ML20236U422
| ML20236U422 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 07/27/1998 |
| From: | Repka D NORTH ATLANTIC ENERGY SERVICE CORP. (NAESCO), WINSTON & STRAWN |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| CON-#398-19364 LA, NUDOCS 9807300134 | |
| Download: ML20236U422 (21) | |
Text
___
i j 00CKETED
% % 7,1998 UNITED STATES OF AMERICA E NUCLEAR REGULATORY COMMISSION OFFKE J s(
RUll? " .
BEFORE THE-ATOMIC SAFETY AND I ICENSING BO[ib JI n the Matter of )
)
North Atlantic Energy Service Corporation ) Docket No. 50-443-LA
)
(Seabrook Station) )
NORTH ATLANTIC ENERGY SERVICE CORPORATION'S SUPPI EMENTAL ANSWER RF STANDING ISSITER I. INTRODUCTION In accordance with the Licensing Board's " Memorandum and Order (Initial Order)"
of June 18,1998, North Atlantic. Energy Service Corporation '("NAESCO")' hereby; files its
' supplemental answer to the amended and supplemented request for hearing and petition to intervene fil'ed on behalf of the Seacoast Anti-Pollution League ("SAPL") and the New England Coalition on Nuclear Pollution ~ ("NECNP")' (collectively, the. " Petitioners"). This supplemental answer
- specifically responds to the Petitioners' filing ofJuly 9,1998 (" Amended Petition")", and' addresses only issues related to the Petitioners' standing to intervene in this matter and the timeliness of
< NECNP's request to become a party. NAESCO originally addressed these issues in an answer dated
- ' Petitioners' Amended I'etition consists of seven affidavits addressing the standing of the two l
L s. organizations, an untitled ' document with four proposed contentions, and. a separate-
" Memorandum of Law"_ addressing the admissibility of Proposed Contentions 2 through 4.
- ' The lateness of NECNP's participation in this matter is addressed only in the cover letter
~"
from Petitioners' counsel. As such, for purposes of'this " Supplemental Answer,"
LNAESCO's focus is on the cover letter and affidavits.
L 9eo7300i34'980727 ; 03 g PDR. ADOCK 05000443 5 0 eD g ) 4 y
na -- :--- - - - - _ - - - - - - - - - - - - _ - - - - - - - - - n-_--- - - - -- O
.e p.
July 2,1998 ("NAESCO's Initial' Answer").# NAESCO will further respond regarding the :
iadmissibilitp of the Petitioners' proposed contentions in accordance with the schedule established L bp the Licensing Board in its Initial Order.
p NAESCO concludes that neither SAPL nor NECNP has satisfied the requirements -
a i for intervention established by 10.C.F.R.' 2.714. The Petitioners have provided affidavits that cure .
procedural defects related to organizational standing. However, the Petitioners have not' defined or
' demonstrated a particularized offsite " injury in fact" that could result from the amendment at issue in'this proceeding. ' In addition, NECNP has still failed to address the criteria for late-filed petitions.
~
, 77 . As a result, the request for hearing and petition to intervene should be denied for both.
i II. BACKGROUND -
i? - As previousif discussed in NAESCO's Initial Answer, the FederalRegister Notice .
'giving rise to this proceeding relates to NAESCO's license amendment application of April 8,1998.
L l .
J As' described in the Notice, the specific proposed change would:
. . . revise [Seabr'ook Station] Technical Specifications n (TSs) 4.4.5.3, Steam Generators -- -' Inspection Frequencies, and 3,4.6.2.c, Reactor Coolant System
- s. - (RCS) Leakage, J and . the associated bases . to accommodate fuel cycles of up' to 24 months with .
respect to the allowed time' interval between steam
- generator inservice inspections, b '
.[63 Fed.~ Reg. 25113, at col ' l'. This is a narrow approval focused on the frequency of steam generator
- inspections. : It would also impose a more restrictive Limiting Condition for Operation'for RCS n
w 1
N 2 The NRC Staff also responded to Petition'ers' initial request for hearing and petition to intervene, by a filing dated June 24,1998 ("StaffInitial Answer").
J
leakage through steam generators. Id. It does not address other changes to Technical Specifications related to 24-month operating ~ cycles.*
In accordance with the Federal' Register Notice, interested parties were to file a
. request for hearing and petition for leave to intervene under 10 C.F.R. { 2.714(a)(1). Such a petition must:
. . . set forth with particidarity the interest of the petitioner in the proceeding, how that interest may be affected by the results oftheproceeding, including the
. reasons why ' petitioner should be permitted to _
intervene, with particular reference to the factors in paragraph (d)(1) of this section, and the specific aspect or aspects of the subject matter of the-proceeding as to which petitioner wishes to intervene.
10.C.F.R. { 2.714(a)(2) (emphasis added). See also Georgia Institute of Technoloov (Georgia Tech
' Research Reactor), CLI-95-12,42 NRC 111,115 (1995).
