ML20040B413
| ML20040B413 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 01/21/1981 |
| From: | Cohn M, Crawford R, Gunnison F SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY |
| To: | |
| Shared Package | |
| ML20040B409 | List: |
| References | |
| RTR-NUREG-0803, RTR-NUREG-803 NUDOCS 8201260025 | |
| Download: ML20040B413 (79) | |
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LA SALLE EVALUtl~10fl REPORT REGARDIf1G INTEGRITY OF SCRAM
) SYSTEM PIPING 4
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Prepared by
)
SCIEllCE APPLICATIONS, INC.
1211 West 22nd Street, Suite 901
) Dak Brook, Illinois 60521
) Principal Investigator R. f t. Crawford With Contributions by li. Cohn F. Gunnison 3 D. Harris !
T. Regenie l l
l 3
Prepared for C0ilt:0NWEALTH EDISON COMPANY STATION NUCLEAR ENGINEERING DEPARTf1ENT D
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D EXECUTIVE
SUMMARY
In August 1981, the l'.5. fluclear Regula tory Commission published flUREG-0803,
" Generic Safety Evaluation Report Pegarding Integrity of BWR Scram System D
piping." This document addressed the possibility of scram system pipe breaks outside the primary containment. Specifically, a generic BWR probabilistic risk assessment in that document indicated that the postulated SDV event is not a dominant contributor to the probability of core damage. However, NRC guidance in Chapter 5 of fiUREG-0803 requires that the assumptions used in the.
risk assessrrent be verified on a plant specific basis.
The plant specific issues in NUREG-0803 are addressed for La Salle County
)
Station (LSCS) Nuclear Units 1 and 2 in this document. It is established that: (1) the SDV piping satisfies all appropriate ASME Codes, is seis-mically qualified, and a probabilistic fracture mechanics evaluation of the scram systems piping demonstrates that the probability of a pipe 3
rupture in this system is not a significant contributor to an accident scenario; (2) for a postulated break in the scram system piping LSCS leak detection equipment and the associated operatino procedures will 3 guide the reactor operators to a prompt and successful mitioation of the event; and (3) all equipment needed to mitioate a postulated SDV pipe break event is environmentally qualified for the environment and will function adequately to bring the reactor system to a safe shutdown condition.
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O TABLE OF CONTENTS PAGE EXECUTIVE SU!7ARY ........................................................
i
,U INTRODUCTION ........................................................
1 l.
I 1.1 BAC KGROUND AllD SUM'iARY OF CONCERtiS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
2 1.2 PLANT-SPECIFIC ISSUES IDENTIFIED BY NUREG-0785 ......................
O 1.2.1 Piping Integrity ............................................
2 2
1.2.2 M i t i g a t i o n C a p a b i l i ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3 1.2.3 Eq u i pmen t Q ual i fi ca ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1.3
SUMMARY
OF RELEVANT PLANT-SPECIFIC FEATURES OF LA SALLE COUNTY 3 O STATION .............................................................
3
- 1. 3.1 Secondary Con tainment (Reactor Buil ding) Layout . . . . . . . . . . . . .
4
- 1. 3. 2 SDV Design Features .........................................
5 1.4 PLANT-SPECIFIC C0f1CLUSIONS FOR LA SALLE COUNTY STATION . . . . . . . . . . . . . .
.O 5
- 1. 4 .1 Piping Integrity ............................................ 5 1.4.2 Mitigation.................................................. 6 1.4.3 Eq u i pme n t Qual i fi ca ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
9
- 2. NUREG-0803 GUIDANCE FOR INDIVIDUAL PLANTS ...........................
O 9 2.1 I N T R O DU C T I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Il 2.2 LA SALLE COUNTY STATION RESPONSE TO NRC GUIDANCE ....................
11 2.2.1 Inservice Inspection and Surveillance of SDV System . . . . . . . . .
12 O 2.2.2 Th r e a de d J o i n ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 2.2.3 Seismic Design ..............................................
14 2.2.4 HCU-SDV Equipment Procedures Review ......................... 15 2.2.5 As-buil t Inspection of SDV Piping and Supports . . . . . . . . . . . . . . 16 2.2.6 Imp ro vemen t o f P ro c edu res . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 2.2,7 Cool an t Iodi ne Ac ti vi ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
,O 2.2.8 Environmental Qualification of Prorpt
- - 18 Depressurization function ...................................
19 2.2.9 Equi pment Quali fi ed for Ua ter Impingement . . . . . . . . . . . . . . . . . . . 20 2.2.10 Equi pmen t Qual i fi ca for We tdown . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 2.2.11 Qual i f ica tion o f Essential Equipment . . . . . . . . . . . . . . . . . . . . . . . . 22 2.2.12 Evaluation of Availabili ty of HPCI and RCIC Turbines . . . . . . . .
O 2.2.13 Verification of Operability of feedwater and 23 Condensate System ...........................................
24
2.3 CONCLUSION
S .........................................................
'O O ij
1 1
O i 1
1 TABLE OF C0flTEf1TS (cont'd)
PAGE
- 3. LA SALLE COUNTY STATION EVALUATION OF SCRAM SYSTEM PIPING ................................................. 25
3.1 INTRODUCTION
........................................................ 25 3.2 PIPING INTEGRITY .................................................... 25
'0 25 3.2.1 Design ............................,.........................
3.2.2 Materials ....................... ........................... 27 3.2.3 Fabrication and Installation ................. .............. 28 3.2.4 Quality Assurance .. ........................................ 28 3.2.5 Preservice and Inservice Inspection, ASME Section XI ............................................. 29
- O 29 3.2.6 As - b u i l t I n s pe c t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3.2.7 Main tenance and Modi fica tion of SDV Piping . . . . . . . . . . . . . . . . . . 29
- 3. 3 MITIGATION CAPABILITY ............................................... 30 3.3.1 I n t ro d u c t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30
- O .....
30 3.3.2 Opera tor Res ponse to a Reactor Scram . . . . . . . . . . . . . . . . . . . . . . . .
3.3.3 O p e ra to r T ra i n i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 3.3.4 Control Room Alarms Associated with a SDV Event ................................................. 31 3.3.5 Alarms with Procedures that Include a Ra pi d Dep res s u ri za ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32
.O 3.3.6 Ope ra to r Res pons e to a SDV Even t . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 3.3.7 Personnel Observa tion of Lea kage . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 3.3.8 Control Rod High Temperature Alarms ......................... 34 3.3.9 Loss-of-Offsi te Power Coincident w i t h t h e S DV Fa i l u re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 0 3.3.10 Impact of SDV Leak Rates Less than the J
Ma x i m u m Po s t ul a te d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 3.3.11 Manual I sol a tion o f the SDV Lea k . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 3.3.12 P r i ma ry Coo l a n t A c t i v i ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 3.3.13 Emergency and Long-Term Cooling Capability .................. 36 3.3.14 Secondary Containment Design and Emergency O Cooling Pump Locations ...................................... 36 3.3.15 Emergency Core Cooling Following a S DV Fa i l u re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 i 3.3.16 Alternative systems Available for Emergency Core Cooling ...................................... 37 0 3.4 ENVIR0f; MENTAL QUALIFICATION ......................................... 37
- 3. 5 CONCLUSIONS ......................................................... 38 APPENDIX A ............................................................... A-l O
- O jjj
)
- 1. INTRODUCTION
)
1.1 BACKGROUND
AND
SUMMARY
OF CONCERNS In March 1981, the U.S. Nuclear. Regulatory Commission (NRC) completed NUREG-0785, " Safety Conceins Associated with Pipe Breaks in BWR Scram
) Systems." This document addressed the possibility of scram system breaks located outside primary containment, and specifically within the Scram Discharge Volume (SDV) subsystem. It is acknowledged in NUREG-0785 that the resulting small leak flow lost from the bottom of
) the reactor vessel could easily be replaced by the High Pressure Coolant Injection (HPCI) system (in La Salle Station the HPCS system performs this function) or by the Reactor Core Isolation Cooling (RCIC) systam.
Alternatively, the reactor vessel could be depressurized and inventory
) replaced by various low pressure systems. However, it was further pos-tulated in NUREG-0785 that water emerging from the break could potentially run across the floor on which the SDV is located, flow down or through various stairwells, and eventually make its way to the
) basement, where essential ECCS pumps are located. Flooding of these pumps could conceivably impair the equipment itself and/or instrumentation systems required to assure long-term core cooling and water inventory replacment.
)
In April 1981, General Electric Company issued NED0-24342, "GE Evaluation in Response to NRC Request Regarding BWR Scram," as a generic
) evaluation of these issues. This document concluded that SDV rupture does not constitute a significant safety problem because:
o Such a pipe break has an extremely low probability
) of occurrence.
o Even if such a break should occur, the event would be detected and terminated by manual operator action. Further, no lower level pump rooms
) would be flooded because manual depressurization would assure that any leak rate would be reduced to a level well within the capability of drain sump pumps.
J l
.O o Even if all ECCS pumps should become unavailable, the long-term expected leak rate of 40-50 gpm could easily be replaced by other pumps.
O In August 1981, the NRC released NUREG-0803, " Generic Safety Evaluation Renart Regarding Integrity of BWR Scram System Piping," which (1) evalu-ates the generic conclusions of NED0-24342, (2) perfonns additional O probabilistic analyses to show that the postulated event is not a dominant contributer to core damage, and (3) outlines certain plant-specific issues to be addressed by BWR owners. Continued plant operatiori is concluded to be acceptable because the risk level is lower O than other events (e.g. Anticipated Transient Without Scram) which are under continuing study.
1.2 PLANT-SPECIFIC ISSUES IDENTIFIED BY NUREG-0785 O
The plant-specific resolution of a SDV problem involves issues falling under three areas of concern: (1) Piping Integrity, (2) Mitigation Capability, and (3) Environmental Qualification.
1.2.1 P_iping Integrity (PI)
The Quality Assurance (QA) for SDV piping and pipe supports is to be checked by verifying that a seismic analysis has been performed and by O'
verifying its as-built configuration. Inservice inspection procedures are to be shown adequate for keeping the pipe failure probability at a low level.
r 1.2.2 Mitigation Ca_pability (M)
Tssuming that an SDV piping break should occur, the capability must be '
shown to exist to mitigate the transient so that no radioactive release occurs in excess of the levels specified by 10 CFR 100. It is to be O
verified that adequate operations procedures and instr mentation exist to ensure that such a condition is diagnosed and that appropriate mitigating action is taken to guarantee that reactor pressure vessel
(" } " "' """ '~ " "" "" " ' ' " "#*'
O building should be low enough to permit operator access to the affected area, and offsite dose levels must at all times remain below limits set by 10 CFR 100. Finally, there must be adequate core cooling capacity,
.O 2
)
through ECCS and other pumps, to maintain water inventory at all times; this question is closely tied to equipment qualification issues
) discussed in Section 1.2.3.
1.2.3 Equipment Qualification (EQ)
On a plant-specific basis, the enuipment should be identified which is
) necessary to detect a SDV pipe break or to mitigate the transient resulting from an unisolable SDV pipe break. The environmental con-ditions, including temperature, humidity, and wetting, are to be asses- l sed and the qualification status of all such equipment determined in
) relation to these er.vironmental conditions.
The above surinary of NUREG-0785 issues was addressed by GE in their NED0-24342 document on a generic basis. The qualifications of those generic considerations for la Salle is included in the following La Salle specific treatment of the SDV issue.
