ML20108D136
| ML20108D136 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/29/1996 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20108D129 | List: |
| References | |
| NUDOCS 9605070290 | |
| Download: ML20108D136 (58) | |
Text
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ATTACIIMENI 10 JCAN039604 PROPOSED TECHNICAL SPECIFICATION ,
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l RESPECTIVE SAFETY ANALYSES ;
l IN THE MATTER OF AMENDING l i
LICENSE NO, DPR-51 !
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ENTERGY OPERATIONS, INC.
ARKANSAS NUCLEAR ONE, UNIT ONE l DOCKET NO. 50-313 ;
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9605070290 960429 PDR ADOCK 05000313 PDR p
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l April 23,1996 ICAN049606 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station PI-137 Washington, DC 20555 l
Subject:
Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 Proposed Technical Specification Change Revising The Pressurizer Code Safety As-Found Setpoint Tolerance Gentlemen:
Attached for your review and approval are proposed Technical Specification (TS) changes to allow Arkansas Nuclear One - Unit 1 (ANO-1) to revise the as-found setpoint tolerance for the pressurizer code safeties described by the Bases associated with Specifications 2.2 and 3.1.1 from +1/-3% to i3%. The changes also increase the relief flowrate of the pressurizer code safeties described in the Bases associated with Specification 3.1.1 from 300,000 lb/hr to 324,000 lb/hr, reword the Bases associated with Specification 3.1.7 to describe the actual value of moderator temperature coefficient used as an input to the startup accident analysis, -
and revise the values for minimum and maximum pressurizer water level specified by Specification 3.1.3.4 to refer to a figure that.will be incorporated in this change. These changes are supported by revised startup accident and rod withdrawal accident analyses.
Proposed changes to the ANO-1 Safety Analysis Report incorporating the new analysis results
- have also been included for your use in reviewing this change request.
The new startup and rod withdrawal accident analyses were performed using the RELAP5/ MOD 2-B&W computer code to justify an increase in pressurizer code safety valve as-found tolerance to +3%. The analyses verified, using conservative assumptions, that a
+3% tolerance is acceptable for two pressurizer code safety valves. The analyses also showed that a maximum pressurizer water level of 259 inches below 15% Rated Power and a maximum level of 320 inches when at or above 15% Rated Power produces acceptable i results.
Currently, when a pressurizer code safety valve setpoint is found to be outside of the +1/-3%
- setpoint tolerance, the other pressurizer code safety valve must be tested and the occurrence must be tracked under the ANO 10CFR50 Apperdix B corrective action program. With this l
.ev. ,
U< S. NRC ]
April 23,1996 1CAN049606 Page 2 change, those occurrences when a pressurizer code safety valve setpoint is found outside of a i
+1% setpoint tolerance, but within the proposed +3% setpoint tolerance, would not require testing of the other pressurizer code safety valve and would not require tracking of the corrective action. The change still requires any valve setpoint found to be outside of a 1%
tolerance be returned to within the il% as-left tolerance as currently described in the Bases associated with Specification 2.2.
Entergy operations currently utilizes the 1980 Edition of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code at ANO-1. Subsection IWV-3512 of this Edition of the Code endorses ASME Performance Test Code (PTC) 25.3-1976 for the testing of safety and relief valves. The pressurizer code safety valves are currently tested in accordance with this standard. i In accordance with 10CFR50.55a(f)(4)(iv), Entergy Operations requests approval to use the 1989 Edition of Section XI of the ASME Code to test the ANO-1 pressurizer code safety ,
valves beginning with testing to be conducted during our next refueling outage which is currently scheduled to commence on September 17, 1996. This Edition, which has been incorporated by reference in 10CFR50.55a(b)(2), endorses ASME/American National Standt.rds Institute (ANSI) Operations and Maintenance (OM) Code, Part 10 [OMa-1988 Addenda to the OM-1987 Edition per 10CFR50.55a(b)(2)(viii)]. This Edition of OM Part 10 ;
endorses OM Part 1 (1987), and allows a i3% tolerance for as-found testing of safety valves.
In adopting the 1989 ASME Code for pressurizer code safety valve testing, Entergy ,
Operations commits to adopt all the related requirements of OM Part 1. The ANO-1 safety analysis was reviewed and determined to be unaffected by this change in tes'ting requirements.
The proposed TS change has been evaluated in accordance with 10CFR50.91(a)(1) using criteria in 10CFR50.92(c) and it has been determined that this change involves no significant hazards considerations. The bases for these determinations are included in the attached submittal. ;
Entergy Operations requests that the effective date for this TS change be within 30 days of approval. Although this request is neither exigent nor emergency, your prompt review is requested prior to our next refueling outage.
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U. S. NRC April 23,1996 ICAN049606 Page 3 Very truly yours, JWY/cws Attachments To the best of my knowledge and belief, the statements contained in this submittal are true.
SUBSCRIBED AND SWORN TO before me, a Notary Public in and for County and the State of Mississippi, this day of ,1996.
i Notary Public My Commission Empires
i .,- ,e U. S. NRC April 23,1996 1CAN049606 Page 4 cc: Mr. Leonard J. Callan Regional Administrator i U. S. Nuclear Regulatory Commission Region IV ,
611 Ryan Plaza Drive, Suite 400 !
Arlington, TX 76011-8064 NRC Senior Resident Inspector l Arkansas Nuclear One P.O. Box 310 London, AR 72847 Mr. George Kalman NRR Project Manager Region IV/ANO-1 & 2 l U. S. Nuclear Regul.atory Commission NRR Mail Stop 13-H-3 i One White Flint North l 11555 Rockville Pike l Rockville, MD 20852
- Mr. Bernard Bevill l Acting Director, Division of Radiation j Control and Emergency Management 1 l Arkansas Department of Health l 4815 West Markham Street Little Rock, AR 72205 j
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l ATTACHMENT l TD 1CAN039604 I
PROPOSED TECHNICAL SPECIFICATION AND RESPECTIVE SAFETY ANALYSES IN THE MATTER OF AMENDING l l
LICENSE NO. DPR-51 ENTERGY OPERATIONS, INC.
ARKANSAS NUCLEAR ONE, UNIT ONE l
DOCKET NO. 50-313 I
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? .e- .o Attachment to l
ICAN049606
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l DESCRIPTION OF PROPOSED CHANGES l l
The proposed changes to the Arkansas Nuclear One - Unit 1 (ANO-1) Technical !
Specifications (TSs) are as follows:
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e The Bases associated with Specification 2.2 were revised to reflect a new pressurizer code )
safety setpoint as-found tolerance ofi3%. '
e The Bases associated with Specification 3.1.1 were revised to reflect a new pressurizer j code safety setpoint as-found tolerance ofi3% and pressurizer code safety valve relief l flowrate of 324,000 lb/hr. i 1
e The pressurizer water level requirements of Specification 3.1.3.4 have been revised to l
refer to Figure 3.1.3-1, Pressurizer Level Acceptable Range. ;
- A new page has been inserted to allow incorporation of a new figure. Figure 3.1.3-1 I shows the acceptable ranges for pressurizer water level as a function of reactor power, as !
required by the revised Specification 3.1.3.4.
- The Bases associated with Specification 3.1.7 were revised to indicate the actual value of moderator temperature coeflicient used as an input in the startup accident analysis instead I of the currently specified range of values reference.
BACKGROUND The reactor coolant system (RCS) serves as a barrier which prevents the release of radionuclides contained in the reactor coolant to the reactor building atmosphere. A pressure safety limit of 2750 psig (110% of design pressure) has been established and is specified by TS 2.2.1. The RCS is protected against overpressure by two pressurizer code safety valves mounted on top of the pressurizer. The ANO-1 pressurizer code safety valves are Dresser mndd 31759A Ap-to-open, spring-to-close pressure relief valves. The required capacity of these valves is determined from considerations of: (1) the reactor protection system, (2) pressure drop (static and dynamic) between the point of highest pressure in the RCS and the pressurizer, and (3) accident or transient overpressure conditions. The pressurizer code safety valves are described in ANO-1 Safety Analysis Report (SAR) Section 4.2.4.1.
TS Table 4.1-2 requires testing of one pressurizer code safety valve setpoint every 18 months.
Currently, the Bases associated with TS 2.2 state that the as-found lift setpoint may be 2500 psig +1/-3%. If the setpoint is found to be outside of a il% tolerance band, it must be reset to 2500 psig il%. If the setpoint is found to be outside of the +1/-3% tolerance band, the remaining pressurizer code safety valve setpoint must be tested in accordance with Section III of the ASME Code (PTC 25.3). ASME/ ANSI OM Part 1 (1987) allows a 3%
l tolerance band for the as-found testing of code safety valves.
