ML20205J894
| ML20205J894 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 09/29/1986 |
| From: | Hernan R NRC |
| To: | Lyons J, Rosztoczy Z, Sheron B Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20205J677 | List: |
| References | |
| FOIA-87-7 NUDOCS 8704010449 | |
| Download: ML20205J894 (1) | |
Text
_
t f
p...
[
UNITED STATES a
I NUCLEAR REGULATORY COMMISSION 5
\\...../
WASHINGTON,D.C. 20065 September 29, 1986 NOTE TO:
ALL NRR DIVISION DIRECTORS B. SHERON Z. ROSZTOCZY J. LYONS H. Denton V. Noonan FROM:
R. HERNAN
SUBJECT:
AIF PAPER ON EPZ VS SOURCE TERM This paper was made available to the staff during an ACRS Subcommittee on September 26 in conjunction with a review of the updated Seabrook Probablistic Safety Study..
I thought it may be of interest to you and your staff.
~
Ron ernan F014-97-7 8704050449 870330 d/ k3 PDR FOIA GHOLLYB7-7 PDR
a.
l dl( t l#
Question 17 - Enclosure 3 a;
' ) *pa alcoq(*,
UNITED STATES
[
g NUCLEAR REGULATORY COMMISSION 5
l WASHINGTON,D.C 20665
/
g gf 386 Docket kos.: 50-443 and 50-444 kr. Robert J. Harrison President & Chief Executive Officer Public Service Company of hew Hampshire Post Office Box 330 Manchester, New Hampshire 03105
Dear Mr. Harrison:
Subject:
liequest for Additional Information for Seabrook Station Units 1 and 1'
- 2. Emergency Planning Sensitivity Study The enclosad Request for Additional Information documents the oral and handwritten questions transt.itted to Public Service Company of New Hampshire personnel ar.d contracturs during our meeting in Bethesda, Maryland on September 23, 1986.
.Please provide your responses promptly to facilitate our review.
4 Questions or additional information regarding this matter should be directed to the Techr.ical Project Manager for the review of the Seabrook Emergency Planning Sensitivity Study S. M. Long (301) 492-8413.
Sincerely,
)~.5 n,
Steven M. Long, Project Manager PWR Project Directorate No. 5 Division of PWR Licensing-A
Enclosure:
As stated cc: See next page F0M - 8'? -7 4(gllZ fh-- lO f}
L/y f
_ ~i___.__-,. _ _,..,._ ___ZTT^ ~ _1_. T ~?~~~ f E _1_ _ __ _ _._.. _. _ -
l h*
3 Mr. Robert J. Harrison Public Service Company of New Hampshire Seabrook Nuclear Power Station cc:
Thomas Dignan, Eso.
E. Tupper Kinder, Esq.
John A. Ritscher, Esq.
G. Dana 81shee, Esq.
Ropes and Gray Assistant Attorney General 225 Franklin Street Office of Attorney General Boston, Massachusetts 07110 208 State Hosue Annex Concord, New Hampshire 03301 Mr. Bruce 8. Beckley, Project Manager Public Service Company of New Hampshire Resident Inspector Post Office Box 330 Seabrook Nuclear Power Station Manchester, New Hampshire 03105 c/o US Nuclear Regulatory Commission Post Office Box 700 Dr. Mauray Tye, President Seabrook, New Hampshire 03874 Sun Valley Association 209 Summer Street Mr. John DeVincentis, Director Haverhill, Massachusetts 01839 Engineering and Licensing Yankee Atomic Electric Company
).
Robert A. Backus, Eso.
1671 Worchester Road O'Neil, Backus and Spielman Framingham, Massachusetts 01701 116 lowell Street Manchester, New Hampshire 03105 Mr. A. M. Ebner, Project Manager United Engineers & Constructors William S. Jordan, III 30 South 17th Street Diane Curran Post Office Box 8223 1
+
Harmon, Weiss & Jordan Philadelphia, Pennsylvania 19101 20001 S Street, NW i
Suite 430 Washington, D.C.
