ML20214E713
| ML20214E713 | |
| Person / Time | |
|---|---|
| Site: | Vogtle, 05000000 |
| Issue date: | 04/30/1986 |
| From: | Vanderbeek R EG&G IDAHO, INC. |
| To: | NRC |
| Shared Package | |
| ML20213F414 | List:
|
| References | |
| CON-FIN-A-6493, FOIA-87-213, RTR-REGGD-01.097, RTR-REGGD-1.097 EGG-EA-6806, NUDOCS 8604210357 | |
| Download: ML20214E713 (24) | |
Text
{{#Wiki_filter:- _ _ _ . __ I
. EGG-EA-6806 l
1 CONFORMANCE TO REGULATORY GUIDE 1.9'7 . V0GTLE ELECTRIC GENERATING PLANT, UNIT NOS. 1 AND 2 j
.- 1 R. VanderBeek
' Published April 1986'.
'b EG&G Idaho, Inc.
Idaho falls, Idaho 83415 J Prepared for the U.S. Nuclear Regulatory Connission . Washington. 0.C. 20555 Under DOE Contract No. DE-AC07-761001570-FIN No.'A6493 e-I f
.,n.n* **'"'' ."'w.,
(p}%9(o(ph@dNkY - . _
4 I ABSTRACT
.l ,
This EG&G Idaho,'Inc., report reviews the submittals for Regulatory o Guide 1.97 for Unit Nos.1 and 2 of the Vogtle Electric Generating Plant j and identifies areas of nonconformance to the regulatory guide. Exceptions ' to Regulatory Guide 1.97 are evaluated. ! i l I ! 1 1 i
~j I
Docket Nos. 50-424 and'50-425 11 i e F f t _ _ _ .. . . . . .
r j
-1 i
)
i 1 I
-i
)
i 1 I FOREWORD i i This. report is supplied as part of the ' Program for Evaluating Licensee / Applicant Conformance to RG 1.97," being conducted for' the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of PWR Licensing-A, by EG&G Idaho, Inc., NRR.and I&E Support -
)
Branch. j
- i i
The U.S. Nuclear Regulatory Commission funded the work under: I authorization B&R 20-19-40-41-3. i W
.. i Docket Nos. 50-424 and 50-425 i
I
. 1 CONTENTS
?
i ABSTRACT ..............................................................- 11 FOREWORD .............................................................. iii i l
- 1. INTR 000CTI0h ..................................................... 1
- 2. REVIEW REQUIREMENTS .............................................. 2 3.- EVALUATION ....................................................... 4 i
l L3.1 Adherence.to Regulatory Guide 1.97 ......................... 4 -4 3.2 Type A Variables ........................................... 4 j 4
)
3.3 Exc'eptions to Regulatory Guide 1.97 ........................ '5 lt
- 4. CONCLUSIONS ...................................................... 18
.]
i
- 5. REFERENCES ....................................................... 19
-l 1
l l i I l k i l I i
. l l
i 1
- J l
+
1 iv-1 i
, 1 1
i CONFORMANCE TO REGULATORY GUIDE 1.97
~
V0GTLE ELECTRIC GENERATING PLANT. UNIT NOS. 1 AND 2 i l
- l
- 1. INTRODUCTION On December 17, 1982. Generic Letter No. 82-33 (Reference 1) was issued by D. G. Eisenhut, Director of the. Division of Licensing, Nuclear Reactor Regulation, to all licensees of operating reactors, applicants for operating licenses and holders of construction permits. This letter included additional clarification regarding Regulatory Guide 1.97, Revision 2 (Reference 2), relating to the requirements for emergency response capability. These requirements have been published as Supplement No. I to NUREG-0737, "TMI Action Plan Requirements" (Reference 3).
