ML20113A786
| ML20113A786 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 06/10/1996 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20113A781 | List: |
| References | |
| NUDOCS 9606250332 | |
| Download: ML20113A786 (126) | |
Text
Summary of Proposed Changes Attachment B Page 1 of 2 B. Summary of Proposed Changes The list of proposed changes summarizes the proposed changes by topic and refers to the following Index of Proposed Technical Specification Changes which are ordered by Technical Specification page i
number.
Topic (s)
Change Description / Summary Index #
Addition of SPC 1,9,4,11 The SPC related definitions of FUEL DESIGN LIMITING RATIO FOR Thermal Limits CENTERLINE MELT (FDLRC), FUEL DESIGN LIMITING RATIO (FDLRX),
FDLRX,FDLRC, and TRANSIENT LINEAR HEAT GENERATJON RATE (TLHGR) are added to 1
TLHGR Table of Contents and Definitions. These reflect the SPC licensed methods for fuel related operational limits (FDLRX, FDLRC), and APRM setpoint adjustment for
' peaking factor' changes (TLHGR).
Definition Changes 3,10,2,8 Deletion of definition of" ROD DENSITY" for use in Recctnity Anomaly surveillance (below), and minor wording change to clarify that "FLPD" and
)
"MFLPD" are applicable to GE fuel oniv.
Reactivity 16,17 Include wording to allow use of SPC licensed methodology ("Kef!", eigenvalue Anomaly deviation) versus less detailed " Rod Density" method.
{
Surveillance APRM Setpoint /
5,7,22, Change the title of Specification 3.11.B from "APRM Setpoints" to " Transient
' Peaking Factor' 24 Linear Heat Generation Rate" (TLGHR), and re-write the Specification to reflect licensed SPC methodology. This is consistent with the current Dresden Specification. It is worded such that minimal change will be needed once core i
reaches 100% SPC fuel.
)
Rx Coolant System 6,27 Relocation of Rx Coolant System data to UFSAR. This change is not required for the Data SPC implementation. It is included for improvement purposes.
1 Methodology 18,19, 25, Revise Bases descriptions of MCPR assurance methodologies, including the CRD
)
Descriptions, 28 scram time data inputs to the MCPR analyses. These changes ensure that the MCPR / Scram descriptions of the methodology or analysis inputs an correct for both the GE and Timing SPC licensing methods.
Methodology Desc, 23 The bases description of the APLHGR limits is revised to reflect vendor specific APLHGR licensing aspects of GE and SPC fuel.
Additional 15,20,21, (Index # 15) expands the Bases notation of the definition of the Top of Active Fuel Descriptions or 26 currently given in the Bases of Specification 2.1.D.
Corrections (Index # 20) corrects literal references to the Rod Drop Analysis licensing analyses for GE and SPC, and improve the description of the implementation of rod sequences using BPWS. (Index # 21) changes the description of Rx vessel pressurization analyses to reflect the fact that vessel overpressurization criteria is applied to more transients than just the expected limiting one.
(Index # 26) adds a descriptive clarification to reflect that the SPC fuel bundle design has an interior " water box" in the place of the GE bundle " water rod" Editorial Changes 12,13,14 These items make minor editorial corrections or wording changes.
Admin Controls 29 Addition of the list of NRC approved SPC analysis methodology submittals. The detailed list is given because the summary document EMF-94-217 is not NRC approved 9606250332 960610 PDR ADOCK 05000254 P
Inder of Proposed Changes Index TSUP Section TSUP Page Description 1
Table of Contents page1 Reference addition of FDLRX and FDLRC Definitions 2
Table of Contents page1 FLPD and MFLPD annotated as " applicable to GE fuel" 3
Table of Contents page 11 Reflect deletion of ROD DENSITY Definition 4
Table of Contents page II Reference addition of TLHGR Definition
~
Table of Contents page XIII Reflect changed title of Specification 3/4.11.B from "APRM 5
Setpoints" to "TLHGR" 6
Table of Contents page XIV Delete reference to descriptions of Reactor Coolant System parameters which are relocated to UFSAR 7
Table of Contents page XXVI Reflect changed title of Specification 3/4.11.B from "APRM Setpoints" to "TLHGR" 8
Definitions page 1-3 FLPD and MFLPD annotated as " applicable to GE fuel" 9
Definitions page 1-3 FDLRX and FDLRC Definitions added 10 Definitions page 1-6 ROD DENSITY Definition deleted 11 Definitions page 1-7 TLHGR Definition added 12 Bases 2-1
- p. B 2-1 Editorial clarifications to MCPR Bases 13 Bases 2.1.B
- p. B 2-2 Editorial clarifications to MCPR Bases 14 Bases 2.2.A.1
- p. B 2-5 Corrected low power RWE description 15 Bases 3/4.2
- p. B 2-8,3/4.2-1 Added clarification for top of active fuel reference 16 Specification 3/4.3.B
- p. 3/4.3-2 Modified reactivity anomaly surv. to enable use of Keff methodology (SPC licensed methods) 17 Bases 3/4.3.B
- p. B 3/4.3-2 Modified reactivity anomaly smv. description to reflect use of Keff methodology 18 Bases 3/4.3.F
- p. B 3/4.3-3 Removed text reflecting GE specific methodology 19 Bases 3/4.3.F
- p. B 3/4.3-4 Modified to reflect SPC methods 20 Bases B3/4.3.L pp B 3/4.3-6,7 Added wording and clarification to reflect implementation and references to BPWS analysis 21 Bases 3/4.6.EAF
- p. B 3/4.6-3 Modified Bases description to reflect that events in addition to MSIV flux scram are verified for pressure response reamts.
22 Specification
- p. 3/4.11-2 Re-named and re-wrote specification for APRM setpoint 3/4.11.B adjustment for ' peaking factor' to reflect SPC licensed methods, with retention of GE 23 Bases 3/4.11.A
- p. B 3/4.11-1 Modified description of APRM setpoint adjustment Bases
("TLHGR") to reflect SPC fuel 25 Bases 3/4.11.C
- p. 5-5 Added water boxes to fuel description to tellect SPC bundle 5.3.A design 27 Design Features
- p. 5-6 Remove description of Rx Pressure vessel to reflect re-location 5.4 of coolant system data to UFSAR 28 Admin Controls p.6-15 Remove specific reference to 20% CRD scram time, to reflect 6.9.A SPC methodology which includes other scram time data.
29 Admin Controls
- p. 6-16 Added NRC licensed references for SPC fuel and analysis 6.9. A methodologies.
TABLE OF CONTENT 3 TOC DEFINITIONS SECTION g
i Section 1 DEFINITIONS 1
\\
4 l
A CTI O N............................................
1-1 AVERAGE PLANAR EXPOSURE (APE) 1-1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) 1-1 i
CHANNEL...........................................
1-1 i
j C H A N N EL C All B R ATI O N................................
1-1 C H A N N EL CH EC K.....................................
1-1 CH ANNEL FUNCTIONAL TEST............................
1-2 l
CO R E ALTER ATI O N....................................
1-2 CORE OPERATING LIMITS REPORT'(COLR) 1-2 CRITICAL POWER RATIO (CPR) 1-2 DOSE EQUlVALENT I-131 1-2 FRACTION OF LIMITING POWER DENSITY (FLPD) (.'// I'$f.h.M b'h I
1-3 FRACTION OF RATED THERMAL POWER (FRTP) 1-3 F_R EQUE_NCY NOTATION K......................
$0EL OgStGN L/h/f/NG N770 (f040Y) l*3 1-3 i
(f'uit des /GN L/tt/T/N& MT/C he CENTReuME MELT (FNRC))
l-7 I D EN TI FI ED LEA K AG E..................................
1-3 LIMITING CONTROL ROD PATTERN (LCRP)...................
1-3 1
LINEAR HEAT GENERATION RATE (LHGR)....................
1-3 LOGIC SYSTEM FUNCTIONAL TEST (LSFT)...................
1-3 MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD)h dd /* Mh 1-3 f
MINIMUM CRITICAL POWER RATIO (MCPR) 1-3 OFFSITE DOSE CALCULATION MANUAL (ODCM) 1-4 1
QUAD CITIES - UNITS 1 & 2 l
Amendment Nos.
l TABLE OF CONTENTS TOC DEFINITIONS SECTION PAGE OPERABLE - OPERABILITY,.............................
1-4 O PE R ATI O N A L M O D E.................................
1-4 PHYS I C S TE STS.....................................
1-4 PRESSURE BOUNDARY LEAKAGE.........................
1-4 PRIMARY CONTAINMENT INTEGRITY (PCI) 15 i
PROCESS CONTROL PROGRAM (PCP)......................
1-5 RATED THERM AL POWER (RTP)..........................
1-5 REACTOR PROTECTION SYSTEM RESPONSE TIME.............
1-5 R EPORTABLE EVENT..................................
1-5 ROD DENSITY 1y l
SECONDARY CONTAINMENT INTEGRITY (SCl)................
1-6 l
SHUTDOWN MARGIN (SDM) 1-6 l
SO U RC E C H EC K.....................................
1-6 i
THERMAL POWER 1-6 TRAw&NT Uh5M ll57 G5,95L1 r/m 04 TE TL Il@) * '.
M TRI P SYSTEM......................................
1-6 UNIDENTIFIED LEAKAGE...............................
1-7 Table 1-1, Surveillance Frequency Notation Table 1-2, OPERATIONAL MODES i
I QUAD CITIES - UNITS 1 & 2 11 Amendment Nos.
h TABLE OF CONTENTS TOC LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.11 POWER DISTRIBUTION LIMITS 3/4.11.A APLHGR i
3/4.11-1 TLHGA 3/4.11.8 A""*/?:tpst:.....................................
3/4.11-2 3/4.11.C MCPR............................................
3/4.11-3 3/4.11.D LHGR 3/4.11-4 3/4.12 SPECIAL TEST EXCEPTIONS 3/4.12. A PRIMARY CONTAINMENT INTEGRITY......................
3/4.12-1 3/4.12.B SHUTDOWN MARGIN Demonstrations......................
3/4.12-2 I
QUAD CITIES - UNITS 1 & 2 Xill Amendment Nos.
TABLE OF CONTENTS TOC DESIGN FEATURES SECTION PAGE Section S DESIGN FEATURES 5.1' SITE 5.1.A Site and Exclusion Area................................
5-1 Figure 5.1.A-1, [lNTENTIONALLY LEFT BLANK]
5.1.B Lo w Po pula tio n Zo n e..................................
5-1 Figure 5.1.B-1, [ INTENTIONALLY LEFT BLANK) 5.1.C Radioactive Gaseous Effluents 5-1 5.1.C Radioactive Liquid Effluents.............................
5-1 5.2 CONTAINMENT 5.2.A Con figura tio n.......................................
5-4 5.2.B Design Temperature and Pressure.........................
5-4 5.2.C Seco ndary Containment................................
5-4 5.3 REACTOR CORE 5.3.A Fu el A s se m blie s.....................................
55 5.3,B Contral Roc' Assemblies................................
5-5 5.4 REACTOR COOLANT SYSTEM n__
..__. C..r.u1Epis0MAlly t EFT B24Ae]
5.4.A
-.w........ui u o u
.mi..rm. m..
56
_._..-....C. I. A.i T. E. xt T./.6.N n. J L P..L G W E 4 f/. M. D.......
5.4.B 56 OUAD CITIES - UNITS 1 & 2 XIV Amendment Nos.
TABLE OF CONTENTS B TOC BASES SECTION M
3/4.11 POWER DISTRIBUTION LIMITS 3/4.11. A APLHGR 71 NGR..........................................
B 3/4.11-1 3/4.11.8
^ ""!/. SGit:
B 3/4.11-1 3/4.11.C MCPR............................................
B 3/4.11-2 3/4.11.D LHGR............................................
B 3/4.11-3 3/4.12 SPECIAL TEST EXCEPTIONS 3/4.12.A PRIMARY CONTAINMENT INTEGRITY......................
B 3/4.12-1 3/4.12.8 SHUTDOWN MARGIN Demonstrations......................
B 3/4.12-1 QUAD CITIES - UNITS 1 & 2 XXVI Amendment Nos.
Definiti:ns 1.0 1.0 DEFINITIONS FRACTION OF LIMITING POWER DENSITY (FLPD)
The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LHGR existing at location divided by the specified LHGR limit for that bundig (opph<d/e' fo6E A</).
FRACTION OF RATED THERMAL POWER (FRTP)
The FRACTION OF RATED THERMAL POWER (FRTP) shall be the measur divided by the RATED THERMAL POWER.
FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1-1.
gg IDENTIFIED LEAKAGE O
IDENTIFIED LEAKAGE shall be: a) leakage into primary containment collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or b) leakage into the primary containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of the leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE.
LIMITING CONTROL ROD PATTERN (LCRP)
A LIMITING CONTROL ROD PATTERN (LCRP) shall be a pattern which results in the core on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPR.
LINEAR HEAT GENERATION RATE (LHGRl LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.
LOGIC SYSTEM FUNCTIONAL TEST (LSFT)
A LOGIC SYSTEM FUNCTIONAL TEST (LSFT) shall be a test of all required logic components, i.e., all required relays and contacts, trip units, solid state logic elements, etc, of a logic circuit, from as close to the sensor as practicable up to, but not including the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping or total system stepu so that the entire logic system is tested.
MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD)
The MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD), shall be the highest value of the FLPD which exists in the coref (aghed/c h 6E Oc/4 MINIMUM CRITICAL POWER RATIO (MCPR)
The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists core for each class of fuel.
QUAD CITIES - UNITS 1 & 2 1-3 Amendment Nos.
l Insert #1 l
FUEL DESIGN LIMITING RATIO FOR CENTERLINE MELT (FDLRC)
The FUEL DESIGN LIMITING RATIO FOR CENTERLINE MELT (FDLR be 1.2 times the LHGR at a given location divided by the product of the TRANSIENT LINEAR HEAT GENERATION RATE limit and the FRACTION OF RATED THERMAL POWER (applicable to SPC Fuel).
l l
I l
l
l Insert #2 i
FUEL DESIGN LIMITING PATIO (FDLRX)
The FUEL DESIGN LIMITING RATIO (FDLRX) shall be the limit used to assure i
that the fuel operates within the end-of-life steady-state design criteria by, among other items, limiting the release of fission gas to the cladding plenum (applicable to SPC Fuel).
l l
- - = _ - -.
Osfinitions 1.0 1.0 DEFINITIONS REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR l
Part 50.
ROD DENSITY ROD DENSITY shall be the number of control rod notches inserted as a fraction of the number of control rod notches. All rods fully inserted is equivalent to 100% ROD DENSITY.
SECONDARY CONTAINMENT INTEGRITY (SCI)
SECONDARY CONTAINMENT INTEGRITY (SCI) shall exist when:
l l
i l
All secondary containment penetrations required to be closed during accident conditions a.
are either:
l l
- 1) Capable cf being closed by an OPERABLE secondary containment automatic isolation l
valve system, or 2)
Closed by at least one manual valve, blind flange, or deactivated automatic damper secured in its closed position, except as permitted by Specification 3.7.0.
b.
All secondary containment hatches and blowout panels are closed and sealed.
The standby gas treatment system is in compliance with the requirements of Specification c.
3.7.P.
1 I
d.
At least one door in each access to the secondary containment is closed.
The sealing mechanism associated with each secondary containment penetration; e.g.,
e.
welds, bellows or 0-rings, is OPERABLE.
j f.
The pressure within the secondary containment is less than or equal to the value required by Specification 4.7.N.1.
SHUTDOWN MARGIN (SDM) shall be the amount of reactivity by which the reae. tor is subcritical or would be suben,ical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the l
reactor is in the shutdown condition; cold, i.e. 68'F; and xenon free.
SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of CHANNEL response when the CHANNEL sensor is exposed to a radioactive source.
i r
j QUAD CITIES - UNITS 1 & 2 1-6 Amendment Nos.
D;finitisns 1.0 1.0 DEFINITIONS THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
hser f Wh TRIP SYSTEM L
\\
i A TRIP SYSTEM shall be an arrangement of instrument CHANNEL trip signals and auxiliary I
eauipment required to initiate action to accomplish a protective trip function. A TRIP SYSTEM may require one or more instrument CHANNEL trip signals related to one or more plant parameters in order to initiate TRIP SYSTEM action. Initiation of protective action may require the tripping of a single TRIP SYSTEM or the coincident tripping of two TRIP SYSTEMS.
UNIDENTIFIED LEAKAGE UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.
i QUAD CITIES - UNITS 1 & 2 1-7 Amendment Nos.
h m
- +
+ w --
e-
1 Insert #3 TRANSIENT LINEAR HEAT GENERATION RATE (TLHGR)
The TRANSIENT LINEAR HEAT GENERATION RATE (TLHGR) limit protects against fuel centerline melting and 1% plastic cladding strain during transient conditions throughout the life of the fuel (applicable to SPC Fuel).
SAFETY LIMITS B 2.1 BASES
.L1 SAFETY LIMITS The Specifications in Section 2.1 establish operating parameters to assure that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). These parameters are based on the l
Safety Limits requirements stated in the Code of Federal Regulations,10 CFR 50.36(c)(1):
" Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity."
(So.lculdc<l&
The fuel cladding, reactor pressure vessel and primary system pipingfare the principal barriers to the release of radioactive materials to the environs. Safety Limits arp established to protect the integrity of these barriers during normal plant operations and anticipi ted transients. The fuel cladding integrity limit is set such that no eeleuteted fuel damageJuneoccur as a result of an AOO. Because fuel damage is not directly observable, a step-ba::k approach is used to establish a Safety Limit for the MINIMUM CRITICAL POWER RATIO (MCPR) that represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.
The fuel cladding is one of the physical boundaries which separate radioactive materials from the environs. The integrity of the fuel cladding is related to its relative freedom from perforations or cracking. Although some corrosion or use-related cracking may occur during the life of the cladding, fission product migration from this source is incrementaty cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the protection system safety settings. While fission product migration from cladding perforations is just as measurable as that from use-related cracking, the thermally caused cladding pe'rforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.
Therefore, the fuel cladding integrity Safety Limit is defined with margin to the conditions which would produce onset of transition boiling (MCPR of 1.0). These conditions represent a significant departure from the condition intended by design for planned operation. Therefore, the fuel c' adding integrity Safety Limit is established such that no calculated fuel damage shall result from an abnormal operational transient. This is accomplished by selecting a MCPR fuel cladding integrity Safety Limit which assures that during normal operation and AOOs, at least 99.9% of the fue: rods in the core do not experience transition boiling.
Exceeding a Safety Limit is cause for unit shutdown and review by the Nuclear Regulatory Commission (NRC) before resumption of unit operation. Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational deficiency subject to regulatory review.
Amendment Nos.
QUAD CITIES - UNITS 1 & 2 B 2-1
SAFETY LIMITS B 2.1 BASES 2.1. A THERMAL POWER, Low Pressure or low Flow This fuel cladding integrity Safety Umit is established tiy establishing a limiting condition on core THERMAL POWER developed in the following method. At pressures below 800 psia (-785 psig),
the core elevation pressure drop (0% power,0% flow) is greater than 4.56 psi. At low powers and flows, this pressure differential is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low powers and flows will always be greater than 4.56 psi. Analyses show that with a bundle flow of 28 x 10' lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.56 psi driving head will be greater than 28 x 108 lb/hr.
Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. At 25% of RATED THERMAL POWER, the peak powered bundle would have to be operating at 3.86 times the average powered bundle in order to achieve this bundle power. Thus, a core thermal power limit of 25% for reactor pressures below 785 psig is conservative.
g g4,.g 4 f,,
kl asse M J.-
\\h 2.1.B THERMAL POWER, Hioh Pressure and Hioh Flow N 'NC //C y
This fuel cladding integrity Safety Limit is set such that no (mechanistic) fuel damage is calculated \\
to occur if the limit is not violated. Since the parameters which resuit in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to 8WR fuel rods, the critical power ratio (CPR) at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity Safety Umit is defined se 9e C"". ;c, ;;.. ;,,,,,;.c.;; S:! xxmb y in c;E.
ore than 99.9% of the fuel rods in the core are expected to avoid boiling transitiorprwn; the power distribution within the core and all uncertainties.
