ML20091J247
| ML20091J247 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 05/31/1984 |
| From: | Woolever E DUQUESNE LIGHT CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 2NRC-4-071, 2NRC-4-71, GL-84-04, GL-84-4, NUDOCS 8406050448 | |
| Download: ML20091J247 (19) | |
Text
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Nuclear Construction Division hiecon 2
2 Robinson Plaza. Building 2. Suite 210 Pittsburgh, PA 15205 May 31, 1984 Dr. Harold R. Denton Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC 20555 BEAVER VALLEY POWER STATION - UNIT NO. 2 Alternate Pipe Break Design Criteria and Safety Balance for Elimination of Large Primary Loop Pipe Ruptures
References:
1.
NRC Generic Letter 84-04 dated 2/1/84.
2.
DLC Letter 2NRC-4-017 dated 2/24/84.
3.
NRC Letter to DLC dated 4/10/84.
Dear Dr. Denton:
In support of our request for exemption to General Design Criteria 4 (Reference 2), please find enclosed the Alternate Pipe Break Design Criteria and the Leak Before Break Safety Balance for Beaver Valley Power Station Unit #2.
These two reports provide an alternative to postulating Double-Ended Cuillotine Breaks (DEGB) which are required in General Design Criteria 4 of.t.0CFR50 Appendix A.
The approach used in developing the BVPS-2 Alternate Pipe Break Criteria and the associated Safety Balance methodology is in accordance with the approach suggested by the NRC in Generic Letter 84-04 dated February 1,1984.
Shus1d you have any questions concerning this matter please contact this office.
Very truly yours, b (-
E.
. Woolever Vice President Nuclear Construction Division RBC:el Enclosures 8406050448 840531 PDR ADOCK 05000412 A
PDR cc:
Mr. G. Knighton (NRC) 7)
/
Mr. E. A. Licitra (NRC)
/
/
Ms. M. Ley (NRG)
[
Mr. G. Walton (NRC-Site)
I l
I ALTERNATE PIPE BREAK DESIGN CRITERIA FOR THE PRIMARY COOLANT LOOP AT BEAVER VALLEY UNIT #2 I.
Introduction Asymmetric blowdown loads on PWR primary systems results from postulated rapid-opening, double-ended guillotine breaks -(DEGB) at specific locations of reactor coolant piping.
This current pipe break criteria requires installation of pipe whip restraints and jet deflectors. Various restraints and deflectors have not yet been either designed, procured or installed at Beaver Valley Unit #2.
As an alternative to postulating DEGB (required in General Design Criteria - 4) the NRC has stated (Generic Letter 84-04) that demonstration of deterministic fracture mechanics
- analysis, as contained in References 1-4 is acceptable.
This report demonstrates the applicability of modeling and conclusions contained in References 1-4 t o the reactor coolant primary piping in Beaver Valley Unit #2.
Conclusions are also provided regarding the susceptibility of the reactor coolant primary loop piping to failure from the effects of intergranular stress corrosion cracking, water hammer and fatigue.
II.
Mechanistic Fracture Evaluation The information contained in topical reports WCAP 9558 Rev. 2, and WCAP 9787 includes the following:
1.
Definition of the primary piping loadings.
2.
Analyses to define the potential for fracture from ductile rupture and unstable flaw extension.
3.
Material tests to define the material tensile and toughness properties.
4.
Predictions of leak' rate from flaws that are postulated to occur in PWR primary system piping.
The folicwing is a brief demonstration of the application of the deterministic fracture mechanics analysis as contained in WCAP 9558 and WCAP 9787.
Input for this analysis was obtained from a review of the maximum dead weight, thermal and seismic load conditions on Beaver Valley Unit #2 primary loop piping.
A.
Loads Maximum loads in Beaver Valley Unit #2 have been determined to be enveloped by WCAP 9558 Rev. 2.
Loads acting on the Reactor Coolant Pressure Boundary (RCPB) piping during various plant conditions include the following:
L..
'l) weight of piping and its contents, system pressure,
- 2) restraint of thermal expansion, operating transients in addition to start-up and shutdown, and 3): postulated seismic events.
