ML20207E818
| ML20207E818 | |
| Person / Time | |
|---|---|
| Issue date: | 07/18/1986 |
| From: | Scarano R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Jordan E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| NUDOCS 8607220432 | |
| Download: ML20207E818 (16) | |
Text
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9 f UL 181986 MEMORANDUM FOR: Edward L. Jordan, Director, Division of Emergency Preparedness and Engineering Response, Office of Inspection Enforcement FROM:
Ross A. Scarano, Director, Division of Radiation Safety and Safeguards, Region V
SUBJECT:
REVISION OF THE EMERGENCY RESPONSE FACILITIES APPRAISAL PROCEDURE Your June 12, 1986, memorandum to me, which transmitted a copy of a draft revision of the subject procedure, asked for our views regarding the scope and depth of the ERF Appraisal and its use in conjunction with IP 82301. We have examined the subject procedure and in response to your solicitation would like to offer the following comments:
Based upon our having completed ERF Appraisals at 67 percent.of our reactor sites, we believe that scheduling such an appraisal to coincide with a licensee's EP exercise will cause several problems. To conduct a minimally acceptable inspection of the exercise requires at least a three (3) day effort. At least four (4) days is required to conduct an ERF Appraisal envisioned by the subject revised procedure. Allowing for the duplicative effort, i.e.,
site access training and exercise / walk-throughs, at least 5.5 days will still be required to perform these activities and there will likely be inadequate time to prepare for the exit interview. Also, a problem is created by mixing two different types of inspections (an appraisal and observing an exercise) and using persons who do not perform inspections on a regular basis. With the proposed team composition, we would expect the effort of satisfying both procedures (82212 and 82301) during a single week's visit would result in less attention to the exercise observation requirements of 82301 than is envisioned by the inspection program. Under the existing staffing restrictions, it is not likely that additional persons will be provided to observe the EP exercise.
The revised procedure does not adequately address the subject of the training required by NRR and contractor personnel who are to participate in the observing of the exercise and the appraisal.
In inspection guidance Paragraph 033.c, the procedure refers to a Manual Chapter 0320; however, this chapter doesn't exist. The procedure needs to include a discussion on the training required by NRR and contractor personnel who will participate in this inspection / appraisal program. This training discussion should also discuss the differences between making an inspection and performing an appraisal.
The matter of material to be assembled and distributed to the various team members has not been adequately covered. The next to the last sentence in guidance Paragraph 033.c identifies some of these materials.
Guidance Paragraph 038 references a licensee submittal to NRR addressin
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JUL 181986 i
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Regulatory. Guide 1.97 data. Some material, i.e.,
description of the dose.
assessment program and the emergency _ data acquisition system, has not been a:Idre: sed. The Region:is in a position to provide copies of applicable previous inspection l reports, an example of the appraisal report ~(format)',Leopies of the emergency plan and,related implementing procedures > and copies of pertinent pa'rts of the.FSAR.
The guidance needs to clearly delineate.who identifies and provides what materials that are to be sent to the; team raembers for study prior to conducting the appraisal.
i4 i
We believe that " deficiencies and "open items" as presently used in the revised pro ~cedure may cause some.misunde'rstanding of these terms.
Deficiencies must refer to those items'that meet the term's use in 10 CFR 50.54(s)(2)(ii): The use of " ope'n items" must be consistent with the definition in,(IE Manual Chapter) 0610-03.06.
In guidance Paragraph 039.d there appears to be an implication that "open items" refers to those. items which the licensee or applicant agrees to correct prior to the issuance of the ERF Appraisal report. Under the inspection / appraisal program envisioned by the revised procedure, it is not likely that all the items to be classified as "open items" will be identified at the exit interview. Thus, the licensee or' applicant will wait for the report to be issued before addressing the subject of corrective action. We suggest that the words " items agreed to be corrected prior to issuance of the ERF Appraisal report" be eliminated.
We believe there are time requirements in this procedure that'are not very workable.
Guidance Paragraph 037.b states that at least two weeks prior to the onsite appraisal contact plant management. This contact needs to'be made at least one month prior to the visit. The time for issuing the ERF Appraisal report in' guidance Paragraph 039 may not be met.
Since inputs covering most of the topical material are not due to the Team Leader until three weeks after the appraisal, the draft report will probably be completed in 5-6 weeks. This only leaves about two weeks for the distribution and review by the EPB:0IE and team members.
In addition, during this period of time the Team Leader will likely be out of the office for at least one week on an inspection. Therefore, we suggest that the report be issued "no later than 90 days after leaving the site."
