ML20080R972
| ML20080R972 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 03/01/1995 |
| From: | George Thomas DUQUESNE LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 TAC-M81201, NUDOCS 9503100166 | |
| Download: ML20080R972 (9) | |
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-GEORGES. THOMAS-
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Otv6sion Vice President
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Nuclear Services
' Nuclear Power Division 7
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.U. Se. Nuclear Regulatory Commission--
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' Attn:. Document Control Desk' LWashington,1DC -20555-0001' subject:
Beaver Valley PoweriStation, Unit No.~_1
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Docket No. 50-334, License.No. DPR-66.
j Response to Baquest for Additional Information Related to Neutron Flux Instrumentation (TAC: No. _M81201).
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Reference:
Duquesne.. Light company letter to the NRC submitting the' report." Evaluation.of.the Adequacy of Existing' Neutron Flux Instrumentation'for NUREG-0737, Supplement 1," date'd May 31, 1991
'l i
This letter responds to the NRC s Request.for ' Additional:
Information (RAI),
dated November 29, 1994.
Duquesne?LightiCompany-
~i (DLC)- has.provided several submittals to the NRC.on'this issue'since-
-l' the' issuance.
of
- NUREG-0737, Supplement ^
1, in._ December
'.1982._
Agreement has.been reached on-each of the techni' cal'. Issues-of-'
1 NUREG-0737, with one exception; i.e.,_ environmental.' qualification"of-neutron flux instrumentation for beyond design basis-. accidents that j
produce
.a harsh containment environment.
In the1 referenced. report 1
submitted on May. 31, 1991, DLC provided technical justification for.
l not upgrading. the neutron flux instrumentation according.to the -
]
. guidance of Regulatory Guide 1.97.(Revision 2)..
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.-The existing-Unit No. 1 neutron flux monitors are appropriately t
qualified for the majority'of postulated accidents.
In thejunlikely event of a
beyond design basis accident causing. harsh containment conditions, the consequences of neutron
. flux.. monitoring
]
unavailability will not prevent the : operator from determining:and-1 taking-appropriate corrective action.-
-Use of the temperature i
monitoring instrumentation
- system, in -the manner describedjin.the H
Emergency Operations -Procedures (EOPs),
allows 'the operator -' to
.]
' predict the appropriate event paths and to take corrective action to a
mitigate the resulting conditions.
j 1
.It was estimated in the referenced report th'at an upgrade of:the-
~j existing neutron flux' instrumentation-will, cost in excess"of $3 i
million.
As-described in the attached response to the' subject RAI,-
.this. expense-provides little or no safety benefit':nor ~ derives 7
increased protection, and an upgrade will not result in_any changes 1,
to the mitigative actions prescribed by existing EOPs.
Therefore,-'we-believe the additional cost is unnecessary because the existing H
neutron flux instrumentation is properly qualified for all its design n
9503100166 950301 PDR ADOCK 05000334 O
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'BOnver. Valley Pow r Station, Unit No. 1 Response to' Request for Additional Information
-Dated. November 29,-1994 Page 2 basis accidents-including the beyond' design basis ATWS event.-
In
-lieu of upgrading neutron flux instrumentation, 'the use of the existing qualified. Reactor Coolant-System; temperature measurement instrumentation during events 'which could generate an adverse containment environment will allow the operator to confirm return to core power conditions.
We believe
.our position to not upgrade. the ' neutron flux instrumentation is. consistent with our licensing basis and pertinent regulatory requirements.
This submittal is the result of a-cooperative effort with Consolidated Edison and Rochester Gas and Electric.
Once the NRC has reviewed our attached response, as well as those from the other two utilities, we would be pleased to meet with you in a technical forum to discuss any additional concerns and to finalize the resolution of this issue.
Sincerely,
'l Ge rge S.
Thomas cc:
Mr.
L. W. Rossbach, Sr. Resident Inspector Mr.
T.
T. Martin, NRC Region I Administrator Mr. G.
E.
Edison, Sr. Project Manager
~
Response to NRC Request for Additional Infcrmation
, Related to Neutron Flur Instrumentation. Dated November 29, 1994 4
1 QUESTION 1 A
major difference between the existing neutron flux measurement and the proposed temperature measurements is the additional delay introduced, during a
reactivity insertion
- accident, by the tima required for the temperature measurement to detect sensible heat.
For slow reactivity insertion rates this delay can become substantial.
In order to evaluate this effect, provide a quantitative estimate of the time delay in identifying a situation in which the reactivity is increasing (assuming, e.g.,
a constant reactivity insertion rate) for a
complete range of reactivity insertions.
