T 14,1966 MAR 02,1971 COM 0 HEALTH EDISON Duane Arnold 05000331 DPR-49 BWR 4 JUN 22,1970 FEB 22,1974 10WA ELECTRIC POWER & LI6HT Fersi 2 05000341 NPF-43 SWR 4 SEP 27,1972 7/15/85(51) DETROITEDISON Fit: Patrick 05000333 DPR-59 IWR 4 MY 20,1970 OCT 17,1974 POWER AUTHORITY OF STATE OF NY SrandSuli1 05000416. DPR-29 IWR & SEP 04,1974 NDY 01,1984 MISSIS $1PPI POWER & LIGHT COMPANY Brand Gulf 2 05000417 DEFERR BWR & SEP 04,1974 IEFERRED MISSIS $1PPI POWER & LISHT COMPANY Hatch 1 05000321 DPR-57 IWR 4 SEP 30,1969 DCT 13,1974 BEDR614 P MER Hatch 2 05000346 NPF-5 IWR 4 DEC 27,1972 JUN 13,1978 SEORGIA ,*0WER HopeCreek 05000354 NPF-50 BWR 4 NOV 04,1974 APR !!,1986 PUBLIC SERVICE ELECTRIC & 6AS
- Lacrosse 05000409 DPR-45 AC MR 29,1963 JUL 03,1967 DA!RYLAND POWER LaSalle1 05000373 NPF-11 BWR 5 SEP 10,1973 AUG 13,1982 COM0 HEALTH EDISON LaSalle2 05000374 NPF-ll SWR 5 SEP 10,1973 MR 23,1994 COM0 HEALTH ED! SON Limerick 1 05000352 NPF-39 BWR 4 JUN 19,1974 AUG 08,1985 PHILADELPHIAELECTRIC Limerick 2 05000353 PENDIN BWR 4 JUN 19,1974 PENDING PHILADELPHIAELECTRIC Millstone 1 05000245 DPR-21 HR 3 MAY 19,1966 0CT 07,1970 NORTHEASTUTILITIES Monticello 05000263 DPR-22 SWR 3 JUN 19,1967 JAN 19,1971 NORTHERN STATES POWER Nine Mile Point ! 05000220 DPR-63 BWR 2 APR 12,1965 AUG 22,1969 N!A6 ARA MOHAWK POWER Nine Mile Point 2 05000410 NPF-54 BWR 5 JUN 24,1974 OCT 31,1986 N!AGARA MOHAWK POWER Oyster Creek I 05000219 DPR-16 BWR 2 DEC 15,1964 AUG 01,1969 BPU NUCLEAR CORPORATION Peach letton 2 05000277 DPR-44 BWR 4 JAN 31,1968 DEC 14,1973 PHILADELPHIAELECTRIC Peach'lettes 3~ 05000278 DPR-56 LWR 4"JAN 31,1968 JUL 02,1974 PHILADELPHIAELECTRIC Perry 1 05000440 NPF-58 IWR 6 MY 03,1981 MAR 18,1986 CLEVELANDELECTRICILLUMINATING Perry 2 05000441 DEFERR IWR 6 MAY 03,1981 DEFERRED- CLEVELANDELECTRICILLUMINATINS Pilgrie1 05000293 DPR-35 IWR 3 AUS 26,1918 SEP 15,1972 BOSTON EDISON BundCities1 05000254 DPR-29 BWR 3 FEB 15,1967 DEC 14,1972 COM 0 HEALTH EDISON Sund Cities 2 05000265 DPR 30 IWR 3 FEB 15,1967 DEC 14,1972 COM0 HEALTH EDISON River Send 05000458 NPF-47 IWR 6 MR 25,1977 NOV 20,1985 6ULF STATES UTILITIES Shorehae 05000322 NPF-36 BWR 4 APR 14,1973 7103/85(511 LONGISLANDLIGHTINGCOMPANY Susquehanna1 05000387 NPF-14 BWR 4 NOV 02,1973 NOV 12,1982 PENNSYLVANIA POWER & LIGHT Susquehanna2 05000388 NPF-22 IWR 4 NOV 02,1973 JUN 27,1984 PENNSYLVANIAPOWER&Ll6HT Versont Yankee 05000271 DPR-28 IWR 4 DEC 11,1967 MAR 21,1972 VERMONT YANKEE MUCLEAR POWER HP-2(Hanford2) 05000397 NPF-21 SWR 5 MR 19,1973 APR 13,1984 WASHIN6TONPUBLICPOWERSUPPLY$VS 9
- ______m_..__ _ _ _ _ _______
Page No. 1 12/02/06
, 80!LIN6WATERREACTORINFORMATIONSYSTEM PROJECT DIVIS!0N INFORMATION PLANT NAME PROJECT DIRECTOR PRrJECT SACKUP PM DIRECT. MANASER.
