ML20136F128
| ML20136F128 | |
| Person / Time | |
|---|---|
| Site: | Vogtle, Comanche Peak, 05000000 |
| Issue date: | 08/07/1984 |
| From: | Lamastra M Office of Nuclear Reactor Regulation |
| To: | Congel F, Muller D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML082840446 | List:
|
| References | |
| FOIA-84-663 NUDOCS 8408160020 | |
| Download: ML20136F128 (3) | |
Text
0 s
m orz r
+
n
\\
gf o
UNITED STATES g * ; s..
g NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 s
j e
%s
- AUG -0 7 1934 MEMORANDUM FOR:
Daniel R. Muller, Assistant Director for Radiation Protection, DSI Frank J. Congel, Chief Radiological Assessment Branch, DSI THRU:
Oliver D. T. Lyr.ch, Leader [. -
Radiation Protection Section,"RAB FROM:
Michael A. Lamastra Radiation Protection Section, RAB
SUBJECT:
V0GTLE - TRIP REPORT AND RESOLUTION OF Q-1 In the morning of July 24, 1984, John Hopkins and I met with Jim Bailey, Mike Kurtzmah, Don Hallman, Mehdi Sheibani, G. Bochtold, and Indira Kochery at the Vogtle Nuclear Power Plant. After a brief introduction and explanation of the purpose of our visit, we took a plant tour with Mr. Hallman and Ms. Kochery.
Since Unit 1 of the Plant was only approximately 65% complete, the Radiation Protection facilities were only partially complete.
Accordingly, a future plant tour by members of RAB staff will be required.
We toured the Unit 1 containment, viewed the spent fuel pool, the control room, the control room simulator, and other areas of the plant.
From the tour and discussions with Vogtle Health Physics staff the following observationc were made:
(1) The plant proposed RPM, Mr. Don Hallman, appears to meet the criteria of R.G. 1.8.
However, appropriate documentation is still required.
In addition, the Sr. Health Physicist, Ms. Kochery appears to be a qualified back-up to the RPM.
~
(2) The plant union contract may restrict the professional Health Physicist at the plant from directing the actions of Health Physicist Technicians.
This needs to be followed carefully by both NRR and IE to ensure that the plant radiation safety program is not compromised.
After the plant tour, we had an afternoon meeting with Vogtle management to discuss our Questions 471.01 - 471.17, and applicant's responses.
From this afternoon discussion it was agreed that additional information would be pro-vided for the following questions:
471.03. Vogtle will re-evaluate the plants health eysics equipment require-ments and submit new Tables 12.5.2-1 through 12.5.2-4.
d & O62D U
)
~.
_D. Muller
.. o r. - 2,, _
AUG 0 71934 F. Congel-
'471.06 Vogtle will submit a new resume for their Health Physics Superinten-dent demonstrating that he meets the criteria of R.G.1.8.
471.10. Vogtle will clarify their intent to meet the recommendations of Section 5, NUREG-0761.
The meeting _with the Vogtle Licensing and Health Physics group was very success-ful. Verbal agreements were reached for all open items.
Upon appropriate docu-mentation on the part of the applicant, we would anticipate no problems in closing all open items in a timely manner.
N s; W A cw Michael A. Lamastra Radiation Protection Section Radiological Assessment Branch, DSI cc:
- 0. Lynch M. Lamastra M. Miller, PM; Vogtle J. Hopkins S
9 e
\\--
4 c*;NRC:
r Michael 'A. La'mastra, DSI, NRR John Hopkins, DL, NRR Vogtle
~'
+v.
Jim Bailey, SCS,. Project Lincensing hidr Mike Kun tzman, GPC, Supr. H.P/ Chem Training Don Hallnian, GPC, HP Supt.
- Mehdi Sheibani, GPC, Regulatory Compliance
- G. Bockhold, GPC, General MGR. VN00
-Indira Kochery, GPC, Sr. Health Physicist t
,...z ;;.~ :..
~
~
- ~ * '
u._
.3_
.:y
> s.*..
.w r -,
4
%.sh e
g ge g
g
,m
.;c -
w
.. ' :. ~
~..
a,
'em # '
- h.
s-W
- D 9
op g,
,,e.*=--
$D '
'g**'
Wk1 g
b
't O
' t,
,,\\.
= *
- "'IOQ UNITED STATES NUCLEAR REGULATORY COMMISSION f,
j
~
WASHINGTON, D.C. 20555 o
+.***
FEB 'l 1984 MEMORANDUMFOR:[Eli Adensam, Chief, Licensing Branch #4 D
ision of Licensing FROM:
Brian W. Sheron, Chief, Reactor Systems Branch Division of Systems Integration
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION - V0GTLE ELECTRIC GENERATING STATION Plant Name:
Vogtle Electric Generating Station, Units 1 and 2 Docket No.:
50-424/425 Licensing Status:
OL Responsible Branch: Licensing Branch #4
'Gg Project Manager:
M. Miller J
Review Status:
Request for Additional Information Enclosed with this letter is a set of questions concerning the Vogtle plant. These questions are a result of a review of those sections of 5.2 and 5.4 of the FSAR for which Reactor Systems Branch has primary review responsibility. RSB is continuing its review and will submit additional questions as the evaluation proceeds through the other areas for which we are responsible.
N.
Brian W. Sheron, Chief Reactor Systems Branch Division of Systems Integration
Enclosure:
As stated cc:
R. W. Houston M. Miller CONTACT:
M. Wigdor X27592 f
'\\
T.7 s
i m_
- h GEORGIA POWER C0f1PANY V0GTLE ELECTRIC GENERATOR PLANT, UNITS 1 AND 2 DOCKET NOS. 424/425 REQUEST FOR ADDITIONAL INFORMATION 1.
An examination of P&ID's shows, in general, that there are no coordinates on the figures; consequently, it is very difficult to locate equipment and interconnections from figure to figure.
This is not acceptable; figures must be corrected so that coor-dinates are available.
Correct all piping and instrument draw-ings (P&ID's), accordingly.
O 2.(5.2.2.1)
What events, other than those listed cculd lead to overpressuri-zation of the RCS if adequate overpressure protection were not provided?
- 3. (5.2.2.8, What kinds of positive position indication are provided for the 5.2.2.10) pressurizer safety valves and PORVs? Discuss compliance with NUREG-0737 items II.D.1 and II.D.3.
4.(5.2.2.10.2)
What are the postulated worst case mass input and heat input events for a Low Temperature /0verpressurization event? Staff position, for previous Westinghouse plants, has been that the I
' t. -
design basis mass addition event is the inadvertent actuation of
\\
a safety injection pump under runout conditions. Justify your selection if other than the above.
- 5. (5.2.2.10.4)
A number of administrative controls have been described in the FSAR to maintain RCS pressure to within allowable limits. What alarms are available in the control room to remind the operator that specific administrative controls are to be effected (e.g.,
maintaining at least one RCP in operation until reactor coolant temperature reaches 160*F and maintaining a full open valve in the bypass line to the letdown orifices during water solid operation)?
.,s l#. ( J (5.2.2.10.4)
In the FSAR it is noted that ECCS actuation on higS pressurizer t
pressure or low steam line pressure is blocked for RCS pressures under 1900 psig.
Provide an analysis that shows that these SI signals are not needed for this condition.
' 7. (5.2.2)
Section 5.4.13.2 describes the loop seals on the pressurizer safety valves.
Has the delay due to the time it takes to
_ discharge the water from these loop seals been accounted for in the limiting pressure transient? If it has not been accounted for, how would this delay affect the conservatism of the result?
