ML20136F271
Text
ywa _._.
m.;
ma r m___.,_
= -, - -
2m_,_
.v, g.F 22it tic &*:.
J W
f":
- sgr
- ur
- Cr N
3 DISTRIBUTION.
DOCKET FILES AEB 9/F MWohl WPasedag grianrrateg Docket No.: 50-424/425 AD/RP RF ~
SEP gg y MEMORE DUM FOR: Thomas M. Novak, Assistant Director for Licensing Division of Licensing FROH:
Daniel R. Huller, Assistant Director I
for Radiation Protection Division of Systems Integration
SUBJECT:
V0GTLE ELECTRIC GENERATING PLANT - ORAFT SER INPUT FROM THE ACCIDENT EVALUATION BRANCH l
PLANT NAME: Vogtle Electric Generating Plant Unit I and Unit 2 DOCKET NO.: 50-424/425 RESPONSIBLE BRANCH: Licensing Branch No. 4 PROJECT HANAGER:
M. A. Hiller REVIEW STATUS: AEB review continuing Enclosed is the draft SER input (Enclosure 1) from the Accident Evaluatior. Branch which was typed by CRESS. This input includes sections on the Radiological Consequences of Accidents (Section 15.X) and Control Room Habitability (Section 6.4).
We have identified an open item dealing with different control room leak rates assumed for protection of control room personnel from toxic gases and from radionuclides following design basis accidents. Additionally, information is needed from the applicant on the data used to estimate the control room dose following a LOCA (Question 450.3) and on the toxic gas evaluation listed in FSAR Table 7.2.3-18.
The evaluation should include data described in Table C-3 of R. G. 1.78. Until this matter is resolved, habitability of the control room following toxic gas releases j
or radianuclide releases is identified as an open item for this draft SER.
In response to staff questions on the icent's steam generator tube rupture (SGTR) accident analysis, the applicant states that the Westinghouse Owners' Group is investigating general SGTR licensing concerns and will address the staff's concerns through a generic resolution in late 1984. Upon receipt of this additional information, d
. the s;aff will complete its review of the SGTR event and the Or'adiciogical consequences thereof.
I.>
Mk
... A %
,,o,
...... d....'1 l O l..i..UD '..
~...................
N y.... -...
.=
................. 7
............. ~......
...........--aa.
~~~~aa.~-aa a a a a a * ~ " *. ""
a"aaaa~~.
m)
. " - ~ ~ - ~ ~ ~
...... ~. - ~. - -.
roaw m no,ecinacw ouo OFFICIAL RECORD COPY
- u.s.am am-.co.u
. ~, -
..n
w -w-8
~ _ -.~.... -,-.
~.
. ~ - -
t::==
TZ 2 M
M g
E..
W T
%~
T
~
Thomas M. Novak.
The plant ESF filter designs have been determined to be acceptable and in accordance with the guidelines of Regulatory Guide 1.52, with the following exceptions, which are still being evaluated:
1.
No local indication or alarn signals are provided for certain flow, pressure drop, temperature, and status sensors; and 2.
No recorded indication or alarm signals are provided at the control room for certain pressure drop and temperature sensors.
Upon resolution of itens (1) and (2) above, the staff will estiablish final verification of those of its accident evaluations involving use of ESF filter efficiencies.
Upon closure of the above items, we will supplement our SER as appropriate. Also, enclosed is a SALP evaluation (Enclosure 2). This review was coordinated and contributed to by Hillard Wohl (X27065).
Frank Akstulewicz, Larry Bell and Ken Dempsey also contributed to the review.
Odginal signed by N
Daniel R. f.iu!'er.
Daniel R. Muller, Assistant Director for Radiation Protection Division of Systens Integration
Enclosures:
As stated cc: ' R. Bernero L. Rubenstein E. Adensam O. Parr B. Sheron.
W. Gamill H. Miller R. W. Ho'uston W. Regan
- 11. Wigdor
./
,c.),DSI:AEB DSI:AEg DSI:AEB @
DSI me).M. Wh....,..,k..........
