June 14,-Tý95

Mr. Lee Liu Chairman of the Board and

Chief Executive Officer IES Utilities Inc. Post Office Box 351 Cedar Rapids, IA 52406

Distribution'-Wencls: Docket File GHilI(2) PUBLIC OGC PD3-3 r/f CGrimes ACRS(4) EAdensam OPA OC/LFDCB WAxelson, RIIl

SUBJECT: AMENDMENT NO. 212 TO FACILITY OPERATING LICENSE NO. DPR-49 - DUANE

ARNOLD ENERGY CENTER (TAC NO. M91940)

Dear Mr. Liu:

The Commission has issued the enclosed Amendment No.212 to Facility Operating License No. DPR-49 for the Duane Arnold Energy Center (DAEC). This amendment consists of changes to the Technical Specifications (TS) in response to your application dated March 28, 1995.

The amendment revises TS Table 3.2-A by clarifying or correcting entries to the table. The amendment also revises the TS Bases to describe more clearly the logic arrangements in Table 3.2-A. TS Table 3.2-A defines trip functions, trip level settings, applicable operating modes, isolation signals, and minimum number of channels.

A copy of the related Safety Evaluation is also enclosed. Notice of issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely,

ORIGINAL SIGNED BY:

Glenn B. Kelly, Project Manager Project Directorate 111-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

Docket No. 50-331

Enclosures: 1. Amendment No. 2 1 2 to License No. DPR-49

2. Safety Evaluation

cc w/encls: See next page DOCUMENT NAME: G:\DUANEARN\DUA91940.AMD

To receive a copy of this document, Indicate In the box: "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure

"N" = No copy

OFFICE NAME DATE

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UNITED STATES o NUCLEAR REGULATORY COMMISSION z WASHINGTON, D.C. 20555-0001

June 14, 1995

Mr. Lee Liu Chairman of the Board and

Chief Executive Officer IES Utilities Inc. Post Office Box 351 Cedar Rapids, IA 52406

SUBJECT: AMENDMENT NO. 212 TO FACILITY OPERATING LICENSE NO. DPR-49 - DUANE ARNOLD ENERGY CENTER (TAC NO. M91940)

Dear Mr. Liu:

The Commission has issued the enclosed Amendment No. 2 12 to Facility Operating License No. DPR-49 for the Duane Arnold Energy Center (DAEC). This amendment consists of changes to the Technical Specifications (TS) in response to your application dated March 28, 1995.

The amendment revises TS Table 3.2-A by clarifying or correcting entries to the table. The amendment also revises the TS Bases to describe more clearly the logic arrangements in Table 3.2-A. TS Table 3.2-A defines trip functions, trip level settings, applicable operating modes, isolation signals, and minimum number of channels.

A copy of the related Safety Evaluation is also enclosed. Notice of issuance will be included in the Commission's next biweekly Federal Register notice.

',i ncerely,

Glenn B. Kelly, Pro, ect Manager Project Directorate. 111-3 Division of Reactor Projects Ill/IV Office of Nuclear Reactor Regulation

Docket No. 50-331

Enclosures: 1. Amendment No. 2 1 2 to License No. DPR-49

2. Safety Evaluation

cc w/encls: See next page

Mr. Lee Liu Duane Arnold Energy Center IES Utilities Inc.

cc:

Jack Newman, Esquire Kathleen H. Shea, Esquire Morgan, Lewis, & Bockius 1800 M Street, NW. Washington, DC 20036-5869

Chairman, Linn County Board of Supervisors Cedar Rapids, Iowa 52406

IES Utilities Inc. ATTN: David L. Wilson Plant Superintendent, Nuclear 3277 DAEC Road Palo, Iowa 52324

Mr. John F. Franz, Jr. Vice President, Nuclear Duane Arnold Energy Center 3277 DAEC Road Palo, Iowa 52324

Mr. Keith Young Manager, Nuclear Licensing Duane Arnold Energy Center 3277 DAEC Road Palo, Iowa 52324

U. S. Nuclear Regulatory Commission Resident Inspector's Office Rural Route #1 Palo, Iowa 52324

Regional Administrator, RIIl U. S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532-4531

Mr. Stephen N. Brown Utilities Division Iowa Department of Commerce Lucas Office Building, 5th floor Des Moines, Iowa 50319

1 0UNITED STATES NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

IES UTILITIES INC.

