ML20070S281

Amend 199 to License DPR-49,revising TS by Making Editorial Changes to Correct Typo & Administrative Errors

ML20070S281
Person / Time
Site:	Duane Arnold
Issue date:	05/18/1994
From:	Pulsifer R Office of Nuclear Reactor Regulation
To:
Shared Package
ML20070S284	List: ML20070S281 ML20070S287
References
NUDOCS 9405230324
Download: ML20070S281 (15)

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UNITED STATES iO

--l NUCLEAR REGULATORY COMMISSION

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WASHINGTON, D.C. 20555-0001 v

g LES VTILITIES INC.

CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET N0. 50-331 QUANE ARNOLD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 199 License No. DPR-49 1.

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

A.

The application for amendment by Iowa Electric Light and Power Company, et al., dated December 22, 1993, 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 l

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 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 9405230324 940518 PDR ADOCK 05000331 7Yor2 b32f p

PDR

(2) Technical Soecification1 The Technical Specifications contained in Appendix A, as revised through Amendment No.199, are hereby incorporated in the license.

The licensee shall operate the facility in acccrdance with the i

Technical Specifications.

3.

The license amendment is effective as of the date of issuance and shall be implemented within 60 days of the dat a of issuance.

FOR THE NUCl. EAR REGULATORY COMMISSION e$

Robert M. Pulsifo/, Project Manager Project Directorate !!I-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: May 18, 1994 j

1 ATTACHMENT TO LICENSE AMEN 0 MENT N0. 199 FACIllTY OPERATING LICENSE NO. OPR-49 DOCKET N0. 50-331 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.

The revised areas are indicated by marginal lines.

gemove lasert 1.0-8 1.0-8 1.1-6 1.1-6 1.1-13 1.1-13 1.1-14 1.1-14 1.1-15 1.1-15 1.2-2 1.2-2 3.3-2 3.3-2 3.5-5 3.5-5 3.5-6 3.5-6 3.6-33 3.6-33 6.8-2 6.8-2 6.11-5 6.11-5 i

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k DAEC-1 27.

FRE0VENCY NOTATION NOTATION FRE0VENCY S

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

W At least once per 7 days.

At least once per 31 days.

d At least once per 92 days.

,t At least once per 184 days.

At least once per year.

4 At least once per 18 months.

,a Prior to each reactor startup.

P Prior to each release.

NA Not applicable.

28. DELETED

29. REACTOR TRIP SYSTEM RESPONSE TIME REACTOR TRIP SYSTEM RESPONSE TIME is the time interval from when the l

monitored parameter exceeds its trip setpoint at the channel sensor untn deenergization of the scram pilot valve solenoids.

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30. REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

31. OFFSilf DOSE ASSESSMENT MANUAL The OF7 SITE DOSE ASSESSMENT MANUAL parameters to be used in the calcula(0 DAM) contains the methodology and tion of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODAM shall also contaD.

(1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Program required by Section 6.9.4 and (Radioactive Material Release

2) descriptions of the information that should be included in the Annual l

Report and Annual Radiological Environmental Report required by the Technical Spe..ification 6.11.1.

12. DELETED

33. PURGE - PURGING PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humicity, concentration or othcr operating condition, in such a manner that replacement air or gas is required to purify the confinement.

AMENDMENT NO. 109.I34.199 1.0-8 t

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DAEC-1 8.

Core Thermal Power Limit (Reactor Pressure s785 psig or Core Flow s10%

of Rated)

At pressures below 785 psig, the core evaluation pressure drop (0 power, O flow) is greater than 4.56 pt1.

At low power and all flows, this pressure differential is maintained in the bypass region of the core.

Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and all flows will always be greater than 4.56 psi.

3 Analyses show that with a flow of 20 x 10 lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, 3

the bundle flow with a 4.56 psi driving head will be greater than 28 x 10 lbs/hr irrespective of total core flow and independent of bundle power for the range of bundle powers of concern.

Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors this corresponds to a core thermal power of more than 50%.

Thus, a core thermal power limit of 25% for reactor pressures below 800 psia or core flow less than 10% is conservative.

C.

Power Transient Plant safety analyses have shown that the scrams caused by exceeding any safety setting will assure that the Safety Limit of Specification 1.1.A or 1.1.B will not be exceeded.

Scram times are checked periodically to assure the insertion times are adequate.

The thermal power transient resulting when a scram is accomplished other than by b expected scram signal (e.g., scram from neutron flux following close of ti,.

.in turbine stop valves) does not l

necessarily cause fuel damage.

However, for this specification a Safety Limit violation will be assumed when a scram is only accomplished by means of a backup feature of the plant design.

The concept of not approaching a Safety Limit provided scram signals are operable is supported by the extensive plant s+fety enalysis.

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2 AMENDMENT N0.-11$,199 1.1-6

DAEC-1 2.

APRM Hiah Flux Scram (Refuel or Startuo & Hot Standby Model for operation in these modes the APRM scram setting of 15 percent of rated power and the IRM High Flux Scram provide adequate thermal margin between the setpoint and the Safety Limit, 25 percent of rated. The margin is adequate to accommodate anticipated maneuvers associated with -

power plant startup.

Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer.

Worths of individual rods are very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise, i

l Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow.

Generally, the heat flux is near equilibrium with the fission rate.

In an assumed uniform rod withdraaal approach to the scram level, the rate of power rise is no more than 5 percent of rated power per minute, and the ApRM system would be more

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4 than adequate to assure a scram before the power could exceed the Safety Limit. The 15 percent APRM scram remains active until the mode switch is placed in the RUN position.

This switch occurs when reactor pressure is greater than 850 psig.

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AMENDMENT NO. M9,199 1.1-13 l

DAEC-1 THIS PAGE INTENTIONALLY LEFT BLANK.

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1 ANENDMENT N0.I20,199 1,j_14 y

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DAEC-1 3.

IRM The IRM system consists of 6 chambers, 3 in each of the reactor protection system logic channels.

The IRM is a 5-decade instrument which covers the range of power level between that covered by the SRM and the APRM.

The 5 decades are covered by the IRM by means of a range switch and the 5 decades are broken down into 10 ranges, each being one-half of a decade in size.

The IRM scram trip setting of 120 divisions is active in each range of the IRM.

For example, if the instrument were on range 5, the scram would be 120 divisions on that range.

Thus, as the IRM is ranged up to accommodate the increase in power level, the scram trip setting is also ranged up.

The most significant sources of reactivity change during the power increase are due to control rod withdrawal.

For insequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing control rods that the heat flux is in equilibrium with the neutron flux, and the IRM scram would result in a reactor shutdown well before any Safety Limit is exceeded.

in order to ensure that the IRM provides adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents has been analyzed.

This analysis included startir.g the accident at various power levels. The i

most severe case involves an initial condition in which the reactor is just subcritical and the IRM system is not yet on scale.

This condition exists at quarter rod density.

Additional conservatism was taken in this analysis by assuming that AMENDMENT N0.720>l99 1.1-15

DAEC-1 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING C. Relief Valve Settings - Low-Low Set Function Valve Group Open Close Low (1 valve) 1020 psig 900 psig High (1 valve) 1025 psig 905 psig.

All settings are i 25 psi.

I D. Safety Valve settings 1240 psig i 12 psi (2 valves) 2.

The reactor vessel dome 2.

The shutdown cooling isolation l

pressure shall not exceed 135 valves shall be closed whenever the psig at any time when reactor vessel dome pressure is operating the Residual Heat 2 135 psig.

Removal pump in the shutdown cooling mode.

i AMENDMENT NO. 102,199 1.2-2

DAEC-1 LIMITING CONDITIONS FOR OPERATION SVRVEILLANCE RE0VIREMENTS d.

