ML20235P276
| ML20235P276 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 10/01/1987 |
| From: | YANKEE ATOMIC ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20235P208 | List: |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 NUDOCS 8710070229 | |
| Download: ML20235P276 (66) | |
Text
_ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _
YANKEE ATOMIC ELECTRIC COMPAhT REGULATORY GUIDE 1.97 NUCLEAR INSTRUMENTATION ANALYSIS
/
Yankee Atomic Electric Company Nuclear Services Division 1671 Gercester Road Framingham; Massachusetts 01701 5670R/23.11 pu"PosnBu8lgg9 P
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i TABLE OF COvTENTS
(
i Page ,
1.0 INTRODUCTION
....................................................... 1 1.1 Background and Report Purpose............................... 1
) 1.2 Issue Review................................................ 2 1.3 Summary......... 3..........................................c 2 s
P 2.0 APPROACH AND RESULTS.............................................. 3 1
2.1 Genera 1..................................................... 3 2.2 Technical Approach.......................................... 3 2.i.1 Establish iandidate Sequences (Step 7 ). . . . . . . . . . . . . . 3 2.2.2 Develop Lohical Categorization (Step 2)............. .} ,
2.2.3 Identify Specific Events (Step 3)................... >
2.2.4 Screen Potential Events (Steps 4 and 5)............. 9 ,
2.2.5 Steps 6 Through 9................................... 10 l 2.3 Potential Improvement...................................... 11 r.
3.0 NUCLEAR INSTRUMENTATION SiSTEM DESCRIPTION........................ 49
}. ,
4.0 CONCLUSION
S AND RECOMMENDATIONS................................... 54
'I 4.1 Conclusions.........i........................................ 54 l 4.2 Recommendations......{...................................... 55 i
f 5.0 R E F ER EN C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ' . '. . . . . . . . . . . . . . . . . . . . . . . . 56 ATTACHMENT A: Nuclear Instrumentation and Reactor Protection System.... 57 s
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, y 1 LIST OF TABLES
~ Number' Title Page 2.1 Gqvital Technic l' Approach 13
, v
-2.2 E4tablishing. Logical Search, Categorization, and 9
Inve'stigation of Risk Attributable to Inadequate g Reactivity Control- 14 l~ \
- 2. 3 ' Candidate Sequences 15- 1 2.4 Summary of Reactivity Addition Methods Considered 20 2.5 Summary of Specific Event Scenarios Exam 2ned 21 2.6 Summary of Screening Criteria 26 2.7-A Basis. for Dispdition of Boron Dilution '
Event Scenarios 27 4 '
2.7-B Basis for Disposition of Control Rod Withdrawal Event 1 Scenarios
- 36 y .
1 .s e 2.7-C. Basis for Disposition of MCS Cooldown Event Scenarios 38
- 3. 4 2.7-D Basis for Disposition cf Xenon-135 Deny Event Scenarios 45 2.8 Sequence of Events for Charging Dilution of MCS 46
- 2. 9 ' ' Termination of the Charging Dilution Event 48 i
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i LIST OF FIGURES '
' Number ' Title -Page l
2-1 Flow Chart'of. Cost-Benefit Analysis Methodology 16 c ',; 2-2, ,- Event Tree; 49
,q w, o.k 2-54)L Boron Dilution Flow Path 50
- 3-1
- YNPS Nuclear Instrumentation and Rod Position Power Supply 52
,3! 3-3' . Safe Shutdown System (SSS) Nuclear Instrument and Power
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Supply- 53 4
I I
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-iv-3670R/23.11
.1. 0 INTRODUCTION 1.1l Background and Report Purpose Btckground In' response to the TMI Action Plan (NUREG-0737, Supplement 1). Yankee
' Atomic Electric Company (YAEC) was required to address Regulatory Guide 1.97 (Rs6 rence 1). Each requirement has been resolved, except those related to dasign and qualification of Nuclear Instrumentation (NI). Reference 2 stated
.tha most recent staff position:
... post-accident neutron flux monitoring is.a ' key variable' for-detecting an uncontrolled approach to criticality, be it inadvertent boron dilution or other reactivity addition situation resulting from accidents, and for the determination that an accident has been and is maintained mitigated. Since key variables are classified as Category 1, the licensee should commit to installation ...
Report Purpose This report reviews the need for changes to NI using a risk-based analysis. This risk-based approach is based on methods previously used to support resolution of SEP Topic III-2, " Wind and Tornado Loadings " and SEP Topic III-4A.. " Tornado Missiles." Reference 3 provided the staff safety evaluation of these SEP Topics and concluded:
- 1. "The licensee has appropriately assessed the risk from wind / tornado events at Yankee."
- 2. "The modifications proposed by the licensee result in an acceptable level of risk from high wind / tornadoes for the Yankee plant."
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1.2 Issue Review NI issues are related-to range, redundancy, and environmental qualification. These issues are:
o Is full range NI necessary to meet the objectives of Regulatory Guide 1.97 for post-accident conditions at the Yankee Nuclear Power Station (YNPS)?
o Is redundant NI necessary to meet the objectives of Regulatory Guide 1.97 for post-accident conditions at the YNPS?
o Is environmentally qualified NI necessary to meet the objectives of Regulatory Guide 1.97 for post-accident conditions at the YNPS?
1.3 Summary The potential for a core damage event induced by inadequate reactivity control was assessed to be extremely low, less than 1E-07 per year. If a
-cost-benefit value of $1,000 per person-rem were used, the maximum cost that could be justified to reduce this frequency to 0.0 is less than $1,000.
Yankee believes that current design and operational characteristics of the YNPS meet the intent of Regulatory Guide 1.97. Because the cost of plant changes required to explicitly meet Regulatory Guide-1.97 exceeds $750,000, thsse changes are not required based on the above assumption. These resources i
would be better used if retained to be applied to other activities aimed at maintaining or improving overall plant safety and reliability. There are, j however, several procedural changes which will be made to ensure the concerns l J
associated with reactivity addition are minimized.
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' 2. 0 APPROACH'ANL RESULTS i
2.1 General Table 2.1 summarizes the general technical approach. Each step and associated results are described in the following sections. Where possible.
