ML20203N064
| ML20203N064 | |
| Person / Time | |
|---|---|
| Site: | Seabrook, 05000000 |
| Issue date: | 08/19/1986 |
| From: | Lyon W Office of Nuclear Reactor Regulation |
| To: | Berlinger C Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20198G688 | List:
|
| References | |
| NUDOCS 8609230123 | |
| Download: ML20203N064 (36) | |
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I MEMORANDUM TOR: Carl Ber!!nger, Chief Reactor Systems Branch Division of PVR Licensing-A THROUCN: Richard Lobel, Section Leader Reactor Systems Branch Division of PVR Licensing-A FROM: Varren C. Lyon, Sr. Nuclear Engineer Reactor Systems Branch Division of PVR Licensing-A
SUBJECT:
REVIEV 0F SEABROOK DOCUMENTS PERTAINING TO CNANCE IN EMERCENCY PLANNINC ZONE SIZE
REFERENCES:
- 1. "Seabrook Station Risk Management and Emergency -
Planning Study", Pickard, Lowe and Carrick, Inc.,
PLC-0432, December 1985.
. 2. "Seabrook Station Energency Planning Sensitivity
- Study", Pickard, Lowe and Carrick, Inc., PLC-04&5, Apr!! 1986.
I have read the reference documents and have complied a number of observations and questions which may be useful in our continuing review. These are documented in Enclosures 1 and 2. I suggest these be transmitted to BNL for their consideration during the formal review.
I find this work to represent a considered investigation of Seabrook Station response to severe accident conditions. It is a logical estension of the Probabi!!stic Risk Assessment work which I helped to review, and it describes the.most detailed investigation of LOCA outside of containment that I have encountered. In general, I believe the work to be we!! founded, although there are exceptions which are identified below. It is not clear if the exceptions would seriously perturb the overall conclusions, and this is an area which should be investigated. If the exceptions are overlooked (for the moment), the reports present reasonable conclusions in the areas in which I am qualified to formulate a professional opinion.
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1 My perception is that we are investigating this area from the viewpoint of a l comparison between what was done several' years ago and what we know now, with the objective of reaching a conclusion with respect to the rationale used in
' formulating the entsting energency planning criteria. A concern is that this approach may result in the esclusion of. items that were not included by the original planners, but which impact risk in areas close to the plant when the overly conservative original basis is corrected.
Clearly, Seabrook Station has a large volume, strong containment which will mitigate almost any realistically postulated severe accident. (1 do not believe the staff should expend extensive resources investigating this, although we should satisfy ourselves of the correctness of the conclusion.)
Risk problems wit! not be found by studying the containment. Nor will they be found by rehashing chemical reaction phenomena, vaporisation of fission
. products, and the like.,,The problta with respect to plant response, if there is one, will be in een'tainannt g
bypass nd it is here that I believe we should concentrate our resouseas 4- also here that I believe the Pickard. Lowe
/ and Carrick (PLC) work is potentia!!y deficient. They have not fully explored ways to bypass containment. Examples which are not obvious which I believe to be valid are:
- 1. Steam Generator 1111 Ight Ruotare. This is the rupture of multiple tubes in response to high temperature which in turn is a result of core uncovery. This accident sequence should be of concern-any time there is a core melt with the Reactor Coolant System at more than a few hundred psi pressure, with no water in the SC secondary side. These conditions lead to a potential for natural circulation transport phenomena to significantly heat the tubes prior to breach of the reactor vessel. The resulting loss of tube strength can lead to tube rupture. Reactor Coolant Pump operation, as outlined in many plant emergency procedures, almost assures this to be a concern. If tube rupture occurs, and any of the secondary side valves are open, the secondary side is breached outside containment, or the reactor coolant system pressure is above the SC relief valve setpoints, containment is bypassed. This has_not been adequately investigated, and is not recognized as a release path in the PQE.