1.3
SUMMARY
OF RELEVANT PLANT-SPECIFIC FEATURES OF LA SALLE COUNTY STATION (LSCS)
The La Salle County Station is a BWR-5 with a Mark II Containment. The SDV subsystem design and the secondary containment (Reactor Building)
) have several key plant-specific features which tend to alleviate the NRC generic concerns.
1.3.1 Secondary Containment (Reactor Building) Layout
) Figures 1 and 2 show representative plan and elevation views of the reactor building. The Hydraulic Control Units (HCUs) are on the 761' level, while the ECCS pumps are located four floors below on the 674' level. The corner rooms containing the ECCS pumps are isolated from the
) basement inner annulus by normally-closed water tight doors, starting from a concrete curb about one foot off the floor. Further, a control room annunciator and lamp are activated whenever any of these doors are not closed. Floor drains on the 761' level drain into a sump in the
) basement inner annulus and not_ into any of the corner rooms containing 3
O ECCS pumps. Thus, the 674' floor level does not provide a conrnon volume for flooding.
O Withir these corner rooms, the ECCS pumps rest on concrete pedestals more than three feet high, and the bottom of each pump motor is approxi-mately nine feet above the floor. SDV leakage does not pose a threat to ECCS pump operability. (See 2.2.9 for other flood pathway n
considerations . )
The secondary containment is vented via check dampers into the steam tunnel, which constitutes a very large sink should the reactor building
.O become slightly pressurized by steam flow from an SDV pipe break. The reactor building integrity is further protected by steam tunnel blow-out panels which, should the steam tunnel be overpressurized, would open briefly to reline pressure and then reclose. Normally the s tandby gas O
treatment system is used to clean up the reactor building and the steam tunnel gases. Gases vented out of the steam tunnel are routed through the turbine building basement to be exhausted through the plant stack.
1.3.2 SDV Design Features Each La Salle reactor has two banks of HCus. Each bank has its own SDV and SDV instrument volume; these bas s are separated physically on the g North and South sides of the reactor. At present, a single drain line from each instrurcent volume goes into a conmon drain line with a 2-inch drain valve and thence to the reactor building equipment drain tank.
Separttrly, the vent lines from each SDV are joined at a common high g point and routed via a single vent line with a single vent valve to the phase separatne room where an atmospheric opening is located. At the first refueling outage, this configuration will be changed so that there are redundant drain valves and redundant vent valves. The present SDV system is designea to appropriate ASME codes and standards, has O
satisfied all appropriate QA procedures, and is seismically qualified.
All 3/4" piping from the HCUs to the SDV is stainless steel. In the g present configuration, the 8" pipe bet, een the HCU bays leads into a coninon 8" header, which in turn integrally connects to the 12" SDV instrument volume. The instrument volume empties into the 2" drain l line.
lO 4
l
-O l Each Scram Instrument Volume has four Magnetrol level switches. The !
l lowest level switch simply initiates a control room alarm, while an J intermediate level switch initiates the rod block to prevent rod with- l
{O drawal. Should water level reach the third and highest monitored level, there are two redundant switches which immediately initiate a reactor scram. At the first refueling outage, two diverse DP switches will be installed to further assure a reactor scram when the instrument volume O is filled. 110 wever, none of the Scram Instrument Volume instrumentation is necessary to mitigate the 59V pipe break event, since a scram has already occurred and other means have been identified for diagnosing the event.
O 1.4 PLANT-SPECIFIC CONCLUSIONS FOR LA SALLE COUNTY STATION Plant-specific issues outlined in Section 1.2 have been addressed for O La Salle County Station. The postulated rupture of Scram Discharge Volume piping has been shown conclusively not to be a significant safety hazard from the points of view of piping integrity, mitigation c4 ability, or equipment qualification.
O 1.4.1 P_iping Integri ty_(PI_)_
Appropriate quality control and quality assurance procedures both in the design and construction stages ensure that all SDV piping meets all "n applicable codes and conforms to the highest standards of quality. A highly conservative fracture mechanics analysis shows that sufficient margin exists for a pipebreak to be a very unlikely event for the scram piping and the Scram Discharge Volume and that the instrument volume
.O along with the vent and drain lines have essentially 7ero probability of cracking.
1.4.2 Mitigation (M1 O
The LSCS abnormal event procedures, backed by specific operator training exercises on leakage detection, will assure early detection of any significant leak outside primary containment. Should such a leak occur and be unisolable, then procedures dictate rapid depressurizaticn
'O of the reactor pressure vessel. Adequate core cooling capacity is available throughout any such event.
O 5
l 1.4.3 Equipment Qualification (EQ)_
Equipment required for leak detection, depressurization, short-term and long-term core cooling is fully qualified for the reactor building environment following a postulated SDV rupture. The isolation valve on the SDV drain line is currently undergoing qualification. Flowever, this i valve does not perfonn an active safety function.
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- 2. NUREG-0803 GUIDANCE FOR INDIVIDUAL PLANTS 3
2.1 INTRODUCTION
The generic response to a postulated pipe break in a BWR scram system 7; is reported in GE NED0-24342. NUREG-0803 presented the NRC staff appraisal of that generic report and listed specific guidance to BUR owners for plant specific evaluations. This guidance was summarized in Table 5.1, Summary of Guidance for Individual Plants, of NUREG-0803 and e is reproduced here as Table 2.1. Each of these issues is briefly discussed for LSCS in the following sumnary of issues. A more detailed discussion is prescoted in Chapter 3.
O 0
O J
D D 9
D Table 2.1 Summary of Guidan:e for Individual Plants D
Area of Concern Guidance PI Periodic inservice inspection and J surveillance for the SDV system PI Threaded joint integrity PI Seismic design verfication D PI HCV-SDV equipment procedures review EQ Environmental qualification of prompt depressurization function PI As-built inspection of SDV piping and G supports M Improvement of procedures EQ Verfication of equipment designed for water impingement EQ Verification of equipment qualified for wetdown by 212 F water EQ Verfication of feedwater and condensate system operation independent of thc
- reactor building environment EQ Evaluation of availability of HPCI-LPCI turbines due to high ambient temperature trips
- EQ Verfication of essential components qualified for service at 212 F and 100% humidity M Limitation of coolant iodine concen-i tration to Standard Technical O Specification salues O
PI = Piping Integrity M = Mitigation Capability EQ = Equipment Qualification 10
i i
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2.2 LA SALLE COUllTY STM!Ott RESP 0flSE TO flRC GUIDAtlCE
)
2.2.1 Inservice Inspection & Surveillance of the SDV System ilRC Finding A program of.. inservice inspection and surveillance for SDV piping is necessary to periodically verify its integrity.
LSCS Responso
) A Preservi< ction of the CRD system was conducted in accordance
~
with the ASMt Boiler and Pressure Vessel Code Section XI, Summer 1975 Addenda. This included ultrasonic testing and liquid penetrant examin-ation of several welds in the scram discharge headers. The results are
)~ described in the la Salle Unit 1 Preservice Inspection Report and in the La Salle Unit 2 Preservice Inspection Report. Future Inservice Inspections are described in the La Salle County Inservice Inspection Programs for each unit.
)
)
)
)
4N I
I i
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, 11
l 2.2.2 Threaded Joints _
f4RC Finding Plants with threaded joints in the SDV piping should assess their structural and leaktight integrity under conditions where severe erosion, crevice corrosion, dynamic events, and vibration can occur.
LSCS Respons_e_
There are no threaded joints in the SDV piping at La Salle.
t
l 2.2.3 Seismic Desi_gn f.LRC Finding Each plant must verify that the SDV piping has been designed for seismic loading.
LSCS Response _
All SUV piping, including the CRD insert and withdraw lines, the scram discharge volume, the scrant discharge instrument volume, and the vent and drain lines, have had seismic loadings included in the stress analyses. The drain line stress level meets the requirements of ASME Code,Section III, Class B, 1974 issue. The remaining lines were designed to satisfy the stress requirements of ASME Code,Section III, Class 2 piping. With the exception of the drain valve all the essential active components in the SDV system have been seismically qualified.
The drain valve will be seismically qualified during February 1982.
13
-. - _ _ _ _ _ _ _ _ _ _ _ l
.O 2.2.4 HCU-SDV Equipment Procedures Review
'O NRC finding The plant-specific reviews should verify that all surveillance, maintenance, inspection, or modification procedures which conceivably
'(J have the potential for defeating SDV integrity contain sufficient guidance to ensure that the loss of SDV system integrity will not occur at a time when such integrity should be available.
O L_SCS Response All maintenance related and modification work on the SDV is controlled by CEC 0 Q. A. Procedures (Topical Report CE-1) and La Salle Station i Administrative Procedures. Any work on the CRDs, HCus, or SDV would be
- C) classified as safety related or ASME Code related and would require multiple levels of review prior to commencing the work. Final approval n any work order is grantad by the Shift Supervisor.
'O O
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_2.2.5_ As-built inspection of SDV Piping and Supports
)
flRC_Findina The staff reconniends that an as-built inspection of SDV piping and its supports be conducted at all BWRs.
)
LSCS Response The as-built walkdown for the LSCS Unit 1 SDV piping is underway and will be completed by February 1,1982. The Unit 2 walkdown will be
) scheduled at an appropriate time before Unit 2 fuel load. The Unit 2 SDV piping configuration is to be equivalent to that installed on Unit 1 during the first refueling outage, i.e. dual vent valves and dual drain valves along with diverse and redundant instruments for measuring level
) in the instrument volumes.
)
15
O 2.2.6 Improvement of Procedures O
NRC Finding Individual plants should implement procedures that include modifications addressing the secondary containment problems anticipated with an SDV g rupture and a scram which cannot be reset.
LSCS Response l
La Salle Operating Annunciator Procedures for alarms which are anticipated f 11 wing the postulated event have been modified to ref5-O the reactor operator to more general La Salle Operating Abnormal Procedures which describe possible reasons for the alarms (e.g. a leak outside primary containment) and which contain explicit directions for O
rapid depresssurization (1) if the primary or secondarj containment integrity should be lost, (?) if the RPV level cannot be maintained, or (3) in order to minimize releases to the environment. In addition, the LSCS version of the BWR Ownerr
- Emergency Guideline Procedures for C.
c Id wn (LGA-02, Cooldown) has been modified to include the same di rections , i .e.
"...The 100 F/hr cooldown rate may be exceeded to conserve RPV inventory , protect primary or secondary containment 9 integrity or limit radioactive release to the envircnment."
Further, LSCS reactor operators have received or will receive similator and/or classroom training on how to interpret the symptoms associated o with pipe breaks outside the primary containment. Specifically, the operators will be trained for the SDV rupture event.
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{
l3 2.2.7 Coolant Iodine Activity 3
NRC finding The Standard Technical Specifications (STS) for coolant activity should be implemented for all operating BWRs unless the licensee can demons-trate, based on analysis of operating history and current and projected fuel performance, that the probability of operating at coolant activity levels in excess of those allowed by STS is less than 10-3 per reactor
, year.
s
.LSCS Response The LSCS Technical Specification for coolant activity limits the activity to less than that allowed by STS.
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i IJ 2.7.8 Environmental qualification for Prompt Depressurization Function C
flRC Finding The prompt depressurization function should be qualified to meet the expected HCU break environmental conditions.
