ML20154N562
| ML20154N562 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 09/30/1988 |
| From: | Odell L SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML20154N559 | List: |
| References | |
| ANF-87-161, NUDOCS 8809290341 | |
| Download: ML20154N562 (148) | |
Text
,
\\
ANF-87-161
.y,, n.
-s QT p,.
- ', t g,.
,7
\\3; ADVANCEDNUCLEARFUELS CORPORATION MILLSTONE UNIT 2 PLANT TRANSIENT ANALYSIS REPORT ANALYSIS OF CHAPTER 15 EVENTS SEPTEMBER 1988 hhR D
00 36 p
\\
ADVANCEDNUCLEARFUELS CORPORATION
)
ANF-87-161 Issue Date: 9/9/88
)
i l
MILLSTONE UNIT 2 PLANT TRANSIENT ANALYSIS REPORT -
ANALYSIS OF CHAPTER 15 EVENTS o
Prepared by vn nad$
' L. D. O' Dell, Team Ldhder PWR Safety Analysis Licensing & Safety Engineering Fuel Engineering & Technical Services Contributor: N. N. Nesselbeck (ENSA, Inc.)
September 1988 9f
1 1
CUSTOMER DISCLAIMEft It0PORTANT M371CE REGAR0tNG CONTENTS AND USE OF THts
~
00CUh8ENT rLEASE READ CAREFULLY 1
Advanced Nucteer Fue6e Corporacon's warrantes and reoresentatens con-Corneng the eWO;ect matter c' this document are those set form in the Agreement
]
tenween Advanced Nucay Fusie Corporaten and the Customer pursuant to which thee document is issued. Accordmgly, except as omervnse expressly pro.
need in such Agreement, norther Advanced NucJear Fuede Corporaten nor any i
person acnng on its bena#f menos any warrenty or representation, expressed or w9phed, with respect to the accuracy, comp 6etenees, or usefulness of the efor.
menon contened in me document, or that the use of any eformaten, accarstus, memod or process a m in mis document wul not etnnge onvateey owned nghts: or assumes any tacehtee wem resoect to tne use of any mformaton, ao-paratus, memoa or process discioned in mis document The informaten contamed heren is for me so6e use of Customer.
in order to avoed imoestment of ngnte of Advanced Nuc! oar Fume Corporaten m patents or inventrns wesen may ee ecluded in tne informaten contamed in this cocument, tN rocceent. by de accoolance of inte document, agrees not to putWeen or make pubiic use On me patent use of the ttfm)of sucn eformation untd to authorued in wrtog by Acranced Nuclear Fue4s Corporaten or untd after six (6) monme followeg terminacon or expiraten of the aforesaid Agreerrent and any extensen thoroof, unsees otrervnse expressly proviced c me Agreement. No ngnte or licensee in or to any parents are impimed by me fumisning of this docu-mont.
I ANF 314SM2A (12/87) 4
s I
j ANF-87-161 Page i
(
TABLE OF CONTENTS i
l Section EA9.t
1.0 INTRODUCTION
1 2.0
SUMMARY
OF DISPOSITION OF EVENTS................
4 3.0 BASIS AND JUSTIFICATION FOR DISPOSITION OF EVENTS 11 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM........
14 15.1.1 Decrease in Feedwater Temperature 14 15.1.2 Increase in Feedwater Flow...................
15 15.1.3 Increase in Steam Flow.....................
17 15.1.4 Inadvertent Opening of a Steam Generator Relief 19 or Safety Valve 15.1.5 Steam System Piping Failures Inside and Outside o f Con t a i nment.........................
20 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM........
33 15.2.1 Loss of External Load 33 15.2.2 Turbine Trip..........................
34 15.2.3 Loss of Condenser Vacuum....................
36 15.2.4 Closure of the Main Steam Isolation Valves (MSIVs).......
36 15.2.5 Steam Pressure Regulator Failure................
38 15.2.6 Loss of Nonemergency A.C. Power to the Station Auxiliaries...
39 15.2.7 Loss of Normal Feedwater Flow 39 J
15.2.8 Feedwater System Pipe Breaks Inside and Outside Containment 41 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW 54 15.3.1 Loss of Forced Reactor Coolant Flow 54 15.3.2 Flow Controller Malfunction 55
r ANF-87-161 Page 11 TABLE OF CONTENTS '(Cont.)
Section EASA 15.3.3 Reactor Coolant Pump Rotor Seizure...............
56 15.3.4 Reactor Coolant Pumo Shaf t Break................
57 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 64 15.4.1 Uncontrolled Control Rod / Bank Withdrawal From a Subcritical or low Power Startup Condition...........
64 15.4.2 Uncontrolled Control Rod / Bank Withdrawal at Power 65 15.4.3 Control Rod Misoperation....................
67 15.4.4 Startup of an inactive Loop 72 15.4.5 Flow Controller Malfunction 73 15.4.5 CVCS Malfunction That Results in a Decrease in the Boron Concentration in the Reactor Coolant..............
73 15.4.7 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position 74 15.4.8 Spectrum of Control Rod Ejection Accidents...........
74 15.4.9 3pectrum of Rod Drop Accidents (BWR)..............
76 15.5 INCREASES IN REACTOR COOLANT SYSTEM INVENTORY 98 15.5.1 Inadvertent Operation of the ECCS That Increases Reactor Coolant Inventory 98 15.5.2 CVCS Malfunction That Increases Reactor Coolant Inventory 98 15.6 DECREASES IN REACTOR C0OLANT INVENTORY,...........
101 15.6.1 Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve 101 l
15.6.2 Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside of Containment 102 15.6.3 Radiological Consequences of Steam Generator Tube Failure 102 1
1
I i
ANF-87-161 i
j Page 111 i
TABLE OF CONTENTS (Cont.)
Section
'f.Allt 15.6.4-Radiological Consequences of a Fain Steam Line Failure-Outside Containment (BWR)
................... 105 15.6.5 Lt - > rf Coolant ' Accidents Resulting from a Spectrum of Postul M ed Piping Breaks Within the Reactor Coolant Presse e Boundary 105 15.7 RADIOACliVE RELEASES FROM A SUBSYSTEM OR COMPONENT......
117 15.7.1 Waste Gas System Failure...................
117 15.7.2 Radioactive Liquid Waste System Leak or Failure (Release to Atmosphere) 117 t
15.7.3 Postulated Radioactive Releases Due to Liquid-Containing Tank Failures 117 15.7.4 Radiological Consequences of Fuel Handling Accident 117 l
15.1.5 Spent Fuel Cask Drop Accidents................
118 i
4.0 MILLSTONE UNIT 2 FSAR EVENTS NOT CONTAINED IN THE STANDARD l
REVIEW PLAN 124 l
4.1 EFFECTS OF EXTERNAL EVENTS..................
124 4.2 FAILURES OF EQUIPMENT PROVIDING JOINT CONTROL AND SAFETY FUNCTIONS
........................... i25 4.3 CONTAINMENT PRESSURE ANALYSIS
................125 4.4 HYDROGEN ACCUMULATION IN CONTAINMENT.............
126 t
4.5 RADIOLOGICAL CONSEQUENCES OF THE DESIGN BASIS INCIDENT (081).
128 1
5.0 REFERENCES
134 l
1
)
t I
l i
2 l
ANF-87-161 Page iv i
LIST OF TABLES I
Table P_ng 1.1 Reactor Operating Modes for Millstone Unit 2..........
3 5
2.1 Disposition of Events Summary 15.1.1-A Available Reactor Protection for Decrease in 23 Feedwater Temperature Event 15.1.1-B Disoosition of Events for the Decrease in 24 Feedwater Temperature Event 15.1.2-A Available Reactor Protection for the Increase in Feedwater Flow Event......................
25 15.1.2-8 Disposition of Events for the Increase in Feedwater Flow Event......................
26 15.1.3-A Available Reactor Protection for the Increase in Steam Flow Event..................
27 15.1.3-8 Disposition of Events for the Increase in Steam Flow Event 28 15.1.4-A Available Raactor Protection for the Inadvertent Opening of a Steam Generator Relief or Safety Valve Event 29 15.1.4-8 Disposition of Events for the Inadvertent Opening of a Steam Generator Reitef cr Safety Valve Event 30 15.1.5-A Available Reactor Protection for Steam System Piping Failures Inside and Outside of Containment Event 31 15.1.5 B Disposition of Events for Steam System Piping Failures Inside and Outside of Containment Event........ 32 i
15.2.1-A Available Reactor Protection for the Loss of External Load Event 42 15.2.1-B Disposition of Events for the loss of External Load Event.....................
43 l
15.2.2-A Available Reactor Drotection for the I
turbine Trip Event.......................
44 i
l ANF-87-161 Page v N
LIST OF TABLES (Cont.)
IAhlt EA9A 15.2.2-8 Disposition of Events for the Turbine Trip Event.......................
45 i
15.2.3-A Available Reactor Protection for the loss of Condenser Vacuum Event.................
46 15.2.3-8 Disposition of Events for the loss of Condenser Vacuum Event.................
46 15.2.4-A Available Reactor Protection for the Closure of the MSIVs Event...................
47 15.2.4-B Disposition of Events for the Closure of the MSIVs Event...................
48 15.2.5 A Available Reactor Protection for the Steam Pressure Regulator Failure Event.............
49 15.2.5-B Disposition of Events for the Steam Pressure Regulator Failure Event.............
49 l
15.2.6-A Available Reactor Protection for the Loss of Nonemergency A.C. Power to the Station Auxiliaries Event 50 15.2.6 B Disposition of Events for the Loss of Nonemergency A.C. Power to the Station Auxiliaries Event................
50 15.2.7-A Available Reactor Protection for the Loss of Normal Feedwater Flow Event 51 15.2.7 B Disposition of Events for the loss of Normal Feedwater Flow Event 52 j
15.2.8 A Available Reactor Protection for the Feedwater System Pipe Breaks Inside and Outside Containment Event..............
53 15.2.8-B Disposition of Events for the Feedwater System Pipe Breaks Inside and Outside Containment Event 53
w 4
ANF-87-161 Page vi LIST OF TABLES (Cont.)
f Table Eagg 15.3.1-A Available Reactor Protection for the Loss of Forced Reactor Coolant Flow Event 58 15.3.1-B Disposition of Events for the loss of Forced Reactor Coolant Flow Event 59 15.3.2-A Available Reactor Protection for the Flow Controller Malfunction Event 60 15.3.2-B Disposition of Events for the Flow Controller Malfunction Event 60 15.3.3 A Available Reactor Protection for the Reactor Coolant Pump Rotor Seizure Event.............
61 15.3.3-B Disposition of Events for the Reactor Cool ant Pump Rotor Seizure Event................
62 15.3.4-A Available Reactor Protection for the Reactor Coolant Pump Shaft Break Event.............
63 15.3.4-B Disposition of Events for the Reactor Coolant Pump Shaft Break Event.................
63 15.4.1-A Available Reactor Protection for the Uncontrolled Control Rod / Bank Withdrawal from a Subcritical or low Power Startup Condition Event..............
77 15.4.1 B Disposition of Events for the Uncontrolled Control Rod / Bank Withdrawal from a Subcritical or Low Power Startup Condition Event..............
78 15.4.2-A Available Reactor Protection for the Uncontrolled Control Rod / Bank Withdrawal at Power Event...........
79 15.4.2 B Disposition of Events for the Uncontrolled Control Rod / Bank Withdrawal at Power Event...............
80 15.4.3(1) A Available Reactor Protection for the Dropped Control Rod / Bank Event 81 15.4.3(1)-B Disposition of Events for the Dropped Control Rod / Bank Event 82
(
i e
ANF-87-161 Page vii J
I LIST OF TABLES (Cont.)
IAhlt EAat 15.4.3(2)-A Available Reactor Protection for the Dropped Part-Length Control P.od Event............
83
]
15.4.3(2)-B Disposition of Events for the 7
Dropped Part-Length Control Rod Event............
83 15.4.3(3)-A Available Reactor Protection for the Malpositioning of the Part-Length Control Rod Group Event 84 15.4.3(3) B Disposition of Events for the Ma1 positioning of the Part-Length Centrol Rod Group Event 84 l
4 15.4.3(4)-A Available Reactor Protection for the Static' ally Misaligned Control Rod / Bank Event..............
85 4
15.4.3(4) B Disposition of Events for the Statically Misaligned Control Rod / Bank Event..............
85 15.4.3(5)-A Available Reactor Protection for the Single Control Rod Withdrawal Event 86 15.4.3(5)-B Disposition of Events for the Single Control Rod Withdrawal Event 87 3
15.4.3(6) A Available Reactor Protection for the Reactivity Control Device Removal Error I
During Refueling Event 88 15.4.3(6)-B Disposition of Events for the Reactivity Control Device Removal Error During Refueling Event 88 a
j 15.4.3(7) A Available Reactor Protection for the Variations in Reactivity Load to be Compensated by Burnup
]
or On Line Re. fueling Event 89 15.4.3(7)-8 Disposition of Events for the Variations in Reactivity Load to be Compensated by Burnup
(
or On Line Refueling Event 89 i
15.4.4 A Availablo Reactor Protection for the Startup of i
an Inactive Loop Event.....................
90 l
i 1
1 I
ANF-87-161 Page viii LIST OF TAj.LES (Cont.)
)
f Table Eagt 15.4.4-B Disposition of Events for the Startup of an 90 A
Inactive Loop Event 15.4.5-A Available Reactor Protection for the Flow Controller Mal function Event..................
91 15.4.5-B Disposition of Events for the Flow Controller Mal function Event..................
91 i'
15.4.6 A Available Reactor Protection for the CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Cool ant Event..................
92 1
15.4.6 8 Disposition of Events for the CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant Event..................
93 15.4.7-A Available Reactor Protection for the Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position Event 94 15.4.7-B Disposition of Events for the Inadvartent Loading and Operation of a Fuel Assembly in an Improper Position Event...........
94 15.4.8 A Available Reactor Protection for the Spectrum of t
Control Rod Ejection Accidents.................
95 15.4.8 B Disposition of Events for the Spectrum of Control Rod Ejection Accidents.................
96 15.4.9 A Available Reactor Protection for the i
Spectrum of Rod Drop Accidents (BWR)..............
97 15.4.9 8 Disposition of Events for the Spectrum of Rod Drop Accidents (BWR) 97 15.5.1-A Available Reactor Protection for the Inadvertent Operation of the ECCS that Increases Reactor Coolant Inventory Event 99 15.5.1 B Disposition of Events for the Inadvertent Operation of the ECCS that Increases Reactor Coolant Inventory Event...
99 I
l
]
)
ANF-87-161 Page ix LIST OF TABLES (Cont.)
(
IAhlt I'192
{
15.5.2 A Available Reactor Protection for the CVCS Halfunction that Increases Reactor Coolant Inventory Event 100 e
15.5.2-B Disposition of Events for the CVCS Halfunction that Increases Reactor Coolant Inventory Event........
100 15.6.1-A Available Reactor Protection for the Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve Event.......
109 15.6.1 B Disposition of Events for the Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve Event 110 15.6.2 A Available Reactor Protection for the Radiological j
Consequences of the Failure of Small Lines Carrying Primary Coolant Outside of Containment Event.........
111 15.6.2 B Disposition of Events for the Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside of Containment Event.........
111 15.6.3 A Available Reactor Protection for the Radiological Consequences of Steam Generator Tube Rupture Event......
112 15.6.3-B Disposition of Events for the Radiological Consequences of Steam Generator Tube Rupture Event......
113 15.6.4 A Available Reactor Protection for the Radiological Consequences of a Main Steam Line Failure Outside Containment (BWR) Event...................
114 15.6.4-8 Disposition of Events for the Radiological Consequences of a Main Steam Line Failure Outside Containment (BWR) Event 114 15.6.5-A Available Reactor Protection for the loss of Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary......................
115 15.6.5 B Disposition of Events for the loss of Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary.........
116
ANF-87-161 Page x
)
I LIST OF TABLES (Cont.)
Table Elst i
15.7.1-A Available Reactor Protection for the Waste Gas System Failure Event...................
119 15.7.1-B Disposition of Events for the Waste Gas System Failure Event....................
119 15.7.2-A Available Reactor Protection for the Radioactive Liquid Waste System Leak or Failure (Release to Atmosphere) Event 120
)
15.7.2-8 Disposition of Events for the Radioactive Liquid Waste System Leak or Failure (Release to Atmosphere) Event.....................
120 15.7.3-A Available Reactor Protection for the Postulated Radioactive Releases Due to Liquid-Containing Tank Failures Event
. 121 15.7.3-8 Disposition of Events for the Postulated 5
Releases Due to Liquid-Containing Tank Failures Event 121 15.7.4 A Available Reactor Protection for the Radiological Consequences of Fuel Handling Accidents 122 15.7.4 8 Disposition of Events for the. Radiological Consequences of Fuel Handling Accidents 122 15.7.5 A Available Reactor Protection for the Spent Fuel Cask Drop Accidents 123 l
15.7.5 8 Disposition of Events for the Spent Fuel Cask Drop Accidents 123 4.1 A Available Reactor Protection for the Effects of External Events..................
129 i
4.1 B Disposition of Events for the i
Effects of External Events..................
129 I
- ~.
I
\\
ANF-87-161 Page xi LIST OF TABLES (Cont.)
lib.11 P.Aat
(
i 4.2 A Available Reactor Protection for the Failures of Equipment Providing Joint Control. and Safety Functions Event 130 4.2-B Disposition of-Events for the Failures of Equipment Providing Joint Control and Safety Functions Event 130 4.3-A Available Reactor Protection for the Containment Pressure Analysis Event
..............~131 i
4.3 B Disposition of Events for the Containment Pressure Analysis Event 131 4.4-A Available Reactor Protection for the Hydrogen Accumulation in Containment Event 132 l
4.4 8 Disposition of Events for the Hydrogen Accumulation in Containment Event..............
132 l
4.5-A Available Reactor Protection for the Radiological t
i Consequences of the Design Basis Incident Event 133 4.5 B Oisposition of Events for the Radiological l
Consequences of the Design Basis Incident Event 133 i
1
[
ANF-87-161 Page 1
1.0 INTRODUCTION
This report provides a review of the Standard Review PlanII) Chapter 15 Events for Millstone Unit 2.
The event review is being performed to support the first reload of Advanced Nuclear Fuels Corporation (ANF) fuel. The review i
also supports Technical Specification changes required to increase plant operating flexibility (e.g., longer cycles).
