ML19326E110
| ML19326E110 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 07/10/1980 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Finfrock I JERSEY CENTRAL POWER & LIGHT CO. |
| References | |
| TASK-03-05.B, TASK-RR NUDOCS 8007250720 | |
| Download: ML19326E110 (30) | |
Text
e wa l
&p>
r
/- [
[he / [k, UNITED STATES g
NUCLEAR REGULATORY COMMISSION l
r,,.; p&gNn j WASHINGTON, D. C. 20555
,y+....J July 10, 1980 Docket No. 50-219 Mr. I. R. Finfrock, Jr.
Vice President - Generation Jersey Central Power & Light Cogany Madison Avenue at Punch Bowl Road Morristown, New Jersey 07960
Dear Mr. Finfrock:
RE: SEP TOPIC III-5.B Pipe Break Outside Containment -
(0yster Creek Nuclear Generating Station)
Enclosed is a copy of our current evaluation of Systematic Evaluation Program Topic III-5.B, Pipe Break Outside Containment (Enclosure 1). This assessment compares your facility, as described in Docket No. 50-219 with the criteria currently used by the reguiatory staff for licensing new facilities. Please inform us if your as-built facility differs from the licensing basis assumed in our assessment within 60 days of receipt of this letter.
This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your facility. This topic assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is completed.
We are also enclosing a request for additional information (Enclosure 2) to enable us to co@lete the review of issues identified in the conclusions section of the above referenced evaluation. Please submit your response to this request within 60 days of receipt of this leti.er.
There are also two staff positions enclosed (Enclosure 3) that requests that you submit schedules by September 1,1980 for the co@letion of certain modifi-cations to your facility.
Sin rely, Dennis i.
rutchfield, Chie Operating Reactors Branch #o Division of Licensing
Enclosure:
1.
Completed SEP Topic III-5.B.
2.
Request for Additional Information 3.
Staff Positions cc w/ enclosure:
See next page 8007250920 k
[
Mr. I. R. Finf rock, J r. July 10, 1980 cc w/ enclosures:
G. F. Trowbridge, Esquire Gene Fisher Shaw, Pi':tman, Potts and Trowbridge Bureau Chief 1800 M Street, N. W.
Bureau of Radiation Protection Washington, D. C.
20036 380 Scotts Road Trenton, New Jersey 08528 GPU Service Corporation ATTN: Mr. E. G. Wallace Mark L. First Licensing Manager Deputy Attorney General 260 Cherry Hill Road State of New Jersey Parsiopany, New Jersey 07054 Department of Law and Public Safety Environmental Protection Section Anthony Z. Roisman 36 West State Street Natural Resources Defense Council Trenton, New Jersey 08525 917 15th Street, N. W.
Washington, D. C.
20006 Joseph T. Carroll, Jr.
Plant Superintendent Oyster Creek Nuclear Generating Steven P. Russo, Esquire Station 248 Washington Street P. O. Box 388 P. O. Box 1060 Forked River, New Jersey 08731 Toms River, New Jersey 08753 Joseph W. Ferraro, Jr., Esquire Director, Technical Assessment Deputy Attorney General Division State of New Jersey Office cf Radiation Programs Department of Law and Public Safety (AW-459) 1100 Raymond Boulevard U. S. Environmental Protection Newark, New Jersey 07012 Agency Crystal Mall #2 Ocean County Library Arlington, Virginia 20460 Brick Township Branch 401 Chambers Bridge Road U. S. Environmental Protection Brick Town, New Jersey 08723 Agency Region II Office Mayor ATTN: EIS COORDINATOR Lacey Township 26 Federal Plaza P. O. Box 475 New York, New York 10007 Forked River, New Jersey 08731 Richard E. Schaffstall Comissioner VJiC, Incorporated Department of Public Utilities 1747 Pennsylvania Avenue, N. W.
State of New Jersey Washington, D. C.
20006 101 Commerce Street Newark, New Jersey 07102 e
ENCLOSURE 1 SEP EVALUATION OF PIPE BREAK OUTSIDE CONTAINMENT TOPIC III-5.B FOR THE OYSTER CREEK NUCLEAR POWER PLANT
TABLE OF CONTENTS Page
1.0 INTRODUCTION
1 2.0 B A C KG R O U N D...........................................................
2 3.0 EVALUATION...........................................................
5 3.1 Onergency Condensers............................................