. SAPL (but not NECNP) first filed a document on this docket on June 5,1998. The foriginal letter timely responded to the Notice and primarily addressed the NRC Staff's proposed "no
.. significant hazards consideration" determination, a matter not properly before the Licensing Board
, (sce 10 C.F.R. { 50.58(b)(6))'. At that time, it was not clear that SAPL was even requesting a formalI hearing under 10 C.F.R.L { 2.714. ' Sub'sequently, the Petitioners collectively filed a Supplemental
,y, I Petition on_ June 18,1998. NAESCO responded to that filing in NAESCO's Initial Answer ---
pointing out three fundamental flaws in Petitioners' request:
ISec License Amendment Request 98-03," Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle Per Generic . Letter 91-04/ Submittal No.
~ 2,'.' NYN-98053, dated April 8,1998. As explained therein, the License Amendment
-. Request is "the second submittal in a planned series of License Amendment Requests which l propose changes to' the Seabrook Station Technical Specifications to accommodate fuel (cycles of up to 24 months."
i
[0 l8 ~
N ,
i < i.
l L __ .
v .. . .
_ _ _ _ _ _ _ _ _ - - _ _ J
'?:
(1) Neither SAPL nor NECNP had demonstrated its organizational
- standing, something typically accomplished by identifying at least one member and providing some." concrete indication" that the 8
member has authorized the' organization to represent him or her in the proceeding. See, c4, Vermont Yankee Nuclear Power Com.
.(2) Even assuming the above requirement was satisfied, neither SAPL nor NECNP had satisfied the requirements of 10 C.F.R. Q 2.714(a)(2) and Commission precedent' regarding a demonstration of a
- particularized injury; that could result 'to a member of the organization. Where standing ~ would be based on _ the nearby.
residence of the member, Petitioners must allege a clear potential for -
, ofTsite consequences resultingfrom the license amendment at issue.
See, e.g., Florida Power & T icht Co. (St. Lucie Nuclear Power Plant, Units 1 and 2), CLI-89-21,30 NRC 325,329-30 (1989).
(3) NECNP is a late petitioner and had not met its burden of making an affirmative showing against the factors of 10 C.F.R. Q 2.714(a)(1) for late-filed intervention petitions.
Petitioners' latest Amended Petition must be evaluated to determine whether Petitioners have made progress on these three issues. In the end, they have not made sufficient
. progress to be granted intervenor status.
III. DISCIISSION A.' . Organi7ational Standing The Amended Petition adequately addresses the previous pleading defect with respect to both SAPL's and NECNP's organizational standing. SAPL pmvides affidavits from four members who live within 10 miles'of Seabrook Station and who have authorized SAPL to represent
.their interests. . Similarly, NECNP provides affidavits from~ three members who reside within i10' miles of the plant : However while these affidavits address the pleading defect Petitioners still l.
s l
, j l
--_i----__-_-----_-_--__----__----------------- - - - - - - - . - - - - - - - - - - - - - - - - - - _ - - - - - - - - - - - - - - - - - -------------d
l 4
L need to show that these individual members have standing with respect to this matter. Georgia Institute of Technology (Georgia Tech Research Reactor), CLI-95-12,42 NRC 111,115 (1995).
i B. Particulars 7ed Offsite Harm is Still Not Demonstrated In a license amendment case where standing would be based on nearby residence, the Commission has held that petitioners must allege a clear potential for offsite consequences resulting from that amendment. Florida Power & Light Co. (St. Lucie Nuclear Power Plant, Units 1 and 2),
CLI-89-21, 30 NRC 325, 329-30 (1989). It is therefore incumbent upon the Petitioners to demonstrate, with particularity, how offsite consequences could result from the amendment here at issue. Cf. Sequoyah Fuels Corn. (Gore, Oklahoma Site), CLI-94-12,40 NRC 64,72-74 (1994)
(focusing on whether alleged injury is " concrete and particularized" and whether there is a " realistic i
l threat" of a direct injury). Vague references and generalized allusions to offsite accidents and
- injuries would not be sufficient. Comnare Lujan v. National Wildlife Fed'n,497 U.S. 871,884-89 i
(1990).
Petitioners, in the Amended Petition, do not directly confront the standing ,
requirements -including the directive of10 C.F.R. 2.714(a)(2). The form affidavits refer only to l
the affiants' " understanding that I and my family are within the area that could not only be impacted by an accident at Seabrook with off site consequences, but within an area in which the licensee is required to provide protective actions in that event." This, however, does not address the present !
amendment at all. This proceeding is not an initial licensing proceeding on Seabrook Station and the generalized threat of an accident with offsite consequences allegedly deriving from plant ;
1
= operation is not sufficient to confer standing here.