..y
~-
4 -
- 1. (5.2.2, Check valv'es in the discharge side of the high pressure safety 5.4.12,6.3) injection, low pressure safety injection, RHR, and charging systems perform an isolation function in that they protect low pressure systems from full reactor pressure. The staff requires that these~ check valves be classified ASME IWV-2000 category AC, with the leak testing for this class of valve being performed to code specifications.
It should be noted that a testing program which simply draws a suction on the low pressure side of the outermost check valves will not be acceptable. This only verifies that one of the series check valves is fulfilling an isolation function. The necessary frequency will be that speci-fied in the ASME Code, e,xcept in cases where only one or two check valves separate high to low pressure systems. 'In these O.
cases, leak testing will be performed at each refueling after lD}
the valves have been exercised.
Identify all check valves which should be classified Category AC as per the position discussed above.
Verify that you will meet the required leak testing schedule, and that you have the necessary test lines to leak test each valve.
Provide the leak detection criteria that will be in the Technical Specifications.
9.(5.2.2)
WCAP 7769, Section 3.4 assumes failure of one steam generator safety relief valve per loop. Provide assurance that your remaining safety valves can provide the required minimum capa-city or justify why your analysis assumes only a single failure in one loop.
l
r.
'O. (5.2.2.2)
It is stated in 5.2.2.2 of the FSAR that the pressurizer and SG
~
safety valves are sized with sufficient capacity to provide overpressure protection for typical worst-case transient condi-tions. What are these conditions? Verify that sufficient margin has been provided to account for uncertainties in the 4
design and operation or the plant and that the maximum instru-mentation and control errors have been assumed.
Discuss the preoperational tests which will verify the accuracy of instru-mentation systems used to initiate overpressure protection.
i
- 11. (5.2.2)
What are the setpoint tolerances for all of the safety and power operated valves? What tolerances are taken credit for in the setpoint analyses? Do the analyses take into account setpoint
-- 7) drift?
- 12. (5.2.2.2)
In Section 5.2.2.2, reference is made to WCAP-7769, Revision 1.
Provide a. comparison of Vogtle parameters with all parameters
~
listed in Table 2.2 of this topical report. Where differences
~
exist, show that these differences will not affect the conserva-tism of the results given in WCAP-7769.
13.(5.2.2)
Provide verification that your analysis of the limiting tran-sient for overpressure protection assumes the reactor trip is initiated by the second safety grade signal.
~
S-14.(5.2.2)
Provide assurance that-the dynamic loading of the PORVS and the code safeties due to water relief has been considered in the piping and support analysis including the passage of a water slug and effects of water hammer. What liquid water relief rates were assumed in the loading analysis? Are these values consistent with experimental results? Are the power operated relief valves and safety valves designed and qualified for liquid relief?
- 15. (5.2.2, Section 5.4.13.2 cites a backpressure compensation feature on 5.4.13) the pressurizer safety valves.
Provide a discussion of this feature which explains how this function is performed.
6.(5.2.2)
, Have the pressurizer PORVs been qualified for the dynamic loads that could be sustained when the maximum liquid flow rate or maximum acceleration of liquid occurs during a low temperature overpressurization?
- 17.(5.2.2.10)
In Section 5.2.2.10, it is stated that the low temperature overpressure protection system is manually armed.
Is there an alarm to alert the operator to arm the system at the correct plant condition during cooldown as required by Branch Technical Position RSB 5-2.B.37 18.(5.2.2)
Provide a description of the design features to be used to mitigate the consequences of overpressure events while
- i 4
m
~0~
operating at low temperatures. Our position regarding overpres-sure protection while operating at low temperatures is presented in the Branch Technical Position RSB 5-2 attached to SRP 5.2.2.
Your description should address each portion of this position.
19.(5.2.2)
The Branch Technical Position RSB 5-2 states the reactor vessel overpressurization protection system should meet the single active failure criterion when the initiating cause of the event is not considered as the single active failure. Provide a failure modes and effects analysis to demonstrate that a single electrical or mechanical component failure will not disable both trains of PORVs from functioning.
,0.(5.2.2.10.1)
In Section 5.2.2.10.1 of the FSAR, you indicate that "an auction-eered system temperature is continuously converted to an allow-able pressure and then compared to the actual RCS pressure. The system logic first annunciates a main control board alarm whenever the measured pressure approaches within a predetermined
~
amount of the allowable pressure, thereby indicating that a pressure transient is occurring, and on a further increase in measured pressure, an actuation signal is transmitted to the l
PORVs when required to mitigate the pressure transients." Our review of the low temperature overpressure protection design for certain other Westinghouse plants indicates that a failure in O
=
l,,
the temperature auctioneer for one PORV (signalling it to remain closed) could also fail the other PORV closed (by denying its,
permissive to open). Address this concern about a potential common mode failure in the low temperature overpressure protec-tion system for Vogtle.
21.(5.2.2)
Provide your limiting Appendix G curve for the first eighteen full power months of operation.
Discuss the operational proce-dures which will minimize the likelihood of an overpressure event.
22,(5,2,2,10,1) The staff is concerned that your proposed low temperature overpressure protection (LTOP) system does not adequately Y)'
protect the reactor vessel during transient events where the vessel wall temperature lags behind the temperature used in the variable setpoint calculator.
For example, starting an RCP in a loop with a hot steam generator when the RCS is water solid causes the RCS pressure and temperature to rise. Your LTOP
~
system would automatically raise the PORV setpoint as a function of auctioneered cold or hot leg temperature, but the vessel wall will not be heated in this transient at the same rate. Thus, due to the LTOP system auctioneering scheme, the part of the RCS most vulnerable to brittle fracture will be protected to a higher pressure than its temperature allcws.
D
-8 If, during a cooldown, the cold leg temperature detector down-stream of the generator (s) being used failed, and a mass input, event occurred, your proposed LTOP system may not protect the coldest location in the vessel since the setpoint would not be based on the coldest fluid temperature.
Address the above concerns by discussing the following:
(1) Show that for all normal events and events in which the RCS fluid temperature is changing, your proposed system suit-ably protects the reactor vessel at its coldest location.
(2) Show data to justify the RCS temperature transients assumed in(1)'above.
(3)
Include in your analyses the most limiting single failure, and justify the choice.
(4)
Include in your analyses the effects of system and compo-
~
nent response times, including; a.
temperature detectors b.
pressure detectors c.
logic circuitry Show the response times that were assumed and the techniques, including surveillance requirements for ensuring their conserva-tism.
(-
.,w e
+ew
?3.(5.4.7.2.4)
Section 5.4.7.2.4 of FSAR states that "Each discharge line from 1
'the RHRS to the RCS is equipped with a pressure relief valve designed.to relieve the maximum possible back leakage through the valves." What is the basis for determining the maximum possible back leakage?
Is this back leakage consistent with a relief flow capacity of 20 gpm at a set pressure of 600 psig?
Show that there are design provisions to permit periodic testing for leak tightness of the check valves that isolate the discharge
. side of the RHRs from the RCS.
24.(5.4.7.2.1)
Is there direct position indication for the isolation valves on l
the suction side of the RHR system?
i i
5.(5.4.7.2.1)
The RHR miniflow bypass lines allow bypass flow when RHR pump f.
discharge flow is insufficient. At what frequency is the operability of these miniflow lines verified? What assurances
[
_ are' available to the operating staff that the miniflow isolation valves are not misaligned? Discuss what testing will be per-I'-
formed to validate that the miniflow lines provide an adequne pump flow path such that damage to these pumps will be precluded i
during this mode of operation.