6 3.2 Re acto r Wa te r Cl e a nu p Sys t em....................................
9 3.3 Co ntrol Rod Driv e Hyd raul ic Sys tem..............................
10 3.4 Ma in S te am and Ma in Fe ed Sys tems................................
11
4.0 CONCLUSION
S..........................................................
13 TABLE 1.
Effects of Pipe Break Outside Conta inment.......................
16 R E F E R E NC E S................................................................
24 4
e e
1
1.0 INTRODUCTION
The safety objective of Systematic Evaluation Program (SEP) Topic III-5.8,
" Pipe Break Outside Containment" is to assure that pipe breaks would not cause the loss of needed functions of " safety-related" systems, structures and com-ponents and to assure that the plant can be safely shut down in the event of such breaks. The needed functions of " safety-related" systems are thoss functions required to mitigate the effects of the pipe break and safely shutdown the reactor plant. The current criteria for review of pipe breaks outside contain-ment are contained in Standard Review Plan 3.6.1 and 3.6.2 including their attached Branch Technical Positions.
2
2.0 BACKGROUND
In December 1972, the staff sent letters (Reference 1) to all power reactor licensees requesting an analysis of the effects of postulated failures of high energy lines outside of containment. A summary of the criteria and requirements in this letter is set forth below:
a.
Protection of equipment and structures necessary to shut down the reactor and maintain it in a safe shutdown condition, assuming a concurrent and unrelated single active failure of protected equipment, should be provided i.'om all effects resulting from ruptures in pipes carrying high energy fluid, where the temperature and pressure conditions o' the fluid exceed 200'F and 275 psig, respectively, up to and including a double-ended rupture of such pipes. Breaks thould be assumed to occur in those loca-tions specified in the " pipe whip criteria." The rupture effects to be considered include pipe whip, structural (including the effects of jet impinge:nent), and environmental.
b.
In additfor, protection of equipment and structures necessary to shut down the reactor and maintain it in a safe shutdown condition, assuming a concurrent and unrelated single active failure of protected equi should be provided from the environmental and structural effects (pment, includ-1 ing the effects of jet impingement) resulting from a single open crack at the most adverse location in pipes carrying fluid routed in the l
vicinity of this equipment. The size of the cracks should be assumed to be 1/2 the pipe diameter in length and 1/2 the wall thickness in width.
In response to cur letter, Jersey Central Power and Light Company (JCP&L), the licensee) submitted Amendment 75 to the Facility Description and Safety Analysis Report (FDSAR) dated July 1,1974 (Ref. 2) and Revisions 1 through 4 to Amend-ment 75 dated December 24,1974 (Ref. 3), March 24,1975 (Ref. 4), April 25, 1975 (Ref. 5), and June 1,1976 (Ref. 6). Additional infonnation was provided in an analysis of jet impingement loads on the torus dated September 21, 1976 (Ref.7). The staff's review of these documents is contained in a Safety Evaluation Report dated December 2),1976 (Ref. 8).
3 Additional information regarding moderate energy line breaks (MicB) has been provided by the licensee in Supplement 6 (Addendum 1) to the ApJ1ication for a Full Term Operating License for Oyster Creek (Ref. 9) and by letters dated March 13, 1974 (Ref. 10) and July 3, 1974 (Ref. 11).
The licensee has proposed an additional analysis concerning tne acceptability of pipe breaks outside containment associated with the isolation condenser steam and condensate lines. The analysis addresses the same concerns identified in this evaluation.
The NRC staff reevaluation of the effects cl pipe breaks outside containment under SEP Topic III-5.8 includes the comparison of Oyster Creek with current criteria for pipe breaks outside containment.
The staff used an " effects oriented" approach to determine the acceptability of plant response to pipe breaks, i.e.,
each structure, system, component, and power supply which must function to mitigate the effects of the pipe break and to safely shut down the plant was examined to determine its susceptibility to the effects of the postulated break.
Break effects considered were compartment pressurization, pipe whip, jet impinge-ment, spray, flooding, and environmental conditions of temperature, pressure, and humidity. This review complements that of SEP Topic III-12, " Environmental Qualification of Safety-Related Equipment."
(The effects of potential missiles generated by fluid system ruptures and rotating machinery were also considered and are evaluated under SEP Topic III-4.C, " Inter-nally Generated Missiles.")