1 Georgia Power Company, the applicant for the Vogtle Electric l Generating Plant (VEGP), provided a response to the generic letter on
, April 14, 1983 (Reference 4). This submittal refers to the Vogtle final l Safety Analysis Report (fSAR, Reference 5) for a review of the instrumentation provided for conformance to Regulatory Guide 1.97. On March 15, 1985 (Reference 6), on June 20,1985, (Reference 7), on l September 6, 1985 (Reference 8) and on January 21, 1986 (Reference 9), the ;
i applicant provided additional information. ' This report provides an evaluation of these submittals 1 i
- 2. REVIEW REQUIREMENTS )
j Section 6.2 of NUREG-0737, Supplement No'. 1, sets forth the
- i documentation to be submitted in a report to the NRC describing how the applicant complies with Regulatory Guide 1.97 as applied to emergency response facilities. The submittal should include documentation that provides the following information for each variable shown in the applicable table of Regulatory Guide 1.97.
- 1. Instrument range
- 2. Environmental qualification
- 3. Selsmic qualification 4 Quality assurance i
- 5. Redundance and sensor location '
l l
- 6. Power supply
)
1
- 7. Location of display i
- 8. Schedule of installation or upgrade ;
The submittal should identify deviations from the regulatory guide and I provide supporting justification or alternatives. Subsequent to the issuance of the generic letter, the NRC held regional meetings in February.and March 1983 to answer licensee and applicant questions and concerns regarding the NRC policy on this subject. At these meetings, it was noted that the NRC review would only address exceptions; taken to Regulatory Guide 1.97. Where licensees or' applicants , explicitly state that instrument systems conform to the regulatory guide it was noted that no further staff review would be necessary. Therefore, this-l 2-
. 1.
i report.'only addresses exceptions-to Regulatory Guide1.97. The following'
. evaluation.is an audit of the applicant's submittals based on the review.
policy described..in the NRC regional meetings. e l 1 1 l
,.i e
a I j
.'y- #
i
~3 t .
- t
+
9
1 1 1
- 3. EVALUATION a
In the applicant's response to NRC generic letter 82-33, Section 7.5
- of the FSAR is identified as containing (a) the description of the .
post-accident monitoring system (PAMS), (b) tables that identify the "] monitored parameters, and (c) identification of compliance to or deviations ] from Regulatory Guide 1.97 along with the supporting justification or alternatives. This evaluation is based on the information provided in d 1 Section 7.5 of the FSAR, and the March 15, 1985, the . lune 20, 1985, the ; 1 September 6, 1985 and the .lanuary 21, 1986 submittals.
-]
l J 3.1 Adherence to Reaulatory Guide 1.97
'j Within Table-7.5.2-1 of the FSAR, the applicant'has identified where the post-accident monitoring instrumentation conforms to Regulatory Guide 1.97, Revision 2, and where deviations exist. Therefore, we conclude that the applicant has provided an explicit connitment on conformance to Regulatory Guide 1.97 prior to startup. Exceptions to and deviations from the regulatory guide are noted in Section 3.3.
1 3.2 Tvoe A Variables Regulatory Guide 1.97 does not specifically identify Type A variables, i.e., those variables that provide the information required to permit the control room operator to take specific manually controlled safety actions. The applicant classifies the following instrumentation as Type A:
- 1. Reactor coolant system pressure (wide range)
- 2. Reactor coolant system wide range hot' leg temperature
- 3. Reactor coolant system wide range cold leg temperature
- 4. Wide range steam generator water level 4
- 5. Narrow. range steam generator water level -)
'I '
- 6. Pressurizer level
~
- 7. Containment pressure
- 8. Steamline pressure '
- 9. Refueling' water stor. age tank level
- 10. Containment water level--narrow range I
- 11. Containment water level--wide range 1
- 12. Condensate storage tank level q
'13. Auxiliary feedwater flow 1 . 1
- 14. Containment radiation level--wide range' <
- 15. Containment radiation level- narrow range '
i j
-16. Steamline radiation monitor !