' previous evaluations in the Yankee Nuclear Power Station Probabilistic Safety Study (YNPS PSS) and those described in Reference 3 were used.
2.2' Establish Current Risk Level Attributable to Inadequate Reactivity Control The issues related to NI requirements are based on ensuring that the risk attributable to inadequate reactivity control is sufficiently cmall.
Therefore, this eve tion concentrated on sequences that could result in inedequate reactivity control by first establishing the current. risk level attributable to such sequences (Step 1 in Table 2.1).
Table 2.2 delineates the nine basic steps used to establish the current risk level.- Each step and the results from each step are provided below.
(Figure 2-1 is a flow chart of the process.)
2.2.1 Establish Candidate Sequences (Step 1)
. Successful core cooling requires adequate reactivity control during all modes of plant operation, from normal, at power conditions to post-accident conditions (e.g., main steam line rupture conditions). Hence, a comprehensive set of sequences, including those that were not initiated by reactivity contro1' problems, were investigated.
A comprehensive set of representative sequences was selected by:
- 1. keviewing event sequences in the Yankee Nuclear Power Station Probabilistic Safety Study (YNPS PSS).
- 2. Reviewing analyses of design basis events.
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- 3. Reviewing the possib2e failure causes of instrumentation or systems that might cause a reactivity control problem or be available to detect and mitigate the problem to determine if they could also cause a plant transient (i.e., dependent-type initiating events).
The set of sequences considered were reviewed for the effects of NI range, redundancy, and equipment qualification. Plant Operating Modes 1-3 were considered.
Table 2.3 lists the candidate sequences (initiating events)' considered and their estimated yearly frequencies (mean values). These sequences provide comprehensive coverage of all possible sequences.
Main Steam Line Rupture This event results in a cooldown of the Main Coolant System (MCS)
(directly challenging reactivity control) and an adverse containment environment if the rupture occurs inside containment.
Loss-of-Coolant Accident This event causes adverse containment conditions, and depending on the rupture size, location, and operations personnel response, the possible need to recirculate MCS and Emergency Core Cooling System (ECCS) fluid from the containment sump.
Normal Plant Trip This relatively frequent event (several times per year on average) was investigated to assess the significance on reactivity control of random and human failures in the absence of abnormal containment conditions.
Loss of Vital Bus No. 1 A loss of Vital Bus No. I will result in a plant trip and loss of power to the normal Nuclear Instrumentation System. In approximately 15 minutes the neutron count rate will be well within the source range, and alternate 5670R/23.11
channels of source-range count rate will be available if containment conditions are not severe and other failures have not occurred.
Loss of 480 Volt Bus No. 1
' Failure of this bus results in a plant trip and loss of power to normal and backup NI. ' Operations personnel could transfer the back-up NI to 480 V Bus No. 2 by repositioning a switch in the Switchgear Room.
, Loss of Off-Site Power This event causes a plant trip and decreases on-site power reliability.
Steam Generator Tube Rupture This event provides a path for unborated water from the ruptured steam generator to the MCS.
2.2.2 Develop Logical Categorization (Step 2) t Prior to identifying hardware or personnel-related problems that might adversely impact reactivity control, a review of design and operational factors'that influence reactivity control was performed to develop a logical categorization of reactivity control parameters, such as temperature, rod position, boron concentration, etc. This logical, "high level" development assures that important reactivity addition events are identified.
The following reactivity addition categories were identified by a consensus of-Transient Analysis, Systems Engineering, and PRA/ Systems Analysts with independent peer review by plant operations personnel.
For each group, the manner in which reactivity could be increased is dsscribed. Methods which cannot substantially reduce shutdown margin were oliminated from further consideration at this level. Table 2.4 summarizes the resetivity addition categories examined and the disposition of each. Each category is explained in more detail below.
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l 2.2.2.1 MCS Void The Yankee core has a negative void coefficient of reactivity and op; rates without voids during normal operation. Voiding in the form of steam bubbles reduces reactivity. Void decrease increases reactivity. Since the core remains subcooled during normal power operation, the collapse of voids formed during any event will only offset the reactivity decrease resulting from the formation of the voids. Therefore, changes in the amount of core voiding cannot decrease the original pre-event shutdown margin and, in fact, will aid in the mitigation of other reactivity increase events. Therefore, this group was eliminated from further consideration in this study.
2.2.2.2 MCS Pressure The Yankee core has a small positive pressure coefficient of rsactivity. Pressure increases would increase reactivity. An MCS pressure increase to greater than 16,000 psi above the pretrip pressure would be required before a complete loss of shutdown margin would occur. The two events which will result in a harsh containment environment, a LOCA or Main Steazz Lice Ereak (MSLB) inside containment, cause pressure to decrease.
Therefore, the effect of pressure on the reactivity state for these events will be beneficial, i.e., an increase in the shutdown margin. Since the two pressurizer safety valves are set at 2485 psig and 2560 psig, along with the four loop relief valves at 2735 psig (not crediting the pressurizer power-operated relief valve which opens at 2400 psig), the pressure coefficient is not a credible positive reactivity addition event. The maximum possible reactivity addition through pressure increase is inconsequential.
Therefore, this group can be eliminated from further consideration in this j study.
2.2.2.3 Samarium-149 Concentration
{
Samarium-149 is a thermal neutron absorber which is produced by the dscay of fission products. Reduction in the Sm-149 concentration will increase reactivity. However, Sm-149 is a stable isotope not subject to j radioactive decay. Its concentration can be reduced only by neutron absorption. Following a reactor trip, the neutron flux is reduced to l
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1 n;gligible levels, eliminating the only means of reducing the Sm-149 concentration. Because fission products decay following a trip, the Sm-149 concentration actually increases. Since this decreases reactivity and increases the shutdown margin, this method of reactivity increase can be
]
eliminated from further consideration.
2.2.2.4 Core Geometry e A change in core geometry, specifically the fuel-to-water ratio, may effect the reactivity state of the core. The fuel-to-water ratio is affected by changes in the spacing between fuel rods (rod pitch). The rod pitch in the Ycnkee core is slightly less than the optimum value for neutron economy in order to provide a negative moderator temperature coefficient.