4 The previous evaluation of pipe breaks outside containment for Oyster Creek was performed using some methods and criteria which are no longer used by the staff in the review of current plants.
For example / the current definition of a high energy fluid system as one that is maintained under conditions where either or both the maximum operating temperature and pressure exceeds 200*F and 275 psig is different from the definition applied in the previous review where a high energy fluid system was one in which both temperature and pressure exceeds 200 F and 275 psig. The SEP reevaluation of this topic was performed using the current criteria in Standard Review Plans 3.6.1 and 3.6.2 and their attached Branch Technical Positions.
Data for this assessment was gathered during a visit to the Oyster Creek plant on January 15-17, 1980.
l I
i
5 3.0 EVALUATION The results of the SEP reevaluation of pipe breaks outside containment for Oyster Creek are provided in Table 1.
The following paragraphs provide additional information used to evaluate certain pipe breaks listed in Table 1.
The safe shutdown systems which were examined from the standpoint of protection from pipe break effects are identified in the SEP Safe Shutdown Review for Oyster Creek (Reference 12). These systems are:
(a) Reactor Control ar,d Protection System, (b) Emergency Condensers, (c) Condensate Transfer System, (d) Automatic Depressurization System, (e) core Spray System, (f) Emergency Service Water System, (g) Instrumentation for Shutdown and Cooldown, (h) Emergency Power (AC and DC) and control power for the above systems and components.
6 3.1 Emergency Condensers The two emergency condensers are located in the reactor building, 95' elevation, east side. The steam supply and condensate return lines for the condensers are routed from containment penetrations on the 75' elevation to the condensers and back. These lines are.naintained at reactor system pressure because the steam supply valves, all of which are outside containment, are open during normal plant operations. The condensate return valve inside containment is also normally open, while the condensate return valve outside containment is shut.
Each condenser's steam and condensate valves close (in approx. 55 sec.)* spon receipt of a high flow signal from sensors in its own steam supply and/or condensate return lines.
The shell side of the emergency condensers is supplied by the condensate transfer system through air operated fill v e ~,*-;hi;h are controlled by the control room operator.
Emergency condenser high energy line breaks on the 95' level of the reactor building were analyzed in detail in Reference 4.
Based on Reference 4, the licensee concluded that there were no pipe break locations in either emergency condenser which interact with the emergency condenser of the redundant system, cable tray 45, or conduit containing or supplying safety related equipment.
Based on our reevaluation of HELBs on the 95' elevation, taking into account Reference 4, we have determined that interactions are possible between the two emergency condenser systems.
These interactions are:
"The 55 second valve closure time consists of 20 seconds for the valve to shut following a 35 second time delay after the shut signal.
7 1.
Jet impingement on cable tray 45 from a longitudinal break in the A cond.enser steam line.
(Tray 45 contains level and control cables for both emergency condensers.)
2.
Pipe whip damage to conde
containing level indication and control signals for the B condenser from a break in the A condenser steam line.
(The remaining potential targets of a HELB, the emergency condenser shells, condensate fill lines and fill valve air supply, are adequately protected by the geometry of piping layout and shielding provided by structures and equipment.)
The two potential interactions above could result in (1) the immediate loss of function of the condenser system suffering the break and (2) the eventual loss, in approximately 40 minutes, of the other condenser when its shell side water is boiled away by core decay heat. The steam line break would be isolated automa-tirally by the high flow sensors. A reactor trip would occur b h ause of either hig i power, low reactor water level, or main steam isolation valve closure on low ; team pressure caused by the steam line break.
In accordance with current criteria, a reactor trip causes an assumed loss of offsite power; therefore, the main reactor feed system and main condenser are l
After the second c.mergency condenser has boiled dry, reactor system pressure would increase to the safety / relief valve setpoints. Pressure would be limited by the relief valves, but reactor system coolant inventory would continue to
8 be lost through the relief valves. To put the plant in a safe condition, the operator must manually initiate the Automatic Depressurization System (ADS) and ensure that at least one train of the Core Spray (CS) System is operating.
Adequate long-term core cooling is accomplished, even assuming the single failure of one of the two emergency diesel generators, with the ADS, CS, cuntainment spray, and emergency service water systems (Ref. 12). These actions are included in the plant emergency procedures.
The availability of these emergency systems to provide safe shutdown capability and sufficient time for operator action to initiate these sytems, even with the loss of both emergency condensers, provide adequate mitigation of the effects of these postulated HELBs.