I
- 17. Core exit temperature ]
- 18. Degrees of subcooling These variables meet the. Category 1 recommendations consistent.with the.
requirements for Type A variables. 3.3 Exceptions to'Reaulatory Guide 1.97 The applicant identified deviations and exceptions from Regulatory-Guide 1.97. These are discussed in the'following paragraphs.
.l 5.
3.3.1' Reactor Vessel Water Level In the initial submittal, the applicant took exception to Regulatory . Guide 1.97 for the reactor vessel water level variable. Category 2 , instrumentation was indicated instead of the recommended Category 1 - instrumentation.
. i In the June 20, 1985 submittal, the applicant states that Table 7.5.2-1 of the FSAR will be revised in Amendment 17 to reflect that the reactor vessel water level system conforms to Regulatory Guide 1.97 Rev. 2, i.e., meeting Category 1 criteria. We conclude that the clarification provided by the applicant in the March 15, 1985 and June 20, 1985 submittals shows that this instrumentation is acceptable.
3.3.2 Containment Isolation Valve Status In the initial submittal, the applicant indicates that the instrumentation for this variable is Category 2 instead of the recommended.
- Category 1 instrumentation.
The applicant provided additional information within the . March 15, 1985 submittal stating that containment isolation valve status conforms to Regulatory Guide 1.97, Rev. 2, Category 1 instrumentation criteria, except for a single indication per valve. From the information _j provided, we find that the applicant deviates from a strict interpretation of the Category 1 redundancy recommendation. Only.the active valves have position indication (i.e.. check valves have no position indication). Since redundant isolation valves are provided, we find that redundant indication per valve is not intended by the regulatory guide. Position indication of check valves is specifically excluded by Table 2 of Regulatory Guide 1.97. Therefore, we find that the instrumentation for this variable is acceptable. 6
r 3.3.3 Pressurizer Heater Status l 1 Within the initial submittal, the applicant does not identify . instrumentation for monitoring current to the pressurizer heaters. The k applicant states that indication of the pressurizer heater breaker position 1 is adequate indication to the operator that the pressurizer heaters are j operable. 4
'In Reference 9 the applicant commits to provide the recomended current monitoring instrumentation to those heater banks that are capable ;
of being supplied emergency power from the diesel generators. We find this instrumentation acceptable. 3.3.4 Accumulator Tank Level and Pressure Regulatory Guide 1.97 reconnends Category 2 instrumentation for-this ! variable with ranges of 10 to 90 percent volume and 0 to 750 psig. The , applicant has identified Category 3 instrumentation for level with a range' l of 59 to 66 percent volume, and Category 2 instrumentation for pressure' with a range of 0 to 700 psig. I The applicant has identified the accumulator pressure as the key f variable to monitor the operation of'the accumulators. The' range is justified because two check valves in series on the discharge line prevents 3 pressures from exceeding the manually controlled pressure. The nitrogen cover gas is manually kept between 564 and 637 psig. l As a backup variable, the category and the range of.the level instrumentation is adequate, as the regulatory guide accepts Category 3 backup variables. The accumulators are passive devices. Their discharge.into the RCS is l actuated solely by a decrease in RCS pressure. We. find that the 7 i
sg . instrumentation. supplied for this variable is adequate to determine that the accumulators have discharged. Therefore, the ranges ofithis l instrumentation are acceptable for this variable. 3.3.5 Containment Atmosphere Temperature Within:the' initial submittal, th'e applicant' specifies Category 3 instrumentation;instead of Category 2. The applicant states that this is based on the fact that the plant emergency response guidelines do not l require _the. operator to take action that would result.in' adverse. consequences.if the containment-temperature were indicating an erroneous j
~value.-
The applicant's justification was expanded in the March 15, 1985 i submittal. The' applicant states that the key variables ~for monttoring - . containment cooling are containment spray flow, containment water level. I containment spray valve status, containment fan cooler damper. position, , _ containment fan' cooler breaker position, and containment pressure. Because the operator will not use the containment atmosphere. temperature.as a. key' t j variable, and the applicant has identified alternate key variables to: j monitor containment cooling, we find the deviation from Category 2.to l Category 3 for this variable acceptable, i 3.3.6 Containment' Sumo Water Temperature
~
Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. The' applicant has instrumentation for- this variable..'The sump , temperature instruments are Category 2 except.for the electrica1' connectors-inside containment that are not qualified for a steam environment. The' l residual heat removal (RHR) heat' exchanger inlet temperature (essentially 4 the same temperature as the sump) instruments'are Category 2 except,for! ., j high radiation levels. 'The appitcant states that further qualification of-l i i
'8 l- i
these instruments will not increase the safety of the station or increase
-the ability to mitigate the consequences of an accident or'to bring the unit to a safe shutdown.
