U-600848, Forwards Proposed Changes to Tech Specs for Full Power OL, Supplementing 870108 & 870204 Submittals.Change Deletes Word Recoupled and Inserts Word Moved,In Table 1.2,Page 1-11 Notations.Amend to License Not Requested

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Forwards Proposed Changes to Tech Specs for Full Power OL, Supplementing 870108 & 870204 Submittals.Change Deletes Word Recoupled and Inserts Word Moved,In Table 1.2,Page 1-11 Notations.Amend to License Not Requested
ML20212K823
Person / Time
Site: Clinton Constellation icon.png
Issue date: 03/03/1987
From: Hall D
ILLINOIS POWER CO.
To: Harold Denton
Office of Nuclear Reactor Regulation
References
U-600848, NUDOCS 8703090425
Download: ML20212K823 (26)
v • d • e


Text

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'lLLJN018' POWER COMPANY

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,, CLfNTON POntR STATION, P.O. BOX 678. CLINTON. fLLINOIS 61727

MAR 31987

- ' Docket No. 50-461

-Mr. Harold R. Denton,' Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.. 20555

Subject:

Clinton Power Station Technical Specifi' cations for Full Power Operating License Dear Mr. Denton n This letter supplements Illinois Power Company's (IP) letters dated January 8, 1987 (U-600785) and February 4, 1987 (U-600817) and provides additional changes that IP requests be included in the Technical Specifications (CPS-TS) which will accompany the full-power operating license for the Clinton Power, Station. These changes represent clarification and sahancements to the CPS-TS and in'part, have been discussed with your Mr. B. L. Siegel, Clinton NRC Project Manager. The justifications and the proposed marked-up phges are attached. None of these changes affect IP s ability to safely operate the Clinton

. -Power Station under its current license. Therefore, no amendment to the present low-p'ower license (NPF-55) is being rs' quested. ,

These changes to the CPS-TS'have been reviewed and are consistent in all material aspects with the FSAR as amended,

.the Safety Evaluation Report and its Supplements Nos. 1-7 and the as-built plant. An affidavit relating to this certification accompanies this letter.

If you have any questions or require additional information, please contact me.

Sinc e ours,

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D. P. all Vice President RPF/ckc Attachment 80*'

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STATE OF ILLINOIS COUNTY OF DEWITT DONALD P. HALL, Being first duly sworn, deposes and says: That he is Vice President of Illinois Power Company; that the information provided in letter U-600848 to certify that the Clinton Power Station (CPS) Technical Specifications are consistent with the CPS - Final Safety Analysis Report, the NRC Safety Evaluation Report and the as-built facility, has been prepared under his supervision and direction; that he knows the contents thereof; and that to the best of his knowledge and belief said request and the facts contained therein are true and correct.

DATED: Thisd day of J4a 1p Signed: ('DonaT&T . Hall Subscribed and sworn to before me this M day of March, 1987.

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/ Notary Public My commission expires:

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Attcchment U-600848 Page 1 of 23 Description of Change Table 1.2, page 1-11 Change Table 1.2 Notations *** to delete the word "recoupled" and insert the word " moved".

Justification The purpose of this note is to allow for testing to determine operability in accordance with Specifications 3/4.1.3.2 and 3/4.1.3.4 In addition, this change is consistent with changes made to other BWR Technical Specifications (LaSalle, River Bend, Perry, and Hope Creek).

The above change request was presented for consideration in letter U-600817 dated February 4, 1987. At the request of Mr. B. Siegel and D.

Katze, the following additional information is provided.

Movement of control rod under the controls of the one-rod-out interlock is considered as part of the design basis of operation of the reactor.