The maximum combination of the cxial and bending loads occurs at
-the crossover leg as shown below:
Loading Conditions Dead Weight Thermal Seismic Total Axial Load (Kips) 15.6 221.0 23.4 260.0 Bending Load-(Ft - Kips) 28.2 1794.6 177.4 2000.2 These loads are enveloped by the loads in WCAP 9558 Rev. 2.
B.
Fracture Mechanics Analysis An elastic plastic fracture mechanics analysis as performed in WCAP 9558 is performed.to demonstrate that significant margins against double-ended pipe break _ would be maintained' for PWR Stainless Steel primary piping that contains a ~1arge postulated crack and is subjected to large postulated loadings.
The analysis includes the potential for growth of an existing crack due to the application of the load be determined.
1.
Postulated Flaw A throughwall crack 7.5 inches long is postulated along the circ umference.
.2.
J Integral The maximum value of the J integral in the Beaver Valley Unit 2 primary loop piping is less than the maximum value of the J integral in WCAP 9558, 3.
Material Properties A comparison of Beaver Valley Unit #2 centrifuga11y cast stainless steel primary piping tensile and fracture toughness properties with those in the Westinghouse materials test program was made.
The comparison showed that Beaver Valley Unit #2 properties were enveloped in Westinghouse's materials test-program.
WCAP 10456 shows that thermal aging of piping will have end of life toughness properties with a. tearing modulus at least a factor of 2 greater than the applied tearing load.
The material chemistry examined in WCAP 10456 is worse than that expected to exist in Beaver Valley Unit #2.
p -
r
'4.'. Loads As shown previously, the loads of WCAP 9558 envelope the loads in Beaver Valley. Unit #2.
9
- 5. " Leak Rate Calculations
.s Since --the pressure and crack length are similar to those in WCAP 9558, specific leak rate calculatio_s are not. required and
-a 10. gpm' or greater leak rate is assumed.
A leak rate of 10
.gpm or. larger is sufficient for detection during normal operation since the leak detection system at Beaver Valley Unit 2 has:been designed to satisfy the requirements of Reg.
Guide 1.45.
- 6.
Summary Based in items 1 through 5, the postulated 7.5 inch flaw is
-both locally and globally stable. and will leak at a sufficiently high level to assure detectability.
III. Fatigue
. Fatigue--evaluation has shown that postulated surface flaws (inner diameter) will not grow significantly during the design life of Eeaver Valley Unit #2.
Thus fatigue effects are insignificant.
-IV.:
Stress-Corrosion Cracking (SCC)-
'Intergranular stress-corrosion cracking (IGSCC) is not expected.'to occur because all the~ conditions required for ICSCC are not present.
Contaminants in :the water are controlled by the following:
- 2) Hydrazine addit ions'-during start-up - and hydrogen overpressure during o'peration control the _ dissolved oxygen.
Thus, the levels 'are always below the technical ~ specification limit for oxygen of 0.10 ppm when the coolant temperature is above 2000 F.
- 3) Other impurities that might cause SCC such as halides and caustics are also rigidly controlled.
' V.
Water Hammer Water hammer is defined as the pressure change in a closed pipe caused lar sudden change in the fluid velocity. or rate of flow.
. transients are considered to include transients. involving steam flow
( s team. hammer) and two phase flow (e.g.
water entrainment in
.steamlines, steamfbubble collapse), in addition to the classical water Lhammer transients such as those. involving valve closing and pump start-(
up in solid water ' systems.
In the primary loop the coolant travels at a constant rate of flow.
Conditions for a sudden change in rate of flow would occur downstream of a safety valve or pilot. operated relief valve. These valves are not located on the' Primary Coolant System being analyzed and therefore, water'hansner-is not a concern.
VI.
. Conclusions
. An evaluation of ' the Beaver Valley Unit #2 primary loop piping system shows that the loads and material properties.(including the effects of thermal ' aging) are within the. limit / criteria of References 1-4 and that
- IGSCC, water hanuner and fatigue effects are insignificant and consequently the postulated reference flaw will be stable and will leak at a detectable rate, t
T 4
REFERENCES 1.