Each control rod shall be coupled d.

When a contr>l rod is withdrawn the to its drive.*

If a control rod first time after refueling, after becomes uncoupled, CR0 maintenance or when required by Specification 3.3.A.2.d(ii),

(i) recouple the control rod coupling integrity shall be verified within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and by observing that the drive does not 90 to the overtravel position when (ii) verify coupling by performing the rod is fully withdrawn, surveillance 4.3.A.2.d.

(iii) If the control rod is not recoupled, declare the control rod inoperable.

The actions stated in Specification 3.3. A.2.e shall be

taken,

e. A control rod that has been declared e.

(not used) inoperable for reasons other than being stuck shall:

(i) be fully inserted,** and (ii) disarm the associated directional control valves electrically.

The control valves may be re-armed to permit testing associated with returning the control rod to OPERABLE status.

(iii) Whenever the reactor is less than 20% power, verify all inoperable control rods not in compliance with BPWS are separated by 2 or more OPERABLE control rods in any direction, including the diagonal.

(iv) Verify that no more than 8 inoperable control rods exist.

(v) If the requirements of Specification 3.3.A.2.e (i)- iv) cannotbemet,beinCOLDSHL(TDOWN 4

within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• This requirement does not apply in

, the refuel condition.

Refer to Specifications 3.9.A.3 and 3.9.A.4 for control rod requirements during refueling.

• • The RWM may be bypassed, if required, to allow insertion of inoperable control rods and continued operation.

AMENDMENT NO.J01,720,J/2,143,JB0,199 3.3-2

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS C Residual Heat Removal (RHR)

C.

Surveillance of the RHR Service Service Water System Water System

1. Except as specified in 3.5.C.2, 1.

Surveillance of the RHR service l

3.5.C.3, 3.5 C.4, and 3.5.G.3 water system shall be as follows:

below, both RHR service water subsystem loops shall be operable RHR Service Water Subsystem whenever irradiated fuel is in the Testing:

reactor vessel and reactor coolant temperature is greater than 212*F.

Item Freauency a.

Pump and Motor Once/3 months operated valve operability.

b.

Flow Rate after major Test-Each pump RHR service maintenance water pump and every 3 shall deliver months at least 2040 gpm at a TDH of 610 ft. or more.

2. With one RHRSW pump inoperable, provided the remaining active components of both RHRSW subsystems are verified to be OPERABLE, restore the inoperable pump to OPERABLE status within 30 days or be in at least H0T 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 />.

3. With one RHRSW pump in each subsystem inoperable, provided the remaining active components of both RHRSW subsystems and the diesel generators are verified to be OPERABLE, restore at least one inoperable pump to OPERABLE status within 7 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 />.

AMENDMENT N0.708,J39,J70,J74,199 3.5-5

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS 4.

With one RHRSW subsystem inoperable, provided the remaining RHRSW subsystem and its associated diesel generator are verified to be OPERABLE, restore the inoperable system to OPERABLE status with at least one OPERABLE pump within 7 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 />.

D.

HPCI Subsystem D.

HPCI Subsystem 1.

The HPCI Subsystem shall be 1.

HPCI Subsystem testing shall be OPERABLE whenever there is performed as follows:

irradiated fuel in the reactor vessel, reactor pressure is Item Frecuency greater than 150 psig, and prior to reactor startup from a COLD a.

Simulated Annual CONDITION, except as specified in Automatic 3.5.0.2 below.

Actuation Test b.

Pump Operability Once/3 Months c.

Motor Operated Once/3 Months Valve Operability d.

At rated reactor Once/3 Months pressure demonstrate ability to deliver rated flow at a discharge pressure greater than or equal to that pressure required to accomplish vessel injection if vessel pressure were as high as 1040 psig.

