ML19339A007
| ML19339A007 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 09/25/1980 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Finfrock I JERSEY CENTRAL POWER & LIGHT CO. |
| References | |
| TASK-05-10.B, TASK-05-11.A, TASK-05-11.B, TASK-07-03, TASK-09-03, TASK-5-10.B, TASK-5-11.A, TASK-5-11.B, TASK-RR NUDOCS 8010310532 | |
| Download: ML19339A007 (65) | |
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TSUIDRY DIE'".'I COPY 0 STRIBUTION Glainas Docket)
DCrutchfield IiRC FDR HSmi th LPDR Project Manager Docket No. 50-219 TERA OELD NSIC Ol&E (3)
SF.PTEMBER ~, *,.A80 NRR Reading ACRS (16)
ORB #5 Reading SEP File M r. 1. R. Fi nf rock, J r.
DEisenhut TNovak Vice President - Generation JRoe RTedesco Jersey Central Power & Light Company SEP BC SEP TF (5)
Macison Avenue at Punch Bowl Road SEP PF (3)
RSchaffstall Horristown, New Jersey 07900
Dear Mr. Finfrock:
RE: SEP TOPICS V-10.8, V-II. A, V-ll.B. VII-3 and IX-3 ( SAFE SrlUTDOWN SYSTEMS) - OYSTER CREEK NUCLEAR GENERATIhG STATION, UNIT N0.1 Encloseo is a copy of our current evaluation of Safe Shutocwn Systems (Revision 1) for Oyster Creek Nuclear Generating Station, Unit No.1.
This assessment Codiparts your facility As described in Docket No. 50-219 with the criteria currently used by the regulatory staff for licensing new facilities.
Please inform us if your as-built facility ciffers from the licensi.ng basis assumed in our assessment within 90 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 neecea to reflect the as-ouilt conditions at your f acility. This assessment t::ay De reviseo in the future if ) car faci 1ity design is changed or if NRC criteria reiating to this suoject is modified before the integrated assessment is completeo.
I am also enclosing Staff Positions regarding the SEP Safe Snutdown Systems review for your facility.
Sincerely, Origi2:31 Signed by A nniS L Crutchfield Dennis M. Crutenfield, Chief Operating Reactors Branch v5 Division of Licensing
Enclosure:
1.
Completed SEP Topics -
Safe Shutdown Systems 2.
Staff Positions i
cc w/ enclosures:
See next page 801033063A
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION y
<3 WASHINGTON. D. C 20555
'p September 25, i980 Docket No. 50-219 Mr. 1. R. Finf rock, J r.
Vice President - Generation Jersey Central Power & Light Company Madison Avenue at Punch Bowl Road Morristown, New Jersey 07960
Dear Mr. Finfrock:
RE:
SEP TOPICS Y-10.9, Y-11. A, Y-ll.B, VII-3 and IX-3 (SAFE SHUTDOWN SYSTEMS) - OYSTER CREEK NUCLEAR GENERATING STATION, UNIT N0.1 Enclosed is a copy of our current evaluation of Safe Shutdown Systems (Revision 1) for Oyster Creek Nuclear Generating Station, Unit No.1.
This assessment compares your f acility, as described in Docket No. 50-219 with the criteria currently used by the regulatory staff for licensing new fa-ilities.
Please inform us if your as-built facility differs from the licensing basis assumed in our assessment within 90 days of receipt of this letter.
This evaluation will be a "asic input to the integrated safety assessment for your f acility unless you identify changes needed to reflect the as-built conditions at your f acility. This assessment may oe revised in the future if your facility design is changed or if NRC criteria relating to this
- Jbject is modified before the integrated assessment is completed.
I am also enclosing Staff Positions regarding the SEP Safe Shutdown Systems review for your facility.
Si cerely, 7#.
ennis 6 Cruten 1 eld, Chi Operating Reactors Branch 5 Divisien of Licensing
Enclosures:
1.
Completed SEP Topics -
Safe Shutoown Systems 2.
Staff Positions cc w/ enclosures:
See next page
M r. 1. R. Finf rock, J r.
-2' Septembe.r 25, 1980 cc w/ enclosures:
G. F. Trowbridge, Esquire Gene Fisher Shaw, Pittman, Potts and Trowbridge Bureau Chief 1800 M Street, N. W.
Bureau of Radiation Protection Washington, D. C.
20036 380 Scotts Road Trenton, New Jersey 08628 GPU Service Corporation ATTN: Mr. E. G. Wallace Comi ssioner Licensing Manager New Jersey Department of Energy 260 Cherry Hill Road 101 Commerce Street Mrsippany, New Jersey 07054 Newark, New Jersey 07102 Natural Resources Def ense Council Plant Superintendent 917 15th Street, N. W.
Oyster Creek Nuclear Generating Washington, D. C.
20006 Station P. O. Box 388 Forked River, New Jersey 08731 Steven P. Russo, Esquire 248 Washington Street Resident Inspector P. O. Box 1060 c/o V. S. NRC Toms River, New Jersey 08753 P. O. Box 128 Forked R.ver, New Jersey 08731 Joseph W. Ferraro, J r., Esquire Deputy Attorney General Director, Technical Assessment Div.
State of New Jersey Office of 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 Tcwnship Branch 401 Chambers Bridge Road U. S. Environmental Protection Brick Town, New Jersey 08723 Agency Region 11 Office Mayor ATTN: EIS COORDINATOR Lacey Tcwnship 26 Federal Plaza P. O. Box 475 New York, New York 10007 Foried River, New Jersey 08731 Richard E. Schaffstal' Commi ssioner KMC incorporated Department of Public Utilities 1747 Pennsylvania Avenue, N. W.
State of New Jersey Washington, D. C.
20006 101 Ccmerce Street Newark, New Jersey 07102
4-ENCLOSURE 1 SEP REVIEW OF SAFE SHUTD0%N SYSTEMS FOR Ti'c OYSTER CREEK NUCLEAR PCWER PLANT REVISICN 1
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TABLE OF CONTENTS
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P_ag g
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1.0 INTRODUCTION
7 2.0 OISCUSSION..................................
7 2.1 Normal Plant Shutdown and Cooldown............
- 2. 2 Shutdown and Cooldown with Loss of Offsite Power...
9 3.0 SHUTOC%N AND CCOLDOWN FUNCTIONS 10 AND METH005.
t.0 COMPARISON OF SAFE SHUTD0hN SYSTEMS WITH 24 CURRENT NRC CRITERIA..
28 4.1 Functional Requirements.......
4.2 Residual Heat Removal System Isolation 30 Requirements.................
32 4.3 Pressure Relief Requirements...........
33 4.4 Pump Protection Raquirements.....
34
- 4. 5 Test Requirements.................
35 4.6 Operational Procedures....................
4.7 Auxiliary ceedwater Supply.
35 Table 4.1 Classification of Safe Shutdown Systems.
37 Table 4.2 List of Safe Shutdown Instruments.
40 Table 4.3 Shutdown Systems Power Supply and Location.
41 5.0 RESOLUTION OF SYSTEMATIC EVALUATION PROGRAM TOPICS.
42 5.1 Topic V-10.8 RHR System Reliability.
42
- 5. 2 Topic V-ll. A Requirements for Isolation of 43 High and Low Pressure Systems...
44
- 5. 3 Topic V-ll.3 RHR Interlock Requirements.
44 5.4 Topic VII-3 Systems Require for Safe Shutdown.
47
6.0 REFERENCES
Accendix A.
Safe Shutdown Watcr Requirements -
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1.0 INTRODUCTION
The Systematic Evaluation Program (SEP) review of the " safe shutdown" subject encompassed all or parts of the #cllowing SEP topics, which are among those identified in the November 25, 1977 NRC Office of Nuclear Reactor Regulation document entitled " Report on the Systematic Evaluation of Operating Facilities":
1.
Residual Heat Removal System Reliabflity (Topic V-10.9) 2.
Requirements for Isolation of High and Low Pressure Systems (Topic V-ll.A) 3.
RHR Interlock Requirements (Topic V-ll.8) 4.
Systems Required for Safe Shutdown (Topic VII-3) 5.
Station Service and Cooling Water Systems (Topic IX-3)
The review was primarily performed during an on-site visit by a team of SEP personnel.
This on-site effort, which was performed on August 8 &
9, 1978, afforded the team the opportunity to obtain current information and the licensee (Jersey Central Power & Light Company) the opportunity to provide input into the review.
The review included specific system and equipment requirements for remaining in a shutdown condition (defined in the Oyster Creek Technical Specifications as the reactor mode switch-being in the shutdown mode position) and for proceeding to a cold shutdown condition (defined as mode switch in shutdown mode position, all operable control rods fully inserted, and the reactor coolant system maintained at less than 212*F
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l and vented).
The review for transition from operating to shutdown I"
considered the requirement that the capability exists to perform this operation from outside the control room.
The review was augmented as necessary to assure resolution of the applicable topics, except as noted below:
Topic V-11.A (Requirements for Isolation of High and Low Pressure Systems) was examined only for application to the Shutdown Cooling System. Other high pressure / low pressure interfaces were not inves-l l
tigated.
The Shutdown Cooling System is the Oyster Creek equivalent of j
an RHR system.
i Topic VII-3 (Systems Required for Safe Shutdown) was completed except for
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determination of design adequacy of the systems.
Topic IX-3 (Station Service and Cooling Water Systems) was only reviewed to consider redundancy and seismic and quality classification of cooling water systems that are vital to the performance of safe shutdown system components.
(No discussion of Topic IX-3 is included in this report.
