ML20090L024
| ML20090L024 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 03/11/1992 |
| From: | Shiraki C Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20090L028 | List: |
| References | |
| NUDOCS 9203200040 | |
| Download: ML20090L024 (24) | |
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,,e UNITED STATES ni NUCLEAR REGULATORY COMMISSION j e,~
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WASHINGTON, D C. 20!65 o,
g =w, + j, IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET NO. 50-311 DUANE ARN0LD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.180 License No. DPR-49 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by lowa Electric Light and Power Company, et al., dated July 6, 1990, revised August 30, 1991, and January 8,1992, 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 regulatf or.s 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 rnended to read as follows:
i 1
l 9203200040 920311 PDR ADOCK 03000331 i
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b A
2 (2)
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.180, 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 Ad Clyde Y. Shira i, Sr. Project Manager Project Directorate Ill-3 Division of Reactor Projects Ill/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications
- Date of issuance: March 11, 1992 L
4 ATTACHMENT TO LICENSE AMENDMENLNO.180 FACILITY OPERATING LICENSE NO. OPR-49 DOCKET NO. 50-331 Replace the following pages of the Appendix A Technical Specifications with the-enclosed pages.
The revised areas are indicated by marginal lines.
Plati i
vii 1.0-9 3.2-2 3.2-2a 3.3-1 3.3-2 3.3-3 3.0-4 3.3-5 3.3-6 3.3-7 3.3-8 3.3-9 3.3.-10 3.3-11 3.3-12 3.3-13 3.3-14 3.3-15 I
3.9-2 I
t
g DAEC-1 TECHNICAL SPECIFICATIONS TABLE Of CONTENTS PfASE NO.
1.0-1 1.0 Definitions LIMITING SAFETY SAFETY LIMITS
.51 STEM SETTING 1,1 Fuel Cladding Integrity 2.1 1.1-1 1.2 Reactor Coolant System Integrity 2.2 1.2-1 SURVEILLANCE (1M1 TING CONDITION FOR OPERA 7.LO.N B.[Ql!LR.E.M. ENT S 3.1 Reactor Protection System 4.1 3.1-1 3.2 Protective Instrumentation 4.2 3.2-1 A.
Primary Containment Isolation Functions A
3.2-1 B.
Core and Containment Cooling Systems B
3.2-1 C.
Control Rod Block Actuation C
3.2-2 D.
Radistion Monitoring Systems D
3.2-2 E.
D ywell Leak Detection E
3.2-3 F.
Surveillance Information Readouts F
3.2-3 G.
Recirculation Pump Trips and Alternate Rod Insertion G
3.2-4 H.
Accident Monitoring Instrumentation H
3.2-4 3.3 Reactivity Control 4.3 3.3-1 A.
Reactivity Limitations A
3.3-1 B.
Scram Discharge Volurc.e B
3.3-3 C.
Reactivity Control Systems C
3.3-4 0.
Scram Insertion Times
'D 3.3-5 E.
Reactivity Anomalies E
3.3-6 F.
Recirculation Pumps F
3.3-6 3.4 Standby Liquid Control System 4.4 3.4-1 A.
Normal System Availability A
3.4-1 B.
Operation with Inoperable Components B
3.4-2 C.
Sodium Pentaborate Solution C
3.4-2 3.5 Core and Containment Cooling Systems 4.5 3.5-1 A.
Core Spray and LPCI Subsystems A
3.5-1 B.
Containment Spray Cooling Capability B
3.5-4 Amendment No. 7#,JEJ,7N,180
DAEC-l' TELHNICAL SPECIFICATIONS LIST OF FIGURES Figure Number litle
- 1.1-1 Power / Flow Map 1.1-2 Deleted 2.1-1 APRM Flow Biased Scram and Rod Blocks 2.1-2 Deleted 3.3-1 Thermal Power vs. Core Flow limits for Thermal Hydraulic Stability Surveillance 4.1-1 Instrument Test Interval Determination Curves 4,2-2 Probability of System Unavailability Vs. Test Interval 3,4-1 Sodium Pentaborate Solution Volume Concentration Rec.uirements 3.4-2 Minimum Temperature of Sodium Pentaborate Solution 3.6-1 DAEC Operating Limits 4.8.C-1 DAEC Emergency Service Water Flow Requirement 6,2-1 Deleted I-Amendment-No. - 720,J42,7E7.7E4,7E6,767,180 vii
Hr. Lee Liu lowa Electric Light and Power Company Dua nt-Arnold Energy Center cc:
Jack Newman, Esquire Kathleen H. Shea, Esquire Newman and Holtzinger 1615 L Street, N.W.
Washington, D.C.
