ML20212D289
| ML20212D289 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 10/23/1997 |
| From: | Barron H DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20212D293 | List: |
| References | |
| TAC-M94491, TAC-M94492, NUDOCS 9710310057 | |
| Download: ML20212D289 (14) | |
Text
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Y Duke Pow *r Comp 2ny A tw twp huyeny
'f0l\\ "'
- a sa 4,
12700 llagers ferry Rd.
Ilumemille, NC 28078-9340 H. R. Beeeea
%cr hnident, MiGirr (704) 875 4800 orItcs Nwicar Gntrazion ikpremens (704) 875-4809 art October 23, 1997 O
U. S. Nuclear Regulatory Commission Attention: Document Control Desh Nashington, DC 20555
Subject:
McGuire Nuclear Station Docket Nos. 50-369, 50-370 License Nas. NPF-9, NPF-17
.(NRC TAC M94491 and M94492)
'Ref.:
- 1) Emergency Plan Change Submittal dated December 27, 1995 2)NRC Request for Additional Information Regarding Emergency Action Level Revision to NUMARC/NESP-007 Methodology dated January 15, 1997
- 3) Duke Power Response to Request for Additional Information Regarding Emergency Action Laval Revision to NUMARC/NESP-007 Methodology, dated April 1, 1997
, Attachment 1 contains additional information requested by Larry Cohen, Bill Meier and Vic Nerses on June 23, 1997, regarding revision to McGuire's Emergency Action Level based on l
NUMARC/NESP-007 Methodology. Please contact Becky Hasty at (704) 875-4662 or Kay Crane at (704) 875-4306 if there are any questions on this information.
Yours truly, H. B. Barron HBB/EMK(j \\EMK7353\\MNS RAI 2. doc) h wen Illllllllilll($l0[llllli 9710310057 971023 PDR ADOCK 05000369 F
-PDR
e UiS. NRC Oct:ber 23, 1997
.Page-)
XC I.
L.;A. Reyes, Regional Administrator USNRC, Ragion II (2 copies)
V. Norses, Project Manager USNRC, ONRR-9.-iM. Shaeffer, Senior Reaident Inspector j'AcGuire Nuclear Station
!>3 i.. ';N '
C e
4
U.S. NRC-October 23, 1997 Page $
b'cc
-H. B. Barron (w/ attachment)
B. J. Dolan (w/ attachment)
R. L. Hasty (w/ attachment)
E. M. Kuhr (w/ attachment)
P. R. Newton (w/ attachment)
K. L. Crane (w/ attachment)
ELL (w/ attachment)
EP File 735 (w/ attachment) i E. M. Ceddie (w/o attachment) i M. T. Cash (w/o attachment)
F G. A. Copp (w/o attachment) l 3
[ \\
D j
g/
t List of Attachments
- 1. Responses to Issues 1-6 from June 23, 1997 conference call
.2. Revised Disposition of NUMARC/NESP-007, Rev. 2 Initiating Conditions and Emergency Action Levels in the McGuire EAL Submittal
- 3. ' Revised McGuire Emergency Plan Section D pages with Revision Marks
- 4. Revised McGuire Emergency Plan Section D pages without Revision Marks S. Revised McGuire Emergency Plan Implementing Procedure P?/0/A/5700/00 pages
~
4' r
e-g Response to NRC Questions pc McGuire Emergency Action Levels submitted December 27, 1995 and Response to nequest for Additional Information discussed in a conference call June 23, 19f7 s
Revision to Fission Product Barrier Matrix Th McGuire Emergency Planning Group performed a self-assessment in July, 1997, which included a review of issues from the March 1997 exercise at Three Mile Island.
As a result of this review, the following note is being added to the Fission Product Barrier
. Matrix:
Note 3: When determining Fission Product Barrier status, the Fuel' Clad Barrier should be considered to be lost or potentially lost if the conditions for the Fuel Clad Barrier loss or potential loss EALs were met previously during the
. event, even if the. conditions do not currently exist.
Qli Have the changes made as a result of the NRC's request for
' additional information (RAI) been discussed and agreed upon with the offsite authorities (State and local)? If so, are there letters of agreement you can submit on the docket to verify that agreement?
