ML20099L849
| ML20099L849 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 12/22/1995 |
| From: | IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT |
| To: | |
| Shared Package | |
| ML20099L847 | List: |
| References | |
| NUDOCS 9601030003 | |
| Download: ML20099L849 (27) | |
Text
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_ _ _.~..._. _._.. _ _ _ _. _. _. _ _
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RTS-269 to NG-95-2985 Page1of3 PROPOSED CHANGE RTS-269 TO THE DUANE ARNOLD ENERGY CENTER TECHNICAL SPECIFICATIONS 1
The holders oflicense DPR-49 for the Duane Arnold Energy Center propose to amend Appendix A (Technical Specifications) to said license by deleting certain current pages and replacing them with the attached, new pages. The List of Affected Pages is given below.
LIST OF AFFECTED PAGES iii 3.7-6 iv 3.7-22 3.7-1 3.7-23 3.7-2 3.7-24 3.7-3 3.7-42 3.7-4 6.11-5 3.7-4a 6.11-7 3.7-5 6.12-1 (new page)
SUMMARY
OF CHANGES-The following list of proposed changes is in the order that the changes appear in the Technical Specifications (TS).
Eage Description of Changes iii, iv Administrative changes were made to reflect the TS revision.
3.7-1 Current TS (CTS) that duplicate Appendix J, Option A were deleted.
Administrative changes (such as renumbering) were also made. A cross reference to the new drywell airlock specification was added.
A surveillance requirement (SR) was added to perform visual examinations and leakage rate testing in accordance with the Primary Containment Leakage Rate Testing Program. These changes are consistent with the guidance provided by the NRC by letter dated November 2,1995 from C. Grimes (NRC) to D.
Modeen (NEI).
9601030003 951222 PDR ADOCK 05000331 PDR p
RTS-269 to 3
i NG-95-2985 Page 2 of 3 4
3.7-2 CTS that duplicate Appendix J, Option A were deleted. CTS Sections 4.7.A.I.a(7) and 4.7.A.l.a(8) were also deleted. These sections contained the DAEC-specific values for P, and L,. This information is contained in the Bases.
i i
3.7-3 CTS that duplicate Appendix J, Option A were deleted. Section 4.7.A.I.c(3) was renumbered and was revised consistent with the ITS. These changes are consistent with the guidance provided by the NRC by letter dated November 2, 1995 from C. Grimes (NRC) to D. Modeen (NEI).
i -
l 3.7-4 CTS that duplicate Appendix J, Option A were deleted.
1 l
3.7-4a CTS that duplicate Appendix J, Option A were deleted and CTS 4.7.A.I.d.4 was renumbered to 4.7.A.l.c.
i j
3.7-5 CTS that duplicate Appendix J, Option A were deleted.
CTS 4.7.A.l.e contains a requirement to replace the T-ring inflatable seals for the 4
18 inch purge valves every four years. This provision is not in the ITS as it is a i
maintenance issue and not a surveillance for operability. It will be relocated to i
plant procedures. CTS 4.7.A.l.e also contains a requirement to verify (during j
Type C testing) that the mechanical modification which limits the maximum opening angle for the 18 inch purge valves is intact. The ITS only require this surveillance if the mechanical modification is not permanent. At DAEC, the 18 inch purge valves are permanently blocked to restrict opening to 30. These CTS 4.7.A.I.e provisions will be relocated to plant procedures. Any change to these requirements will require an evaluation in accordance with 10 CFR 50.59 j
CTS 4.7.A.l.e also contains a stipulation that the Cycle 6/7 refueling outage establishes the baseline for replacement of the T-ring inflatable seals for the I
containment purge valves. Siuce this is a one time provision that has been completed, its deletion is considered an administrative change.
l 3.7-6 CTS that duplicate Appendix J, Option A were deleted.
Requirements for primary containment air lock operability were added. Primary containment air lock leakage rate requirements were added as supporting surveillances for primary containment air lock operability. These changes are consistent with the guidance provided by the NRC by letter dated November 2, 1995 from C. Grimes (NRC) to D. Modeen (NEI).
3.7-22 The Bases were revised to be consistent with the changes to the TS pages.
RTS-269 to NG-95-2985 Page 3 of 3 3.7-23 The Bases were revised to be consistent with the changes to the TS pages, d
3.7-24 The Bases were revised to be consistent with the changes to the TS pages.
3.7-42
- The References were revised to reflect the changes to the Bases.
I i
6.11-5 The requirement to report the results of the Reactor Containment Integrated Leakage Rate Test was deleted. The recordkeeping requirements of Option B will be followed.
I 6.11-7 The Table of Routine Reports was revised to reflect the elimination of the requirement to report the results of the Reactor Containment Integrated Leakage Rate Test.
6.12-1 The Primary Containment Leakage Rate Testing Program was added to TS Chapter 6. This change is consistent with the guidance provided by the NRC by letter dated November 2,1995 from C. Grimes (NRC) to D. Modeen (NEI).
s DAEC 1 SJRVEILLOACE LIMitlNC CON 0!TIONS FOR OPERAfl0NS
- fQUIREMENTS P ACE NO.
3.7 Plant Contairunent Systems 4.7 3.71 AnhVn(NLry A.
Primary Containment L.... a,
A 3.7 1
% WQft, h T
O.