Emergency condenser HELBs on the 75' elevation of the reactor building coula result in damage to the emergency condenser isolation valves and controls, and cable trays V22, VA? 41, 42, and 43.
The motor operators are susceptible to jet impingement damage from both steam and condensate line cracks and breaks and from pipe whip of the A condenser condensate lines.
The conduit containing isolation valve control and power cable are susceptible to pipe whip and impinge-ment effects from both steam and condensate line breaks. The steam supply line motor operated isolation valves for both emergency condensers are outside contain-ment and are normally open. The steam line isolation valves could be prevented from automatically closing by.Pe effeci of a break in the isolation condenser piping. Considering the W.ngle.
lure issumption, only one valve needs to be damaged by the break effects to
. "l'
,.1 an unisolable break.
l l
i
9 Cable trays V22, V23, 41, 42 and 43 carry electrical cables for one train of the core spray (CS) system, standby liquid control system (SLCS) and the emergency condenser system. As indicated previously in the discussion of HELBs on the 95' elevation, the CS is needed to cope with a loss of both emergency condensers.
The dmaage to cable trays could prevent the opening of the emergency condenser condensate return valves which must open to initiate emergency condenser operation.
In this case, the operator would not have the 40 minutes emergency condenser boil-dry time to initiate ADS and CS for core cooling.
Recall that a loss of offsite power has been assumed because of a reactor trip.
In addition the postulated break has damaged one CS train.
Based on this discussion, we have determined that there is inadequate protection from the effects of postulated emergency condenser line breaks on the 75' eleva-tion at Oyster Creek. ner position on this is provided in the CONCLUSIONS l
section of this repcrt.
l 3.2 Reactor Water Cleanup System The Reactor Water Cleanup (RWCU) system high energy piping is located on the 51' elevation of the reactor building, south side. The isolation vaives for the system include (1) inside the contrinment drywell:
a check valve on the RWCU system return line and a 40V on the letdown line, and (2) outside containment:
a MOV on the return line and a MOV on each leg of the letdown line whic., branches into two lines just outside of the containment penetration. The ir lation MOVs automatically shut on a containment isolation signal on low-low reactor vessel level indication.
10 HELBs in the RWCU system could affect the motor operated isolation valve elec-trical power or the operator itself. This damage in combination with a single failure of the isolation valle inside containment could result in an unisolable break path from the reactor system.
This is addressed further in the CONCLUSIONS section of this report. The HELB could also damage cables in trays 13A and 14A. Damage to cable tray 14A was assumed and analyzed in Reference 2, however, potential damage and effects of damage to cable tray 13A were not addressed. Additional information regarding the cables in tray 13A is required to assess the effects of this damage. The licensee will be requested to supply this information.
3.3 Control Rod Drive Hydraulic System Pipe breaks associated with the CRD hydraulic centrol units, on the 23' elevation of the Reactor Building, could involve (1) the drive insert and withdraw lines which lead through containment penetrations to the drives, (2) the CRD hydraulic drive and cooling water lines, (3) the CRD charging water line, and (4) the CRD return line. A break in the drive withdraw line would cause its associated control rod to insert (scram). A break in any of the other lines would cause the rod to remain in position but the rod could still be inserted by the operator or by a reactor protection system scram signal.
Loss of electrical power to the CRD hydraulic control unit would also result in a rod insertion.
Therefore, potential pipe break damage to the CRD hydraulic control units would not prevent control rod trip (scram) by the operator or the reactor protection system.
11
- 3. 4 Main Steam and Main Feed Systems The effects of main steam (MS) and main feed (MF) HELBs in the Turbine Building Mezzanine area on the control room, cable spreading room, main steam isolation valves (MSIV), main feedwater piping and isolation valves, and cable trays 12, 13, 14, 15, 30, 31, and 32 were evaluated in Reference 2.
Information regarding the h5 line break detection system, MS and MF isolation valve supports and jet impingement effects on the torus is provided in Reference 3.
The SEP reevaluation of this area of the turbine building has determined that interactions between postulated MS and MF HELBs and one train of emergency service water system (ESWS) piping (loop II) is possible.
The interactions could result in the loss of function of this loop.
Since MS and MF breaks result in turbine and reactor trips with concurrent assumed loss of offsite power, the single failure of diesel generator #1 would result in the loss of function of ESWS loop I.
