ML20211L047
| ML20211L047 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 08/31/1985 |
| From: | Stoffel J EG&G IDAHO, INC. |
| To: | NRC |
| Shared Package | |
| ML20211L029 | List: |
| References | |
| CON-FIN-A-6483, RTR-REGGD-01.097, RTR-REGGD-1.097 TAC-51069, NUDOCS 8612150285 | |
| Download: ML20211L047 (21) | |
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, ;. CONFORMANCE TO REGULATORY GUIDE 1.97 ARKANSAS NUCLEAR ONE, UNIT NO. 1
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'. j J. W. Stoffel' i
I Published August 1985
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- EG&G Idaho, Inc.
- Id.nho Falls, Idaho 83415 l * :? t u:
i' Prepared for the
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'i; U.S. Nuclear Regulatory Comission
- j. .j Washington, D.C. 20555 Hi Under DOE Contract No. DE-AC07-76ID01570 l .si FIN No. A6483 1
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ABSTRNCT
,] This EG&G Idaho, Inc., report reviews the submittals for Regulatory
- Guide 1.97 for Unit No. 1 of Arkansas Nuclear One and identifies areas of
..., nonconformance to the regulatory guide. Exceptions to Regulatory
,! Guide 1.97 are evaluated and those areas where sufficient basis for I'l acceptability is not provided are identified.
4 1 FOREWORD q p
' il
- This report is supplied as part of the " Program for Evaluating Licensee / Applicant Conformance to R.G.~1.97," being conducted for the U.S.
':4: Nuclear Regulatury Commission, Office 'of. Nuclear Reactor Regulation .
.] Division of Systems Integration by EG&G Idaho, Inc., NRC Licensing Support
'I Section.
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,l The U.S. Nuclear Regulatory Cannission funded the work under l authorization 20-19-10-11-3.
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-i CONTENTS 1
I ABSTRACT .....................'......................................... 11
.1 ii 7',,1 FOREWORD ..............................................................
a.
.j l. INTRODUCTION ..................................................... 1
- -i 2
j 2. REVIEW REQUIREMENTS ..............................................
i 3. EVALUATION ........................:.............................. 4 3.1 Adherence to Regulatory Guide 1.97 ......................... 4 3.2 Type A Variables ........................................... 4 L
3.3 Exceptions to Regulatory Guide 1.97 ........................ 5 f:
': ;. 4. CONCLUSIONS ...................................................... 16 I .
- 5. REFERENCES ....................................................... 17
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i CONFORMANCE TO REGULATORY GUIDE 1.97 i ARKANSAS NUCLEAR ONE, UNIT NO. 1 ij 4 . 1. INTRODUCTION' x].
li On December 17, 1982, Generic Letter No. 82-33 (Reference 1) was 1 issued by D. G. Eisenhut, Director of the Division of Licensing, Nuclear
'I Reactor Regulation, to all licensees of operating reactors, applicants for operating licenses and holders of construction permits. This letter M included additional clarification regarding Regulatory Guide 1.97, f Revision 2 (Reference 2), relating to the requirements for emergency
- i response capability. These requirements have been published as Supplement
.i No. I to NUREG-0737, "TMI Action Plan Requirements" (Reference 3).
s j Arkansas Power and Light Company, the licensee for Arkansas Nuclear i
One, Unit No. 1, provided a response to the Regulatory Guide 1.9,7 portion of the generic letter on June 25, 1984,,(Reference 4). Additional information was submitted on May 31, 1985 (Reference 5).
z This report provides an evaluation of those submittals.
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- 2. REVIEW REQUIREMENTS 1
- Section 6.2 of NUREG-0737, Supplement No.1, sets forth the documentation to be submitted in a report to the NRC describing how the Ml -
ll licensee complies with Regulatory Guide 1.97 as applied to emergency l:. response facilities. The submittal should include documentation that
,j provides the following information for each variable shown in the applicable table of Regulatory Guide 1.97.
'! 1. Instrument range t
?j 2. Environmental qual.ification
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' Seismic qualification
'd 3.
s I 4. Quality assurance
- 5. Redundance and sensor location"
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- 6. Power supply
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- 8. Schedule of installation or upgrade s ,.
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'-- Furthermore, the submittal should identify deviations from the regulatory l guide and provide supporting justification or alternatives.
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Subsequent to the issuance of the, generic letter, the NRC held
+] regional meetings in February and March 1983, to answer licensee and 4
gj applicant questions and concerns regarding the NRC policy on this subject.