26.(5.4.7)
Branch Technical Position RSB 5-1 specifies in Table 1, Item 1-C that the steam generator atmospheric dump valves (ADVs), their operators, and their power supplies shall be safety grade. FIAR l(
Section 10.3.1 states that the power operated atmospheric relief valves are part of the safety design basis. Are the ADVs, the,ir operators, and their power supplies considered safety related and therefore are designed to safety grade standards?
If not, why not?
27.(5.4.7.2.2.'1) Does the RHR pump performance curve take into account instrumen-tation uncertainties used in deriving the curve?
28.(5.4.7.2.3.4) What precautions and procedures are there that will preclude the cooldown rate from exceeding the Technical Specification limit upon loss of instrument air to both the RHR heat exchanger outlet and bypass flow control valves?
C=..,1 What indications and alarms are available in the control room to intonn the operator of a potential excessive cooldown rate and how excessive can this rate become before the situation is turned around?
~
29.(5.4.7.2.3.5) Section 5.4.7.2.3.5 states that the steam generator power-oper-ated relief valves can be used to attain a primary side cooling rate of 35"F/h. Section 10.3.2.2.3 states that these valves can be used to attain a rate of 50*F/h.
Explain this discrepancy.
Is the control system used to maintain this rate considered a l
safety related system and designed to safety grade standards?
l
. _ _... =. _...
11 30.(5.4.7.2.2.3) The FSAR states that " valves that perform a modulating function are equipped with two sets of packing and an intermediate leakoff connection that discharges to the drain header."
Please identify these valves.
Do both trains of the RHR system share the same header?
31.(5.4.7.2.3)
With the RHR flow high enough such that the miniflow bypass flow valve is closed, is it possible, considering a single failure, tha' both the residual heat exchanger ' outlet and bypass flow control valves will be closed? How much time does it take for the miniflow bypass flow valve to open and what is the possible damage to the RHR pump in the interim?
.1)32.(5.4.7.2.3)
Discuss the possibilities for air getting trapped in any part of the RHRS during startup and for the air causing water hammer and damage to the RHRS.
~ 33. (5.4.7.1)
Provide the calculations of the cooldown times given in Section 5.4.7.1.
What values were assumed for the component cooling water temperature, heat transfer surface area and heat transfer coefficient? Show in the, calculations that fouling of the heat exchanger was taken into account.
34.(5.4.7.2.3)
In Section 5.4.7.2.3.5 it is stated that local manual actions could be performed if permitted by the prevailing environmental
T.
12 conditions in order to achieve cold shutdown. Would any of these actions become necessary considering a single failure coupled with a loss of either onsite or offsite power and if so what are the actions and where may these actions occur?
- 35. (5.4.7)
Provide detailed information on the sizing criteria used to determine the relief capacity of the RHR system suction line pressure relief valves.
Did the version of the ASME code to which these relief valves were sized require establishing liquid or two-phase relief capacity with testing? If so, describe in detail the test program and results.
If the liquid or two-phase relief capacity
&--J was not established by test, show that the difference between lU the rated and maximum required capacity is more than sufficient
';o bound liquid and two-phase relief rate uncertainties.
In the absence of liquid relief valve testing, describe why you believe
~
these valves can reliably pass water without damage to the valve.
- 36. (5.4.7)
Provide additional information regarding the power sources I
supplied to the RHR isolation valves. The staff's position is that a single failure of a power supply or interlock will not prevent isolation of the RHR when RCS pressure exceeds it's design pressure. Additionally, loss of a single p~ower supply cannot result in the inability to initiate at least one 100 percent RHR train.
l What is the design pressure of the RPRS? Section 5.4.7.2.4 h.
7.-(5.4.7.2.4) states that each RHR relief valve has the capability to maintain the RHRS to within maximum code limits.
Identify those design basis events that were excluded from the analysis that determined the relief valve capacity. Provide the bases for the exclusions. Specifically address the capability of the valves to provide relief for the discharge of the charging pumps as well as thermal expansion. Describe the postulated accident events and their sequences, including the discharge of the accumulators, and the combined flow of the safety injection pumps which exceeds the charging pump flow at lower pressures.
- 38. (5.4.7.2.2.3) Describe the design for the RHRS isolation valves and the tests R
performed to demonstrate that they will operate properly for the postulated pressure transients and environments.
~39.(5.4.7.2.1)
What indicates loss of component cooling water to RHR pumps? Do all of these instruments meet IEEE 279 requirements? How long could the pumps continue to run following a loss of component cooling water without damage? Provide date to support your response 40.(5.4.7)
Provide or reference a discussion of your compliance with each item of RSB BTP 5-1 in NUREG-0800. Justify any deviations from j
this Branch Technical Position.
(.
14 It is the staff's position that all operator actions necessary k
to take the plant from normal operation to cold shutdown should be performed from the control room.
Indicate whether there are any systems or components needed for shutdown cooling which are de-energized or have power locked out during plant operation.
If so, indicate what actions have to be taken to restore oper-ability to the components or systems.
In particular, address the accumulation system.
i 41.(5.4.7.2.2)
In the event the RHS relief valves'open, describe the means available to alert the operator of the situation. Verify that procedures will be available to the operator for responding to this event.
O
~42. (5.4.7)
Provide the following information related to pipe break or leaks in high or moderate energy lines outside containment associated with the RHR system when the plant is in a shutdown cooling
~
mode:
1.
Determine the time available for operator action based on the maximum discharge rate from a pipe break in the systems outside containment used to maintain core cooling.
2.
Describe the alarms available to alert the operator to the event, the recovery procedures to be utilized by the operator.
f i
13.(5.4.7)
Recent plant experience has identified a potential problem regarding the loss of shutdown cooling during certain reactor.
coolant system maintenance evaluations.
On a number of occa-sions when the reactor coolant system has been partially drained, improper reactor coolant system level control, a partial loss of ieactor coolant inventory, or operating the RHR system at t.: inadequate NPSH has resulted in air binding of the RHR, pumps with a subsequent loss of shutdown cooling.
Regarding this potential problem, provide the following additional infor-mation.
1.
Discuss the design or procedural provisions incorporated to maintain adequate reactor coolant system inventory, level control, and NPSH during all operations in which RHR r<
1 ":t cooling is required.
O*
2.
Discuss the provisions incorporated to ensure the rapid restoration of the RHR system to service in the event that the RHR pumps become air bound.
3.
Discuss the provisions incorporated to provide alternate methods of shutdown cooling in the event of loss of RHR cooling during shutdown maintenance. These provisions should consider maintenance periods during which more than one cooling system may be unavailable, such as loss of steam generators when the reactor coolant system has been
~
partially drained for steam generator inspection or mainte-
[
nance.
~'
l 44.(5.4.7.2)
Describe the consequences of a failure associated with the
)
isolation valves in the suction line from the hot leg to the RHR pumps during normal shutdown cooling.
Evaluate this event assuming that"only one RHR train is operating at the time.of the failure. Describetheconsequencesofthiseventassuming(a) the reactor vessel is closed, and (b) the reactor vessel head has been unbolted. The failure could be caused by operator error or a passive failure such as the gate separating from the stem. These failures could cause pump dama'ge due to cavitation end loss of core cooling.
Discuss the operator actions required to mitigate the consequences, describe the alarms available to alert him to the situation and the time frame available to perform the required action.
--3)
- 45. (5.4.7.2.3)
Describe your proposed program for varification of adequate mixing of borated water added to the RCS under natural circulation conditions and confirmation of natural circulation
~
cooldown ability, in accordance with the criteria of BTP RSB 2_
5-1.