The information gathered during the safe shutcown review will be used to resolve this topic later in the SEP.)
The criteria against which the safe shutdown systems and components were compared in this review are taken from the:
Standard Review Plan
($RP) 5.4.7, " Residual Heat Removal (RHR) System"; 3ra.ch Technical Position RSB 5-1 Rev. 1, " Design Requirements of *:.e Residual Heat l
.. Removal System" and; Regulatory Guide 1.139, " Guidance for Residual Heat i
Removal" These documents represent current staff criteria and are used in the review of facilities being processed for operating licenses.
4 This comparison of the existing systems against the current licensing criteria led naturally to at least a partial comparison of design criteria, which will be input to SEP Topic III-1, " Classification of Structures, Components and Systems (Seismic and Quaiity)"
This recort will also be reviewed for its apr'ication to the resolution of other topics.
As noted above, the five tonics were examined while neglecting possible interactions with other topics and other systems and components not directly related to safe shutdown.
For example, Topics II-3.3 (Flooding Potential and Protection Requirements), II-3.C (Safety-Related Watar d
Supply), III-4.C (Internally Generated Missiles), III-5.A (Effects of Pipe Break on Structures, Systems, and Components Inside Containment),
III-6 (Seismic Design Considerations), III-10. A (Thermal-Overload Protection for Motors of Motor-Operated Valves), III-11 (Component Integrity), III-12 (Environmental Qualification of Safety-Related Equicment) and V-1 (Compliance with Codes and Standards) are among several topics which could be affected by-the results of the safe snutdown review or could have a safety impact upon the systems which were reviewed. These effects will be determined by later review.
- Further, this review did not cover in any significant detail the reactor protection
=4-system nor the electrical power distribution, both of which will also i
)
be reviewed later.
The staff considers that the ultimate decision concerning the safety of any of the SEP facilities depends upon the ability to withstand the Design Basis Events (DBEs).
The SEP topics provide a major input to the SEP DBE review, both from the standpoint of assessing the probability of l
As the event and that of determining the consequences of tne event.
examples, the safe shutdown topics pertain to the listed DBEs '.the extent of applicability will be determined during plant-specific review):
l-Impact Upon Probability Tooic DBE Group Or Consecuences of OBE i
V-10.B VII (Spectrum of Loss of Coolant Consequences (Accidents)
V-11.A VII (Defined above)
Probability V-11.8 VII (Defined above)
Probability VII-3 All (Defined as a generic topic)
Consequences IX-3 III (Steam Line Break Inside Consequences Containment)
(Steam Line Break Outside Containment)
IV (Loss of AC Power to Station Consequences Auxiliary)
(Loss of all AC Power)
V (Loss of Forced Coolant Flow)
Probability (Primary Pump Rotor Seizure)
(Primary Pump Shaft Break)
VII (Defined above)
Consequences
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. Completion of tne safe shutdown topic review (limited in scope as noted above), as documented in this report, provides significant input in assessing the existing safety margins at Oyster Creek.
Piping System Passive Failures The NRC staff normally postulates piping system passive failures as
- 1) accident initiating events in accordance with staff positions on piping failures inside and outside containment, 2) system leaks during long term coolant recirculation following a LOCA, and 3) failures resulting from hazards such as earthquakes, tornado missiles, etc.
In this evaluation, certain piping system passive failures have been assumed beyond those normally postulated by the staff, e.g.,
the catastrophic i
f ailure of mcderate energy systems.
These assumptions were made to i
i demonstrate safe shutdown system redundancy given the complete failure of these systems and to facilitate future SEP reviews of DBEs and other topics
'wnich will use the safe shutdown evaluation as a source of data for the SEP facilities.
SRP 5.4.7 and BTP RSB 5-1 do not require the assumption of piping system passive failures.
Credit for Ooerating Procedures For the safe snutdown evaluation, the staff may give credit for facility operating procedures as alternate means of meeting regulatory guidelines.
Those procedural requirements identified as essential for acceptance of an SEP topic or OBE aill be carried through the review process and con-1 sidered in the integrated assessment of the facility.
At that time, we will:
(1) decide which procedures are so important that they snould be
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I included in technical specifications and (2) establish an administrative l
procedure (e.g., FSAR changes) for ensuring that the other operating l
l procedures are not changed without appropriate ;onsideration of their j,
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importance to the topic or OBE evaluations.
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. 2.0 OISCUSSION 2.1 Normal Plant Shutdown and Cooldown
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Power is reduced from its operating value during commencement of shutdown by first simultaneously reducing recirculation flow in all loops to a specified value, about 50,000 gpm.
Reactor coolant system pressure is i
controlled by the electrical pressure regulator / mechanical pressure regulator (EPR/MPR) and maintained between 980 and 1020 psig to the hot shutdown condition. Thus, as power is reduced, the tuibine control valves are correspondingly closed to " hold" reactor pressure.
After achieving the desired recirculation flow, control rods are selectively inserted in a prescribed pattern to continue the power decrease.
Power is reduced at a rate compatible with the load dispatcher's requirements.
Feedwater heaters are removed from service wnen power reaches about 200 MWe.
Feedwater is controlled in " automatic" with the master controller until high flow is no longer neeced.
As steam ficw is reduced, feedwater matches and when all 3 pumps are no longer needed, a feedwater train is placed on its individual flow controller and manually controlled until flow is stopped and the train is secured.
Flow is reduced until one pump operates and its controller is placed in " manual",
controlling via the aw flow control valve around the main feedwater control valve wnen flow reaches about 1000 gpm.
Finally, curing cooldown when tne last operating feed pump is no longer needed for reactor water level c caol, the icw flow control valve is fully closed and the pump is
o.
3 tripped.
Then as RCS pressure is lowered, the vessel can be fed via a condensate pump. While shutting down, water continues to be added, about 70 gpm, to the RCS inventory by the control rod drive (CRD) hydraulic pumps.
When power reaches about 100 MWe, the station electrical feed is switched from the auxiliary transformer (station generator) to the startuo trans-formers.
Power is reduced further and the turbine generator is removed from service. With the turbine no longer extracting energy, steam is bypassed to the main (;ondenser at very low power and the heat is trans-ferred to the circulating bay water.
The reactor is at pressure
(>860 psig) and critical at low power, minimum RCS recirculation flow is on, steam is being bypassed to the condenser with a feedwater train in service and thus the reactor is at hot standby.
Cooldown is now accomplished, if desired, by continuing with control rod insertion and with a feedwater train on controlling reactor water level in " manual" via the lowflow valve and bypassing steam to the main condenser.
This is continued, establishing a cool down rate not to exceed 100 F/hr. or a metal to flange AT (vessel or head) of 2C0 F.
When RCS temcerature reaches 350*F,the Shutdown Cooling System (SCS) is placea in service.
Reactor Building Closed Cooling Water (RSCCW) flow to the SCS heat exchangers is established and service water flow to the RSCCW heat exchangers is already established therecy cr9ating the heat transport path to the bay.
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. Normally all recirculation pumps continue to run until vessel cooldown is complete then shutdown as desired. When all control rods are fully inserted, RCS temperature is less than 212*F, the mode switch is in
" shutdown", and the reactor is in ccid shutdown.
2.2 Shutdown an'.fcoldown With Loss of Offsite Power On loss of offsite power the main condenser is unavailable for heat removal following reactor trip.
The reactor can stay in the hot condition briefly while pressure is controlled with relief valves.
The two isolation condensers activate on sustained high RCS pressure or may be "sanually" activated.
The single closed valve in each condenser system, in the condensate return lines, is opened and main steam passes througn the isolation condenser tubes and boils off the secondary side water in the condenser.
Makeup water is provided to the condensers from the condensate storage tank by transfer pumps powered from onsite sources or Dy the fire protection system using diesel fire pumps.
The reactor is cooled by the isolation condenser until the SCS interlock temperature is reached.
The SCS may then be put in service as atove since it, the RBCCW, and Service Water Systems are powered by onsite electrical sources.
Cooldown is accomp11shed as described in Section 2.1.
If the isolation condensers were unavailaele, depressurization of RCS by operation of relief valves and activation of core spray at the lowered pressure would provide an alternate means of decay heat removal and cooldown to the cold shutdown condition.
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3.0 SHUTOOWN AND CCOLDOWN FUNCTIONS AND METHCOS This section will describe the existing systems available at Cyster Creek to accomplish the necessary functions for the safe shutdown of the reactor following either the loss of offsite power or the loss of onsite AC power.
Seismic and Quality Group Classifications of the pertinent equipment (based upon USNRC Regulatory Guides 1.26 and 1.29) will be discussed in Section 4.0.
The minimum list of safe shutdown systems is also provided in Section 4.0.
The losses of offsite and onsite AC power are tat considered to be concurrent or sequential events, but rather, for the purposes of this evaluation, are taken as wholly independent occurrences.
Offsite power is supplied for startup through startup transformers SA and 58, each of which is supplied from a separate substation.
These trans-formers are capable of supplying all electrical auxiliaries with the main generator producing full power. When the turbine gene.ator output reaches approximately 50 MWe during the startup, electrical power for
" house" loads is transferred from the startuo transformers to the auxiliary transformer, which has one p*imary winding and two secondary windings.
Should this transformer fail (loss of onsite power) power is automatically transferred back to the startup transformers, preserving power to essential auxiliaries and minimizing the consequences of a loss of onsite power.
Transfer to either startup transformer will also occur upon loss of its " companion" secondary winding on the auxiliary trans f o rmer.