There are three more specific comments we would like to make:
1.
The fourth sentence of guidance Paragraph 034.c seems to imply there are some team members who will be appraising specific areas on an independent. basis and, thus, not under the direction and control of i
l the Team Leader. ~The wording should be changed to eliminate this possible interpretation because all members of the team are to be directed and controlled by the Team Leader.
I 2.
The material covered in guidance Paragraph 037.e is oriented almost exclusively toward the ERF Appraisal. Because the observation of the EP exercise required by IE Inspection Procedure 82301 is also
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- JUL 181986
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being accomplished by the appraisal team, the subjects in the
" Discussion" section of Enclosure 1 to Procedure'82301 should be discuss'ed as well.during the' meeting held prior'to the start of the
' inspection / appraisal.
.3.
Guidance' Paragraph 038.a mentions a_ licensee report to NRR concerning their implementation-of Regulatory Guide 1.97.
The NRR evaluation (SER)~of this licensee submittal is not identified. We suggest the words "an'd the related SER" be inserted after the word
" report" on the eleventh line of the paragraph.
Our suggested changes to Appendix 1 have been attached in the form of marked-up pages.
We appreciate the opportunity to comment on the revision before it is implemented. Further clarification or questions you may have concerning our comments should be directed to Ray Fish of my staff, FTS 463-3761.
}D Ross A! Scarano, Director Division of Radiation Safety and i
Safeguards
Enclosure:
As stated cc w/ enclosure:
T. Martin, Region I i
l J. P. Stohr, Region II J. Hind, Region III R. Bangart, Region IV D. Matthews, EPB, IE E. Williams, EPB, IE bec w/ enclosure:
RSB/ Document Control Desk (RIDS)
G. Cook B. Faulkenberry J. Martin Region V M.
m>
RFish/horma FWenslawski RAS arano 7/17/86 7/fl/86 7//7/86
R:v. I pig > 1 cf 42 APPENDIX 1 APPRAISAL CRITERIA AND REQUIREMENTS 1.0 TECHNICAL SUPPORT CENTER (TSC)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS 1.1 PHYSICAL FACILITIES 1.1.1 Desfon 10 CFR SO.47(b)(8)(11), Supplement 1 to NUREC-0737 (TSCrequirementsa,b,c,dc,[,[andk) 1.1.1.1 Size and Layout g
p a.)
coveukey 1.1.x a.
Is the size of the TSC adequate to accommodate and support NRC and : ; ;;;;?;n t:f Itcensee predesignated personnel, equipment and documentation to perform the intended functionst e.g.,
a work space for Individual and o
working group functions e work space for document and drawing review
- space for equipment and Instru-mentation use, maintenance and
. n e w a.
v b.
Is the layout designed to
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provide unimpeded personnel traffic and information flow?
c.
How were the TSC size and layout verified as adequate?
1.1.1.2 Location is the TSC located within the plant site protected area and situated to facilitate the necessary interaction with the Control Room, OSC, EOF, and other personnel involved in the emergency?
Note 1: The " bullets" fn11owing the questions are g requirements and need not be addressed to demonstrate compilance with either Supplement 1 to NUREC-0737 or the regulations (see items 033a and b).
m" A b f) 1
Rev. I Pag 2 2 af 42
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1.0 TECHNICAL SUPPORT CENTER (TSC) (contd) i CRITERIA ACCEPTANCE REQUIREMENTS / COP 94ENTS 1.1.1.3 Structure a.
Was the TSC built in accordance with the Uniform Building Codef b.
Is it capable of operating uninter-ruptepduringperiodsofactivation?
e.g., a e adverse weather conditions loss of offsite power e
systems reifability o
1.1.1.4 Habitability /Enitronment y
a.
goes the TSC provide r:d':?:;'::? protection :-d
'" 'n; #
.' ut to assure that any person working in the TSC would not receive radiation exposure in excess of 5 rem whole body or its equivalent (25 rem to the thyroid) fo{ the duration of the accident 7 e.g.
gamma radiation shleiding e
e isolated ventilation ventilation filtering for airborne e
lodine and particulates reserve air supply e
positive pressure atmosphere e
b.
