ML20065M525
| ML20065M525 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 10/15/1982 |
| From: | Daltroff S PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0803, RTR-NUREG-803 NUDOCS 8210210297 | |
| Download: ML20065M525 (10) | |
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PHILADELPHIA ELECTRIC COMPANY 23ol M ARKET STREET P.O. BOX 8699 PHILADELPHIA. PA.191o1 SHIELDS L. D ALT ROFF rLecNic PYoEc ion October 15, 1982 Re: Docket Nos. 50-277 50-278 Mr. John F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Stolz:
Your letter of August 24, 1982 forwarded a request for additional information with regard to our response of December 29, 1981 (S. L. Daltrof f, PECO, to J. F. Stolz, NRC) to NUREG 0803, " Safety Concerns Associated with Pipe Breaks in BWR Scram Systems".
Your questions are restated below, followed by our responses to these questions.
In addition, we have attached the report forwarded by the BWR Owners Group (T. J. Dente, BWROG, to i
D. G. Eisenhut, US NRC, dated August 22, 1982) addressing the l
probability of such an event occurring.
ASB 1.
Threaded Joint Integrity In your response (1), you noted that a review of plant specifications revealed that the only threaded joints specified for either Peach Bottom Unit 2 or Unit 3 were l
those for non-safety related air supply piping (compression fittings) and limited test connections.
In addition, you reported that you would conduct a walkdown of the Unit 2 piping during the upcoming refueling outage for Unit 2 to confirm the results of your review and that you would not conduct a similar walkdown of Unit 3 if no threaded joints were revealed in Unit 2.
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PDR ADOCK 05000277
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Mr. J. F. Stolz Page 2 Provide information showing the location of the limited test connections which are threaded, together with the size of these connections for Unit 2 and 3.
In addition, provide a commitment to conduct a similar walkdown of Unit 3 SDV process piping since the walkdown of Unit 2 is apparently intended to ascertain the presence of threaded connections not in accordance with specifications, and the assurance that Unit 2 has been built in accordance with specifications does not provide similar assurance for Unit 3.
Finally, commit to provide us with the results of your walkdown to assure no threaded joints other than those permitted by plant specifications as a result of your walkdown during the February 1982 refueling outage.
Response
It is important to note that the threaded joint connections described in our letter were for non-safety related air supply piping and a limited number of calibration and test connections on instruments.
These threaded joints are outside the scope of interest of NUREG 0803.
In all cases, the connections are 1/2" pipe size or smaller.
In our letter of December 29, 1981, we i
proposed a walkdown to confirm that there were no threaded joint connections in the area of interest, i.e.,
threaded joints existed only in non-safety related air piping and instrument calibration connections which are valved away from the scram system during operation.
Our Engineering Design Division has performed this walkdown using acceptable QA methods and has confirmed our engineering review.
Specifically, an inspection of a 10% sample of the lines between the drive housing and scram discharge header and 100% of the header and associated drain, vent, and instrument lines was completed on Unit 2.
Furthermore, to add additional confidence that our construction methods and engineering reviews were correct, a complete inspection was made of the header and associated drain, vent and instrument lines on Unit 3.
Based on this review, we confirmed that no threaded j
piping connections are part of the pressure boundary of the scram discharge volume.
In our December, 1981, letter wherein we proposed a walkdown to gain confidence that our engineering review was correct, we specifically declined a similar inspection for Unit 3 if our Unit 2 expectations were confirmed.
Such a walkdown presents personnel radiation exposure which is contrary to the principles of our ALARA program.
In light of the fact that our Unit 2
Mr. J. F. Stolz Page 3 walkdown confirmed our engineering review coupled with the fact that the threaded piping being discussed is outside the scope of NUREG 0803, we do not believe a walkdown of our Unit 3 is appropriate from both a cost and radiation exposure viewpoint.
ASB 2.
HCU-SDV Equipment Procedures Review In your response (1) you state that the procedures already reviewed "...do not specifically address the maintaining of the scram system boundary integrity as discussed in NUREG 0803 (2).
However, it is thought that sufficient steps are taken to assure the postulated problem is avoided."
This is a rather vague response to our recommendation that procedures be reviewed in order to eliminate possible errors leading to a defeat of SDV integrity at a time when SDV integrity is required.
Verify that plant procedures for surveillance, maintenance, inspection, and modification which have the potential for defeating SDV integrity at a time when SDV integrity is required have been reviewed to assure that proper procedural controls are maintained in all cases as to prevent a breach of SDV integrity.
Provide a list of any procedures which have to be modified to prevent a breach of SDV integrity together with a schedule for such modifications.
Response
A review of plant procedures for surveillance, maintenance, inspection and modification has been conducted by a headquarters and a plant technical engineer to assure that SDV integrity is not breached at a time when the integrity is required.
No procedures have been identified which require revision.
ASB 3.
Improvement of Procedures Your response (1) noted that you would support a preliminary study by the BWR Owners Group (BWROG) to determine the best approach to carry out the guidance of NUREG-0803 (2) in addressing scram system pipe breaks and that the BWROG will then determine whether to initiate specific actions to modify the Emergency
Mr. J. F. Stolz Page 4 Procedure Guidelines, accordingly.
You expected the BWROG study to be completed during the first quarter of 1982.
Based upon the current status of this study, provide us with a schedule to provide emergency procedures to address a break in the scram discharge volume piping, together with summaries of the procedures for our review.
Response
The BWROG has recently initiated work on a secondary containment control guideline which addresses the problem of scram system pipe breaks.
The current schedule is for the completion of this work early in 1983.
Upon completion of this task, and approval of the guideline by the NRC, PECo will adopt the guideline (after appropriate internal review).
ASB 4.
Verify that the temperature trip monitors for the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) pump turbines are located sufficiently remote from the scram system and SDV to prevent initiation of turbine trip signals because of high ambient temperature resulting from the postulated scram system pipe break.
Your analysis should account for the potential leakage path from the pipe break and air flow within the reactor building with normal ventilation systems in operation in order to determine if the temperature at the location of these monitors increases to the point where trip is initiated.
(Refer to NUREG-0803 Section 4.3.1.3).
Response
The purpose of the area temperature monitors is to detect a broken steam line to either HPCI or RCIC and isolate the affected steam line.
Accordingly, these monitors have setpoints in the 190-195 degrees F range.
The analysis performed to determine the temperature response in the reactor building to a postulated break in the SDV has determined that the maximum temperature will not reach the setpoint of these monitors.
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MEB 5.
Seismic Design Verification i
In your response to NUREG-0803 (1) it was stated that j
the SDV piping has been reviewed to verify that it has been designed to seismic loadings as part of IE Bulletin 79-14.
Because IE Bulletin 79-14 does not provide c
I coverage of small diameter piping (less than 2 1/2 inches nominal pipe size), you are requested to verify that for small diameter piping in the SDV system:
a.
the piping and supports have been designed for seismic loadings, and 1
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the actual piping and support installation have been verified to assure the validity of the seismic j
analysis.