O LSCS Response The prompt depressurization function is qualified to meet expected HCU break environmental conditions. The ADS system is located inside
.(D primary containment and is safety related equipment. Thus, it is appropriately qualified and would not be exposed to a harsh environment in the event of a SDV rupture or leak in the reactor building. In other sections of this chapter the qualification of the emergency shutdown and
. C) long-term cooling equipment is also shown to be adequate.
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D 2.2.9 Eguiprpent Qualified for Water Impingement E
Ii.RC Finding Eoch licensee should verify that any emergency equipment that could be sprayed with water frem dripping or splattering of overflow leakane down 3 open stairnells is designed to operate with water impingement.
l.SCS_ Response Eighteen floor drains in the HCU-SDV area make it unlikely that there 3 will be overflow leakage because they are adequate for the maximum postulated leakage. Further, there is only one stairwell with a direct flow path from the HCb-SDV area to an emergency cooling equipment room, i.e. the LPCS/RCIC cubicle. The elevation and location of the D equipment in this room, is such that water impingement is not a problem.
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l E.2.10 E_quipment Qualified for Wetdown O
flRC Finding Licensees should verify that any equipment needed for mitigation that could be wetdown from leakage through equipment hctches is qualified for g wetdown by 212 F water.
LSCS Response With the exception of the large equipment hatch which terminates at the o 710' level in the northeast corner of the reactor building, all hatch covers will normally be in place during operation. To reach the emergency core cooling and long-term cooling pumps the water would have to drip through four consecutive hatches. The resistance to such flow
'O is very high. The eighteen floor drains in the HCU-SDV area would minimize water leakage through equipment hatches. Further, if the water finally reached a motor through an equipment hatch it would be cooler than 212 F. fievertheless, the ECCS motors are qualified for wetdown by O 212 F water from the top, i.e. flow around the hatch covers.
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2.2.11 gualification of Essential Equipment
)
f[RC Finding Each licensee should verify that all the components of systems required for safe shutdown and long-term core cooling are qualified for service U
) at 212 F and 100% humidity.
_L_SCS Respo_nse Because the ECCS systems are not in the imediate vicinity of the
) postclated leak but are 88 feet below the HCU-SDV level, (only the ,
LPCS/RCIC room is directly coupled to the HCU-SDV level via stairwells),
the environment cannot be 212U F and 100% humidity. The estimated maximum temperature is 140 F and 100% humidity for a steam instrument
) line break in this part of the reactor building. However, the motors and the electrical systems of the LPCS and the RHR systems are qualified for serlice at greater than 212 F and 100% humidity. The remote shutdown panel is located in the auxiliary electric equipment room of
) the auxilliary building which is physically separated from the reactor building. Controls and instrumentation for normal operation of long-term cooling and safe shutdown are located in the Control Room. l The RCIC system at La Salle is not an ESF system and it is not relied
) upon for mitigation of transients or accidents.
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) i 2.2.12 Evaluation of Availability of HPCI and RCIC Turbines
)
NRC Finding Licensees shuuld conservatively detennine whether the temperature trip monitors for the riPCI and RCIC turbines would cause the turbines to trip
) because of temperature buildup in the areas where the sensors are located.
LSCS Response
) There is not a HPCI system at LSCS. The HPCS system at LSCS performs the equivalent function to the llPCI system at other plants. The HPCS system is powered by an electric motor not a steam turbine. The HPCS notor is qualified for the anticipated SDV event environment, and its
) operation is not dependent on a room temperature trip. The temperature trip for the RCIC turbine is 200 F. Because the RCIC system is located ,
eighty-eight feet below the HCU-SDV level and the reactor building is a large heat sink, it is not likely that temperatures in tha RCIC room
) will exceed 140 - 150 UF before prompt depressurization. Thus, the RCIC turbine will continue to function during the SDV event. The RCIC system at La Salle is not an ESF system and it is not relied upon for mitigation of transients or accidents.
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'O 2.2.13 Verification of Operability _ of feedwater and Condensate System
- O ttRC Finding Even though the feedwater and/or condensate systems are located outside '
the reactor building, the licensees should verify tnat operation of g these systems is independent of any systems or components contained in the reactor building.
LSCS Response
- O The feedwater and/or condensate systems are independent of any systems a
or components in the reactor building. Thus, they will be available for core cooling provided the scram prior to the SDV event does not initiate a primary isolation. If such an isolation occurs the reactor operators g are trained in how to restore these systems following such a scram. The symptomatic emergency procedures for water inventory control were originally developed at La Salle by the BWR Owners Group.
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l 2.3 C0f!CLUSI0 tis
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Each plant-specific requirement cited in NUREG-0803 has been incorporated or verified for the LSCS SDV system. The piping has been constructed and fabricated to ASME,Section III, Class 2. The pre-service and inservice inspections are consistent with the ASME Section
)
XI code. The entire system is seismically supported. There are no threaded connections in the system and a final walkdown inspectice will be completed prior to February 1,1982. Further, CECO's Q.A. and Q.C.
procedures in conjunction with LSCS administrative procedures assure
)
that the system will be operational for all necessary reactor condi-
, tions.
The LSCS LOA procedures have incorporated specific directions for the reactor operator to initiate prompt depressurization when a significant unisolable leak outside primary containment is detected. In addition the reactor operators and shift supervisors will receive SDV event
)
training. The LSCS Technical Specification for coolant activity limits the coolant activity to the Standard Technical Specificatior.3.
The equipment needed for safe-shutdowr, and long-term cooling is
) environmentally qualified or is in the La Salle EQ program for eventual qualification for the SDV event reactor building environment. Water impingement has been shown not to apply because of the tSCS equipment placement. While water wetdown is not likely, all potentially affected
) equipment is fully qualified. The feedwater and condensate systems are not dependent on systems located in the reactor building. Thus, they will be available for mitigation of this event.
) Based upon these results, the assumptions made in NUREG-0803 for the generic integrated SDV risk assessment are fully applicable to LSCS.
Therefore, the postulated SDV event does not contribute significantly to the probability of a degraded core at La Salle.
)
) 24 1
D
- 3. LA SALLE COUt4TY STATI0fl EVALUATIOf4 0F SCRAM SYSTEM PIPIfiG D
3.1 IflTRODUCTI0f1 flVREG-0803 requires plant specific information in three general areas -
3 piping integrity, mitigation, and environmental qualification to support the assumptions of a generic probabilistic risk assessment which indicates that an SDV ruptt.re following a scram which cannot be reset is not a dominant contributor to core damage. Here, the concerns J delineated in fiUREG-0803 will be addressed for LSCS Unit 1. LSCS Unit 2 is essentially identical to Unit 1. The construction and hence the appropriate Q.A. and Q.C. documentation for Unit 2 are several months further from completion than Unit 1. However, the conclusions reached D for Unit 1 will also apply to Unit 2. The Unit 2 Q. A. and Q.C.
documentation will be completed at the appropriate time before fuel load of that unit.
9 3.2 PIPJf4G If4TEGRITY lhe scram discharge volume (SDV) and control rod drive (CRD) piping systems have been designed to the required ASME and AfiSI applicable 9 codes and standards. In fact, the La Salle requirements in many cases are beyond the normal standards to reduce the system failure probability to the lowest practical value.
] La Salle piping integrity requirements are compared to industry stan-dards in Table 3.1. La Salle's requirements for the SDV and CRD piping systems meet all of the ASME B & PV Code Section III Class 2 require-ments. Details of the satisfied requirements are presented below.
D 3.2.1 Design _
Jetailed stress analysis reports on the SDV and CRD piping systems de-lineate the load combinations for the piping systems which include seis-O mic, hydrodynamic, thermal, and dead weight loads. Sargent & Lundy (AE) has completed their analysis of the drain lines Ito the requirements of
'*) 25
id Table 3.1
.O La Salle SDV and CRD Piping Systems Comparison of La Salle Piping Integrity Requirements and Industry Standards O
ASMC E&PV Code la salle Sect. Ill AN51 E&PV Code Ciass :
ciais r 40 e3i.i (ie qs t r e r+ n t s) ( Af plication)
Piping Integrity Fequirement (Rewirerents)
! Yes No Yes DESIGN Enign hecification 5 tress Analysis Yes tes Yes 5 tress Frport Yes ho Tel o _ _ _ - _ -
PA flRI AL Peterial 5;-ec ification ASME A51M A51M b ter 'e1 t aemiration Ter Fat') Fer Fat'l Fer %t'l Sr.ec S t,ec Spec l ull Ter.e tr a tion Fipe > 2* Ptpe >3* Fipe > 2*
fvtt held Klhi
'O 5x h e t Weld Pipe 12* Pipe 1 3* Pipei?*
Throded ho ripe 13* No I n 41 CAT 10N _ _ _ . _.
Pr oc e ss Sect. II Sect. II Sect. 11 I"1 AL L Ai!Cs. p),, p g ,g, -
Wild GTAW Root G1AW Ecot Weld Frep B10.25 bi6.25 216.25 wit o e ti 'eio a' ni 5 PT O Is M Socbet Leld P1 Or MT v
y PT Ol'41 T V Ceal, t.Dt ier'sonal tes Yes* its AS$UG%([ PJterial s co'Os e %es Tes* Ves -
I it.ric a t ion f rs e kre its Yes+ Tes fab & [sas becords Tes 1es* Ves Q4 Frogree its Ves+ Ves 0 - - - - - - -
O GTAW - Gas Tungsten Arc Weld MT - Magnetic Patricle Examination RT - Radiographic Examina tion V - Visual Examination PT - Liquid Penetrant Examination l + - Per Owner / Manufacturer Practices O
M 26 O t
- m. ,
O the ASME Code,Section III, Class B, 1974 issue. All the piping stresses and valve accelerations are within the allowables. Cygna Energy Services has reanalyzed the insert, withdraw, scram discharge 2
volume and the vent lines piping systems to a recently revised required response spectra. The stress analysis report, which meets the requirerrents for an ASME,Section III Class 2 piping analysis, will soon O
be completed. It has been concluded by GE and the NRC staff that the SDV header piping is not sensitive to fatigue failure and would be exempted from fatigue analysis even if it were Code Class 1 piping .
O The rules of NB-3630 permit Class 2 analysis of Class 1 piping if:
]
(1) the normal piping size is 1 inch or less, or (2) fatigue exemption i rules are satisfied for larger diameter piping. Nevertheless, fracture mechanics verification of piping integrity is presented in Appendix A.
O Seismic qualification reports on the active essential equipment in the SDV/CRD system have been reviewed by Sargent & Lundy. The active O essential mechanical hardware are:
- 1) MPL Cll-0001, HCU #761E5000001,
- 2) MPL Cll-F009, solenoid valse for instrument air #21 A9317AC,
- 3) MPL Cll-F010, 1" globe valve, A0 #21A1750P001, (pressure
.O boundary only), and
- 4) MPL Cll-F0ll, 2" globe valve, A0 #21A1750P002, (pressure boundary only).
- O Items 1 through 3 have been qualified to the 3QRT criteria. (Sargent &
Lundy file number EMD-032130, CQD-000255, and CQD-000256). Item 4 will be tested in February 1982. The SDV level switch #145C3047 is qualified in Sargent & Lundy file number EMD-029805. The future dP level switch
,0 is undergoing qualification now.
3.2.2 Materials l La Salle has purchased all of the piping system material to the specifi-
!O cations in the Summer 1975 addenda of the 1974 Section III ASME Class 2 Code. Each nuclear part includes a tranufacturer's data report, certifi-1 27 0
D cate of design specification, and certification of shop inspection.
