ML20003B235
| ML20003B235 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 01/30/1981 |
| From: | Baynard P FLORIDA POWER CORP. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.1.1, TASK-1.C.1, TASK-2.B.1, TASK-2.K.3.02, TASK-2.K.3.07, TASK-2.K.3.13, TASK-2.K.3.30, TASK-3.D.3.4, TASK-TM 3-11-15, NUDOCS 8102100538 | |
| Download: ML20003B235 (54) | |
Text
__
__g u
?!
\\
~
ic
,,7
~
!/.
Florida r=
s
=d Power S
m m
January 30, 1981 File:
3-3-30 Mr. Darrell G. Eisenhut Director Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Subject:
Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 hUREG-0727, Post-TMI Requirements
Reference:
Florida Power Corporation (FPC) Letter, 3aynard to Eisenhut; December 15, 1980
Dear Mr. Eisenhut:
The referenced Florida Power Corporation (FPC) letter provided proposed implementation and licensing submittal schedules as requested by your letter of October 31, 1980.
By this letter FPC provides those items of the referenced FPC letter which indicated a January 31, 1981, schedule l
for submittal.
Item I.A.1.1 Shift Technical Advisor In compliance with the short-term requirements of NUREG-0578 and the subsequent clarification dated October 30, 1979, FPC is presently uti-lizing a group of interim STAS.
The qualifications, training, duties and shif t rotation of the interim STAS have been accepted by the NRC Staff (see your May 5,1980, " Evaluation of NUREG-0578 Category A Imple-mentation"; Reid to Hancock).
l As was delineL:ted in our letter of December 31, 1980, (Baynard to j
Eisenhut), FPC nas developed and is presently implementing a program for permanent STA training based upon the document included in NUREG-073' (INP0 Guidelines, Rev. O, April 18,1980).
This training program is be-ing conducted utilizing the University of Florida, Nuclear Engineering Department; NUS; B&W and FPC.
The permanent STAS are expected to re-place the present group of interim STAS by December 31, 1981.
902100 $
General Office 3201 Thirty fourth Street South P O Box 1404;, st Petersburg. Fionda 33733 813 - 866 5151
4 v
Mr. Darrell G. Eisenhut Page Two January 30, 1981 I
As committed by FPC (FPC letter; Baynard to Eisenhut; dated December 31, 1980) to address the requirements of this item, FPC has enclosed five (5) copies of its " Nuclear Operations Technical Advisor Training Program" for your review.
FPC is meeting the recommendations of the INP0 guidelines and in some cases have exceeded these (for example, in the area of STA qualifications). The STAS will be requalified vis-a-vis the present SR0 requalification program.
FPC has considered three alternatives for the long-term STA utilization or phase-out.
The following are offered to aid future guidance in this area and should not be interpreted as a commitment:
Continue the STA program indefinitely for the life of the plant.
To satisfy the long-term requirement, require a Bachelors Degree for the Shift Supervisor, phase-out the STAS, and utilize in this position.
Phase-out the STAS, and utilize these personnel in a supervisory position on the operating staff (this does not include the Shift Supervisor position).
Item I.C.1 - Short-Term Accident and Procedures Review The Abnormal Transient Operating Guidelines (AT0G) Program of the Babcock & Wilcox Owners Group was discussed with the NRC Staff on December 16, 1980.
The draft ANO-1 operator guidelines which had been provided to the Staff are representative of the guidelines which are in preparation for Crystal River Unit 3.
In order to facilitate confirmation by the Staff that all plant specific guidelines are essentially identical, FPC will provide the draft Crystal River Unit 3 guidelines when available.
Item II.B.1 - Reactor Coolant System Vents As committed in our December 15, 1980, letter (Baynard to Eisenhut),
five (5) copies of our position report on venting the reactor vessel head are herein.provided for your review.
This report concludes the reactor vessel head vent is not necessary for removal of noncondensible gases in order to establish and maintain natural circulation for long-term core cooling following a small break LOCA. Specifically, should significant quantities of noncondensible gases collect in the reactor vessel head region, natural circulation is still assured as this l
gas does not inhibit loop flow.
The gas which does flow into the hot legs, either during generation of the gas or subsequent cooldown and expansion of the reactor vessel head gas, will be removed by the hot leg vents. Therefore, as shown in our report, a reactor vessel head vent is not needed _ to assure natural circulation and core cooling.
4 e
Mr. Darrell G. Eisenhut Page Three January 30, 1981 Item II.K.3.2 - Report on PORY Failures The attached generic report (" Report on Power - Operated Relief Valve Opening Probability and Justification for Present System and Setpoints",
Babcock and Wilcox Document No. 12-1122779 - Rev.1) was prepared at the request of FPC and other operating plants with Babcock and Wilcox de-signed reactors to address this Action Plan item.
Based on the analy-sis provided therein, and with the existing reactor high pressure trip and PORV pressure setpoints, an automatic block valve closure system is not necessary.
Five (5) copies of this report have been enclosed with this letter for your review.
Itam II.K.3.7 - Evaluation of PORV Opening Probability The report referenced in Item II.K.3.2 above, also addresses this Action Plan item and supports the contention that an automatic block valve clo-sure system (described in Action Plan item II.K.3.1) is not necessary to reduce the probability of a small-break LOCA from a stuck open PORV with the existing reactor high pressure trip and PORV pressure set-points.
Item II.K.2.13 - Thermal-Mechanical Report As requested in your October 31, 1980 letter, and as committed in our December 31, 1980 letter, five (5) copies of a generic thermal-mechani-cal report are enclosed for your review.
This report, BAW-1648; "Ther-mal-Mechanical Report - Effect of HPI on Vessel Integrity for Small Break LOCA Event with Extended Loss of Feedwater", discusses the generic evaluation of the reactor vessel brittle fracture concern during re-covery from a small break LOCA with extended loss of all feedwater.
Based upon our evaluation of the conservatisms assumed and the completed and ongoing efforts to significantly increase the reliability of the emergency feedwater system, FPC contends that Crystal River Unit 3 is t
l safe for continued operation.
Worthy of note in our evaluation are the facts that the Crystal River Unit 3 reactor vessel will not accummulate a radiation exposure equivalent to 3.8 Effective Full Power Years (EFPY) l until at least mid-1983, and the BWST water temperature has historically been maintained above 60 F.
The assumptions made in the attached report are very conservative for Crystal River Unit 3.
t Item II.K.3.30 - Small Break LOCA Methods to Show Compliance with Ap-I pendix K to 10 CFR 50 FPC is evaluating the development of a generic program to address the l
small-break LOCA model concerns identified in the applicable B&O Task l
Force reports.
To fully evaluate these concerns and incorporate the discussions with your staff in a meeting held on December 16, 1980, we l
will require until March 1,1981, to evaluate the forthcoming B&W pro-posal to perform the model modifications and to determine FPC's course of action.
You will be advised by March 1,1981, of our decision on l
4 F
Mr. Darrell G. Eisnehut-Page Four January 30, 1981 this issue.
FPC states the existing small-break LOCA model, as approved i
by the NRC staff, meets the rquirements of 10 CFR 50, Appendix K.
Item III.D.3.4 - Control Room Habitability Requirements A comprehensive habitability re-evaluation of the Crystal River Unit 3 control room has been performed. The results of this evaluation and,
proposed modifications are discussed in our report entitled, " Crystal River Unit 3 Control Room Habitability Requirements." Five (5) copies of this report are hereby provided for your review.
Very truly yours, FLORIDA POWER CCRPORATION d[ G[![
T s
Patsy Y. Baynara Managel Nuclear Support Services Enclosures Legendre(T02)D105-1
+
n e-.
--p w
\\
Position Paper on Reactor Vessel Head Vents
\\
't POSITION PAPER ON REACTOR VESSEL HEAD VENTS 1.
Introduction The NRC has required the installation of vents in the high points of the reactor coolant system, i.e., the hot legs, the pressurizer, and the RV head for the purpose of removing noncondensible gases which may collect in the system in order to enhance satisfactory long-tenn cooling,2 This paper l
presents an assessment of the uses of the vents for achieving long-term cooling and demonstrates that the RV head vents are not necessary.
Section 2 of this report provides a summary of this paper.
Section 3 pro-vides a description of the expected course of events for a small break LOCA.
Included in Section 3 is a description of the types of scenarios necessary to lead to core uncovery and the subsequent generation of noncon-densible gases. Section 4 of the report provides a discussion of the abili-ty of the hot leg vents to remove noncondensibles for a variety of plant conditions.
2.
Summary and Conclusions During a small break transient, depressurization of the primary system can cause gases to accumulate in the RV head and in the upper regions.of the hot legs.
With proper operator action and functional ECCS equipment, only small quantities of noncondensible gases would be generated and the voided por tions of the RCS would be filled with mostly steam.
In order to generate large quantities of noncondensible gases, substantial core uncovery must The potential scenarios necessary to lead to core uncovery require occur.
operator errors and/or multiple failures of systems.
In light of the up grades in operating procedures and equipment following the TMI-2 accident, small break transients leading to the generation of ~.large quantities of non condensible gases are not expected to occur.
l The ability to remove gases, including large quantities of noncondensibles, from the primary system following a small break LOCA has been assessed.
By starting the RC pumps and/or by opening the hot leg vents, gases which may collect in the upper regions of the hot legs can be removed and forced or natural circulation can be established.
There is no need to vent. gases which are trapped within the RV head as they will not prohibit the estab-lishment of natural or forced circulation.
Subsequent plant depressuriza-tion to cold shutdown conditions can be performed, even with a gas bubble in the RV head, without interrupting natural circulation.
Thus, RV head vents are not necessary to maintain natural circulation, and therefore to remove best from the core (i.e., maintain core cooling).
l.