MsNuu (tws:dch. of The margin between a MCPR of 1.0 (onset of transition boiling) and the Safety Limit, is derived from a detailed statistical analysis which considers tne uncertaintics in monitoring the core operating state, including uncertainty in the critical power correlation. Because the transition boiling correlation is based on a significant quantity o' practical test data, there is a very high confidence that operation of a fuel assembly at the condition where MCPR is equal to the fuel cladding integrity Safety Limit would not produce n.2:x. i;:"$1n addition, during single recirculation loop operation, the MCPR Safety Limit is increased by 0.01 to conservatively account for increased uncertainties in the core flow and TIP measurements.
&{ c{c.ddM 4bbe d4 to OWA** kh However, if transition boiling were to occur, cladding perforation would not necessarily be expected. Significant test data accumulated by the NRC and private organizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative Amendment Nos.
QUAD CITIES - UNITS 1 & 2 B 2-2
LSSS B 2.2 BASES
-y 2.2 LIMITING SAFETY SYSTEM SETTINGS The Specifications in Section 2.2 establish operational settings for the reactor protection system i
instrumentation which initiates the automatic protective action at a level such that the Safety Limits will not be exceeded. These settings are based on the Limiting Safety System Settings requirements stated in the Code of Federal Regulations,10 CFR 50.36(c)(1):
" Limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions. Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded. "
2.2.A Reactor Protection System Instrumentation Setooints The Reactor Protection System (RPS) instrumentation setpoints specified in the table are the values at which the reactor scrams are set for each parameter. The scram settings have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and assist in mitigating the consequences of accidents. Conservatism incorporated into the transient analysis is documented by each approved fuel vendor. The bases for individual scram settings are discussed in the following paragraphs.
1.
Intermediate Ranoe Monitor. Neutron Flux - Hioh 4
The IRM system consists of eight chambers, four in each of the reactor protection system log CHANNELS. The IRM is a 5 decade,10 range, instrument which covers the range of power level between that covered by the SRM and the APRM. The IRM scram setting at 120 of 125 divisions is active in each range of the IRM. For example,if the instrument were on Range 1, the scram setting would be 120 divisions for that range; likewise, if the instrument were on Range 5, the scram would be 120 divisions on that range. Thus, as the IRM is ranged up to accommodate the incrbase in power level, the scram sotting is also ranged up.
The most significant soerces of react vity change during the power increase ara due to c I
i withdrawal. In order to ensure that the IRM provides adequate protection against the single rod I
withdrawal error, a range of rod withdrawal events has been analyzed. This analysis included starting the event at various power levels. The most severe case involves an which th actor is just sube and the IRM system is not yet on scale.
7 6" W onservatism was tal;en in this nalysis by assuming that the IRM CHANNEL closest to ou Additional The resu s of this analysis show that the reactor is scrammed the withd wn rod is bypassed f rated power, thus maintaining MCPR above the fuel cladding and peak power is limited to integrity Safety Limit. Based on the above analysis, the IRM provides protection ag i
l Amendment Nos.
I QUAD CITIES UNITS 1 & 2 8 2-5 f
LSSS B 2.2 1
l BASES l
decrease as power is increased to 100% in comparison to the level outside the shroud, to a maximum of seven inches, due to the pressure drop across the steam dryer. Therefore, at 100%
power, an indicated water level of + 8 inches water level may be as low as + 1 inches inside the shroud which corresponds to 144 inches above the top of active fuel and 504 inches above vessel l
& }.y a { schee -)Cc/ is d'cbN h $s %O > lel Qbo't V&S5*l# W 5.
Main Steam Line Isolation Valve - Closure Automatic isciation of the main steam lines is provided to give protection against rapid reactor depressurization and cooldown of the vessel. When the main steam line isolation valves begin to close, a scram signal provides for reactor shutdown so that high power operation at low reactor pressures does not occur. With the scram setting at 10% vawe closure (from full open), there is no appreciable increase in neutron flux during normai or inadvertent isolation valve closure, thus providing protection for the fuel cladding integrity Safety Limit. Operation of the reactor at pressures lower than the MSN closure setting requires the reactor mode switch to be in the Startup/ Hot Standby position, where protection of the fuel cladding integrity Safety Limit is provided by the IRM and APRM high neutron flux scram signals. Thus, the combination of main steam line low pressure isolation and the isolation valve closure scram with the mode switch in the Run position assures the availability of the neutron flux scram protection over the entire range of applicability of fuel cladding integrity Safety Limit.
l 1
1 6.
Main Steam Line Radiation - Hioh High radiation levels in the main steam line tunnel above that due to the normal nitrogen and oxygen radioactivity are an indication of leaking fuel. When high radiation is detected, a scra initiated to mitigate the failure of fuel cladding. The scram setting is high enough above l
background radiation levels to prevent spurious scrams yet low enough to promptly detect gross failures in the fuel cladding. This setting is determined based on normal full power background l
(NFPB) radiation levels without hydrogen addition. With the injection of hydrogen into the feedwater for mitigation of intergranular stress corrosion cracking, the full power background i
may be significantly increased. The settir.g is sufficiently high to allow t'.le injection of hydrogen without requiring an increase in the setting. This trip function provides an anticipatory scram to limit offsite dose consequen~es, but is not assumed to occur in the an' lysis of any design basis l
event.
I i
Amendment Nos.
QUAD CITIES - UNITS 1 & 2 B 2-8
INSTRUMENTATION B 3/4.2 BASES 3/4.2 INSTRUMENTATION 1
In addition to reactor protection instrumentation which initiates a reactor scram (Sections 2.2 and l
3/4.1), protective instrumentr-jon has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or which terminates operator errors before they result in serious consequences. The objectives of these specifications are to assure the effectiveness of the protective instrumentation when required and to prescribe the trip settings required to assure adequate performance. As indicated, one CHANNEL may be required to be made inoperable for brief intervals to conduct required surveillance. Some of the l
settings have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. It should be noted that the setpoints of other instrumentation, i
l where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inradvertent actuation of the safety system involved and exposure to abnormal situations. Smillance requsements for the instrumentation are selected in order to demonstrate proper function and OPERABILITY. Additional instrumentation for REFUELING operations is identified in Sections 3/4.10.B.
Suf WV l 3/4.2.A Isolation Actuation Instrumentation l
The isolation actuation instrumentation automatically initiates closure of appropriate isolation i
valves and/or dampers, which are necessary to prevent or limit the release of fission products from l
the reactor coolant system, the primary containment and the secondary containment in the event of a loss-of-coolant accident or other reactor coolant pressure boundary (RCPB) leak. The parameters which result in isolation of the secondary containment also actuate the standby gas treatment system. The isolation instrumentation includes the sensors, relays, and switches that l
are necessary to cause initiation of primary and secondary containment and RCPB system isolation.
Functional diversity is provided by monitoring a wide range of dependent and independent parameters. Redundant sensor input signals for each parameter are provided for initiation of isolation (one exception is standby liquid control system initiation).
1 The reactor low level instrumentation is set to trip at greater than or equal to 144 inches above the top of active fue'. (whicn is defined to be 360 inches above vessel zero). This trip initiates closure of Group 2 and 3 primary containment isolation valves but does not trip the recirculation pumps.
l For this trip setting and a 60-second valve closure time, the valvee will be closed before l
perforation of the cladding occurs, even for the maximum break.
3/4.2.8 Emeroency Core Coolino System Actuation Instrurnentation i
The emergency core cooling system (ECCS) instrumentation generates signals to automatically actuate those safety systems which provide adequate core cooEng in the event of a design basis transient or accident. The instrumentation which actuates the ECCS is generally arranged in a j
one-out-of-two taken twice logic circuit. The logic circuit is composed of four CHANNEL (s) and each CHANNEL contains the logic from the functional unit sensor up to and including all relays QUAD CITIES - UNITS 1 & 2 B 3/4.2-1 Amendment Nos.
i 1
e insert #4 Current fuel designs incorporate slight variations in the length of the active fuel, and thus the actual top of active fuel, when compared with the original fuel i
designs. Safety Limits, instrument water level setpoints, and associated LCOs refer to the top of active fuel. In these cases, the top of active fuel is defined as 360 inches above vessel zero. Licensing analyses, both accident and transient, utilize this definition for the automatic initiation and manual intervention associatt d i
l with these events.
l l
l l
l l
l 1
REACTIVITY CONTROL Anomalies 3/4.3.B 3.3 - LIMITING CONDITIONS FOR OPERATION 4.3 - SURVEILLANCE REQUIREMENTS B.
Reactivity Anomalies Cr'Ya'/ ('"f/ ^'
B.
Reactivity Anomalie C^'*"# *d*/
f co fry refun ro o cc.,f.ymf;c, The reactivity equivalenedf the difference The reactivity quivalencehf the difference between the actualqiOO DENP!? and the between th ctual*" ^""r'"f and the predictedf00 CENC'O'shall not exceed predicted 6uu cr=:= shall be verified to 1 % Ak/k.
be less than or equal to 1% Ak/k:
c, r), a./ cc6c/
f o/ (*' b'7"'
'h 1.
During the first startup following CORE APPLICABILITY:
ALTERATION (s), and OPE.qATIONAL MODE (s) 1 and 2.
2.
At least once per 31 effective full power days.
ACTION:
With the reactivity equivalence difference exceeding 1% Ak/k, within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> perform an analysis to determine and explain the cause of the reactivity difference; operation may continue if the difference is explained and corrected.
With the provisions of the ACTION above not met, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
i
~
%s QUAD CITIES - UNITS 1 & 2 3/4.3-2 Amendment Nos.
Ructivity Control B 3/4.3 l
BASES Daring MODE 5, adequate SDM is required to ensure that the reactor does not reach criticality during control rod withdrawals. An evaluation of each in-vessel fuel movement during fuelloading (including shuffling fuel within the core) is required to ensure adequate SDM is maintained during l
refueling. This evaluation ensures that the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern. For example, bounding analyses that demonstrate j
adequate SDM for the most reactive configurations during the refueling may be performed to demonstrate acceptability of the entire fuel movement sequence. These bounding analyses include additional margins to the associated uncertainties. Spiral offload / reload sequences inherently
~
satisfy the SR, provided the fuel assemblies are reloaded in the same configuration analyzed for the l
new cycle. Removine fuel from the core will always result in an increase in SDM.
.. 7.,./ ad A l +erna+;we l
... +.e 4 k eff e.,
k y,
3/4.3.B Reactivity Anomaliesl b " " ' f f* * * #
/
c ert Jim s Iatt c c.. ga.
j
( 3-th l
During each fuel cycje, excess operati g reactivity varies as fuel depletes and as any burnable poison in supplementary control is bu ed. The magnitude of this excess reactivity may be inferred from the critical rod configuration. A. fuel burnup progresses, anomalous behavior in the excess i
reactivity may be detected by comparison of the critical rod pattern selected base states to the predicted rod inventory at that state.VPower 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 j
l exceeds 1% Ak/k. Deviations in core reactivity greater than 1% Ak/k are not expected and require l
thorough evaluation. A 1% Ak/k 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 l
system.
j 3/4.3.C pontrol Rod OPERABILITY Control rods are the primary reactivity control system for the reactor. In conjunction with the Reactor Protection System, the control rods provide the means for reliable control of reactivity changes to ensure the specified acceptable fuel design limits are not exceeded. This specification, along with others, assures that the performance of the control rods in the event of an accident or transient, meets the assumptions used in the safety analysis. Of primary concern is the trippability of the control reds. Other causes for inoperability are addressed in other Specifications following this one. However, the inability to move a control rod which remains trippable does not prevent the performance of the control rod's safety function.
The specification requires that a rod be taken out-of-service if it cannot be moved with drive pressure. Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control i
rods. Control rods that are inoperable due to exceeding allowed scram times, but are movable by i
QUAD CITIES - UNITS 1 & 2 B 3/4.3 2 Amendment Nos.
Reactivity Central B 3/4.3 1
BASES control rod drive pressure, need not be disarmed electric if the shutdown margin provisions are met for each position of the affected rod (s).
If the rod is fully inserted and then disarmed electrically or hydraulically, it is in a safe position of maximum contribution to shutdown reactivity. (Note: To disarm the drive electrically, four amphenol-type plug connectors are removed from the drive insert and withdrawal solenoids, rendering the drive immovable. This procedure is equivalent to valving out the drive and is preferred, as drive water cools and minimizes crud accumulation in the drive.), if it is disarmed electrically in a non-fully inserted position, that position shall be consistent with the SHUTDOWN MARGIN limitation stated in Specification 3.3.A. This assures that the core can be shut down at all times with the remaining control rods, assuming the strongest OPERABLE control rod does not insert. The occurrence of more than eight inoperable control rods could be indicative of a generic control rod drive problem which requires prompt investigation and resolution.
In order to reduce the potential for Control Rod Drive (CRD) damage and more specifically, collet housing failure, a program of disassembly and inspection of CRDs is conducted during or after each refueling outage. This program foRows the recommendations of General Electric SIL-139 with nondestructive examination results compiled and reported to General Electric on collet housing cracking problems.
The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.
j 3/4.3.D Control Rod Maximum Scram insertion Times:
3/4.3.E Control Rod Aversoe Scram Insertion Times: and 3/4.3.F Four Control Rod Group Scram insertion Times These specifications ensure that the control rod insertion times are consistent with those used in the safety analyses. The control rod system is analyzed to bring the reactor subcritical at a rate fast enough ta prevent fuel damage, i.e., to prevent the MCPR from becoming less than the fuel cladding integrity Safety Limit. The analyses demonstrate that if the reactor is operated within the limitation set in SpeAciftion 3.11.C, the negative reactivity insertion rates associated with t5e scram performanceqas adjusted for statFstical variation in the observed dat@ result in protection oi i
the MCPR Safety Limit.
M l
Analysis of the limiting power transient shows that the negative reactivity rates, resulting from the scram with the average response of all the drives, as given in the above specification, provide the required protection, and MCPR remains greater than the fuel cladding integrity SAFETY LIMIT. In the analytical treatment of most transients, 290 milliseconds are allowed between a neutron sensor reaching the scram point and the start of motion of the control rods. This is adequate and conservative when compared to the typically observed time delay of about 210 milliseconds.
Approximately 90 milliseconds after neutron flux reaches the trip point, the pilot scram valve OUAD CITIES - UNITS 1 & 2 B 3/4.3-3 Arr.endmem.Nes.
Ructivity Centrol B 3/4.3 BASES solenoid de-energizes ar.d 120 milliseconds later the control rod motion is estimated to begin. However, 200 milliseconds rather than 120 milliseconds is conservatively assumed for time interval in the transient analyses and is also included in the allowable scram insertion tim specified in Specifications 3.3.D, 3.3.E, and 3.3.F. fin t statisti treatm f th itin ticalAesfributiop of totaljseram de is us rather the ndin luyf I
The performance of the individual control rocl Arives is monitored to assure that scram perform is not degraded.gbserved nt data or chnical Mcation-timits e used tydetermi h?
(tha aver se
~ perform ce used i e transi analyse tid the cycle ar <js its of epeh set of ntrol l
ser tests p rmed dur' the cur ompare gainst eariidr results verify
[
e perf ance of t control insertion em has t changed Ignifican If te esults sh d be de ined to f utside of t statistical lation fining th scram ormanc aracter' cs used in transient a yses, a r termmati of ther margi
/
require ts is rtaken as r ired by Sp tion 3.11. A small test sa e than at A
re
' ed by specificat' s is not stati ally signifi t and shy not be used in t re-JFf h
t etermina of therm aroinsdTndividual control rod drives with excessive scram times can be i
fully inserted into the core and de-energized in the manner of an inoperable rod drive provided t silowable number of inoperable control rod drives is not exceeded. In this case, the scram spe of the drive shall not be used as a basis in the re-determination of thermal margin requirements.
For excessive average scram insertion times, only the individual control rods in the two-by-two srray which exceed the allowed average scram insertion time are considered inoperabla.
The scram times for all control rods are measured at the time of each refueling outage. Experie with the plant has shown that control drive insertion times vary little through the operating cycle; hence no re-assessment of thermal margin requirements is expected under normal conditions. The j
history of drive performance accumulated to date indicates that the 90% insertion times of new l
l and overhauled drives approximate a normal distribution about the mean which tends to become l
skewed toward longer scram times as operating time is accumulated. The probability of a drive
(
not exceeding the mean 90% insertion time by 0.75 seconds is greater than 0.999 for a normal l
distribution. The measurement of the scram performance of the drives surrounding a drive, which exceeds the expected range of scram performance, will datect local variations and also provide assurance that local scram time limits are not exceeded. Continued monitoring of other drives exceeding the expected range of scram times provides surveillance of possible anomalous performance.
l The test schedule provides reasonable assurance of detection of slow drives before system deterioration beyond the limits of Specification 3.3.C. The program was developed on the basis of
{
the statistical approach outlined above and judgement. The occurrence of scram times within the limits, but significantly longer than average, should be viewed as an indication of a systematic problem with control rod drives, especially if the number of drives exhibiting such scram times exceeds eight, which is the allowable number of inoperable rods.
4 i
l QUAD CITIES - UNITS 1 & 2 B 3/4.3 4 Amendment Nos.
Insert #5 Transient analyses are performed for both Technical Specification Scram Speed (TSSS) and nominal scram speed (NSS) insertion times. These analyses result in the establishment of the cycle dependent TSSS MCPR limits and NSS MCPR limits presented in the COLR. Results of the control rod scram tests performed during the current cycle are used to determine the operating limit for MCPR. Following completion of each set of scram testing, the results will be compared with the assumptions used in the transient analysis to verify the applicability of the MCPR operating limits. Prior to the initial scram time testing for an operating cycle, the MCPR operating limits will be based on the TSSS insertion times.
i 1
Rtactivity Control B 3/4.3 BASES 3/4.3.J Control Rod Drive Housina Suoport 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 of 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 Section 4.6.3.5 of the UFSAR. 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.
3/4.3.K Scram Discharae Volume Vent and Drain Valves The scram discharge volume is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a reactor scram and will isolate the reactor coolant system from the containment when required. The operability of the scram discharge volume vent and drain valves assures the proper venting and draining of the volume, so that water accumulation in the volume does not occur. These specifications designate the minimum acceptable level of scram discharge volume vent and drain valve OPERABILITY, provide for the periodic verification that the valves are open, and for the testing of these valves under reactor scram conditions during each refueling outage.
de L P's P )
(i~ rower (up t._
' Y' I'
- '7#'
l 3/4.3.L Rod Worth Minimizer hl.t us are cal /p li e is ei + tree.Jal.
n e.
G.,, e swiete ec. m:+y -.cA]
N Control rod withdrawal and insertion seque s are established to assure that the maximum insequence individual contrgl rod or control rod egments which are withc rawn at any time during the fuel cycle could not b; =:05 :n;;;h to resu in a peak fuel enthalpy greater than 280 cal /grn in the event of a control rod drop accident. These sequences are d;=::p;d p ::::: lr':::l Oper::!:r Of the un't f:!!:=ln; :ny ::fu ':n; ;;;;;; :nd the requirement that an operator follow these sequences is supervised by the RWM or a second technically qualified individual. These sequences are developed to limit reactivity worth of control rods and, together with the integral rod velocity limiters and the action of the control rod drive system, limit potential reactivity insertion such that the results of a control rod drop accident will not exceed a maximum fuel energy content of 280 cal /gm. The peak fuel enthalpy of 280 cal /gm is below the energy content at which rapid fuel dispersal and primary system damage have been found to occur based on experimental data. Therefore, the energy deposited during a postulated rod drop accident is significantly less than that reqdred for rapid fuel dispersal.