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Note that the above discussion omitted use of the diesel generators for supolying power to the 4160V emergency switchgear buses IC and 10 (which in turn supply 460V unit substations carrying additional essential electrical loads). Either diesel is capable of supplying sufficient power for the safe shutdown of the reactor.
Additionally, although the total 2nd simultaneous loss of offsite and onsite AC power (including diesel generators) is considered an extremely low probability event, Jersey Central Power and Light Company (JCP&L) management has prepared for such an occurrence by providing a procedure to De folicwed by plant operators in the event of complete l'oss of AC power.
Assuming a total loss of offsite power (this has never occurred at Oyster Creek) with the reactor at full power (1930 MWt, 650 MWe), a reactor scram would follow due to interruption of the protection system power supply. At 1050 psig, two electromagnetic relief valves (OC-cowered) would li ft to relieve the pressure.
Each valve is rated at 600,000 counds/
hour capacity. At 1060 psig, the two isolation condenser systems automatically would initiate after a time delay of less than 3 seconds, raoidly decreasing reactor coolant system cressure as natural circulation flow from the reactor through the condensers returns cold water to the reactor vessel.
Each isolation condenser shell contains a minimum (as per Technical Specifications) of 22,730 gallons, wnic? represents 11,060 gallons of water acove the tubac.
Both condensers in operation can absorb reactor decay heat for one (1) hour and forty (40) minutes witnout reolenishment a
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of the water supply, while one (1) concenser alone can remove decay heat for forty-five minutes without replenishment. These figures are based upon an analyzed scram from 1950 Wt, which is greater than maximum allowable power and is therefore conservative.
Makeup to the isolation condensers is provided by either the condensate transfer system (normal e
source) or the fire protection system.
The condensate transfer pumos (two) receive their power from the same vital" bus.
Power can be supplied by the diesel generators, but the pumps will not start auto-Even in the matically and must be brought onto the diesel bus manually.
unlikely event that these pumps fail, the two diesel-driven fire pumps can be utilized. These diesels have separate battery supplied starter All valves in systems and can be relied upon to provide makeup water.
the makeup path from either source (condensate or firewater) are local manual valves with the exception of the inlet valves to the condensers themselves.
These valves, wnich are normally air-operated remote-manually from the control room, can easily be overridden locally to prov'oe makeup flow to the condensers.
4 Each isolation condenser is prov'Jed steam inlet f rom its own single-use no::le in the reactor vessel.
This line contains,in series,two motor-operated valves (one AC, one OC) which are maintained open during i
i operation at reactor coolant temperature greater than 212*F (~ecnnical Specifications allow one isolation condenser to be out of service for a i
l j
period not to exceed seven days, provided augmented surveillance of the l
operable condenser is cerformed).
Thus,the piping and isolation condenser The return line from tubes are always pressurized at reactor pressure.
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. each condenser to the reactor (cne to recirculation loco A, one to loop E) contains two valves, the first of whicn is DC-powered and is normally closed (opened only to initiate flow), the second of which is AC-powered, normally open, and is the only one of the four valves of each ccndenser system to be inside containment.
Power to the two AC-powered valves for each condenser is provided from vital ector control center (MCC) 1AB2, which can ce powered from either emer;ancy diesel generator through an automatic bus transfer switch to eitner motor control center IA2 or 182.
The adequacy of tnis electrical Vital arrangement will be further examined under SEP topic VII-7.
I MCC 1 AB2 is important in that it powers not only the isolation condenser AC valves, but also provides power to shutdown cooling system and core 1
spray system valves, among others.
The failure of this MCC will have no effect upon the normal operation of the isolation condensers since the AC valves are normally open and will f ail "as is" upon loss of electrical power.
Each condenser will be totally isolated from the reactor, and thus inoperable, with closure of all valves (two AC and two DC), upon receipt of a high flow signal from sensors in its own steam supply and/or con-densate return lines.
Actuation of the sensors, and subsequent isolation, has occurred in the past uoon initiation of isolation condenser flow.
This condition was apparently due to the sensitivity of the flow sensors and has since been corrected, with the result that during tne one condenser initiation event since the alter tion no problem occurred.
s
The isolation of both condensers upon initiation, already an unlikely event after the modification mentioned above, ould only become a problem Loss of this MCC wit:. che highly unlikely concurrent loss of MCC 1AB2.
would result in the inability to open the one valve in each system which is located inside containment and is thus inaccessible.
Jersey Central Power & Light Co. is prepared for this unusual circumstance by having provided a procedure for reactor shutdown with attendant loss of normal reactor cooling mechanisms, including loss of the isolation condensers.
This is discussed later.
Power to the two DC-operated valves for condenser A and the CC-operatec Power to the inlet valve for condenser B is provided from MCC OC-1.
MCC DC-2 DC-operated outlet valve of condenser B is provided by MCC OC-2.
receives power from DC distribution center C.
MCC CC-1 receives power from DC distribution center A or B through an automatic transfer switch.
All other DC isolation valves, such as those for the shutdown cooling system, are on this MCC.
Modifications to the DC electrical distribution These center have been completed and are under review by the NRC staff.
t 1978 modifications are described in the licensee's letter of April 4, (Reference 4).
k.
both isolation condenser systems function Assuming that one or properly, the next system of concern during a shutdown following loss of offsite power is the shutdown cooling system.
This system has s single suction line from recirculation loco E and a single discharge line to recirculation loco E.
Very little of this system is inside containment.
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However, two AC motor-operated isolation valves, one on the suction side, the other on the discharge, are located inside containment, receive power from vital MCC 1 A82, and upon the unlikely loss of the MCC would be This sas also noted acove in the discussion of the isolation inoperable.
Also, a single f ailure of one of these valves to open condenser systems.
would result in inoperability of the entire shutdown cooling system.
Outside the drywell, the shutdown cooling system Oranches into three headers, each containing (as major equipment) a DC motor-operated suction valve, pump, heat exchanger, and a DC-mot.or-operated discharge valve.
These headers then return to a common line through the AC discnarge valve The mentioned above.
(Thus diversity of isolation power is provided).
DC power-neersted valves are powered from MCC CC-1.
This system was j
designed for 1250 psig (Reactor Coolant System pressure) at 350 F, which j
is less than reactor temperature. However, it would take multiple i
i f ailures of the valves (all of whicn are normally shut) and interlocks to initiate flow at temperatures greater than 350*F.
The interlocks prevent AC valve opening at temperatures greater than 350*F and will isolate the Each of the five system if it is in operation, upon reacning 350*F.
recirculation loops must be less than 350*F to satisfy tnis interlock.
Additionally, each pump is interlocked such that starting is prohibited unless suction pressure exceeds 4 psi and temcerature in tne suction line is less tnan 350*F.
! In order to accommodate decay heat after shutdown and cooldown to its initiation limits, the shutdown cooling system requires only two of the three branches (i.e., two pumps and two heat exchangers) to be in service.
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Cooling flow to the shutdown cooling system heat exchangers is provided by the Reactor Building Closed Cooling Water (RBCCW) system whicn is in i
turn cooled by the Service Water (SW) system.
Only one of two R8CCd pumcs and heat exchangers need be in service to provide shutdown cooling, proiided no RSCCW flow is being supplied to the reactor water cleanup (RWCU) system non-regenerative heat exchanger and the five reactor coolant recirculation pumps.
Power to shutdown cooling pump A is provided from 460 volt unit substation IA2, with pumps B and C being supplied from substation 182.
Upon loss of offsite power, these substations are provided power from the emergency diesel generators.
The RBCCW system, which provides cooling water to the shutdown cooling system, consists of two pumps and two heat exchangers, which are in turn cooled by the SW system as noted above.
Power to the pumps is provided f rom 460 volt unit substations l A2 (pumo 1-1) and 182 (pump 1-2).
These substations are supplied by the emergency-diesels, assuring that even in the event of single bus or single diesel failure, one pump will be available.
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The RBCCW system has a common suction line for both pumps and a common discharge line from the heat exchangers.
Passive failure of either line would result in loss of RBCCW and subsequent heatup of the reactor coolant to the point of shutdown cooling system isolation.
Alternate means for core cooling for this circumstance are discussed below.
There is only one motor-operated valve unich isolates RBCCW flow from the shutdown cooling sy-tem.
This valve, which is AC-powered, is on the common discharge of the three shutcown cooling heat exchangers.
It is outside containment and is accessible for manual operation.
The need for such operation is unlikely, since valve power is supplied from MCC 1921 A, which can be supplied from emergency diesel generator number 2.
The RBCCW system is cooled by the SW system, which consists of two pumps and associated valving and piping.
The SW system provides flow primarily to the RBCCW system heat exchangers and is also utilized to maintain the emergency service water system (for containment spray heat exchanger cooling) filled.
However, it is also an alternate means of cooling for the turbine building closed cooling water heat exchangers and is used as such when main condenser circulating water pumps are not operating.
All i
valves in the SW system are manually operated, and power to the pumps is provided by 460 volt unit substations I A3-(pumo 1-1) and 133 (pumo 1-2).
These substations are powered respectively from 4160 voit emergency switchgear buses 1C and 10, wnich are provided cower from the diesel generators (CGI to IC; CG2 to 10) in an emergency.
.. The shutdown cooling system operates at a higher pressure than the RBCCW, which in turn is at a higher pressure than the SW system.
A leak in both the shutdown cooling and RSCCW heat exchangers and a failure to take proper action would be required to release radioactivity to the environ-ment. The RBCCW and SW systems both incorporate radiation detectors to alert operators to leakage, and the R8CCW system includes a surge tank with high and low level alarms.