-~
\\
r 3.
Small Break LOCA Response The response of the primary system to a small break differs greatly depending on the break size, its location in the system, operation of the reactor coolant pumps, the number of ECCS trains functioning, and the avai! ability of secondary side cooling.
This section will provide a general discussion of the expected course of a small break LOCA, i.e., a
" normal" small break.
Additionally, types of scenarios necessary for a transient to progress to core uncovery and substantial noncondensible gas generation, i.e., inadequate core cooling situation, are also discussed.
3.1
" Normal" Small Break System response to a small break can generally be characterized as follows:
- a. Breaks small enough to be mitigated by the makeup system:
for these breaks, if emergency feedwater (EFW) remains available, the primary system loops will remain full of coolant and the operator can initiate a normal plant cooldown.
- b. Breaks which result in automatic initiation of the HPI pumps and ~ are within the capability of the high pressure injection system without resulting in an interruption of primary system flow:
the primary sys-tem pressure will be balanced at a value where the coolant outflou through the leak equals the feed rate of the high pressure injection system.
By using the HPI to naintain inventory and subcooling margin, the primary system loop will remain full of coolant and the operator can initiate a normal cooldown.
- c. Breaks which result in automatic initiation of HPI and also results in primary system voiding:
breaks in this category are those which are not large enough to remove the energy added to the primary system fluid by the core decay heat. Steam generator heat removal is requir-ed.
For smaller size breaks within this category, void formation in the hot leg could result in an interruption of natural circulation and a system pressure increase would occur.
This repressurization would be terminated when the primary side liquid level falls below the ele-vation of the EFW injection nozzles at which time steam in the primary system would be condensed.
For larger sized breaks within this category, the transition between the loss of natural circulation and the establishment. of condensation heat removal would be a smooth process and primary system repressuriz-ation would not occur.
Analyses of small break LOCA transient response have been performed andt are reported in references 3, 4, 5 & 6.
As shown by these analyses, cladding temperatures will be maintained below 1100 F and no cladding ruptures nor significant metal water reaction will take-place unless multiple equipment failures occur.
Under these situations, large quantities.of noncondensible gases are not expected to be produced.
t
(,
4 i
3.2 Inadequate Core Cooling Inadequate core cooling has been defined as situations which lead to core uncovery and excessive cladding temperatures. To aid the opera-tor in minimizing the consequences of such an event, inadequate core cooling guidelines have been developed.7,8,9 As described pre-viously, the small break analyses which have been performed do not predict large core uncovery nor excess lives cladding ' temperatures.
The types of scenarios which potentially could lead to inadequate i
core cooling situations are described below.
Small break LOCA mitigation relies upon the availability of the ECCS equipment.
Lack of HPI due to operator intervention is a potential scenario which could lead to inadequate core cooling.
This is not expected to occur due to the upgrade in emergency procedures for handling small break LOCA's.
Total failure of the ECCS can also lead to inadequate core cooling.
In light of the fact that the ECCS is a saf aty-grade system and is redundant, operable with both on and offsite power, seismically and environmentally qualified, total failure is not expected.
Addition-ally, the inadequate core cooling guidelines provide backup means for handling the consequences of multiple failures in order to further minimize the potential for the generation of large amounts of noncon-2 densible gases.
Mutiple equipment failures can also result in core uncovery.
For ex-ample, a total and extended loss of feedwater, main and emergency, and a single failure in the HPI system may lead to core uncovery for-certain break sizes and locations.
Since the HPI is a separate sys-tem from the feedwater systems, this combination of multiple failures would not be expected to occur simultaneously. Additionally, efforts are underway to upgrade the EFW systems to increase this reliability.
Failure of the operator to promptly trip the reactor coolant pumps-during a LOCA could also lead to an inadequate core cooling situa-tion.
As shown in reference 10, loss of the reactor ' coolant pumps-during a small break LOCA at a time where the primary system has reached a high void fraction could result in core uncovery and exces-c l
sive cladding temperatures.
However, as'shown in reference 10, the peak cladding temperatures would not. violate the criteria of 10CFR50.46, and substantial quantities of noncondensible gases are not expected to be-generated if realistic assumptions are used.
Additionally, operator' training has been conducted and procedural l
modifications-have been 'made to require prompt tripping of the RC pumps for a.small break LOCA, thereby minimizing the potential for this sequence of events.
As described above, there~are scenarios which could leac to an inade-quate. core cooling situation.- the above discussion was r.ot developed to be a complete set, but. rather to provide a general description of l
l l
l the possible scenarios.
Based on the analyses which have been performed and the discussions above, it is readily apparent that operators errors and/or multiple equipment failures are required before an inadequate core cooling situation could occur.
Several actions have been taken to minimize the potential for the occurrence of an inadequate core cooling situation.
These include:
- a. The designing of a reliable ECCS system such that small break LOCA's would be mitigated without excessive cladding temperatures occurring.
- b. Upgrading of the small break LOCA procedures and additional operator training.
- c. Incorporation of inadequate core cooling guidelines into the small break operating procedures, thereby minimizing the consequences of core uncovery should it occur.
- d. Upgrading of the EFW system to provide additional assurance that steam generator heat removal will be available when required.
In light of these actions, if a small break LOCA should occur, the transient is not expected to result in core damage or generation of large amounts of noncondensible gases.
4.
Assessment of Vent Usages Depressurization of the primary system during a small break transient can cause gases to accumulate within the RV head and in the upper regions of the hot legs.
During a " normal" small break, only small quantities of non-condensible gases would be generated.
This section will discuss ways in which the hot leg vents may be utilized in order to remove gases from the primary system to allow the reestablishment of natural circulation in the event natural circulation is lost.
Additionally, possible methods to depressurize the plant with a gas bubble trapped within the RV head are l
described.
1 4.1 Vent Usages for " Normal" Small Break The need to vent gases from the primary system during a " normal" small break is limited to the class of breaks wherein energy removal via the break itself is insufficient to allow plant depressurization and which also leads to void formation in the primary system.
These breaks are the Category C breaks de-fined in Section 3.1.
The subsequent paragraphs describe how gases can be removed from the primary system for various combinations of equipment avail-ability and demonstrate the RV head vents are not necessary.
4.1.1 RC Pumps, EFW and HPI Available The plant response for Category C breaks with EFW available can generally be l
characterized as follows. The initial system depressurization will result in I
a reactor trip and low pressure ESFAS initiation.
Operator action would then.
be taken to trip the operating RC pumps.
The continued system depressuriza-tion will ultimately result in saturated fluid conditions in the RCS while the ' upper regions of the hot legs and the reactor vessel will void.
Natural !
1 circulation will be lost if the void in the hot leg becomes sufficient to fill the 180* U-bend.
With the loss of circulation through the loops, the steam generator energy removal will be Icst and a primary system repressuriz-ation would commence.
Once sufficient primary system inventory is lost to cause the primary side liquid level to decrease below the elevation of the EFW injection nozzles, condensation of steam in the primary system would occur thereby initiating " boiler-condensor" circulation.
As a consequence, the primary system pressure will decrease and the HPI flow will increase to establish a stable core cooling node.
Later in the transient, a system re-fill will commence.
Mswever, a gas bubble will remain trapped in the upper regions of the hot leg and in the RV head.
Once the level in the primary system rises above the EFW injection nozzle elevation, heat removal via the steam generators will once again be icst and the primary system pressure would start to increase.
By operating the RC pumps consistent with the small break guidelines, i.e., by " bumping" or starting the RC pump, voids in the upper regions of the hot leg will be swept out and forced circulation will be established.
It is important to note that any voids present in the RV head at this time would not prohibit the reestablishment of circulation within the primary system.
With forced circulation established in the primary system, the operator can then turn his attention to removal of the bubble in the RV head.
By spraying into the pressurizer, and then using the pressurizer heaters to strip the gas from solution, the gases will collect in the pressurizer.
Ultimately the gases can be vented by using the pressurizer vent.
Following the removal of the bubble from the RV head, the operator can then initiate a plant cooldown.
4.1.2 EFW and HPI Available The initial primary system response for this case would be essentially the same as the described in Section 4.1.1 up to the point of using the RC pumps to sweep out the bubble in the hot legs. Without the RC pumps available, the hot leg vents can be opened to remove the trapped gases in the upper regions of the hot leg.
Due to the continued injection of HPI, the liquid level would rise in the system and natural circulation would be restored. As noted before, gases trapped within the.<V head will not prohibit the restoration of natural circulation.
A concern existed about the ability to depressurize the plant to cold shut-down conditions with a bubble trapped within the RV head.
As the plant is depressurized, there was concern that expansion of the gas bubble from the RV head into the hot legs may cause an interruption of natural circulation.
As described below, this problem has been examined and it has been concluded that plant depressurization could be performed without RV head vents, while-maintaining nctural circulation.
For non-pressurizer breaks, the primary system could be depressurized at a controlled rate using the pressurizer vent with the SGs beng utilized to maintain subcooled conditions within the primary _ system..
The plant depressurization would allow the bubble in the RV head to expand into the hot leg.
The depressurization rate would be controlled such that the volume of gas expanded into the hot legs is less than or equal to the volume of gas that could be removed by the open hot leg vents.
In this man-ner, no significant gas accumulation will occur within the hot legs and natural circulation would be continuously maintained.
For breaks in the pressurizer, the pressurizer would remain full and the sys-tem pressure would stabilize at a value at which the injected ECC fluid mat-ches the leak flow.
The system could be depressurized at a controlled rate, as described above, by throttling the HPI while maintaining an adequate sub-cooling margin within the primary system.
As described above, the gases ex-panded into the hot legs from the RV head would be bled out of the system through the open hot leg vents.