The analysis of the control rod drop accident was originally presented in Sections 7.9.3,14.2.1.2, and 14.2.1.4 of the original SAR. Improvements in analytical capability have allowed a more refined analysis of the control rod drop accident which is discussed below.
Every operating cycle the peak fuel rod enthalpy rise is determined by comparing cycle specitic parameters with the results of parametric analyses. This peak fuel rod enthalpy is then compared QUAD CITIES - UNITS 1 & 2 B 3/4.3-6 Amendment Nos.
I l
Rcsctivity Contral B 344.3 i
s BASES to the analysis limit of 280 cailgm to demonstrate compliance for that operating cycle if the cycle specific parameters are outside the range used in the parametric study, an extension of the enthalpy may be required. Some of the cycle specific parameters used in the analysis are:
maximum control rod worth, Doppler coefficient, effective delayed neutron fraction and maximum four bundle local peaking factor.8:The ZO eppieved n eihedelegy lleied lr. Opeelf;eet pt:>!d : : det:!!:d d::::!pti;r. 0 50.:$2de!:;y :::d 5 p:-'::-5; $: : d &cp ent!;:::.
The rod worth minimizer provid s automatic supervision to assure that out-of-sequence control rods will not be withdrawn or i sorted, i.e., it limits operator deviations from planned withdrawal j
sequences (reference UFSAR ction 7.7.2). It serves as a backup to procedural control of control rod worth, in the event that tL e rod worth minimizer is out-of-service when required, a second licensed operator or other technically qualified individual who is present at the reactor console can manually fulfill the control rod' pattern conformance function of the rod worth minimiter. In this case, the normal procedural controls are backed up by independent procedural controls to assure conformance.
m me+L.J. logy vss a S.,-
+1, < c.,4r.g r,)
Je, y res;de +
sest s;>
- > A c.
sepe.ve d s a ;,
p,+
,(
y Rod Block Monitor h~ tic. se.
3/4.3.M 6sses re 4ce e./
1, S r** ; #i c"+l"
'9A'-
The rod block monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided, and one of these may be bypassed from the console for maintenance and/or testing. Tripping of one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the operator, who withdraws control rods according to a written sequence. The specified restrictions with one channel out-of-service conserv1tively assure that fuel damage will not occur due to rod withdrawal errors when this condition exists.
3/4.3.N Economic Generaticn Control System Operation of the facility with the economic generation control system (EGC) (automatic flow control) is limited to the range of 65% to 100% of rated core flow. In this flow range and above 20% of RATED THERMAL POWER, the reactor could safely tolerate a rate of change of load of 8 MWe/sec (reference UFSAR Section 7.7.3.2). Limits within the EGC and the flow control system prevent rates of change greater than approximately 4 MWe/sec. When EGC is in operation, this fact will be indicated on the main control room console.
QUAD CITIES - UNITS 1 & 2 B 3/4.3-7 Amendment Nos.
PHiMAHY SYSTEM BOUNDARY b 3/4.6 BASES 3/4.6.E Safety Valves hh 50f gg' 3/4.6.F Relief Valves The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief sy:, tem, the size and number of safety valves are selected such that peak pressure in the nuclear system will nnt avemari the ASME Code limits for the reactor coolant pressure boundary. /pletverorhre ar4tiondy'stenuffust Godphe,Jer6st s_sverrepessuyaffon trentiaMva6 dons hs6detarfnined thef the acc
~
os eraartinsienfis the1!foserf3 of 3Mne mjurt'steanfline)stilation;.6en wrnwed bVin resce6r
[pefa'm perfiiq4rr6utropffux. Jh6 analysis results demonstratp4 hat thdesign gety valfe canadity is ablK mainesinma cemetyssure)p61owffe ASMFrCode linGt of 118% of t(e renc4 J resspre ves,perdesigp6essupr.'
/
/
The relief valve function is not assumed to operate in response to any accident, but are provided to remove the generated steam flow upon turbine stop valve closure coincident with failure of the turbine bypass system. The relief valve opening pressure settings are sufficiently low to prevent the need for safety valve actuation following such a transient.
Each of the five relief valves discharge to the suppression chamber via a dedicated relief valve discharge line. Steam remaining in the relief valve discharge line following closure can condense, creating a vacuum which may draw suppression pool water up into the discharge line. This condition is normally alleviated by the vacuum breakers; however, subsequent actuation in the presence of an elevated water leg can result in unacceptably high thrust loads on the discharge piping. To prevent this, the relief valves have been designed to ensure that each valve which closes will remain closed until the normal water level in the relief valve discharge line is restored.
The opening and closing setpoints are set such that all pressure induced subsequent actuation are limited to the two lowest set valves. These two valves are equipped with additionallogic which functions in conjunction with the setpoints to inhibit valve reopening during the elevated water leg duration time following each closure.
Each safety / relief valve is equipped with diverse position indicators which monitor the tailpipe acoustic vibration and temperature. Either of these provide sufficient indication of safety / relief valve position for normal operation.
3/4.6.G Leakuoe Detection Systems The RCS leakage detection systems required by this specification are provided to monitor and dstect leakage from the reactor coolant pressure boundary. Limits on leakage from the reactor coolant pressure boundary are required so that appropriate action can be taken before the integrity of the reactor coolant pressure boundary is impaired. Leakage detection systems for the reactor coolant system are provided to alert the operators when leakage rates above the normal bzckground levels are detected and also to supply quantitative measurement of leakage rates.
QUAD CITIES - UNITS 1 & 2 B 3/4.6-3 Amendment Nos.
l I
Insert #6 The overpressure protection system must accommodate the peak transient pressure during the most severe licensing basis pressurization transient. This includes, but is not limited to, the licensing basis ASME Section III compliance event which is the closure ofall MSIVs with no credit for relief function or direct scram from valve position. For the purpose of the ASME Section 111 analysis, the SRV (combination safety / relief valve) is assumed to operate only in the safety mode. The ASME Section III analysis demonstrates that the design capacity of the SVs and SRV is capable of maintaining the r:: actor pressure below the ASME code limit. The licensing basis pressurization transients are evaluated for each reload to assure compliance with the ASME Code limit of 110% of vessel design pressure.
This LCO ensures the acceptance limit of 1375 psig is met dudng the most severe licensing basis pressurization transient.
t
& Aa 1
WER DISTRIBUTION LIM
[d6 APRM Setpoints 3/4.11.
(3.11 - LiMlTING CONDITIONS FOR OPERATION4.11 - SURVEILLANCE REQUIREMEN B.
Average Powerdange Monitor Setpoi B.
Average' Power Range Monitor Setpoints
\\
The Averag/
,/
/
\\
e Power Range Mon' r (APRM)
The value of MFLPD shall be verified:
MAXIMUM FRACTION OF MITING
[ 1. At least once er/'
f' gain or,setpoints shall be set ch that the
/
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, l
POWER DENSITY (MFI,PD) shall be lessCTION OF RATED'/
[
than or equal to the 2.
Within 1;2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after completion of a j
THERMAL POWER FRTP).
THERMAL POWER increase of at'least j
157o 'of RATED THERMAL POWER, and 1
3.. dnitially and at least onc/
APPLICABIL e per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> I
l l
/
,/ when the reactor is operating with MFLP OPERATIONAL MODE 1,[when THERMAL l
greater than.ellual to FRTP.
POWER is greater than,o equal to 25% of RATED THERMAL P WER.
4.
The provisilons of Specification 4.0.D are
/
not applicable.
l
./
ACTION:
With MFL greater than FRTp initiate correcti ACTION within 15 minutes and withi hours either-
/.
Restore MFLP o within its limit,
/
7 2.
Adjust the ow biased APRM etpoints
,. /
specifie in Specifications
.A and 3.2.E y FRTP/MFLPD,
/
3.
Adjust
/
/
With the provisions of the ACTION above not
/
met, reduce THERMAL POWER ss than f
i 25% of TCD THERMAL PO R within the
\\
next 4 ours.
/
\\
wEided that the adjusted APbading does not exceed 100% of RATE
{
c ES - UNITS 1 & 2 Amendment Q4.11-2p -
~
l I
POWER DISTRIBUTION LIMITS TLHGR 3/4.11.8 3.11 - LIMITING CONDITIONS FOR OPERATION 4.11 - SURVEILLANCE REQUIREMENTS B.
TRANSIENT LINEAR HEAT GENERATION B.
TRANSIENT UNEAR HEAT GENERATION RATE RATE i
The TRANSIENT LINEAR HEAT The value of FDLR all be verified: -
GENERATION RATE (TLHGR) shall be 4
(ct) maintained such that the FUEL DESIGN 1.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
' nu"7'Mn RATIO for CENTERLINE MELT (FDLRdis less than or equal to 1.0.
2.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a Where FDLRC is equal to:
THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and (LHGR) (1.21 (TLHGR) (FRTP) 3.
Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with 4
APPLICABILITY:
FDLRC greater than or equal to 1.0.
OPERATIONAL MODE 1, when THERMAL 4.
The provisions of Specification 4.0.D i
POWER is greater than or equal to 25% of are not applicable.
RATED THERMAL POWER.
ACTION:
With FDLRC greater than 1.0, initiate corrective ACTION within 15 minutes and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:
1.
Restore FDLRC to less than or equal to 1.0, or 2.
Adjust the flow biased APRM setpoints specified in Specifications 2.2.A and 3.2.E by 1/FDLRC, or
- e(ac]h APRM gain such ha b
3.
Adjust the APRM readings are 210 ames the FRACTION OF RATED THERMAL POWER (FRTP) times FDLRC.
With the provisions of the ACTION above not met, reduce THERMAL POWER to less than 25% of RATED THERMAL POWER 9,j p., Fp ec. Aajush edr oss desce/e.,
l within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
/$. ?'Ap h Scpvr 9 cd AMM mb"] WIN-) E"^ Ye f"#
'h d
et
_4 e
~
3 gProvided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel.
oq 3/4.11-2 Amendment Nos.
0"CCOEN - VisT5 210 -
$UAD CITf65 - ()N/T5I&w
POWER DISTRIBUTION LIMITS B 3/4.11 BASES 3/4.11. A AVERAGE PLANAR LINEAR HEAT GENERATION RATE
_G E fue/
This specmcation assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46. The specification also assures that fuel rod mechanical integrity is maintained during normal and transient operations.
The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axiallocation and is dependent only secondarily on the rod-to-rod power distribution within an assembly. The peak clad temperature is calculated assuming a LINEAR HEAT GENERATION RATE (LHGR) for the highest powered rod which is equal to or _less than the design LHGR corrected for densification.
The APLHGR limits specified are equivalent to the LHGR of the highest powered fuel rod assumed in the LOCA analysis divided by its local peaking factor. A conservative multiplier is applied to the LHGR assumed in the LOCA analysis to account for the uncertainty associated with the measurement of the APLHGR.
T11~s w d # 9 {
\\
fThe calculational procedure used to establish the maximum APLHGR values uses NRC acoroved i
calculational models which are consistent with the requirements of 4dhtfWff)tif4f)A$8#affAHb The approved calculational models are listed in Specification 6.9.
l h__ff @Mbl L/0 C The daily requirement for calculating APLHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when i there have not been significant power or control rod changes. The requirement to calculate APLHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while still allotting time for the power distribution to stabilize. The requirement for calculating APLHGR after initially determining a LIMITING CONTROL ROD PATTERN exists ensures that APLHGR will be known following a change in THERMAL POWER or power shape, that could place operation above Qthermal limit.
s I
3/4.11.8
.",PF'."OCTFOiTO kfS4NS/E#T W N N I M N #
h h
f be t nI r 'gh scram setting and control rod blo.:k functions of the APRM instruments for both two recirculation loop operation and single recirculation loop operation must be adjusted to ensure that the MCPR does not become less than the fuel cladding safety limit or that 21% plastic strain does not occur in the degraded situation. The scram settings and rod block settings are adjusted in accordance with the formula in this specification when the value of MFLP rydicates a higher peaked power distribution to ensure that an LHGR transient would not be ir1creased in the degraded condition.
D Se' l W
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QUAD CITIES - UNITS 1 & 2 B 3/4.11-1 Amendment Nos.
l
insert #8 SPC Fuel This specification assures that the peak cladding temperature of SPC fuel following a postulated design basis loss-of-coolant accident will not exceed the Peak Cladding Temperature (PCT) and maximum oxidation limits specified in 10CFR50.46.
The calculational procedure used to establish the Average Planar Linear Heat Generation Rate (APLHGR) limits is based on a loss-of-coolant accident analysis.
The PCT following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod-to-rod power distribution within the assembly.
The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits for two-loop and single-loop operation are specified in the Core Operating Limits Report (COLR).
i
1 l
i insert #9 SPC Fuel l
The Fuel Design Limiting Ratio for Centerline Melt (FDLRC) is incorporated to protect j
the above criteria at all power levels considering events which cause the reactor power l
to increase to 120% of rated thermal power.
The scram settings must be adjusted to ensure that the TRANSIENT LINEAR HEAT l
GENERATION RATE (TLHGR) is not violated for any power distribution.
This is accomplished using FDLRC. The scram setting is decreased in accordance with the formula in Specification 3.11.8, when FDLRC is greater than 1.0.
l The adjustment may also be accomplished by increasing the gain of the APRM by l
FDLRC. This provides the same degree of protection as reducing the trip setting by l
1/FDLRC by raising the initial APRM reading closer to the trip setting such that a scram l
would be received at the same point in a transient as if the trip setting had been l
reduced.
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POWER DISTRIBUTION LIMITS B 3/4.11 BASES 3/4.11.C MINIMUM CRITICAL POWER RATIO The required operating limit MCPR at steady state operating conditions as specified in Specificatio 3.11.C are derived from the established fuel cladding integrity Safety Limit MCPR, and an analy of abnormal operational transients. For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Umit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2.
Are' To assure that e fuel cladding integrity Safety Limit is not exc ded during any anticipated abnormal ope tional transient, the most limiting transients 5:_ h= analyzed to determine which result in the I rgest reduction in the CRITICAL POWER RATIO (CPR). The type of transients svaluated hange of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest delta MCPR. When added
{
to the Safety Limit MCPR, the required minimum operating limit MCPR of Specification 3.11.C is cbtained and presented in the CORE OPERATING LIMITS REPORT.
The steady state values for MCPR specified were determined using NRC-approved methodology listed in Specification 6.9.
g fi urp of CP ultip
- ive f or sp ied i eCO PER G IJMITS O
s to ine PR o tin its a er t rated re flo onditi
. A)t elIs th 10 ed f
,the uired PR i e pro et of MCP d the ratepffow R
ti k. The CPR tiplie ssures at the fety it MCP ill no)4e vio ed.
/
At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant cxperience indicates that the resulting MCPR value has considerable margin. Thus, the demonstration of MCPR below this power levelis unnecessary. The daily requirement for csiculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. Thr. requirement for calculating MCPR after initially determining that a LIMITING '.:ONTROL ROD PATTERN exists ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation above O thermallimit.
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QUAD CITIES - UNITS 1 & 2 B 3/4.11-2 Amendment Nos.
Insert #10 MCPR Operating Limits are presented in the CORE OPERATING LIMITS REPORT (COLR) for both Nominal Scram Speed (NSS) and Technical Specification Scram Speed (TSSS) insertion times. The negative reactivity insertion rate resulting from the scram plays a major role in providing the required protection against violating the Safety Limit MCPR during transient events.
Faster scram insertion times provide greater protection and allow for improved MCPR performance. The application of NSS MCPR limits utilizes measured data that is faster than the times required by the Technical Specifications, while the TSSS MCPR limits provide the necessary protection for the slowest allowable average scram insertion times identified in Specification 3.3.E. The measured scram times are compared with the nominal scram insertion times and the Technical Specification Scram Speeds, The appropriate operating limit is applied, as specified in the COLR.
For core flows less than rated, the MCPR Operating Limit established in the specification is adjusted to provide protection of the Safety Limit MCPR in the event of an uncontrolled recirculation flow increase to the physicallimit of the pump. Protection is provided for manual and automatic flow control by applying the appropriate flow dependent MCPR limits presented in the COLR. The MCPR Operating Limit for a given power / flow state is the greater value of MCPR as given by the rated conditions MCPR limit or the flow dependent MCPR limit. For automatic flow control, in addition to protecting the Safety Limit MCPR during the flow run-up event, protection is provided to prevent exceeding-the rated flow MCPR Operating Limit during an automatic flow l
increase to rated core flow.
I
REACTOR CORE 5.3 5.0 DESIGN FEATURES i
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5.3 REACTOR CORE g
gM/;ej p o7 (oofat-tvcherr0S Of b#
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Fuel Assemblies 5.3.A The reactor core shall contain 724 fuel assemb es. Each assembly consists of a matrix of Zircaloy clad fuel rods with an init' omposition of natural or slightly i
enriched uranium dioxide as fuel material,
- ^-
Limited substitutions of j
zirconium alloy, in accordance with NRC-approved applications of fuel rod
)
configurations, may be used. Fue! assemblies shall be limited to those fuel designs i
that have been analyzed with applicable NRC staff-approved codes and methods, and t
shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.
Control Rod Assemblies 5.3.8 The reactor core shall contain 177 cruciform shaped control rod assemblies. The control material shall be boron carbide powder (B.C) and/or hafnium metel. The control rod assembly shall have a nominal axial absorber length of 14311ches.
QUAD CITIES - UNITS 1 & 2 5-5 Amendment Nos.
REACTOR COOLANT SYSTEM 5.4 5.0 DESIGN FEATURES 5.4 REACTOn OOOL^MT SYSTEM
[J//76A/7/0//AA/ I 8/8//#
Desian Pressure and Temperature s
5.4.A The reactor coolant system is designed and shall be maintained:
i 1.
In accordance with the code requirements specified in Section 5 of the UFSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, 2.
For a pressure and temperature of:
1175 psig at 565'F on the suction side of the recirculation pump.
a.
b.
1450 psig at 575'F from the recirculation pump discharge to the outlet side of the discharge shutoff valve.
1325 psig at 580'F from the discharge shutoff valve to the jet pumps.
c.
Volume 5.4.B The total water and steam volume of the reactor vessel and recirculation system is approximately 15,679 cubic feet at 68'F.
I QUAD CITIES - UNITS 1 & 2 5-6 Amendment Nos.
Riporting RIquirrments 6.9 ADMINISTRATIVE CONTROLS 4.
Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the facility during the previous calendar year shall be submitted prior to April 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the facility. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix ! to 10 CFR Part 50.
5.
Monthly Operating Report Routine reports of operating statistics and shutdown experience, including documentation of all challenges to safety valves or safety / relief valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S.
Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator of the NRC Regional Office, no later than the 15th of each month following the calendar month covered by the report.
6.
CORE OPERATING UMITS REPORT Core operating limits shall be established and documented in the CORE a.
OPERATING UMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:
(1) The Control Rod Withdrawal Block Instrumentation for Table 3.2.E-1 of i
i Specification 3.2.E.
(2) The Average Planar Unear Heat Generation Rate (APLHGR) Umit for Specification 3.11.A.
(3) The Unear Heat Generation Rate (LHGR) for Specification 3.11.D.
(4) The Minimum Critical Power Operating Umit (including 20 scram insertion time) for Specification 3.11.C. This includes rated and o.-rated f'ow conditions.
b.
The analytical methods used to determine the operating limits shall be those previously reviewed and approved by the NRC in the latest approved revision or supplement of topical reports:
(1) NEDE 24011-P-A, " General Electric Standard Application for Reactor Fuel,"
(latest approved revision).
(2) Commonwealth Edison Topical Report NFSR-0085, " Benchmark of BWR Nuclear Design Methods," (latest approved revision).