20009 f
Mr. Philip Ahrens. Esq.
Assist' ant Attorney General 6
1 !
State House, Station #6 Augusta, Maine 04333 Carol S. Sneider, Esq.
Office of the Assistant Attorney General Environmental Protection Division Mr. Warren Hall One Ashburton Place Public Service Company of Boston, Massachusetts 02108 New Hampshire Post Office Box 330 D. Pierre G. Cameron, Jr., Esc.
Seabrook, New Hampshire 03874 General Counsel
.f Public Service Company of New Hampshire i
Seacoast Anti-Pollution i.eague Post Office Box 330 Ms. Jane Doughty Manchester, New Hampshire 03105 1
5 Market Street Portsmouth, New Hampshire 03801 Regional Administrator, Region I U.S. Nuclear Regulatory Courission Mr. Diana P. Randall 631 Park Avenue 70 Collins Street King of Prussia, Pennsylvania 19406 Seabrook, New Hampshire 03874 Richard Hampe, Esq.
j New Hampshire Civil Defense Agency n
i 107 Pleasant Street
^
Concord, New Hampshire 03301 i
.... ~.
-n----,----r vw"'
e v-. mvm'-
w'----wv w--
-'fww
-'=--**rv-w
-r--r-w--m-r e--r--+-w--r-
-et,.
- - - *- * -er w-----ws'-
vr ew'-we------
f l,-
'l i
i i
Public Service Company of Seabrook Nuclear Power Station New Hampshire cc:
Mr. Calvin A. Canney, City Manager Mr. Alfred V. Sargent.
City Hall Chairman 126 Daniel Street Board of Selectmen Portsmouth, New Hampshire 03801 Town of Salisbury, MA 01950 Ms. Letty Hett Senator Gordon J. Humphrey Town of Brentwood ATTN: Tom Burack
^
RFD Dalton Road U.S. Senate Brentwood, New Hampshire 03833 Washington, D.C.
20510 Ms. Roberta C. Pevear Mr. Owen 8. Durgin. Chairman Town of Hampton Falls, New Hampshire Durham Board of Selectmen Drinkwater Road Town of Durham 4
- l,)
Hampton Falls, New Hampshire 03844 Durham, New Hampshire 03824 g
Ms. Sandra Gavutis Charles Cross, Esq.
Town of Kensington, New Hampshire Shaines, Mardrigan and RDF 1 McEaschern East Kingston, New Hampshire 03827 25 Maplewood Avenue Post Office Box 366 Portsmouth, New Hampshire 03801 thairman, Board of Selectmen RFD 2 South Hampton, New Hampshire 03827 Mr. Guy Chichester, Chaiman Rye Nuclear Intervention Mr. Angie Machiros, Chairman Counittee
[
Board of Selectmen c/o Rye Town Hall for the Town of Newbury 10 Central Road r
i Newbury, Massachusetts 01950 Rye, New Hampshire 03870 l
r Ms. Cashman, Chairman Jane Spector Board of Selectmen Federal Energy Regulatory Town of Amesbury Connission Town Hall 825 North Capital Street, NE Amesbury, Massachusetts 01913 Room 8105 Washington, D. C.
20426 Honorable Peter J. Matthews Mayor, City of Newburyport Mr. R. Sweeney Office of the Mayor New Hampshire Yankee Division l
City Hall Public Service of New Hampshire Newburyport, Massachusetts 01950 Company 7910 Woodmont Avenue l
Mr. Donald E. Chick, Town Manager Bethesda, Maryland 20814 Town of Exeter i
4 10 Front Street Mr. William B. Derrickson Exeter, New Hampshire 03823 Senior Vice President Public Service Company of 1
New Hampshire Post Office Box 700, Route 1 Seabrook, New Hampshire 03874
- = 2 -
- L :... -
- ~X.
L
~
~
, h1
'_ v.
l,
?