Primary Contairunent'PW0perated B
3.7 7 Isolation Valves C.
Drywell Average Air femocrature C
3.7 9 D.
Pressure Suppression Chaneer Reactor 0
3.7 10 Suilding vacuun treamers j
E.
Drywell Pressure suppression Chancer E
3.7 11 Vacuun treamers F.
Main Steam Isolation valve Leakage F
3.7 12 Control States (MSIV LCS)
G.
Smaression Pool Level and Teeperature G
3.7 13 N.
Contairment Atmospheric Dilution N
3.7 15 I.
Oxygen concentration 1
3.7 16 J.
3.7 17 K.
Secondary Containment Automatic K
3.7 18
! solation Danpers i
L.
Standby Gas Treatment System L
3.7 19 M.
Mechanical Vacuun Puip M
3.7 21 3.8 Auxiliary Electrical systems 4.8 3.8 1 A.
AC Power Systems A
3.8 1 B.
DC Power Systems 3
3.8 3 C.
Onsite Power Olstribution Systems C
3.8 5 0.
Aunillary Electrical Equpment -
0 3.8 5 CORE ALTERAfl0NS E.
Emergency Service Water System E
3.8 6 3.9 Core Alterations 4.9 3.9 1 A.
Refuteing Interlocks A
3.9 1 B.
Core Monitoring a
3.9 5 l
C.
Spent Fuel Pool Water Level C
3.9 6 D.
Auxiliary Electrical Ecpgsment -
0 3.9 6 CORE ALTERATIONS 3.10 Additional Safety Related Plant capabilities 4.10 3.10 1 A.
Main Control Room Ventilation A
3.10 1 8.
Resnote Shutdown Panels a
3.10 2a C.
Control Building Chillers C
3.10 2a 3.11 River Level Specification 4.11 3.11 1 RT5-2G3 AMENOMENT No.4e;.W/,,pe,y.2&f, 209 iii la/ W
. ~
DAEC *.
PAGE NO 5.0 Design Features 5.1-1 5.1 Site 5.1-1 5.2 Reactor 5.2-1 5.3 Reactor Vessel 5.1-1 5.4 cc.ntainment 5.4-1 5.5 spent and New Fuel Storage 5.5-1 5.6 seismic Design 5.6-3 6.0 Administrative controls 6.1-1 6.1 Management - Authority and Responsibility 6.1-1 6.2 organization 6.2-1 6.3 Flant staff Qualifications 6.3-1 6.4 Retraining and Replacement Training 6.4-1 6.5 Review and Audit 6.5-1 6.6 Reportable Event Action 6.6-1 6.7 Action to be Taken if a Safety Limit is Exceeded 6.7-1 6.8 Plant Operating Procedures 6.8-1
)
6.9 Radiological Procedures and Programs 6.9-1 l
6.10 Records Retention 6.10-1 6.11 Reportin Requirements leted.fri m ci)(1 McAh f
6.12 kro h(#
\\
"TC.eb 6.13 Deleted j
6.14 offsite Dose Assessment Manual 6.14-1
)
6.15 Process Control Program 6.15-1 Amendment No.
J$$,J/$,184 ly Jut 2 21992 RTs - u 9 is/9s
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DAEC-1 f
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 PLANT CONTAINMENT SYSTEMS 4.7 PLANT CONTAINMENT SYSTEMS l
Ane11erbility:
Aeolicability:
Applies to the operating status Applies to the primary and of the primary and secondary secondary containment system containment systeem.
integrity.
Obioetivet Obinetive:
To assure the integrity of the To verify the integrity of the primary and second y ontapnment primary and secondarg g e -
systems.
W Sgr containments.
M lbimete soecifientions soecifleation:
I:t : _h c fM l, I..ter r ity t
/ A.
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f.)1. pew,Primarv ContaiM), cc,m ts.nD h-s m
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PRIMARY'CONTAtNMENT INTEGRITY 1.
PRIMARY CONTAINMENT 4N3EGEEEE-A shall be maintained at all times chill M ' - --trated :- f ell:::::
when the reactor is critical or when the temperature is above a.
Type A Test l
212'F and fuel is in the reactor vessel except while performing P
ary Reactor Contaipatent low power physics tests at Into ted Leakage to Test atmospheric pressure at power f
}
w(t)
( 1)
The inter s3r aces of the
/_els not to sycessi-5 phance wifh subsections,.\\
drywell and us shall be y
3.7.B.2 satisfies the requirements visually pe d each operating to maintain PRIMARY CONTAINMENT cycle f evidenc f
deter,ioration.
In a tion, the INTEGRITY. p v
ax rnal surfaces of th torus d' /.
ithout PRIMARY CONTAINMENT low the water level sha be
(
INTEGRITY, restore PRIMARY inspected on a routine basis fpr h
f CONTAINMENT INTEGRITY within 1
-evid:::: Of t:rt: :;;;;; ion g [
hour or be in at least HOT le ' ;r.
SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> lN SERT A and in COLD SHUTDOWN within th.
following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Ex::;t f:: th: initial Ty; test, all Type A tests shall be performed without any prelimi ry
-vC
/
leak detection surveys and ak
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pairs immediately prior the c
c-c a.u.a.o.v.u,
2 If a A test is e plated but 3,7 A.z. d W
/
th. ac peance crite a of Specific ion 4.7..1.a.(8) is not v-
%d satisfied nd re rs are necessary, e
A test need not be repea provided locally measured le reductions, achieved b rep rs, reduce the containme
's ove 11 measured
[
1eakage ate suffi ntly to meet (
the ac eptance crite a.
2)
C1 ure of containment olation v vos for the Type A tes shall e accomplished by normal de of
}
actuation and without any
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RTS-2(o9 12/96 AMENDMINT NO. M.$,201 3.7-1
s
+
INSERT A Perform required visual examinations and leakage rate testing in accordance with a.
the Primary Containment Leakage Rate Testing Program.
4 l2/95
- -,. - -. - -. ~. -.. _
.-.~.- -.
DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS fww n v
evy yg i
Th7 :::::in;;;; 0;;; p;;;;;;;
uhall be allowed to stabilize to a period of about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior o the start of a leakage rate tes.
4 The reactor coolant pressure boundary shall be vented to t e containment atmosphere prior o the test and remain open dur ng h,
the test.
5) est methods are to comply with SI N45.4-1972.
6)
T accuracy of the Type A test sh 11 be verified by a
(.
su lamental test.
An cceptable met d is described in ppendix c of SI N45.4-1972.
7)
Perio c Leakage Rat Tests Periodi leakage ra e tests shall be perf d at or bove the peak pressure (Pa) of 4 peig.
4 8)
Acceptanc criter a The maxim all able leakage rate (Lam) is 0.
, where La is defined as t e design basis accident lea e rate of 2.0 weight pereen of contained air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a 43 psig.
9)
Additional equ ements If any per odic A test fails to meet t appli le acceptance criteria he test chedule applicab e to subse ont Type A tests w 1 be revie d and approve by the comm selon.
If tw consecutive pe odic Type A
(
tests fail to meet the acceptance crit la of 4.7.A.1.a.( ) aTypeAl test shall be performed ach ope ating cycle, or appr ximately ev.
18 months, whicheve occurs fi st, until two consecut e Type
(
A ests meet the subject captance criteria after ich ime the retest schedule of
.7.A.1.d may be resumed.
b.
Type B Tests Type B tests refer to penetrati no with gasketed seals, expansion 7
bellows or other type of resilien,t-(L t LCWTs -u9 AME W MENT NO. /@S,)@&,201 3.7-2
/1$6
DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEfttANCE REOUIREMENTS YYYNv Cp(L Y
. )
Test Pressure All Type B tests shall be performed by local pneumatic pressurization of the conta ne penetrations, either indivi ually or in groups, at a pressur not less than Pa.
2) cceptance criteria
}
The combined leakage rat of all strations subject to 5 and c eats shall be less t 0.60 La.
c.
Type c' Testa 1)
Type C tests shall performed on s containment isolatio valves.
<)
[
Each v lve to be to ed shall be closed y normal o ration and
),,
withoutanyprelimnaryexercising}
or adjus nts.
j INSERT B 42)
Ace.ptane, crite. - The combin.d )
leakage rate for all penetrations
/!
subject to\\ Type and c tests shall be le o t 0.60 La.
3)
The leakage ir any one main steam isolati valve shall not exceed 100scf/hrata. test pressure of 4 peig.*
The combined max pathway leakage rate for al fo e main steam lineed shall not e cae 200 scf/hr at a test press e of 24 poig.
g 4)
The leakacye rate pom any containmept isolation valve whose seating rface reaisins water i
covered st-LOcA, And which is
='
hydrostatically Type c tested, l
shall bh included in the Type c test t'tal.
If a main steam isolation va ve
'xceeds 100 sef/hr, it will be e
restored to s 11.5 scf/hr.
1/V rs-u s AMENDMENT NO.//),il91,207 3.7-3 17-f 6
D INSERTD-b.
Verify leakage rate through each MSIV is s 100 scfh when tested at 2 24 psig and that the combined maximum pathway leakage rate for all four main steam lines is s 200 scfh when tested at 2 24 psig in accordance with the Primary Containment Leakage Rate Testing Program.*
- If the leakage rate through an individual MSIV exceeds 100 scfh, the leakage rate will be restored to s 11.5 scfh.
l R rs - 2q
( 2/ cis
DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS CN
[
Periodic Retest SchJdule
(
~
1 Type A Test After the preoperational leak ge d,
rate tests, a set of three A
tests shall be perforsned, at approximately equal interva e
/
uring each 10-year servic riod.
(These intervals y be ended up to eight mont if n essary to coincide wi re unling outages.)
The third te of each set shall con c.ted when the p1 is shut down for the 10-year p ant in-servi e inspections.
}
The pe formance of A tests shall limited to riods when the pla facility a nonopera ional and secured in the shutdown onditior under administe ive co trol and in accordance with a plant safety procedures.
2)
Type B Tests
(
a)
Penetrations seals of this type (except locks) shall be leak tested t rester than or equal to 43 si (P.) during each )
reactor sh down or major refueling e othe convenient
/*
interval t in n case at
(
intervals greater an two years.
b)
The per nnel airlo shall be pressu sed to great than or equal o 43 peig (P,)
nd leak test at least once ery six (6)
)
month. This test int al may be g exte ed to the next re ualing
(
out e (up to a max 2 mum nterval bet, en P tests of 24 no the) i pecyvided there have been o j
airlock oranings since the last 1
rVccessfu) tout at P..
j j
/
/
c) ithin three (3) days after securing the airlock when k
- "*" ""*"" 1"*"#1"Y
- #*9" 3-0 N
the airlock gaskets shall be 1 ak tested at a pressure of P..