The design basis for the one-rod-out interlock is such that the single, highest worth control rod can be removed from the reactor core at any time while still preserving the required shutdown margin. Shutdown margin is demonstrated in Specification 3/4.1.1. A sufficient shutdown margin ensures that the reactor can be made suberitical from all operating conditions, the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

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Attechmtnt U-600848

~Page 2 of 23 TABL'E 1.2 r,3

  • OPERATIONAL CONDITIONS

~

MODE SWITCH CONDITION AVERAGE REACTOR POSITION COOLANT TEMPERATURE

1. POWER OPERATION Run Any temperature
2. STARTUP Startup/ Hot Standby Any temperature
3. HOT SHUTDOWN Shutdown #'*** > 200*F
4. COLD SHUTDOWN Shutdown #'##'*** 1 200*F
5. REFUELING
  • Shutdown or Refue1**'# $ 140*F M , TABLE NOTATIONS l
  1. The reactor mode switch may be placed in.the Run or Startup/ Hot Standby position to test the switch interlock functions, provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.
    1. The reactor mode switch may.be placed in the Refuel. position while a single control rod drive is being removed from the reactor pressure vessel per Specification 3.9.10.1.
  • Fuel in the reactor fully tensioned vessel or with with the the head vessel head closure bolts less than removed.
    • See Special Test Exceptions 3.10.1 and 3.10.3.

j y *** The reactor mode switch may be placed in the Refuel position while a single control rod is being(r;;;;;?:C OPERABLE.

ed provided that the one-rod-out interloc l

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4 CLINTON - UNIT 1 1-11 l

, ,. Attachment

'. U-600848 Page 3 of 23 Description of Change Table 3.3.2-2, Item 4.1, page 3/4 3-23.-  !

Change the inequality sign from 1 to f . ,

1 Justification Typographical Error.

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TABLE 3.3.2-2 (Continued)

P E CRVICS INSTRUMENTATION SETPOINTS 3

z TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE E

I Z 4. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION (Continued) i e. RCIC Equipment Room Ambient .

l Temp. - High 1 222.5'F 1 233.1*F

f. RCIC Equipment Room a Temp. - High 1 34.5'F i 40*F l g. Main Steam Line Tunnel Ambient j Temp. - High i 165'F i 176*F i
h. Main Steam Line Tunnel t

w a Temp. - High < 54.5*F

_ < 60*F s

b

w i. Main Steam Line Tunnel e

, 4 w

Temp. Timer -d5 min. 1 28 min. l i

! j. Drywell Pressure - High i 1.68 psig i 1.88 psig l k. Manual Initiation NA NA j

j 1. RHR/RCIC Steam Line Flow - High 1 179.5 in. H20** 1 188 in. H20**

m. RHR Heat Exchanger A, B Ambient Temperature - High 1 138.5*F i 149.6*F
n. RHR Heat Exchanger A, B a Temp. - High 5 74.2*F 1 79.6*F , , y c: g oo e n
5. RHR SYSTEM ISOLATION -

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! a. RHR Heat Exchanger Rooms A, B *$"

Ambient Temperature - High 1 138.5*F i 149.6 F g

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b. RHR Heat Exchanger Rooms A, B a Temperature - High 1 4.2*F 7 i 79.6*F

-s Attcchm:nt U-600848 Page 5 of 23 Description of Change Table 3.3 3-2. Item A.1.c, page 3/4 3-39 and Item B.1.c, page 3/4 3-40 Change the trip setpoint to 472 psig and allowable value range to 1 452 psig,S. 478 psig.

Justification The Instrument Setpoint Methodology Program (ISM) sponsored by the BWR Owners Group authorized General Electrical Company to perform validation calculations of the Technical Specification operating limits and instrument setpoints. The ISM calculation, for the LPCS/LPCI injection valve interlock, has determined that the instrument (Rosemount 1153B pressure transmitter) loop accuracy is significantly affected by a high environmental radiation dose equivalent to that in a LOCA. To remedy this concern, the Rosemount 1153B pressure transmitters are being replaced with Rosemount 1154R pressure transmitters. These new pressure transmitters have higher radiation tolerances and low inaccuracies due to radiation effects. The replacement of the 1 transmitters and revisions to the CPS-TS bring the CPS-TS into compliance with the ISM requirements.

