ML20148L428
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| ML20148L428 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/14/1988 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20148L418 | List: |
| References | |
| 50-423-88-03-EC, 50-423-88-3-EC, NUDOCS 8804010272 | |
| Download: ML20148L428 (46) | |
See also: IR 05000423/1988003
Text
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.: 50-423/88-03
Docket No.: 50-423
License No.: NPF-49
Licensee: Northeast Nuclear Energy Company
P. O. Box 270
Hartford, Connecticut 06101-0270
Facility: Millstone Neclear Power Station, Unit 3
Meeting Location: NRC Region I, King of Prassia, Pennsylvania
Meeting Date: March 8, 1988
Approved by: $0 N, 9/d/88
_ _ _E. C. McCabe, Chief, Reactor Projects Section IB Date
Meeting Summary: The Enforcement Conference was held to discuss the findings of
Inspection Report 50-423/88-03. The topics discussed at the meeting dealt with
the root cau;e analysis and corrective actions for an overpressure transient that
occurred at low temperature without the required overpressure protection systems
being in service. Additional topics were discussed, including generic implications
of the event and predicted plant response without operator action.
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8804010272 880323 !
PDR ADOCK 05000423 i
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DETAILS
1. Participants
Northeast Nuclear Energy Company (NNECo)
W. Romberg, Vice President, Nuclear Operations
S. Scace, Station Superintendent
C. Clement, Millstone 3 Superintendent
G. Van Noordennen, Supervisor, Generation Facilities Licensing
J. Ely, Supervisor, Component Engineering
J. Harris, Acting Operations Supervisor
R. Joshi, Licensing
U.S. Nuclear Regulatory Commission
W. Kane, Director, Division of Reactor Projects
L. Bettenhausen, Chief, Projects Branch No. 1
E. McCabe, Chief, Reactor Projects Section IB
W. Raymond, Senior Resident Inspector
S. Barber,' Resident Inspector
V. Pritchett, Reactor Engineer, Reactor Projects Section IB
R. Ferguson, Millstone 3 Project Manager, NRR
J. Strosnider, Chief, Materials and Processes Section
J. Gutierrez, Regional Counsel
D. Holody, Enforcement Coordinator
.
2. Purpose
The Enforcement Conference was initiated by NRC Region I to discuss the Janu-
ary 19, 1988 overpressure event without the required overpressure protection
systems operable. Northeast Nuclear Energy Company was requested to identify
and describe the event's root cause; broader implications; generic concerns;
lessons learned; and, both their short and long term corrective actions. To
ensure that the licensee was aware of the specific NRC concerns, Attachment
1 to this report hsd previously been provided to the licensee during the
February 18, 1988 exit meeting for Special Inspection 50-423/88-03,
3. Discussion
NRC management made a few brief opening remarks. NNECo then made a presenta-
tion based on the outline included as Attachment 2. NNECo agreed that the
events and technical issues described in the NRC January 18-29, 1988 special
inspection report were accurate.
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ATTACHMENT 1
17 PAGE 1
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During the RHR isolation event on January 19, a potentially significant event !
existed when the operating charging pump continued to run after the established
RHR letdown path was secured. Operator actions at the time were prompt and
appropriate. A relief valve on the letdown line downstream of the orifices
with a setpoint of 600 psig was capable of relieving RCS pressure during the
RHR isolation. The engineering evaluation discussed above demonstrates that
adequate margins remained to the reactor vessel limits. Although the RHR
isolation and RCS pressure increase event on January 19 is a significant event
that highlighted weaknesses in operating procedures, operator and technician
knowledge, drawings, and controls, the actual safety impact of the event was
minimal.
One item that warrants further licensee and NRC staff consideration because
of the potential significance of the isolation event is the question of "what
would the outcome of the transient have been if the operators had not acted
to mitigate the pressure increase." This question was presented to the lic-
ensee for consideration in his evaluation of the event.
8.0 Summary of Findings
Listed below are some issues to be discussed by the NRC and the licensee at
an enforcement conference. .
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Evaluation of what the January 19th pressure transient outcome would have !
been assuming no operator action.
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Adequacy of drawings and of training personnel in their use, especially
in interrelationships between drawings from dif ferent suppliers (e.g. ,
architect-engineer and NSSS vendor drawings).
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Assuring that work affecting SSPS and other safety-related equipment is
performed by personnel having appropriate training, qualifications, and ,
experience.