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m 1
g (1
Jp n y}~.)y 8 75-2.0 i
W_/15 I
AMENDMENT NO. ffi,#f,:df4 207 3.7-4
s DAEC-1 LIMITING CONDITIONS FOR OPERATION
_ SURVEILLANCE REOUIREMENTS Y Y^YPnom%
)
C Tests Type C to shall perforised during eac~
or shutdown for major refu g
other conven interva t in no case at evals greater t two } "
PA.
Additional Periodic Tests Additional purge systest isolation valve leakage integrity testing shall be perforsned at least once every three months in order to detect excessive leakage of the purge isolation valve resilient seats.
The purge system isolation valves will be tested in three groups, by penetrations drywell purge exhaust group (CV-4302 and CV-4303), torus purge exhaust group (CV-4300 and CV-4301), and drywell/ torus purge supply group (CV-4307, CV-4308 and CV-4306).
1 i
i l
l l
l l
RTS-267 AnanonanT wo.ns na,2s,207 3,7_4.
s DAEC-1 LIMITING CONDITIONS FOR JPERATION SURVEILLANCE REOUIREMENTS
.p(y 4
n Seal Replacement and Mechanical l
Limiter k
The T-ring inflatable seals for purge isolation valves CV-4300, CV-4301, CV-4302, CV-4303, CV-4306, CV-4307 and CV-4308 shal be replaced at intervals not to exceed four years.
ring Type C testing, it sh 11 be i
v rified that the mechanica ification which limits t e m
um opening angle for urge iso ation valves CV-4300,
-4301, CV-02, CV-4303, CV-4306 CV-4307 and
-4308 is intact.
The b eline for this r quirement shall e established du ing the 3
Cycle 7 refuel outag.
/
f.
Contai nt Modificat on Any major modificati n, replacemen of a e nent which is part of the pr ry reactor containment bounda y, or resealing
),
a seal-weld door, performed after the pr oper tional leakage rate test sha 1 followed by
[
either a Type
, Type B, or Type C test, as appli le, for the ares
(,
affected by the modification. The s
/
measured leaka from this test shall be inclu e in this test report.
The cc tance criteria
)1 as appropria s all be met.
/
Minor modifi ation, replacements,
/
or ressalin of se -welded doors, (N
(
performed d rectly ior to the conduct of a schedu d Type A test 4 do not re ire a separate test.
g.
Reporting 4
Periodi tests shall be the
' r subject of a summary te nical report ubmitted to the ommission approx tely 3 months af er the condu of each test.
Th report s,
/
will e titled " Reactor cont inment Integrated Lea ge Rate Test."
[
Th results of the periodic 1,
tapting performed to satisfy e
sallbereportedwiththesum\\
r irements of 4.7.A.l.d.(4) 3,
/
mary s 1
(
, chnical report prepared to
\\
provida the r6sults of the testibg
/ performed in accordance with
/Section 4.7.A.1.d.(3).
i J
- 846i,
. ANENDMENT NO. Ah&,AAA,201 3.7-5 whJ 12/ts
=
- - - - ~
~.
. - -. ~
DAEC-1 l
LIMITING CONDITIONS FOR OPERATION SUEVEILLANCE REOUIREMENTS l
g he report shall include a chamatic arrangement or scription of the leakage rat m asurement system, the in trumentation used, the su lemental test method, th test pro ram selected, and all sub quant periodic tests.
he repo t shall contain an an ysis and i torpretation of the oakage rate et data for the T A test result to the extent ne saary to demons ate the acceptab lity of
)
the con ainment's leaka rate in meeting he acceptance riteria.
For each riodic tes, leakage test resu a from A,
B, and C tests shal be repor d.
The report shal contai an analysis i
and interpr tation f the Type A test results and a ummary analysis of rio e Type B and
') j Type C tests hat ce performed
/,
since the las A test.
Leakage test r e its from Type A, B, and C tests at failed to meet the acceptance riteria shall be reported in a arate accompanying ry report.
The Type A test s report shall include an a lys and interpretat n of he test data, l
the least-s ares t analysis of 1
the test d a, the strumentation error anal sis, and he structural
[
condition of the co ainment or I
(
componen,
if any, w ich contribu ed to the fai ure in meeting he acceptance criteria.
Result and analyses o the supple ntal verificati test emplo d to demonstrate he valid ty of the leakage to test meas.ements shall also b<
)
inc ded.
Th Type B and C tests e ry re et shall include an ana sis a
interpretation of the da a and i
e condition of the compone a phichcontributedtoanyfail e
in meeting the acceptance
/ criteria.
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-u" i2-/15 AM3DMENT NO. Q,/M,M/,201 3.7-6
INSERT C 2.
Primary Containment Air Lock When in RUN, STARTUP, or HOT SHUTDOWN MODE, the primary a.
containment air lock shall be OPERABLE.
b.