CENTRAL IOWA POWER COOPERATIVE

CORN BELT POWER COOPERATIVE

DOCKET NO. 50-331

DUANE ARNOLD ENERGY CENTER

AMENDMENT TO FACILITY OPERATING LICENSE

Amendment No. 212 License No. DPR-49

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by IES Utilities Inc., et al., dated March 28, 1995, 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 operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety 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 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 Specifications 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:

9506290146 950614 PDR ADOCK 05000331 P PDR

-2-

(2) Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 212 , are hereby incorporated in the license. The licensee shall operate the facility in 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

"Gail H. Marcus, Pre'e• Director Project Directorate 111-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

Attachment: Changes to the Technical Specifications

Date of issuance: June 14, 1995

ATTACHMENT TO LICENSE AMENDMENT NO. 212,

FACILITY 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 vertical lines.

Remove Insert

3.2-3 3.2-3

3.2-4 3.2-4

3.2-7 3.2-7

3.2-44 3.2-44

Table 1.2-A ISOLATION ACTUATION INSTRUMENTATION

TRIP FUNCTION TRIP LEVEL SETTING

MINIMUM OPERABLE

APPLICABLE CHANNELS OPERATING PER TRIP

MODE SYSTEM )

C-. z 0~

I ý0

ro

Reactor Water Level-Low

Reactor Water Level - Low-Low-Low

Drywell Pressure - High

Main Steam Line Isolation

Main Steam Line Pressure - Low

Main Steam Line Flow - High

Condenser Backpressure - High Main Steam Line Tunnel Temperature - High

Turbine Building Temperature - High

Main Steam Line Radiation - High

L 170 Inches

S18.5 Inches

S2.0 psig

* 850 psig

* 140% of Rated Steam Flow

* 20 In. Hg

* 200"F

1,2,3 1,2,Y, and

1,2,3

1,2,3 1,2,3

1,2,3 1,2,3 1,2,3 1,2,3

1

1,2,3

1, 2**, 3**

1,2,3

S200°F 1,2 ,3

• 3 x Normal Rated 1,2,3 Power Background"j)

2 2 2

2 4(n)

2 2 2 1(f)

2

2/line

2 4(0)

4

2

VALVE GROUPS

ISOLATED BY

SIGNAL ACTION

2 3 (c) 4,5

1 7

2 3 (c) 4 9

1 1

1

1

1 l(b)

20 26 23

21 20

20 26 23 23

22

20

21

21

21

21

Common Isolation-Sianals

(A

(

I

Table 3.2-A Coptinued) ISOLAT1ON ACTUATION-INSTRUMENTATION

TRIP FUNCTION TRIP LEVEL SETTING

APPLICABLE OPERATING

MODE

(D

CL 0 (D

VALVE GROUPS

ISOLATED BY

SIGNAL ACTION

Secondary containment

Refuel Floor Exhaust Duct High Radiation

Reactor Building Exhaust Shaft - High Radiation

Offgas Vent Stack - High Radiation

RHR System Shutdown Cooling

Reactor Vessel Pressure - High

Reactor Water Cleanup

RWCU Differential Flow - High

RWCU Area Temperature - High

RWCU Area Ventilation Differential Temperature - High

Standb Liquid Control System InitiatLon

RWCU Area Near TIP Room Ambient Temperature - High

< 9 mr/hr

< 11 mr/hr

Note k

5 135 psig

5 40 gpmd

< 130OF

A 14oF(d)

NA

: 1l1.5 0F

1,2,3 and *

1,2,3 and *

Note m

1,2,3

1,2,3

1,2,3

1,2,3

Note i

1,2,3

MINIMUM OPERABLE CHANNELS PER TRIP SYSTEM"'0

�71

V

41

3 (c)

3 (c)

3 (c)

4

1

1

1

1

.1 1

1

1/SBLC System

1

26

26

27

23

23

23

23

5

5

5 (

5(*)

5

23

23

(

DAEC -1

Table 3.2-A (Continued)

ISOLATION ACTUATION INSTRUMENTATION

ACTION

ACTION 20 - Be in at 1last HOT SHUTDOWN within 12 hours and in COLD SHU DOWN within the next 24 hours.

ACTION 21 - Be in at least STARTUP with the associated isolation valves closed within 6 hours or be in at least HOT SHUTDOWN within 62 hours 4nd in SOLD SHUTDOWN within the next 24 hours.

ACTION 22 - in at least STARTUP within 6 hours. ACTION 23 - Close the affected system isolation valves within one hour

and declare the affected system inoperable. ACTION 24 - Not Used ACTION 25 Restore the manual initiation function to OPERABLE status

wjithin 8 hours or close the affected system isolation valves within the next hour and dec are the affected system inoperable.

ACTION 26 - Isolate Secondary Coptainment with the Standby Gas Treatment A Nstm operating w~tinn one h9ur.

ACTION 27 thin one hour close the primary containment vent and purge valves, or estaolish adrainistrative control of those valves with continuous monitoring of alternate instrumentation.

NOTES When handling irradiated fuel in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

** When any turbine stop valve is greater than 90% open and/or when the key-locked bypass switch is in the NORM position.