QUAD CITIES - UNITS 1 & 2 6-15 Amendment Nos.
i
RIporting Requir; mints 6.9 ADMINISTRATIVE CONTROLS
_m (3) Commonwealth Edison Topical Repart NFSR-0085, Supplement 1,
" Benchmark of BWR Nuclear Design Methods - Quad Cities Gamma Scan Comparisons," (latest approved revision).
[s seef 77 (4) Commonwealth Edison Topical Report NFSR-0085, Supplement 2,
" Benchmark of BWR Nuclear Design Methods - Neutronic Licensing Analyses," (latest approved revision).
~
The core operating limits shall be determined so that all applicable limits (e.g., fuel c.
thermal-mechanical limits, core thermal hydraulic limits, ECCS limits, nuclear limits such as sbittdown margin, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector.
6.9.B Special Reports Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.
QUAD CITIES - UNITS 1 & 2 6-16 Amendment Nos.
Insert #11 (5) XN-NF-80-19(P)(A)," Exxon Nuclear Methodology for Boiling Water Reactors."
(6) XN-NF-85-67(P)(A)," Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel."
(7) XN-NF-82-06(P)(A)," Qualification of Exxon Nuclear ?uel for Extended Burnup:
Extended Burnup Qualification of ENC 9x9 BWR Fue".."
(8) ANF-89-014(P)(A)," Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advance Nuclear Fuels Corporation 9x9-IX and 9x9-9X BWR Reload Fuel."
(9) ANF-89-98(P)(A)," Generic Mechanical Design Criteria for BWR Fuel Designs."
(10) XN-NF-79-71(P)(A)," Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors."
(11) ANF-1125(P)(A),"ANFB Critical Power Correlation."
(12) ANF-524(P)(A)," Advanced Nuclear Fuels Critical Power Methodology for Boiling Water Reactors."
(13) ANF-913(P)(A), Volume 1,"COTRANSA 2: A Computer Program for Boiling Water Reactor Transient Analysis."
(14) ANF-91-048(P)(A)," Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model."
(Is) C e. we l+k E A 4..
T. g; c. i Aep.,+
9 Fr R -oo q i, us...&.mek
.4 1
CMM o/Mi c Ao Bunr3 BWR Nuclear bast$n M e de t.s
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l Attachment C i
I Evaluation of Significant Hazards Considerations 4
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Evaluaten cf Significant Hazards Ccnsiderations Attachment C Page1of5 C. Evaluation of Significant Hazards Considerations The fuel supplier for Quad Cities is being changed from General Electric (GE) to Siemens Power Corporation (SPC). As a result, certain items in the Technical Specifications are being revised. These j
j changes can be classified in three categories: (a) fuel thermal limits, (b) miscellaneous, and (c) minor l
changes not related to the SPC transition. Each is discussed below.
- a. Fuelthermallianits The fuel thermal limits in the Technical Specifications are LHGR, APLHGR, and MCPR. Each fuel vendor provides LHGR and APLHGR limits for their fuel. As required by the Technical Specification Surveillance Requirements (SR's), each fuel type will continue to be monitored via its vendor supplied LHGR and APLHGR limits. As such, the change to the Technical Specifications Bases for LHGR and APLHGR will be the addition of background information related to the SPC LHGR and APLHGR. The Limiting Conditions for Operation (LCO), Action Statements and SR's are unaffected since they refer to the Core Operating Limits Report for the fuel type dependent limits.
The CPR is calculated using a Nuclear Regulatory Commission (NRC) approved CPR correlation. The GE correlation (GEXL) is being replaced by the SPC correlation, ANFB The i
co-resident GE fuel will be monitored by the ANFB correlation supplemented with bundle geometry dependent factors to ensure the calculated CPR data is conservative with respect to that which would have been calculated by the GEXL correlation. This mixeJ core treatment of CPR j
is being documented in EMF-ll25(P) Supplement 1 Appendix C,"ANFB Critical Power Correlation Application for Co-Resident Fuel," November 1995. The Technical Speci5 cations and Bases related to MCPR are being modified to better reflect the use of scram timing results in determining the operating limit.
- b. Miscellaneous changes The Basis for the Reactivity Anomaly surveillance is being modined to enable the use of monitored Keffin addition to control rod density. The fuel description in Specification 5.3.A is being modified to tellect the water box in the SPC fuel design. The Bases for the Rod Worth Minimizer B 3/4.3.L was changed to better reflect control rod sequence development practices, referenced methodology, and wording.
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3 Evaluati:n cf Significant Hazards Ccnsiderations Attachment C Page 2 of 5
- c. Minor Change Not related to SPC Transition l
A wording change in the Bases (page B 2-5) clarifying the power level at which the IRM system terminates the low power control rod withdrawal error event. The MCPR Bases are clarified via editorial changes. Clarification concerning the top of active fuel (360") for various fuel types was added to pages B 2-8 and B 3/4.2-1. The reactor coolant system temperature, pressure and i
j volume in Specification 5.4 is being re-located to the UFSAR as a line item from Improved i
Technical Specifications, per NUREG-1434. The reference for the calculational procedure used to establish the maximum APLHGR for LOCA was broadened from reference to Appendix K of.
10 CFR Part 50 to 10 CFR Part 50.46.
Comed has evaluated the proposed Technical Specification amendment and determined that it does nct represent a significant hazards consideration. Based on the criteria for defining a significant hazard consideration established in 10CFR50.92 (c), operation of Quad Cities Units 1 and 2 in accordance with the proposed amendment (s) will not represent a significant hazards consideration for the following reasons:
3 These changes do not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated.
The probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences Limits will be established consistent with NRC approved methods to ensure that fuel pe formance during normal, transient, and accident conditions is acceptable. The proposed Technical Specifications amendment reflects previously approved SPC methodology used to analyze normal operations, including anticipated operational occurrences (AOOs), and to determine the potential consequences of accidents.
Licensine Methods and Models The proposed amendment is to support operation with NRC approved fuel and licensing methods supplied from Siemens Power Corporation. In accordance with FSAR Chapter 15, the same accidents and transients will be analyzed with the new fuel and methods as were analyzed by GE for GE fuel. The analysis methods and models are NRC approved. These approved methods and models are used to determine the fuel thermal limits (e.g., LHGR, APLHGR, MCPR). The SPC core monitoring code enables the site to monitor k aas well as rod density to perform the reactivity anomaly surveillance. This is consistent with GE methodology. The support systems for minimizing the consequences of transients and accidents are not affected by the proposed amendment. Therefore, the change in licensing analysis methods and models does not significantly increase the probability of an accident or the consequences of an accident presiously identified.
New Fuel Design The use of ATRIUM 9B fuel at Quad Cities does not involve a significant increase in the probability or consequences of any accident previously evaluated in the FSAR. The ATRIUM 9B fuel is generically approved for use as a reload BWR fuel type
(
Reference:
ANF-89 014(P)(A) Rev.1 Supplement 1. Generic Mechanical Design for Advanced Nuclear Fuels 9X9-IX and 9X9-9X BWR Reload Fuel).
Limiting
- - ~..
Evaluauon of Significant Hazards Ccasiderati:ns Attachment C Page 3 of 5 postulated occurrences and normal operation have been analyzed using NRC-approved methods for the ATRIUM 9B fuel design to ensure that safety limits are protected and that acceptable transient and accident performance is maintained.
The reload fuel has no adverse impact on the performance of in-core neutron flux instrumentation on CE response. The ATRIUM-9B fuel design will not adversely affect performance of neutron instrumentation nor will it adversely affect the movement of control blades relative to the GE fuel. The exterior dimensions of the ATRIUM-9B fuel have been evaluated by Comed; the SPC fuel provides adequate clearances relative to the GE10 fuel installed at Quad Cities. Thus, no increased interactions with the adjacent control blade and nuclear instrumentation are created.
Additionally, given the above mentioned overall envelope similarities, no problems are anticipated with other station equipment such as the fuel storage racks, the new fuel inspection stand and the spent fuel pool fuel preparation machine. Therefore, the probability of adverse interactions between the Siemens fuel and components in the core and fuel handling equipment is not significantly increased.
The ATRIUM 9B design is neutronically compatible with the existing fuel types and f
core components in the Quad Cities core. SPC tests have demonstrated that the j
A'IRIUM-9B fuel design is hydraulically compatible with the GE9/GE10 fuel. The bundle pressure drop characteristics of the ATRIUM 9B bundle are similar to those of the GE9/GE10 fuel design, hence core thermal-hydraulic stability characteristics are not adversely affected by the ATRIUM 9B design. Cycle stability calculations are performed by SPC. Therefore, the probability of thermal hydraulic instability is not significantly increased An evaluation of the Emergency Procedures is being performed to ensure that the use of the ATRIUM-9B fuel at Quad Cities does not alter any assumptions prniously made in evaluating the radiological consequences of an accident at Quad Cities 1
Station. Therefore, the radiological consequences of accidents are not significantly increased ~
Methods approved by the NRC are being used in the evaluation of fuel performance during normal and abnormal operating conditions. The Comed and SPC methods to be used for the cycle specific transient analyses have been prmiously NRC approved.
The proposed methodologies are administrative in nature and do not significantly affect any accident precursors or accident results; as such, the proposed incorporation of the SPC methodologies for Quad Cities does not significantly increase the probability or consequences of any previously maluated accidents.
The description of the fuel is modified to include the water box design of the NRC approved ATRIUM-9B fuel. This change is administrative.
Roiew of the above concludes that the probrJMlity of occurrence and the consequences of an accident proiously evaluated in the safety analysis report have not been significantly increased.
i
Evaluation of Significant Hazards Considerations Attachment C Page 4 of 5 Comed has evaluated the proposed Lic-nse -endment and determined that it does not represent a significant hazards consideration. Based on the criteria for defining a significant hazard consideration establisLd in 10CFR50.92 (c), operatioa of Quad Cities Units I and 2 in accordance with the propcsed amendment (s) will not represent a sig ti'icant hazards considcration for the following reasons:
These changes do not:
- 2. Creste the possibility of a new or different kind of accident from any accident previously evaluated:
Creation of the possibility of a new or different kind of accident would require the creation of one or more new precursors of that accident. New accident precursors may be created by modifications of the plant configuration, including changes in allowable modes of operation.
Licensine Methods and Models The proposed Technical Specification amendmem refleas previously approved SPC methodology used to analyze normal operations, including AOOs, '.nd to determine the potential consequem es Of accidents. In accordance with FSAR Chapter 15, the same accidents and transients veill be analyzed with the new fuel and methods as were analyzed by GE for GE fuel. As stated above, the proposed changes do not permit modes of reactor operation which C Ter fron. those currently permitted; therefore, the poesbility of a new or diffetent kind cf accident is tmt created. Plant support equipment is not aff cted by the poposed cht.ag-s, therefore, no nes !ailure modes are created.
New Fuel Desien The bash design concept of a 9x9 fuel pin array with an internal water box has been used in vad dj assembly programs and ir reload quantities in Europe since 1986. WNP-2 has loaded renad gutities since 1991. Approximately 650 water box assemblies have been irradiated in the Un ited States through 1995, with a subster. dally higher number being irradiated overseas. The Nhc has reviewed and appn ved the ATi l'JM-9D fuel design (
Reference:
ANF-89-014(P)(A)
Rev.1 Supplement 1. Gern a dechanican IMsign for Advanced Naclear Fuels 9X9-IX and 9X9-9X BWR Reload Fuel). The m::.rities in M1 design and opecti3n between GE and SPC, and the previous Boiling Water Restor aperience with both vendors' fuel, indicate there would be no new or different types of acddents for Quad Cities than have in en considered for the existing fuel. Therefore, the use of ATRiiTM-vB fuel at Quad Cities doemot create the possibility of a new or different kind of accident from any accident previously enluated.
i
Evaluation of Significant Hazards Considerations Attachment C Page 5 of 5 Comed has evaluated the proposed License amendment and determined that it does not represent a significant hazards consideration. Based on the criteria for defining a significans hazard consideration established in 10CFR50.92 (c), operation of Quad Cities Units I and 2 in accordance with the proposed amendment (s) will not represent a significant hazards consideration for the following reasons:
These changes do not:
- 3. Involve a significant reduction in the margin of safety for the following reasons:
The existing margin to safety is provided by the existing acceptance criteria (e.g.,10CFR50.46 timits). The proposed Technical Specification amendment reflects previously approved SPC methodology used to demonstrate that the existing acceptance criteria are satisfied. The revised methodology has been previously reviewed and approved by the USNRC for application to reload cores of GE BWRs. References for the Licensing Topical Reports which document this methodology, and include the Safety Evaluation Reports prepared by the USNRC, are added to the Reference section of the Technical Specifications as part of this amendment.
Licensing Methods and Models i
i The proposed amendment does not involve changes to the existing operability criteria. NRC approved methods and established limits (implemented in the COLR) ensure acceptable margin is i
maintained. The Comed and SPC reload methodologies for the ATRIUM-9B reload design are consistent with the Technical Specification Bases. The Limiting Conditions for Operation are taken into consideration while performing the cycle specific and generic reload safety analyses.
NRC approved methods are listed in Section 6 of the Technical Specifications.
Analyses performed with NRC-approved methodology have demonstrated that fuel design and licensing criteria will be. met during normal and abnormal operating conditions. The same l
margins of safety are utilized by SPC as GE (e.g., limits on peak cladding temperature, cladding oxidation, piastic strain). Therefore, there is not a significant reduction in the margin of safety.
New Fuel Design The exterior dimensions of the ATRIUM-9B fuel assembly result in equivalent clearances relative to the GE108. Thus, no increased interactions with the adjacent control blade and nuclear instrumentation are created. The change does not adversely impact equipment important to safety; therefore, the margin of safety is not significantly reduced.
Guidance nas been provided in " Final Procedures and Standards on No Significant flazards Considerations," Final P.ule,51 FR 7744, for the application of standards to license change requests for determination of the existence of significant hazards considerations. This document provides examples of amend nents which are and are not considered likely to involve significant hazards considerations. This proposed amendment most closely fits the example of a change which may either result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable eriteria with respect to the system or component specified in the Standard Review Plan.
This proposed amendmert. does not involve a significant relaxation of the criteria used to establish safety limits, a significant raaxation of the bases for the limiting safety system settings or a significant relaxation i
of the bases for the limiting conditions for operations. Therefore, based on the guidance provided in the Federal Register and the criteria established in 10 CFR 50.92(c), the proposed change does not constitute a significant hazards consideration.
Attachment D Environmental Assessment Applicability Review i
Environm:ntal Assessment Applicability Review Attachment D PageIofI D. Environmental Assessment Applicability Review Comed has evaluated the proposed amendment against the criteria for identification oflicensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. It has been determined that the proposed changes meet the criteria for categorical exclusion as provided for under 10 CFR 51.22(c)(9). This conclusion has been determined because the changes requested do not pose significant hazards censiderations or do not invclve a significant increase in the amt unts, and no significant changes in the types of any effluents that may be released off-site. Additi >nally, this request does not involve a significant increase in individual or cumulative occupational radiat'on exposure.
1
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Attachment E Boiling Water Reactor Licensing Methodology Summary, EMF-94-217 t
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SIEMENS EMF-94-217(NP)
Revision 1 Boiling Water Reactor Licensing Methodology Summary November 1995 g-Eq (h
t I(l Siemens Power Corporation Nuclear Division 406W6fl8V
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Sismens Power Corporation - Nucisar Division EMF-94-217(NP) l Revision 1 l
Issue Date:
11/13/95 l
Boiling Water Reactor i
Licensing Methodology Summary i
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D. E. Hershberger Product Licensing t
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NUCLEAR REGULATORY COMMISSION REPORT DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Siemens Power Corporation. It is being submitted by Siemens O
Power Corporation to the U.S. Nuclear Regulatory Commission as part of a technical contribution to facilitate safety analyses by licensees of the U.S. Nuclear Regulatory Commission which utilize Siemens Power Corporation-fabricated reload fuel or other technical services provided by Siemens Power Corporation for light water power reactors and it is true and correct to the best of Siemens Power Corporation's knowledge, information, and belief. The g
information contained herein may be used by the U.S. Nuclear Regulatory Commission in its review of this report, and under the terms of the respective agreements, by licensees or applicants before the U.S. Nuclear Regulatory Commission which are customers of Siemens Power Corporation in their demonstration of compliance with the U.S. Nuclear Regulatory Commission's regulations.
O Siemens Power Corpt ration's warranties and representations concerning the subject matter of thir. document are those set forth in the agreement between d
Siemens Power Corpustion and the customer to which this document is issued.
Accordingly, except as otherwise expressly provided in such agreement, neither Siemens Power Corporation nor any person acting on its behalf:
A.
Makes any warranty, or representation, express or implied, with respect to the accuracy, completeness, orusefulness of the information contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned g
rights, or B.
Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this document.
O O
EMF-94-217(NP) p Revision 1 Pageii TABLE OF CONTENTS h
Section Pagg
1.0 INTRODUCTION
1-1 1.1 Fuel Assembly Design Objectives 1-1
)
1.2 Reload Fuel Licensing Bases..............................
1-2 1.3 Fuel System Design Criteria 1-3 1
1.3.1 Fuel Rod 1-3 1.3.2 Fu el System.................................
1-3 1.3.3 Fuel Coolability...............................
1-4 3.
1.4 Thermal and Hydraulic Design Criteria 1-4 1.5 Nuclear Design Analysis.................................
1-5 1.6 Testing, inspection, and Surveillance........................
1-6 1.7 Fuel Assembly Reconstitution.............................
1-7 2.0 MECH ANICAL DESIGN AN ALYSES..............................
2-1
)
2.1 Fuel System D amage...................................
2-1 2.1.1 Stress.....................................
2-1 2.1.2 Strain 2-2 2.1.3 Strain Fatig u e................................
2-3 2.1.4 Fretting We ar................................
2-3
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2.1.5 Oxidation and Crud Buildup 2-4 2.1.6 Rod Bowing.................................
2-5 2.1.7 Axi a l G ro wt h.................................
2-5 l
2.1.8 Rod Internal Pressure 2-6 2.1.9 Assembly Liftoff..............................
2-6 2.1.10 Fuel Assembly Handling.........................
2-7
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2.1.11 Miscellaneous Component Criteria 2-7 2.1.11.1 Compression Spring Forces 2-7 2.1.11.2 Lower Tie Plate Seal Spring......................
2-8 2.2 Fue l R od Failure.......................................
2-8 2.2.1 Internal Hydriding.............................
2-8 3
2.2.2 Cladding Collapse.............................
2-9 2.2.3 Overheating of Cladding
.......................2-10 2.2.4 Overheating of Fuel Pellets
.....................2-10 2.2.5 Pellet / Cladding Interaction......................
2-11 2.2.6 Cladding Rupture
............................2-11 2.2.7 Fuel Rod Mechanical Fracturing..................2-12
)
2.2.8 Fuel Densification and Swelling
..................2-13 2.3 BWR Fuel Coolability..................................
2-13 l
2.3.1 Fragmentation of Ernbrittled Cladding..............
2-13 2.3.2 Violent Expulsion of Fuel.......................
2-14 2.3.3 Cladding Ballooning...........................
2-15
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2.3.4 Fuel Assembly Structural Damage from External Forces.
2-15
)
EMF-94-217(NP)
O-Rsvision 1 Page lii TABLE OF CONTENTS (Continued) 2 4;
l Section Paae 3.0 THERMAL AND HYDRAULIC DESIGN ANALYSES....................
3-1 3.1 Hydraulic Compatibility..................................
3-1 l
3.2 Thermal Margin Performance 3-2 8
3.3 Fuel Centerline Temperature..............................
3-5 3.4 RodBow............................................
3-5 3.5 B yp a s s Fl o w.........................................
3-5 1
4.0 NUCLEAR DESIGN AND ANALYSES 4-1 g
i l
4.1 Nuclear Design and Analysis Methodology....................
4-1 4.2 Nuclear Design Analysis.................................
4-1 l
l 4.2.1 Fuel Rod Power History.........................