Response
All of the small piping associated with the scram discharge-volume has been verified to be adequate for seismic loadings.
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As a part of our compliance with IE Bulletin 79-14, all piping which had had a computer stress analysis, was walked-down and verified.
This included the insert and withdrawal lines.
As a result of IE Bulletin 79-02, many of the hangers on these lines had been previously upgraded, t
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- The remainder of the small piping associated with the scram i
discharge volume has been verified to be seismically adequate as a result of the scram discharge volume modifications undertaken in response to IE Bulletin 80-173 l
AEB 6.
Limit of Coolant Iodine Concentration to Standard Technical Specification Valve
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The radiological consequences of a scram discharge-volume failure are analyzed generically in NUREG-0803 with respect to onsite occupational exposure to workers entering the scram discharge volume area, as well as offsite doses, and were found to be within the relevant 1
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guidelines for plants with General Electric Standard Technical Specifications (GE STS) for reactor coolant iodine concentration; while worker exposure and offsite consequences were found to exceed the guidelines for coolant iodine technical specifications similar to Browns Ferry.
We note that you have neither proposed to adopt the
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General Electric Standard Technical Specifications (GE STS) for reactor coolant iodine activity and i
surveillance requirements, nor calculated occupational or offsite dose consequences for the scram discharge volume break, using your technical specifications in the analysis.
Also, we find that you have not provided clear evidence to provide that the probability of the reactor coolant iodine concentration exceeding the GE STS is 0.001 per reactor year or less.
As noted on p.
5-5 of NUREG-0803, " Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping", 1981, a scram discharge volume break which causes a rupture of the blow-out panels may result in excessive offsite doses in addition to causing an exposure problem for workers (for instance, those workers who might enter the scram discharge volume vicinity to manually close valves).
Therefore, you should either: 1) propose GE i
STS for reactor coolant iodine activity, or 2) provide us with an evaluation of radiological dose consequences, using calculative methods described in NUREG-0803, and demonstrate that the doses from this fission product release do not exceed occupational or offsite dose guidelines.
The assumptions used should include the proposed or existing technical specifications on reactor coolant iodine concentration and an iodine spike caused by the accident.
Response
A request for an amendment to Technical Specifications which utilizes the General Electric Standard Technical Specifications as a model will be prepared and submitted with the next appropriate submittal.
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Mr. J. F. Stolz Page 7 EQB 7.
Equipment Qualification a.
Identify all systems and equipment that would be used to detect a break and/or leak in the SDV system and state that this equipment is, or provide a commitment that it will be 1) included in the environmental qualification program established in response to IE Bulletin (IEB).79-OlB, and 11) qualified for service either in a 212 F and 100%
humidity environment, or in a plant specific SDV break environment.
b.
Identify all systems and equipment needed for the prompt depressurization function and all emergency systems and equipment, i.e.,
systems and equipment needed for mitigation of an SDV system pipe break, safe shutdown of the plant, and long-term core cooling.
State that this equipment is, or provide a commitment that it will be i) included in the environmental qualification program established in response to IEB 79-01B, and 11) qualified for service either in a 212 F and 100% humidity environment, or a plant specific SDV break environment.
c.
Identify any emergency systems and equipment that could be sprayed with water from dripping or splattering of overflow leakage down open stairwells following a break in the SDV system, and state that this equipment is, or provide a commitment that it will be 1) included in the environmental qualification program established in response to IEB 79-OlB, and 11) designed to, or qualified to, operate with water impingement.
d.
Identify all systems and equipment needed for mitigation of an SDV system pipe break that could be wet down from leakage through equipment hatches following the break, and state that this equipment is, or provide a commitment that it will be 1) included in the environmental qualification program established in response to IEB 79-OlB, and 11) qualified for wet down by 212 F water.
Mr. J. F. Stolz Page 8 e.
If any equipment needed i) to detect a break and/or 3
leak in the SDV system, 11) for mitigation of an SDV system pipe break, iii) for safe shutdown of the plant and iv) for long-term core cooling is not qualified for service in an environment that could exist following a break in the SDV system, provide justification for interim operation pending qualification of the equipment or replacement with qualified equipment.
Response
PECo was notified by the NRC Peach Bottom Project Manager on September 10, 1982, (telecon M. Fairtile, NRC, to W. M. Alden, PECo) that this question should not be addressed at this time and that this particular issue will be the subject of future correspondence from the NRC.
MTEB 8.
Periodic Inservice Inspection and Surveillance for the SDV System You made the following statement (1) concerning the periodic inservice inspection and surveillance of the l
Scram Discharge Volume (SDV) System:
"The NUREG recommends that the SDV piping i
should, as a minimum, be subjected to the ASME Section XI Inservice Inspection (ISI) l requirements for class 2 piping.
We shall
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inspect the piping on Unit 3 equivalent to-Class 2 piping for ISI purposes.
Upon completion of the scheduled modifications on i
the Unit 2 Scram Discharge System, that piping shall also be treated as equivalent to Class 2."
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l Later you committed (3) to upgrade the SDV inspection j
program in accordance with the requirements for Class 1 l
piping specified in Section XI of the ASME Code.
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To evaluate the adequacy of the inservice inspection'and surveillance program for the SDV system, the additional information listed below is required.
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Mr. J. F. Stolz Page 9 a.
What Code Edition and Addenda of Section XI will be used to perform the required examinations and tests on the SDV System?
b.
What are the pipe schedule numbers and diameters and from what materials are the discharge header and instrument volume fabricated?
c.
Will any portion of the SDV System subject to examination be exempted from examination by any of the criteria given in IWB-1220 of Section IX of the ASME Code?
If so, please state which portion and the criteria used to establish the exemption.
d.
Will any relief from Code requirements be requested in the inservice inspection program for the SDV System?
If so, please state the relief and the basis for requesting it.
Response
a.
The SDV Systems, on both Units 2 and 3 at Peach Bottom Atomic Power Station, will be examined and tested in accordance with the requirements of Subsection IWC of the ASME Boiler and Pressure Vessel Code,Section XI, 1974 Edition with all addenda through Summer of 1975.
The SDV systems will be treated as class II components for the purpose of these examinations.
This is a revision of our previous commitment resulting from the completion of our review of ISI requirements for the SDV system.
b.
Peach Bottom CRD SDV/IV pipe sizes are as follows:
UNIT 2 UNIT 3 Scram Discharge volume 6"-Sch 80 8"-Sch 80 Instrument Volume 12"-Sch 80 12"-Sch 80 Material Carbon Steel Carbon Steel A-106-B A-106-B c.
Our Inservice Inspection Group advises that the SDV system components will be exempted from the Section XI examination requirements of IWC 2520 as allowed by the exemptions contained in IWC-1220(b).