D These documents are contained in Reactor Control Inc. file La Salle #45, Unit #1, Scram System.
3.2.3 Fabrication and Installation D All of the SDV and CRD piping having a diameter greater than 2" was full-penetra tion butt-welded. This specifically includes the 8 inch scram discharge header piping and 12 inch instrument volume piping systems. These welds were perfont.ed to the requirements of ASME Section 3 III Class 2. Construction welding inspection for these systems included radiographic and liquid penetrant examinations. Documentation is available on: (1) field and shop weld identification, (2) type of weld, (3) type of joint, (4) pipe size and schedule, (5) material 8 specification, and (6) type of weld examination for each weld.
The SDV and CRD piping runs having a diameter 2 inches or less were socket welded. There are no threaded connections in these systems. Each 3 of these welds was inspected in accordance with ASME Section III, Class 2 requirements using liquid penetrant examination.
Since the full penetration welds were volumetrically examined and the
- socket welds were exair,ined by surface methods, there is a low prob-ability of having a fabrication defect of a size large enough to result in a piping failure. Even if a fabrication defect is hypothesized, it is shewn in Appendix A that the probability of a piping failure due to D the growth of such defects is very low.
3.2.4 Quality Assurance Quality Assurance procedures are being used at La Salle as illust;ated D in tne material specification and welding certification documents des- 1 l
cribed above. The Quality Assurance programs use established design and installation procedures (Ref.4) as well as a walkdown procedure (Ref.5) to assure piping integrity and appropriate documentation. Nondestruc-tive test personnel are required to be formally trained. All fabrica-tion procedures require approval in advance.
9 28
O As explained in Reference 6, Reactor Controls, Inc. has totally reviewed o the QA/QC program and improved their QA manuals to appropriately address the areas of organizational interfaces, design control, document control, installation, and inspection. CEC 0 has now established a site QA moni'.oring program with monthly status reports to assure timely O completion of comnitted corrective action. In addition, the CEC 0 QA de-partment has been reorganized to include three supervisory level personnel at the construction site so that on-site and off-site quality items will be more closely followed.
- O 3.2.5 Preservice and Inservice Inspection, ASME Section XI A Preservice Inspection of the CRD system was conducted in accordance with the ASME Boiler and Pressure Vessel Code Section XI, Summer 1975 O Addenda. This includes ultrasonic examination and liquid penetrant i examination of several welds in the scram discharge headers, and is described further in the La Salle Unit 1 Preservice Inspection Report and in the la Salle Unit 2 Preservice Inspection Report. Future
.O Inservice Inspections are described in the La Salle County Station, Inservice Inspection Program for Unit 1 and for Unit 2 (separate documents).
O 3.2.6 As-built Inspection An as-built inspection of the SDV and CRD piping systems and their sup-ports will be completed at la Salle by January 31, 1982. At that time, the stress analysis reports will be reviewed to verify that they reflect O the latest revisions of the applicable piping layout drawings.
3.2.7 Maintenance and Modification of SDV Piping All maintenance related and modification work on the control rod drives, CO hydraulic control units, and SDVs is controlled by CEC 0 QA procedures, (Topical Report CE-1) and La Salle Station administrative procedures.
Work on CRDs, HCus, or SDVs requires multiple levels of review nrior to conrnencing the work. Work orders specify the plant conditions re-
- O quired prior to starting the work. Final approvai an any work order is granted by the Shift Supervisor.
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3.3 141TIGATION CAPAL!LITY
) 3.3.1 Introduction As discussed in Section 3.2 and Appendix A a loss of integrity of the BWR Scram System Piping is extremely unlikely. Nevertheless, in consideration of a SDV failure or some other postulated pipe failure
) outside the primary containment at the LSCS, Commonwealth Edison Company (CECO) has adopted several policies to accomplish a timely and success-ful mitigation of the event. This section considers the operator response to a normal scram and to a postulated scram which cannot be
) reset followed by an SDV failure. The following topics are considered:
- 1. Operator training;
- 2. Control room alarms associated with the SDV event;
)
- 3. LSCS reactor operator procedures for the activated alarms;
- 4. The impact of loss-of-offsite power on the operator's response; and 3 5. Manual isolation of the failure.
This information demonstrates that with the available leak-detection systems and the LSCS emergency procedures the reactor operators are able to determine if a significant radioactive water / steam leak exists out-3 side primary containment such that a rapid depressurization - cooldown at a rate greater than 100 F per hour - is necessary.
3 Next, this sectinn considers the specifications for primary reactor coolant activity and the associated doses for a postulated SDV event.
This is followed by an analysis of the systems available to maintain reactor pressure vessel (RPV) inventory in the event of a postulated SDV 3 fail ure.
3.3.2 Operator Response to a Reactor Scram To identify operator actions in the unlikely event of a SDV failure 3 first consider the normal progression of events after a reactor scram.
Af ter either a manual or an automatic scram the reactor operator implements LSCS Reactor Scram Procedure, LGP 3-2. Step 12 in Section F 30
O of LGP 3-2 instructs the operator to bypass the scram discharge volume g high level signal and reset the scram. The operator can confirm that the reset is successful by monitoring the indicator lights on the full core rod display. The blue lights on this panel indicate that scram valves are open. Resetting the scram closes the valves and extinguishes g the blue panel lights. A successful scram reset immediately terminates an SDV event. The time required to reset a scram depends on the problem which initiated the scram, but for most cases the reset would be 1 attempted in about ten minutes or less.
O 3.3.3 Operatar Training All LSCS shift supervisors and licensed operators receive three days of l
j simulator training per year. Many of the LSCS operators have had the O SDV event included in their simulator training. Further, the SDV event has been discussed in the classroom training of the LSCS operators.
Simulator and/or classroom training for a SDV event will be included in the training of all LSCS shift supervisors and licensed operators.
O
(,EC0 does not have a specific procedure for an SDV event or a pipe failure outside primary containment. Instead, the operators are trained via the Symptomatic Emergency Frocedures to arrive at a timely and O successful mitigation of any such event.
3.3.4 Control Room Alarms Associated with a SDV Event As pakted out in IEDO-24312 and flVREG-0803, there are several abnormal O event signals which occur in the control room as a result cf a SDV break. These include:
- 1. Reactor Building Area Radiation Monitor Alarm
- 4. Reactor Building Ventilation High Radiation Alarm
- 5. Reactor Btalding Differential Pressure Alarm
- 6. Personnel Observation of Leakage !
O Any one of these signals by itself would provide little or no informa-tion on SDV status, but the anticipated combination of signals quickly j identifies a significant leak outsiae of primary containment.
O 31
3 3.3.5 Alarms With Procedures That Include a Rapid Depressurization The LSCS Reactor Building Area Radiation Monitor Alarm and the Reactor Building Floor Drain Sump Alann annunciator procedures include steps which reference procedures that describe when a rapid depressurization is warranted. The location. operation and calibration of the area radiation monitors are described in LOP-AR-01, LRP-ll50-4, and 3
LIP-AR-01 The sump and sump pump operability is covered in LOS-ZZ-Q2.
3.3.6 Op. erat _or Response to a SDV Event 3 The reactor operators' response to a SDV event following a scram which cannot be reset can be anticipated by examining the LSCS procedures for the alarms associated with the event. Imrediately after a the scram and a postulated SUV failure, the Reactor Building Area Radiation Alarm
, would be initiated. The operator would them proceed according to LSCS LOA 1(2)H13-P601 8110. This procedure directs the operator to:
determine the affected area evacuate the area if necessary 3 -
direct the Rad / Chem Department to validate the high radiation alarm and determine the source " possible notify the Shif t Supervisor refer to LOA-AR-02 3 Thus, the operator will determine the area of high radiation and direct personnel to locate the source and refer to LOA-AR-02 which pertains to the detailed quantification of the radiation in the affected area, e The operator respor:se per .tation procedures for the sump alarms enables him to determine the extent of leakage from piping or equipment, the nature of the leak (fluid), the contamination level, the approximate size of the leak, and the affect of the leak on associated systems or D vital plant equipmeat. From reference to these two station procedures and the results therefrom, the operator receives an indication and ,
confirmation that a rapid dopressurization in excess of the 100 F per hour cooldown limit may be in order. Further the operator will have 7
8 referred to LGA-01, Level Control and LGA-02, Cooldown and the Shift Supervisor will have classified the event for severity and possibly a
J 32
O initiated GSEP. The time frame for these actions is greater than five minutes but probably less than fifteen minutes. In this initial period O
the operator will have considerable additional information available from alarms such as the Reactor Building Ventilation High Radiation Alarm and the Reactor Building Differential Pressure Alarm. Thus, the g operator is aware of the relative size of the break with respect to the integrity of tM secondary containment. Numerous high radiation alarms would indicate the need for immediate action to minimize radiation re' eases in case of potential loss of secondary containment integrity.
- O Bec use n ither the ECCS leak detection alarms nor the primory coolant isolation valves for the emergency cooling systems or the safe shutdown systems have been activated, the viable conclusion is that there is a break in a line with high temperature radioactive water, i.e. primary O system coolant sater, outside the primary containment. Further, this break either has compromised or might compromise the integrity of the secondary containment0 . Because it is assu
- ned that the operator cannot successfully reset the scram, the operator has enough information in O c njunction with the directions in the two station procedures to initiate a rapid depressurization and effectively terminate the accident.
3.3.7 Personnel Observation of Leakage
,o The physical location of the operational support center, the control room and the shift engineer's office facilitate personnel observa-tion of a SDV failure. Each of these facilities is on the 768' level of the Auxiliary Building, i.e. they are immediately adjacent to the 761' O level of the Reactor Building where the HCU-SDV area is located. Oper-ating personnel, not on rounds or attending equipment, are normally in the operational support center. If operating personnel are directed to, or have other reason to enter the reactor building the most probable o point of entry would be the 761' level because of its proximity to their usual location. Thus, an operator entering the reactor building in response to an alarm would immediately see the unusual environment.
Further, it is anticipated that the acoustic ooise associated with the O event will be loud, and it is likely that it would be noticed prior to entry into the Reactor Building.
O 33
'O 3.3.8 fontrol Rod High Temperature Alarms _(Alternate Dectection of_
Problem)
-O The information cited above is sufficient for the operator to determine that a rapid depressurizatiors was necessary in this situation. However, as pointed out by GE in response to fikC questions on fiEDO-24342 for the SDV event, the Control Rod High Temperatura Alarm would be activated at O
i a greater than normal rate, e.g. GE estimates that nearly all the CRD's would exceed the high-temperature setpoint instead of the 5-15% that are usually seen af ter a scram. The LSCS high temperature scan samples at a rate of 5 seconds per point. Thus, a scan of all the control rods is O
completed every 16 minutes. When the control rod high temperature alarm sounds the operator would refer to procedure LOA 1(2)H-13P603 A403.
This procedure directs the operator to go to the Control Rod Drive O Hydraulic Temperature Recorder to determine which of the CRD's ir.itiated the alarm and then if all the CPD's are reading high temperatures to verify that the LRD system cooling water flow is operating. Upon completing this verification the operator is directed to go to the
.O affected CRD Hydraulic Control Units to verify that the CRD cooling water stop valve is fully open. Thus, if for some reason either the area radiation or the floor tunp alarms failed, the operator still goes into the Reactor Building and hence would detect the problem.