Big Rock Point 1 DH 1 ZWOLINSKI ROTELLA Browns Ferry 1 BC 2 MULLER SEARS GROTENHUIS Broans Ferry 2 HD 2 MULLER GEARS BROTENHUIS Bromns Ferry 3 l@ 2 MULLER BEARS BROTENHUIS Brunsuick 1 BH 2 MULLER SYLVESTER Brunsultk 2 SWD 2 MULLER SYLVESTER Clinton SWD 4 luTLER SIE6EL Cooper Station BC 2 MULLER LON6 Dresden2 BC 1 ZWOLINSKI STAN6 Dresden3 BC 1 IWOLINSKI STANS Duane Arnold HD 2 MULLER GILBERT Feral 2 le 3 ADENSAM STEFAND Fit: Patrick 30 2 MULLER ABELSDN R00NEY BrandBulf1 34 4 BUTLER KININER Grand Gulf 2 BWD4 BUTLER KINTNER Hatch i BWD 2 MULLER RIVENBARK 6 EARS Match 2 BWD 2 MULLER RIVEGARK BEARS HopeCreek HD 3 ADENSAM NA6 NERD.
Lacrosse HD1 IWOLINSKI BEVAN LaSalle! SWD3 ADENSAM 900RNIA LaSalle2 BWD 3 ADENSAM 900RNIA Limerick 1 BC 4 BUTLER MARTIN,R. CARUSO Limerick 2 BWD 4 luTLER MARilW,R. CARUSO Millstone ! Pl!A 6 RIMES SHEA AKSTULEWICI -
Monticello IND 1 IWOLINSKI LYNCH Nine Mile Point ! IND 1 IWOLINSKI KELLY Nise Mile Point 2 HD 3, ADENSAM HAU6 HEY Oyster Creek i IND I IWOLINSKI DON 0 HEW Peach lottos 2 WD 2 MULLER CLARK Peach Bottos 3 BWD2 MULLER CLARK Perry!..,, MD 4 BUTLER LEECH ,
~ Perry 2 HD 4 IUTLER LEECH Pilgrie! SWD 1 IWOLINSKI AULUCK ,
GuadCities1 BWD 1
~
IWOLINSKI ROTELLA Guad Cities 2 SWDi ZWOLINSKI ROTELLA River Bend HD 4 luTLER STERN C.orehas BWD 4 IUTLER LO,R. CARUSO Susquehannai BWD 3 ADENSAM THADANI Susquehanna 2 BWD 3 ADENSAM THADANI VersontYankee In 2 MULLER ROONEY W P-2 (Hanford 2) BWD3 ADENSAM BRADFUTE
~--. , -
- ~ - ~ -
- w. .
.........-.~...~:u....a.-.....,...~~.~......'..-...~...
- Pete uc. 1,
'. 12/02/86 * *
. B0!Ll"3 CTER REACTOR INFORMATIr! SYSTEM SITEINFORMATION _. ,_ _
PLANT NAME CONSTRUCTOR NEAREST DIST A/E OFFSAS SliE LOW EICL POPULATION (MI) RELEASE BDRY POP RADIUS CENTER HEIGHT DIST ZONE (Fil (METERS) (METERS) (METERS) .
