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Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML20211A9881999-08-18018 August 1999
Rev 5 to DPC Nuclear Security Training & Qualification Plan
ML20204B4141999-03-27027 March 1999
Revised Oconee Nuclear Station Selected Licensee Commitments, List of Effective Pages
ML20155C6391998-10-26026 October 1998
Proposed Tech Specs Tables 4.1-1 & 4.1-2,revised Surveillance Frequency of Refueling Outage to 18 Months
ML20154L7191998-10-0505 October 1998
Rev 8 to DPC Nuclear Security & Contingency Plan, for Oconee,Mcguire & Catawba Nuclear Stations
ML20154C1021998-09-30030 September 1998
Proposed Tech Specs Changing UFSAR by Increasing Max Rod Internal Pressure in Spent Fuel Pool from 1,200 to 1,300 Psig
ML20236V9611998-07-27027 July 1998
Corrected TS Bases Pages 3.5-30b of 961211 & 3.16-2 of 970205,correcting Amend Numbers
ML20236T5171998-07-20020 July 1998
Proposed Tech Specs Re one-time Extension to Functional Test Frequencies for Hydraulic & Mechanical Snubbers
ML20236R1901998-07-16016 July 1998
Proposed Tech Specs Extending TS Surveillances TS 4.20.2(a), Table 4.20-1,items 1,2 & 4,TS 4.1-1,Table 4.1-1,items 8,9,10 & 11 & TS 4.5.1.2.1,item b(1)
ML20217F2841998-04-20020 April 1998
Rev 7 to DPC Nuclear Security & Contingency Plan, for Oconee,Mcguire & Catawba Nuclear Stations
ML20141F1521997-06-25025 June 1997
Rev 4 to Nuclear Security Training & Qualification Plan
ML20134K5401996-10-31031 October 1996
Rev 6 to Chemistry Manual 5.1, Emergency Response Guidelines
ML20117J0181996-08-15015 August 1996
Revised Chapter 16 of Oconee Selected Licensee Commitments Manual
ML20095H1291995-12-0505 December 1995
Rev to ONS Selected Licensee Commitments (SLC) Manual, Revising SLC 16.6.1, Containment Leakage Tests to Reflect Current Plant Configuration & Update Testing Info
ML20091P2461995-08-21021 August 1995
Rev to ONS Selected Licensee Commitments Manual
ML16141A8111994-04-20020 April 1994
Proposed Tech Spec Table 4.1-2, Min Equipment Test Frequency.
ML15224A3521993-10-26026 October 1993
Safety Assurance Directive 6.1, Oconee Nuclear Site Safety Assurance Emergency Response Organization.
ML20044C9581993-05-0404 May 1993
Proposed TS 4.7.1, Control Rod Trip Time Test.
ML20045F0721993-04-0707 April 1993
Rev 9 to Corporate Process Control Program Manual.
ML20097D9451992-05-28028 May 1992
Rev 11 to Training & Qualification Plan
ML20096C2561992-04-30030 April 1992
Public Version of Revised Crisis Mgt Implementing Procedures (Cmip),Including Rev 46 to CMIP-1,Rev 39 to CMIP-4,Rev 45 to CMIP-5,Rev 49 to CMIP-6,Rev 48 to CMIP-7,Rev 42 to CMIP-9 & Rev 3 to CMIP-15
ML20096D5451992-04-0707 April 1992
Public Version of Revised Crisis Mgt Implementing Procedures (Cmip),Including Rev 45 to CMIP-1,Rev 27 to CMIP-13 & Notification of Deletion of CMIP-8.Procedure CMIP-8 Reserved for Future Use
ML20092M5891992-02-0606 February 1992
Public Version of Revised Crisis Mgt Implementing Procedures,Including Rev 43 to CMIP-1,Rev 43 to CMIP-5,Rev 33 to CMIP-8,Rev 25 to CMIP-13,Rev 6 to CMIP-18,Rev 4 to CMIP-22 & Rev 38 to CMIP-4
ML20094H2391992-01-0101 January 1992
Rev 33 to McGuire Nuclear Station Odcm
ML20094H2511992-01-0101 January 1992
Rev 34 to Catawba Nuclear Station Odcm
ML20087F3931991-12-11011 December 1991