In accordance with ANF methodology (2) all events described in the Standard Review Plan (SRP) have been reviewed and placed (dispositioned) into one of the following four categories:
(1) The event initiator or controlling parameters have been changed from the analysis of record so that the event needs to be reanalyzed for the current licensing action; (2) The event is bounded by another event which is to be reanalyzed; (3) The event causes and principal variables which control the results of the event are unchanged from or bounded by the analysis of record; or (4) The event is not in the licensing basis for the plant.
The review of the current plant safety analysis for the event disposition process incorporated the following revisions to acceptable plant operating conditions:
(1)
Increased maximum rad % peaking factor.
(2)
Extension of the cycle length to 18 months.
(a)
Increased both positiva and negative bounds on the moderator temperature coefficient.
(b)
Increased shutdown margin requirements.
(3) Operation over an inlet temperature range.
e ANF-87-161 Page 2
'I Section 2.0 provides a summary of the event disposition.
In order to facilitate review, the events are numbered in accordance with the SRP and cross referenced to the pertinent updated FSAR sections.
Section 3.0 presents
}
the results of the analysis and justifications.
Section 4.0 presents the references used in this report.
j In the event disposition, all of the reactor operating conditions allowed by the plant Technical Specifications (3) are examined to insure that the bounding subevents are identified for each SRP event category.
This insures that the subsequent safety analysis will support the complete range of allowable operating conditions.
The reactor operating modes allowed for Millston1 Unit 2 by the plant Technical Specifications are listed in Table 1.1.
i
(
l e
e i
l i
1 Table 1.1 Reactor Operatin9 Modes for Millstone Unit 2 Reactivity
% Rated Average Coolant Mode Condition. K Thermal Power
- Temperature ggy 1.
Power Operation
> 0.99
> 5%
> 300*F 2.
Startup
> 0.99 1 5%
> 300*F-3.
Hot Standby
< 0.99 C
> 300*F 4.
Hot Shutdown
< 0.99 0
300*F > T**9
> 200*F 5.
Cold Shutdown
< 0.98 0
5 200*F 6.
Refueling **
s 0.95 0
5 140*F l
l l
l l
l l
Excluding decay heat.
2-
- Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with M
the head removed.
45Y
- E w~
s ANF-87-161 Page 4 2.0 SufMARY OF DISPOSITION OF EVENTS I
Table 2.1 presents a sumary of results of the event dispJsition. In accordance with ANF methodology, the events are placed in one of the four categories identified in Section 1.0.
l i
i F
I 1
[
l i
i
}
2 1
d i
Table 2.1 Disposition of Events Summary SRP Event Event Updated Classifi-Desig-Bounding FSAR Difoosition.
Event Desianation f
cation nation Name 15.1
' INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.1 Decrease in Feedwater Temperature Bounded 15.1.3 14.1.4, 14.11.2 i
15.1.2 Increase in Feedwater Flow l
1)
Power Bounded 15.1.3 14.1.4, 14.8, 14.11.2 2)
Startup Bounded 15.1.3 15.1.3 Increase in Steam Flow Analyze 14.8, 14.11.1 l
15.1.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve Bounded 15.1.3 14.8, 14.11.1 15.1.5 Steam System Piping Failures Inside and Outside of Containment Analyze 14.1.4, 14.12 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 14.9 15.2.1 Loss of External Load Analyze 15.2.2 Turbine Trip Bounded 15.2.1 14.9-15.2.3 Loss of Condenser Vacuum Not in Licensing Basis k
15.2.4 Closure of the Main Steam Isolation Valves (MSIVs)
Analyze 14.8 os EY 15.2.5 Steam Pressure Regulator Failure Not applicable;
- g BWR Event
Table 2.1 Disposition of Events Susunary (Cont.)
SRP Event Event Updated Classifi-Desig-Bounding FSAR cation nation Name Disposition Event Desionation 15.2.6 Loss of Nonemergency A.C. Power to the 'tation Auxiliaries Not in Licensing Basis 15.2.7 Loss of normal Feedwater Flow Analyze 14.8, 14.10, 14.A 15.2.8 feedwater System Pipe Breaks Inside and Outside Containment.
Not in Licensing Basis 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW 15.3.1 Loss of Forced Reactor Coolant Flow Analyze 14.6 15.3.2 Flow Controller Malfunction Not Applicable 15.3.3 Reactor Coolant Pump Rotor Seizure Analyze 14.6 15.3.4 Reactor Coolant Pump Shaft Break Bounded 15.3.3 14.6 l
15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1 Uncontrolled Control Rod / Bank Withdrawal from a Subcritical or Low Power Condition Analyze 14.2 y
o co 15.4.2 Uncontrolled Control Rod / Bank a?
Withdrawal at Power Analyze 14.2
- J 15.4.3
- Control Rod Misoperation
- 1) Dropped Control Rod / Bank Analyze 14.4
- 2) Dropped Part-Length Control Rod Not Applicable 14.4
Table 2.1 Disposition of Events Summary (Cont.)
SRP Event Event Updated Classifi-Desig-Bounding FSAR cation nation h
Disposition Event Desianation
- 3) Ma1 positioning of the Part-Length Control Rod Group Not Applicable 14.4.5
- 4) Statically Misaligned Control Rod /8ank Not in Licensing Basis
- 5) Single Control Rod Withdrawal Analyze 14.2
- 6) Reactivity Control Device Removal Error During Refueling Not Applicable 14.1.4
- 7) Variations in Reactivity Load I
to be Compensated by Burnup or On-Line Refueling Not Applicable 14.1.4 15.4.4 Startup of an Inactive loop Not App.icable 14.7 15.4.5 Flow Controller Malfunction Not applicable;-
No Flow Con-i troller 15.4.6 CVCS Malfunction that Results
- Analyze, 14.3 l
in a Decrease in the Boron Con-Modes 1-6 centration in the Reactor Coolant
'15.4.7 Inadvertent Loading and Operation Not in Licensing Basis of a Fuel Assembly in an Improper E'
I Position 7'
25
%L J
=
Table 2.1 Disposition of Events Summary (Cont.)
l SRP l
Event Event Updated Classifi-Desig-Bounding FSAR cation __
nation Name Disposition Event Desianation 15.4.8 Spectrum of Control Rod Ejection Analyze 14.13 Accidents l
l 15.4.9 Spectrum of Rod Drop Not applicable; l
Accidents (BWR)
BWR Event 15.5 INCREASES IN REACTOR COGLANT INVENTORY i
l l
15.5.1 Inadvertent Operation of the Not in Licensing Basis ECCS that Increases Reactor Coolant Inventory 15.5.2 CVCS Malfunction that In-Not in Licensing Basis creases Reactor Coolant Inventory l
15.6 DECREASES IN REACTOS COOLANT INVENTORY l
15.6.1 Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve Analyze 14.5 15.6.2 Radiological Consequences of the Not Applicable 14.1.4 Failure of Small Lines Carrying Primary Coolant Outside of Con-tainment g
n 15.6.3 Radiological Consequences of Bounded 14.14 14.14 ms3 Steam Generator Tube failure ds1"
- U~
15.6.4 Radiological Consequences of a Not applicable; Main Steamline Failure Outside BWR Event i
Containment
Tabic-2.1 Disposition of Events Summary (Cont.)
SRP Event Event Updated Classifi-Desig-Sounding FSAR cation nation M4mg Disoosition Event Desianition 15.6.5 Loss of Coolant Accidents Analyze 14.15.3, 14.15.4 Resulting from a Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary 15.7 RADIDACTIVE RELEASE FROM A SUBSYSTEM OR OMPONENT 15.7.1 Waste Css System failure Bounded 14.17 14.17 15.7.2 Radioactive Liquid Waste System Leak or failure (Release to Atcosphere)
Not in Licensing Basis 15.7.3 Postulated Radioactive Releases Not in Licensing Basis due to Liquid-Containing Tank Failures 15.7.4 Radiological Consequences of Fuel Bounded 14.19 14.19 l
Handling Accidents i
15.7.5 Spent fuel Cask Drop Accidents Not in Licensing Basis l
2
=i ma>
3?
- 3" e-c
Table 2.1 Disposition of Events Sumary (Cont.)
SRP Event Event Updated Classift-Desig-Bounding FSAR cation nation flame Disposition Event Desionation FSAR EVEfiTS fiOT 00flTAltiED Ill IllE STAllDARD REVIEW PLAtt (1)
Effects of External Events Bounded 14.21 14.21
~
(2)
Failures of Equipment Which Bounded 14.1.4 14.1.4 Provide Joint Control / Safety functions (3)
Contai.e nt Pressure Analysis Bounded 14.16 14.16 (4) flydrogen Accumulation in Bounded 14.18 14.18 Containment (5)
Radiological Consequences of Bounded 14.20 14.20 the Design Basis Incident (DBI)
')
W sE
l ANF-87-161 3.0 BASIS AND JUSTIFICATION FOR DISPOSITION OF EVENTS This section presents the basis and justification for the disposition of
- events.
The section numbers and event names are in accordance with those events described in the SRP.
Each event described in the SRP (and events in the FSAR but not in the SRP) is considered in accordance with the plant licensing basis and dispositioned inti one of the four categories described in Section 1.0.
Events which are not bounded by other events or by existing accepted analyses, and are in the plant licensing basis, are dispositioned to be analyzed.
In the event disposition process, the event initiator is identified for each event. The magnitude of the initiator for each event is calculated and compared to the magnitude of the initiator for other events.
The comparison basis includes all the plant operating conditions.
This allows, in several cases, a ranking of the event initiators as to severity, allowing the lesser events to be dispositioned a-bounded by the greater event.
Similar logic is applied in determination of the applicability and bounding nature for existing accepted analyses.
The licensing basis for Millstone Unit 2 is as stated in the Updated Final Safety Analysis Report.(4)
The formulation of event scenarios to be considered in the safety analysis depends on single failure criteria established by the plant licensing basis.
Examination of the Millstone 2 licensing basis yields the following single failure criteria:
(1)
The' Reactor Protection System (RPS) is designed with redundancy and independence to assure that no single failure or removal from service of any component or channel of a system will result in the loss of the protection function.
(2)
Each Engineered Safety Feature (ESF) is designed to perform its intended safety function assuming a failure of a single active component.
ANF-87-161 Page 12 (3) The onsite power system and the offsite power system are designed such that each shall independently be capable of providing power for the ESF assuming a failure of a single active component in either power system.
The safety analysis is structured to demonstrate that the plant systems design satisfies these single failure criteria.
The following assumptions result:
(1) The ESFs reo,uired to function in an event are assumed to suffer a worst single failure of an active component.
(2) Reactor trips occur at the specified setp int within the specified delay time assuming a worst single active failure.
(3) The following postulated accidents are considered assuming a concurrent loss of offsite power; main steam line break, control rod ejection, and LOCA.
(4) The loss of normal feedwater, an anticipated operational occurrence, is analyzed assuming a concurrent loss of offsite power.
The requirements of 10 CFR 50, Appendix A, Criteria 10, - '
25 and 29 require that the design and operation of the plant and the reactor protective system assure that the Specified Acceptable fuel Design Limits (SAFDLs) not be exceeded during Anticipated Operational Occurrences (A00s).
As per the definition of A00 in 10 CFR 50, Appendix A.
"Anticipated Operational Occurrences mean those conditions of normal operation which are expected to occur one or more times during the life of the plant and include but are not limite<1 to loss of power to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power." The Specified Acceptabla fuel Design Limits (SAFDLs) are that: 1) the fuel shall not experience centerlir.c melt (-21 W/ft); and 2) the departure from nucleate boiling ratio (DNBR) shall have a minimum allcwable limit such
ANF-87-161 Page 13 that there is a 95% probability with a 95% confidence interval that departure from nucleate boiling (DNB) has not occurred (XNB DNBR of 1.17).
As indicated, three revisions to acceptable plant operating conditions are planned.
The three revisions are:
(1) The maximum Technical Specification radial peaking factor is being increased from the current limit of 1.537 to 1.61.
l (2) Plant cycle length is being increased from 12 to 18 month cycles.
(a) The moderator temperature coefficient (MTC) in the Technical Specification is being changed in order to accommodate the increased cycle length.
The HTC change for power less than or equal to 70%
is from +5 to +7 pcm/*F.
The HTC change for power greater than 70%
is from +4 2 MTC 1 -24 ocm/*F to +4 A MTC 2 -28 pcm/'F.
(b) The shutdown margin requirements are being changed to offset the more negative end of cycle MTC.
The change in MTC results in the modes 1, 2, 3 and 4 shutdown margin going from the current limit of 2 2.9% to 2 3.6%.
(3)
In order to cover end of cycle coastdowns, the analysis will also support plant operation at reduced inlet temperatures.
The current nominal inlet temperature is 549'F and the analysis will support up to a 12*F inlet temperature reduction at full power.
Greater temperature reduction is acceptable if concurrent with reduced power and pressurizer level during an EOC coastdown.
The event review and event analyses will be performed to insure that thes) revisions to acceptable plant operating conditiens are supported.
ANF-87-161 Page 14 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.1 Decrease in Feedwater Temoerature 15.1.1.1 Event Initiator A decrease in feedwater temperature may be caused by loss of ono of several feedwater heaters.
The loss could be due to the interruption of steam extraction flow or to an accidental opening of a feedwater heater bypass line.
The worst case loss of feedwater haaters would occur if all of the low pressure heaters were bypassed.
The effects of any decrease in the feedwater temperature due to flow increases (Main or Auxiliary Feedwater) are discussed in Section 15.1.2.
i l
15.1.1.2 Event Description - Due to a malfunction in the feedwater heater system, the enthalpy of the feedwater being injected into the steam generators is reduced.
The increasel subcooling of the feedwater reduces the secondary l
system average fluid enthalpy and increases the energy removal rate from the l
primary system.
The increase in primary to secondary heat transfer causes the reactor coolant temperature at the outlei of the steam generator to decrease.
l This causes a corresponding decrease in the core inlet coolant temperature.
With a negative moderator reactivity temperature coefficient, the reactor core power will begin to increase as the cooler moderator fluid reaches the core.
15.1.1.3 Reactor protection - Reactor protection is provided by the variable overpower, thermal margin / low pressure, local power density, and low steam generator pressure trips.
Reactor protection for the decrease in feedwater temperature event is summarized in Table 15.1.1-A.
15,1,1.4 Disoosition and Justification For operating Modes 1-3, the response of the nuclear steam supply system is governed by the magnitude of the overcooling introduced by the initiating event.
There is no extraction to the feedwater heaters for operating Modes 4-6.
As such, there is not a credible event for these reactor operating conditions.
l l
l l
I ANF-87-161 l
Page 15 l
The most limiting case for Mode 1 is from rated power conditions because the feedwater flow rate and heater duty decrease with load.
Also, at rated power conditions, the initial DNBR margin is minimized.
The consequences of the event in Model 2 av.i 3 are bounded by those of Mode 1 because the magnitude of tha itsf uati.r.( event in Modes 2 or 3 is much smaller than in Mode l
1.
l This cc + ; 9t
' to bypassing the feedwater heaters is bounded by
' M ceu' r the limiting event postulated in Section 15.1.3.
the cooldown a
As such, the cu xqucrcs of the Increase in Steam Flow event (15.1.3) bound the consequences foi the Decrease in Feedwater Temperature event discus, sed in this section (15.1.1).
The disposition of events for the Decrease in Feedwater Temperature event is st"nmarized in Table 15.1.1-B.
15.1.2 Increase in Feedwater Flow 15.1.2.1 Event Initiator - This event is initiated by a failure in the feedwater system which causes an increase in the feedwater flow to the steam j
generators.
The initiators considered are complete. opening of a feedwater control valve, overspeed of the feedwater pumps, inadvertent start of a second feedwater pump at low power, startup of the auxiliary feedwater system, and inadvertent opening of the feedwater control valve bypass line.
The increased flow to the steam generators 15.1.2.2 Event Descriotion causes an increase in the energy renoval capability of the steam generators by reducing the average fluid enthalpy in the steam generators.
The increased energy removal from the primary system causes the reactor coolant temperature at the outlet of the steam generator to decrease.
The core inlet temperature will correspondingly be reduced, which will cause the core power to increase if the moderator temperature coefficient is negative.
Because this event is characterized as a primary system overcooling event, the primacy system pressure initially decreases along with the core
--n..,
n.,..
....,c-
. _ _ _,,, -.. -. - ~ - -..
ANF-87-161 Page 16 l
inlet temperature.
There is also a ;:ossibility for a core power increase in the presence of a negative moderator reactivity feedback coefficient.
Increased reactor power rer:uces the core DNB margin.
A potential exists that the net effect of tht.se three factors will represenc a challenge to tha core DNB margin.
15.1.2.3 Reactor protection - Reactor protection for the rated power and power operation conditions (Mode 1) is provided by the variable overpower trip, local power density trip, thermal margin / low pressure trip, low steam generator pressure trip, and reactor trip on turbine trip due to high steam generator water level.
For Modes 2 and 3, protection is provided by the low steam generator pressure trip, safety injection actuation signal, and the variable overpower trip.
Reactor protection for the increase in Feedwater flow event is summarized in Table 15.1.2-A.
15.1.2.4 Discosition and Justification - The event consequences at ' rated power operating conditions will bound the consequences from all other power operating conditions.
At rated power operating conditions, the initial thermal margin (DNBR) is minimized. Maximizing the increase in feedwater flow maximizes the load demand.
This results in the maximum rate of moderator cooldown which, in the presence of a
negative moderator temperature coefficient, results in the maximum challenge to the specified acceptable fuel design limits (SAFDLs). Therefore, the limiting consequences of the increase in feedwater flow will occur at the full load *ated power conditions and will bound all other power operating conditions due to the initial steam generator inventory and initial margin to DNB.
The greatest cooldown which can be postulated due to feedwater addition at full power is the inadvertent startup of all three auxiliary feedwater pumps.
This cooldown is larger than that due l
to the full opening of both feedwater control valves but less than that I
calculated for Event 15.1.3, Increased Steam Flow.
1
I ANF-87-161 Page 17 The main feedwater system is off-line in Modos 4-6 but may be on-line in hode 3.
For Mode 3 operating conditions, the potential cooldown in conjunction with a negative moderator temperature coefficient may result in a return to power at reduced crimary oressure, elevated all-rods-in peaking, and less than four reactor coolant pump conditions.
This case may pose a greater challenge to the SAFDLs than the full power case, and would bound zero power operation in Mode 2.