The only significant failure mode of the SW system would be passive failure of the common pump discharge header, which would entirely disable the system.
As with the loss of R8CCW discussed before, this subject will be treated below.
The acove discussion has dealt with systems which would normally be used for the safe shutdown of the reactor upon loss of offsite power.
There are, of course, alternate means utilizing other equipment, which will permit safe shutdown should any or all of the above-mentioned systeas fail.
As a review of the above, loss of offsite power results in turbine trip, reactor scram, loss of condenser circulating water pumps (rendering the condensers useless for cooling), and loss-of feedwater and condensate pumps.
In such a situation, the redundant isolation condenser systems would be relied upon to reduce reactor 3ressure and temperature to the point of shutdown cooling initiation.
.o However, even when hypothesizing the f ailure of coth independent isolation condensers, Oyster Creek has sufficient capability in the reactor pressure relief system to assure depressurization.
There are five electromatic relief valves (EMRVs) which are located on the main steam lines in the drywell and which discharge into the pressure suppression pool (torus).
Although these valves would function automatically to relieve pressure (two valves lift at 1050 psig, three at 1070 psig), they can be actuated by remote-manual means from the control room.
They are DC-powered and receive power from either 125 voit DC power panel 0 (battery room) or panel F (460 switchgear room).
As is noted in a JCP&L Oyster Creek crocedure concerning loss of reactor cooling mechanisms during reactor shutdown, the control rod drive hydraulic system may be used to maintain vessel level while maintaining only enough blowdown to limit reactor pressure to acceptable levels.
However, as noted in another procedure concerning complete loss of AC power, the EMRVs may be opened to reduce vessel pressure and temperature to tne levels at anich fire water to the core spray system can be utilized, assuming shutdown cooling and core scray were not availacle (.nich is not the Case here). Oyster Creek FDSAR Amendment 10 states tnat tne automatic depressuri:ation system would function upon receiving the aporopriate low low low reactor water level, high drywell pressure, and core spray booster pump discharge pressure signals. This system will allow depressurization in sufficient time to add a substantial amount of core spray and prevent fuel clad temperature f rom exceeding 2C00*F.
?DSAF figure VI-6-6 coes show that the core is temporarily and partially uncovered.
L 20 -
Under the circumstances of this analysis we nave assumed, in addition to In l
the loss of offsite power, the loss of both isolation condensers.
this case, the operator, as noted above, could choose to remain at hot 5
standby, maintaining level with the control rod drive system while If plant conditions relieving pressure througn the relief valves.
dictated the need to immediately decrease pressure and cool the system, the use of the relief valves would serve this purpose, and would probably accomplish the necessary depressurization prior to uncovering the top of However, even were the level to decrease to the low-low-low the core.
setpoint prior to blowdown initiation, the FOSAR analysis concludes that no clad melting would occur. We find tne temporary and partial uncovery of the core, in this scenario, to be an acceptacle event, given first that we have assumed.3ultiple failures to achieve the scenario and second that no fuel melting would occur, as previously calculated, since a large influx of :coling water would be available upon completion of the depressurization.
Torus cooling may be desired by the operator utili:ing tne containment The cumps and motor-operated spray and emergency <ervice water systems.
valves of th3se systems are powered from sources wnich can be supplied by the emergency diesel generators.
As was discussed before, Cyster Creek to remain has the capability, through use of either isolation condenser, The hot while removing core decay heat upon loss of of fsite AC power.
isolation condensers may also be used to cool down.
The discussion immediately above shows that there is redundant capability for
decressurization and cooldown, utilizing the EMRV's and tne core spray / fire water systems.
Taken to an extreme, the method above (EMRVs and Core Soray) could function as a closed loop by filling the vessel with Core Spray, overflowi9g hot watar to the pressure-suppression chamber (torus) through The torus provides water to the Core Spray system and the relief valves.
cooling for such water would be provided by the containment spray system.
The cycling of the water through the core and through the relief valves to the torus and back again would only be limited by the design of the relief valves themselves.
These valves incorporate a spring which must The spring will shut the valve at be overridden by system pressure.
approximately 50 psig (FDSAR Page VI-6-1) and will hold it shut until the core heats up again and raises pressure or until pressure is increased by the Core Spray pumps.
The core spray system contains four main cumps and four bcoster pumps.
These Each of the main pumps provides 1700 gpm each to the reactor.
pumps and all motor-operated valves of the system are powered from e
Like the other systems discussed above, the component emergency AC buses.
trains and their power supplies are arranged such that failure of one emergency diesel generator will not disable the entire system.
The core spray system is interlocked such that it will not orovide water to tne reactor until reactor pressure nas decreased to 285 psig.
1
reactor water However, the system automatically initiates upon a low-lcw level signal, with a minimum flow line returning flow to the torus until the pressure interlock is satisfied and the system discharge valves open.
Although the torus (nominally 85,000 cubic feet chromated water) is the preferred source of supply, water can also be drawn from the condensate storage tank (250,000 gallons minimum) or the demineralized water tank (30,000 gallons). Vessel level can be maintained by manual oceration of easily-accessible valves, if required.
If, for any reason, the core spray pumps are inoperable, fire water system flow can be supplied to core spray and the reactor kept cooled.
The interconnection valves are readily-accessible for manual operation, flow is provided by diesel-driven pumps which are battery-started, and JCP&L has even provided a procedure which includes using a fire truck from a local fire department to pressurize the core spray system.
As an alternative to the core spray system, the shutdown cooling system could be utilized to provide cooling after the EMRV depressurization.
RBCCW and SW systems would then be necessary. If either of these systems fail, rendering shutdown cooling inoperable, a means still exists to maintain the core covered and cooled.
Letdown from the reactor can be l
accomplished through the reactor water cleanup demineralizer system, although a temperature interlock to prevent resin damage would have to be disabled.
Flow of cool water to the vessel could be obtained through core spray, as noted above.
The letdown flow from the cleanup 1
l
., j domineralizer could be routed to the concensate storage tank or to the radioactive waste processing system.
I If the RBCCW and SW systems are operable but shutdown cooling is not, i
some cooling could be maintained by increasing RSCCW flow to the cleanup This capability demineralizer system's non-regenerative heat exchanger.
has been accounted for in an Oyster Creek procedure.
CONCLUSION As can be readily seen from the foregoing discussion, Oyster Creek has the ability to withstand multiple failures and still retain the capability Problems with systems to be to depressurize and cool the reactor core.
primarily relied upon have been noted, as was the yet unreviewed change to the 125 volt DC system.
We are satisfied that Oyster Creek can be safely shutdown upon loss of onsite or offsite AC power, even considering f ailure of a single major component.
. 4.0 CCMPARISCN OF SAFE SHUT 00kN SYSTEMS WITH CURRENT NRC CRITERIA The current criteria used in the evaluation of the design of systems required to achieve cold shutdown for a new facility are listed in the Standard Review Plan (SRP) Section 5.4.7 and 3 ranch Technical Position I'
RSB 5-1 (or proposed Regulatory Guide 1.139).
This section discusses the I
comparison of these criteria with the safe shutdown systems of the Oyster Creek nuclear power plant.
This comparison will be done by quoting a sec-tion of the Branch Technical Position RSB 5-1 and then discussing the degree to which Oyster Creek meets the requirements of that particular section.
"A.
Functional Requirements The system (s) wnich can be used to take the reactor from normal operating conditions to cold shutdown
- shall satisfy the functional requirements listed below.
1.
The design shall be such that the reactor can be taken from normal operating conditions to cold shutdown
- using only safety-grade systems. These systems shall satisfy General Design Criteria 1 through 5.
2.
The system (s) shall have suitable redundancy in comconents and features, and suitable interconnections, leak detection, and isolation capabilities to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system function can be accomplished assuming a single failure.
3.
The system (s) shall be capable of beirg operated from the control room with either only onsite or only offsite oower available with an assumed single failure.
In demonstrating that the system can perform its function assuming a single failure, limited operator action outside of the control room would be considered acceptable if suitably justified.
' Processes involved in cooldown are heat remcval, depressurization, flow circulation, and reactivity control.
The cold shutdown conditions, as described in the Standard Technical Specifications, refers to a sub-critical reactor with a r+ actor coolant temperature no greater than 200*F for a PWR and 212*F for a SWR.
eg 4.
The system (s) shall be capable of bringing the reactor to a cold shutdown condition, witn only of f site or onsite power available, within a reasonable period of time following shutdown, assuming the most limiting single f ailure."
Backaround A " safety grade" system is defined, in the NUREG 0138 (Reference 1) i discussion of issue #1, as one which is designed to seismic Category I (Regulatory Guide 1.29), quality group C or better (Regulatory Guide 1.26),
and is operated by electrical instruments and controls that meet Institute of Electrical and Electronics Engineers Criteria for Nuclear Power Plant Protection Systems, (IEEE 279).
Oyster Creek nuclear power plant was constructed prior to the issuance of Regulatory Guides T.26 and 1.29 (as Safety Guides 26 and 29 on 3/23/72 and 6/7/72 respectively).
Also Proposed IEEE 279, dated August 30, 1968, was issued late in the construction phase of the facility.
General Design Criteria 1 requires that these systems be designed, fabricated, erected and tested to quality standards, that a quality assurance (QA) program be implemented to assure that these systems perform their safety functions and that an appropriate record of design, fabrication, erection and testing be kept.
At the time that Oyster Creek was licensed, the NRC (then AEC) criteria for QA nere under develcpment.