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TABLE 3.3.3-2 ,

b EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS 5

E

, ALLOWABLE e TRIP FUNCTION TRIP SETPOINT VALUE A. DIVISION I TRIP SYSTEM

[

1. RHR-A (LPCI MODE) AND LPCS SYSTEM -
a. Reactor Vessel Water Level - Low Low Low, > -145.5 in.* 2 -147.7 in.

Level I

b. Drywell Pressure - High $ 1.68 psig $ 1.88 psig
c. Reactor Vessel Pressure-Low (LPCS and LPCI 452 9 78 Injection Valve Permissive) y72M psig > M sig, < ig
d. LPCI Pump A Start Time Delay Lggic Card 5 sec. 5 i 0.5 sec.-
e. LPCS Pump Discharge Flow - Low 2 875 gpm 2 750 gpm w f. LPCI Pump (A) Discharge Flow - Low,, 3 1100 gpm 3 900 gpm

) g. Manual Initiation NA NA

[ 2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "1"

  • ADS LOGIC "A" AND "E"
a. Reactor Vessel Water Level - Low Low Low, -> -145.5 in.* -> -147.7 in.

Level 1

b. Drywell Pressure - High $ 1.68 psig $ 1.88 psig
c. ADS Timer < 105 sec. < 117 sec.
d. Reactor Vessel Water Level-Low, level 3 I 8.9 in.* I 8.3 in.
e. LPCS Pump Discharge Pressure-High [145psig [125psig
f. LPCI Pump A Discharge Pressure-High 2 125 psig 2 115 psig
g. ADS Drywell Pressure Bypass Timer < 6.0 min. < 6.5 min.
h. Manual Inhibit ADS Switch NA NA
1. Manual Initiation NA NA . , ycg

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TABLE 3.3.3-2 (Continued) ..

n C EMERGENCY CORE COOLING SYSTEM ACTUATION INS 1RUMENTATION SETPOINTS 5

E ALLOWABLE

. TRIP FUNCTION TRIP SETPOINT VALUE e

5 B. DIVISION II TRIP SYSTEM H

1. RHR 8 AND C (LPCI MODE)
a. Reactor Vessel Water Level - Low Low Low, > -145.5 in.*

> -147.7 in.

Level I

b. Drywell Pressure - High 1.68 psig i 1.88 psig
c. Reactor Vessel Pressure-Low (LPCI Injection 4 72. 452 478 Valve Permissive) M sig > 469# psig, 5 d psig
d. LPCIPump(B)StartTimeDelayLogjgCard 5 sec. 5 2 0.5 sec.
e. LPCI Pump (B) Discharge Flow - Lowg > 1100 gpa > 900 gpm
f. LPCI Pump (C) Discharge Flow - Low > 1100 gpm > 900 gpm w g. Manual Initiation NA NA k

w 2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "2" 1 ADS LOGIC "B" AND "F" o

a. Reactor Vessel Water Level - Low Low Low, -> -145.5 in.* ~> -147.7 in.

Level 1

b. Drywell Pressure - High 5 1.68 psig 5 1.88 psig
c. ADS Timer < 105 sec. < 117 sec.
d. Reactor Vessel Water Level-Low, Level 3 5 8.9 in.* I 8.3 in.
e. LPCI Pump (B and C) Discharge Pressure-High [125psig. [115psig
f. ADS Drywell Pressure Bypass Timer s 6.0 min. i 6.5 min.
g. Manual Inhibit ADS Switch NA NA
i. Manual Initiation NA NA 2?R

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,. t. Att:chment U-600848 Page 8 of 23 Description of Change Specification 3/4.3.7.12, Table 3.3.7.12-1, pages 3/4 3-102, 3/4 3-103, and 3/4 3-104.