....W..P..a.d. 8 ".......
.....L..G.H..u..l.m..a.n.....
....DM m) 9/...JJ /.8. 4
.. 9 /............/.8. 4
.. 9../...)..,.$...../. 8 4
... 9/.g..../.84 ionu se no,somacu ouo OFFICIAL RECORD COPY
- us no in3-4 2 7
.. - - c # -- -
-~,
ng,w.
m m w-w
~ = -- -
~ ~
._.u.
..r ENCLOSURE 1 DRAFT SER INPUT FROM THE ACCIDENT EVALUATION BRANCH FOR V0GTLE UNITS 1 AND 2 15.4 Radioloaical Consecuences of Accidents To evaluate the effectiveness of the ESFs proposed for the Vogtle plant and to ensure that the distances to the exclusion area boundary (EAB) and low popula-tion zone (LPZ) are adequate, the applicant and the staff have analyzed the radiological consequences of a number of accidents, selected to challenge specific design features of the plant. The severity of the accidents analyzed range from minor leakage of coolant to a hypothetical release to the contain-ment of a fraction of the core's fission products associated with substantial core melting, as required by 10 CFR 100.11. The calculated doses for these accidents are'given in Table 15.1.
15.4.1 Loss-of-Coolant Accident The limiting fault postulated as the design basis for the containment and its associated engineered safety features, and as a demonstration of the adequacy of the distances to the exclusion area boundary and the low population zone (LPZ) is a loss-of-coolant accident (LOCA) in conjunction with the release of a substantial fraction of the fission product inventory of the core, as set forth in 10 CFR 100.11(a). The analysis has included the sources and radio-activity transport assumptions specified in Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors," as well as additional guidance contained in the Standard Review Plan.
This-postulated event involves,the assumed availability for release from the c:ntainment atmosphere of 100% of the core's inventory of noble gases, and 25%
09/26/84 15-1 V0GTLE SER SEC 15
.. e y-
.,,n.-
. ~
of the iodine inventory. Although the containment is assumed to be intact, two pathways for slow leakage of these fission products to the environment were identified:
(1) leakage from containment (2) leakage from ESF systems outside containment The contribution of these two leakage pathways to the calculated offsite doses are discussed below.
15.4.1.1 Containment Leakage Pathway The safety features of the Vogtle plant include a containment designed to minimize the leakage of fission products from postulated accidents involving the failure of the first two barrlers against a release of fission products, i.e. the fuel cladding and primary pressure boundary.
The containment consists of a post-tensioned concrete primary containment vessel with a carbon steel i f ner. Another engineered safety feature (ESF) is the containment spray system with an Na0H additive to achieve a slightly basic pH in the water collecting in the containment sump following a LOCA. The staff's calculation of the consequences of the hypothetical LOCA used the conservative assumptions of Positions C.1.a through C.I.e of Regulatory Guide 1.4, Revision 2, "Assump-tions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors." The primary containment was assumed to leak at a rate of 0.2 percent per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 0.1 percent per day after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The analysis took into account radiological decay during holdup in the containment, mixing in the containment, iodine decontamination by the ESF spray system, and conservative estimates of disper-sion of the fission products in the environment.
A list of assumptions used in the calculation of the LOCA doses is given in Table 15.2.
15.4.1.2 Post-LOCA Leakage from ESF System Outside Containment During the recirculation mode of operation of the ECCS following a LOCA, the sump water is circulated outside containment to the auxiliary building for 09/26/84 15-2 V0GTLE SER SEC 15
y,.
m.
m y,
,y.,, _ _
y,_ _.r
.a..
./
cooling and re-injection via the ECCS or containment spray system.
Normal leakage of equipment in this fluid system, or malfunctions,'such as a pump seal failure, would result in an airborne release of any volatile forms of fission products carried by the sump water to the auxiliary building.
For Vogtle, the ECCS area in the auxiliary building is served by an ESF air filtra-tion system (the auxiliary building exhaust system) which would be expected to collect and process such releases through filters. Therefore, doses resulting from passive failure of equipment carrying this fluid were not explicitly considered (as specified in SRP Section 15.6.5, Appendix B) in our calculations.