4.
- 46. (5.4.13) -
The FSAR states that the PORVs provide a s'afety related means for RCS depressurization to achieve cold shutdown.
Does this mean that the entire PORV system is designed to safety related v
criteria?
~
s e
o I
e i
l '.,,
_17 o
'7. (5.4.15)
The FSAR description of the Reactor Vessel Head Vent system (RVHVS)isincomplete.
Please provide the following information to show compliance with the requirements of Action Item II.B.1.
(1)
Please amend the RCS P&ID (Figure 5.1.2-1) to show the
- ping, valves, and instrumentation, including control and indication, for the RVHVS.
Include in this drawing the identification associated with these items.
(2) Describe the means for venting the pressurizer.
Describe the procedures that would assure sufficient coolant can enter the U-tube region so that sufficient decay heat can be removed from the RCS.
~
(3) Describe RVHVS testability.
(4) Describe the position indication for the RVHVS valves that is available in the control room.
(5)
Describe the control system th'.t operates the RVHVS valves.
Describe its compliance with safety related requirements.
(6)
Identify the power sources for the modulating valves.
~
(7) Describe the capability of the system to vent the RCS hot and cold legs.
(
- ..-g,.
)
.r (8)' Provide a reliability analysis consisting of a failure mode
{
and effects analy' sis t(FMEA) or equivalent qualitative analysis that showe that no single active component fail-ure, human error, or test and nefn,tenance action could i
result in inadvertent opening or failure to close after 3
' intentional opening of an RCS vent path.
Include in the analysis components in the associated power, instrumenta-tion, and control systems as well as the electrical and mechanical components of the RCS vent system (reference NUREG-0737ItemII.B.1ClarificationA.(7)and(8)).
- 48. (5.2.2)
Will the PORV setpoints be adjusted over time for low temperatureoverpreNureprotectioninordertoaccountfor
.i")
vessel embrittlement? Justify your response.
4 e
O-t O
'\\
?M 3
o UNITED STATES g
NUCLEAR REGULATORY COMMISSION o
h WASHINGTON, D. C. 20555
% w..... /
FEB 141984 Docket Nos. 50-424/425 MEMORANDUM FOR: Elinor G. Adensam, Chief Licensing Branch No. 4 Division of Licensing FROM:
George Lear, Chief Structural and Geotechnical Engineering Branch Division of Engineering
SUBJECT:
REVIEW QUESTIONS - GE0 TECHNICAL ENGINEERING Plant Name: Vogtle Electric Generating Plant, Units 1 and 2 Licensing Stage:
OL Docket Number:
50-424/425 Responsible Branch: Licensirg Branch No. 4, M. Miller, LPM We have reviewed Section> 2.5.4 of the Vogtle Electric Generating Plant (VEGP), Units 1 and 2 FSAR submitted by Georgia Power Company in support p.
R-7 of their application for an Operating License for VEGP. On the basis of V
this review we have identified the additional infonnation needed to complete our safety evaluation. The enclored questions prepared by Joseph Kane and Dinesh C. Gupta, Geotechnical* Engineering Section, Structural and Geotechnical Engineering Branch, Division of Engineering, have been prepared for your transmittal to the applicant. identifies reports referenced in the FSAR which need to be provided by the Applicant in order to complete our safety review.
O_
a). 7/d&
g Geo e Lear, Chief v
Structural and Geotechnical Engineering Branch Division of Engineeting
Enclosure:
As stated cc:
J. Knight J. Kane T. Novak S. Chan
. G. Lear D. Gupta R. Jackson M. Miller L. Heller G. Staley 0
I. Alterman l
A. Ibrahim M
I 1
y ~-
Enclosure
- j Vogtle Electric Generating Plant, Units 1 and 2 Docket Nos. 50-424/425 FSAR Review Questions Geotechnical Engineering Section Prepared by J. Kane, DE, SGEB, GES with Input from D. Gupta, DE, SGEB, GES 241.1 Table 1.3.2-1 should identify the foundation design change from
~(SRP2.5.4)
'the PSAR to place one of the seismic Category 1 structures (the Radwaste Solidification Building) on drilled caissons.
241.2 The FSAR does not provide a plan that clearly identifies all (SRP2.5.4) seismic Category 1 structures, piping and conduits in relation to their foundation conditions. We recommend a plan similar to Figure 2.5.1-23, Sheets 2 and 3 be developed that provides the following minimum infonnation:
1.
Outline of all seismic Category 1 structures including tanks and tunnels and the location (alignment) of seismic Category 1 piping and conduits.
2.
Location of' foundation excavations (Top and bottom elevations, A
slopes) including the outline of the deeper excavation to Q
Elev.108.6 ft.
3.
Sufficient bottom foundation elevations of piping and conduits
.to understand the depth of fill beneath them and elevations of piping penetrations into structures.
4.
The extent of riprap placement and the excavation slopes that slumped which are described on Page 2.5.1-24.
If clarity of drawing permits, the extent of the eroded Category 1 backfill areas that occurred in November 1979 should also be shown
,(refer to Q 241.23).
5.
The location of borings and test pits used to define geologic and foundation conditions.
marl bearing stratum after completion The borings drilled in the clay (See page 2.5.4-4) indicate poor 241.3 (SRP.2.5.4) of the power block excavation core recovery in nine of the 36 borings completed. No explanation or discussion on this poor core recovery is offered in the FSAR. To assist the Staff in its assessment of foundation adequacy we request-that representative cores be made available for inspection at the planned site visit. The selection of recovered cores for display should include boreholes.where recovery was poor as well as good and cover the entire depth of the marl layer and should be c.
m--
m,
' 241.3 for boring locations in the vicinity of important seismic Category 1 (Continued) structures. To better understand the procedures followed in the selection of important soil and foundation design parameters (shear strength, soil modulus, etc.) for this highly variable clay marl stratum, we request that engineers knowledgeable *in this selection procedure that was completed be available for discussions during the site visit.
241.4 Provide a sumary of the actual results for control testing
.(SRP2.5.4) completed on compacted Category I backfill. The summary should pemit the location and elevation of backfill material tested to be recognized and graphically demonstrate how PSAR commitments on Category 1 backfill requirements were fulfilled (gradation, placement moisture content, in situ density, moisture-density relationsandpercentcompaction).
l 241.5
-Provide a table with the~as-built dimensions (length and width)
(SRP 2.5.4) for all seismic Category 1 structural foundations and 'ndicate the bottom elevations of foundation slabs. To understand the magnitude of actually applied bearing stresses provide the applied gross and net loading stresses (dead, live and seismic loading) for all seismic Category 1 structures including valve house, pumphouses and tanks. Table 2.5.4-12 needs to be revised to include the results of bearing capacity analysis for all seismic Category 1 sk structures. The maximum pemissable foundation pressures listed
~~
in the last column of Table 2.5.4-12 appear to be in error and appropriate corrections should be made. The factors of safety under dynamic loading conditions should also be provided.
241.6 Please identify the' location of observation wells 101 A, 247, (SRP2.5.4) 248, 806B and 807A on Figure 2.4.12-6.
Verify that the water level measurements presented for these wells on Table 2.4.12-7 are in agreement with the water table contours and piezametric 1
surfaces shown on Figures 2.4.12-6 and 2.4.12-7.
Table 2.4.12-7 should also identify the bottom elevations of the cbservation wells.
l 241.7 Describe water level measurements made at observation wells T-1, (SRP2.5.4)
MU-1, MU-2, 138 and 181 and show that these measurements are consistent with the shown contours on Figures 2.4.12-6 and 2.4.12-7.