The Final Safety Analysis Report (FSAR) analysis documents that a coolable core geometry is maintained following a LOCA. Thus, the changes in rod pitch would not be large enough to significantly affect the core's reactivity state. Even if a significant fraction of the core were to be dsgraded or melt, the tendency would be toward a reduction in the rod pitch.
This would reduce the reactivity state of the core. Therefore, this potential mathod of reactivity addition can be eliminated from further consideration.
2.2.2.5 Boron Concentration (Boron Dilution)
A reduction in the Main Coolant System (MCS) boron concentration increases reactivity. Since a reduction in the boron concentration can occur through the injection of unborated water, this method of increasing reactivity is addressed directly.
2.2.2.6 Control Rod Position (Control Rod Withdrawal / Inadequate Insertion)
Withdrawal of control rods increases reactivity. This can be secomplished by normal rod withdrawal, via a rod ejection from the core, or through stuck rods failing to insert. These events are addressed directly.
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- 2.2.2.7 MCS Temperature (Cooldown Events)
Due to the negative moderator temperature coefficient of reactivity and th2 negative fuel temperature coefficient of reactivity, at End-of-Cycle (EOC) condition a significant amount of positive reactivity can be added by reducing ths MCS temperature. Therefore, this method of reactivity addition is eddressed directly.
2.2.2.8 Xenon Concentration (Xenon Decay)
Xenon-135 is a thermal neutron absorber. Reduction of the Xe-135 concentration increases reactivity. Reduction occurs uiturally following a reactor trip due to radioactive decay. (Note that Xe-135 is not a stable isotope.) Since the equilibrium worth of Xe-135 during full power operation exceeds three percent delta rho, complete decay of Xe-135 can significantly increase reactivity. Therefore, this method of reactivity addition is addressed directly.
2.2.3 Identify Specific Events (Step 3)
For each of the four reactivity addition categories retained in Step 2, plant design and operational characteristics were reviewed to determine failures'that could cause an increase in reactivity. Specific event scenarios were then developed for each category. The basis for the development of the individual event scenarios is given below:
- 1. MCS Boron Concentration - Plant drawings were reviewed to determine all connections to the Main Coolant System (MCS) through which unborated water may be added. In addition, for post-LOCA l conditions, connections to the Vapor Container (VC) which could supply unborated water to the VC sump were evaluated. With the Safety Injection (SI) System in the recirculation mode following a LOCA, any dilution of the VC sump may eventually result in a dilution of the core region boron concentration.
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- 2. ~ Control Rod Position - All possible means of withdrawing control rods were considered, including manual withdrawal by the operator, and rod drive system malfunctions resulting in'a continuous witSdrawal of the selected rod group. A. control rod ejection and a failure of one or more of the control rods to insert on a reactor trip signal were also considered.
- 3. MCS Cooldown - Plant drawings were reviewed to locate all connections to the main steam lines'through which steam may be d rawn .' Systems capable of injecting cold water into the MCS were also considered.
- 4. Xenon-135 Concentration - A variety of plant power maneuvers prior to the initiating event'were considered to ensure that the most limiting post-trip Xenon transient was included.
Table 2.5 summarizes the event scenarios considered, giving a dancription of each scenario. The event scenarios are grouped by reactivity categary.
2.2.4 Screen Specific Events (Step 4) e The reactivity categories and the specific event scenarios were screened separately. As described earlier, the reactivity categories were first screened following their identification on the basis of whether the maximum possible change in reactivity in each category could result in a significant reduction in shutdown margin. Categories incapable of a significant reduction in shutdown margin were eliminated from further consideration at this level of screening.
Specific event scenarios were then developed for the remaining categories, as described in previous sections. These event scenarios were thsn screened, based on the conditions necessary for the event to occur (e.g., equipment failures, operator errors, etc.).
9_
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Table 2.6 provides a summary of the screening criteria used at both stages in the screening process. Table 2.7A lists the conditions required for sech boron dilution event to occur, as well as the basis for disposition for osch event considered. Tables 2.7B, 2.70, and 2.7D provide the same information for the control rod, MCS cooldown, and Xenon-135 decay events, respectively.
In Tables 2.7A through 2.7D, the initiating event listed as " normal" considers: (1) normal plant trip,'(2) loss of Vital Bus No. 1, (3) loss of 480 volt Bus 1, (4) loss of NI, and (5) loss of off-site power candidate sequences listed in Table 2.3. These candidate sequences include the various wrys in which power can be lost to the NI. The "LOCA" initiating event listed in the tables includes both the loss-of-coolant accident, and steam generator tube rupture candidate sequences in Table 2.3. Lastly, the "MSLB" initiating event in the tables corresponds to the main steam line break candidate sequence in Table 2.3.
Figure 2-2 depicts the event tree used through the screening and analysis process to evaluate the probability of core melt for those selected scenarios.
2.2.5 Steps 6 Through 9 Following the final screening of the reactivity addition event scenarios, none of the scenarios were found to pass the screening criteria.
No " credible" event scenarios were identified. (The Total Frequency of Core Malt was estimated to be less than 1.0E-7 per year.) The most likely event, while still highly improbable, was found to be a charging dilution of the MCS j initiated from normal plant conditions with a coincident loss of NI.
Hence, formal implementation of Steps 6 through 9 of Table 2.2 was not required. However, to demonstrate the process, the charging dilution event is described qualitatively below.
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i Charging Dilution of MCS ,
Table 2.8 provides the sequence of events that would be expected as the charging dilution progresses. The dilution is assumed to occur sometime following the reactor trip caused by the loss of NI. It begins when the valves necessary to align the charging pump suction to a source of unborated water are inadvertently opened.
As shown in Table 2.8, numerous alarms and main control board indications are available to alert the operator that the charging system is in en off-normal lineup, and that a dilution should be expected. These indications and alarms are spread out in time and, therefore, are readily discernable. The only operator action credited in this scenario is that a boiler feed pump be started (either normal or emergency) upon receipt of a steam generator low level alarm. This would be a normal response to that clarm. Core melt is conservatively assumed when the core power rises above the capacity of an auxiliary feed pump to provide water to the steam generators.