Add (#) to the Minimum Channels Operable column of Table 3.3.7.12-1, pages 3/4 3-102 and 3/4 3-103. Add note i to the Table Notations on page 3/4 3-104 as follows:

"A channel may be placed in an inoperable status for up to I hour for the purpose of performing surveillances of Specification 3/4.11.2.1 and this Specification."

Justification Sampling system down time and resultant associated monitoring system instrumentation inoperability, due to unavoidable perturbations to the system configuration, is acceptable for small periods of time. The NRC recognizes that performing sampling requirements of CPS-TS 3/4.11.2.1 will affect the system operability requirements of CPS-TS 3/4.3.7.12.

Performance of surveillances of this Specification (4.3.7.12) in accordance with the requirements of Table 4.3.7.12-1 will also result in instrumentation inoperability for small periods of time. Since performance of surveillances is the normal method of determining

" operability" there should be no penalty (i.e., entry into ACTION statement) for their performance. A one hour provision for surveillance performance is well within the ACTION limitations for appropriate systems and Specifications 3.0.3 and 3.0.4 are not applicable as referenced in this Specification.

. This change has been discussed with the NRC Messrs. J. Lee and B.

Siegel.

4

TABLE 3.3.7.12-1 .

b RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 5

INSTRUMENT MINIMUM CHANNELS OPERABLE d APPLICABILITY ACTION e

5

1. Station HVAC Exhaust PRM

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a. High-Range Noble Gas Activity 1 121 Monitor
b. Low-Range Noble Gas Activity 1 121 Monitor
  • 122
c. Iodine Sampler 1 l d. Particulate Sampler 1 122 i

$ e. Sample Flow-Rate Measuring 1

  • 123 l' w Device O
  • 2 f. Effluent System Flow Rate 1 123

, Measuring Device ,

2. Standby Gas Treatment System I txhaust PRM
    • 126
a. Medium-Range Noble Gas 1 Activity Monitor l
    • 126
b. Low-Range Noble Gas 1 l
Activity Monitor

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RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION z

e MINIMUM g INSTRUMENT CHANNELS OPERABL APPLICABILITY ACTION

2. Standby Gas Treatment System Exhaust PRM (Continued) .

Low-Range Iodine Sampler ** 122

d. 1
    • 122
e. Particulate Sampler 1 Sample Flow-Rate Measuring 1 ** 123 f.

Device .

Effluent System Flow-Rate **

g. 1 123 Measuring Device

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T 3. Main Condenser Off gas 5

Treatment System Explosive Gas Monitoring System Hydrogen Monitor *** 124

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Attschment

., U-600848 Paga 11 of 23 TABLE 3.3.7.12-1 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION TABLE NOTATIONS

  • At all times.

=ff ACTION ACTION 121 - With the number of channels OPERABLE.less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and analyzed for gross noble gas activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 122 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the channel has been declared inoperable, samples are contin-uously collected with auxiliary sampling equipment as required in Table 4.11.2-1.

ACTION 123 - With the number of channels OPERABLE less than required by the 5 Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

ACTION 124 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of the main condenser off gas treatment system may continue provided grab samples are collected at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and analyzed within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

ACTION 125 - Deleted ACTION 126 - With the number of channels OPERABLE less than required by the

! Minimum Channels OPERABLE requirement, suspend release of radio-l activity effluents via this pathway.

l cfanaeI may b< Place d 51 d'l I0'Y" 5 lclu5 (c, af k illour f" S /"@S '

of ,auf,w),19 sarveillGtKC5 d f S3eCY54bb'l j 3/ q. it. 2. / a ad His Sp ci Aca (n i .

. CLINTON - UNIT 1 3/4 3-104 I

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. Attcchment U-600848 Page 12 of 23 Description of Change Specification 4.4.3.2.1.a. page 3/4 4-11.

Delete surveillance requirement.