,*t jd At these meetings, it was noted that the NRC review would only address exceptions taken to Regulatory Guide 1.97. Furthermore, where licensees or 1
3 Ej applicants explicitly state that instrument systems conform to the
.tj regulatory guide it was noted that no further staff review would be J
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- necessary. Therefore this report only addresses exceptions to Regulatory
, ,I , Guide 1.97. The following eva.luation is an audit of the licensee's ;:
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submittals based on the review policy described in the NRC regional ,
/ meetings.
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! 3. , EVALUATION s
j The licensee provided a response to NRC Generic Letter 82'-33 on
.; June 25, 1984. Additional information was submitted on May 31, 1985. This
~d evaluation is based on those submittals.
3.1 Adherence to Regulatory Guide 1.97
'N. !
u The licensee states that, based on the information presented in their submittal, Arkansas Nuclear One, Unit No.1, will conform with the
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recommendations of Regulatory Guide 1.97, Revision 3 (Reference 6), by the
- end of the next two refueling outages. Therefore, we conclude that the licensee has provided an explicit connitment on conformance to Regulatory
,. 4
! Guide 1.97. Exceptions to and deviations from the regulatory guide are -
d.j noted in Section 3.3.
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'" 3.2 Type A Variables i
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i Regulatory Guide 1.97 does not specifically identify Type A variables,
- i.e., those variables that provide information required to permit the
< a ol control room operator to take specific manually controlled safety actions.
The licensee classifies the following instrumentation as Type A:
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[.-j 1. Reactor coolant system (RCS) hot leg. water temperature l] :
. 2. RCS pressure d 3. Containment hydrogen concentration
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Hj 4. Steam generator level l
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- 5. Steam generator pressure
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- ,.I _ L--- 6. . Condensate storage tank level '. 2 _ '[j . 1, 7. Borated' water storage tank level. - _. 4 These variables meet the Category I recomendations consistent with the l'ji . requirements for Type A variables. %)
'l 3.3 Exceptions to Regulatory Guide 1.97 r4 e, The licensee identified deviations and exceptions from Regulatory 1 Guide 1.97. These are discussed in the following paragraphs.
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? x, . 3.3.1 RCS Soluble Boron Concentration q J . 1 The licensee has not provided indication with the recomended range q 5I for this variable. The licensee states that the letdown line for the 1
! boronometer is isolated follcwing an accident, and that the Post-Accident h Sampling System (PAS'S) for ANO-1 was designed to provide boron 1 concentration measureme,nt capability. : The licensee states that the range C and the design criteria for the PASS is consistent with guidance provided for Item II.B.3 of NUREG-0737. ,
The licensee deviates from Regulatory Guide 1.97 with respect to this
! variable. This deviation goes beyond the scope of this review and is being j addressed by the NRC as part of their review of NUREG-0737, Item II.B.3.
- l 3 3.2 RCS Cold leg Water Temperature
. The licensee has taken exception ,to the Category 1 and the range ,1 (50 to 700*F) recomendations of Regulatory Guide 1.97 for this variabie.
The licensee's instrumentation is Category 3 with a range of 50 to 650*F.
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L el The licensee states that there must be reactor coolant system (RCS) flow E, 4
.; through the steam generators for this variable to represent core
.,li conditions. Also, due to the proximity of the cold leg resistance v1
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li temperature detectors (RTD) to the high pressure injection (HPI) nozzles, 1 HPI flow may significantly affect the cold leg temperature indication d particularly in the absence of forced RCS flow. Incore temperature r
,j monitors are defined by the licensee as the key variable for this function as it provides a more direct indication of core cooling independent of whether or not coolant flow exists in the loops. The licensee states that )]
[.l RCS cold leg temperature serves as a backup variable and is accordingly I qualified to Category 3 recommendations. A.
.i Based on the licensee's justification, we find the existing Category 3 -
RCS cold leg water temperature instrumentation adequate. The licensee justified the range deviation by stating that the maximum saturation temperature *for the steam generators is 600*F (maximum pressure
': of 1200 psig, based on the highest main steam safety valve setting of e '4 1000 psig), an upper range'of 650*F for the RCS cold leg water' temperature instrumentation should provide adequate margin to account for any
- h t, reasonable variation in the maximum pres's'ure,of the steam generators.
-9 We find the existing temperature range adequate for the intended j purpose with the core exit temperature instrtaentation used as a diverse 'I method of determining RCS heat removal. Therefore, this is an acceptable i deviation from Regulatory Guide 1.97.