This exemption is based upon the
Mr. J. F. Stolz Page 10 fact that the scram discharge volume is not required to function during normal reactor operation and is not an emergency core cooling system.
This exemption applies to the entire scram discharge volume system and includes all components and welds therein.
d.
No relief from Code requirements is required in the Inservice Inspection program for the SDV system other than that discussed in paragraph c above.
Very truly yours,
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Site Inspector Pench Bottom
e T. J. Dente. Choirmon P.O. Box 270 e Hortford. Connectkut 06101 e (203) 6664911 I 5489 r,.
BWROG-8254 August 23, 1982 U. $. Nuclear Regulatory Commission Division of Licensing Office of Nuclear Reactor Regulation Washington, D.C. 20555 Attention: Darrell G. Eisenhut. Director Gentlemen:
SUBJECT:
Analysis of Scram Discharge Volume System Piping Integrity NED0-22209 (prepublication fann)
Reference:
NUREG-0803: Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping August 1981 The enclosed report. " Analysis of Scram Discharge Volume System Piping Integrity". NED0-22209, documents the results of a BWR Owners' Group study to determine the probability of the loss of SDV piping integrity.
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and to evaluate the contribution of such a loss to a core melt.
J It is the position of the BWR Owners' Group, substantiated by the results of these analyses, that the probability of core damage initiated by a failure of the scram discharge volume piping integrity is sufficiently low so as to preclude the necessity of qualifying equipment to detect and/or mitigate the consequences of such an integrity loss. Consistent with these conclusions, it is also the position of the BWR Owners' Group that no further action is required as regards the equipment qualification and system design modification recomendations of the reference NUREG.
The enclosed document and the conclusions drawn from the results of these analyses have been endorsed by a substantial number of the members of the BWR Owners' Group; however, it should not be interpreted as a connitment of any individual member to a specific course of action. Each member must formally endorse the BWR Owners' Group position in order for that position to become the member's position.
fr$gbibA$ 3Y
4 U. S. Nuclear Regulatory Commission Subj: Analysis of Scram Discharge Volume System Piping Integrity.
NED0-22209 (prepublication form)
August 23, 1982 Page 2 Should you have any questions on the enclosed material, please feel free to contact F. R. Hayes of the General Electric Company at (408) 925-2140.
Sixty copies of the published version of the subject report will be transmitted to you shortly under separate cover.
Very truly yours, L_
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,,3 T. J. Dente, Chairman BWR Ovners' Group TJD:WHP:na Enclosure cc: BWR Owners' Group K. Eccleston (NRC)
J. F. Schilder (GE)
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NED0-22209 82NEBD88 CLASS I August 1982 ANALYSIS OF SCRAM DISGARGE VOLUME SYSTEM PIPING INIEGRITY G. Alesii
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F.R. Hayes P.P. Stancavage
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Approved by:
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R.J. 'Br// don, Manager Nucleaf Services Engineering Operation Approved by:
Approved by:
M) r! % irk, Manager J.F. Schilder, Nanager 1 Systems Licensing BTR Generic Programs l
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DISCLAIMER OF RESPONSI8ilJTY This document was prepared by or for the Generet Elecmc Company. Neother the Gerreral Elecmc Company nor any of me contnbutors to tha document-A.
Makes arry warranty or representation, espress or impioed, wMtr respect to the accutacy, completeness. orusefulness of me informa00n conteined on thes 6ocu.
ment or that the use of any nnformabon disclosed on ttus document may not infrege pavately owned nghts; or B. Assumes any responsaboley forliabelty or damage of any httd wfuch may result kom me use of any unformatron disclosed on ttus document.
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TABLE OF CONITNTS 0.1 Tables 0.2 Figures 0.3 Abstract 1.0 Introduction
1.1 Background
2.2 Purpose 2.0 Anal; sis 2.1 Description of SUV Systen 2.2 Fanit Tree Diagram 2.2.1 General Description 2.2.2 SUV Pipe Break Probability
(.2.2.1ReviewofNEDO-24342 Approach I
2.2.2.2 Review of NUREG-0803 Approsch 2.2.2.3 Reevaluation of Break Probability Using Plant Data 2.2.2.3.1 Evaluation Procedure 2.2.2.3.2 Discussion of Results 2.2.2.4 Fracture Mecharic's Approach 2.2.3 Probability of Stuck Open Valves 4
2.2.3.1 Fallare Rate of Drain and Vent Valves 2.2.3.2 Fallare Rate of SUV Relief Valve 2.2.3.3 Other Considerations j
l 2.2.4 Probability of Breach of SDV In egrity l
3.0 Summary and ConcInsions l
l 4.0 References
' ppendia'A Perth ipating Utilities
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O.1 TABIES 2.1 Characteristics of The SDV System For Varlons Plants 2.2 Break Probabilities Using Experience Approach l
0.2 FIGURES 2.1 Simplified Schematic of Control Rod Drive System 2.2 Typical Scram Discharge Volume Configuration 2.3 Fanit Tree fer Breach of SUV Integrity 2.4 Slap 11fied Diagram of a Typical SUV Instrument Air Control (E
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0.3 ABSTRACT Analyses of the Boiling Water Reactor (BTR) scram system piping integrity have been performed. The purpose of these analyses is to determine the probability of a loss of SDV piping integrity and to evaluate the contribution of such a loss to a core melt.
The likelihood of a loss of piping integrity was calculated based on a consideration of pipe length, scram frequency and vent and drain valve reliability. Conservative values for the key input values were selected based on BWE plant data and on generic reliability data. Pipe breal probabilities were estimated based on the experience data used in the Reactor Safety Study and on a fracture mechanics analysis of the piping system.
The results of these analyses show that the probability of an unisolatable loss of c
scram system piping integrity for an average plant is 3 x 10-7 per plant year.
The probability of core damage resniting from a loss of SDV pipe integrity is approximately 4 x 10 81 events per reactor year. This is significantly below the proposed NRC safety goal for core melt events of 10 8 per plant year.
Consequently, the probability of a loss of scram system piping integrity leading (i
to core damage is sufficiently low to preclude the necessity of qualification or design modifications of equipment required to detect and/or mitigate the consequences of such an integrity loss.
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.o Introduction 1.1 Backaround Im Assust 1981, the NRC issued the resnits of a generic review of pipe breaks in the BTR scram system piping in NUREG-0803 Generic Safety Evaination Report Rosarding Integrity of BWR Scram System Piping. (Ref.1). The NRC concluded that for Mark I and Mark II containment plants the scram system piping is coceptable provided that steps be taken to:
(1) ensure the piping integrity, (2) citigate the consequences of a scram discharge volnae (SUV) break, and (3) l onvironmentally qualify the equipment required to detect and/or mitigate the ocasoquences of the break.
The need for mitigation measures and equipment qualification was predicated on an ostimated probability of SUV pipe break being sufficiently high that it could not bo dismissed.