20036 Chairnan, Linn County Board of Supervisors Cedar Rapids, Iowa 52406 lowa Electric Light and Power Company ATTN: David L. Wilson Post Office Box 351 Cedar Ripids, Iowa 52406 U.S. Nuclear Regulatory Commission Resident inspector's Office Rural Route #1 Palo, Iowa 52324 Regional Administrator, Region 111 U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137 Mr. Stephen N. Brown Utilities Division Iowa Department of Commerce Lucas Office Building, 5th Floor Des Moines, Iowa 50319 l
I
o.
DAEC-1
- 34. yD TING VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during the process. Vent, used in system names, does not imply a VENTING process.
- 35. PROCESE_ CONTROL PROGRAM (PQEJ The PROCESS CONTROL PROGRAM shall generally describe she essential process controls and checks used to assure that a process for,olidifying radioactive waste from a liquid system produces a product that is acceptable for burial according to 10 CFR Part 61.56.
- 36. MEMBER (S) 0F THE PUBLIC Member (s) of the Public are persons who are not occupationally associated with lowa Electric Light and Power Company and who do not normally frequent the DAEC site..The category does not include contractors, contractor employees, vendors, or persons who enter the site to make deliveries or to service equipment.
- 37. SITE BOUNCARY The Site Boundary is that line beyond which the land is neither owned, nor leased, nor otherwise controlled by lELP.
UFSAR rigure 1.2-1 identifies the DAEC Site Boundary.
For the purpose of implementing radiological ef fluent technical specifications, the Unrestricted Area is that land (offsite) beyond the Site Boundary.
- 38. ANNUAL Occurring every 12 months, For the purpose of designating surveillance test frequencies, annual surveillance tests ar( to be conducted at least once per 12 months.
- 39. CORE OPERATING LIMITS rep 0RT The Core Operating Limits Report is the DAEC-specific document that provides cycle-specific operating limits for the current operating reload cycle. These cycle-specific operatirg limits shall be determined for each reload cycle in accordance with TS 6.11.2.
Plant operation within these limits is addressed in individual technical specifications.
- 40. SHUTDOWN MARGIN l
Shutdown margin is the amount of reactivity by wh'ch the reactor is subtritical or would be subcritical assuming all control rods are inserted, except for the analytically strongest worth control rod, which is fully withdrawn, with the core in its most reactive state during the OPERATING CYC;;
l i
Amendment No JO9,Jf 3,J67,180 1.0-9
-. ~
.DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT-
.C. -Control-Rod Block Actuation C.
Control Rod Block Actuation 1.
Instrumentation shall be Discharge ~ Volume Rod Blocks functionally tested, calibrated and checked as indicated in Table The Limiting Conditions of 4. 2-C.-
Operation for the instrumentation that initiates'these control rod System logic shall be' functionally
-block-are given in Table 3,2-C.
. tested as indicated in Table-4.2-C.
2.
Rod Block Monitor (RBM)
} -(a) The RBM control rod block setpoints are given in Table 3.2-C. - The upscale High-Power Trip Setpoint shall be applied when the core thermal power is-greater than or. equal to 85% of rated-(P>85%). The upscale Intermediate Power Trip Set-point shall be applied when the core thernal-power is greater than or equal to 65% of rated and less-thnn 85% of-rated-
_(65% 1 P < 85%). -_T_he upscals low Power _ Trip Setpoint shall
-be applied when the core thermal power is greater than or equal-to 30% of rated and Y
-less than 65% of rated (30% 1 P < 65%).- The RBM can be bypassed when core thermal power
-isiless than 30% of ratad. The RBM bypass time delay (,g )
shall be less than or eq0al to 2.0 seconds.
L -
l-F Amendment No. J70, 180 3.2-2 l-
'%W+A44-
-4:e l-ACg-4
.Jd Wex& 4 r A44+-4.
u,A DAEC-1 LIMITING' CONDITIONS-FOR OPERATION S'JRVEILLANCE REQUIREMENT 1
D.
Radiation Monitoring Systems-D.
Radiation Monitoring Systems-Isolation & Initiation Functionsf Isolation & Initiation Functions 1;
Steam Air Ejector Offgas System 1.
Steam Air Ejector Offgas System a) At least one post-treatment Instrumentation shall be steam air ejector offgas functionally tested, calibrated
. system radiation monitor and checked as indicated in shall be operable during Table 4.2.D.
reactor power-operation.
The monitors shall'be set System logic shall be to initiate immediate-functionally tested as indicated
. closure of the charcoal bed in Table 4.2-0.
bypass valve and the air ejector offgas isolation valve at a setting equivalent to or below the dose rate limits in Speci fication 3.15.2._1.
b) 'In the event no post-treatment monitor is operable, gases-from the
- steam air ejector offgas system may be released to the environment for up'to
- 72. hours provided (1) the charcoal bed of the offgas system is not bypassed.-and (2)-.the offgas: stack noble gas activity monitor is operable.