4-1 4.2.2 Kinetics Parameters 4-2 l
4.2.3 S ta bilit y....................................
4-3 4.2.4 Core Reactivity Control.........................
4-4 9) 4.3 G e n e ric An a lys e s......................................
4-5 4
4.3.1 Control Rod Withdrawal Error, BWR/6...............
4-5 4.3.2 Loss of Feedwater Heating 4-6
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4.4 Criticality Saf ety Analysis................................
4-6 0
5.0 EVALUATION OF ANTICIPATED OPERATIONAL OCCURRENCES 5-1 5.1 Analysis of Plant Transients at Rated Conditions................
51 5.2 Analyses for Reduced Flow Operation.......................
5-2 5.3 Analysis for Reduced Power Operation.......................
5-3 5.4 ASME Overpressurization Analysis 5-3 5.5 Limiting Transient Events................................
5-4 5.6 Control Rod Withdrawal Error 5-6 i
i 5.7 Determination of The mal Margins..........................
5-7 l
5.8 Supporting Documentation...............................
5-8 i
l 6.0 ANALYSIS OF POSTULATED ACCIDENTS.........................
6-1 g;
6.1 Loss of Coolant Accident Analysis..........................
6-1 l
6.1.1 Break Location Spectrum........................
6-1 1
6.1.2 Break Size Spectrum...........................
6-2 6.1.3 Worst Single Failure 6-2 6.1.4 M APLH G R Analyse s...........................
6-3 6.1.5 Supporting Documentation.......................
6-3 9) 6.2 Control Rod Drop Accident Analysis.........................
6-4 6.3 Fuel lo ading Error.....................................
6-4 6.3.1 Mistoaded Fuel Bundle..........................
6-5 i
6.3.2 Misoriented Fuel Bundle.........................
6-5 6.4 Fuel Handling Accident During Refuelir.g 6-6 9
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EMF-94-217(NP)
)
Ravision 1 Page iv TABLE OF CONTENTS (Continued)
)
Section g
7.0-REFERENCES,...,,,..,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,
71
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APPENDIX A LISTING OF SIEMENS POWER CORPORATION NRC-APPROVED LICENSING TOPICAL REPORTS SUPPORTING BWR METHODOLOGY.
A-1
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EMF-94-217(NP)
Rsvision 1 Page v LIST OF TABLES Table Enga 2.1 Steady State Stress Design Limits..............................
2-17 2.2 Summary Description of Fuel System Design Criteria.................
2-18 3.1 Summary Description of Thermal and Hydraulic Design Criteria...........
3-7 4.1 Summary Description of Nuclear Design Criteria.....................
4-7 LIST OF FIGURES Fiaure Eggg 3.1 Calculational Flow Used by the THERMEX Thermal Limits Methodology 3-8
~.. ~ ~
EMF-94-217(NP) p Revision 1 l
Page 1-1 i
1.0 INTRODUCTION
)
The introduction of nuclear fuel fabricated by Siemens Power Corporation - Nuclear Division (SPC) into the core of a boiling water reactor (BWR) requires assurance that the reactor will continue to meet accepted safety criteria during normal operation and accident conditions.
Additionally, assurance must be provided that SPC fabricated fuel is compatible with the l
existing fuel in the reactor core. In providing this assurance, SPC performs analyses for normal operations, anticipated operational occurrences, and postulated accidents which l
D confirm or modify operating procedures, setpoints, and limits.
l l
l The results of theses analyses are provided to each customer in a report which follows a standard, NRC-approved, format.M3) The methodology used for these analyses is described in licensing topical reports issued by SPC and approved by the NRC. A complete bibliography of these reports is provided in Appendix A of this report. This report presents generic information covering the fuel design and analysis of BWRs for which SPC supplies reload fuel.
l This report also describes the application of the SPC BWR methodology for generic and plant D
specific analyses. In some cases, the design basis analyses for a plant, as documented in the FSAR, may differ from the approved SPC methodology. For these special cases, SPC will continue to support the original design basis as documented in the FSAR for that plant.
)
1.1 Fuel Assembiv Desian Objectives SPC builds fuel assemblies to specific design criteria in order to assure that:
D The fuel assembly shall not fail as a result of normal operation and anticipated operational occurrences. The fuel assembly dimensions shall be designed to remain within operational tolerances and the functional capabilities of the fuel shall be established to either meet or exceed those assumed in the safety analysis.
Fuel assembly damage shall never prevent control rod insertion when it is required.
The number of fuel rod failures shall be conservatively estimated for postulated accidents.
D Fuel coolability shall always be maintained.
O
EMF-94-217(NP) s U
Rsvision 1 Page 1-2 The mechanical design of fuel assemblies shall be compatible with co-resident fuel and the reactor core internals.
g Fuel assemblies shall be designed to withstand the loads from in-plant handling and shipping.
The first four objectives are those cited in the Standard Review Plan.m The last two O
objectives assure structuralintegrity of the fuel and the compatibility of the fuel with existing reload fuel.
9 1.2 Reload Fuel Licensina Bases The SPC design criteria are approved by the NRCM8) and are consistent with Chapter 4 of the Standard Review Plan. These criteria are chosen to provide assurance that all SPC BWR fuel g:1 designs will perform satisfactorily throughout their design lifetimes. Compliance with the design criteria is demonstrated by:
Documenting the fuel system description and fuel assembly design drawings.
O Performing analyses with NRC-approved models and methods.
Testing significant new design features with prototype testing a'nd/or lead test assemblies prior to full reload implementation.
g:
Continued irradiation surveillance programs including post irradiation examinations to confirm fuel assembly performance.
Using the NRC-approved QA procedures, QC inspection program, and design control requirements identified in the SPC Quality Assurance Manual.A 9:
SPC BWR fuel designs which are demonstrated to meet the NRC-approved design criteria documented in Reference 48 do not need to be submitted to the NRC for explicit review and approval.
Demonstration that a BWR fuel design meets the NRC-approved criteria is equivalent to formal NRC approval of the design.
O.
O
EMF-94-217(NP) h Ravision 1 Page 1-3 As required for future designs, the design criteria presented in this report and Reference 48
'O will be evaluated and updated, as necessary. Any changes to the criteria will be submitted to the NRC for review and acceptance.
1.3 Fuel System Desion Criteria O
Chapter 4.2 of the Standard Review Plan contains criteria specified to provide assurance that the fuel system is not damaged as a result of normal operations or anticipated operational O
occurrences, that fuel system damage is never so severe as to prevent control rod insertion when it is required, that the number of fuel rod failures is not underestimated for postulated l
accidents, and that cootability is always maintained. These design criteria are necessary to meet the requirements of 10 CFR Part 50 (50.46: GDC 10, 27, and 35; Appendix K) and 9
10 CFR Part 100. The fuel system design enteria are summarized in Table 2.2 and discussed f
l in further detail in Section 2.0 as well as in the referenced topical reports.
l 1.3.1 Fuel Rod O
The detailed fuel rod design establishes such parameters as pellet diameter and density, cladding-pellet diametral gap, fission gas plenum size, and rod pre-pressurization level. The p
design also considers effects and physical properties of fuel rod components which vary with burnup. The integrity of the fuel rods is ensured by designing to prevent excessive fuel temperatures, excessive rod internal gas pressures, and excessive cladding stresses and strains. This end is achieved by designing the fuel rods to satisfy the design criteria during O
normal operation and anticipated operational occurrences over the fuel lifetime. For each l
design criterion, the performance of the most limiting fuel rod shall not exceed the specified limits.
O 1.3.2 Fuel System j
Fuel system criteria are established to assure that fuel system dimensions remain within
'O perational tolerances and that functional capabilities of the fuel assembly (system) are not l
i O
l
' EMF-94-217(NP)
O Revision 1 Page 1-4 reduced below those assumed in the safety analysis. The criteria apply for normal operation and for anticipated operational occurrences. The SPC criteria for the fuel system include O!
those topics identified in the Standard Review Plan, as discussed in Section 2.0 and summarized in Table 2.2.
9 1.3.3 Fuel Coolability To meet the requirements of General Design Criteria 27 and 35 as they relate to control rod insertability and core coolability for postulated accidents, fuel coolability criteria are g
established for all severe damage mechanisms. Coolability, or coolable geometry, has traditionally implied that the fuel assembly retains its rod-bundle geometry with adequate coolant channels to permit removal of residual heat. Reduction of coolability can result from cladding embrittlement, violent expulsion of fuel, generalized cladding melting, gross structural O'
deformation, and extreme fuel rod ballooning.
1.4 Thermal and Hydraulic Desian Criteria O
Fuel designs are evaluated relative to the thermal and hydraulic design criteria to determine and provide thermal operating limits with acceptable margins of safety during normal reactor operation and anticipated operational occurrences. To the extent possible, these analyses are g
performed on a generic fuel design basis. Because of reactor and operating differences, however, many of the analyses supporting these thermal and hydraulic operating limits are performed on plant and cycle specific bases.
O SPC uses NRC-approved methods and models(3A,0 in the thermal and hydraulic design and analysis of new fuel designs and new fuel design features. In the event the proposed design features are determined to be outside the range of the methods and models, applicable documentation will be submitted to the NRC for review and approval.
O O
O EMF-94-217(NP)
Rsvision 1 Page 1-5 1.5 Nuclear Desion Analvsis O
The nuclear design analyses are subdivided into two parts: a nuclear fuel assembly design analysis and a core design analysis. The fuel bundle nuclear design analysis is assembly specific and changes o ily as features affecting the nuclear characteristics of the fuel change, i.e., rod enrichments, burnable absorber content, etc. The core nuclear design analysis is specific to the core configuration and changes on a cycle basis. Nuclear fuel and core analyses are performed using NRC-approved methodologyW to assure that the new fuel O
assembly and/or design features meet the nuclear design criteria established for the fuel and core.
The fuel bundle nuclear design characteristics are considered for each SPC fuel bundle design 0
added to the core. The key characteristics affecting the nuclear design analysis include the following items:
Assembly average enrichment; Radial and axial enrichment distribution; Burnable absorber content and distribution; and Nature and location of non fueled rods or water channels.
O These key characteristics establish the fuel (local) and core power distributions and the kinetic parameters which are used in the thermal hydraulic, mechanical, and nuclear safety O
evaluations. The key neutronic design characteristics are selected such that fuel design limits are not exceeded during either normal operation or anticipated operational occurrences, and that the effects of postulated reactivity accidents will not cause significant damage to the reactor coolant pressure boundary or impair the capability to cool the core. These fuel.
\\O assembly characteristics are evaluated on a reload cycle specific basis during the neutronic and thermal hydraulic safety analyses.
O O
1 EMF-94-217(NP) n Rsvision 1 v
Page 1-6 l
The core nuclear characteristics are evaluated during the core design analysis. These analyses include evaluation of the power distributions, kinetic parameters, control rod worths, etc.
O~
These characteristics are calculated for the reference core loading configuration for each operating cycle, as discussed further in Section 4.2.
1.6 Testina, insoection, and Surveillance The SPC testing and inspection requirements are essential elements in assuring conformance to the design criteria. The component parameters either directly demonstrate compliance with g
the design criteria or are input for the design calculations. Therefore, the components must be as specified.
The SPC Quality Control program provides assurance that the components satisfy the product specifications. The SPC Quality Assurance Manual (2) controls and maintains this program.
The NRC has reviewed and accepted this manual as being in compliance with Appendix B of 1 0 C F R 5 0.I7I O
The specific QC inspections performed by SPC include component parts, pellets, rods, and assemblies, as well as process controlinspections. In addition, SPC reviews and overchecks inspections performed by vendors. These SPC and vendor inspections provide verification that g!
the manufactured fuelis consistent with the fuel design.
New SPC fuel designs incorporate proven design features in combination with new design features which improve some fuel performance characteristics. SPC introduces new fuel 9;
designs through Lead Fuel Assembly programs which include surveillance of the in-reactor performance of the new design features.
The particulars of a Lead Fuel Assembly surveillance program depend on the specifics of the i
new fuel design features and are developed on a case specific basis in cooperation with the l
utility in whose reactor the Lead Fuel Assemblies are irradiated.
O O
O EMF-94-217(NP)
Revision 1 Page 1-7 SPC can perform detailed visual inspection of fuel assemblies and make poolside h
measurements of the following parameters depending on the design changes being introduced:
1 Rod Diameter Profilometry (Cladding Creepdown, Ovality, and Ridging)
Oxide Thickness (Rods & Spacers) l
'O Cladding Wall Thinning Rod Length Fission Gas Release (Nondestructive and Destructive)
Pellet Column Slump Pellet Column Gaps Axial Power Profile
.O Isotope Redistribution Cladding Defects l
l Rod Bow and Rod-to-Rod Spacing Channel Closure l
Crud Sampling (Chemical Composition)
'O Spacer Spring Relaxation Assembly Length Surveillance programs of the irradiated fuel provide confirmation of the design adequacy. SPC
'O has performed extensive poolside examinations of irradiated fuel designs. These suryeillance programs have confirmed the good performance of the SPC fuel. Post irradiation surveillance programs will continue to be an important part in assuring and confirming the adequacy of current and future SPC fuel designs.
O 1.7 Fuel Assembly Reconstitution The application of SPC-approved design methodology to justify reinsertion into a reactor core irradiated fuel assemblies which have been reconstituted with replacement rods is described in Reference 49. Replacement rods can be fuel rods containing natural uranium pellets, water rods, and inert rods containing Zircaloy or stainless steel inserts.
O Removal and replacement of suspected or known leaking fuel rods from irradiated assemblies has obvious advantages. Radiation levels in the plant are reduced with the removal of leaking fuel rods. Valuable technicalinformation regarding fuel performance may be gathered through O
inspections. Where use of reconstituted fuel assemblies shows no significant impact on l
o
EMF-94-217(NP) g Rsvision 1 Page 1-8 assembly and core performance, core loadings previously analyzed may be preserved and uranium that would otherwise be lost to energy production may be utilized.
O A variety of replacement rod types may be used to reconstitute irradiated fuel assemblies depending on customer preference and application. In addition to fuel rods containing natural uranium, SPC uses either Zircaloy or stainless steel filled inert rods as well as water rods.
Replacement fuel rods containing natural uranium pellets have the advantage of more closely matching neutronic and mechanical characteristics of the remaining fuel rods in an irradiated e
fuel assembly.
However, using fuel rods adds cost and complexity in manufacturing, transportation, and material accountability relative to using non-fuel replacement rods. In addition, a fuel replacement rod is not used in cases where there is a potential for failure of the replacement rod in subsequent operating cycles (e.g., spacer damage, etc.).
Zircaloy or stainless steel filled rods and water rods are used depending on customer preference. These types of replacement rods allow fast response to customer requirements e
since they can be easily manufactured and transported. Neutronic, thermal-hydraulic, and mechanical aspects of the reconstituted fuel assembly are reviewed for each replacement rod application. The review, performed with NRC-approved methodology, assures that design limits are applicable.
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EMF-94-217(NP) l Rsvision 1 l
Page 2-1 l
2.0 MECHANICAL DESIGN ANALYSES k
1 2.1 Fuel System Damaae
)
The design criteria for Fuel System Damage should not be exceeded during normal operations, l
including Anticipated Operational Occurrences (AOOs).
l 2.1.1 Stre_gs
)
Desian Criteria.
The design criteria for evaluating the structural integrity of the fuel assemblies follow:
Fuel assembiv handlina - The assembly must withstand dynamic axial loads approximately 2.5 times assembly weight.
For all aoolied loads for normal operation and anticioated operational events - The fuel l
assembly component structural design criteria are established for the two primary
)
material categories, austenitic stainless steels (tie plates) and Zircaloy (tie rods, grids, l
spacer capture rod tubes). The stress categories and strength theory for austenitic stainless steel presented in the ASME Boiler and Pressure Vessel Code, Section ill, are used as a general guide.
Steady state stress design limits are given in Table 2.1. Stress nomenclature is per the
)
ASME Boiler and Pressure Vessel Code, Section 111.
Loads durina costulated accidents - Deflection or failure of components shall not interfere with reactor shutdown or emergency cooling of the fuel rods.
!)
Bases. In keeping with the GDC 10 SAFDLs, the fuel damage design criteria for cladding i
stress assure that fuel system dimensions remain within operational tolerances and that func-l tional capabilities are not reduced below those assumed in the safety analysis. Conservative j
stress limits are derived from the ASME Boiler Code, Section Ill, Article Ill-2000;W and the specified 0.2% offset yield strength and ultimate strength for Zircaloy.
l l
The structuralintegrity of the fuel assemblies is assured by setting design limits on stresses and deformations due to various handling, operational, and accident loads. These limits are
!)
-.=--..-.
I i
EMF-94-217(NP) 0:
Revision 1 Page 2-2 applied to the design and evaluation of upper and lower tie plates, grid spacers, tie rods, O
spacer capture rod, water rods, water channels, fuel assembly cage, and springs where applicable. The allowable component stress limits are based on Appendix F of tne ASME Boiler and Pressure Vessel Code, Section 111, with some criteria derived from component tests.
Cladding stress categories include the primary membrane and bending stresses, and the secondary stresses. The loadings considered are fluid pressure, internal gas pressure, thermal gradients, restrained mechanical bow, flew induced vibration, and spacer contact. Table 2.1 gives the ASME stress level criteria.
(
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The stress calculations use conventional, elasticity theory equations. A general purpose, finite I9 element stress analysis code such as ANSYS )is used to calculate the spacer spring contact stresses. The fuel assembly structural component stresses under faulted conditions are O
evaluated using primarily the criteria outlined in Appendix F of the ASME Boiler and Pressure Vessel Code, Section Ill.
t l
l The SPC analysis methods for calculating fuel rod cladding and assembly steady-state stresses i
01 are discussed and approved in References 10 and 11.
l 2.1.2 Strain l
9
_r3slan Criteria. The SPC design criteria for fuel rod cladding strain is that the transient-induced deformations must be less than 1 % uniform cladding strain.
l l
Bases. The design criteria for cladding strain is intended to preclude excessive cladding 8 !
deformation and failure from normal operations and AOOs. SPC uses the NRC-approved RODEX2A coden2) to calculate steady-state cladding strain during normal operation. The j
i RAMPEXU3) code is used by SPC to calculate cladding strain during transient operation.
Oi i
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b EMF-94-217(NP)
Revision 1 Page 2-3 2.1.3 Strain Fatiaue b
i Desian Criteria. The SPC design criteria for strain fatigue limits the cumulative fatigue usage l
factor to less than 1
b Bases. Cycle loading associated with relatively large changes in power can cause cumulative damage which may eventually tend to fatigue failure. Therefore, SPC requires that the cladding not exceed a cumulative fatigue usage factor The fatigue usage factor is b
the number of expected cycles divided by the number of allowed cycles. The total cladding usage factor is the sum of the individual usage factors for each duty cycle.
The SPC methodology for determining strain fatigue is based on the use of the RAMPEXM 3) h code and the O'Donnell and Langer fatigue design curves.04) The fatigue curves have been l
adjusted to incorporate the recommended safety factor of two on stress amplitude or 20 on number of cycles, whichever is more cranservative. The RODEX2 code is used to provide initial steady-state conditions for SPC transient and accident analysis. Consequently, the RODEX2 code provides input to the RAMPEX code for each power change, and RAMPEX provides stress amplitudes for the various power cycles, h
2.1.4 Frettina Wear Desian Criteria. The SPC design criteria fr t fretting wear requires that fuel rod failure due to s
fretting shall not occur.
'O Sases. SPC controls fretting wear by use of design features, such as a spacer spring dimple system, which assure that fuel rods are positively supported by the grid spacers throughout the expected irradiation period. Spacer grid spring systems are designed such that the
.g SPC performs fretting tests to verify consistent fretting performance for new spacer designs.
Examination of a large number of irradiated BWR rods has i
- O substantiated the absence of fretting in SPC designs.