Reading the affidavits in their most favorable light, the only assertion of harm directed at license amendments, as opposed to plant operation, is a statement that "[a]ny license L
l exemption, which, if granted, would have the effect of reducing safety margins at Seabrook, would have an impact on me should an accident resulting from reduced safety margins, release radioactivity off site." However, this assertion is still general to any license " exemption" (or presumably, ;
amendment). It does not explain what " safety margins" are impacted by the amendment in this case, how those margins are reduced, or how any such reduction would lead to offsite releases of radioactivity. In NAESCO's view, this fails to satisfy the Commission's requirement for a showing af particularized aff-site harmfrom the amendmentproposed in this case.
l Looking beyond the affidavits and giving Petitioners credit for the petition more holistically (including the Supplemental Petition ofJune 18,1998 and the Amended Petition), it is apparent that Petitioners seek to litigate issues related to the adequacy of the proped steam generator surveillance frequency, issues related to the impact of 24-month operating cydes on fuel integrity, issues related to the potential for increased reliance on on-line maintenance, and issues related to other smveillance requirements (iA, not related to steam generators, such as for detection of mispositioned valves). However, with the exception of the arguments related to steam generator surveillance frequency, and as discussed in NAESCO's Initial Answer (at page 5), none of these matters relates to the specific license amendment at issue in the FederalRegister Notice that gave rise to this proceeding. The Notice does not address 24-month operating cyclo generally, and should not give rise to a proceeding on all aspects of that issue.
With respect to steam generator surveillance frequency, nothing in the Supplemental Petition or the Amended Petition shows "with particularity" how the proposed amendment will lead to offsite consequences that would create standing. As discussed in the FederalRegister Notice, the
[ NRC Staff has proposed a "no significant hazards consideration determination" regarding the L _ __ _ __ _ __
proposed amendment. 63 Fed. Reg. 25113. As discussed there (at col. 3), and among other conclusions, the Notice states (emphasis added):
While the proposed changes will lengthen the interval between surveillance, the increase in interval has been evaluated; and based on the reviews of the steam generator tube ECT inspections, it is concluded that the wear growth rate of the o.71y active degradation mechanism (AVB wear)identifici to date at Seabrook Station is such that sufficient margin exists between the plugging criteria and structural limit such that no tubes are predicted to exceed the stmetural limit even with the longer surveillance interval. Therefore, extension of the current surveillance intervals to accommodate a 24 month cycle will not significantly degrade the ability, the availability or the reliability of the steam generators to perform their intended safety function, thus, it is concluded that there is no significant reduction in a margin ofsafety.
Moreover, as is discussed in the Seabrook Station Updated Final Safety Analysis Report, Revision 4, Section 15.6.3 (copy attached), the steam generator tube rupture is an analyzed design basis event for Seabrook. The analysis includes an assessment ofradiological consequences and concludes (at page 15.6-12, Section 15.6.3.4) that:
The offsite doses from a postulated steam generator tube rupture at Seabrook Station are well within the exposure guideline values. Thus, the occurrence of this postulated accident will not result in an undue hazard to the general public.
Therefore, even ifthere were a reduction in margin as a result oftheproposedamendment, this does not equate to a potential for offsite harm.
In summary, Petitioners have not satisfied the statutory and regulatory standing f-requirements. Reliance on nearby residence alone is insufficient to confer standing where, as here, V
there is no obvious potential for offsite consequences resulting from the specific amendment as proposed.
C. NECNP Still Hns Not Justified Its I ate Petition NECNP's request for hearing is not timely. As such, NECNP was required to address ,
the factors of 10 C.F. R. { 2.714(a)(1) for a late-filed intervention petition. NECNP has still not done so. The burden is on NECNP to make an affirmative showing, and as a result ofits own failure, NECNP's petition should be dismissed.. See, e4, Boston Edison Co. (Pilgrim Nuclear Power Stationj. ALAB-816,22 NRC 461, 466 (1985) ("the burden of persuasion on the lateness factors is on the tardy petitionef'); Detroit Edison Co. (Enrico Fermi Atomic Power Plant, Unit 2),
LBP-82-96,16 NRC 1408,1432 (1982) ("the burden of showing good cause is on the late i petitioner").
In the cover letter transmitting the Amended Petition, Petitioners make only a token effort at addressing the lateness issue -- and this effort is purely misdirection. NECNP argues that "NECNP is mmly seekingjoinder pursuant to [ Federal Rule of Civil Procedure] 20." However, as NECNP acknowledges, the NRC Rules of Practice have no rule onjoinder. The NRC Rules of L Practice have a requirement for late filed petitions: 10 C.F.R. { 2.714(a)(1). The requirement for late l
petitions was specifically cited in the Federal Register Notice offering the opportunity for hearing
( in this case. See 63 Fed. heg. at 25102, col. 3. NECNP is a late petitioner and, just as the late I j; . petitioner in the Pilgrim case, it "must come to grips with those factors [of Sution 2.714(a)(1)] in the petition itself." Pilgrim: ALAB-816,22 NRC at 466 (denying a petition that was eight days late, precisely because the petitioner did not confront the lateness Getors).
p i
s.