The methods described above allow for some of the gases trapped in the RV head to be expelled through the hot leg vents.
However, they do not result in the complete elimination of the RV head gas bubble.
Therefore, long term arrangements are needed for the removal of tha reactor vessel bubble.
This can be accomplished by many means, a few of which are:
1.
Allow the reactor vessel head to cooldown naturally and condense the bubble.
2.
Manual operation of the control rod drive mechanism (CRDM) to vent the noncondensible gases.
3.
Use the letdown line to bleed fluid from the RCS, while providing makeup fluid to the system.
This process will result in a decrease in the non-condensible concentration within the fluid thereby causing some of the trapped noncondensible gases in the RV head to dissolve in the coolant.
Since the bubbles in the RV head do not prohibit natural circulation nor operation of the decay heat removal system, longer term actions are accept-able.
For contrast, an assessment of operator actions with RV head vents has been performed.
Since a bubble in the RV head does not prohibit reestablishment of natural circulation, only opening of the hot leg vents, as described pre-viously, should be performed in order to reestablish natural circulation as soon as possible.
Once natural circulation has been established and the ori-mary system has regained subcooled margin, the operator could close the hot leg vents, open the RV head vents, and throttle HPI to raintain system pres-sure.
In this_ manner, gases in the RV head can be expelled while the water level increases in the vessel.
If the-HPI is insufficient to maintain system pressure during this operation, the RV head vent should be closed with the system subcooling deecreases to y 50 F.
The primary system should then be repressurized to ~ 1000 F subcooling and.the procedure' repeated until the bubble in the RV head is' eliminated as indicated by a pressure increase in the system.
Based on hand calculations for a.-187 inch ID vent, the
~
P expulsion of a bubble which completely fills the RV head would take at least 30 minutes for a pure hydrogen bubble and three hours for a steam bubble.
Followir.g the return of a completely solid primary system, with the exception of the pressurizer, the operator can then initiate a normal plant cooldown while using the HPI to maintain subcooling margin.
4.1.3 Just HPI Operating The plant response for Category C break sizes without EFW available is some-what different than described previously.
The system will initially depres-surize and initiate a reactor trip.
The system depressurization would con-tinue until the steam generator inventory is boiled-off and then a system re-pressurization would occur.
Automatic actuation of the HPI may not occur prior to the generators boiling dry and manual initiation of HPI is requir-ed. The system pressure would ultimately increase to a point where the break alone is capable of removing the energy or it may reach the pressurizer relief and/or safety valve setpoints and be maintained at that pressure.
For these cases use of the hot leg vents to discharge steam could aid in the ultimate depressurization of the system.
Depressurization of the RCS will be controlled by operator throttling of the HPI while maintaining subcooling in the system. A trapped gas bubble in the RV head does not prohibit plant de-pressurization nor will it interrupt natural circulation because-natural cir-culation cannot be established without feedwater to the steam generator.
This mode of operation would continue until the operator can reestablish EFWA at which time he would revert to the actions listed in either Section 4.1.1 or 4.1.2 depending on RC pump status.
4.2 Vent Usages for Inadequate Core Cooling Situations During an inadequate core cooling situation, non-condensible gases may be re-leased due to cladding rupture or metal-water reaction.
Since the steam generators are utilized to depressurize the primary system and lead to subse-quent actuation of the core flooding and/or low pressure injection systems, non-condensible gas concentrations within the steam generator should be mini-mized.
It is expected that some of the non-condensible gases which may be generated collect within the RV head while the remainder will flow towards the steam generators.
The trapped gases within the RV head need not be vent-ed at this time as they do not interface with the SG heat removal.
The gases which flow towards the SG can be removed by opening of -the high point vents thereby minimizing the concentration of 'non-condensibles which reach the SG.
Following core recovesty, the operation of the hot leg vents returns to the
" normal" small break guidelines, described in Section 4.1.
Mardis(Vents)D105-1.
i i
REFERENCES 1.
NUREG-0578, "Three Mile Island lessons Learned Task Force Status Report and Short Term Recommendation," July, 1979.
2.
Letter, H.R. Denton (NRC), " Resumption of Licensing Reviews for Nuclear Power Plants," August 20, 1979.
3.
BAW-10103A, Rev. 3 "ECCS Analysis of B&W's 177 FA lowered-loop NSS,"
July, 1977.
4.
Letter, J.H. Taylor (B&W) to S.A. Varga (NRC), July 18, 1978.
5.
BAW-10075A, Rev.1 "Multinode Analysis of Sraall Breaks for B&W's 177-Fuel-Assembly Nuclear Plants with Raised Loop Arrangement and Internal Vent Valves," March, 1976.
6.
" Evaluation of Transient Behavior and Small Reactor Coolants Systsem Breaks in the 177 Fuel Assembly Plant," Babcock & Wilcox, May 7, 1979.
7.
69-1106001-00 "Small Break Operting Guidelines for Oconee 1, 2, and 3; Three Mile 1 and 2; Crystal River 3; and Rancho Seco 1," Babcock &
Wilcox, November, 1979.
8.
69-1106002-00 "Small Break Operating Guidelines for Arkansas Nuclear 1," Babcock & Wilcox.
9.
69-1106003-01 "Small Break Operating Guidelines for Davis Besse 1,"
Babcock & Wilcox.
10.
Letter, S.H. Duerson (B&W)_ to D.G. Slear (GPU) " Task 48, Task 62 and Task 63," GPd-80-0480, October 21, 1980.
Mardis(Vents)DN105-1
!!UCLEAD CPEffATICf!S TECl'IliCAL t.LVI.':CR TRAll!II:C PRCCRA!'
frepored by: FLCRICA PC',IR CChPCBATICl!
Eubmitted: January 31, 19P1
l t
1:UCLEAR OPEPATICSS TEGI ICAL ADVI50R TPAlt;I!!G PROGFJd!
GCAL:
This program is designcd to train Cperations Technical Advisors able to pron.ptly assess a cc, lex array of indications and alarms associated with any of f-norral cperating events or accidents and immediately analyze the necessary actions to terminate or mitigate the consequences of th ' event. The Operations Technical Advisor must have the technical and analyti~ul capability to recognize and react to a wide range of off-normal situations including multiple equipment f ail ure s, complex transient response, inadequate core coc3ing, and operator errors.
His correct diagnosis and recovery recommendations will drastically reduce the likelihood of an event similar to Three Mile Island Unit 2.
OBJECTIVES :
..(#5 Upon completion of this program the Nuclear Operations Technical Advisor must be capable of the following:
1.
Evaluate the operating history of the plant (equipment failures, design problems, operator errors, etc.) and initiate corrective measures to prevent rec urren ce.
2.
Review Licensee Event Reports from other nuclear plants similar in design l
l to CR-3 and assure suitable dissemination of the evaluation to other plant staff members.
3.
Analyze plant conditions required for maintenance activities and testing to assure adherence to Technical Specification requirements.
4.
Develop and revise procedures to comply with state and federal regulations and to increase plant safety and efficiency.
. 5.
Evaluate the adecuacy of operating Procedures to assare safe, continuous, normal cperations and of Abnormal and Emergency procedures for mitigatien capabilities during of f-normal plant operations cr accidents.
6.
Review the rcccmmendations of shift operating personnel and initiate modifications to increase plant safety and efficiency.
7.
Provide diagnostic capabilities during of f-normal cperating events and advise the Shif t Superviser of actions necessary to mitigate or terminate the event.
8.
Provide overall coordination of maintenance and refueling activities during planned and forced cutages to minindre unit downtime and reduce man hours expended on outage tasks.
9.
Prepare special reports in response to quality Programs Audits, Ccmpliance Audits, NRC Inspections cr othcr cperational reports as directed by the Operations Engineering Supervisor.
10.
Evaluace the continuing adequacy of plant cperations with regard to the assurance of quality and safety of cperation consistent with Regulatory equirements.
11.
Prepares Unusual Operating Events Reports after thorough review of strip chart recordings, computer alarm summary, and annunciator alarm summary and recommends changes to prevent recurrence.
PROGRAM CCSTENT:
In general the program is to be presented in five phases, some running concurrently.
The five phases are listed below as a general course content follcwed by detailed description of the program.
3_
GE: ERAL CONTENT:
Phase I.
College Level Fundamental Education.
Five quarters taught on site of basic academic matcrial required as background for eperations.
Phase II.
Management Supervisory Skills and Administrative Controls.
Provided to bring the OTA's up to speed in Plant Administration and Management of personnel skills.
Phase III.
Plant Systems (Prima ry, Secondary, Instrumentation and Control.
Provided to teach prospective OTA's plant systems, instruments controls and protection systers with which he can analyae plant status and the systems or tools with which to terminate transients er mitigate accidents.
Thase IV.
General Operating Procedures, Emergency Procedures, Transient / Accident Imalysis.
Provided so that the OTA has the capabilities to analyae plant transients and plan his way to get the plant back into a safe configuration.
Phase V.
Simulator Training.
The culmination of the other phases allows the perspective OTA's to put their newly acquired skills to practice therefore driving them home.
DETAILED CONTENT:
Phase I.
College Level Fundamental Education.
STAP FIVE QUARTER COUPSE SEQUENCE 41 CREDIT HOUPS FI.US S HPS. PRACTICAL WORK FU';DA."ENTALS OF PENCIPLES OF NUCLEAR REACTCR S AFETY POWER REACTOR OPERATIONS Sth Qtr.
gPN EMU RXSAF, 6 CR.
10 CR.
EMU 5176L, 4 CR.
A
a n__
+$+
+'+,e
,.e...
<e.1,.