This is due to the potential for prompt criticality in Mode 3.
The greatest increase in feed flow would result from j
the startup of an idle pump with both control valves full open.
The conidown rate is less than the rate computed for Event 15.1.3 in Mode 3,
and consequently Event 15.1.2 in Mode 3 is bounded by Event 15.1.3 initiated from Mode 3.
In Modes 4-6, the only increased feed flow event initiator is inadvertent startup of an auxiliary feedwater pump since the main feedwater system is off-i line.
The startup of all three auxiliary feedwater pumps results in an increased energy removal rate, less than that computed for the Increase in Steam Flow event (15.1.3) for Modes 4-6.
The disposition of events for the Increase in Feedwater Flow event is l
summarized in Table 15.1.2-B.
15.1.3 Increase in Steam Flow I
This event is initiated oy a failure or 15.1.3.1 Event Initillar misoperation in the main steam system, which results in an increase in steam flow from the steam generators.
This event could be caused by the rapid opening of the turbine control valves, the atmospheric steam dump valves, the steam bypass to condenser valves, or the safety relief valves (SRVs).
l 15.1.3.2 Event Descriotion The increase in steam flow creates a mismatch between the energy being generated in the reactor core and the energy being removed through the secondary system.
This mismatch results in a cooldown of
ANF-87-161 Page 18 the primary system.
A power increase will occur if the moderator temperature reactivity feedback coefficient is negativo.
The power increase will cause a decrease in the DNB margin.
t The main steam system is designed to 15.1.3.3 Reactor Protection accomodate a 10% increase in load (step increase).
Reactor protection against a main steam flow increase greater than a 10% step is provided by the following trip signals: variable overpower trip, thermal margin / low pressure trip, local power density trip, low steam generator water level trip, and low secondary pressure trip.
Reactor protection for the Increase in Steam Flow event is sumarized in Table 15.1.3-A.
15.1.3.4 Disposition and Justification - The atmospheric steam dump valves are sized to accomodate 15% of the steam flow at 2700 MWt.
The steam dump to condenser valves and the turbine bypass valve are sized to accomodate 41% of the steam flow at 2700 MWt, and each SRV will pass 6.75% of the steam flow at 2700 MWt.
The capacities of the control valves for the main feedwater and i
auxiliary feedwater pump turbines are significantly less.
As such, the simultaneous opening of the steam dump valves and the turbine bypass valve could result in an increased load as great as 41% of the steam flow above the rated power operating condition of 2700 MWt.
This energy removal rate bonds the rated power operating conditions for Events 15.1.1, 15.1.E, and 15.1.4.
Therefore, this event will be analyzed as part of the plant transient analysis i
for Millstone Unit 2.
The consequences of this event for all other power i
operating conditions, including Mode 2,
are bounded by the rated power operating condition due to the increased margin to DNB at the other power operating conditions.
Since the Mode 3 operating condition has a higher average coolant temperature and a larger potential for cooldown than the Modes 4-6 reactor operating conditions, this condition represents the bounding event for the
[
"zero power" initial conditions.
It is evaluated to assess the potential for a return to power at reduced pressure conditions.
Whereas the steam dump t
-_- _ - -,. -.-,-.-__~__,_.- - _____.
i I
l ANF-87-161 J
Page 19 l
l valves are available in Mode 3, MSIV closure results in opening of the l
atmospheric dump valves being the only increased steam flow event initiator in l
Modes 4-6.
The analysis is performed from a hot zero power, all-rods-in plant l
state assuming a 3.6", Ap shutdown margin and one to four operating reactor l
coolant pumps as allowed by the Technical Specifications (Reference 3),
i The disposition of events for the Increase in Steam Flow event is summarized in Table 15.1.3-B.
The event analysis is performed to support increased radial peaking and a more negative EOC moderator temperature coeffscient.
15.1.4 Inadvertent Ooenina of a Steam Generator Relief or Safety Valve i
15.1.4.1 Event initiator - This event is initiated by an increase in steam flow caused by the inadvertent opening of a secondary side safety or relief valve.
15.1.4.2 Event Descriotion - The resulting mismatch in energy generation and removal rates results in an overraoling of the primary system.
If the moderator temperature coefficient is negative, the reactor power will increase.
I 15.1.4.3 Reactor Protection - Reactor protection is provided by the variable l
overpower trip, local powar density trip, thermal margin / low pressure trip, low secondary pressure trip, and low steam generator water level trip.
Reactor protection for th*e Inadvertent Opening of a Steam Generator Relief or Safety Valve event is summarized in Table 15.1.4 A.
i l
15.1.4.4 Disoosition and lustification - The inadvertent opening of a steam generator safety valve would result in an increased steam flow of approximately 6.75% of full rated steam flow.
Each dump (relief) valve is sized for approximately 7.50% steam flow with the reactor at full rated power.
As such, the consequences of any of these occurrences will be bounded by the i
l
ANF-87-161 Page 20 events in Section 15.1.3.
The disposition of events for the inadvertent Opening of a Steam Generator Relief or Safety Valve event is summarized in Table 15.1.4-B.
15.1.5 Steam S/ stem Pioina Failures inside and Outside of Containment 15.1.5.1 Event Initiator - This event is initiated by a rupture in the main steam piping upstream of the MSIVs which results in an uncontrolled steam release from the secondary system.
The increase in energy removal through the 15.1.5.2 Event Descriotion secondary system results in a severe overcooling of the primary system.
In the presence of a negative moderator temperature coefficient, this cooldown causes a decrease in the shutdown margin (following reactor scram) such that a return to power might be possible following a steam line rupture.
This is a potential problem because of the high power peaking factors which exist, assuming the most reactive control rod to be stuck in its fully withdrawn position.
15.1.5.3 Reactor Protection - Reactor protection is provided by the low steam generator pressure and water level trips, variable overpower trip, local power density trip, thermal margin / low pressure trip, high containment pressure trip, and safety injection actuation signal.
Reactor protection for the Steam System Piping Failures inside and Outside of Containment event is summarized in Table 15.1.5 A.
15.1.5.4 Disoosition and Justification At rated power conditions, the i
stored energy in the primary coolant is maximized, the available thermal margin is minimized, and the pre trip power level is maximized.
These conditions result in the greatest potential for cooldown and provide the greatest challenge to the SAFDLs.
Initiating this event from rated power also i
results in the highest post-trip power since it maximizes the concentration of i
ANF-87 161 Page.21 delayed neutrons providing for the greatest power rise for a given positive reactivity insertion.
Additional thermal margin is also provided at lower power levels by the automatically decreasing setpoint of the variable overpower trip.
Thus, this event initiated from rated power conditions will bound all other cases initiated from at power operation modes.
For the zero power and suberitical plant states (Modes 2-6), there is a potential for a return-to-pow >r at reduced pressure conditions.
The most limiting steam line break event at zero power is one which is initiated at the highest temperature, thereby providing the greatest capacity for cooldown.
This occurs in Modes 2 and 3.
Thus, the event initiated from Modes 2 and 3 will bound those initiated from Mades 4-6.
Further, the limiting initial conditions will occur when the core is just critical.
These conditions will maximize the available positive reactivity and produce the quickest and largest return to pcwer.
Thus, the steam line b eak initiated from critical conditions in Mode 2 will bound the results of the event initiated from suberitical Mode 3 conditions.
The Technical Specifications (Reference 3) only require a minimum of one reactor coolant pump to be operating in Mode 3.
One pump operation provides the limiting minimum initial core flow case.
Minimizing core flow minimizes
[
the clad to coc1&nt heat transfer coefficient and degrades the ability to remove heat generated within the fuel pins.
Conversely, however, a maximum loop flow will maximize the primary to secondary heat transfer coefficient, thus providing for the greatest cooldown.
Higher loop flow will sweep the cooler fluid into the core faster, maximizing the rate of positive reactivity addition and the peak power level.
The worst combination of conditions is achieved for the four pump loss of offsite power case.
In this situation, the initial loop flow is maximized resulting in the greatest initial cooldown, while the final loop flow is minimized providing the greatest challenge to the DNB SAFDL.
Since the natural circulation flow which is established at the end of the transient will nw-r.c,----v,,---------s---mm r-
,-,-----m-,-,~
ANF-87-161 Page 22 be the same regardless of whether one or four pumps were initially operating, the results of the four pump loss of offsite power case will bound those of the one pump case.
Thus, only four pump operation need be analyzed for the l
Mode 2 case.
The event is analyzed to support a more negative moderator temperature coefficient.
This event must be analyzed both with and without a coincident loss of-offsite power.
Typically, there are two-single failures which are f
considered for the offsite power available case.
The first is failure of a HPSI pump to start. The second is failure of an MSIV to close, resulting in a continued uncontrolled cooldown.
- However, Millstone 2 has combination MSIV/ swing disc check valves.
A double valve failure would thus be required for steam from the intact steam generator to reach the break.
This is not l
deemed credible. Thus, the single failure to be considered with offsite power i
available is failure of a HPSI pump to start.
For the loss-of-offsite power case, the limiting single failure is the failure of a diesel generator to s t art,.
This is assumed 'to result in the less of one HPSI pump and one charging pump. The disposition of events for the Steam System Piping Failures l
Inside and Outside of Containment event is summarized in Table 15.1.5 B.
i i
i i
1 1
l I
i
l ANF-87-161 Page 23 TABLE 15,1,1-A AVAILABLE REACTOR PROTECTION FOR DECREASE IN FEEDWATER TEMPERATURE EVENT Reactor Operating Conditions Reactor Protection 1
Variable Overpower Trip Thermal Margin / Low Pressure Trip local Power Density Trip low Steam Generator Pressure Trip 2
Variable Overpower Trip low Steam Generator Pressure Trip l
1 3
Variable Overpower Trip 3
46 Not a credible event for i
these reactor operating conditions
[
since there is no extraction steam l
to the feedwater heaters l
I i
i I
1 i
f
z ANF-87-161 1
Page 24 TABLE 15.1.1-8 DISPOSITION OF EVENTS FOR THE DECREASE IN FEEDWATER TEMPERATURE EVENT Reactor Operatirq Conditions Disoosition 1
Bounded by Event 15.1.3, Increase in Steam Flow Event 2, 3 Bounded by the above 4-6 No analysis required; not a credible event
ANF-87-161 Page 25
)
i TABLE 15.1.2-A AVAILABLE REACTOR PROTECTION FOR THE INCREASE IN FEEDWATER FLOW EVENT l
1 Reactor Operating Condition Reactor Protection 1
Variable Overpower Trip Local Power Density Trip Thermal Margin / Low Pressure Trip Low Steam Generator Pressure Trip i
Safety In.jection Actuation Signal Reactor Trip on Turbine Trip due to High Steam Generator Water Level j
f 2
Low Steam Generator Pressure Trip Variable Overpower Trip Saftty Injection Actuation Signal 3
Variable Overpower Trip l
l Safety Injection Actuation Signal 1
4 Tech. Spec Requirements on Shutdown Margin Inherent Negative Doppler Feedback 5, 6 No analysis required; no significant
, consequences t
ANF-87-161 Page 26 TABLE 15.1.2-8 DISPOSITION OF EVENTS FOR THE INCREASE IN FEE 0 WATER FLOW EVENT Reactor Operating Condition Disposition 1
Bounded by Event 15.1.3 (Increase in Steam Flow) 2 Bounded by the Mode 3 case 3-6 Bounded by Event 15.1.3 I
l 1
ANF 87-161 Page 27 TABLE 15.1.3 A AVAILABLE REACTOR FROTECTION FOR THE INCREASE IN STEAM FLOW EVENT 1
Reactor Operating Condition Reactor Protection 1
Low Steam Generator Pressure Trip Low Steam Generator Water Level Trip 1
Thermal Margin / Low Pressure Trip local Power Density Trip Variable Overpower Trip I
Safety injection Actuation Signal j
2 Low Steam Generator Pressure Trip i
j Low Steam Generator Water Level Trip i
1 Variable Overpower Trip t
Safety injection Actuation Signal 3
Variable Overpower Trip l
j Safety Injection Actuation Signal l
4 Tech. Spec Requirements on Shutdown Margin j
Inherent Negative Doppler Feedback i
5, 6 No analysis requiredt no significant consequences i
l
f ANF 87-161 i
I Page 28 TABLE 15.1.3-8 DISPOSITION OF EVENTS FOR THE INCREASE IN STEAM FLOW EVENT Reactor Operating Condition Disoosition 1
Analyze 2
Bounded by the above 3
Analyze 4
Dounded by the above
(
5, 6 No analysis required; no significant consequences s
i I
1 l
l ANF-87-161 Page 29 l
TABLE 15.1.4 A AVAILABLE REACTOR PROTECTION FOR THE INADVERTENT l
OPENING OF A STEAM GENERATOR REllEF OR SAFETY VALVE EVENT l
Reactor Operating Conditions Reactor Protection 1
Low Steam Generator Pressure Trip
)
Lcw Steam Generator Water Level Trip Yariable Overpower Trip Local Power Density Trip l
i Thermal Margin / Low Pressure Trip Safety injection Actuation Signal i
2 Low Steam Generator Pressure Trip l
Low Steam Generator Water Level Trip j
Variable Overpower Trip Safety injection Actuation Signal i
3, 4 Tech. Spec. requirements on shutdown margin, inherent negative Doppler f
feedback 5, 6 No analysis required; not a credible i
event l
U
f i
ANF 87-161 Page 30 TA8LE 15.1.4 8 DISPOSITION OF EVENTS FOR THE INADVERTENT OPENING 0F A STEAM GENERATOR RELIEF OR SAFETY VALVE EVENT 1
i Reactor Operating Conditions Disposition 1-4 Bounded by analyses presented for Event 15.1.3 i
i 5, 6 Not a credible event; no analysis required l
t 1
i 1
i I
I i
t l
l r
1 l
l l
I I
i l
l
I ANF-87-161 Page 31 TABLE 15.1.5-A AVAILABLE REACTOR PROTECTION FOR STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE OF CONTAINMENT EVENT Reactor Operating Conditions Reactor Protection 1
Low Steam Generator Pressure Trip Low Steam Generator Water level Trip Variable Overpower Trip Local Power Density Trip Thermal Nargin/ Low Pressure Trip High Containment Pressure Trip Safety Injection Actuation Signal 2
Low Steam Generator Pressure Trip low Steam Generator Water level Trip Variable Overpower Trip High Containment Pressure Trip Safety injection Actuation Signal 36 Tech. Spec, requirements on shutdown margin, inherent negative Doppler feedback O
l ANF-87-161 l
Page 32 TABLE 15.1.5-B DISPOSITION OF EVENTS FOR STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE OF CONTAINMENT EVENT Reactor Operating Conditions Discosition 1
Analyze 2
Analyze 36 Bounded by the above L
f ANF 87 161 I
I Page 33 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM i
15.2.1 Loss of External Load l
15.2.1.1 Event Initiator A major loss of load can be initiated as the j
result of a loss of external electrical load or a turbine trip.
Turbine stop valve closure is assumed as the initiator of this event because this is the fastest load rejection which can be postulated which will challenge the plant overpressure and SAFDL protection features.
The assumed fast valve closure
[
time (0.1 sec.) and the assumed unavailability of the steam dump system allow
[
this event to bound the effects of Events 15.2.2 (Turbine Trip - Steam bypass system available) and 15.2.4 (Closure of the MSIV - Valve closure time >0.1 l
l sec.).
4
]
15.2.1.2 Event Descriotion - For a full load reduction at power, the primary l
to secondary heat transfer would be severely diminished because of the
{
increase in secondary side temperature.
Initially, in response to the load f
reduction and diminished energy removal through the secondary system, the t
primary system temperatures begin to increase.
The increasing primary system i
i average temperature causes an insurge into the pressurizer due to the l
expanding primary fluid.
The primary system pressure increases as the pressurizer steaai space is compressed by the insurging liquid.
Primary system
[
overpressure protection is afforded by the pressu'izer power operated relief f
valves and the primary safety valves.
Eventually, the secondary system pressure reaches the opening setpoint of the secondary side safety valves and l
steam discharge occurs to limit the secondary side pressure rise.
Energy
{
removal through the steam generator and pressurizer safety valves mitigates l
the consequences of the load reduction.
- However, in analyzing the
{
I i
overpressurization aspects of this event, no credit is taken for the power l
operated relief valves on the primary or secondary systems.
l 15.2.1.3 Reactor Protection - Reactor protection is provided t,y the high j
pressurizer pressure, variable overpower, thermal margin / low pressure, and low l
t I
f i
ANF-87-161 Page 34 l
steam generator water level trips.
If the turbine is tripped at.the initiation of this event, a direct reactor trip signal would be generated and
]
the effects of this event would be mitigated. However, no credit is taken for l
a direct reactor trip on turbine trip.
Additionally, reactor protection is provided by the primary and secondary safety valves.
Because of the potential for increasing the primary system temperatures, with small increases in pressure, this event can challenge the SAFDLs as well as the overpressure criteria mentioned above.
Reactor protection for the Loss of External Load j
event is summarized in Table 15.2.1 A.
l l
15.2.1.4 Disoosition and Justification This event is only credible for rated power and power operating conditions because there is no load on the l
l turbine at other reactor conditions.
The consequences of this event for rated
)
power operation bound the consequences for other rea: tor conditions because of l
the maximum stored energy in the primary coolant, minimum initial thermal margin and maximum power to load mismatch which occurs upon load loss.
[
]
This event will be analyzed for Millstone Unit 2 from rated power
[
conditions to support the proposed increased radial peaking and a more positive BOC moderator temperature coefficient. Additionally, the analysis of l
the overpressurization aspects must consider plant operating conditions f
i I
representative of the end of cycle (EOC) coastdown.
4 l
There is no single failure considered which could worsen the results, i
The disposition of events for the loss of External Load event is summarized in Table 15.2.1-B.
15.2.2 Turbine Trio 15.2.2.1 Event initiator This event is initiated by a turbine trip which
]
results in closure of the main steam stop valves and a rapid reduction in energy removal through the steam generators.
)
l
f ANF-87-161 r
Page 35
(
f 15.2.2.2 Event Descriotion - The reactor protection system is designed to generate a reactor trip signal automatically when the turbine is tripped.
Following reactor trip, there would be a rapid decrease in the energy being generated in the primary system.
This would mitigate the consequences of the turbine trip event.