Otherwise be in at least HOT STANDBY within the.
following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Amendment No. 109,J26,J/B, 180 3.2-2a a
1 DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3 REACTIVITY CONTROL 4.3 REACTIV!TY CONTROL I
Applicability:
Applicability:
3 Applies to the operational status Applies to the surveillance of the control rod system.
requirements of the control rod system.
Objective:
Qhlective:
To assure the ability of the control To verify the ability of the control rod system to control reactivity.
rod system to control reactivity.
Specification:
1plcification:
A. Egattivity_ limitations A.
Rpactivity limitations
- 1. Reactivity marain - cqne loadina
- 1. Reactivit.y_ margin - core loadina A sufficient number of control rods Prior to or during the first startup shall be OPERABLE such that a following CORE ALTERATIONS, verify SHUTDOWN MARGIN of at least 0.38%
that the required SHUTDOWN MARGIN Ak/k exists or be in COLD SHUTDOWN exists by measurement during within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, control rod withdrawal.
2.
Reistly11.y marain - inonerable 2.
Reactivity margin'- inocerable control rods g.ontrol rods a.
If one control rod scram accumulator a.
At least once per week, during 5
E'""
Reactor Power Operation, verify the (i) verify reactor Pressure is pressure and level alarms for each greater than 950 psig and OPERABLE scram accumulator are not (ii) restore the accumulator to in the alarmed condition.
OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (iii) If the requirements of Specification 3.3.A.2.a (i) or (ii) cannot be met or greater than one accumulator is inoperable, the control rod (s) shall be declared inoperable and the actions stated in Specification 3.3.A.E.e shall be
- taken, b.
If a control rod (s) position cannot b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, be determined.he actions stated in determine the position of each declare the rod inoperable, l control rod' Specification 3.3.A.2.e shall be taken.
c.
Control rods with scram times c.
(not used).
l greater than those permitted by Specification 3.3.0.3 shall be declared inoperable.
A.nendment No.19,180 3.3-1
['*
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIRLMENT d.
Each control rod shall be coupled
- d. When a control rod is withdrawn the to its drive.* If a control rod first time after refueling, after becomes uncoupled, CRD maintenance or when required by (i) recouple the control rod within Specification 3.3.A.2.d(ii),
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and coupling integrity shall be (ii)verifycou$,A..d.by performing verified by observing that the drive lin surveillance 4.
does not go to the overtravel position when the rod is fully (iii) If the control rod is not withdrawn, recoupled, declare the control rod inoperable. The a.:tions stated in Specification 3.3.A.2.e shall be I
taken.
e.
A control rod that has 'een declared e.
(not used) l inoperable for reasons other than being stuck shall:
l (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) cannot be met, be in COLD SHUTDO.WN 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.5 and 3.9.A.6 for control rod requirements during refueling.
- The RWM may be bypassed, if required, to allow insertion of inoperabia control rodt and continued operation.
l l
Amendment No. 10J,720,JA2, 3.3-2 J43,180
DAEC-1 4
LIMITING CONDITION FOR OPERATION SURVEllLANCE REQUIREMENT
- f. A control rod which is not
- f. Vhenever the reactor is moveable with drive or scram operating greater than 201 pressure (stuck) shall be declared inoperable and the power
- following actions shall be taken.
(i) Disarm the associated control (i) each partially or fully rod drive and withdrawn operable control rod shall be demonstrated to be moveable by exercising it one notch at least once per week.
(ii) verify compliance with (ii) if a control rod cannot be Specification 3.3.A.1.
moved with drive or scram pressure, e8 pa W al h or M h M & awn (iii) Whenever the reactor is iess than 20% power, verify all OPERABLE control rod shall be inoperable control rods not in exercised one notch at least once compliance with BPWS are separated each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, unless it has been by 2 or more OPERABLE control rods determined that the failure is not in any direction, including the a failed control rod' drive i
- diagonal, mechanism collet housing, (iv) within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, verify that
-Se cause of the failure is not due to a failed control rod drive mechanism collet housing.
(v) if the requirements of Specification 3.3,A.2.f (1)-(iv) cannot be met or more than one control rod is stuck, be in COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.
Control Rod Driye Housina 3.
Control Rod Drive Housing Succort Succort The control rod drive housin9 The control rod drive housing support system shall be suprort system shall be in place inspected after reassembly and whenever the reactor vessel is the results of the inspection pressurized above atmospheric recorded.
pressure with fuel in tne reactor vessel.
l B.
Scram Discharae Volume B.
Scram Discharae Volume l
l (Not Used)
- 1. At least once per month, verify the SDV vent and drain valves are open.
