ML20057A109
| ML20057A109 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 09/08/1993 |
| From: | Dyer J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20057A110 | List: |
| References | |
| NUDOCS 9309130063 | |
| Download: ML20057A109 (12) | |
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2+i UNITED STATES
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.j NUCLEAR REGULATORY COMMISSION 1
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f WASHINGTON. D.C. 20555-0001 9..u
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1 ILLINDIS POWER COMPANY l
50YLAND POWER COOPERATIVE. INC.
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DOCKET NO. 50-461 i
CLINTON POWER STATION. UNIT NO 1 l
j AMENDMENT TO FACILITY OPERATING LICENSE l
Amendment No. 83 l
License No. NPF-62 1.
The Nuclear Regulatory Commission (the Commission) has found that: e i
A.
The application.s for amendment by Illinois Power Company' (IP),
l and Soyland Power Cooperative, Inc. (the licensees) dated i
February 17, 1993, and April 16, 1993, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the l
Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I-I B.
The facility will operate in conformity with the application, the I
provisions of the Act, and the rules and regulations of the i
j Commission; i
C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health i
and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; j
i D.
The issuance of this amendment will not be inimical to the common defer.se and security or to the health 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 applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical j
Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-62 is hereby amended to read as follows:
9309130063 930908 A
DR ADOCK 05000461 f' PDR L
1 Illinois Power Company is autnorf ed to act as agent for Soyland Power Cooperative, Inc. and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.
2 Pages 5 and 6 are attached, for convenience, for the composite license to reflect this change.
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(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 83, are hereby incorporated into this license.
Illinois Power Company shall operate the facility in accordance with the Technical Specifications an6 the Environmental Protection Plan.
i D.
The facility requires exemptions from certain requirements of 10 CFR Part 50 and 10 CFR Part 70. These include:
(a) an i
exemption from the requirements of 10 CFR 70.24 for the criticality alarm monitors around the fuel storage area; (b) an exemption from the requirement of paragraph III.D.2(b)(ii) of Appendix J, substituting the seal leakage test at Pa of paragraph III.D.2(b)(iii) for the entire airlock test at Pa of paragraphe III.D.2(b)(ii) of Appendix J when no maintenance has been performed in the airlock that could affect its sealing capability (Section 6.2.6 of SSER 6); (c) an exemption from the requirement of paragraph III.C.3 of Appendix J, exempting the measured. leakage rates from the main steam isolation valves from inclusion in the combined leak rate for the local leak rate tests (Section 6.2.6 af SSER 6); (d) an exemption from the requirements of paragraph III.B.3 of Appendix J, exempting leakage from the valve packing and the body-to-bonnet seal of valve IE51-F374 associated with containment penetration IMC-44 from inclusion in the combined leakage rate for penetrations and valves subject to Type B and C t
tests; and (e) an exemption from the requirement of paragraph III.D.I.(a) to conduct the third Type A test of each 10-year i
service period when the plant is shut down for the 10-year plant i
inservice inspections. The special circumstances regarding each exemption, except for Items (a) and (d) above, are identified in i
the referenced section of the safety evaluation report and the supplements thereto.
An exemption was previously granted pursuant to 10 CFR 70.24.
The exemption was granted with NRC materials license No. SNM-1886, issued November 27, 1985, and relieved IP from the requirement of having a criticality alarm system.
IP is hereby exempted from the criticality alarm system provision of 10 CFR 70.24 so far as this section applies to the storage of fuel assemblies held under this license.
The special circumstances regarding the exemption identified in Item (d) above as identified in the safety evaluation accompanying Ament cent No. 62 to this license. The special circumstances rega. ding the exemption identified in Item (e) above are identified in the safety evaluation accompanying Amendment No. 83 to this license.
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These exemptions are authorized by law, will not present an undue i
risk to the public health and safety, and are consistent with.the l
common defense and security. The exemptions in items (b) and (c) j above are granted pursuant to 10 CFR 50.12. With these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the i
provisions of the Act, and the rules and regulations of the i
Commission.
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3.