~
The applicant points out the following justifications.
- 1. There are no operator actions that are dependent on sump '
temperature. 2.- Recirculation flow is. initiated based on the refueling water.' storage tank level.
- 3. The RHR system is the only system that can reduce'the. sump water temperature. 'The decision to.use this' system is. based on reactor coolant' system (RCS) temperature, inadequate cooldown rate or RCS pressure.
1
- 4. The RHR pumps will have' adequate net positive' suction head regardless of the sump temperature.
'5. Should a quantitative measure of heat removal be desired -the supplied instrumentation can be used.
The applicant does not have instrumentation or alternat'- e instrumentation for this variable that is fully qualified to:the Category 2 requirements. Thus, in a post-accident situation, a quantitative measure; of the heat removal by way of the containment sump would notL.necessarily be-available.- Because of this, the applicant's justification is not acceptable. The applicant should therefore provide instrumentation for:this vartable that is. environmentally qualifled in accordance with:the-- provisions of 10 CFR 50.49 and Regulatory Guide-1.97. L -
r i j i 3.3.7 Heat Removal by the Containment Fan Heat Removal System ; l The applicant indicates that this variable is not required. The l applicant states that (a) the containment spray flow indication (b) the containment spray system valve status indication, (c) the containment pressure indication. (d) the containment water level indication (e) the containment spray pump status indication, (f) the containment fan cooler
- l damper position indication, and (g) the containment fan cooler breaker.
position, all provide indication to allow the operator to determine operability. These variables are identified by the applicant as Category 2. Based on the above diversity, we find the alternate j instrumentation acceptable. 3.3.8 Accident Samplina (primary Coolant. Containment Air and Sump) i I Within the initial submittal the applicant notes an exception for this variable. The applicant did not provide the informab on required by " Section 6.2 of Supplement No. 1 to NUREG-0737.
. f In the applicant's June 20, 1985 submittal, the information required by Section 6.2 of Supplement No. 1 to NUREG-0737 is'provided. The applicant states that this information will be included in Table 7.5.2-1 of the FSAR by Amendment 17. Review of the information provided by the applicant on the accident sampling capability indicates that the applicant conforms with Regulatory Guide 1.97 for this variable.
J 3.3.9 Boric Acid Charaina Flow The applicant' indicates that this variable is not required. The justification is that (a) the refueling water storage tank level indication. (b) the high-head safety injection flow indication, (c) the . low-head safety injection flow indication, (d) the containment water level indication, and (e) the emergency core cooling system (ECCS) valve ~ status, monitor the performance of the ECCS. The normal charging flow'and reactor i coolant system (RCS) sampling is used to demonstrate that the RCS is being 10
o I borated. The licensee' states that the boric acid charging flow is not a safety injection system',' nor is it used for emergency boration. Therefore, we find that this var.iable is not applicable at the Vogtle Station. 3.3.10' Dearees of Subcoolino Within the initial submittal, the applicant has identified this as a Type A variable. 'As such Table 2 of Regulatory Guide 1.97 requires Category 1. instrumentation.- The applicant is supplying Category 2 .- i instrumentation. The applicant's March 15, 1985. submittal stated that the category will be revised in. Amendment 16 and that the instrumentation is:in . , compliance with the Category 1 recommendations of Regulatory Guide 1.97. 3.3.11 Condenser Air Elector Radiation In the initial submittal, the applicant identified this as a Type A variable. As such, Table 2 of Regulatory Guide l'.97 requires-Category 1
. instrumentation. The applicant is supplying Category 2. instrumentation.