Primary and secondary system overpressurization protection is provided by the code safety valves on both the primary and secondary systems and thc secondary atmospheric dump valves.
Also, if the condenser was available, the steam bypass system would be activated to reduce the secondary system pressure.
15.2.2.3 Reactor Protection Reactor protection is provided by the high pressurizer pressure
- trip, variable overpower trip, thermal margin / low pressure trip, low steam generator water level trip, and a nonsafety grade reactor trip on turbine trip.
Additional protection is also provided by the primary and secondary side safety valves.
Reactor protection for the Turbine Trip event is summarized in Table 15.2.2 A.
This event is only credible for 15.2.2.4 Discostion and Justification rated power and power operating conditions since the turbine will either be in a tripped condition or there will be no load on the steam generators for other reactor operation conditions.
The consequences of this event for rated power operation bound the event consequences for other operating conditions because of the higher initial stored energy in the primary system, maximum power to load mismatch potential, and the reduced SAFDL margin for rated power operation.
Because of the limiting assumptions used in the analysis o' the consequences of the Loss of External Load event (15.2.1), the consequences of the Turbine Trip event are bounded by the consequences of Event 15.2.1, which is to be analyzed for Hillstone Unit 2.
The major assumptions used in Event l
15.2.1 are the conservatively rapid turbine stop valve closure time, the failure to trip the reactor on turbine trip, and the assumed unavailability of l
the atmospheric steam dump system.
The disposition. of events for the Turbine i
Trip event is summarized in Table 15.2.2 B.
i i
I ANF-8T 161 Page 36 15.2.3 Loss of condenser Vacuum This event is not in the current licer. sing basis for Millstone Unit 2 and therefore is not analyzed.
This is shown in the Available Reactor Protection and Event Oisposition Summary Tables 15.2.3 A and 15.2.3 B, respectively.
15.2.4 Closure of the Main Steam Isolation Valves (MSIVs) r 15.2.4.1 Event Initiator - The event postulated is the loss of controi air to the MSIV valve operator.
Either one or both MSIVs may inadvertently close.
The valves are swinging disc type check valves, installed in a reversed position and held open against steam flow by a pneumatically operating
)
cylinder assembly.
The valves are spring-loaded to the closed position.
15.2.4.2 Event Descriotion - The inadvertent MSIV closure is primarily of concern in boiling water reactors as indicated in the SRP (Ref. 1), but closure of the MSIVs in a PWR would cause a drastic reduction in the load on the reactor.
As such, the consequences of a dual MSIV closure are similar to the consequences of Event 15.2.1. Although the valve closure time for the MSIVs is less than 6 seconds, this is much longer than the turbine stop valve closure time assumed in Event 15.2.1 (0.1 seconds); as such, the transient events will proceed somewhat slower and be less severe than in the case of Event 15.2.1.
A single MSIV closure will result in an asymmetric primary temperature i
distribution.
Upon cessation of steam flow, the pressure in the affected steam generator will increase to the opening setpoint of the steam line gafety j
valves.
The primary to secondary heat transfer will be diminished, resulting in a heatup of the associated primary side loop.
In response to the drop in steam flow through the turbine control valves, the steam flow out of the unisolated steam generator will increase.
Dyressurization of the steam generator will result, causing an increase in heat removal from the primary.
The associated primary side loop will thus experience a cooldown. The side of 4
p
1
(
ANF-87-161 Page 37 the e
,ected to the cooldown will experience a power rise in the pre:
negative moderator temperature coefficient.
- 15..-
2 actor Protection - Reactor protection for the dual MSIV closure is provided by the high pressurizer pressure trip, variable overpower trip,
)
thermal margin / low pressure trip, low steam generator pressure trip, and low steam generator water level trip.
Additional protection i' provided by the primary and secondary side safety valves.
Reactor pro'.ection for the single MSIV closure is provided by the low steam generator level and low steam generator pressure trips.
Due to the location of the excore detectors and the asymmetries associated with this event, the variable overpower and TM/LP trip may not get the required 2 out of 4 channels tripped.
Further, since one loop will be cooling down and one will be heating up, the pressure may be either increasing or decreasing.
- Thus, this event cannot take credit for the variable high power, TM/LP or high pressure trips. Additional protection continues to be provided by the primary and secondary side safety valves.
15.2.4.4 Disoosition and Justification - This event is not credible in Modes 4-6 as the MSIVs are closed.
For simultaneous closure of both MSIVs, the event will progress very similarly to Event 15.2.1.
As such, the limiting case is obtained when the event is initiated from rated full power conditions.
The turbine stop valve closure time employed in the 15.2.1 analysis (0.1 sec) is much small,er than the MSIV closure time (6 sec). Thus, the consequences of Event 15.2.1 will bound those of +he dual MSIV closure event.
l The asymmetric conditions resulting from the closure of only one of the two MSIVs is similar to that predicted for a steam line break, That is, the primary coolant loop associated with the closed MSIV experiences a heatup due to the loss of heat sink and the primary coolant loop associated with the open MSIV experiences a cooldown due to the perceived load increase.
The temperature increase seen by the hot loop will be limited by the actuation of l
l
ANF-87-161 Page 38 the steam generator safety valves.
The temperature decrease seen by the cooling loop will continue until such time as a reactor trip is generated.
Since the loop experiencing the cooldown will see the larger temperature change, the limiting conditions for the event are at end of cycle. The end of cycle moderator temperature coefficient (MTC) is larger in absolute magnitude f
than the beginning of cycle MTC.
When the larger HTC is coupled with the larger temperature change in the cooling loop, a larger overall increase in core power will be predicted.
This larger increase in core power will produce the limiti.;g DNB conditions for the event.
}
i Since the asymmetries associated with the event preclude taking credit for the high pressure or variable overpower trip, the single MSIV closure cannot be bounded a priori by the *.oss of load, Event 15.2.1.
Further, since the loop that is experiencing the heatup is actuall.y driving the event instead of the loop experiencing the cooldown, the event can.ot be bounded a priori by the steam line break.
Thus, it is concluded that the event will be analyzed.
The limiting single MSIV closure case is that which is initiated from rated power in Mode 1.
There is no Mode 3 concern, as in this mode the magnitude of interruption in steam flow is not great enough to cause any significant consequences.
There is also no potential for a post-trip return to pr since the remaining MSIV and the turbine stop valves provide redunes ideans for terminating the remaining steam flow.
There is no single failure considered which could worsen the results.
The disposition of events for the Closure of the Main Steam Isolation Valves event is summarized in Table 15.2.4-B.
15.2.5 Steam Pressure Reaulator Failure Millstone Unit 2 does not have any steam line pressure regulators, so this event is not credible for this plant. No analysis needs to be con'sidered for this event.
ANF-87-161 Page 39 I
15.2.6 Loss of Nonemercency A.C. Power to the Station Auxiliaries This event is not in the current licensing basis for Millstone Unit 2 and therefore is not analyzed.
This is shown in the Available Reactor Protection f
and Event Disposition Summary Tables 15.2.6-A and 15.2.6-8, respectively.
15.2.7 Loss of Normal Feedwater Flow t
15.2.7.1 Event Initiator - The Loss of Normal Feedwater Flow transient is initiated by a trip of the main feedwater pumps or a malfunction in the feedwater control vaives.
)
15.2.7.2 Event Description - The loss of Normal Feedwater Flow event results in a total loss of all main feedwater flow to the steam generators.
Because the main feedwater system is supplying subcooled water to the steam generators, the loss of main feedwater flow will result in a reduction of the secondary system heat removal capability. The decrease in energy removal rate from the primary system causes the primary system fluid temperature to increase.
The resulting primary system fluid expansion results in an insurge into the pressurizer, compressing the steam space and causing the primary system pressura to increase.
The RCS pressure and temperature rise until a reactor trip occurs either due to low steam generator water level or high pressurizer pressure. Assuming
{
the steam bypass control system is in the manual mode of operation, l
l termination of main steam flow due to closure of the turbine stop valves l
following reactor trip temporarily causes steam generator and RCS j
pressurization.
The decrease in core heat rate after insertion of the CEAs, i
in combination with the main steam safety valvet-opening, resteres the RCS to j
a new steady state condition.
Auxiliary feedwater flow is automatically initiated on a low steam generator water level assuring sufficient steam i
generator inventory to prevent steam generator dryout and provide for core decay heat removal.
i t,
I
f ANF-87-161 1
Page 40 15.2.7.3 Reactor Protection - System overpressure protection is provided by the primary and secondary system safety valves.
A reactor trip occurs on low steam generator level with additional reactor protection provided by the high pressurizer pressure trip, variable overpower trip, and the thermal margin / low pressure trip.
Reactor protection for the loss of Normal Feedwater Flow event is summarized in Table 15.2.7-A.
f 15.2.7.4 Disoosition and Justification - This event is only credible for rated power and power operating conditions because the main feedwater system is not required to provide feedwater to the stehm generators for other reactor operating conditions.
The consequences of this event for rated power operation bound the consequences for other conditions because of the higher
(
initial stored energy in the primary system, the minimum steam generator inventory, and the greater impact of the loss of feedwater flow on the secondary system.
The near term pressurizatien and DNB aspects of this event are bounded by those of Event 15.2.1.
This is due to fact that in the analytical methodology for Event 15.2.7 given in Reference 2, it is indicated that reactor trip occurs at time zero coincident with turbine trip on a low steam generator water level signal.
In Event 15.2.1, reactor trip is delayed until a high pressurizer pressure signal is received.
This results in a higher power level at trip, greater pressurization and greater challenge to the SAFDLs than in Event 15.2.7.
Long term pressurization, if it occurs, is very gradual and is arrested by opening of the pressurizer code safety valves.
The loss of Normal Feedwater event will be analyzed to assess the maximum expected pressurizer level swe;l and the long term adequacy of the auxiliary feedwater system to restore and maintain steam generator inventory and prevent steam generator dryout.
The maximum level swell will be examined to assure that the pressurizer does not become water solid.
The analysis must consider
ANF-87-161 i
Page 41 plant operating conditions representative of the end of cycle coastdown as described for Event 15.2.1.
The analysis will support a more positive BOC moderator temperature coefficient.
The single failures considered in this analysis are failure of an auxiliery feedwater pump to start, or a loss of offsite power resulting in coastdown of the reactor coolant pumps.
The disposition of events for the loss of Normal Feedwater Flow event is summarized in Table 15.2.7-8.
15.2.8 Feedwater System Pioe Breaks inside and Outside Coatainment This event is not in the current licensing basis for Millstone Unit 2 and therefore is not analyzed.
This is shown in the Available Reactor Protection f
and Event Disposition Summary Tables 15.2.8-A and 15.2.8 8, respectively, b
i I
i I
\\
?
ANF-87-161 Page 42 TABLE 15.2.1-A AVAILABLE REACTOR PROTECTION FOR THE LOSS OF EXTERNAL LOAD EVENT
]
Reactor Operating Conditions Reactor Protection 1
High Pressurizer Pressure Trip Variable Overpower Trip l
Thermal Margin / Low Pressure Trip Low Steam Generator Water Level Trip i
2 High Pressurizer Pressure Trip Variable Overpower Trip Low Steam Generator Water Level Trip 3-6 No analysis requirea; not a credible event.
j j
i a
l ANF-87-161 Page 43 u
l i
TABLE 15.2.1-B DISPOSITION OF EVENTS FOR THE LOSS OF EXTERNAL LOAD EVENT Reactor Operating Conditions Disposition L
i 1
Analyze I
l 2
Bounded by the above, no analysis required.
t 3-6 No analysis required; not a credible event.
)
b r
1 1
1 I
I i
l l
r l
i l
i i
i i
AN7-87-161 Page 44 TABLE 15.2.2-A AVAILABLE lFACTOR PROTECTION FOR THE TURBINE 7'iP EVENT Reactor Operating Conditions Reactor Protection 1
High Pressurizer Pressure Trip Nonsafety Grade Reactor Trip on Turbine Trip Variable Overpower Trip i
Thermal Margin / Low Pressure Trip Low Steam Generator Water Level Trip 2
High Pressurizer Pressure Trip Variable Overpower Trip low Steam Generator Water Level Trip 36 No analysis required; not a credibic event.
i I
1 a
I
{
ANF-87-161 Page 45 TABLE 15.2.2-B DISPOSITION OF EVENTS FOR THE TURBINE TRIP EVENT 6
Reactor Operating Conditions Disposition 1
Bounded by Event 15.2.1 for the rated power operating condition (#1).
2 Same as above.
3-6 No analysis required; not a credible event.
4 I
l l
4 1
ANF-87-161 8
Page 46.
TABLE 15.2.3-A AVAILABLE REACTOR PROTECTION FOR THE LOSS OF CONDENSER VACUVM EVENT j
f i
Reactor Operating Conditions Reactor Protection j
1-6 Not in licensing basis; not analyzed.
j I
TABLE 15.2.3-B DISPOSITION OF EVENTS FOR THE LOSS OF CONDENSER VACUUM EVENT I
Reactor Operating Conditions Disoosition 16 Not in licensing basis; not analyzed.
h
..,,...-,--.....,--,.--.n..-----.,.,..--,,.--...,,-,.,,-..n---.-.
ANF-87-161 Page 47 TABLE 15.2.4-A AVAILABLE REACTOR PROTECTION FOR THE CLOSURE OF THE MSIVS EVENT b
i Reactor Operating Conditions Reactor Protection I
High Pressurizer Pressure Trip Variable Overpower Trip Thermal Margin / Low Pressure Trip Low Steam Generator Water Level Trip 2
High Pressurizer Pressure Trip Variable Overpower Trip Low Steam Generator Water Level Trip 3
Variable Overpower Trip 46 fio analysis required; not a credible event.
l i
4
J ANF-87-161 Page 48 TABLE 15.2.4-B DISPOSITION OF EVENTS FOR THE CLOSURE OF THE MSIVs EVENT
]
)
Reactor Operating Conditions Disoosition 1
Dual MSIV closure: bounded by Event j
15.2.1.
Single MSIV closure: analyze i
2, 3 Bounded by Mode 1.
1 i
i 4-6 No analysis required; not a credible event.
l i
P i
\\
~
I
)
1
,n.
,..,n,. - - _ - _,., -
---n.,.,_
,. ~... -. - - -
,.v..nn,,,-,,~na,n,
,n,--w-,,,,,
--,.,.,nn,,--,n,~.r.r-.
(
ANF-87-161 Page-49 TABLE 15.2.5-A AVAILABLE REACTOR PROTECTION FOR THE STEAM PRESSURE REGULATOR FAILURE EVENT i
Reactor Operating Conditions Reactor Protection
+
l-6 None required, not a credible event for this plant, e
TABLE 15.2.5 B DISPOSITION OF EVENTS FOR THE STEAM PRESSURE REGULATOR FAILURE EVENT t
l Reactor Operating t
I Conditions Discosition l
1 l
l-6 No analysis required.
l l
t I
l L
)
ANF-87 161 L
Page 50 J
TABLE 15.2.6-A AVAILABLE REACTOR PROTECTION FOR THE LOSS OF NONEMERGENCY A.C. POWER TO THE STATION AUXILIARIES EVENT Reactor Operating Conditions Reactor Protection I
1-6 Not in licensing basis; not analyzed.
l i
l TABLE 15.2.6-B DISPOSITION OF EVENTS FOR THE LOSS OF NONEMERGENCY A.C. POWER TO THE STATION AUXILIARIES EVENT Reactor Operating Conditions Discosition 16 Not in licensing basis; not analyzed.
(
ANF-87-161
[
Page 51 TABLE 15.2.7-A AVAILABLE REACTOR PROTECTION FOR THE LOSS OF NORMAL FEEDWATER FLOW EVENT Y1 Reactor Operating Conditions Eg3ctor Protection 1
Low Steam Generator Water Level Trip High Pressurizer Pressure Trip
. Thermal Margin / Low Pressure Trip Variable Overpower Trio 2
High Pressurizer Pressure Trip Variable Overpower Trip Low Steam Generator Water Level Trip 3
Variable Overpower Trip i
4-6 No analysis required; not a credible event.
t
(
ANF-87-161 Page 52 l
TABLE 15.2.7-B DISPOSITION OF EVENTS FOR THE LOSS OF NORMAL FEEDWATER FLOW EVENT Reactor Operating Conditions Disposition t
1 Analyze to assass maximum pressurizer level swell and long term adequacy of AFW.
Pressurization and DN8 aspects bounded by Event 15.2.1.
2, 3 Bounded by the above, no analysis required.
4-6 No analysis required; not a credible I
event.
ANF-87-161 l
Page 53 I
TABLE 15.2.8-A AVAILABLE REACTOR PROTECTION FOR THE FEEDWATER SYSTEM PIPE BREAKS INSIDE AND OUTSIDE CONTAINMENT EVENT i
Reactor Operating Conditions Reactor Protection 16 Not.in licensing basis; not analyzed.
4 TABLE 15.2.8-8 DISPOSITION OF EVENTS FOR THE FEE 0 WATER SYSTEM PIPE BREAKS INSIDE AND OUTSIDE CONTAINMENT EVENT Reactor Operating Conditions Dispositign i
l-6 Not in licensing basis; not analyzed.
t i
k.
ANF-87-161 I
Page 54 15.3 DECREASE IN REACT 0k C00LAN7 SYSTEM FLOW 15.3.1 Loss of Forced Reactor Cc)lant Flow 15.3.1.1 Event Initiator - The loss of forced reactor coolant flow in the primary system may result from a me-hanical or electrical failure in a main reactor coolant pump or in the power supply to these pumps.
Forced coolant 5
flow may be completely or partially lost.
The limiting event initiator is that which results in the trip of all four reactor coolant pumps.
15.3.1.2 Event Descriotion - The immediate result of the loss of forced coolant flow is an increase in the coolant temperature as it flows through the r64ctor core.
If the reactor is at power, this temperature increase could challenge the specified acceptable fuel design limits.
DNB could result if the reactor is not tripped.
15.3.1.3 Reactor Protection - Reactor protection is provided by the following reactor trips:
(1)
Low reactor coolant flow; (2) Low reactor coolant pump speed; (3) Thermal margin / low pressure; and (4) High pressurizer pressure trip.
Reae, tor protection for the loss of Forced Reactor Coolant Flow event is summarized in Table 15.3.1-A.
i 15.3.1.4 Discosition and Justification The power sources for the main reactor coolant pumps are the most likely initiator for a loss of flow event I
. involving more than one pump.