1
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EMF-94-217(NP)
O Rsvision 1 Page 2-4 2.1.5 Oxidation and Crud Builduo O
Desion Criteria. There is no specific limit for oxide thickness or crud buildup. The effects of oxidation and crud buildup are included in the thermal and rod internal gas pressure analysis.
O Bases. The SPC fuel design basis for cladding corrosion and crud buildup is to prevent
- 1) significant degradation of the cladding strength, and 2) unacceptable temperature increases. Cladding corrosion reduces cladding wallthickness and results in less cladding load carrying capacity. At normal light water reactor operating conditions, this mechanism is not g.
limiting except under unusual conditions where Ngh cladding temperatures greatly accelerate the corrosion rate. Because of the thermal resistance of corrosion and crud layers, formation of these products on the cladding results in an elevation of temperature within the fuel as well as the cladding.
O!
There is no specific limit for crud buildup. However, SPC fuel performance codes 02,20 reviewed and approved by the NRC include the crud buildup in the fuel performance O
predictions. That is, the crud and oxidation models are a part of the approved models and therefore impact the temperature calculation.
SPC data show that even at higher exposures and residence times, cladding oxide thickness g
is relatively low. Mechanical properties of the cladding are not significantly affected by thin oxide or crud layers. For the thermal and rod internal gas pressure analyses, the effect of ox.'detion is included. For steady-state strain, transient strain and cyclic stress, the effect of wall thinoing is insignificant since cladding deformation is strain dependent. That is, the O
change in cladding diameter during a power change is primarily determined by the change in the pellet diameter since pellet-cladding contact occurs at higher exposures. For the cladding end-of-life stress analysis, the wall thickness is reduced consistent with the peak oxide thickness.
O O
EMF-94-217(NP) g R:; vision 1 Page 2-5 2.1.6 Rod Bowino S
Desian Criteria. The SPC design criteria for rod bowing is that S
Bases. Differential expansion between the fuel rods, and lateral thermal and flux gradients can lead to lateral creep bow of the rods in the spans between spacer grids. This lateral creep bow alters the pitch between rods and may affect the peaking and local heat transfer.
9 SPC has established a rod-to-rod clearance closure limit 0 13 below which a penalty is addressed on the minimum critical power ratio (MCPR) and above 9
which no reduction in MCPR is necessary.
SPC uses NRC approved models01) which are based on empirical data to calculate minimum EOL rod to rod spacing. The potential effect of this rod bow on thermal margin is negligible.
Rod bow at extended burnup does not affect thermal margins due to the lower powers achieved at high exposure.
p 2.1.7 Axial Growth Desian Criteria. SPC requires that the fuel assembly be compatible with the channel throughout the fuel assembly lifetime. In addition, SPC reqthes the fuel rods and other assembly components to maintain clearances and engagements i1the fuel assembly structure throughout the lifetime of the fuel.
Bases. SPC evaluates fuel channel-fuel assembly differential growth to assure that the fuel l
)
channel to lower tie plate engagement is maintained to the design burnup. Another concern for BWR fuel assemblies is to maintain engagement between the fuel rod end cap shank and the assembly tie plates, i.e., to prevent fuel rod disengagement from the tie plates. The 3
change in BWR fuel rod-to-tie plate engagement (and possible disengagement) is due to the D
EMF-94-217(NP)
OI Rsvision 1 Page 2-6 growth rate of the tie rods that connect the bottom and top tie plates being greater than the O
growth rate of the fuel rods.
The SPC analysis method for evaluating rod-to-tie plate engagement is based on a statistical upper bound of measured differential rud-to-tie plate growth from both 8x8 and 9x9 9.
designs.(11,15,25) e, O
2.1.8 Rod Internal Pressure Desian Criteria. SPC limits maximum fuel rod internal pressure relative to system pressure.
In addition, SPC requires that when fuel rod pressure exceeds system pressure, the pellet-clad gap has to remain closed if it is already closed or that it should not tend to open for steady state or increasing power oprations.
Bases. Rod internal pressure is limited to prevent unstable thermal behavior and to maintain g
the integrity of the cladding. Outward circumferential creep which may cause an increase in pellet-to-cladding gap must be prevented since it would lead to higher fuel temperature and higher fission gas release. The maximum internal pressure is also limited to protect against 8
embrittlement of the cladding caused by hydride reorientation during cooldown and depressurization conditions. A proprietary limit above system pressure has been justified by SPC in Reference 15.
el 2.1.9 Assembly Liftoff s
Desion Criteria. SPC requires that the assembly not levitate from hydraulic or accident loads.
l Ol O'
J EMF-94-217(NP)
Rsvision 1 Page 2-7 Bases. Levitation of a fuel assembly could result in the assembly becoming disengaged from 9
the fuel support and interfering with control rod movement. For normal operation, including AOOs, the submerged fuel assemHy weight, including the channel, must be greater than the hydraulic loads. The criteria is spp!icable to both cold and hot conditions and uses the Technical Specification limits on total core flow. For accident conditions, the normal hydraulic loads plus additional accident loads shall not cause the assembly to become disengaged from the fuel support. This assures that control blade insertion is not impaired.
G 2.1.10 Fuel Assembly Handlina Desian Criteria. The assembly design must withstand all normal axialloads from shipping and fuel handling operations without permanent deformation.
O Bases. SPC uses either a stress analysis or testing to demonstrate compliance. The analysis or test uses an axial load of 2.5 times the static fuel assembly weight. At this load, the fuel assembly structural components must not show any yielding.
g O
The rod plenum spring also has design criteria associated with handling requirements.
The component drawing specifies the fabricated cold spring force.
O 2.1.11 Miscellaneous Component Criteria 2.1.11.1 Comoression Sorina Forces J
Desian Criteria.
O
EMF-94-217(NP)
Ravision 1 O
Page 2-8 Bases. The compression springs must support the weight of the upper tie plate and channel throughout the design life of the fuel. Therefore, SPC has a requirement on the minimum O;
compression spring force.
4 2.1.11.2 Lower Tie Plate Seal Sorina 0:
Desian Criteria. The seal accommodates the channel deformation to limit the leak rate of coolant between the lower tie plate and channel wall.
O Bases. The lower tie plate seal spring limits the leak rate of coolant between the lower tie i
plate and the channel wall. The seal shall have adequate corrosion resistance and be able to withstand the operating stresses without yielding. The design also considers the differernial axial growth between the channel and the bundle. Flow testing of prototypic components Ol verifies the leakage rate and fretting resistance. A stress analysis provides the seal stresses.
2.2 Fuel Rod Failure O!
The fuel' rod failure criteria and bases cover normal operation conditions, including AOOs, and postulated accidents. When the fuel rod failure criteria are applied in normal operation including AOOs, they are used as limits (Specified Acceptable Fuel Design Limits) since fuel failure under those conditions should not occur according to GDC 10.(16) When the criteria are used for postulated accidents, fuel f ailures are permitted, but they must be accounted for in the dose calculations required by 10 CFR 100.(17) 9 2.2.1 Internal Hvdridina Desian Criteria. SPC limits internal hydriding by imposing a fabrication limit for total hydrogen O'
in the fuel pellets of less than 2.0 ppm.
Bases. The absorption of hydrogen by the cladding can result in cladding failure due to reduced ductility and formation of hydride platelets.
Hydriding, as a cladding failure O
9
7
- O EMF-94-217(NP) 1 Rsvision 1 Page 2-9 mechanism, is precluded by controlling the level of moisture anc, er hydrogenous impurities O
during fuel pellet f abrication. SPC assures that the hydrogen evncentration criteria is met through careful moisture control during fuel fabrication.(15) 2.2.2 Claddino Collaose
-Q Desian Criteria. Creep collapse of the cladding is avoided in the SP'. fuel system design by
.O Bases, if axial gaps in the fuel pellet column were to occur due to handling, shipping, or fub!
densification, the cladding would have the potential of collapsing into the gap. Because of the large local strains that would result from the collcase, the cladding is assumed to fail.
O Creep collapse of the cladding and the subsequent pote..tial for fus) failure'is avoided in the SPC fuel system design by O
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)
O i
SPC uses the NRC-approved RODEX2A(12) and COLAPXIl8I
. O codes to predict creep collapse. The RODEX2A code is used to provide initial in-reactor fuel rod conditions to COLAPX, e.g., radial fuel-cladding ge =ize, fill gas pressure, and cladding temperatures. The COLAPX code calculates oval'tv avs and creep deformation of the g
cladding as a function of time.
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2.2.3 Overheatino of Claddina O
Deggn Criteria. The SPC design basis to preclude fuel rod cladding overheating is that 99.9%
of the fuel rods shall not experience transition boiling.
O Baser it has been traditional practice to assume that fuel failures will occur if the them.at margin criteria is violated. Thermal margin is stated in terms of the minirnum value of the critical power ratio (CPR) for the most limiting ',uel assembly in the cort. Prevention of potential fuel failure from overheating of the cladding is accomplished by mir:imizing the gi probability of exceeding thermal margin limits on limiting fuel rods during normal operation and anticipated operational occurrences. SPC confirms compliance with this criteria as part of the reload thermal hydraulics anyiv,% as discussed in Section 3.2 of this report. Experimentally based critical heat flux correlations which have be3n accepted by the NRC are used (see 8
Section 3.2).
2.2.4 Overheatino of Fuel Pellqis 0,
l Desian Criteria. Fuel f ailure from overheating of the fuel pellets is not allowed. The centerline temparature of the fuel pellets must remain below melting during normal operations and AOOc.
4 Bases. SPC establishes steady state and transient design LHGR limits for each fuel system which protect against centerline melting. Operation within these LHGR limits prevents centerline melting during normal operation and anticipated operational occurrences throughout 0;
the design hfetime of the fuel.
SPC utilizes a correlation for the fuel melting point that accounts for the effect of burnup and W
gadolinia content. This fuel melting limit has been reviewed and approved by the NRC l
with respect to application to fuel and gadolinia bearing fuel at extended burnup levels because the limit conservatively accounts for the decrease in fuel melting point with l
increasing burnup.
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.O EMF-94-217(NP)
Ravision 1 Page 2-11 SPC uses the RODEX2A computer coden2) to calculate maximum possible fuel centerline O
temperature for normal operations. Conservative LHGR power histories are used to perform the centerline temperature calculations. For AOOs, SPC uses the RODEX2A code to calculate maximum possible fuel centerline temperatures with an LHGR history which is higher than the i
steady-state LHGR history used for normal operations by a conservative factor.
O 2.2.5 fellet/Claddina Interaction O
Desian Cr teria. The Standard Review PlanUI does not contain an explicit criteria for pellet / cladding interaction. However, it does present two related criteria. The first one is that transient-induced deformations must be less than 1% uniform cladding strain. And, the second one is that fuel melting cannot occur. SPC requires that: 1) transient-induced O
deformations must be less than 1% uniform cladding strain, and 2) fuel centerline melting i
cannot occur.
D_a_ sis. The cladding strain requirement is addressed in Section 2.1.2 of this document. The g
centerline temperature requirement is addressed in Section 2.2.4 of this document.
)
2.2.6 Claddina Ruoture O
Desian Criteria. 10 CFR 50 Appendix KU 9I requires that cladding rupture must not be underestimated when analyzing a lor.a of coolant accident.
O Bases. Zircaloy cladding will burst (rupture) under certain combinations of temperature, heating rate, and differential pressure conditions during a LOCA. While there are no specific 1
UI design criteria in the Standard Review Plan associated with cladding rupture, the requirements of Appendix K to 10 CFR 50 must be met as those requirements relate to the incidence of rupture during a LOCA: therefore, a rupture temperature correlation must be used during the LOCA emergency core cooling system (ECCS) analysis.
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EMF-94-217(NP)
O Revision 1 Page 2-12 SPC includes the effects of cladding rupture as an integral part of the ECCS evaluation model.
SPC uses the cladding ballooning and rupture models presented in NUREG-0630(20) for Oi cladding rupture evaluation.
These models are described in XN-N F-82-07(P)( A),
Revision 1,(21) which has been reviewed by the NRC and found acceptable for use in LOCA analyses. The link between the cladding deformation and rupture models, and the LOCA 9.
ECCS analysis is described in Reference 22.
1 2.2.7 Fuel Rod Mechanical Fracturina l
9 Desian Criteria. SPC limits the combined stresses from postulated accidents to the stresses given in the ASME Code, Section Ill, Appendix F(8) for faulted conditions, O
f,gs. A mechanical fracture refers to a defect in a fuel rod caused by an externally applied force, such as a hydraulic load cr a load derived from core plate motion induced by seismic /LOCA events. The design bases and criteria for mechanical fracturing of SPC BWR l
reload fuel are presented in Reference 23, which describes SPC's LOCA-seismic structural G:
response analysis. LOCA-seismic structural response analysis covering SPC's 9x9 fuel designs is presented in Reference 24. The design basis is that the channeled fuel assemblies must withstand the external loads due to earthquake and postulated pipe breaks without fracturing the fue; rod cladding. The stresses, due to postulated accidents in combination gi with normal steady-state fuel rod stresses, should not exceed the stress limits given in References 23 and 24. The allowable stresses are derived from the ASME Boiler and Pressure Vessel Code, Section ill, Appendix F, for fau' tad conditions. The design limits have been S'
reviewed and approved by the NRC.
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2.2.8 Fuel Densification and Swellina O
Desian Criteria. Fuel densification and swelling are limited by the design criteria specified for fuel temperature, cladding strain, cladding collapse, and internal pressure criteria.
I I
O Bases. SPC uses the NRC reviewed and r.ccepted densification and swelling models in the fuel performance codes.02,25) 2.3 BWR Fuel Coolability O
For accidents in which severe fuel damage might occur, core cootability and the capability to l
insert control blades are essential. Normal operation or anticipated operational occurrences must remain within the thermal margin criteria. Chapter 4.2 of the Standard' Review PlanUI provides several specific areas important to the coolability and the capability of control blade insertion. The sections below discuss these areas.
O 2.3.1 Fraomentation of Embrittled Claddino Desian Criteria. SPC ECCS evaluations meet the 10 CFR 50.46(26) limits of 2200 *F peak cladding temperature, local and core-wide oxidation, and long term coolability.
O Bases.
The requirements on cladding embrittlernent relate to the LOCA requirements of 10 CFR 50.46. The principal cause of cladding embrittlement during severe accidents such as LOCA is the high cladding temperatures that result in severe cladding oxidation.
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Ravision 1 l
Page 2-14 i
l The models to compute the temperatures and oxidation are those prescribed by Appendix K l
of 10 CFR 50.08) The SPC methodology for evaluating cladding oxidation and embrittlement O
l during a LOCA is incluosd in the approved topical reports for LOCA-ECCS analysis.(22,2n The LOCA analysis is performed on a plant specific basis.
9 2.3.2 Violent Exouision of Fuel Desion Criteria. SPC limits the radially-averaged enthalpy deposition at the hottest axial location to 280 cal /gm for severe reactivity initiated accidents.
g Bases, in a severe reactivity initiated accident (RIA), large and rapid deposition of energy in the fuel could result in melting, fragmentation, and dispersal of the fuel. SPC has adopted the guideline in Regulatory Guide 1.77(28) and the Standard Review PlanUI that restricts the O
radially-averaged energy deposition. SPC uses the 280 cal /gm radially-averaged enthalpy deposition at the hottest axial location as the design criteria.
I 4
l The limiting RlA for SPC fuel in a boiling water reactor is the control rod drop accident. SPC calculates the maximum radially averaged enthalpy for the CRDA for each reload core in order to assure that tne maximum calculated enthalpy is below the 280 cal /gm limit. The SPC control rod drop calculation methodology approved by the NRC is described in Reference 29.
g The parameterized SPC control rod drop methodology determines maximum deposited enthalpy as a function of dropped rod worth, effective delayed neutron fraction, Doppler j
coefficient, and four-bundle local peaking factor.
The CRDA analysis is not part of the normal fuel assembly mechanical analysis but is part of
]
i the cycle specific safety analysis performed for each BWR.
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Page 2-15 i
2.3.3 Claddina Balloonina l
Desion Criteria. There are no specific design limits associated with cladding ballooning, other l
than a 10 CFR 50 Appendix KU 8I requirement that the degree of swelling not be underestimated.
Bases. Zircaloy cladding will balloon (swell) under certain combinations of temperature, heat rate, and stress during a LOCA. Ciadding ballooning can result in flow blockage; therefore,
)
the LOCA analysis must consider the cladding ballooning impacts on the flow, i
l The SPC cladding ballooning model is an integral part of the cladding rupture temperature l
model for the LOCA ECCS analysis. The cladding rupture temperature modelis addressed in f
Reference 21. SPC, in Reference 21, has adopted the NUREG-0630(20) data base and modeling, which specifies a method acceptable to the NRC for treating cladding swelling and rupture during a LOCA. These models have been approved by the NRC for extended burnup levels.0 5)
The RODEX2 fuel performance code (25) is used to provide burnup dependent input to the l
LOCA analysis, e.g., stored energy and rod pressures, that are a function of the initial steady-state operation of the fuel. This initial steady-state fuel condition is also important to cladding ballooning.
i 2.3.4 Fuel Assembiv Structural Damaae from External Forces l
Desion Criteria. The SPC design criteria for fuel assembly structural damage from external forces are covered in Sections 2.1.1, Stress, 2.1.9, Assembly Liftoff, and 2.2.7, Fuel Rod Mechanical Fracture.
Bases. Earthquakes and postulated pipe breaks in the reactor coolant system would result in UI external f arces on the fuel assembly. The Standard Review Plan states that fuel system
)
coolability should be maintained and thst damage should not be so severe as to prevent i
)
EMF-94-217(NP)
Ol Revision 1 Page 2-16 control blade insertion when required during these accidents. The SPC design basis is that the fuel assembly will maintain a geometry that is capable of being cooled under the worst case accident Condition IV event and that system damage is never so severe as to prevent control blade insertions. SPC assures these design bases are met by placing ASME design limits on the stresses that critical fuel assembly components can experience. These limits g;
have been approved for SPC 8x8 and 9x9 fuel assemblies in References 23 and 24, respectively.
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j EMF-94-217(NP) l Rsvision 1 Page 2-17 Table 2.1 Steady State Stress Design Limits
- Stress intensity Limits *
- Ultimate Yield Tensile
)
Strength Strength (a )
(a )
l y
u J
General Primary Membrane Stress 2/3 a 1/3 a y
u Primary Membrane Plus Primary Bending Stress 1.0 a 1/2 a y
u Primary Plus Secondary Stress 2.0 a 1.0 o y
u l
l Characteristics of the stress categories are defined as follows:
I l
a)
Primary stress is a stress developed by the imposed loading which is necessary to satisfy the laws of equilibrium between external and internal forces and l
moments. The basic characteristic of a primary stress is that it is not self-limiting. If a primary stress exceeds the yield strength of the material through the entire thickness, the prevention of failure is entire ly dependent on the strain-hardening properties of the material.
2 b)
Secondary stress is a stress developed by the self-constraint of a structure. It l
must satisfy an imposed strain pattern rather than being in equilibrium with an j
external load. The basic characteristic of a secondary stress is that it is self-i limiting. Local yielding and minor distortions can satisfy the discontinuity I
conditions due to thermal expansions which cause the stress to occur.