A1:
No, these changes have not been discussed and agreed
~"
upon with the offsite authorities. Duke Energy will review the NRC. approved NUMARC/NESP-007 based classification scheme 6
with'the offsite authorities and receive their agreement q
prior.-to implementation, Q2, NUMARC IC AUl,' AAl; Duke IC 4.3.U.1, 4.3.A.1: The NUMARC ICs address effluent releases that are multiples of the Radiological
. Tech. Specs.
The basis for the NUMARC ICs allows the use of the e
instantaneous release rate limit contained in the Offsite Dose l Calculation' Manual (ODCM) for plants that have eliminated the RETS.
The Duke plants list Selected Licensee Commitments (SLC) fas the source of the referenced release rate limit although the limit is also contained in the ODCM, which is an NRC controlled
,-document and part of the license.
What is the reason for
A2: Duke Power amended our Technical Specifications and s
relocated'the Radiological Effluent Technical Specifications g
(RETS) to the respective stations Selected Licensee Commitment (SLC) Manual.
This amendment was requested for
- ;;j L
all'three sites on February 7, 1990, and was approved on May 22, 1991, in Tech. Spec. Amendment 119 for McGuire Unit 1
-rL
((
q.
.ej one:
(jj t 0
h And 101 for McGuire Unit 2,-and'on December 9, 1992, in o
9-Techt spec. Amendment 97 for Catawba Unit 1 and 52 for b
Catawba. Unit,2.
The SLC manual.is Chapter 16 of the site's
[
"FSAR and is controlled by the 10CFR50.59 process.
L_
.,Q3, NUMARCIIC AS1, AGl; Duke IC14.3.S.1, 4.3.G.1: The bases for F
ithe NUMARC ICszstates that 'the FSAR source terms applicable to
- each monitored pathway should be used in conjunction with annual
-)]. average meteorology in determining indications for the monitors in that pathway."
A draft NEI white paper, dated 9/10/94, stated
- that-the use~of the FSAR accident source term without 4G
. modification could result.in_ conservative EAL thresholds and ipossible1 overlap with the Alert. threshold.
This white paper.
H recommends the use of-a reasonable" source term.-
The
> information: submitted in the response to the RAI indicates that the' source term. considered in the derivation of the Duke thresholds for' monitors is from the ODCM methodology (assumed Lyalue. for :"K"). What. is the rationale for using-the ODCM isotopic
. mix fornthe-higher levelEEALs that-are associated with some
~'(accident) level of core damage? Has_this assumed isotopic mix been~ modified in any'way?
V j{ '.A3: Duke Energy 1 originally-looked at accident source terms in performing. detailed monitor response calculations for the
. Site Area Emergency and General Emergency classifications,
.but. experienced'.significant overlap between the Alert and Site Area Emergency classification monitor readings.
The 6
primary-reason for che' monitor reading overlap was the
' assumed difference'in source term isotopic spectrum in calculating site boundary dose. rates that would cause an
' Alert-_or Site Area Emergency classification.
The NUMARC EALs work backwards from a site boundary dose rate to determine monitor readings, and the monitor response is not
.ednq 111nearly proportionalLto d0se.
For example, the Total LBodyl dose conversion ~ factor for Ar-41 (1.8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> half-life) is about 30 times higher than Xe-133, but the monitor response-is less-than 3. times' higher..If you a assume a
' hotter" (higherJenergy) spectrum,-it takes significantly fewerscounts on the monitor to get an equivalent dose at the 4
site' boundary, and,1therefore, monitor reading overlap can
-occur.~
EG
'To eliminateMthe' overlap, Duke Energy decided to use a
~ consistent, source term-throughout.
Xe-133 was chosen since titLis1the predominant noble gas isotope released during
-normal operations (e.g.,
97% of Catawba noble gas in 1996)
'and1the NUMARC EALs use site normal operating release rate
'limitsLas its starting point.
Our attempt at a " reasonable" l
lFSAR source: term assumption-resulted in significant monitor reading overlap, and we decided not to artificiully adjust 2
uz
w 4s the PSAR source terms without any real basis to " solve" the overlap problem.