. TEST TARGET (MT-3)
' tu EM s
1.0
= y @ EE I.l l 'e EM
-~
l 1.8 1.25 1.4 1.6
==
j
$ge 4tk /
<$ A
- Sj)$
e.s x
+
0
$+++%
$s't h
i 4
,. e. Ev <e _
. TEST TARGET (MT-3) 1.0 E HH I!M 83 I 5 m !laa m
l,l D bb j l.8
'~
1.25 1.4 1.6 6"
4%
++#b d$tk /
4;,f)[4
, ~ -
_4 l
I.,-.-.-_.-.%
-. n., U -n
-..,.-......n..-.a 0-
.fv a n.3 r...L r2 ui r L.. & n.
r.r O T
,,9
- s. m-
. L 'm..t p *, v.e ;. m..+ C,,,.
k.
p.y
.ns u
---sU.
win w
c...
r,. ; s., c,
u..
- m c c.:,.
t r. n. ~~1,
- a. r r..
- w -
-. c A
/\\
>Uc.en,n ~r, G T -- r e vl - -,
. u.. _r.e.t.n D. ~,, c.... CS,
z:._.n 2
.s 1
+1
...-.m u
- n.,. D ea
. r,c..-.-,,.-
2.,.
.rc n'tr.
.e : v r.i
-..., '. 3 C,,
C.,r..
= ' "
t,,,v, u,, _ _,- -
i n c,, S A
8 CR.
ENU 4134, 4 CR.
.A
..., - = nn-z...7.
r - 7 s, 1 1aaa. n, D c, = n. S.
.v. B.
- s. v u.
.a r-n
.n.. d Q ~-..
c.,. U.,. l v.,,
- .,.., %o052_,
c Cn.
u C,.
t...
n A
y.
.... s. a u.. n., C:
ro. 22 n.A G,.,.
..n 1 n e.a r.:n. - -wiva c.nirn1 ale n.
y, _r..u. v ~. r..o,
,- = _r -= p 3 s., Q ~,..
..u
.._ ~
z,,.,J
,, e n e,
C n.,.
7 Cn,.
.r,.,, : v: u,
, Cn.
v-v M LOCATIC: OF STAP CFIDITS TO PEET INPO ACADEMIC FZQUIFEMENTS (Quarter Syster. Credits)
SU3 JECT AFIA I!??O CFIDIT STAP PROGRAM FIQUIFIMENTS CFIDIT AI. LOCATION Mathe=atics 9
(6)*
Reactor Theory 10 10 as follows:
ENU 4103 (4/4)
E!;U 4104 (3/4)
Et;U 4 60'5 (1/4)
EI;U 4192 (1/4)
ENU F2SAF (1/6)
Reactor Cherdstry 3
3 as follows:
ENU 5176L (2/4)
E:iU FUEL (1/3)
(
Nuclear ::aterials 4
4 as follows:
E:;U 4192 (1/4)
ENU FUEL (2/3)
ENU R.XSAT (1/6)
I
. II:FO CFIDIT STAP PROGRA".
SUSJECT AREA FZQUI RE:E::TS CFIDIT ICLOCATION Therral Sciences 12 8 as follows:**
ENU 4134 (4/4)
ENU 4192 (2/4)
ENU RXS AF (3/6)
Electrical Sciences 6
3 as follows:***
EMU 5176 (1/4)
ENU 4905L (1/4)
ENU 4605 (1/4)
Nuclear Instrumentation and Centrol 4
8 as follows:
ENU 4605 (2/4)
EMU 5176L (1/4)
ENU R):SAF (1/6)
ENU 4104 (1/4)
ENU 4905L (3/4)
Nuclear Radiation Protection and Health Physics 4
ENU 4241 (4/4)
Phase II.
Management Supervisory Skills and Administrative Controls.
A.
Management Supervisory Skills 1.
Leadership 2.
Interpersonnel Communication 3.
Motivation of Personnel Prcblem Analysis 4.
Decisional Analysis 5.
Command Responsibilities and Limits 6.
Stress B.
Administrative Controls 1.
Responsibilities for Safe Operation and chutdown
- f 2.
Equipment Outages and Clearance Procedures
.3 3.
U3e Cf PrCCedures
~.1..+.
..,a.i.c.: C3 1C n.r r u
-. a m..2
.2
.t.
.c w ; s. : p.1.; C_ s,
.u,,.,.-..n.
3
.u.
..3 n-s us
.O wC..;....
e
...e-
..ax.
C.
..w. 3..
..e r..g-p- s.. ;
,. - c.
C.# r 1 m*.*. *.
-s.
~
C*
. o c...
-. a '. u' S 8.
U...
.T.r.. o...e. e w.. +.- r 1 3 ( ~.. 7 * :.
..e t.
.i,. w g w.
k..
n
- c..
c w
r.413
...a, C v,
. a-..., -. C.,
C. '_. O
-n e
L.,,.-....o aux
.c.
..u-m.
s vol b e -,w.. :....,
- o..
- r. u.,;. r
- n.
rv-.. ~ v. v 1 2,-.
- c. L-C. o.c m
ou v
J-.
- C e n n..a.: u.:.3.:s:es C:. u. n.
S r~.
e..
- e 5
- u. a.-
al.
n
~
ex 12 o c.C1 3.#
...a_. c. a..n w y'
- o. l c..
r
~
v-vox
'3.
Code of Federal Regulatiend (a.r. crc.triate secticns)
O
..e.c-4r
.a-t e1 u-....e.w...:
a1 c_e 2
1 4
'.". a" *-.n '~..o..'.
- a' t x C.". a." d CC".'."O l S ). -
-- e C '.7d ":',
nk.a s o.
.T w T.
0 1... *.
-ya c.3 r' t.
.~, v..,
c q..c+
~;
z.
c 1.
Reacter Ccciant System 2.
'dancup and Purification 3.
Decay Heat F.enoval 4.
Chemical Addition and Sarpling 5.
- uclear Service and Decay Heat closed cycle cooling water and raw water svstems 6.
Waste Disposal a.
Gaseous b.
Licuid c.
Solid (PN 7.
S.
Main, auxiliary and reheat steam 9.
Main turbine, Lubcoil and EHC 10.
Condensate, Feedwater, and Emergency Fcedwater 11.
Secendar) Service Cooling 12.
Heater drains and vents 13.
Air Systems a.
Instrument b.
Ecuseservice 14.
Plant vantilation 15.
Fire Service 16.
Nuclear and Non-Nuclear Instrumentaticn 17 Plant cc=puters f*\\
18.
Integrated Centrol System 19.
Reactor protectica system 20.
Engineered safeguards actuaticn system 21.
Energency Diesel Generators 22.
Plant Distribution a.
500 KV to 480 Vac b.
480 V to 120 Vac c.
DC Distributicn and Batteries d.
Protective relaying Phase IV.
General Cperating Procedures, Energency Procedures, Abnormal Procedures, Transient Analysis.
1.
Review of selected normal plant operating precedures, a.
Startup, heatup b.
Pcwer Cperations c.
Plant shutdown and cooldown I
I
_g.
.t.....n y-p ". C V, u'.". d s'.b. w^ '~... 2 1 0 - ^ " ' ' ' * " C.' ^w ^ C # " - e s.
-~
sg..--a.;
3, Trar.s ient/ Acciden t Analysis
- a. c.,. e
. S. c e.s c..-
n.e
,..,..2.,,
.-vo
-c.
o; b.
Plannins and Use of Systens fer Mitigaticn of Accidents c.n s_; c-
.v,e-. ; c. co l s u.u.. e.
-.a
, n
- c.. a m-~
c -
c.
o, Lne c. ne a.x, s
v.
.c u.c _ n_ %,.
.c.....,1, n,.
u,.:.,..:n.
.-e
.u.
y
- = v. 1
^
3.
T...
a.,..,,
.o w..a_ Ct,...s n.
4.-
... - - -... s a.
Role of the Shif t Technical Afviser
- t. w.<., C a.,
.".Y. C.
- " e < b..'.' '. ~ e w"k.....' C.2 '
. n'-..s w
.4 vy m.
~c_ c....' c = 1 n' #_'. 4 e c.-
k.
c.
""^.1-
.eedad kv ' h.a.
o.
=.a. n c ~. o. c ' =.- '. ".~ m.* s - "- a s '. c..
f Review of applicable limits and precauticns a.
b.
Review of technical specifications and their bases inclrding minimum conditions for critical operations c.
Review of Reactor Startup Procedure 3.
Simulator Session a.
Control Eccm familiarication
_o b.
Reactor Startup - all rods in to 10 " amps (repeat as time permits)
Day 2 1.
Plant Heatup and Cooldown Review of plant heatup and cooldewn curves including the basis of the a.
curves and applicable limits and precautions b.
Revicw plant heatup procedures c.
Review plant shutdcwn and cocidewn precedures (PN
_o_
[
2.
Review cf tne Integrated Cent cl 5) :ter a.
Control theory b.
Easic subsy,tems 3.
Simulator Session a.
Reactor startup/ plant heatup - Safety rods out to 15% power including turbine startup b.
Plant shutdown and cocidewn Day 3 1.
Revicw of the Integrated Centrol Syste.-
a.
Unit Load Demand 5
Integrated Master c.
Feedwater Subsystem f'
d.
Reactor Subsysten 2.
Plant Maneuvering Response Using the ICS Operation during a Load Change a.
b.
Manual / Automatic Operations 3.
Sinulator Session a.
Plant maneuver 15% to 100%, 100% to 15%
b.
Manual and Automatic cperation of the ICS stations during plant maneuvers Cay 4 1.
Review of Reactor Physics a.
Subtritical Multiplication and Res,:cnse b.
Critical Respense including Calculations 2.
Reactivity Balance hs a.