The contours drawn in the area of wells 27 and 157 are not consistent with the water level measurements presented in Table 2.4.12-7.
Explain this inconsistency or revise contours.
241.8 Provide water level measurements available in the plant area (SRP 2.5.4)
(Detail A - Figure 2.4.12-6) after 1974 in the aquiclude and confined aquifer.
j
('
a
~
i
_41.9 The staff has difficulty in understanding the statement "The (SRP2.5.4) marl contains no free ground water and no springs, etc."
(Page-2.5.4-21, third paragraph) in view of the water level measurements submitted in Table 2.4.12-7 for the marl aquiclude.
A groundwater pressure diagram should be provided for depths that extend into the lower confined aquifer. The diagram should.
reflect <the hydrostatic pressures used in the design of structures founded on or in the clay marl stratum. Address the possibility that the water head loss across the marl layer results from openings or cavities in the lower portions of the clay marl layer.
We request a figure be provided that presents a sectional view of the typical installation details for the observation wells installed in the upper water table zone, the aquiclude and the confined aquifer.
241.10 Please identify the wells /piezometers that are to be measured (SRP 2.5.4) as part of the groundwater monitoring program (Page 2.5.4-23) during years of plant operation.
Provide the pertinent infomation for these wells /piezometers (top and bottom elevations, type, typical installation details, etc.) and discuss the monitoring program requirements and objectives (frequency of readings, expected range of piezometric levels, field controls, etc.).
Identify the controls to be required in technical specifications.
./241.11 Compare the results of your field geophysical surveys and (SRP 2.5.4) laboratory strain-controlled dynamic triaxial testing with the adopted curves of strain dependent shear modulus and damping ratios (Figures 3.7.B.2-5 thru 3.7.B.2-7 and Figures 3.7.B.1-8 thru 3.7.B.1-10) f.or the compacted backfill, clay marl and lower sand layer which were used in your soil-structure interaction analysis.
Discuss the bases for your selection of the adopted curves.
In view of the wide range in engineering properties exhibited by the foundation materials (Table 2.5.4-1), indicate what reasonably conservative variations in dynamic soil properties were used in soil-structure interaction studies.
241.12 Provide the input values to permit the Staff to verify that you (SRP2.5.4) have used a consistent set of soil properties and soil profiles below plant grade in your finite-element and lumped parameter studies. Your description of techniques used to obtain.the impedance functions for layered medium, provided in the Appendix 3E of the FSAR, is inadequate. Give the depth of soil profile and values of soil parameters you considered while using this approach.
Provide design assumptions and sufficient details of your calculative procedures and results to justify your proper use of soil stiffnesses and damping values for soil springs used in your lumped-parameter analysis.
i
4
- i-41.13 Although the depth to bedrock below the plant finished grade is.
l 2
(SRP2.5.4) approximately 950 ft., your soil-structure interaction model uses a 219 ft depth of soil thickness (Figures 3.7.B.2-3 and 3.7.B.2-4).
Provide details of your assumptions and justify the basis for selecting this depth of soil.
241.14
-Provide the values of Category I backfill properties used in the (SRP2.5.4) seismic analysis of the underground piping and conduits.
Explain and reference your procedure for calculating dynamic axial and bending stresses including the seismic input used for this-analysis. Verify that you have adequately accounted for the effect of reasonable variations in soil properties in your analysis.
241.15 Sufficient information and details for the foundation design (SRP2.5.4) under static and dynamic loading has not been provided for seismic Category 1 tunnels, water storage tanks and the Diesel i
Fuel Oil Storage Tank Pumphouses. This information is needed along with the engineering soil properties adopted in foundation design with the supporting basis for that selection.
241'.16 Please provide an explanation on how the range of estimated total (SRP2.5.4) settlements shown on Figure 2.5.4-8 was established.
4
. re-*41.17 The'FSAR does not provide any records of actual settlements H. -)RP 2.5.4) measured to date.
Provide up-to-date plots of settlement versus D
time for seismic Category 1 structures. These plots should also i
reflect significant contruction activities (foundation excavation and heave, dewatering events, magnitude of structure loading stresses, etc.) in order to pemit an understanding of the effect of these activities on settlement behavior and structure 1
performance.
~
241.18 Discuss and compare total and differential settlement allowed for i - (SRP 2.5.4) in design.with actual settlement records at specific structure and buried piping locations.
Provide sufficient plan and sectional views of involved structures and' buried piping as needed for meaningful discussion and comparison.
Provide a table of maximum stresses for the required loading combinations that includes infomation indicating the magnitude of stresses induced by differential settlements as allowed for in design.
241.19 Provide sufficient infomation and details (specific locations,
.(SRP2.5.4) frequency of readings, allowable settlement limits, etc.) to permit an evaluation of the settlement monitoring program to be required during years of plant operation.
Identify the controls (e.g.
allowable limits) and criteria to be required in the technical specifications for seismic Category 1 structures and piping.
25 -
I 41. 2 0 The discussions and information provided on caisson foundation (SRP2.5.4) design (Section 2.5.4.10.3) should be expanded to include description of actual field installation (layout, typical sectional views, method for drillir.g and casing, results of down-the-hole inspection, placement procedures, etc.) and any construction problems encountered and actual settlement versus time plots.
Provide a description of the seismic analysis completed on this caisson supported foundation with adopted dynamic soil and caisson properties and include response estimates to demonstrate an adequate safety margin is available for foundation stability under SSE loadingeconditions.
241.21 Paragraph 2.5.4.10.5 is inadequate in describing the design (SRP2.5.4) procedures used to establish lateral earth pressures. Provide a detailed discussion covering all seismic Category 1 structures on the procedures used in design to determine passive earth pressures and dynamic earth pressures and include supporting pressure diagrams and actual soil parameters adopted in design with the basis for selection.
Give the values of soil friction used in studies of sliding resistance and present, in tabular fonn, the calculated factors of safety against overturning, sliding and flotation for all seismic Category 1 structures for applicable loading combinations.
~ 41.22 There is insufficient information in Section 2.5.4.14 of the FSAR 1 JRP 2.5.4) which documents the actual field work completed to repair the eroded areas of Cattgory 1 backfill (November 1979). Describe the actual field procedures and activities performed to establish the extent of disturbed soil backfill in the eroded areas.
Provide the results of field and laboratory tests which were completed to verify the competency of the unaffected fill.
Identify with supporting figures the extent of repair excavations (grades, limits and slopes) and describe backfill operations including working space limitations. Describe procedures used to overcome these limitations.
Define the 11mits1(areilaandsdepth) where mounds of loosely placed temporary fill were placed to prevent further seepage and erosion after the 1979 incident.
Describe the steps taken to remove and backfill these areas.
Describe future monitoring planned to demonstrate the adequacy of the completed repair work.
241.23 Describe the measures taken to assure foundation stability of (SRP2.5.4) e.ffected safety related structures and piping in the areas subjected to extensive excavation slope slumping andrriprap placement (Page 2.5.1-24). Coordinate this discussion with the limits identified in the requested plan in question Q241.2 item 4.
O I
l l
l
I Vogtle Electric Generating Plant, Units 1 and 2 Docket Nos. 50-424/425
Subject:
Referenced FSAR reports required for completion of geotechnical engineering safety review Required Reference No. and' Title Page Identifying Reference No. 64, " Report of Marl Investigation:
2.5.1-45 December 1974 No. 65, " Report on Stratigraphic Irregularities 2.5.1-46 Exposed in Auxiliary Building Excavation" February 1978 No. 66, " Report of Geology and Foundation 2.5.1-46 Conditi6ns', Rower Block Area" September 1979 No. 7,
" Test Fill Program, Phase II" 2.5.4-35 October 1978
~+1 3
No.15, " Final Report on Dewatering and Repair 2.5.4-36 in Category 1 Backfill in Power Block Area" August 15, 1980.