The fact is that post-accident source terms will be continually changing, and will be extremely accident and time dependent.
Duke Energy also looked at this problem fro.n the context o.
that it.is very unlikely that a Site Area Emergency would Ie classified based on a radiation monitor reading. Accident classification is made primarily based on plant conditions.
In addition, we have implemented on-shift dose assessment.
Our on-shift dose assessment tool takes into account real
' time meteorology and accident source term spectra.
The radiation monitor readings will be used to classify the higher emergency classifications only if dose assessment cannot be completed within 15 minutes.
4
-The NUMARC EALs start with the 500 mrem /yr (= 0.05 mrem /hr) basis and then use multiples of that relcase rate for the four classifications (Unusual Event, Alert, Site Area Emergency, General Emergency).
To account for the higher energy FSAR source term for a Site Area Emergency and
~
General Emergency, we have made the multiple from the Alert to the Site Area Emergency a factor of 5 rather than 10.
This adjusts the_ monitor reading for a higher energy
.ci'
. accident spectrum and prevents overlap, while removing the e
non-conservatism of using the factor of 10 multiple between
.each of the four emergency classes. See attacnments 3 and 4 for revised Emergency Action Levels.
Q4, NUMARC IC AA2; Duko IC 4.3.A.2: The disposition cross-reference lists Duke EAL 4.3.A.2-3 as being equivalent to both
~
AA2.3 and AA2.4 in the NUMARC scheme.
Since EAL 4.3.A.2-3 only lists.a level-for the NC System and only applies to cases in which KF-122 cannot be closed, it is unclear how this EAL is equivalent to NUMARC EAL AA2.4 for cases in which KF-122 is 1 closed.
I recognize that Duke EAL 4.3.A.2-2 is listed as equivalent to NUMARC AA2.2, but there daesn't seem to be an equivalent-EAL in the Duke seneme that gives a preemptive spent fuel pool level for making a timely declaration for cases in v
Lwhi~ch the transfer valve is closed.
A4: McGuire agrees.
The reference to KF-122 has been deleted from EAL 4. 3. A. 2-3.
EAL 4.3.A.2-4 has been added to address NUMARC EAL AA2.4.
See attachments 3 and 4 for the 77 revised-EALs.
l i
-05, NUMARC IC AA3; Duke IC 4.3.A.3: For EAL 4.3.A.3-3, it is l
unclear how a. classifying official (especially one on-shift) can
. etermine, in a timely manner, when the EAL is satisfied.
Is the l
d
' declaration of this EAL contingent on the personnel expected to p
3 i
perform the safe shutdown activities, since the margin to the TEDE-limit is dependent on the existing annual occupational dose?
O-Is the classifying official expected to estimate the time to putform the necessary tasks?
Why isn't the EAL threshold based on a radiation level that is derived from estimated task duration and an assumed pre-existing dose?
y A5: We agree that the EAL needs to be clear to a classifying official and should not be personnel dependent.
McGuire has 0
Lchosen for EAL 4.3.A.3-3 to use a generic emergency action level of greater than or equal to 5000 mrad /hr for areas in the' plant that are normally accessible as low dose rate jy areas that have equipment installed, operated, and used for safe operation or safe shutdown of the unit.
This value has been chosen for the following reasons: (1) to be consistent
-r with equivalent NUMARC/NESP-007 based Emergency Action
~
Levels implemented at Oconee, and (2) radiation levels at or above this range may make it difficult to complete tasks necessary for safe operation of the plant or to establish or maintain cold shutdown without exceeding normal occupational 4
dose limits of 5 Rem per year TEDE.
The EAL has been modified to' include area monitor numbers and the location of the area monitor.
Also, similar information has been added to EAL 4.3.U.2-3.
See attachments i and 4 for revised EALs and basis.
Q6, NUMARC ES4; Duke IC 4.4.S.2: I can accept the use of the L
.CSFSTs as the indicators for this IC, but there appears to be a conflict inlthe classification of a Core Cooling - RED path between t ais EAL and 4.1.F.1 on the Fission Product Barrier matrix.