This license amendment is effective as of its date of issuance,.to be '
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implemented during the fourth refueling outage.
f FOR THE NUCLEAR REGULATORY COMMISSION i
e f/AA/
James E. Dyer, Directo i
Project Directorate III-2 j
Divisi m of Reactor Projects.- III/IV/V Office of Nuclear Reactor Regulation I
Attachment:
1.
License pages 5 and 6 2.
Changes to the Technical Specifications j
Date of Issuance:
September 8,'1993 i
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(8)
Post-Fuel loadina Initial Test Procram (Section 14. SER. SSER 5 i
and SSER 6) i i
Any changes to the initial test program described in Section 14 of i
the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of j
such change.
a l
l (9)
Emeroency Response Capabilities (Generic letter 82-33. Sucolement 1
)
to NUREG-0737. Section 7.5.3.1. SSER 5 and SSER 8. and Section 18.
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SER. SSER 5 and Safety Evaluation Dated April 17. 1987) i I
a.
IP in accordance with the commitment contained in a letter dcted December 11, 1986,. shall install and have operational i
separate power sources for each of the fuel zone level l
channels as provided for in Regulatory Guide 1.97 prior tos i
startup following the first refueling outage.
b.
IP shall submit a detailed control room design final supple' mental summary report within 90 days of issuance of the. full power license that completes all the rer-ining items identified in Section 18.3 of the Safety Evaluation dated
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April 17, 1987.
D.
The facility requires exemptions from certain requirements of 10 CFR Part 50 and 10 CFR Part 70. These include:
(a) an exemption from the i
requirements of'10 CFR 70.24 for the criticality alarm monitors around a
the fuel storage area; (b) an eu: ption from the requirement of paragraph III.D.2(b)(ii) of Appendix J, substituting the seal leakage test at Pa of paragrart III.D.2(b)(iii) for the entire airlock test at i
Pa of paragraph III.D.2(b)(ii) of Appendix J when no maintenance has been performea in the airlock that could affect its sealing capability (Section 6.2.6 of SSER 6); (c) an exemption from the requirement of paragraph III.C.3 of AprMix J, exempting the measured leakage rates a
from the main steam isi 4
- valves from inclusion in the combined leak i
1 i
rate for the local leak C asts (Secticn 6.2.6 of SSER 6); (d) an l
exemption from the require % Ls of paragraph III.B.3 of Appendix J, i
exempting leakage from the valve packing and the body-to-bonnet seal of valve IE51-F374 associated with containment penetration IMC-44 from inclusion in the combined leakage rate for penetrations and valves j
subject to Type B and C tests; and (e) an exemption from the requirement l
of paragraph III.D.I.(a) to conduct the third Type A test of each 10-year service period when the plant is shut down for the 10-year plant inservice inspections. The special circumstances regarding each exemption, except for Items (a) and (d) above, are identified in the referenced section of the safety evaluation' report and the supplements thereto.
Amendment No. 83
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An exemption was previously granted pursuant to 10 CFR 70.24. The exemption was granted with NRC materials license No. SNM-1886, issued November 27, 1985, and relieved IP from the requirement af having a criticality alarm system.
IP is hereby exempted from the criticality alarm system provision of 10 CFR 70.24 so far as this section applies to the storage of fuel assemblies held under this license.
The special circumstances regarding the exemption identified in Item (d) above are identified in the safety evaluation accompanying Amendment No. 62 to this license. The special circumstances regarding the exemption identified in Item (e) above are identified in the safety evaluation accompanying Anendment No. 83 to this license.
These exemptions are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security.
The exemptions in items (b) and (c) above are granted pursuant to 10 CFR 50.12. With these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.
i E.
The licensees shall fully implement and maintain in effect all provisions of the Commission-approved physical security plan, guard -
training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments
- and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).
The plans, which contain Safeguards Information protected under 10 CFR i
73.21, are entitled:
"Clinton Power Station Physical Security Plan,"
with revisions submitted through May 27, 1993; "Clinton Power Station Training and Qualification Plan," with revisions submitted through May 27, 1993; and "Clinton Power Station Safeguards Contingency Plan,"
I with revisions submitted through May 27, 1993.
Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.
F.