1
- 2. At least once per quarter verify that the 50V vent and drain valves close a.
i within 30 seconds after receipt of Amendment No. J9,180 3.3-3
.,1 DAEC-1 y
LIMlilNG CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT a close signal and l
- b. after removal-of the close signal, the vent and drain valves are open.
- 3. At least once per OPERATING CYCLE, verify that a.
the SOV vent and drain valves close within 30 seconds after receipt of a signal for the control rods to scram and b.
the SDV vent and drain valves open when the scram signal is reset.
I C. REALTJylTY CONTROJ. SYSTEliS C.
REACTIyJTY LONTROL SYSTEMS l
1' Whenever the reactor is operating
- 1. Prior to the start of control rod l
at less than 20% power, the Rod withdrawal towards criticality and prior to obtaining 20% RATED POWER l
Worth Minimizer (RWM) shall be during rod insertion at shutdown, OPERABLE or the capability of the RWM shall be verified by the following checks.
- a. With the RWM inoperable after the a.
The correctness of the Banked first 12 reds are fully withdrawn.
Position Withdrawal Sequence (or operation may continue provided that a second licensed operator equivalent) input to the RWM verifies control rod. movement and computer shall be verified.
compliance with the prescribed control rod pattern.
b.
The RWM computer on-line diagnostic l test shall be successfully
- b. With the RWM inoperable before the
- wrformed, first 12 control rods aref ully f
withdrawn,onestartuppedfalendar c.
Proper annunciation of the l
year may be performed provided that selection error of at least one-a second licensed operator verifies out-of-sequence control rod in each control. rod movement and compliance with.the prescribed control rod fully inserted 9roup shall be pattern.
verified.
- c. Otherwise, with the RWM inoperabk.
d.
The rod block function of the.RWM control rod movement shall not be permitted except by a scram.
shall be verified by demonstrating the inability to withdraw an out-of-sequehce control rod.
c l 2. Control rods shall not be withdrawn 2.
Prior to control: rod. withdrawal in l l
in STARTUP or REFUEL modes unless STARTUP or REFUEL modes, verify that at least two Source Range Monitor l
Channels have an observed count rate at least two Source Range Monitor l
equal to or greater than.three Channels have an observed count rate counts per second.
of at least three counts per second.
'1 ' 3 - 4 Amendment No. 77 U2.180 l
+r--
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT l 3.
During operation with Limitin Control Rod Patterns, either:g 3
When a Limiting Control Rcd Pattern exists and one RBM channel is I a, both RBM channels shall be OPERABLE, inoperable, an Instrument or Functional Test of the operable RBM channel shall be performed within l b. withoneRBMchannelinoberable, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to rod withdrawal, control rod withdrawal s all be blocked within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, unless OPERABILITY is restored within this time period, or I c. with both RBM channels inoperable, control rod withdrawal shall be blocked until OPERABILITY of at least one channel is restored.
I D. ic_r.,3m I n se r t i on T i me s D. Ecram inser_tlpn Times l
1.
The average scram insertion time.
- 1. After each refueling outage all based on the deenergization of the OPERABLE rods shall be scram time scram pilot valve at time zero, of tested from the fully withdrawn all OPERABLE control rods in the position to the drop-out of the reed reactor power operation condition switch at the rod position required shall be no greater than:
by Specification 3.3.D.
The nuclear i system pressure shall be abeve 950 psig (with saturation temperature). [
This testing shall be completed Average Scram prior to exceeding 40% power.
Rod Insertion During all scram time testing below Position T imes (.}.el) 20% power, the Rod Worth Minimizer shall ba OPERABLE or a second 46 0.35 licensed operator shall verify that 38 0.937 the operator at the reactor console 26 1.86 is following the control rod 06 3.41 frogram.
2.
The average scram insertion times for the three fastest control rods of all groups of four control rods in a 2 x 2 array shall be no greater than:
Average Scram Rod Insertion Position Times (Sec) 46 0.37 38 1.01 26 1.97 06 3.62 3.
Maximum scram insertion time to rod position 04 of any OPERABLE control L
rod thould not exceed 7.00 seconds.
Amendment No. J20,JA2, 180 3.3-5
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 4.
If Specification 3.3.0.1, 2 or 3 cannot be met, be in COLD SHVTOOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
[ E. Reactivity Anomalies E.
Reactivity Anomalies i
The reactivity difference between the actual rod density and predicted The rod density shall be predicted rod density shall not exceed and compared to the actual rod 1% Ak/k, density:
- 1. If the reactivity is different by
- 1. during the first startup following more than 1% Ak/k, perform an analysis to determine and explain CORE AL1 ERAT 10NS and f rence operabn ontinue 2.
at least once per full power month. l if the differen:e-is exp ained and corrected.