Big Rock Point 1 MCHTEL TRAVERSE CITY 45 BECH 808 2020 808 BrownsFerry1 TVA DECATUR,ALA 10 TVA 600 3218 1465 Broens Ferry 2 TVA MCATUR,ALA 10 TVA 600 3218 1465 Browns Ferry 3 TVA KCATUR. ALA 10 TVA 600 3218 1465 Brunseitk 1 BRRT WILMINGTON 16 UEC Brunswick 2 BRRT WILMINGTON 16 UEC Clinton BALDWIN MCATUR,ILL 22 $6L VENT 975 4018 975 Cooper Station BLR LINCOLN, NEB 60 B&R 808 1609 746 Dresden2 UEC JOLIET,ILL 14 66L 310 805 8045 Dresden 3 UEC JOLIET,ILL 14 S6L 310 805 8045 Duane Arnold IECHTEL CEDAR RAPIDS, IONA B BECH 328 9656 440 Forei 2 DANIEL MONRDE,MICH & S&L VENT 4B50 915 Fit: Patrick S&W OSWE60 7 S&W 385 976 5500 976 Srand Gulf 1 SECHTEL VICKSBUR6,MISS 25 BECH 31.5tVI 3218 696 Grand Gulf 2 VICKSBUR6,MISS 25 MCH 31.5tV)
Hatch ! 6PC NAYCROSS,6A 48 MCH 1250 1250 Hattb2 BPC WAYCROSS,6A 48 BECH 1250 1250 HopeCreek ECHTEL WILMINGTON, DEL 18 MCH VENT 8045 901 ,
Lacrosse MAION LACROSSE,UlS 20 $6L LaSalle! CWE OTTOWA,ILL !! $6L 370 6400 509 LaSalle2 CNE OffAWA,ILL 11 S&L 370 6400 509 Limerick 1 DECHTEL POTTS00NN 2 BECH VENT 2043 731 Linerlek2 NCHTEL POTTSDOWN 2 KCH VENT 2043 731 Millstone! ESASCO NEWLON00N, CONN 3 E9ASCO375 Monticello IECHTEL ST. CLOUD, MINN 22 MCH Nine Mile Point i S&W OSWEBO,NY 7 NMPC 342 4400 1255 Nine Mile Point 2 S&W ' OSWE60,NY B S&W 170(V) 1400 6400 1400 Dyster Creek 1 B&R TDMS RIVER 9 B6R Peach Bottes 2 IECHTEL LANCASTER,PA 18 SECH 500 1060 7300 716 Peachlottos3 KCHTEL LANCASTER,PA 18 SECH 500 1060 7300 716
- Perryi KAISER PAINESVILLE', OH10 6 BIL VENT 4002 B63 Perry 2 PAINESVILLE, OHIO 6 YENT 4002 863 Pilgris! MCHTEL PLYMOUTHCENTER 2 KCH 400' 700 5000 300 GuadCities1 UEC CLINTON,IONA 7 S&L 310 2424 380 OuadCities2 UEC CLINTON,IONA 7 56L 310 2424 300
' River lend $6W BATON R006E, LA 24 66W 190(V) 4003 194 Shorehas $6W ISLIP,NY 21 S6W VENT 3218 285 Susquehanna1 NCHTEL HAILETON 12 KCH 200(V)
Susquehanna2 BECHTEL HA!LETON 12 BECH 200(V)
Vermont Yankee EBASCO NORTHAMPTON, MASS 30 EBASCO318 NNP-2(Hanford2) MCHTEL RICHLAND 12 86R 4800 1950
, .. .... ... . > ... .. - . . a . - .. - .. ..~. .
a.
- Page C . !
, 12/02/86 BOILING WATER REACTOR INFORMATION SYSTEM PLANT SYSTEMS DATA PLANT NAME CONT CONDENSER ULTIMATE OFF6AS SRV SRV CONF 16 C00 LING HEAT PROCESS CAPACITY REFERENCE SINK SYSTEM AT REF P PRESSURE .
(LB/HR) (PS!6) lig Rock Point 1 DRY ONCETHRU LAKE MICHI6AN 30MINHLDUP Brones Ferry 1 MK I ONCE THRU TENNESSEE RIVER F62 Broens Ferry 2 MK I DNCETHRU TENNESSEE RIVER F62 BrownsFerry3 MKI DNCE THRU TENNESSEE RIVER N-62 Brunswick 1 MK I ONCE THRU CAPE FEAR RIVER CRYO + RECOMB 829,000 1080 Brunsuitt 2 MK I ONCE THRU CAPE FEAR RIVER CRYO + RECOMB B29,000 1080 Clinton MK !!! ONCE THRU LAKECLINTON N-66 Cooper Station MK I ONCE THRU MISSOURIRIVER RECOMB + CHAR Dresden 2 MK I COOLINGLAKE KANKAKEE RIVER F62 Dresden3 MK I COOLINGLAKE KANKAKEE RIVER F62 Duane Arnold MK 1 COOLING TOWER CEDAR RAPIDS RIV N-62 Fere! 2 MK I *