- 2. LQ(& Outside Containment. PLC has done seme of the most innovative work I have seen in this area. In many respects, it is excellent. But they didn't ce far enough. Common cause failures appear to be weakly investigated. Some failure paths ar, ignored (were they investigated and found negligible?). I also question the data base and its application. ,
- 3. Other Vhat have we missed? I believe a careful investigation is in i order to assess if there are uniden_tified_ bypass paths. This will be a f difficult task to accomplish properly since one is looking for the
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My final concern with the reported work, and the risk assessment approach, is that we are ignoring the potential risk impact of sabotaan. This is just as real as an earthquake or any of the equipment failures which can initiate a severe accident, and it will impact risk. I do not believe the rational of limiting the comparison to topics used by the formulators of the old emergency e
planning is sufficient when one is potentially considering one or two mile
- radius sones. We are already violating this rational with the large number of
- potential accident paths considered in the Seabrook PRA and with Jelsmic considerations. Of importance here, there are i t _e a s that were_ negligible with t2La old, higM y consy v at,ive assumptions that b e c o,m e s i g rt i f,t.c a n t, w i,,t h , r eno,v,a l of conservatism M ated y Jn r e_c e n t_k n,ow l e d g e . These may impact risk assessments when a reduced zone size is considered. Steam generator tube rupture due to overheating is one such item. Sabotage is another.
W Varren C. Lyon Senior Nuclear Engineer Reactor Systems Branch, DPL-A
Enclosures:
As Stated cc: T. Novak C. Rossi B. Sharon R. Ballard V. Benaroya J. Milhoan C. Bagchi B. Doolittle V. Nerses i
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, e-ENCLOSURE 1 REVIEV COMMENTS ON ,
"SEABROOK STATION RISK MANAGEMENT AND EMERGENCY PLANNING STUDY" l (PICKARD, LOVE AND CARRICK, INC., PLC-0432, DECEMBER 1985) i SECTION 3 Page 2 The assumption is made that a loss of instrument air will allow the purge valves to close prior to core damage or uncovery. What is the justification for this assumption? (How long does it take for the instrument air pressure to decrease to the point that that valves close?) I What likelihood is assigned to mechanical failure which results in the purge valves falling to close?
What is the likelihood that human error, such as failure to properly close the personnel hatch, will provide containment bypass?
Deliberate attempts to create a release have not been addressed. Obvious reasons for this omission are the difficulty in assigning a likelihood of occurrence and the need not to publicly draw a map for potential saboteurs. However, this is a real consideration, and one which the I believe should be addressed before reaching a favorable finding on any request for a decrease in the energency planning sone radius.
Operation at conditions other than full power and accident conditions while at shutdown have not been addressed. Although accidents under these conditions are generally considered to represent less risk than accidents at full power, I believe they should be considered before one reaches a favorable finding in regard to decrease of t' h e energency planning zone radius.
As far as I know, no PRA has considered rupture of steam generator tubes during the approach to or the progression cf core melt accidents. This is of concern due to the high temperatures in the reactor vessel, the possib111ty that the approach to melt occurs at high reactor coolant system (RCS) pressure, and the presence of mechanisms which may exist for 4
e transporting hot fluid into the steam generator tubes, thus significantly reducing their strength. (Mechanisms of concern involve both natural circulation and the use of reactor coolant pumps as a "last ditch" effort to present or delay core melt.) I believe we should consider such containment bypass paths before reaching a favorable finding in regard to decrease of the emergency planning sone radius.
I consider it necessary that all reasonable release paths be considered either directly or via a suitable allowance (for unknowns) in order for PRA associated work to be used as part of the basis for a reduction in the radius of the emergency planning sone. In !!ght of the above, this has not been accomplished. Therefore, I recommend a careful evaluation ~
of possible containment bypass paths and a search for areas which have
, not been considered in the FRA associated work that is being used as a ,
basis for a potential request to reduce the radius of the planning zone.