- 2. Otherwise be in COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
1 F. Recirculatien Pumns F.
Recirculation Pumpj l
When the~ reactor mode switch is in j
STARTUP or RUN position, the reactor With two recirculation pumps in shall not be operated in the natural operation and with core thernal circulation flow mode, power greater than the limit With two recirculation pumps in operation and with core-thermal c re flow less than 45% of rated, power greater than the limit establish baseline APRM and LPRM*
specified in Figure 3.3-1 and total neutron flux noise levels within 2 core flow less than 45% of rated, hours, provided that baseline the APRM and LpRM* neutron flux noise levels shall be determined values have not been previously within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and:
established since the last core refueling.
. I 1.
if the APRM and LPRM* neutron flux noise levels are less than or equal to three times their established baseline levels, continue to determine the noise levels 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 also within 30 minutes after the completion of a core. thermal power inc. ease of at least 5% of rated core thermal power while operating in this region of the power / flow map, or i 2.
if the APRM and/or LPRM* neutron flux noise levels are greater than three-times their established baseline levels, immediately initiate corrective action and restore the roise levels to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by increasing core flow, and/or by-initiating an orderly reduction of Amendment No. JJA,JJ7,J AJ,180
DAEC-1 l
SURVEILLANCE REQUIREMENT LIMITING CONDITION FOR OPERATION core thermal power by inserting control rods.
See Specifications 3.6.F.2 for SLO.
A recirculation pump shall not be started while the reactor is in natural circulation flow and reactor power is greater than 1% of RATED POWER.
i~
l
- Detector levels A and C of one LPRM string per core octant plus detector levels A and C of one LPRM string in the center of the core shall be monitored.
l 1
Amendment No.19,33,JJ9,180
~
DAEC-1 3.3 and-4.3 BASES A.
Reactivity Limitation
- 1. Reactivity biargin - Core Loading I
lhe requirements for the control rod drive system have been identified by evaluating the need for reactivity control via control rod movement over the full spectrum of plant conditions and events. As discussed in Subsection 4.6.1 of the Updated FSAR, the control rod system design is intended to-provide sufficient control of core reactivity that the core could be made subcritical with the strongest red fully withdrawn.
This reactivity characteristic has been a basic assumption in the analysis of plant performance.
Verification of this required shutdown margin is performed prior to or during the first startup after core alterations by measurement during control rod withdrawal.
This demonstration is performed with the reactor core in the cold, xenon-free condition and will show that the reactor is subcritical by at least R 4 0.38% Ak/k with l the analytically determined strongest control rod fully withdrawn.
(Shutdown margin can also be verified, when actual demonstration is not feasible, by an analytical determination of the highest rod worth and core reactivity.)
The value of "R", in units of % M/k, is the amount by which the core i
reactivity, in the most reactive condition at any time in the subsequent operating cycle, is calculated to be greater than at the time of the demonstration.
"R", therefore, is the dif ference between the calculated value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity.
The value of "R" must be positive or zero and must be determined for each fuel cycle.
In determining the " analytically strongest" rod, it is assumed thet every fuel assembly of the same type has identical material properties.
In the actual core, however, the control cell material properties vary within allowed manufacturing tolerances, and the strongest rod is determined by a combination of the cnntrol cell geometry and local k m.
Thereforo, an additional margin is included in the shutdown margin test to account for the fact that the " analytically strongest" rod is not necessarily the strongest rod in the core. Studies have been made which compare experimental criticals with calculated criticais.
These studies have shown that actual criticals can be predicted within a given tolerance band.
For gadolinia cores the additional margin required due to control cell material manufacturing tolerances and calculational uncertainties has experimentally been determined to be 0.38% ok/L. When i this additional margin is demonstrated, it assures that the reactivity control requirement i.s met.
2.
Reactivity margin - inoperable control rods l
Control rod operability (the capability to insert the control rods) ensures that the assumptions for scram reactivity in the safety analyses are not violated.
Operability of an individual control rod 4 t based on a combination of factors, primarily the scram insertion times, control rod scram accumulator status, control rod cnupling integrity, control Amendment No. JJ A,141,180 3.3-8
~
DAEC-1 s
rod movability and the ability to determine a control rod's position.
When a rod is declared inoperable, strict control over the number and distribution of inoperable control rods is required for the assumptions of the safety analyses and to provide early indication of potential generic problems in the CRD system.
The control rod drive scram acctmulators are rart of the CRD system and are provided to ensure adequate control rod s; ram under varying reactor conditions.
The safety analyses assume that all of the control rods scram at a specified insertion rate.
Surveillance of the control rod scram accumulator provides assurance (along with the other surveillances of control rod operability and scram insertion times) that the scram reactivity assumed in the safety analysr.s can be met.