Justification Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," states that the source of reactor coolant leakage should be identifiable to the extent practical. Leakage detection and collection systems should be selected and designed according to Position C.5, such that leakage to the containment building from unidentified sources should be collected, and the flow monitored with an accuracy of I gallon per minute in one hour (gpm/hr.), or better.

In Section 5.2.5.10 of the CPS FSAR, " Regulatory Guide 1.45 Compliance,"

IP commits to comply with Regulatory Guide 1.45 and describes the details of that compliance. In the Clinton Safety Evaluation Report (SER), Section 5.2.5, the NRC finds the Clinton compliance to Regulatory Guide 1.45 acceptable and Item (5) of the SER states, "The sensitivity and response time of the systems . . . for the detection of unidentified leakage are adequate to detect a leakage rate of 1 gpm in less that I hour." The Lenk Detection System will comply with Regulatory Guide 1.45 as currently specified in the FSAR except that the drywell atmospheric particulate and gaseous radioactivity monitoring systems are not capable of measuring a leakage rate of 1 gpm within an hour as required by Position C.5 of Regulatory Guide 1.45. Therefore, the airborne radioactivity monitoring systems will be considered as a secondary detection method, along with the monitoring of pressure and temperature, to detect gross unidentified leakage.

On January 12, 1987 (U-600809), and February 16, 1987 (U-600840), IP submitted letters describing revisions to the Clinton Power Station Final Safety Analysis Report (FSAR). This change and the FSAR revisions have been discussed with the NRC's Messrs. J. Ridgely and B. Siegel.

Attcchment

. U-600848 Pegs 13 of 23 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.3.2 Reactor coolant system leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE.
b. 5 gpm UNIDENTIFIED LEAKAGE.
c. 25 gpm IDENTIFIED LEAKAGE (averaged over any 24-hour period).

] d. 0.5 gpm leakage per nominal inch of valve size up to a maximum of 5 gpm

from any reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1, at rated reactor pressure.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

a ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT 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 next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b. With any reactor coolant system leakage greater than the limits in b and/or c, above, reduce the leakage rate to within the limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 SHUTOOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c. With any reactor coolant system pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two other closed manual or deactivated automatic valves, 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 REQUIREMENTS 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be within each of the above limits by: i Deleted.

i Monitoring-the-drywell-atmospheric-particulate and gaseous-radle d

a. 3 activity-et-least-once-per-12 hours;- *'

j

b. Monitoring the drywell floor and equipment drain sump level and sump flow rate at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,
c. Monitoring the drywell air coolers condensate flow rate at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and CLINTON - UNIT 1 3/4 4-11

. Attachment U-600848 Page 14 of 23 Description of Change Specification 4.6.1.4.c.2, page 3/4 6-7.

Change the surveillance commensurate to the markup.

i Justification Experience gained in the testing and startup programs indicate that the present Technical Specification is burdensome to manage and implement as written. As an operational enhancement. IP proposes to change this i specification to accommodate the as-built configuration of the system.

The control logic of the outboard blowers is such that both blowers i operate at the same time. Testing the blowers on an individual basis requires that the non-testing blower be physically removed from its

foundation and a blind flange installed in place of the removed blower i- each time this surveillance test is performed. The proposed change l' allows both blowers to be tested simultaneously and as configured by the system logic and construction. This change is typical of other BWR's and the Standard Technical Specifications (NUREG-0123).