- MB 870,000 1090 Fit 2 Patrick MK I ONCE THRU LAKEONTARIO REC + HOLDUP 829,000 1000 Grand Sulf 1 MK 111 COOLING TOWER MECH DRAFT TOWERS N-64 Srand Suli 2 MK !!! N-64 Hatch ! MK 1 COOLING TOWER ALTAMAHA RIVER F62 789,000 1080 Hatch 2 MKI C00LIN6 TOWER ALTAMAHA RIVER N-62 B29,000 10B0 Hope Creek MK ! e e CRYO + RECOMB 870,000 1090 Lacrosse DRY ONCE THRU MISSISS!PFI RIVER 72 HR T G K LaSalle ! MK 11 POND RESERVOIR N-62 LaSalle2 MK !! POND RESERVOIR F62 Limerick 1 MK 11 COOLING TOWER SCHUYKELL RIVER CATRECOM8 170,000 1090 Limerick 2 MK!! COOLING TOWER. SCHUYKELL RIVER CATRECOMB - -
Millstone ! MK I ONCETHRU LON6 ISLAND SOUND RECOMI+ CHAR Muticello MKI COOLING TOWER MIS $1SSIPPI RIVER RECOMI+ CHAR Nine Mile Point ! MK 1 ~ ONCE THRU LAKE ONTARIO RECOM8+ CHAR Nine Mile Point 2 MK !! COOLING TOWER LAKE ONTARIO RECOMI+ CHAR Oyster Creek 1 MKI ONCETHRU BARNEGAT BAY 30 MIN HOLDP
, Peach lottos 2 MK I ONCE THRU SUSQUEHANNARIVERREC+HLOUP 800,000 1080
- - Peach Octtos 3 MKI ONCETHRU SUSOUEHANNACIVERREC+HLDUP 900,000 1980 '
Perry 1 MK!!! COOLING TOutR LAKE ERIE N-64 ,
Perry 2 MK III N-64 Pilgris1 MK I ONCETHRU CAPE C00 BAY F62 l GuadCities1 MK1 ONCE THRU MIS $15$1PPI RIVER E 62 GuadCities2 MK I ONCE THRU MISSISSIPPI RIVER F62 River Bend MK 111 COOLIN6 TOWER COOLING TONER F64 Corehas MK !! ONCETHRU LON6 ISLAND SOUND AMB 870,000 1090 Susquehanna! MK!! CC,HNDCT SUSQUEHANNARIVERRECOMB B31,105 1080 Susquehanna2 MK !! CC,HNDCT SUSQUEHANNARIVERRECOM8 831,105 1080 l Versont Yankee MK 1 COOLIN6 TOWER CONNECTICUT RIVER RECOMB + CHAR C~P-2 iManford 2) MK!! COOLING TOWERS MECHANICAL TOWERS F64
-,.,.v-., - - - - - - - - . . _ - . - . - . , - - - - - - --- - - - - -nn, --.--m . ---- , - - - - - - - - - - - - - , -- - - . - --- - ---A
- 4 .
,, Page No, l' 30! LING NATER REACTOR INFORMATION SYSTEM
. ENGINEERED SAFEff FEATURES PLANT NAME ECCS LPCI SPRING / ADS $6TS CONF 16URAT!DN LOOP RELIEF / FLON LO61C SRV RATE (CFM)
Big Rock Point ! IC/LPCS NA 0/6/0 4 trowns Ferry i RCIC/HPCl/LPCl/LPCS N 0/0/13 6 Browns Ferry 2 RCIC/HPC1/LPCl/LPCS N 0/0/13 6 Browns Ferry 3 RCIC/HPC!/LPC1/LPCS N 0/0/13 6 Brunssick l RCIC/HPCI/LPCl/LPCS N 0/0/11 9 Brunswick 2 RCIC/HPCI/LPC1/LPCS N 0/0/11 7 Clinton RCIC/HPCS/LPC1/LPCS N 0/0/16 7 CooperStation RCIC/HPCI/LPCI/LPCS N 3/0/8 6 Dresden2 IC/HPCI/LPC1/LPCS Y B/4/1 5 Dresden 3 IC/HPCl/LPCl/LPCS Y 8/4/1 5 DuaneArnold RCIC/HPCI/LPCI/LPCS Y 2/0/6 4 Feral 2 RCIC/lFC!/LPC1/LFCS Y 0/0/15 5 Fit Patrick RCIC/HPCI/LPC1/LPCS N 0/0/11 7 Brand Buli 1 RCIC/HPCS/LPC1/LPCS N 0/0/20 8 2300 Grand Gulf 2 RCIC/HPCS/LPC1/LPCS N 0/0/20 B Hatch ! RCIC/HPCI/LPCl/LPCS N 0/0/11 7 Hatch 2 RCIC/HPC!/LPC1/LPCS N 0/0/11 7 HopeCreek RCIC/MPC!/LPCI/LPCS N 0/0/14 5 La Crosse HPCS/ACS N 3/0/0 M LaSalle1 RCIC/HPCS/LPC!/LPCS N 0/0/18 7 4000 LaSalle2 RCIC/HPCS/LPCl/LPCS N 0/0/1B 7 4000 Limerick 1 RCIC/HPCI/LPCI/LPCS N 0/0/14 5 1500 Limerick 2 RCIC/HPCl/LPCI/LPCS N 0/0/14 5 1500 Millstone i IC/FNCl/LPCl/LPCS Y 0/0/6 3 !!00 Montitello RCIC/HPC!/LPCI/LPCS Y 0/0/B 3 Nina Rile Point ! IC/LPCS N 16/6/0 6 Nine Rile Point 2 RCIC/HPCS/LPCI/LPCS N 0/0/1B 7 3500 Dyster Creek i IC/LPCS N 16/5/0 5 Peach Bottos 2 RCIC/HPC!/LPCl/LPCS W 2/0/11 5