O 3.3.9 Loss-of-Offsite Power Coincident with the SDV Failure (Alternate Detection of Problem)
The two area radiatica monitors in the HCU-SDV area are not connected to o an emergency peser source. Thus, if offsite power is lost, these detectors would not operate. However, the sump's HI-HI alarm is not dependent on offsite power. Moreover, area radiation monitors on other floors are connected to emergency power. The maximum delay associated
'O with the loss-of-offsite power would be either the time required to trip ,
the sump alarm, or the time required for the escaping fluid to reach an active radiation monitor. Further, the reactor building ventilation high radiation alarm would also be active. These alarms in conjunction
-Q with operator entry into the reactor building in response to the sump or area radiation alarms or to conditions caused by the loss-of-offsite power still assures an early detection of the problem.
O 34
i
)
3.3.10 Impact of SDV Leak Rates Less Than the Maximum Post _ulated if the SDV leak rate is less than the maximum postulated flow, then
) there may be an associated delay in alarm activation and operator response. However, if the flow rate is lower, any potential problems associated with the leakage will take a longer time to develop. Her ce, the possible delayed alarms and response will not have detrimental
) consequences.
3.3.11 Manual Isolation of the SDV Leak Af ter a rapid depressurization, the flow through any SDV crack would be I reduced to subcooled water so that only residual steam pockets wot!1d present high temperature problems. Further, because LSCS primary coolant activity is limited to the Standard Technical Specifications, radiation levels following depressurization would not preclud9 personnel !
entry into the HCU-SDV area. At LSCS, entry to manually itolate the leak would be made in accordance with the Generating Station Emergency I I
Plan (GSEP).
3.3.12 Primary Coolant Activity _
The technical specifications for LSCS reactor coolant system currently specifies that the specific activity of the primary coolant shall be limited to less than or equal to 0.2 microcuries per gram DOSE EQUIVALENT l-131, and less than or equal te 100 / E microcuries per gram.This is the Standard Technical Specifications (STS) cited in the plant cuidance of t'UREG-0803, Chapter 5. Thus, the LSCS radiation levels in the reactor building would permit operator access following an
)
SDV leak provided that routine LSCS precautions for entering potentially high-radiation areas are followed. Offsite doses would remain within regulatory limits.
As noted in NUREG-0803, the equilibrium iodine activiQ in the reactor coolant is generally considerably lower than the STS limits. LSCS FSAR calculations estimate that the equilibrium concentration of I-131 will be 0.013 microcuries per gram. Thi". value is fifteen times smaller than
)
the STS activity limit.
)
35
l
) i The LSCS secondary containment design is likely to prevent long-term radioactive releases to the atmosphere. The secondary containment
) (reactor building) is vented into the steam tunnel. Thus, before the secondary containment integrity can be comprised, the steam tunnel, )
which constitutes a large volume, would have to be pressurized.
) Further, should the steam tunnel volume become pressurized, blow-out panels would relieve the pressure to the turbine building thence to the station stack. Thus, the possibility of the SDV cuent permanently compromising the secondary containment integrity is significantly less
) likely than a containment design with reactor building blow-out panels as referenced in the generic evaluation.
3.3.13 Emergency and Long-Term Cooling Capability D
in flVREG-0785, the emergency and long-term cooling of the reactor core following an SDV event was questioned on the basis that the corner rooms wh!ch house the emergency and long-term cooling pumps would be flooded and the operability of these pumps may then be questionable. In D
flVREG-0803 prompt depressurization followed by manual isolation of the leak is presented as a solution which mitigates general flooding.
Notwithstanding these fach, the LSCS secondary containment design precludes submergence of the corner rooms and relegates flooding to very D
remote probability when compared to the GE generic solution. This design is documented in the FSAR.
3.3.14 Secondary Containment Design and Emergency Cooling Pump Locations At LSCS there is only one stairwell with a direct path into a corner basement room (LPCS/RCIC cubicle). In addition at LSCS the basement g corner rooms are isolated from the basement inner annulus by water-tight doors. These doors are normally closed; a control-room alarm is activated v9 en a door is not closed. The floor drains on the 761' level drain into sumps a:hich are located in the basement enrular region. These 3 floor drains can carry almost all the leakage from a SDV failure into these sumps. If the sumps overflow and/or the sump pumps fail, leakage J
36
1 0
into the corner rooms will not increase; correct operation of floor dra'ns was verified via pre-operational tests. The LPCS and the RCIC are
.O located in the corner room with the open stair-way to the HCU-SDV level.
The LPCS pump motor is set on a pedestal with the motor being approximately nine feet above the floor.
- O 3.3.15 EmerSency Core Cooling.Following_a SDV_ Failure HPCS, LPCS, RCIC or any one of the LPCI subsystem pumps could maintain the RPV level during an SDV event with a maximum leakage of 550 gpm.
i
.O 3.3.16 Alternative Sy_ stems Available for Emergency _C_ ore Cooling If it is postulated that the normal ECCS are not operable, then several other systems are available to maintain RPV inventory. These include the feedwater pumps, the condensate systems and the RHR service
)
water. The availability of these systems has been discussed in flEDO-24342 and accepted in !1UREG-0803. These systems are located outside the reactor ballding and the impact of the degraded reactor building environment on th< se systems is discussed in the foliowing C)
Section.
lO 3.4 ErlVIR0fiMENTAL QUALIFICATION l If an SDV event is postulated, the reactor building environment may perhaps reach 212 F, 100% humidity only in the immediate vicinity of the
() SDV rupture. The overall building environment may reach 140 - 150 F, with a maximum of 100% humidity in environmental zone H2A for a maximum of four hours after initiation of the event. The precise conditions will depend on the availability of ventilation systems, the exact time
!C) until prompt depressurization, the magnitude of the leakage flow, and the status of the CRD pumps. The ability of LSCS systems to successfully mitigate this event in the harsh environment has been reviewed.
O 1
1
- O 37
O In addition to the emergency and long-term core cooling equipment -
g discussed in Chapter 2, local instrument racks containing instruments that control reactcr level and pressure, or provide opr.rator information -
1 were investigated. It was found that there are also four instrument racks for reactor level and pressure located within environmenal zone g H2A which is common with the Scram Discharge Header (SDH) and Scram Discharge Volume (SDV). A review of the instruments on the local racks shows that an adaquate number of instruments were qualified to the local environment to provide t eactor level information to the operator (see g Table 3.2). Cold shutdown of the La Salle Station BWR/5 can be achieved with only level information if available or even without any level infonnation if the symptomatic emergency procedures are used.
lg The safe shutdown and long-term cooling functions have been designed to -
assure operability even if gr ,ups of instruments and/or subsystems in a given plant are rendered inoperable by local environmental conditions.
Chapter 9.5 and Appendix H.4 af the LSCS FSAR fully documents the Safe o Shutdown analysis and assures a generalized capability for safe shutdown and long-term cooling.
3.5 CONCLUSI0flS
- O The LSCS SDV system has been designed and fabricated to ASME Section III, Class 2 code. The welds have been tested according to the code by radiographic and liquid penetrant techniques. Further, the system has
.O been preservice inspected according to ASME Section XI code. The preservice weld inspections used ultrasonic and penetrant techniques.
The stesses on the system have been analyzed and are within the I allowables. The stress analysis included seismic loadings. CECO Q.A.
l
.O procedures and LaSalle Administrativa procedures assure that the SDV system integrity will not be degraded by modification, inspecticn or similar work while it is required for safe reactor operation.
Complementing these results is a probabilistic fracture mechanics O evaluation of the SDV piping waich shows that the probability of a significant fracture is less than 7x10-6 per year.
38
U U kl U U U V U V \d \/
1 l
TAR E 3.2 LOCATION FthCTION 7(ALIFICATI04 PUMP MOTORS Time Terp Pres. Hum e LJ.!C/LPCS Pump Room LPCS Pur;: Drive GE SERIES f Sri 1111 Reactor B1dg.
t M Vertical Motor a R4R A P ep Drive hema WP-1 RuR Pump Rcom Steam A Reactor Bldg. 0-6 hr 2120F twice
B/C Reactor Bldg.
HPCS Pump Room bPCS Pump Do ve Re&ctor Bldg.
INSTRUMENTS Reactor Bldg. Elv. 710' Recirc. pump trip ' hr IS$F 7C0%
- 1. Level Indicating Switch Yarway 4418C Local Panels H22-P026 on low water fevel . I hr 2MF ICC:
{$ Peactor water level I hr 0 235 p jgg; and H22-P027 narrow range. I hr 253"F 100 ECCS initiation trip I hr 4DCF ATH - 20%
- 2. Level Ir.dicating Reactor Bldg. Elv. 710' 1.0* WC Switch Barton 288A Local Panels H22-P004, and RPS SCRAM 1 br 700F ATit ATH H22-P005. H22-P026 and 1 hr 100c F AT/ ATH H22-P027 1 hr 2120F 7.c' WC 1001 Reactor Bldg. Elv. 710' Senses drywell pressure
- 3. Pressure Switch and init' *es ECCS on Static 0-Ring Local Panels H22-P004 The switch is functional in H22-PD05. H22 >026. and 2 psig signal. enviryments of: 0
'io F/92t RH 12N-AAA-(x9)TT H224027 and 2:2'F/1001 RH
- 4. Pressure Transmitter Reactor Bldg. Elv. 710' The irstreent must Local panels H22-P004 maintain pressure N/A
- Bailey 556 integrity only.
and H22-P027. ACIC/LPCS Purro Room Local Parel Elv. 663' H22-P017. RHR A Pump Pa. Local Panel H22-P018.
- The base of all pump motors is 9' above the Pump Poom Flaor. The f.ema WP-1 frane is capable of direct impingenent of water on the top of the motor without affect'ng the motor operation
TABLE 3.2 - continued
' QUALIFICATION IN3TRLFENTS LOCATION FLACT 10N Time Temp Pres. Hurd.
- 5. Differential Dressure Reactor Bldg. Elv. 710' Provides control infor- ' 10 min 350DF 60 PSIG Dry Transmitter Rosemont Local Parel H22-P004 mation for operation 1 tr 3160F 70 PSIG Saturated Model 1151 DP H22 - P005, and H22-P027 system initiation. 7 nr 3030F 55.4 PSIG Saturated Reactor Bldg. Elv. 663' 42 hr 2300F 6 PSIG Saturated RCIC/LPCS Puma Room Local Panels H22-P001 and F22-P017. RHR A Pump acc-m Local Panel H22-P018.
RdR B/C Pump Room Local Panel H22-P021. HPCS Pump Room Local Panel H22-P024.
Reactor Bldg. Elv. 710' Senses reactor pressu r e 700F tn 2120F in 200F
- 6. Pressure Switch increments for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Barksdale Local Panels H22-P304 initiates RPS on high BIT-M12SS-GE H22-P005. H22-P026, and vessel pressure H22-P027
- 7. Dif ferential Pres. Reactor Bldg. Elv. 663' Provides nermissive signal 6 hr 2120F 7.0"hC 100%
g, Switch. Barton 2B8 RCIC/LPCS Pump Room for LPCS injection valve c3 Local Panels H22-PC01/P029 when a P across valve is RHR A Pump Room Local low enough Parel H22-P018. RHR B/C Pump Room local Panel H22-P021. HPCS Pump Room Local Panel h22-P024
- 8. Differential Pres. Reactor Bldg. Elv. 663' 6'hr 2120F 7.0" WC 1001 Transmitter Barton RCIC/LPCS Pump Room Local M0D. 289 Panel H22-P001 and H22-P017.