Implicit in this approach is the argument that if the probability of a break in the SDV piping is sufficiently low, then consideration need not be given to mitigation features and equipment qualification for that particular break.
Using a defect rate of 3 x 10 5 per foot of pipe per year and an estimated SUV piping length of 2500 f t, the NRC calen1sted an SDV failure rate of 10-4 per plant year.
It noted that this value is extremely conservative since the SDV would be under load less than 1% of the time.
[ 4arlier report, NEDO-24342, GE Evaluation In Response To NRC Request Rogarding BTR Scram System Pipe Breaks (Ref. 2) used WASH-1400 (Ref. 3) vaines to ovaluate the SUV break probability.
It calculated the ratio of the SDV pipe longth to the LOCA sensitive piping length and took into consideration the dianster of the pipes.
(LOCA sensitive piping is that piping inside the containment that wonid result in a loss of reactor coolant in case of a break.)
This cyproach yielded a break probability of 3 x 10 8/ plant year taking into I
account the fraction of time the SUV piping is pressurized. Both NEDO-24342 and NUREG 0803 used estimated conservative generic plant data.
l 1.2 Persose It is the purpose of this report to perform a more detailed analysis of the failure probability of the SUV taking into account plant specific data, in order to d2=onstrate that an SUV failure resulting in a substantial leak which could threaten equipment required to detect and/or mitigate the leak is not a credible cycat.
Three different approaches will be used:
- 1) the NEDO-24342 approach
- 2) the NUREG-0803 approach tho fracture mechanics approach The last approach evaluates break probabilities by analysing the mechanism of crack growth while under repeated stress.
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h 2.0 Analysis 2.1 Descrintion of SDV System The scram discharge system receives the wat6r exhausted from the control rod drives (CRD) during a reactor scram. For a short time during and following each reactor scram, it contains reactor coolant at in11 reactor pressure. This section briefly describes the fundamentals of operation of the system.
The scram discharge system, which is depicted in Figure 2.1, consists of the CRD, the CED withdraw lines, the scram discharge volnae and the va.ves associated with the discharge volume.
During a scram, water from the volumes above the CRD pistons is discharged to the CRD withdraw lines.
It flows through the scram valves to the scram discharge volume. The scram discharge volume vent and drain valves are open during normal operation, and close automatically on receipt of a scram signal.
The discharge voInme partially fills with the water discharged from the CRDs.
Upon completion of a reactor scram, with all control rods in11y inserted, water leaking past the CRD seals from the reactor and water from the CRD pump continues to flow into the scram discharge volume. This flow continues antil the pressure in the scram discharge volnae is equal to the reactor pressure.
I When the scram signal is reset by the operator, the scram valves close and the scram discharge volnae vent and drain valves open. The scram discharge volume empties and returns to atmospheric pressure, configuring it for normal operation.
The scras valves and the scram discharge volume vent and drain valves are diaphragm actuated. These valves are designed to move into their scram positions when air pressure is removed. Motive air from the reactor building instrument air system is supplied to these valves via solenold-operated pilot valves actuated by the reactor protection system. TWo normally open manual isolation valves are provided at each hydraulic control unit to isolate the scram discharge volume from i
the CRD.
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The system, because of its simple design, provides a high reliability to scram:
and because the valves assume their scram positions when air pressure is removed, the reactor will be shut down autoretically if the air supply becomes unavailable.
l Figure 2.2 shows additional deta!!s of the scram discharge volnae itself. To comply with the SDY Safety Evaluation Report (Ref. 4) all SDV have or will hava two vont valves in series and two drain valves in series. Also, some systems currently have a relief valve. Table 2.1 sammarizes the details of each plant incInding pipe lengths as a function of diameter, design code used, number and types of joints and scram history. The piping system which is of interest for this study is that portion which extends from the check valves upstress of the SDV header up to and incInding the vent and drain valve piping.
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2.2 Fanit Tree Dinaram 2.2.1 31aeral Descrintion Figure 2.3 shows a fanit tree diarras for the SDV system shown in Figure 2.2.
The top event consists of any violation of the integrity of the SDV incInding pipe breaks and valve malfunctions that would result in water spilling into the reactor building. Teo events need to ocent; the SDV integrity must be breached and the reactor must be scrammed (i.e., the SDV and associated piping must be pressurized)
There are several ways that the SDV integrity can be breached: (1) a break in the pipe, (2) the relief valve fails open, and (3) two drain and/or two vent valves l
are stuck open. The relief, drain and vent valves are typically all piped to snaps in the basement. Depending on the size of the snap (s) and capacity of the snap pump (s), stuck open valves during a scram that are not or cannot be reset con 1d lead to eventual overflow of the snap. For this reason, the stuck open walves are considered as a failure of SDV integrity. However, the consequences are expected to be considerably less significant than those for a break.
2.2.2 SDV Pine Break Probability 2.2.2.1 Review of NEDO-24342 Anoroach D e SUV pipe break probability has been previously addressed in NEDO-24342
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(Ref.2). NEDO-24342 followed the approach used in Appendiz 3 of WASH-1400. It used the assessed break probability for a LOCA. However since the piping length for the SDV is different than the length of LOCA sensitive piping, the l
probabilities were modified by the ratio of SDV piping length to LOCA sensitive l
piping length. His approach resulted in a break probability of 3x10 4 por year assuming the SUV is constantly pressurized.
It estimated that a reactor is scrammed (SDV pressurized) It of the time. Thus an overall breat probability of 3x10 8/ plant year resnited.
2.2.2.2 Review of NUREG-0803 Anoroach NUREG-0803 need a different approach than that used in NEDO-24342. It estimated an SDV piping length of 2500 f t and mu* tiplied it by a failure rate of 3 x 10 'per foot per year to obtain a break probability of 10 4 per plant per year.
It also noted that the SDV is only pressurized 1% of the time but it did not factor it directly into the break probability.
If it were included, the resnIt would have been very r' ter to that of NEDO-24342.
2.2.2.3 fustion of Break Probability Usina Plant Soecific Data 2.2.2.3
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effic data, the SUV break probability was reevaluated following both M03 and the NEDO-24342 approaches.
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"Ic data that are being considered are the actual piping diameters, lengths, and sc. as histories. Following NEDO-24342 the SDV piping was first grouped into three diameter eines- (2', 2, 2
to 6 and )
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(See Table 2.1).
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(7 The ratio of these lengths to the length of LOCA sensitive piping of the same diameter grouping were evaluated. The total length of LOCA sensitive piping was taken to be 6000 ft (Ref. 5).
Following WASH 1400, the total length was equally apportioned among the three pipe groups. Thus each group consists of 2000 f t of pipe.
The median probabilities for a break in 2000 ft of LOCA sensitive piping from TASH 1400 are:
1/2 to 2 diameter 1 x 10 8/ plant year 2 to 6 diameter 3 x 10 */ plant year
>6 diameter 1 x 10 4/ plant year Using these vaines and plant specific data from table 2.1 the probability of a break was evaluated.