Enclosure 2
i,;
REQUEST FOR ADDITIONAL INFORMATION SEABROOK STATION, UNITS 1 AND 2 DOCKET N05.: 50-443 AND 50-444
~
EMERGENCY PLANNING SEN5ITIVITY STUDY 1.
Describe how the overpressurization calculations made by SMA were checked 1
or design reviewed.
2.
A meeting should be arranged with the originator of these calculations to assist the BNL reviewers in following these calculations and understanding the assumptions.
4 i
3.
Document the basis for the assumptions in the calculations.
In particular, explain the uncertainty factors assigned to various pressure capacities.
4.
Explain *the mechanism for transferring the load from the penetration sleeves to the containment wall, in particular, the equipment hatch, when subjected to high strain conditions. Explain how the rebars around the penetrations were assessed to assure that they can resist these loads in addition to the primary pressure induced loads.
5.
The calculations use a rebar ultimate stain value of 4.7%, i.e., more than 21 feet of linear extension for the hoop bars. This linear extension under the high pressure lead will be acconsnodated by formation of cracks in the concrete totaling approximately 21 feet in width. Justify the assumption i
that the pressure loads will be carried proportionately by the linear plate and the rebars (similar to the elastic condition) in this highly cracked condition. Also address the potential for developing a' crack large enough for the local extension of the liner plate to lead to its failure at that I
point.
6.
Was compatibility of strains in the rebars and the liner plate satisfied in the calculations? For example, the outennost hoop bars will fail before the inside bars and the liner plate reach their respective ultimate strengths.
Was this fact reflected in the calculations? In addition, how is the biaxial stress-strain state of the liner plate considered.
7.
The combined tension, shear and bending effect at base and spring line levels was not considered in the calculations (Ref. p. 35, assumption 6).
Verify that the combined effect does not change the conclusions of the l
analysis.
i
(
,._-,.,_m
,.-,._..m.
._y.,. _, -.
,,-__.c__
n.,.,m-..
..,.,e-,,,er,
..m,-_.~
m,
-.m,.-,
___g v.*
..T ! '( c.
" [.
8.
Since 31 cadwelds out of a total of 169 test samples failed at'a stress lowr than the rebar ultimate strength and there was apparently a construction problem concerning staggering of these welds, provide justification for not using a reduced ultimate strength for the rebar.
i
)
9.
The containment analysis is based on an axisymmetric geometry and loading.
l This is not the case due to the presence of adjoining structures such as the fuel building and main steam and feedwater pipe chase.
Identify these l
axisymetric conditions and assess their impact on the failure modes and l
analysis.
1
- 10. Only a sample of pipe penetrations are considered in scoe detail (X-23 j,
X-26andX-71). The justification to consider only these should be provided.
- 11. A structural evaluation of electrical penetrations should be provided.
i
- 12. The basis for the leakage area assigned to the flued head at failure should be provided.
- 13. A more detailed evaluation of the impact of punching shear at the Fuel Transfer Building should be provided.
- 14. Clarify the extent to which double ended piping failures have been considered in the overall containment perfomance assessment. Provide isometric drawings of all piping attached to containment penetrations.
.{
15.
In PLG-0465, page 2-10, Figure 2-3, the conditional frequency of exceeding whole body dose vs distance appears to be driven by the 52 i
source tem.
If this is the case, please describe all accident sequences (internal and external events) that contribute to the frequency of the SE source tem given in Table 4-2, pg. 4-7.
in particular, define how the timing and size of containment leakage was detemined for each of il-these classes of accident sequences. Justify the appropriateness of the binning of each of the accidents into this particular source tem.
- 16. Provide justification for the liner yield stress increase from the i
specified yield stress of 32 ksi to a mean yield stress of 45.4 ksi.
- 17. Indicate the correlation between containment failure sequences and the containment failure modes.
- 18. Provide the basis for concluding that the sight glasses in the hatches j
will not fail under high containment temperature and pressure conditions.