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Adequacy of controls allowing work on turbine bypass control circuits '
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without tags for pulling fuses, and the duty senior control room operator .
on January 19th approving work without reviewing the applicable AWO. )
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Adequacy and utility of Technical Specification curves prescribing
pressure-temperature limits.
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Adequacy of procedures and training on the COPS design and interface with I
the SSPS. )
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Use of alarn response procedures in post-transient reviews.
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Appropriateness of provisions that allow up to 31 days to elapse between
surveilling a system for operability and placing it in operation, and
assuring adequacy of supporting equipment configuration in the interim
and afterwards,
.
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ATTACHMENT 1
18. PAGE 2
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Adequacy of the surveillance used to determine COP 3 operability.
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Lack of positive indication of COPS arming. l
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Development of a specific low temperature overpressure protection proce-
dure, or providing additional guidance in existing procedures to assure {
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that operators are properly aware of low pressure overprotection fea-
tures' status and that they follow-up quickly, appropriately, and fully
to transients and to losses of overpressure protection.
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Assumption by the operators that COPS was the available overpressure
system when, in fact, it was not operable. This resulted in the viola-
tion of TS 3.4.9.3 which requires at least one overpressure system to
be available at all times when less than 350 degress F (VIO 88-03-02).
The failure to notify the NRC about the overpressure transient via the
--
Emergency Notification System (ENS) within four hours (VIO 88-03-03).
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ATTACHMENT 2
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Docket No. 50-423
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NRC ENFORCEhENT CONFERENCE
.
Millstone Unit 3
Cold Overpressure Protection System
Falls to Operate During a Pressure
Transient
(Incident Date: January 19,1988)
Presented By:
Northeast Nuclear Energy Company
March 8,1988
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NRC ENFORCEMENT CONFERENCE
EVENT SUMMARY
PLANT CONDITIONS, JANUARY 16,1988
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ARMING COPS, JANUARY 16,1988
PRESSURE TRANSIENT, JANUARY 19,1988
POST TRANSIENT REVIEW
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SAFETY SIGN'FICANCE
ANALYSIS - RHR ISOLATION !
CORRECTIVE ACTION - RHR
ANALYSIS - ARMING COPS
CORRECTIVE ACTION - COPS
REPORTABILITY
GENERIC IMPLICATIONS
CONCERNS
SUMMARY
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EVENT SUMMARY .
:
On January 19,1988 with the plant solid in cold shutdown,
temperature 135 *F, pressure 350 psia, a pressure transient
occurred which challenged the COLD Overpressure Protection
System (COPS). COPS failed to operate when required. The
pressure transient was mitigated by manual operator action.
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PLANT CONDITIONS, J ANU ARY 16,1988
COLD SHUT DOWN,135 *F,350 psia
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RCS solid
Loops B & C filled and vented
- Loops A & D drained and isolated for RCP locking cap repair
- Reactor Coolant Pump C running
Residual Heat Removal (RHR) A & B operable
- Overpressure protection provided by RHR relief valves (440 psia)
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ARMING COPS, JANUARY 16,1988
- Technical Specification surveillances reviewed
. Protection set Analog Channel Operational Test (31 day)
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Reactor Protection Channel Calibration (18 month)
PORV block valve open (72 Hour)
- Verified support systems operable by discussion with l&C tech
. Armed COPS by placing the PORV switches to ARM position
- Overpressure protection provided by COPS
- RHR train A tagged for system outage
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PRESSURE TRANSIENT. JANUARY 19,1988
. I&C calibrating the Main Steam Dump Valves
- P-12 Interlock must be disabeled to stroke the steam dumps
- l&C tech A researched the fuse to be pulled to disable P-12
- l&C Tech B reviewed and concured with the fuse pull i
- l&C tech A discussed the fuse pull with the SCO
- Fuse 61 pulled in the reactor protection system
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RHR suction valve starts to close
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RHR low flow alarm
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Operator action to mitigate the pressure transient !
RHR pump stopped
- RCS pressure increasing
- Stopped the charging pump
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Increased letdown flow
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Reestablished RHR
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POST TRANSIENT REVIEW
. The fuse pulled not only supplied power to the steam...
- Deenergization of the relays closed the RHR suction valves...
. Investigation showed COPS setpoints were attained but the...
