ML20039F682
| ML20039F682 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 01/04/1982 |
| From: | Kay J YANKEE ATOMIC ELECTRIC CO. |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| TASK-03-03.C, TASK-03-05.A, TASK-03-05.B, TASK-15-04, TASK-15-08, TASK-15-19, TASK-15-4, TASK-15-8, TASK-3-3.C, TASK-3-5.A, TASK-RR FRY-82-1, NUDOCS 8201130217 | |
| Download: ML20039F682 (19) | |
Text
.
YANKEE ATOMIC ELECTRIC C0DPANY f'
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h 1671 Worcester Road, Framingham, Massachuset:s 01701 2.C.2.1
,Yaluxe,s ryR 82-1
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January 4,1982 du
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RECE!VED United States Nuclear Regulatory Commission 2 IOO2 b Wa shing ton, D. C.
20555 k~
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im Attention:
Mr. Dennis M. Crutchfield, Chief Operating Reactors Branch #5 0
Division of Licensing
Reference:
(a) License No. DPR-3 (Docket No. 50-29)
Subject:
SEP Topic Assessment Completion
Dear Sir:
Enclosed please find our assessments of *.he following topics:
III-3.C In-Service Inspection of Water Control Structures.
XV-4 Loss of Non-Emergency A-C Power to the Station Auxiliaries.
XV-8 Control Rod Misoperation (System Malfunction or Operator Error).
XV-19 LOCA Resulting From Spectrum of Postulated Piping Breaks f
Within the Reactor Coolant Pressure Boundary.
Schedules for submittal of information on Topics III-5. A and III-5.B have been revised to 2/15/82 and 1/15/82, respectively. We trust this information is satisf actory; however, if you have any, questions, please contact us.
Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY
- 0. K J. A. Kay Senior Engineer - Licensing JAK: dad Enc losure s Q
8201130217 G20104 f
bi \\
gDRADOCK 05000029 PDR
g YANKEE NUCLEAR POWER STATION ROWE, MASSACHUSETTS SEP Topic III-3.C, In-service Inspection of Water control Structures
1.0 INTRODUCTION
The objective of this topic is to assure that water control structures of a nuclear power facility (i.e., dams, reservoirs, conveyance facilities) are adequately inspected and maintained so as to preclude their deterioration or failure which could result in flooding or in jeopardizing the integrity of the ultimate heat sink for the facility.
2.0 CRITERIA Regulatory Guide 1.127 states that:
In-service inspection should be performed at periodic intervals to check the condition of the water control structures and evaluate their structural safety and operational adequacy.
3.0 ASSOCIATED SEP TOPICS o
II-3.A Hydrologic Description o
II-3.B Flooding Potential and Protection Requirements o
II-3.C Safety-Related Water Supply (Ultimate Heat Sink) 4.0 DISCUSSION Water control structures at the Yankee plant are limited to the intake and discharge structures with associated features; the flood protection dike and stop logs; and Sherman Dam.
A formal inspection program for this site under Regulatory Guide 1.127 would include inspection of the concrete surfaces of the intake structure and discharge structure for deterioration, structural cracking, horizontal or vertical movement including abnormal settlements, heaving, deflections, or l
lateral movements or seepage. The interior surfaces of the intake conduits would be examined for erosion, corrosion, cavitation, cracks, joint separation, and leakage at cracks or joints. The intake and discharge structures and all features including the trash racks, would be examined for any conditions that may impose operational constraints such as silt or debris accumulation. The flood protection structures would be examined for damage or exposure of less wave resistant materials.
Sherman Dam is owned and operated by New England Power Company.
It is licensed and regulated by the Federal Energy Regulatory Commission. In l
accordance with the terms of the Federal Energy Regulatory Commission license, l
Sherman Dam has a formal inspection program including periodic inspection raports, It is not necessary for Yankee or the NRC to implement an inspection
SEP Topic III-3.C Page 2 program on Sherman Das because the owner has an ongoing inspection program under the jurisdiction of the Federal Energy Regulatory Commission.
The inspection program for the Yankee plant consists of a routine preventative maintenance program performed on the intake structure to detect any signs of deterioriation that could result in failure. Specifically, the.
intake racks are frequently inspected and cleaned throughout the year. Divers normally check for silt buildup.and inspect the conduits and surrounding areas and perform any maintenance required during refueling outages. During more than 20 years of plant operation, no significant operational problems have -
occured involving either the intake or discharge structures. The flood protection structures are high and dry except during the hypothesized Probable Mszimun Flood and therefore are not subjected to erosional or wave damage.