Over 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> of continuous dilution is required to cause a return to criticality. Thus, at least two different operating shifts would have to fail to notice the indications available. (Since this is an event that could staff the Technical Support Center, it could also be credited but is not for conservatism.) In addition, the LPST level, LPST pressure, and charging temperature are recorded hourly on the Control Room primary log sheet. The PWST level is recorded once per shift.
The charging dilution of the MCS can be terminated quickly and eas.ly 4 from the Main Control Board. The necessary operator actions to accomplish this, both before and after criticality, are given in Table 2.9.
2.3 Potential Improvements While no " credible" event scenarios were discovered, several procedure improvements were identified in the process of evaluating the different ;
scenarios which could enhance maintenance of the reactivity control critical safety function in the absence of NI indication. These candidate improvements ;
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l ara not justified on a cost-benefit basis. .The changes, however, are considered prudent and will'be made within three months of the NRC's concurrence with the conclusions.of this report.
- 1. OP-3000. " Emergency Shutdown From Power" - Insert an additional step in the procedure to check for a loss of NI. If two or more NI j
(
indicators are lost, the operator would be directed to take the )
following actions, assuming the reactor trips:
- a. Borate the MCS to the Mode 5 (cold shutdown) shutdown margin requirements.
- b. Energize the refueling source range channel to provide alternate source range indication.
- 2. OP-3106, " Loss of Main Coolant" - Insert a caution following the step directing the operator to restore steam generator levels. In performing this task, the operator should monitor for excessive feed water flow. Excessive flow would be indicative of a leaking steam generator, which could dilute MCS boron concentration if the event (LOCA) caured steem generator tube leakage.
- 3. OP-3201, " Steam Line Break" - Relocate the step directing the operators to perform an emergency boration upon a steam line break. It would be prudent to move this step to a more prominent location to stress the desirability of emergency boration for large steam line breaks.
- 4. OP-2153, " Boric Acid Control" - Add a step to the prerequisites !
beneath the Shift Supervisor permission, directing the Nuclear
. Auxiliary Operator to remain in the vicinity of the manual demineralized water supply valve, while a dilution or make up is in progress, to close the valve in the event of any off-normal condition. This places additional redundancy on the preclusion of an inadvertent boron dilution event.
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TABLE 2.1 General Technical Approach i
- 1. Establish current risk attributable to inadequate reactivity control.
- 2. On the basis of the current risk level, identify candidate changes to reduce the probability and/or consequences of the key sequences.
- 3. Perform an analysis of candidate changes, including those required to I
explicitly meet Category I requirements.
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TABLE 2.2 Establishing Logical Search, Categorization, and Investigation of Risk Attributable to Inadequate Reactivity Control
- 1. Establish candidate sequences for evaluation using the YNPS PSS as a basis.
- 2. Develop a logical categorization of possible ways of changing core reactivity.
- 3. Identify specific ways of changing core reactivity for each category developed in Step 2 by reviewing plant design and operation.
4 Qualitatively screen possible reactivity addition events identified in Step 3 on the basis of probability and consequences to identify candidates for quantitative evaluation.
- 5. Develop a core melt event tree for examining each candidate reactivity addition event from Step 4.
- 6. Develop fault trees for each event tree top event established in Step 5 considering:
- 7. Quantify each event tree sequence considering dependencies among events.
- 8. Establish the consequences of each core-melt sequence. Consequences are assessed on a person-rem basis so that cost-benefit evaluations can be performed.
- 9. Establish current risk level by combining results of Steps 7 and 8.
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l TABLE 2.3 Candidate Sequences l
1 Sequence Yearly Frequency
-0
- 1. Main Steam Line Break 3 x 10
- 2. Loss of Coolant Accidents 2 x 10-
- 3. " Normal" Plant Trip 3
-2
- 4. Loss of Vital Bus No. 1 10
-2
- 5. Loss of 480 Volt Bus No. 1 10 1
-2
- 6. Loss of NI 10
- 7. Loss of Off-site Power 0.05
- 8. Steam Generator Tube Rupture .02 5670R/23.11
FIGURE 2-1 Flewchert of Cost Banafit Analysis Methodology CSTART) l u (
STEP 1 Identify all, Initiating Events- .)
(based on events.which could- l challenge.the reactivity control critical safety function) sr STEP 2 Identify all possible methods of increasing reactivity (based on Technical Specification shutdown margin requirements and expert knowledge)-
sr Determine maximum impact of each reactivity addition method' C on pre-event shutdown margin Can Eliminate from further the reactivity consideration those
- reactivity addition addition method '
methods incapable of a substantially reduce NO hutdown margi significant reduction in shutdown margin
?
YES Select single initiating Event (, STEP 3 for investigation v . _
Select single reactivity addition method for~ investigation from ,
those methods not previously eliminated from consideration Identify specific scenarios for increasing core reactivity via current reactivity addition jg method in conjunction with current initiating event (based on review of plant design and operation)
Bave all reactivity addition methods >
been considere NO
?
YES l5670R/23.11 ( ()
r
I FIGURE 2-1 (continued) have all initiatin events been '
considered NO
?
YES w
Evaluate the potential for. STEP 4 criticality for each specific ::
event scenario Eliminate from further Is consideration those event criticality s scenarios for which criticality cannot be possible NO
?
achieved y YES Qualitatively evaluate remaining event scenarios for probability 4 of core melt Is Eliminate from further probability consideration those event o f core melt i > scenarios for which 10-8 per year YES probability of core melt 7
is 1 10-8 per year NO
%/ _
From remaining event scenarios determine the minimum set of bounding scenarios s/
Tor the bounding scenarios: STEPS 5-7
- 1. develop core melt event trees
- 2. develop fault trees for each top event
- 3. quantify each event tree 3r Establish consequences of each core melt STEP 8 L
sequence, based on person-REMS v
Establish current risk level by STEP 9 combining results of last two steps @
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TABLE 2.4 0
1 1
Summary of Reactivity Addition Methods Considered n; -------------------------------------------------
1 Applicable' 1 Report:
Section Reactivity Addition Method Disposition 2.2.2.1 MCS Voids Eliminated from further consideration. Since the core is subcooled during normal operation, void reactivity is incapable of reducing.the pre-event shutdown margin.