. The failure of COPS to operate was do to the lack of input from...
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SAFETY SIGNIElCANCE
:
- COPS failure challenged overpressure limits per 10 CFR 50...
- Actual Peak pressure was 526 psia.
- Actual low temperature was 134 F.
COPS setpoint curves are conservative. They include... i
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- Appendix G limits not exceeded when conservatisms removed. l
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ANALYSIS - RHR ISOLATION !
- Electrical Schematic for steam durep does not show branch loads. ,
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- SSPS vendor drawing showing all fuse load is complex.
- I&C Technician A not qualified in SSPS.
- l&C Technician A did request assistance, but not formally.
- I&C Technician B did not fully review the consequences of the...
- l&C Technician A did not request assistance from foreman.
- The procedure for setting steam dump positioners was a generic...
ROOT CAUSE
- Incorrect use of drawings by a technician not qualified in...
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CORRECTIVE ACTION - RHR
- Memo issued to discuss proper use of drawings.
- Tail board discussion of incident with all departments.
- Training to incorporate lessons learned - April 15, 1988.
- Policy set to ensure only qualified personnel work on SSPS.
- Procedure for setting steam dump positioners - May 15, 1988.
- Conduct of Maintenance procedure to inicude correct line of...
- Train additional I&C personnel for SSPS qualification.
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ANALYSIS - ARMING COPS !
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- GOP to arm COPS was not adequate for plant condition. !
- Survellances to arm COPS did not ensure pre requisites.
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Relationship between SSPS and COPS not fully understood. l
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No direct indication that COPS is armed.
- Support system reviews for arming COPS were inadequate.
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CORRECTIVE ACTION - COPS
- Procedure for arming COPS - complete.
- Annunciator response for SSPS General Warning - complete.
- Review feasability of providing direct COPS armed indication ...
. Operator Training will be revised to re emphasize the...
- Tag SSPS test switch to SS when COPS is armed.
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REPORTABILITY
- Initially classified as not reportable.
- Reclassified as reportable per 10 CFR 50.73(a)(2)(vii).
Should have been reportable via ENS per 10 CFR 50.72.
. Resident was initially informed on January 19, 1988.
. Resident was kept informed during investigation.
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GENERIC IMPLICATIONS l
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- CY has mechanical relief valves for cold overpressure.
- MP2 - COPS is a single setpoint for PORV's.
- Use of drawings for pulling fuses at MP1, MP2 and CY,
. Qualified personnel only to remove fuses in I&C systems with...
- Review other potential procedures for fuse pulls.
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CONCERNS
- Pressure transient outcome assuming no operator action.
- Adequacy of drawings and of training personnel in their use.
- Work on safety related equipment performed by trained...
- Adequacy of controls of work.
- Adequacy of curves prescribing pressure temperature limits.
- Procedures / training on the COPS design / interface with SSPS.
- Surveillance interval / equipment configuration control.
- Are procedures inadequate to insure COPS operability?...
- Development of a specific low temperature overpressure...
- Use of alarm response procedures in post transient reviews.
- Lack of positive indication of COPS arming.
- The failure to notify the NRC within four hours.
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SUMMARY-
- COPS not fully operable for 63 hours January 16 - 19.
- RCS was protected against overpressure until RHR suction valve...
- Max RCS pressure 526 psia.
- Corrective Action extensive and prompt.
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OVERVIDi OF nnrn OVun'6 PICI1ETIO4 DURING *11E MTTLN UNIT 3 GJEAGE .
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Jan. 16
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COPS armed,
Mxle 4 entered
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Nov. 13 Dec. 4
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(Note 3)
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Notes: 1) On Nov.1,1987, mode 4 was entered at '
1730; COPS had been armed at 1655. But,
there is an SS log entry that states
"SSPS restored" at 1910. '1herefore, COPS
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may not have been available for 3 hours
and 40 minutes.
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2)' Time period of the events umaxining loss
' of Mm, SSPS, and COPS.
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3) 3 loops isolated (after both trains of
RHR available) .,
2
FIGURE 2
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STRUCTURAL INTEGRITY ISSUES FROM THE
JANUARY 19 MILLSTONE 3 PRESSURE TRANSIENT
o ADEQUACY AND UTILITY OF TECHNICAL SPECIFICATION
CURVES:
--
BACKGROUND--DEPRIVATION AND MARGINS.