The flood protection stop logs are installed, from time to time, to ensure that the flood protection capability can be maintained.
5.0 TECHNICAL EVALUATION
The inspection program for Sherman Dam is under Federal Energy Ragulatory Commission jurisdiction and thus, subject to a formal inspection and reporting program. The program for inspection of water control structures in the Yankee plant has provided for the proper maintenance of these structures through many years of operation. Based on this review, Yankee concludes that the in-service inspection programs for water control structures at Yankee are acceptable and meet the intent of Regulatory Guide 1.127.
YANKEE NUCLEAR POWER STATION SEP Topic XV-4: Loss of Non-Emergency Power to the Station Auxiliaries I.
INTRODUCTION A.
Description of YNPS's Energency Power System The similarly-titled USNRC Standard Review Plan (SRP) applicable to SEP Topic XV-4 is Section 15.2.6 of Reference 1.
In the SRP scenario for Topic XV-4; an assumed complete loss of non-emergency ac power results in loss of all power to station auxiliaries. According to its definition, the safety objective for this SEP Topic is to ensure (1) that the reactor coolant pressure boundary is protected from overpressurization damage, and (2) that thermal margin for fuel integrity is maintained. In the Topic XV-4 scenario, for Yankee Nuclear Power Station (YNPS) a loss of ac power eventually results in loss of (1) all operating reactor coolant pumps (does not occur simultaneously), (2) steam-bypass to the condenser, and (3) steam generator feedwater pumps. Each diesel-generator has a fast-starting, two qrcle, sixteen-cylinder engine that is highly reliable and is coupled directly to its synchronous generator.
Under these conditions, emergency power is provided by three 500 kVA, 480 V diesel-engine-driven generators. The three station batteries provide independent starting power for the three diesel-enginea.
Only one of the three diesel-generators, however, is required to power the minimum set of shutdown equipment.
Each of the three diesel-generators provides standby power to a 4SO V emergency bus, and each emergency bus is normally supplied from a 480 V station service bus by two circuit breakers connected in series. Loss of normal voltage to an emergency bus will result in (1) automatic starting of the diesel-generator associated with that bus and (2) isolation of the affected emergency bus by tripping the two circuit breakers that tie the buses together. When the diesel-generator has reached its rated voltage, its breaker closes automatically to provide power to the emergency bus.
l Subsequently, by procedura and manual actior., any load within the capacity of the diesel-generator may be supplied by the emergency bus.
If I
e.ccessary, in order to pick up larger loads, the diesel-generator units may be synchronized and paralleled. The emergency electrical distribution system is discussed in SEP Topic VII-3, which has been addressed in the Reference 3 NRC draf t Safety Evaluation Report on syetems requirod for safe shutdown.
The immediate and simultaneous loss of all reactor coolant pumps is not
)
considered to be a credible event. Upon loss of all non-emergency power, the main generator will disconnect from the 115 kV of fsite transmission system.
Tko reactor coolant pumps will be operated for 30 to 60 seconds during generator coastdown.
B.
Loss of Non-Emergency Power A complete loss of the station power supply would occur if the 115 kV switchyard were disabled or if power is lost to both 115 kV transmission linee.
If power cannot be restored from the 115 kV lines, the emerger.cy I
l 1 [
diesel generators will be utilized to provide power (1) for charging to the main coolant system, (2) for providing service water to the heat sxchangers, (3) for supplying component cooling water, (4) for operating essential instrumentation and controls, and (5) for maintaining a full-charge in the station batteries. Operation of the auxiliary feedwater system, to remove decay heat f ollowing reactor scram, is discussed in the next section.
Recently, relevant information on this subject was provided to the NRC via e.he Reference 4 description of the auxiliary feedwater system modifications at YNPS.
Following a loss of offsite ac power, main coolant system pressure control is necessary to maintain an adequate subcooling margin and to ensure natural circulation cooling flow. Once natural circulation is achieved, pressure control would be accomplished by using the chemical and volume control system or pressurizer heaters powered by the emergency power system.
One full-capacity group of heaters is capable of maintaining a hot shutdown condition, as discussed in the Reference 3 evaluation of safe shutdown systems.