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u 3.3.3 Degrees of Subcooling ,e : t t,1 = :. ! Regulatory Guide 1.97 reconsnends instrumentation with a range of 200*F u ,y subcooling to 35'F superheat for this variable. The licensee has * [y instrumentation with a range of 0 to 2,00*F subcooled for this variable. The licensee states that during situations that may result in superheated p:'] conditions in the RCS, incore temperature and hot leg RTD's will be r;,
- . monitored against RCS pressure to determine the degree of superheat or subcooling, and that the safety parameter display system (SPUS) computers ,
I will plot the core exit temperature versus pressure on a grid with the fmj saturation curve so that the operator can tell at a glance the . L> 6 [: .s
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l -:' . +. thernrJdynamic status of the reactor coolant. Therefore, the licensee . ;,i , st,ates that a range of 0 to 200*F subcooled on the subcooled margin monitor
. , is adequate to meet the intent of Regulatory Guide 1.97. .
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The NRC is reviewing the acceptability of this variable as part of .I'k their review of NUREG-0737, Item II.F.2. d J
.' 3.3.4 Radiation Level in Circulating Primary Coolant I The l'icensee has chosen the following alternate instrumentation to 1 monitor the radiation level in the RCS during accident armi post-accident
. conditions:
lii . j o letdown line radiation monitor
.] o radiochemistry analyses a .
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, ; o post-accident sampling system
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'E Based on the alternate instrumentation supplied by the licensee, we
'I conclude that the instrumentation supplied for this variable is adequate and, therefore, acceptable.
d: 3.3.5 Containment Effluent Radioactivity--Noble Gases from Identified
!j
Release Points
".* 8 Effluent Radioactivity--Noble Gases from Buildings or Areas Where
'I Penetrations and Hatches are Located r
Condenser Air Removal System Exhaust _. _
.h Connon Plant Vent y -
i I;j In Reference 4 the licensee states that the power supply for the (j gaseous effluent radiation monitoring system (GERMS) does not comply with MJ l
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.l the strict definition of'a Category 2 power supply. In Reference 5 the r'I liceosee re-evaluated the existing power supplies. Based on this
-l evaluation and operational experience with the system to date, the licensee considers the GERMS power supply to be a high-reliability power source,
;., meeting the regulatory guide power supply requirements.
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f The GERMS auxiliary power is supplied from either ANO-1 or ANO-2 1
,. j normal AC through an automatic, power-seeking transfer switch. Battery
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. backup is provided for the system electronics. Based on this information,
\j we conclude that the power supply for this variable is highly reliable and,
.] therefore, acceptable.
$. I j 3.3.6 Accumulator Tank (Core Flood Tank) Level and Pressure d ,
l' The licensee has taken exception to supplying Category 2 y
. instrumentation for the core flood tank pressure. The licensee states that
.j the core flood tank pressure is the key variable for pre-accident Status for assuring that this passive safety system is prepared to serve its j function. The licensee states that the key variable necess,ary to determine i whether the core flood tanks have fulfilled their safety function is core flood tank level. Therefore, core flood tank pressure is a backup variable
-I and has accordingly been provided Category 3 instrumentation. 4 i j The accumulators are passive devices. Their discharge into the j reactor coolant system (RCS) is actuated solely by a decrease in RCS 6 i pressure. We find that the instrumentation provided for this variable is
- j.
- adequate to determine that the accumulators have discharged. Therefore,
'j the instrumentation for this variable is acceptable. p 3.3.7 Boric Acid Charging Flow [., 14 The licensee does not have instrumentation for this variable. Their ..j nuclear steam supply system design does not include the charging system as @q part of the emergency core cooling system. The high pressure injection and [j low pressure injection are the flow paths to the RCS that are monitored and . 'Q 8
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- this is done with Category 2 instrumentation. Therefore, we find that this N variable is not applicable at Arkansas Nuclear One, Unit No.1.