No. 16 Letter from C. J. Dunnicliff to J. D. Duffin 2.5.4-36 October 12, 1977 e
e e
O (3
A
'o,,
UNITED STATES y
g g
NUCLEAR REGULATORY COMMISSION yj WASHINGTON, D. C. 20555
$,7% 4,/
April 16, 1984 Docket Nos.
50-424 50-425 MEMORANDUM FOR:
Elinor G. Adensam, Chief Licensing Branch No. 4 Division of Licensing Office of Nuclear Reactor Regulation FROM:
David B. Matthews, Acting Chief Emergency Preparedness Branch Division of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement
SUBJECT:
REVIEW 0F THE V0GTLE ELECTRIC GENERATING PLANT EMERGENCY PLAN
.We have completed our review of the Vogtle Electric Generating Plant Emergency Plan which is not dated, but was received for our review on September 16, 1983.
The plan was reviewed against the requirements of 10 CFR 50.47(b); 10 CFR 50, p~.
r
)
Appendix E; Supplement 1 to NUREG-0737 (Generic Letter 82-33); and the guidance C7 criteria in NUREG-0654, Revision 1, which has been endorsed as Regulatory Guide 1.101, Revision 2.
Our review has indicated that the plan is incomplete and a large amount of additional infonnation and mar additional' comitments are re-quired from the applicant before we can find tr4is plan acceptable.
It is requested that the enclosed coments and a letter similar to. the enclosed draft be sent to the applicant.
Please provide this Branch with a copy of the final correspondence.
In addition, it is suggested that a meeting between the applicant and the staff of this Branch be established to review and discuss the b-enclosed coments to ensure that the applicant understands the requirements for ar. acceptable emergency plan.
The Emergency Preparedness Branch contact is Ed Williams (492-7611).
I David B. Matthews, Acting Chief Emergency Preparedness Branch i
Division of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement Enclosures.
- 1. Coments on Vogtle EP
-[
- 2. Draft Ltr cc:
See Attached
- 7, _ "I &
L Elinor G. Adensam -
cc: _ E. L. Jordan, IE J. N. Grace, IE S. A. Schwartz, IE
.C. R.; Van'Niel, IE F..Kantor, IE E. F. Williams, IE
.:-h m
ao r
7 1
l l
r t
f
~~
L t
V DRAFT Docket Nos. 50-424 50-425 Mr. D. O. Foster Vice President and General Manager Vogtle Project
~
Georgia Power Company Post Office Box 4545 Atlanta, Georgia 30302
Dear Mr. Foster:
We have completed our review and evaluation of your Emergency Plan submitted on August 29, 1983 for the Vogtle Electric Generating Plant, Units 1 and 2.
The acceptance criteria used as the basis for the staff's review of your Emergency Plan are specified in Section 13.3 " Emergency Planning" of the Standard Review Plan (SRP), NUREG-0800 dated July 1981 and include the Planning Standards of 10 CFR 50.47(b); the requirements of 10 CFR 50, Appendix E; the-specific guidance criteria of NUREG-0654/ FEMA-REP-1, Revision 1 " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Prepared-ness in Support of Nuclear Power Plants", dated November 1980 and Supplement 1 to NUREG-0737, " Requirements for Emergency Response Capability (Generic Letter 82-33)",'datedDecember 17, 1982. The guidance criteria of NUREG-0654 have been endorsed in Regulatory Guide 1.101,. Revision 2. " Emergency Planning and Prepared-ness for Nuclear Power Reactors", dated October 1981.
Enclosed are the staff's comments on your Emergency Plan indicating the need for
' additional information and commitments which are necessary before we can find your 5mergency Plan acceptable.
We request that you provide this office written l
~
D. O. Foster -
responses to these comments, along with page changes for your Emergency. Plan reflecting your connitments within 90 days of the date of this letter.
In addition, it is requested that your staff meet with the NRC staff on May 1984 to review and discuss these connents and your plans for addressing each item.
Sincerely, Elinor G. Adensam, Chief Licensing Branch No. 4 Division of Licensing Office of Nuclear Reactor Regulation
. :: Y
.i o
e.
6
[
/
L ENCLOSURE 1 Review Coments On The Alvin W. Vogtle Electric Generating Plant, Units 1 and 2 Emergency Plan (August 1983)
Docket Nos. 50-424 and 50-425
^ The following comments apply to the Vogtle Electric Generating Plant Units 1 and 2 Emergency Plan (herein after referred to as the plan) and identify in parentheses, the ap(plicable evaluation criteria of 10 CFR 50 or Regulatory G 1~.101, Revision 2 NUREG-0654/ FEMA-REP-1, Revision 1) or Supplement I to NUREG-0737.
A.
Assignment of Responsibility 1.
The plan does not identify the State and local organizations in the State of South Carolina, al.1 the local organizations in the State of Georgia and the principal Federal organizations that are intended to be part of the overall. response.
(A.1.a & App.5) 2.
The-plan does not specify the concept of operations of many of the organiza-tions and suborganizations having an operational role or their relationship h,:
to the total effort.
For those organizations and suborganizations for which t
their role and relationship to the total effort is described, the description is not adequate to determine their concept of operations or their exact relationship to the total emergency effort.
( A.1.b & A.3) 3.
-The block diagram provided in the plan does not include all the organizations and suborganizations which have a role in the overall response and does not adequately illustrate.the interrelationships between those organizations and suborganizations which are included.
(A.1.c & A.3) 4.
The plan does not identify all the specific individuals by title who are in charge of the emergency response.
(A.1.d) 5.
The plan does not specify adequately the functions and responsibilities for major elements and key individuals by title including command and control, alerting and notification, communications, public information, accident assessment, public health and sanitation, social services, fire and rescue, traffic control, law enforcement, transportation, protective 7
response and radiological exposure control.
(A.2.a) 6.
The plan does not contain the legal basis for emergency response authorities l
by reference to specific acts, codes or statutes.
(A.2.b) 7.
The plan states that the written agreements with the various Federal, State
[
and local organizations having an emergency response role are provided in Appendix 1, but no written agreements are provided. These agreements are not described adequately in the plan and no signature page format is provided.
(A.3)
I L
H. ~.-. _ _
- - ~ -..
1 s
' 8.
The plan does not describe how each principal organization is capable of continuous (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) operations nor does it provide the specific title of individual for assuring the continuity of resources for the licensee, the States of Georgia and South Carolina and the local government agencies.
(A.4) 9.
The plan provides for a plume exposure pathway EPZ 5 miles in radius which is not in compliance with the regulatory requirements.
The basis for establishment of this 5 mile plume exposure pathway EPZ is not provided and the factors used in determining the exact size of this EPZ are not identified.
[10CFR50.47(c)(2)]
10.
The radiological response plans of State and local government within the plume exposure oathway EPZ are not provided in accordance with regulatory requirements.
[10CFR50.33(g)]
B.
Onsite Emergency Organization 1.
The plan does not specify the relationship of the normal plant staff for all shifts to the onsite emergency organization.
(B.1)
Cg 2.
The plan does not identify the specific conditions for higher level utility officials assuming the function of emergency director.
(B.3) the functional responsibilities The plan does not establish adequately (B.4) 3.
assigned to the emergency director.