The Reference 5 letter to the NRC indicated that operating experience has demonstrated that the operation of one group of pressurizer heaters (37.5 kW) is required to compensate for heat losses from the pressurizer with normal spray flow at hot standby conditions. The ability to maintain natural circulation cooling flow under emergency conditions would require less capacity than for normal operations.
Reference 3 also discusses a Westinghouse study performed to determine minimum heater requirements without offsite power.
Extrapolation of the results of this study confirmed that the power to supply this required heater capacity is available from emergency power sources.
The Westinghouse study also determined that the capability to supply emergency power to the heaters within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> would prevent loss of subcooling in the primary system following a loss of offsite power. The time required to accomplish this, giving due consideration to all requirements of plant operating procedures, is 15 minutes from the occurrence of the loss of offsite power.
II.
DECAY HEAT REMOVAL Following loss of offsite power, decay heat energy removal will occur by supplying auxiliary feedwater to the steam generators. Reference 3 calculations show that the Technical Specifications' requirement of 85,000 gallons total stored water inventory is sufficient to maintain a hot shutdown condition for 73 hours8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br />. The time required to achieve shutdown cooling system operation, however, is only about 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (72,000 gallons) according to Reference 3 results. Steam release occurs automatically via the main steam safety valves or manually via the atmospheric steam dump system.
Operation of one of the two safety-class 150 gpm, motor-driven auxiliary feedwater pumps provides sufficient feedwater for decay heat removal. These pumps can be si:pplied power from the emergency diesel-generators by remote-manual operation of circuit breakers. The charging pumps can also be made available in this manner, to supply feedwater following a loss of offsite ac power. Steam generator inventories are more than sufficient to provide time for establishing auxiliary feedwater flow before loss of secondary cooling capacity, as discussed in the Reference 4 letter to the NRC.
~2 r-
i
'III.
PROCEDURAL CONSIDERATIONS Following a total loss of ac power, operator actions are specified by.
operating procedurer so that emergency electric power is supplied to vital equipment.. - These procedural measures have been discussed in the Reference 3 ~
NRC draf t Safety Evaluation Report on saf e shutdown systems.
IV.
SINGLE ACTIVE FAILURE CONSIDERATIONS l
Single active failures considered for this event are (1) failure of a' diesel generator unit, and (2) fai16re of an atmospheric steam dump valve.
The auxiliary feedwater system is designed so that a single-failure will not prevent its functioning. One diesel-generator unit is capable of providing power to the minimum set of shutdown ' equipment.
Failure of a single atmospheric steam dump valve is discussed for Topic XV-1, Increase in Steam Flow Events.
V.
REACTOR PROTECTION ANALYSIS SEP Topic evaluations for loss-of-load (Topic XV-3), loss-of-feedwater (Topic XV-5), and loss-of-coolant flow (Topic XV-7), were submitted previusly via Reference 2.
The initial stages of a loss of non-emergency power to station auxiliaries is similar to a loss-of-load event with failure of the steam bypass system to function, which necessitates steam release via safety relief valves or the atmospheric steam dump system. However, the reactor scram can be credited as a direct result of low main coolant flow rate, and process variables such as main coolant pressure (increasing) are not relied upon for the scram. Thus, the event is less thermally-limiting than a locs-of-load.
without direct reactor scram.
Because the two reactor coolant pumps powered' by offsite lines are lost immediately, main coolant flow from-the two remaining pumps powered from the generator is briefly available while the generator coasts down after turbine-trip. This aspect of the loss-of-power event is similar to the 2/4 partial loss-of-flow event, except that reactor scram precedes the Icss-of-flow. Thus, the consequences of a loss-of-non-emerEency power are less limiting. Finally, until the delivery of auxiliary feedwater begins, the event is similar to a loss-of-feedwater.
Again, however, reactor scram occurs directly and is not delayed, for example, until a low steam generator level trip occurs. Thus, the loss-of-feedwater event is more limiting.. Reference 2 supplies the protection analyses for each of these evente, which meet the acceptance criteria of Section 15.2.6 of '
Reference 1 for loss of non-emergency ac power events. Thus, the consequences of loss of non-emergency ac power are bounded by the consequences of other analyzed events; namely, the loss-of-load, loss-of-flow, and loss-of-feedwater events.
