ML20033E144
| ML20033E144 | |
| Person / Time | |
|---|---|
| Issue date: | 02/26/1990 |
| From: | Beckjord E NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Morris B NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| References | |
| REF-GTECI-063, REF-GTECI-NI, TASK-063, TASK-63, TASK-OR NUDOCS 9003090143 | |
| Download: ML20033E144 (7) | |
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-MEMORANDUM FOR: Bill M. Morris, Director, Division of Regulatory Applications, Office of Nuclear Regulatory Lesearch FROM:_
Eric S. Beckjord, Director, Office of Nuclear Regulatory Research
SUBJECT:
- GENERIC ISSUE 63, "USE OF EQUIPMENT POT CLASSIFIED AS ESSENTIAL TO SAFETY IN BWR TRANSIENT ANALYSIS" The prioritization of Generic Issue 63, "Use of Equipment not Classified 2
as Essential-to Safety in-BWR Transient Analysis," shows that the. issue has little safety significance and will be DROPPED from further consideration.
The enclosed prioritization evaluation will be incorporated into NUREG-0933,
.t "A Prioritization of Generic Safety Issues," and is being sent to the regions, other offices :the ACRS, and the PDR, by copy of this memorandum'and.its enclosures. to allow others the opportunity to comment on the evaluation., All comments should be sent to the Advanced Reactors and Generic Issues Branch, DRA,RES-(MailStopNL/S-169).
Should you have any: questions pertainin the contents of this memorandum, please contact Ronald Emrit (492-3731)g.to s.
/
Eric S. Beckjord, Director Office of Nuclear Regulatory Research
Enclosure:
.Prioritization Evaluation cc: 'T. Murley, NRR E. Jordan AE0D
-W.. Russell, Reg. I
'J. Grace, Reg. II A. Davis, Reg. III R. Martin, Reg. IV s
J. Martin, Reg. V TPORih ACRS
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.9003090143 900226 PDR, GTECI GNIO p
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ENCLOSURE PRIORITIZATION EVALUATION Generic Issue 63:
Use of Equipment not Classified as Essential to Safety in BWR Transient Analysis 7
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ISSUE 63: USE OF EQUIPMENT NOT CLASSIFIED AS ESSENTIAL T0 SAFETY IN BWR TRANSIENT ANALYSIS
. DESCRIPTION-Historical Background BWRs are required to be operated within set thermal limits to maintain the-
. integrity of the fuel cladding during postulated events.
One of the i
. established thermal limits is the minimum critic.:1 power ratio (MCPR).
Th7 critical power-ratio _(CPR) is'the~ ratio of the fuel bundle power at which l boiling transttion begins to the actual' bundle power..By maintaining the CPR above a predetermined safety limit everywhere in the core, boiling transition i
and possible fuel failure.can be precluded. 'Several postulated transients 4
result in a reduction of CPR.
In order to' assure that the CPR safety limit is
.not. violated,'an operating-limit must be set.
The operating limittis based l
upon the_ CPR safety limit and the change. in CPR from the most limiting.or i
severe transient.
By operating above the operating limit, a plant will-not violate the:CPR safety limit for any of the-postulated transients.
Since the i
CPR operating limit is related-to core thermal output, a higher operating limit may. result in restricted output.
Therefore, it is advantageous'to show-the
-least change in CPR for all transients-analyzed..
Safety' Significance
..In1the past, applicants have been required to: assume failure of certain-
.equipmen.t and only take credit for the ope _ ration of other equipment in the analyses of transierts. ass The combination of these assumptions has often i
idictated the.most severe transient with respect to CPR.
In 1981, the Reactor-Systems Branch of.WRR expressed concerns with the' credit given for equipment which is not classified as safety-related. ass Possible Solution The' solution proposed.for this issue would be to. require that the analysis of
. transients only rely on equipment classified as safety-related.
It is likely that this'would result in_. increased costs to industry either due to penalties
!in derating plants or in new equipment to meet the safety requirements.