With one primary containment air lock door inoperable, verify the OPERABLE door is closed within I hour; lock the OPERABLE door closed within the followin 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />; and verify the OPERABLE door is locked closed once per 31 days.t.2.g4 c.
With the primary containment air lock interlock mechanism inoperable, verify an OPERABLE door is closed within I hour; lock an OPERABLE door closed within the followin 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />; and verify an OPERABLE door is locked closed once per 31 days.gdd 1
d.
With the primary containment air lock inoperable for reasons other than 3.7.A.2.b or c above, immediately initiate action to evaluate primary containment overall leakage rate per 3.7.A.1, using current air lock test results; verify a door is closed within I hour; and restore air lock to OPERABLE status within the following 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />.i. 2 e.
With Specifications 3.7.A.2.b,3.7.A.2.c or 3.7.A.2.d not met, be in HOT I
SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.i.2 Note 1: Entry and exit is permissible to perform repairs of the air lock components.
1 Note 2: Take actions per Specification 3.7.A.1, " Primary Containment," when air lock leakage results in exceeding overall containment leakage rate acceptance criteria.
Note 3 Entry and exit is permissible for 7 days under administrative controls.
Note 4: Air lock doors in high radiation areas or areas with limited access due to inerting may be verified locked closed by administrative means.
Note 5: Entry into and exit from containment is oermissible under the control of a dedicated individual.
RT s - 2(?t t 7. /9 5
INSERT D 2.
Primary Containment Air Lock Perform required primary containment air lock leakage rate testing in accordance a.
with the Primary Containment Leakage Rate Testing Program.6' 7 b.
Once per 184 days, verify only one door in the primary containment air lock can be opened at a time.8 Note 6: An inoperable air lock door does not invalidate the previous successful perfonnance of the overall air lock leakage test.
4 Note 7: Results shall be evaluated against acceptance criteria applicable to SR 4.7.A.l.a.
Note 8: Only required to be performed prior to startup following entry into primary containment when the primary containment is de-inerted.
1 1
Rf$-2dch r2 /95
1 s
A 0h frt fY10 N cr k O.t A W h*
IcLoek 3.7.A & 4.7.A BASES:
Primary Containment I.t;;ri:.
/
The integrity of the & x x s m s w A w % >and operation of the core standby primary containment cooling system in combination, limit the offsite doses to values less than those suggested in 10 CFR 100 in the event of a break in the primary system piping.
Thus, containment integrity is specified whenever the potential for violation of the primary reactor system integrity exists.
Concern about such a violation exists whenever the reactor is critical and above atmospheric pressure. An exception is made to this requirement during initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required.
There will be no pressure on the system at this time, thus greatly reducing the chances of a pipe break.
The reactor may be taken critical during this period; however, restrictive oporating procedures will be in effect again to minimize the probability of an i
accident occurring.
Procedures and the Rod Worth Minimizer would limit control worth such that a rod drop would not result in any fuel damage.
In addition, in the unlikely event that an excursion did occur, the reactor building and standby gas treatment system, which shall be operational during this time, offer a sufficient barrier to keep offsite doses well below 10 CFR 100 limits, In the event primary containment is inoperable, primary containment must be restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time provides a period of time commensurate with the importance of maintaining primary containment and also
)
ensures that the probability of an accident requiring primary containment during this time period is minimal.
The primary containment preoperational test pressures are based upon the calculated primary containment pressure response corresponding to the design basis loss-of-coolant accident.
The peak drywell pressure would be about 43 psig which would rapidly reduce to 27 psig within 30 seconds following the pipe break.
Following the pipe break, the suppression chamber pressure rises U5-167 AMENDMENT NO. 201 3.7-22 12L/16I 1
t DAEC-1 to about 25 psig within 30 seconds, equalizes with drywell pressure shore 19 thereaft r gpp geylghe rg ell prpsp rp ecpyq(gferepc he-r e rn es c f C ant o-\\ p e.A is dtCi e
LU 8 M A Nmi m{;.> m aVlo ua a_b te
( en k. a c_.
- I',
rgh Q.d oC E.6 \\' C b
welht o f % e., ccm b a me f c.tr l'19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> c.vled d (fica a drywell and (suppress on chamber rt_ Giu f' O C h o k
- A u fn C nt t[n Wutft The d Agn pressure of t s 56 psig, g
(Reference 2).
T;.e desig.- Lesie e;;.__...
...h;;;._.. i; 2.T./d;3 r,-a- (
M_ALA
_; ::: :: Of 12 7:i; _ - a pointed out above, the drywell and suppression chamberpressurefol)lowinganaccident
't L - b would equalize fairly rapidly.
Based on the primary containment pressure response and the fact that the drywell and suppression chamber function as a unit, the primary containment will be tested as a unit rather than the individual components separately.
The design basis loss-of-coolant accident was evaluated by tne AEC staff incorporating the primary containment design basis accident leak rate of 2.0%/ day, (Ref. 3).
The analysis showed that with this leak rate and a standby gas treatment system filter efficiency of 90% for halogens, 90% for particulate iodine, and assuming the fission product release fractions stated in TID-14844, the maximum total whole body passing cloud dose is about 2 rem and the maximum thyroid dose is about 32 rem at the site boundary over an exposure duration of two hours. The resultant thyroid dose that would occur over the course of the accident is 98 rem at the boundary of the low population zone (LPZ).