Immediately following a loss of ac power with accompanying turbine trip and reactor. scram, no operator' action is required.except for verification of automatic responses to ensure reactor protection. Subsequently, if power cannot be restored, the operator controls plant cooldown using the atmospheric steam dump system and borates the main coolant system to prepare for cooldown to shutdown cooling system entry conditions.
VI.
CONCLUSION In conclusion, a review of (1) the emergency power system's design and (2) operating procedures shows that even if eingle failures are considered there exista sufficient system redundancy for decay heat removal in the event of loss of non-emergency power to the station auxiliaries.
Reactor protection is ensured for this event, and its consequences are bounded by the consequences of loss of-load, loss-of-cooling flow, and loss-of-feedwater events, which were provided to the NRC in Reference 2.
The SEP safety objective is satisfied.
VII. REFERENCES 1.
NUREG-0800, USNRC Standard Review Plan, Revision 1, July 1981.
2.
FYR 81-95, Letter from YAEC to NRC, SEP Topic Assessments, 30 June 1981.
3.
Docket No. 50-29, LS05-81-09-020, Letter from NRC to YAEC, Topic VII-3, Safe Shutdown Systems (Contractor's Evaluation), 6 September 1981.
4.
FYR 81-98, Letter from YAEC to NRC, Response to Action Items, 30 June 1981.
5.
Letter from YAEC to NRC, Resolution of TMI " Category A" Implementation Audit Outstanding Items, 9 April 1980.
-4
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YANKEE NUCLEAR POWER STATION SEP Topic XV-8:
Control Rod Misoperation (System Malfunction or Operator Error)
I.
INTRODUCTION SEP Topic IV-2, Reactivity Control Systems Including Functional Design and Protection Against Single Failures, is closely related to Topic IV-8.
Ihe assessment for Topic IV-2 was submitted to the NRC via Reference 1.
The events labeled " control rod misoperation" for Topic IV-8 include (1) inadvercently withdrawing one or several rods; (2) leaving one or several rods behind during bank withdrawal, and (3) inserting one or several rods inadvertently. In addition to these misoperation events, Section 15.4.3 of the Standard Review Plan, Reference 2, states that " single failures in equipment or errors in operation" should also be considered. Thus, the assessment submitted for Topic IV-8 is made directly applicable to the assessment required for Topic IV-2 via the Standard Review Plan language.
Single failures in the rod control system could produce the following occurrences, whose consequences were discussed in the Reference I submittal to the NRC:
1.
A single control rod may drop; 2.
A r.ontrol rod group may drop; 3.
I single control rod may fail to move upon command; 4.
A control rod group may fail to move upon command; 5.
A single control rod may move inadvertently; and 6.
Control rods may be malpositioned within their group.
Safety evaluations of reload cores are performed for the following events that are more limiting than the single failure events cited for Topic IV-2:
(1) uncontrolled group rod withdrawal, (2) single rod drop, and (3) single rod ejection. Most recently the Core XV Performance Analysis submitted via Reference 3 and approved in Reference 4, describes analyses demonstrating compliance with applicable design criteria for these events.
In addition, evaluations were made of Core XI continuous single rod withdrawals confirming that DNB ratios exceeded 1.30 in all cases.
Similar results can be expected for Core XV.
The methods used for the favorable Core XI scoping calculations of a single rod withdrawal event, were approved by the NRC in the Reference 5 amendment pursuant to the Reference 6 reload licensing submittal and are essentially the same methods used for Core XV licensing calculations.
Section II below discusses reactor protection considerations for the Topic XV-8 control rod misoperation events.Section III below provides the safety analysis for each event, including assumptions where applicable.
II.
RFACTOR PROTECTION CONSIDERATIONS Protection against the subject single failure events is provided by automatic scram on a high power trip signal, without crediting operator intervention to terminate the event in advance of scram.
Reactor scrams are also produced for high main coolant pressure and high pressurizer level indications.
In addition, the dropped rod system alerts the operator in advance of a core power return so that intervention is possible. Operator intervention is not required, however, to demonstrate acceptable performance for this event. In all cases, the substantial steady-state operating margin to DNB, manifest in design-value DNB ratios near 3.30, ensures that no violation of Condition II design criteria requiring that DNB ratios exceed 1.30 occurs for these events.