AMENDMENT N0. JJB,JA3,J60,77A,199 3.5.

l DAEC-1 3.6.F & 4.6.F BASES:

Jet Pump Flow Mismatch The LPCI loop selection logic has been previously described in the Updated FSAR Section 7.3.1.1.2.4.

For some limited low probability accidents with the recirculation loop operating with large speed differences, it is possible for the logic to select the wrong loop for injection.

For these limited conditions the core spray itself is adequate to prevent fuel temperatures from exceeding allowable limits.

However, to limit the probability even further, a procedural limitation has been placed on the allowable variation in speed between the recirculation pumps.

The licensee's analyses indicate that above 80% power the loop select logic could be expected to function at a speed differential up to 14% of their average speed.

Below 80% power the loop select logic would be expected to function at a speed differential up to 20% of their average speed.

This specification provides margin because the limits are set at 110% and 115% of the average speed for the above and below 80% power cases, respectively.

If the reactor is operating on one recirculation pump, the loop select logic trips that pump before making the loop sel ec tiori.

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- AMENDMENT NO. )f4,)f2,199 3.6-33

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i DAEC-1 8.

Procedures required by the plant Security Plan.

9.

Operation of oactive waste systems.

10.

Fire Protectit Program implementation,

11. A preventive maintenance and periodic visual examination program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient to as low as practical levels.

This program shall also include provisions for performance of periodic systems leak tests of each system once per operating cycle.

12.

Program to ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions, including training of personnel, procedures for monitoring and provisions for maintenance of sampling and analysis equipment.

13. Administrative procedures for shift overtime for Operations personnel to be consistent with the Commission's June 15, 1982 policy statement.

14.

OFFSITE DOSE ASSESSMENT MANUAL.

15.

PROCESS CONTROL PROGRAM.

i 16.

Quality Control Program for effluents.

6.8.2 Procedures described in 6.8.1 above, and changes thereto, shall be reviewed by the Operations Committee as indicated in Specification 6.5.1.6 and approved by the Plant Superintendent-Nuclear prior to implementation, except as provided in 6.8.3 below.

6,8.3 Temporary minor changes to procedures described in 6.8.1 above which do not change the intent of the original proceaure may be made with the concurrence of two members of the plant management staff, at least one of whom shall hold a senior operator license.

Such changes shall'

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be documented and promptly reviewed by the Operations Committee and by the Plant Superintendent-Nuclear.

Subsequent incorporation, if necessary, as a permanent change, shall be in accord with 6.8.2 above.

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AMENDMENT N0.JM,J24,J2A,JA3,JE/,199

-6.8-2 a

DAEC-1 d.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

6.11.3 UNIOVE REPORTING RE0VIREMENTS Special reports shall be submitted to the Director of Inspection and Enforcement Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification, a.

Reactor vessel base, weld and heat affected zone metal test speiimens (Specification 4.6.A.2).

b.

deleted Inservice inspection (Specification 4.6.G.).

c.

d.

Reactor Containment Integrated Leakage Rate Test (Specification 4.7.A.2.g).

e.

deleted f.

deleted 1

9 deleted h.

Radioactive Liquid or Gaseous Effluent - calculated dose exceeding specified limit (ODAM Sections 6.1.3, 6.2.3 and 6.2.4).

i.

Off-Gas System inoperable (00AM Section 6.2.5).

j. Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of ODAM Table 6.3-3 when averaged over any calendar quarter sampling period (00AM Section 6.3.2.1).

k.

Annual dose to a MEMBER OF THE PUBLIC determined to exceed 40 CFR Part 190 dose limit (0 DAM Section 6.3.1.1).

l 1.

Radioactive liquid waste released without treatment when activity concentration is equal to or greater than 0.01 pci/ml (00AM Section 6.1.4.1).

Explosive Gas Monitoring Instrumentation Inoperable (Specification m.

3.2.I.1),

Liquid Holdup Tank Instrumentation Inoperable (Specification n.

3.14.B.1).

AMENDUENT NO. (s6,Is6,199 6.11-5