IP shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report as amended, for the Clinton Power Station, Unit No. 1, and as approved in the Safety Evaluation Report (NUREG-0853) dated February 1982 and i
Supplement Nos. I thru 8 thereto subject to the following provision:
IP may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
G.
Except as otherwise provided in the Technical Specifications or Environmental Protection Plan, IP shall report any violations of the requirements contained in Section 2.0 of this license in the following manner:
initial notification shall be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the NRC Operations Center via the Emergency Notification System with written followup within thirty days in accordance with the procedures described in 10 CFR 50.73(b), (c), and (e).
Amendment No. 83
ATTACHMENT TO LICENSE AMENDMENT NO. 83 FAClllTY OPERATING LICENSE NO. NPF-62 i
DOCKET t/0. 50-461 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages.
The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
The corresponding overleaf pages, as indicated by asterisk, are provided to maintain document completeness.
Remove Paoes Insert Paaes
- 3/4 6-1
- 3/4 6-1 3/4 6-2 3/4 6-2 3/4 6-3 3/4 6-3 3/4 6-4 3/4 6-4 B 3/4 6-1 B 3/4 6-1
- B 3/4 6-2
- B 3/4 6-2 i
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3 /4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT PRIMARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be maintained.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2*, and 3.
ACTION:
Without PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY within I hour or be in at least HOT SHUTDOWN within the next 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 />.
SURVEILLANCE RE0VIREMENTS 4.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be demonstrated:
After each closing of each penetration subject to Type B testing, except a.
the primary containment air locks, if opened following Type A or B test, by leak rate testing the seals with gas at Pa, 9.0 psig, and verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Surveillance Requirement 4.6.1.2.d for all other Type B and C penetrations, the combined leakage rate is less than or equal to 0.60 La.
b.
At least once per 31 days by verifying that all primary containment l
penetrations ** not capable of being closed by OPERABLE contair. ment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in positien, except as provided in Specification 3.6.4.
l By verifying each primary containment air lock is in compliance with the c.
requirements of Specification 3.6.1.3.
d.
By verifying the suppression pool is in compliance with the requirements of Specification 3.6.3.1.
l
- 5ee Special Test Exception 3.10.1 1
- Except valves IHG016 and 1HG017 and valves, blind flanges, and deactivated automatic valves which are located inside the primary containment, steam tunnel, or drywell, and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD j
SHUTDOWN except such verifi:ation need not be performed more often than once per 92 days.
CLINTON - UNIT 1 3/4 6-1 Amendment No. #6, 68 DEC 2 41992
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Primary containment leakage rates shall be limited to:
f a.
An overall integrated leakage rate of less than or equal to La, 0.65% by l
weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at Pa, 9.0 psig.
l b.
A combined leakage rate of less than or equal to 0.60 La, for all penctra-l l
tions and all valves subject to Type B and C tests when pressurized to l
Pa, 9.0 psig.
c.
Less than or equal to 28 scf per hour for any one main steam line through the isolation valves when tested at Pa, 9.0 psig.
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d.
A combined leakage rate of less than or equal to 0.08 La, for all penetra-l l
tions that are secondary containment bypass leakage paths when pressurized to Pa 9.0 psig.
e.
A combined leakage rate of less than or equal to 1 gpm times the total number of containment isolation valves in hydrostatically tested lines which penetrate the primary containment, when tested at 1.10 Pa, 9.9 psig.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2**, and 3.
ACTION:
With:
The measured overall integrated primary containment leakage rate a.
exceeding 0.75 La, or b.
The measured combined leakage rate for all penetrations and all valves subject to Type B and C tests exceeding 0.60 La, or c.
The measurad leakage rate exceeding 28 scf per hour for any one main steam line through the isolation valves, or d.
The combined leakage rate for all penetrations which are secondary containment bypass leakage paths exceeding 0.08 La; or e.
The measured combined leakage rate for all containment isolation valves in hydrostatically tested lines which penetrate the primary containment exceeding 1 gpm times the total number of such valves, restore:
- Exemption to Appendix J of 10 CFR 50.
,,See Special Test Exception 3.10.1.