Furthermore, individual control rods are not routinely withdrawn during steady-state operation within the power range. Operator manipulation of a single rod, in violation of Technical Specifications, is unlikely during operations in the power range. During low power physics testing usually conducted at approximately 2% rated power level, when rod worth measurements require individual rod control, and also at power levels below 15 MWe by procedure, the low poser scram switch is set to LOW (neutron flux trip reduced i
from 108% to 35%) and the high startup rate trip at < 5.2 DPM is available to limit the consequence of single failures in the rod drive control system.
In contrast, analyses of group rod withdrawal accidents and rod ejection accidents are allowed to progress to the high power trip condition (108%) for purposes of calculating limiting DNB ratios.
Analysis performed since Core XI using currently approved methods demonstrates DNB ratios greater than 1.30 for the group rod withdrawal and single rod drop events. An alarm system is designed to sense a dropped rod and to alert the operator; however, DNB ratios will exceed 1.30 without the need for a reactor trip when a rod drops. The group rod drop event exhibits less limiting consequences than the single rod drop because local fuel power peaking is comparatively lower. These events are discussed in the Reference 3, Core XV Performance Analysis. Rod ejections, which are a type of limiting single rod withdrawal event, were discussed in the SEP Topic XV-12 submittal in Reference 7.
The analysis of the so-called limiting fault rod ejection events demonstrates that no fuel pin failures occur and that design criteria are not exceeded.
The large margin to DNB at steady-state conditions reported in Reference 3 ensures that results for the less limiting single failure events will satisfy design criteria for Condition II moderate frequency events. The group rod withdrawal and single rod drop analyses bound the consequences of single failure Topic IV-2 events, including control rod mispositioning and rods failing.to move upon command.
Design features that limit rod malpositioning and reactivity insertion, in addition to alarms, interlocks, and operating procedures that provide protection from control rod misoperation events are discussed in Section D of the Reference 1 submittal to the NRC of the Topic IV-2 assessment.
III.
REACTOR PROTECTION ANALYSIS A.
Group Rod Withdrawal Accident Section 7.2 of the Core XV Performance Analysis, Reference 3, reported that results provided in Reference 6 were still limiting. The Reference 6 analysis was conducted with (1) the GEMINI-II computer code for determining plant response, and (2) the COBRA-III-C subchannel analysis code for determining core thermal performance. The following assumptions were made for conservatism:.
1.
Design peaking factors were used including engineering conservetsses; 2.
Core average power was assumed to be at the high power trip point, 112% of 618 MWt; 3.
Main coolant pressure was assumed to be at the low end of the operating band, 1940 psia; 4.
Coolant temperatures were assumed to be at values consistent with steady-state operation at the high-flux trip setpoint.
The minimum DNB ratio calculated for the Core XI event was 2.30 for a reactivity insertion rate of 0.02 (10)~4 Ap /sec. The NRC approval of the Core XI analysis was obtained via Reference 5.
The minimum DNB ratio corresponding to Core XV for this event was determined to be greater than 2.30 because.of more favorable peaking factors, even when the higher core inlet temperature (5150F vs. 5110F) and lower steady-state design-value DNB ratio (3.13 vs. 3.24) for Core XV are accounted for.
Thus, for this event, the acceptance criteria that the DNB ratio must exceed 1.30 is satisfied.
B.
Control Rod Drop The single rod drop accident, also evaluated for each reload core, can produce increased power peaking due to local distortions in the core radial power distribution. The design of the reactor core ensures that DNB ratios will exceed 1.30 without the need for reactor trip when a rod drops. An alarm system is designed to sense a dropped rod and to alert the operator, who then attempts retrieval according to procedure. Analyses performed since Core XI have demonstrated DNB ratios greater than 1.30 for this single failure event without crediting operator action to manually reduce core power. The group rod drop accident will exhibit less limiting consequences than a single rod peaking factor increases are comparatively lower. Also, drop, because Fny for the most limiting DNBR conditions during operation within the power range, the control rod group worths throughout core life are sufficiently large that a group drop would result in a reactor scram on low loop pressure.
A bounding analysis for the single rod drop accident was performed for Core XI, Reference 6, and approved via NRC Reference 5.
The consequences of this event are an initial decrease in core power, system pressure, and coolant temperatures.
Depending on the worth of the dropped rod, a distorted power distribution could exist when reactor power subsequently increases to equal the steam flow power with the coolant temperature reduced in proportion to the dropped rod worth and moderator temperature coefficient. The dropping of a low worth rod can result in a faster power level recovery at a relatively higher reaperature, but the drop of a high worth rod is the more severe event because of its higher radial peaking effect.