Testing results since IM89 are summarized in Figure #1 attached to this submittal) for all three ANO-1 pressurizer code safety valves (two valves a e in service and one is a spare).
[ . a. - .v l i Attachment to I
! 1CAN049606 l Page 2 of 5 l
As shown in Figure #1, two of the twelve tests conducted since IM89 were not within the l proposed setpoint tolerance of+3/-3%. The high setpoints both occurred during IRI1 and were attributed to the practice of" jack and lap" after setpoint testing. This process allowed the valve to be partially disassembled, leaving the spring intact, in order to lap the seats to eliminate post testing leakage. AAer the process was completed, the valve was re-assembled l without further setpoint testing. Based on recent information, this repair method can not be l l used on the Dresser valves without re-verifying the setpoint because the valves utilize four spiral wound gaskets between the body to bonnet interface. Since the gaskets may not i l- compress to the same degree aAer re-assembly, the spring compression could change thus l l affecting the setpoint. After both valves lifted out of tolerance during IRI1, the spare valve l which had been in storage since IR10 was tested, and also lifted out-of-tolerance. Since this ;
valve had also been " jack and lapped" without re-verifying the setpoint in IR10, the practice i of" jack and lap" without subsequent setpoint verification was determined to be questionable.
Now, if a valve is " jack and lapped," its setpoint must be re-verified. !
During IR12, PSV-1001 lifted 1.5% above setpoint. Because the valve lifted out of the current setpoint tolerance of+1/-3%, PSV-1002 also had to be tested to meet code requirements. PSV-1002 was found 0.64% above it's setpoint. Expanding the setpoint i tolerance range to i3% would reduce the likelihood of a valve being found out of tolerance. l This in turn would reduce the probability of subsequent valve testing during each outage. !
DISCUSSION OF CHANGE ;
1 The two limiting accidents identified in the TS 3.1.3.4 Bases with respect to pressurizer code safety valve response are the startup accident (SAR Section 14.1.2.2) and the rod withdrawal accident (SAR Section 14.1.2.3). Analyses have been performed to demonstrate the acceptability of a +3% pressurizer code safety valve setpoint tolerance in the event of a 1 I
startup accident or a rod withdrawal accident. The methodology for analyzing these accidents is identical to that employed in the ANO-1 Safety Analysis Repon using an improved computer code - RELAP5/ MOD 2-B&W. The acceptance criteria for these analyses are: (1)
Peak RCS pressure must remain below the safety limit of 2750 psig, and (2) peak reactor thermal power must remain below 112% Rated Power. All computer analyses were performed using the RELAPS/ MOD 2-B&W computer code. The RELAP5\ MOD 2 code has been previously submitted to the NRC for review in B&\ opical Report BAW-10193P, "RELAP5/ MOD 2-B&W For Safety Analysis of B&' Jesigned Pressurized Water Reactors," dated August 14,1995. A comparison of the RELAPS/ MOD 2 - B&W prediction of the startup accident with that of CADDS, an approved code for analyzing this event for B&W-designed pressurized water reactors, was provided in BAW-10193P.
- The analysis demonstrates that a startup accident from hot zero power with a pressurizer code ,
- safety valve setpoint tolerance of 3% above the pressurizer code safety valve setpoint of 2500 l
! psig will not result in a peak RCS pressure greater than 2750 psig or a reactor thermal power ;
) greater than 112% Rated Power. This analysis included additional sensitivity studies that !
demonstrated acceptable results in the event of a stanup accident assuming one pressurizer j code safety valve lifted at 5% above the pressurizer code safety valve setpoint of 2500 psig t
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1CAN049606 Page 3 of 5 and the other pressurizer code safety valve failing to actuate to relieve RCS pressure. A w bounding value for moderator temperature coefficient, +0.9 x 10" AK/K/ F, was assumed in ,
the analysis in place of the range of coefficients referred to in the Bases associated with TS , 3.1.7.
The Bases associated with Specification 3.1.3.4 indicate that the specified pressurizer levels assure that the reactor coelant system cannot become solid in the event of a rod withdrawal accident or a startup accident and that the water level is above the minimum detectable level.
The Bases do not, however, specifically state which pressurizer levels are analytically justified l for any specific power levels. In other words, the Bases do not indicate what initial pressurizer level was assumed in either the startup accident or control rod withdrawal accident. The original analyses did, however, employ conservative methods and setpoints l while utilizing nominal values for the operationa: parameters.
It was recognized that a more appropriate requirement for pressurizer level was necessary to accommodate the thermal expansion associated with the reactivity addition and the conservative assumptions used in the startup and rod withdrawal event analyses. The l operational range for pressurizer level is approximately 140 inches at 0% Rated Power, and l approximately 220 inches at 100% Rated Power. The startup accident design analysis, using j conservative input assumptions, justified a maximum pressurizer level of 259 inches. Since l postulated rod withdrawal events at higher power levels are considered to have less severe
- consequences due to the effects of the assumed power level on the input assumptions, this l limit was considered unnecessarily restrictive for operation above 15% Rated Power.
A control rod withdrawal analysis was performed at a power level of 15% Rated Power to support the proposed setpoint tolerance change. This analysis is considered to be bounding from 15% Rated Power to 100% Rated Power due to the ramping of the moderator l temperature coefficient from a value of +0.9x10" Ak/k/ F at 0% Rated Power to a value of
+0.0x10" Ak/k/ F at 95% Rat:d Power. The analysis provided acceptable results, assuming i an initial pressurizer level of 320 inches when the unit is above 15% Rated Power.
Proposed changes to the ANO-1 SAR incorporating the new analysis results have been l included for your use in reviewing the proposed TS changes. Based on these analyses, i ANO-1 proposes to revise the as-found pressurizer code safety valve setpoint tolerance to i3%. If found outside of a il% tolerance band, the pressurizer code safety valve setpoint will continue to be reset to 2500 psig il%, as required by Section III of the ASME Code and as described in the Bases associated with Specification 2.2. The Bases associated with !
Specification 3.1.1 have been revised to reflect the change in as-found tolerance, and to reflect the pressurizer code safety valve relief flowrate of 324,000 lb/hr used in the reanalysis of the startup and rod withdrawal accidents. The pressurizer code safety valve relief flowrate was revised from 300,000 lbm/hr to 324,000 lbm/hr to reflect the actual relief capacity of the l pressurizer code safety relief valve and to remove excess conservatism from the analyses. The Bases associated with Specification 3.1.7 have been revised to describe the value of l moderator temperature coefficient used in the startup accident analysis as a bounding value.
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r-l . ,. , o l Attachment to l 1CAN049606 l Page 4 of 5 A new figure, Figure 3.1.3-1, has been added on inserted page 21a specifying the acceptable range for pressurizer level as a function of reactor power. The minimum pressurizer level for all power levels remains at the currently specified 45 inches. From 0% to 15% Rated Power, the pressuri:.er maximum water level is 259 inches. From 15% to 100% Rated Power, the pressurizer maximum water level is 320 inches.
Figure 3.1.3-1 also contains a note to clarify that the specified pressurizer levels and reactor power levels do not contain an allowance for instrument error. The previous pressurizer level requirements were specified as " indicated" levels. No reference was made in the associated I Bases to indicate whether instrument error was included in these limits. Since the values for pressurizer level and reactor power used as inputs to the startup and rod withdrawal analyses were not corrected by the inclusion ofinstrument error, this note indicates that values used for I controls in the plant procedures should be corrected for the instrument error allowance. l l
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION l I
An evaluation of the proposed change has been performed in accordance with 10CFR50.91(a)(1) regarding no significant hazards considerations using the standards in 10CFR50.92(c). A discussion of these standards as they relate to this amendment request follows: 1 Criterion 1 - Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.
The startup accident and the rod withdrawal accident have been reanalyzed to justify the proposed increase in pressurizer code safety valve as-found tolerance. The analyses establish more appropriate boundaries and re-analyze the same initiators as are currently found in the ANO-1 Safety Analysis Report. Changing the as-found setpoint tolerance does not change how the pressurizer code safety valve operates as it will continue to be reset to 2500 psig il% prior to reactor startup.
The acceptance criteria for these analyses are that the reactor coolant system (RCS) pressure shall not exceed the safety limit of 2750 psig (110% of design pressure) and that the reactor thermal power remains below 112% Rated Power. The analyses using the proposed setpoint tolerance have shown that the acceptance criteria were met and that the consequences of the events were essentially the same as those in the ANO-1 SAR. Analyses were performed to determine the pressurizer maximum water level that would prevent the RCS from exceeding i the safety limit of 2750 psig in the event of either a startup accident or a rod withdrawal I
accident. More appropriate pressurizer level requirements have been incorporated in accordance with these analyses.
Therefore, this change does n_qt involve a significant increase in the probability or consequences of any accident previously evaluated.