2.2.2.2 MCS Pressi;re Eliminated from further i consideration. The maximum reduction of the pre-event shutdown margin by pressure reactivity is inconsequential.
2.2.2.3 Samarium-149 Concentration Eliminated from further consideration. Since Sm-149 is a stable isotope, it cannot reduce the pre-event shutdown margin by post-trip radioactive decay.
2.2.2'4
. Core Geometry Eliminated from further consideration. Since FSAR analyses show that a coolable geometry is maintained, the maximum reduction in-the pre-event shutdown margin is inconsequential.
2.2.2.5 MCS Boron Concentration Examined possible event scenarios (see Table 2.5). Addition of boron-reactivity via MCS dilution could significantly reduce the pre-event shutdown margin.
2.2.2.6 Control Rod Position Examined possible event scenarios I (see Table 2.5). Addition of rod reactivity could significantly reduce the pre-event shutdown margin.
2.2.2.7 MCS Cooldown Examined possible event scenarios (see Table 2.5). Addition of-Moderator / Doppler reactivity could l significantly reduce the pre-event shutdown margin.
2.2.2.8 Xenon-135 Concentration Examined possible event scenarios (see Table 2.5). Post-trip decay of Xe-135 could significantly reduce the pre-event shutdown margin. ;
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Table 2.6 Summary of Screening Criteria A two step screening process was utilized. First, all possible methods for adding positive reactivity to the core (e.g. MCS boron concentration, control rod position, etc.)
were identified. These reactivity addition categories were then screened:
Screening Criterion for Reactivity Addition Categories o Will the maximum possible reactivity added cause a significant reduction of the pre-event shutdown margin?
Specific reactivity addition event scenarios were then developed for the remaining reactivity addition categories, t The remaining categories were MCS Boron Concentration, Control Rod Position, MCS Cooldowns, and Xenon-135 Concentration. These specific event scenarios were then screened:
Screening Criteria for Specific Event Scenarios o Is the probability of core melt less than lE-08 per year?
o Is it possible for the event scenario to result in criticality?
o Is the probability of core melt for the event scenario clearly bounded by another event scenario?
o Is the event scenario included in or bounded by the YNPS i Probabilistic Safety Study?
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Table 2.8 Sequence of Events for Charging Dilution of MCS Tins:(HH:MM). Event 00:00 o Valves'DW-MOV-655 and DW-V+662, 634, or 635 are inadvertently opened. They are assumed to remain for the duration of the event. The unplanned dilution begins.
Plant Initial Conditions: (normal post-trip) o All control rods inserted o MCS Boron = 1415 ppm (BOL) o T-avg = 515 degrees (controlled by steam dump) i o Charging = Letdown = 30 GPM o' Charging Temperature = 450 degrees o LPST Level - 44 inches (50% full) l o.LPST Pressure = 10 psig o PWST Level = 30 feet o All Steam Generator Levels = 23 feet 00:05 o Charging temperature has decreased to approximately 400 degrees due to the cold PWST water supplied to j L
the charging pump suction. The charging temperature will remain at this value for the duration of the event.
00:15 o LPST high level alarm occurs at 50 inches. The j operator will be unable to reset the alarm since level is still increasing, o LPST Pressure = 17 psig 00:27 o LPST high pressure alarm occurs at 25 psig. The operator will be unable to reset the alarm since pressure is still increasing due to the level l l
increase.
00:30 o LPST Level = 56 inches o LPST Pressure = 28 psig 00:45 o LPST Level = 62 inches ,
o LPST Pressure = 48 psig l 00:56 o LPST safety valve opens at 75 psig o LPST safety valve discharge high pressure alarm occurs.. I I
o LPST pressure will remain relatively constant at about 75 psig for the remainder of the event. The safety valve will open repeatedly over the next 37 j minutes until the high level dump valve opens. This will also result in repeated high safety valve discharge pressure alarms.
I
_ _ - _ - - - l
=. - - - - - - -
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Table 2.8 (cont
________________'d) 1 Sequence.of Events for Charging Dilution of MCS Tipo (HH:MM) Event j 01:00 o LPST Level = 68 inches 01:15 o.LPST' Level = 75 inches 01:33 o The LPST is full.
o The LPST high level dump valve opens to dump the excess water to Waste Disposal.
03:00 o PWST Level = 28.9 feet 06:00 o PWST Level = 27.8 feet
'09:00 o PWST Level = 26.7 feet 12:00 o PWST Level - 25.6 feet 15:00 o PWST Level - 24.5 feet 18:00 o PWST Level = 23.4 feet 20:23 o The reactor returns to critical.
O PWST Level = 22.6 feet o' Steam. Generator Levels = 23 feet 20:30 o Steam Generator Levels = 20.7 feet 20:35 o Steam Generator low level alarm occurs at 19 ft.
o The operators are assumed to respond to the alarm by initiating auxiliary feed.
o Steam Generator levels will initially increase after auxiliary feed is started. The rate of increase will ,
drop to zero as the assumed point of core melt is reached.
20:45 o Core melt is conservatively assumed to occur, since core power exceeds the capacity of the auxiliary ;
feed pump to maintain Steam Generator level, o Steam Generator levels will decrease continuously until the secondary heat sink is lost, unless operator action is taken to shut down the reactor or re-establish normal feed with the boiler feed and i condensate pumps.
i 2
TABLE 2.9 Termination of the Charging Dilution Event Thn' dilution event can be terminated quickly from the main control board, or
'as directed from the Control Room,'either prior to criticality or between criticality and core melt. The following operator actions are necessary to accomplish this:
Prior to Criticality 1.. . Trip all operatir.g charging pumps f rom the main control board, or
- 2. Close DW-MOV-655 on main control board, or 3.- Have the auxiliary operator close either DW-V-662, 634, or 635, whichever flow path was in use.
t l Following Criticality. But Prior to Core Melt I
{
- 1. Terminate the dilution by either Method 2 or 3 above.
- 2. Begin an emergency boration from the main control board by:
o Open CS-MOV-529, charging suction to boric acid mix tank, and o Start all charging pumps at maximum flow, and o Close CH-MOV-521, CS-MOV-540, PU-FICV-202, and DW-MOV-655. This
. isolates the charging pump suction from all sources except the boric acid mix tank.