What are the bases for determining thea{equacyofTSChabitability, e.g., a design basis for outside environ-e ment (10 CFR S0, Appendix A, NUREC-0578,etc.)
documentation of calculations e
and :-asurements Note la IBID page 1.
j Rev. t P:g2 4 ef 47
.4 -
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1.0 TECHNICAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS 1.1.2 Radiolog! cal Equipment. and 10 CFR 50.47(b) f1); 10 CtR 50, Appendix E ilV.E.1, Supp1fes Supplement 1 t6'NUREC-0737 (TSC requirement f) 1.1.2.1 Radiation Monitoring a.
Does the TSC have adequate equipment and Instrumentation (quantity, type, range and sensitivity) to determine dose from direct radiation exposures and airborne concentrations underaccidentcondit{ons A x ". '. ' '. ) ?
e.g.,
e rate measuring gamma Instrumenta-e tion (portable or fixed) af r sampling equipuent
)
high range dosimeters (([pg/~M h )
e TLDs or film badges b.
Is the equipment adequately maintained, calibrated and inventoried?
c.
Are adequate radiological instruments and equipment of appropriate quantity, type, range and sensitivity available in the TSC for l
use in traveling outside the TSC under accident conditions 7 d.
Is a procedure available for tracking dose to Individual TSC personnel throughout the course of an I
accident?
I 1.1.2.2 Protective Suppifes I
i Are there adequate and dedicated j
protective supplies available or l
readily accessibly to all TSC and augmentation personnel sufficient to l
suppo{ttheirassignedfunctions?
e.g.,
t respiratory protection e
e decontamination supplies and equipment protective clothing
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Rev. I P g) 5 ef 42 1.0 TECHNICAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS /COPO4ENTS N) 1.1.3 Non-Radf ologleal Ecutoment and 10 CFR 50.47(b
- 10 CFR 50, Appendix E fiV.E.9, Supplies Supplement 1 to REC-0737 (TSC requirements a,g,h and I) 1.1.3.1 Records / Drawings Are appropriate records and drawings adequatelystored, main {ainedandupto date in the TSC7 e.g., a as-built plant drawings e
e FSAR
/(h e emergency plan lO e EPIPs
[A
'N schematic diagrams e
J!, procedures [
(L gh e notification lists equipment manuals.
e 1.1.3.2 Support Supplies Are adequate supplies and information available, sto{ed, and maintained in '
the TSC7 e.g. :
maps (10 and 50-mile EPZ) e l
e plant floor diagrams isopleths e
e calculators e means for data trending pens, pencils, paper e
e conversion tables and other references availability of emergency vendor e
assistance comp $ N & paper and ribbons.
l fQ M uter o
},/,33,fPowerSupplies M
1.1.3./
Do the power supplies assure that the TSCwillfunctionwithoutipterruption during an emergency?
e.g.,
t diesel power suppif es e
battery power supplies emergency lighting e
e alternate sources of of fsite power.
l NOTE 1: IBID page 1.
Rev. I P:g3 8 sf 42 1.0. TECHNICAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE REOUIREMENTS/CONNENTS d.
Are the available 1.97 variables adequatetodetermir.ergactorcoolant system integrity 7 e.g. :
e pressurizer level or reactor vessel level letdown and makeup flow rates e
high radiation levels or fission product activity in containment high activity in the steam system e
(PWR) a physical break or crack in the e
coolant system piping e failure of relief valves to reseat.
e.
Are the available 1.97 variables adequatetodetermineprimary containment integrity 7 e.g.
e a physical break or crack in a containeent penetration containment overstressing by high e
temperature and pressure e failure of a containment isolation logic or valve to operate or inability to isolate o H concentration in containment hfghradiationorradioactivity e
levels in containment interfacing systems LOCA e
f.
Are the available 1.97 variables adequate to determine the operability, ca city, and integrity of the Ifquid, soli and gaseous rad waste systems 7 e.g.
- fypto m h mk h
a failure of waste gas holdup tanks or their relief valves hydrogen recombiner or offgas if piping explosions N
inadvertent discharge of untreated or concentrated wastes.
Note la IBID page 1.
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l R;v. I P:gs to cf 42 h
a 1.0 TECHNICAL SUPPORT CENTER (TSC) (contd) y CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS
'A 1.2.2 Data Acquisition 10 CFR 50.47(b)(9); Supplement 1 to NUREC-0737 (TSC requirement h). General Design Criteria (CDC) 24 Regulatory Culdes 1.97, Rev. 2 and 1.75.
p 1.2.2.1 Data Collection Method a.
How are the data acquired? e.g. :
video techniques digital or analog instruments computerized acquisition system e
- voice communication b.