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1CAN0496% ,
Page 5 of 5 Criterion 2 - Does Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated.
The proposed changes introduce no new mode of plant operation. The reanalysis of the l startup accident and the rod withdrawal accident were performed using methodologies f identical to that employed in the ANO-1 SAR and an improved computer code (RELAP5/ MOD 2). The pressurizer code safety valve setpoint will continue to be reset at 2500 psig il% prior to reactor startup and will continue to function to maintain RCS pressure below the safety limit of 2750 psig. Analyses were performed to determine the pressurizer maximum water level that would prevent the RCS from exceeding the safety limit of
- l. 2750 psig in the event of either a startup accident or a rod withdrawal accident. More l . appropriate pressurizer level requirements have been incorporated in accordance with these analyses.
l Therefore, this change does nat create the possibility of a new or different kind of accident i from any previously evaluated. ,
l Criterion 3 - Does Not Involve a Significant Reduction in the Margin of Safety.
The safety function of the pressurizer code safety valves is not altered as a result of the proposed change in setpoint tolerance. The reanalysis of the startup accident and rod withdrawal accident have shown that with a i3% setpoint tolerance,' the pressurizer code safety valves will function to limit RCS pressure below the safety limit of 2750 psig. The sensitivity studies for the startup accident showed the acceptance criteria would still be met l even if one pressurizer code safety valve lifted at 5% above 2500 psig at startup conditions.
l Additional analyses were performed to determine the pressurizer maximum water level that l would prevent the RCS from exceeding the safety limit of 2750 psig in the event of either a startup accident or a rod withdrawal accident.
Therefore, this change does nnt involve a significant reduction in the margin of safety.
f Therefore, based upon the reasoning presented above and the previous discussion of the amendment request, Entergy Operations has determined that the requested change does nat j involve a significant hazards consideration.
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Figure 1 of 1CAN049606 .'
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SET POINT DEVIATION OF UNIT 1 PRESSURIZER VALVES ,
5.00 W BR-06763 4.00 B BM-04914 4.32%
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PROPOSED TECHNICAL SPECIFICATION CHANGES i
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2.2 SAFETY LIMITS - REACTOR SYSTEM PRESSURE l Applicability l
Applies'to the limit on reactor coolant system pressure. ,
i i objective !
l To maintain the integrity of the reactor coolant system and to prevent i U.e release ,of significant amounts of fission product activity. ;
specification ,
2.2.1 The reactor coolant system pressure shall not exceed 2750 l psig when there are fuel assemblies in the reactor vessel.
2.2.2 The setpoint of the pressurizer code safety valves shall-be !
l in accordance with ASME, Boiler and Pressurizer Vessel Code, l Section III, Article 9, Summer 1968. !
l l Bases ,
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j The ' reactor coolant. system ( 3 ) serves as a barrier to prevent radionuclides in the reactor coolant from reaching the atmosphere. In the event of a ,
fuel cladding failure, the reactor coolant system is a barrier against the ;
release of fission products. Establishing a system pressure limit helps to assure the integrity of the reactor coolant system. The maximum transient ;
! pressure allowable in the reactor coolant system pressure vessel under the j
! ASME code,Section III, is 110 percent of design pressure. (8) The maximum [
transient pressure allowable in the reactor coolant system piping, valves, .j and fittings under ANSI Section B31.7 is 110 percent of design pressure. ,
! Thus, the safety limit of 2750 psig (110 percent of the 2500 psig design -
l pressure) has been established. (8) The settings for the reactor high ,
- pressure trip (2355 psig) and the pressurizer code safety valves (2500 psig i j 11%) ( 5 ) have been established to assure that the reactor coolant system j
pressure safety lindt is not exceeded. When testing the pressurizer code safety valves, the "as found" lift setpoint may be 2500 psig 13%. However, l ,
if found outside of a 11% tolerance band, they shall be reset to 2500 psig i1%. l
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The initial hydrostarf- test is conducted at 3125 psig (125 percent of design l pressure) to verify *;ie integrity of the reactor coolant system. Additional j assurance that the reactor coolant system pressure does not exceed the safety !
limit is provided by setting the pressurizer electromatic relief valve at i 2450 psig. (d) l l
REFERENCES (1) PSAR, Section 4 (2) FSAR, Section 4.3.11.1 (3) FSAR, Section 4.2.4 (4) FSAR, Table 4-1 f
i Amendment No. 44,-1-04 10 REVISED BY "n0 SEM ER 0?.TEO : DECEMBER-10 1991
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The plant is designed to operate with both reactor coolant loops and at least one reactor coolant pump per loop in operation, and maintain DNBR above 1.30 (for the BAW-2 correlation) and 1.18 (for the BWC correlation) during all nornal operations and anticipated transients. (1) l Whenever the reactor coolant average temperature is above 280*F, single l failure considerations require that two loops be operable.
l The decay heat removal system suction piping is designed for 300*F thus, l the system can remove decay heat when the reactor coolant system is below this temperature. (2,3)
One pressurizer code safety valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greater than that required by the sum of the available heat sources which are pump energy, pressurizer heaters, and reactor decay heat.
(4) Both pressurizer code safety valves are required to be in service prior to criticality to conform to the system design relief capabilities.
The code safety valves prevent overpressure for a rod withdrawal accident.
(5) The pressurizer code safety valve lift setpoint shall be 2,500 psig il percent allowance for error and each valve shall be capable of relieving 324,000 lb/h of saturated steam at a pressure not greater than 3 percent l above the set pressure. When testing the pressurizer code safety valves, the "as found" lift setpoint may be 2500 psig i3 percent. However, if l found outside the il percent tolerance band, they shall be reset to 2500 psig il percent.
The internals vent valves are provided to relieve the pressure generated by steaming in the core following a LOCA so that the core remains sufficiently covered. Inspection and manual actuation of the internal vent valves (1) ensure operability, (2) ensure that the valves are not open during normal operation, and (3) demonstrate that the valves begin to open and are fully !
open at the forces equivalent te the differential pressures assumed in the l safety analysis. )
1 The reactor coolant vents are provided to exhaust noncondensible gases and/or steam from the primary system that could inhibit natural circulation core cooling. The operability of at least one reactor coolant system vent path from the reactor vessel head, the reactor coolant system highpoints, and the pressurizer steam space ensures the capability exists to perform this function. The valve redundancy of the vent paths serves to minindze the probability of inadvertent actuation and breach of reactor coolant pressure boundary while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path. Testing requirements are covered in Section 4.0 for the class 2 valves and Table 4.1-2 for the vent paths. These are consistent with ASME Section XI and Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plan 1 Requirements," 11/80.
REFERENCES (1) FSAR, Tables 9-10 and 4-3 through 4-7 (2) FSAR, Section 4.2.5.1 and 9.5.2.3 l (3) FSAR, Section 4.2.E.4 l (4) FSAR, Section 4.3.10.4 and 4.2.4 l (5) FSAR, Section 4.3.7 i
l Amendment No. 41,64,94 17 nEVISED SY "nc LETTEn D?.TED : 12/15/01, 9/15/96
3.1.3 Minimum Conditions for criticality specification 3.1.3.1 The reactor coolant temperature shall be above 525F except for portions of low power physics testing when the requirements of Specification 3.1.8 shall apply.
3.1.3.2 Reactor coolant temperature shall be to the right of the criticality limit I of Figure 3.1.2-2. !
l l 3.1. 3. 3 - 5.5ca the reactor coolant temperature is below the minimum temperature I specified in 3.1.3.1 above, except for portions of low power physics testing l when the requirements of Specification 3.1.8 shall apply, the. reactor shall be suberitical by an amount equal to or greater than the calculated i reactivity insertion due to depressurization.
3.1.3.4 The reactor shall be maintained subcritical by at least 1 percent Ak/k until a steam bubble is formed and a pressurizer water level within the limits of Figure 3.1.3-1 is established.
3.1.3.5 Except for physics tests and as limited by 3.5.2.1, safety rod groups shall
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be fully withdrawn and the regulating rods shall be positioned within their i
position limits as defined by Specification 3.5.2.5 prior to any other reduction in shutdown margin by deboration or regulating rod withdrawal during the approach to criticality.
3.1.3.6 The reactor shall not be made critical until at least 2 of the 3 emergency-powered pressurizer heater groups are operable. With less than 2 of the 3 required heater groups operable, restore the required heater groups to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the required heater groups are not restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, be in hot shutdown within the I
following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
I j 3.1.3.7 With.any of the above limits violated, restore the reactor to within 'he c limit in 15 minutes or be in at least Hot Shutdown within the next 15 minutes.