The break probability was also evaluated using an approach similar to that in NUREG-0803. This involves unitiplying the SUV pipe length by a defect rate of 3 x 10 ? per foot per year.
(Ref.
3).
The final break probability is evaluated by unitiplying this preceding product by the fraction of time the plant is scrammed, (i.e., that SUV is pressurized) based on the scram history for that plant.
2.2.2.3.2 Discussion of Resnits s'.'
\\.
The SUV pipe break probability was evaluated for the average plant and for the
limiting plant. The average plant refers to a plant having the average pipe lengths, number of scrans and scram duration from the data in Table 2.1.
The limiting plant is defined as the plant with the longest pipe lengths, the largest number of scrans and longest average scram duration based on the data complied in Table 2.1.
The resnits appear in Table 2.2; the following observations can be made:
a) Both the NEDO-24342 and the NUREG-0803 approaches yield very similar resnits.
Since the WASH 1400 break probability numbers used in NEDO-24342 are in part derived from the number of defects per foot per year (Ref. 3), the similarity of the two rosnits might have been anticipated, b) The break probabilities are about two orders of magsitude lower than those obtained in NEDO-24342 and NUREG-0803 This resnits from the fact that plant specific data show that the SUV system is pressurized much less than the 1% assumed in the previous analyses. Table 2.2 indicates the fraction of time scrammed (i.e., pressurized) for the average and limiting plant. This is the biggest contributor to the reduction in the break probability.
s) The dominant contributor to the break probability are pipes of less than 2
in diameter.
This is because most of the SUV piping length is small diameter piping; typically 70% or more is less than l
in diameter, with resulting low leakage flow rate.
If the conseguences of a small pipe break could be dismissed this would reduce the senseguential pipe break probability by at least another factor of 10.
i
. ~.. _. _ _ _.
1 t
~
However, even including small pipes, the resniting break probability based on either the GE or NRC approachen is, on the average, less than 2 x 10 ? per plant year.
Note that no credit has been taken for installation esaminations, the design code and piping class, the seismic class and inservice inspection. As indicated in Tr t e 2.1, these factors are present in all plants and would further reduce the bi : t probability.
2.J.2.4 Fracture Mechanics Anoroach The two previous methods used to determine the break probabilities are based on l
accanulated experience. An alternate method is the fracture mechanics approach which examines the failure of pipes due to growth of crack-like defects that may be introduced into welds during fabrication of the pipe. (Ref. 6,7) This method i
[
will be used to support the resnits from the experience approaches.
The fracture mechanics approach is described in Reference 6 and has been applied in Reference 7 to analyze the probability of a pipe break in an SUV.
It was found that the small pipes bound the large pipes in probability of failure. The small pipes are analyzed in this report following the method used in Reference 7, but using the SUV stress vaines from NEDD-24342 (Ref. 2).
l
. The fracture mechanics approach investigates the probability of low-cycle fatigue (1;[
causing through-wall crack propagation in the SDV piping system over the plant lif e t ime. This method assumes that piping failures ocent due to the growth of defects introduced into welds during fabrication of the pipe. These initial defects are considered to be randomly distributed in both the number of defects and their size. The fallare probability during a stress cycle equals the probability of a crack being larger than the critical crack size, given that a crack exists.
The stress levels assumed for this evaluation are the peak cyclic stresses in the SUV piping. The maximum stresses are (Ref. 2):
l Pressure 1.5 Esi Temperature 1.2 Esi Total 2.7 Esi l
Deadweight stresses are not included because they do not sontribute to fatigue.
Solamic stresses are not accounted for because they contribute a small number of cycles. Typically only one operating basis earthquake can be expected during plant lif e (p ( 10-8/ry) and the probability of a safe-shutdown earthquake is less than 10 8 per reactor year. Tater hammer effects on the SUV are not expected to be significant. Fast opening of the scram valve will esult in a simple compression (Ref. 4) of the SUV since it is empty or r*
. empty of water at the start of a scram. Opening of the drain or vent valvse is also not expected to produce significant stresses since they drain into air filled pipes at atmospheric j
pressure. This will result in simple decompression of the SDV.
Intergranular stress corrosion cracking, as pointed out in NUREG-0803 is not espected to be a potential failure mechanian, because the SUV is pressurized for only a short period of time.
. Scram frequencies of 9 (average) and 17 (maximum) per year are used (from Table 2.2).
His amounts to 360 and 680 cycles over the plant life, respectively.
D e initial crack distribution accounts for the probability that a crack exists and the size distribution of cracks given that a crack exists, n e crack probability in a weld of volume, V, is Poisson distributed according to
~
P, = 1 - e (1) where:
A = crack existence frequency 10 4/in8 V = 2x(ID)h8, inch 8 ID = Pipe ID, inch h = Pipe thickness, inch H e size distribution of cracks, given that a crack asists, is distributed j
exponentially with a complementary ensnistive distribut*cn i
P,f, = 0 x>h
~*
~
s/c =
e
-e 0Ixfh (2)
(f
- hh i
g_,
where k = average crack size, inch, and h here represents the maximum crack size.
D e SDV's undergo preservice proof testing. Positive resnits from this test insure that no cracks above a certain size, a, exist.
(If they omisted the pipe would fail during the proof test) Equation (I), thus, becomes:
P,f, = 0 x>h pM e
-a 01xIa P* *(a>x) =
P 3,,-h/k i
where:
l a
is the largest crack size that von 1d survive proof testing.
Each stress cycle increases the size of the cracks. He crack growth rate per cycle for stainless steel is given by: (Ref. 7):
gg = 10-* (AK)
- l da where:
11 dm = crack growth rate, inches / cycle AK = cycle stress intensity factor, ksi-in1/s l
l g,37, g 3
2 S **
- E ** )
2+Ce+C***U 4
(1-a)*/8 1
a=
g Ao = cyclic stress k
C = -1.H250 C = -6.21135 3
3 C = 4.79463 C = 1.79864 2
4 h e SUV consists of both stainless and carbon steel. D e above relationship applies to stainless steel but it will be applied to ' carbon steel as well for conservatism.
(
The critical crack size is given Sy, at which point n e crack continues to grow natil it reaches a critical size, a (Ref. 7):
the pipe is assaned to fail.