5670R/23.11
FIGYTRE 2-2 EVENT TREE REACTIVITY PRIOR TO CRITICAL PRIOR TO CORE MELT END STATE IE ADDITION EVENT DETECTED l MITIGATED DETECTEDlMITIGATED NA NA OK K l CM NA CM 1
OK l
NOTE: CM = CORE MELT ZE = ORIGINAL INITIATING EVENT OFF-NORMAL / ACCIDENT CONDITION.
NA'= STEP NOT APPLICABLE FOR THIS SEQUENCE 5670R/23.11
4 .. !
l 1
i 1
i FIGURE 2-3 BORON DILUTION FLOW PATH (NO)
X-->TO EMERGENCY BOILER FROM FEED PUMP NO.1 PRIMARY j WATER > '
STORAGE. p (NC)9 g
X->TO EMERGENCY BOIL TANK (NC)
U MAKEUP I b (NO) FEED PUMP NO.2 (TK-39) 01 PUMP 2O DW-V-662 ]t(NC)l l X (NO) (NO) X FROM y- SI TANK DISCHARGE HEADER
'TO (NC) '{
BORIC (NC)DW-V-648 DEMINERALIZED WATER HEADER
- ACID < 7 ', 7I l X (NO)
O DW-V-640 MIX-TO
-TARK CHARGING PUMP <-X-)h[<-[<-X Y j SAFETY INJECTION SUCTION. HEADER (NC) 'TO CHARGING SUCTION
- l. DW-MOV-655 NOTE:(NO) = NORMALLY OPEN j X = NORMALLY OPEN VALVE (NC) = NORMALLY CLOSED I
l 1
I i
l 5670R/23.11 k---_ mum.
3.0 NUCLEAR INSTRUMENTATION SYSTEM DESCRIPTION j
A Westinghouse type FN design using magnetic amplifiers and vacuum tube electronics for signal conditioning is provided to serve as the primary means I of monitoring the level of and rate of change of reactivity in the reactor core.- This design does not readily lend itself to a redundant power supply I
design, since the outputs of the magnetic amplifiers (bistables) are used to input both of the scram amplifiers and, therefore, the power supplies must be i from the same source.
Three of the four source range channels are located in the Main Control Room and'two of these (the normal source range channels) are displayed on the main control board. The third channel is normally de-energized and is described below. Both normal channels of instrumentation are provided power from a single safety class, vital ac bus which, in turn, is powered from an Uninterruptible Power Supply (UPS) fed from two independent and diverse sources. Safety Class Emergency Motor Control Center No. 5 provides the standby source and the No. 1 Station Battery de bus provides the normal source of power to the Uninterruptible Power Supply (UPS). Selection of the power source is controlled by a static switch in the UPS which automatically transfers the vital bus to the standby source (ultimately, safety class i emergency 480 V ac Bus No. 1) upon sensing a degrading condition on the normal supply. A manual switch on the UPS is provided if a transfer is so desired.
For the other two channels of source range instrumentation, the refueling channel and the Safe Shutdown System (SSS), independent power supplies are used. The refueling channel is powered from transformer "A" bus off the Emergency Motor Control Center No. 1 (EMCC No. 1). For this EMCC, the normal source of power is ultimately from the same source as the standby source for the normal source range channels (Safety Class Emergency 480 V ac Bus No. 1). It can be readily transferred to the alternate standby source by means of a manual throwover switch located in the Switchgear Room directly beneath and immediately accessible from the Main Control Room, to safety class emergency 480 V ac, Bus No. 2. For the SSS source range channel, the instrument is normally supplied from a non-safety class 480 V ac MCC which is 5670R/23.11
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! bicked by a dedicated SSS diesel generator. This diesel generator is totally independent of, and redundant to, the non-SC source upon local manual startup.
Thus, the operators and supervisors have available three redundant channels of source range NI powered from safety class power supplies (as discussed above) and one additional remote source. range channel powered from a dsdicated, redundant, and diverse power source and instrument. Note that the refueling source range NI power supply can be transferred to the No. 2 1
'l 480 V ac emergency bus via a manual throwover switch (see Figure 3-1).
A description of the NI and Reactor Protection System is provided in i
Attachment A.
The power supply configuration for the critical components is contained in Figures 3-1 and 3-2.
In addition to this NI, other normal instruments and procedures provide the operating crew with the status of the reactivity control critical safety function continually.. Among these, the most important are:
'o Rod position indication i'
o Indicated on the Main Control Board o Powered from the transformer "A" Bus, as is the refueling nuclear instrument o Core exit thermocouple (s) and saturation monitor o Indicated on the Main Control Board o Indicated and trended on the Safety Parameter Display System o Pressurizer level o Indicated on the Main Control Board o Indicated and trended on the Safety Parameter Display System
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.o ~ Direct boron sampling periodically.or as directed by the Operating l Crew or Technical Support Staff l
o A standard plant procedure is performed routinely on every shutdown-or in the event of a LOCA, rod ejection, stuck rod (s), or excessive ;
cooldown events. This routine procedure and the other emergency procedures direct tha operator on how much 12-weight percent boric acid to inject for greater than or equal to 5% delta-rho shutdown margin during 1) normal shutdown, 2) off-normal events (listed
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above), and 3) any. time the operator and his supervisor deem that an emergency boration is necessary.
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TO 480VAC BS4-1
'480VAC BS6-3 1 480VAC EBUS 2 480VAC EBUS 1 '
MANUAL THROWOVER SWITCH EMER. SUPPLY
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NORMAL SUPPLY O
_( SWITCHGEAR ROOM )
EMER. MCC NO. 3 EMER MCC NO. 1
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TRANSFORMER A BUS EMER. MCC NO. 5
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I +- ROD POSITION EN SW. A DC BUS NO. 1 INDICATION LIGHTS O --- I ( MAIN CONTROL BOARD )
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NUCLEAR INSTRUMENTATION o TN 1- j AO ( MAIN CONTROL ROOM ) l V ;
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1 NUCLEAR INSTRUMENTATION B
REFUELING CHANNEL
( MAIN CONTROL ROOM ) i U- 1 S 1 i
FIGURE 3-1 YNPS NUCLEAR INSTRUMENTATION AND ROD POSITION POWER SUPPLY 5670R/23.11 f
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I SSS DIESEL I .1 SSS f GENERATOR g
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NORMAL FEEDER EMERGENCY FEEDER f
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NON-SC MCC - -
7 (PAB)- MOTOR CONTROL CENTER I
SSS SOURCE RANGE CHANNEL i' _\
FIGURE 3-2 4
SAFE SHUTDOWN SYSTEM (SSS) NUCLEAR INSTRUMENT AND POWER SUPPLY i
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4.0 CONCLUSION
S AND REC 0!91ENDATIONS T '
4.1 Conclusions g For the issues being reviewed, the following was concluded:
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- 1. Is full" range'NI necessary to meet the objectives of Regulatory Anide 1.97 for' post-accident conditions?