- 19. Document the effect that the recent update in seismic fragilities will i
have on the conclusions of the PSA results.
W J..J.-.- ----
3 1\\-
r ne* '
3-g l
- 20. Assess the impact on risk of using the assumption of ultimate containment capability predicted by VE8C analysis (150 psig).
- 21. What is the impact on risk from accidents during shutdown and refueling j
when the containment function may not be available?
i
- 21. It is the staffs understanding that preexisting violations of containment integrity were " included" in the PSA by assuming the average effect was to raise the containment leak rate to the design basis value of 0.1%/ day.
Compare this assumption with the containment integrity violation a.
data presented in hDREG/CR 4220.
1 b.
What contributions would these containment integrity violation data make to the probabilities for each of the release categories (Assume 3
the $5W category is redistributed over all the appropriate categories by the conditional probabilities of preexisting leakage paths of the sizeappropriatetoeachcategory).
q 23.
a.
Provide a narrative description that quantitatively delineates the i
dominant contributors to the dose probability vs distance curves and
)
the early fatality probability curves. The dominant release categories should be specified and the dominant accident sequences i
contributing to each of these release categories should be specified.
The probability of occurrence of each release category should be stated. These data should be provided for the current study and for the original PSA results. Changes between the two studies should be attributed to specific differences in the analysis.
j j
b.
Provice a set of early fatality conditional probability curves for each release category, assuming evacuation distances of 1 mile and
)
- a 2 miles.
c.
Provide the conditional mean risk of early fatality for each of f
the curves provided in b.
- 24. Provide a quantitative description of the effects of the following differences between the original PSA and the current study:
a.
reduction in probability of core-melt V sequences b.
factor of 1000 scrubbing of releases through RHR seals c.
change of release category (56 to S1) for unscrubbed event V sequences.
~
The effects should be described in terms of differences in risk curves for early fatalities and fur 200 rem vs distance.
- 25. Provide a list of all paths for loss of RCS inventory outside containment.
Show how these have been considered with respect to LOCA and with respect to containment bypass for radioactive materials following core damage.
i
-4e N*MM****4 4'*
D +e#
e'
--%9.%-
.--.w__
._y,,
O 89eMy&a Sp 6
6m S
4
'Y, g,-.
~
- 26. Indicate the extent to which the effect of local deflagration / detonation of hydrogen gas concentration in localized areas both inside and outside the containment has been considered in the assessment of risk. Include a discussion of how weak areas of containment have been considered in your assessment, for example, the containment is considerably weaker in its resistance to pressure loading from outside the containment.
- 28. Identify any penetrations connected directly into the containment atmosphere which rely on any remote manual or manual valves for isolation.
e l
4
.i
)
m QJ cl8l%
horen, huoco, Lc) icc_h L.our t clo nt q ONL.
wt so frocycih cr -nol much ofhe r on in PA( folc 1o frwrc eca c y
A yding.coi %6 etc.g.wv civ e rn.: i4 v3 L. uvn -
c3v.ne me c.rn 9e rticac+
ci. +
n > n =Oi i
1 y
J
$cc) bach inf o +Ru Tt6 LU644 bu how Mo cicyclogo. V,c koovA: bosJio vAd'.u @ hw fcxo,cn a ;..
muui cp 4,a 6coc evo. io cu, B
V.c - (cccornce ci 40 -e vul,q - O_v;na.nito Mos nravenc2 3 be encun.
voc 6. parn ( no gocic guen Ato ao; C o w n n.')
Cx h Alcaecy cboot mcctq Czo.nicacr<.> >
W cn duay/>.45 33 lo +#ctn g LUcticre dlc.ingeetw,3 w rnen-he;n cico 6rk. /w2C Acctmco
.t or eno.e. pat e Lococ-n4.qon buc.
- 9. %9atued -
I or 2 y
- DNL,
/2c 6arpn.
7
- 19. A, pc. 3,, e fa,[
Videco ae..c 3
i d..'