. Peach lettos 3 RCIC/HPCI/LPC1/LPCS N 2/0/11 5 Perry 1 RCIC/HPCS/LPC1/LPCS N 0/0/19 8 Perry 2 RCIC/HPCS/LPCl/LPCS N 0/0/19 8 '
Pilgriei RCIC/HPCl/LPCl/LPCS Y 2/0/4 4 GuadCities1 RCIC/HPC!/LPC1/LPCS Y 9/4/1 5 GuadCities2 RCIC/HPCl/LPCI/LPCS Y I/4/1 5 River tend RCIC/HPCS/LPCl/LPCS N 0/0/16 7 Shorehas RCIC/HPCI/LPCl/LPCS N 0/0/11 7 lusquehanna1 RCIC/HPC!/LPC1/LPCS N 0/0/16 6 Susquehanna 2 RCIC/HPC!/LPCl/LPCS N 0/0/16 6 Vereont Yank u RCIC/HPC!/LPCI/LPCS N 2/0/7 4 NNP-2 (Hanford 2) RCIC/HPCS/LPCl/LPCS N 0/0/18 7 l
l
. w. - .a.. n .... . .. .. - .: ..w.~.au.a-_-_--...-.c.~.-._
' ' Page No. a 12/Q2/86 .
- BOILING BATER REACTOR INFORMATION SYSTEM
. NSSS DATA PLANT NAME Llc CONT FUEL FUEL RECIRC FEED TUR8 T-6 COOLANT NON-CDINE PWR RODS CHANN BUND FLOW PUMP / BYPASS DESIGN ACT I-133 ACTIVITY (MW)' THlCK CONT NUMBER CAPAC (MWT) (TECH SPEC) (TECH SPEC)
~
(MIL) (1)
Big Rock Point i 240 32 100 e Valve Motor /a 100+ 158 0.2 e Browns Ferry 1 3293 185 100 764 M6 Turb./3 26 3440 3.2 e Browns Ferry 2 3293 185 100 764 M6 Turb./3 26 3440 3.2
- Browns Ferry 3 3293 185 80 764 M6 Turb./3 26 3440 3.2
- Brunswick 1 2436 137 80 560 M6 Turb./2 85 2535 0.2 100/E Brunswick 2 2436 137 80 560 MS Turb./2 26 2535 0.2 100/E Clinton 2894 145 120 624 Valve M+T/s 35 3016 0.2 100/E Cooper Station 2381 137 80 548 M6 Turb./2 25 2486 3.1 e Dresden2 2527 177 80 724 M6 Motor /3 40 2527 0.2
- Dresden 3 2527 177 80 724 MS Motor /2 40 2527 0.2
- Duane Arnold 1658 89 80 368 MS Motor /2 26 1658 1.2
- Feral 2 3292 185 100 764 M6 Turb./2 26 3435 0.2 100/E Fit: Patrick 2436 137 100 560 M6 Turb./2 30 2535 3.1 e Brand Gulf 1 3833 193 120 800 Valve Turb./2 35 3995 0.2 100/E Brand Gulf 2 e e 120 e +LFM6 /2 3995 - -
Hatch 1 2436 137 100 560 M6 Turb./2 25 2535 0.2 e Hatch 2 2436 137 100 540 M6 Turb./2 25 2535 0.2 100/E HopeCreek 3293 185 100 764 MS Turb./3 26 2436 0.2 100/E -
Lacrosse 165 29 e 72 e Motor / 100 60 0.2 100/E LaSalle1 3323 185 100 764 Valve Turb./3 25 3458 0.2 100/E LaSalle 2 3323 185 100 764 Valve Turb./3 25 3458 0.2 100/E Limerick 1 3293 185 100 764 M6 Turb./3 25 3293 0.2 100/E Limerick 2 3293 185 100 764 M6 Turb./3 25 3293 - -
Millstone 1 2011 145 80 580 M6 Motor /3 100+ 2011 0.2 100/E Monticello 1670 121 80 484 M6 Motor /2 15 1670 5.0
- Nine Mile Point 1 1850.129 80 532 M6(5) T6 Shit 40 1850 25. e Nine Mile Point 2 3323 185 100 764 Valve Turb./3 25 3467 - -
Oyster Creek 1 1930 137 80 560 M6(5) Motor /3 45 1933 8.0 e PeachBottos2 3293 185 100 764 M6 Turb./3 30 3440 0.2 e PeachBottoe3 3293 185 100 764 M6 Turb./3 30 3440 0.2 e
~ Perry 1 3579 177 120 748 Valve M+T/l+2 35 3730 0.2 100/E Perry 2 3579 177 120 748 Valve M+T/1+2 35 3730 - -
Pilgriei 1998 145 80 580 M6 Motor /3 26 1998 20.