Bases
, At the beginning of life of the initial fuel cycle, the moderator temperature
! coefficient is expected to be slightly positive at operating temperatures with the
! operating configuration of control rods. (1) Calculations show that above 525F the positive moderator coefficient is acceptable.
l Since the moderator temperature coefficient at lower temperatures will be less
! negative or more positive than at operating temperature, (2) startup and operation of the reactor when reactor coolant temperature is less than 525F is prohibited except where necessary for low power physics tests.
The potential reactivity insertion due to the moderator pressure coefficient (2) that could result from depressurizing the coolant from 2100 psia to saturation pressure of 900 psia is approximately 0.1 percent Ak/k.
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'During physics tests, special operating precautions will be taken. In addition, the .
strong negative Doppler coefficient (1) and the small integrated Ak/k would limit the I
, magnitude of power excursion resulting from a reduction of moderator density. I e
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l 360 340 - Operation Not Permitted In This Region 320 -
N 320 inche's 320 inches 100% Power
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i 300 15% Power l
! 280 - l 1 - 259 inches 15% Power
- 260
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.C 240 -[ 259 inches ;
l 09 Power l
~220 l e .
l e 200 l 4 Acceptable Region of Operation i
u180 e :
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W140 e
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E100 e -
W -
m 80 I 45 inches 45 inches 6 0 --
F 0% Power 100% Power 7 I I k 40 -
- Operation Not? Permitted'-In[This[Redion1
~
20 -
0 , , , ,;,,,,,,,,,;,,,,,, ,,,
0 20 40 60 80 100 Reactor Thermal power (%)
Figure 3.1.3-1 ANO-1 Pressurizer Level Acceptable Region of Operation NOTE: The values specified for pressurizer level and reactor power do not contain an allowance for instrument error.
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Amendment No. 21a l
... . i' 3.1.7 Moderator Temperature Coefficient of Reactivity Specification l
3.1.7.1 The moderator temperature coefficient (MTC) shall be non-positive whenever thermal power is 295% of rated thermal power and shall be less positive than 0.9 x 10-4 Ak/k/'F whenever thermal power is <95% of rated thermal power and the reactor is not shutdown.
l 3.1.7.2 The MTC shall be determined to be within its limits by confirmatory measurecents prior to initial operation above 5% of rated thermal power after each fuel loading. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the limits in 3.1.7.1 above.
I 3.1.7.3 With the MTC outside any one of the above limits, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Bases A non-positive moderator coefficient at power levels above 95% of rated power is specified such that the maximum clad temperatures will not exceed the Final Acceptance Criteria based on LOCA analyses. Below 95% of rated power, the Final Acceptance Criteria will not be exceeded with a positive j moderator temperature coef ficient of +0.9 x 10-4 Ak/k/*F corrected to 95% of rated power. The most limiting event for positive MTC, the Startup Accident, has been analyzed for a bounding moderator temperature coefficient of +0.9 x 10-d Ak/k/*F.
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Amendment No. M,M, M,-1-M 30
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MA.RKUP OF CURRENT ANO-1 TECHNICAL SPECIFICATIONS (FOR INFO ONLY) l I
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i 2.2 SAFETY LIMITS - REACTOR SYSTEM PRESSURE l Applicability Applies to the limit on reactor coolant system pressure. ,
. objective I To maintain the integrity of the reactor coolant system and to prevent the release of significant amounts of fission product activity.
Specification 2.2.1 The reactor coolant system pressure shall not exceed 2750 psig when there are fuel assemblies in the reactor vessel. ,
l 2.2.2 The setpoint of the pressurizer code safety valves shall be (
in accordance.with ASME, Boiler and Pressurizer Vessel Code,Section III, Article 9, Summer 1968.
i Bases ;
The reactor coolant system (1) serves as a barrier to prevent radionuclides >
in the reactor coolant from reaching the atmosphere. In the event of a fuel cladding failure, the reactor coolant system is a barrier against the release of fission products. Establishing a system pressure limit helps to. ,
assure'the it.tegrity of the reactor coolant system. The maximum transient (
pressure allowable in the reactor coolant system pressure vessel under the ASME code,Section III, is 110 percent of design pressure. (8) The maximum ,
transient pressure allowable in the reactor coolant system piping, valves, !
and fittings under ANSI Section B31.7 is 110 percent of design pressure. l Thus, the safety limit of 2750 psig (110 percent of the 2500 psig design .
pressu e) has been established. (8) The settings for the reactor high pressure trip (2355 psig) and the pressurizer code safety valves (2500 psig 11% ) ( ' ) have been established to assure that the reactor coolant system pressure safety limit is not exceeded. When testing the pressurizer code safety valves, the "as found" lift setpoint may be 2500 psig :1, -312%. l !
However, if found outside of a i1% tolerance band, they shall be reset to ;
2500 psig il%. The initial hydrostatic test is conducted at 3125 psig (125 i percent of design pressure) to verify the integrity of the reactor coolant system. Additional assurance that the reactor coolant system pressure does ,
not exceed the safety limit is provided by setting the pressurizer electromatic relief valve at 2450 psig. (4) l REFERENCES ;
(1) FSAR, Section 4 (2) FSAR, Section 4.3.11.1 (3) FSAR,.Section 4.2.4 (4) FSAR, Table 4-1 Amendment No. 49,444 10 REVISED SY "nc LETTED 0".TEO : DECEMEED 15, 1991 i
j
.,a . s' BASES:
The plant is uesigned to operate with both reactor coolant loops and at least one reactor coolant pump per loop in operation, and maintain DNBR above 1.30 (for the BAW-2 correlation) and 1.18 (for the BWC correlation) during all normal operations and anticipated transients. (1)
Whenever the reactor coolant average temperature is above 280*F, single failure considerations require that two loops be operable.
The decay heat removal system suction piping is designed for 300*F thus, the system can remove decay heat when the reactor coolant system is below this temperature. (2,3)
One pressurizer code safety valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greater than that required by the sum of the available heat sources which are pump energy, pressurizer heaters, and reactor decay heat.
(4) Both pressurizer code safety valves are required to be in service prior to criticality to conform to the system design relief capabilities.
The code safety valves prevent overpressure for a rod withdrawal accident.
(5) The pressurizer code safety valve lift setpoint shall be 2,500 psig il percent allowance for error and each valve shall be capable of relieving 300,000 324,000 lb/h of saturated steam at a pressure not greater than 3 percent l above the set pressure. When testing the pressurizer code safety valves, the "as found" lift setpoint may be 2500 psig 11, 3 11_ percent. However, if l found outside the il percent tolerance band, they shall be reset to 2500 psig il percent.
The internals vent valves are provided to relieve the pressure generated by steaming in the core following a LOCA so that the core remains sufficiently covered. Inspection and manual actuation of the internal vent valves (1) ensure operability, (2) ensure that the valves are not open during normal operation, and (3) demonstrate that the valves begin to open and are fully open at the forces equivalent to the differential pressures assumed in the safety analysis.
The reactor coolant vents are provided to exhaust noncondensible gases and/or steam from the primary system that could inhibit natural circulation core cooling. The operability of at least one reactor coolant system vent path from the reactor vessel head, the reactor coolant system highpoints, and the pressurizer steam space ensures uhe capability exists to perform this function. The valve redundancy of the vent paths serves to minindze the probability of inadvertent actuation and breach of reactor coolant pressure boundary while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path. Testing requirements are covered in Section 4.0 for the class 2 valves and Table 4.1-2 for the vent paths. These are consistent with ASME Section XI and Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plan Requirements," 11/80.
l REFERENCES (1) FSAR, Tables 9-10 and 4-3 through 4-7 l (2) FSAR, Section 4.2.5.1 and 9.5.2.3 (3) FSAR, Section 4.2.5.4 l
i (4) FSAR, Section 4.3.10.4 and 4.2.4 (5) FSAR, Section 4.3.7 Amendment No. 31,66,94 17 REVISED BY Mnc LETTER D.":ED : 12,'15,'91, 9,'25,'95 I
.v .
3.1.3 Minimum conditions for criticality Specification 3.1.3.1 The reactor coolant temperature shall be above 525F except for portions of low power physics testing when the requirements of Specification 3.1.8 shall apply.
3.1.3.2 Reactor coolant temperatura shall be to the right of the criticality limit of Figure 3.1.2-2.
3.1.3.3 When the reactor coolant temperature is below the minimum temperature l
specified in 3.1.3.1 above, except for portions of low power physics testing when the requirements of Specification 3.1.8 shall apply, the reactor shall be suberitical by an amount equal to or greater than the calculated reactivity insertion due to depressurization.
3.1.3.4 The reactor shall be maintained suberitical by at least 1 percent Ak/k until a steam bubble is formed and an indicated aster level between 05 and 305 inche is catchliched in the pressurizer water level within the limits of Ficure 3.1.3-1 is established.