A mechanical or electrical fault in one of the
)
pumps will only result 1.1 a single pump loss of forced coolant flow transient.
The normal power supplies for the pumps are from two buses which receive power from the main generator.
Two pumps, in opposite loops, are powered from each
ANF-87-161 Page 55 bus.
If there is a generator trip, the pumps are automatically transferred to a bus supplied from the external power lines.
A generator trip with the failure of this transfer could result in a loss of power to all four pumps.
In the case of four pump operation, two situations must be considered:
two pump coastdown and a total loss of forced coolant flow.
Considering first the total loss of flow cases, the consequences of this postulated event are bounded by rated power operation. The rated power case is bounding because of i
the reduced DNB margin for this initial state combined with tne highest power to flow ratio during coastdown.
For the two pump loss of flow cases, the magnitude of the coastdown is less severe than the four pump coastdown, and the consequences of this event are bounded by the four pump loss of flow event. For the two pump flow coastdown cases, there is always some degree of forced reactor coolant flow.
These events are, therefore, not as challenging as the four pump coastdown events.
A comparison of the governing parameters indicates that these events are bounded by the four pump loss of flow event from full rated power conditions.
In summary, the four pump loss of flow event is the bounding event for the 15.3.1 events in all modes of operation.
It will be analyzed to support increased radial peaking and a more positive BOC moderator temperature coefficient.
The only active system challenged is the reactor protection system which is redundant and single failure proof.
The disposition of events for the Loss of Forced Reactor Coolant Flow event is summarized in Table 15.3.1-B.
15.3.2 Flow Controller Malfunction There are no flow control devices on the primary reactor coolant system of Millstone Unit 2.
This event is therefore not credible and need not be analyzed.
l
I f
ANF-87-161 Page 56 15.3.3 Reactor Coolant Pumo Rotor Seizure 15.3.3.1 Event Initiator - This event is initiated by an instantaneous seizure of a reactor coolant pump rotor.
l Flow in the affected loop will be rapidly 15.3.3.2 Event Descriotion reduced causing the core flow to also decrease rapidly.
As in the 15.3.1 events, the reduction in primary reactor coolant flow will result in the increase in primary coolant temperatures and a challenge to the ONB margin. A pressurization of the primary system will also occur due to the heatup of the primary coolant which causes a rapid insurge into the pressurizer.
A low reactor coolant flow trip will be generated.
15.3.3.3 Reactor Protection - Reactor protection for the reactor coolant pump rotor seizure event is provided by the low reactor coolant flow trip, thermal margin / low pressure trip, and the high pressurizer pressure trip.
Reactor protection for the Reactor Coolant Pump Rotor Seizure event is summarized in Table 15.3.3-A.
i 15.3.3.4 Disoosition and Justification - This event is a concern for only rated power and power operating conditions because for other reactor operating conditions there is sufficient thermal margin so there will not be a challenge I,
to the fuel design limits.
The core heat flux to flow ratio is an excellent indicator of the potential DNB challenge for a loss of flow event.
The highest ratios for this event are predicted to occur during the first few l
seconds of the transient from full power rated operating conditions.
The l
consequences of this event will therefore be bounded by a pump rotor seizure l
event initiated from full power rated conditions.
There is no single failure considered which could worsen the results.
The event is analyzed to support j
increased radial peaking and a more positive BOC moderator temperature i
coefficient.
)
ANF-87-161 Page 57 The disposition of events for the Reactor Coolant Pump Rotor Seizure event is sumarized in Table 15.3.3-8.
15.3.4 Reactor Coolant Pumo Shaft Break This event is not in the current licensing basis for Millstone Unit 2 and is, therefore, not analyzed.
This is shown in the Available Reactor Protection and Event Disposition Sumary Tables 15.3.4-A and 15.3.4-8.
i l
j
s
/
ANF-87-161 Page 58 i
TABLE 15.3.1-A AVAILABLE REACTOR PROTECTION FOR THE LOSS OF FORCED REACTOR COOLANT FLOW EVENT i
t Reactor Operating Conditions Reactor Protection 1 (4 pump operation)
Low Reactor Coolant Flow Trip Low Reactor Coolant Pump Speed Trip Thermal Margin / Low Pressure Trip High Pressurizer Pressure Trip 2 (4 pump operation)
High Pressurizer Pressure Trip Tech.' Spec. requirements on number of operating pumps 3-6 (less than 4 pump High Pressurizer Pressure Trip operation)
Tech. Spec. requirements on number of operating pumps 4
J p
n l
l l
l
'l ANF-87-161 '
i Page 59 1
TABLE 15.3.1-B DISPOSITION OF EVENTS FOR THE LOSS OF FORCEO REACTOR COOLANT FLOW EVENT l
l Reactor Operating Conditions Oisoosition J
1
. Analvze i
2-6 Bounded by the above, no analysis 3
required.
1 i
l l
4 i
4 l
I i
i i
4 i
l a
s
_.- _ _ _,~.J
(
Af1F-87-161 r
Page 60 f
TABLE 15.3.2-A AVAILABLE REACTOR PROTECTION FOR THE FLOW CONTROLLER MALFUNCTION EVENT l
Reactor Operating Conditions Reactor Protection 1-6 No analysis required, event not credible.
1 TABLE 15.3.2-B OISPOSITION OF EVENTS FOR THE FLOW CONTROLLER MALFUNCTION EVENT i
Reactor Operating Conditions Disoosition 1-6 Not a credible event, no analysis required.
1
]
ANF-87-161 Page 61l TABLE 15.3.3-A AVAILABLE REACTOR PROTECTION FOR THE f
REACTOR COOLANT PUMP ROTOR SEIZURE EVENT Reactor Operating Conditions Peactor Protection 1
Low Reactor Coolant Flow Trip Thermal Margin / Low Pressure Trip High Pressurizer Pressure Trip i
2 High Pressurizer Pressure Trip Available Thermal Margin l
l 3-6 Available Thermal Margin 1
l e
4 i
j
f ANF 87-161 Page 62 r
I TABLE 15.3.3-B DISPOSITION OF EVENTS FOR THE REACTOR COOLANT PUMP ROTOR SE!ZURE EVENT Reactor Operating Conditions Disposition 1
Analyze 2
Bounded by the above.
3-6 No analysis required.
t i
i f
i i
I i
l i
1 l
J
(
l f
ANF-87-161 Page 63 TABLE 15.3.4-A AVAILABLE REACTOR PROTECTION FOR THE i
REACTOR COOLANT PUMP SHAFT BREAX EVENT s
t Reactor Operating Conditions Reactor Protection 1-6 No analysis required; not in J
licensing basis.
l l
l i
TABLE 15.3.4 B OISPOSITION OF EVENTS FOR THE REACTOR COOLANT PUMP SHAFT BREAK EVENT I
l Reactor Operating Conditions Discosition 1
16 No analysis required; not in l
H licensing basis.
l l
l l
l L.
L ANF-87-161 Page 64 I
15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1 Uncontrolled _ Control Rod / Bank Wi+hdrawal From a Suberitical or Low Power Startuo Condition Initiated by the uncontrolled withdrawal of a 15.4.1.1 Event Initiator control rod / bank, this event results in the insertion of positive reactivity and consequently a power excursion.
This event could be caused by a malfunction in the reactor control or rod control systems.
The consequences of a single bank withdrawal from operating Modes 3 6 are considered it.- this event category;.the consequences at rated power and power operating initial conditions are considered in Event 15.4.2.
The control rods are wired together into preselected bank configurations.
These circuits prevent the control rods from being withdrawn in other than their respective banks.
Power is supplied to the bank, in such a way that no more than two banks can be withdrawn at the same time and in their proper withdrawal sequence.
15.4.1.2 Event Descriotion - The neutron flux rises very rapidly in response to the continunus positive reactivity insertion.
The initial rapid rise is terminated by the reactivity feedback effect of the negative Coppler coefficient.
The number of reactor coolant pumps in operation can significantly affect the ability to remove the heat generated in the fuel due to the power increase.
15.4.1.3 Reactor Protection The power transient is eventually terminated (as j
well as the control rod withdrawal) by the reactor protection system on one of the following signals:
(1) Variable overpower trip; (2) High pressurizer pressure tript or (3) Variable overpower pre trip alarm, which initiates Rod Withdrawal
)
Prohibit Action.
l 1
,q
.ANF-87-161 Page 65 Reactor protection for the Uncontrolled Control Rod Bank Withdrawal from a Subcritical.,er Low Power Startup Condition euat is summarized in Table 15.4.1-A.
15.4.1.4 Disoosition and Justification The Technical Specifications for Millstone Unit 2 require that the control rod drives be de-energized in Modes 4 6 whenever the Reactor Coolant System boron concentration is less than the refueling concentration of 1720 ppm (Reference 3, pg. 3/4.1-31).
A rod withdrawal from these modes is therefore not considered a e edible s rent.
Whenever the rod control system is energized. Technica aecifications require 4 operating reactor ecolant pumps, although pressere is allowed to be as low as 2000 psia in Mode 3 per Reference 3, pg. 3/4.1-3.
The greatest power rise for this event is obtained when it is initiated frsm the lowest power.
I 4
Therefore, the event initiated from a low power critical Mode 2 condition at 2000 psia will bound all other low power or suberitical cases.
The only active system challenged in this event is the reactor protection system, which is redundant and single failure proof.
The event will be ra ure coeff cient The disposition of events for the Uncontrolled Control Rod / Bank Withdrawal from a Suberitical or low Power Startup-Conditioa event is I
suanarized in Tabic 15.4.1-B.
15.4.2 LLn. controlled Control Rod / Dank Withdrawsl at power i
15.4.2.1 EyJLa.t., Initiator - This event is initiated oy an uncc trolled control l
rod / bank withdrawal from power operating conditions.
i I
i
[
f ANF-87 161 Page 66 i
t 15.4.2.2 Event Descriotion - Positive reactivity is added to the reactor core due to the uncontrolled bank withdrawal resulting in a power transient.
The increase in core powar results in an increase in the core heat flux creating a challenge to the ONB margin.
The challenge to the DNB margin is further accentuated by the mismatch between the energy removal from the steam j
generaters and the power produced in the core.
This mismatch in power causes j
the primary system temperatures to rise, reducing the DNS margin.
r i
i 15.4.2.3 Reactor Protection - The challenge to the fuel design limits is terminated by the automatic action of the reactor protection system which
[
terminates the bank withdrawal and inserts negative reactivity to terminate i
l the power transient. The automatic action of the reactor protection system is l
initiated as the result of one of the following signals:
i (1) Variable overpower trip;
{
(2) local power density trip; (3) Th.ermal margin / low pressure trip; or
{
j (4) High pressurizer pressure trip.
i
)
Reactor protection for the Uncontrolled Control Rod / Ban Withdrawal at Power event is summarized in Table 15.4.2 A.
i t
i 15.4.2.4 Diseosition and Justification - This event is designed to address l
1.
the safety challenge posed by an uncontrolled control rod / bank withdrawal l
i transient from power conditions.
This event addresses all the power operating conditions and the rated power operating conditions.
It is performed to test 4
{
the adequacy of the variable overpower and thermal margin / low pressure (TM/LP) i trip setpoints in mitigating the challenge to the SAFDLs.
i j
A rod withdrawal initiated from lower powers will provide less of a
)
challenge to the SAFDLs due to increased initial thermal margin, a lesser i
amount of setpoint overshoot, and a decreased variable overpower trip setpoint resulting in a greater thermal margin at trip.
The event initiated from full l
l
s ANF 87-161 Page 67 power will then bound those initiated from' lower power conditions.
This event will therefore be analyzed at full power for conditions ranging from BOC to EOC, for a spectrum of reactivity insertion rates.
The only active system challenged by this event is the reactor protection system, which is redundant and single failure p,oof. This analysis will supnort increased radial peaking and a more positive BOC ar.d a more negative EOC moderator temperature coefficient.
l The disposition of events for the. Uncontrolled Control Rod / Bank Withdrawal at Power event is summarized in Table 15.4.2 B.
15.4.3 Control Rod Mitoceration 15.4.3.1 E,ent Initiator - The control rod misoperation event encompasses a number of transients resulting from different event inittstors.
The specific events addressed under this event' category include the following:
(1) Dropped control rod or control rod ba,.k; l
(2) Dropped part-length control rod; l
(3) Ma1 positioning of the part-length control rod group; (4) Statically misaligned control rod / control rod bank; t
l (5) Single e.ontrol rod withdrawal; l
(6) Reactivity control device removal error during refueling; and
('/) Variations in reactivity load to be compensated by burnup or on line refueling.
l (1)
Droceed Control Rod / Bank 15.4.3.1(1)
Event Initiator - The dropped control rod and dropped control bank events are initiated by a de-energized control rod drive mechanism or by a malfunction associated with a control rod bank.
15.4.3.2(1) [yent Descriotion - In the dropped control rod event, the reactor power initially drops in response to the insertion of negative reactivity.
ANF 87 161 Page 68 However, the local peaking increases due to the local effect on the power e
distribution. The reactor core will attempt to return to a new equilibrium at the original power level as a result of moderator and Doppler reactivity feedback.
Because of the increased peaking and the potential return to the initial power level, the dropped control rod event poses a severe challenge to the DNB margin.
15.4.3.3(1)
Reactor Protection If the amount of reactivity is large enough to cause a significant reduction in core power, a reactor trip would be generated by the variable overpower trip.
Reactor protection for the Control Rod Misoperation (Dropped Control Rod / Bank) event is summarized in Table 15.4.3(1) A.
Since the control rod drive
(
15.4.3.4(1)
Disoosition and Justification mechanisms are de-energized in Modes 4 6 and reactor power is limited to zero percent with k,ff <.99 in Mode 3, there will be no consequences of this event for these modes.
Ultimately, the consequences of this event are a return to power at elevated peaking conditions.
Thus, the worst case is obtained when the final power level, increased peaking, and core inlet temperature are maximized. This occurs for dropped rod / bank events initiated from full power.
The full power case thus bounds all other power operation conditions.
In general, a bank drop will cause a reactor trip and, as such, poses no challenge to the DNB margin.
However, the event is analyzed from full power to assure that even the minimum worth bank when dropped will cause a reactor trip prior tv a significant return to power.
For a single dropped control rod, a reactor trip is not expected.
Thus, a DNB evaluation assuming a return to full power at maximum dropped rod peaking will be performed to demonstrate that the SAFDLs are not violated.
The analysis will support increased radial peaking and a more negative EOC moderator temperature coefficient.
It should be noted that the operator will have multiple indications that a dropped red / bank has occurred via CEA deviation circuit alarms and rod bottom signals.
The only active system challenged in this event is the reactor protection
ANF-87-161 Page 69 system, which is reduviant and single failure proof.
The disposition of events for the Control Rod Misoperation (Dropped Control Rod / Bank) event is summarized in Table 15.4.3(1)-B.
(2) Drooned Part-Lenath Control Rod, and (3) lidloositionina of the Part Lenath Control Rod Groun All part length control rods have been removed from the Millstone Unit 2
)
core.
Therefore, these events are not applicable.
This is shown in the Available Reactor Protection and Event Disposition Summary Tables 15.4.3(2).A.
15.4.3(2) 8, 15.4.3(3)-A, and 15.4.3-(B).
(4) Statically Misalianed Control Rod / Bank These events are not in the current licensing basis for Millstone Unit 2 f
and therefore are not analyzed.
This is shown in the Available Reactor f
Protection and Event Disposition Summary Tables 15.4.3(4) A and 15.4.3(4) 8, respectively, f
1 t
d (5) Sinale Control Rod Withdrawal l
l 15.4.3.1(5)
Event Initiator This event is initiated by the inadvertent withdrawal of a single CEA from the core.
No single electrical or mechanical
{
failure in the Rod Control System could cause the accidental withdrawal of a j
single CEA from the inserted CEA bank during full power operation.
Procedures
[
l are available to permit the operator to withdraw a single CEA in the control bank since this feature is necessary in order to retrieve an assembly should one be accidentally dropped.
The event can occur only as the result of multiple wiring failures cr multiple operator actions in disregard of l
available event indication.
I In the extremely unlikely event of simultaneous electrical failures which could result in single CEA withdrawal, the rod position indicators and I
deviation alarms would indicate the relative positions of the assemblies in the bank.
Withdrawal of a single CEA by operator action, whether deliberate L
l
(
l
s ANF 87-161 Page 70 or by a combination of errors, would similarly result in the same visual indications.
The CEA Motion Inhibit prevents further rod control motion upon detection of CEA malpositioning.
15.4.3.2(5)
Event Descriotion - The withdrawal of a single full length CEA is initiated by the inadvertent withdrawal of a single control rod.
The ensuing reactivity insertion causes core po..er to increase.
In the event that the secondary steam dump control system does not respond to the increased power production, secondary system temperature and pressure will increase, causing a corresponding increase in primary coolant temperature.
This increase in primary coolant temperature occurs slowly enough that the pressurizer pressure control system, if available, is capable of suppressing f
the primary pressure increase.
The degradation of coolant conditions coupled with the power increase is essentially the same as expected for CEA bank j
withdrawals at power, and may approach DNB conditions in the hot channel.
The single CEA withdrawal is distinguished from the withdrawal of a CEA j
bank by a severe radial power redistribution.
High radial power peaking is localized in the region of the single withdrawn CEA and may, in severe cases, l
surpass the design limits.
Thus, assemblies in the immediate vicinity of the I
withdrawn CEA may experience boiling transition.
Sen exposure would be limited to short time periods.
Some fuel damage might occur.
]
15.4.3.3(5)
Reactor Protection - The challenge to the fuel design limits is terminated oy the automatic action of the reactor protection system which terminates the CEA withdrawal and inserts negative reactivity to terminate the i
power transient.
The automatic action of the reactor protection system is initiated as the result of one of the following signals:
(1) Variable overpower trip; (2) local power density trip; l
(3)
Thermal margin / low pressure trip; i
(4) High pressurizer pressure trip; or j
1 ANF-87-161 Page 71 (5) Variable overpower pre-trip alarm, which initiates Rod Withdrawal Prohibit action.
}
Reactor protection for the Uncontrolled Control Rod Bank Withdrawal at Power event is summarized in Table 15.4.3(5) A.
15.4.3.4(5)
Disoosition and Justification - The overall system response to I
the withdrawal of a single CEA will be identical to the response to a withdrawal of a CEA bank.
The only difference will be that the core will experience localized peaking in the vicinity of the withdrawn CEA that is not present if an entire bank is withdrawn.
