ML20090L124
| ML20090L124 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 03/11/1992 |
| From: | Shiraki C Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20090L128 | List: |
| References | |
| GL-91-08, GL-91-8, NUDOCS 9203200075 | |
| Download: ML20090L124 (17) | |
Text
- -
f
[pS CE@
g
- o UNITED STATES
!" s m [,h NUCLEAR REGULATORY COMMISSION 7 /
t W ASHINGTON, D. C. 20565 Q..U
/
IOWA [1ECTRIC TIGHT AND POWER COMPANY CENTRAL IOWA POWER [00PERATlyL
[0RN @lLT PQEB _(_00 PERATIVE DOCKET l(Qu 50QU QVANE ARNOLD ENERGLCQ{1[B MENDMENT TO FALlilTY OPERATING LICENSE Amendment No. 181 License No. DPR-49 1.
The Nuclear Regulatory Commission (the Commission) ha.1 found that:
A.
The application for amendment by Iowa Electric Light and Power Company, et al., dated September 20, 1991 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will oparate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
Tiiere is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and sa fety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the healtn and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all aPr!icable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-49 is hereby amended to read as follows:
9203200075 920311 PDR ADOCK 05000331 P
e
.m...
V (2)
Itchnical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.181, are hereby incoraorated in the license.
The licensee shall operate the facility ' n accordance with the Technical Specifications.
3.
The license amendment is effective as of the date of issuance and shall be implemented within 60 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 4
Cly e Y. Shi akt, Sr. Project Manager Project Directorate 111'3 Division of Reactor Projects Ill/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: March 11, 1992 l
l l
l
\\
l t
m-
~.
ATTACHMENT TO LICENSE AMENDMENT NO. 181 FACillTY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.
The revised areas are indicated by marginal lines.
Remove Insert vi vi 3.2-36a 3.2.-36a 3.2-38 3.2-38 3.2-39 3.2-39 3.7-2 3.7-2 3.7-5 through 3.7-7 3.7-5 through 3.7-7 3.7-18 through 3.7-29a 3.7-18 through 3.7 90 3.7-38 3.7-38 3.7-47 3.7-47 3.7-48 3.7-48 1
t l
L
M DAEC 5 TECHNICAL SPECIFICATIONS LISTOFTABLES(Continued)
Table Numbey Title Page 3.7-1 Deleted Deleted
- 3. 7-2 3.7-3 Deleted 4.7-1 Summary Table of New Activated Carbon Physical Properties 3.7-50 4.10-1
-Sunrnary Table of New Activated Carbon Physical Properties 3.10-7 3.12-1 Deleted 3.12-2 Deleted 3.13-1 Fire Detection Instruments 3.13-11 3.13-2 Required Fire Hose Stations 3.13 3.14-1 Radioactive Liquid Effluent Monitoring Instrumentation 3.14-5 4.14 Radioactive Liquid Effluent Monitoring Irutrumentation Surveillance Requirements 3.14-7 4.14-2 Rndioactive Liquid Waste Sampling and Analysis Program 3.14-9 3.15-1 Radioactive Gaseous Effluent Monitoring Instrumentation 3.15-7 4.15-1 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 3.15-9 4.15-2 Radioactive _ Gaseous Waste Sampling and Analysis Program 3.15-11 3.16-1_
Radiological, Environmental Monttoring Program 3.16-6 3.16-2 Maximum Values of the Lower Limit of Detection for Environmental Sample Analysis 3.16-8 3.16-3' Reporting Levels for Radioactivity Concentrations in Environmental Samples 3.16-10
- 6. 2-1_
Minimum Shift Crew Personnel and License Requirements 6.2-3 6.9-1
. Del eted -
6.11-1:
Reporting Summary - Routine Reports-6.11-8 6.11.-2
-Deleted 6.11-3a Semiannual Radioactive Material Release Report Liquid Effluents 6.11-10 6.11-3b Semiannual Radioactive Material Release Report Gaseous Effluents 6.11-11 e
i L
vi Amendment No. JJ/,JJ7,JEJ,NI#,181
~.... -
DAEC-1 The low water level instrumentation set to trip at 170" above the top of the active fuel closes all isolation valves except those in Groups 1, 6, 7 and 9, For valves which isolate at this level this trip setting is Y
F l
l Amendment No. 59,181 3.2-36a L
l
.. ~. -.., ~
~..
L*
i DAEC-1 1
1 The high drywell pressure instrumentation is a diverse signal for malfunctions to the water leve1' instrumentation and in addition to initiating CsCS, it causes isolation of Group 2 and 3 isolation valves.