Review of reactivity coefficients b.
Balance procedures c.
Problem solving J
i 1 3.
Calculatien of Estinated Critical Position 4.
Simulator Session - Demonstration of Plant Transients a.
Reactor Trip b.
Load Rejection c.
Loss of Normal Feedwater d.
Solid Plant Operation 5.
Loss of Cff-site Power - Demonstratien of Natural Circulation Day 5 1.
Review of Abnormal and Emergen;y Procedures a.
Reactor Trip b.
Loss of Coolant c.
Loss of Feedwater d.
Loss of Feedpump e.
Loss of One or "cre Reactor Coolant Pumps f.
Steam Leaks g.
OTSG Tube Leaks and Tube Ruptures 2.
Simulator Session - Demonstration of Plant Transients a.
Large RCS Leak b.
Intermediate RCS Leak c.
OTSG Tube Rupture d.
Steam Leak e.
Steam Line Rupture Day 6
. 1.
Thermodynamics Review
' f**
i Review of Definitions (1)
Types of Energy
F.
(2)
Types of Systems (3)
Steady State and Equilibrium b.
Propcrties of Water (1)
Phase (2)
Saturation (3)
Superheat (4)
Subcooled (5)
Steam Tables (6)
Latent Heat of Vaporiration (7)
Quality and Void Fraction 2.
Heat Transfer Review a.
First Law of Thernodynanics and General Energy Equation f'
b.
Use of the General Energy Equation to demonstrate the steady state heat balance of the:
(1)
Primary System (2)
Secondary System (3)
OTSG (4)
Pressuriz r (5)
Main Condenser note:
These demonstrations should be developed by the student as guided by the instructor to demonstrate the fundamental concepts introduced above.
Day 7 Heat Transfer and Fluid Flow PN 1.
Discuss heat transfer concepts q
h = $Dp3T a.
b.
h = UAST or HALT (CT5G and fuel assemblics) h=?dh c.
2.
Analyze the following plant transients using thermodynamics and heat transfer concepts (Instructor guided analysis) a.
Less of feedwater with and without emergency feedwater b.
Reactor trip without turbine trip c.
Steam line rupture 3.
Discuss heat transfer mechanisms a.
Forced convection - subcocled b.
Forced convection - nucleate boiling - subcooled f
Forced convection - nucleate boiling - bulk boiling c.
d.
Film boiling e.
Flow profiles (1) Annular flow (2) Slug flow (3) Chug flow (4) Flcw with film boiling (5) Dryout f.
Critical Heat Flux (1) D:!3 (2) Dryout (3) Variation of CHF with O
(a)
Fressure X
00)
Temperature (c)
Flow (d) Location
4 g.
- atural Circulatien (1) 1:echanics of natural circulation (2)
Factors affecting natural circulation
.lant examples demonstrating heat transfer mechanisru 4.
a.
Analyae small break LCCA to demonstrate (Instructor guided analysis)
(1)
Energy 1 css thru the break (2)
Effects cf iiFI (3)
Effects of ECP b.
Analyze effects on natural circulatien due to (1)
RCS saturation cend'.tions (2)
OTSG level (3)
Aux g ncrmal feed path f'
(4) i RCS voiding (5)
RCS temperature Day 8 1.
Detailed Review of ICS a.
Signal flow paths b.
System respcnse to normal cperations c.
Calibrating integrals 2.
Transient responses of the ICS a.
Load changes b.
Reactor trip c.
Rod drop d.
Loss of feedwater (Less of MFP and Icss of all feedwater)
Ps e.
RCP pump trip - RC flow degradation f.
Effects on transients with staticns in manual
3.
Reliability assessment of the IC5 4.
Analyze less of power transients (Instructor guided analysis) a.
Off site b.
ICS/h";I Day 9 1.
Derivation of Cperating Limits 2.
Derivation of EPS Se; points 3.
Safety Analysis Day 10 1.
Large LOCA analysis 2.
Small break LOCA analysis 3.
Inadequate core cccling analysis f
Day 11 1.
Transient Analysis a.
Loss of feedwater with stuck open PCRV (Instructor guided analysis)
(1)
With RCP's running (2)
Without RCP 's running 2.
LOCA Guidelines 3.
Simulator Session Reactor startup including ICS operation a.
b.
Si steam leak Day 12 1.
Discussion of B&W Assessment Reports of Cperating Plant Events and B&W development of Abnornal Transient Operating Guidelines.
P%
2.
Review of selected Operating Plant Events (Instructor guided analysis) a.
Integrated Control System component failure b.
Turbine Throttle Valve Transients c.
Reactor Trip - Overcooling Event
-I'-
Y n,y
,.3
-c, 1.
Analysis cf TP.I-2 Accident (Instructcr guided analysis) 2.
Simulator Sessiens - Pcwcr creration with unannounced casualties a.
Overfeed transient b.
OTSG tube rupture vay 3 1.
Cc.Tplete T!*I-2 Accident Ana2ysis 2.
Transient Analysis of Loss of Feedwater - Ecactcr Trip - Loss of !??;I Fcwcr Event (Instructor guided analysis) 3.
Sinulator Sessicn - Power Cperaticn with unannounced casualties a.
Large stcan break b.
Loss of all feedwr.ter fN Day 15 1.
Shifc Technical Advisor Overvicw a.
Analysis o.? selected plar.t parameters b.
Use cf plant instr umentat ion c.
Limiting plant conditions d.
Evaluation of heat sinks and core cooling 2.
Drill Critiques 3.
Siculater Session - Power Cperaticn with unar.nounced casualties 2
a.
0.01 ft EC lei; b.
Failed pressurizer spray valve
9 M
e ac 5.
N (t
C M
T m
Ca O
C N
w w
ac D
M rt C
M v2 G
C 5
C
.T
~
ee W
e c
v.
o 4
L C
C I
u M
H H
l 3
M H
1 C
H M
{
D c
w A
MM Cu M
ft C
M a
h t
}
D M
C v3 H
c I
o M
et C
M tp
_5 U
M C
- [
4 i
l n
kat k
N i
l REPORT ON POWER-0PERATED RELIEF VALVE OPENING PROBABILITY AND JUSTIFICATION FOR hm PRESENT SYSTEM AND SETPOINTS
>/
- Submitted to Satisfy Requirements of NUREG-0737, Items II.K.3.2 and II.K.3.7 n
Docunent No. 12-1122779 - Rev. 1 70 January 1981 p
"JW t&
e-M Ell W
L i-F-
W k
~
[
y.
n--
r-i e
6 mm ff(
Table of Co ients
.j 1.0 Intrcduction & Summary i
2.0 Discussion 2.1 Evaluation of PORV Opening Probability During C/'
Overpressure Transient 2.2 Evaluation of PORV and Safety Valve Reliability 2.2.1 Safety Valve Failure Rate History L--
s,-
h' 2.2.2 Evaluation of 9nall Break LOCA Probability /Need for PORV Isolation System N
3.0 Conclusions r' -v is e UY?
!*Il
..c l
k
'Y er
- P.1' ssn r
s
,.,A f';*
f 1.0 INTRODUCT Rv AND
SUMMARY
I NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980, required that a report be submitted which provides the information identified in Items II.K.3.2 and II.K.3.7.
Specifically, NUREG-0737 requested the following information/ justifications:
F 1.
II.K.3.2 e Compile operational data regarding pressurizer safety valves to determine safety valve f ailure rates l-
[
e Perform a probability analysis to determine whether the modifica-i tions already implemented have reduced the probability of a small 6
break LOCA due to a stuck-open PORV or safety valve a sufficient i;(
amount to satisfy the criterion (<10-3 per reactor year), or whether the automatic PORV isolation system specified in Task Item II.K.3.1 is necessary.
C; ar'g 2.
II.K.3.7 e Perform an analysis to assure that the frequency of PORV openings h
F.'
is less than 5% of the total nunber of overpressure trant;ents.
This report is submitted in compliance with NUREG-0737 and demonstrates that the W#
requirements of NUREG-0737 are met with the existing Power-0perated Relief Valve ks pF (PORV), Safety Valves and 5"gh Pressure Trip Setpoints and that no automatic QC isolation system is required.
w
- w
['
m
'I 2.1 Evaluation of PORV Opening Probability During an Oeerpressure Transient e
An evaluation of the probability of PORV opening has been performed.
Two separate analyses have been performed.
The first is an analytical estimate, the second is an analysis based upon operating experience.
c
{',
2.1.1 PORV Openir.g Probability Basea Upon Analyses A series of calculations have been completed using best estimate ntnbers f
L to Mi:nate the probability of PORV opening.
Wherever possible, these calculations were based on operating plant data in an attempt to provide F
realistic estimates for the analyzed events.
The following paragraphs I
stamarize the results and calculational basis for the analysis.
f, G
9 The probability of the PORV lif ting during a loss of feedwater (LOFW) or A
turbine trip is approximately 3.9x10-6/Rx-Yr for plants with a PORV setpoint of 2450 psig and 3.9x10-3/Rx-Yr for plants with a PORV setpoint of 2400 psig.
The latter setpoint is presently applicable only to Davis-Besse 1.
These probabilitie; are based on the assumptions that
-a the high pressure trip setpoint is 2300 psig with a standard deviation of 1.4 psi and that the actual setpoint at which reactor trip occurs is a random variable which is normally distributed.
The small standard deviation is based on the fact that the PORV and RPS actuation points are
? '-
g.
not completely independent; i.e., they share a comon source; i.e.,
sensor and instrtment string.
Thus, ther 5 parts of the string errors are h
L perfectly correlated and cancel one another in the analysis.
Other parts p
of the relevant string error are not correlated and it is upon these that L
i the 1.4 psi standard deviations are based.