Therefore, the disposition of the single CEA withdrawal will be identical to that of the CEA bank withdrawal.
The disposition of the low to zero power bank withdrawal is addressed in Event 15.4.1.
The disposition of the bank withdrawal from power operating conditions is addressed in Event 15.4.2.
The disposition of events for the Single Cortrol Rod Withdrawal event is summarized in Table 15.4.3(5)-B.
(6)
Reactivity Control Device Removal Error Durino Refuelino
.rh11 stone Unit 2 has no reactivity control devices which are used during refueling and could inadvertently be removed.
Boron dilution during refueling is considered in Event 15.4.6.
Therefore, this event is not applicable.
(7) Variations in Reactivity load to be Comoensated by Burnuo or On-line Refuelino This event considered the anticipated variations in the reactivity load of the reactor, to be compensated by means of action such as buildup and burnup of xenon poisoning. fuel burnup, oa-line refueling, fuel followers, temperature moderator and void coefficients.
Provisions for xenon changes and fuel burnup are described in Section 3 of Reference 4.
On-line refueling will not be performed on Hillstone Unit 2.
l The core design does not include fuel followers.
The safety analyses are based upon the most adverse combination of temperature, moderator and void
s ANF-87-161 Page 72 ccefficients.
Therefore, this event has no significant consequences and is f
not analyzed.
15.4.4 Startuo of an Inactive Looo 15.4.4.1 Event Initiator - This event is initiated by the startup of an inactive reactor coolant pump.
15.4.4.2 Event Descriotion - Each primary coolant loop is equipped with two single suction centrifugal pumps, one per cold leg, which are located between the steam generator outlet and the reactor vessel inlet nozzles.
A nonreversing mechanism is provided to prevent reverse rotation of the pump rotor.
This feature also limits backflow through the pump under non-operating conditions.
Note: there is no backflow in the hot leg (or steam generator) associated with the side of the plant that has the inactive reactor coolant pump.
The inadvertent actuation of an inactive pump would therefore lead to a decrease in moderator temperature and, with a negative moderator coefficient, an increase in core reactivity with a potential increase in core power level.
15.4.4.3 Reactor Protection - Reactor protection for this event is afforded by Technical Specification requirements on shutdown margin and reactor coolant pump operation.
Reactor protection for the Startup of an Inactive loop event is summarized in Table 15.4.4-A.
i This event,is not credible in 15.4.4.4 Disoosition and Justification 1
operating Modes 1 and 2 because Technical Specifications require all four reactor coolant pumps to be oparating (Reference 3, pg. 3/4.41).
It is not credible in Mode 6 due to administrative procedures requiring that the pumps be prevented from starting, j
l Technical Specification requirements on shutdowr margin in Modes 3-5 are l
such that any reactivity insertion due to an. inactive loop start is not great l
ANF-87-161 Page 73 J
enough to reach criticality.
Thus, the consequences of this event in Modes 3-5 are minimal and no analysis is required.
The disposition of events for the Startup of an Inactive Loop event is summarized in Table 15.4.4-B.
15.4.5 Flow Controller Malfunction Millstone Unit 2 does not have any flow control devices on the primary reactor coolant loops so this event is not credible and does not need to be analyzed.
15.4.6 CVCS Malfunction That Results in a Decrease in the Boron Concentration in the Reactor Coolant A dilution of the primary system boron 15.4.6.1 Event Initiator concentration can occur as a result of adding primary grade water into the l
reactor coolant system via the Chemical Volume and Control System (CVCS) or the Precise Control of Reactivity System (PCRS).
The greatest dilution rate occurs for operation of the CVCS charging pumps. The three available charging pumps can inject water into the primary system at a maximum rate of 132 gpm.
15.4.6.2 Event Descriotion Dilution of the primary coolant Boron
)
concentration results in the insertion of positive reactivity.
For reactor operation Modes 3-6, the event can result in a gradual erosion of available shutdown margin which, if unchecked, can cause a return to criticality.
In the case of a boron dilution at rated power and power operation reactor operating conditions, the consequences are very similar to the consequences of a slow control rod withdrawal.
15.4.6.3 Reactor Protection - Reactor protection for the boron dilution event during operating Modes 3-6 is provided by Technical Specification Shutdown Margin requirements, Administrative procedures, and sufficient time for the operator to take the appropriate action in the unlikely event that a boron dilution should occur.
Reactor protection for the reactor critical, power operation, and rated power operating conditions is provided by various trips 1
1 l
ANF-87-161 Page 74 i
i and operator response time.
Reactor protection for the CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant event is summarized in Table 15.4.6-A.
15.4.6.4 Disoosition and Justification - For baron dilutions in reactor Modes 1-6, the challenge to the SAFDLs is very similar to that of slow control rod withdrawals and can be bounded by the consequences of control rod withdrawal events as analyzed for events 15.4.2 and 15.4.1.
A spectrum of control rod withdrawal reactivity addition rates is considered for Events 15.4.2 and 15.4.1, so the range of reactivity addition rates will be established to encompass the predicted reactivity addition rates for boron dilution events in Modes 1-6.
The operator must have sufficient time to terminate the dilution prior to reaching Tech. Spec.
shutdown margin requirements and/or losing minimum shutdown margin. These response times will be calculated for Millstone Unit 2 l
to bound the predicted critical boron concentrations.
The disposition of events for the CVCS Malfunction that Results in a Decrease in the Baron Concentration in the Reactor Coolant event is summarized in Table 15.4.6-B.
1 l
15.4.7 Inadvertent loadina and Ooeration of a Fuel Assembly in an Imorocer Position v
This event is not in the current licensing basis for Millstone Unit 2 and therefore is not analyzed.
This is shown in the Available Reactor Protection l
and Event Disposition Summary Tables 15.4.7-A and 15.4.7-B, respectively.
15.4.8 Soectrum of Control Rod Eiection Accidents l
1 f
i 15.4.8.1 Event Initiator - This accident is initiated by a falure in the I
control rod drive pressure housing which could result in the rapid ejection of a control rod.
(
i l
- 15. 4. is. 2 Event Descriotion - Ejection of the control rod from the reactor i
t
l l
ANF-87-161 j
Page 75 core results in a rapid loss of negative reactivity causing a nuclear power transient.
In addition to the power transient, the ejected rod results in a highly perturbed power distribution which, coupled with the power transient, could possibly lead to localized fuel damage.
Also, the rapid nuclear power excursion can result in a significant short term heatup of the coolant with a resultant reactor coolant system pressure increase, although on the long term the reactor coolant system will depressurize due to the break in the reactor coolant pressure boundary.
15.4.8.3 Reactor Protection - Reactor protection for the Spectrum of Control Rod Ejection Accidents is summarized in Table 15.4.8-A.
Doppler feedback inherent in the fuel also limits the nuclear power excursion.
15.4.8.4 Disoosition and Justification - This event is not a concern in Modes l
4-6 as all control rods are required to be fully inserted per Technical l
Specifications and no one control element assembly possesses enough reactivity l
worth to overcome the niinimum allowed shutdown margin.
The fuel energy f
content is maximized by starting from rated power initial conditions, so the consequences of this event are bounding for power operating initial i
conditions.
However, because of the complex interaction of the ejected rod I
worth, and ejected peaking factor (which are maximized at het critical operating conditions), and Doppler feedback effects, it is difficult to a l
priori bound the consequences of the avent for either rated power or hot critical operating conditions.
Therefore, the consequences of this event are analyzed for both rated power and hot critical operating conditions.
The analysis is performed to support a more positive BOC and a more negative EOC moderator temperature coefficient, l
In addition to the rod ejection, this event is characterized by a small break LOCA (SBLOCA) as the failure of the pressure housing is assumed to j
result in a breach of'the primary coolant pressure boundary.
The short term aspects of the event are dominated by the rod ejection, while the long term aspects are dominated by the SBLOCA.
The lin.iting SBLOCA is evaluated in j
I
ANF-87-161 l
Page 76 l
Event 15.6.5 and is typically a cold leg break.
In the rod ejection, the break is more characteristic of a hot leg break and therefore will be bounded by the SBLOCA.
Also in the rod ejection, a much earlier reactor trip occurs, resulting in lower powers and temperatures than in Event 15.6.5.
It is I
concluded that the long term aspects of the rod ejection are bounded by those of Event 15.5.5 for small breaks.
Thus, only the short term rod ejection consequences r.eed be evaluated.
Note also that the limiting 15.6.5 event l
occurs for rated power operating conditions.
The disposition of events for the Spectrum of Control Rod Ejection Accidents is sunmarized in Table 15.4.8-B.
i 15.4.9 Soectrum of Rod Dros Accidents (BWR)
Millstone Unit 2 is not a Boiling Water Reactor (BWR) and as such this l
1 event is not applicable.
I j
l f
i
]
)
l i
i i
i l
l
ANF 87-161 Page 77 TABLE 15.4.1-A AVAILABLE REACTOR PROTECTION FOR THE UNCONTROLLED CONTROL R00/ BANK WITHDRAWAL FROM A SUBCRITICAL OR LOW POWER STARTUP CON 0! TION EVENT
)
Reactor Operating Conditions Reactor Protection 1
Not considered in this section 2
Variable Overpower Trip High Pressurizer Pressure Trip Rod Withdrawal Prohibit on Variable Overpower Pre-Trip Alarm
[
i l
j 3
Variable Overpower Trip Rod Withdrawal Prohibit on Variable Overpower Pre-Trip Alarm 4-6 Not a credible event; no analysis required I
I I
l j
1 i
1
(
s s
ANF-87-161 Page 78 TABLE 15.4.1-B DISPOSITION OF EVENTS FOR THE UNCONTROLLED CONTROL ROD / BANK WITHDRAWAL FROM A SUBCRITICAL OR LOW POWER STARTUP CONDITION EVENT l
Reactor Operating Conditions Disposition 1
Not considered in the section 2
Analyze at 2000 psia 3
Bounded by the ebove 4-6 Not a credible event; no analysis required l
ANF-87-161 Page 79 TABLE 15.4.2-A AVAILABLE REACTOR PROTECTION FOR THE UNCONTROLLED CONTROL ROO/ BANK WITHDRAWAL AT POWER EVENT Reactor Operating Conditions agpctor Protection 1
Variable Overpower Trip Local Power Density Trip Thermal Margin / Lou Pressure Trip i
High Pressurizer P, essure Trip Rod Withdrawal Prohibit Action on Variable Overpowtr 01: TM/LP Pre-Trip Alarm 2-6 Not considered in this section 4
1 a
r j
t l
i l
i l
t h
l t
l
ANF-87-161 Page 80 TABLE 15.4.2-B DISPOSITION OF EVENTS FOR THE UNCONTROLLED CONTROL ROD / BANK WITHDRAWAL AT POWER EVENT Reactor Operating Conditions Disnosition 3
Analyze at rated power 2-6 No analysis required; not considered in this section
1 1
ANF 87-161 i
Page 81 i
TA8LE 15.4.3(1)-A AVAILABLE REACTOR PROTECTION FOR THE DROPPED CONTROL ROO/ BANK EVENT i
t Reactor Operating Conditions Reactor Protection 1
Variable Overpower Trip 4
i Thermal Margin / Low Pressure Trip i
Local Power Density Trip l
i Available Thermal Margin
)
i r
j 2
Variable Overpower' Trip
?
\\
Available Thermal Margin 3
i 3-6 No significant consequences for these reactor operating i
conditions i
l 1
i I
i I
i I
4 l
ANF-87-161 Page 82 TABLE 15.4.3(1) B OISPOSITION OF EVENTS FOR THE DROPPED CONTROL R00/ BANK EVENT 1
l Reactor Operating Conditions Disnosition i
1 Analyze i
2 Bounded by the above; no analysis required 36 No analysis required j
i i
l i
i k
[
.i j
- i J
)
r l
I e
I 4
I I
ANF 87-161 Page 83 TABLE 15.4.3(2)-A AVAILABLE REACTOR PROTECTION FOR THE DROPPED PART-LENGTH CONTROL R00 EVENT Reactor Operating Conditions Reactor Protection 16 Not a credible event; part-length control rods have been removed TABLE 15.4.3(2) B DISPOSITION OF EVENTS FOR THE DROPPED PART LENGTH CONTROL R00 EVENT Reactor Operating Conditions Disoosition I
l6 No analysis required; part-length control rods have been removed l
\\
ANF-87-161 Page 84 TABLE 15.4.3(3) A AVAILABLE REACTOR PROTECTION FOR THE MALPOSITIONING OF THE PART LENGTH CONTROL R00 GROUP EVENT i
nd!o Reactor Protection 16 Not a credible event; part-length control rods have been removed TABLE 15.4.3(3) B O!SPOSITION OF EVENTS FOR THE MALPOSITIONING OF THE PART-LENGTH CONTROL ROD GROUP EVENT Reactor Operating Conditions Discosition 16 Not a credible event; no analysis required
ANF-87-161 Page 85 TABLE 15.4.3(4)-A AVAILABLE REACTOR PROTECTION FOR THE STATICALLY MISALIGNED CONTROL ROD / BANK EVENT f
Reactor Operating Conditions Reactor Protection 16 No analysis required; not in licensing basis TABLE 15.4.3(4)-B DISPOSITION OF EVENTS FOR THE STATICALLY MISA1.1GNED CONTROL ROD / BANK EVENT Reactor Operating Conditions Discosition 16 No analysis required; not in licensing basis
I
\\
ANF 87-161
(
Page 86 1
TABLE 15.4.3(5)-A AVAILABLE REACTOR PROTECTION FOR THE SINGLE CONTROL R00 WITH0RAWAL EVENT a
L h
l Reactor Operating Conditions Reactor Protection 1
Variable Overpower Trip l
Local Power Density Trip Thermal Margin / Low Pressure Trip High Prossurizer Pressure Trip Rod Withdrawal Prohibit Action on i
l Variable Overpower or TM/LP Pre-Trip Alarm a
2 Variable Overpower Trip l
l High Pressurizer Pressure Trip Rod Withdrawal Prohibit on Variable
[
Over' power Pre Trip Alarm l
l I
3 Variable Overpower Trip Rod Withdrawal Prohibit on Variable l
Overpower Pre-Trip Alarm j
l 46 Not a credible event; no analysis i
required l
ANF 87-161 Page 87 TABLE 15.4.3(5) B DISPOSITION OF EVENTS FOR THE SINGLE CONTROL ROO WITH0RAWAL EVENT l
Reactor Operating Conditions Disnosition l
1 Analyze at rated power 2
Analyze at 2000 psia 3
Bounded by the above 46 Not a credible event; no analysis required l
l l
i i
1 ANF-G7-161 Page 88 TABLE 15.4.3(6) A AVAILABLE REACTOR PROTECTION FOR THE REACTIVITY' CONTROL DEVICE REMOVAL ERROR DURING REFUELING EVENT Reactor Operating Conditions Reactor Protection 16 Not a credible event TABLE 15.4.3(6) B DISPOSITION OF EVENTS FOR THE REACTIVITY CONTROL DEVICE REMOVAL ERROR DURING REFUELING EVENT Reactor Operating Conditions Disposition 16 No analysis required; not a credible event 1
l i
(
l i
ANF-87 161 Page 89 TABLE 15.4.3(7)-A AVAILABLE REACTOR PROTECTION FOR THE VARIATIONS IN REACTIVITY LOAD TO BE COMPENSATED BY BURNUP OR ON LINE REFUELING EVENT Reactor Operating Condition Reactor Protection 1-6 No analysis required 1
1 i
i s
TABLE 15.4.3(7)-B DISPOSITION OF EVENTS FOR THE VARIATIONS IN REACTIVITY LOAD TO BE COMPENSATED BY BbANUP l
OR ON LINE REFUELING EVENT l.
Reactor Operating Condition Discosition i
16 No analysis required; no significant consequences i
I.
~
i
[
ANF 87-161 l
Page 90 TABLE 15.4.4 A AVAILABLE REACTOR PROTECTION FOR THE STARTUP OF AN INACTIVE LOOP EVENT Reactor Operating i
Conditions Reactor Protection 1, 2, 6 Not applicable 35 Technical Specification requirements on shutdown margin and reactor coolant pump operation TABLE 15.4.4 B OISPOSITION OF EVENTS FOR THE STARTUP OF AN INACTIVE LOOP EVENT d
]
Reactor Operating Conditions Disposition 1, 2, 6 Not applicable l
35 No analysis required; minimal consequences m.
i ANF 87 161 Page 91 l
i i
TABLE 15.4.5 A AVAILABLE REACTOR PROTECTION FOR THE FLOW j
CONTROLLER MALFUNCTION EVENT n
Reactor Operating Conditions Reactor Protection 16 Event is not credible
,i i
l i
l 1
i
)
{
i i
1 i
TABLE 15.4.5 B DISPOSITION OF EVENTS FOR THE FLOW q
CONTROLLER MALFUNCTION EVENT j
f f
l l
j Reactor Operating Conditions Disoosition i
l6 Event is not credible; no analysis required i
a i
4
4 1
l ANF 87 161 l
Page 92 l
l TABLE 15.4.6 A AVAILABLE REACTOR PROTECTION FOR THE CVCS MALFUNCTION THAT RESULTS IN A DECREASE IN THE B0RON CONCENTRATION i
IN THE REACTOR COOLANT EVENT l
I Reactor Operating Conditions Reactor Protection 1
Local Power Density Trip l
1l Variable Overpower Trip i
Thermal Margin / Low Pressure Trip 1
j High Pressurizer Pressure Trip j
2 Variable Overpower Trip
[
3 l
High Pressurizer Pressure Trip f
36 Technical Specification Shutdown i
Margin Requirements i
i Adrinistrative Procedures i
l Operator Response Time l
l 4
i i
l l
I l
l 4
i i
l l
t r
]
t l
i
l
.I
,o ANF-87-161 Page 93 TABLE 15.4.6-B DISPOSITION OF EVENTS FOR THE CVCS MALFUNCTION THAT RESULTS IN A DECREASE IN THE BORON CONCENTRATION IN THE REACTOR COOLANT EVENT Reactor Operating Conditions Disposition 1-6 Analyze for loss of shutdown margin l
u.