The applicant states, his' March 15, 1985 submittal, that the! condenser air.
- ejector radiation instrumentation is located in a non-seismically designed building and is nonseismically qualified equipment.1 Since it-servesLa- -j backup function to the steam. generator level monitors and the main; steamline radiation monitor, and is not a Type A variable, the variable.is-
~
revised to a Type E, Category 3. The revision and justification are acceptable. l 3.3.12 Areas of Additional Noncompliance l
- a. Reactor Coolant System Soluble Boron Concentration.
In the applicant's. initial submittal, this variablelwas not. )
- addressed or indicated in. Table 7.5.2-1, nor.was the information.
required by Section 6.2 of NUREG-0737. Supplement No.1 4 provided. In the March 15, 1985 submitt'l, a the applicant states, 11 ! 4
p that the reactor coolant system soluble boron concentration is part of the post-accident sampling system, ~ rather than continuous monitoring instrumentation. . The appitcant deviates from Regulatory Guide 1.97 with respect to . this variable. This deviation goes beyond the scope of this review and is being addressed by the NRC as part of their review , of NUREG-0737, Item 11.8.3.
- b. Radiation Level in Circulating Coolant.
In the applicant's initial submittal, this variable was not addressed nor indicated in Table 7.5.2-1, nor was the information required by Section 6.2 of NUREG-0737, Supplement No.1 provided. In the March 15 1985 submittal, the applicant states that this-variable is part of the post-accident sampling system (PASS). Based on the alternate instrumentation associated with the PASS, - we conclude that the instrumentation, supplied for this variable is adequate and, therefore, acceptable.
- I
- c. Analysis of Primary Coolant (Gamma Spectrum)
.i In the applicant's initial submittal..this variable was not addressed nor indicated in Table 7.5.2-1, nor was the information required by Section 6.2 of NUREG-0737 Supplement No. 1 provided. In the March'15, 1985 submittal, the' applicant states that the variable is part of the post-accident sampling system.
The applicant deviates from Regulatory Guide 1.97 with respect to post-accident sampling capability. 'This deviation goes beyond- - the scope of this review and is being addressed by the NRC as ' ' ' part of their review of NUREG-0737 Item B.II.3.
. I i
12-
. -, : - 1
- d. Containment Effluent Radioactivity
. In the applicant's initial submittal, this variable was not addressed nor indicated in Table 7.5.2-1, nor was the information
]
required by Section 6.2 of NUREG-0737, Supplement' No.1 provided. In the March 15, 1985 submittal, the applicant states that this ~ - 1 variable is monitored by the plant vent monitor which receives.. discharge from the containment purge system, auxiliary building. . .- control building and the fuel handling building. The required-- -4 information is provided in the June 20, 1985 submittal and the instrumentation is acceptable.
- e. Accumulator Isolation Valve Status I I
In the applicant's initial submittal, this variable was not addressed nor indicated in Table 7.5.2-1,-nor was the information. ,
. required by'Section 6.2 of NUREG-0737, Supplement No'. 1 provided.
.j
-i In the March 15, 1985 submittal, the applicant states that this variable.is part of the ECCS valve status. The required information is provided in the June 20,'1985 submittal:and the-instrumentation is acceptable.
- f. Reactor Coolant Pump Status In the applicant's initial submittal, this variable was~not addressed nor indicated,in Table 7.5.2-1,'nor was the-information required by Section 6.2 of NUREG-0737, Supplement _No.'-1 provided.