Provide a detailed evaluation of the effect of this delay on plant safety analyses, accident consequences and required operator action, relative to the case where the increasing neutron flux is detected by the flux 4
instrumentation at approximately 10 % power.
Any comparison to the excore neutron flux instrumentation should only be made for the condition where the water level is measured to be above the hot
- leg, and the neutron flux provides a proportional indication of the core neutron flux.
BfSPONSE TO OUESTION 1 It is important to understand three fundamental points with respect to the proposed use of Reactor Coolant System (RCS) temperature measurements (core
- exit, hot
- leg, and cold leg) in lieu of ers.onmentally qualified neutron flux instrumentation:
1.
9.te temperature instrumentation only replaces neutron flux instrumentation under harsh containment conditions.
These conditions exist for the design basis accidents described in our May 1991 submittal (Section 3.1), and a number of beyond design basis accidents which were considered during Emergency Response Guideline (ERG) development (Section 3.2).
2.
As is implied in the NRC question, neutron flux instrumentation is not always proportional to reactor power, and therefore may provide anomalous indications which can potentially mislead the operator.
Excore neutron flux instrumentation response is dependent on the location of voiding in the core and/or downcomer, the degree of core uncovery, and detector location.
These are usually local, non-homogeneous conditions which may not be detected by the neutron flux instruments.
This is particularly likely for accidents which produce harsh containment environments, since reactor vessel voiding may be occurring.
Anomalous neutron flux indication (i.e.,
indication not proportional to reactor power) was observed at the Three Mile Island accident (as shown in Appendix Recrit of NSAC-1), and has been demonstrated in NRC funded experiments at the pennsylvania State University Breazeale nuclear reactor and LOFT facility (Nuclear Technology, Vol. 96, December 1991).
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1Racponco. to' NRC RequOct fcrl Additien21 Inform 2 tion:
tRelated.to Neutron Flux Instrumentation, Dated November 29, 1994 K,
.Page 2 L
RESPONSE TO OUESTION 1 (Continued) 3.
Post-accident criticality is only monitored as part of the Emergency-Operating Procedures (EOP)
Critical' Safety Function Status Trees
'CSFST).
The CSFSTs are only. intended to determine
(
.if an imminent threat to a critical safety function exists.
The CSFSTs act. as a
decision point in order to determine if the i
, operator should-immediately suspend the performance of optimum-F recovery procedures, in order to address this challenge.
Westinghouse. Owners Group (WOG) guidance suggests that this:
decision should-be made only if there is a continuous challenge to the safety function.
As described in our previously submitted 1
- report, a
continuous challenge does not exist unless the core is i
producing enough power that a temperature rise is indicated on the. RCS temperature instrumentation.. This rise results when core i
power. exceeds decay heat removal capability, well above the power levels suggested for analysis by the NRC question.
j Based on these three
- points, time response of the temperature measurements is not a
concern for the intended function, nor is it necessary, under post accident conditions, to be able to detect core criticality at the power' level suggested in the question.
l L
OUESTION 2 Describe any unique plant-specific design features or. operating conditions that support the use of temperature measurements for criticality, rather than the existing neutron flux instrumentation..
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JtESPONSE TO OUESTION 2 The proposal is to use temperature measurements for gross indication of core power production, rather than.the existing neutron flux instrumentation, and then only under conditions where-the existing i
instrumentation is not qualified.
There are no features unique to the Peaver Valley plant that support this position.
The studies referenced in Item 2
of the response to Question No. 1 support the use of RCS temperature indication in lieu of neutron flux instrumentation under harsh environmental conditions at any PWR.
?
OUESTION 3 i
Since the temperature measurements only determine that a critical i
state exists and' sufficient power is being generated to be measured on the temperature instrumentation, describe how the proposed temperature measurements will determine the subcritical states of the
)
core as suggested in Section-I.
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Rarp;nco to NRC R:qu;;t for Additional Inforantion Related.to Neutron Flux Instrumentation, Dated November 29, 1994 Page 3 RESPONSE TO OUESTION 3 As described in the response to the first question, the WOG emergency response guidelines for use of CSFSTs, and consequently the Beaver Valley EOPs, only implement safety function restoration procedures if a
continuous challenge to that function exists.
This implies the need to monitor gross changes in core power, not the subcritical state of the core.
A precise measurement of neutron flux, even if it is a
valid measurement, is not necessary to determine if a challenge to a
safety function exists.
Only a gross indication of core power is required which would be evidenced by an RCS temperature increase.