(O
\\A.I ni O k O "h'u d d k IM))J f.(\\,
c n ( (, I h, [ J gr.
( I-1 Y.'C,
}Q C)Q lo bedC.
L.U h O CE(Yi n L. O n, cic f01N ~Y 9~ 7 6/'is"
F-
. Y.
u,,t
~ CN LuCAfI m % c#t< wiA MO gc/ mcenq +o cf,ocun chho ter) ushet4 rcoccu ch Ao cloq M W / M w ro E :
who-goi~(6~2 9
[
i i
!(:
h i
I i
/
/
r e
4 COMMENTS AND QUESTIONS PERTINENT TO EMERGENCY PLANNING ZONE EVALUATION REACTOR SYSTEMS BRANCH OCTOBER 9, 1986 The recently reported work by Pickard, Lowe and Garrick (References 1 and 7) has significantly reduced conservatisms associated with containment bypass, which makes previously ignored (and untreated) bypass situations more j
important. Considerations associated with the potential impact of a reduced emergency planning zone size add to the importance of assuring adequate consideration of containment bypass accidents. Neither the Seabrook PRA, nor the more recent investigations, address the possibility of steam generator l
tube rupture as a consequence of phenomena associated with inadequate core cooling of sufficient duration that core damage results. This possibility can no longer be ignored since the tacit assumption that the potential i
consecuences are adequately compensated by other conservatisms may no longer be applicable, j
l Steam generator tube rupture during the approach to a core melt accident is of concern due a combination of:
4 1.
High temperatures in the reactor vessel, 2.
The possibility that approach to melt occurs at high reactor coolant system (RCS) pressure with the secondary side of the steam generators dry, and, 1
)
I 3.
eThe presence of M cLar.!sms which could transport hot fluid into the steam generator tubes, thus significantly reducing their strength.
(Mechanisms of concern involve both natural circulation and use of reactor coolant pumps as a "last ditch" effort to prevent or delay core melt.)
The situation with heat transport out of the reactor vessel seldom arises when analyses are performed with the suite of accident analysis codes originally applied to release analyses because these codes use one dimensional simulations of RCS fluid behavior. Fluid flow and transport of fission F018 7 1
4%
1
~. -..
products are based on a steam flow rate that is determined by the water boiloff rate in the lower vessel. This flow rate is low, and the energy transported into the upper vessel and out of the vessel toward the steam generators is correspondingly low. Multidimensional modeling of the same problem typically will show circulatory flow rates within the vessel of the order of a factor of ten higher than with one dimensional modeling. Heat and fission product transport are correspondingly affected. High RCS pressure results in a strong heat transport mechanism due to the high steam density, the high steam heat capacity, and the large density change with temperature.
A similar situation exists with the hot legs, where multidimensional analyses are necessary to model flow away from the vessel in the upper portion of the
~
legs, and return flow along the bottom of the legs. Similarly, muJtidimensional modeling is required to simulate behavior in the inlet and outlet plena regions of the steam generators. Single dimensional modeling is probably acceptable for individual steam generator tubes.
Wa have mentioned fission product transport several times in the above discussion. Elaboration is necessary. Most accident codes consider some sort of fission product transport.
Indeed, this is necessary in order to arrive at a source term. To our knowledae, as with energy transport, none of the popular computer programs utilize a multidimensional model to determine where the fission products go. This is just as important as with heat transfer phenomena. Many of the high heat production radioisotopes are volatile at the temperatures attained during a core melt accident.. They will i
be carried by the flowing fluid to a lower temperature surface, where they will'be deposited. This will cause localized heating, raising the temperature with the possibility that the radioisotopes will again enter the fluid stream and repeat the migration process. The popular one dimensional codes that are in use do not consider this movement of the heat heat generation location.
This is an important effect. Of the order of a quarter of the decay heat can be involved.
Although References 1 and 2 do not address the multidimensional aspects and the influence on fission product transport in general, there is a mention of 4
i i
?