i 3.3.8 Quench Tank' Temperature
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f'.l . r'y; In Reference 4, the licensee indicates that quench tank temperature
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instrumentation does not exist. In Reference 5, the licensee states that
': this variable is monitored, however the existing range is 0 to 200*F. ,. Regulatory Guide 1.97 reconenends instrumentation for quench tank 'l temperature with a range of 50 to 750*F. The licensee has comitted to
['.] upgrade the instrumentation to measure a range of 0 to 350*F. e ',i gj The range covery the anticipated requirements for normal operation, l anticipated operational occurrences and accident conditions. This range s
. relates to the tank's rupture disk and the 100 psi tank design pressure j that limits the temperature of the tank contents to saturated steam
$ conditions under 350*F. Thu's, we find..this deviation from the regulatory guide acceptable. i
,i 4 3.3.9 Steam Generator Pressure The licensee has instrumentation for this variable that does not meet
- 1) the range recomended by the regulatory guide (0 to 20 percent above the
;j lowest safety valve setting). '-
<}. 8 ls j [: Control room indication of main steam pressure at the turbine as well as at the steam generators is provided. The steam generator outlet
'. 'I .ti pressures are measured over the range .from 0 to 1200 psig. Since the
.a l,] lowest safety valve setting is 1050 psig, the range should be to 1275 psig. id A 7 'j. The main steamlines are provided with relief valves, atmospheric dump Pi . valves, and condenser dump valves to prevent overpressurization of the , :j. steamlines as well as for pressure control capability. The licensee has [- fj approximately 40 percent excess steam relief capacity with all safety u1 1. 9 H m
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a valves and condenser' dump valves operable. Technical specifications
.! requjre 14 of the 16 main steam safety valves to be operable during power
- operation. The licensee states that fourteen safety valves will relieve ,
.j approximately 110 percent of the rated steam flow. Combined with the ,
j condenser dump valves, these 14 safety valves will provide a total of
. approximately 25 percent excess steam relief capacity.
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;j Based on the highest safety valve setting being 1100 psig, and an j excess r.elief steam capacity of approximately 25 percent being maintained 1 when as many as two safety valves are inoperable, we find the existing
?: range to be adequate to monitor the steam generator pressure during all
.j accide.it and post-accident conditions. Therefore, this is an acceptable deviation from Regulatory Guide 1.97.
f i 3.3.10 Containment Atmosphere Temperature s { The licensee deviates from the Category 2 and range (40 to 400*F) recommendations of Regulatory Guide 1.97r The instrumeritation provided is
;l Category 3 with a range of 0 to 300*F. The lice'nsee states that the
. reactor building atmosphere temperature is not a key variable for accident 1 monitoring; that the key variable for reactor building monitoring is reactor building pressure which is measured by Category 1 instrumentation; that the reactor building atmosphere temperature is a backup variable for d reactor building accident monitoring and as such is measured by Category 3 j instrumentation with a range of 0 to 300*F. This range is justified by the 8
!I licensee based on a safety analysis which demonstrates that the worst case I peak reactor building temperature would be 286.5'F [ Final Safety Analysis' ..j Report (FSAR), Table 6-9]. l- d 'n . The licensee indicates that the maximum containment temperature will ,' be less than 287'F during all accident and post-accident conditions. L'd, Therefore, the range of 0 to 300*F is acceptable. Ed L ') 2 p-
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3.3.11 Containment Sump Water Temperature j: . 1 . [ The licensee ~has not provided instrumentation for this variable. In
.+
"! Reference 4 the licensee states that the FSAR accident analysis assumes saturated conditions for sump water during sump recirculation, and that
]- because of this conservative assumption, adequate net positive suction head
.,j 1 exists for reactor building spray and safety injection pumps at all
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..j feasible sump water temperatures. In Reference 5, the licensee indicates l 'that the key variables for monitoring operation of the containment cooling systems are rtactor building pressure and spray flow. Backup 1 instrumentation is provided by reactor building temperature, cooling fan circuit breaker status and flow rate of cooling water to .the cooling 1 ,
units. The licensee states that sump water temperature is not recorded or i j indicated because it,is not required to mitigate the consequences of a A design basis accident. m 1 t The temperature of the sump water is useful to the operaton in determining the amount of containment heat removed during. recirculation. ,
'I Therefore, a means of determining the: containment sump water temperature should be provided by the licensee.
3.3.12 Makeup Flow-In i
.. Letdown Flow-Out
..- i
! Volume Control Tank Level I
The licensee takes exception to the Category 2 recommendations of the
'j regulatory guide for these variables. , Category 3 instrumentation is j provided by the licensee. The licensee states that for accidents in which M harsh environments are a result, the system containing this instrumentation 9n (letdown and makeup portion of the makeup and purification system) is not ,4 required. Letdown is automatically isolated upon an engineered safety G.