O EVALUATION OF THE JANUARY 19 PRESSURE TRANSIENT
ASSUMIBG_tio_0P_ERATOR ACIron: !
--
USE OF ASME SECTION XI, APPENDIX E VERSUS
"0PERATIONAL" TECHNICAL SPECIFICATION CURVES.
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REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 10 EFPY
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VESSEL
g
180' : 0
-
k
PLN4 VIEW : b 0:
VESSEL i
-
f wAu.
s
P
hi
.
; s CAPSULE
C O.'" .
s/ ASSEMBLY
* '
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c 3 ..- CORE
s
f_.__ a
s
s
MIDPLANE
Cl
FIGURE 1-1. LOCATION OF THE IRRADIATION ' '
TEST CAPSULES IN THE f - i
$
NEUTR W
D
MILLSTONE UNIT NO. 3 ) N '
CORE BARRll re
.
REACTOR VESSEL f
th
no
1
ELEVATION VIEW
14
.. .-
.
<
l
,
l
SECTION 4 l
, POSTlRRADIATION TESTING
41. CAPSULE REMOVAL
The first capsule (Capsule U) should be removed at the end of the first core cycle (1st
refueling) as shown in Table 41. Subsequent capsules should be removed at 5,9, and 15
EFPY (Effective Full Power Years) as indicated. Each specimen capsule, removed after
exposure, will be transferred to a postirradiation test facility for disassembly and testing
of all the specimens.
TABLE 41
SURVEILLANCE CAPSULE REMOVAL SCHEDULE
_
. Orientation
Capsule of Lead Removat Expected Capsule
identification Capsules I*I FactorIbl Time Fluence (n/cm )
2
U 58.5* 4.00 ist Refueling 3.6 x 10
Y 241 * 3.69 5 EFPY 1.3 x 10I'l
V 61 * 3.69 9 EFPY 2.4 x 10Idl
X 238.5' 4.00 15 EFPY 4.3 x 10
W 121.5* 4.00 Stand By --
Z 301.5' 4.00 Stand By --
a. Reference Irradiation Capsule Assembly Drawing, Figure 2-4.
b. The factor by which the capsule fluence leads the vessels maximum inner wall fluence.
c. Approximate Fluence at % wall thickness at End of Life. '
d. Approximate Fluence at vesselInner wall at End-of Life. l
'
.
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,
TABLE A 4
TNDT.RTNDT AND UPPER SHELF ENERGY FOR
THE MILLSTONE UNIT NO 3 REACTOR
PRESSURE VESSEL CORE REGION SHELL PLATES
AND WELD METAL
NDT[e][bj
Upper Shelf! 'll*l
RTNDT Energy
Material
('C) (*F) ('C) ('F) (J) (f t Ib)
Intermediate
Shell Plates:
89805 1 -40 -40 16 60 126 93
B9805 2 - 51 - 60 -12 10 122 90
89805 3 -40 -40 -18 0 145 107
Lower
Shell Plates:
B98201 -46 - 50 -12 10 104 77
89820 2 - 34 - 30 4 40 103 76
B9820 3 - 34 - 30 -7 20 108.5 80
'
a. Data obtained from Combustion Engineering, Inc. Reactor Vessel Material
Certification Reports.
b. Drop weight data obtained from the transverse material properties (normal to the
major working direction).
c. From impact data obtained from the transverse material properties (normal to the
major working direction).
II
THDT RTNDT Energy
Material
.
('C) ('F) (*C) (*F) (J) (f t Ib)
Intermediate and Lower '
Shell Longitudinal Weld
Seams and Closing Girth
Wold Seam (Weld Wire Heat
-46 -50 -46 - 50 271 200 i
'
No. 4P6052, Linde 0091 Flux,
l
Lot No. 0145)
d. Data obtained from Combustion Engineering, Inc. Wire / Flux Weld Deposit Material
Certification Test No.1332.
A5
.
_. . . . -.
.. .-
. .
i
l
TABLE 3 5
SUMMARY OF MILLSTONE UNIT NO. 3
._ REACTOR PRESSURE VESSEL IMPACT TEST RESULTS FOR
INTERMEDIATE SHELL PLATE B98051 AND
CORE REGION WELD AND HEAT AFFECTED4 . MATERIAL
.._
Upper Shelf 41 J 68 J 0.89 mm
Energy (30 ft Ib) (50 ft Ib) (35 mils)
Material Index Temp Index Temp Index Temp
(USE)
(J) (f t Ib) (*C) (* F) ('C) (' F) ('C) (* F)
Plate B98051
(Longitudinal 180 133 -15 5 2 35 -1 30
.