- Quad Cities 1 2511 177 80 724 M6 Motor /3 40 2511 5.
- Quad Cities 2 2511 177 80 724 MS Motor /3 40 2511 5. e l River Bend 2894 145 120 624 Valve Motor /3 10 3039 0.2 100/E Shorehas 2436 137 100 540 M6 Turb./2 30 2535 0.2 100/E Susquehanna 1 3293 185 80 764 M6 Turb./3 30 3439 0.2 100/E Susquehanna 2 3293 185 80 764 MB Turb./3 30 3439 0.2 100/E VersentIankee 1593 89 100 368 M6 Motor /3 100+ 1664 0.2 e UNP-2 (Hanford 2) 3323 185 100 764 Valve Turb./2 25 3468 0.2 100/E O
l l
' ~~ "
,, . . .. ~ .. ..-........:
d 4 ,
er .
[ Page No.. . I 12/02/86
- BOILIN6 NATER REACTOR INFORMATION SYSTEM
, CONTAIMENT DATA PLANT NAME CONT ERYWELL NETNELL NETNELL CONT FREE SEC CONT CONT DES NETNELL CONT ULT CONFI6 VOLUME AIR VOL NATER VOLUME VOLUME PRESSUE DES PRESS PRESSURE (CUFil (CU FT) (CU FT) (CUFil (CU Fil (PSIB) (PSIB) (PSIB) ligRockPoint1 DRY 1,150,350 N/A N/A 940,000 27.0 N/A 72.0 BrownsFerry1 KI 159,000 129,300 123,000 288,300 56.0 56.0 greens Ferry 2 KI 159,000 129,300 123,000 288,300 5&.0 36.0 ,
, Browns Ferry 3 MK I 159,000 129,300 123,000 288,300 56.0 56.0
, transaick i K1 164,100 124,M0 87,600 288,H0 62.0 62.0 transmick 2 KI 164,100 124,000 87,600 288,000 62.0 62.0 Clinton MK111 246,000 1,550,000 131,920 1,796,000 30.0 15.0 63.0 Cooper Station MKI 132,465 106,850 87,660 242,550 56.0 56.0 Dresden 2 MK 1 158,236 !!6,645112,203 274,881 62.0 62.0 tresden 3 MKI 158,236 !!6,645 112,203 274,881 62.0 62.0 Duane Arnold K1 118,000 94,270 58,900 212,270 56.0 56.0 Fersi 2 KI 163,730 130,900117,450 294,630 56.0 56.0 Fit: Patrick K1 154,5M 113,089 105,600 264,000 56.0 56.0 70.0 Brand Gulf 1 K !!! . 270,0M 1,400,000137,000 1,670,000 30.0 15.0 56.0 8 rand Buli 2 MK!!! 270,000 1,400,000 137,000 1,670,000 30.0 15.0 56.0 Hatch 1 MK 1 146,010 112,900 87,300 258,910 56.0 56.0 Hatch 2 K1 146,266 109,800 87,300 256,066 56.0 56.0 NopeCreek MK 1 169,000 133,500!!8,000 302,500 4. E6 62.0 62.0 100.0 Lacrosse DRV h/A N/A 262,000 52.0 N/A LaSalle 1 MK!! 221,513 e 142,000 e 47.0 47.0 151.0 LsSalle 2 K !! 221,513 e 142,000 e 47.0 47.0 151.0 Limerick 1 K 11 243,500 159.540128,300 412,540 1.8 E6 55.0 55.0 140.0 Limerick 2 K !! 243,5M 159.540 128,300 412,540 55.0 55.0 i ,
Millstone 1 KI 146,900 109,900 98,M0 256,800 1.7 E6 62.0 62.0 Monticello KI 134,200 103,510 68,000 237,700 _56.0 62.0 Nine Mile Point i K I - 180,000 120,000 89,000 300,000 62.0 35.0 Nine Mile Point 2 MK 11 340,900 e i e 68.0 Oyster Creek i KI 180,000 127,000 83,4 M 307,000 62.0 35.0