Public Version of Rev 13 to Crisis Mgt Implementing Procedure CMIP-11, Emergency Classification - Mcguire. W/ 920121 Release Memo
ML20086K8831991-11-18018 November 1991
Public Version of Revised Crisis Mgt Implementing Procedures (Cmips),Including Rev 42 to CMIP-1,Rev 29 to CMIP-2,Rev 42 to CMIP-5,Rev 12 to CMIP-11 & Rev 37 to CMIP-21
ML20086G7711991-10-16016 October 1991
Public Version of Revs to Crisis Mgt Implementing Procedures (Cmip),Including Rev 41 to CMIP-1,rev 37 to CMIP-4,rev 41 to CMIP-5,rev 46 to CMIP-6 & CMIP-7,rev 32 to CMIP-8,rev 40 to CMIP-9,delete CMIP-12 & Rev 24 to CMIP-13
ML20082C7441991-06-11011 June 1991
Public Version of Rev 13 to Crisis Mgt Implementing Procedure CMIP-12, Classification of Emergency for Oconee Nuclear Station
ML20076A7241991-06-10010 June 1991
Public Version of Crisis Mgt Implementing Procedures, Including Rev 39 to CMIP-1,rev 28a to CMIP-2,rev 44 to CMIP-7,rev 39 to CMIP-9 & Rev 10 to CMIP-11
ML20081J9841991-06-10010 June 1991
Rev 3 to EDA-1, Procedure for Estimating Food Chain Doses Under Post-Accident Conditions
ML20081K0101991-06-0606 June 1991
Rev 8 to EDA-3, Offsite Dose Projections for McGuire Nuclear Station
ML20072S0521991-03-15015 March 1991
Public Version of Revised Crisis Mgt Implementing Procedures,Including Rev 9 to CMIP-11, Classification of Emergency for McGuire Nuclear Station & Rev 11 to CMIP-12, Classification of Emergency for Oconee Nuclear Station
ML20066G3621991-02-0101 February 1991
Public Version of Crisis Mgt Implementing Procedures, Including Rev 28 to CMIP-2,Rev 34 to CMIP-4,Rev 38 to CMIP-5,Rev 43 to CMIP-6,Rev 42 to CMIP-7,Rev 29 to CMIP-8, Rev 37 to CMIP-9 & Rev 34 to CMIP-21
ML20082P7611991-01-0101 January 1991
Rev 30 to Odcm,Catawba Nuclear Station
ML20072S9621991-01-0101 January 1991
Public Version of Rev 12 to Crisis Mgt Implementing Procedure CMIP-12, Classification of Emergency for Oconee Nuclear Station
ML20082P7711991-01-0101 January 1991
Rev 31 to Odcm,Mcguire Nuclear Station
ML20072P9621990-11-0808 November 1990
Rev 9 to Crisis Mgt Implementing Procedure CMIP-12, Classification of Emergency for Oconee Nuclear Station
ML20028H2211990-10-31031 October 1990
Public Versions of Revised Crisis Mgt Implementing Procedures,Including Rev 37 to CMIP-1,Rev 27a to CMIP-2,Rev 33 to CMIP-4,Rev 37 to CMIP-5 & Rev 42 to CMIP-6
ML20059F4301990-08-22022 August 1990
Public Version of Rev 27 to Crisis Mgt Implementing Procedure CMIP-2, News Group Plan
ML20063Q2721990-08-14014 August 1990
Public Version of Revised Crisis Mgt Implementing Procedures,Including Rev 36 to CMIP-1,Rev 32 to CMIP-4,Rev 36 to CMIP-5,Rev 41 to CMIP-6,Rev 40 to CMIP-7,Rev 27 to CMIP-8,Rev 35 to CMIP-9 & Rev 7 to CMIP-11
ML20043F4841990-05-23023 May 1990
Public Version of Crisis Mgt Implementing Procedures, Consisting of Rev 20 to CMIP-13 & Deletion of CMIP-15 & CMIP-17
ML20043B2651990-05-0909 May 1990
Revised Crisis Mgt Implementing Procedures,Including Rev 35 to CMIP-1,Rev 26 to CMIP-2,Rev 31 to CMIP-4,Rev 35 to CMIP-5,Rev 40 to CMIP-6,Rev 39 to CMIP-7,Rev 26 to CMIP-8, Rev 33 to CMIP-9,Rev 2 to CMIP-14 & Rev 10 to CMIP-16
ML20043F4621990-04-20020 April 1990
Rev 5 to Oconee-specific Process Control manual.W/900606 Ltr
ML20006C0571990-01-18018 January 1990
Public Version of Rev 33 to Crisis Mgt Implementing Procedure CMIP-1, Recovery Manager & Immediate Staff & Rev 24 to CMIP-2, News Group Plan.