(a) When a channel is placed in an inoperable status solely for the performance of required surveillances, entry into associated Limiting Conditions for Operation and required Actions may be delayed as follows: (1) for up to 6 hours for RWCU Differential Flow- High, RCIC Manual Initiatjon, HPCI Manual Initiation; and (2) for up o 6 hours for the remaining Trip Functions.provided the associated Trip Funytion maintains isolation aability.

(b) Operates Group I va ves exgcpt Main Steam Iso a ion Valvyes. Also trips Mechnical Vacuum Pump which results in a subsequent Isolation of the Mechanical Vacuum Pump suction valves.

c) Also starts the Standby Gas Treatment System. Actua setpoint shall Be 14°F above the 100% operation ambient temperature conditions as determined by DAEC plant test procedure. Closes MO-2701 and MO-2740 only. Requires system steam su ply pressure-low coincident with drywell ressure-high to close HPCINRCIC exhaust vacuum breaker valves.

(g) manual isolation closes MO- 2401 only, if RCIC initiation signal resent.

(h) Manual isolation closes MO-2239 only, if HPCI initiation signal resent.

(i) Khen the.itandby Liquid Control System is required to be OPERABLE Ser Spec•ifcation 3.4.A.

(j) within 24 hours prior to the planned start of the hydroqen injection test with the reactor power at greater than 20% rated power, the normal full-power radiation background level a7d associated trip setpoins may be changed based on a calculated value of the radiatio? leyel expecteg during.the test. The background radiation evel and assocoated trip setpoints may be adjusted during the test program base, on either calculations or measurements ot actual radi• aion levels resulting from hydrogen injection. The background radiation lev?] ,hall be determined and associated trip setpoints shal be set within 24 hours of reestabliihing normal radiatiog.levels after com letion of the hydrogen injection test or with n 12 hours of es ablishing reactor power levels below 20% rated power, while these functions are required to be operable.

(k nThe.mooitors sha l be set to initiate isolation at a setting equivalent to or below the dose rate limits in ODAM section 6 2,2.1.

(m) Durinq VENTING or PURGING of primary containment at any time when PRJMARY CQNTAINMENT INTEGRITY is required.

ýn3 On I y Trip 5ystem. Sequir~s a miniimum of 2 temperature sensors (instrument channels) per main steam line.

Amendment No.89,131,193,209,212 3.2-7

DAEC-1

emergency diesel generators. These trip level settings were chosen to be high enough to prevent spurious actuation but low enough to initiate ECCS operation and primary system isolation so that post accident cooling can be accomplished and the guidelines of 10 CFR 100 will not be exceeded. For large breaks up to the complete circumferential break of a 22-inch recirculation line and with the trip setting given above, ECCS initiation and primary system isolation are initiated in time to meet the above criteria. Reference Sections 6.3 and 7.3 of the Updated FSAR.

The high drywell pressure instrumentation is a diverse signal for malfunctions to the water level instrumentation and in addition to initiating ECCS, it causes isolation of Group 2 and 3 isolation valves. For the breaks discussed above, this instrumentation will generally initiate ECCS operation before the low-low-low water level instrumentation; thus the results given above are applicable here also. The water level instrumentation initiates protection for the full spectrum of loss-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 the main steam line. There are four differential pressure instruments on each venturi, yielding 16 total instrument channels. Because of the number of instrument channels, they are arranged in a unique manner to achieve both electrical and spatial separation. One instrument channel from each of the four venturis, i.e., one from each Main Steam Line (MSL), is combined into one logic channel. The four logic channels are then combined into the standard "one-out-of-two-twice" trip actuation arrangement. This allows one logic channel to be out-of-service and still provide isolation capability in the event of a steamline break in any MSL. The Allowed Outage Time (AOT) for required surveillance is applied at the instrument channel level, since the associated trip function and isolation capability are maintained via the companion logic channel. This is consistent with the "normal" trip arrangements with one instrument channel feeding each logic channel. Thus, a six-hour AOT is applied to each instrument channel undergoing required testing.

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 1000 0 F, and release of radioactivity to the environs is below 10 CFR 100 guidelines. Reference Subsection 15.6.5 of the Updated FSAR.