I PRIORITY' DETERMINATION
]
Assumptions 6
Based on the concernsass expressed by RSB, PNL did an assessment 4 of this
'iss' e and: concentrated on those-transient events which involve the high-water u
level trip and the turbine bypass system which are not classified as
~ safety related.- Two approaches were used in analyzing this issue.
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-The first approach was to examine three major transients and evaluate changes in CPR due,to failure of the high-water level. trip (L8) and/or failure of the
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~ bypass valves (TBP) to oper.
The transients analyzed included:
(1) turbine i
. trip without bypass (100% power); (2) loss 'of one feedwater string _(100% power);
and (3)^feedwater control failure (high - 50% power).
Based on examining these-transients with the Browns Ferry Simulator,64 it was concluded that:
(1) the i
CPR did not exceed the. fuel cladding integrity limit; and (2) in each case 1
- other Reactor Protection System signals provided a scram and in no case examined-was the reactor vessel coolant inventory compromised due-to failure of the l
non-safety grade equipment.
Therefore, the potential for fuel damage and core-melt was consideret. negligible.
The second approa:h was a quantitative analysis to estimate an upper bound on
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Lpotential risk reduction assuming an upgrading (to safety system requirements)
.of the high-water level trip and the turbine bypass system.
Grand Gulf was used as the representative BWR.
- The rationale for the use of the. second approach to evaluate public risk is based on the premise that this fault combination comprises the majority of the public risk.
While the probable failure result is a departure from nucleate boiling (DNB) which may result in rupture of the fuel cladding, the
-consequences'are constrained by the' limited amount and form of radioactivity which is released.
Conversely, while the probability of failing to achieve-subcriticality is much less'than the probability of DNB, the consequences are
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so'much greater that it is~judgnd that the risk to the public is dominated by l
this failure sequence.
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Frequency / Consequence Estimate lT s2 c.(T s are those transients other than loss-of-offsite pos;e that require 2
emergency reactor shutdown' and C is the failure to achieve suberiticality) is the only affected Grand Gulf dominant accident sequence for this issue. T s was 2
redefined to include only those transients associated with failure of--L8 trip i
-and TBP.
Redefining T s involved--identification.of transient initiators -
2 associated with L8 trip and TBP.
These initiators and associated raean total frequencies of occurrence taken from the EPRI ATWS reappraisa1307 are as follows:
(1) ' Turbine trip with TBP valve-failure 0.01/RY (2) _ Feedwater--increasing flow at power 0.16/RY (3) -Loss of feedwater heater-0.04/RY 0.21/RY
.The percentage contribution of these non-loss-of-offsite power transients (0.21/RY) to the mean total frequency of all BWR non-loss-of-offsite power 1
transients-(8.78/RY) taken from1the EPRI report 307 is [(0.21/RY)/(8.78/RY)] x
-100% = 2.4% where:
' Total BWR transients 8.90/RY Loss of'offsite power 0.12/RY Total:
8.78/RY 7
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2-j The frequency of:the'above transients normalized to Grand Gulf (7/RY)C4
' produces the-redefined value of Tra= (0.024)(7/RY) = 0.168/RY.. The failure to. achieve subcriticality (C) is defined by C = (RPLS + CR)(MANSD) where:
~
1.9'x 10 6, the failure rate of the Reactor Protection RPLS
=
Logic System 5.8 x 10 6, the failure rate of the Control Rod CR
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System not inserting given a trip signal MANSD is-the failure rate of the recirculation pumps failing to trip, or.the operator fails to manually insert the control rods, or the operator fails ta initiate the Standby Liquid Control System-(SLCS). The MANSD failure is dominated by the SLCS failure rate, 0.1.
Hence,. for Grand Gulf, i
C = (1.9 x 10 6 + 5.8 x 10 6)(0.1) = 7.7 x 13 7 Thus, for Grand Gulf, the non-loss-of-offr.ite power transient accident sequence is given by:
Ts2 c = (0.168/RY)(7.7 x 10 7) = 1.29 x 10 7/RY
.The base-case, affected cere-melt frequency is given by F = 1.29 x 10 7/RY.