WCAP. 9558, Revision ' 2' (May 1981) " Mechanistic Fracture Evaluation of Reactor ~ Coolant-Pipe Containing a Postulated Circumferential Throughwall Crack" 2.
'.WCAP 9787 _(May 1981) " Tensile. and Toughness Properties of Primary Piping Weld. Metal for Use in Mechanistic Fracture Evaluation" 3.
. Letter Report NS-EPR-2519, E.
P.
Rahe to D.
G. Eisenhut (November 10, 1981)' Westinghouse Response to Questions and Comments Raised by Members of ' ACRS Subcommittee on. Metal Components During the Westinghouse Presentation on September 25, 1981.
4.
WCAP 10456 (November 1983) "The Effects of Thermal Ageing on the Structural Integrity of Cast Stainless Steel Piping for Westinghouse NSSS"
E LEAK BEFORE BREAK SAFETY BALANCE i1.
. Introduction-Beaver -Valley Unit #2 requests' an exemption from General Design. Criteria 4 ' (GDC-4) on 'the basis that - the avoidance of operational occupational
. exposure associated with the. use' of pipe whip restraints _ and jet deflectors-far -outweighs the small increas~e in public risks and potential, accident exposure. A safety balance was performed for sixteen
- Westinghouse A-2 plants -.as provided in an attachment to' Enclosure 2 of Generic Letter 84-04.
Using this methodology, a plant specific safety balance was' performed for Beaver _ Valley Unit #2..This report provides a pla'ntJspecific safety balance which estimates the public risk and the avoided occupational exposure - for, not; installing pipe whip restraints Land : jet ~ deflectors to mitigate the consequences of asymmetric blowdown
, loading-in'the primary system.
The estimated reduction in public risk.for installing pipe whip
. restraints and jet deflectors as ' necessary to mitigate or withstand asymmetric pressure blowdown loads is very small, 0.84 man-rem total for the nominal ~ case.
The estimated reduction in accidental occupational exposure due to installation is 0.12 man-rem total for the nominal case.
These-small l changes - result from the estimated small reduction in core-
- melt frequency. of.1.4x10-7 events / reactor year.that would result from installation.of-the protective structures.
On the other hand, the
~
- .occ upa t ional-exposure estimated for maintaining the protective
~
structures would increase.by 80 man-re:ns.
Conseq'uently, the savings in exposure. by: granting the exemption far exceed the potentially small
- increase'- in public' risk and - avoided accident exposure associated with
. pipe restraint devices..
II.
Scope-A.
Jet Impingement Protection
~
. The essential' tagets of reactor-coolant line breaks can be divided into three categories:
1.
Structural,
- 2.
Westinghouse piping scope, and 3.
Stone &. Webster (S&W) pi' ping sco'pe.
Among. the -many assumptions.in this analysis, it' is first assumed that the breaks would not be shielded at their source.
Therefore,'
protection would be shielded. at the target any-target. requiring ~
itself.
Secondly, it ;is assumed that af t er ' the stress analyses. on structural. targets are completed the result would be that there are no unacceptable structural interactions.
Therefore, no shields ' are
~
. postulated - to - protect structures from the reactor coolant line breaks.
, 'As for the.above -Westinghouse scope piping targets, S&W has
- completed the-stress. analysis for the primary loop.
Loads must be
, analyzed to determine if. jet ' load protection is required.
It will' be assumed for this Analysis: that. no jet bumpers will be required.
These assumptions have a conservative effect on both the value ; and
_ impact results.
e The S&W; target / break interactions are identified below:
Target Line No.
Break No.
2CHS-002-97-1 3,5,8,12 '.
~
2RCS-002-045-1 3,4,5,8,12
-2 SIS-006-266-1 4,5,8,12 2 SIS-006-270-1 4,5,8,12 2BDG-025-10-2 5,8-
-2RCS-008-40-1
. 5,8 2RCS-008-41-1 5,8,12.
-2 SIS-012-71-1 5,8 2CHS-002-141-1 12 2CC-002-PB3
-12 2CC-002-PB4 12 18 Total S&W Scope Break-Target Interactions have been identified for
- Loop B.