ML16152A8951990-01-0202 January 1990
Rev 33 to Public Version of Crisis Mgt Plan for Nuclear Stations.
ML15264A1571990-01-0202 January 1990
Revised Crisis Mgt Implementing Procedures,Including Rev 32 to CMIP-1,Rev 29 to CMIP-4,Rev 33 to CMIP-5,Rev 38 to CMIP-6,Rev 37 to CMIP-7,Rev 32 to CMIP-9,Rev 1 to CMIP-14 & Rev 30 to CMIP-21
ML20012A3801990-01-0101 January 1990
Rev 28 to, Offsite Dose Calculation Manual,Oconee,Mcguire & Catawba Nuclear Stations.
ML20012A3791990-01-0101 January 1990
Rev 27 to, Offsite Dose Calculation Manual,Oconee Nuclear Station.
ML20011D2441989-12-0101 December 1989
Crisis Mgt Implementing Procedures.
ML20012A3731989-11-15015 November 1989
Rev 4 to, Process Control Program Oconee Nuclear Station.
1999-08-18
[Table view]
Category:TEST REPORT
MONTHYEARML20212G1041987-01-0606 January 1987
Cycle 9 Startup Test Rept
ML20198S5141986-05-29029 May 1986
Station,Oconee 3,Cycle 9 Startup Testing Rept,Part 1 Zero Power Physics Test,Part II Power Escalation Test
ML20094C1511984-07-12012 July 1984
Extension of Retainer Lifetime to 4 Cycles
ML20087M4431984-03-26026 March 1984
Cycle 7 Startup Testing Rept
ML20040B0711981-12-31031 December 1981
Analysis of Capsule OCII-A from Duke Power Co Oconee Nuclear Station,Unit 2,Reactor Vessel Matl Surveillance Program.
ML20040B0681981-10-31031 October 1981
Analysis of Capsule OCIII-B from Duke Power Co Oconee Nuclear Station,Unit 3,Reactor Vessel Matls Surveillance Program.
ML20147B6681978-09-29029 September 1978
Integrated Leak Rate Test of Reactor Containment Bldg
ML19340A1351978-05-31031 May 1978
Unit 3,Cycle 3,Startup Testing Summary.
ML19316A5431978-04-30030 April 1978
Cycle 3 Startup Testing Summary.
ML19312C8071977-10-0606 October 1977
Integrated Leak Rate Test of Reactor Containment Bldg.
ML19340A1661977-03-0101 March 1977
Unit 3,Cycle 2,Startup Testing Summary.
ML19312C8481976-10-0101 October 1976
Jocassee Dam QA Test Results. List of Info Submitted on 761001 Encl
ML19316A1151976-06-22022 June 1976
Integrated Leak Rate Test of Reactor Containment Bldg.
ML19312C9261975-03-20020 March 1975
Reactor Bldg Post-Tensioning Sys Initial Surveillance.
ML19322C1301975-03-14014 March 1975
Startup Rept,Unit 3,for 750314.
ML19317F2811974-10-29029 October 1974
Reactor Bldg Post-Tensioning Sys,End Anchorage Surveillance.
ML19322C1711974-08-0909 August 1974
Startup Rept.
ML19322C1291974-08-0505 August 1974
Structural Integrity Test Rept of Reactor Containment Bldg. Prepared for Util
ML19316A4581973-11-16016 November 1973
Startup Rept.
ML19317E4791973-10-12012 October 1973
Rept of Testing of Two Atmosphere Supplying Suits.
ML19322C1681973-09-20020 September 1973
Structural Integrity Test Rept of Reactor Containment Bldg. Prepared for Util
ML19317F0811973-08-23023 August 1973
Reactor Coolant Flow Evaluation, Preliminary Rept
ML19322C1801973-02-15015 February 1973
Results of Oconee 1 Hot Functional Testing,Internals Vibration Monitoring Program.
ML19316A2401971-10-29029 October 1971
Structural Integrity Test Rept of Reactor Containment Bldg. Prepared for Util
ML19317F2451971-08-0505 August 1971
Integrated Leak Rate Test of Reactor Containment Bldg.
ML19316A2271970-07-14014 July 1970
Integrated Leak Rate Test of Reactor Containment Bldg. Prepared W/Technical Assistance by Bechtel Corp
ML19317E7741968-02-26026 February 1968
Keowee-Toxaway Project Jocassee Development;Summary of Matls Testing & Stability Design.