RHR A Pump Room Local Panel H22-P018. RHR B/C Pump Room Local Panel H22-P021 HPCS Pump Poom Local Panel H22-P024
- 9. Pressure Switch '** Reactor Bldg. Elv. 663' Monitors pressure of the Not type tested to be Barksdale RCIC/LPCS Pump Room Local RCIC stm. turbine Environmental Envelope D2H-M8055 Panel H22-P017 and H22- exhaust.
P029
- 10. Pressure Switch Reactor Bldg. Elv. 663' Provides permissive 6 hr 2120F 7" WC Saturated.
Barksdale RCIC/LPCS Pump Room signal to ADS logic PlH-M340SSY Local Panel H22-P001 and H22-P017
- This switch is similse to Barksdale BIR-M12SS-GE in construction & function and therefore is considered to meet the environnental requirenents.
Ts?tE 3.2 - continued l
00ALIflCAT10N INSTRUMENTS LOCATION FUNCTION Time Temp Pres. Humd.
11, Pressure Switch Reactor Bldg. Elv. 663' Provides a Signal to 6 hr 2120 " 7" WC 100:
Statif 0-P.ing RM1 A Pump Room Local ADS Interlocks to pre-SN- AA3-( x10)-S ITT Panel H22-P0lB. RHR B/C vent A05 bicwd .n Pump Room , Local Panel without LPCI/RHR H22-P021. HPCS Pump Room availability Local Panel H22-P024
- 12. Pressure Switch Reactor Bldg. Elv. 663' Same as #11 Same as til Static 0-R;.ig RCIC/LPCS Pump Room 6N AA21V Local Panel H22-P017 RHR B/C Pump Room Local Panel H22-P021. HPCS Pum? Ucom Local Panel H22-224. Local Panel H22-P029 RCIC/LPCS Pump Room.
[$
- 13. Pressure Switch Reactor Bldg. Elv. 663' N/A Robert Shaw RCIC/LPCS Pump Room SP-222-C Local Panel H22-P001.
RHR A Pump Room Local Panel H-22-P018. RHR B/C Pump F.;om local Panel H22-P021 HPCS Pump Room Local Panel H22-P024
- 14. Level Indicating Reactor Bldg. Elv.710'-0" Reactor water level Transmitter Switch Local Panel H22-P004 wide range Barton Mod. 760 H22-P005. H22-P026, and H22-P027 1 br 1000F ATH ATH 1 hr 2120F 7.0 WC 100%
O The LSCS reactor operators and shift supervisors have or will receive O SDV event training. The presently installed leak detection equipment and the LSCS LOA procedures assure an early detection of a postulated SDV rupture and a subsequent prompt depressurization.
- O The reactor coolant activity limit is the STS limit. Analyses in NUREG-0803 show that for the STS limit doses to personnel entering the HCV-SD7 area are not prohibitive if reasonable precaustions are ta<en.
Further, any release to the environment will be less than those
'O specified by the current regulations.
There are numerous emergency and long-term core cooling optiens available to the reactor operators. The design of the secondary O containment and the equipment placement are such that water impinge-ment or flooding are extren:ely unlikely even if depressurization should be delayed. All of the equipment needed for safe shutdown after the SDV event is qualified for the anticipated accident environment. In O addition there are independent systems - feedwater and condensate -
outside the reactor building which can be used to maintain RPV water level.
O Thus, the LSCS SDV piping integrity, mitigation capability, and equipment environmental qualification assures that the postulated SDV event would be successfully mitigated.
- O O
O 42
O REFEREfiCES_
O
- l. Pi_ ping Stress Analysis, Commonwealth Edison Company, La Salle County -
Unit 1, 4266-10, Control Rod Drive, RD-05, performed by Engineering IEchanics Division of Sargent & Lundy, DID-032-611, Rev. 2, 8-20-81.
~
- 2. La Salle Unit 1 Piping Reanalysis of the Control Rod Drive Hydraulic O System (CRDHS), by Cygna Energy Services, to be published. ,
- 3. Generic Safe _ty_ Evaluation Repor_t Regardina Integrity of BWR Scram System Piping, USNRC, NUREG-0803, Auaust 1981.
- 4. Quality Assurance lianual, Reactor Controls Inc. , 11-21-77.
- O s
- 5. "QA18-4, Procedure for Final Walkdown," QAI fianual, Reactor Controls Ir, .
- 6. Letter from C. Reed, Commonvealth Edison, to J. Keppler, NRC,
Subject:
La Salle County Station flRC Inspection Report 50-373/80-48 and 50-374/80-30 NRC Docket fios. 50-373/374, dated February 3,1981.
O
- 7. Section C4 of LGA-02, Cooldown says, "... The 100 F/hr cooldown rate may be exceeded to conserve RPV inventory, protect primary or secondary containment integrity, or limit radicactive release to the environment."
g" 8. GE has estimated that the reactor building check dampers would open approximately 5 minutes after activation of the Low Differential Pres-sure Alarm. However, the LSCS design is such that the Reactor Building air is vented into the steam tunnel. The steam tunnel has blow out panels which are designed to relieve an overpressurization. Thus, the probability that this event would cause the loss of the Reactor o Building roof is considerably less than for the design analyzed by GE in NED0-24342.
- 9. An alternate BWR coolant concentration technical specification is under consideration to replace this PWR gener ic specification.
O
- O
-O l
O 43 )
i 1
O O
O O
APPENDIX A 0
PROBABILISTIC FRACTURE !!ECHANICS EVALUATION OF SCRAf1 SYSTEl!S PIPIT 1G
'O O
O O
1 O
O A-1
O FRACTURE MECHAf1ICS ANALYSIS OF SCRAM PIPING RELIABILITY O
D.0. Harris SAI - San Jose Janua ry 11, 1982 "n
INTRODUCTION In order to address concerns regarding the integrity of BWR scram system piping (1,2), and to assess the consecuences of failure of such piping, it is necessary to estimate the probability of failure of various line sizes. This has been accomplished in a preliminary manner (3) using procedures developed for the reactor safety study (4). However, such~
procedures do not take into account specifics of the pipe design and oper-0 ation that are known to influence the piping integrity. Such factors in-clude operating stress levels, number of stress cycles, and frequency of inspection and proof testing. Additionally, no estimates are made of the reliability of pipes of diameter less than 2 inches. Much of the scram piping is 3/4 inch diameter. In order to estimate the piping reliability for a variety of pipe sizes ud to account for specific operating conditions of the piping system, an analysis was performed to estimate the failure
,0 probability of the scram piping at La Salle Station. The procedures employed and results obtained will be presented in the following sections.
Basically, the methodology assumes that piping failures occur due to the gr wth o' crack-like defects introduced into welds during fabrication of O
the pipe. These initial defects are considered to be randomly distributed in both the number of defects and their size. The as-fabricated defect distribution is altered by pre-service inspection according to detection g probabilities associated with the inspection procedures. The post-inspection defect distribution then serves as initial conditions for fracture mechanics calculations of crack growth that occurs as a result of service conditions.
The probability of failure at a given weld location at a given time is equal O
to the probability of a crack larger than the critical crack size existing at that location and time.
l l
l O
A-2
Such procedures are generally referred to as "probabilistic fracture mechanics" and have been widely applied to nuclear reactor pressure vessels and piping. Reference 5 provides a comprehensive review of work in this area and also serves as an example of the current state-of-the-art. Refer-ences 6 - 8 provide additional discussions in this area. Figure 1 schemat-ically shows the various steps involved in the analysis.
The scram piping under consideration is seamless, and all welds are therefore circumferential. Interior surface part-circumferential cracks, such as shown schematically in Figure 2, are therefore the crack geometry of most concern.
Co:nplex calculations of crack growth can be performed for such cracks (5,8).
)
The calculations can be greatly simplified if it is assumed that the crack is very much longer than it is deep (b/a 1). In such a case the crack becomes one-dimensional and the analysis is greatly simplified. Reference 6 provides I an example of a 1-D approach, which will be used in the analysis of the scram piping system. i Proof testing can be very beneficial in increasing piping reliability (5,9),
because the fact that 3 weld joint survived a proof test indicates that no
)
cracks larger than the critical size were present during the proof. This allows the crack-size distribution to be truncated at the critical size corresponding to the proof conditions. This will be discussed in more detail later, and can have a significant impact on the calculated failure prob-
)
abilities.
REVIEW OF PIPING INPUTS
) .
Pipe failures (leaks or complete pipe severances) that can produce appre-ciable leak rates and can not be isolated by valves are of concern. Leak rates due to failure of a 3/4 inch scram discharge line between the hy-draulic control unit (HC!)) and reactor pressure vessel will be limited to
)
the leak rate past the control rod seal. Such failures are therefore not of concern. However, failure of such a line between the HCU and header can result in higher leak rates that are limited only by the 3/4 inch pipe dia-meter. Therefore, attention will be concentrated on all lines downstream of
)
the HCU, including lines out to the first isolation valve. This results in
) A-3
O pipe lines out to the valve in the 2 inch drain line (Cll-F010) and the ;
valve in the 1 inch header vent (Cll-F0ll) being considered. Table 1 !
summarizes the piping'sub-systems considered and their corresponding O sizes and wall thicknesses. Also included is the piping material and the number of welds in each line. These numbers were determined either l directly from piping drawings that indicate weld locations, or from piping layouts with a bend (elbow) comprising two welds, a valve two welds and a
" tee" three welds. This will tend to overestimate the number of welds,
>O because some 1ines are bent rather than having welded fittings. This i procedure is therefore secewhat conservative.
i O The stresses in the various piping systems listed in Table 1 are required as inputs to the fracture mchanics analysis. The scram piping is subjected to a number of transient types. Only normal operating stresses will be considered here, and only the stresses due to pressure, (o ), deaducight, (oDW)and restraint of thermal expansion (oI E) will be considered. Maximum stress levels or loads were obtained from various references with results summarized in Table 2. The axial component of the stress due to internal pressure (o p) was ,
calculated from the following expression -
<O -
o = E U D)_ (1) p 4h where h is the pipe wall thickness and p is the reactor design pressure of 1250_ psi. .
The deadweight ( DW) and restraint of thermal expansion stress (oTE w re evaluated from the corresponding moments by use of the following expression ,
(DW or TE) - 1 The following stress components are also of interest loaa controlled stress =
DW + p (3)
O LC cyclic stress = Ao = o
+oE (4) p T max. proof ; tress = =
1.25 op +cW D (5) prf The 1.25 coefficient in equation 5 is due to the proof pressure being 1.25 lO times the operating pressure (reference 2, page 3-2).
The number of times during the plant lifetime that the pipes are subjected to the stresses shown in Table 2 is also required for the fracture mechanics O analysis. This number of stress cycles will equal the number of times the .
A-4
._ ___, _ _ 1_ _ - . _ _ - . _ .
reactor is scrantned. Reference 1 (page A-1) suggests a rate of twice per year, which would result in 80 cycles during a 40 year plant lifetime. However, this estimate may be somewhat optimistic. Reference 3 (page A-2) suggests a rate of 6.1/yr. or 244 per plant lifetime, and reference 14 suggests 320 per plant lifetime. Values of 200 and 320 will be considered, with the latter value serving as an upper bound estimate.