NECNP offers no support for its "joinder" theory " The two NRC cases cited in the cover letter on the subject have nothing to do with late intervention petitions. At most, they stand for the proposition that the Federal Rules of Civil Procedure can serve as a guide to NRC Licensing Boards. However, neither case supports the proposition that a late petitioner can be added as a party summarily -- even if, as Petitioners claim, the net effect would be only that this proceeding would
" bear the names of two petitioners."5' Regardless of how NECNP would contribute to the proceeding, or how it envisions its role, NECNP is a late petitioner who has failed to make its required showing.
Neither NAESCO nor the Licensing Board should be obligated to refute whether NECNP meett the factors of 10 C.F.R. @ 2.714(a)(1), where NECNP has made no affirmative showing to refute. NAESCO notes, however, that the lack of a showing of good cause (Section 2.714 (a)(1)(i)) is particularly damaging to NECNP's position. In addition, NECNP's own indication that it will do no more than offer the same contentions and witnesses as SAPL (see the June 18,1998 Supplemental Petition, at Paragraph 9, for example), and its own argument that the Moreover, the Supplements! Petition of June 18,1998. where NECNP first appeared, makes no mention of"joinder."
3/
Sec Georgia Power Co. (Vogtle Electric Generating Plant, Units 1 and 2), LBP-96-16,44 i NRC 59, 62 (1996) ("the Federal Rules of Civil Procedure are considered suggestive with
, respect to NRC procedures," but Rule 23(e) requiring a court to approve a settlement in a class action has "no direct applicability" to the NRC context); Cincinnati Gas and Electric J
Co. (Zimmer Nuclear Power Station, Unit 1), LBP-82-47,15 NRC 1538,1542 (1982)
(stating only that an NRC rule ofprocedure for depositions is " adapted from" Rule 30(c) of the Federal Rules of Civil Procedure, and also noting that the Commission has not adopted the federal rules in toto).
In a case not cited by Petitioners, Public Service Co. of New Hnmnshire (Seabrook Station, Unit 2), CLI-84-6,19 NRC 975 (1984), the Commission specifically declined to express an l opinion on the procedural validity of a similar " motion forjoinder."
J i
9 l
4
~
, I effect ofits "joinder" will be merely to add a name to the case caption (Amended Petition, at cover letter), certainly suggest that NECNP will not assist in developing a scund record (Section 2.714(a)(1)(iii)) and that SAPL, if admitted, could certain'y represent NECNP's interests (_Section 2.714(a)(1)(ii) and (iv)). NECNP's late petition should be denied.
IV. CONCIIEION x,-
For reasons set forth above, neither SAPL nor NECNP has established standing to
. intervene in this proceeding. In addition, NECNP still has not established its basis to intervene as a late petitioner. The request for hearing and petition to intervene, as applied to both Petitioners, should be denied. l Respectfully submitted, k \
k David A. Repka '
WINSTON & STRAWN 1.'00 L Street, N.W.
- Washington, D.C. 20005-3502 (202)371-5726 Lillian M. Cuoco NORTHEAST UTILITIES SERVICE COMPANY
, 107 Selden Street .-
Berlin, Connecticut 06037
- ATTORNEYS FOR NORTH ATLANT,IC ENERGY SERVICE CORPORATION
. Dated in Washington, D.C.
~-
- ' this 27th day ofJuly,1998
~
1 l
p ,[
g .-
1 1
i
- ATTACHMENT l
i i
l t.
t
?.
J .,
I.
- NS Transmission Informs fon Avallcble fo* 'KJOHN' f JUL-21-1MB 15:34 WESCO LICD451NG-WORK $tV1 printed FAX 3584830D113E on 07/21/f598 E.?7PM
- Pg 2/28 CO3 773 7A10 P 02/29 SEABROOK UPDATED FSAR REVISION 4 15.6.2.4 canelusion e
I The desas which have been calculated for the sceidant of a small line break ;
outside .the guideline Containment are within a small fraction of the 10 CFR Part 100 values. '
l
/
15.6.3 - Staam Generator Tube Rt.mture 15.6.3.1 IdentDatten of caun c and Accidene Damerintion The tube.accident examined is the complete severance of a single, steam generator j This event is considerad an ANS Condition IV event. ' a = limiting fault - '
(see subsection 15.0.1). The accident is assumed to take pl, ace at power with the reactor coolant activity co. responding to continuous operation with a limited amount of defective fuel rods. The accident leads to an increase'in activity in the secondary systen due to leakage of radioactive coolant from the Reactor coolant system.
In the avant'of a coincident loss of offsite power, or failure of the condenser dump system, discharge of activity' to the atmosphere relief valves.takes place via tha steam generator safaty and/or power-operated ;
1 l
Since the steam generator tube material is Inconel, a highly ductile material, !
the assumption of a complete severance is' conservative.' The more probable
' mode of tube failure would be one or more minor leaks of undetermined origin.