5 Reference is made to Fig. 4-8. This figure shows the Baron Injection Tank (BIT) as part of the Engineered Safety Features (EST) System. The BIT is no longer a part of the Seabrook EST. This should be corrected.
In general, we should be assured that the information used in the report is up to date.
A minimal number of paths are presented as leading to pressurization of the low pressure portions of the RHR system. Figure 4-8 shows many more l
potential paths. I suggest that all potential paths be listed, and that we be provided with the reason for the refection of each (such as a likelihood of the path being open and the contributors to that likelihood). Consideration should be given to operating procedures, the
!!kelihood that the procedures will not be followed, p'ossible errors (such as the recently discovered difficulty with insufficient head to supply the 51 and charging pumps in the recirculation mode and the impact upon LOCA outside of containment, if any), and interlock behavior.
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Other risk contributors that are often overlooked in PRAs are design and construction errors. The prior comment in regard to insufficient head is a good example. Please address how these are contained in the data base which is being provided to the staff to justify the decrease in emergency planning sone radius.
5 Seabrook has recently proposed changes in the operating procedure for initiation of the recirculation mode of 51 operation. Do these changes have any influence on the probability of satisfactory switchover from the injection to the recirculation mode?
4 The TSAR gives the relief rate as 900 gpa with a set pressure of 450 psi. -
The flow rate does not agree with the value used here. Please explain.
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What is the mechanism for assuring that plant changes and new knowledge are promptly factored into the technical considerations which form a part of the foundation for staff consideration of a reduced emergency planning sono radius?
I note that FSAR Figure 4.3-2 appears to be inconsistent between the actual figure and the tout which describes valve positions. This also impacts upon proper operation of the SI system. See, for example, Valves 14, 21, and 22. Is this correct? If so, what are the implications with respect to the issue under consideration here?
7 What is a sump low level alarm? (Fifth paragraph) 7 What consideration has been given to relatively small breaks and interaction with the fusible links in the ventilation'systen? (Is the potential impact worth consideration?) The concern is that enough hot water may be released to activate the links, thereby terminating ventilation and indirectly causing failure of the pumps due to overheating of the pump motors, and that this could occur at a time earlier than might occur due to flooding.
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t I do not understand the conclusion that presence of water in the reactor cavity will decrease (significantly?) the revaporisation of' fission products from RCS and perhaps RHR surfaces. I anticipate that a significant quantity of heat producing radioisotopes will remain in the wreckage of the reactor vessel, and this may be effective in heating whatever ga'ses or vapor are flowing toward the break. Has this been investigatedt 10 What is the justification for the statement that the first sign of trouble w!!! be pressuriser low level or low pressure alarmst I anticipate a number of other indicators may be first, such as abnormal indications from the PRT or even a smoke alarm. -
11 There have been a number of indications that there is a good chance of containment spray being actuated due to a high RHR relief valve release ~ ,
rate into containment. What is the justification for this conclusion?
Inetude the effect of containment heat sinks and containment cooler operation in the response.
11 The statement "As soon as the pumps begin to produce flow to the RCS, valves in the miniflow lines close and all RHR pump flow is infected into the reactor vessel via the RHR cold leg injection lines" is not correct.
The sensors are not located at the RCS to detect flow at that location.
Further, one is postulating a break in the RHR system, and a significant portion of the pump flow may never reach the RCS (as is stated in the following paragraph).
I 11 The last paragraph contains a number of timing of event statements.
Please provide justification of each. Plots of plant behavior showing ;
suitable parameters and indicating the event points are sufficient for most. Operator response information, in addition to RC5 parameter j information, is necessary for the statement that RCPs will be tripped within about 21 seconds of break initiation. (I personally observed a simulator run in which RCPs were not tripped for several minutes in a 7
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large break simulation. The operator missed the step, and I didn't say
.. .anything because 1 wanted to see how long it would take before someone caught the mistake.)