The following assumptions for the bounding Core XI analysis are conservative for purposes of calculating minimum DNB ratios for this event:
1.
Maximum design radial peaking factors for the core, with calculational uncertainties were used; -
i l
l 2.
No credit for power reduction due to rod drop was assumed, so
'618 MWt was used (600 MWt + 3% uncertainty);
3.
Main coolant pressure was assumed to be at the low pressure trip setpoint, which is adverse for DNB ratio calculations; and 4.
The design core inlet temperature with uncertainty was assumed.
The minimum DNB ratio predicted for this event was 2.18, which is well in excess of the minimum allowable DNB ratio of 1.30.
Results of this analysis Core XV provided in Reference 3 showed that the lower peaking factors and linear heat generation rates for Core XV resulted in more favorable DNB ratios compared to Core XI results.
C.
Control Rod Malpositioning During critical operations in Modes 1 and 2, the Technical Specifications limits to 13 inches the malpositioning of control rods within a group. The basis for this limitation is to ensure the applicability of F H and Fq peaking factors assumed in safety analyses when calculating DNB ratios. Even if all control rods in the most worthy group were each misaligned by 18 inches from their assumed position, the reactivity and power peaking effects would not be more thermally limiting than those produced by a single rod drop accident.
D.
Control Rods Failing to Move-Upon Command Neither the failure of a single rod nor the failure of a rod group to move with manipulation of the rod movement switch are explicitly addressed for licensing submittals. The consequences of these events are bounded in the analysis of group rod withdrawal and single rod drop accidents.
E.
Single Rod Withdrawal The uncontrolled single control rod withdrawal accident (rod runaway) is not considered to be a design basis event and it has never been required or requested for NRC licensing submittals. However, scoping calculations for the uncontrolled single rod withdrawal accident were performed for Core XI.
Withdrawal of a single rod produces more localized power peaking than does withdrawal of a group of rods.
Thus, DNB ratios for the single rod withdrawal are comparatively low. Results of the Core XI scoping calculations showed that DNB ratios were greater than 2.0 for the single rod withdrawal accident.
In addition to routine parameter variations, shutdown margin requirements, power-dependent-insertion-limits, and allowable peak-rod LHGR limits have all changed favorably since Core XI analysis. Thus, DNB performance for single rod withdrawal events is bounded by the Core XI scoping analysis.
In order for a single f ailure to result in an inadvertent single rod withdrawal to a high power trip condition the operator must disregard numerous indications and the high power alarm while manipulating the rod movement -
switch. This could occur within the power range when repositioning a misaligned control rod.
Thus, operator error in failing either to terminate the withdrawal or to manually scram the reactor must occur before the single failure event. progresses to an uncontrolled single rod withdrawal accident resulting in a high power trip condition. At power levels greater than 130 MWe (about 75% power), the rod movement switch must be returned to the OFF position af ter each rod movement step of about 3/8 inch. Otherwise, interlock circuitry prevents continuous outward movement of rods during high power opera tion.
Under the ANSI N18.2 (1973) classification system used in the NRC Standard Review Plan, the uncontrolled single rod withdrawal event resulting from a single failure plus operator error is not a Condition II single f ailure event (moderate frequency; once per year), for which a DNB ratio in excess of 1.30 must be demonstrated. Rather,'a single failure event coupled with operator error is arguably a Condition III event (infrequent incident; once per plant lifetime). Ihe ANSI N18.2 design requirement for Condition III events is that they "shall not cause more than a small fraction of fuel elecents in the reactor to be damaged..."
Since this Condition III design requirement is satisfied for both rod ejection accidents, which are the extreme variety of a single rod withdrawal event, and by the Core XI scoping analysis, adequate protection is thus demonstrated for a single failure in the rod drive system coupled with operator error.
Westinghouse follows a similar rationale for the single rod withdrawal accident caused by a single failure in the rod control system, as discussed in Reference 7, Page 6-8.
The Westinghouse analysis indicates that DNB might occur for a single rod withdrawal due to localized power peaking, but that it would be limited to less than 5% of all fuel pins.
Likewise, the Core XI analysis predicted no occurrence of unacceptable DNB conditions.