Orientation)
Plate B98051
(Transverse 151 111 - 15 5 -16 60 2 35
Orientation)
Weld
194 143 - 37 - 35 - 26 -15 - 29 - 20
_
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Heat
Affected 193 142 - 90 - 130 - 71 - 95 - 76 - 105
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!40 ATHEAST UTILlilEl $ERYlCE COMPANY
'
HUCLEAR ENGINEERING AND OPERATIONS GROUP
GENERATION MECHANICAL ENGINEERING
A5ME SECTION til CLA55 2 AND 3
AND ANSI B31.1.0 PlPING ANALY115
PROJECT ASSIGNMENT: 84-00 %
CALCULATION NUMBER: bi*C03 A73 6P
PLANT: fnMc7 /f of fAvfrF
TITL E: //f 9& TT'"N'#FW lW / 7
(vevet Rc ?? (FPr
QA CATEGORY 1
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;.CONECTL CT.MtqE JhIT CURVE FOR HYDf0 STATIC AND EAX STSTING . . - . . _
APPLICABIE FOR 22.0 IVE FULL KMER YEARS.
-
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(T e 10 ,P
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- ---2200 -- -- - - - - -- . - - . -- ... -- -. - - - - - - - - - -- .
FJ i
!2 SYSTD1 HYDROSTATIC EST - - - -
e PRESSURE (P=2160)
y e
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- SYSTD1 IZAKAGE 'IIST
b ~.~~ PRESSURE (P=2000) ' '
_.1800-- --- - - - - - --- -- - --- --- - - . - - - - - - --- ---- -- - - - - - - -
,
_ _ . . . _ .. . _ _ . - . _ . . . . . _ . _.
D'PY MDD1 MINDUI SYSTDI _
E ~1600~ - - - - - - - - - - . .
HYDFO TDP IIAR TEST TDP
c. -
b '- ~ --~~
~ ~ ~ ~ ~ ' ' - ' ~~
22 245 F - - - -
-235 F --- --
._1400,, --
. . . _ _ . . _ ._. . . _ . . _ . _ _
!
. _ . _ _ _ .
i .-. . I
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i
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. _ . . . . . . . _ . . _ . . . .
s- F 600.. -
,
-fDR RESIDUAL HEAT RDD/AL--- ----- -
- - -
.N. 'p . _ . . . . , _. .. SYSTD1 OPEPATIQi . .... .__. . . _ . . . _ . ._
m m, . . . - - . ..
_ . _ . _ . . . . . _ . . _ .
-
_ . _ _
_ _,,; . '
/ - * -
' '
2i_.200- _MINDut PPESSURE (P = 300 PSIG). TOR ' !
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RE/CIOR CDOIA'TT PmP OPEPATIQi'
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' ._ OCNNDCTICOT. YANKEE REACICR COOLvc SYS'IDi HEA20P _ - =-= _ . . .
I : LIMITATICES FOR 22.0
' tuRIVE FitT.L POWER YEARS. :
i (TERIOR = -10 , p RIOR- = 150 psIg) -
.
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b _1400_ __- _ . - . . ... ._ .-. . .l PATE Og J50D/HR UP 'IO_.
I
- , 20 A';D 100 F/HR UP 'IO
550
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.
; (P = 300 PSIG) IVR RESIDUAL HEAT RDWAL SYSTD1 CPEPATICt1 ;
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_ i LIMITATICNS .*tR._.22.0 ptLC.T_IV.E . FULL . POWER.._YE_A.RS.
i +
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, P ..= .. _.
;
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;
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.
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..
_ . . . - . _ . . . . _ _ . . . . .. .
,__240L , '. . . _ _ . . . _ . . . . - . . . . . _ _ . . . . . . . - _ . . .
___. . . _ . . . . . . .. _
... 2200 . _ . . _ . . . . . . . . . . _ . _ . . _ _ -
_.. _ ...____.
R
g b .. .....).... . . . . - . . .c . . . . . ._ : _ .
$ . 2000. ... ... .- . . . . . . _ . _ _ . . . - . . . - . - -
L
w
p ~.. . _- .