I CLINTON - UNIT 1 3/4 6-2 Amendment No. 83
i CONTAINMENT SYSTEMS l
PRIMARY CONTAINMENT LEAKAGE i
LIMITING CONDITION FOR OPERATION (Continued)
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3.6.1.2 ACTION (Continued):
l a.
The overall integrated leakage rate (s) to less than or equal to 0.75 La, I
and b.
The combined leakage rate for all penetrations and all valves subject to l
Type B and C tests to less than or equal to 0.60 La, and L
c.
The leakage rate to less than 28 scf per hour for any one main steam line through the isolation valves, and i
I d.
The combined leakage rate for all penetrations which are secundary i
j containment bypass leakage paths to less than or equal to 0.08 La, and r
The combined leakage rate for all containment isolation valves in hydr,o-e.
statically tested lines which penetrate the primary containment to less than or equal to 1 gpm times the total number of such valves i
i prior to increasing reactor coolant system temperature above 200*F.
SURVEILLANCE RE0VIREMENTS 4.6.1.2 The containment leakage rates shall'be demonstrated at the following
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test schedule and shall be determined in conformance with the criteria speci-fied in Appendix J of 10 CFR 50 using the methods and provisions of ANSI N45.4-1972 and BN-TOP-1 and verifying the result by the Mass Point Methodclogy described in ANSI /ANS N56.8-1981.
a.
Three Type A Overall Integrated Containment Leakage Rate tests shall be conducted at 40. 10 month intervals during shutdown at Pa, 9.0 psig l
during each 10-year service period.
b.
If any periodic Type A test f ails to meet 0.75 La the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission.
If two consecutive Type A tests fail to meet 0.75 La a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 La at which time the above test schedule may be resumed.
c.
The accuracy of each Type A test shall be verified by a supplemental test which:
1.
Confirms the accuracy of the test by verifying that the difference between the supplemer.tal data and the Type A test data is within 0.25 La. The formula to be used is :
[Lo + Lam - 0.25 La] s Lc s
[Lo + Lam + 0.25 La] where Lc - supplemental test result, Lo =
superimposed leakage and Lam - measured Type A leakage.
CLINTON - UNIT 1 3/4 6-3 Amendment No. 83
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CONTAINMENT SYSTEMS PRIMARY CONTAINMENT LEAKAGE SURVEILLANCE REQUIREMENTS (Continued)
I 4.6.1.2 (Continued) 2.
Has duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test.
3.
Requires the quantity of gas injected into the primary containment or bled from the primary containment during the supplemental test to l
be between 0.75 La and 1.25 La.
d.
Type B and C tests shall be conducted ***," with gas at Pa, 9.0 psig, at l
intervals no greater than 24 months except for tests involving:
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1.
Air locks, i
2.
Main steam line isolation valves, s
3.
Penetrations using continuous leakage monitoring systems, 3
4.
All containment isolation valves in hydrostatically tested lines-which penetrate the primary containment, and 5.
Purge supply and exhaust isolation valves with resilient material seals.
Air locks shall be tested and demonstrated OPERABLE per Surveillance e.
Requirement 4.6.1.3.
f.
Main steam line isolation valves shall be leak tested with gas at Pa, 9.0 psig, at least once per 18 months.
g.
Type B tests for genetrations employing a continuous leakage monitoring system shall be conducted at Pa, 9.0 psig, at every other reactor shutdown for refueling, but in no case at intervals greater than 3 years, h.
All containment isolation valves in hydrostatically tested lines which penetrate the primary containment shall be leak tested at 1.10 Pa, i
9.9 psig, at least once per 18 months.
i.
Purge supply and exhaust isolation valves with resil;en;..rterial seals shall be tested and demonstrated OPERABLE per Surveillance Require-ment 4.6.1.8.3.
j.
The provisions of Specification 4.0.2 are not applicable to Specifica-tions 4.6.1.2.a, 4.6.1.2.b, 4.6.1.2.d, and 4.6.1.2.g.
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- Except as provided in NRC-approved exemption to Appendix J to 10 CFR 50 for centainment penetration IMC-44.