'The base-case, affected public risk is given by W r (1.29 x 10 7/RY)(7.1 x 106 man rem) u 0.92 man-rem /RY, where the dose in man-rem is that for a BWR-2 type l
release, as defined in Section 2.2 and Appendix 0 of NUREG/CR-2800.64 For the adjusted case,,it was assumed that improving the high water level trip and the turbine bypass system would increase the availability of the RPLS by a factor of 10 (at most); this would produce extremely conservative results.
- C*-(adjusted) =:[(1.9 x 10 6)(0.1) + (5.8 x 10 6)](0.1) = 6 x 10 7
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'T aC*
.= (0.168/RY)(6.0 x 10 7) = 1.01 x 10 7/RY 2
.The adjusted-case,-affected core-melt frequency is given by F* = 1.01 x
-10 7/RY.D The adjusted-case, affected public risk is given W* = (1.01 x 10 7/RY)(7.1 x 106 man-rem) = 0.72 man rem /RY.
The reduction:in core-melt l --
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frequency.(6F) is (F - F*) = (1.29 x 10 7/RY) - (1,01 x 10 7/RY) = 2.8 x 10 8/RY.
l The' reduction in public risk per' plant-(AW) is (W - W*).
Therefore, AW = (0.92 - 0.72) man-rem /RY = 0.20 man-rem /RY.
The total.public risk reduction is (44 BWRs)(27.4 yr) (0.20 man-rem /RY) = 240 man-rem.-
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-o Cost Estimate LA rough estimate of costs was made based on an upgrade (to safety system
. requirements) trip and the turbine bypass system.
It was estimated that equipment costs would be on the order of $500,000 for each system.- The engineering' and-installation costs would be double this cost.
Therefore, industry. costs would be on the order of $3M/ plant.
Annual operation-and maintenance was~ estimated'at about 2 man-weeks / year; this amounts to
. ($2270/ man-week)(2 man-weeks / year)(27 years / plant) or $150,000/ plant. The total industry cost was therefore estimated at $3.1M/ plant or (44 plants) x
($3.1M/ Plant) = $136.4M-for the affected plants.
NRC costs were considered negligible when compared to these costs and were not estimated.
Value/ Impact Assessment Based on an estimated risk reduction of 240 man-rem and a cost of $136.4M, the value/ impact score is given by:
240 man-rem 3,
$136.4M 2 man-rem /$M
=
Other Considerations The following items will also have an effect on the reliability of non-safety grade systems-and will impact this issue.
(1). The high-water level' trip is also being investigated as part of. USI A-47
" Safety Implications of Control Systems."
(2) Present licensing practice has been to require technical specifications-for the-surveillance of the high-water level trip and the turbine-bypass system.
(3) On future BWRs, (BWR-6) the high water level trip is part of the reactor protection system.
-(4) RES'is embarking on an Accident Management Research program and will be
- looking at existing and alternate strategies for arresting core damage and/or radioactivity release during a severe accident.
Credit is being:
considered for all equipment:
safety grade, non-safety grade, and even offsite equipment (i.e., fire trucks).
CONCLUSION Based on the calculated risk reduction, value/ impact score, and the other considerations, we believe that this issue should be DROPPED from further consideration, i
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- REFERENCES
. 64.
NUREG/CR-2800, " Guidelines for Nuclear Power Safety Issue Prioritization
'Information Development," U.S. - Nuclear Regulatory Commission, February 1983.
- 307. EPRI-NP-2230 "ATWS:
A Reappraisal, Part 3," Electric Power Research Institute, 1982.
385.. Memorandum for T. Murley from D. Ross "Use of Equipment n6t Classified as-Essential to Safety in BWR Transient Analysis," March 10, 1981.
386. Memorandum for T. Novak from R. Frahm, " Summary of Meeting with General Electric-on the Use of Non-Safety Grade Equipment," March 1979.
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