In the March 15,.1985 submittal, the. applicant states that'the.- reactor coolant pump status would be monitored by the; reactor.- , s O coolant: pump circuit breaker position'andithe'information-
~!
pertainingLto the instrumentation will be added'to/ Table 7.5.2-1 in a subsequent amendment. It.will be designated asia Type.D
.q
- 13. ,
( 1
. ,: c
., 1 ,
- ' ' - ' ' ' "' ' ' ' 'I* ' '
1 L j Ca'tegory 2 variable. -The applicant'did not' provide justification
~
for providing pump breaker ' position instead of the recensnended j
' ~
motor current indication in that. submittal. The applicant provides additional ~information_in his . lune 20, 1985 submittal. The justification for.providing pump breaker position-is that the emergency operating procedures do not. l utilize motor' current to assess the operation of.the reactor coolant pumps but pump' circuit breaker positions are..LReactor- 'I coolant pump motor current is available in the control room by- h i means of the emergency response facility computer CRT display. ~1 This' instrumentation conforms to Category 3 criteria. Based on the current display' capability in the control; room, we find the deviation acceptable,
- g. Pressurizer Relief Tank Level ,.
In the applicant's' initial' submittal this' variable was'not)
~
addressed or indicated in Table 7.5.2-1, nor was the information-required by Section 6.2 of NUREG-0737 Supplement No. 1.provided. The required information for this variable was provided in thel i appitcant's March 15, 1985 submittal and the instrumentation-is- ' acceptable. e
- h. Pressurizer Reitef Tank Temperature
- In the applicant's initial submittal,zthis variable was not' addressed nor indicated in Table-7.5.2-1,.nor:was.the information required by Section :6.2 of.. NUREG-0737, Supplement No.1 provided.
~
.In the March 15, 1985 submittal,lthe a'pplicant'provided t'he- ,
1 required information. However.the. range.lforfthepressurizer~ ;
' relief tank? temperature is 50' to.300*FCinstead-~of'the.: H '
- 1 recommended,50' to 750*F. 1 H
1n p 14 q a
. ; f ~y
' i t
, -{ ,J _
g
, - q a
=
The applicant stated, in his January 21 1986 submittal,-that the
~
. pressurizer relief tank (PRT) temperature range would be
.,- increased to 50 to 350*F, and.that this. range is adequate't'o monitor any expected conditions in the tank. This;rangerel'ates to the-tank's rupture disk that relieves pressures.in exces's'of 91'psig. This pressure relief limits the temperature of the. tank' contents to saturated steam conditions under 321*F. Thus, we find this deviation from the regulatory guide acceptable.
- i. High Level Radioactive' Liquid Tank L'evel In the applicant's initial submittal, t'is h variable was not' addressed nor indicated in Table 7.5.2-1, nor was the information' ]
required by-Section 6.2 of.NUREG-0737 Supplement No'. 1 provided.- H The required information for the variable was provided in the applicant's June 20, 1985 submittal and the. instrumentation is
]
. accer. table. !
W I
'1
- j. Radioactive Gas Holdup Tank Pressure j l
In the applicant's initial submittal, this-variabletwas not addressed nor indicated in Table 7.5.2-1, nor was'the-information- ; i required by Section 6.2 of.NUREG-0737, Supplement No. 1'provided.
. 1 The required information for this variable was provided in the 1 applicant's June 20, 1985 submittal. However,-the applicant ' !
takes exception to the range. The range is 0 to'100, percent -j design pressure instead of the recommended 0.to 150' percent of. i design pressure. The applicant states that,the design pressure
~
q of each of-the waste gas decay tanks (WGOTs)>is 150 psig. There are seven WGDTs per unit'and'two shared: shutdown decay tanks.