"The leakage rate for containment penetration IMC-44 is not required to be determined until startup from the fifth refueling outage in accordance with an approved exemption to Appendix J of 10 CFR 50.
CLINTON - UNIT 1 3/4 6-4 Amendment No. 83
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3.4.6 CONTAINMENT SYSTEMS l
BASES
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3/4.6.1 PRIMARY CONTAINMENT l
i 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive materials fra. the containment atmosphere will be restricted to those leakage-i i
i paths and associated leak rates assumed in the accident analyses.
This restriction, in conjunction with the leakage rate limitation, will limit the j
site boundary radiation doses to within the limits of 10 CFR Part 100 during i
accident conditions.
3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE l
The limitations on containment leakage rates ensure that the total. containment l
leakage volume will not exceed the value assumed in the accident analyses >t the peak accident pressure of 9.0 psig, Pa. As an added conservatism, the i
measured overall integrated leakage rata is further limited to less than o{
equal to 0.75 La during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage.
tests.
Operating experience with the main steam line isolation valves has indicated j
that degradation has occasionally occurred in the leak tightness of the valves; therefore the special requirement for testing these valves.
The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix J to 10 CFR 50 (with the exception of exemption (s) 3 granted by the NRC).
In addition to exemptions related to testing of individual components, the following exemption has been granted from the requirements of Appendix J of 10 CFR 50:
Section III.D.I.(a) - an exemption that rerroves the requirement that the third Type A test for each 10-year service period be conducted when the plant is shut down for the 10-year plant inservice inspection.
(Reference NRC letter dated September 7,1993.)
4 3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and the containment leakage rate given in Specifications 3.6.1.1 and 3.6.1.2.
The specification makes allowance for the fact that there may be long periods of time when the air locks will be in a closed and secured position during reactor operatior:. Only one closed door in each air lock is required to j
maintain the integrity of the containment.
The surveillance testing for measuring leak rate for the containment air locks is consistent with the requirements of Appendix J to 10 CFR 50 with the excep-tion of exemption (s) granted for the containment air lock leak testing.
CLINTON - UNIT 1 B 3/4 6-1 Amendment No. 83 i
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CONTAINMENT SYSTEMS BASES i
3/4.6.1.4 MSIV LEAKAGE CONTROL SYSTEM l
1 Calculated doses resulting from the maximum leakage allowance for the main steam line isolation valves in the postulated LOCA situations would be a small l
l fraction of the 10 CFR 100 guidelines, provided the main steam line system from the isolation valves up to and including the MSIV-LCS motor operated boundary valve remains intact. Operating experience has indicated that degradation has occasionally occurred in the leaktightness of the MSIV's such that the specified leakage requirements have not always been maintained continuously. The requirement for the leakage control system will reduce the untreated leakage from the MSIV's when isolation of the primary system and containment is required.
3/4.6.1.5 CONTAINMENT STRUCTURAL INTEGRITY j
This limitation ensures thzt the structural integrity of the containment 4111 be maintained comparable to the original design standards for the life of the
-i unit. Structural integrity is required to ensure that the containment wiJl withstand the maximum pressure of 15 psig in the event of a steam line break accident.
A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.
3 /4.6.1. 6 CONTAINMENT INTERNAL PRESSURE l
The limitations on containment to secondary containment differential pressure i
ensure that the containment peak calculated pressure of 9.0 psig does not exceed the design pressure of 15.0 psig during design basis steam line break j
conditions or that the external pressure differential does not exceed the design maximum external pressure differential of 3.0 psid. The limit of -0.25 i
to +0.25 psid for initial containment to secondary containment pressure will limit the containment pressure to 9.0 psid which is less than the design pressure and is consistent with the safety analysis for containment tiesign pressure.
3 /4. 6.1. 7 PRIMARY CONTAINMENT AVERAGE AIR TEMPERATURE The limitation on containment average air temperature ensures that the containment peak air temperature does not exceed the design temperature of 185'F during steam line break conditions and is consistent with the safety analysis.
CLINTON - UNIT I B 3/4 6-2 revised by letter dated 5/21/93 3 474M M#[
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