3.1.3.5 Except for physics tests and as limited by 3.5.2.1, safety rod groups shall be fully withdrawn and the regulating rods shall be positioned within their position limits as defined by Specification 3.5.2.5 prior to any other reduction in shutdown margin by deboration or regulating rod withdrawal during the approach to criticality.
3.1.3.6 The reactor shall not be made critical until at least 2 of the 3 emergency-powered pressurizer heater groups are operable. With less than 2 of the 3 required heater groups operable, restore the required heater groups to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the required heater groups are not restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, be in hot shutdown within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
3.1.3.7 With any of the above limits violated, restore the reactor to within the l limit in 15 ndnutes or be in at least Hot Shutdown within the next 15 ndnutes .
Bases l
At the beginning of life of the initial fuel cycle, the moderato'r temperature coefficient is expected to be slightly positive at operating temperatures with the l operating configuration of control rods. (1) calculations show that above 525F the positive moderator coefficient is acceptable.
Since the moderator temperature coefficient at lower temperatures will be less negative or more positive than at operating temperature, (2) startup and operation of the reactor when reactor coolant temperature is less than 525F is prohibited except j where necessary for low power physics tests.
The potential reactivity insertion due to the moderator pressure coefficient (2) that could result from depressurizing the coolant from 2100 psia to saturation pressure of 900 psia is approximately 0.1 percent Ak/k.
During physics tests, special operating precautions will be taken. In addition, the strong negative Doppler coefficient (1) and the small integrated Ak/k would limit the magnitude of power excursion resulting from a reduction of moderator density.
Amendment No. G,34,&O,67 21
)
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360 340 - Operation 1Not'. Permitted'/-In' This ; Region ~
k _ 320 inches 320 1'nches /
300 - 15% Power 100% Power 280 - f
- 259 inches E 260 / 15% Power 1
A \
U
.$240 lu259 inches ,
0% Power I
s220 l as .
as 200 4 Acceptable Region of Operation
[
u180 a) :
4J . 1 to 160 1 3: -
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~
M140 a) - ,
u -
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~d120 l j
- 3 - ,
E100 l l e . l 4 -
a, 80 ,
45 inches 45 inches j 60 'F 0% Power 100% Power 7 i I l k l 40 -
! 20 - Operation!Not PermittedlIn.This4 Region:
0 , , , , , , , , , , , , , , , , , , , ; , , , ,
, 0 20 40 60 80 100 Reactor Thermal power (%)
Ficure 3.1.3-1 A_N0-1 Pressurizer Level Acceptable Re.gion of Operation NOTE: The values specified for pressurizer level and reactor power do not contain an allowance for instrument error. )
l Amendment No. p l
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l' 3.1.7 Moderator Temperature Coefficient of Reactivity Specification 3.1.7.1 The moderator temperature coefficient (MTC) shall be !
non-positive whenever thermal power is 195% of rated thermal power and shall be less positive than 0.9 x 10-i Ak/k/*F whenever thermal power is <95% of rated thermal power and the reactor is not shutdown. l 3.1.7.2 The MTC shall be determined to be within its limits by ,
confirmatory measurements prior to initial operation above 5% of rated thermal power after each fuel loading. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the limits in 3.1.7.1 above.
3.1.7.3 With the MTC outside any one of the above lindts, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. 4 l
Bases l .
A non-positive moderator' coefficient at power levels above 95% of rated power is specified such that the maximum clad temperatures will not exceed the Final Acceptance Criteria based on LOCA analyses. Below 95% of rated ,
power, the Final Acceptance Criteria will not be exceeded with a positive moderator temperature coefficient of +0.9 x 10-8 Ak/k/*F corrected to 95% of rated power. The most limiting event for positive MTC, the Startup Accident, has been analyzed for a boundinar ng: Of moderator temperature coefficient includingof +0.9 x 10-4 Ak/k/*F.
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! Amendment No. M, M,W,H4 30
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t PROPOSEI) ANO-1 SAFETY ANALYSIS REPORT CHANGES l (For Use in Review of Proi > sed TS Changes) i l
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ARKANSAS NUCLEAR ONE j Unit 1 I
! 3A.7.2.1 Startus Event ^
l
- The hng-of-cycle mod A _ \ tnt)
- t. .
temp === woucient at hot zero power (HZP) for Cycle 13 is d
j given u +0. 10 Table 3A-8. This parameter is used in the startup event analysis. A
- sensitivity study rmed in the SAR that varied the moderator temperature coefBeient up to d
! +0.90 x 10 t the remaining startup event analyses in the SAR considered an MTC of
! zero. An evalua ' done to verify that the results of the analyses in the SAR do validate d
l the use of a crator t ture coefBeient of +0.90 x 10 Ak/k/F at hot zero power. The l analysis, refore, bounds the cle 13 parameters.
l 3A.7.2.2 Steam Line Failure l
The steam line break (SLB) accident was evaluated based on the reactivity feedback, termed the reactivity deficit, at conditions below HZP (532F and 2200 psia). The reactivity deficit for the steam line break analysis is 0.93702 %Ak/k. This value includes the effects of both fuel and moderator temperature changes. The reactivity de6 cit predicted for Cycle 13 using the same SLB system conditions is 1.07 %Ak/k (Table 3A-8). The Cycle 13 value, calculated by NEMO, is larger than the SLB analysis value, indicating a greater reactivity feedback for the Cycle 13 core. The cross section library used by NEMO to calculate the Cycle 13 reactivity deficit has not been benchmarked to the final SLB temperature and pressure of the moderator and temperature of the fuel. For conservatism, an uncertainty of 0.2 %Ak/k has been applied to the above Cycle 13 NEMO reactivity deficit calculation to bound the cross section data uncertainties. The rod insertion limits have been verified to accommodate the difference between the NEMO reactivity de6 cit for Cycle 13 and the TRAP 2 reactivity deficit used for the MSLB analysis 3A.7.2.3 Non-LOCA Safety Analysis Conclusions The key cycle-specific parameters for each of the events in chapter 14 of the ANO-1 SAR were reviewed. It has been concluded that the non-LOCA safety analyses remain bounding for Cycle 13 operation.
3A.7.3 LOCA EVALUATION The emergency core cooling system (ECCS) evaluation model (EM) reported in BAW-10103A, Rev. 3 (reference 12) has been approved for the analysis oflarge break loss-of-coolant accidents (LOCA) for the B&W-designed plants. The EM has been upgraded with the B&W-modified version of FLECSET (reference 13). The application of the EM to the B&W-designed,177-fuel assembly, lowered-loop nuclear steam supply (NSS) system is reported in BAW-10104PA, Rev. 5 (ref-rence 14). The fuel performance data input to the EM have been provided by TACO 2 and current TACO 3 computer codes (references 15 and 4).
The analyses are performed generically, using the limiting values of key parameters for all of the operating B&W-designed 177-fuel assembly lowered-loop plants. The LOCA linear heat rate (LHR) limits include the combined effects of the NUREG-0630 cladding swell and rupture model, the BWC CHF correlation, reduced fuel rod pre-pressure, and the B&W-modified version of FLECSET.
Amendment No.13 3A.7-2
. .. ..- ' ~
ARKANSAS NUCLEAR ONE Unit 1 TABLE 3A-8 COMPARISON OF KEY PARAMETERS FOR ACCIDENT ANALYSIS Parameter Safety Analysis Cycle 13 y_alue Value BOC (a) Doppler coefficient, -1.17(t) -1.61 10-5, Alotf'F (b pler coefficient, -1.30 -1.80 BOC moderator coefficient (HFP), 0.0 -0.22 10-4, Ak/k/'F EOC moderator coefficient (HFP), -4.0 -3.23 10-4, Ak/k/'F BOC moderator coefficient (HZP), +0.9 +0.36 10-4, Ak/k/*F SLB reactivity deficit, 0.93702(c,d) 1.07(d)
%Ak/k All rod bank worth 12.90 7.56 (HZP), %Ak/k Maximum single group worth Nominal 2.59 (HZP), %Ak/k 3.0 Inverse boron worth 140 152 (HFP), ppm /%Ak/k Maximum ejected rod worth 0.65 50.65 (HFP), %Ak/k Maximum dropped rod worth 0.65 50.20 (HFP), %Ak/k Initial boron concentration 2270 2042
~
(HFP), ppm (a) BOC denotes beginning ofcycle.
(b) EOC denotes end of cycle.
(c) Used in the steam line break analysis.
(d) Calculated over a moderator temperature range of 532F to 477.51F, a fuel temperature range of $32F to 650.7F, and a core pressure range of 2200 psia to 735.87 psia.
@ %nhr w L'..A uJ %r swu etus as -t.3x / Lt/a/ F.