~
Each tank is provided with a pressure transmitter:and a high: pressure alarm in the control' room as part of_the gaseous waste
-processing. system trouble alarm. An~ operator would be dispatched I
, r q
l to the. local control panel.to clear the annunciated signal. The l 1 alarms for the WGDTs are set at 100 psig. All of the WGDTs are provided with relief valves set at or below the tanks design. , pressure. The relief valves for the WGDTs discharge to shutdown- l decay tanks which are normally at low pressure. Should an - extended discharge to the shutdown decay tank occur, a nigh alarm would'be received prior to lifting of the shutdown decay tank , relief valve. The' relief valves for the shutdown decay tanks ! discharge to the. plant vent which is monitored by the plant' vent monitor. Based on the protection afforded by the installed tank relief valves and the potential eventual release being to.the , 1 plant. vent, the gaseous waste processing system trouble alarm and ; pressure indication to 150 psig in the control-room is adequate l to provide information concerning the status of the WGDis. Based'on the relief valves being set below the tanks design I pressure, we find the deviation from the reconsnended range acceptable. *
- k. Emergency Ventilation Damper Position i In the applicant's initial submittal, this variable was not-addressed nor indicated in Table.7.5.2-1, not was the information required by Section 6.2 of NUREG-0737, Supplement No. 1 provided.
In the' March 15, 1985 submittal, the applicant states that this vartable is a part of the heating, ventilation, and~ air-conditioning system status. The required information is provided and the instrumentation is acceptable.
- 1. Auxiliary Building Radiation Level Portable Samp' ling ,
In the March 15, 1985 submittal, the applicant added the information required by NUREG-0737,' Supplement No. 1, for the "4 auxiliary building radiation level. From the information 16 .
4 provided, it is assumed that this instrumentation is for the ! Type C variable, radiation exposure rate. The applicant deviated
, both in category and range. The applicant provided Category 3 l instead of Category 2 instrumentation with no. range specified.
Within the June 20, 1985 submittal, the applicant provided justification for this deviation. The applicant stated that the non-containment area' monitors originally specified as a part of the process and effluent radiation monitoring system have been redesignated as Type E, Category 3 post-accident monitors. While the ranges of these monitors have not been expanded'or increased, they provide usable surveillance data. It should be noted that Revision 3 of Regulatory Guide 1.97 (Reference 10) has reclassified this variable as Category 3. Because post-accident servicing of safety-related equipment is not expected immedi'ately after an accident, permanently installed area monitors have not ! been provided. Because of the unpredictable nature of.
, maintenance requirements post-accident, plant health physicists ,
equipped with portable monitors will precede personnel into any
- areas where radiation may restrict access to service safety-related equipment. Portable radiation survey instrumentation with the capability to detect gama radiation )
4 over the range of 10 to 10 R/hr wil1~be maintained at the health physics station. From a radiological standpoint, if'the radiation levels reach or exceed the upper limits of the range, personnel would not be permitted into the areas without portable monitoring except for life saving. Based on the alternate instrumentation used by.the applicant for this variable, we find the deviations.for the. radiation exposure rate monitors' acceptable.
\
17 ' i
~
- 1 i
- 4. CONCLUSIONS Based on our review, we find that the app 11 cant either conforms to or ,
is justified in deviating from the recommendations of Regulatory Guide 1.97. except for the following: .
- 1. Containment sump water temperature - the applicant should provide ,
. fully. qualified Category 2 instrumentation for this variable to l provide a quantitative measure of the heat removed by way of the containment sump (Section 3.3.6),
y l l e J
'4 L
t t I i
~
j i 1 l 3 18 i s N
4., ur y 3 :.: .
.,':. TC; :\
( j-
!. f (Y t
- 5. REFERENCES
.. 1. D. G.-Eisenhut, NRC letter to all Licensees of Operating Reactors.
" Applicants for Operating Licenses, and Holders of Construction Permits', " Supplement No. 1 to NUREG-0737--Requirements for' Emergency.
Response Capability (Generic Letter'No. 82-33),' December. 17, 1982.
- 2. -Instrumentation for Licht-Water-Cooled Nuclear Power Plants-to Assess Plant and Environs Conditions Durina and Fo lowine an Aceident, Regulatory Guide-1.97, Revision 2, NRC, Office of Standards .