1987-01-06
[Table view] |
Text
' UNIT 1 E ,IOR COOLANT FLOW EVALUATION c '
Preliminary Report
- d.
_ Introduction .
Oconee Unit 1 was designed for a minimum primary coolant flow rate of 131.32x10 6 p:unds per hour. A greater flow rate than the minimum is expected, however.
Wile this will afford excess DNB protection, a flow rate of 110.8% design flow has been specified by the Babcock & Wilcox Company as the upper limit to avoid c:re lift at the end of life.
A test was performed during the Power Escalation Sequence at the 75% full p:ver plateau to verify that the magnitude of the primary system flow is within ceceptable limits.
The details of this test are delineated herein.
,F=aluation The ters.
basis of the flow calculation is a calorimetric around the two steam genera-Thermal-hydraulic data was monitored for an hour on July 29, 1973, properly i cveraged, and substitued into the heat balance equation described below to provide primary flow.
k Figure 1 is a schematic of a steam generator with its associated coolant flow '
1 cops; the dotted line represents the control. volute for the darivation of the cciorimetric equatien. Since the enersy entering the volume must leave it in scme form, the following balance for the A generator can be made. s 1(+ 1 =
1(+ + Id r
A similar equation exists for the B steam generator. Both can be scived for primary coolant system flow and are presented below.
~
P
=
(l( - Ih) + K^ ( 11 - If) + K
~
~
Precision thermocouples and dead-weight gages were installed on the feedwater and '
ateam lines to measure Precision manometers were used to measureemperatures and pressures to calculate enthalpies.
the pressure drop across the cali-brated Bailey flew nozzles for the'feedwater and steam flow determination. The plcat process computer was used to monitor the primary side temperatures and prcssures and feedvater temperature. -
Manometer readings were taken every two minutes for the duration of the test.
Stsam secondary side temperatures and feedwater pressures were recorded on a five minute interval while primary side temperatures and pressures arid feed- ~
-water temperature were =onitored on a 15 second basis. The data was averaged cnd'the flow and enthalapies were calculated. .
(
\ \
- .---_ - - - - - - - - ~~
..-7 -7 j 8001080 ( /(
The heat lors term rey ents tha surf ace radiation a- or convection from the (uricen, cf the piping and the steam generators. This term has minor signifi-ccnca but is included for completeness. Its magnitude is taken as 0.724 and 0.787,million BTU /hr for loops A and B, respectively.
Table 1 is a listing of the average values of the data collected during the test. The calculated enthalpics and flows are displayed in Table 2. The flow equation is shown below with the proper values inserted and the primary flow noted.
W = (1251.03 - 415.28) 4.0815 + 0.724
+ (1251.69 - 415.28) 3.9642 + 0.787 x 10 6 609.27 - 561.20
= 140.34 M lbm/Hr The error analysis for the above flow value is derived in Appendix A. The result of the error analysis yielded a band of + 1.146 M lbm/Hr.
Since minimum design flow is 131.32 M lbc/hr at rated power unich corresponds 6 to 130.2 M lba/hr at 75% power, the censured flow and experimental error is 107.8 + .82 as expressed in percent.
Safety /.nalysis The minimum RC system flow rate shall be the FSAR basis of the 100% (131.32 x 106 lb/hr, minimum design flow at rated pcver) plus 2.3% excess for bypass due to removal of 44 orifice plugs. This flow rate is established as the
_ minimum flow rate to meet the DNBR requirements stated in the FSAR. Therefore, the minimum flow shall be 134.34 x 100 lb/hr at rated power.
, . i The maximum reactor coolant system flow rate is 110.8% of the minimum design flow rate based on fuel assembly lift limitations. This 10.8% excess flow design limit is determined by utilizing experimental evidence of fuel assembly' hydraulic resistance characteristics and the maximum expected flow rate for any fuel assembly based on flow distributions from the Vessel Model Flow Test. This maximum allowable flow rate is bas'ed on the more limiting end-of-life conditions.
The measured system pressure loss is lower than predicted and represents a design conservatism. Also, the modification of the reactor vessel and internals resulted in a reduction c the reactor vessel unrecoverable pressure loss. The reduction in reactor vest 1 pressure loss due to the internals changes is approximately 4 psi at the design flow rate. (Reference BAW-10037, Rev. 2, November 1972, " Reactor Vessel Model Flow Tests.") These two points account for the actual RC system flow rate being above minimum l design flow rate. -
Therefore, the reactor coolant system flow including possible measur.cment error for Oconee 1 is within acceptabic limits.