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Attcchment

. U-600848 CONTAINMENT SiSTEMS Page 15 of 23 MSIV LEAVAGE CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.4 Two independent MSIV leakage control system (LCS) subsystems shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

With one MSIV leakage control system subsystem inoperable, restore the inoper-able subsystem to OPERABLE status within 30 days 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 REQUIREMENTS 4.6.1.4 Each MSIV leakage control system subsystem shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying:
1. Blower OPERABILITY by starting the blowers from the control room and operating the blowers for at least 15 minutes.
2. Heater OPERABILITY by demonstrating electrical continuity of the heating element circuitry.
b. During each COLD SHUTDOWN, if not perforined within the previous 92 days, by cycling each remote, manual and automatic motor operated valve through at least one complete cycle of full travel,
c. At least once per 18 months by:
1. Performance of a functional test which includes simulated actuation of the subsystem throughout its operating sequence, and verifying that each automatic valve actuates to its correct position, the blower (s) start.
2. Verifying that Nbloweridevelops at least the required vacuum at l

the rated capacity and each heater unit draws 7.4 to 9.2 amperes per phase.

a) Inboard system, 15" H 2O vacuum at > 100 scfm.

b) Outboard system, 15" 2H O vacuum at > 100'scfm for-each-blo d 2co l CLINTON - UNIT 1 3/4 6-7

k

.. Attrchment

D-600848

. Page 16 of 23 Description of Change Specification 4.6.2.2, page 3/4'6-15.

. Add the words "or during each refueling outage" as indicated in the markup.

i Justification This change will bring the CPS-TS into agreement with the NRC position 1

as described in SSER-5, page 6-1, paragraph 6.2.1.7. This change has

! been discussed with the NRC's B. Siegel, NRC Clinton Licensing Project Manager. _This change amends the change presented in letter U-600817 1 dated February-4, 1987.

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Attachment U-600848 CONTAINMENT SYSTEMS Page 17 of 23 DRYWELL BYPASS LEAKAGE ,

LIMITING CONDITION FOR OPERATION 3.6.2.2 Orywell bypass leakage shall be less than or equal to 10% of the minimum acceptable A/4 design value of 1.18 ft2 ,

~

APPLICABILITY: When ORYWELL INTEGRITY is required per Specification 3.6.2.1.

1- ACTION:

With the drywell bypass leakage greater than 10% of the minimum acceptable A/4 design value of 1.18 ft ,2restore the drywell bypass leakage to within the limit prior to increasing raactor coolant system temperature above 200'F.

, j SURVEILLANCE REQUIREMENTS 4.6.2.2 The drywell bypass leakage rate test shall be conducted at least once

> I per 18 months e at an initial differential pressure of 3.0 psi and the A/4 shall g i tFe calculated from the measured leakage. One drywell airlock door shall remain open during the drywell leakage test such that each drywell door is leak tested j during at least every other leakage rate test) i U CIf any drywell bypass leakage test fails to meet the specified limit, the schedule for subsequent tests shall be reviewed and approved by the Com-mission. If two consecutive tests fail to meet the limit, a test shall be performed at least every 9 months until two consecutive tests meet the

< limit, at which time the 18 month test schedule may be resumed. .

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l CLINTON - UNIT 1 3/4 6-15 l

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. Attcchment U-600848 Page 18 of 23 Description of Change Specification 3/4.9.12, page 3/4.9-19 Add the information per the markup.

Justification Illinois Power Company letter U-600785 dated January 8, 1987, submitted a change to this Specification. After NRC's review, and in accordance with discussions held with the NRC's Messrs. B. Siegel and J. Ridgely,

, additional changes to Specification 4.9.12.1 are proposed to ensure sufficient precautions are taken to prevent personnel from entering into areas adjacent to where irradiated fuel is being handled with the Inclined Fuel' Transfer System (IFTS). This action is considered

- necessary because, unlike similar BWR's, Clinton does not have visible warning lights which would warn individuals of IFTS operation.

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Attcchment

.. U-600848 REFUELING 0PERATIONS Page 19 of 23 3/4.9.12 INCLINED FUEL TRANSFER SYSTEM LIMITING CONDITION FOR OPERATION -

3.9.12 The inclined fuel transfer system (IFTS) may be in operation provided that:

a. The access doors
  • of all rooms through which the transfer system penetrates are closed and locked.

b.