- Activity in the Steam and Power Conversion system is subject to continual- i surveillance, and en accumulation of minar leaks which exceed the limita !
established operation, in the Technical Specifications is not permitted during unit
.The operator will determine that s~ steam generator tube rup*%u. has occurred, l and will identify and isolate the faulty steam generator w minimize contamination' of the secondary systes and ensure termination of radioactive >
releare to the ' atmosphere from the faulty unit. The recovery procedure can be-i carried out on a time. scale which ensures that break flow to the secondary system is , terminated before water level in the affected steam generator rises into the main steam pipe, sufficient inainations and controls are provided so _i l : the operator can carry out these functions satisfactorily. l'
' Assuming normal' operation of the various plant control syst' ens, the following !
sequence of events is initiated by a tube rupture:
4 15.6 5
- 1 p .
t -
ba________---_--------------------------------------~~-------- ~ ~ - ~ . _ - _ _ _ _ _ _ _ _ - _ - _ _ _ _ - _ _ _ _ - -
N3 Transmission Information Avallrble for 'KJOHN' -
WORKSRV1 printed fax 35B4B300113E on 07/21/1993 03:37PM
- Pg 3/28 AL-21-1999 . 158 35 tFESCO LICENSING CO3 773 7740 P.03/29 SEA 3100K tJPDATED FSAR REVISION 4
.a. Pressurizer low-pressure and low-level alarms are actuated and charging pusap flow increases.in an attempt to maintain pressurizer level. On the secondary side thera is a ataan flow /feedwater flow mismatch before trip as feedwater flow to the affected steam generator is reduced due to the additional break flow which is now being supplied to that unit,
- b. Continued loss of reaccor coolane inventory leads to a reactor trip signal generated by Dvertemperature AT, low pressurizer pressure, or manual operator action. Resultant plant cooldown following reactor trip ' leads to a rapid change of pressurizer level, and the safety injaation s1 5nai, -initiated manually by the operator or automatically by low pressurizer pressura, follows soon after the reactor. trip. . The safety injection signal automatically tarainates normal feedwater supply and initiates emergency feedwater addition.
- c. -The Jteam generator blowdevn liquid monitor and the condenser off- !
gas radiatien monitor will alarm, indicating a sharp increase in radioactivity in the secondary mystem.
j- d. The reactor trip automatically trips the turbine and, if offsite i power is available,~the.stemia dump valves open permitting steam dump to the condanser and atmosphere. In the event of a coincident station blackout, the steam dump valves would automatically close to protect the condenser. - The steam generator pressure would rapidly.
increase.resulting in steam discharge to the atmosphers (when the l .. pressure. reaches to the sstpoint) through the steam generator safety and/or power-operated relief-valves,
- e. Following reacter trip, the continued action of emergency feedwater supply and borated safety injection flow (supplied from the refool- .
'ing water storage tank) provida s heat sink which absorbs some.of ,
the decay heat. Thus, steam bypass to the condenser, or. in 'the casa of loss of,offsite power, steam relief to the atmosphere, is attenuated during the cima interval in which the rooovery proce' dure leading to identification of the ruptured steam generator is being carried out, l
The sequence of events for the steam Generator tube rupture with a I
. loss of.offsite power:is: given in Table 15.6-7. . l
- f. -safety injsecion flow results in increasing preeeuriser water level The time after trip at which the operator can clearly see returning level in the pressurizer is dependent upon the amount of operating auxiu.ary equipment.
j l
!. The : time - dependant parameters for' Steam Generator Tube Rapture event l, are listed in'tho'following figures: 1 p .
i 15.6 -I l
[.
4 z
e I
' ien Transmission Information AY:ltrble for. 'KJOHN'
^
WORK $RV1 orinted FAX 35848300113E on 07/21/1998 03:37PM
- Pg 4/28
! JUL-21-1998 15t35 ffESCO LICENSING 603 773 7740 P.04/29 l
l t
l s- SEABROOK UPDATED FSAR REVISION 4 t .
l l Figure 15.6 Pressurizer Pressure l Figure 15.6 Reactor coolant System Temperature l - Figure 15.6 Steam Cenerator Pressure (Faulted Steam l Generator) l Figure 15.6-$3 - Primary Coolant Flashing (Faulted Steam r Generator) f l Figure '15.6.54 -. Pressurizer Water Level l- Figure 15.6 Steam Flow Rate (Faulted Steam Generator) l Figure 15.6 Feedwater Flow to Faulted Steam Generator Figure 1$.6 Faulted Steam Generator Steam Flow Rate to Atmosphere l Figure 15.6 Faulted Steam Generator Break Flow Rate l Figure 15.6 Steam Cenerator Mass l Figure 15.6.60 - Faulted Steam' Cenerator 1.iquid Volume 15.6.3.2 Analysis of Effects and Consequences
! a. Mathed of Analysia .