12 Ta!!ure to diagnose a LOCA outside containment is identified in the third paragraph. What actions are under way with Seabrook Emergency Procedures to correct.this situation? What notifications have been given by Seabrook to the NRC, Vestinghouse, lHPO, or others in regard to this deficiency? (See also the discussion and recommendations later in the report, such as pages 3-34 and 3-35) 15 This discussion is not clear with respect to what is in the present study ,
and what was not considered in the 55PSA but is considered in the present study. A number of LOCA outside containment events are missing. For esample, the following do not appear:
- a. Inadvertent opening of the two hot leg suction valves due to common cause failure such as improper maintenance, malfunction of the interlock system, design error, or unidentified means.
- b. Failure of the stem or other internal connections in valves equipped with limit switches or fa!!ure of a limit switch (including improper maintenance such as reversing indication).
- c. Malfunction due to fire or other electrical short circuits.
(includes testing operations of all types, such as testing that involves jumpers which could be incorrectly connected to cause a LOCA outside containment. The testing is not necessarily limited to testing of valves.)
- d. Common cause failure associated with improper maintenance such as installation of taproper components (gaskets, seats, or valve disks) which may fall almost immediately or at a later time.
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15 The comment is made "The disc rupture failure mode has not been reported in the nuclear industry data base,." The staff notes a number of disc valve failures have been reported in the nuclear industry for non-SI or
-RNR systems, including some where there were common cause failures.
Vere these considered in the work being reported here? (Note several valve failures at San Onofre Unit i may be too recent.)
15 Are the individual check valves leak tested after each RCS depressurisation er are they checked at once in series? If they are checked individually, pleasa describe how this is accomp!!shed, include checking of the hot leg infection valves in the response. (Note the hot leg injection. path was escluded from consideration because there are -
three valves in series.)
15 Please describe the valve inspections that are promised each time the plant goes to cold shutdown or is refueled. (The imp!! cation is that the valves are opened up for inspection.)
16 The largest leak rate in Figure 3-3 is of the order of 200 gpa, whereas the arena of interest ranges to 45,000 gpa. Please justify this estrapolation. In addition, include consideration of"the valve designs, operation modes, and sizes used in the data base as contrasted .o Seabrook.
14 Several references have been provided to not failing the RHR system at pressures of roughly 2250 psi. What cateulations substantiate these statements or assumptions? What temperatures were involved while the RHR was considered to be esposed to 2250 psi?
17 What likelihood of failing to test after cold shutdown or failure to test after refueling was included in this work? Vhat likelihood was assigned to an incorrectly conducted test that would impact the conclusions?
Vhat is the Itkelihood that a LOCA outside containment can be terminated if initiated? Include consideration of all possible paths and address the dynaale interaction of closing the various valves against water flowing from the RCS at high pressure.
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e 17 In the discussion of the BVR events, would it be accurate to state that none of the events were included in Table 3-8 and Figure 3-3?
17 Inclusion of accumulator check valve leakage is viewed as a conservatism according to the report authors. Vere accumulator check valves also incorporated in the counting to obtain the total number of valves in the RCS and ECCS systenst !! so, how was this a conservatism?
17 There are large differences between check valves. What is th'e impact of this on Seabrook's estimated results?
An item under consideration for advanced nuclear power plants is the
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ability to monitor pressure on the low pressure side of check valves.
This could provide early warning cf check valve leaks and would provide monitoring capability to help assure check valves were operating properly. The same monitoring capability with respect to RHR suction line valves could identify if individual valves were mispositioned or malfunctioning. Would such a system for Seabrook be of significant benefit in reducing risk in a reduced size emergency planning zonet 21 The first MOV in the RHR suction line is identified as not having a stem mounted limit switch. What is the impact of this on plant risk and what would be the cost of adding appropriate instrumentation so that valve position would he indicated?