Because of the large number of control rods available for scram and the assumed single failure of a control rod to scram in the Safety Analysis, a specified amount of time is allowed to restore the accumulator to OPERABLE status.
This time is only allowed, however, if reactor pressure is sufficient it scram the rod without help of the accumulator.
If the accumulator cannot be restored to OPERABLE Status, the associated control rod could potentially have a degraded scram insertion speed and therefore must be declareu inoperable.
Control rod position information is required to ensure adequate information is available to the operator for ;ietermining CR0 operability and controlling rod patterrs.
However, if a rod's position is not displayed, a control rod's position can be determined by moving the control rod to a position with an OPERABLE indicator or by the use of other appropriate metheds.
Control rod drop accidents as discussed in the Updated FSAR can lead to significant core damage, if coupling integrity is maintained, the possibility of a rod drop accident is eliminated.
The overtravel position feature provides a positive check as only uncoupled drives may reach this position.
Continued operation with an uncoupled control rod should not be allowed because of the increased probability of a CRDA.
Therefore, only a short period of timo is allowed to establish and verify coupling.
Since the allowable time with an uncoupled control rod is short, and control rods do not always couple on the first try, multiple attempts to recouple a control rod may be necessary.
Specification 3.3. A.2.e requires that a rod be taken out of service if it is declared inoperable.
It is required to be fully inserted and disarmed electrically
- to ensure it is in a safe position of maximum contribution to shutdown margin and to prevent inadvertent withdrawal during subsequent operations.
Consideration of the control rod drop accident (CRDA) requires that the inoperable inserted control rods not
- To disarm the drive electrically, four Amphenol type plug connectors are removed from the drive insert and withdrawal solenoids rendering the rod incapable of withdrawal.
This procedure is equivalent to valving out the drive and is prefer ed because, in this condition, drive water cools and minimizes crud accumulation in the drive. Electrical disarming does not eliminate position indication.
Amendment No. O,#3,180 3.3-9
DAEC-1 s
in compliance _with BPWS (out-of-sequence) be separated by at least two J
operable control rods in any direction including the diagonal and that j
no more than 8 control rods are declared inoperable.
Controlling the distribution of out-of-sequence control rods limits the potential reactivity worth of adjacent control rods. -Limiting the number of inoperable rods to less than or equal to 8 ensures that possible generic 4
3 problems are investigated and resolved.
If a control rod is declar?d inoperable and cannot be fully inserted (stuck), its position must be verified to be in compliance with the required SDM.
This assures that the core can be shutdown at all times 4
with the remaining control rods assuming the strongest operable control rod does not insert.
In addition, the control rod should be isolated from a scram to protect the CRD and surrounding fuel assemblies should a scram occur. The CRD can be isolated from scram by isolating the hydraulic control unit from scram and normal insert / withdraw pressure yet still maintain cooling water to t he CRD. Once this is done, the accumulator should be depressurized.
If the control rod is immnvable and damage within the control rod drive l mechanism and, in particul0r, cracks in the drive internal housings cannot be ruled out, then a generic problem affecting a number of drives l
cannot be ruled out.
Circumferential cracks resulting from stress i-assisted intergranular corrosion have occurred in the collet housing of drives at several BWRs.
This type of cracking could occur in a number of drives and if the cracks propagated until severance of the collet housing occurred, scram could be orevented in the affected rods.
Limiting the period of operation with a potentially severed collet housing and requiring increased surveillance after detecting one stuck rod will assure that the reactor will not be operated with a large number of rods with failed collet housings.
3.
Control Rod Drive Housing Support I
}
The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the extremely remote event of a housing failure.
The amount of reactivity which could be added by this small amount oi rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage to the primary coolant system. The design basis is given in Subsection 4.6.1 of the Updated FSAR and the safety evaluation is given in Subsection 4.6.2 of the Updated FSAR.
This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing.
B.
1.
To ensure that a volume exists to accept discharge water from the control rods during a reactor scram, the scram discharge volume (SDV) vent and drain valves are required to undergo surveillance testing.
For the Amendment No..T9, 180 3.3-10
DAEC-1 monthly verification of SDV vent and drain valve status, observation of control room indicating lights is acceptable.
C. Reactivity Control Systems l
- 1. The RWM restricts withdrawals and insertions of control rods to l
prespecified sequences.
These sequences are established such that the drop of any in-sequence control rod from the fully inserted position to the position of the control rod drive would not cause the reactor to sustain a power excursion resulting in a peak fuel enthalpy in excess of 280 cal /gm.
An enthalpy of 280 cal./gm. is well below the level at which rapid fuel dispersal could occur (i.e., 425 cal./gm.).