The stress intensity is defined as twice the maximum shear stress and is equal to the largest algebraic difference between any two of the three principal stresses.
l
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EMF-94-217(NP) g Rsvision 1 Page 2-18 Table 2.2 9'
Summary Description of Fuel System Design Criteria Section Descriotion Generic Desion Criteria 9
2.1 Fuel System Criteria 2.1.1 Stress, Strain, or Loading Table 2.1 Steady State Limits on Assembly Components (See 2.3.4 below for accident) 9:
2.1.2 Cladding Strain 1% strain 2.1.3 Fatigue Cumulative usage factor 2.1.4 Fretting Wear No significant fretting wear 8
2.1.5 Oxidation, Hydriding and Acceptable metal loss due to oxidation Crud Buildup 2.1.6 Rod Bow Protect Thermal Limits g
2.1.7 Axial Irradiation Growth Seal Spring compatible with channel Maintain end cap engagements in UTP 2.1.8 Rod internal Pressure Radial gap does not open, no hydride
- i reorientation 2.1.9 Assembly Liftoff Maintain assembly engagement in core support piece n.nd maintain positive holddown force g:
2.1.10 Fuel Assembly Handling Assembly withstand 2.5 times static weight as axial load 2.1.11 Miscellaneous Components 9:
2.1.11.1 Compression Spring Force 2.1.11.2 Lower Tie Plate Seal Spring Accommodate channel deformation adequate corrosion resistance withstand operating stresses g;
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EMF-94-217(NP)
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Ravision 1 Page 2-19 Table 2.2 Summary Description of Fuel System Design Criteria (Continued)
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Section Description Generic Desian Criteria l
2.2 Fuel Rod Criteria 2.2.1 Internal Hydriding
< 2 ppm H 2
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2.2.2 Cladding Collapse l
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2.2.3 Overheating of Cladding 99.9% of rods not to exceed CHF 2.2.4 Overheating of Fuel Pellets No centerline melting 2.2.5 Pellet / Cladding Interaction 1% strain Ne fuel melting 2.2.6 Cladding Rupture Not underestimated during LOCA and used in determination of 10 CFR 50.46 l
criteria (2m 2.2.7 Mechanical Fracturing ASME Section 111, Appendix F(8)
Limits 2.2.8 Densification and Swelling included in 2.1.8, 2.2.4, and 2.2.5 l
2.3 Fuel Coolability 2.3.1 Cladding Embrittlement included in LOCA Analysis 2.3.2 Violent Expulsion of Fuel
< 280 cal / gram 2.3.3 Fuel Bellooning Consider impact on flow blockage in LOCA analysis 2.3.4 Structural Deformations Cootable geometry, control rod insertability
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EMF-94-217(NP)
Rsvision 1 Page 3-1 3.0 THERMAL AND HYDRAULIC DESIGN ANALYSES
)
Thermal-hydraulic analyses of the fuel and core are performed to verify that design criteria are satisfied and to establish an appropriate value for the MCPR fuel cladding integrity safety limit.
3.1 Hydraulic Comoatibility 1
Desion Criteria.
The hydraulic flow resistance of the reload fuel assemblies shall be sufficiently similar to existing fuel in the reactor that there is no significant unplanned impact on total core flow or the flow distribution among assemblies in the core, i
Bases. The Standard Review PlanW does not contain an explicit criterion for fuel assembly
)
hydraulic compatibility. However, flow differences between assembly types in a mixed core i
need to be accounted for in assuring that all design criteria are satisfied.
The component hydraulic resistances in the reactor core are determined by a combination of
)
both analytical techniques and experimental data.
l D
D The SPC thermal-hydraulic methodology implicitly includes the impact of assembly differences on the individual assembly Wow. The overall criterion for acceptability is that individual fuel I
types must be in compliance..ith the thermal hydraulic limits.
)
The purpose of these evaluations is to better define the
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EMF-94-217(NP) 0:
Rsvision 1 Page 3-2 core stability behavior with this mismatch in flow. The MCPR performance remains protected by the compliance with the safety and operating limits.
O 3.2 Thermal Maroin Performance Desion Criteria. The fuel design shall fall within the limits of applicability of the approved critical heat flux (CHF) correlation. New fuel assembly designs and/or changes in existing assembly designs shall minimize the likelihood of boiling transition during normal reactor operation and AOOs. The applicable critical power correlation will be used to determine the g;
operating limits and for consistency be used to monitor the core.
Bases. SPC fuel and reload cores are designed such that operation consistent with Technical Specification limits will result in <0.1 % of the rods experiencing boiling transition during O
AOOs. An NRC-approved CHF correlation is used by SPC to determine operating and safety limits during the design of a reload core, and, for consistency, the same CHF correlation is used to monitor the core during operation.
Gl Operation of a BWR requires protection against fuel damage during normal reactor operation and AOOs. A rapid decrease in heat removal capacity associated with boiling transition can potentially result in high transient temperatures in the cladding, which may cause cladding degradation and a loss of the fuel rod integrity. Protection of the fuel against boiling transition assures that such degradation is avoided. This protection is accomplished by determining the operating limit minimum critical power ratio (OLMCPR) for each fuel bundle in the reactor core for each cycle.
9 The SPC approach to thermal limits analysis, THERMEX, is described in Reference 3. This THERMEX thermallimits methodology consists of a series of related analyses which establish S
an operating limit minimum critical power ratio (OLMCPR). The OLMCPR is determined from two calculated values, the safety limit MCPR (SLMCPR) and the limiting transient ACPR. The overall methodology is comprised of four major segments, which are 1) reactor core hydraulic i
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O EMF-94-217(NP)
Revision 1 Page 3 3 methodology, 2) an NRC-approved critical power correlation, 3) plant transient simulation O
methodology, and 4) critical power methodology.
SPC fuel assembly pressure drop methodology is presented in Reference 31.
The methodology addresses part of the calculational method used by SPC to determine assembly pressure drop that is used to calculate assembly flows for a BWR core. The pres.sure drop.
methodology report identifies the constitutive relationships to determine void fraction and two-phase pressure losses which are in turn used as input to the calculation of the assembly O
pressure drop by the XCOBRA computer code.(3)
The calculation of the fuel assembly critical power performance is established by means of an empirical correlation based on results of boiling transition test programs. SPC currently O
employs two NRC-approved critical power correlations. The ANFB Critical Power Correlation is described in Reference 32. The older XN-3 Critical Power Correlation is described in References 33 and 34.
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The methodology and computer codes for SPC BWR plant transient analyses are described in detailin References 4 and 5. This plant transient methodology is supplemented by the NRC-approved XCOBRA-T code (37,38) and the COTRANSA2 code.(39) The COTRANSA2 code is used to calculated BWR system behavior for steady-state and transient conditions. This O
O
EMF-94-217(NP) g Rsvision 1 Page 3-4 behavior is then used to provide input to the XCOBRA-T and XCOBRA codes, from which critical power ratios are determined for limiting transients.
O in the generation of the limiting transient ACPR in the THERMEX methodology, consideration may be given to the statistical convolution of uncertainties associated with the calculation of the thermal margin. The statistical uncertainty analysis methodology, GSUAM, is described on a generic basis in Reference 40. Each plant specific GSUAM application must contain the variances and distributions of the predictor variables used in the response surface fitting and sufficient data must be presented to identify the mean and statistical variation of each 3
predictor variable. All plant parameters not treated statistically and any predictor variable which is eliminated from the response surface fitting most be set at their limiting value.
The critical power ratio methodology is the approach used by SPC to determine the margin-to-O thermal limits for boiling water rea.: tors. The SPC critical power methodology for boiling water reactors is presented in References 41 and 42. The Reference 41 topical report addresses the critical power methodology associated with the XN-3 CPR correlation while the M
Reference 42 topical addresses the more recent ANFB correlation.
The ANFB-based topicalreportM2) provides the basis for the SPC methodology for determining the operating safety limit for minimum critical power (SLMCPR) which ensures that 99.9%
g) of the fuel rods are protected from boiling transition. This determination is carried out by a series of Monte Carlo calculations in which the variables affecting the probability of boiling transition are varied randomly and the total number of rods experiencing boiling transition is determined for each Monte Carlo trial.
O!
The methodology and component uncertainties needed 9-to calculate power distribution uncertainty are given in Reference 6 and the assembly flow calculational uncertainty is given in the Reference 42 topical report. The SPC methodology for analysis of assembly channel bow effects is also given in Reference 42. Figure 3.1 shows the relationship between the SPC transient and critical power methodologies.
g.
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O EMF-94-217(NP)
Revision 1 Page 3-5 The THERMEX analytical process requires the input of certain external data, a portion of which O
may be plant-dependent. The specific input sources are described Reference 3.
3.3 Fuel Centerline Temoerature O
Desian Criteria. Fuel design and operation shall be such that fuel centerline melting is not projected for normal operation and AOOs.
lO Bases. This design criteria is addressed during the fuel type specific mechanical design analysis. The bases is discussed in Section 2.2.4 of this document.
(
3.4 Rod Bow l
O Desion Criteria. The anticipated magnitude of fuel rod bowing under irradiation shall be accounted for in establishing thermal margins requirements.
O Bases. The bases for rod bow are discussed in Section 2.1.6 of this document. Rod bow magnitude is determined during the fuel type specific mechanical design analyses. Tho need for a thermal margin rod bow penalty is evaluated on a plant and cycle specific basis. Post-O irradiation examinations of BWR fuel fabricated by SPC show that the magnitude of fuel rod bowing is small and the potential effect of this bow on thermal margins is negligible. Rod bow at extended burnups does not affect thermal margins due to the lower powers achieved by high exposure assemblies.
O 3.5 Bvoass Flow Desian Criteria.
g O
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EMF-94-217(NP)
O Revision 1 Page 3-6 Bases. The Standard Review PlanMI does not contain an explicit criterion for fuel assembly bypass flow characteristics. However, significant changes in bypass region flow may alter O!
the response characteristics of the incore neutron detectors, in order to avoid altering the incore neutron detector response characteristics, SPC evaluates bypass flow fraction on a plant and cycle specific basis Ol 4:
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O Table 3.1 Summary Description of Thermal and Hydraulic Design Criteria O
Section Description Generic Desian Criteria 3.1 Hydraulic Compatibility Hydraulic flow resistance similar to resident fuel assemblies O
3.2 Thermal Marga Performance
< 0.1 % for rods in boiling transition 3.3 Fuel Centerline Temperature No centerline melting 3
3.4 Rod Bow Protect thermal limits 3.5 Bypass Flow O
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Nuclear Safety Analysisj Neutronics niethxiology
}
Limiting Transient
?
DeltaCPR Plant
?
M Initial Transient Operating Conditions an T ient
?
Limit Thermal m
Ana i i
Lat
~ Critical Power liethodology m
K
?
Figure 3.1 o$
2Eb Calculational Flow Used by the THERMEX Thermal Limits Methodology Ej$
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EMF-94-217(NP)
Rsvision 1 Page 41 l
4.0 NUCLEAR DESIGN AND ANALYSES
)
l 4.1 Nuclear Desion and Analysis Methodoloav SPC's NRC-approved nuclear design and analysis methodology is centered around five computer codes:(29) 1) XFYRE - fuel assembly depletion model, 2) XTGBWR - core simulator l
model, 3) COTRAN - reactor kinetics model, 4) XDT - multi-group diffusion theory, and 1
- 5) XMC - Monte Carlo neutron diffusion. In Reference 6, SPC provided a basis for, and the j
NRC approved, the direct replacement of the fuel assembly depletion model and core simulator i
model with the CASMO-3G and MICROBURN-B codes, respectively. The SPC BWR neutronic l
methods for design and analysis (29) provide NRC-approved methodology for calculation of steady-state core parameters, evaluation of control rod drop accidents, analysis of fuel assembly mislocation, evaluation of control rod withdrawal errors and determination of l
neutronic reactivity parameters.
l l
4.2 Nudeer Desian Analysis l
The nuclear design analyses are divided into two parts: a nuclear fuel assembly design analysis and a core design analysis. The design anclysis should demonstrate operating margin
~
to design limits, including MCPR, maximum average planar linear heat generation rate (MAPLHGR), and LHGR.
j l
4.2.1 Fuel Rod Power Historv
)
Desian Criteria. The nuclear design analysis must be consistent with the exposure dependent j
LHGR limit established during the mechanical design analysis for each fuel assembly design.
Two LHGR limits are established for each fuel design. One is a steady state limit, the other a transient limit. Both limits are a function of fuel planar burnup. The transient LHGR design limit protects strain and fuel overheating design criteria discussed in Sections 2.1.2 and 2.2.4.
)
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EMF-94-217(NP)
O Ravision 1 Page 4-2 The design margin between the steady state and transient LHGR limits is sufficient to account for increases in the LHGR during transients.
Bases. An exposure dependent LHGR limit is established for each fuel assembly design as part of the mechanical design analysis.(10,15) The LHGR limit bounds the power history 4
established to perform the mechanical analyses. Therefore, operation of the fuel assembly within the LHGR limit is necessary to ensure that the power history assumption used in the mechanical design analyses remains valid. The specific mechanical design criteria are discussed in Section 2.0 of this document. During the nuclear design process, and in the g
i reactor, the fuel is required to operate within this power history used to perform the mechanical analyses.
Exposure dependent Linear Heat Generation Rate (LHGR) limits are established for each fuel 8
design. These limits are established based on nuclear design analyses and the fuel design criteria established in Section 2.0. Specifically, conservative power histories are generated based on proposed LHGR limits.
O 1
This assumption leads to conservative estimates of fission gas release.
O.
9 4.2.2 Kinetics Parameters Desian Criteria. The design criteria for the reactivity coefficients are as follows:
j 91 Void Reactivity Coefficient due to boiling in the active channel shall be negative; J
l Doppler Coefficient shall be negative at all operating conditions-G~l Power Coefficient shall be negative at all operating conditions.
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3 EMF-94-217(NP)
Revision 1 Page 4-3 t
Bases. Design of fuel assemblies such that the less moderation and/or higher temperatures
)
reduce the reactivity of the core results in an automatic shutdown mechanism. Thus, prompt critical reactivity insertion events such as the control rod drop accident have an inherent shutdown mechanism. SPC calculates the reactivity coefficients on a plant and cycle specific basis through application of the standard neutrorics design and analysis methodology.W,29) s l
4.2.3 Stability l
D Desian Criteria. New fuel designs and new fuel design features must be stable (core decay ratio <1.0) and exhibit channel decas; ratio characteristics equivalent to existing SPC fuel designs.
I l
Bases. Determination of the effect of all fuel designs and design features on core stability l
is made on a cycle specific basis. Associated with these licensing calculations is confirmation i
of existing exclusion regions or redefinition of the regions, as necessary.
l D
SPC uses the NRC-approved STAIF code (30) for stability evaluations. STAIF is a frequency domain code that simulates the dynamics of a boiling water reactor. STAIF's main output is a series of transfer functions that define the linear dynamic behavior of: 1) the channel D
thermal-hydraulics,2) the fundamental-mode coupled neutronics and thermal-hydraulics, and
- 3) the subcritical-modes coupled neutronics and thermal-hydraulics. STAIF estimates the decay ratios for the above three modes of oscillation using a mathematical procedure that 1
j applies to the computed transfer functions.
D l
SPC will confirm that the stability performance of a new BWR fuel design is equivalent to or better than that of an approved SPC fuel design.
The stability performance will be demonstrated by comparative analytic stability calculations for both the new fuel design and i
the existing approved SPC designs using the approved stability analysis code.
k Because the stability performance of fuel is dominated by the power distribution, tirtd to a g
lesser extent by fuel design variations, it is necessary to establish consistent boundary
)
l EMF-94-217(NP) 0 Ravision 1 Page 4-4 i
conditions when comparweg for fuel design variations. Therefore, SPC will use the same 8
power distribution and full core equilibrium cycle loading pattern for the fuel design comparison. The neutronic characteristics will remain unchanged with the exception of the void coefficient, which is directly dependent on the fue! design. Fuel mechanical design parameters, e.g., component dimensions, spacer loss coeificients, etc. Will be input for each S
fuel design in the comparative analysis.
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The BWR Owners Group is actively developing a generic resolution of the stability concerns.
g' As the BWROG results become available and accepted by the NRC, and as they are implemented by the utilities on an individual basis, the SPC cycle specific stability i
determinations will likely be modified. These modifications will be included in the reactor cycle specific evaluations and are not expected to affect the comparative fuel design related 9
stability performance.
4.2.4 Core Reactivity Control 9l l
Desian Criteria. The design of the assembly shall be such that the Technical Specification I
shutdown margin will be maintained. Specifically, the assemblies and the core must be designed to remain subcritical by the Technical Specification margin with the highest reactivity j
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EMF-94-217(NP)
Rsvision 1 Page 4-5 worth control rod fully withdrawn and the remaining control rods fully inserted. Calculated
)
shutdown margin is verified using startup critical data. At a minimum, this verification is performed at BOC for each reactor.
Bases.
Shutdown margin is calculated on a cycle specific basis using NRC-approved
)
methodology.(6,29) If necessary, shutdown margin is calculated at several cycle exposure points in order to determine the minimum shutdown margin for a cycle. The calculated chutdown margin is reported on a plant and cycle specific basis as required in Reference 43.
')
SPC also confirms the worth of the standby lic.
antrol system on a cycle specific basis using the Technical Specification values of boron concentration.
1 4.3 Generic Analyses When practical, SPC performs generic analyses of normal operation anticipated operational occurrences and postulated accidents.
SPC currently has two NRC-approved generic analyses, Control Rod Withdrawal Error for BWR/6(44,45) and Loss of Feedwater Heating.(46) 4.3.1 Control Rod Withdrawal Error BWR/6 3
SPC had performed, and the NRC has approved, a BWR/6 generic control rod withdrawal error analysis.(44) The generic analysis has been extended to covu MEOD operation.(45) SPC demonstrated that at a 95% confidence level, that MCPR safety limits are not violated for at least 95% of the rod withdrawal events.
)
The SPC generic analysis of the rod withdrawal error event assumes the presence of the Technical Specification on rod withdrawal limits as a function of power which have been established by the reactor designer. These limits are one foot for core powers greater than
)
70% of rated power and two feet for core powers between 20% and 70% of rated. The generic analyses were performed to establish the values of operating limit MCPR as a function 1
of core power which are required to assure that the Safety Limit MCPR will not be violated
)
for the rod withdrawal event.
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EMF-94-217(NP)
O Rsvision 1 Page 4-6 The calculational methods and procedures used to simulate the rod withdrawal event are
'I essentially the approved methods and procedures described in Reference 29. The only difference being the different means of affecting the rod block.
4.3.2 Loss of Feedwater Heatina SPC had developed, and the NRC has approved, a generic methodology for evaluating the loss of feedwater heating (LFWH) transient in BWRs.M The generic methodology is a parametric description of the critical power ratio response that was developed using the results of many ei applications of the currently approved plant / cycle specific methodology.W,29) SPC developed a critical power ratio function Applying this function yields a conservative MCPR operating limit for the LFWH event.
9 4.4 Criticality Safety Analvsis SPC performs criticality safety analyses of new fuel storage vaults and spent fuel storage pools. Storage array k-eff calculations are performed with the KENO.Va Monte Carlo code, which is part of the SCALE 4.2 Modular Code System (Reference 50). The CASMO-3G (Version 4.1) bundle depletion code with the H library (Reference 6) is used to calculate k, values for fuel assemblies at beginning of life (new fuel storage) or as a function of exposure, e
void, and moderator temperature for incore geometry (spent fuel storage).
The KENO.Va and the CASMO-3G computer codes are widely used throughout the nuclear O
industry. They are used primarily for criticality safety and core physics calculations, respectively. SPC has broad experience using both of these codes.
SPC uses criteria given in plant technical specifications and References 51 through 55 to i
Gl assess the acceptability of the criticality safety analyses.
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Summary Description of Nuclear Design Criteria
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Section Descriotion Generic Desian Criteria 4.2.1 Fuel Rod Power History LHGR Limits 1
j 4.2.2 Kinetic Parameter Void Reactivity Coefficient Negative Doppler Reactivity Coefficient Negative Power Coefficient Negative 4.2.3 Stability Decay Ratio < 1.0 4.2.4 Core Reactivity Control Technical Specification - Margin Maintained
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Rsvision 1 Page 5-1 5.0 EVALUATION OF ANTICIPATED OPERATIONAL OCCURRENCES D
Analyses are performed to demonstrate that the fuel performs within design criteria for boiling transition during infrequent and moderately frequent anticipated operational occurrences and to establish appropriate operating limits for the reactor. The methodology used for the analysis of these anticipated events has been reported in References 3, 4, 5, 6, 29 and 40.