In a siailar fGshion, the hy -
actual opening setpoint of the PORV is also assumed to be a random variable wlth a normal distrtbution.
The assumption of normality for the actuatio1 of either the hi,h pressure trip or the PORV is just an L
assumption; no data is available to justity or deny the validity.
The RCS pressure rise above the RPS high pressure trip setpoint (hence referred to N
as " pressure rollover") during a LOFW or turbine trip was determined by a combination of plant data and engineering analysis.
Pressure rollover Eisn data from the operating plants (Table 2.1-1) was compiled from available g.;
ce data.
- However, these data points represent situations in which the PORY amma could open, thus decreasing t1e amount of pressure overshoot.
Therefore, it was necessary to correct for the PORV opening, since we are interested in the situation in which it remains closed.
This was M
accomolished by benchmarking the CADD code to a transient in which the amma PORV was isolated.
After shtisfactory duplication of this transient, the code was rerun modeling proper functioning of the PORV.
The resulting pressure correction to the rollover data was 17.4 psi.
The rollover data itself was tested and is stat stically acceptable as normally distributed.
It has a mean of 9.2 and a standard deviation of 27.52 psi.
The presence
-6 of negative values in this data set indicates that the RPS trip setpoints
/
have frequently been set low.
Since the data reflects actual operating y
expm-ience, the use of the nei;ative values can be justified in the G
analysis.
T p
.sg Using the above data and asstrnations, a Monte Carlo simulation of the h,_
relation p
PORV - RPS - EXCESS - BIAS = SAMPLE
...s r-
cas conducted.
The terms in the above relation are defined as follows:
PORV - PORV setpoint, a normally dttributed random variable W
RPS - High pressure trip setpoint, also a normally distributed random variable Q
v,..
c' ' -
'L EXCESS - Pressure rollover, a randenty-distributed normak l'
variable BIAS - A constant (17.4 psi) defined by analysis which g
compensates the rollover data for the fact that the PORV will remain closed.
I Six thousand sample values of the above alogrithn expre:, ion were j
e=,
calculated using the SAMPLE code.
A negative value of the above expression implies the PORV opens.
In the compuw. trials, no negative values in 6000 instances were observed.
It was then asstmed that the random variables described above are L
independent in the probabilistic sense, so an analytic approach was WI applied.
The stm or difference of several independent normal D
distributions is also a normal distribution with mean equal to the algebraic sta of the means and standard deviation equal to the square root
$C[
of the sta of variances.
In this case, the mean is p,,,
s, t.4 2450 - 2300 - 9.23 - 17.4 = 123.37 (except DB-1, = 73.37) 4 '.
1.
and standard deviation is
~]
t'.-
pm u
(1.4)
+ (1.4)2 + (27.52)2 = 27.59 (for D8-1,= 27.59) i m
The probability that the PORV t9ill op!n during an overpressure transient is 3.9X10-6/Rx-Yr (for 08-1 tnis value is 3.9X10-3/ Rx-Yr).
The i
statistics show that we can be 99% confident that at least 99.99% of all LOFW and turbine trip high pressure transients will not open the PORV for r
the P0hv set at 2450 psig.
For a setpoint of 2400 psig, the statistics t
indicatt a 99% confidence that more than 99.4% of the overpressure I
transients will not result in opening the PORV.
l mum 2.1.2 PORV Opening Probability Based Upon Operational Data g
NUREG-0667, " Final Report of the B&W Reactor Transient Response Task Force," contained a listing of reactor trips (148) with PORV actuations prior to the TMI-2 accident.
Since the accident at TMI-2 approximately 59 trips have occurred on B&W designed plants.
Approximately 42 of these g
ips would have lifted the PORV with the old setpoints.
Of the 190 m
I trips that would have lifted the PORV with old setpoints, three of these events would have lif ted the PORV with thc new setpoints.
In addition the modifications that have been made to the plants since those transients would have precluded PORV actuation given the same initiating events on i
those plants and the new setpoints.
Based on these data, it is estimated that the present PORV opening probability is less than 1.6% for an
-t overpressure transient, which is less than the 5% requirement stated in bZ II.K.3.7 of NUREG-0737.
P
-R$
p = P,,
a 5.tv w
V!
f.k Peu l
t i,
em
TABLE 2.1-1 i
, PRESSURE ROLLOVER DATA Trip #
_P_o.ve r, %
Peak Pressure, psig Rollover, psia _
mv ct' 1
95 2355 0
5 2
90 2385
+30 t
f.
3 25 2400
+45 4
20 2385
+30 5
90 2390
+40 6
32 2345
-10 7
40 2360
+5 8
40 2352
-5 9
92 2375
+20 10 15 2365
+10 11 35 2400
+45 mum 12 13 2370
+15 13 14 2355 0
14 38 2380
+25 15 98 2410
+55 16 72 2400
+45 17 100 2340
-15 C
18 100 2340
-15 h
19 100 2390
+35 e
20 100 2330
-25 M
21 98 2325
-30 i
fr 22 15 2355 0
f.-
23 9
2370
+15 k-
- n..
24 30 2345
-10 M.;
25 99 2350
-5 I
h',$
26 16 2295
-60
- 6.
!b C
2.2 Evaluation of PGRV and. Safety Valve Reliability 2.2.1 Safety Valve Failure Rate History There have been three cases where pressurizer safety valves were lif ted on B&W plants.
None of these cases resulted in failure of the safety valve to reseat.
Because of the few data points, no estimate was nade of the safety valve failure rates.
2.2.2 Evaluation of Small Break LOCA Probabilities /Need for PORV Isolation System The contribution to the probability of a SB LOCA from an open PORV was pa-estimated by two nethods.
The first was an analysis effort, the second l
was based strictly upon operational data.
The results are discussed below:
6
$N r"
2.2.2.1 Small Break LOCA Probability Calculations The probability of a stuck open PORV is the product of the probability of being demanded open times the probability of failing open on demand.
L The raising of the PORV setpoint has reduced the nunber 'of demands and hf thus the probability of being in the stuck open state.
The point
.~
estimate for PORV SB LOCA probability (variation not estimated) is M.
calculated to be 5.04 x 10-4 per reactor year (5.48 x 10-4 for Davis Besse) which is less than the II.K.3.2 requirement of 1 x 10-3 CT W.?
per reactor year.
The initiators of PORV actuations have been grouped h "
p{/
into five categories along the associated frequency of each category.
.c,
Details on how the values are calculated are contained in Table DT P/.[
2.2.2-1.
tN w
- ,.s
' Jc'
- l. PORY opening on overpressure transient 3.9 x 10-6/ Rx-Y r
[-
2.
PORV opening on transient with delayed 1.4 x 10-3/Rx -Yr aux. feed M
3.
PORV opening on operator action under 1.58 x 10-2/Rx-Yr g
ATOG guidelines y
4.
PORV opening due to instrurienta, tion 5 x 10-3/Rx-Yr control faults 5.
PORV opening from additional 1.8 x 10-3/ Rx-Yr consideration from II.K.3.7 6-is TOTALS 2.40 x 10-2/Rx-Yr a
2.61 x 10-2/Rx-Yr( CB) k 4
This total is then multiplied by the probability of the PORV sticking open on demand.
~d e
Note that all plants except Davis Besse (Crosby PORV) have Dresser valves; h
zum however, the entire B&W operating plant experience was us.ed to arrive at a generic PORV sticking open probability as follows:
There have been ten stuck open PORV events, five of which could be classified as mechanical f ailure of the PORV (the other five were basically installation errors).
Using all these fiva f ailures in determination of future frequency is 7
considered conservative since two of the failures (OC-3,6/13/75 and CR-3,
!97 m
V 11/75) were rectified by design changes, another (TMI-2, 3/28/79) cause is y
C':
unknown.
OC-2, 11/6/73 could be considered as a burn-in failure and the
- p ft.
W.
.t
- .n W. 4
- '[e
,s-.
M w
L,
08-1 10/13/77 event is a Crosby valve.
Using five failures in 250 0
demands results ir. 'a value of 2 x 10-2 to fail to reclose on demand.
This value is considered conservative not only due to the inclusion of all five f ailures but also the ntnber of demands is O
probably much higher than 250.
There have been 148 doctnented PORV openings on reactor trips; however, there is not a listing of PORV demands wnen the reactor did not trip (e.g., ICS runback) nor is consideration given to transients that could have actuated the PCRV mum nunerous times during an event.
The value of 250 demands is conservatively tesed here.
An analysis was also performed to include i
values for other than mechanical f ailure that keep the PORV open.
The t
results of this analysis is sunmed with the mechanical contributor (2 x mas 10-2/d) to arrive at the value for failure to reclose on demand g
(2.1 x 10-2fd),
Probability of PORV small break LOCA equals:
L (2.4 x 10-2)(2.1 x 10-2/d) = 5.04 x 10-4/ Rx-Y r (2.61 x 10-2)(2.1 x 10-2/d) = 5.45 x 10-4/Rx-Yr (DB) paie 2.2.2.2 Small Break LOCA Probability Based Upon Operational Data (J;
t' As discussed in Section 2.1.2, there have been three events which with
$g.
p..+
the revised setpoints wculd have actuated the PORV.
However, the
(',s plants have been reconfigured (e.g., upgrades on aux, feedwater, h)
, :7-control circuitry of PORV, NNI power sources, AC power sources) so as y
to reduce the probability of these PORV actuations.
Conservatively L._-
I estimating that one event could occur in the 45 years of B&W plant operation, yields a probability of occurrence of 2.22 x 10-2/ Rx -Y r.
-9_
The previous section gave a PORV f ailure probability of 2.1x10-2/d.