A person classifying a Core Cooling - RED condition by
^
referring to the matrix would classify it as an Alert condition,
'instead'of the SAE called for in 4.c.S.2.
This conflict needs to l
be resolved.
A6: McGuire agrees.
We have deleted EAL 4.4.S.2-2,
" Core Cooling CSF - RED" and renumbered the remaining EALs.
Ultimate heat sink and reactivity control are specifically 91 mentioned in the NUMARC IC SS4 basis, so that the core cooling-reference is not needed here. A Core Cooling CSF -
RED condition is classified es an Alert condition under E1.L 4.1.F.1. This eliminates the conflict.
See attachments ?
and 4 for the revised EALs.
,+
4 1
l
a o
t.
ATTACHMENT 2 Revised Disposition of NUMARC/NESP-007, Rev. 2 Initiating Conditions and Einergency Action Levels in the McGuire EAL Submittal c
Y w
' k_w, f,
y t
i.. '1 p
M
.+s M
l k'.
+
f
.g McGuire Nuclev Site Disposition of NUMARC/NESP-007, Rev.' 2 IC/EAL in McGuire EAL Submittal g
NUMARCIC/EAL McGuire EAL Recognition Category sad EAL Number Recognkion CategoryA
- AUl/l' Abnormal Rad Levels / Radiological Emuent 4.3.U.l.1 and -2
-AUl/2 Abnormal Rad Levels / Radiological Emuent 4.3.U.1 3 and -4 AUl/3 Deleted - McGuire does not have perimeter monitors AUl/4 -
Deleted - McGuire does not have real time dose assessment t
L AU2/1.
' Abnormal Rad Levels / Radiological Emuent 4.3.U.2 1
'AU2/2 Abnormal Rad Levels / Radiological Emuent 4.3.U.2 2 AU2/31 Deleted -- McGuire does not have dry fuel storage p
AU2/4 Abnormal Rad Levels / Radiological Emuent 4.3.U.2 3 o
AAl/l -
Abnormal Rad Levels / Radiological Emuent 4.3.A.1 1 and -2 AAl/2,
~ Abnormal Rad Levels / Radiological Emuer.t 4.3.A.1 3 and -4 P
~AA1/3 Deleted - McGuire does not have perimeter morJtors
- AAl/4 Deleted - McGuire does not have real time dose assessment f
AA2/1 -
Abnormal Rad Levels / Radiological Emuent 4.3.A.2 1 I
7.A2/2:
Abnormal Rad Levels / Radiological Emuent 4.3.A.2 2 AA2/3 Abnormal Rad Levels / Radiological Emuent 4.3.A.2 3 l! AA2/4 Abnormal Rad Levels / Radiological Emuent 4.3.A.24 L
- AA3/1 -
Abnormal Rad Levels / Radiological Emuent 4.3.A.3 1 and AA3/2 -
Abnormal Rad Levels / Radiological Emuent 4.3.A.3 3 ASl/l Abnormal Rad Levels / Radiological Emuent 4.3.S.1-1 ASl/2 --
Deleted - McGuire does not have perimeter monitors
-ASl/3 Abnormal Rad Levels / Radiological Emuent 4.3.S.1-2
/ ASl/4' Abnormal Rad Levels / Radiological Emuent 4.3.S.1-3 AOl/l Abnormal Rad Levels / Radiological Emuent 4.3.0.1-1 lAOl/2; y
Deleted - McGuire does not have perimeter monitors E
o AOl/3 Abnormal Rad Levels / Radiological Emuent 4.3.0.1 2 7 AOl/4 :
- Abnormal Rad Levels / Radiological Emuent 4.3.0.1-3 m
5 I
.s y
h
w JQ
!?
f n0 McGuire Nuclear Sita b
y Disposition'of NUMARC/NESP-007, Rev. 2 IC/EAL in McGuire EAL Submittal a
f+
MUMARCIC/EAL McGuin EAL Recognition Category and EAL Number Ascognitden CategoryF_
- P P B P uol/l Fission Product Barrier 4.1.F.1 u
{
Q:
? PPB Puol/2
- - Fission Product Barrier 4.1.F.2 p!c DPPB Puol/3 Deleted - Core exit thermocouples are monitored as part of
- E p;
f7 Critical Safety Function (CSF) status. _ If the OAC is not b
%y% " "
available, manual monitoring is donc per EP/l or 2/A/5000/10 dh l and will be used to classify.