Thus, these doses are the maximum that would be
{
expected in the unlikely event of a design basis loss-of-coolant accident.
I These doseu are also based on the assumption of no holdup in the secondary containment, resulting in a direct release of fission products from the primary containment through the filters and stack to the environs.
- NOTE: The initial leak rate testing performed during plant startup was conducted at a pressure of 54 psig in accordance with the original FSAR analysis of peak containment pressure (Pa).
RTs - 2 M l1[M AMENDMENT No.201 3.7-23
6 DAEC-1 Therefore, the specified primary containment leak rate is conservative and provides additional margin between expected offsite doses and 10 CFR 100 guidelines.
(%D s
j f
)
)
psig is 2.0 weight percent per day.
To allow a margin for possib%e
/
leakage det loration during the interval between Type A tests,,the maximum allowable contai nt operat tonal leak rate (L.), is 0.75
\\
/
l Type B and Type C tests are rformed on testab penetrations and isolation
)
valves during the interim period tween A tests.
This provides j
assurance that components most likel undergo degradation between Type A
,.s tests Laintain leaktight intog y.
A conte ed list of the testable penetrations and isolatio alves subject to Type and Type C testing is
/
located in the plan dministrative control Procedures.
The cent nt leakage testing program is based on NRC guideline or
/
dev opment of leak rate testing and surveillance schedules for reactor
...:..i m le, u
g-
[J' 3.7.B and 4.7.5, Bases.Js Primary containment Power Ooerated Isolation Valves Automatic isolation valves are provided on process piping which penetrates the containment and communicates with the containment atmosphere. The maximum closure times for these valves are selected in consideration of the design intent to contain released fission products following pipe breaks inside containment.
Several of the automatic isolation valves serve a dual role as both reactor coolant pressure boundary isolation valves and containment isolation valves.
The function of such valves on reactor coolant pressure boundary process piping which penetrates containment (except for t. hose lines which are required to operate to mitigate the consequences of a less-of-coolant accident) is to provide closure at a rate which will prevent AMENDMENT No. 201 3.7-24 gg - gg i 2_/q s
e INSERT E Primary containment OPERABILITY is maintained by limiting leakage to less than or equal to 1.0 L, except prior to the first startup after performing a required Primary Containment Leakage Rate Testing Program leakage test. At this time, applicable leakage limits must be met.
Maintaining the primary containment OPERABLE requires compliance with the visual examinations and leakage rate test requirements of the Primary Containment Leakage Rate Testing Program. Failure to meet air lock leakage testing, purge valve leakage testing, or main steam isolation valve leakage does not necessarily result in a failure of surveillance requirement 4.7.A.I.a. The impact of the failure to meet these SRs must be evahtated against the Type A, B, and C acceptance criteria of the Primary Containment Leakage Rate Testing Program.
One double door primary containment air lock has been built into the primary containment to provide personnel access to the drywell and to provide primary containment isolation during the process of personnel entering and exiting the drywell.
The air lock is designed to withstand the same loads, temperatures, and peak design internal and external pressures as the primary containment. As part of the primary containment, the air lock limits the release of radioactive material to the environment during normal unit operation and through a range of transients and accidents up to and including postulated DBAs.
Each air lock door has been designed and tested to certify its ability to withstand a pressure in excess of the maximum expected pressure following a DBA in primary containment. Each of the doors contains a single gasketed seal to ensure pressure integrity. To effect a leak tight seal, the air lock design uses pressure seated doors (i.e.,
an increase in primary containment internal pressure results in increased sealing force on each door).
The t.ir lock is nominally a right circular cylinder,12 ft in diameter, with doors at each end that are interlocked to prevent simultaneous opening. During periods when primary containment is not required to be OPERABLE, the air lock interlock mechanism may be disabled, allowing both doors of the air lock to remain open for extended periods when frequent primary containment entry is necessary. Under some conditions, as allowed by the primary containment air lock LCO, the primary containment may be accessed through the air lock, when the interlock mechanism has failed, by manually performing the interlock function.
The primary containment air lock forms part of the primary containment pressure boundary. As such, air lock integrity and leak tightness are essential for maintaining primary containment leakage rate to within limits in the event of a DBA. Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assumed in the safety analysis.
8TS-1M es
For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door to be opened at a time. This provision ensures that a gross breach of primary containment does not exist when primary containment is required to be OPERABLE.
Closure of a single door in the air lock is sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for normal entry and exit from primary containment.
The air lock interlock mechanism is designed to prevent simultaneous opening of both doors in the air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum-expected post accident primary containment pressure, closure of either door will support primary containment OPERABILITY. Thus, the interlock feature supports primary containment OPERABILITY while the air lock is being used for personnel transit into and out of the containment.
Maintaining the primary containment air lock OPERABLE requires compliance with the leakage rate test requirements of the Primary Containment Leakage Rate Testing Program. The acceptance criteria were established during initial air lock and primary containment OPERABILITY testing. The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall primary containment leakage rate. The frequency is required by the Primary Containment Leakage Rate Testing Program.