Therefore the probability of a PORV small break LOCA equals:
(2.22x10-2d/Rx-Yr)(2.1x10-2/d) = 4.7x10-4/Rx-Yr which is less than the 1.0x10-3/Rx-Yr criterion.
i
3.0 CONCLUSION
S Both the analytical prediction and the estimate based on historical data result in values for a stuck open PORV from all causes which meet the N
requirements given in II.K.3.2.
Note that no credit has been assigned for the operator closing the block valve given an open PORV.
Analytical l
predictions (given proper auxiliary feedwater response) result in a value
[
1ess than.01% of PORV openings for overpressure transients (taking into account the most limiting non-anticipatory trips) and tistorical data yg r
shows the frequency to be less than 1.6% which satisfies the criterion sum (less than 5%) specified in II.K.3.7.
Since the requirements of II.K.3.2 and II.K.3.7 are met with the current Ef PORV configuration and set point it is not necessary to address the requirement for an automatic block valve Closure system per II.K.3.1.
LT g;;f.
k.'
h:
pk Pfiw EL, F
$l r-- umu t
M
('
i
.1
Table 2.2.2-1 1.
The probability of a PORV opening on an overpressure transient from Section 2.1.1 for plants witn PORV setpoint of 2450
I 2.
The PORV opening probability in a transient witn celayed aux. feed g
A value of 1.0 was assigned for PORV 1.4 x 10-3/ Rx-Yr g
opening probability if aux. feedwater was not supplied.
A value of 1.4 x 10-3/Rx-Yr for loss of all feedwater f
as referenced f om a B&W calculation g
which used average unavailability as pg
.a,
calculated in the generic aux. feedwater P
reliability studies (BAW-1584) in conjunction with generic EPRI data on loss of main feedwater frequency and loss of offsite power frequency.
sc b
On completion of the ongoing aux. feedwater reliability analysis ( AP&L, SMUD, FPC) more
~
specific values can be applied to those plants.
N 3.
The PORV opening probability on operator action p
under ATOG guidelines W
There are 3 events that call for 4 *.-
operator opening of the PORV:
a) Loss of All e
Feedwa ter.
This contribution is already counted be n 2 above; b) Small LOCA.
Not applicable to y.
i-
~
Y' m,
j.P Table 2.2.2-1 (Cont'd) this calculation since the plant is already in a small LOCA; c) Steam Generator Tube 4
Rupture (considered smaller than small m
LOCA as defined in II.K.3.2 so arg ment of l
t b) does not hold):
The demand on the PORV
[
given a tube rupture varies depending on be.
whether offsite power is available or lost.
hj If offsite power (Reactor Coolant Pmps) is available, only one PORV opening is required, whereas in the loss of offsite power scenario as many as 23 PORV openings are required.
E:=
The value calculated assumes that the probability of Steam Generator fube Rupture considered with a LOOP event is small (no causal effect of LOOP or Steam Generator Tube Rupture) and therefore, the WASH-1400 of 1 x 10-3 for a LOOP given a reactor trip is used in the calculations.
There have not been any tube ruptures in the cmulative B&W experience, W
(45 Rx-Yrs) due to the limited n mber of years Q
experience.
A Chi-square 50% confidence value b
with 0 failures is rather high (1.54 x 10-2 Rx-Yr).
.E in
- s, p (,'c i
',* [w i - ~
~
t-I Table 2.2.2-1 (Cont'd) i 1.54 x 10-2/Rx-Yr x 1 demand (offsite power available) 1.54 x 10-2/Rx-Yr 1.54 10-2'Rx-Yr x 10- Of fs te Power Loss / Event m
x 23 demands (offsite pcwer lost) 3.54 x 10-4/Rx -Yr
}
+
b 1.58 x 10-2/Rx-Yr In the final calculation of probability b
to reclose, it should be noted that no adverse effects of the 23 demands in the l
l-loss of offsite power case on PORV i
operability is assuned.
4.
PORV opening due to instrumentation control faults e
This has been estimated at 5 x 10-3/
5 x 10-3/Rx-Yr I
reactor year.
This value assunes that power supply faults and other control deficiencies have been corrected by each
- ~
utility.
0+
5.
PORV opening probability from additional N
consiaerations from II.K.3.7 k-"
r~
T"ar- 'ra overcooling transients
[?
r,.
that initiate HPI and operator failure to throttle or terminate flow before the PORY N
gb setpoint is reached.
There have been k
u 8 overcooling transients that initiated
!NN
Table 2.2.2-1 (Cont'd) w HPI in 392 reactor trips.
The current
~
'h~
(
frequency of reactor trips is 6 trips /
^/j Rx-Yr per plant.
In this event sequence, paa the operator has approximately 4 minutes 3
from time of HPI initiation until PORV setpoint is reached.
The operator E
e.;
failure rate to terminate or throttle b5 HPI flow is based on having AT0G in place (1.5x10-2/d - based on NUREG-CR-1278 with moderately high h
stress).
The overall probability of this sequence is therefore estimated MIM to be 6 trips /Rx-Yr x 8/392 overcooling events / trip x 1.5x10-2 =
1.8 x 10-3 Rx-Yr
~
N.A. for DB TOTALS 2.40 x 10-2/Rx-Yr 2.61 x 10-2/Rx-Yr(DB) hh Note that these values are dominated by the conservative analysis of steam tt:
,an-generator tube rupture.
Analytical studies could be performed to obtain a i,F W
more realistic value.
Also note that the calculation for category 4 did not
.i include operator or maintenance induced faults, such as the DB event of g.-
10/27/80.
W L
L r..
mus -
Rf
-~
m e-
+5 kA-h-1--L
-.. _am
.A4 a
_ A h
a a
2-.
a-.
4 4
i a
4 1
N 1
?
CONTROL ROOM i
HABITABILITY EVALUATION I
d 1
P J
1 5
- a,
i 4
1.C Introducticn j
Wis report is to cynluation of the hrbitebility of the Crystal Elver Unit 3 centrol roen in complience with UCCEG 0737 Item III.P.?.li 'Ih is cvaluation utilizcs tbc results frcm the rnalyses of centrol rocn concentration for postuleted occidentc1 release of toxic cases, control rcen operetcr rrdiction exposures free airborno radioactive nate:-ial, rnd direct radiation resultinr fron desirn besi.c accidents.
In edditien to providing the results of the Ptove rnalyses, the input data is fresented in ficure rnd trble ferr with subsequent discussion of results end, where required, ccrrectivo measures.
2.0 tnalyses 2.0.1 f.nclyses of Toxic Cas l'a7ards Analyses of ptential toxic gas hazards were performed in accorcance with Etandard Ecview P)rn (FHP) 6. 11 and applicab?e parts of fcctions P.P.1-P.P.2 and P.23 We 3ccation of Crystal Piver chemical stornce fccilities is show i on Figure 1 end qucntities cre detni]cd on Table 1.
We results of these analyses for Iccations 1-11 resulted in the control rocr. air intoke concentrations belou:
Chemic 11 Concentration at Toxicity Corrents Control Pocn Limit Intake Carbon Lioxide 0.157,by Volume
- 17. by Voltre Flash coeff.
0.472 l
Hydrogen of plirce, con-centration expected 20 Mitregen 0.17. by Volume Asphyxiant l
<337 Volume Eulfuric
%0 2 mg/m o
10C Vapor i
- ressure <1 j
torr.
Failure of a liquid chlorine or anhydrous entronia storcge tenk could result in a control room air intake concentration cbove toxicity limits, if worst esse conditions are uti]ized.
Accordingly, the design for installation of ammonia and ch]orine detectors and _ rn additional intake isolation dampcr is required to derronstrate habitability.
The renmininn leertions are under evalucticn; hcwever, it is not cxpected that. them cher'cals, uith the exception of trvenia, v.culd rose e barrrd to con. col roon bcbitsbility in the centext of Eegulttcry Guide 1.7f requircr.cnts.
Fcr locatien 1h, this rcronia storene would nct irpose ony rdditional requircernts above that of locatien 1.
As shom
- above, cnly failurcs of the chlcrine er enhydrcus crronic storacc vessels have the rotential to result in toxic cas ccncentrrtien in the centrol rcom that
- cculd, under discrete atrospheric conditions, excced regulatory guldence.
Posed upon these findings, FFC has undertrken a design deve3creent prorrrn to:
(1) install chlorine dctceters, (2) install err.cnin detectors, rnd (3) upgrade the intake isolation darpers.
Estimates of procurcrrent and installation schedules has identified ccrpletion during our 1983 refueling outage.
2.0.2 t.nalysis of Airborne Padioactive Paterial The present rnelysis utilized source terms for a LCCA per Feculatory Cuide 1.3 EFF Icakcce per ERP 15.r56.5 App. F uas considered and found to be a negligibic centributor in cerrperison to containment lerkrce.
Other design basis cecidents were revicued rnd fcund to be less severe hartrds.
Strndard tethods of calculatien as found in the 13th AEC Air Clenning Conference paper, F.A. Purphy and K. V. Ccepe (reference 2) were used to perform the enalyses.
The final rsults of the cnalysis indiccte the follcuing for the 30-day durstion of the accident:
Pose Cc]culated Linit rose (PEP)
(FEM) khole-Pedy Gmma 5
2.67 Thyroid Inhalatien 30 911.7 with 9000 CFM 28.7 with 2500 CFl' Eeta Skin 30 1P.1 As can be seen, the dose limit criteria of CCC 19, I.ppendix A of 10 CFR Part 50 and SRP 0.11, " Habitability Systcms" are met provided air flow into the control room is throttled to 2500 CFM versus the present 9000 CFM.
Engineering work has been initiated to design a rodification to limit
~
air intche flow into the control room.