The purpose of this section is to identify the potentially limiting events which require evaluation for each operating cycle.
D To prevent or minimize boiling transition, the operating limits established by the evaluation of anticipated operational occurrences consist of a limiting transient ACPR, which in turn defines the MCPR operating limit, and a reduced flow MCPR limit function which adjusts the MCPP D
operating limit at reduced flow settings to allow for the consequences of events which are more severe at reduced flow. In some instances, these analyses also require a reduced powor MCPR limit function which protects the fuel from the consequences of events which are more severe at reduced power.
5.1 Analysis of Plant Transients at Rated Conditions p
Anticipated operational occurrences involving the entire core and the recirculation system are evaluated at full power and flow conditions to determine the nominal MCPR limit. Some events are also analyzed at off-rated conditions as discussed in Sections 5.2 and 5.3. The limiting transient event (or events) is (are) evaluated using the plant transient methodology E
described in References 4 and 5.
The evaluation of anticipated operational occurrences at rated conditions considers events in the following classifications:
Rapid vessel pressurization; Decrease in recirculation flow rate; D
Increase in recirculation flow rate; b
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EMF-94-217(NP)
Rsvision 1 Page 5-2 Decrease in core inlet subcooling; 9:
Increase in core inlet subcooling; Decrease in vessel coolant inventory:
Increase in vessel coolant inventory; and 4i Combination events.
l Events under the classifications of rapid vessel pressurization, increase in recirculation flow O
rate, and increase in core inlet subcooling are potentially limiting events. Events in the categories of increase in vessel coolant inventory and combination events are evaluated on a generic basis for each major BWR plant type. Representative analyses of potentially limiting events in the above classifications for BWR/3 plants are contained in Reference 4.
9:
i 5.2 Analyses for Reduced Flow Operation The transient events described in the preceding section are most severe at full power (except gl for those which result in an increase in recirculation flow and possibly those which result in an increase in vessel coolant inventory). Protection of the MCPR Fuel Cladding Integrity Safety Limit is assured during reduced flow operation through application of a flow dependent O
MCPR operating limit which is established independently of the full flow MCPR limits through analyses of the flow dependent transients from reduced power and flow settings.
The reduced flow MCPR limit is established to perform two protective functions. During 9
operation in the Automat.c Flow Control (AFC) mode, the limit assures that MCPR will not be below the MCPR operating limit if the flow control system demands an increase to full power and full flow. During operation in the Manual Flow Control (MFC) mode, the limit assures that 1
MCPR will not be below the MCPR Fuel Cladding Integrity Safety Limit if the recirculation flow gj is inadvertently increased to the maximum allowed by the physical settings of the equipment.
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the adequacy of the flow-dependent MCPR limit to provide the desired degree of protection of the MCPR limits.
A special case of operation at less than rated power and flow is operation with a single recirculation loop out of service. It may be desirable to operate the reactor on a single loop if one component should require extensive maintenance. Analysis of single loop operation is performed on a plant specific basis, where needed.
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i 5.3 Analysis for Reduced Power Ooeration Reactor operation at points below full power is evaluated to determine the adequacy of the
)
operating limits to protect against fuel f ailures during events initiated from high flow and low power conditions. If a need for reduced power operating limit augmentation is shown, results of these analyses are used to establish a power dependent MCPR limit function which protects the MCPR Fuel Cladding integrity Safety Limit during the occurrence of anticipated events
)
from power-flow states below the nominal 100% flow controlline.
5.4 ASME Overoressurization Analysis
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An overpressurization ana!ysis is performed to assure that the vessel p essure requirements of the ASME Code are satisfied. This analysis, which presumes failure of all non-safety grade components, does not contribute to the determination of thermal margin requirements.
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The turbine trip or generator load rejection transient is generally more limiting in regard to thermal margin requirements than the containment isolation event. However, the Main Steam isolation Valve (MSIV) closure event with the assumption of failure of the direct position
)
scram may result in a more severe calculated overpressurization. Determination of the most severe combination of overpressurization event and active component failure is accomplished generically for each major classification of BWR plants. The ASME overpressurization event
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is analyzed with COTRANSA2.
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l 5.5 Limitino Transient Events O
The loading of fresh fuel, regardless of design,into s reactor core may alter the characteristics of both steady state core performance and plant transient response throughout each subsequent cycle of operation. Limiting criteria for plant operations are established to assure that acceptable thermal operating margins are maintained during all anticipated operations.
Application of SPC's methodology as described in Reference 4 provides a basis for the determination that plant operation will meet appropriate safety criteria.
O Several of the potential events are characteristically self-limiting, but others will require reevaluation of the margins for safe operation. The following have been identified as potentially requiring reanalysis by SPC.
O turbine / generator trip w/o bypass loss of feedwater heating recirculating flow increase events O
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feedwater flow increase inadvertent ECCS high-pressure subsystem startup turbine or generator trip w/o bypass and w/o direct scram' MSIV closure with indirect scram.'
Other events including loss of recirculating coolant flow or increase of recirculating coolant O
enthalpy are inherently non-limiting due to the characteristic negative void reactivity feedback of a BWR. The events described above were also found to be the most limiting of their type due to the severity of the initiating cause.
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- Evaluated for compliance with the provisions of the ASME code.
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Primarily because of the strong void reactivity feedback characteristic of a boiling water
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reactor, incidents involving decrease in vessel coolant inventory and flow and events involving a decrease in vessel coolant subcooling are not expected to result in a limiting ACPR. Events involving either an increase in core inlet subcooling or rapid pressure increases are considered
)
potentially limiting transients for all BWR designs.
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A decrease in feedwater enthalpy may result in a gradual core heatup cate the high neutron flux scram setpoint is exceeded. Since the gradual nature of the power excursion assures that
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the fuel thermal response will not greatly lag the neutronic response, this event can be evaluated with either a transient code or a steady-state code. The possible mitigation of this event with an effective flow control system would not normally be assumed in the analysis.
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Rapid pressure increases may be a thermal margin limiting event for some designs and conditions. The severity of the event is strongly dependent upon the reactivity state of the core, the valve closure characteristics initiating the event, and the performance of the scram shutdown system. Thus, specific event sequences at some reactor conditions may emerge as consistently most limiting in nature. Lead plant analyses willinitially evaluate a spectrum of pressurization events for each general class of JP-BWRs. Each potentially limiting event will be considered in the determination of cycle limiting conditions for operation. The scope
)
of plant specific analyses may be narrowed to those conditions determined to be consistently more limiting.
The remaining two single event categories which involve increases in either coolant flow rate
)
or inventory are dependent upon plant design and conditions. Both involve potentially limiting conditions at partial power and flow conditions, where the augmentation of flow (either recirculatory or feed) to the maximum physical capacity of equipment is greatest. Effective designs and/or reactor protection systems may substantially mitigate the rate and potential acceleration of power production in the core or terminate the transient prior to serious degradation of thermal margin. Current technical specifications for JP-BWRs provide an augmentation of the CPR operating limit to protect against potentially greater transient
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reduction of CPR at partial flow conditions. SPC will verify that the existing augmentation
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EMF-94-217(NP) g:
Revision 1 Page 5-6 procedure (i.e., K, curves) for a plant will provide adequate thermal margin for the cycle design at applicable conditions.
Once the applicable set of limiting transients for thermal margin has been identified for a specific reactor, the evaluation of each event at limiting reactor conditions will provide the basis for determining the MCPR Operating Limit which is applicable to all other anticipated operating conditions.
1 es 5.6 Control Rod Withdrawal Error Withdrawal of the highest worth control rod in the core until its movement is blocked by the control system is evaluated with MICROBURN-B, which is described in Reference 6.
O For the analysis, the reactor is assumed to be in a normal mode of operation with the control rods being withdrawn in the proper sequence and all reactor parameters within the Technical Specification limits and requirements. The most limiting case is when the reactor is operating at power with a high reactivity worth control rod fully inserted.
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EMF-94-217(NP) 4 Ravision 1-Page 5-7 Results for the control rod withdrawal error analysis include maximum control rod withdrawal O
distance, change in thermal margin, and the limiting control rod pattern used for the analysis.
For reactors utilizing reduced power augmentation to MCPR limits, the existing reduced pow er limit functions are revised as necessary and verified for operation with SPC-fabricated fuel.
A detailed description of the SPC Control Rod Withdrawal Error evaluation methodology is g
given in Reference 29.
5.7 Determination of Thermal Marains 20 The results of the anticipated operational occurrences evaluated under this chapter are compared for the greatest change in MCPR for full power operation. (If the event causing the greatest change in MCPR is one which is sensitive to the uncertainties in plant measurement O
data, the statistical methods described in Reference 40 are used to calculate the thermal margin requirement. If necessary, additional statistical analyses are used to determine which of the considered transient events should define the MCPR operating !imit.)
- O The limiting transient ACPR which is used to define the MCPR operating limit is used to select the rod block monitor setting from the tabulated results of the control rod withdrawal error analysis. Observance of the operating MCPR limit and rod block monitor settings determined
-O in this fashion provides protection of the MCPR Fuel Cladding integrity Safety Limit during operation at rated conditions.
i The results of the reduced flow and reduced power analyses are used to establish the proper O
values for the MCPR limit functions required for operation at lower than rated power and flow conditions. Reactor operation within the power-and flow-dependent limits defined in this f ashion assures adequate protection of MCPR limits throughout the power-flow operating map.
'O The scram insertion time used for the transient analyses may be based on either the Technical Specifications or plant measurement data. If plant measurement data were used to determine the scram performance used to define any of the limiting safety system settings or limiting
-O' O
EMF-94-217(NP)
O Revision 1 Page 5-8 conditions for operation, surveillance procedures are specified to determine the continued applicability of the data base and to modify limits to assure applicability of the analysis.
O The cote power and exposure distributions are monitored by the licensee throughout the cycle to assure that the EOC axial power shape assumed in the licensing analysis will bound the 9
actual EOC s.xial power shape.
5.8
!b>poortino Documentation l
j Results of the transient analyses at rated conditions are documented in the plant transient l
analysis report, which also reports the calculation of the MCPR Fuel Cladding Integrity Safety Limit. If epplicable, results from reduced flow and power analyses are also reported in the plant transient analysis report. Control rod withdrawal error analyses at reduced power O'
conditions are performed and reported on a generic basis for the classifications of BWR plants which utilize reduced power augmentation to MCPR limits.
9 Results of the cycle analyses described in this section are documented in the reload analysis report, which may also incorporate the plant transient analysis report referenced above.
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Hypothetical Loss of Coolant Accidents (LOCAs) are analyzed in accordance with Appendix K modeling requirements using the ECCS models described in References 21, 22, 25, and 27.
Postulated Control Rod Drop Accidents are analyzed using the COTRAN methodology described in Reference 29.
6.1 Loss of Coolant Accident Analysis D
The ECCS analyses provide Peak Cladding Temperature (PCT) and peak local metal-water reaction (MWR) values and are used to define MAPLHGR limits in accordance with 10 CFR 50.46. For each SPC fuel type and limiting single failure, limiting break calculations I
are undertaken to determine the MAPLHGR, PCT, and MWR values over the expected exposure lifetime of the fuel. The limiting break is determined generically for each BWR type by evaluating a spectrum of potential break locations and sizes. A representative fuel type is used for the break spectrum evaluation. Once the limiting break is identified, PCT analysis for specific fuel types is performed.
6.1.1 Break Location Soectrum D
Representative LOCA analyses for piping breaks in the recirculation system piping form the basis for the location spectrum evaluation, which is accomplished on a generic basis for each major class of BWR plants. A figure of merit is drawn from MAPLHGR, PCT, and MWR values I
calculated for consistent exposure conditions in the fuel at each of the break spectrum locations. Analyses performed by the NSSS supplier are used as guidelines to narrow the sco.ne of the analyses.
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O,I Ravision 1 Page 6-2 6.1.2 Break Size Spectrum ei Once the location of the limiting break has been established, representative analyses are undertaken to establish the size of the limiting break. These analyses are performed on a generic basis for each major class of BWR plants.
O' Hypothetical piping system breaks are evaluated up to and including those with a break area equal to twice the cross-sectional area of the largest pipe in the limiting break area. Due to physical phenomena observed during the blowdown phase of the LOCA analysis, the e
difference in results between guillotine pipe breaks and split pipe breaks is not significant for total break areas greater than 40% of the maximum guillotine break area. Smaller breaks are evaluated to assure that the largest break is also the most severe. If the largest break is not shown to be the most severe, the most severe break is selected for further analysis. As with O
the location spectrum, the determination of the limiting break size is based on a comparison of MAPLHGR, PCT, and MWR values for consistent exposure conditions in the fuel.
9 6.1.3 Worst Sinale Failure SPC works closely with the licensee to identify the most damaging single failure of ECCS equipment. The most damaging single failure of ECCS equipment is that failure which results
.g in the largest increase in PCT for the limiting break location and size.
Previous evaluations by the NSSS vendor identifying the worst case single failure of ECCS equipment are used in determining the worst single failure. The applicability of the previous 0,
evaluations will be reviewed on a case-by-case basis. Additional calculations may be performed if plant modifications have occurred.
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After the location and size of the limiting break have been determined, analyses are undertaken to characterize the maximum steady-state nodal power at which the fuel may be operated prior to the postulated design basis LOCA without exceeding the ECCS limits g
specified in 10 CFR 50.46.
O
.O The blowdown phase is evaluated with RELAX (References 22 and 27). Refill and reflood are g
evaluated with FLEX (References 22 and 27).
Fuel heatup is analyzed with HUXY (Reference 47). Stored energy and fuel response characteristics are determined with RODEX2 (Reference 25).
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6.1.5 Sucoortino Documentation The break spectrum analyses are performed and repor'.ed generically for each major O
classification of BWR plants. The cycle specific reload ana', als references the generic break O
EMF-94-217(NP) g Revision 1 Page 6-4 spectrum analyses in specifying the limiting break location and size for the reactor in question.
O Detailed MAPLHGR analyses may be incorporated in the reload analysis or may be reported separately and referenced in the reload analysis.
6.2 Control Rod Droo Accident Analysis O
Analysis of the postulated Control Rod Drop Accident (CRDA) is performed on a generic basis in Reference 29. Because the behavior of the fuel and core during such an event is not dependent upon system response, a single generic CRDA analysis can be applied to all BWR e
types. The applicability of the generic CRDA analysis is verified for each application.
The results of the generic CRDA analysis consist of deposited fuel enthalpy values parameterized as a function of effective delayed neutron fraction, Doppler coefficient, O
maximum (dropped) control rod worth, and four-bundle local peaking factor. Each cycle-specific application includes the values for each of the parameters, which are compared to the generic analysis and curves and the resulting deposited fuel enthalpy determined.
O 6.3 f_uel Loadino Error Two separate incidents are analyzed as part of the fuel misloading analysis. The first incident, g
which is ter.ned the fuel mislocation error, assumes a fuel assembly is placed in the wrong core location during refueling. The second incident, the fuel misorientation error, assumes that a fuel assembly is misoriented by rotation through 90 or 180 from the correct orientation when loaded into the reactor core. For both the fuel mislocation error and the fuel 9
misorientation error, the assumption is made that the error is not discovered during the core verification and the reactor is operated during the cycle with a fuel assembly mistoaded.
The limiting parameter of interest for the fuel misloading error is the MCPR for the mistoaded fuel assembly. The fuel misloading analysis determines the difference between the MCPR for the correctly loaded core and the MCPR for the core with a mistoaded fuel assembly. Criteria O
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event shall not exceed a small fraction of the 10 CFR 100 limits (Reference 17).
6.3.1 Mistoaded Fuel Bundle D
The inadvertent misloading of a fuel bundle into an incorrect core location is analyzed with the MICROBURN-B methodology described in Reference 6.
D D
Detailed application of the misloaded fuel bundle methodology is described in Reference 29.
D 6.3.2 Misoriented Fuel Bundle The inadvertent rotation of a fuel bundle away from its intended orientation is evaluated with D
the CASMO-3G methodology described in Reference 6. The analysis describes a minimum value of MCPR and a maximum LHGR associated with the orientation error.
B Detailed application of B
the misoriented fuel bundle methodology is described in Reference 29.
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EMF-94-217(NP) n'#
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The introduction of a new mechanical fuel design into a reactor core must be supported by an evaluation of the fuel handling accident for the new fuel design. When required, SPC performs an incremental evaluation of the impact of the new fuel design on the fuel handling 9
accident scenario defined in the reactor's Final Safety Analysis Report (FSAR). Using the boundary conditions and conservative assumptions given in the FSAR and the relevant characteristics of the new fuel design, SPC calculates a conservative number of fuel rods expected to fail as a result of a fuel handling accident with the new fuel design.
g The radiological consequences of a fuel handling accident for a new mechanical fuel design are assessed
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7.0 REFERENCES
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1.
Standard Review Plan for the Review of Safety Analvsis Reoorts for Nuclear Power Plants, NUREG-0800, U.S. Nuclear Regulatory Commission, July 1981.
2.
QA Proaram for Nuclear Fuel Desian and Fabrication. EMF-1, Siemens Power Corporation, Richland, WA 99352.
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7.
" Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,"
l Code of Federal Reaulations. Title 10 " Energy," Part 50, Appendix B.
8.
" Rules for Construction of Nuclear Power Plant Components," ASME Boiler and l
Pressure Vessel Code. Section 111, 1977.
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l 9.
Swanson Analysis System, "ANSYS-Engineering Analysis SystemTheoreticalManual,"
1977, and "ANSYS-User's Guide," 1979.
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W. J. O'Donnell and B. F. Langer, " Fatigue Design Basis for Zircaloy Components," In Nuc. Sci. Ena., 1964, 20:1.
g 16.
" General Design Criteria for Nuclear Power Plants," Code of Federal Reaulations, O
Title 10 " Energy," Part 50, Appendix A.
17.
" Reactor Site Criteria," Code of Federal Reaulations, Title 10 " Energy," Part 100.
9 19.
"ECCS Evaluation Models," Code of Federal Reaulations, Title 10 " Energy," Part 50, Appendix K.
20.
Claddina Swellina and Ruoture Models for LOCA Analvsis. NUREG 0630, U.S. Nuclear Regulatory Commission, April 1980.
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26.
" Acceptance Criteria for Emergency Core Cooling Systems for Light Water Reactors,"
9 Code of Federal Reaulations. Title 10 " Energy," Part 50.46.
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" Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized O
Water Reactors," Reaulatorv Guide 1.77. U.S. Atomic Energy Commission, Washington, D.C., May 1974.
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50.
A Modular Code Svstem for Performina Standardized Computer Analyses for Licensina Evaluation, SCALE 4.2, Oak Ridge National Laboratory, revised December 1993.
51.
Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power 9l Plants, NUREG-0800, Section 9.1.1 (New Fuel Storage), U.S. Nuclear. Regulatory Commission, July 1981.
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Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power p
Plants. NUREG-0800, Section 9.1.2 (Spent Fuel Storage), U.S. Nuclear Regulatory Commission, July 1981.
53.
Soent Fuel Storaae Facility Desian Basis. Regulatory Guide 1.13, Proposed Revision 2, U.S. Nuclear Regulatory Commission, December 1981.
D 54.
Desian Reauirements for Licht Water Reactor Spent Fuel Storaae Facilities at Nuclear Power Plants. ANSl/ANS American National Standard 57.2-1983, American Nuclear Society, October 1983.
55.
Criticality Safety Criteria for the Handlina, Storaae and Transportation of LWR Fuel p
Outside Reactors. ANSI /ANS American National Standard 8.17-1984, American Nuclear Society, January 1984.
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APPENDIX A i
O LISTING OF SIEMENS POWER CORPORATION NRC APPROVED LICENSING TOPICAL REPORTS SUPPORTING BWR METHODOLOGY O
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Boiling Water Reactor Licensing Methodology Summary
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