Thus, the HELB could result in total loss of ESWS function.
The ESWS is used, as described in Reference 12, for long term cooling of the reactor by cooling the containment spray system which cools the torus water which is circulated through the reactor by the core spray and automatic depressuri-zation systems.
For the above described scenario in which the ESWS system is lost, the shutdown cooling system is available for long term core cooling after reactor system l
6
12 tempera!.ure is reduced by the emergency condensers and/or the automatic depressuri-zation system. Therefore, the ESWS is not essential for safe shutdown following a MS or MF line break which disables ESWS in the turbine building mezzanine area; and the plant is adequately protected from these potential breaks.
As described in Ref. 2, the MS and MF breaks in the mezzanine area can also damage electrical control cables used fer the control of core spray (CS), contain-ment spray, ESWS, diesel generators, MS line break detection, automatic depressuri-zation system (ADS) and control rod drive (CRD) hydraulic pumps.
Damage to these cables would not prevent the functioning of the diesel generators, MS break detection, ADS, and CRD hydraulic pump. However, with the assumed loss of offsite power and a single failure, damage to even one train of the CS, containment spray, and ESW control systems could result in the complete loss of these system functions. Again, the shutdown cooling system could be used for long term core cooling and so the containment spray and ESW functions are not essential.
Loss of the CS system function, however, would severely restrict the ability of the plant operator to keep the reactor core covered with coolant during plant recovery l
from the postulated bieak.
In Ref. 2, the licensee stated that the CRD hydraulic system was required to cope with the postulated breaks. The implied requirement of the CR0 hydraulic pumps is to maintain reactor vessel coolant level during the plant cooldown following initiation of the emergency condensers. However, the CR0 hydraulic system was not designed as a safety system, and no credit is given for its capability to inject water into the reactor coolant system. This evaluation depends on the availability of the CS system for reactor system makeup during cooldown.
13
4.0 CONCLUSION
S Based on the information submitted by the licensee and obtained during our site visit to Oyster Creek, we have determined that the following review areas have not been addressed adequately in previous staff safety evaluations and shcild be resolved with the SEP:
1.
Inadequate protection exists for postulated HELBs in the emergency condenser steam and condensate lines on the 75' elevation of the reactor building.
The licensee is currently preparing a report to addre,ss HELBs in this area of the plant. The NRC staff position is that the licensee should submit a schedule by September 1,1980, for modifications to be effected in this area of the plant to provide adequate protection from the effects of these postulated HELBs.
The modifications to be installed must be in accordance with the acceptance criteria of Standard Review Plan 3.6.1 and provide protection for the emergency condenser isolation valves and controls and cable trays V22, V23, 41, 42, and 43.
Justifi-catic. nr continued operation of the facility while the modifications are developed and implemented is based on the extremely low probability that (1) the HELB will occur in the time required to effect the modifi-cations and (2) the postulated HELB would have the proper orientation to cause the worst case damage described above.
2.
Postulated pipe breaks outside of the primary containment between the containment penetration and the first containment isolation valve have not been evaluated for the main steam lines, emergency condenser steam and condensate lines, and reactor water cleanup suction and discharge lines.
14 Currently the staff applies the provisions of Branch Technical Position MEB 3-1 (Reference 10) section B.l.b., to the review of the postulated break areas. The licensee will be required to compare the design of the Oyster Creek plarit systems with these current regulatory provisions.
3.
The effects of postulated pipe breaks in certain systems could result in damage to the containment isolation valves or power supply and control cables to the containment isolation valves for those systems. The combi-nation of the single active failure provision and damage to the containment isolation valve could result in an unisolable break flow path. The systems of concern are the emergency condenser system (steam lines only), and reactor water cleanup system letdown and return lines.
The staff currently applies the provisions of Branch Technical Position ASB 3-1 (Reference 11), Section B.2.c., to the review of these break areas.
The licensee will be required to compare the design of the Oyster Creek systems with these current regulatory provisions.
4.
The postulated break of certain high energy reactor water cleanup lines could damage cable tray 13A on the 51' elevation of the reactor building.
The effects of damage to this tray have not been previously evaluated.
The licensee will be required to provide an analysis of the effects of HELB damage to this cable tray.
1 5.
A MS or MF liELB in the Turbine Building Mezzanine area could damage control cables for the CS system (and for other systems as previously described).