Additionally, the nhutdown margin is verified by the operators monitoring the boron concentration in the containment sump or, as applicable, the RCS and confirming it to be above the minimum shutdown value.
OUESTION 4 Regulatory Guide 1.97, Rev.
3, recommends measurements that a) provide a direct measurement of the desired variable (flux in the case of criticality),
and b) minimize the development of conditions which could cause the measurements to give anomalous readings that would be potentially confusing to the operator.
In view of these recommendations, discuss in detail the ability of the core exit thermocouples and the hot and cold leg temperature measurements to provide an accurate indication of criticality in the presence of large uncontrolled and potentially unknown variations in the core flow and heat removal rate during accident conditions.
RESPONSE TO OUESTION 4 As indicated in the response to Question No.
2, the existing neutron flux instrumentation will be used for all events which do not generate an adverse containment environment.
Potential variations in core flow and heat removal rate during accident conditions affect the accuracy of neutron flux instrumentation as well as RCS temperature indication.
The important point is that the EOPs do not require a highly accurate indication of criticality.
The indication is only required to determine if there is an imminent threat by the critical state of the core to one of the barriers such that a release of radioactive materials to the environment could occur.
This only requires a gross indication of core power.
The proposed temperature indication is more than adequate for this purpose, and is less likely than excore neutron flux instrumentation to provide ambiguous and misleading information to the operator under the conditions described in this question.
t Re~p naa t3 NRC R:qu20t fcr Additienal Inform; tion.
Related.to Neutron Flux Instrumentation, Dated November 29, 1994 Page 4 I
4 OUESTION S In certain situations, the critical boron versus fuel-burnup curve is
~
used to determine if the coolant boron concentration is adequate to insure subcriticality during accident conditions.
How does this curve account for the range of beyond Design Basis Accident (DBA) core conditions.
i RESPONSE 'IU OUESTION 5 Environmental qualification of equipment for beyond design basis accidents is beyond the scope of NUREG-0737 requirements.
For events beyond design basis but addressed by the EOPs, the EOPs are designed to prevent core damage and/or melting.
The EOPs currently use boron values corresponding to minimum shutdown margin for time in core life.
These boron concentration curves are utilized throughout the
It is expected that the required suberitical boron concentrations are adequate for any conceivable post accident core configuration.
f OUESTION 6 What are the qualified temperature limits of the plant core exit thermocoupl s and hot and cold leg temperature measurements?
How will criticality be determined when the plant is outside these limits?
1 RESPONSE TO QUESTION 6 The qualified limits for the core exit thermocouples are 32 to 2300*F; the hot and cold leg temperature measurements indicate a range of 0
to 700'F.
These limits are consistent with the NRC
+
approved Environmental Qualification Program.
These ranges: a) are sufficient to determine core criticality for all conditions where core criticality is a concern, b) are consistent with EOP utilization of these instruments, and c) are consistent with Regulatory Guide 1.97 (Revision 2) guidance for these instruments.
A temperature in excess of the 700*F value indicates that either decay heat removal has been
- lost, or that the reactor is not subcritical.
(Refer to Question 11, and to Section 4 of our submittal report.) With the exit thermocouples indicating a temperature in excess of 2300*F, the core I
void fraction would be such that core criticality most likely could not be achieved even with the control rods completely withdrawn.
n R3;pon03 to NRC Rcquact for Additional Inform 2 tion Related to Neutron Flux Instrumentation, Dated November 29, 1994 Page 5 OUESTION 7.
Under what specific conditions will the neutron flux instrumentation
[
and the (core exit thermocouple and hot and cold leg) temperature measurements be used to determine criticality?
If the neutron flux instrumentation will not be used during conditionc of a hostile environment, how will these conditions be identified?
How will it be assured that the Category 3 neutron flux instrumentation is not used l
under conditions for which the instrumentation system is not qualified?
i RESPONSE TO OUESTION 7 l
RCS temperature indication is only used in lieu of excore neutron flux instrumentation for determination of the critical status of the core when harsh containment conditions exist.
These conditions are i
called out in the EOPs and are periodically monitored by operators, e
as well as reported on the Safety Panel Display System (SPDS).
EOP instructions will direct the operator to the correct CSFST based on the environmental conditions in containment.
This is consistent with l
other EOPs where direction is given on the use of post-accident instrumentation based on whether normal or harsh conditions exist in containment.
Instruments used to complete the EOPs are qualified to the necessary criteria.
i OUESTION 8 Have any special interpretations been made in the application of the t
Westinghouse Owners Group Emergency
Response
Guidelines to accommodate the use of the temperature measurements for the subcriticality function?
