ML17158C148
| ML17158C148 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 05/07/1997 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Allegheny Electric Cooperative, Pennsylvania Power & Light Co |
| Shared Package | |
| ML17158C149 | List: |
| References | |
| NPF-22-A-139 NUDOCS 9705140439 | |
| Download: ML17158C148 (21) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PENNSYLVANIA POWER 5 LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.
DOCKET NO 50-388 SUS UEHANNA STEAM E ECTRIC ST TION UNIT 2 AMENDMENT TO FACILIT OP R TI G LICENSE Amendment No. 139 License No. NPF-22 I.
The Nuclear Regulatory Commission (the Commission or the NRC) having found that:
A.
The application for the amendment filed by the Pennsylvania Power 5
Light Company, dated December 18, 1996 as supplemented on February 26, March 12 and 27, April 3, 9, 16, 18, and 24,
- 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C.
There is reasonable assurance:
(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-22 is hereby amended to read as follows:
(2) Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.139, and the Environmental Protection Pl'an con-
.tained in Appendix 8, are hereby incorporated in the license.
PP&L shall operate the facility in accordance with the Technical Specifica-tions and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and is to be implemented upon receipt by the licensee.
FOR THE NUCLEAR REGULATORY COMMISSION J
F. Stolz, Dire r
oject Directorate
-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
~~ >> >>97
ATTACHMENT TO LICENSE AMENDMENT NO.
139 FACILITY OPERATING LICENSE NO.
NPF-22 DOCKET NO. 50-388 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.
The revised pages are identified by Amendment number and contain vertical'ines indicating the area of change.
REMOVE Xxii INSERT Xxii 2-1 B 2-1 B 2-2 3/4 4-1c 5-6 6-20b 2-1 2-2a B 2-1 B 2-2 3/4 4-1c 3/4 4-1g 5-6 6-20b 6-20c
I
INDEX LIST OF FIGURES 2.1.2-1 MCPR SAFETY LIMIT.~.....
2-2a 3.1.5-1 SODIUM PENTABORATE SOLUTION TEMPERATURE/
CONCENTRATION REQUIREMENTS.....,.....................
3/4 1-21 3.1.5-2 SODIUM PENTABORATE SOLUTION CONCENTRATION....~.......... 3/4 1-22 3A.1.1.1-1 THERMALPOWER RESTRICTIONS 3/4 4-1b 3A.'I.1.2-1 SINGLE LOOP MCPR SAFETY LIMIT....................................... 3/4 4-1g 3.4.6.1-1 MINIMUMREACTOR VESSEL METALTEMPERATURE VS.
REACTOR VESSEL PRESSURE
. 3/4 4-18 4.7.4-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONALTEST.. ~ ~.~... ~.. ~. ~.. 3/4 7-15 B 3/4 3-1 REACTOR VESSEL WATER LEVEL....... ~. ~.~....
~.~..........
B 3/4 3-8 B 3/4A.6-1 FAST NEUTRON FLUENCE (E) 1MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE
. B 3/4 4-7 SUSQUEHANNA - UNIT 2 XXII B,9$,
Amendment No. i39
1.0 DEFINITIONS The following terms are defined so that uniform interpretation of these specifications may be achieved..The defined terms appear in capitalized type and shall be applicable throughout these Technical Specifications.
hGIIQH 1.1 ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.
'l.2 The AVERAGE BUNDLE EXPOSURE shall be equal to the total energy produced by the bundle divided by the total initial weight of uranium in the fuel bundle.
The AVERAGE PLANAR EXPOSURE at a specified height shall be equal to the total energy produced per unit length at the specified height divided by the total initial weight of uranium per unit length at that height.
1.3 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle at the height.
I 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST.
Calibration of instrument channels with resistance temperature detector IRTD) or thermocouple sensors may consist of an in-place qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel.
The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.
1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.
1.6 A CHANNEL FUNCTIONALTEST shall be:
a.
Analog channels
- the injection of a'simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.
b.