21 Why is it conservative to assume that MOV valve leakage and failure upon demand due to a sudden pressure loading are the same as for check valves, particularly if the check valves are already closed whem esposed to the sudden pressure loading?
-l l 21 An item that concerns ne is dependent failure of the RHR suction valves.
This is not mentioned. Why?
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f 22 The sp!!t fraction in which the flow is greater than the re!!af capacity is given as 0.09. Please provide justification for this with consideration given to the previously identified concerns regarding RHR section !!ne valve failure modes.
23 This is a cursory list of actions an operator can take to mitigate the accident. ,
23 Vhat is the frequency of failures in the pipe tunnel that led the authors to conclude they are very low?
24 As identified ear!!ar, little thought appears to have been given to the --
real world of altigation. For esample, this discussion of potential
, actions presumes the charging pumps still work. What proceduras exist to turn them off prior to damage.
24 The meaning of the following is not clear: "The second consideration is contingent on the interfacing LOCA being located in the RHR vault at an elevation higher than that required to get significant scrubbing due to flooding. !! such a leak has occurred, the configuration of the vault is such that the leaking primary coolant itself will flood the vault.
External sources for flooding the vault could also be employed." If the LOCA is located high enough that flooding will not provide release nitigation, what does all of this meant t
i 24 Vith respect to modeling all of the important failure modes, I have already commented above.
25 Are the emergency procedures at Seabrook updated as identified in " Top J
Event 01"? See also the middle of page 3-26 for the same issue. l i
b 27 There are a number of cases where the combined sump pump capacity is sufficient to remove leaks and keep the vaults from flooding. In these l cases, the RHR, 51, and CS pumps are assumed not to be lapacted by flooding. No consideration is given to failure of one (or both) sump pumps.
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h 30 "To our knowledge, the IDCOR evaluations discussed above have not been refuted by the NRC." This statement adds little to the report if a knowledgeable reader is involved. There are many things that have not been refuted by HRC which NRC has not reviewed, or which, for whatever reason, nothing has been said.
30 The test indicates a pipe failure probability of about 6 s 10
~' if the low pressure piping is esposed to 2250 ps!. This is based on an assumed probability of fa!!ure at the material yield strength of 0.01 and a
, probability of failure at the ultimate strength of 0.99. As previously identified, temperature is not specified.Is it the 350 F design temperature? Please provide substantiating information. Cover all ,
components which are exposed to a pressure or temperature which is above the design value. This should also include items such as the RHR pump seals and seat material response at RCS temperature.
What is the mariana flow rate that can be infected into the R"P pump seals? (Of potential interest since it may be an alternate path for injection into the RCS.)
35 Shutting an RHR system crosstie valve is identified as an action to help isolate a LOCA outside containment involving the RHRISI systems. Has a careful evaluation of these systems been performed to assess isolation strategy?
35-36 Relative water levels in the RHR vavits and the RCS are mentioned.
What are the water volumes in these regions as a function of elevation?
(Of particular interest is the level at the top of the core and at the elevation of the hot leg connections to the RHR.)
34 Vhat is the justification for the statement that the water level in the vaults will be approximately the same as that in the RCS? (I do not agree.)
37 What is the likelihood that actor operated valves can be used in a flooded RHR vault?
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s 37 "End state DLOC contains sequences in which the interfacing LOCA has been terminated, and the ECCS has been degraded (D) (RHR or 51 pumps have
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failed). ... The point estimate frequency of DLOC is 4.0 s 10 per year.
The additional failures required to achieve core melt would lower this frequence by at least one order of magnitude." What is the justification
,for this conclusion? (Ve have already lost a portion or all of the ability to inject water into the RCS via the usual paths.)