Amendment 13 3 A.ll-ll
ARKANSAS NUCLEAR ONE Unit 1 TABLE 3A-9 ANALYSIS STATUS OF NON-LOCA SAFETY ANALYSIS Cycle-Speci6c Effective Cycle For Parameters gygg Analysis OfRecord
_ Bounded?
Startup Event -+-- it -GeetierrWett-4es Rod Withdrawal at Power Event 1 Tes Moderator Dilution Event At Power 12 Yes During Refueling 12 Yes Cold Water Event 1 Yes Loss of Coolant Flow System Response
- Locked Rotor Event 1 Yes Four-Pump Coastdown Event 1 Yes Four-to-Two Pump Coastdown Event 1 Yes Dropped Rod Event 1 Yes Loss ofElectric Power Events Loss ofLoad Event 1 Yes Complete Loss of AC Power Event 1 Yes Turbine Overspeed Event 1 Yes FuelHandling Accident 1 Section 3A.7.1 Steam Line Failure Event 12 Section 3A.7.2 Steam Generator Tube Failure Event" l Yes Rod Ejection Event 1 Yes Loss-of-Coolant Event Section 3 A.7.3 Section 3 A.7.3 MM-= HypotScal Accident Section 3 A.7.1 Section 3A.7.1 Waste Gas Decay Tank Rupture Event Section 3A.7.1 Section 3A.7.1 (a) The plant system response (including swer, RCS flow, core inlet temperature, and em pressure) has been shown to be bounc ing for cycle 13. The DNB analysis is discu separately in section 3A.6.
(b) For dose consequences of the steam generator tube rupture event, refer to section 3 A.7.1.
Amendment 13 3 A.ll-12
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ARKANSAS'NUCMAR ONE Unit 1
) D. A short-period withdrawal stop and alarm are provided in the intermediate range.
.i l E. A high flux level and a high pressure trip are provided in the power range.
! 14.1.2.2.2 Reactor Protection Criteria e
i The criteria for reactor protection for this accident are:
l A. Reactor thermal power shall not exceed 112 percent of rated power.
B. RCS pressure shall not exceed code pressure limits.
I 14.1.2.2.3 Methods of Analysis \ot..C17. Y j g w eco d W t b ^*'ng
A B&W digital computer model of the reactor and S was used 'o determine the
- characteristics of this accident. This model used Sow but no beat transfer out j o em and no sprays in the pressurizer. Dgppler_coef5eient was used
- twc.L.4dl the Doppler coefEcient is much larger (more negative) than thi u-r PW @fer ab, i p.A eptAw The rods were assumed to be moving out along the steepest part of the rod worth i versus rod travel curve. The values of the principal parameters used in this analysis are listed in i
Table 14-3.
i l In addition, the criterion for minimum movable control rod worth is that a shutdown margin of r
- l one percent Ak/k at the hot standby condition is required (Section 3.1.2.2). The startup accident l has been analyzed using the minimum tripped rod worth with the mari== worth stuck rod as j l part of the analysis. The stanup accident was analyzed from 0.5% Ak/k subcritical at the hot, f pressurized condition.
! M McA rde. M reO S I N i
j 14.1.2.2.4 Results of Analysis {ne "*'"A g3.D& peaje pec h re ydeM a r.rw g
' ~
i Figure 14-1 shows the results of rth co f l "m rpo m . cal. tid)
- vi im fa 1 -_N a_ nho Doppler effect terminates the neutron power (neutron i
power is defined u the total energy release from Ession) rise, but the hest input to the reactor coolant increases the pressure past the trip point and the transient is terminated by the high i pressure trip. l k sb tNeEpoM i
Figure 14-2 shows [resuhs of withdrawing all Control Rod Assemblies (CRAs) at the i l maximum speed froniWB& Awes This results in a mari=m possible reactivity
! addition rate. The total rod worth used in this analysis is slightly greater than the calculated worth I l (Table 3-5). The power rise is terminated by the negative Doppler effect. The high neutron Sux trip takes effect after the wer is reached and terminates the transient. The peak thermal j heat flux is signi6cantly less the rated power heat Sux.
At w.4. eon
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Amendment No.13 14.1-3 4 - . , , - _ -
n[:}
d 9:k %
.u w,,.as rs dk Warm **% *M"Pt>r.v. M A sensitivity analysis was determine the esset of varying several key parameters.1 Variation of thej np 9#to 0.7 second resultedin a change in peak thermal power ofless thang N g eQS( g 4 ,,,, g .
Figursp(14-3 Mahowythe effect of varying the b rate on the peak thermal power < .z, - n n.. o g. This reactivity rate was varied from more than an order of magnitude below the single rod group rate 6to a rate slightly above that for simultaneous withdrawal of d rods. The slower rates will result in the pressure tdp being A
- actuated. Only the very fast rates actuate the high neutron flux level trip.
7$ Fi A q u k pressste 14-6 showythe peak thermal powerhon as a function ofh4) moderator coefficients forggp996dMGM a. ro.%e- of reacM st4b *AAh re+cs
- re.swW b worsi we peak peasawee a d G er M r wer.
MjMrbedMV4tWMg5MeAhe643rgspMgng $ptsdAgrme af#-i" ARsGWd6 Table 14-4 summanzes the results of the postulated startup accidents.
It is concluded that the reactor is completely protected against any startup accident involving the withdrawal of any or all control rods, since in no case does the thermal power approach the design overpower condition and the peak pressure never exceeds code allowable limits.
14.1.2.3 Rod Withdrawal Accident at Rated Power Operation 14.1.2.3.1 Identification of Cause A rod withdrawal accident pre-supposes an operator error or equipment failure resulting in accidental withdrawal of a control rod group while the reactor is at rated power. As a result, the power level increases, the reactor coolant and fuel rod temperatures increase, and, if the withdrawal is not terminated by the operator or the protection system, core damage would eventually occur.
The following provisions are made in the design for the indication and termination of this accident.
A. High reactor coolant outlet temperature alarms.
B. High RCS pressure alarms.
C. High pressurizerlevel alarms D. High reactor coolant outlet temperature trip.
E. High RCS pressure trip.
F. High power level, i.e., neutron flux level, trip.
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Amendment No.12 14.1-4
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.; gag a gg; e 4b,, [
Insert A '^-
De high pressure trip seapoint was varied for the peak pressme'and thermal power c5se resulting 6cm a reactivity addition rate of 1,73 E 4 (AK/K)/sec. An increase of the high pressure trip seapotat by 5 psi resulted in the peak pressure increasing by less than two psi and the peak thermal power increasing by less than one percent.
Variation of the assumed effective delayed neutron fraction (pef!) changes the reactisity addition rate which results in the peak pressure and thermal power. A decrease in the eff from 0.007 to 0.006 resulted in a reduction of the peak pressure by two psi and an increase in the peak thermal power by less than one percent based on reactivity addition rates that result in peak pressure and thermal power.
Variation of the assumed axial peaking factor changes the reactivity addition rate wluch results in the peak pressure and thermal power. Analysis of the results with axial peaking factors of 1.0,1.7, and 2.0 showed the axial peakmg factor of 1.5 used for the analyses dised for this event results in the peak RCS pressure. Although different axial peaking factors result in differet peak thermal powers the margin available for thermal power is less limiting than the margin avadable for pk pressure The effect of varying the initial power level has shown that lower initial powe in conjunction with high reactivity addition rates can result in higher peak thermal powers. nese same studies show there is still margin to the rated thermal power even if all rods are simultaneously withdrawn at the maximum rate of withdrawal from an initial power of 1 E-9 watts. The power rise is terminated by the negative Doppler effect. The high neutron flux trip terminates the event. The pressure increases slowly until the PSV lifts.
The resultant peak pressure in the RCS will be dictated by the PSV liA pressure plus any pressure differential between the PSV and the peak RCS pressure location.
The effect of varying the number of RCPs operating at the onset of the event show that the reactisity addition rate that results in the peak pressure and thermal power will change due to the different initial conditions. The resultant change in peak pressure ofinitiating the event with 3 RCPs versus 4 RCPs operating is an increase in the peak pressure by about 6 psi, while the peak thermal power remains approximately the same or slightly lower than results initiated from 4 RCP initial conditions.
Figure 14-4 shows the effect of varying the pressurizer safety valve (PSV) liA setpoint tolerance (accumulation) from 3% to 5% (assuming all other inputs remain constant).
Figure 14-5 shows the effect of varying the pressurizer safety valve flow rate from a single PSV flow rate of 300,000 lbm/hr to 2 PSVs with a flow rate of 324,000 lbm/hr/ valve.