If the operator terminates the event prior to reaching a high power trip condition, however, the single f ailure does not result in an uncontrolled single rod withdrawal accident. In this case, the Condition II design criterion requires that DNB ratios in excess of 1.30 must be demonstrated.
This requirement is satisfied because even the uncontrolled withdrawal of any entire group of rods to a high power trip condition does not result in a DNB ratio of less than 2.30.
Therefore, an inadvertent single rod withdrawal during manual shimming operations is an acceptable Condition II event assuming operator termination of the event prior to a high power trip. Without crediting operator termination, the Core XI studies predicted DNB ratios of not less than 2.0.
In conclusion, adequate reactor protection for single failures or misoperation that can result in inadvertent single rod withdrawals can be demonstrated by:
(1) the group rod withdrawal analysis for Core XV if operator response to terminate the event is credited, or (2) analysis of single rod withdrawals similar to the Core XI studies (similar methods and expected results) if operator error permits the accident to progress to a high power trip condition. Condition II design requirements are met in either case.
IV.
SUMMARY
AND CONCLUSIONS The possibilities for control rod misoperation events, including operator error and single failures, have been reviewed and analyzed. In all cases, fuel thermal design margins are adequately protected by automatic protective action. The substantial steady-state operating margin to DNB, near 3.30 for Core XV fuel, ensures that no design criteria violations occur and that DNB ratios exceed 1.30 as required for these events. _ -_ ______ _ _ _. -.....
V.
REFERENCES 1.
FYR 81-136, Letter from YAEC to NRC, SEP Topic Assessment Completion (Topics IV-2 and IX-3),17 September 1981.
2.
NUREC-0800, USNRC Standard Review Plan, Rev. 2, July 1981.
3.
FYR 81-52, Letter from YAEC to NRC, Core XV Refueling, 26 ' March 1981.
4.
Docket No. 50-29, LS05-81-07-71, Letter from NRC to YAEC, 22 July 1981.
5.
DOL /AEC Letter to YAEC, Amendment No. 9, 30 July 1974.
6.
Docket No. 50-29, Letter from YAEC to NRC, Proposed Change No.115 Core XI Refueling, 29 March 1974.
7.
FYR 81-95, Letter from YAEC to NRC, SEP Topic' Assessments, 30 June 1981.
8.
NUREC-0138, Staff Discussion of Fif teen Technical Issues Listed In Attachment to November 3,1976 Memo from Director, NRR to NRR Staff, November 1976.
2
YANKEE NUCLEAR POWER STATION XV-19 LOSS OF COOLANT ACCIDENTS RESULTING FROM SPECTRUM OF POSTULATED >
PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY -
+
-RADIOLOGICAL PORTION I.
INTRODUCTION The safety objective of this topic is to assure that the offsite doses.from a loss of coolant accident are within the guideline values of 10 CFR Part 100.
T II.
REVIEW CRITERIA Section 50.34 of 10 CFR Part 50, " Contents of Applications: Technical Information,"
requires that each applicant for a. construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, i
and components of the facility with the objective of assessing the risk:to' public health and safety resulting from operation of the facility. The loss of coolant-accident and subsequent release of radioactive material is one of the postulated j
accidents used to-evaluate the adequacy of-these' structures, systems, and components with respect to the public health and safety.
l 10 CFR Part 100 provides the acceptable dose consequences for siting of nuclear power plants.
III. RELATED SAFETY TOPICS 4
4 Topic II-2.C, " Atmospheric Transport and Diffusion-Cha'racteristics for Accident Analysis" provides the meteorological data used for' calculating the offsite dose 4
consequences.
IV.
REVIEW GUIDELINES The design basis loss of coolant accident was reviewed following the assumptions and procedures outlined in Standard Review Plan (SRP) Section 15.6.5 - (Appendices e
A, B & C) and Regulatory Guide 1.4.
The dose to an individual from a postulated loss of coolant accident should be within the exposure guidelines of 110 CFR 100.
V.
EVALUATION The doses resulting from all postulated pathways nre calculated and cumulatively compared with the appropriate exposure guideline values to confirm the accepta--
bility of the nearest exclusion area boundary (EAB) and low population zone (LPZ) outer boundary and the adequacy of the engineered-safety features (ESP) provided for the purpose of mitigating potential accident doses.
The following exposure; pathways have been evaluated:
1.
Containment' leakage,. including the contribution.from containment purge valves
' during. closure.