7:! feature (ESF) system actuation. di A 11 3 N"' [ T reg
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- l. j 3.3.13 Component Cooling Water Temoerature to ESF System
- l
.,. 5 The licensee has not provided a readout in the control room for this a
variable. The licensee states that the service water system is used for !' .d v . ') this variable. H
..i The inlet temperature of the service water, by design, is based on a
'I ' maximum temperature of 129.5'F frcn the emergency cooling pond. The average temperature of the pond (June through September) is 85'F. Furthermore, there is no control over the temperature of the service water. The justification submitted by the licensee for this deviation is
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! adequate. Therefore, this is an acceptable deviation from Regulatory 2
- N Guide 1.97.
3.3.14 Component Cooling Water Flow to 'ESF System
.h Regulatory Guide 1.97 recomends a range of 0 to 110 percent design flow for this variable. The licensee uses pump pressure and valve position
to nionitor the operation of this system. The licensee states that the
; design flow to various ESF components varies from 6 gpm for the reactor
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building spray pump bearing coolers to 3000 gp'n for the decay heat removal \f coolers. The licensee states that due to this wide range of design flows to ESF components, total loop flows would not be indicative of overall
,l
.j system performance. The licensee states proper system operation is shown
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by the correct service water header pr, essure and by knowing that the remote
.[,j actuated valves supplying service water to ESF components are in their ;O proper positions. Service water header pressure and remote actuated valve ;l position indications are available in the control room and meet Category 2 g 1 requirements.
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i Based on the justification provided by the licensee, we find this j instrumentation acceptable. .
.j 3.3.15 Radioactive Gas Holdup Tank Pressure
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"j The licensee has not provided control room instrumeritation for this 4 variable. The licensee states that in the event of an accident which results in significant failed fuel or significant radioactive gas release, the manual transfer of radioactive gases to the radioactive gas holdup j tanks would not be attempted. There ar'e no automatic transfer operations I involving the radioactive gas holdup tanks. The licensee states that the
, monitoring of the radioactive gas holdup tanks during post-accident i conditions is not necessary since these tanks are not utilized for accident
.} mitigation. ,
1 - Since these tanks are not used for accident mitigation at this l station, this exception from Regulatory Guide 1.97 is acceptabla 3.3.16 Radiation Exposure Rate i
, The licensee deviates from the instrument range recommended by 4
) Regulatory Guide 1.97 (10-I to 10 R/hr) for this variable. ANO-1 currently has an area radiation monitoring system consisting of 20 area (l
monitors: four with a range of 10'2 to 10 3R/hr and 16 with a range of 10-4 to 10I R/hr. The licensee states thatthese ranges are based on
] a background readings in the areas in which they are located. Should :
, personnel entry be required in areas where these monitors have gone off h scale or indicate a high radiation, the licensee states that a health
. .; physics escort would accompany personnel into these areas using portable
% instrumentation to assess radiation levels. The high range for portable instrumentation at ANO is 103 R/hr. The licensee does not anticipate, p1 even under emergency conditions, sending personnel into radiation fields of 4 this magnitude.
'4 .
n From a radiological standpoint, if the radiation levels reach or 1 exceed the upper limit of the range (103 R/hr), personnel would not be
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.I l 3.3.17 Plant and Environ Radiation
- l Regulatory Guide 1.97 recommends portable instrumentation with a range
~4rads /hr, beta radiation and j of 10-3 to 104 R/hr, photons; and 10-3 to l0 low-energ.y photons. The licensee does not comply with the recommended range for this instrumentation and stated that the existing portable instrumentation can detect gama dose rates from 10-3 to 103 R/hr and
< beta dose rates from 10-3 to 50 rad /hr. The licensee does not anticipate j encouncering radiation fields greater than those which can be measured by
~
their current equip. ment except under severe accident conditions. Even
]
under accident conditions, the licensee does not anticipate sending - 3 individuals into greater than 10 R/hr fields. i This instrumentation is portable and would not be used to assess 1evels of radiation greater than the range provided by the licensee.
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; Therefore, this is an acceptable deviation from Regulatory Guide 1.97.
3.3.18 Plant and Environs Radioactivity The licensee has not provided portable instrumentation for isotopic l analysis as recommended by Regulatory Guide 1."97 and stated that gama -
, spectroscopy can be performed using equipment in the health physics department and the radiochemistry department at the station, and in the j Technical Analysis Laboratory in Little Rock. In addition, the licensee ' 'i has an NO-60 spectrometer in the Emerg.ency Offsite Facility which can be lf used for less defined analysis. The licensee states that it is not
] appropriate for this instrumentation to be portable due to rough handling
} it would encounter in the field and the limited amount of time field teams ;;j have to assess the release.