Testing of the air lock requires the installation of a strongback on the inner door to keep it closed during testing, since the air lock is tested by pressurizing the space between the inner and outer doors. Without the strongback, the inner door could be forced open by the pressure against it in the non-accident direction. Opening the air lock door h remove the strongback (or other tut equipment), does not require further leak testing, as long as the inner door seal is not disturbed.
The primary containment air lock surveillance requirements have been modified by two notes. One note states that an inoperable air lock door does not invalidate the previous successful performance of the overall ali lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. The other note requires the results of air lock leakage tests be evaluated against the acceptance criteria of the Primary Containment Leakage Rate Testing Program (TS Section 6.12). This ensures that the air lock leakage is properly accounted for in determining the combined Type B and C primary containment leakage.
~
i iz/%
s DAEC-1 3.7.A & 4.7.A REFERENCES 1.
"Duane Arnold Energy Center Power Uprate", NEDC-30603-P, May, 1984 and
- to letter L. Lucas to R.E. Lessly, " Power Update BOP Study Report," June 18, 1984.
2.
ASME Boiler and Pressure Vessel Code, Nuclear Vessels,Section III, maximum allowable internal pressure is 62 psig.
3.
Staff Safety Evaluation of DAEC, USAEC, Directorate of Licensing, January 23, 1973.
f m [ [ c f Y V~% O,m w ~fs m m p~,
4.
-1
-Part-50 -Appendi:: J, Reactorhainment Testing
- T'i
*s 3
e Delg g y rede. el ".cgister, 7.pil 19, 1975.
5.
De1It 6.
Deleted 7.
General Electric Company, Duane Arnold Enerav Center Suoeression Pool Temperature Response. NEDC-22082-P, March 1982.
i 4
l AMEnDMEur no. A,201 3.7-42 i
l
(?TS-2G1 i2/95
DAEC-1 d.
The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
6.11.3 UNIOUE REPORTING REOUIREMENTS Special reports shall be submitted to the Director of Inspection and Enforcement Regional Office within the time period specified for each j
report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification.
a.
Reactor vessel base, weld ar.d heat aft'ected zone metal te2t specimens (Specification 4.6.A.2).
l b.
deleted c.
I e
s ection (Specification 4.6.G.).
l d.
see er Ce.,tei,..uc,t :,,te;, c.ted Leekage Cete Test (Cpeeificetier, 4.7. A).
e.
deleted 1
f.
deleted g.
deleted l
h.
Radioactive Liquid or Gaseous Effluent - calculated dose exceeding specified limit (00AM Sections 6.1.3, 6.2.3 and 6.2.4).
i 1.
Off-Gas System inoperable (ODAM Section 6.2.5).
j.
Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of ODAM Table 6.3-3 when averaged over any calendar quarter sampling period (ODAM Section 6.3.2.1).
i k.
Annual dose to a MEMBER OF THE PUBLIC determined to exceed 40 CFR Part 190 dose limit (00AM Section 6.3.1.1).
1.
Radioactive liquid waste released without treatment when activity concentration is equal to or greater than 0.01gci/ml (ODAM Section 6.1.4.1).
m.
Explosive Gas Monitoring Instrumentation Inoperable (Specification 3.2.I.1).
n.
Liquid Holdup Tank Instrumentation Inoperable (Specification 3.14.B.1).
AMENDMENT NO. Ml6,JD5,JD9,201 6.11-5 1
RTs-zd7
/ 2 / '75 l
i DAEC-1 TABLE 6.11-1 (cont)
REPORTING
SUMMARY
- ROUTINE REPORTS Reauirement Renort
.Timino of Submittal
$50.59(b)
Changes, Tests, Within 6 months after each and Experiments REFUELING OUTAGE.
570.53 Special Nuclear Within 30 days after March 31 Material Status and September 30 of each year.
570.54 Transfer of Special Promptly upon transfer Nuclear Material 670.54 Receipt of Special Within 10 days after Nuclear Material material is received Appendix G Fracture Toughness On an individual-case basis to 10 CFR at least 3 years prior tu Part 50 the date when the predicted fracture toughness levels will no longer satisfy section V.B. of Appendix G to 10 CFR Part 50.
Appendix H Reactor Vessel Completion of tests after to 10 CFR Material Surveillance each capsule withdrawal.
Part 50 Appendix I Annual Radioactive On or before May 1.
to 10 CFR Material Release Part 50 Report Appendix I Annual Radiological On or before May 1.
to 10 CFR Environmental Report Part 50 aaandiv 1 "0 :t:r C::;t:ina..i Appruximei.eiy 3 montns t: 10 CF",
hildin; Int:;r:ted fe!!cein; redct of test-
":rt 50 Leak Data T==+
Amendment No. 109,J79,JS4,196 6.11-7 l
g rs -201 l 2 /96 l
L______-____--
DAEC-1 6.12 Primary Containment Leakage Rate Testing Program i
A program shall be established to implement the leakage rate testing of the primary containment as required by 10 CFR 50.54 (o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163," Performance Based Containment Leak-Test Program," dated September 1995.
The peak calculated containment internal pressure for the design basis loss of coolant accident, P;, is 43 psig.
The maximum allowable primary containment leakage rate, L,, at P, shall be 2.0 %
of primary containment air weight per day.
Leakage Rate acceptar.ce criteria are:
Primary Containment leakage rate acceptance criterion is s 1.0 L,.. During 1
a.
the first startup following testing in accordance with this program, the leakage rate acceptance criteria are: s 0.60 L, for the Type B and Type C tests; and, s 0.75 L, for the Type A tests; b.