4.
The plan does not specify the positions or titles and major tasks to be perforced by persons to be assigned to functional areas of emergency
'~
activity. Specific assignments are not made for all shifts and plant staff members both onsite and away from the site.
The plan does not demonstrate how the assignments shall cover the minimum on-shift staffing levels as indicated in Table B-1 of NUREG-0654 and it is not clear if the minimum on-shift. staffing meets this criteria.
The plan does not specify whether the augmentation of the on-shift staff will meet the recommended emergency staffing levels and does not indicate the exact size of the emergency staffing levels within the plant emergency organization.
(B.5 & Table B-1) 5.
The plan does not specify adequately the interfaces between and among the onsite functional areas of emergency activity, licensee headquarters support, and State and local government response organizations to determine if they will be effective. The block diagram provided does not illustrate adequately how these interfaces and interrelationships wil. operate.
(B.6) 6.
The plan does not specify the corporate management, administrative and technical support personnel who will augment the plant staff in the areas C-of logistics support, technical support for planning and reentry / recovery operations, management level interface with governmental authorities and
~
+
~
j -
release of information to the news media coordinated with governmental authorities during an emergency.
(B.7) 7 The plan does not specify all the local agencies or identify all of the services to be'provided by local agencies.
Copies of the arrangements and agreements reached with contractor, private and local support agencies delineating (authorities, responsibilities and limits are not appended to
.the plan.
B.9)
C.
Emergency Response Support and Resources 1.
The plan does not specify by title the individuals authorized to request Federal assistance.
(C.1.a) 2.
The plan does not specify the Federal resources expected, including expected arrival times at the vicinity of the plant site.
(C.1.b) 3.
The plan does not specify the command posts, telephone lines, radio fre-quencies and the telecommunications centers available from State and local 4
resources to support the Federal response.
(B.I.c)
~
4.
The plan does not provide for.the dispatch of Georgia Power Company representatives to principal offsite governmental emergency operations centers.
(C.2.b) 5.
The plan does not identify radiological laboratories, their general capa-bilities and expected availablility to provide radiological monitoring and analyses services which can be used during an emergency.
(C.3) 6.
The plan does not adequately identify nuclear and other facilities, organi-zations and individuals which can be relied upon to provide assistance during an emergency. Appropriate letters of agreement are not provided in Appendix 1 of the plan as indicated.
(C.4) i D.
Emergency Classification System 1.
The emergency classification system provided in the plan does not identify the specific instruments and does not always specify the necessary(parameter values for each emergency class for each emergency action level.
0.1) 2.
The plan does not describe why the initiating conditions in all the postu-lated accidents in the Final Safety Analysis Report (FSAR) are classified l
as meeting a specified emergency action level.
(D.2)
\\
i
7'
.m 4
p
~
_4 E.
Notification Methods and Procedures 1.
The plan does not state that notifications of offsite licensee, Federal, State and local officials will be verified or the method to be used for this verification.
(E.1) 2.
The plan does not provide sufficient information on the method to be used for alerting, notifying and mobilizing emergency plant personnel during nonworking hours and emergency corporate personnel to determine if the methods will be effective.
(E.2) 3.
The plan does not establish the contents of the initial emergency messages and whether the messages were developed in conjunction with State and local officials.
(E.3) 4.
The plan does not provide adequate information about the followup emergency messages to determine if they will include the information required by the criteria.
(E.4) 5.
The plan does not provide sufficient information on the administrative and i
physical means of notifying and providing prompt instruct'ons to the public
'* r within the plume exposure pathway EPZ to demonstrate that the system meets the criteria.
(E.6 & Appendix 3) 6.
The plan does not contain examples of draft messages intended for the public consistant with the Georgia Power Company's classification scheme giving instructions with regard to specific protective actions for specific locations within the plume exposure pathway EPZ.
(E.7)
F.
Emergency Communications 1..
The plan does not. provide for comunications with the local emergency operating centers.
(F.1.d) 2.
The plan does not provide the titles and alternates for those in charge at both ends of the communications links.
(10 CFR 50, Appendix E,lV..9) 3.
The plan does not specify the backup power source for onsite and offsite communications.
(10 CFR 50, Appendix E,IV.E.9) 4.
The plan does not provide for testing the comunications with the NRC Head-quarters and the NRC Regional Office Operations Center from the control room, TSC and EOF on a monthly basis.
(10 CFR 50, Appendix E,iV.F.9.d) e.
s
G.
Public Information 1.
The plan does not provide sufficient details on the types and topics of public information to be disseminated to determine if it will meet the criteria.
(G.1) 2.
The plan does not designate the point of contact and the spokesperson for the Georgia Power Company who will have access to all the necessary infor-mation for dissemination to the media.
(G.3.a & G.4.a) 3.
The plan does not provide sufficient information on the arrangements for the timely exchange of information among designated spokespersons to evaluate these arrangements against the criteria.
(G.4.b) 4.
The plan does not provide adequate details on arrangements for coordination in dealing with rumors.
(G.4.c)
H.
Emergency Facilities and Equipment 1.
The plan does not provide adequate details with regard to the TSC size,
. 3 location, layout, staffing, functions, habitability, equipment, data acquisition system and technical support to the control room and the E0F to determine if it meets the criteria.
(H.1) 2.
The plan does not provide adequate details with regard to the EOF and back-up E0F size, location, layout, staffing, functions, equipment, habitability, data acquisition system and evaluation and coordination of all Georgia Power Company activities to determine if it meets the criteria.
(H.2 & Supplement 1 to NUREG-0737) 3.
The plan does not provide sufficient details on the timely activation and staffing of facilities and centers to determine if the goals are met.
,~
(H.4 & Table B-1) 4.
The plan does not provide adequate information with regard to onsite monitoring systems and instrumentation to determine if the criteria are met for initiating emergency measures and conducting accident assessment.
(H.5) 5.
The plan does not provide sufficient information on provisions to acquire or access offsite data for emergency monitoring and analysis to determine if the access of this data meets the criteria. The plan does not discuss laboratory facilities other than to mention the normal plant laboratory.
(H.6) 1 6.
The plan does not provide adequate information on provisions for offsite
[
radiological monitoring equipment and instrumentation for emergency use s
7 to determine if the criteria are met.
(H.7) 7.
The plan does not provide sufficiient information on meteological diffusion and transport models and methods of projecting offsite radiological exposure to determine if the criteria are met.
(H.8 & Supplement 1 to NUREG-0737) 8.
The plan does not provide adequate details with. regard to the size, location,
~
icyout, staffing, equipment and instrumentation for the OSC to determine if it meets the criteria.
(H.9)
The plan does not provide sufficient informatiob on the reserves of instru-9.
mentation and equipment for emergency kits to determine if the criteria are met.
(H.10) 10.
The plan does not contain an appendix and the information is not provided on the identification and contents of emergency kits to determine if they meet the criteria.
(H.11) 11.
The plan does not contain provisions for a central point or the capability to receive, analyze and coordinate all field monitoring data other than the normal plant laboratory.
(H.12)
I.
Accident Assessment 1.
The plan does not identify in sufficient detail the plant systems, the effluent parameter values, the kinds of instruments and other information which will be used to characterize off-normal conditions and accidents in the plant to determine if the criteria are met.
(I.1) 2.
The plan does not provide adequate information on the capabilities and resources for accident assessment including post-accident sampling, effluent monitors, in-plant iodine and particulate measuring instrumentation and radiation monitoring in containment to determine if the criteria are met.
(I.2) 3.