(
L' whre a
= load contro11ad stress = a + a g
dw i
e
= stress due to pressure f
l
.,, = stress due to deadweight i
a
= critical stress (flow stress)
(i
= (yield strength + tensile strength)/2
= 45ksi for stainless and carbon steel (Ref.7)
To evaluate the pipe failure probability consider the tolerable initial crack size, a,(n),
nis represents an initial crack size that won 1d just grow to the criticaI size after a stress cycles. D e probability of failure within a cycles is then equal to the probability of having a crack larger than s (n) at time t
zero. His is given by P (cond)(. "
I" > *t(*
f
-a,(s) / A
-a,/A 01ag(n) I a,
- e
-e 1-e 0
Otherwise
=
n e tolerable initial crack sizes, at*****
'8*
a ( " *t *-I ~U a = a (n-1) t da t
Finally, the seconditional average failure rate for the SDV systen can be found using f (cond)
- f" e
where L is the aanber of welds in the SDV, t is the life of the plant and P (cond) is evaluated over the life of the plant.
f I
{~,
This approach resulted in no failures for the aforementioned cyclic stresses (2.7 i
kai) for both the average and maximum number of scrans cases. The reason for Ohis is that the cyclic stresses are not sufficient to increase a crack from aThe minimum stresIss(the proof test orack size) to the critical crack size, a.
that would accomplish this are ~6.5 kai for 9 scrams / year'and ~5.5 kai for 17 scram /
year. This is over twice the peak cyclic stress orpected for a typical SDV. This result was obtained even with the use of the following conservative assumptions.
- 1) The influence of in-service inspection was ignored.
- 2) Only pre-service proof test was considered. In-service proof tests were ignored.
I 3)
Stress intensity factors were conservatively estimated assuming all cracks to be in117 circumferential.
I
- 4) The initial crack depth distribution for thick piping was used. This has a l
significant effect on the probability of having cracks greater than tolerable j
depth.
- 5) Upper bound estimate on f atigue crack growth characteristics were employed.
- 6) Conservative estimate of the flow stress was used.
(t
These fracture mechanics results support the outcome of the experience approaches which show that the probability of an SDV pipe failure is insignificant.
2.2.3 Probability of Stuck Open Valves As pointed out in section 2.2, water from the SUV could spill onto the reactor building basement floor if the two drain valves or the two vent valves or the relief valve (if the plant has one) were to remain open af ter a scram that could not be reset. This event would not be as serious as a break since no water would be sprayed at the equipment. Typically instead, the water would simply flow to the samp. At this time the reactor building is assumed to be accessible, allowing personnel to close the manual SUV isolation valves. Depending on the actual samp desist, flooding may eventually occur.
In sammary, the consequences of stuck open SUV vent and drain valves are not as severe as those for a break. Timely operator action before the flooding reaches vital equipment levels will ensure the operability of equipment for detection and mitigation of the valves' fallare.
Iowever, since flooding from such an event is conceivable the probability of stack open valves will be addressed. A typical configuration where the vent, drain and relief valves (if any) are piped to snaps, will be analyzed.
. {
2.2.3.1 Failure Este of Drain and Vent Valves Both the drain and vent valves are air actuated globe valves which close upon loss of air. The air is controlled by solenoid operated valves. The vont and drain valves con 1d remain open while the reactor is scrammed if (1) they stick open, (2) the air in them cannot vent, or (3) air from the instrument line is not cut off.
The probability of an air operated valve sticking open is 6.6 x 10 8/ demand (F.f. 8).
The probability, then, of two drain or vent valves in series sticking open is 4.4 x 10-1 per demand. For the average of 9 scrans per year the probability is 3.9 x 10 8 per reactor year; for the maximum of 17 it is 7.4 x 10 8/ry.
The air to the vent and drain valves are normally controlled by two solenoid operated valves configured as shown in figure 2.4.
Solenoid valves V3 and V4 each I
controls one vent and one drain valve. Under normal operating conditions the exhaust port is closed and the other two ports are open. This maintains air pressure on the vent and drain valves to keep them open. When a scram occurs, the air supply port should close and the exhaust port open. This would allow the air from the drain and vent valves to escape and thus close. A f ailure, however, can be postnisted where both the air supply and exhaust ports are plugged. This would (e1 prevent the air from the drain and vent valves from escaping and keep them in the open position.
The median probability of a solenoid valve being piogged is 8 x 10 s/ demand (Ref. 3).
In order for two drain or two vent valves to fail open (1) both solenoid valves need to be plugged or (2) one solenoid valve must plug and one drain or vent valve, not controlled by the piogged solenoid valve, must stick open. The som of the probabilities for the various combinations is 2.2 x 10 '/
demand. For 9 scrams / year it becomes 2 x 10 8/ry, for 17 scrams / year it is 3.7 x 10 8/ry.
Given a scram signal, the air to two drain or two vent valves is maintained only if all four valves fail in the so-scram position. The median probability for a solenoid valve to fail to operate is 1 x 10 8/ demand (Ref. 3).
The probability for four valves to not operate is thus 1 x 10 ss/ demand. Given 9 (17) scrans per year, the probability of the air mot being cut off is 1 x 1011 (2 x 10 51).
In summary, then, the probability of either two drain or two vent valves failing open is 6 x 10 8/ry for nine scrans a year and 1 x 10 8/ry for 17 scrans a year.
1
,___,..s_
-. ~. #
,__m__
m.
( !
2.2.3.2 Failure Rate of SDV Relief Valve Some plants are equipped with an SDV relief valve as shown in figure 2.2.
It was originally installed to comply with ANSI B31.1 for occasional over pressurizations. It was not, and is not specifically required for this system because the SDV pressure is limited to that of the reactor, which has its own pressure relief valves. The typical nominal opening set point is 1250 psig with a discharge capacity of 75125 spa at 1375 psig. This flow rate is within the capability of most (if not all) snap pumps..For the valve to fail open, the pressure would have to exceed its setpoint and then it wonid have to fail to ressat. Events that will cause the pressure to exceed 1250 psig are transients such as closure of all Main Steam Isolation Valves (MSIV) with flux scram (i.e.,
failure of four scram position switches), or failure of several relief valves during a pressurization transient such as turbine trip without bypass.
To estimate the probability of stuck open SUV relief valve, consider the closure l
of all MSIV transient. The frequency of all MSIV closure with position switch scram is ~ 0.5/ year (Ref.
9).
The probability of a position switch failing is estimated to be 10 8/ demand (Ref. 3).
Scram will not occur if two switches fail simultaneously; the probability. is 10 d/ demand or 5 x 10 8/ year. The probability l
that a relief valve won't ressat is ~5 x 10 s/ demand (Ref.8) (It is assumed to be I
similar to that for a primary relief valve) or ~ 2 x 10 8/ year. Thus the
?-
probability that the relief valve will stick open is ~1 x 10 5/ year for closure of (3
all MSIV with finz scraa.
The probability of a stuck open SUV relief valve for other events such as Turbine Trip without bypass with failure of several primary relief valves to open is even l ow er.
The probability of the SUV sticking open is thus conservatively estimated to be 1 x10 '/ year.
I 2.2.4 Other Considerations Figure 2.2 shows that the SUV system has several calibration valves that are normally locked closed. In addition, the end of each calibration line is capped.