- 2.. Post-LOCA leakage from ESF systems outside containment.
3.
Post-LOCA hydrogen purge from containment.
4.
Direct radiation from the containment.
mm
~.
4 The postulated consequences are given in Table XV-19-A.
Tha assumptions tnd input parameters used in calculating the potential consequences are given in XV-19-B.
VI. -
CONCLUSIONS Based on the above evaluation; we conclude that the radiological consequences of
-loss of coolant accident are within the guidelines of 10 CFR 100 and the Standard Review Plan criteria are met.
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REFERENCES 1.
NRC to YAEC, September'4, 1981, SEP Topic II-2.C, Atmospheric Transport and Diffusion Characteristics for Accident Analysis - Yankee ~.
2.
YAEC to NRC, December 31, 1979, Imssons Learned Short-Term Requirements Item 2.1.6a Systems Integrity for Containing Radioactive Materials Outside of Containment.
3.
YR OP-2658, Operation of the Post Accident Vapor Containment Hydrogen Control System.
Rev. No. 9
TABLE XV-19-B ASSUMPTIONS USED FOR THE LOSS OF COOLANT ACCIDENT ANALYSIS Power Level 600 Mwt Operating Time 3 years Fraction of Noble Gases Available for Release 100% @ T = 0 Fraction of Halogens Available for Release 50% @ T = 0 Iodine Fractions Elemental 917 Organic 4%
Particulate 5%
Halogen Plateout Time Constants Elemental 2.5 hr-l*
Organic 0.0 Particulate 2.5 hr-l*
3**
Containment Volume 8.6 (+5)ft Containment Leak Rate 0.2% per day 0<t<24 hr 0.1% per day t>24 hr Distances to Applicable Receptors Exclusion Area Boundary 3100 ft up and downstream Low Population Zone 2 miles up and 6 miles downstream y/Q's (From Ref. 1) 0-1 hr EAB = 2.8 (-4) sec/m3 1-2 hr EAB = 2.3 (-4) sec/m3 3
0-8 hr LPZ = 2.8 (-5) sec/m,
8-24 hr LPZ = 1.9 (-5) sec/m3 24-96 hr LPZ = 1.6 (-5) sec/m3 96-720 hr LPZ = 1.0 (-5) sec/m3 3
Breathing Rate 0-8 hr 3.5 (-4) M /sec 3
8-24 hr 1.8 (-4) M /sec 3
24-720 hr 2.3 (-4) M /sec Containment Purge Valve Closure Dose from purge valve closure is zero ESF Leakage Long Term 20 gallons / day for 30 days (Ref. 2)
Short Term 50 gallons / min for 30 minutes 0 T =
Passive Failure 24 hr Fraction of Iodine released 10%
- A = 2.5 hr-1 Until Airborn Containment Concentration Reduces by a Factor of 100.
5
- 8. 6 (+5) = 8. 6 x 10
_ TABLE XV-19-B ASSLHPTIONS USED FOR THE LOSS OF COOLANT ACCIDENT ANALYSIS (Continued)
Hydrogen Purge The Hydrogen concentration is not expected to reach 4% until 139 days after the LOCA (Therefore the 2-hour EAB and 30-day LPZ dose contribution from Hydrogen purge are zero)
(Ref. 3) 1 l
._-__e..-
O TABLE XV-19-A s.
CALCULATED DOSES FOR LOSS OF COOLANT ACCIDENT Doses, (Rem)
Thyroid Whole Body Exclusion Area Boundary (EAB) 2-Hour Doses:
1.
Containment Leakage
- 6. 2 (+1) -
5.7 (-1)-
2.
ESF System Leakage outside 2.7 (+0) 8.5 (-3)^
Containment 3.
Post LOCA Hydrogen Purge N/A N/A 4.
Shine (Direcc Radiation)
N/A 1.8 (-1)
Total 6.5 (+1).
7.5 (-1)
L Low Population Zone (LPZ) 30 day Doses:
1.
Containment Leakage 2.1-(+1) 1.8 (-1) 2.
ESF System Leakage Outside 7.8 (+1) 6.0 (-2)
Containment 3.
Post LOCA Hydrogen Purge N/A N/A 4.
Shine (Direct Radiation)
NEGLIGIBLE Total LPZ 9.9 (+1) 2.4 (-1)l
- 6.2 (+1) = 6.2 x 10I = 62.
_