Temperature monitoring instrumentation is provided in the main steam line tunnel and turbine building to detect leaks in this area. There are four resistance temperature detectors (RTDs) located along each MSL in the MSL tunnel area, yielding 16 total instrument channels. Because of the number of instrument channels, they are arranged in a unique manner to achieve both electrical and spatial separation. The four instrument channels from each MSL are combined into one logic channel. Eight instrument channels from two MSLs (2 logic channelsk make up a trip system. To ensure spatial coverage is provided by these channels, a minimum of two instrument channels per MSL is required. Since the MSLs outside the MSL tunnel (Turbine Building) have fewer RTDs monitoring them (two RTIDs per line), all eight instrument channels (four per trip system) are required to ensure spatial coverage. As with the

Amendment No. W9,129,12,18 3, 3.2-44 212

DAEC-1 1_

differential pressure transmitters on the MSL venturis, the AOT for required surveillance is applied at the instrument channel level, since the trip function and isolation capability are maintained via the companion logic channel. Thus, a six-hour AOT is applied to each instrument channel undergoing required testing. Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves. The setting is 200°F for the main steam line detectors. For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation.

High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in the control rod drop accident. A trip setting of 3 times normal full-power background is established to close the main steam line drain valves, recirculation loop sample valves, and trip the Mechanical Vacuum Pump. For changes in the Hydrogen Water Chemistry hydrogen injection rate, the trip setpoint may be adjusted based on a calculated value of the radiation level expected. Hydrogen addition will result in an increase in nitrogen (N-16) activity in the steam due to increased N-16 carryover in the main steam. Reference Subsection 15.4.7 of the Updated FSAR.

Pressure instrumentation is provided to close the main steam isolation valves in the 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 Mode 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 the RUN Mode is not required.

Pressure instrumentation provides a permissive function for opening RHR shutdown cooling isolation valves and provides automatic isolation of RHR shutdown cooling on increasing system pressure. This instrumentation provides assurance that the RHR piping transient pressure limits are not exceeded during shutdown cooling operation.

The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentation results in

Amendment No. 149,19•42,12•,1 3.2-44a

212

7P A• UNITED STATES

NUCLEAR REGULATORY COMMISSION t WASHINGTON, D.C. 20555-0001 Soo.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

RELATED TO AMENDMENT NO. 212 TO FACILITY OPERATING LICENSE NO. DPR-49

IES UTILITIES INC.

CENTRAL IOWA POWER COOPERATIVE

CORN BELT POWER COOPERATIVE

DUANE ARNOLD ENERGY CENTER

DOCKET NO. 50-331

1.0 INTRODUCTION

By letter dated March 28, 1995, IES Utilities Inc. (licensee) submitted a request for revision of the technical Specifications (TS) for the Duane Arnold Energy Center (DAEC). Amendment 193 to the Technical Specifications incorporated extended allowable out-of-service times and surveillance test intervals for various instrumentation and made other changes to improve clarity and provide consistency with the NRC Standard Technical Specifications (NUREG-0123) and the rest of the DAEC TS.

During a review of maintenance activities planned on the Main Steam Line Tunnel Temperature - High trip switches by the licensee, a question arose as to whether TS Table 3.2-A was inconsistent with the existing (i.e., original) plant design. Specifically, the TS MINIMUM OPERABLE CHANNELS PER TRIP SYSTEM requirement is "2/line." It was not clear that with the physical design and the TS definitions of channel and logic, the "2 instrument channels/line/trip system" requirement could be met. Subsequent investigation by the licensee determined that the existing design of the system has not been modified since the original plant startup.

Also, the licensee decided to evaluate other entries in TS Table 3.2-A to determine if they were appropriate. This evaluation identified other needed corrections or clarifications that improve TS clarity and consistency within the Table.

2.0 EVALUATION

The licensee checked each entry in Table 3.2-A against the current plant design and against the Architect/Engineer's design documentation to determine if (1) the as-built plant was configured as assumed in the design and (2) the TS clearly and accurately reflected the actual design and logic configurations. The licensee also checked for internal consistence among entries in Table 3.2-A. Based on this evaluation, the licensee submitted

9506290151 950614 PDR ADOCK 05000331 P PDR

-2-

various modifications to Table 3.2-A and its Bases. The TS changes do not alter the physical design or operation of the plant; they serve to describe more accurately and clearly the actual logic configurations and to improve consistency of the information in the Table. The licensee found the existing logic designs are in accordance with the Architect/Engineer's design documentation and are configured to provide the intended function. The clarifications do not affect the ability of any of the isolation logics to perform their intended functions, nor do they affect any assumptions contained in the plant safety analysis. The staff finds these changes to Table 3.2-A acceptable.

The changes to the TS Bases support the submitted logic corrections to Table 3.2-A and provide additional information to describe more clearly the logic configurations. The staff finds the changes to the TS Bases to be acceptable.

3.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Iowa State official was notified of the proposed issuance of the amendment. The State official had no comments.

4.0 ENVIRONMENTAL CONSIDERATIONS

This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes a surveillance requirement. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluent that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (60 FR 20519). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

5.0 CONCLUSION

The staff has concluded, based on the considerations discussed above, that: (I) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: Glenn B. Kelly

Date: June 14, 1995