, FRACTURE MECHANICS MODEL INPUTS a
The inputs to a simplified one-dimensional crack f racture mechanics model will be summarized in this section. These inputs will be combined with in-formation provided in the previous section to perform the fracture mechanics analysis presented in the next section.
Failure Criterion _: A failure criterion is required in order to define the 3
critical crack size for the various piping systems considered. The pipes are fabricated from SA106B carbon steel, except for the 304 stainless steel scram discharge lines (See Table 1). Both of these materials are tough and ductile, and will not fail in a brittle manner. !ikewise they will not fail due to a tearing instability. Reference 5 provides a detailed discussion on this topic for 304 SS. However, a catastrophic failure can occur when a crack of suffi-cient size (or area) exists to reduce the remaining cross-sectional area of the pipe to the point when it is not sufficient to sustain the load controlled component of the applied stress. Hence, a net section stress failure criterion 3
is applicable (12). This criterion was applied to 304 SS reactor piping in reference 5 and to SA106B carbon stael reactor piping in reference 3 and can be expressed as D
(A p -A )o fjg =A c LC (6) of)g is the critical net section stress, which is equal to (yield strength +
3 tensile strength)/2 (12). A pis the cross-sectional area of the pipe and A r is the critical crack area. A value of o f)g of 45 ksi will be used for both ma terial s (3,5) .
3 If the crack is taken to be one-dimensional, it can be conservatively assumed to be complete circumferential with a depth a. Equation 6 then reduces to the following expression (see Reference 6).
O A-5
O a
c = h (1 - oLC / flo)
(7)
Subcritical Crack Growth Characteristics: Subcritical crack growth in O
reactor piping can occur due to stress corrosion cracking (SCC), fatigue crack growth or environmentally enhanced fatigue crack growth. SCC has not been observed in carbon steel lines, but has been observed in sensitized welds of 304 stainless steel (15). However, SCC is a time dependent process, and O
the scram discharge line is under load only for a short period (s200 hrs. ,
see reference 2). Therefore SCC is ruled out as a significant contributor to crack growth. Environmentally enhanced fatigue crack growth then remains as the dominant contributor to subcritical crack growth. The following relations
-g provide conservative estimates for the naterials under consideration in an operating reactor environment 304 SS (ref. 16) $=10-9(AK)4 (8)
'O SA106B (ref. 17) h=1.68x10-9 (AK)2.37 (g) 5 n AK - cyclic stress intensity factor, ksi - in v
da/dn - crack growth rate, inches / cycle These relations are conservative and are applicable to high mean stresses. In fact, information in reference 5 suggests that, for stainless steel, the prob-O -9 -6 ability that the coefficient in equation 8 exceeds the value of 10 is 3x10 ,
Thus, the use of equation 8 is indeed very conservative.
g Stress Intensity Factors: In keeping with the conservative use of a failure criterion for complete circumferential cracks to represent the behavior of part-circumferential cracks, the stress intensity factor relation for complete circumferential cracks subjected to uniform stress will be employed here. This
- O relation (which is applicable for pipes with ID/H = 10) is available from refer- i ence 18 and is as follows 2
K 1"+- 2" 3" 4"
= (10) oa l (1 - n)I
!O i l
a = a/h C 3
= -6.21135 C = -1.00250 C = 1.79864 1 4 C = 4.79463 O 2 A-6
O Initial Crack. Distribution: The initial crack distr ibution consists of two components: (i) the probability of a crack existing at the weld location, and (ii) the size distribution of cracks given that a crack is present. Following I
o the approach of reference 5, cracks will be assumed to be Poisson distributed 3
with a crack existence frequency per unit volume of 10- /in . This value is denoted as p*. The probability of having a crack ir a weld of volume V, which is denoted as p*, is then given as 0 -V p*V p* = 1 - e (11)
The volume of weld, V, is taken to include a distance h on each side of the O weld, and is given by 2
V s n(ID) h (2h) = 2n (ID)h (12)
O The size distribution of cracks, given that a crack is present, is denoted as pcond.
The complementary cumulative conditional crack depth distribution will be assumed to be exponential with a parameter, A , of 0.246 inch. This value was used in the Marshall report (7) and was employed for the marginal O distribution of crack depths in reference 5. The crack size distribution must be adjusted to account for the impossibility of having a crack deeper than the wall thickness h. The following expression is obtained (5) f
- -x/A *-h/A 0>x>h O h/A (13)
P cond (o ' *) * < !O .
otherwise 1
(
g 1 = 0.246 inch This is considered to be the initial as-fabricated crack depth distribution.
This distribution should be conservative for the relatively thin wall pipes g under consideration, because it was estimated for reactor pressure vessels -
which are much thicker.
Detection Probability: The as-fabricated crack depth distribution is modified
- 9 by the detection probability of the pre-service inspection employed. The l
detection probability will be conservatively taken to be zero. This is l equilivalent to not considering the pre-service examination, and simplifies i
'O A-7
O the following analysis.
FRACTLi'E MECHANICS ANALYSIS Ai'D RELIABILITY RESULTS O
The input components necessary to perform the fracture mechanics analysis of piping reliability have now been presented. This section will present the procedures involved and the results obtained.
v Effect of_ Proof Testing: The proof testing performed on the piping can have a strong influence on the calculated reliability. This is because the fact that the piping has survived the proof means that no cracks larger than the v
critical size corresponding to the proof conditions (ap ) existed at the time of the proof test - otherwise the pipe would have failed during the test.
Hence, the crack size distribution can be truncated at a ,p which can have a marked ef fect on the calculated reliability. The following modification of J
equation 13 is then applicable
-ap/A
[e-x/A j
-e 0<x<a g Pcond ( >x) = 1-e -h/A p
[74) 0 otherwise
(
The value of a p for the piping systems considered is obtainable from g equation 7 with LC taken equal to oppf (ea. 5) which is provided in Table 2. Such values will be preented along with other relevant crack sizes discussed in the next section.
g Subcritical Crack Growth Calculations: The size distribution of cracks re-maining af ter the proof test, as given in equation 14, will change during operation of the plant due to the cyclic stresses imposed. Hence, the prob-ability of pipe failure will be time-dependent and equal to the probability g of a crack larger than the critical size existing at a given time. The critical crack ,ize is obtainable from equation 7 in conjunction with information from Table 2.
g An alternative vie vpoint is to consioer " tolerable initial crack sizes". For a given pipe weld and time, t, this is the crack size at t = 0 that would just grow to critical size in t. Denoting this as atol (t), the probability of failure within time t is then equal to the probability of having a crack larger O l A-8 )
1
l l
O '
than atol ( t) a t t = 0. Hence, once atol (t) is known (which is strictly a frat _ture mechanics calculation) the conditional cumulative failure probability is given by v
Pf (cond)(t) = P[a > atol(t)]
f -atol(t)/A -ap/A O = 1-e -h/
0 5atol(t) 5 a p I (15) 1 0 otherwise
(
O The value of atg)(t) at t = 0 is ac (by definition). The value of a p will be less than a , because the proof stress is higher than the load controlled stress c
during normal operation. Therefore, the failure probability will be zero for the time it would take a crack to grow from a p to a .
c O
The value of atol(t) con be calculated on a cycle-by-cycle basis by the following procedure.
O a tol(0) = a c a -
tol(I CYCI") " #c - !a=a c
c a =a c
C "tol( cycle) = atol( cyc e) - C[AKlatol(1 cycle)
! (16) etc.
O The appropriate values of C and m follow from equations 8 and 9. K for a l
l given c and a is obtained by use of equation 10 with ao obtainable from Table 2.
10 l A colservat.iva bourd on a tol aner n Tcycles is obtained by asnming that da/dn j
remains constant et the value corresponding to a .
c This provides the following expression q
l a (17) tol ("T) = a c ~"T C[aKla=ac ]
O A-9 l
l l
0 This simple estimating procedure provides a screening of the piping systems considered. If a g) (nT ) (for n T= total life) is greater than a , pthen failure of the piping system will not occur during nT cycles. This allows several of the piping systems to be immediately eliminated from consider-ation - as shown in Table 3, which also includes other relevant fracture mechanics infonnation. In cases where the piping system can not be immediately removed from consideration, the tolerable initial crack depth for n T = 200
,)
and 320 cycles can be calculated by the more involved procedure given in equation 16. The results of such calculations are also shown in Table 3 which reveals that most of the piping systems can be eliminated from consid-eration due to the influence of the proof test that was mentioned above. Only g
the scram discharge, header, and instrumentation lines remain to be consid-ered.
Failure Probabilities: The cumulative conditional probability of failure O
within nT cycles (plant lifetime) can be calculated from the atol ("T) results in Table 3 along with ap and equation 15. Table 4 summarizes the results of such calculations along with other closely related information. The conditional g- probability of failure of a given weld can be converted to a non-conditional value by multiplying by the probability of having a crack present in the weld (p*)
g Pf (t) = p* Pf(cond) (t) (18)
The probability of failure in the system of L welds is then obtained by con-servatively assuming that all welds in the systen have the same failure prob-O abilities but are independent of one another. The following expression pro-vides the cumulative system failure probability within time t Pf (sys)(t) = 1 - [1-Pf (t)] = 1 - [1 - p* P f(cond)(t)]L O
m p*L Pf (cond)(t) (19)
The average failure rate during the period t is obtained as O
f(sys)(t)/t (20) f O A-10
O The time t for the 200 (or 320) scrams is tnen to be estimated plant life-time of 40 years.
- Table 4 presents the average system failure rates as estimated by the above procedures. These can be expressed as a function of pipe diameter as follows 1109!!ze,jn. avgragg gyptem faj}g[g ratg, fr.N l
4.0 x 10 -4 3/4 (scram disch. &
side instr. line) 2 3.0 x 10 -7
, 6 7.0 x 10 -6 a
>6 0 The values from Table 4 for n = 320 were conservatively used in the above T
g tabulation. Combining the results for lines > 2" results in
-0 -1 pf (>_ 2") s 7 x 10 Yr D
A pipe is considered to have failed when a through-wall crack exists. Hence, both leaks and double guillotine breaks are included in the above results.
Estimates of leak rate probabilities can not be obtained from these results O unless some assumption is made regarding the failure mode - such as the con-servative assumption that all pipe failures are sudden and complete double guillotine failures. However, such an assumption may be overly conservative, expecially for the larger size lines. More complex and sophisticated analyses O based on the procedures presented in reference 5 could be employed to discrim-inate between leaks and sudden complete severances. However, such refinements are felt to not be warranted at this time.
O The above results are considered to be conservative for the following reasons:
quantities of weld conservatively estimated, especially for the scram discharge line, O - influence of in-service inspection ignored,
- influence of in-service proof tests ignored, only the pre-service , roof test was considered,
- stress intensity factors conservatively estimated assuming O
A-ll
w all cracks to be very long relative to depth 2
- initial crack depth distribution for much thicker material utilized, O - upper bound estimates on fatigue crack growth t
characteristics employed,
- conservative estimate of flow stress used,
- all welds in a piping system assumed to have stresses
'O equal to those for the highest stress joint, i.e.
the most adverse weld (see Reference 19).