Insert B The peak RCS pressure was found to be dependent on the initial pressurizer level. Higher initial pressunzer levels result in less volume to acummodate the expansion of the RCS volume due to the heat input caused during the startup event. Figure 14-7 shows the resultant peak pressure corr-ama- to the reactivity addition rate that results in peak pressure Figure 14-8 shows the effect of varying the reactisity addition rate on peak pressure I
l I
l j ARKANSAS NUC12ARONE Unit 1 Table 14-3 i
STARTUF ACCIDENT FARAMETERS ;
- T N* 9 E .t Y Ji z_ . ~ - - - ,
i Maximum Rod Speed,inimin 30 4
Maximum Number ofCRAs 4 O t 1
Mtvimum Rod Worth, All Rods, % Ak/k 12.9 M=imum Reactivity Addition Rate,
! Allpf Rods at Max Speed,(Ak/k)/s 9.27 x 10'4 i 60 i Maximum Rod Worth of Single Group j When Reactor is Critical, % Ak/k 3.0 i
Maximum Reactivity Addition Rate for l
j Single Rod Group,(Ak/k)/s -
2.15 x 10 4 i '
Doppler CoefficientM ~'
,y l 4.!? -I 3 >r t o
- (Ak/k)/*F
- Moderator Coefficient 95N5GiRG 3 Eertr 4 c.9 x tdi i
4 (M./t4f*F Peak Thermal Power Permitted (Design Overpower), % rated power 112 l
i I Trip Parameters ag( ht>$we T/.e Sei .'.id t , psia. 7.40o Delay for High Pressure Trip, s -6:6- c.6 l l
); Delay for High Flux Trip, s 4+ o.3
! ControlRod Traveltime to 2/3 Ire,ertion, s 1.4 l
DelageA 0e Mr.~ N.dt.s(p,qqh o. col i O kdee c0 T6V s .2.
TSV 144 T terme (Ac.c.mdd;.h t1 % h6 p5h TW Nw hk h /k, /vdve 3 324,000 2.m Qx rA N power)l.
Lh\ Towe.r 3 web L W.d Tren wcwer Leve.\ , kd.n \80 4de.c <R 'it,c.V.s 's e p ,.d :.s . 4 Co re. I'l ww A d'i d 7 e.a. U q N.edor I.f Ameedment No.13 14.5-4
..- ~ - .
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ft i
Unit i Table 144 1
{
SUMMARY
OF STARTUF ACCIDENT ANALYSIS i
i 1. Peak thermal power for withdrawal rates less than that corresponding to
- the withdrawal of all rods is always less than rated power.
- 2. Average fuel temperature in the average fuel rod never exceeds l
-t;699'F.
gh e 515 te c. kNea c. k lnurf C-l :
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l Amendment No.12 14.5-5 l
, _ . - - . . = _ _
pym. y
\ .. , , < -[ $
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! Insert C l
The pank RCS pressure was assured to be less than 2750 psig using a pressurizer level of 180 inches (minus any applicable uncertainty) with two pressunzer safety valves (PSVs) relieving at a 2590 psia
( setpoint and a flow rate of 324,000 lbm/hr/ valve. The peak RCS pressure was also assured to be less than
- 2750 psig with only a single PSV relieving at a 2640 psia setpoint and a flow rate of 300,000 lbm/hr at a
- pressunzer level of 180 inches (minus any applicable uncertainty).
l I
1 l
i I
~
ins tas 8 .....
e...<. s 3, /
~
T l
l e , , , I N
l 63
\W I**' sg x
500 Ingrk C 4as ^
,ki l: l" u, m [ \
g , , , ,
N is N
T w.A h /
cia. .. t ,
4 ist -
2500 Iritsa 4g r,.ii.... .. .
N STAR M P ACCIDENT FROM 10~9 RATED POWER USE ARKANSAS P0ffER & LIGHT CO. A
- HIGH PRESSURE FIG. NO.
A REACIOR TRIP IS AC WATED ARKANSAS NUCLEAR ONE-UNIT 1 1 14 _ 1 bdLikg b N'd'sen N C. f 0.03x16 kj '
S t.L
neutron power
~
Neutron Power Versus Time for Startup Accident From 1E-09 Rated Power Using A Reactivity Addition Rate of 1.73 E-04 (DK/K)/sec; High Pressure Reactor Trip is Actuated 30E+08
)
25E+08 __
g 20E+08 i
0 15E+08 5
a:
$ 10E+08
=+.
A 05E+08 1
00E+00 >
0 8 16 24 32 40 48 56 64 Time (sec) s e
Page 1
. t THERMAL POWER Thermal Power Versus Time for Startup Accident From 1E-09 Rated Power Using A Reactivity Addition Rate of 1.73 E-04 (DK/K)/sec; High Pressure Reactor Trip is Actuated 1.0 0.8 i
N .
0.6 2
4 5
0.4 e f 3
sg 0.2 / \ -
0 -
E ' %
S 0.0 5
M [
. ,n
-0.2 0 8 16 24 32 40 48 56 64 Time (sec)
Page 1
AVG FUEL TEMP -
Fuel Temperature Change Versus Time for Startup Accident From 1E-09 Rated Power Using A Reactivity Addition Rate of 1.73 E-04 (DK/K)/sec; High Pressure Reactor Trip is Actuated 500 400 -
u.
O A Y e
?
!n-d E
~ 100 t x h ,
O i 0 8 16 24 32 40 48 56 64 f Time (sec) A Page 1
MODERATOR TEMP chinge '
Average Core Moderator Temperature Change Versus Time fcr Startup Accident From 1E-09
~
Rated Power Using A Reactivity Addition Rate of 1.73 E-04 (DK/K)/sec; High Pressure Reactor Trip is Actuated 30.0 25.0 O
- W 20.0 IT 1
b 15.0 h
O 10.0 l O '
2 m --;.
$ S.0 ;IL o M!
2 %
$ k w 0.0 '
N g tr
-5.0 0 8 16 24 32 40 48 56 64 s Time (sec)
Page 1 i
PRESSURE Reactor System Pressure Versus Time for Startup Accident From 1E-09 Rated Power Using A Reactivity Addition Rate of 1.73 E-04 (DK/K)/sec; High Pressure Reactor Trip is Actuated 2750 2700 A 2650
'~ \ _
g-
\ /
g 2500 f
\/
$ 2450 E
- 2400 s
" 2350 j c 2250 D
R 2200 m 2150 0 8 16 24 32 40 48 56 64 },
,v '
Time (sec) '3 Page 1
' " ~ '
" 'W '
~
)
1 l
1 1
5 00 -
A isw*F;,:"", '" -
l\
700 -
100 0
l 40 kA k ine,ael ~
l 1 M r.
lll _
7 -
cr 'l*
Q ,: , 2 see I
l 1
Lwd 7- ,,,,,,, c. i
. .,.i., -
l Itett,atu,4 4 -
- e. . . . , , _
l , i i i I 20H
~
$ Reeste, _
sritse P,tSSW,0. DO 3
i t
i _-
STARTUP ACCIDEfT FROM 10-9 RATED POWFL ARKANSAS POWER & LIGHT CO, USI1OG ALL RODS WIM ADIDEWidWX/sd : FIG. NO.
ARKANSAS NUClE.AR ONE-UNIT 1 HIGH FIUI REACTOR TRIP IS ACTUA'Tra 11+-2 l
i b M.dtc'i k N ' b b e k \X h [g/ K/gg,c, I
neutron power -
Neutron Power Versus Time For A Startup Accident From 1E-09 Rated Power Using A Reactivity Addition Rate of 1 E-03 (DK/K)/sec; High Flux Reactor Trip is Actuated 2.4E+10 2.0E+10 g 1.6E+10 h
i m
0 1.2E+10 8
15 g ;
z 8.0E+09 3
4.0E+09 T
0.0E+00 O 4 8 12 16 20 24 28 32 Time (sec)
Page 1
THERMAL POWER -
Thermal Power Versus Time For A Startup Accident From 1E-09 Rated Power Using A Reactivity Addition Rate of 1 E-03 (DK/K)/sec; High Flux Reactor Trip is Actuated 1.0 I 5-
.u, 0.8 E
0.6 4
1 0.4 S
g 0.2 k
0.0 1 +
, e
-0.2 0 4 8 12 16 20 24 28 32 i Time (sec) r Page 1
Avg Fuel Temp Fuel Temperature Change Versus Time For A Startup Accident From 1E-09 Rated Power Using A Reactivity Addition Rate of 1 E-03 (DK/K)/sec; High Flux Reactor Trip is Actuated 400 350 u.
250 5 i 5
g 2m E
1 w 150 t
100 e
50 -
s 0 ~
50 0 4 8 12 16 20 24 28 32
[
Time (sec) i.
Page 1
MODERATOR TEMP change Average Core Moderator Temperature Change Versus Time For A Startup Accident From 1E-09 Rated Power Using A Reactivity Addition Rate of 1 E-03 (DK/K)/sec; High Flux Reactor Trip is Actuated .