The CS is needed to provide makeup water to the reactor system during a plant cooldown following the postulated HELB. The licensee will be
4 15 required to move or protect all CS control cables from the effects of these potential breaks.
6.
A MELB in the cable spreading room could flood the room to some level before the floor drains could accommodate the flow. The TBCCW system leakage flow rate of 118 gpm is the largest potential MELB for this room. The licensee will be required to determine the depth of the flood-ing and what equipment would be affected by the flooding.
The staff is continuing this reevaluation of pipe breaks outside containment and will update this report as additional information is provided and conclusions are reached.
TABLE 1.
EFFECTS OF PIPE BREAK OUTSIDE CONTAINMENT Affected Mitigating Affected Safe Adequacy of Zone Pipe Break System Shutdown System Protection Remarks Intake Structure SWS, Screen None ESWS Adequate. Spray from MELB systems Wash System, would not affect the ESWS pumps which New Radwaste are designed for outdoor use. The SWS, CW (MELB)*
open-air intake structure precludes flooding of ESWS. However, if the structure is enclosed in the future, flood warning and protection for the ESWS must be considered.
Condensate Trans-Fire system None Condensate Trans-Adequate. A fire system MELB in fer Pump Area (MELB) fer System the cond. transfer pump enclosure (267 gpe) could result in flooding the pumps.
Loss of the transfer pumps would not cause any plant transients or LOP. The pumps are designed for outdoor use.
Loss of ether pump would result in a " pump" tripped" alarm in the control room to warn the operator of the flooding condition.
Reactor Build.
Fire System None None Adequate. Hatches and floor
'(119')
(MELB) drains are adequate to remove this leakage. Fire system MELB enve-lopes other MEL8s ire this zone, e.g., demin. water.
- See last page of Table 1 for list of abbreviations.
16
TABLE 1.
(Continued)
Affected Mitigating Affected Safe Adequacy of Zone Pipe Break System Shutdown System Protection Remarks Reactor Build.
Emergency Con-None Emergency Con-Adequate. Potential targets of (95')
denser Steam densers HELB effects are the Energ. Cond.
Line (HELB) shells, condensate supply lines, fill valve air supply line, level instruments, and cable tray 45.
These interactions are discussed in the EVALUATION section.
Fire System None None Adequate. Hatches and floor CfELB) drains are adequate to remove leakage from a fire system HELB which envelopes other HELBs on this elevation.
None None Adequate. A SLCS HELB outside containment would result in the containment isolation check valve inside containment seating with reactor system pressure. This would isolate the flow path from the reactor recirculation system to the p'pe break.
Reactor Build.
None None Adequate. See above remarks.
(75')
Emergency Con-Emergency Conden-Emergency Con-Inadequate. Emerg. cond. steam denser Steam ser Isolation densers or condensate HELBs may result in
& Condensate Valves damage to the emerg, cond. con-Lines (HELB) tainment isolation valves, and these HELBs could damage elec-
'.-ical cables in trays V22, V23,
'1, 42 and 43.
These breaks are discussed in the EVALUATION section.
17 j
M TABLE 1.
(Continued)
Affected Mitigating Affected Safe Adequacy of Zone Pipe Break System Shutdown System Protection Remarks Reactor Build.
RBCCW, 2400, None None Adequate. RBCCW MELB envelopes (75' cont.)
Fire System all other MELB's on 75' elev.
(MELB)
(approx. 200 gpm). Sufficient drainage via hatches and floor drains exists to prevent flood-ing.
Reactor Build.
RWCU Isolation Cable Trays 13A, Potentially inadequate. HELB (51')
Valves 14A nay damage the RWCU system isolation valve motor operator or electrical power. This could prevent operation of the normally open valves. Effects of break in RWCU system could damage elec-trical cables in trays 13A and 14A. Trays 13A and 14A carry cables for the CS system, the ADS system, and the RPS. These effects are discussed in the EVALUAfl0N section.
SWS, RBCCW, None None Adequate. Adequate drainage Fire System exists via hatches, stairs, etc.
(MELB) to prevent flooding of equipment by the largest MELB on this elevation (a 20" SWS line break of appro::imately 550 gpm).
18
J 9
TABLE 1.
(Continued)
Affected Mitigating Affected Safe Mequacy of Zone Pipe Break System Shutdown System Protection Remarks Reactor Build.