~
4 the latter in Reference 1 (Section 4, page 12). A generalization is drawn to the effect that not-through-the-insulation heat losses are large as compared to the heat generation rate. Then, the conclusion is reached that "...the primary system heat losses are sufficiently great that the potential for lono-term revaporization within the primary system for a PWR with a large, dry containment is negligible. Therefore, this issue does not influence the Seabrook Emergency Planning Study." The intent appears to be to establish that after the initial phases of the accident, temperatures will become lower, and one need not be concerned with revaporization of fission products.
Certainly, this will be true at some point in time as the decay heat rate decreases; but as an overall generalization, it is incorrect. The fallacy 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 during melt, nor immediately afterward since large portions of core probabiy will remain together.
It similarly may not be the case for the upper plenum because of the multidimensional aspects of fluid flow and fission product transport as previously identified. Additional justification by the applicant will be necessary to establish an approximate time after which temperatures are sufficiently lowered that the conclusion is reasonable.
Reference 3 identified much of the above, and also posed a number of questions and observations based upon a cursory review of References 1 and P.
A number of these are being considered by Brookhaven National Laboratory as a part of j
their investigation of this issue as a contractor to the Staff. A number of otherg should be addressed by the Applicant. The items which should be addressed by the Applicant are provided in the following list:
1.
Please address the issues and phenomena identified above insofar as they are applicable to the Seabrook Station response to core damage accidents
)
and potentially impact upon risk.
i e
2.
Most of the work pertinent to severe accidents has addressed plant i
behavior ~at full power, and many other potential accidents which could result in releases have been assumed negligible in light of the conservatisms believed to exist in the full power treatment. Similarly, the nuances of containment bypass have not been carefully addressed in light of the conservatisms believed to exist in other aspects of x
containment response to severe accident conditions. This approach may not be adequate for a reduced Emergency Planning Zone Size for a 4
situation where some of the prior conservatisms have been removed.
Therefore:
a.
Please address the possibility of accidents inside the containment building while in Modes 2 - 6 (Startup, Hot Standby, Hot Shutdown, Cold Shutdown, and Refueling) insofar as these accidents could impact upon risk with respect to the issue under consideration here.
b.
Event V and steam generator tube rupture provide a direct path from I
the RCS to the environment during severe accidents. Please describe the Seabrook work which identifies any other direct paths.
c.
Please provide further information and/or specific references pertinent to release of radioactive material originating outside of the containment building insofar as the magnitudes are large enough to impact upon the issue under consideration-here.
3.!of450 psi The FSAR gives RHR relief valve flow rate as 900 gpm with a set pressure -
The flow rate does not agree with the value used in r
Reference 1, section 3, page 6.
Please explain, i
4.
Please describe the mecharism 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 R
(
emergency planning zone radius?
l 5.
Ref. 1 (page 3-7, 5th paragraph) references both high and low level sump alarms. What is a sump low level alarm?
4 l
U
4 t
6.
Page 3-7 contains a discussion of vault behavior in response to RHR system breaks. The emphasis is upon loss of equipment due to flooding.
What consideration has been given to breaks which are small enough that 4
the vault is not flooded, but there is a significant themal energy release that may impact equipment operation? Please include consideration that enough energy may be released to activate the fusible links in the ventilation system, thereby terminating ventilation and I
indirectly causing failure of pumps due to overheating of pump motors, and that this could occur at a time earlier than might occur due to flooding.
7.
Reference is made on page 3-7 to the RHR system crosstie line and RHR system response due to flow in this line as well as in the miniflow
~
bypass lines. The conclusion is drawn that the RHR system pressure will tend to be uniform as a reselt. Are flow conditions such that this is realistic? What is the impact of this assumption on conclusions l
pertinent to the discussion?
l 8.