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j:t \ i The existing laboratory equipment available at'this station is L adequate to provide isotopic analysis and a timely assessment of radioactive releases. Therefo're, this is an' acceptable deviation from j Regulatory Guide 1.97. w I'
! 3.3.19 Estimatien of Atmospheric Stability
.c}
.' Regulatory Guide 1.97 recomends instrumentation for this variable with a range of -5 to 10'C or an analogous range for alternative stability estimates. The licensee has instrumentation with a range of -3 to 5'C
-C temperature differences and 0 to 40' wind direction sigma. The licensee states that atmospheric stability is derived from the temperature
- differential indicated between 34 and 180 feet; that the -3 to 5'C f temperature range covers the seven Pasquill stability classes vs. Delta-T i as derived from Reguiatory Guide 1.23 as specified in the ANO-1 FSAR. In addition to temperature differential, atmospheric stability can also be calculated for all seven classes using wind direction sigma. ,
[ '. Table 1 of Regulatory Guide 1.23 fReference 7) provides seven
., atmospheric stability classifications based on the difference in temperature per 100 meters elevation change. These classifications range i
from extremely unstable to extremely stable. Any temperature difference
.a beyond -2 or +4'C does nothing to the stability classification. The licensee's instrumentation includes this range. Therefore, we find this an l acceptable deviation from Regulatory Guide 1.97.
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- 4. CONCLUSIONS
' Based on our review, we find that the licensee either conforms to or is , justified in deviating from Regulatory Guide 1.97, with the following
.,n
., exception:
- v.;
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.,} 1. Containment sump water temperature--the licensee should provide
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instrumentation to monitor this variable (Section 3.3.11). 4 e y I
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- 5. REFERENCES
.; - 1. . NRC letter, D. G. Eisenhut to All Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction Permits, " Supplement No. I to NUREG-0737--Requirements for Emergency 3
, Response Capability (Generic Letter No. 82-33)," December 17, 1982.
- 2. Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, j Regulatory Guide 1.97, Revision 2, U.S. Nuclear Regulatory Commission (NRC), Office of Standards Development, December 1980.
- 3. Clarification of TMI Action Plan Requirements, Requirements for Emergency Response Capability, NUREG-0737, Supplement No.1, NRC, Office of Nuclear Reactor Regulation, January 1983.
- 4. Arkansas Power and Light Company letter, John R. Marshall to Darrel G. Eisenhut, NRC, "NUREG-0737 Supplement 1, Regulatory Guide 1.97," June 25, 1984. -
- 5. Arkansas Power and Light Company letter, J. Ted Enos to Director of Nuclear Reactor Regulation, " Response to Position Document Open }
Items," May 31, 1985, ICAN058507
- 6. Instrumentation for Light-Wat. .,coled Nuclear Power Plants to Assess
', Plant and Environs Conditions During and Following an Accident, Regulatory Guide 1.97, Revision 3, U.S. Nuclear Regulatory Connission (NRC), Office of Nuclear Regulatory Research, May 1983.
- 7. Onsite Meteorological Programs, Regulatory Guide 1.23 (Safety Guide 23), NRC, February 17, 1972, or Meteorological Programs in Support of Nuclear Power Plants, Proposed Revision 1 to Regulatory Guide 1.23, NRC, Office of Standards Development, September 1980.
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7.'S'" ^ EGG-EA-6847
.o.e rm......BIBUOGRAPHIC DATA SHEET k . e .rauCre. ..
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i Conformance to Regulatory Guide 1.97, Arkansas Nuclear One, . ..r. .. .r co u n o O Unit No. 1 . j , s .or ,. August 1985
! e o.r...po e'.3W.o 1
e.o8.r '8'a i J. W. Stoffel August 1985
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.; , ,, , o a r.o . .s.
EG&G Idaho, Inc. * " aa a"*"' "*""" Idaho Falls, ID 83415 }L A6483
". "a a* a**oa' 2 o uro . .. .~ e. c Division of Systems Integrition Technical Evaluation Report
" Office of Nuclear Reactor Regulation , , , , , , , , , , , _ _
U.S. Nuclear Regulatory Comission - Washington, DC 20555 _[ _
,3 . . .o,.u . r... .or u 5
i ., . ar..cr ,u. , . I This EG&G Idaho, Inc. report reviews the submittals for the Arkansas
.. Nuclear One, Unit No. 1, and identifies areas of nonconformance to Regulatory
,, Guide 1.97. Exceptions to these guidelines are evaluated and those areas where sufficient basis for acceptability is not provided are identified.
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