l Two scenarios are cionsidered, cne leading to minimum margin to overfill' of the ruptured steam generator, and the other to maximum radiological consequences, i In estimating the mass transfer from the Reactor Coolant System l l 'through the broken tube for the scenario with maximum radiciogical consequences the. follow 1.ng assumptions are made:
i
- 1. Reactor trip occurs automatically as a result of lov l pressurizer pressure. I f
- 2. Following the initiation of the safety injection signal, two centrifugal, high head safety injection and two charging pumps l are actuated and continue to deliver flow until the emergency l instructions for a tube rupture accident indicate that the operator should switch off all but one pump vben he has l identtfied the accident and has pressurizer level indication. ;
I 15.6-7
j ta transmission Information Avsflable for 'CJOHN' -
WORK $RV1 printed FAX 35B48300113E on 07/21/1 7 8 03:37PM
- Pg 5/28 JUL-21-1999. it3:36 tVESCO LICENSItc 603 773 770 P.05/29 SEABROOX UPDATED FSAR REVISION I.
- 3. After reactor trip the break flow rasches aquilibriurn at the r
point where incomins safety injection flow is balanced by L
outgoing break flow. The resultant break flow persists from j; plant trip until reactor coolant and steam generator pressures are assumed to be equilibrated.
L
, 4. The power-operated relief valve on the main steam lina from tha ruptured steam generator fails in the full open position during h 'the initial attempt by the operator to isolate steam flow from i
,the ruptured steam generator.
l- -5 The operators identify the open power-operated relief valve and i manually isolate it by locally closing the upstream lock valve L vithin 30 minutes of the initial ~ attempt to close th4 valve.
The-implementation of further recovery procedure actions is
- f. delayed until the power operated relief valve on the ruptured j _. sesem generator has been isolated.
- 6. The break flow.is terminated by cooldown of the rasetor coolant l- system opening the power operated relief valves on the intact steam generators. reducing reactor coolant system pressure to a pressure below the pressure of the ruptured steam generator at the end of the cooldown by opening' one of the pover-operated '
. relief valves on the pressurizer, and stopping safety injection flow.
.l .
The assumptions for the overfill scenario are similar except that a power-operated relief viiva' on a main staan line from one of the intact steen generators is ' assumed te fail to open on demand, reducing the rate of reactor coolant system cooldown, and the power-
- operated relief valve on the main steam line from the ruptured steam
(- generator is. assumed to function normally and close' upon operator -8 i demand When the;operacors attempt to isolate steam flow from the .
ll , . ruptured" steam generator.
I b.- Recovery Procedure
- Symptoms of a tube ' rupture such as falling pressurizer pressure and level and increased charging pump ~ flow are also symptoms of small steamline breaks and loss-of coolant accidente. .It is therefore Limportant to determine that the accident is a rupture of a' ateam
(' ganarator tube to carry out the correct recovery procedura. The accidentfunder discussion can be identified by the following method.
In the event of a complete cube'ruptura, the level in one steam generator will rise more repidly than in the.others. 'This is a i uniqua indication of c. babe rupture accident. .Also, this accident could be identified by a steam generator blowdown radiation alare.
. The recovery procedure includes. isolation of the faulty sene generator and unit' cooldown.
15.6-8 ,
.6 L
3 l l
to Transmission'information Av11trble for. 8KJOHN' WO2KSRV1 printed fax 3584B300113E on 07/21/1998 03:37Pl4
- Pg 6/28 Jt.i.-21-1990 15 36 tFESCO L! CENSING . 603 773 7740 P. 06/29
,f. , r SEABROOK UPDATED FSAR REVISION 4 j-After the Residual Heat Removal System is placed in operation. the condenasce accumulated in the secondary system can be analyzed and processed as required.
There is ample ting available to carry out the above recovery procedures so that isolation of the affected stoaa generator is established before water level rises. into the main steam pipes. The available time scale is improved by the termination of emer5=ncy feedwater flow to the faulty steam generator and the regulation of pressurizer water level with only one charging pump operatin5 Normal operator viE11ance therefore assures that excessive water
- level will not be attained.
- c. Results the results of the scenario leadin5 to minimus margin to overfill of
.the ruptured steen generator show that operator implementation of
' the steam generator tube rupture recovery proceduraa results in -
termination of the break flow before water level in the ruptured
, steam genersor rises-into the main steam pipes. The results of the scenario leading to maximum radiological consequences are described in Section 15.6.3.3 below. Tha thermal hydraulic: results of this accident are less severe than that for a _ LOCA small break (see Subsection 15.6.5).
15.6.3.3. Radf ainef eal consmenene==
- a. -Assumptions and Par ==tarm l The conservative analysis of the postulated steam generator tube
~
rupture assumes the loss of offsite power. and hence involves the release of steam from the secondary system. This analysis assumes ,
primary and secondary activity levels-based on Technical Specification limits prior to the event. Parameters used in the
-analysis are presented in Table 15.6-6 Detail.d parameters and assumptions which are not presented in Tabis 15. 5-6 are discussed in this section.