For the breaks discussed above, this instrumentation will generally initiate Cscs operation before the low-low-low water level instrumentation; thus the results given above are applicable here also. The water level instrumentation initiates protection l
1 for the full spectrum of logs-of-coolant accidents and causes isolation of all isolation valves except Group 6.
Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. The primary function of the instrumentation is to detect a break in tho main steam line.
For the worst case accident, main steam line break.outside the drywell, a trip setting of 140% of rated steam flow in conjunction with the flow limiters and consequently main steam line valve closure, limits the mass inventory loss such that fuel is not-uncovered, fuel clad temperatures peak at approximately 2000*F and release of
- radioactivity to the environs is below 10 CTR 200 guidelines.
Reference-Subsection 15.6.5 of the-Updated TSAR.
Amendment No. JM,181 3.2-38 g
wev w-
=)r-w m
4 re -<er w-*
4w-=
ae-s-s
- r
+.
0 l
.n DAEC-1 Temperature monitoring instrumentation is provided in the main steam line tunnel'and turbine building to detecc leaks in this area.
Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves.
The setting is 200'F for the main steam line tunnel detector.
For large breaks, the high steam flow instrumentation is a backup to the temperature instrumeatation.
High radiation monitors in the main steam line tunnel have been provided to detect groes fuel failure as in the control rod drop accident. With the established setting of 3 times normal background, and main steam line isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines are not exceeded for this accider.t.
For the performance of a Hydrogen Water Chemistry pre-implementation test, the scram setpoint may be changed based on a calculated value of the radiation level expected during the test.
Hydrogen; addition will result in an approximate one-to five-fold increase in the-nitrogen (N-16) activity in the steam due to increased N-16
~
carryover.it.~the main steam. Reference subsection 15.4.7.of the Updated FSAR.
Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steam line' pressure drops below 850 psig.
~
The Reactor Pressure Vessel thermal transient due to an inadvertent opening of the turbine bypass valves when not-in the RUN Hode is less severe than the loss of
-feedwater analyzed in subsection 15.6.3 of the Updated FSAR, therefore, closure of the Main Steam Isolation valves for thermal transient protection when. not in RUN Hode is not required.
\\-
l Amendment No. JJf,J27,181 3.2-39 L
l
,,,,,s_.
~_
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILIANCE REQUIREMENT (3)
The reactor shall be scrammed from any operating condition if the pool temperature reaches 110'F. Power operation shall not be resumed until the pool tLmperature is reduced below the normal power operation limit specified in (1) above.
(4)
During reactor isolation conditions, the reactor shall be depressurized to less than 200 psig at normal cooldown rates if the pool temperature reaches 120'F.
2.
Primary containment integrity shall 2.
The primary containment integrity be maintained at all times when the shall be descnstrated as f ollows:
reactor is critical or when the temperature is above 212'F.and fuel a.
Type A' Test is in the reactor vessel except while performing low power physics Primary Reactor Containment tests at atn.ospheric pressure at Integrated Leakage Rate Test power levels not to exceed 5 Mw(t).
Compliance with Subsection 3.7.D.2 1)
The interior surfaces of the satisfies the requirement to drywell and torus shall be maintain primary containment visually inspected each operating integrity.
cycle for evidence of deterioration.
In addition, the external surfaces of the torus below the water level shall be inspected on a routine basis for evidence of torus corresion or
. leakage.
Except for the initial Type A test, all Type A tests shall be priormed without any preliminary leak detection surveys and leak repairs immediately prior to the test.
If a Type A test is completed but the acceptance criteria cf Specification 4.7.A.2.a.(9) is not satisfied and repairs are necessary, the Typs A test need not be repeated provided locally measured leakage reductions, achieved by repairs, reduce the containment's overall measured leakage rate sufficiently to meet the acceptance criteria.
Amendment No. 108, 181
- 3. 7- ?.
--,_,s_,
a.
n,,
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT b.
Type B Tests i
Type B tests refer to penetrations with gasketed seals, expansion bellows or other type of resilient seals.
l 1)
Test Pressure All Type B tests shall be performed by local pneumatic pressuragation of the containment penetrations, either individually or in groups, at a pressure not less than Pa.
2)
Acceptance Criteria The combined leakage rate of all penetrations subject to Type B and C tests shall be leer than 0.60 La.
c.
Type C Testa 1)
Type C tests shall be performed on containment isolation valves.
Each valve to be tested shall be -
closed by normal operation and without any preliminary exercising or adjustments.
2)
Acceptance criteria The combined leakage-rate for all penetrations subject to Type B and C teste shall be less than-0.60 La.