25.0 Y
k 20.0 -
5 E
E .
$ 15.0 g E
i'u v
u.10.0 8
2 5.0 E t 8 '
8 1 I 0.0
? W s%
4
-5.0 (;
O 4 8 12 16 20 24 28 32 $
Time (sec)
Page1
PRESSURE -
~.
Reactor System Pressure Versus Time For A Startup Accident From 1E-09 Rated Power Using A eacHvity Addition Rate of 1 E-03 (DK/Kysec; High Flux Reactor Trip is Actuated 2750 2700 2650
/
/ \
\
28 "
2550 g2,00
,/ .
=
g
/
a /
" 2350 2300 2250
/ ?
3 2200 2150 x 4 g 12 20 24 28 32
% Time (sec) c 9
4 Page 1
"#-## -- -- r ---_ _,_
- > p i
i
! l 10
. . . . ,,, , I . I 1 g g y y Pressure pg.,,
Trip g
I8
/
,, \ ,
linile
\ / Central
/'-
- 3 ., \ -
l 2 I 5 30 all see l
t
<0
.. . \
10 6 ,g.$ 10 10 3 Red setMrasal tale. (at/ti/sec M. E kkC, - -
O IA h p had K PFAK DIERMAL POWER VS ROD WITHDRANAL RATE FOR A STARTUP ACCIDENT ARKANSAS P0hER E LIGHT Co.
ARKANSAS NUCLEAR ONE-UNIT 1 MM d m m H{- 1 -3
power vs RIR -
Peak Thermal Power VS Reactivity Addition Rate For A Startup Accident From 1 E-09 Rated Power; 3% Accumulation on PSV 80 70
/
,o / \ /
/ /
/ /
- 50
- 40 >
! /
- 30 /
t 20 10 Y 0 1.00E-05 1.00E-04 1.00E-03 4 Reactivity insertion Rate, (dK/K)/sec N
x Page 1
t 1000 . . . . . . . . ..... . . . . . . . ..
7 I
l r .
/
l l
j All Reds g l g 100 ,
E -
. \
c 5 -
f -
E ~
~
g Slagle Control
- Red Group -
E t i e e 'Eii e iiiii. - i i.ie eii 10 10 5 to 10*3 10'8 Red Withdrassi Rats (ak/k)/$ec h.A L @ n of ule
%- E - oo 6% - o t)
( p g pressue Vs PW Au w*daMon
\ ""' " - T ' Ivw m v .a avu = A ~ ns ARKANSAS POWER 8 LIGHT Co. WWWR FOR A gTARTUP ACCIDM FIG. NO.
FROM 10- RATED PolGR 14-4 ARKANSAS NUCLEAR ONE-UNIT 1
press vs lift tol :
Peak Pressure VS PSV Accumulation For A Startup Accident From 1 E-09 Rated Power Two PSVs 2750 4 2745 2740 /
" 2735
- l. .[
- e
."w2730 a
2725 2720
% 2715 2.5 3 3.5 4 4.5 5 5.5
% Lift Tolerance (Accumulation)
R Page 1
s To
~ Nonenst,
, So
- t High r High N g
- - Flus f Pressure v Trip Trip 5
' N ~
U /
/
' 4e /
3D /
-o.: o.s .i.o .i i , .i , ,,,3 , , ,, , ,, ,
ooppler coef ficierit (ak/k)/F : 1o5 1A14rk k\ Ch c.cd,c_ H-E-N4A-ol) m
% ,,s r.. Vs TSV Rowcde For A Skebe w a.a un x w ma:,a s u m w oc s%
AK/ K /st<- F r.w 16 L .4 ,. A. h e c t-1.B x i6' }
N PFAK TfD!RMAL PFLER CODTICIDfT ARKANSAS POWER & LIGHT CO, FOR A STARTUP AC 3 0% ok/k FIG. NO.
ARKANSAS NUCLFAR ONE-UNIT 1 ROD GROUP AT 2. )fsFROM10-9 14-5
press vs psv flow ;
Peak Pressure VS PSV Flowrate For A Startup Accident Using A Reactivity Addition Rate of 1.73 E-04 (DK/K)/sec From 1 E-09 Rated Power; 5% Accumulation on PSV(s) 2765 2760
.5 g 2755
= ,
5 I 2750 ~
E 2745
$ 2740
$% 200000 300000 400000 500000 600000 700000
% PSV Flowrate, gpm Page 1
I i
I j
1
,4 N
l ,, N /
(
/
i.
I 66 s !
62 I /
a.
I
- 1 0 '
= 58 \
$ [
Nominal 54 - I
/ /
50
/
0.5 0.5 0.4 0. 2 0.0 . 0. 2 0.4 0. + 0. 8 .l.0 Nederator Coef ficient, (ak/k)/F : 1 nuer+$%.scod.c.
g ,u . - .. .
%- E - oo 69 -o D
.. - - -9 . _ , ,. A s __. we TrM W. OJ t3 %MT t av s g 36,a -.- f .
I E MWsi Ca.ns Aca4.bvikddle'i/ sit Nafe
?canwa uA 1 I
PFAEgTHERNAL POWER VS )CDERATOR CODTICIENT ARKANSAS POWER & LIGHT CO. M A STARTUP ATIDENT US TIG' NO' N
ARKANSAS NUCLEAR ONE-UNIT 1 vmm 10-Y RATED POWER 14-6 ,
i
l press & power vs mic ;
Peak Pressure And Thermal Power VS Moderator Coefficient For A Startup Accident Using The Worst Case Reactivity Addition Rate From 1 E-09 Rated Power; 5% Accumulation on One PSV 80 2760 75 2755 70 2750 g- 65 2745 j
.$ . e I 60 2740 E I 55 2735f E
l N 50 2730 i i 45 V 2725 t
40 2720
-2.00E-04 -1.50E-04 -1.00E-04 -5.00E-05 0.00E+00 5.00E-05 1.00E-04 Moderator Temperature Coefficient (DK/K/F) -*--Peak Thermal Power, %
-e-Peak Pressure, Psia Page 1
. . , ~ . .
60
/
50 A \
x
. N l W ncainal 2 30 Ng l
w 30 / A N
w ,
\
-o.8 9 -1.o -1.1 -1.2 -1 3 \ -1.4 -1 5 -1.6 DOPPLER CODTICIENT (Ok/k)/F x L r+ kt7 h Q c.dc. S4-E-co64-01).
m '
jd itt s wswee. 4s t44t\ 'PreaswrM r La<el Fer b b b y A cc.!A. J R $t ~ g A b cA b t h A k 1 ; O e n % .k e.
4 (of t.n x i6 u./ g/sec g row gg S LA.1 rower h i PFAK THERMAL DOPPLER COEFFICIDIT ARKANSAS POWER & LIGHT CO. FOR A S y~ USING FROM10,pRATEDPOWER RODS AT FI . NO.
ARKANSAS NUCLEAR ONE-UNIT 1 9 27 x 10 1 -7
press vs pzrIvl =
i.
i Peak RCS Pressure VS Initial Pressurizer Level For A Startup Accident Using A Reactivity Addition Rate of 1.73 E-04 (DK/K)/sec From 1 E-09 Rated Power; 3% Accumulation - 2 PSVs 2770 2760 2750 i
E
- n. 2740
=
{ 2730
% 2720 E
2710 2700 -
I 2690 100 150 200 250 300 350 I initial Actual Pressurizer Level, inches Page1
9
.. \
% f
/ i l
\ / .
E
/
E 5
5-Jid j .0
\
Nomi i I
36
\
0.8 0. 0.4 0.2 0.0 +0.2 +0.4 +0.6 +0.8 +1.0 Nederator Coef ficient, (ak/k)/F : 1 N I L. \ ugnd k 1.% d f.edc. % E-co69 _
l v e. m -
T M.r l
l
[y.ea.k. .s+ % AaA.a Treasse. A mso- V.s % KeM d %W .y.rNO MODERATOR COEFFICIENT hPRK TIERMAL ENT USING ALL RODS AT ARKANSAS POWER & LIGHT CO. FOR A S y *
- ARKANSAS NUCLEAR ONE-UNIT 1 9.27 x 10 TROM 10-9 RATED POWER g,g l
1 <
pressure vs RIR 1
r a
Peak Pressure VS Reactivity Addition Rate For A Startup Accident From i E-09 Rated Power; 3% Accumulation on PSVs 2720 2710 2700
/
m /
l j/
.u 2690 /
S E
/
.x 2680 E
2670
- 2660
?
h 1 2850 1.00E-05 1.00E-04 Q
~
Reactivity insertion Rate, (dK/K)/sec Page 1
_ _ _ _ _ _ _ _ - _ - _ _ - _