The air lock testing acceptance criterion is overall air lock leakage rate s 0.05 L, when tested at 2P,.
The 25% extension, per definition # 26 for Surveillance Frequency, does not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.
6.12-1 RTS -2 M t z/95
IRTS-269 to NG-95-2985 j
Page 1 of 2 i-7 SAFETY ASSESSMENT 1.
INTRODUCTION
'By letter dated December 22,1995, IES Utilities Inc. submitted a request for revision to the Technical Specifications for the Duane Arnold Energy Center (DAEC). The proposed change adopts the guidance provided in NUREG 1433, Improved Standard Technical Specifications (ITS), for the performance of tests in accordance with the Primary Containment Leakage Rate Testing Program.
2.
ASSESSMENT Information already contained in 10 CFR 50, Appendix J was deleted and references to the Primary Containment Leakage Rate Testing Program were added. These are administrative changes to allow the use of performance based containment leakage testing methods. The proposed amendment requires compliance with the regulatory requirements of 10 CFR 50, Appendix J, Option B. Any excmptions to the requirements of 10 CFR 50, Appendix J require prior NRC approval, i
The proposed Technical Specification change does not involve any change to the configuration or method of operation of any plant equipment that is used to mitigate the consequences of an accident, nor do they affect any assumptions or conditions in the accident analysis. No changes in either plant design or operational strategies will be made as a result of this revision. The use of Option B will significantly reduce the frequency ofleak testing for highly reliable components provided their performance remains acceptable. This will result in reduced occupational radiological exposure, while at the same time assuring the performance of the containment safety functions as a barrier to the release of radioactivity to the environment. The addition of drywell air lock surveillance requirements provides further assurance that primary containment integrity will be maintained.
The proposed revision does not involve any change to the configuration or method of operation of any plant equipment that is used to mitigate the I
consequences of an accident, nor does it afTect any assumptions or conditions in any of the accident analysis. The proposed revision does not degrade any existing plant programs, nor modify any functions of safety related systems or accident mitigation functions DAEC has previously been credited with. The proposed changes do not impact initiators of analyzed events. They also do not impact the l
RTS-269 to i
NG-95-2985 Page 2 0f 2 assumed mitigation of accidents or transient events. These TS changes will not alter assumptions made in the safety analysis and licensing basis.
The proposed changes are consistent with NUREG-1433 which was approved by the NRC Staff and with NRC guidance provided for the implementation of Option B. Therefore, revising the CTS to reflect the NRC accepted level of detail and requirements ensures no reduction in a margin of safety.
Based upon the above assessment, we conclude that this request is acceptable.
I '
RTS-269 to l
NG-95-2985 Page1of2 ENVIRONMENTAL CONSIDERATION 1
1 1
a 10 CFR Section 51.22(c)(9) identifies certain licensing and regulatory actions which are l
eligible for categorical exclusion from the requirement to perform an environmental i
assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration; (2) result in a f
significant change in the types or significant increase in the amounts of any effluents that
)
may be released offsite; and (3) result in a significant increase in individual or cumulative occupational radiation exposure.
IES Utilities Inc. has reviewed this request and determined that the proposed amendment j
meets the eligibility criteria for categorical exclusion set forth in 10 CFR Section 51.22(c)(9). Pursuant to 10 CFR Section 51.22(b), no environmental impact statement or
- environmental assessment needs to be prepared in connection with the issuance of the amendment. The basis for this determination follows:
Basis The change meets the eligibility criteria for categorical exclusion set forth in 10 CFR.
Section 51.22(c)(9) for the following reasons:
1.
As demonstrated in Attachment I to this letter, the proposed amendment does not involve a significant hazards consideration.
2.
The proposed amendment includes changes which delete information already contained in 10 CFR 50, Appendix J and adds references to the Primary Containment Leakage Rate Testing Program. These are administrative changes to allow the use of performance based containment leakage testing methods. The proposed amendment requires compliance with the regulatory requirements of 10 CFR 50, Appendix J, Option B. Any exemptions to the requirements of 10 CFR 1
50, Appendix J require prior NRC approval. No change in either plant design or operational strategies will be made as a result of this revision.
Thus, there will be no significant change in the types or significant increase in the amounts of any efliuents that may be released offsite.
.. e '
+
RTS-269 to NG-95-2985 Page 2 of 2 3.
The proposed amendment includes changes which delete information already contained in 10 CFR 50, Appendix J and adds references to the Primary Containment Leakage Rate Testing Program. These are administrative changes to allow the use of performance based containment leakage testing methods. The proposed amendment requires compliance with the regulatory requirements of 10 CFR 50, Appendix J, Option B. Any exemptions to the requirements of 10 CFR 50, Appendix J require prior NRC approval. The use of Option B will significantly reduce the frequency ofleak testing for highly reliable components provided their performance remains acceptable. This will result in reduced occupational radiological exposure, while at the same time assuring the performance of the containment safety functions as a barrier to the release of radioactivity to the environment. No changes in ei$er plant design or operational strategies will be made as a result of this revision.
Thus, there will be no significant increase in either individual or cumulative occupational radiation exposure.
1