^ :
It is. assumed that these eight S&W. scope piping targets for this
? loop are 2 typical for the other. two loops. Therefore, there would be approximately 24 shields to be adddd.
All shields would require periodic inspection and maintenance.
15. -For RC Equipment-Restraints.
k
.This' ' analysis applied to.the -following restraints which are-designed to prevent. damage ~ to RC loop equipment af ter. a postulated
- DECB:
for each of 3 loops there are two restraints on;the crossover leg elbows, 1. restraint.on _ the hot leg, and i restraint on the cold
'legi III. Assumptions-
' The' above estimated changes in -public risk and' avoided accident : exposure -
iwere obtained by f utilizing applicable portions of the plant risk model
. developed for : the calculation of severe reactor accident risks provided in' the Beaver Valley Unit. 2 Environmental' Report.
~t
~
'The following major assumptions were utilized:
-1.. A Double-Ended Guillotine Break' (DEGB) was assumed. l This a'ssumption
.is very. conservative ; in light of the mechanistic fracture analysis which has :shown that the postulated reference flaw is stable and Lwill leak at a-detectable rate.
2.
The ' installation Lof pipe whip restraints is. assumed to eliminate the
. possibility that a DEGB inside the reactor-cavity would Lead directly to a large LOCA proceeding to an early core melt. : This -
also is a conservative. assumption.
K3.
If ia DECB were to occur outside the reactor cavity, it.could lead to
' core melt through the additional failure of emergency core cooling.
L.k-
[ '
4.
Estimates of DEGB frequencies for large primary system piping were developed from two sources of data, a.
.The upper-estimate of 10-5 per reactor-year is based on_a paper on nuclear and non-nuclear pipe reliability data
" Reliability of Piping in Light Water Reactors" dated 1977 by S. H. Bush as quoted in Generic Letter 84-04, b.
Due nominal estimate of 6x10-7 per reactor year for primary system-piping outside the reactor cavity and 1.4x10~7 per reactor year inside the reactor cavity are' the values for a three loop plant derived from Report SAI-001-PA dated June 1976 prepared by D. O. Harris and R.
R. Fullwood as quoted in Generic Letter 84-04.
c.
Both the upper and nominal estimated DEGB frequencies are less than the WASH-1400 large LOCA median frequency of
.1x10-4 per reactor year.
The following table identifies several. factors associated with Beaver Valley Unit #2, Westinghouse A-2 Plants, and the data base used l
for WASH 1400 that support the use of a lower pipe break frequency.
Beaver Valley Unit kASH-1400 Large
(
[
Factor
- 2 And W A-2 Plants LOCA Pipe Size
> 30" diameter
> 6" diameter Pipe material Austentic stainless Carbon steel and-steel-
, stainless steel System and Class Only Class I primary Miscellaneous of pipe system pipe with
. primary and nuclear grade QA and secondary system ISI piping of various-classifications Type of failure Doubled-ended Circumferential Guillotine Break and longitudinal (DEGB) only.
breaks, large cracks Failure location Selected primary Random system system break break locations locations Leak detection LDS capability to No requirement or system (LDS) detect leak in a provision for timely manner to leak detection maintain large margin'against unstable crack extension
w y<-
- s 5.
Public ' dose estimate for the release categories were derived using the CRAC-2 code.
Inputs to the code include the release categories used in the Beaver Valley Unit 2 (BV-2) severe accident impact evaluation, site specific populat ion and umteorology, and 10 mile population evacuation.
These are based on a 350 mile radius release model.
6.
The change in occupational exposure associated with accident avoidance assumes 20,000 man-rem / core melt to clean up the plant
'and recover from the accident as indicated in NUREG/CR-2800.
7.
The estimated occupational exposure associated with maintaining the-protective structures is also considered.
8.
Financial considerations have not been addressed in this
. report.
IV.
Results A.
Public Health
- A.1 Description of Methodology The Value-Impact Analysis attached to NRC generic letter 84-04 contains analysis for 16 Westinghouse PWR plants in assessing the, average man-rem release from DEGB LOCA events.