Since that time, various QA related regulations and criteria have been instituted by the NRC, and the QA program for operation of the clant war approved by the staff on November 5, 1976.
O a
l 1 l
-l The Oyster Creek Technical Specification and QA program require appropriate I
QA records to be kept.
l General Design Criterion 2 requires that structures and equipment important to safety be designed to witnstand the effects of natural phenomena without loss of capability to perform their safety function.
The original Staff SER (Reference 2) states tnat the Oyster Creek power plant can safely survive a flood level of 23 feet above mean sea level (MSL). The maximum flood height at the Oyster Creek site has been 4.5 f t.
The licensee's seismic design bases specify that for a ground acceleration of 0.22g, there will be no loss of function of critical structures and components necessary to ensure a safe and orderly shutdown.
These seismic design bases were aporoved by the Staff in the original SER and will be re-reviewed as part of several SEP tasks.
?
General Design Criterion 3 requires that structures, systems and components important to safety be designed and located to minimize the effects of fires and explosions.
The Staff has Completed an evaluation of the fire safety requirements of the Oyster Creek nuclear power plant.
The results of this evaluation are given in Reference 3.
l
General Design Criterion 4 requires that equipment important to safety be designed to withstand the effects of environmental conditions for normal ope rati c.1, maintenance, testing and accidents.
Equipment should also be I
protected against dynamic effects such as internal and external missiles, pipe whip and fluid impingement.
l' i
The SEP will evaluate the extent to wnich Oyster Creek conform.s to GDC 4 j
when reviewing topics III-12, " Environmental qualification of Safety-Related Equipment," III-5.A, " Effects of Pipe Breaks Inside Containment,"
III-58, " Pipe Breaks Outside Containment," and III-4, " Missile Generation and Protection."
General Design Criterion 5 relates to the sharing of structures, systems and components important to nuclear safety among nuclear units.
Since the Oyster Creek nuclear plant is the single unit at the site, GOC 5 does not apply.
The BTP RS3 5-1 functional requirements focus on the safety grade systems that can be used to take the reactor fiom operating conditions to cold shutdown.
The staff and licensee developed a " minimum list" of systems necessary to perform this task.
Although otner systems may be used to perform shutdown and cocidown functions, the following list is the minimum number of system required to fulfill the STP RS8 5-1 criteria:
a
-w_
.. 1 l
1.
Reactor Control and Protection System l
2.
Isolation Condensers Condensate Transfer System (for isolation condenser makeup) 3.
Electromatic Relief Valves (all 5 of which consititute the Automatic 4.
Depressurization System of the ECCS) 5.
Core Spray System Emergency Service Water System and Containment Spray System (for 5.
containment cooling) 7.
Instrumentation for shutdown and cooldown" Emergency Power (AC and DC) and control power for the above systems 8.
and equipment.
i In addition to these systems, other systems may function as backup for the above systems and components.
The preceding discussion in Section 3
{
described both these systems and the systems which may function as backup.
1 Table 4.1 lists the minimum safe shutdown systems for the Oyster Creek
[
Nuclear Power Plant along with the comparison of present criteria with the criteria to which these components and subsystems were designed.
Table 4.3 provides the power suppiles and locations of these systems.
4.1 Functional Requirements The Reactor Control anc Protection System (RCPS) is designed on a channeli:ed basis to provide physical and electrical isolation between
- Saf e snutcown instruments are identified in Table 4.2.
h t
f 5
redundant reactor trip channels.
Each channel is functionally independen 1
of every other channel and receives power from two independent sources.
7 i
The power source for the RCPS is the instrument buses wnich can receive
{
l power from either onsite or offsite sources.
The RCPS fails safe (tripped) 1 J
on loss of power.
The system can be manually tripoed both from the control room and from other locations outside the control room.
The RCPS is designed so that a single failure will not prevent a reactor trip.
Initiation of a reactor trip causes the insertion of sufficient reactor control rods to make the core subcritical from any credible operating condition assuming the most reactive control rod remains in the fully withdrawn position.
The design of the RCPS, as well as safe shutdown related electrical control and power systems, will be evaluated later in the SEP.
The normal shutdown systems and alternate systems have been "eviewed in Section 3.
The isolation condeaers would be relied upon for cooling from full power conditions upon loss of the main condenser which is not available upon loss of offsite power.
The isolation condensers are capable of cooling the reactor to near c'id snutdown conditions.
After reactor system pressure is reduced to the cut-in pressure of the core spray system, this system could be manualty initiated and would take the reactor to cold shutdown conditions.
The reactor can be maintained in cold shutdown conditions using the core spray, AOS, emergency service water, and containment spray systems.
I The functional requirement to acnieve cold shutdown conditions within a reasonable period of time is evaluated in Appendix A.
l 4.2 RHR System Isolation Requirements The RHR system shall satisfy the isolation requirements listed below.
1.
The following shall be provided in the suction side of the RHR system to isolate it from the RCS.
(a) Isolation shall be provided oy at least two power-operated valves in series.
The valve positions shall be indicated in the c7ntrol room.
(b) The valves saall have independent diverse interlocks to prevent the valves from being opened unless tne RCS pres-sure is below the RHR system design pressure.
Failure of a power supply shall not cause any valve to change position.
(c) The valves shall have independent diverse interlocks to protect against one or both valves being coen during an RCS increase above the design pressure of tne RHR system.
The purpose of tnese requirements is to provide assurance that a low pressure shutdown cooling system will not be exposed, either througn a single operator error or failure of a single valve, to a pressure greater than design pressure.
4 I
The Oyster Creek Shutdown Cooling Syrtem is designed, as stated in Section 3, for reactor coolant system design pressure, 1250 psig.
The design temcerature is 350*F which is lower than the reactor coolant system design temperature (575 F).
The licensee is evaluating the ability of the SCS to withstand exposure to these high temcerature conditions on a one-time basis; newever, as pointed out in Section 3,
, multiple f ailures of valves (all of which are normally shut) and interlocks would be necessary in order for this situation to exist.
Section 3 described the interlock on the RHR system which prevents opening of the suction and discharge valves on the SCS if the reactor coolant temperature in any of the five coolant loops is greater than i
350'F.
Also noted was the fact that system isolation will occur upon I.
increase in temperature to 350*F.
Although redundant diverse pressure interlocks do not control the SCS isolation valves, the temperature interlock on the AC valves and the high design pressure of the SCS provide adequate protection foe the SCS.
The valves are motor operated and would fail in their "as-is" condition (which would be closed unless the SCS were in operation).
Thus, the Cyster Creek SCS acceptably meets the present criteria for SCS system isolation.
2.
One of the following shall be provided on the discharge side of the RHR system to isolate it from the RCS:
(a) The valves, position indicators, and interlocks described in item 1 (a) - (c).
(b) One or more check valves in series with a normally closed power-operated valve.
The powee-operated valve position shall be indicated in the control room.
If the RHR system discharge line is used for an ECCS function the power-operated valve is to be opened upon receipt of a safety injection signal once the reactor coolant pressure has decreased below the ECCS design pressure.
(c) Three check valves in series, or
f (d) Two check valves in series, provided that there are design provisions to permit periodic testing of the check valves for leak tightness and the testing is performed at least annually.
The isolation on the discharge side of the SCS is identical to that on the suction side, and, as discussed above, adequately meets the present criteria for SCS system isolation.
- 4. 3 Pressure Relief Recuirements The RHR system shall satisfy the pressure relief requirements listed below.
1.
To protect the RHR system against accidental overpressuri:ation when it is in operation (not isolated from the RCS), pressure relief in the RHR system shall be provided with relieving capacity in accordance with the ASME Boiler and Pressure Vessel Code.
The most limiting pressure transient during the plant operating condition when the RHR system is not isolated from the RCS shall be considered when selecting the pressure relieving capacity of the RHR system.
For example, during shutdown cooling in a PWR with no steam cuccle in the pressurizer, inadvertent operation of an additional charging pump or inadvertent opening of an ECCS accumulator valve should be considered in selection of the design bases.
2.
Fluid oischarged through the RHR system pressure relief valves must be collected and contained sucn that a stuck open relief valve will not:
a.
Result in flooding of any safety-related eouipment.
b.
Reduce the capacility of the ECCS below that neeced to mitigate the consequences of a postulated LOCA.
c.
Result in a non-isolable situation in which the water providec to the RCS to maintain the core in a safe condition is discharged outside of the contaif. ment.
3.
If interlocks are provided to automatically ciose the isolation valves when tne RCS pressure exceeds tne RHR system design pressure, adequate relief capacity snall be provided during the time period wnile the valves are closing.
i i.
The SCS relief valves discharge to the reactor building equipment drain tank (RBEDT). Overflow from this tank is dire:ted to the reactor building drain sump. Therefore, a stuck-open relief valve would not result in the flooding of any safety related equipment.
A hign level alarm in the RBEDT would alert the operator to a potential proclem which could be then diagnosed and corrected.
{
i f
At Oyster Creek the SCS is independent of tne ECCS.
Therefore a failure of the SCS would not affect the ECCS.
Since the Shutdown Cooling System is designed for reactor design pressure, the reactor vessel relief valves will provide adequate SCS overpressure protection.
4.4 Pumo Protection Requirements The design and operating procedures of any RHR system shall have provisions to prevent damage to the RHR system pumps due to overheating, cavitation or loss of adequate pump suction fluid.
1 The SCS pumps are provided with bypass lines which return the pump discharge flow to the pump suction.
Thus, even if the downstream valve were closed while the pump was running, the pump would be protected from overheating.