The only credible way that a severe leak con 1d occur from this line is either from a in11 break or from failure to fully close the valve and recap the line. The former event has already been included nader pipe break. The latter depends on the quality of inservice inspection. The NRC through NUREG-0803 has mandated that
aurveillance, maintainance, inspection or modification procedures which l
conceivably have the potential for defeating SDV integrity be reviewed (or modified, if necessary) by licensee on a plant-by-plant basis. These plant-specific reviews should verify that all such procedures contain sufficient guidance to ensure that the loss of SUV system integrity will act ocent at times when such integrity shon1d be available. These actions should preclude the valve being left open and the end of the pipe being nacapped.
O e
_...m- _ _ _ _ _, _ _ _ _ _.. -, _ -
f i
2.2.5 Probability of Breach of SDV interrity The probability of loss of SUV integrity is the sum of the probabilities of pipe failure and valve failure. Based on the calenistions previously discussed these probabilities are:
Failure mode Probability / Reactor year Averare Plant Limitina Plant l
l Pipe Break 1 x 10 5 6 x 10 '
(Table 2.2) i Vent valve open 6 x 10 8 1 x 10 s (Section 2.2.3.1)
Drain valve open 6 x 10 8 1 x 10 s (Section 2.2.3.1)
Relief valve open 1 x 10 '
1 x 10 '
(Section 2.2.3.2)
All other Nen11mible Nen11mible (Section 2.2.3.3)
Total
~1.2 x 10 8
~2 x 10 8 These vaines are based on the scrans not being reset.
NUREG-0803 conservatively estimated the probability of failure to reset scram in 30 minutes at ~.5.
This high vaine was used because of the nacertainty in the post-leak environment that might contribute to the inability to reset.
(I This argument, however, is not as applicable in the case of stuck open vent or drain valves as it is to pipe break, since valves are not spraying nacentrollably in the air. Rather, they are discharging into snaps. In this case the operator f ailure to reset will most likely be the dominant failure-to-reset.
NUREG-0803 used an apper bound vaine of 0.02 for operator f ailure to reset.
l Thus, using a fa!!are to reset probability of 0.5 in the case of pipe breaks and 0.02 in the case of valve fallares, the probabilities of non-isolatable leaks are:
l Failure Mode Probabilitv/ Reactor year Averare Plant Limitina Plant Pipe Break 5 x 10 8 3 x 10 '
Vent Valve Open 1.2 x 10 '
2 x 10-5 Drain Valve Open 1.2 x 10 5 2 x 10-7 Relief *.1ve Open 2 x 10
- 2 x 10 8 Total
~3.0 x 10 5
-7 x 10 7
'e.
(
3.0 Summary and Conclusions NUREG-0803 requires the equipment used to detect and/or mitigate the consequences of a loss of SUV integrity event be qualified for the environmental conditions of that event. This study concIndes that environmental qualification is not necessary due to the low probability of a breach in SDV integrity. It also follows that there is a low probability of core damage resulting from such a breach.
The loss of SUV integrity can occur from any of four failure modes: (1) rupture of the SUV piping systream of the vent and drain valves, (2) failure of the redundant vent valves to close following a scram, (3) failure of the redandant drain valves to close following a scram or (4) failure of the SDV relief valve. The first failure mode was investigated using methods similar to those used in NUREG-0803 and NEDD-24342. Actual plant data on SDV pipe size and scram frequency was i
considered for these two approaches. The calculated break probabilities from those two approaches was compared to the calculated probability using a fracture mechanics approach and the resnits were shown to be consistent.
The probabilities associated with failure of the vent or drain valves to close were calculated based on previous operating history with this type of valve. The probability of an SDV relief valve failure to close was small relative to the l
other failure modes due to the relatively low frequency of challenge to this valve.
k.7 Consideration was given in the probability analysis to the ability of the operator I
to reset the scram. Due to the more severe environmental conditions, that probability is lower for the SDV pipe break than for the vent or drain valve failure.
The total probability of a breach in SUV integrity is the sum of the individual probabilities for each failure mode. That total probability was determined to be approximately 3 x 10 7 per reactor year.
The probability of a core melt event given the breach in SDV integrity was previously calentated and reported in Section 7.8 of NEDO 24342 and was determined to be 1.2x 10 d per plant year. Therefore, the probability of a breach in SDV integrity leading to a core melt is approximately 4 x 10 18 per plant year. This is significantly below the NRC proposed safety goal for core melt events which is 10 8 per reactor year.
The NRC, in NUREG-0803, stated that it was agreed that if the probability of core damage from the postnisted scenario (i.e., loss of SUV pipe integrity) was shown to be sufficiently small, no farther review, beyond verification of plant-specific response applicability, would be necessary. They further noted that
as the review progressed, it became evident that a sufficient data base did not exist to conservatively terminaf s the generic review on the basis of a quantitative risk assessment. Iowever, seasidering that the estimated core melt frequency following a loss of SUV integrity is considerably below the proposed NRC safety goal (by ~6 orders of magnitude), this significar". margin should be sufficient to aceonat for any perceived sparsity in the data base.
Therefore, it is conc 1sded that the breach of SUV integrity need not be considered for environmental quellfication of equipment in the reactor building.
rs T-bla 2.1 - Charmatorist!*e of t.
SUV System fer ths Varie"s Placts Parameter Fermi PB PB Duane Line-Fitz Pil-WNP Hatch Oyster Susque-Monti-NNP B run s,,
2 2
3 Arnold rick grim 2
2 Creek hanna cello 1
wick 1+2 Length of Pipe (ft) 1/2 -(2
1700 2023 2053 997 1439 1037 1015 1670 1684 1548 1992 1108 949 1761 2-6
120 5 82 9
158 140 18 370 293 123 278 181 244 327 303
)
6
2 90 11 414 188 170 257 18 147 274 100 289 71 94 241 Instal. Exam. Class 2
1 1
2 2
B31.1 B31.1 1
B31.1 (5) 2 2
Dess Code + Class 2
B31.1 B31.1 1
2 B31.1 2 Safety 2 2
(3) 2 B31.1 B31.1 + B31.1
+ GE Qual. 1
+ GE Class 1 + GE S31emie Design Class 1
1 1
1 1
1 2
1 1
(4) 1 1
1 1
In Serv. Insp. Class 2(1) 1 1
1 2
2 ASME ASNE 2
Surveill 2 1
1 None XI II for water Colded Joints 1044 941 ~905 1044 974 1205 683 957 1097 833 1024 Threaded Joints 0
0 0
0 0
0 0
0 0
0 0
0 0
0 Average Scram /yr (2) 4.3 7.5 8.2 (2) 7.3 9.5 (2) 17 (2) 6.8 12.6 17*
Average Scram Dur.
(2) 17.5 17.5 5.83 (2) 30 (2)
(2) 16 1
4 min.
( ) - Number la parenthesis refers to Note.
- Not Available
- Average scram /yr for both Brunswick 1 and 2
. Notes For Table 2.1
- 1) Visual test all piping while at hydrostatic pressure. Ultrasonic test scram discharge volume and instrument volume (25% of stress welds over 10 years).