All access doorfinterlocks are OPERABLE. l

c. The blocking valve located in the fuel building IFTS hydraulic power unit is 0PERA8LE.**
d. At least one IFTS carriage position indicator is OPERA 8LE at each carriage position and at least one liquid level sensor is OPERA 8LE.**
e. Any keylock switch that provides IFTS access control-transfer system lock-out is 0PERABLE.

, APPLICA8ILITY: When the IFTS containment blank flange is removed.

ACTION:

With the requirements of the above specification not satisfied, suspend IFTS operation with the IFTS at either termTnal point. The provisions of Specifi-cation 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS @d d/ /td5/ one per/2 hours t .

4.9.12.1 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> prior to the startup he F hat no personnel are in areas immediately adjacent to the IFTS and that all access doors to rooms through which the IFTS penetrates are closed and locked.

4.9.12.2 Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to the operation of IFTS and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter, verify that; a.

AllaccessdoorfinterlocksareOPERA8LE. l

b. The blockin OPERA 8LE.**g valve in the Fuel Building IFTS hydraulic power unit is
c. At least one IFTS carriage position indicator is 0PERA8LE at each carriage position and at least one liquid level indicator is OPERA 8LE.**

" Includes removtble shields.

    • The blocking vdive in the fuel building IFTS hydraulic power unit and the liquid level indicator are not required to be OPERA 8LE for the purposes of these specifications until after fuel loading, but before exceeding 5% of RATED THERMAL POWER or before removal of the reactor pressure vessel head after the initial criticality.

CLINTON - UNIT 1 3/4 9 19

U-bc0 7FD Att:chment U-600848 -

Page 20 of 23 REFUELING OPERATICKS ,  ;-;

3/4.9.12 INCLINED FUEL TRANSFER SYSTEM . .

  1. {

.5 LIMITING CONDITION FOR 0PERATICN. t l .

3.9.12 The inclined fuel transfer system (IFTS) may be in operation pr: viced that: .-

a. The access doors
  • of all rooms through which the transf.er system penetrates are closed and locked. .
b. All access door interlocks are OPERABLE. -
c. The biceking valve loc'ated in the fuel building IFTS hydraulic pcwer unit is 0PEMBLE.'"
d. At least one IFTS carriage position indicator is OPEMELE at each carriage position and at least ene liquid level sens' ro is CPEMBLE.'"
e. Any keylock switch.that provides IFTS access control-transfer system lock-

-, out is OPE MBLE.

. APPLICABILITY: When the IFT5 containe.ent blank flange is removed.

i ACTION:

With the requirements of the above specification not satisfied, sus;end IFT5 operation with the IFT5 at either termTnal point. The provisions of Specifi-cation 3.0.3 are not applicable.

SURVEILLANCE REOUIREu!NTS ,

4.9.12.1 Within I hour prior to the startup of the IFTS, verify that no personnel are in areas ircediately adjacent to the IFi$ and that all access s doors to rooms through which the IFTS penetrates are closed and locked.