Cavitation protection is provided by the interlock anich trips the pump if the suction pressure falls below 4 psig.
The pumo also trips on a coolant temperature greater than 350*F.
4.5 Test Requirements The isolation valve operability and interlock circuits must be designed so as to permit on line testing when operating in the RHR mode.
Testability shall meet requirements of IEEE Standard 338 and Regulatory Guide 1.22.
(This is discussed in Section 5 of this report.)
The preoperational and initial startup test program shall be in conformance with Regulatory Guide 1.68.
The programs for PWRs shall include tests witn supporting analysis to (a) confirm that adequate mixing of borated water added prior to or during cooldown can be achieved under natural circulation conditions and permit estimation of the times required to achieve such mixing, and (b) confirm that the cooldown under natural circulation conditions can be achieved within the limits specified in tne emergency operating procedures.
Comoarison with performance of previously tested plaats of similar design may be substituted for these tests.
Regulatory Guide 1.68 was not in effect w9en Oyster Creek was designed and built.
However, the licensee committed to preoperational tests to confirm operability and many uses have shown the system to be reliable for removing decay heat.
As part of the startup testing for Oyster Creek, the isolation condensers were placed in operation and the heat removal rates were measured and found to be in excess of design capacity.
Similarly, the relief valve capacities were measured and found to be within tolerance of their design flow rates.
The licensee does not perform tests of SCS isolation feature for reactor coolant temperature greater than 350*F.
The isolation of the SCS due to low-low water level is tested during the refueling outage.
l
- O e
) !
4.6 Ooerstional Procedures The operational procedures for bringing the plant from normal operating power to cold shutdown shall be in conformance with Regulatory Guide 1.33.
For pressurized water reactors, the coerational procedures shall include specific procedures and information for cooldown under natural circulation conditions.
The licensee has procedures to perform safe shutdown operations including shutdown to hot standby, operation at hot standby, hot shutdown, operation at hot shutdown and cold shutdown including long-ters decay heat removal.
The licensee has also provided the operating staff with procedures covering of f-normal and emergency conditions for reactor shutdown and decay heat removal under conditions of loss of systems or parts of system functions normally needed for shutdown and cooling the core.
Drocedures for opera-tion of systems used in safely shutting down the reactor are also included in the plant operating procedures.
These procedures include provisions identified in Regulatory Guide 1.33.
These procedures were reviewed and are in conformance with Regulatory Guide 1.33.
Certain operations were identified to the reviewers which constitute alternate ways and paths to achieve cooling water source alignment or neat sink alignment.
Some of these methods are not included in their procedure system.
I 4.7 Auxiliary Feedwater Sucoly The seismic Category I water supply for the auxiliary feedwater system for a PWR shall have sufficient inventory to cermit coeration at hot shutdown for at least 4 nours, followed by cooldown to the conditions
~
permitting operation of the RFR system.
The inventory needed for cooldown shall be cased on the longest cooldown time needed with either only onsite or only off site power available aith an assumed single failure.
i
)
Oe 36 -
4 i
Boiling Water Reactors such as Cyster Creek do not have an auxiliary feed However, the cooling water inventory requirements for a safe system.
shutdown of the facilit.y, using the systems identified in Section 4.0, i
1 are evaluated in Appendix A.
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5 5
8 4
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I ABLL 4.1 CLASSif ! Call 0N Of SAFE SilulDudN SYSIEM5 OYSlER Ci1EIK Seismic Quality Group Plant Plant R.G.
l.26 Desion R.G.
1.29 Design Remarks Components / Subsystems Except for the reactor ASME 111 ASME Ill Category 1 Class I coolant pressure boundary, isolation Condensers the system boundaries (tube side)
Class I Class C are defined as those ASME II1 ASME VilI (sheiI side) portions of the system Class 2 required to accomplish Iso. Condenser piping &
ASME Ill ASME I the system safety Class 1 1965 function and connected valves piping up to and Core Spray System including the first valve that is either ASME III ASME III normally closed or pumps (4)
Class 2 Class C Cdpdble Of automatic closure when the safety ASA B31.1 Piping and valves in the function is required.
system boundary (See R.G. 1.26) 1 ASME Vill A Containment lorus NucIear Code Cases Automatic DepressurizaL &
Sys tem Reliet valves (5)
ASML 111 ASME I if Class I Supplies condensate f or-Condensate Iransfer Isolation Condensers.
System Class 11 ASME Ill
?
pumps (2)
Class 3 I
1AB1f 4.I (Continued)
Seisinic Qua1ity Groisp Plant Plasit Components / Subsystems R.G.
1.26 Design H.G.
1.29 Design Remarks ASME 111
?
Category I Class il Piping and valves
?
Condensate Storage tank Leergeucy Service Water ASHL III
?
pawnps (4)
Class 3 ASA B31.I Piping and valves heat excliasigers (4)
U
?
(sliell side)
(tube side)
ASML 111
?
(contaisi spray)
Class 2 t
Containment Sgray System A5ME Ill
?
Class 1 pumps (4)
Class 2 ASA B31.1 Class I Piping and valves I
e a
IhJ J 4.1 (Continued)
Quality Group Seismic Plant Plant Components / Subsystems R.G.
l.26 Design R.G.
1.29 Design Remarks Emergency Power System Category 1 Class 1 Diesel generators N/A DC batteries Distribution lines, switch-9edr Control Dodrds, motor
,I control centers Diesel meclianical auxiliaries ASME 111
?
?
Class 3 Reactor Control & Protection N/A Category 1 Class 1 System Safe Sliutdown System N/A
_ Instrumentation &
Control 1
- '4"M ew
IABtL 4.2 Lli,' Of SAFE SitulDOWN INSTRUMLNIS Component / System Instrument instrument 1ocation Reterence Reactor Recisculation Reactor Vessel level LI & LIIS - Reactor Building DWG 148f712 System (LI 1Al2, llIS RE05-19 (RK01 & RK02)
A&B 11 1A13, L1 RE21A&B)
LI-Control Room (Sf/6f)
Reactor Vessel pressure Pli - Reactor Building Plant Descriptica (Pli 1D45, Pil ID46A&B, (RKOL & RK02)
Manual (PDM)
PR/IR 10 71, PRIA08)
PR/fR, PR - Control Room Isolation Condenser Secondary level (L1 IG06 LI - Reactor Building DWG 148f262 A&B, LI IG01 A&B)
Li - Control Room (2f)
Condensate Iransfer System Pump disciiarge pressure
(
)
Cond. Storage Tank level LT - At base of tank, Site visit east side (LI
)
LI - Control Room Core Spray Sy!, tem CS tlow (IIRV 2/A&B, fl - Reactor Building PDM 11 RV 2/A&B) fl - Control Room (If)
Pressure Suppression Cont. Spray suction IE - Reactor Building PDM Sy:, tem ( f orus) temperature IR - Control Room (IE 1903-40A, 1R IP01) tmergency Service LSW - Cont. Spray D/P dPI - Reactor Building PDM Water System (dPI IPOSA,11, C, D &
dPI - Control Roosi dPI IPO6A,11, C, 0)
Containsuunt Spray Cont. Spray riow II - Reactor Building PDM System (f ilP03A & li, il fl - Control Room IPO4 AA B)
Diesel Generator No. 1 Diesel Gen. output Control Room dnd voltage and curresit Diesel Generator No. 2 DC Power Div. I Voltagit and current, Control Room DWG U-3033-IA dild Div. I - Batteries A & B (8t/9f, 9Xt)
DWG 3028-IIA Site visit UC Power Div. 11 Div. 2 - Battery C
~
- '-~
SAF E SilullWri SYSILMS POWL R SUPPLY.iND lOf AIluti 1ABII 4.3 Locationi Power Sijply Cgiipunent/Sys tem Main trasistornier and condensate 480V MCC 1832 via area (west of turb. build., 23')
Condensate t ratis ter 183 from 41BOV lius ID System pumps Inside Drywell 12S VDC Control Power Electromatic Reliet Valves Reactor Build.
(pumps A, B, C, D (NW, SW Corner Core Spray System A, D - 4160V Bus 1C pumps A, B, C, D B, C - 4160V Bus 10 Rooms)
(booster pumps A, C 51', B, D 23' )
booster pumps A,it, C, D B, C - 460V Substa. 182 A, D - 460V Substa. lA2 Reactor Build. (NE, SE Corner Containment Spray System A, B - 460V Substa. lA2 pump s A, 11, C, D Rooms)
C, D - 460v Substa. 182 Reactor Build. (23' NE, SE) heat excharigers imergency Service Water System Cir. Water intake Structure pimips A, B, C, D A, B - 4160V ilus. IC C, D - 4160V Bus. 10 Diesel Generators No. I arid 2 Diesel Generator Rooms lurbine Build. Mezzatiine Oftsite power or Diesel generator 41bOV Bus IC No. 1 lurbine Build. Mezzanine Oftsite power or Diesel generator 4160V tius 10 No. 2 Of f ice Build. North (23')
4160V Bus IC 460V Substation lA2 Office Build. North (23')
4160V Bus ID 460V Substation IB2 12SV Batteries (A, B, C)
A,il - Battery Room (office build. 35')
C - Enclosure lurbine Build.
Mezzanine
- 5. 0 RE3OLUTION OF SEP TOPICS The SEP topics associated with safe snutdown nave been identified in the INTRODUCTION to this assessment.
The following is a discussion of how t
Oyster Creek meets the safety objectives of these topics.