In FSAR Table 15.6.5-4, the applicant has identified a value of 50 gal / min as the maximum amount of leakage from ECCS equipment following an accident.
Following the Standard Review Plan, the staff evaluated the potential radio-logical consequences from this release pathway assuming a leakage rate of twice the applicant's value. The resultant radiological consequence estimates were 71 rems to the thyroid at the exclusion area boundary and 148 rems to the thyroid at the low population zone (LPZ). The staff considers 50 gal / min to be a larger than necessary leakage value, which could be substantially reduced.
The applicant has committed to implement a program to minimize such leakage in accordance with the TMI action plan requirements.
15.4.1.3 Conclusions The staff's calculated thyroid and whole-body doses from the hypothetical LOCA are given in Table 15.2..The staff concludes that the distances to the exclu-sion area and to the LPZ boundaries of the Vogtle site, in conjunction with the ESFs of the Vogtle plant design, are sufficient to provide reasonable assurance that the total radiological consequences of a postulated LOCA will be within the exposure guidelines set forth in 10 CFR 100.11.
This conclusion is based on the staff review of the applicant's analyses and on the independent analysis performed to verify that the total calculated doses are within the guidelines.
09/25/84 15-3 V0GTLE SER SEC 15
y.
~ ~.
m.v
~ -
_c
~
n.
15.4.2 Main Steamline Break Outside Containment Both the staff and the applicant have evaluated the radiological consequences of a postulated steamline break accident occurring outside containment and upstream of the main steam isolation valve. Although the contents of the secondary side of the affected steam generator would be vented initially to the atmosphere as an elevated release, the staff has conservatively assumed that the entire release throughout the course of the accident occurs under ground level conditions. During the course of the accident, the shell side of the affected steam generator was assumed to stay dry since auxiliary feedwater flow to the affected steam generator would be blocked off under the conditions of this accident.
Because of the dryout condition in the affected steam generator, all iodine transported to the secondary side by leakage (1 gpm) was assumed available for release to the atmosphere with no reduction from holdup or attenuation.
.( l -
The staff investigated three scenarios.
For Case 1, the most reactive control rod was assumed to be stuck in the fully withdrawn position. The applicant has indicated, and the staff agre'es, that no departure from nucleate boiling is expected to occur and, therefore, no fuel cladding failure was assumed in the calculation. With no additional fuel failures occurring, Case 1 becomes identical to Case 2, and no radiological consequences are presented for Case 1.
For Case 2, the staff assumed that a temporary increase in the coolant iodine concentration (iodine spike) occurred as a result of the power and pressure transient caused by the accident. Before the accident, the plant was assumed to be operating at the Westinghouse Standard Technical Specification equili-brium primary coolant limit of 1 pCi/gm dose equivalent fodine-131 (DEI-131).
The iodine spike generated during the accident is assumed to increase the release rate of iodine from the fuel by a factor of 500.
This increase in the release rate results in an increasing iodine concentration in the primary I
coolant during the course of the accident. The radiological consequences for this case have been calculated using assumptions given in Table 15.3 and the
~
consequence values are given in Table 15.1 of this SER.
09/26/84 15-4 V0GTLE SER SEC 15 1
7pm 7=+-
- r~
3 For Case 3, the staff assumed that previous reactor operation had resulted in a primary coolant concentration equal to the maximum transient full power Westinghouse Standard Technical Specification limit (60 pCi/gm DEI-131)
As in Case 2, the radiological consequences were calculated using assumptions found in Table 15.3 and the consequence values are given in Table 15.1.
Based on its findings, the staff concludes that there is reasonable assurance that the calculated radiological consequences of a postulated main steamline failure outside the containment of the Vogtle plant can be controlled by technical specifications to remain within (1) the dose guidelines of 10 CFR 100.11 for the case that the failure occurs with a primary coolant iodine concentration corresponding to a preaccident iodine spike; and (2) 10 percent of these guidelines for the case that the failure occurs with a primary coolant activity corresponding to the equilibrium concentration of the Westinghouse Standard Technical Specifications..