Frequency is refueling cycle and Class 2 program.
- 2) Plant has not started up yet, so there is no scran data.
- 4) Uniform Building Code with following acceleration values:
.433 Horiz.
.293 Vert.
- 5) VT/PT for withdrawl lines, VT/RT for headers and instrument volume.
(
4
(.
TABLE 2.2 - BREAK PROBABILITIES USING EXPERIENCE APPROACH Parameter Avernme Plant Limitina Plant Length of SUV pipe (ft) 1/2 to 2 diam.
1496 2023 2
to 6
dias.
225 582
>6
dian.
183 11 Scrams / year 9
17 Total time to reset 91 285 per year (min)
(n Fraction of time scrammed 1.7 x10 d 5.4 x 10 4 1
Probability (NEDO-24342)(
1.3 x10 '/ reactor year 6x10 5/ reactor year Probability (NUREG-0803)II' 1.0 x 10 '/ reactor year 4.2x10 '/ reactor year
( ) - refers to Notes.
i G
e Notes For Table 2.2
- 1) Fraction of time scrammed is the Total time to reset per year divided by the number of minutes in a year.
- 2) Probability (NEDO-24342)
=[(L x 10 8) + (L x 3x10 4) + (1 I I W z F /2000 3
2 3
3 where: L = Length of SUV piping of 1/2 to 2 diameter 3
g = Length of SUV piping of 2
to 6 diameter i
L = Length of SDV piping of
>6 diameter 3
F = Fraction of time scrammed 3
- 3) Probability (NUREG-0803)
- ~' *
= (L3+L2 3
1
(!
i l
[:
i
-6 o
NORM AL POSITION
(-
REACTOR ONTAINMENT SOUND ARY
~
O i
"7
(MARKI/II)
A MYDR AULIC CONTROL UNIT SOUND ARY j
6/
CRD WITMDR AW LINE
=
P STON =%
i SCRAM h " ""1
- J VALVE 102 DISCHARGE fe VALVE 101 llSQLATION VALVE)
VALVE g
g, usOLATION VALVEl I
g4 l
pa SCRAM J g g
PILOT
=
s lp VALVE
- FINLET SCRAM 3 NE' l
4 VALVE W
l:
l s
~
l 1r 11 11 1r ts 11
'9 er si so it is n
=
guggg l
l lN l"2h SCRAM VALVE 2
l l
RISER l
VALVE 112 l
VALVE 113 (ISOLATION VALVE) i L__
_aso.w a~v3v.s _ _ _ _ _ _ _
___._ _J l
({
CHARGING WATER OTHER HCUs %
VENT.
DISCHARGE SCRAM HEADER 37 PILOT VALVE (Typical of two)
AIR SUPPLY
% g ff
<k NE b
[.
~
DRAIN 1 Q
h 11 ft 11 ff 11 ff if fi if If II fl 14 II L BeORMAL PLOWPATH Solenoid Valve
~De-Energized
- Typical of two
" Pos,ition
~(Dot Indicates
-Energized Position Figure 2.1 Simplified Schematic of Control Rod Drive System
~~
I
.o o
g g
Vent
(
Z l
I HCU
~Q I
I 1
I L _y J
l Other Italf l
}
_ _ _ _ of SDV X
m
=
a u
a l
lO Other Half of SDV Relief b alve V
]N H
l Drain I,
Figure 2.2 Typical Scram Discharge Volume Configuration (Simplified)
(
Less of SDV Integrity while Reactor Scramed f
Reactor Scram Drain Valves are elief Vent Valves are Pipe y,),
Open During l
l
{penDuring Scram 63 Cl
(:
Drah Air ir Vent Va'1ves\\
Not cut of Not cut of Valves
' Stick I
uring Scram During Scram Stick Open l
l l
Figure 2.3 - Fault Tree For Loss of SDV Integrity 9
-p
5
(
Backup Scram Valve Scram Signal Scram Signal Ins trument l
V3 V2 l
V3 To one vent and one drain Ai r kgd kd g%
y valve g
( l' g
Signal Exh aus t
(
V4 To one vent and one drain 4
k valve Exhaus t Figure 2.4 Simplffed Diagram Of A Typical ~SDV Instrument Air Control. The position showi is the no-scram position. The dot represents the port thet will close upon receipt of the scram signal.
r I
(
- 4.0 References t
- 1) ' Generic Safety Evaluation Report Regarding Integrity of BTR Scram System Piping', NL h 0803, August 1981.
- 2) L.F. Fidrych, R.L. Gridley, 'GE Evaluation In Response To NRC Request Regarding BTR Scram System Pipe Break', NEDO-24342, April,1981
- 3) ' Reactor Safety Study', WASH-1400, (NUREG-75/014) October 1975.
- 4) NRC Nemorandna, Generic Safety Evsination Report - BTR Scram Discharge System', December 1980.
- 5) Farmer, F.G. et.a1., ' Screening Values For National Reliability Evaluation Program Reliability App 11casions', Preliminary Draf t, April 1982.
- 6) ' Review and Assessment of Research Relevant to Design Aspects of Nuclear Power Plants Piping Systems', NURED-0307, July, 1977.
l
- 7) J.S. Abel, 'Onad Cities Station Units 1 and 2, Dresden Station Units 2 and 3 Plant Specific Response to NUREG-0803', letter to T.J. Rausch, Jan. 25, 1982.
I *c:
- 8) ' Data Summaries of Licensee Event Reports of Valves at U.S. Commercial Nuclear Power Plants', NUREG/CR-1363 vol.3.
- 9) 'ATTS: A Reappraisal, Part 3: Frequency of Anticipated Transients,' EPRI NP-2230, January 1982.
- 10) ' Safety Gos1s For Nuclear Power Plants: A Discussion Paper', NUREG-08SO, February 1982 (Draft).
i 1
l l
l l
-,,--,w,-
--,---e
--, -we.
~
e.
,e
~.
o r-w-
-~~ --
w---
=
s o e t
{
APPENDIX A This report applies to the following plants whose owners participated in the report's development.
Boston Edison Co.
Pilgrim Carolina Power + Light Co.
Brunswick 1 and 2 Detroit Edison Co.
Fe rmi 2 Georgia Power Co.
Batch 2 GPU Nuclear Oyster Creek Iowa Electric Light and Power Co.
Duane Arnold Niagara Mohawk Power Co.
Nine Nils Point 1 Northeast Utilities Millstone Northern States Power Co.
Monticello
( ;.
PASNY Fitzpatrick Pennsylvania Power + Light Co.
Susquehanna 1 and 2 Philadelphia Electric Co.
Peach Botton 2 Peach Bottom 3 Limerick 1 and 2 Public Service Electric + Gas Co.
Hope Creek 1 Wastington Public Power Supply System WNP-2 l
+
._-._m
... '