The analysis is carried out for plants which do not have pipe whip restraints.
In the' analysis initiating events are divided into two categories:
Category 1.
- DEGB LOCA occurs in the reactor cavity and leads to early core melt.
Category II. - DEGB LOCA occurs outside the reactor cavity and leads to challenges to the ECCS and containment safeguards.
'The nominal and upperbound initEating event frequencies for the above events are given below:
1.
nominal frequency
= 9.0E-08/py. ( for two-loop plants) upperbound frequency
= 2.0E-06/py.
II. nominal frequency
= 4.0E-07/py. ( for two-loop plants) upperbound frequency
= 8.0E-06/py.
An updated WASH-1400 analysis is used to calculate the average man-rem exposure from the above events.
t-
~
-5 y:
_-In. the ' following analysis for Beaver Valley Unit 2, the same initiating. event' frequencies are used and the same major
.modeling assumptions are utilized.
The severe accident risk analysis carried out-for the Beaver Valley Unit 2 environmental report fis - used to estimate the average man-rem exposure from the above events.
This analysis is based on a 350 mile radius around the plant, whereas a 50 mile radius is considered in the
. attachment to Generic Letter 84-04.
1Ruis, the results of the present analysis tend to be conservative.
A.2 DEGB LOCA' Incremental Risk Analysis.
The.1init iating event frequencies used in this analysis are egiven below:
- 1. Breaks in reactor cavity:
1.4E-07/py (nominal)
-2.0E-06/py (upperbound)
II.~ Breaks outside reactor cavity:
6.0E-07/py (nominal) 8.0E-06/py (upperbound)
The. nominal. frequencies have been adjusted by a factor of 3/2 to account for the three. loops of the present plant.
The s'even release. categories associated with DEGB LOCAs are defined in Table.1.
The mean value for the total whole: body.
man-rem, for.each release category is given in Table 2.
The incremental risk is calculated by using the same model as was used -in the attachment' to Generic. Letter 84-04. -Namely the
~
-following equation is used:
_ EQUATION 1)
(
dRISK = RISKY + 0.2
- RISK r; y
where-RISKY;= man-rem risk from ' in-cavity breaks,
i,
- RISKyy
= man-rem risk from breaks outside of the reactor cavity.
~
L.
~
'The' factor of 0.2 is. used to account for.the effect of sys t em -
i, interactions leading.to additional risk without ? pipe whip restraints.
This model conservatively assumes that-the removal of pipe whip restraints results in. asymmetric blowdown from in-cavity breaks causing all cases to result in' core; melt.
(;
i u
l L
' ~
RISK is calculated as RISK =
Fi
- Ri (EQUATION 2) i=1 where Fi = frequency of being in the ith release category
._(summarized in Table 3);
Ri = average man-rem ass'ociated with the ith release category-(Table 2)
A.2.1
' Calculation'of Nominal Risk from DEGB LOCAs a.
Initiating event Category I.
In-this category, the ECCS is modeled to be insufficient to cool the core -and early core melt is assumed.
The frequencies of. the release categories have been calculated and are displayed in column 1 of Table 3.
Using Table 2, column l_ of Table 3 and Equation 2, the man-rem risk is calculated to be RISKY = 2. lE-02 man-rem /py.
b.
Initiating event Category II.
In this' category,.. the _already calculated LOCA event consequences have been modified to calculate the release category frequencies with the given initiating event frequency.
The re'sults are summarized in column 3 of Table 3 and the risk is calculated ' to be
. RISK y= 3.4E-04 man-rem /py.
y Combining the two initiating events-for. inside and outside of _ the reactor' cavity, the stotal risk is - found (using Equation 1) as follows:
dRISKN _ = 2. lE-02 + 0.2.
- 3.4E-04 = 2. lE-02 man-rem / year dRISKN = 2. lE-02 + 0.2 *: 3.4E-04 = 2.1E-02 man-rem / year _ -
A.2.2 Calculation of Upperbound Risk from DEGB LOCA Events The upperbound estimates are made using the upperbound initiating event frequencies given above, a.