3)
~ The leakage from any one main steam isolation valve'shall not exceed 11.5 scf/hr at an. initial test pressure of 24'psig, 4)
The leakage rate from any
> containment isolation valve whose seating surface remains water covered post-LOCA, and which is hydrostatically Type C tested,.
shall be included-in the Type C test total.
l-l=
l'-
Amendment No. J06, 181 3.7-5
m 4.
?'
DAEC-1 LIMITING CONDITION FOR OPERATION SURVdILI.ANCE REQUIREMENT d.
Po,riodie Rotent Schedule 1)
Type A Test Af ter the preoperational 3eakage rate tests, a set of three Type A tests shall be performed5 at approximately equal intervals during each 10 year service period.
(These intervals may be extended up to eight months if necessary to coincide with refueling outages.) The third.
test of each set shall be conducted when the plant is shut down for the 10-year plant in-service irspections.
The performance of-Type A tests shall be-limited to periods when the plant facility is nonoperational and secured in the shutdown condition-under administrative control and in accordance with the plant safety procedures.
2)
Type B Tests a)
Penetrations and seals of this type (except air locks) 6 hall be leak tested at greater than or equal to 43 psig (P,) during each reactor shutdown for major fueling or other convenient interval but Lin no case at intervals greater than two years.
l b)'
The personnel airlock shall-be pressurized to greater than or equal to 43 psig (P ) and leak tested at.least once every six (6)-
months. This test interval may be extended to the next refueling outage (up to a maximum interval between P tests of 24 months) provided there have been no airlock openings since the last successful test at P,.
L
' ment No. ~ JJE,/Jpp,181 3.7-6
DAEC-1
. LIMITING CONDITION FOR CPERATIC.,
SURVEILLANCF REQUIRDtENT 4
3)
Type C Tests Type C tests shall be performed during each reactor shutdown for major refueling or other convenient interval but in no case at intervals greater than two years.
l 4)
Additional Periodic Tests Additional purge system isolation valve leakage integrity testing shall be performed at least once every three months in order to detect excessive leakage of the-purge isolation valve resilient osats.
Th. purge systea 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),
e.
Seal Replacement & Mechanical Limiter The T-ring inflatable seals for purge isolation valves CV-4300, CV-4301, CV-4302, CV-4303, CV-4306, CV-4307 and CV-4308 nhall be replaced at intervals not to exceed four years.
l During Type C testing, it shall.be-verified that the mechanical modification which limits the maximum opening angle for purge isolation valves CV-4300, CV-4301, CV-4302, CV-4303, CV-4306, CV-4307 and CV-4308 is intact.
The baseline for this requirement shall be established during the Cycle 6/7 refuel outage.
f.
Containment Hodification l
Any major modificution, replacement of a component which is part of the primary reactor containment boundary, or ressaling a seal-welded door, performed after the preoperational-leakage rate test shall be followed by either a Type A, Type 3, or Type C test, as applicable, for the area affected by the modification.
Amendment No. 100,7f0, 181 3.7-7
DAAC-1 LIM 138NG CDND3 TION FOR OPERAT;DN
$URVE!L1JJICI REQg!RkNENT '
+,
maintain the remainder of the secondary containment at 1/4 inch of water negative pressure under calm wind conditions.
2.
If specification J.7.c.1 cannot be mets a.
suspend reactor building fusa cask and irradiated f uv1 movement, and b.
Restore secondary containmer.e integrity withLn one court or, c.
Se 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 />.
D.
Primary containment Power ooetated D.
Prinary containttent rever operated Tiolatton Valveg 2solatten Valves 1.
During reactor power operating 1.
The prLeary contdinment isolation conditions, all prLmary contair. ment valves surveillance shall be i
isolation valvs: and all instrument performed as f ollows:
line flow check valves shall be CPERABLE except J specified in 4.
At least once per operating cycle 3.7.D.2.
the CPERA3LI isolation valves
- l that are power operated and automatically initiated shall be tested for simulated automatic initiation and closure times.
b.
At least once por quarter:
1)
All normally open power operated isolation valves ** shall be fully l
closed and rsopened.
2)
With the reactor power less than 754, trip main steam isolation valves individually and verify closure tLae.
c.
At least once per operatLng cycle the operabilits of the reactor coolant system (notrument line flow check valves shall be verified.
- Due to operation ilmitations, the Main Steam Line Isolation valves are exempt from subsection 4.7.D.1,a.
t
- Due to plant operational 1Laitations, the Well cooling Water supply / Return valves, Reactor Building closed cooling Water Supply / Return Valves and the Containment compressor Discharge and suction valves are exempt from the requirements of Subsection 4.7.D.1.b.