Bistable channels
- the injection of a
simulated signal into the sensor to verify OPERABILITYincluding alarm and/or trip functions.
The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.
SUSQUEHANNA - UNIT 2 n, res, Amendment No. I39
.~
2.0 SAFETY LIMITSAND LIMITINGSAFETY SYSTEM SETTINGS 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10 million Ibm/hr.
- OPERATIONAL CONDITIONS 1 AND 2.
hQIIQM'ith THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10 million Ibm/hr., be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
2.1.2 The MINIMUMCRITICAL POWER RATIO (MCPR) shall not be less than the value shown in Figure 2.1.2-1" with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10 million Ibm/hr.
OPERATIONAL CONDITIONS 1 AND 2.
hQ:QQN:
With MCPR'less than the value shown in Figure 2.1.2-1" and the reactor vessel steam y¹ dome pressure greater than 785 psig and core flow greater than 10 million Ibm/hr., be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome,
'shall not exceed 1325 psig.
~JaaaUZD hQIIQH:
With the reactor coolant system
- pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification'.7.1.
See Specification 3.4.1.1.2.a for single loop operation requirement.
Only applicable for Unit 2 Cycle 9 operation.
SUSQUEKANNA - UNIT 2 2-1 Amendment No.gg, 189, lN 139
1.40 1.35 1.30
~ 1.25
~g 1.20 M
a.
1.15 O
1.10 1.05 o o o e e o eye o o o o o o ~ o o o o eg 30,1.27
~to o e.. 50,1.22.
40,1.24 ~ oo ~ ~ ooo O4 e o
~ ~
o
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- ~ 70,1 ooo+ooo
~ ~ o ceo o
~e ~ ~ ~
~ ~ ~ Ho ~ o
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17 o
I
~ oo e o o o o o OIO o o o o o o o o OI~ o
.19 J
ooo e
ooeiooo o ~ oo Joo o
Jo oooo e
.... 90,1.13..
80,1.15 o o o IP ~ o o o Po o ~
1 o o ooooaoo 60i1 o ~ o o ~ o o ~ e o le O SO 108,1.
00,1.1 1.00 30 40 50 60 70 80 90 100 110 Cote Flow (Mlb/hr)
Figure 2.1.2-1 MCPR Safety Limitvs Core Flow SUSQUEHANNA - UNIT 2 2 28 Amendment No. i3~
BASES The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients.
The fuel cladding integrity Safety Limit is set such that no fuel damage is expected to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than the limit specified in Specification 2.1.2 for SPC fuel.
MCPR greater than the specified limit represents a
conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs.
The integrity of this cladding barrier 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 incrementally cumulative and continuously measurable.
Fuel cladding perforations,
- however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use related
- cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.
Therefore, the fuel cladding Safety Limit is defined with a 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. The MCPR fuel cladding integrity Safety Limit assures that during normal operation and during anticipated operational occurrences, at least 99.9% of the fuel rods in the core do not experience transition boiling (ref. ANF-524(P)(A) Revision 2).
The use of the ANFB correlation is valid for critical power calculations at pressures greater than 585 psig and bundle mass fluxes greater than 0.1 x 10 Ibs/hr-ft. For operation at low 6
2 pressures or low flows, the fuel cladding integrity Safety Limit is established by a limiting condition on core THERMAL POWER with the following basis:
Provided that the water level in the vessel downcomer is maintained above the top of the active fuel, natural circulation is sufficient to assure a minimum bundle flow for all fuel assemblies which have a relatively high power and potentially can approach a critical heat flux condition. For the SPC 9x9-2 fuel design, the minimum bundle flow is greater than 30,000 Ibs/hr.
For the SPC ATRIUM-10 design, the minimum bundle flow is greater than 28,000 Ibs/hr.
For both the SPC 9x9-2 and ATRIUM-10 fuel designs, the coolant minimum flow and maximum flow area is such that the mass flux is always greater than 0.25 x 10 Ibs/hr-ft.
Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at 0.25 x 10 Ibs/hr-ft is 3.35 Mwt or greater.
At 25%
thermal power a bundle power of 3.35 Mwt corresponds to a bundle radial peaking factor of approximately 3.0 which is significantly higher than the expected peaking factor. Thus, a
THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressures below 785 psig is conservative.
SUSQUEHANNA - UNIT 2 8 2-1 Amendment No. B N$, fN.> 139
BASES Onset of transition boiling results in a decrease in heat transfer from the clad and, therefore, elevated clad temperature and the possibility of clad failure. However, the existence of critical power, or boiling transition, is not a directly observable parameter in an operating reactor.
Therefore, the margin to boiling transition is calculated from plant operating parameters such as core
- power, core flow, feedwater temperature, and core power distribution. The margin for each fuel assembly is characterized by the critical power ratio (CPR), which is the ratio of the bundle power which would produce onset of transition boiling divided by the actual bundle power. The minimum value of this ratio for any bundle in the core is the minimum,critical power ratio (MCPR).
The Safety Limit MCPR assures sufficient conservatism in the operating MCPR limit that in the event of an anticipated operational occurrence from the limiting condition for operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. The margin between calculated boiling transition (MCPR = 1.00) and the Safety Limit MCPR is based on a detailed statistical procedure which considers the uncertainties in monitoring the core operating state.
One specific uncertainty included in the safety limit is the uncertainty inherent in the critical power correlation.
Section 6.9.3.2 contains the methodologies used in determining the Safety Limit MCPR.
The ANFB critical power correlation is based on a significant body of practical test data.
As long as the core pressure and flow are within the range of validity of the correlation (refer to Section 8 2.1.1), the assumed reactor conditions used in defining the safety limit introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition.
These conservatisms and the inherent accuracy of the ANFB correlation provide a reasonable degree of assurance that during sustained operation at the Safety Limit MCPR there would be no transition boiling in the core. If boiling transition were to occur, there is reason to believe that the integrity of the fuel would not necessarily be compromised.
Significant test data accumulated by the U.S. Nuclear Regulatory Commission and private organizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach. Much of the data indicates that LWR fuel can survive for an extended period of time in an environment of boiling transition.
SPC fuel is monitored using the ANFB Critical Power Correlation.
The effects of channel bow on MCPR are explicitly included in the calculation of the ANFB MCPR Safety Limit. Explicit treatment of channel bow in the ANFB MCPR Safety Limit addresses the concerns of NRC Bulletin No. 90-02 entitled "Loss of Thermal Margin Caused by Channel Box Bow."
SUSQUEHANNA - UNIT 2 B 2-2 Amendment No.N >
IN, 139
0
LIMITINGCONDITION FOR OPERATION 3.4.1.1.2 One reactor coolant recirculation loop shall be in operation with the pump speed 5 80% of the rated pump speed and the reactor at a THERMAL POWER/core flow condition outside of Regions I and II of Figure 3.4.1.1.1-1, and a.
the following revised specification limits shall be followed:
1.
Specification 2.1.2:
the MCPR Safety Limit shall be increased to the value shown in Figure 3A.1.1.2-1 2.
Table 2.2.1-1:
the APRM Flow-Biased Scram Trip Setpoints shall be as follows:
':i!j~'~)+!g<Q~~,";.:$'~-&Tiip".Setpoint>>:N:.-,":-'4':=::
5 0.58W + 54%
0.58W + 57%
3.
Specification 3.2.2: the APRM Setpoints shall be as follows:
S 5 (0.58W + 54%) T SRs 5 (0.58W + 45%) T S 5 (0.58W + 57%) T SRB ~ (0.58W + 48%) T 4.
Specification 3.2.3: The MINIMUM CRITICAL POWER RATIO (MCPR) shall be greater than or equal to the applicable Single Loop Operation MCPR limit as specified in the CORE OPERATING LIMITS REPORT.
5.