,1. Conservative Analysis (a)- The primary and secondary coolant activities correspond to the specific activity limits given in the Technical Specifications. The primary' coolant' activity is 60 pCi/gm of dose l equivalent I-131 due to a pre-existing
. iodine spike, and 100E pCi/gm total coolant activity
. (conservatively assumed to be comprised entirely of noble ses activity). The anoondary coolant activity is
.0.1'pci/5e of dose equivalent I-131. No. noble gases are L
contained or dissolved in the secondary coolant; i.e..
I all noble gases leaked to ' the secondary coolant are 15.6 9 j
L -
l t __
f Ro Tri.nsmission Information Avallible' for 'KJOHN' WC::KSRV1 printed fax 355483DD113E on 07/21/1998 03:37PM
- Pg 7/28 JUL-21-1998-'15 37 tVESCO L! CENSING 603 773 7740 P.07/29 o
SEA 3 ROOK UPDATED FSAR REVISION 4 continuously released through the condenser off Ges System. The iodine and noble 6as isotopic breakdowns for l the primary and secondary coolant are given in Table 15.6-8.
(b) A. conservative analysis with a coincident iodine spike instead of the pre existing iodine spika has also been investigated. The coincident spiko is based on ,
(
. multiplying the iodine escape rata coefficient by a factor of 500 and using this revised escape rate j l
coefficient for the initial four bours following the j accident. Initial primary coolant activity is based on a ;
Technical specification limit of 1.0 pci/ga dose i equivalent 1+131 and 100E yCi/gm total coolant activity (conservatively assumed to be comprised entirely of noble gas activity). secondary side activity is based on 0.1 p.cl/gm of dose equivalent I-131. The Lodine and l.
noble gas isotopic breakdowns for primary and secondary coolant are presented in Table 15.6-8 i (c) Light hours after the accident the Residual Heat Removal system starts operation to cool down the plant, ;
terminating accident-asacciated releases.
-l
.(d) Thare is no release from the condenser vacuum pumps and l no steam generator blowdown during the accident.
.(e) The mass. of reactor coolant dischargad' into the secondary system through the rupture and the mass of steam released from the intact and ruptured stama generators to the atmosphere are presented in Table 15.611 (f) . The time-dependent fraction of rupture flow that flashes to stern .ad is insnediately available for release' to the environment wes determined based on thermodynamic considerations. '
-(g) The iodine removal efficiency for scrubbing of steam
- bubbles as they rise from the leak' site (assumed to be at
'the top of the tube bundle) to the water surface was alsn deteruiped based on NUREC-0409 (Reference (25))
rect a ndations. The iodine removal afficiency is a
- function of the bubble rise time, partition factor for iodine, and tha water. level above. the- top of the tube bundle.- The iodine partition factor used was ppm a 100.
. A bubble' riae time of 30 cm/sec (rise time for largest stab 1w bubble of 3.6 cm) was used (Reference (25)). A conservative scrubbing efficiency of 0.0 is used prior to the reactor trip, while ' two-phase water exists above the
- l. . break 1ccation. - The iodine removal efficiency used once 15,6-10 e
-1
(
[
f-q i
C_: _ = - 2-------
% transmission information'Avallebte for 'KJOHNi WCEKSRV1 printed FAX 3584B3DD113E on 07/21/1993 03:37PM
- Pg 8/28
-Jll.-21-193Gf 15:37 tFE5C0 LICENS1r0 603 773 7740 P. 0349 1
SEABROOK UTDATED FSAR
- REVISION 4 the collapsed steam generator.wate.; level is at a minimum j of approximately 3 feet above the bzeak location is 0.40 (h)
The maximum allowable preaccident primary to secondary laak ' race _ of 1 gym is assumed to be avanly dividad among the intact steam 8enerators.
(1) All noble gas activiry in the re. actor coolant which is t
. transported to che, secondary system via the tube rupture is assumed to be immediately available for release to the atmosphere.
(j) The iodine partition factor between the liquid and stamm of tha ruptured and intact steam generators is assumed to be 100 on a mass basis (Reference -(23)). The iodine partitioning factor for flashing primary coolant is..
conservatively assumed to be equal to 1.0 (Reference (25)). This meane that the fraction of primary coolant iodine volatized at any point in time equals the' flashing fraction.
(k)- No credit was taken for radioactive decay during release and transport, or for cloud depletion by ground.
deposition during transport to the site boundary or outer boundary of the low population sons. .
- (1) A series of tests were conducted at the MB+2 test facility (Raference (26)) to determine the primary coolant bypessing and secondary coolane carryover fraction is during a SGTK event. The results of chess
. tests were published by the Electric Power Research Institute . (Reference (26)). This report recommends that i 0.001% of the (primary coolant) break flow be assumed to be released directly from an open. valve. This 0.0011 i would result in an insignificant incraase 'in this iodine-release ocapared to the iodine release fros' primary ~
{
' 8 ,; coolant flashing as discussed in Paragraph No. 3 above. !