V 4 PPB Puol/41 gp-
.i
^
Deleted - Reactor Vessel level is monitored as part of CSF gWw status. If the OAC is not available, manual monitoring is done
- per EP/l or 2/A/5000/10 and will be used to classify.
1
_ 2 P P B P uol/5 Fission Product Barrier 4.1.F.3.
b?
PP Deleted -- McGuire does not believe additional indications are djg 74 BPuol4 cc t W to hk Fiel Clad Barrier status. Air sampling -
4...
(PAOS) results are not used, as containment monitors can be Y*
% used to determine Fuel Clad status.
3 f ;/
P 1 PB Puol/7
%. Fission Product Barrier 4.1.F.4
,44
,TPPB RCS/l Fission Product Barrier 4.1.N.1 aw l PPB RCS/2
- Fission Product Barrier 4.1.N.2-
[%
7,
_ PPB RCS/3 Fission Product Barrier 4.1.N.3 Note: The classification procedure does not provide site specific criteria that a steam.
h ',
diagnostic information of this type belongs in the Emergency g"
generator is ruptured. Duke Power's philosophy is that i*
Operating Procedures. Once the Fmergency Coordinator / EOF r
14:; Ab-
~,. _
Dinoctor ibLmines that a steam generator is ruptured, the -
W T'
i classification procedure can be used to determine the correct Wn 9
~-'
classification.
dP f r 1 PPB RCS/4 Deleted - Containment process monitors alarm at too low a leak 1
f rate for NCS barrier loss and uto-isolate on safety injection.
oi Reactor Building monitors cannot indicate clearly activity levels
[%
from loss of coolant with normal activity.
) PPB RCS/S Deleted - McGuire does not believe additional indications are ruw w -
needed to determine Reactor Coolant System Barrier status. Air PE-
)
c 1 sampling (PAOS) results are not used, as containment monitors t,'^
- can be used to determine Reactor Coolant System Barrier status.
PPB RCSM Fission Product Barrier 4.1.N.5 VN, y~5!w PPB t'antal l PPB t'antaiammat/l Fission Product Barrier 4.1.C.1
_,o amant/2 Fission Product Barrier 4.1.C.2
&T FPB Containmaat/3
" Fission Product Barrier 4.1.C.3
? FPB Containment /4 -
Fission Product Barrier 4.1.C.4 e FPB Containment /5 ;
Fission Product Barrier 4.1.C.5 e
4 f*
6
-&9 L
-24 l -
A,-
- T k
- 1.
1.1..
McGuire Nuclear Site Disposition of NUMARC/NESP-007. Rev. 2 IC/EAL in McGuite EAL Submittal
[ FPB C*J'- +3t/6 Fission Product Barrier 4.1.C.6 Core exit thermocouples are 1
~'
monitored as past of CSF status. If the OAC is not available, manual monitoring is done per EP/l or 2/N5000/10 and will be used to classify.
- FPB Containment /7, Deleted -- no additional indications needed FPB Containment /8 Fission Product Barrier 4.1.C.7
'f y
i p
[."