! {
i! l l
1
l
.ABLE 1. AVERAGED DATA Loop A Loop B s
Main Steam, Temperature, 'F 590.34 590.80 Pressure, psia 911.73 912.22
' Feedwater, Temperature, 'F 436.47 436.25 Pressure, psia 942.61 939.00 AP,'ppi: Tap 1 35.64 35.25 Tap 2 35.95 33.32
. Hot Leg, Temperature, *F 596.60 596.86 Pressure, psia 2122.0 2141.7 Cold Leg, Temperature, 'F 560.997 560.945
- . Pressure, psia '
2089.4 2109.1 1
TABLE 2. HEAT BALANCE DATA Enthalpies (BTU /lbs) Loop A Loop B Hain Steam 1251.03 1251.69 Feedwater '
415.28 ,
415.28 Hot Leg 609.00 609.27 Cold Leg 561.31 561.20 Feedwater Flow Gi lbm/Hr) 4.0815 3.9642 Heat Losses 0.724 0.787 l
1 l
g6 e 4
I i
. t -.3-i 1
,, - ,, c > ~ =
9
~
, - TIGURE 1 LOOP i STEAM GENER' A 6
MP I!OT LEG W = Total Primary Coolant Flow p
M = Loop i Primary Flow 11 11 P f ~7 w 1
T g= Loop i ilet Leg Temperature / \ STEM I 1>4i P = Loop i llot Leg Pressurc $ T , P,
' i l li HEAT LCSSES g = Loop i llot Leg Enthalpy FEEDh'ATER i
Tc"L P i Cold Leg' Temperature ( l 1 \ ' I I P = Loop 1 Cold Leg Pressurc c FPi-F
.\ /
i N /
H = Loop i Cold Leg Enthalpy '
_.. c T
i- P i COLD LEG c c V
Tf=LoopiSteamTemperaturc
P, = Loop i Steam Pressure i
11 8
= Loop i Steam Enthalpy ,
i l Tp= Loop i Feedwater Temperature t 1 Py= Loop i Feedwater Pressure j i
i 11 7 = Loop i Feedwater Enthalpy My = Loop i Feedwater Flow
i K = Loop i lleat Losses
. l 4_
l~, i t
APPENDIX A The, basic flow equation from Figure 1 is as follows:
Wp = (1 -1 ) + }[ +
(H - 1() dp+ K 1(- - 11 or W = W(X , X ' * * *
y 2 n and dW = "
6W
$=1 6Xil Therefore dW = 1 -I d + d + d Ifg - II 1( - 1 _ (6 l !
6P^
)
F f_6th T A d T~ ~
p + 2[
61 dP A\
-1 6 A 6P p p }1
__ (I -I )k + 1 (0I dT^ + I dP^
(1 -1 ) A (6 6P
+ ~
} +
dT^
- * - + aP A
C (1 - Il^}
C 0 C
k
, dK H - 11F NF 1 - 11^
__ S l - 11 d[ + _ f6H
__S_ dTg B
+ 6H S
_ dP g B
g -H C \6T3 6P
~
dT + B 0 H -H F F C ( 61 6P dP),- dTH+ dP '
(1(-11 )2 6T 6B p
)
+ ("S
+ H dT + dP + dK (1 - 11 ) 2 _
} 6T 6P C) -1, O
h' t
- an b2 replaced by finite diff nces,AT[, representing
.The diffsrentials, dT The measurement
- th3 trcuret.ent. tolertr. for each variable'substituto tol;rcncca ara giv:n balows
'^ -
+ 0.5'F
- Main Steam Temperature .
, 4
- 1 psi Main Steam Pressure Feedwater Temperature + 0.5'F Feedwater Pressure
+ 1 psi i + 0.5%
Feedwater Flow RC Hot Leg Temperature + 0.25'F
+ 25 psi RC Hot Leg Pressure RC Cold Leg Temperature + 0.25'F RC Cold Leg Pressure + 35 psi Ambient Heat Losses +- 50%
for the feedwater flow and
! The heat balance The values data from forTable 2 isof the rate substituted change with respect to the differential j enthalpies.
are substituted for the partial derivation.
The terms of AWp are the squared, summed,and the square root taken. The terms represent the error in f eedwater flow, steam temperature, steam pres: 0:c, feedwater temperature, feedwater pressure, reactor coolant hot leg tey erature, reactor ecclant hot leg pressure, reactor coolant cold leg temperature, reactor coolant cold leg pressure, and ambient heat loss measurements.
4 %
4 e
e e- g
-t .
- 6-L_