Is the capacity of the data collection equipment sufficient to access all of the data to be transmitted to the TSC7 1.2.2.2 Time Resolution a.
Is the sampling frequency of each variable adequate to ensure detection of significant changes, particularly accident conditions?
l b.
Is the time resolution for the transmission of each of the available variables adequate to assure that no significant data is lost?
1.2.2.3 Isolation a.
Is the signal isolation performed l
for those variables obtained from safety systems adequate to assure that the safety systems will not be degraded by the data acquisition system?
. b.
Was the isolation of the installed system verified and validated? If so, how?
l l
Note 1: IBID Page 1.
1 Rev. I Pag) 11 cf 42 1.0 TECHNICAL' SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS 1.2.3 Data Comunfeations 10 CFR S0.47(b)(9); Supplement 1 to NUREC-0737 (TSC requirements g and h) 1.2.3.1 Capacity a.
What is the channel capacity of the data system 7 b.
Is it adequate to meet the needs of the data system under peak load and accident conditions?
1.2.3.2 Error Detection a.
What techniques are used for error detection / correction?
b.
Does this technique assure error detection from sensor to CPUT 1.2.3.3 Transmission Between ERFs a.
What methods are used for data transmission?
b.
Is data transmission adequate between the TSC, the Control Room, and the EOF 7 1.2.4 Data Analysis 10 CFR S0.47(b)(4 (b)(9); 10 CFR S0, Appendix E flV.E.2 and flV
.3 Supplement 1 to NUREC-0737 (TSC requirement 1.2.4.1 Reactor Technical Support a.
Is the data analysis adequate to h,h.3 d support the TSC functions?
b.
Will the data analysis facilitate determinationofreactorsta{uspast, present and projected 7 e.g. :
Forecasting (trending) j e
containment pressure vs. time j
containment temperature vs.
time Note 1: IBID Page 1.
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Rev, l Pag 2 14 ef 42 1.0 TECHNICAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE REOUIRENENTS/ COMMENTS b.
Does the Ifeensee have the capability to provide dose projections
' __;- at the site boundary 4e.- k
'p T du?
. int: c:'-
f r y 15 minutes 7 c.
Can the licensee make timely plume exposure dose projections to about 10 miles from the site for adequate protective action reccomendations?
d.
Based on the variables available and calculational methods used, Is there adequate information to determine source terms fo{ all potential release pathways?
e.g. :
e effluent monitors e containment monitors e containment leak rate fuel storage e
e Ifve time environmental monitor post-accident sampling results e
in plant radiological monitoring e
inoperable or offscale monitoring e
instruments.
e.
Do source term methods provide for cont;ibution by radionuclides? If so, what radionucifdes and what.
contributions and how are they determined?
e.g.
l e real time measurements e default values simp 11fyIng assumptions e
l e
laboratory analysis i
f.
Are the meteorological variables and calculational methods adequate to l
characterize the meteorological conditions to about 10 alles from the site for release pathways (ground level and elevated releases)7 l
i Note l a IBID Page 1 l
Psv. 1 Pag 2 20 of 42 1.0 TECHNICAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS /CDMMENTS 1.3.2 Initial EOF Functions Supplement 1 to NUREC-0737 (TSC requirement a) is the TSC staff able to perform he EOF functions during an Aler* Nergency classification a ' '- '-
T: :;;- ;
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- untti the EOF is functional including the following?
e dose assessment protective action decisionmaking e
coordination of radiological and e
environmental assessment offsite.
e
Rev. l Page 21 of 42 i
2.0 OPERATIONAL SUPPORT CENTER (OSC)
CRITERIA ACCEPTANCE REOUIREMENTS/ COMMENTS 2.1 PHYSICAL FACILITIES 2.1.1 Deslon 10 CFR S0.47(b)(8),,'_ :^ ^ ;; Supplement 1 to NUREC-0737 (OSC functions and OSC requirement b) 2.1.1.1 OSC Location is there an OSC located onsite, separate from the Control Room and TSCF (Note Each assembly staging or other location used for support of the OSC must be evaluated.)
2.1.1.2 Alternate OSC Location (s)
Are provisions made to perform OSC functions elsewhere if the primary OSC becomes unInhabftable?
2.1.1.3 Size Layout and Environment a.
Are the size and layout of the OSC and alternate OSC adequate to provide an assembly area for all assigned support personnel and to facilitate performance of support functions and tasks?
b.
How was the OSC layout verifled as adequate?
c.
Are environmental conditions (temperature, lighting,noiselevel) acceptable for operations?