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FullranghNIwouldnotyieldanyfurtherinformationmore If'- useful than that provided by source range instrumentation. Full range instrumentation would only increase the stress level on the operating crew if it failed high during conditions when
[f reactivitywasbeingadequateltycontrolled. Operations personnel are in a better situation with source range indicators j .in this case. An abundance of secondary indicators to power c level are available end the indicators are direct indications of the primary critical safety functions which must be controlled to prevent core damage.
3 Therefore, full range N% is not necessary to ensure that the Reactor Systcm is proparly controlled following an accident and it may be detrimental in that it could place unnecessary stress p( on the operators attempting recovery actions. Full range NI is kp not a key variable in mitigating a return to criticality event.
- 2. Is redundant NI necessary to meet the.Tbjectives of Regulatory Guide 1.97 for post-accident conditions a.t the Yankee Nuclear Power Stationf and
- 3. Is environe.entally qualified NI needed?
f From the results of this studyg no " credible" events were identifiedthatcouldresultiNaninadvertentincreasein reactivity leading;to criticality; Source range NI is not necessary to detect the occurr'nce e of such an event. Due to the simplicity of the YNPS, Main Control Board, its de ign for numan control and d
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V the diversity and redundancy of indications, including their motive ;
force (somearepneumaticaswellaseleebric),theactcual.needfob source range NI is' greatly diminished.
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In view of the results and conclusions of this investigation and the i
cost involved in changihI;L, we conclude that the resource in ter.tus of hardware and personnel costs gould (and should) be more appropriately applied to other vetivities aimed at maintaining and improving overall plant safety. f 4.2 Recommendations 9
Section 2.3 identifies possible changes that would reduce the risk associated with postulated, reactivity control problems. Although these. ?
changes would not " pass a cott--benefit test," they should be implemented if-agreement ev.n be reached with the NRC on the need for NIs.
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5.0 REFERENCES
- 1. Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following
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an Accident."
- 2. Letter, USNRC to YAEC, " Yankee Nuclear Power Station - Regulatory Guide 1.97" (Revision 2), NYR 86-273, dated December 9, 1986.
- 3. Letter, USNRC to YAEC, " Yankee Nuclear Power Station - Integrated Plant Safety Assessment Report (IPSAR), NUREG-0825 Section 4.5, Wind and Tornado Loading, and 4.8, Tornado Missiles," NYR 87-86, dated May 13, 1987.
- 4. Swain, et. al . , NUREG/CR-4772, SAND 86-1996, February 1987, " Accident Sequence Evaluation Program Human Reliability Analysis Procedure."
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ATTACHMENT A Nuclear Instrumentation and Reactor Protection System General The NI and Reactor Protection System monitors the nuclear reactor flux from source levels to above maximum power levels and provides the necessary indications, alarms, and controls for safe and efficient operation of the reactor. This equipment incorporates provisions for initiating a reactor and turbine shutdown in the event of conditions which may be hazardous for plant operation.
Reactor shutdown signals which originate in the turbine-generator protection equipment and main coolant flow (main coolant pump current) trip system are connected through relay contacts directly to the control rod scram air circuit breakers. The relay contacts are open below 15 MWe.
The NI and Reactor Protection System equipment contains ten nuclear information channels, two scram amplifiers, and various auxiliary equipment, all mounted in *.icee cabinets in the Control Room. The nuclear detectors themselv'es a;c installed in the neutron shield tank around the reactor vessel. The detectors are connected directly to the equipment in the Control Room by trinaial and coaxial cables. Each nuclear channel is indicated locally at the three cabinets as well as on the nuclear section of the Main Control Board (MCB). A two-pen recorder on the Main Control Board may be used to record any two of the source, intermediate, and power range channels. Also included on the nuclear section of the MCB are various selector and reset switches, indicating lights, and two reactor shutdown push buttons.
The ten nuclear information channels consist of two source range, two intermediate range, three intermediate power range, and three power range channels. The source and intermediate channels are designated as the startup range.
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Source Range Nuclear Instrumentation o
The source range (Channel Nos. 1 and 2) detectors are high sensitivity BF3 proportional counters. There are four detectors, located around the ,
reactor vessel, any two of which may be connected to Channel Nos. 1 and 2.
The third detector is connected to Safe Shutdown System instrumentation during normal operations. During refueling periods, the third and fourth detectors are used in the Control Room to indicate source range count level and alarms on an increaming level. The third and fourth detector channels are normally de-energized and must be turned on by the operator when needed.
The detectors have a sensitivity of approximately 40 counts / neutron cm -second. Counter output of 1 to 100,000 counts per second correspond to 3
a flux range of 2.5 x 10~ to 2.5 x 10 neutron /cm -second, and the l
counter high voltage is automatically cut off by the intermediate range channels above this flux level to prevent counter burn-out.
The signal output of the BF3 c unter consists of pulses which are proportional in number to the neutron and gamma flux present at the detector location. These pulses are fed over a triaxial cable to the panel unit in the Control Room. The first panel unit (pulse integrator) separates the pulses from the high voltage, amplifies the pulse, provides an sdjustable discriminator which rejects the gamma pulses and pulses resulting from noise, and converts the neutron pulses to a direct current which is proportional to the reactor neutron flux.