These conclusions are based on (1) our review of the plant design and the applicant's analysis of this postulated accident, (2) our independent calcula-tion using appropriately conservative assumptions (including atmospheric diffusion factors as discussed in Section 2 of this report), and (3) the specific Technical Specifications for the iodine concentration in the reactor coolant (which consistr, of a maximum allowable limit and a limit for the equilibrium concentration for continued plant operation) and the limit on primary-to-secondary leakage in the steam generators. The staff will review the Vogtle plant's technical specifications to ensure that these cperating restraints are incorporated.
15.4.3 Steam Generator Tube Failure The applicant has provided an analysis of the systems response and radiological consequences of a steam generator tube rupture (SGTR) accident. The staff has requested justification for the asserted ability to isolate the affected steam generator within 30 minutes and for the assumed mitigative capability of systems to reduce the radiological consequences of the accident.
In response to the staff request, the applicant states that the Westinghouse Owners' Group is investigating several SGTR licensing concerns and will address the staff's 09/25/84 15-5 V0GTLE SER SEC 15
_J
~~n_.__.
~.m.
m _ m___
.__.n
.~.
[
concerns through a generic resolution in late 1984.
Upon receipt of this additional information, the staff will complete its review of the SGTR event and the radiological consequences thereof.
The staff's review and conclusions will be reported in a future supplement to this SER.
15.4.4 Control Red Ejection Accident A nonsechanistic rupture of a control rod drive housing is postulated.
Ber w cf the resultant opening in the pressure vessel, primary coolant is lost a the containment with concurrent rapid depressurization of the reactor pressure vessel.
Reactor trip, initiated by one of several trip signals, is assumed to occur rapidly.
Ejection of a control rod results in rapid reactivity insertion. The applicant has conservatively assumed that 10 percent of the fuel elements will experience cladding failure, releasing the volatile fission products in the fuel-cladding gap.
In addition, the applicant stated that 0.25 percent of the fuel rods may experience fuel melting. The fission products released as a result of this damage to the fuel are assumed to be released with the primary coolant.
The release to the environment may occur by either of two pathways. The first pathway involves a release of coolant carrying fission products to the primary containment, which is then assumed to leak to the atmosphere at the design leak rate of the containment (0.2 percent per day).
In the second pathway, activity would reach the secondary coolant via steam generator tube leaks. A maximum of 1 gpm primary-to-secondary leak rate is assumed (as limited by technical specifications). With loss of offsite power (assumed to occur as a result of reactor / turbine trip) and subsequent steam venting, some of the iodine transferred to the shell side is available for leakage to the environment.
In considering the consequences of this postulated event, the staff calculated the doses from the activity available for release separately for each of the above pathways.
The staff would expect the actual consequences to be some combination of these pathways. ' The assumptions used in calculating the radio-logical consequences are presented in Table 15.4, and the resultant doses for cach pathway are given in Table 15.1 of this SER.
09/26/84 15-6 V0GTLE SER SEC 15
- 7 -~ -~
~ ~ - - -
. w. -.
The staff has reviewed the applicant's analysis of the radiological consequences following a postulated control rod ejection accident. The staff concludes
~
that the proposed operation of the Vogtle plant within the limits of the technical specifications assumed above provides reasonable assurance that the calculated radiological consequences are well within (less than 25 percent) the dose guidelines of 10 CFR 100.11.
The staff's conclusion is based on (1) review of the applicant's analysis of the radiological consequences, (2) independent dose calculations using the recommendations of Appendix 8 of Regulatory Guide 1.77 and the atmospheric dispersion factors as diccussed in Section 2 of this report, and (3) the Westinghouse Standard Technical Specifications for the primary-to-secondary leakage in the steam generators.
15.4.5 Fuel-Handling Accident B
In the evaluation of the fuel-handling accident, the methodology used by the staff is cased on Positions C.I.a through C.1.f of Regulatory Guide 1.25 and j
SRP Section 15.7.4.