0
" __1800. _ . . _ . .
C001DCQ1 PATE OP 100 F/HR DN1 TO 200 F
'..__._ . . .
AND 307/HR TO 707.
..
_ 1600 . . . . . _ . . _ . _ _ . .__ _ . _. . .. . . . . . . _ . . _ _ . . - . . _ _ . . _ _ _ . . . _
__ _ _ _ . ._. - - - -
a OPFP77D1Ci 2CUF - - -
h__l400. ._ _ _ . . . . _ . . . . . . _ _ . _ _
. . . . .. .. .
.. . . . .
. 1200. . . _ _ _ . . . . . . __ .. . . _ _ _ _ _ . . _
i <
g _~ _ . _ . . - . . - . . . _ .
. . . . . ._. . .. . _ .
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b) - - . _ . . . . . _ . . . . . . ._ . _. . .+.. .. .. .. . ._ ,
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_ BOR _ ._ .____ . _.._.... . _ _ . . _ . . . .
t=- m
-
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-
.
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, . . . :
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h L 600.. - ... __ __ -. ... .. . - ..___ ..
y
.. M'XDU1n PRDSSURE (P = 350 PSIG) .
_ . __ .. .
g ... . . ._ . _ . _ .
.. ,
9 _ ADO .
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f . .; ,.
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PT LIMIT CURVE INSTRUMENT ERROR INL.,' APcHA'I' ION
BOO
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GO PS
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10 0 11 0 12 0 13 0 I40 15 0 16 0 17 0 18 0 19 0 200
TEMPERATURE
(DEG. F)
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EVALUATION OF JANUARY 19 TRANSIENT
ASSUMING K0_DEERATOR ACTION
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o USE SECTION XI, APPENDIX E, EVALUATION OF l
UNANTICIPATED OPERATING EVENTS.
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0 SECTION III, APPENDIX G CONTAINS ASSUMED BOUNDARY
CONDITIONS:
1
--
1/4 T FLAW, ASPECT RATIO OF 1/6.
l
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--
,
"BOUNDED" HEAT-UP OR COOLDOWN RAMP RATES.
--
KIR CURVE.
J
--
,
FLUENCE THROUGH THE END OF AN OPERATING
'
!
PERIOD.
!
O TRANSIENT CONDITIONS, EVEN DESIGN BASIS CCIDENTS
'
FROM CHAPTER 15, ARE HQI CONSIDERED OR CONSISTENT i
i WITH THE P-T LIMIT CURVES. l
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ARTICLE E-1000
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INTRODUCTION
L1100 SCOPE TABLE E1
. .
This Append.ts prondes acceptance criteria and MAXIMUM ALLOWABLE PRES $URE AS A FUNCTION
0F T, - AT.,, FOR !$0 THERMAL PRES $URE
guidance for performing an engineering evaluation of
TRANSIENTS (AT/At < 10*F/ht)
the effects of an out oflimit condition on the structur. For design pressures greater than 2400 psis
al integrity of the reactor sessel beltline region.
Showing compliance with the criteria in either E.1200 T,- #Ts,, WadngnAApaW
or E 1300 assures that the beltline region has adequate m Prniet (80
structura] integrity for the unit to return to aerdce. + 25 and prester 1.1 x Destem
+ 15 fa00
+10 2250
0 2000
-10 17S0
-25 1500
-S0 1200
E.1200 ACCEPTANCE CRITERIAu ,,,,
' 3,,,
Adequate structural integrity of the reactor vessel
beltline region is assured if the fs!]oming applicable 200
$
criterion is satisfied throughout the event: ,750
*
(a) For isothermal pressure transients (i e., AT/At ' I h E "* ' N O T E "' **' ** t* " ***"'d'
4 10*F/hr). the maxirnum pressure does not exceed
the allowable salues of Table L1 at any value of
T, - R Tur. '
(b) For pressurized thermal transients (i.e., AT/At
> l'fFAr), the masimum pressure does not enesed
the design pressure and T, - RTuris not less than 'ALUATION BY ANALYSIS
, $5'F.
(a) Adequate structural lategrity of the reactor
if compliance with the above applicable criterion is
' sessel beltline region is assured ifit can be shoma by
( not shown, adequate structural integrity can be ana1 3 sis using the input of Table E.2 that the -
assured by satisfying the guidelines and criteria fo!!owing eriterion is met throughout the event:
specifsed in E lM)0. ~
,
,
1.4 (K. + K.) + KfK.