The direct current signal from the pulse integrator is fed into a second panel unit (log microammeter). This circuit converts the linear input l signal to an output which is proportional to the logarithm of the neutron flux level. An output is also provided which is proportional to the rate of change of the logarithm of the neutron flux level. The source level meter is calibrated from 1 to 100,000 counts per second and the startup rate meter is calibrated from -1 to +10 decades per minute.
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The log level (counts per second) and the rate of change signal (decades per minute) are indicated at the nuclear section of the MCB as well as at the NI cabinets. The log level signal may be switched to the nuclear recorder. The startup rate meter at the MCB is calibrated -0.2 to +2.0 decades per minute.
Startup and Power Range Auxiliary Panel The startup and power range auxiliary panel receives signals from the NI to provide the necessary signals for the appropriate annunciator circuits and the Rods Stop Signal circuit. This panel also contains the power range coincidence-single scram switch, the turbine load cutback relaying (not used),
and the source range BF 3 high voltage disconnect relays.
The high startup rate annunciator circuit is normally set to trip when j
the reactor startup rate reaches 1.0 decade per minute (adjustable between 0.5 l and 5 decades per minute) and the rods stop circuit is set to trip at 1.5 decades per minute (adjustable between 0.5 and 5 decades per minute). These circuits are of the manual reset type and must be reset by operating the manual reset switch, which is located on the nuclear section of the MCB. The source range and intermediate range signals actuate the 1.0 decade per minute annunciator circuit, but only the intermediate range signals operate the 1.5 decades'per minute rods stop circuit.
The startup rate scram and alarm (Channel Nos. 3 and 4 intermediate range only) is normally set to trip at 15.2 decades per minute and is adjustable from three to ten decades per minute. The source range and intermediate range signals can actuate individual channel bistable magnetic amplifiers to initiate the scram; however, Channels 1 and 2 are normally not in use.
In addition to these signals, there exists from each one of the log microammeter units in the intermediate range channels an automatic signal, which disconnects the high voltage from the source range BF3 Proportional 5670R/23.11
counters when the reactor neutron flux is increasing between 5 x 10' and 105 nV and reconnects the high voltage on decreasing flux at approximately the same value. A source range high voltage light is mounted on the nuclear section of the MCB. The light is off when the high voltage is off. A manual switch disconnecting the BF3 source range high voltage is also available at the NI cabinets.
The coincidence feature makes it necessary for two out of six power range channels to initiate high neutron flux level cignals in order to cause the scram amplifiers to trip. The high neutron flux level trip setpoint is adjustable for various reactor operation conditions. For reactor 100 percent _ - .-
full power operation (i.e., 600 MWt), with four loops in service, the level trip set point is set at 1108 percent. For reactor operation between 0 and 15 MWe the level trip setpoint is manually adjusted to 135 percent of full power. A power range coincidence single switch is provided to allow for coincidence scram or any single channel scram. The low power scram set switch is located on the nuclear section of the MCB.
Signals not fed through the high neutron flux level coincidence circuit but operating on the scram amplifiers through the alarm and scram panel are those initiated from high startup rate. Provision is made in the alarm and scram panel to accommodate additional signals for memory light indication only.
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Additional scram signals which directly actuate the control rod breaker shunt trip coils are provided from the nonreturn valve (NRV) trip, the low Main Coolant System pressure trip or the high Main Coolant System pressure trip. The NRV trip is actuated by either a main steam line low pressure trip or a Containment Isolation Signal (high containment pressure).
A permissive relay circuit is provided which is operated by two millivolt bistables activated from the thermal converters which monitor the generator output. Operation of the circuit occurs at a generator output equivalent to 15 MWe. This circuitry provides for an optional manual bypass .
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for the low steam generator level scram, low flow scram and turbine-generator 1
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scram signals when the power is below 15 MWe. At 15 MWe and above, the scram bypass is automatically removed. The high startup rate scram signal is automatically connected at 15 MWe and below and automatically bypassed above 15 MWe, A second permissive relay circuit is activated at approximately 130 MWe power level, which provides for automatic cut-in of a manual single step rods-out reret circuit. At power levels of approximately 130 MWe and above, the reset circuit requires the control switch to be returned to the neutral or reset position before making each additional rods-out step. Below approximately 130 MWe output, the reset circuit becomes ineffective and thus a controlled but continuous rods-out motion may be effected.
Alarm and Scram Panel The scram signals for high intermediate range startup rate, and high neutron flux levels, are connected to the magnetic amplifier alarm and scram panel that acts as the control center for indicating the individual signals that may have caused the scram and for operating the scram amplifiers directly.
Scram Amplifiers The two scram magnetic amplifiers operate individual scram relays whose contacts are connected to the shunt trip coil circuitry of the rod scram circuit breakers. The scram relays are energized at all times except when a scram signal is sent to the scram amplifiers. When the scram amplifier ,
outputs are zero, the scram relays are de-energized and the contacts in the control rod breaker shunt trip coil circuits close causing both breakers to open. The circuit is such that any one scram relay can trip either circuit breaker.
Meter and Test Panel Two meter and test panels provide local indication for Channel Nos. 1, 2, 3, and 4 level and rate at the NI cabinets, and the 60 cps test calibration signals to accurately test the source range channels.
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Auxiliary Meter Panel The auxiliary meter panel contains three meters which were used during early plant life to measure the uncompensated signal from the three intermediate power range ion chambers. They have been short circuited because intermediate power range channels now measure this signal with greater accuracy.
The switches used to select (from the low voltage power supply or from the power supply on the power range panel) which high voltage is to be used for the intermediate range detectors are located on the auxiliary meter panel. In addition, the relay to select the high or low power scram setpoints of Channels Nos. 6, 7, and 8 is also located on the auxiliary meter panel.
Recorder The recorder used with the NI is a two-pen, twc-speed instrument with two switching circuits. The switching allows for full-scale deflection for a two-decade change in reactor flux. This recorder is mounted on the front of the nuclear section of the MCB.
Power Supply The instrument bus power supply is 120 volts and 60 cycles, normally supplied by the vital bus No. 1.
The instrument bus power supply for the refueling channel source range count meter is supplied from Class IE Bus - Transformer A.
The instrument bus power supply for the Safe Shutdown System (SSS) source range instrument is the SSS Motor Control Center which is powered by a dedicated diesel generator.
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