, Peach lottos 2 KI 175,000 127,700122,900 293,500 56.0 62.0 Peach lottes 3 KI 175,000 127,700122,900 293,500 56.0 .62.0 Perry 1 K 111 276,500 e 120,000 e 1.16 E6 30.0 15.0 56.0 Perry 2 K !!! 276,500 e 120,000 e 't.16E6 30.0 15.0 56.0 Pilgria1 MKI 147,000 120,0 M 84,000 267,000 56.0 56.0 had Cities 1 K1 158,236 116,645112,203 274,881 (2.0 62.0 BuadCities2 K1 158,236 116,645!!2,203 274,881 62.0 62.0 River lead MK111 234,000 e 127,930 e 2.21E6 25.0 20.0 53.0 Shorehas K !! 192,000 184,000 81,350 326,000 48.0 Susquehanna i K !! 239,600 153,900127,000 393,500 53.0 53.0 Susquehanna2 MK!! 239,6% 153,900127,000 393,500 53.0 53.0 Vermont Yankee MKI 134,000 112,200 68,000 246,200 56.0 62.0 MP-2 (Hanford 2) MK11 202,900 144,000112,000 346,900 45.0 e
4
~4 .
4 s* PageNo. *1
, 12/0I/86
- B0! LING NATER REACTOR INFORMATION SYSTEM ECCS DATA PLANINAME HPClilFCSLPCI LPCI LPC1 LPCS LPCS COND STOR COND SiDR lift 00D FLON t RATED RATED PUMPS RATED RATED VDL (MIN) VOL (DESN) 09 1000 PSI FLON PRE'lSURE FLON PRESSURE (CUFil (CUFT) (1N) ,
, (OPM) (BPM) (PSID) (NUM) 16PM) (PS!D1 ft) (d) lig Rock Point 1 N/A N/A e 400 L. MICH. N/A Brsans Ferry 1 5000 10000 4 6250 122 10,036 207 Broens Ferry 2 5000 10000 4 6250 122 18,036 207 Brouns Ferry 3 5000 10000 4 6250 105 18,036 207 Bronseick1 4250 5775 4 4625 113 13,370 178 Bransmith 2 4250 5775 4 4625 113 13,370 178 Clinton 1400 5050 3 5010 113 RC1CTK 185 Cooper Station 4250 7700 4 4500 113 7,948 178 Dresden2 5600 3625 4 4500 90 12.031 207 Dresden3 5600 3625 4 4500 90 12,031 207 BuaneArno1J 3000 3600 4 3020 113 12,090 145 Forsi2 5000 7500 4 6250 100 13,400 207 Fit: Patrick 4250 4375 4 4625 113 13,370 178 trendButi1 2500 7330 3 7000 122 22,667 215 Orand Gulf 2 2500 7330 3 7000 122 22,667 215 Natch ! 4750 5775 4 4625 !!3 13,370 17B Hatch 2 4250 5775 4 4625 113 13,370 178 NopeCreek 5600 10000 4 6250 105 18,04B 207 Lacrosse 50(al ---
900 b LaSalle! 6250 10600 2 4000 122 207 LaSalle2 6250 10600 2 4000 122 11,380 207 Limerick 1 5600 10000 4 6350 105 11,380 207 ,
Limerick 2 5600 10000 4 6350 105 207 Millstone 1 N/A 7500 165 4 3600 90 30,083 60,000 114 Montite!!o 3000 3000 4 3020 130 10,028 167 Nine Mile Point 1 3800
- N/A 0 3400 365 105,000 179 Nine Mile Point 2 6250 10600 2 6250 122 135,000 450,000 207 Oyster Creek 1 e N/A 0 3400 110 32,100 179 Peach Bottes 2 5000 19900 4 6250 105 18,000 207
- Feach Bottos 3 5000 19900 4 6250 105 18,000 '207 PortI1 6000 6500 3 6000 122 ,
202 Perry 2' 6000 6500 3 6000 122 202 Pilgris1 4250- 3600 4 3600 104 20,055 184 Ouad Cities 1 5600 3625 4 4500 90 12,031 207 Ouad Cities 2 500 3625 4 4500 90 12,031 207 River Bend 4900 5000 3 4900 113 202 Shorehas 4250 4075 4 4625 113 15,746 178 Susquehanna 1 5000 15000 2 6250 105 207 Susquehanna 7 5000 15000 2 6250 105 207 Vereont Yankee 4250 7200 4 3000 120 13,370 169 WNP-2 (Haniord 21 6250 10600 2 6250 122 207 4 e l
l s
- -g v-,em ---,-wv--- ~ - - - e 4 w -
---*~da v'
~^
Y.