37 (Bottom of page) Why does failure of one' charging pump lead to core melt? The perception is that sufficient flow could be provided by alternate means to keep'the core covered (two other charging pumps, perhaps the reactor makeup water pumps). -
39 " Containment recovery is assumed successful once the containment spray and recirculation functions have been accomp11'shed." Please address t h'a likelihcod that initiation of containment spray could reduce the water content in the containment atmosphere, thereby making possible a deflagration involving hydrogen and osygen. In addition, address the behavior of core melt in the reactor cavity and whether late addition of water prevents containment meltthrough.
45 What is to be the status of the " temporary" 34.5 kV power lines?
44 Vhat is to be the status of the mobile power supplies?
46 Vhat capaht!!ty has been provided to connect etternal pumps as identified in the second and third paragraphs? (The brief mention on page 3-48 may laply that little has been accomplished.) Use of a pump to simply infect water into containment via the sprays on a short term basis (no recirculation) does not appear to be identified. Has this been l considered?
44 This page identifies a number of possibilities. What are the specific plans?
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48 There have been several references to purchase of a mobile electric generator by pooled resources. What is the likelihood that such a generator would be needed by several plants at the same time?
SECTION 4
- Chapter 4 contains little elaboration on the accident sequences which I have identified earlier as receiving inadequate consideration. Indeed, the material in this chapter appears to substantiate my concern that potentially important items have been missed in the technical evaluation I in the Seabrook (PLC) document.
12 A generalisation is drawn to the effect that not-through-the-insulation heat losses are large as compared to the heat generation rate. Thus the conclusion is drawn that ..the primary system heat losses are sufficiently great that the potential for long-term revaporisation within the primary system for a PVR with a large, dry contalnpent is negligib!s. Therefore, this issue does not influence the Seabrook Emergency Planning Study." This is incorrect. Lets start with core melt. If we apply this argument to the core rather than to the upper plenua, we can argue that heat loss from the primary systea is sufficiently great that all generated heat is balanced by heat loss.
Therefore, THE CORE VILL NOT ATTAIN HICH TEMPERATURES AND VILL NOT MELTI (Thus, we don't have to worry about any of these phenomena, and all of the work is unnecessaryl)
The fa!!acy in the argument pertains to the location of energy generation and energy transport. One must be able to transport the heat to the location of loss without attaining a high temperature., This is clearly not the case in the core. It similarly is not the case for the upper ;
plenum, but the situation is not as clear. There are two contributors to the problem:
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- 1. Most calculations of upper plenum behavior involve one dimensional modeling of flow, and any fluid (!! quid, vapor, or gas) that passes through the core is assumed to flow through the upper plenum and out the hot leg. This modeling is incorrect for upper planum behavior under severe accident conditions where a major portion of the core has been uncovered or the core is being vapor or gas cooled. Strong recirculation patterns w!!! develop tahich link the core and upper plenum temperatures. At pressures in the range of normal operation (typically 7250 psi), the linkage is strong, and some of the upper plenus t,*p+sent te9peratures,can be espected to closely fo!!cw core temperature. The strength of the !!nkaga clataishes with decreasing pressure. Although some calculations have been performed which .,
indicate a qualitative coupling. I AM NOT AVARE OF SUBSTANTIATED QUANTITATIVE INFORMATION VHICH CAN BE APPLIED TO THIS ISSUE. '
- 2. A significant quantity of heat producing radioisotopes probably has left the core under the conditions of interest, and substantial deposits can be expected in the upper plenua structure. Most calculations incorrectly do not include thus effect. The influence of the incorrect modeling is the same as in item 1. THE UPPER PL'ENUM TEMPERATURES ARE SUBSTANTIALLY UNDERCALCULATED.
THIS ISSUE CANNOT BE DISMISSED AS READILY AS IS DONE IN THE SEABROOK i (PLC) DOCUMENT.
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12-13 The report authors are arguing that that little material will be released with the exception of noble gases, even in the case of eventual high pressure containment failure. I suspect the containment atmosphere i
will b& fcpleted of onygen early in the accident. I wonder what happens if the containment is suddenly opened with all of that hot stuff around -
inside containment that would love to get hold of some osygen. I am particularly interested in regions where there are large quantities of l
material, such as under the vessel and within the vessel.