(J:
e y
' ' ~
r 4
I f
a
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~
McGuire Nuclear Site Disposition of NUMARC/NESP-007. Rev. 2 IC/EAL in McGuire EAL Submittal g
NUMARC IC/EAL McGuire EAL Recognition Category and EAL Number -
R:cognit!vn Category H 1101/1 Natural Disasters, llazards, etc. 4.7.U.1-1 and.2 HUl/2 Natural Disasters, llazards, etc. 4.7.U.1 3 HUl/3 Deleted -- McGuire considers this redundant to Emergency Director Judgement HU1/4 Natural Disasters, Itazards, etc. 4.7.U.1-4 HUl/5 Fire and Security 4.6.U.l.2 HUl/6 Natural Disasters, Hazards, etc. 4.7.U.1 5 11U1/7 Deleted - McGuire believes no additional items are needed based on currently approved EALs HU2/1 Fire and Security 4.6.U.l.1
. HU3/1 Natural Disasters, Hazards, etc. 4.7.U.2 1 HU3/2 Natural Disasters, Hazards, etc. 4.7.U.2 2
. HU4/1 Fire and Security 4.6.U.21 HU4/2 Fire ad Security 4.6.U.l 2 and.3
- HU5/1 Natural Disasters, Hazards, etc. 4.7.U.3 1 L
1HAl/l Natural Disasters, Hazards, etc. -4.7.A.1 1 s-
- HA1/2 Natural Disasters, Hazards, etc. 4.7.A.1 2 g
- HA1/3 Natural Disasters, Hazards, etc. 4.7.A.1-3 4(
HAl/4 Natural Disasters, Hazards, etc. 4.7.A.1 3
' HAli5 Natural Disasters, Hazards, etc. 4.7.A.1 3 is HAl/6 Natural Disasters,11azards, etc. 4.7.A.1-3 m
HAl/7 Natural Disasters, Hazards, etc. 4.7.A.1-3
~ '
HA2/1 Fire and Security 4.6.A.1 1 and 4.6.A.2 1 HA3/1 Natural Disasters, llazards, etc. 4.7.A.2 1 e
HA3/2 Natural Disasters, llazards, etc. 4.7.A.2-2 HA4/1 Fire and Security 4.6.A.3 1 HA4/2 Deleted -- McGuire believes no others necessary based on currently approved EALs HA5/1 Natural Disasters, Hazards, etc. 4.7.A 3 1
- HA6/1 Natural Disasters, Hazards, etc,4.7.A.4-1 HSl/l Fire and Security 4.6.S.1 1 HSI/2.
Fire and Security 4.6.S.12 and -3 HS2/1.
Natural Disasters, Hazards, etc. 4.7.S.1 1 HS3/1 Natural Disasters, Hazards, etc. 4.7.S.2 1 HGl/l Fire and Security 4.6.G.l.1 HOl/2 -
Fire and Security 4.6.G.12 HG2/1 Natural Disasters, Hazards, etc. 4.7.0.1 1 9
l
L,:*
McGuire Nuclear Site Disposition of NUMARC/NESP-007 Rev. 21C/EAL in McGuire EAL Submittal NUMARC IC/EAL'.
McGuire EAL Recognition Category and EAL Number
. Recognition CategoryS
_ SUl/l Loss of Power 4.6.U.1-1 SU2/1 System Malfunction 4.2.U.1-1 SU3/1 System Malfunction 4.2.U.2 1 4
SU4/1..
Deleted - McGuire has no failed fuel monitor. Reactor coolant sampling is used to determine fuel clad degradation.
SU4/2 System Malfunction 4.2.U.3 1 - Applies in Modes 1 5 only because the Technical Specification applies only in these modes.
SU5/1 System Malfunction 4.2.U.4-1, 2, and -3 L
SU6/1 System Malfunction 4.2.U.5-1 and -2 p
_SU7/l.
Loss of Power 4.5.U.2-1 F
SM/1-Loss of Power 4.5.A.1-1 i.
4A2/1 Loss of Shutdown Function 4.4.A.1 1 h
SA3/1 '
Loss of Shutdown Function 4.4.A.2-1 SA4/1
= SAS/1^
System Malfunction 4.2.A.1-1 Loss of Power 4.5.A.21 SSl/l:
Loss of Power 4.5.S.1 1 SS2/1 Loss of Shutdown Function 4.4.S.1-1
. SS3/1 Loss of Power 4.5.S.2-1
. l SS4/1
_ Loss of Shutdown Function 4.4.S.2-1.andr2, ed 3 g
. SS5/1 Loss of Shutdown Function 4.4.S.3 1, -2, and -3
. SS6/1 System Malfunction 4.2.S.1-1
. sol /l Loss of Power 4.5.0.1-1
' SO2/1 and 2 -
Loss of Shutdown Function 4.4.0.1-1
?
81
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