E *4'PigeNo. 1
[ 12/Q2/84 .
DOILIN6 NATER REACTOR INFORMATION SYSTEM
. .. REAtiCRDAfn PLANTNAME RPV CORE STEAM POWER HT SEN JET NORMAL HI LEVEL Le Lo Lo 30TTOM 70P OF 10 FLOW FLOW DENS AVER PUMPS NATER TRIP LB L1 TRIP OF CORE JET
(!N) (MLB/HR) (MLB/HR) (KN/L) (KW/FT) LEVEL SETPOINT SET POINT PUMPS
~
(IN) (!N) (!N) (IN) (!N)
Big Rock Point i 106 10. 0.606 45.00 7.00 0 BrownsFerry1 251 102.5 13.33 50.70 7.05 20 Browns Ferry 2 251 102.5 13.33 50.70 7.05 20 troens Ferry 3 251 102.5 13.33 50.00 5.40 20 Branssick 1 218 77. 10.47 50.51 5.45 20 Iranswick 2 218 77. 10.47 50.21 7.10 20 Clinton 21B 84.5 12.45 52.40 5.75 20 Cooper Station 218 73.5 9.55 51.10 7.09 20 Dresden 2 251 98. 8.6 36.60 5.10 20 Dresden 3 251 98. E. 6 36.60 5.10 20 DuaneArsold 183 49. 7.17 51.21 7.08 16 Forsi2 251 100. 14.16 48.71 5.34 20 Fit: Patrick 218 75.6 10.47 51.21 7.12 20 Grand Sulf 1 251 !!2.5 16.49 54.15 5.93 24 Scand Suli 2 251 112.5 16.49 54.15 5.93 24 Hatch! 218 78.5 10.03 51.20 7.10 20 Hatch 2 218 77.0 10.47 49.15 5.38 20 Hope Creek 251 100. 14.16 50.70 5.34 20 -
La Crosse e 11.0 0.6 0.00 0.00 0 LaSalle1 251 106.5 14.3 50.00 5.40 20 LaSalle2 251 106.5 14.3 50.00 5.40 20 Limeritt1 251 100. 14.13 48.70 5.30 20 Liserick2 251 100. 14.13 .48.70 5.30 20 Millstone 1 224 69. 8.00 40.80 4.70 20 Monticello 205 57.6 5.9 33.60 4.97 20 Nine Mile Point 1 213 67.5 6.0 41.00 5.91 0 Nine Mile Point 2 251 108.5 14.3 50.00 5.40 20 Oyster Creek 1 .213 61. . _. 7.25 33.60 4.70 .O _ _
f iath tottes 2 .. ~ 1024=_-' 13.15i'==50.76--605:n 5%Zi.-L i.5 ~ . !!-i=
E E ~~ --
~
_ Peach Bottos 3 ' - 1 102.5J:13.4 9 i0.04 ~ 6.95 -20 2 .:
Perry 1
! T 238 104. 15.4 54.10 5.93 20
- Perry 2 238 104. 15.4 54.10 5.93 20 Pilgris 1 224 69. 7.6 38.80 5.42 20 GuadCities1 251 98. 8.6 36.37 5.08 20 GuadCities2 251 98. 8.6 36.37 5.08 20
!!iver Bend 218 84.5 12.45 54.10 5.93 20 Shorehas 218 77. 10.47 49.15 5.39 20 l Susquehanna 1 251 100. 13.48 48.71 5.34 20 Susquehanna 2 251 100. 13.48 48.71 5.34 20 l Vermont Yankee 205 48. 6.43 50.95 7.08 20 j NNP-2 (Hanford 2) 251 108.5 14.3 50.00 5.40 20 e .
l
,