The staff assumed that a single fuel assembly is dropped into the fuel pool during refueling operations and that all.of the fuel rods in the assembly were damaged, releasing radioactive materials in the fuel gaps f
into the pool.
i 5
For the case of a fuel handling accident in the fuel building, the applicant estimates the time for the radioactive materials that escape from the pool to travel from the detector to the isolation damper to be slightly over 0.6 second. The closure time of the isolation dampers in the normal exhaust system following receipt of an isolation signal is estimated at 6 seconds.
Because the travel time is less than'the isolation time by over 5 seconds, the i
staff assumed that the entire activity release escapes from the fuel building without reduction by the ESF filters.
{
In the case of a fuel handling accident occurring inside containment, 25 percent of tha containment volume was assumed for mixing with a containment isolation time of 15 seconds.
No filtration credit is assumed for activity that escapes prior to containment isolation.
09/25/84 15-7 V0GTLE SER SEC 15
w;g.w..n.-, m _ _ _ - m _v n,.-.
.r The estimated offsite doses for the postulated fuel handling accidents inside containment and inside fuel building are shown in Table 15.1.
The list of j
assumptions and parameters used in the analysis are given in Table 15.5 a/b.
The potential doses for the fuel handling accidents are well within the guide-l line value given in 10 CFR 100. Therefore, the staff concludes that the applicant has provided a design that meets GDC 61.
i j
With regard to a potential spent fuel cask drop accident, a Type 1 single f
failure proof crane designed according to NUREG-0554 and Branch Technical Position APCSB-9-1 is used in handling the cask, which is prevented, by inter-locks, from being moved over the fuel. With these safety measures, the staff concludes that the likelihood of a spent fuel cask drop accident js suffi-j ciently small that no radiological consequence analysis is required.
15.4.6 Failure of a Small Line Carrying Primary Coolant Outside Containment
~
There are a number of small lines carrying primary coolant outside the j
containment. The applicant has provided an analysis of an accidental break in j
the CVCS ietdown line outside containment, but downstream of the containment isolation valves as a worst-case failure of one of these lines.
This break 4
j would release 194 gpm of primary coolant to the auxiliary building before isolation could be expected.
The break would cause a low level in the volume control tank, and the operator could diagnose the break and shut the appropriate j
isolation valve to isolate the leak.
The staff has performed an independent assessment of the potential dose conse-quences of the release of primary coolant outside the containment. The staff l
assumed that 20 min will elapse before operator action to isolate a CVCS line break in response to receipt of the low-level signal. Thus, a total of 3880 gal sf primary coolant could be released. The staff estimated that 39 percent of the hot reactor coolant would flash into steam upon entering the auxiliary building atmosphere, and assumed that a proportional f' action of the iodine r
c.. solved in the coojant would become airborne in gaseous or particulate form.
IntheabsenceofESFsdesignedtodetectandmitigatetheconsequencesof such a release, the staff assuaed that this airborne iodine can escape directly j
to the environment at ground level, without delay or effective filtration.
l Other assumptions are given in Table 15.6.
09/25/84 15-8 V0GTLE SER SEC 15
. J
~r~T.2:.: '. K.T ' :-
?Y ^ "Y:LTXX 1
./
The staff concludes that the consequences of a postulated small line failure outside the containment, assuming the primary coolant equilibrium iodine concentration permitted by the Standard Technical Specifications, in combination with an accident generated iodine spike, do not exceed a small fraction of the exposure guidelines of 10 CFR 100.11. The results of,the staff's calculations are given in Table 15.1.