2.1.1.4 Display interface
.a.
la information adequately displayed or made available for use in theOSCforplanningpriortogl'spatch of teams into the p1 ant?
e.g. s e situational status site environmental data e
- radiological conditions e other.
Note 1: IBID page 1.
f Rev. I t"
Paga 24 ef 42 r/
3.0 EMERCENCY OPERATIONS FACILITY (EOF)
CRITERIA ACCEPTANCE REOUIREMENTS/ COMMENTS
- f. 1 PHYSICAL FACILITIES 3.1.1 Design 10 CFR 50.47(b)(8); Supplement 1 to NUREC-0737 (EOF requirements b,c,d,e and k and Table 1) 3.1.1.1 Size and Lavout a.
Is the size of the EOF adequate to accommodate and support Federal, State, local, and licensee predesignated personnel, equipment, and documentationtogerformtheintended functions?
e.g., :
work space for individual and e
working group functions work space for document and e
drawing review space for equipment and instru-e mentation use, maintenance and repair.
b.
Is the layout adequately designed to me. ~hd-suppo-t Ne-9, 3,f.f of e A '
S t:te, 2~:!,
...J
' Str:::
p.a.. I g n a + ;J.._.--- ~ t, :p -- 4 r
d~ r ::::t'-, 2nd_ h; provide unimpeded personnel traffic and information flow?
c.
How were the EOF size and layout verified as adequate?
3.1.1. 2 Location a.
Is the EOF located as described in Table 1 in the Supplement 1 to NUREC-07377 l
l b.
Which option (1 or 2) was chosen and does the EOF meet all the criteria I
in the specified option?
c.
Have adequate provisions been made for the NRC site team near the plant if theEOFisbeyond{0milesfromthe plant sito?
e.g.,
Note 1: IBID page 1.
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Rev. I Pag) 31 cf 42 3.0 ENERCENCY OPERATIONS FACILITY (EOF) (contd)
CRITERIA ACCEPTANCE REOUIREMENTS/CC4HENTS c.
Are the available 1.97 variables adequatetodetermineprimag containment integrity?
e.g.
- containment overstressing by high temperature and pressure hydrogen concentration in e
containment pressure changes in containment e
high radioactivity levels in offgas systems high radiation or radioactivity e
1evels in auxiliary buildings, other plant systems, and site factifties high offsite radiation or e
radioactivity levels.
d.
Are the available 1.97 variables adequate to determine the operability, ca city, and integrity of the Ifquid, solf, and gaseous rad waste systems?
by ombiner or offgas crn e
e na d arge of untreated i1 N
waste high radiation or radioactivity o
levels in auxiliary buildings and other site facilities high offsite radiation or e
radfoactivity levels.
l e.
Are the available 1.97 variables adequate to determine the extent of damageresultingfromargfuelingor fuel pool accident?
e.g. :
loss of water level in fuel pool or vessel e 'high radiation or radioactivity levels in the fuel handling area, fuel pool, containment, or other auxiliary l
plant areas, f.
Can an evaluation be conducted of both the existing and projected status of the core / containment and environs to support determination of proper Note la IBID page 1.
ii R:v. I Peg) 34 cf 42 EMERCENCY OPERATIONS FACILITY (EOF) (contd) 3.0 j
l ACCEPTANCE RE'0UIREMENTS/ COMMENTS CRITERIA 3.2.3.3 Transmission Between ERFs What methods are used for data a.
transafssion?
b.
Is data transmission edequate between the TSC, the Control Room, and the EOF 7 10 CFR 50.47(b)(4) (9); 10 CFR 50, Appendix E
$1V.E.2 and $1V
.3 Supplement 1 to @UREC-0737 3.2.4 Data Analysis (EOF requirement g)
Mh
,h 3.2.4.1 Reactor Technical Support Is the data analysis adequate to
- [
a.
support the EOF functions?
b.
Will the data analysis facilitate determinationofreactorstagus,past, present and projected?
e.g. :
Forecasting (trending) e containment pressure vs. time containment temperature vs.
time containment radiation or radioactivity levels vs. time containment H, concentration vs. time reactor coolant radioactivity vs. time offgas radioactivity levels vs.
time radiation or radioactivity levels in various plant systems locations or systems vs.
time.
Precalculated relationships of e
variables to accident conditf ons?
containment radiation levels to core conditions (with and without abnormal coolant system leakage) coolant radioactivity levels to core conditions H level in contairwent to containment f ailure Note la IDID pcge 1.
.