,
where
- e
'
, K i.,= stress intensity factor due to membrane stress
t *
K,=i stress intensity factor due to thermal stress
K,=i stress intensity factor due to residus) stress
K,=
i fracture toughness per A2ticle A 9000
'r, b & wh n w emtsat te=rentun, and Ar,/Ar b the (b) If complianc4 with the atose criterion cannot be
sanmum sansme et tementun r,in mar o** bout senat
eA Tor h * Wahat ad;*stad nfere== ta=Fntun (foe weld or shown, additional analyws or other actions shall be
<
b.w matenar) et m inia, ..rf.= or the emw . : taken to assure that acceptable margins of safety *d. l
drw===: t> Reruhtory on i to Rev. 2. be maintained during subsequent operation. ;
358.3
.
.
2 o*
.
*
TaMeE2 SECTION XI-DIVISION 1 1986 Edeles
. TABLE E2
EVALUATION INPUT FOR PLANT AND i
EVENT SPECIFIC UNEAR
ELASTIC FRACTURE MECHANICS ANALY$t$
ver w e vehe
<
Prosawee *
Eveat penaure time history
Torgerstwo Event Wwerstwee tarve hietary
eteet transke Eveat/p' ant specmc Aee/mishe
se4tions
O r t type Semi-elfiptical wfect Ase
einmwm Wtletion 0.0 < a 5 1.0 h (Nete (1))
track see
4
Crock e4atsuen Long%dinal i
K./K, location Swiface and run'ewm depth
.
Clad eMeets Cted to be considered in tlw
I
(Nrmal, stress, and fractwo
mech.sMcs 4Wyses (hete (2))
Tra%'tu teve% ness K. per Article A 9000
Upper theT towthness (In cowrse of preparetken)
Nace Flwence at the time of the
trans%t
$Nfttwrw Rep /atory Gwiet 1.99 Rev 2
pesidd strns Appropriate W:strtDvtion for the
fabr4aUen process, se bnear
;
d.stnbet% mith +10tsJ et the
,
*
,
Ins 4e Serface and .30441 et the
eJth 6t w fact
'
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# -' m a . the muim mu cract destMn the tase meta! '
! (2) The stresses dse to the dSereve t.eteten the kse meta' and
Ela$$Ag thermat espa*Lsion Leeff.cients need ret De ten 64ered
,
Wi the 66cthermal presve trans'ee.t . ovatwation (Le
A 7,/ A t < 107 / hr).
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MTNOT
JL
0 e a t t
-100 -50 0 +50 +100 +150 +200
(T~ RTgof.*F
FIG. A-4200-1 LOWER BOUND K,, AND Ke TEST DATA FOR SA-533 GRADE B CLASS 1, SA-508 CLASS 2, AND SA-508
CLASS 3 STEELS
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.
ANALYSIS--APPENDIX E
0
o CURRENT ADJUSTED RTNDT ~ 90 F:
0
--
ORIGINAL 600 AND 30 F SHIFT MEASURED IN f
SURVEILLANCE CAPSULE.
I
!
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--
CONSERVATIVE DUE To CAPSULE LEAD FACTOR ~ 4.0.
0
o Tc--RTNDT = 134-90 = 44 F.
;
o TABLE E-1, ALLOWABLE PRESSURE = 1.1 x DESIGN @
.
Tc--RTNDT = 2.
I
o TRANSIENT PRESSURE WoULD BE LIMITED To 2485 PSI BY
. CODE SAFETIES.
o PALLoW > PTRANriENT - OKAY.
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-_. . - - - - - - . _ _ . - __
_ _ _ _ _ _ _ _ _ _ _ Q
-
_-
,
. ...
..
.
- - -
.
I
CONCLUSIONS
O PRESSURE TRANSIENT OF JANUARY 19 HAS NO IMPACT ON
MILLSTONE 3 STRUCTURAL INTEGRITY.
O CHEMISTRY AND FLUENCE LIMIT RANGE OF ISOTHERMAL
TRANSIENTS OF CONCERN ~ 1500 F OR LESS.
0 OPERATIONAL LIMITS HAVE SIGNIFICANT SAFETY MARGINS
TO PRECLUDE COMPONENT DAMAGE.
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