Specification 3.2.4: The LINEAR HEAT GENERATION RATE (LHGR) shall be less than or equal to the applicable Single Loop Operation LHGR limit as specified in the CORE OPERATING LIMITS REPORT.
6.
Table 3.3.6-2: the RBM/APRM Control Rod Block Setpoints shall be as follows:
- a. RBM-Upscale
- b. APRM-Flow Biased k~<~:,"::i9>Trip:,Setpoint,.:;:: '";.5,;
60.63w + 35%
",";::-,'P~Tr(p;Setpoirit";.;:;;::.-,',:::,.
5 0.58W + 45%
).',:::.,':':;-'.; Allowable,'.,Valu'e:":.,:'.:.<'jl 063W + 37
";:::;:;!-:Allowable';.Vayu'ei.".';:-":!'0.58W
+ 48%
hQIIQH:
OPERATIONAL CONDITIONS 1 and 2 +, except during two loop operation.
a.
In OPERATIONAL CONDITION 1:
1.
With a) no reactor coolant system recirculation loops in operation, or b)
Region I of Figure 3.4.1.1.1-1 entered, or c)
Region II of Figure 3.4.1.1.1-1 entered and core thermal hydraulic instability occurring as evidenced by:
Only applicable for Unit 2 Cycle 9 operation.
f SUSQUEHANNA - UNIT 2 3/4 4-1 c Amendment Nodal >
9S s EN ~
139
1.40 1.35 IM E
1.30 I
1.25 o 120 o
1.15 1.10 V) 1.05 30 1o28o~o o o o o.
~ ~ ~
~ o o o o I'
o o o o
~o o ~ oOo ~ o o 52)1.22 o Jo o 4
.... 40,1.26...
o o o
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~ o o o o o o
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~ oi
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~
o o o o oo oio o o o ~ o o o o o o ~ ~
O ~ o o o o o o 1.00 30 40 60 60 Vo 80 00 100 110 Core Flow (Mlb/hr)
Figure 3.4.1.1.2-1 MCPR Safety Limitv's Core Flow Single Loop Operation SUSQUEHANNA - UNIT 2 3/4 4-1g Amendment No. 139
II'
DESIGN FEATURES 5.3.1 The reactor core shall contain 764 fuel assemblies.
Each assembly consists of a matrix of Zircaloy clad fuel rods with an initial composition of non-enriched or slightly enriched uranium dioxide as fuel material and water rods or water channels.
Limited substitutions of Zirconium alloy filter rods. for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used.
Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff-approved codes and methods, and shown by test or analyses to comply with all fuel safety design bases.
A limited number of lead use assemblies that have not ')
completed representative testing may be placed in non-limiting core regions.
Reload fuel shall have a maximum lattice average enrichment of 4.5 weight percent U-235.
5.3.2 The reactor core shall contain 185 cruciform shaped control rod assemblies.
The control material shall be boron carbide powder (B4C), and/or Hafnium metal.
The
~ control rod shall have a nominal axial absorber length of 143 inches.
Control rod assemblies shall be limited to those control rod designs approved by the NRC for use in BWRs.
5.4.1 The reactor coolant system is designed and shall be maintained:
a.
In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.
For a pressure of:
1250 psig on the suction side of the recirculation pumps.
2.
1500 psig from the recirculation pump discharge to the jet pumps.
c.
For a temperature of 575'F.
5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,400 cubic feet at a nominal T,, of 532 F.
SUSQUEHANNA - UNIT 2 5-6 79l%, LN, Amendment No. 139
0 1
ADMINISTRATIVECONTROLS (Continued) 14.
ANF-1125(P)(A) and ANF-1125(P)(A), Supplement.
1, "ANFB Critical Power Correlation," April 1990.
15.
16.
NEDC-32071P, "SAFER/GESTR-LOCA Loss of Coolant Accident Analysis,"
'GE Nuclear Energy, May 1992.
R NE-092-001A, Revision 1, "Licensing Topical Report for Power Uprate With Increased Core Flow," Pennsylvania Power 5 Light Company, December 1992.