Irplementation of the cbove is expected to coincide with Section 2.0.1 rodifiations.
2.0 3 Analysis of Firect Padir tion The analysis of control room direct rrdiction was presented in our "05000302/LER-1979-109-03, /03L-0:on 791221,following Reactor Trip,Chem/Rod Sampling Showed Reactor Coolant Dose Equivalent I-131 Exceeded 1.0 Uci Per G Contrary to Tech Specs.Caused by Leaking Fuel Pins & [[Atomic Element" contains a listed "[" character as part of the property label and has therefore been classified as invalid. Transient|January 11, 1980, letter]] in response to !!UEEC-0578 Item 2.1.6h.
This analysis identified a dose rate of 11 cr/hr at time : 0 from direct radiation from the centainment.
Page 2
- ~ - - -. -. -. - - - - _. _.
1 j
[
j
=
i i
TAFLE 1 I
[
l.CCATICi!
CilEf'ICAL Cunt?ITY DISTAffCF Frrli Cil j
If!ini'.E ( f t. ) -
i 1
Anhydrcus Arrcnic f,500 1bs
?C T9 P
Carbon oicxide 1?,000 lbs
?fGr 3
Chlorine 16 centainers 2000 j
(20C0 lbs ca.)
11 Chlorine P containers 11000 i
(150 lbs. ca.)
5
!!ydrecen 12 Tuben.
11076
( 11 0. 11 lbs. ca.)
I i
1 6
l'ydrogen 12 lubes ll075 k
(40.11 lbs. ec.)
I l
7
!!itrcccn 11 centainers 4075 (561 lbs. ca.)
.l C
Cu] furic feid 11n,P0g!p Ibs. of 4000 6( Pc I S0 I
p 9
Eulfuric Acid 7tl 2001hs. of 3200 3
3 6f Pc !! PC 2 4 10 fulfuric /.cid 74 ?r0)Rs. of
?300 6
06 Pc if 00 2 g i
i 11 Ful furic Acid 11g,De FLSO,i7Cg lbs, of 3f00 66 i
c i,
12 Fulfuric Acid 1000 cal)ons 114 0 13 Sodiun liydroxide 90,0c0 lbs.
440 1 11 Anbydrous ler.onia (2) 766 Ft.' ca.
hl10 15 fulfuric Acid P000 gallons 500 16 Ecdium !!ydroxide 8000 cellens 560 7
17
!!itrogen 141,000 Ft-204 7
l 18 l'ydrogen 40,000 Ft.-
500 l
19 ifydrogen 40,000 Ft.3
-730
,,. _,.. _ _ _ _. ~ _ _,..
,d.._
e
_.r-s,_
-,--, w -
AttacPrent 1 Centrol Eccn Itabitchi]ity Evalucticn (1) Centrol Poem rodc of crcration Zeno isolation, with filtered recirculttcd air, nnd a positive presr.ure rnintained in the zone.
(I') Centrol Pect chnrocteristics (a) Air Volunc Contro] Poem 243,000 ft3 (free)
(b)
Centrol Pcom Emergency Zcne Elevation 145' 0" of the control Compicx (c) Control Poom venti]ation fce Attachment P system scheratic (d)
Infiltration lenkege rate c600 CFM (existinc) 2500 CFM (after rodification)
(e)
Filter Efficiencies 1.
Charcoal filters 95" for all species of iodince 2.
ITEPA filters (f) Closest distence between PF ft.
conteir. ment and air intrke (g) Layout drawing See Ficure 3 (h)
Control Econ shielding 11 rr/hr at T = 0 (Pef. Jan. 11, ISf0 submittal)
(1) Automatic isolation capability 1.
Damper closing tire 3-5 sec.
2.
Camper Icakage 55 of eir intake p
3 Damper crea
~24 ft-(j)
Chlorine detectors
!!cne (k) Self-continued breathing 2
apparatus availability (1)
Ecttled air supply 2 spare for chove j
(m) Emerccncy food supply 7 dcys Emcrgency portable water supply
!!one.
(n)
Control Poom personnel capacity 4 to 20 -
(o)
!!one w
Page 2 i
i (3) Cn-site storage of chlorinc cnd other hazrrdous chemicals (c) Total trount and size fee Tctle 1 (b) Closest distrnce from Ccntrol Tcc IT'ble 1 Poon air intoke (11) Off-site manufacturing, storoge or transportation facilities of hazardcus chemicals I
(a)
Identify facilities with a 5-rile rcdius (b) Distrnee from Control Boom fee FS/'R recticn F.P.'
(c) Cucntity of hazarJous chemicals in one container (d) Frequency of harcrdous chemical transportatien traffic (truck, rail, end barge)
(5) Technical specifications (refer to standard technien1 specifications)
(a) Chlorine detcetion system
!'cre (b) Control Poca filtraticn system including the capability to naintain the Control P.com pressurization at 1/F in. water J
gucce, verification of isointion by text signals and dorper closure times, and filter testing recuirements.
l
.~ $
i
. j
' SOyod[g g n i
s STLAALL A 7.L A
'~
SE C Y#C Cf.T E 3 g
mu m, y,Suttuh5C &CLD N
b k
STOR AM TA%s
?l C Angom OsJasoE
$7otAGE 4MA N Cav3TAL nivth U44 T S 4 D 6 0
<* STCn AGE TAqs I-N Aid MCM B A N
2 +v
'l d
se C th CUL ATIN 3 WATER
{
f i
(
/
p p
- Tonace taw y'
j
./
SULFWABC ACeQ m
,,/
\\
r i
C,.wuf,.
I Uo'nic?/="cT'
\\
-1
\\
s
\\
\\
.l y cometms Af t pouseta l
fulf URIC AClo
.j O
0 i
- g sto4A63 TAhR$
.o.
y
\\
l 1s
- 1
- y 3
1 4
.j s
3:j A" J
- b s
1 d
1 N
~ - -
4
- ,isenanst cA AL m
_ _ 8T'" AGE
- 8888 sutrynic Acio 3
j is scoivu avonoxion 20**
sTon Aer Ans A
- i. Annvonous Aumonia tan j f tg
}
9, -
h
[
c
-)
~
.D Mf ono0E8e STOR AGE UNIT 3 E--
C "a'm"a
" 0 0 "
AREA inf i
g
~ nn*
~~
AUI. SLO. VENT PMo a t vthT j j
@ Acio : Ton Aer __ -, f **f UNIT I UNIT l 1
f N
5 Ano a
goo
- h CAutTIC ston AGE
\\
uvonoorn g,
a _g sTonaes AncA J
i.
1
)
b4 f
)
?
?
HAZARDOUS CHEMICALE h
4 STORAGE LOCATIONS CRYS ~AL RIVER PLAN' l
FIGURE I
I.-
Sn 2c t
g.t 5 s i
- sn I
a$'~i
- 4 I
ls!
jvu q t
,1 n
8-e f
3 t
I u
Ii sh l
i 3-i 6-t 6!
QC p!!-~.
h k
t n;
'e,.w. j I is _
1 L'l 4'
&. {..I ",I:
i l'
i' s
l L _J I
t,L,I.33 r
a 14, t
- i 9'
I, p:i,j ---. w j! --
rm e _.
4 I
- r.._.
i i
' t*L g,',,
i LL
_t
~
-Q i
1
..g.
.a 4 I t.
- AI r
9 L * *_.
L
.f = LE:
-G &
-G fl,
.O. b:
,.e-i 1-1 4
w::
Ill l,;d,'!_
e', j g
p-
.r[!."9
=
, L__.J j $1, O
Q, &'y I
gJ
.trif r
L d
I
- 't" p ' ii
.;;h
.I G.Y..
j
- c t
d i,~ I T. ~
l j
E i,hi i,$
,I d.
~
+
4 3l '
f l
L ;
rl o
~
r, 9,.T y; irs "l-I to..
g:
_O
=4 B
Fn J:
,t i
g',l' J -,
el
.w q:4; r} -
i r,:
48 o
g f (b, - - r.f7 I
a
.t h
l
. _ t.
Fral 9'
, $ r n !tra =
3 l r
t s
l 11**
- r
. ?. Ljp; o
L,ng I
t I I
e;3]m tl-G
!Q l~n i, b.n,I i r,. y c-D 0 01 ti-c; Ed
' !"qi, L !
j L J; L. j !
f;.
c__
O l J
G a.U G D ". D ' & - U i h*.I
' f'
- !il..!
4 h,
q
, fil ':. e, (tII.t "i
qs as a,1,t 1 @
s i
i
.t+ L;l =J r
i b rT::
t 4
s I
I b'
~
i e
o
=
.. n,
,y i 4 i
l f
l
- I I3 E -.f i-'
R t
4 r,' l' l9 St i
IIgI 4
I.
l
/ L8d i
}
- i_t %s LJ l l
=
3 m
9 O
e 5-(
., ; l.m e
a
+
7 n-G
(, Flh._.
t *L_l P i
7 1:
i i
,3, N
!!N p
9
, [th}-[. 8[
n, o u d,
} }. Q -- } ' Q :
}l att i
i
. _4,gy i
II
~
' L -~-@
r 1
eg g
h'- J H
yw f
gg,
dB,, {
$fp. II"
_i
<. _ ~
s _-
b
+
I BAW-1648 November 1980 THERMAL-MECHANICAL REPORT - EFFECT OF HPI ON VESSEL INTEGRITY FOR SMALL BREAK LOCA EVENT WITH EXTENDED LOSS OF FEEDWATER Applicable to Babcock & Wilcox 177-Fuel Assembly Nuclear Steam Syntems m-D' 3
9c
.m
. J S/ O/6 761fr 4 BABCOCK & WILCOX Nuclear Power Group Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 Babcock & Wilcox 1