The staff's conclusion is based on (1) review of the applicant's classification and identification of small lines in accordance with GDC 55, " Reactor Coolant Pressure Boundary Penetrating Containment," and Regulatory Guide 1.11. "Instru-i ment Lines Penetrating Containment"; (2) review of the applicant's analysis of radiological consequences of a failure in the CVCS line; (3) independent dose calculations using Regulatory Position C.I.b. of Regulatory Guide 1.11 and i
conservative atmospheric dispersion factors as discussad in Section 2 of this report; and (4) the Westinghouse, Standard Technical Specifications for the
(
equilibrium iodine concentrations in the primary coolant system. The staff will review the Vogtle specific Technical Specifications to ensure that the coolant activity limits assumed abcVe are not exceeded.
6.4 Control Room Habitability i
The requirements for the protection of the control room personnel under accident conditions are specified in General Design Criterion 19.
The applicant proposes to meet these requirements by incorporating shielding, emergency heating, ventilating and air conditioning (HVAC) systems, and self-contained breathing apparatus in the control room habitability design. The habitability systems I
also provide storage for food and water, sanitary facilities, and fire protec-I tion that includes a remote shutdown capability.
}
The design of the control room habitability systems relative to the following
\\
areas is discussed in separate SER sections as indicated:
1 Explosion, fire and toxic gas in vicinity of plant - Sections 2.2.1-2,2.3; a.
f 09/25/84 15-9 V0GTLE SER SEC 15 l
i
n.
- n,,w b.
Protection from wind and tornado effects - Section 3.3; c.
Flood Design - Section 3.4; d.
Missile protection - Section 3.5; Protection against dynamic effects associated with postulated ruptures of e.
piping - Section 3.6; f.
Environmental qualification of equipment - Section 3.11; g.
Filter efficiencies - Section 6.5.1; h.
Radiation protection aspect of GDC 19 and NUREG-0737, II.B.2; shielding; TSC - Section 12.3; 1.
HVAC systems analysis - Section 9.4.1 (includes seismic review);
j.
Fire protection and remote shutdown capability - Section 9.5.1; and k.
Human engineering, control room environment, and communications -
Section 18.
The staff evaluation indicates an inconsistency in the applicant's estimate of the control room leak rate.
Relative to toxic gas protection, the applicant states that the air leakage is no greater than 185 fts/ min from all pathways based on 1/8 inch W.G. pressure differential.
This conflicts with the estimated 1500 ft8/ min air intake at the same pressure differential assumed in the evaluation of radiation doses to control room personnel following design basis accidents. Until this matter is resolved, habitability of the control room following radiation and toxic gas release accidents is an open item.
In additiow, information is needad from the applicant in two areas:
(1) Response to Question 450.3 on the data used to estimate the control room dose following a LOCA.
~
09/25/84 15-10 V0GTLE SER SEC 15 4
. mab h
a A
i (2) Toxic gas evaluation for the chemicals listed in Table 7.2.3-18.
The evaluation should include data described in Table C-3 of RG 1.78.
Based upon the foregoing, the staff concludes that the applicant has not demonstrated that the control room habitability systems will adequately protect the control room operators in accordance with the requirements of 10 CFR Part 50, Appendix A, General Design Criterion 19 and, therefore, compliance with NUREG-0737, Item III.D.3.4 cannot be established.
0 e
9 O
e 0
0 09/25/84 15-11 V0GTLE SER SEC 15
w p m w n m,= q _
-m--
- - ~ - - - - - - - ~ - - ~ - ' ~ ~
p,mn mm&
-_~
, = = + --
+.
T.
N g
Table 15.1 Radiological consequences of design-basis accidents s,
e Exclusion area Low population boundary dose, rems
- zone dose, rems *
.,l
', ~
Postulated accident Thyroid Whole Body Thyroid zWhcir Body,
Loss of coolanti' Containment leakage 0-2 hr/'-
A 98 2.6 1
0-8 hr 33
' O. 9 8-24 hr '
12 0.2 24-96 hr
.i.