Initiating event Category I.
The frequencies of the release categories have been calculated and are displayed in column 2 of Table 2, Using this and Table 2, the man-rem risk is calculated to be RISKY = 0.31 man-rem /py.
b.
Initiating event Category II.
are summarized in column 4 The release category frequencies of Table 3 and the risk is calculated to be RISK y = 4.5E-03 man-rem /py.
y Combining the two initiating events for inside and outside of the reactor cavity, the total risk is found as shown:
dRISKg = 0.31 + 0.2
- 4.5E-03 = 0.31 man-rem /py A.3 Results Taking the plant life to be 40 years, the nominal and upperbound risk estimates are calculated below.
RISKN = 40 X 0.021 = 0.84 man-rem RISKg = 40 x 0.31 = 12.4 man-rem B.
Core Melt Frequency The increase in core melt frequency is dominated by the in-c avity events.
This occurs because the ECCS functions in most cases and only a small fraction of the LOCA event s outside of the cavity proceed to core melt.
The nominal and upperbound estimates of the change in core melt frequency are therefore given by the following expressions:
dCMN = 1.4 x 10-7/py dCMU = 2.0 x 10-6/py
C.~
Occupational Exposure - Accidental
- The - increased occupational exposure - from accidents can be estimated as the product of the - change in total core melt frequency and the major occupationall exposure likely to occur in the event of a accident. The change in core melt frequency was estimated as 1.4E-7 events / year.
The occupational exposure in the event of a major accident:has'two components.
The first is the "immediate" exposure
~
to 'the personnel onsite during the span of the event and its short tenn ~ control.-
The second is the longer term exposure associated.
with the cleanup and recovery from the accident.
The total ~ avoided occupational exposure is calculated as follows:
DI0A = N*T*DOA, where DOA = P(Dro + DLTO) where DTOA'= Total. avoided occupational dose N = Number of affected facilities = 1 T = Average remaining lifetime = 40 years
&4
.DOA = Avoided occupational dose per reactor year P.
=. Change'in core-melt frequency Drg = "Immediate" o'ccupational dose DLTO = Long-term' occupational dose.
'Results of the' calculations are shown below.
Uncertainties are conservatively propagated by the use of extremes (e.g.- upper bound DTO + upper; bound DLTO)*
Increase in Immediate(a) Long Term (a)
Total Core Melt Occupational Occupational Avoided
-Frequency Dose Dose Occupational (events /
(man-rem /
(man-rem /
Exposure reactor yr) event)-
event)
(man-rem)
Nominal 1.4E-7 IE3 2E4
.12 Estimate Upper 2E-6 4E3'
-3E4 2.72 Estimate
-(a) Based on cleanup and decommissioning estimates, NUREG/CR-2601 T (Murphy 1982) as quoted in Generic Letter 84-04.
D.
Occupational Exposure - Operational By not installing the jet deflectors and pipe whip restraints, operational exposure dose is reduced.
The amount of occupational exposure dose would be incurred due to the following:
1.
Slowing down of normally anticipated work activities.
a.
Increased inaccessibility for personnel and equipment to and from primary system componets, b.
Increased congestion between adjacent compartments, c.
More difficult housekeeping in areas of high contamination potential, d.
Occasional jet shield removal and replacement (often requiring polar crane time).
2.
Additional routine-maintence of jet deflectors.
a.
Periodic paint inspection.
b.
Maintenance of jet deflectors (requiring support from scaffold
- builders, laborers, inspectors and radiation protection personnel).
To estimate the potential exposure _ amount, it was assumed that two additional man-weeks per plant year would be spent inside containment if the jet _ deflectors and pipe whip restraints were installed.
It was assumed that all workers will receive an average 25 mr/hr.
dose rate. Therefore the total dose is estimated below.
Operational dose averted = (80 man-hr/py)*
(40 plant years) * (0.025 R/ man-hr) 80 man-rem
=
Total avoided occupational doses due to implementat ion, operation and maintenance are shown below.