The authors conclude on page 3-9 that presence of water in the reactor
~
cavity will decrease (significantly?) the revaporization of fission products from RCS and perhaps RHR surfaces. We 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 gases or vapor are flowing toward the break. Has this been i
f nvestigated?
i 9.
What is the,iustification for the statement on page 3-10 that the first sign of trouble will be pressurizer low level or low pressure alams? We 1
l suspect a number of other indicators may be first, such as abnomal indications from the PRT or even a smoke alam.
- 10. There have been a number of indications (prior to and including page j
3-11) that containment spray may be actuated due to RHR relief valve 5
i
. ~.
. ~
. ~.. -
i l
release into containment. What is the,iustification for this conclusion?
4 Include the effect of containment heat sinks and containment cooler operation in the response.
I
- 11. The statement on page 3-11 that "As soon as the pumps begin to produce flow to the RCS, valves in the miniflow lines close and all RHR pump flow is injected 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 PCS (as is stated in a later paragraph).
t
- 12. The last paragraph on page 3-11 contains a number of timing of event l
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 RCS parameter information, is necessary to substantiate the statement that k
RCPs will be tripped within about 21 seconds of break initiation.
j
- 13. An item under consideration for advanced nuclear power plants is the i
ability to monitor pressure on the low pressure side of check valves.
j This could provide early warning of check valve leaks and would provide I
monitoring capability to help assure check valves were operating properly. The same monitoring capability with respect to RHR sucticn i
line valves could identify if individual valves were mispositioned or a1 functioning. Would such a system for Seabrook be of significant l
benefit in reducing risk in a reduced size emergency planning zone?
j
- 14. Please elaborate on the page 3-23 list of actions an operator can take to I
mitigate the accident. This list appears to be short.
Include identification of what has been incorporated into operator training and procedures at Seabrook.
- 15. What is the frecuency of failures in the pipe tunnel that is mentioned on l
page 3-23, and which led the authors to conclude they are very low?
1 i
I 1
i 6
4 i
- 16. Page 3-27 references situations where the combined sump pump capacity is sufficient to remove leaks and keep the vaults from flooding. In these cases, the RHR, SI, and CS pumps are assumed not to be impacted by flooding. What consideration was given to failure of one (or both) sump pumps?
- 17. What is the maximum flow rate that can be injected into the RCP pump seals? (Of potential interest since it may be an alternate path for injection into the RCS.)
- 18. Shutting an RHR system crosstie valve is identified on page 3-35 as an action to help isolate a LOCA outside containment involving the RHR/SI systems. Has a careful evaluation of these systems been perfomed to assess isolation strategy? If so, are procedures in place at Seabrook Station which reflect the work?
- 19. Relative water levels in the RHR vaults and the RCS are mentioned on pages 3-35 and 3-36. 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.)
- 20. What is the justification for the statement on page 3-36 that the water i
level in the vaults will be approximately the same as that in the RCS?
(We do not agree because of the potential that pressure in the vaults and containment are not the same, and water temperature in the two locations ay differ.)
- 21. Page 3-37 contains the wording "End state DLOC cnntains sequences in which the interfacing LOCA has been terminated, and the ECCS has been degraded (D) (RHR or SI pumps have failed).... The point estimate frequency of DLOC is 4.0 x 10~7 per year. The additional failures reouired to achieve core melt would lower this frequence by at least one order of magnitude." What is the justification for this conclusion? (We have already lost a portion or all of the ability to in.iect water into the RCS via the usual paths.)
l 7
f
- 22. The bottom of page 3-37 contains a statement to the effect that failure of one charging pump will lead to core melt. Why is this the case? Our perception is that sufficient flow might be provided by alternate means to keep the core covered, such as use of the remaining two charging pumps, and perhaps the reactor makeup water pumps).
- 23. What is to be the status of the " temporary" 34.5 kV power lines which are identified on page 3-45?
i
- 24. What is to be the status of the mobile power supplies which are identified on page 3-467
- 25. What capability has been provided to connect external pumps as identified in the second and third paragraphs of page 3-467 (The brief mention on j
page 3-48 may imply that little has been accomplished.) Use of a pump to simply inject water into containment via the sprays on a short term basis (no recirculation) does not appear to be identified. Has this been considered?