10 0.1 96-720 hr
_j 0.1 Total containment leakage 98 2.6
' 64 1.3 t
ECCS component leakage 71
- 0. 2 148-0.2 Total 169 2.8 212 1.5 Steamline break outside containment:
Long-term operation > case
> J, N
(DEI-131 at'l pCi/gm)**
2.7
<1.0
- 1. 6 1
'<1.0 Short-term operation case (DEI-131 at 60 pCf/gm) 3.1
<1.0 1.5
<1.0 Control rod ejection:'
/
Containment leakage pathway 26
<1. 0 43
<1.0 Secondary. system release pathway 9.7
<1.0 1.8
<0.1 Fuel-handling accident 3
in fuel-handling area 53
- 0. 7 8
~
0.2 Fuel-handling accident N inside containment 0.3
<0.1 0.1
<0.1
~
' Small line break 4.6
<0.1 0.8
< 0.1 "The short-term diffusion estimate (X/Q's) used in the analysis are those presented and discussed in SER Section 2.3.4.
The meteorological models described in regulatory guides referenced in these analyses are modified by l'
those presented in Regulatory Guide 1.145.
See Section 2.3.4 for further discussiori of the meteorological models.
C* del-131 is 'the dose-equivalent iodine-131 concentration, as defined in the Standard Technical Specifications.
\\
i 09/25/84 15-12 V0GTLE SER SEC 15 sI
- v_,,.a
_ _- %_2.,
..o
- =
-3 Table 15.2 Assumptions used in the calculation of loss-of-coolant accident doses Parameter and unit of measure Quantity Containeer:t leakage contribution Power level MWt 3,565 Operating time, yr 3
Fraction of core inventory available for containment leakage, %
Iodine 25 Noble gases 100 Initial iodine composition in containment atomsphere, %
Elemental 91 Organic u4 Particulate 5
Containment leak rate, %/ day 0-24 hr 0.2 After 24 hr 0.1 Containment volume, ft3
- Sprayed volume 2.15 x 108 Unsprayed volume 6.05 x 10s Containment mixing rate from cooling fan operation, cfm 174,000 Containment spray system Maximum elemental iodine decontamination factor 100 Spray removal coefficients, hr 1 Elemental iodine 10 Particular iodine-0.45 Organic iodine O
Relative concentration values, sec/m3*
0-2 hr at' the exclusion area boundary (EA8) 1.8 x 10 4 0-8 hr at the low population zone (LPZ) boundary 3.1 x 10 s 8-24 hr at the LPZ boundary 2.2 x 10 s 24-96 hr at the LPZ boundary 1.0 x 10 5 96-720 hr at the LPZ boundary 3.4 x 10 s ECCS leakage outside containment Power, MWt 3,565 Sump volume, gal 905,080 Frac' tion of iodine assumed volatilized and/or airborne as aerosol 0.1 l.
09/25/84 15-13 V0GTLE SER SEC 15
,.... -- a mmen.=w= = -m-w =. :c -.
- :.:--=
- a -- -
9 Table 15.2 (continued)
Parameter and unit of measure Quantity ECCS leakane outside containment (continued)
Leak rate, gph (twice the maximum operational. leakage defined in FSAR Table 15.6.5-4) 6000 Leak duration, h'r 720 Delay time, hr 0.50 Filter efficiency for iodine, %
Elemental and particulate 99 Organic iodine 99
- Used for all accidents.
..t..
e e
O I
09/25/84 15-14 V0GTLE SER SEC 15
.. --m nn + -
m;
-- e ~ ~ v
.s
~
Table 15.3 Assumptions used to evaluate the radiological consequences following a postulated main steamline break accident outside containment Power, MWt 3565 Preaccident dose equivalent I-131 in primary coolant, pCi/gm 1.0 (Case 2)*
Preaccident dose equivalent I-131 in primary coolant, pCi/gm 60.0 (Case 3)**
Primary-to-secondary leak rate, as limited by Technical Specifications, gpm 1.0 Leakage in the affected steam generator, gpm 1.0 Fraction of iodine entering shell side of the steam generator released to the environment 1.0 Ratio of iodine release rate from fuel during iodine spike to release rate from fuel during steady state operation 500
- Long-term operation case
- Short-term operation case e
e 09/26/84'.
15-15 V0GTLE SER SEC 15
. ' ::: L*:TT=-Y..-. ^ <-t. T-
-^ --... -.