5.1 Tooic V-10.3 RHR System Reliability The safety objective for this topic is to ensure reliable plant cooldown capability using safety grade equipment subject to the guidelines of 3RP 5.4.7 and BTP RSB 5-1.
The Oyster Creek systems have been compared with these criteria, and the results of these comoarisons are discussed in Section 4.0 of this assessment.
Based on these discussions, we have concluded that the Oyster Creek systems fulfill the teoic safety cojective with the following comments:
1.
The Shutdown Cooling System is not a safety grade system by our definition. However, various ECCS systems, including ADS and core spray, can be utilized to effect reactor cooldown.
2.
Comoonent redundancy and single-failure-proof requirements are not met in the case of the shutdown cooling system, in that failure of either AC powered valve inside containment (system suction or discharge) would result in loss of the system.
However, the ECCS systems would still ce available.
3.
Component redundancy (and single-failure proof) requirements are also not met in the case of tne two isolation condensers.
Eacn condenser's disenarge isolation valve inside containment is supplied s
l AC power from the same vital bus.
However, these valves are As noted normally open and fail open on loss of electrical power.
in Section 3, it would take simultaneous scurious isolation of both condensers plus loss of the power supply to create any problem.
Additionally, even if this hignly-unlikely scenario were to occur, the ECCS systems would still be available.
No procedure exists to perform a shutdown and cooldown to cold 4.
conditions with the systems identified in Section 4.0.
The licensee will be required to develop such a procedure.
5.2 Tooic V-11.A Recuirements for Isolation of High and Low Pressure Systems l
The safety objective of this topic is to assure that adequate measures t
are taken to protect low pressure systems connected to the primary system i
J from being subjected to excessive pressure which could cause failures and in some cases potentially cause a LOCA outside of containment.
As noted in Section I, only the shutdown cooling system was examined.
The shutdown cooling system is designed for full reactor pressure but less than full reactor temperature.
Therefore interlocks (with the exceDtion of the pump suction low pressure interlock) are based uoon temperature considerations.
System operation cannot begin until temperature in all five reactor coolant recirculation loops and the shutdown cooling system suction lines is less than 350*F (and pump suction pressure exceeds 4 psig).
This will enable system inlet and outlet valve-and pumo permissive interlocks and
. allow the system to be started.
Additionally, the inlet and outlet valves will shut, isolating the system, if temperature should rise to 350*F when the system is in operation.
Because of the systems full pressure design and the incorporated interlocks (even though they are temperature-based), we consider the applicable requirements to have been met.
Howe'ver, there are no testing requirements for those interlocks.
The need for such requirements will be addressed in the integrated assessment at the completion of tne SEP review.
- 5. 3 Tooic V-ll.3 RHR Interlock Requirements The safety objective of this topic is identical to that of Topic V-11. A.
The staff conclusion regarding the Oyster Creek Shutdown Cooling System valve interlocks, as discussed in Section 5.2, is that adequate interlocks exist.
- 5. 4 Tooic VI!-3 Systems Required For Safe shutdown The Safety objectives of this topic are:
1.
To assure the design adequacy of the safe shutdown system to (a) initiate automatically the operation of accrocriate systems, including the reactivity control systems, such that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences or postulated accidents, and (b) initiate the operation of systems and components required to bring the plant to a safe shutdown.
, 2.
To assure that the required systems and equipment, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown are located at appropriata locations outside the control room and have a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.
3.
To assure that only safety grade equipment is required for a plant to bring the reactor coolant system ' rom a high pressure to a low pressure cooling condition.
Safety objective 1(a) will be resolved in the SEP Design Basic Event reviews.
These reviews will determine the acceptability of the plant response, including automatic initiation of safe shutdown related systems, to various Design Basis Events, i.e., accidents and transients.
Cbjective 1(b) relates to availability in tne control.oom of the control and instrumentation systems needed to initiate the operation of the safe shutdown systems and assures that the control and instrumentation systems in the control room are capable of following tne plant shutdown from its initiation to its conclusion at cold shutdown conditions.
The ability of Oyster Creek to fulfill objective 1(b) is~ discussed in the preceding sections of this report.
Based on these discussions, we conclude that safety objective 1(b) is met by the safe shutdown systams at Oyster Creek subject to the findings of related SEP Electrical, Instrumentation, and Control topic reviews.
I
i 46 -
t j
Safety objective 2 would require the capability to shutdown to both hot a
I shutdown and cold shutdown conditions using systems, instrumentation, and l.
controls located outside the control room. An Oyster Creek procedure i
i concerning fire in the control room provides the steps for operation of i
the necessary equipment to shut the plant down, initiate the isolation condenser, and monitor necessary parameters.
It does not include sDecific steps for achieving cold shutdown conditions.
The ongoing fire protection review will require the licensee to develop a procedure to achieve cold shutdown conditions from outside the control room.
The adequacy of the safety grade classification of safe shutdcwn systems at Oyster Creek, to show conformance with safety objective 3, will be completed in part under SEP Topic III-1, " Classification of Structures, Components, and Systems (Seismic and Quality)," and in part under the Design Basis Event reviews. Table 4.1 of this report will be used as input to Topic III-1.
3 I
l i
i
6.0 REFERENCES
l
. i i
Staff Discussion of Fifteen Technical Issues listec in Attachment to 1.
November 3,1976 Memorandum from Director, NRR to NRR Staff, 4
NUREG-0138, November 1976.
Letter from USAEC to Jersey Central Power and Light transmittirg 2.
Safety Evaluation Report by Division of Reactor Licensing, December 23, 1978.
Oyster Creek Fire Protection Safety Evaluation Report, Marcn 3, 3.
1978.
Letter from JCP&L to NRC, transmitting 125 VDC electrical 4.
modifications description, April 4,1978.
_m.
i l !'!
l*
APPENDIX A 1
SAFE SHUTOCkN WATER REGUIREMENTS h
~
t
'l Introduction Standard Review Plan (SRP) 5.4.7, " Residual Heat Removal (RHR) System" and Branch Technical Position (BTP) RS8 5-1, Rev. 1, " Design Requirements of the Residual Heat Removal System" are the current criteria used in tne Systematic Evaluation Program (SEP) evaluation of systems required for safe shutdown.
l BTP RSB 5-1 Section A.4 states that the safe shutdown system shall be capable of bringing the reactor to a cold shutdown condition, with only offsite or onsite power available, within a reasonable period of time following shutdown, assuming the most limiting single failure.
STP RSB 5-1 Section G, which applies specifically to the amount of auxiliary feed system (AFS) water of a pressurized water reactor available for steam generator feeding, requires the seismic Category I water supply for the AFS to have suf ficient inventory to germit operation at hot shutdown for at least four hours, followed by cooldown to the conditions permitting operation of the RHR system.
The inventory P
needed for cooldown shall be based on the longest cooldown time needed with either only onsite or only offsite power availacle with an assumed single failure. A reasonable period of time to achieve cold shutdown conditions, as stated in SRP 5.4.7 Section III.5, is 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. For a reactor plant cooldown, the transfer of heat from the plant to the envfrons is accomplished by using water as the heat transfer medium.
Two modes of heat removal are available.
The first mode involves the use of reactor plant neat to boil water with the resulting steam vented to the atmosphere.
The water for this process is typically demineralized, " pure" water stored onsite and, tnerefore, is
- A available only in limited quantities.
The systems designed to use tnis type of heat removal process (boiloff) are the steam generators for a pressurized water reactor (PWR) or the emergenc'y (isolation) condenser for a boiling water reactor (BWR).
The second heat removal mode involves the use of power operated relief valves to remove heat in the form of steam energy directly from the reactor coolant system.
Since it is not acceptable to vent the reactor coolant system directly to the atmosphere following certain accidents, the steam is typically vented to the containment building from where it is removed by containment heat removal systems.
The containment heat removal systems are in turn cooled by a cooling water system which transfers the heat to an ultimate heat sink - usually a river, lake, or ocean. When using the blowdown mode, reactor coolant system makeup water must be continuously sue 71ied to keep the reactor core covered with coolant as blowdown reduces the coolant inventory.
Systems employing the blowdown heat removal mode have been designed into or backfitted onto most BWR's.
The efficacy of the blowdown mode for PWR's has received increased staff attention since the Three Mile Island Unit 2 accident in March 1979.
Additional studies of the viability of this mnde for PWR's are in progress or planned.
This evaluation of cooling water requirements for safe shutdown (and cooldown) is based on the use of the system identified in the SEP Review of Safe Shutdown Systems which has been completed for each SEP facility.
The Review of Safe Shutdown Systems used SRP 5.4.7 and BTP R$8 5-1 as a review basis.
It should be noted that the SEP Design Basis Events (DBE) reviews, wnich are currently in progress, may require the use of systems other than those which are evaluated in this report for reactor plant shutdcwn and cooldown.
In
{
)
t.
- A f I'
i those cases, the water requirements for safe shutdown will have to be evaluated using the assumptions of the DBE review.
Discussion The requirement that a plant achieve cold shutdown conditions within approximately 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, as proffered in BTP RS8 5-1 and SRP 5.4.7, is based mainly on the fact that the amount of onsite-stored water for the AFS of a PWR is limited, and it is desirable to be able to place the RHR system in operation and transfer the plant heat to an ultimate heat sink prior to the exhaustion of the onsite-stored pure water supply.
Remaining in a not snutdown condition, with reactor coolant system temperature and pressure in excess of RHR initiation limits, requires the continued expenditure of pure water via the boiloff mode to remove reactor core decay heat.
A BWR relying on the emergency condenser system for cooldown would also be susceptible to the potential exhaustion of orsite-stored pure water.