Primary system damage in this accident is not possible unless a significan* amount of fuel is rapidly dispersed.
Ref.
Subsections 4.3.1, 7.7.4.9, and 15.4.7 of the Updated FSAR.
These control' rod patterns arc in accordance with the Danked Position Withdrawal Sequence (BPWS) (Ref. 1).
The BPWS has the advantage of having been proven statistically to have such low individual control rod worths that the possibility of a control rod drop accident (CRDA), which exceeds the 280 cal /gm peat fuel enthalpy limit, is precluded (Ref. 2).
The Reduced Notch Worth Procedure (RNWP) may be used in place of BPWS because the RNWP is an extension of BPWS (Ref. 3).
In performing the function described above, the RWM is not required to impose any restrictions at core power levels in excess of 10% of rated.
Material in the cited references shows that it is impossible to reach 280 cal /gm in the event of a controi rod drop occurring at power greater than 10%, regardless of the rod pattern.
This is true for all normal and abnormal patterns including those which marinize the individual rod worth.
Power level for automatic cutout of the RWM function is sensed by feedwater and steam flow and is set nnminally at 20% of rated power to account for instrument error.
The RWM provides automatic supervision to assure that oat-of-sequence I
control rods will not be withdrawn or inserted; i.e., it limits operator deviations from planned withdrawal sequences. The RWM serves as a backup l to procedural control of control rod sequences, which limit the maximum reactivity worth of control rods, in the event that the Rod Worth Minimi7er is out of service at less than 20% of rated power, a second Licensed Operator or other qualified technical plant employee whose qualifications have been reviewed by the NRC shall verify control rod movement and compliance with the prescribed control rod pattern.
'The functions of the RWM make it unnecessary to specify a license limit I on rod worth to preclude unacceptable consequences in the event of a CRDA.
At low powers, below 20%, this device forces adherence to l
acceptable rod patterns. Above 20% of rated power, no constraint on rod pattern is required to assure that tha consequences of a CRDA are acceptable.
I Amendment No.19,J42,180 3.3-11
~
t DAEC-1 Functional testing of the RWM prior to the start of control rod withdrawal at startup, and prior to attaining 20% rated thermal power during rod insertion while shutting down, will ensure reliable operation.
If a rod is declared inoperable, adherence to the BPWS (and RNWP) is maintained by performing those actions required for an inoperable rod.
These actions require fully inserting the inoperable rod, disarming it electrically, ensuring it is separated from other inoperable rods by at least two operable rods in any direction (if it is out-of-sequence) and having a maximum of 8 inoperable rods.
The operability requirements for the RWM have been established to minimize reactor operations without the RWM.
If the operability requirements of the RWM are not satisfied, i.e., RWM is inoperable without the second licensed operator or the BPWS (RNWP) requirements for inoperable rods are not met (below 20Y rated), then further rod movement is not permitted, except by a scram (manual scram or mode switch to SHUTDOWN).
This is done to ensure that high rod worths, with the potential to exceed 280 cal /gm during a CRDA are not generated. However, limited rod movement shall be permitted solely for the purpose of troubleshooting and/or testing the RWM for OPERABILITY.
Limited rod movement is defined as the movement of control rod (s) only to the extent necessary to determine that the rod inhibit functions of the RWM are working properly.
2.
The $ource Range Monitor (SRM) system performs no automatic safety system l function; i.e., it has no scram function.
It does provide the operator with a visual indication of neutron level.
The consequences of reactivity accidents are functions of the initial neutron flux.
The requirement of at least 3 counts per second assures that any transient, should it occur, becins at or above the initial value of 108 of rated power used in the ahalyses of transients in cold conditions.
One I
operable SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal.
A minimum of two operable SRM's are provided as an added conservatism.
- 3. The RBM provides le W protection of the core; i.e.,
the prevention of l
boiling transition in a local region of the core, for a single red withdrawal error from a Limiting Control Rod pattern.
The trip point is referenced to power.
This power signal is provided by the APRMs.
A statistical analysis of many single control rod withdrawal errors has
-been performed and at the 95/95 level the results show that with the specified trip settings, rod withdrawal is blocked at MCPRs greater than the Safety Limit, thus allowing adequate margin.
This analysis assumes a steady state MCPR of 1.20 prior to the postulated rod withdrawal error.
The RBM functions are required whan core thermal power is greater than 30's and a Limiting Control Red Pa uern exists. When both RBM channels are operating either channel will assure required withdrawal blocks occur even assuming a single failure of one channel.
When a Limiting Control Rod Pattern exists, with one RBM channel inoperable for no more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, testing of the RBM prior to vithdrawal of control rods assures that improper control rod withdrawal will be blocked (Reference 4).
l Requiring at least half of the normal LPRM inputs to be operable assures that the RBM response will be adequato to protect against rod withdrawal errors, as shown by a statistical failure analysis.