- 26. Page 3-46 identifies a number of possibilities. What are the specific plans?
3
- 27. There have been several references to purchase of a mobile electric ge,nerator by pooled resources on the pages prior to page 3-49. What is the likelihood that such a generator would be needed by several plants at the same time, and hence might not be available to Seabrook Station when needed? Similarly, where is the generator to be stored, and how is it to be transported to Seabrook? Include consideration of post seismic and 2
post severe storm conditions in the response.
i i
?
- 28. A tacit assumption appears to be incorporated into References 1 and ?
that check valves are always closed.
In reality, many check valves require a (substantial) reverse flow to force them to close, and they additionally often require a significant reverse pressure to keep them 4
8 l
i 4
closed.
Is this the case for any of the valves of interest here? If so, please discuss the implications.
If not, what is the justification for the conclusion?
REFERENCES 1.
"Seabrook Station Risk Management and Emergency Planning Study", Pickard, Lowe and Garrick, Inc., PLG-0432, December 1985.
2.
"Seabrook Station Emergency Planning Sensitivity Study", Pickard, Lowe and Garrick, Inc., PLG-0465, April 1986.
3.,
"Lyon, Warren C., Review of Seabrook Documents Pertains to Change in Emergency Planning Zone Size," NRC Memorandum for Carl Berlinger,
/
/
l l
9 1
h PURPOSE AND AGENDA TRIP TO SEABROOK STATION AND TO BROOKHAVEN NATIONAL LABORATORY WARREN C. LYON, REACTOR SYSTEMS BRANCH, DPL-A SEABROOK STATION, OCTOBER 15, 1986 Purpose.
Study and evaluate systems which potentially impact risk as being considered in assessment of change in the Emergency Planning Zone si:e.
Acenda.
Plant visit will involve walk down of major piping, valves, pumps.
and heat exchangers associated with the following systems and components,:
1.
Residual Heat Removal / Low Pressure Safety injection System 2.
High Pressure Safety Injection System 3.
Charging System 4.
Refueling Water Storage Tank 5.
Reactor Coolant System, including major connections, surge line, RTD manifolds and piping, letdown, and RCP seal connections and region 6.
Steam Generators, including blowdown and feed connections 7.
Steam lines from the Steam Generators to the Main Steam Isolation Valves, including steam line to the Auxiliary Feedwater Turbine 9.
Component Cooling Water insofar as piping can be a conduit for radioactive material under severe accident conditions 9.
Under vessel region of containment, including in-core instrumentation line region and potential travel paths for core debris during Reactor Coolant System blowdown through the bottom of the Reactor Vessel under core melt conditions 10.
Other potential release paths from the immediate within-the plant environment to regions outside the plant.
BR00hHAVEN NATIONAL LABORATORY, OCTOBER 16, 1986
/
Purpose.
Program management and technical review of Brookhaven's investigation of the Seabrook Emergency Planning Sensitivity Study doenda.
Progress will be studied and evaluated, and guidance will be provided on the following:
Task 1 (System Evaluation), including Subtask 1A (Interfacing System LOCA), Subtask 1B (Accidents During Shutdown and Refueling Conditions),
Subtask 1E (Completeness)
BROOKHAVEN NATIONAL LABORATORY, OCTOBER 17, 1986 Purpose.
Technical review of Steam Generator Tube Rupture as a result of Severe Accident conditions
/20CM'I7'T Aaenda.
Seabrook, Brookhaven, and NRC personnel will jointly discuss and b/ Y[
review Seabroot progress pertinent to potential release of radioactive material via the Steam Generator under Severe Accident Conditions.
Previously published questions (Lyon Nemorandum of October 6, 1986), and applicant respenses pertinent to the subject, will be covered in depth.
.