17.
NRC SER on PPRL Power Uprate LTR (November 30, 1993).
18.
PL-NF-90-001, Supplement 1-A, "Application of Reactor Analysis Methods for BWR Design and Analysis: Loss of Feedwater Heating Changes and Use of RETRAN MOD 5.1," September 1994.
19.
PL-NF-94-005-P-A, "Technical Basis for SPC 9x9-2 Extended Fuel Exposure at Susquehanna SES," January 1995.
20.
21.
NEDE-24011-P-A-10, "General Electric Standard Application for Reactor Fuel," February 1991.
PL-NF-90-001, Supplement 2-A, "Application of Reactor Analysis Methods I
to BWR Design and Analysis: CASMO-3G Code and ANFB Critical Power Correlation."
22.
ANF-89-98(P)(A)
Revision 1
and Revision 1
Supplement 1,
"Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corporation, May 1995.
23.
XN-NF-81-58(P)(A) Supplements 1 and 2 Revision 2, "RODEX 2 Fuel Rod Thermal-Mechanical Response Evaluation Model," May 1986.
24.
XN-NF-85-74(P)(A),
"RODEX 2A (BWR)
Fuel Rod Thermal-Mechanical
Response
Evaluation Model," August 1986.
25.
XN-NF-82-06(P)(A) and Supplements 2, 4, and 5 Revision 1, "Qualification of Exxon Nuclear Fuel for Extended Burnup," October 1986.
26.
XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity," November 1986.
27.
28.
29.
ANF-90-082(P)(A) Revision 1 and Revision 1 Supplement 1, "Application of ANF Design Methodology for Fuel Assembly Reconstitution," May 1995.
~
ANF-91-048(P)(A),, "Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model," January 1993.
Il ANF-CC-33(P)(A) Supplement 2, "HUXY: A Generalized Muitirod Heatup Code with 10CFR50 Appendix K Heatup Option,." January 1991.
SUSQUEHANNA - UNIT 2 6-20b Amendment No.
9>~ 8@, OP,/44, V.P,A%, LN, >39
30.
XN-CC-33(P)(A) Revision 1, "HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option Users Manual," November 1975.
31.
XN-NF-80-19(P)(A),
Volumes 2,
2A, 2B, and 2C "Exxon Nuclear Methodology for Boiling Water Reactors:
EXEM BWR ECCS Evaluation Model," September 1982.
32.
XN-NF-80-19(P)(A), Volumes 3 Revision 2 "Exxon Nuclear Methodology for Boiling Water Reactors Thermex:
Thermal Limits Methodology Summary Description," January 1987.
33.
XN-NF-79-71(P)(A) Revision 2, Supplements 1, 2, and 3, "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," March 1986.
34.
ANF-1358(P)(A), Revision 1, "The Loss of Feedwater Heating Transient in Boiling Water Reactors," September 1992.
35.
ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3, and 4, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," August 1990.
36.
XN-NF-84-105(P)(A), Volume 1
and Volume 1
Supplements 1
and 2,
"XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," February 198/.
37.
XN-NF-84-105(P)(A), Volume 1 Supplement 4, "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, Void Fraction Model Comparison to Experimental Data," June 1988.
38.
EMF-97-010, Revision 1,
"Application of ANFB to ATRIUM'-10 for Susquehanna Reloads," March 1997.
I 39.
PLA-4595, "Response to NRC Request For Additional Information On
¹ Siemens'eport EMF-97-010, Revision 1," Mar'ch 27, 1997.
6.9.3.3 The core operating limits shall be determined such that ail applicable limits (e.g.,
fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, transient analysis limits and accident analysis limits) of the safety analysis are met.
ln addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.
6.10.1 The following records shall be retained for at least 5 years:
a.
Records and logs of unit operation covering time interval at each power level.
Only applicable for Unit 2 Cycle 9 operation.
SUSQUEHANNA - UNIT 2 6-20c Amendment No.139