L Amendment tio. J27,J/3,J/J,181 3.7 18
9 DAEC-1 LINT.71No coND1720N FOR OPERATION SURVE!LLANCE REQUIREMENT 2.
With one or more of the primary containment isolation valves inoperable, naintain at least one isolation valve OPERABLE' or 3&OLATED** and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> eithers a.
Restore tha inoperable valve (s; to
)
OPERABLE status, or b.
1solate each affected penetration I
by use of at least one automatic valve locked or electrically deactivated in the isolated position,** cr c.
Isolate each affected penetration i
by use of at lorS* one manual valve locked in the isc>ated position or blind fiange.**
3.
If speeltication 3.7.D.1, and 5
3.7.D.2 cannot be met, an order 1) shutdown shall be initiated and toe reactor shall be in the Cold j
Shute'm'n condition within 24 ho'.rs.
I
-s e
i
- This valve may be locked or electrically
' deactivated as noted in sabsection
-3.7.D.2.b.
- ! solation valves closed to satisfy these requirements may be reopened on an intermittent basis under administrative control.
Amendment No.-7E.JA).7/E,771. 181 J.7-19
.-......a.
w j
i e
i i
i t
PAGES 3.7-20 THROUGH 3.7-29a THAT CONTA1NtD TABLES 3.7-1, 3.7-2. AND 3.7 3 I
i ARE DELETED IN THElR ENTIRETY j
i i
Next page is 3.7-30 t
a r
i f
i r
3 Amendment No.181 3.7-20
v_-.
I e
I DAEC-1 Type B and Type C tests are performed on testable penetrations and i
isolation valves during the interim period between Type A tests.
1 This provides assurance that components most likely to undergo degradation between Type A tests maintain leaktight integrity.
A controlled list of the test 6ble penetrations and isolation valves subject to Type B and Type C testing is located in the plant Administrative Control Procedures.
The containment leakage testing program is based on NRC guidelines for development of leak rate testing and surveillance schedules for I
reactorcontainment_ vessels,(Reference 4).-
5.
Drywell Interior The interiors of the drywell and suppression chamber are coated to prevent corrosio' and for ease of decontamination. The inspec-tion of the coating during each major refueling outage, t
l l
l
' Amendment No. 106, 181 3.7-38 l-
- u. _. - _. _, _. _
DArc-a l
s s
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 l process piping which penetrates containment (escept for those lines which are required to operate to mitigate the consequences of a loss-of-coolant accident) is to provide closure at a rate which will prevent core uncovery following pipa breaks outside primary containment.
A controlled list of the primary contair. ment power operated isolation valves is located an the plant Administrative Control Procedures.
In order to assure that the doses that may result f rom a steam line break are within 10 CFR 100 guidelines, it is necessary that no fuel rod perforation results from the
- accident occur prior to closure of the main steam line isolation valves.
Analyses j
indicate the fuel rod cladding perforations would be avoided for main steam valve closure times, including instrument delay, as long as 10.$ seconds.
The test closure time limit of 5 seconds for these main steam isolation valves provides sufficient margin to assure that cladding perforations are avoided.
Redundant valves in each line insure that isolation will meet the single failure criteria.
t P
Amendment No. JU,-181 3.7-47 4
4
..---,-.--,e1yra,
._,,g
.,-,..--..,,.,o_,_mm.-,-,,.-,,.y--..----,,,~..--_--,.,+_
m,,_.-.,...-_
,,m...mm-,,.-__,-,._,.-,.em,.,-.--.-e
D >.Z C-8 The main steam line isolation valves are functionally tested on a more frequent interval to establish a high degree of reliability.
The containment is penetrated by a large number of small diameter instrunent 11n6e.
The excess flow check valves in these lines shall be tested once each operating cycle.
l Containment vent / purge valves (CV-4300, ev-4301, CV-4302, CV-4303, CV-4306, CV-4307, and CV-4308) have been n>echanically codified to limit the maximum opening angle to 30 degrees.
This has been done to ensura these valves are able to close against the maximum dif ferential pressure expected to occur during a design basis accident.
The oper.ing of locked or saaled clos J :not t)
Et.
iation valves on an intermittent basis under administrativ
- 44. y s tr.caade
- the following considerationst (1) stationing an operater, oho is in cus.etant communication with nontrol room, at the valve controle, (2) instructing this operator to close there valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to riose the velves and that this action will prevent the release of radioactivity outside the containment.
Amendment No. J#3.IA6. 181 3*7~48 i
.