Development, December 1980.' 3.. Clarification of TNI Action Plan Reautrements. Recutrements for Emeroency Response Capability, NUREG-0737, Supplement No. 1, NRC, 4.0ffice of Nuclear Reactor Regulation, January 1983.
- 4. Georgia' Power. company letter D. Dutton to E. G. Adensam, NRC.-
" Supplement'.) to NUREG-0737 (Generic Letter No. 82-33)," April 14 1983.
- 5. Final. Safety Analysis' Report (FSAR) for Vogtle Electric Generating Plant, Unit l'and Unit 2.
' 6. Georgia Power Company letter, J. A. Bailey to E. G. Adensam,.NRC, a
' Request for Supplemental Informatton, OSER Open Item 62--Regulatory 1
. Guide 1.97, Rev. 2," March 15,1985, File: X7BC35, Log: GN-548.
1
- 7. ' Georgia Power Company. letter, J. A. Bailey to E.'G. Adensam, NRC,
* ' Request for Additiona1'Information: OSER Open Item 62 ' June 20, 1985, File: X7BC35. Log:. GN-646.
I
- 8. Georgia Power Company letter, J. A. Bailey to Director of Nuclear l Reactor Regulation NRC, 'SER Open Item 6: Regulatory Guide ~1.97 q Rev. 2. September 6,'1985 File: X78C35, Log: CN-703.
- 9. Georgia Power Company letter, J. A. Bailey to Director of.Nucleari .
Reactor Regulation, NRC, 'SER Open Item 6: ' Regulatory Guide'1,97, J Rev. 2,* January 21, 1986 ' File: X78C35, Log: GN-784. j
- 10. Instrumentation for Licht-Water-Cooled Nuclear Power Plants to Assess' Plant and Environs Conditions Durina and ;o' 10wina an Accident, :l Regulatory Guide 1.97, Revision 3 NRC, Of fice of Nuclear Research,: 'l May 1983.' j q
.\
i 38006 i
.i q
f 19: !
- w. .:i fi j
q
)
- y. .. .. .. . .. . .. .. .
e .'*l t-
] geoa==
wa .uc6u....uu , = co .. . .. .o. r -. . .. . r ,oc . .. . .. , E'," "" "
. (SISUOGRAPHIC DATA SHEET EGG-EA-6806
- u. . . wet.o. o ... . ! Alu Revision 2.
4 r,ta . o w. . a , a m. .g.. Conformance tr/ Regulatory Guide 1.97, o ;Vogtle Electr:c Generating Station
. Unit'Nos. T and42 s-aa n "'o " = ='u.o oor u.
....o...,
u -
-Apri1.
' l 1986-
***o"' "- "
R. VanderBeek v ..'" . g ....
)'AprilL o! '1986
, . . c .. .. o. c... .o r . . . . . u . ... . ,. ,. c , . .c,.,u .. .. .. w .
EG&G Idaho, Inc. Idaho Falls ID 83415 A6483
. . . o..o.... o.u.. . . .. ..o ... duo.. . ,, e. c , ~,e m o ...o., .
Division of PWR Licensing - A' Technical Evaluation Report-Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comission - ,- Washington, DC 20555 >! m
.i w.
' 3 3's,*t3Mdge.. 10785 lN l
o .ui..cr . , m o This EG&G Oaho, Inc.- espp reviews the submittals for the Vogtle Electric Generating Station Unpt Nos.1 and .2, 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 ls s 1 r s y , p > 3 d.o
.p s.
a
/:
. m.~.,..o....... .c.one...2% ,
'.m . . . . .~e.
la n... .. .. . - W. ' l, Unlimited Distribution- l -
- tf. -
n ac.a.+=n.ss,ocat ;,s
- e. ,
.s e,sve=cee n ..,s-
, on.,,..
Unclassified
,, .~~~~
eUncla i 'S
.w,,u.su si fied I
..;.u I #
s te 7..gg rn 6
- b. 'B
\~ , ?n.
T Y w , ,}}