ANF-87-161~
Page 94 TABLE 15.4.7-A AVAILABLE REACTOR PROTECTION FOR THE INADVERTENT LOADING AND OPERATION OF A FUEL ASSEMBLY IN AN IMPROPER POSITION EVENT Reactor Operating Conditions Reactor Protection 1-6 Not in licensing basis TABLE 15.4.7-B DISPOSITION OF EVENTS FOR THE INADVERTENT LOADING AND OPERATION OF A FUEL ASSEMBLY IN AN IMPROPER POSITION EVENT Reactor Operating Conditions Discosition l
16 Not in licensing basis l
i l
L ----
ANF-87-161 Page 95 l
TABLE 15.4.8-A AVAILABLE REACTOR PROTECTION FOR Tile SPECTRUM OF CONTROL R00 EJECTION ACCIDENTS Reactor Operating Conditions Reactor Protection 1
Variable Overpower Trip Thermal Margin / Low Pressure Trip High Pressurizer Pressure Trip 2
Variable Overpower Trip High Pressurizer Pressure Trip 3
Variable Overpower Trip 4-6 No reactor protection required; ejected rod worth less 'than the Technical Specification minimum shutdown margin.
No significant consequence for this operating condition.
\\
ANF-87-161 i
Page 96 i
TABLE 15.4.8-8 DISPOSITION OF EVENTS FOR THE SPECTRUM 0F CONTROL R0D EJECTION ACCIDENTS i
Reactor Operating l
Conditions Disoasition 1
Analyze for short term response.
Long term bounded by Event 15.6.5.
2, 3 Analyze 4-6 No analysis required i
l i
l i
1 r
l
ANF-87-161 Page 97 TABLE 15.4.9-A AVAILABLE REACTOR PROTECTION FOR THE SPECTRUM 0F R0D DROP ACCIDENTS (BWR)
Reactor Operating Conditions Reactor Protection 1-6 Eeent is not applicable.
TABLF 15.4.9-B DISPOSITION OF EVENTS FOR THE SPECTRUM 0F R00 DROP ACCIDENTS (BWR)
Reactor Operating Conditions Discosition 1-6 Event is not applicable; no analysis required.
l s
ANF-87-161 Page 98 15.5 INCREASES IN REACTOR COOLANT SYSTEM INVENTORY 15.5.1 Inadvertent Ooeration of the ECCS That increases Reactor Coolant Inventory This event is not in the current licensing basis for Millstone Unit 2 and therefore is not analyzed.
This is shown in the Available Reactor Protection and Event Disposition Summary Tables 15.5.1-A and 15.5.1-5, respectively.
15.5.2 CVCS Malfunction That increases Reactor Coolant Inventory This event is not in the current licensing basis for Millstone Unit 2 and therefore is not analyzed.
This is shown in the Available Reactor Protection and Event Disposition Sumary Tables 15.5.2-A and 15.5.2-B, respectively.
The potential consequences of diluting the primary system boron concentration are addressed in Event 15.4.6.
l l
l l
i
(
j ANF-87-161 Page 99
]
TABLE 15.5.1-A AVAILABLE REACTOR PROTECTION FOR THE INADVERTENT OPERATION OF THE ECCS THAT INCREASES REACTOR COOLANT.
INVENTORY EVENT P
1 Reactor Operating Conditions Reactor Protection 1-6 Not in licensing basis; not analyzed.
l TABLE 15.5.1-B DISPOSITION OF EVENTS FOR THE INADVERTENT OPERATION OF THE ECCS THAT INCREASES REACTOR COOLANT INVENTORY EVENT Reactor Operating C.qaditions Disoosition 1-G Not in licensing basis; not analyzed, t
I l
{
ANF-87-161 l
Page 100 l
TABLE 15.5.2-A AVAILABLE REACTOR PROTECTION FOR THE CVCS MALFUNCTION THAT INCREASES REACTOR COOLANT INVENTORY EVENT Reactor Operating Conditions Reactor Protection 1-6 Not in licensing basis; not analyzed.
TAbuE 15.5.2-B DISPOSITION OF EVENTS FOR THE CVCS MALFUNCTION THAT INCREASES REACTOR COOLANT INVENTORY EVENT t
Reactor Operating Conditions Disoosition 1-6 Not in licensing basis; not analyzed.
l i
l
- =_
s f
ANF-87-161 Page 101 15.6 DECREASES IN REACTOR COOLANT INVENTORY 15.6.1 Inadvertent Ooenina of a PWR Pressurizer Pressure Relief Valve 15.6.1.1 Event Initiator - The event is postulated to occur as a result of the inadvertent opening of a pressurizer pressure relief or safety valve due to an electrical or mechanical failure.
The limiting event is obtained by assuming the inadvertent opening of both pressurizer power operated relief valves.
15.6.1.2 Evec.t Descriotion - The event initiator results in a blowdown of primary coolant as steam through the faulted valves.
Primary system pressure drops rapidly until the pressurizer liquid is depleted, and then quite rapidly to a pressure determined by the saturation curve at the temperature of the coolant in the upper vessel head.
Reactor scram will occur on thermal margin / low pressure before the pressurizer liquid is depleted, terminating the challenge to SAFDLs.
In this initial stage, pressurizer heaters would actuate in an attempt 3 maintain pressure, but would be turned off on a low level signal before the heater elements were uncovered.
15.6.1.3 Reactor Protection - The thermal margin / low pressure trip provides initial protection against loss of thermal margin and possible fuel damage.
Reactor protection for the Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve event is summarized in Table 15.6.1-A.
15.6.1.4 Disoosition and Justification The event proceeds as a
depressurization of the primary coolant system with a loss of inventory. The core power and primary loop temperatures are relatively unaffected by the pressure drop.
Thus, a short term challenge to the SAFDLs exists due to the depressurization prior to scram.
There is also a long term conce n in that if primary inventory cannot be restored and maintained, core uncowry may result.
ANF-87-161 Page 102 The greatest challenge to core uncovery exists at rated power conditions-when the core power and primary coolant stored energy are maximized.
The greatest challenge to the SAFDLs occurs for the event initiated at rated power where the margin to DN8 is minimized.
This analysis will support increased radial peaking.
An evaluation of the SAFDL challenge will also be made for 5% power operating conditions in Mode 2 when the TM/LP trip may be bypassed.
In this mode, the primary system may depressurize below the TM/LP setpoint pressure without an automatic reactor trip occurring.
The Safety Injection System will, however, be available to inject boron and provide for inventory makeup.
The disposition of events for the Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve event is summarized in Table 15.6.1-B.
15.6.2 Radiolooical Conseauences of the Failure of Small lines Carrvina Primary Coolant Outside of Containment The disposition of this event is provided in Section 14.1.4 of the Updated Millstone Unit 2 FSAR, Reference 4.
This disposition is not dependent on either fuel
- type, power distribution, or reactor protection system l
modifications.
It is therefore not affected by the current licensing action j
and remains valid for this event.
This is reflected in the Available Reactor Protection and Event Disposition Summary Tables 15.6.2-A and 15.6.2-B, respectively.
15.6.3 Radioloaical Conseauences of Steam Generator Tube Failure 15.6.3.1 Event initiator - This event is initiated by the complete severance of a single generator tube.
15.6.3.2 Event Descriotion Experience with nuclear steam generators indicates that the probability of complete severance of a tube is small.
The more probable modes of failure are those involving the occurrence of pinholes
ANF-87-lG1 Page 103 or small cracks in the tubes, and of cracks in the seal welds between the tubes and tube sheet.
A leaking steam generator tube would allow transport of primary coolant into the main steam system.
Radioactivity contained in the primary coolant would mix with shell side water in the affected steam generator.
Some of this radioactivity would be transported by steam to the turbine and then to the condenser.
Noncondensible radioactive materials would then be passed to atmosphere through the condenser air ejector discharge via the Plant stack.
The radioactive products would be sensed by the condenser air. ejector radiation monitor or the stack radiation monitor.
These monitors have audible alarms that will be annunciated in the control room to alert the operator to abnormal activity levels so that corrective action could be taken.
The behavior of the systems will vary depending upon the size of the steam generator tube failure. For small leaks the chemical and volume control charging pumps will be able to maintain the necessary primary coolant inventory and an automatic reactor trip will not occur.
The gaseous fission products will be released from the main steam system at the air ejector discharge and will be discharged via the Plant stack.
Nonvolatile fission products will tend to concentrate in the water of the steam generators.
For leaks larger than the capacity of the charging pumps, the pressurizer water level and pressure will decrease and a reactor trip will occur.
Upon i
reactor trip, the turbine will trip and the steam system atmospheric dump j
valves and the turbine bypass valve will open.
In this case it is possible l
that in addition to the noble fission gases a substantial amount of the j
radioiodines contained in the secondary system may also be released through j
the steam dump valves.
The amount of radioactivity released increases with break size.
For this analysis, a double-ended break of one tube was assumed.
The selection of one
ANF-87-161 Page 104 double-ended break as an upper limit is conservatively based upon the experience obtained with other steam generators.
No double-ended failure has ever occurred.in such units.
15.6.3.3 Reactor Protection - The leak rate through the double-ended rupture of one tube is greater than the maximum flow available from the charging pumps; therefore, the Primary Coolant system pressure will decrease and a low pressurizer pressure trip or thermal margin / low-pressure trip will occur.
The thermal margin trip has a low-pressure floor, set at 1,750 psia, below which trip will always occur.
Following the reactor trip the Primary Coolant System is cooled down by exhausting steam through the atmospheric steam dump valves and turbine bypass valve.
The radioactivity exhausted through the steam dump valves passes directly to atmosphere.
The radioactivity exhausted through the bypass valve flows to the condenser where the gaseous products remaining are vented to the atmosphere through the condenser air ejector and Plant stack.
Reactor protection for the Radiological Consequences of Steam Generator Tube Failure event is summarized in Table 15.6.3-A.
15.6.3.4 Disposition and Justification - The radiological consequences of a steam generator tube rupture incident are maximized at rated power operation due to the stored energy in the primary coolant which must be removed by the intact steam generator in order to bring the primary and secondary systems into pressure equilibrium terminating the primary to secondary leak.
The only proposed licensing action which would impact the radiological conse-quences of this event is the allowance of a slightly more negative E0C moderator temperature coefficient.
However, since this event is basically a depressurization of the primary at power, there is very little reactivity feedback to affect power.
- Thus, the change in moderator temperature coefficient will have an insignificant effect on the transient results.
The radiological consequences of record will therefore remain bounding for this event.
ANF-87-161 Page 105 The challer,3e to the SAFDLs exists due to the depressurization prior to scram.
As such, this challenge is very similar to that which exists due'to the inadvertent opening of a pressurizer relief valve (Event 15.6.1).
Since the depressurization rates associated with Event 15.6.1 are substantially larger than those encountered for this event, the corresponding pressure undershoot will also be greater.
Event 15.6.1 will thus be characterized by lower pressures at the time of MONBR than those obtained for this event.
Therefore, the DNB aspects of this event will be bounded by those of Event 15.6.1.
The disposition of events for the Radiological Consequences of Steam Generator Tube Failure event is summarized in Table 15.6.3-B.
15.6.4 Radioloaical Conseauences of a Main Steam Line Failure Outside Containment (BWR)
This event is only applicable to Boiling Water Reactors (BWRs). As such, this event is not applicable to Millstone Unit 2.
15.6.5 Loss of Coolant Accidents Resultino from a Soectrum of Postulated Pioina Breaks Within the Reactor Coolant Pressure Boundary 15.6.5.1 Event Initiator - This event is initiated by a breach in the Primary Coolant System pressure boundary.
Basically, a range of break sizes from small leaks up to a complete double-ended severance of a Primary Coolant System pipe must be considered.
Typically, these breaks are classified as Large Breaks or Small Breaks.
l 15.6.5.2 Event Descriotion (1)
Laroe Breaks. The large Break LOCA events are characterized by four sequential phases: 1) blowdown, 2) refill, 3) reflood, and 4) long tenu
(
cooling.
l The blowdown phase immediately follows the initiation of a large break.
ANF-87-161 Page 106 Primary system water is discharged through the break into containment.
The' system pressure decreases rapidly during the initial subcooled blowdown.
As the saturation pressure is approached, local boiling and flashing takes place j
in the core and the reactor goes subcritical via the negative moderator reactivity feedback.
The blowdown flow becomes a water-vapor mixture.
The depressurization rate is reduced when core pressure falls below the saturation pressure. The water level continues to decrease until a large amount of water from the ECC1 passive accumulators reaches the lower plenum.
The refill chase starts when the accumulator water begins to fill the lower plenum.
At this time, the core is uncovered by water and the fuel rods are c]oled primarily by thermal radiation.
The reflood chase begins when the water level reaches the bottom of the core.
The lona term coolino Dbase starts after the core has quenched to the point where the zircaloy-water reaction is suppressed, or the water level covers the active fuel.
During this phase, the water inventory is controlled by the safety injection pumps.
The continuous operation of these pumps ensures the long term dissipation of the decay heat.
(2) imall Breaks. The small break LOCA, as generally defined, includes any 2
break in the pressure boundary that has an area of 0.5 ft or less.
The principal PWR design feature for mitigating the consequences of a small break LOCA is the ECCS which maintains the water inventory.
Its major subsystems for restoring water inventory are the high pressure safety injection (HPSI) system, and the low pressure safety injection (LPSI) system and the safety injection (accumulator) tanks.
A small break LOCA is characterized by slow RCS depressurization rates and mass transfer rates within the RCS relative to similar parameters calculated for large break LOCA.
If the break area is large enough that the HPSI pumps cannot maintain the reactor coolant inventory and allow RCS
ANF-87-161 Page 107 pressure control, the RCS will depressurize.
The depressurization produces a low pressurizer pressure (therma, margin / low pressure) reactor trip and a safety injection actuation signal (SIAS).
The rate of RCS depressurization following SIAS depends on the break area and the HPSI shutoff head.
With a combination of a very small break and a sufficiently high HPSI shutoff head, the depressurization may be arrested.
If the break area is sufficiently large to allow continued depressurization and net loss of coolant inventory even with the HPSI pumps in operatien, the coolant level in the reactor vessel may recede below the top of the reactor core.
If sufficient steam is produced in the RCS, natural circulation (the reactor coolant pumps will have been tripped by this time to reduce coolant loss out of the break) around the RCS loop:; will cease.
Eventually, loss of reactor coolant inventory is arrested by ECCS flow exceeding the flow out the break.
In either case, the coolant level within the reactor vessel will rise, and the RCS will eventually be refilled (although leaking).
15.6.5.3 Reactor Protection (1) Larae Breaks.
Basically no credit is taken for a reactor trip by the Reactor Protection System (RPS) due to the rapid depletion of the moderator which shuts down the reactor core almost immediately, followed by ECCS injection which contains sufficient boron to maintain the reactor core in a subcritical configuration.
Technical Specification limits on hot rod power serve to limit the peak cladding temperature.
(2) Small Breaks.
Primary reactor protection for this event is provided by the low pressurizer pressure (low pressure / thermal margin) trips and the Safety injection Actuation Signal (SIAS) on a low pressurizer pressure signal.
ANF-87-161 Page 108 Reactor protection for the Loss of Coolant Accidents Roulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary event is summarized in Table 15.6.5-A.
Section 15.6.5 of Reference 1 15.6.5.4 Disoosition and Justification indicates that the primary acceptance criteria for this event are to limit offsite doses, to limit fuel clad oxidation, and to keep peak cladding temperatures below 2200'F.
Offsite doses are maximized by assuming the highest concentration of radionuclides contained within the fuel pins at event initiation.
This is accomplished by assuming steady state radionuclide concentrations characteristic of long term operation of the plant at full power.
Fuel pin cladding temperatures and oxidation rates are maximized by initiating the event with th'e highest cladding temperatures and linear heat generation rates.
Thus, the most limiting results for this event (both large and small break sizes) are obtained with the plant operating at full power in Mode 1.
These results will bound those from Modes 2-6.
The parameters which are changing from those which were used in the reference LOCA analysis presented in Sections 14.15.3 and 14.15.4 of the Updated Millstone Unit 2 FSAR (Reference 4) are the radial peaking factor and the loading of ANF supplied reload fuel.
The radial peaking factor is increasing.
Both the large and small break LOCA events will be reanalyzed.
Disposition of events for the loss of Coolant Accidents Resulting from a Spectrum of Postulited Piping Breaks Within the Reactor Coolant Pressure Boundary event is summarized in Table 15.6.5-B.
l ANF-87-161 Page 109 TABLE 15.6.1-A AVAILABLE REACTOR PROTECTION FOR THE INADVERTENT OPENING OF A PWR PRESSURIZER PRESSURE RELIEF VALVE EVENT Reactor Operating Conditions Reactor Protection 1
Thermal Margin / Low Pressure Trip Safety Injection Actuation Signal 2, 3 Safety Injection Actuation Signal Available Thermal Margin 4
Available Thermal Margin 5, G No significant consequences for these reactor operating conditions
ANF-87-161 Page 110 TABLE 15.6.1-B DISPOSITION OF EVENTS FOR THE INADVERTENT OPENING OF A PWR PRESSURIZER PRESSURE RELIEF VALVE EVENT Reactor Operating Conditions Discosition 1, 2 Analyze for DNBR 3-6 Bounded by the above
ANF-87-161 Page 111 Table 15.6.2-A AVAILABLE REACTOR PROTECTION FOR THE RADIOLOGICAL CONSEQUENCES OF THE FAILURE OF SMALL LINES CARRYING PRIMARY COOLANT OUTSIDE OF CONTAINMENT EVENT Reactor Operating Conditions Reactor Protection 1-6 None required; not a credible event for this plant.
TABLE 15.6.2 8 DISPOSITION OF EVENTS FOR THE RADIOLOGICAL CONSEQUENCES OF THE FAILURE OF SMALL LINES CARRYING PRIMARY COOLANT OUTSIDE OF CONTAINMENT EVENT Reactor Operating Conditions Discosition 16 Disposition of record provided in Section 14.1.4 of the Updated Millstone Unit 2 FSAR, Reference 4; not a credible event.