Upper and lower estimates were developed using the following model (Andrews et al. 1983):
3X dose Dose
=
upper expected Activity Dose Avoided (man-rem)
Operation, Maintenance Nominal 80 Upper Estimate 240 i
V.
Conclusions The results of the safety balance are summarized below.
The table clearly shows that there is a net safety gain as a result of not installing protection against asymmetric dynamic loads.
Leak Before Break Value Summary Dose (man-rem)
Factors.
Nominal Estimate Upper Estimate Public Health-
.84
-12.4 Occupational
.12
-2.72 Exposure (Accidental)
Occupational 80 240 Exposure (Operational)
Values Subtotal 79.04 224.88 As can be seen from the Values. Subtotal a net safety gain of 79 to 225 man-rem is achieved by not installing the primary loop pipe whip restraints and the associated jet impingement shields.
e
TABLE 1.
RELEASE CATEGORY DEFINITIONS BV2-2 RELEASE CATEGORY FOR CORE MELT SEQUENCES WHICH COULD LEAD TO AN EARLY OVERPRESSURE OF THE CONTAINMENT WITH NO SPRAYS OPERATIONAL AND SHORT WARNING TIME FOR EVACUATION.
BV2 RELEASE CATEGORY FOR CORE MELT SEQUENCES WICH CAN LEAD TO INTERMEDIATE CONTAINMENT FAILURE WITHOUT SPRAYS WITH LATE CORE MELT.
BV2-5~
RELEASE CATEGORY FOR CORE MELT SEQUENCES WHICH CAN LEAD TO INTERMEDIATE CONTAINMENT FAILURE WITHOUT SPRAYS WITH EARLY CORE MELT.
BV2 RELEASE CATECORY FOR CORE MELT SEQUENCES WHICH CAN LEAD TO LATE CONTAINMENT FAILURE AND NO SPRA7S OPERATIONAL.
BV2-7 RELEASE CATEGORY FOR CORE MELT SEQUENCES WHICH CAN LEAD TO INTERMEDIATE CONTAINMENT FAILURE WITH SPRAYS OPERATIONAL.
BV2-8 RELEASE CATEGORY FDR CORE MELT SEQUENCES WHICH CAN LEAD TO LATE CONTAINMENT FAILURE WITH SPRAYS OPERATIONAL.
BV2-9 RELEASE CATEGORY FOR CORE MELT SEQUENCES WHICH COULD LEAD TO
.BASEMAT MELT THROUGH.
m
l' e;.
TABLE 2.
TOTAL WHOLE BODY MAN-REM FOR RELEASE CATEGORIES (Ri)
RELEASE CATEGORY..
MAN-REM (MEAN VALUE) t BV2-2 4.6 X 107 BV2-4 3.7 X 107
- BV2-5 3.7 X 107 BV2-6 2.9 X 107 i
BV2-7 8.9 X 104
- BV2-8 3.8 X 104 BV2-9 4.0 X 103 i-u i
l
'1 p
I f
d
?-
' S q.
1 e
1 9
y
~-v---e-,.-
,.9.,
14w.m,..,-9.m.,-,pw ww-
,--,me.,-,45--,-
-,9
,,e.,
.e..e-p.,,m,e-c.,
,-.,---.--.~,%
m...,_
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TABLE.3.
RELEASE CATEGORY FREQUENCIES FOR DEGB LOCAS (F )
i RELEASE CATEGORY COLUMN 1 COLUNH 2 COLUMN 3 COLUMN 4 INSIDE CAVITY-OUTSIDE CAVITY NOMINAL UPPERBOUND NOMINAL UPPERBOUND BV2-2 1.54E-10 2.20E-09 3.80E-12 5.07E-11 BV2-4 0.0 0.0 1.30E-13 1.73E-12 BV2-5 1.58E-10 2.26E-09 1.32E-12 1.76E-11 BV2-6 3.72E-11 5.32E-10 3.40E-13 4.54E-12 BV2-7 3.96E-08 5.65E-07 9.25E-10 1.23E-08 BV2-8 1.00E-07 1.43E-06 2.37E-10 3.16E-09 BV2-9 1.00E-10 1.43E-09 2.10E-09 2.80E-08