The EPRI report also recommands that secondary moisture carryover be taken as 0.005% of total scss flow from the ruptured steam generator. This value would:also'not significantly increase the lodine roleses during the accident. For.the analysta discussed here .it was conservatively assumed that 190 lb/hr of secor.dary liquid carryover.was released through the ruptured steam generator ASDV. This is equivalent to approximately 0.2%
of the total ralense and would more than account for
' coolant bypass and carryover, i
15.6411
- 4 l
l
ho TEnnsmission Information Av$ltable for 'KJOHN' WC.RKSRV1 printed FAX 35848300113E on 07/21/1998 03:37PM
- Pg 9/28
. JLA.-21-1938 15:38- NAESCO LICENSING a 773 7740 P.09/28 1
. a SEABROOK 'JPDATED FSAR REVISION 4 j
( (a) One hundred percene of the iodine activity in the l secondary liqui.d carryover and coolant bypass is assumed l
to be released. 2his conservatively accounts for flashing and atoa.ization.
(n) . has of off-site power is assumed to cause loss of main condanser availability and its assigned Iodine
!.< j
?
j .
decontamination fraction of 99%. '
- b. Activity Relammad I
The isotopic breakdowns of the activity released dua to the 1 postulated steam generator tube rupture arm givne in Tables 15.6-9 and 15.6 10. 3 l- I
- c. Results l
' Tha~ doses resulting from the postulated steam Benerator tube rupture ;
are presented in Table 15.6 12. The calculated doses are within i ia small fractions of 10 CFF. Part 100 exposure guidelines for the !
L coincident 'iodina spike analysis, and within 10 CIR Part 100 guidelines for the pre-accident iodine spike analysis.
Control room doses have been calculated for this postulated accident l and are given in Table 15.6-12.- These doses are based onihe releases given in Tables 15.6 9 and 15.6-10 and the control room parameters described in Subsection'15.6.5.4 and Appendix 155, 15.6.3.4 Conclusion 1-The offsite doses from a postulated steam generator tube rupture at Seabrook station are well within the exposure guideline values. Thus, the ocourrence '8 of this postulated accident will not result in an undue hazard to the general public.
- 15.6.4 Aggetrum of EWR 5tm Svgam Pintne Failures ours,ide of Containment
[-
Not' applicable to seabrook.
!. 15.6.5
- Less-of-Coolant Accidents ResultInt from a Snacermn.cf_ Postulated Pinine Bre=b vithin the 1taaetor Coolant Pressure Boundary 15.6 12
__._____._s._ _______m..m__ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . . _ _ .__._________________m_ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . . . _ _ _ _ _ _ _ _ _ _
____________________m_______a
a DGCKETED UNITED STATES OF AMERICA USNRC NUCLEAR REGULATORY COMMISSION
% JUL 28 P4 :29 BFFORE THE ATOMIC SAFFTY AND I.TCENSING BOARD OFFIC!. U 70- ...y RUUN . 2O In the Matter of ) ADJUDICAh vi SiAFF
)
North Atlantic Energy Service Corporation ) Docket No. 50-443-LA
)
(Seabrook Station) )
CERTIFICATE OF SFRVICE I hereby certify that copies of " NORTH ATLANTIC ENERGY SERVICE
' CORPORATION'S SUPPLEMENTAL ANSWER RE: STANDING ISSUES" in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, this 27th day ofJuly,1998. In addition, a courtesy copy has been sent by e-mail this same date to those parties designated by an asterisk (*).
Office of the Secretary B. Paul Cotter, Jr.*
U.S. Nuclear Regulatory Commission Chairman Washington, D.C. 20555 Atomic Safety and Licensing Board Attn: Docketing and Service Station U.S. Nuclear Regulatory Commission (original + two copies) Washington, DC 20555-0001 Steven R. Hom, Esq.* Dr. Charles N. Kelber*
Office of the General Counsel Administrative Judge U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board
. Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Robert A.Backus Linda W. Little*
Backus, Meyer, Solomon, 5000 Hermitage Drive Rood & Branch Raleigh, NC 27612 116 Lowell Street j P.O. Box 516 '
Manchester, NH 03105-0516 1
I l l I
4 I
i- !
l )
L _ _ _ _ _ _ _..____________________o
a
.-1 Adjudicatory File - _
Office ofCommission Appellate Adjudication Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission
. U.S. Nuclear Regulatory Commission - Washington, DC. 20555
- Washington, DC 20555 I
k h\Ce -
David A. Repka \- . .l Winston & Strawn
, Counsel for North Atlantic Energy Service Corporation l
i E i l
l
(. !
l
,. .i i
1 i .I i: . '
b '.
- r. -
(' . .
p ,
1 l= '
c i,' l w 1 a._--________--.---_______---_-____-_---___-- _ - _ - - - - - - - - - - = _ _ . _ _ _ - - - . - . "