Should the onsite-stored water supply at a plant be expended, the capacility usually exists to use raw water from a river, lake, or ocean, for examole, to supply the boiloff systems.
However, use of raw water can lead to the degradation, through corrosion, of the boiloff system materials, i.e.,
steam and emergency condenser tuces.
This degradation can occur rapidly generat v
even if fresh water makeup is used.
If seawate,r were used, chlorice stress corrosion cracking of the tubes could occur well within one week
- If raw fresh water were used, caustic stress corrosion cracking of tuce materials could occur in less than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for both stainiess steel and inc:nel tuce materials through NaOH concantration.* A plant coolcown and depressurization
- "vanRooyen, Daniel and. Martin W. <endig, ' Impure Water in Steam 3enerators and Isolation Generators.' 3NL-NURFG-28147, Informal Recort. June 19c0.
t
- A would help reduce the rate of tube cracking by reducing the stresses in the tube materials.
Also, the leakage rate of reactor coolant thrcugh potential cracks in the tubes would be reduced if the plant were in a cool, depressurzied state.
The original design criteria for the SEP facilities did not require the ability to achieve cold shutdown conditions.
For these plants, and for the majority of operating plants, safe shutdown was defined as hot shutdown.
Therefore, the design of the systems used to achieve cold shutdown condition was determined by the reactor plant vendor and was not based on any safety Our safe shutdown reviews have pointed out a difference in the concern.
vendor approach to system design for cold shutdown.
This difference is reflected in the Standard Technical Specification definition of cold shutdown.
For a BWR, cold shutdown requires reactor coolant temperature to be 1212 degrees Farenheit.
For a PWR, cold shutdown requires reactor coolant temperature to me 1200 degrees Farenheit.
These differences in cold shutdown temperatures require the use of additional systems to achieve cold shutdown for a PWR over and above the systems needed for a BWR.
For example, a SWR could use an isolation condenser alone to reach 212 degrees Farenheit (although the approacn to 212 degrees Farenheit would be asymptotic); but a PWR, in addition to the steam generators, must use an RHR and suoporting systems to get below 200 degrees Farenheit.
Evaluation Table 1 provides plant specific data and assumptions used in tne staff calculation of safe shutdown water requirements for tne Oyster Creek nuclear
- A [
r 9
1 plant.
Table 2 provides the results of the calculation.
The temoerature profile for the cooldown is shown in Figure 1.
After the reactor trip, the reactor system pressare and temperature increase to J
l the relief valve pressure setpoint because tne main condenser is not operable l
following an assumed loss of offsite power.
The emergency condensers are t
automatically initiated af ter reactor pressure exceeds 1060 psig for 15 seconds; however, one of the condensers is assumed to be inoperable because this single failure assumotion results in the longest cooldown time and is most limiting from tne standpoint of pure water consumption.
The operator is assumed to maintain reactor system pressure near normal operating pressures, by cycling one of the emergency condenser condensate valves, for a period of four hours prior to commencing the cooldown.
The four hour delay is based on BT? RSS 5-1 Section G and again is intended to maximize pure water consumption.
Emergency condenser pure water makeup is supplied by the condensate transfer system; the
'avel of makeuc water in the emergency condensers is controlled by the control room operator by means of water level transmitters and remotely controlled, ai r-operated makeup valves.
The cooldown data are presented in Table 2 (and on Figure 1).
Since the plant compressed air systems are not on the safe shutdown system list, control of the emergency condenser makeup valves sculd be accomplished by manual operation of the handwneels on the valves.
As the cooldown progresses, the reactor system fluid contracts and the need for reactor system makeup exists to keep the level of coolant above the fuel in the reactor core.
The reactor feed system, which is normally used to inject water into the reactor system at nigh pressures (greater than 235 psig),
i 3
.A6-is not available because it depends on offsite power.
The Control Rod Drive is not (CRD) hydraulic system, which can also supply hign pressure water, considered to be available because it was not designed as a safe shutdown (safety) system and, therefere, is not included on the safe shutdown system list for Oyster Creek. Without these high pressure reactor makeup systems, the operator must rely on the core spray (CS) system to supply reactor coolant.
The CS system is a low pressure system (shutoff head 285 psig); and, therefore, if reactor pressure is not below 285 psig, the operator must initiate the Automatic Depressurization System (ADS) to lower the pressure sufficiently for CS flow into the reactor system to occur.
In fact, the ADS can be manually initiated at any time during the cooldown sequence following reactor trip, provided the reactor vessel level at ADS initiation is at or above the tripic-low level (4'8" above the core); and the CS system will provide adequate core cooling. Thus, the A05 and emergency condensers are redundant to each other for the function of plant cooldown.
The main reason
- nat the emergency condensers are included on the safe chutdown list is to provide a core cooling method which does not reduce the reactor system coolant inventory since Oyster Creek does not have the high pressure coolant injection capability that most other boiling water reactors have.
In the course of the cooldown, the operator must eventually use the ADS, CS, and containment heat removal systems (containment spray and emergency service water) for long term cooling of the plant.
The core heat and stored heat in the reactor system materials is transferred to the containment by the CS and i
ADS.
The containment heat removal systems transfer the heat to the ultimate heat sink.
Normally, long term heat removal would be accomolished by the
- A Shutdown Cooling System (SCS).
If this system and its auxiliary systems are available, it could be started at a reactor system temperature below 350 F (approx. 8.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> on Table 2).
However, since the SCS and its auxiliaries were not designed and constructed with the quality of the plant safety J
systems, the ADS, CS, and containment cooling systems are relied on for long I
term cooling of the plant.
i The quantity of makeup water consumed by the operaole emergency condenser is a function of the time at which the long term cooling neat removal mode begins transferring heat to the ultimate heat sink.
The condensate storage tank contains 2,085,000 lbs of water. At the end of the four hour delay period before cooldown starts, 334,000 lbs. of water would be expended.
To cooldown to the conditions required for CS initiation (285 psig), 750,000 lbs. of condensate would be expended.
And, if the SCS were available, it could be
- tarted at a reactor system temperature of 350*F af ter 955,000 lbs. of censate were consumed.
Based on our calculations, sufficient emergency condenser makeup inventory capacity is available in the condensate storage tank to conduct a plant cooldown in accordance with BTP RS8 5-1.
However, the Oyster Creek plant tecnnical specifications should be modified to require the plant operators to maintain suf ficient condensate storage tank inventory to conduct the cooldcwn (approximately 750,000 lbs) in addition to the inventory requirements for the emergency core cooling systems.
5 f
- A 1 I
i I
TABLE 1 1
Plant: Oyster Creek Power (MW) 1930 Normal Operating Temp. (*F):
547 Safety valve lift (psig):
1070 Initial secondary inventory (1bm):
92240 (in each of 2 emergency condensers)
Secondary makeup water temo. (*F):
80 5/RV flow area (ft 2): NA Emerg. Condenser total ht. xfer coeff. (STU/hr-F):
6.1E5 at 1070 psig Stored sensible heat (BTV/*F):
fuel -
27,000 metal -
224,000 water - 1,540,000 RHR Parameters:
Not applicable j
Pure water onsite (1bm):
2,085,000 in the Condensate Storage Tank
- 250,200 in the Demineralized Water Storage Tank" C cldown assumptions:
1.
At t=0 reactor trips.
2.
Decay heat is in accordance with proposed ANS 5.1 (1973).
3.
Plant remains at hot shutdown for four hrs. prior to coolcown.
4 The secondary (steam generator or emerg. condenser) is considered dry when the initial secondary inventory is boiled away.
5.
Emergency condenser total heat transfer coefficient is as sumed to be constant.
.These quantities are not included in plant technical specifications.
- A,
TABLE 2 Plant: Oyster Creek Phase I (reactor trip to safety lift):
Time to safety valve lift (sec):
approx. O Phase II (safety valve lift to cooldown start):
Time to boil secondary dry, assume no makeuc (min):
40 (for one isolation condenser)
Decay heat generated prior to cooldown start (STU):
324E6 Feedwater expended prior to cooldown start (1bm):
334,000 Phase III (cooldown):
(1 emergency condenser)
Time (hrs)
Temperature (*F)
Pressure (psia)
Cecay heat cenerated (BTU) 4 553 1035 324E6 5
478 563 380E6 6
425 327 443E6 6.5 403 258 459E6 8
357 147 531E6 8.5 345 129 554E6 10 320 94 620E6 12 299 79 703E6 24 267 55 1170E6 4
t
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1 600 FIGUHL 1 REACTOR SYSlEM TEMPERATURE-VS 11ME i
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4 44.
O 400
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taa n:c SCS initiation temperature l
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300 I
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S to 15 20 25 l
TIME (HOURS)
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BACXGROUND
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INFCRMAL REPOR.
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!hM 3i IMPURE WATER IN STEAM GENERATCRS A)
AND ISCLATION GE?iERATORS
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l Daniel van Rooyen anc Martin W..<end i g 9
gll el June 1980 A,
k Corrosion Science Group Ceoartment of ?luclear Energy Breckna en ita:icnal La:cra:Ory l
I Uoton, New York 11973 F
S r I?minary infor"'ation and yas crecarec NOT;r.
' 713 documer- ~
3j orimarily for in eri[,J5e.
since t ay ':e s.+<g+.
% 31:n or nculd no: ce ;j;gc e
n.
Correction 3ne,- 5 not as re -
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DUPLICATE DOCUMENT l
Entire document previously
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entered into system under:
l S 56M7hd%
l ANO No. of pages: J6 gW
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