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, l 15 The question is not the temperature at which circaloy can liquefy the fuel, but'the temperature of a significant quantity of the melt at the time of release. What is this temperature and what are the vapor pressures of the constituents at that temperature? iSee, for example, Figure 4-16.)
19 The discussion of the feedwater pipe penetrition indicates a probability of 17% that the penetration fa!!s and 83% that the pipe fails. Pipe failure is stated not to be a problem because the esternal feedwater pipe and valve will maintain containment integrity. Ve should check to see what valves are involved (checkst) to assess the situation if the applicant is taking credit for this behavior. ,
36 Vhat is the status of the decontamination factor of 1000? .
36 For the case where there is no suppression pool, what is the !!kelihood that the filtration system can hold the fission product inventory from a thermal aspect?
43 "The containment atmospheric purge line is the only penetration that can be opened during normal operation and provide a dieset release path."
Vhat is the likelihood that one of the large purge lines will have failed to seat following a refueling shutdown, and that the failure w!!! not have been discovered?
70 (I've only glanced at portaons of this table.) is the basemat thickness the overall thickness, or the thickness under the reactor cavity sump?
(These represent a significant depression in both Zion and Seabrook.)
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o ENCLOSURE 1 OBSERVATIONS AND QUESTIONS PERTAINING TO "SEABROOK STATION EMERGENCY PLANNING SENSITIVITY STUDY",
PICKARD, LOVE AND CARRICK, INC., PLC-0445, APRIL 1984 I have few comments that were not made in Enclosure 1. Such comments are not duplicated.
SECTION 4 COMMENTS 1 The second paragraph does not provide the background for the conclusion that the release via basemat seltthrough at Seabrook does not provide
- i. attenuation as would be the case with a soll based plant. Seabrook's containment rests on bedrock, but is not bonded to the bedrock.
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Therefore, !! there is a basemat seltthrough, the contents of the containment are assumed to vent directly to the atmosphere via a gap between the bedrock and the containment. This may be a conservative ,
assassment.
1 The last paragraph contains "For consistency with NUREC-0396, only releases during the acute accident time of it hours were considered."
Although I have no trouble with a consistent comparison, as I have commented previously, I don't believe we should limit the assessment to a completely consistent basis from the : int of public risk. (Further, there are inconsistencies anyway.) A pubite risk assessment should be achieved, and items such as the above should be justified in order to be escluded.
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. 8/28/G6 Seabrook EPZ Sensttvity Study Review Plao Goals of Review:
c'
- 1. To-provide a technical assessment of the adequacy of the y
Jf y ?' Seabrook Station support its Emergency conclusion Planning that the degreeSensitivity of public Study to protection f} afforded by a 1 mile emergency planning radius around the
/- Seabrook that was the time Station is' equivalent to the degree of protectior.
perceived for a 10 mile emergency planning radius rit the 10 mile generic planning radius was established
( in NijREG-0396.
- 2. In the event it is concluded that the Study does not -
adequately support its conclusion at the 1 mile radius, to determine the radius at which the study can support a conclusion of_ equivalent protection.
1,'
- ca.4 , d W ca4 a ioc ie c>.s< o M (4 c.;d beo kruck , JM C 'M iI& wt 4t - Ecn, &
f6lcc-(1 LH
&ckA Scope o FReview Effort I. Establish Technical Criteria-for Comparing the Degren of Protection to the Public A. NllREG-0396 bases
- 1. DBA-LOCA considerations
- a. PAG dose levels
- 1. whole body
- 2. thyroid
- b. early fatalities
- c. early in. juries
- d. latent health effects 1
- 2. WASH-1400 considerations
- a. PAG dose levels
- 1. whole body
- 2. thyroid *
- b. early fatalities
- c. early injuries
- d. latent health effects
-