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RESPONSE TO OUESTTON 8 No special interpretations of the WOG ERGS have been made.
The EOPs instruct the operator to use RCS temperature indication in lieu of excore neutron flux indication under harsh environmental conditions.
As detailed in Section 4.2 of our submittal report, the plant remains i
within the technical basis of the ERG guidelines, and differences are fully consistent with the EOP diagnosis and mitigative strategies.
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OUESTION 9 The Chapter-3 evaluation of the beyond DBAs considered the loss of l
reactor
- coolant, loss of secondary coolant and steam generator tube rupture events.
How are the other beyond DBAs included in the safety evaluation?
f i
R Sp:nO3 to NRC R qu 3t fer Additional Information Related to Neutron Flux Instrumentation, Dated November 29, 1994 Page 6 f
RESPONSE TO OUESTION 9
[
f The design basis events evaluated in Section 3.1 of our submittal report and the beyond design basis, but within EOPs, events evaluated
[
in Section 3.2 are the events which result in harsh containment i
environmental conditions, i.e.,
the events in which RCS temperature indications would be used in lieu of excore neutron flux instrumentation.
The intent of Section 3.2 of the submittal report is to demonstrate that should certain design basis events degrade due l
to multiple equipment failures (i.e., beyond the design basis of the plant) and produce harsh containment conditions, proper EOP guidance would be provided by use of the temperature monitoring instrumentation.
For those other design basis events which degrade but do not produce harsh conditions, the existing -neutron flux instrumentation would be used.
Other beyond design basis events that fall below the probability threshold of the ERGS are outside the guidance of Regulatory Guide 1.97, the requirements of Supplement 1 to NUREG-0737, and the scope of these submittals.
t OUESTION 10 j
Discuss how the proposed core exit thermocouple and the hot and cold-leg temperature measurements satisfy the very strong recommendation j
of ANSI /ANS-4.5 that:
a) the criticality measurement should be made with a
flux detector which spans the range from 1x10*8 to 1x10'3 of full power or an equivalent or better i
alternative, and b) to the extent possible, the selected measured variables shall be those that most directly monitor subcriticality.
j Any comparison to the excore neutron flux instrumentation should only be made for the condition when water level is measured to be above i
the hot leg, and the neutron flux provides a proportional indication j
of the core neutron flux.
RESPONSE TO QUESTION 10 Our plant is not designed nor committed to ANSI /ANS-4.5 (or Regulatory Guide 1.97 Revision 3);
- however, ANSI /ANS-4.5 l
...specifically does not preclude use of other variables or combinations of variables."
Qualification criteria for instrumentation is established based on the safety function of the system whose variable qualification category is based upon whether monitoring of system parameters is needed during and following an 1
I accident and whether subsequent operator actions are dependent on the information provided by this instrumentation.
The DLC report, dated May
- 1991, analyzes event scenarios to determine the consequences of neutron flux monitoring unavailability and concludes that the l
failure of this instrumentation will not prevent the operator from i
determining whether mitigative action is required.
As described in
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R Sp;nr3 to NRC R:qusat for Additional Information Rel,a,ted to Neutron Flux Instrumentation, Dated November 29, 1994 Page 7 RESPONSE TO QUESTION 10 (Continued) the responses to Questions No.
1 and No.
4, we believe that the alternate parameter indications (RCS temperature) provide a better alternative than excore neutron flux instrumentation for monitoring the EOP CSFSTs for criticality during accidents which produce harsh containment conditions.
This is consistent with the stated qualification criteria as subsequent operator action is dependent on RCS temperatures or boron concentration.
This is discussed in our submittal report in Section 4.
QUESTION 11 Describe the method used to determine the specific threshold values for the (core exit thermocouple and hot and cold leg) temperature measurements and the boron concentration that are used to protect from the effects of reactivity insertion events.
RESPONSE 'IT) OUESTION 11 The temperature value (700*F) is chosen to be consistent with the value used for the Core Cooling
Background
Document recommendations.
It is intended to indicate the presence of superheated conditions in the reactor
- vessel, an indication that either decay heat removal has been lost, or that the reactor is not subcritical.
The conservative action is to address the critical status of the core
- first, unless it is known to be subcritical (i.e.,
adequate boron concentration).
The subcritical boron concentration curves are established during each Reload
- Analysis, with margin added to compensate for uncertainties and stuck control rod (s).
These curves are consistent with those currently used throughout the
challenge to core subcriticality.
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