CRD Hydraulic None Control Rods Mequate. Ruptures of high (23')
Control Units (RPS) energy portions of CRD control (HELB) units or damage to units resulting from pipe dip would result in either a tripped rod (scram) or loss of CRD supply to the affected control rod. In the latter case, the control rod could still be scrammed manually or automatically by the RPS.
Mequate. See renarks above for and Cable Trays CRD modules. The effects of CRD 15, 16, 17, 18 HELBs on this elevation were 19, 20, 21, 22, previously analyzed in Reference 2.
and 23.
These effects are discussed further in the EVALUATION section.
Fire System, SWS, None None Mequate. Floor drains and (MELB) hatches provide adequate drainage to prevent equipment flooding.
Reactor Build.
CRD Supply Line None Torus Mequate. Although, the torus may
(-19' )
(HELB) be damaged by(and assumed con-pipe whip, no reactor trip current LOP) would result; and the plant cuuld be shut down in an orderly manner. Damage would be restricted to the upper portions of the torus and none of the torus water volume would be lost.
19
TABLE 1.
(Continued)
Affected Mitigating Affected Safe Mequacy of Zone Pipe Break System Shutdown System Protection Remarks Reactor Build.
Torus (MELB)
None Torus Adequate. Flooding from the torus
(-19' cont.)
would not affect any safe shutdown equipment other than the torus itself. The reactor build. corner rooms are separated from the torus area by watertight doors. The torus area is designe<l to contain the leaked water volume of the torus without loss of the ability of the torus to function as a shut-down heat removal system.
ESWS, RBEDT, None CS, Containment Adequate. Flooding of individual MELB's from Spray reactor build. corner rooms could 1evels above disable pumps in that room:
-19' elev.
Room Pumps NW CS - B D NE CS - A, C SW Contain spray - 3, 4 SE Contain. spray - 1, 2 Loss of a corner room would not i
result in a reactor trip or pisnt transient event, and redundant pumps are available in other corner rooms. SE and NE rooms have sump levels alamed in control room.
RBEDT is in NW corner room and has high and low level alanns in the l
control room. Flooding conditions in the torus area are indicated in the control room by high sump level alarms.
)
20
TABLE l_.
(Continued)
Affected Mitigating Affected Safe Adequacy of Zone Pipe Break System Shutdown System Protection Remarks Turbine Build.
MS, MF MSIV, MS Break Control Room, Cable Potentially inadequate. Potential (Mezannine,23')
(HELB)
Detection System Spreading Room, targets of HEl.B effects are the Torus, Cable Trays control room and cable spreading 30, 31, 32, 12, 13, room structures, torus shell, ESW 14, 15, and ESW piping (one loop) and cable trays
- line, which contain control and instru-ment cables for CS, contain. spray, ESW, RPS, and energency diesel systems. These interactions are discussed in the EVALUATION section.
Fire System, None 4160 V Switchgear
. Adequate. Spray from MELB could Demin. Water Panels IC, ID, impact Bettery C switchgear and System (MELB)
Battery C Switch-both 4160 V switchgear panels 10, gear
- 10. Spray would not enter the DC switchgear panel; and an enclosure is being constructed around the 4160 V switchgear (as a result of the fire protection review) and will prevent spray from HELB's from impacting the switchgear.
Cable Spread-TBCCW, Fire None AC and DC Emer-Potentially inadequate. A TBCCW ing Room System, Demin.
gency Power, RPS HELB 118 gpm could flood the room Water (MELB) to a depth of (to be determined).
The effects of flooding on redun-dant emergency power and RPS motor generator sets must be determined.
21
TABLE 1.
(Continued)
Affected Mitigating Affected Safe Adequacy of Zone Pipe Break System Shutdown System Protection Remarks Turbine Build.
CW, ESW, None None Adequate. CW MELB have been (Basement)
TBCCW (MELB) previously analyzed in Refs. 9 (question #10) and 10.
Flooding from smaller MELB in the condenser room is enveloped by the CW break. HELBs outside the con-denser room would be alanned in the control room by drain sump alarms, and no safe shutdown or break mitigating systems are a ffected. Operator action is require' to stop MELB's in this area. ESW function, if required, would not be lost in either train from a MELB.
MF, Condensate None None Adequate. HELB causes reactor (HELB) trip, but no mitigating or safe shutdown systens are affected by the break.
22
TABLE 1.