The failure probability for the scram system piping was estimated in f1ED0-24342 "O
(Appendix B, reference 3) to be
-4 F7 (> 2") = 3.0 x 10 /Yn O Ti.ere is a 1 1/2 order of magnitude difference between this value, and the estimate derived above. The results from reference 3 were obtained by use of estimates from the reactor safety study (4) which were averaged out over many piping systems and plants. Hence, they do not reflect the proof testing O and inspection schedules used in the scram pf ping. Additionally, the stresses summarized in Table 2 are far below code all.cwables, which would tend to re-i duce the failure probabilities relative to lines designed more closely to code limits. Another factor is the relatively small number of stress cycles O imposed on the scram piping and the small time spent under load. Therefore, the values obtained above are felt to be reascnable and representative fo .he relatively low stressed limited length runs of piping employed in the scram system.
!O
SUMMARY
Ar4D C0tiCl.USI0 tis 4 A fracture mechanics analysis of scram piping reliability was performed in order to assess concerns regarding the integrity of such piping under operating i O
reactor conditions. The fracture mechanics analysis of tolerable crack sizes was tambined with estimates of the initial crack size distribution to obtain estimmtes of the probability of failure due to the subcritical and catastrophic growth of crack-like defects introduced during fabrication. This mode of fail-ure is believed to be the dominant one for the r; pes considered. Environmentally enhanced fatigue crack growth was considered to be the mode of subcritical crack i
- O A-12
.O growth. Stress carrosion cracking (SCC) was ruled out for the carbon steel lines because such a crack growth mechanism has not been observed in this r iaterial . Additionally, SCC was not considered for the 304 stainless steel O lines because of the small time they spend under stre. - The influence of a pre-service proof test was found to be large, and eliminated a nurber of the piping systems from consideration for failure. Fracture mechanics calculations of the remaining lines led to the following estimates on the most adverse weld O for the average failure rate for different size lines in the scram piping.
~4 3/4 inch pf = 4.0 x 10 /Yr
> 2 inches pg = 7.0 x 10 /Yr These values are believed to be conservative for reasons detailed in earlier sections. Comparison of the results for lines 2 inches in diameter and above with earlier esti. nates reveal the above value to be 1 1/2 orders of magnitude O
lower than the earlier values. This is felt to be reasonable because the values obtained herein reflect the beneficial influence of the proof testing employed in these pipes and their relatively benign stress history.
.O 4
- O O
O O
O A-13
O i
REFERENCES O
NUREG-0785, U.S. Nuclear Regule. tory Commission, Washington, D.C. ,
May 1981.
O 2. " Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping", NUREG-0803, U.S. Nuclear Regulatory Commission, Washington, D.C., August 1981.
- 3. "GE Evaluation in Response to NRC Request Regarding BWR Scram System Pipe Breaks", General Electric Company, Nuclear Power System Division, O Report NED0-24342, San Jose, California, April 1981.
- 4. Reactor Safety Study, An Assessment of the Accident Risks in U.S. Com-mercial Nuclear Power Plants, WASH-1400 (NUREC-75/014), U.S. Nuclear Regulatory Commission, Washington, D.C., October 1975.
O 5. D.0. Harris, E.Y. Lim and D.D. Dedhia, " Probability of Pipe Fracture in the Primary Coolant Loop of a PWR Flant - Vol. 5: Probabilistic Fracture hechanics Analysis", NUPEG/CR 2189, Vol. 5, U.S. Nuclear Regulatory Commission, Washington, D.C., August 1981.
- 6. 0.0. Harris, "The Influence of Inspection and Crack Growth Kinetics O on the Integrity of Sensitized BWR Piping Welds", EPRI NP-1163, Elec-tric Power Research Institute, Palo Alto, California, September 1979.
- 7. "An Assessment of the Integrity of PWR Pressure Vessels", Report by j a Study Group Chaired by W. Marshall, available from H.M. Stationary Office, London, England, October 1976.
- 8. D.O. Harris and E.Y. Lim, " Applications of a Fracture Mechanics Model of Structural Reliability to the Effects of Seismic Events on Reactor Piping", submitted for publication in Progress _in Nuclear Energy.
- 9. D.0. Harris, "A Means of Assessing the Effects of Periodic Proof Testing O and KDE on the Reliability of Cyclically Loaded Structures", Journal _of Pressure Vesse_1_ Technology, Vol. 100, No. 7, (May 1978), pp. 150 - 157.
- 10. " Piping Stress Analysis, La Salle County - Unit 1 4266-10 Control Rod Drive RD-05"., Sargent and Lundy, Chicago, Illinois, Accession No. EMD-032611, August 1981.
0
- 11. "La Salle Unit 1 Piping Reanalysis of the Control Rod Drive Hydraulic System", prepared by Cygna Energy Services, San Francisco, California, Report B-R-LA-80010-1, November 1981.
- 12. " Scram Header", no drawing number, Reactor Controls, Inc. , San Jose, O California, 1978.
O' A-14
O
- 13. " Scram Volume Level Station", Drawing No. LA- J31, Reactor Controls, Inc. ,
San Jose, California, July 1981.
- 14. ' tetter from R.L. Gridley, GE - San Jose to J.H. Hannon, NRC, No. MFN-C) 126-81, dated June 30, 1981.
- 15. H.H. Klepfer, et al., " Investigation of Cause of Cracking in Austenitic Stainless Steel Weldments", Report NE00-21000, General Electric Company, Nuclear Energy Division, San Jose, California , July 1975.
C) 16. D.A. Hale, J. Yuen and T. Gerber, " Fatigue Crack Growth in Piping and RPV Steels in Simulated BWR Environment", General Electric Company, Nuclear Energy Engineering Division, Report GEAP-24098, NRC-5, San Jose, California , January 1978.
- 17. M.L. Vanderglass and B. Mukherjie, " Fatigue Threshold Stress Intensity C) and Life Estimation of ASTM-A106B Piping Steel", American Society of Mechanical Engineers Paper 79-PVP-86, New York, New York, 1979.
- 18. D.0. Harris and E.Y. Lim, " Stress Intensity Factors for Complete Cir-cumferential Interior Surface Cracks in Hollow Cylinders", Fracture
,s Mechanics: Thirteenth Conferenc_e, American Society for Testing and
'> Materials Special Technical Publication No. 743, Philadelphia, Penn-sylvania, 1981, pp. 375 - 386.
- 19. This analysis very conservatively assumes that a weak point failure ioses the entire SDV, but a 3/4" line break will not do this.
O
- C)
O O
O l
O A-15
TABLE 1
SUMMARY
OF PIPING SYSTEMS CONSIDERED Line Nom. ID, Wall Number of Welds Source, Sched Thickness, Matl. Ref.
Name Diam in in N(90 ) S(270 ) E Scram Discharge 3/4 80 0.742 0.154 8 x 185 8 x 185 -- 304SS. 1, p. 16
?
Header 8 80 7.625 0.500 37 40 --
SA106 12 Ir.strument Volume 12 80 11.376 0.687 2 2 --
SA106 13 Header Vent 1 80 0.957 U.179 70 52 70 SA106 11 Drain 2 160 1.689 0.343 7 3? --
SA106 10 19239umgglapigg(jggy Top 2 160 1.689 0.343 11 11 --
SA106 13 Side 3/4 160 0.614 0.218 72 72 --
SA106 13 Sensors 1/2 160 0.466 0.187 25 25 --
SA106 13
O O O O O O O O O O O TABLE 2
SUMMARY
OF STRESSES IN PIPING SYSTEMS (all stresses in ksi) i i Line Su stained Source oTE o
prf t> o Name p LC + TE Scram Discharge 1.51 5.24(1} 28.88 11, Table 8-2 ---
5.62 25.15
> lleader 4.77 6.33 k1) 29.30 11, Table 9 ---
7.52 27.74 Instrument Volume 5.18 15.18(2) --- ---
7.51(5) 6.48 12.69 Header Vent 1.67 6.28 18.58 11, Table 10 ---
6.70 13.97 Drain 1.54 s3.0 (3) --- ---
7.14(6) 3.39 8.68 i
Igsggegtatjon (jggs Top 1,54 5.30 22.98(4) 11, Table 10 ---
5.69 19.22 Side ! 0.92 5.30 54) 22.98(4) --- ---
5.53 18.68 Sensors 0.78 0.78(2) s0.78 --- ---
low low (1) Stress due to sustained load.
s0 (2) Assumeda$ds1.5ksi.
(3) o assum (4) A@dumed same as " top" .
(5) Calculated by assuming bending moment same as for header.
(6) Calculated from bending moment due to deadweight + thermal expansion 1 from fleference 10.
O O O O O O O O O O O TABLE 3
SUMMARY
OF CRACK SIZES B eliminate a i ,ine a , in AKla=a c ll a
tol (nT) from con- [y) tol (nT) a , in c p siderat ion Name ksi-ini n =200 T
320 200 320 200 320 0.1361 0.1348 45.4 <0 <0 No No 0.0643 0.0558
, Scram Discharge 1
0.4297 0.4167 83.3 0.4177 0.4105 Yes No 0.4131
?
$ 0.5881 48.6 0.6046 0.6026 Yes Yes Instrument Volume 0.0079 ---- ----
Header Vent 0.1540 0.1523 25.2 0.1533 0.1529 Yes Yes ---- ----
Drain 0.3201 0.3172 24.6 0.3194 0.3190 l Yes Yes ---- ----
i Instrumeg}apigg(jges Top 0.3026 0.2996 51.5 0.2988 0.2965 No No (2) (2)
Side 0.1923 0.1912 39.7 0.1902 0.1890 No No (2) (2)
Sensor line eliminated due to very low stresses (see Table 2)
LB (1) Eliminate from considerstica if a (n > a l
(2) Conservatively taken equal to lo b bodn)d va10e. s a t 1 ("T)
- TABLE 4
SUMMARY
OF RESULTS FOR FAILURE PROBABILITIES System P (n )( }
Line Voi me p* ds Pf (cond)("T)( ) f T f ' Y#
Name (eg.12) (eq. 11) (Tabic 1) n =200 T
320 200 l 320 200 320 ,
in
-5 -4 Scram Discharge 0.1106 1.11x10 2960 0.412 0.471 0.0135 0.0154 3.36x10-4 3.84x10
-3 -4 -6 Header 11.98 1.20x10- 77 ---
3.12x10 ---
2.88x10 ---
7.21x10
-4 -3 -3 -6 -5 -8
$ Top Instr. Line 1.248 1.25x10 22 1.28x10 4.99x10 3.52x10 1.37x10 8.80x10 3.43x10-
-5 -3 9.28x10
-6 2.04x10
-5 2.32x10-7 5.12x10
-7 Side Inst.'. Line 0.1833 1.83x10 144 3.52x10-3 7.76x10 (1) For single joint, from equation 15.
(2) From equation 19.
O O O O O O O ..O. ____ O O O stress levels and cycle initiil cra:k post-inspection size cistribution crack size distributica subcritical y failure crao; growth criterion
< characteristics "a 7A u g lI C"
" k ,
a q critical crack g size i
crack tolerable c detection crack sizes probability
> , y v <
b ! 2
% 8 -
U
- 3 stress intensity fj failure = factor solution L a; probability
"" C , __ lime a "
- S 8 3* 2 g ,- -
n time a/h Figure 1: Schematic Representation of Various Components of Analysis of Probability of Failure of a Single Weld Joint.
1 O
l l
l O
i 1
[
1 O 1 w
O mn .
r-O 2b q i R.
1
'O 1 O
l O .
I Figure 2 Geometry of Part-Circumferential Internal Surface '
Crack Considered in this investigation.
1 10 l 1 I
i l
l 10
- O A-21