2:..
. " ^ ~ -
-.2 '
~
Table 15.4 Assumptions used for estimating the radiological consequences following a postulated control rod ejection accident Power, MWt 3565 Primary-to-secondary leak rate, gpm 1.0 Fraction of th.e fuel rods' experience cladding failure 0.1 Fraction of noble gas and iodine inventory in gap of failed rods 0.1 Fraction of the fuel rods experiencing fuel melting 0.0025 Fraction of fodine inventory released from rods experiencing melting 0.5 Fractio' n of iodine entering S.G. secondary side released to environs 0.1 Time of primary and secondary system pressures equalization, sec. 3300 Fraction of iodine plated out in containment 0.5 Design leak rate of containment, % per day 0.2 Iodine concentration (DEI-131) in the secondary coolant, pCi/gm 0.1 09/26/84 15-16 V0GTLE SER SEC 15
.... e..
- -. _ u u a-m.. c-x
- m. s
=
~~-.-...n
. - ~...,. ~.. _ -
. -.. -- - ~..-.+ m
- -.. ~ - - - -. -...
?
Table 15.5a Assumptions used for estimating the radiological consequences following a postulated fuel handling accident in fuel-handling area Parameter and unit of measure Quantity Power level, MWt 3,565 Number of fuel rods damaged 264 Total number of fuel rods in core 50,952 Radial peaking factor of damaged rod 1.65 Shutdown time, hour 100 Inventory released from damaged rods (iodines and noble gases), %
10 Pool decontamination factors Iodine 100 Noble gases 1
Iodine forms in atmosphera above pool, %
Elemental 75 Organic 25 Iodine removal efficiencies for ABGTS (spent fuel pool area), %
Elemental no filters Organic assumed 09/26/84 15-17 V0GTLE SER SEC 15
.. w,- _,
-,- - ~~-;_ n-~m~n m sen w.w n s
y Table 15.5b Assumptions used for estimating the radiological consequences following a postulated fuel handling accident inside containment
- Parameter and unit of measure quantity Power level, Wt 3,565 Number of fuel rods damaged 264 Total number of fuel rods in core 50,952 Radial peaking factor of damaged rod 1.65 Shutdown time, hour 100 Inventory released from damaged rods (iodines and noble gases), %
10 Reactor water cover decontamination factors Iodine Noble gases 100 1
Iodine fractions released f. rom re' actor water cover, %
. Elemental Organic 75 25 Iodine removal efficiencies for containment effluent, %
Elemental Organic no filters assumed
- 25% con'tainment free volume mixing, 15,000 CFM flow rate for 15 sec. assumed.
e O
09/26/84 15-18 V0GTLE SER SEC 15
ww--, n = ~- -
mwr
-?
,.;, 7 _
.n.
r Table 15.6 Assumptions used in accidents involving small line breaks outside the containment Parameter and unit of measure Quantity Coolant released, Ib 23,500 Fraction of coolant released flashed to steam, %
39 Coolant contaminant concentration, pCi/gm 1.0
' Spiking factor (iodine release rate multiplier) 500 e
0 09/25/84 15-19 V0GTLE SER SEC 15
- - ~ ~
~
~
W ^.. T JU.T.iC.:_?. :;.?"
.f.1_.. _ liff. C'fi-f:1..
I'" ". " 1. _.
l
~~
c ENCLOSURE 2 SALP INPUT FOR V0GTLE DRAFT SER Plant: V0GTLE ELECTRIC GENERATING PLANT, UNITS 1 and 2 A.
Functional Areas:
Licensing Activities 1.
Management involvement in assuring quality.
N/A 2.
Approach to resolution of technical issues from a safety standpoint.
Rating: Category,2 3.
Responsive to NRC initiatives.
See 2 above Rating:
Category 2, 4.
Staffing (including Management)
N/A 5.
Reporting and analysis of reportable events.
N/A 6.
Training and qualification effectiveness.
N/A
' 7.
Overall Rating for Licensing Activity Functional Area:
2 t