4.9.12.2 Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to the operation of IFT5 and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter, verify thatf* '79 3

~~~?g,i'

a. All access d;or interlocks are OPEMBLE. t~
b. The blocking valve in the Fuel Building IFT3 hydraulic c:wer unit is

_ O P E RAB LE. "" -

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.sr*' t least one IFT5 carriage p:sition indicat:r is CP!RAEL! at en:n car-iage j' position and at least one liquic level incicator is OE!!AILI."

"Incluces remova01e shields.

'"The blocking valve in the fuel b.silding lFis nycrawlic =cwer unit anc the liquid level incicator are not recuirec to be CP!PABL! s f r tne cur;:ses of

.these s:eciff cations until after fuel leading, but sf:re ex:eecing M of RATED TFIRwAL MwER or tef:re re eval of the rea::ce ;ressee sessel neac after tne initial criticality.

CL:NTON = UN:7 1 3/4 9 *.9

M-kco785 Attcchment -

U-600848 Page 21 of 23 .

REFUELING CPERATIONS . . .

.' 3 i

  • INCLINED FUEL TRANSFER SYSTEM . .

5 '*

5URVEILLANCE REQUIREMENTS [ Continued) ,

s 4.9.12.2 (Continued) ,% y w

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. v -

, hi-dThe keylock switches which provide' IFTS access or control-transfer system W

lockout are OPERABLE.

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g -CFT5 and cd lea.si ence Pec 7 dap -K ercaf4<r, vtci G3 Mat p 3//1-20 +'.' 9. o. 2 o-, b N s 7A .

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  • . Att:chment U-600848 Page 22 of 23 Description of Change BASES 3/4.4.3, page B 3/4 4-3 Revise paragraph 3/4.3.1 as shown in the markup.

Justification The purpose of this revision is to update the BASES in accordance with the described revision to the FSAR as contained in letter U-600840 from F. A. Spangenberg to Dr. W. R. Butler (NRC) dated February 16, 1987.

This change has been discussed with the NRC's Messrs. J. Ridgely and B. I Siegel and is consistent with BASES of other operating BWR's. Please l see the change and justification to page 3/4 4-11 (page 12 of 23 of this  ;

Attachment) for additional details.

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.- , Attcchment U-600848 page 23 of 23 REac7OR C00LaNT SYSTEM LASES .

3/.1.4.1 RECIRCULATION SYSTEM (Continued) the recirculation flow control failures on increasing and decreasing flow are ,

j presented in Sections 15.3 and 15.4 of the FSAR respectively.

The required surveillance interval is adequate to ensure that the flow control I valves remain OPERABLE and not so frequent as to cause excessive wear on the i system components.

i 3/4.4.2 SAFETY / RELIEF VALVES 4

The safety valve function of the safety / relief valves (SRV) operate to prevent the reactor coolant system from bein<1 pressurized above the Safety Limit of 1375 psig in accordance with the ASMB Code. A total of 11 OPERABLE safety- ,

relief valves is required to limit reactor pressure to within ASME III allowable l values for the worst case upset transient. Any combination of 5 SRVs operating 1 l

i in the relief mode and 6 SRVs operating in the safety mode is acceptable.

1 i

, Demonstration of the safety relief valve lift settings will occur only during <

shutdown and will be performed in accordance with the provisions of Specifica-l tion 4.0.5.

[ I The,10w low set system ensures that safety /reifef valve discharges are minimized I for a second opening of these valves, following any overpressure transient.

i' l This is achieved by automatically lowering the closing setpoint of 5 valvesInand lowering the opening setpoint of 2 valves following the initial opening.

) this way, the frequency and magnitude of the containment blowdown duty cycle is T substantially reduced. Sufficient redundancy is provided for the low-low set system such that failure of any one valve to open or close at its reduced set-point does not violate the design basis, f l .

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3/4.4.3 REACTOR COOL ANT SYSTEM LEAGGE Med M* Ioldd l 3/4.4.3.1 LEA UGE DETECf!ON SYSTE f I J

The RCS leakage detecti n systems required by this specification are provided These

' to monitor and det et ea age from the reactor coolant pressure boundary.

..... ..,. ....~ . ~ .,

Jof Regulatory Guide detection systems - -. m...,.

i 45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May ,

i 2 nece*wirfort-w wA*-renmeer du%ww-* F-h u s 2-m r e s %  ?

, e #*/*^4# 8M*/**/-AW'-3**'dary 488V 3/4.4.3.2 OPERATIONAL LEAUGJ Desube9N482 e r I The allowable leakage rates from the reactor coolant system have beenThe based on f

i the predicted and experimentally observed behavior o i capability of the instrumentation for determining system

! sidered.

i what greater than that specified for UNIDENTIFIED l 8 3/4 4-3

]

CLIN 7CN

  • UNIT 1 4
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