Amendment No. JJ#,JA1,180
- 3. N 2
4 OAEC-1 The RBM bypass time delay is set low enough to assure minimum rod movement while upscale trips are bypassed.
i A Limiting Control Rod Pattern for rod withdrawal error (RWE) exists when (a) core thermal power is greater than or equal to 30% of rated and less than 90%'of rated (30% 5 P < 90%) and the MCPR is less than 1.70, or (b) core thermal power is greater than or equal to 90% of rated (P 2 90%) and the MCPR is less than 1.40.
During the use of such patterns, it is judged that testing of the RBM-channel (when one channel is inoperable) prior to withdrawal of such rods l to assure its operability will assure that improper withdrawal does not occur.
D. Scram Insertion Times l
The control rod system is designed to bring the reactor subtritical at a rate fast enough to prevent fuel damage; i.e., to prevent the MCPR from becoming less than the safety limit.
After initial fuel loading and subsequent refuelings when operating above 9b0 psig, all control rods shall be scram tested within the constraints in, posed by the Technical Specifications and before the 40% power level is reached. The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.
E. Reactivity Anomalies i
During each fuel cycle excess operative reactivity varies as fuel depletes and as any burnable poison in supplementary control is burned. The magnitude of this excess reactivity may be inferred from the critical ro '
configuration, As fuel burnup progresses, anomalous behavior,n the excess reactivity may be detected by comparison of the critical rod pattern at selected base states to the predicted rod inventory at that state.
Power operating base conditions provide the most sensitive and directly interpretable data relative to core, reactivity.
Furthermore, using power operating base conditions permits frequent reactivity comparisons.
Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds 1% Ak/k.
Deviations in core reactivity greater than 1% Ak/k are not i
expected and require thorough evaluation. One percent reactivity limit is considered safe since an insertion of the reactivity into the core would not lead to transients exceeding design conditions of the reactor system.
F. Recirculation Pumps l
APRM and/or LPRM oscillations in excess of those specified in section 3.3.E could be an indication that a condition of thermal hydraulic instability exists and that appropriate remedial action should be taken.
These specifications are based upon the guidance of GE SIL #380., Rev. 1, 2/10/84 Amendment No. J M, 180
.m
,,,, a r,e 4
7g.yj, g
DAEC-1 3.3 and 4.3-REFERENCES
[La.nked Position Witt drawal Seave,ltcg, NEDO-21231, January 1977.
l 1.
t
- 2. General Electric Standard Application for R actor Fye.1, NEDE-240ll-P-A*
1 l
3.
General Electric Service Information Letter (SIL) No. 316, Reduced Nottb yorth Proq3 dure, November 1979.
l 4.
Averace Power Range Monitor. Rod B h;.k Monitor and Technical lpecification improvement (ARTS) Program for lhe Ovang Arnold Enerm Center, NEDC-30813-P, December 1984.
- Latest NRC-approved revision.
L Amendment No.180 3.3-14
DAEC-1 l
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i 20 20 --
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- 50 60 70 80 Total Core Flow (t of Rated)
DAEC Iowa Electric Light and Power Company l
Technical Specifications Thermal Power vs Core Flow limits for Thermal Hydraulic Stability Surveillance Fiqure 3.3-1 Amendment Nu 180 3.3-15
[
DAEC-1 l
l LIM 111NG CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT Ak at any time during the maintenance with the strongest remaining operable control rod fully withdrawn and all other operable rods fully inserted.
Alternatively if the remaining control rods are fully inserted and have had their directional control valves electrically disarmed, it is sufficient to dersonstrate that the core is subcritical with a margin of at least 0.38% At at any time during the maintenante. A control rod on which maintenance is being performed shall be considered inoperable.
3.
The fuel grapple hoist load switch 3.
Observe that any drive which has shall be set at 5 400 lbs.
been uncoupled from and subsequently coupled to its control rod does not go to the overtravel position.
=
4.
If the frame-mounted auxiliary hoist, the monorail-mounted auxiliary hoist, or the service platform hoist is to be used for handling fuel with the head off the reactor vessel, the load limit switch on the hoist to be used shall be set at 5 400 lbs.
5.
A maximum of two nonadjacent control rods may be withdrawn from the core for the purpose of performing control rod and/or control rod drive maintenanc.e, provided the foilowing conditions are satisfied:
a.
The reactor mode switch shall be locked in the " refuel" position.
The refueling interlock which prevents more than one control rod from being withdrawn may be bypassed for one of the control rods on which maintenance is being performed. All other refueling interlocks shall be operable.
3.9-2 Amendment No.180 5? tiff