ANF-87-161 Page 112 TABLE 15.6.3-A AVAILABLE REACTOR PROTECTION FOR THE RADIOLOGICAL CONSEQUENCES OF STEAM GENERATOR TUBE RUPTURE EVENT Reactor Operating Conditions Reactor Protection 1
Thermal Margin / Low Pressure Trip Safety Injection Actuation Signal 2, 3 Safety injection Actuation Signal 4-6 No significant consequences for these reactor operating conditions
ANF-87-161 Page 113 TABLE 15.6.3-B DISPOSITION OF EVENTS FOR THE RADIOLOGICAL CONSEQUENCES OF STEAM GENERATOR TUBE RUPTURE EVENT Reactor Operating Conditions Disposition 1
Bounded by the analysis of record and by that of Event 15.6.1
(
26 Bounded by the above
ANF-87-161 Page 114 TABLE 15.6.4-A AVAILABLE REACTOR PROTECTION FOR THE RADIOLOGICAL CONSEQUENCES OF A MAIN STEAMLINE FAILURE OUTSIDE CONTAINMENT (BWR) EVENT Reactor Operating Conditions Reactor Protection 1-6 Not applicable; a Boiling Water Reactor (BWR) event TABLE 15.6.4-8 DISPOSITION OF EVENTS FOR THE RADIOLOGICAL CONSEQUENCES OF A MAIN STEAMLINE FAILURE OUTSIDE CONTAINMENT (BWR) EVENT Reactor Operating Conditions Diseosition 1-6 Not applicable; a Boiling Water Reactor (BWR) event i
ANF-87-161 Fage 115 TABLE 15.6.5-A AVAILABLE REACTOR PROTECTION FOR THE LOSS OF COOLANT ACCIDENTS RESULTING FROM A SPECTRUM 0F POSTULATED PIPING BREAXS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY Reactor Operating Conditions Reactor Protection 1
Larae Breaks -
No credit taken for reactor trip by the Reactor Protection System (RPS)
ECCS - short and long term cooling 1
Small Breaks -
Thermal Margin / Low Pressure Trip low Reactor Coolant Flow Trip Safety Injection Actuation Signal 2
Larae Breaks -
No credit taken for reactor trip by the Reactor Protection System (RPS)
ECCS - short and long term cooling 2
Small Breaks -
Safety Injection Actuation Signal 3-6 No significant consequences for these reactor operating conditions
ANF-87-161 Page 116 TABLE 15.6.5-B DISPOSITION OF EVENTS FOR THE LOSS OF COOLANT ACCIDENTS RESULTING FROM A SPECTRUM 0F POSTULATED PIPING BREAXS WITHIN THE REACTOR COOLANT PRESSURE B0UNDARY Reactor Operating Conditions Discosition 1
Larae Break -
{
Analyze 1
Small Break -
Analyze 2-6 Larae Break -
Bounded by the event initiated from Mode 1 2-6 Small Break -
Bounded by the event initiated from Mode 1
ANF-87-161 Page 117 15.7 RADI0 ACTIVE RELEASES FROM A SUBSYSTEM OR COMPONENT 15.7.1 Waste Gas System Failure The results of this event are not dependent on either fuel type, power distribution, or reactor protection system modifications.
The reference analysis is therefore not affected by the current licensing action and remains the bounding analysis for this event.
The reference analysis is provided in the Updated Millstone Unit 2 FSAR, Reference 4.
15.7.2 Radioactive Liouid Waste System Leak or Failure (Release to Atmosohere), and 15.7.3 Postulated Radioactive Releases Due to Lioutd-Containina Tank Failures These events are not in the current licensing basis for Millstone Unit 2 and therefore are not analyzed.
Further, Event 15.7.2 has been deleted from the SRP, Reference 1.
15.7.4 Radioloaical Conseauences of Fuel Handlina Accident 15.7.4.1 Event Initiator The event is initiated by a mishap either in containment or in the auxiliary building.
The mishap results in the fuel assembly being dropped, causing damage to the fuel pins.
15.7.4.2 Event Descriotion - For the purpose of defining the upper limit on fuel damage as the result of a fuel handling incident, it is assumed that the fuel assembly is dropped during handling.
Interlocks, procedural and administrative controls make such an event unlikely.
However, if an assembly is damaged to the extent that a number of fuel rods fail, the accumulated fission gases and iodines in the fuel element gap could be released to the surrounding water.
Release of the fission products to the surrounding water is considered negligible as a result of reduced diffusion through the fuel due to the low fuel temperature during refueling.
ANF-87-161 Page 118 15.7.4.3 Reactor Protection No reactor protection is required for this event.
This is summarized in the reactor protection for the Radiological Consequences of Fuel Handling Accident event in Table 15.7.4-A.
15.7.4.4 Discosition and Justification Limiting analyses for the radiological consequences of fuel handling accidents both in the containment building and in the auxiliary building are presented in Section 14.19 of the FSAR (Ref. 4).
The evaluation of the limiting modes of fuel bundle failure are unaffected by a fuel reload and thus remain bounding for Cycle 10.
The radiological consequences of a fuel bundle failure were evaluated using methodology given in TID-14844(7) and are thus only a function of core power.
None of the inputs to this analysis are impacted by a fuel reload, and thus the radiological consequences of the analysis of record also remain bounding for Cycle 10.
The disposition of the Radiological Consequences of Fuel Handling Accident event is summarized in Table 15.7.4-8.
15.7.5 Soent Fuel Cask Droo Accidents This event is not in the current licensing basis for Millstone Unit 2 and therefore is not analyzed.
This'is shown in the Available Reactor Protection and Event Disposition Summary Tables 15.7.5-A and 15.7.5 8, respectively.
ANF-87-161 Page 119 TABLE 15.7.1-A AVAILABLE REACTOR PROTECTION FOR THE WASTE GAS SYSTEM FAILURE EVENT Reactor Operating Conditions Reactor Protection 1-6 Reactor Protection System (RPS) action not required TABLE 15.7.1 B OISPOSITION OF EVENTS FOR THE WASTE GAS SYSTEM FAILURE EVENT Reactor Operating Conditions Discosition 16 Bounding analysis presented in Section 14.17 of the Updated Hillstone Unit 2 FSAR, Reference 4
ANF 87-161 Paga 120 TABLE 15.7.2-A AVAILABLE REACTOR PROTECTION FOR THE PADI0 ACTIVE LIQUID WASTE SYSTEM LEAK OR FAILURL-(RELEASE TO ATMOSPHERE) EVENT.
Reactor Operating Conditions Reactor Protection 4
1-6 Not in lic6nsing basis; nn analysis required'.
l 1
TABLE 15.7.2-8 DISPOSITION OF EVENTS FOR THE RADIOACTIVE LIQUID WASTE SYSTEM LEAK OR FAILURE (RELEASE TO ATMOSPHERE) EVENT i
Reactor Operating Condition Disoosition 1-6 Not in licensing basist no analysis required.
1 a
ANF-87-161 Page 121 TABLE 15.7.3-A AVAILABLE REACTOR PROTECTION FOR THE POSTULATED RADIDACTIVE RELEASES DUE TO LIQUID-CCNTAINING TANK FAILURES EVENT Reactor Operating Conditions Reactor Protection 16 Not in licensing basis; no analysis required.
i 4
TABLE 1S.7.3-8 DISPOSITION OF EVENTS FOR THE POSTULATED Rt' LEASES,D'UE TO LIQUID CONTAINING TANK FAILURES EVENT 1
I Reactor Operating Condition Disoosition 1-6 Not in licensing basis; no analysis required.
ANF-87-161 Page 122 i
TABLE 15.7.4-A AVAILABLE REACTOR PROTECTION F0P IHE RADIOLOGICAL CONSEQUENCES OF FUEL HANDLING ACCIDENTS Reactor Operating Conditions Reactor Protection 1-6 Reactor Protection System (RPS) action not required 1
J l
l
]
TABLE 15.7.4-B DISPOSITION OF EVENTS FOR THE RADIOLOGICAL CONSEQUENCES OF FVEL HANDLING ACCIDENTS i
A Reactor Operating Conditions Disposition 16 Bounding analytig presented in Section 14.19 of the Updated Millstone Uni' 2 FSAR, Ref. 4 1
ANF-87-161 Page 123 TABLE 15.7.5-A AVAILABLE REACTOR PROTECTION FOR THE SPENT FUEL CASK DROP ACCIDENTS Reactor 0porating Conditions Reactor Protectiqn 1-6 Not in licensing basis; not analyzed.
TABLE 15.7.5-B "ISPOSITION OF EVENTS FOR THE SPENT FUEL CASK DROP ACCIDENTS Reactor Operating Conditions Discorition l
l-6 Not in licensing basis; not l
analyzed.
I 1
ANF-87-161 Page 124 4.0 NILLSTONE UNIT 2 FSAR EVENTS NOT CONTAINED IN THE STANDARD RFVIEW PLAN 4.1 EFFECTS OF EXTERNAL EVENTS i
3 4.1.1 Event Initiator The external events affecting the plant which were included in this category are as follows:
High and Low Water; Storm:;
1 Tornadoes; and e
j Earthquakes.
I 4.1.2 Event Descriotion The location of the buildings in this installation on the shore of Long Island Sound exposes them to several different natural events of varying intensity. These events include storms and tornadoes and their effects on the water level in the Sound.
Also included, are the effects of seismic tremors which could occur in this region.
Normal fluctuations in the bay water level caused by tides are very 1
predictable and insignificant.
Coupled, however, with strong winds, the water j
level could rise or fall appreciably more than usual.
This may affect the 1
)
intake to the safety-related service water cooling pumps.
i l
The meteorology of the Millstone site is basically that of a seacoast transition zone which lies along a major storm track of extra-tropical cyclones and an occasional storm of tropical origin. These storms, along with seasonal thunderstorms, will produce intense rainfall and (during the winter) snow and freezing rain storms.
These storms can produce high wind and snow loads on structures.
Tornadoes have been reported due to occasional severe storms.
l
ANF 87-161 Page 125 The earthquake history of the Southern New England area has been compiled to ascertain the maximum expected seismic acceleration.
4.1.3 Reactor Protection No reactor protection is required for these events.
4.1.4 Disoosition and Justification l
A fuel reload will not affect any of the inputs to this analysis.
Therefore, the analysis of record remains bounding for this event.
1 4.2 FAILURES OF E0VIPMENT pROVIDING JOINT CONTROL AND SAFETY FUNCTIONS l
There is no equipment at Millstone Unit 2 which provides a joint control function and safety function.
Therefore, this event is not credible.
l 4.3 CONTAINHENT PRESSURE ANALYSIS l
l l
4.3.1 Event initiator This event is initiated by a breach in the primary coolant system pressure boundary.
Basically, a range of break sizes and types in both the l
hot legs and cold legs are considered.
The event initiator is that break l
which results in the greatest mass and energy release to the containment.
4.3.2 Event Descriotion In the event of a LOCA, the release nf primary coolant from the rupture area will cause the high pressure, high temperature liquid to enter the containment, rapidly flashing to steam and water within the containment.
The addition of this mass and energy to the containment will result in a rise in both the pressure and temperature of the containment atmosphere.
The containment building design is based on the mass and energy absorption capabilities of the volume enclosed by the containment structure.
A spectrum of break sizes for both the hot and cold primary coolant legs has been considered in the evaluation of the containment design to detemine
ANF-87-161 Page 126 the most severe combination of reactor system mass and energy releases, sensible heat sources, and shutdown heat sources during the blowdown and reflood phases of the LOCA.
Following the time of peak containment pressure, the safety injection systems and containment cooling systems reduce the containment pressure and temperature until the Refueling Water Storage Tank (RWST) water supply is exhausted.
4.3.3 Reactor Protection l
This event considers the adequacy of the containment design to a LOCA event and, as such, does not address the need for reactor protection.
The adequacy of available reactor protection is addressed in Event 15.6.5.
t 4.3.4 Discosition and Justification t
A fuel reload will not affect the critical parameters of primary pressure, average primary coolant temperature or power which govern the mass and energy release to the containment.
Since the fuel bundle design will not t
change substantially, the fuel performance during the event will be unaffected.
Therefore, the analysis of record will remain bounding for this
)
event.
i 4.4 HYDROGEN ACCUMULATION IN CONTAINMENT I
i 4.4.1 Event Initiator l
The event initiator is the Design Basis Incident (DBI).
The DBI is the
^
j limiting Section 3 event producing the greatest mass and energy release to the l
containment, i
i i
4.4.2 Event Descriotion The significant sources of hydrogen following the Design Basis Incident are radiolysis of water from the decay of fission products, the zirconium-water reaction of the fuel cladding and corrosion of containment metals.
If i
unchecked, the hydrogen generation may become great enough so that a flammable concentration would be present in the containment atmosphere.
An electric l
ANF-87-161 Page 127 4
hydrogen recombiner system is provided to reduce containment hydrogen concentrations.
4.4.3 Reactor Protection No reactor protection systems are required for this event.
4.4.4 Disoosition and Justification The cladding thickness for ANF supplied fuel is slightly greater than that for either Combustion Engineering or Westinghouse fuel. Thus, the amount of zirconium clad in the core will increase slightly with ANF fuel (by less than 8%), resulting in a slight increase in the amount of hydrogen assumed to be generated from the zirconium-water reaction.
The analysis of record has shown that with one of two hydrogen recombiners in operation, the maximum hydrogen concentration is well below the flammability limit.
This limit will not be challenged by the expected increase in hydrogen concentration due to the reload fuel.
I The analysis of record also evaluated the expected offsite doses resulting from a postulated containment purge operation, assuming failure of both hydrogen recombiners.
Again, the resulting peak concentration level was well below the flammability limit and will not challenge the limit with ANF reload fuel.
The offsite dose analysis was performed using TID-14844(7) methodology.
The proposed changes for Cycle 10 will not impact this methodology, and the current offsite dose analysis remains valid for Cycle 10.
We conclude that offsite doses will not exceed those in the analysis of record and that hydrogen flammability limits will not be challenged.
A fuel reload will not affect any of the other inputs to this calculation.
Thus, no reanal'ysis is performed.
ANF-87-161 2
Page 128 4.5 RADIOLOGICAL CONSEOUENCES OF THE DESIGN BASIS INCIDENT (DBI) 4.5.1 Event Initiator The event initiator is the Design Basis Incidant (DBI).
The DBI is the limiting Section 3 event producing the greatest mass and energy release to the containment.
4.5.2 Event Descriotion The DBI is assumed to result in a gross release of radioactivity from the fuel to the containment. The activity is assumed to leak from the containment directly to the atmosphere, into the surrounding enclosure building and into the control room. This results in offsite doses to the general population and can have an adverse effect on control room habitability.
l 4.5.3 Reactor Protection No reactor protection systems are required for this event.
4.5.4 Disoosition and Justification This event assumes a gross release of radioactivity from the fuel to the containment building and ultimately to the atmosphere.
Thus, the consequences of the event are driven by the assumed fission product inventory contained within the fuel.
The analysis of record presented in Section 14.20 of Reference 4 evaluated the radiological consequences using methodology given in
(
TID 14844(7), and are thus only a function of core power.
None of the inputs to this analysis are impacted by a fuel reload, and thus the radiological consequences of the analysis of record remain bounding 4r Cycle 10.
l
i i
ANF 87-161 Page 129 TABLE 4.1-A AVAILABLE REACTOR PROTECTION FOR THE EFFECTS OF EXTERNAL EVENTS l
Reactor Operating Condition Reactor Protection 1-6 None required.
i l
i r
TABLE 4.1-B DISPOSITION OF EVENTS FOR THE EFFECTS OF EXTERNAL EVENTS i
Reactor Operating Condition Disposition 1-6 Bounded by the analysis of record, l
1 l
ANF-87-161 Page 130 TABLE 4.2-A AVAILABLE REACTOR PROTECTION FOR THE FAILURES OF EQUIPMENT PROVIDING JOINT CONTROL AND SAFETY FUNCTIONS EVENT Reactor Operating Condition Reactor Protection 1-6 Not a credible event; no analysis required.
i 4
I f
l l
l TAB' E 4.2 8 DISPOSITION OF EVENTS FOR THE FAILURES OF EQUIPMENT PROVIDING JOINT CONTROL AND SAFETY FUNCTIONS EVENT 1
I i
Reactor Operating i
Condition Disposition I
I-6 Not a credible event; no analysis j
required.
i l
l l
1
1 ANF-87-161 Page 131 TABLE 4.3-A AVAILABLE REACTOR PROTECTION FOR THE CONTAINMENT PRESSURE ANALYSIS EVENT Reactor Operating Condition Reactor Protection 1-6 Addressed in Event 15.6.5 1
i i
TABLE 4.3-B DISPOSITION OF EVENTS FOR THE i
CONTAINMENT PRESSURE ANALYSIS EVENT Reactor Operating Condition Disoosition
(
l-6 Bounded by analysis of record.
ANF-87-161 Page 132 TABLE 4.4-A AVAILABLE REACTOR PROTECTION FOR THE HYDROGEN ACCUMULATION IN CONTAINMENT EVENT Reactor Operating Condition Reactor Protection 16 None required.
TABLE 4.4-8 DISPOSITION OF EVENTS FOR THE HYDR 0 GEN ACCUMULATION IN CONTAINMENT EVENT Reactor Operating Condition Disoosition l-6 No analysis required; analysis of i
record provides offsite dose calcula-l tion; flammability limits not challenged
i ANF-87-161 Page 133 I
TABLE 4.5-A AVAILABLE REACTOR PROTECTION FOR THE RADIOLOGICAL CONSEQUENCES OF THE DESIGN BASIS INCIDENT EVENT Reactor Operating Condition Reactor Protection 1-6 None required.
TABLE 4.5-8 DISPOSITION OF EVENTS FOR THE RADIOLOGICAL CONSEQUENCES OF THE DESIGN BASIS INCIDENT EVENT i
Reactor Operating Condition Disnosition 16 Bounded by analysis of record.
ANF-87-161 Page 134
5.0 REFERENCES
1.
"Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," NUREG-0800, U.S. Nuclear Regulatory Commission, July 1981.
2.
"Exxon Methodology for Pressurized Water Reactors - Analysis of Chapter 15 Events," ANF-84-73fP). Rev. 3. Advanced Nuclear Fuels Corp.
3.
Technical Specifications for Millstone Unit 2 Docket No. 50-336, Updated i
through Amendment No. 116.
4.
Millstre Unit 2. Updated Final Safety Analysis Report.
5.
"Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations,"
XN NF-82-21(A),
Exxon Nuclear Company, i
September 1983.
6.
"XCOBRA-!!!C: A Computer Code to Determine the Distribution of Coolant During Steady-State and Transient Core Operation,"
XN-NF-75-21(A).
Revision 2 Exxon Nuclear Company, r
7.
"Calculation of Distance Factors for Powar and Test Reactor Sites,"
TID 14844, Reactor Technology, Division of Licensing and Regulation, U.S.
Atomic Energy Commission, March 23, 1962.
j i
f i
)
i l
l i
l 1