(Continued)
List of Abbreviations ADS - Automatic Depressurization System (part of the energency core cooling systems)
CRD - Control Rod Drive CS - Core Spray System (part of the energency core cooling systems)
CW - Circulating Water System demin. - demineralized ESWS - Emergency Service Water System HELB - High Energy Line Break LOCA - Loss of Coolant Accident LOP - Loss of Offsite Power MELB - Moderate Energy Line Break MF - Main Feed MS - Main Steam RBCCW - Reactor Building Closed Cooling Water System RBEDT - Reactor Building Equipment Drain Tank RPS - Reactor Protection System RWCU - Reactor Water Clean-Up System SLCS - Standby Liquid Control System SWS - Service Water System TBCCW - Turbine Building Closed Cooling Water System 23
24 REFERENCES 1.
NRC letter, A. Biambusso to CE Co., dated December 18, 1972.
2.
Oyster Creek FDSAR Amendment 75, dated July 1, 1974.
3.
Revision 1 to Oyster Creek FDSAR Amendment 75, dated December 24, 1974.
4.
Revision 2 to Oyster Creek FOSAR Amendment 75, dated March 24, 1975.
5.
Revision 3 to Oyster Creek FDSAR Amendment 75, dated April 25, 1975.
6.
Revision 4 to Oyster Creek FDSAR Ame dment 75, dated June 1, 1976.
7.
Revision 5 to Oyster Creek FDSAR Amendment 75, dated September 21, 1976.
8.
NRC letter, G. Lear to I. Finfrock, Jr., dated December 27, 1976.
9.
Oyster Creek Supplement No. 6 (Addendum 1) to the Application for a Full Term License, Docket No. 50-219, dated November 21, 1973.
- 10. JCP&L, I. Finfrock, Jr. to D. Ziemann, dated March 13, 1974.
- 11. JCP&L letter, I. Finfrock, Jr. to D. Ziemann, dated July 3,1974.
- 12. SEP Review of Safe Shutdown Systems for the Oyster Creek Nuclear Power Plant.
(SEP Topics VII-3, V-10.3, V-11.A, V-11.8, X).
EiiCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION SEP TOPIC III-5.B. PIPE BREAK OUTSIDE CONTAINMENT OYSTER CREEK 1.
Provide a comparison of the design of the containment penetration piping outside containment between the containment and the outermost containment isolation valves for the main steam lines, emergency condenser steam and condensate lines, and reactor water cleanup lines with the provisions of section B.l.b of Branch Technical Position MEB 3-1 (appended to Standard Review Plan 3.6.2) in sufficient detail to identify the degree of conformance with and deviations from these provisions.
2.
Provide a comparison of the design of the containment penetration piping outside containment for the emergency condenser steam lines and reactor water cleanup lines with the provisions of section B.2.C of Branch Technical Position ASB 3-1 (appended to Standard Review Plan 3.6.1) in sufficient detail to identify the degree of conformance with and deviations from these provisions.
3.
Provide an evaluatisn of the potential effects of damage to cable tray 13A on the 51' elevation of the reactor building from a postulated break in the reactor water cleanup system. Consider the effects of pipe whip, jet impingement, and high temperature on the electrical cables.
4.
Provide an evaluation of potential flooding in the cable spreading room from a postulated break in the fire water system or turbine building closed cooling water system. Determine the depth of flooding, what equipment could be flooded, and the effects of loss of that equipment.
\\
- m
)
ENCLOSURE 3 STAFF POSITIONS ON SEP TOPIC III-5.B PIPE BREAKS OUTSIDE CONTAINMENT 0YSTER CREEK NUCLEAR PLANT 1.
Because inadequate protection exists from the effects of postulated breaks in the emergency condenser steam and condensate lines on the 75' elevation of the reactor building, the licensee should submit, by September 1,1980, a schedule for modifications to be installed to provide adequate protection from these postulated breaks. The modifi-cations must be in accordance with the acceptance criteria of Standard Review Plan 3.6.1 and provide protection for the emergency condenser isolation valves and controls and for cable trays V22, V23, 41, 42, and 43.
2.
To provide adequate protection from the effects of postulated main steam and main feed line breaks in the turbine building mezzanine area, the licensee should move or provide protection for all core spray system control cables in that area.
By September 1,1980, the licensee should provide a schedule for resolution of this issue.
l l
I