The plan does not describe adequately the methods and techniques to be used for determining the source tenn for releases of radioactivity) within plant systems to evaluate whether they meet the criteria.
(I.3.a 4.
The plan does not provide sufficient information on methods and techniques for determining the magnitude of the radioactive release based on plant parameters and effluent measurements to evaluate whether they meet the
' cri te ria.
(I.3.b) 5.
The plan does not provide information to establish the relationship between effluent monitoring readings and onsite and offsite radiological exposures C.
and contamination for various meteorological conditions.
(I 4)
,M
6.
The plan does not provide an adequate description of the acquiring and evaluating of meteorological information and how this information will be used by the control room, TSC', E0F and State authorities to prov'ide radiological dose projections.- (I.5 & Supplement 1 to NUREG-0737) 7.
The plan does not provide the methodology to be used for determining release rates and dose projections if the instrumentation used for these assessments are offscale or inoperable.
(I.6) 8.
The plan does no,t describe the capability and resources for field monitoring within the plume exposure pathway EPZ.
(1.7) 9.
The plan does not describe the methods, equipment, expertise, activation, field team composition, transportation, monitoring equipment and estimated deployment times for making rapid assessments of the magnitude and locations of radioactive liquid or gaseous releases.
(I.8)
- 10. The plan does not indicate whether }he capability to measure radiciodine concentrations in air as low as 10- iiiferocuries per cubic centimeter in the presence of high levels of noble gases will be provided under field 1
conditions and how these measurements will be obtained.
(I.9)
V
- 11. The plan does not describe adequately the means for relating various measured parameters to dose rates for key radionuclides and gross radio-activity measurements to estimate projected and actual dose rates and for comparing them with the PAGs.
(I.10)
J.
Protective Response 1.
The plan dose not provide sufficient details on evacuation routes, assembly areas, transportation for onsite personnel, alternative evacuation routes and offsite relocation areas to determine if they meet the criteria.
(J.2) 2.
The plan does not provide an adequate description of the radiological monitoring methods and techniques to be used on personnel evacuated from i
the site to determine if the criteria are met.
(J.3) 3.
The plan does not describe the redological decontamination capability for personnel evacuating the site.
(J.4) 4.
The plan does not provide an adequate description of the accounting methods and procedures to ascertain the names of missing individuals onsite at the start of an emergency, to find the missing individuals if any, and to account for all onsite individuals thereafter.
(J.5)
~
1
.,Y 5.
The plan does not describe in sufficient detail the provisions for individual respiratory protection, protective clothing end the use of thyroidal blocking techniques and radioprotective drugs for individuals arriving or remaining onsite during the emergency.
(J.6) 6.
The plan does not describe adequately the mechanism for recomending protective actions to be appropriate State and local authorities to determine if it meets the criteria.
(J.7) 7.
The plan does not contain evacuation time estimates for the plume pathway EPZ.
(J.8 & Appendix 4)
- 8.
The maps provided in the plan are illegible and do not provide the informa-tion on evacuation routes, evacuation areas, preselected radiological sampling and monitoring locations including designators, relocation centers, shelter areas, and the population distribution by evacuation areas and sectors.
(J.10.a & J.10.b) 9.
The plan does not describe adequately the means for notifying all segments
(^z of the transient and resident population.
(J.10.c)
U 10.
The plan does not provide the bases for the choice of recommended protective actions for the plume exposure pathway EPZ, the expected protection afforded by residental units or other shelter for direct and inhalation exposure and evacuation time estimates.
(J.10.m)
K.
Radiological Exposure Control 1.
The rationale is not given for the establishment, of the emergency personnel exposure guide values provided in the plan so that their adequacy can be evaluated.
The values for nonlifesaving assessment actions and nonlife-saving first aid appear to be excessive and the lifesaving ambulance service
-and medical treatment values appear to be overly restrictive.
2.
The plan does not provide sufficient details about the onsite radiation protection program to be implemented during emergencies to determine if it meets the criteria. The plan does not identify individual (s) by position or title who can authorize emergency workers to receive doses in excess of 10 CFR, Part 20.
The plan does not contain the procedures for permitting onsite volunteers to rec'eive radiation exposures while carrying out life-saving and other emergency activities or the procedures for making these decisions and estimating the relative risks.
(K.2)
~
3.
The plan does not provide for 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day dose assessment capability for emergency personnel and does not describe adequately the distribution and sensitivity range of personnel dosimetry.
(K.3.a)
k 4.
The plan does not indicate the frequency for reading dosimeters or the methods for maintaining dose records for emergency workers involved in a nuclear accident.
(K.3.b) 5.
The plan dec; not specify the action levels for determining the need for radiological decontamination during emergencies.
(K.5.a) 6.
The plan does not describe the means or methods for radiological decon-tamination of personnel, wounds, supplies, instruments and equipment and provisions for waste disposal during emergencies.
(K.5.b) 7.
The plan does not specify the measures for onsite radiological contemination control, area access control, use of food and drinking water supplies and the criteria for returning areas, items and equipment to normal use.
(K.6) 8.
The plan does not provide for the decontamination of relocated onsite personnel or the provisions for extra clothing and methods of decon-tamination.
(K.7)
L.
Medical and Public Health Support
,s 1.
The plan does not provide for a specific backup hospital and does not describe the capability for evaluation of radiation exposure and uptake of radioactive materials. The plan does not describe adequately the pre-paration or services to handle radiologically contaminated personnel who are injured to determine if the criteria are met.
(L.1) 2.
The plan does not provide sufficient detail on the onsite first aid capability to determine if it meets the criteria.
(L.2) 3.
The plan does not describe the arrangements for the transport of radio-logical accident victims other than to state that this service will be provided by the Buke County Ambulance Service.
(L.4)
M.
Recovery and Reentry Plenning and Post-Accident Operations 1.
The plan does not provide sufficient details on the plans and procedures for recovery and reentry to determine if the criteria are met.
The plan does not describe the means for making decisions to relax protective measures as needed to carry out recovery operations.
(M.1) 2.
The plan does not provide the authority and responsibilities of the individuals who will fill the key positions in the recovery organization a.nd does not indicate who these key individuals will be by~ position and title.
The organizational chart of the recovery team does not provide sufficient detail to determine how the organization will function and is incomplete in that all the components are not included (e.g., Recovery Review Board).
(M.2)
((,
3.
The plan does not provide adequate detail on the methods to be used to estimate total population dose to determine if they meet the criteria.
The plan does not provide for the continuation of total population ~ dose estimates during the reentry and recovery phase of operations.
(M.4)
N.
Exercises and Drills 1.
The plan does not describe the methods and procedures for conducting emergency exercises in sufficient detail to determine if they will meet the criteria. The plan does not state that exercises shall be conducted in accordance with current NRC and FEMA rules.
(N.1.a) 2.
The plan does not provide for unannounced exercises.
(N.1.6) 3.
The plan does not describe how communications drills will be conducted to test the understanding of the contents of messages.
(N.2.a) 4.
The plan does not describe how fire drills will be conducted or the methods used to ensure that they comply with the technical specifications.
(N.2.b) g 5.
The plan does not describe how medical emergency drills will be conducted
/
to simulate the trea'tment of a contaminated individual.
(N.2.c) 6.
The plan does not describe how radiological monitoring drills will be conducted to test the collection and analysis of sample media, comunica-tions and record keeping.
(N.2.d) 7.
The plan does not describe how health physics drills will be conducted to test the analysis of airborne radioactivity measurements, liquid samples and direct radiation measurements. The plan provides for these drills on an annual rather than a semi-annual basis as indicated by the criteria.
[N.2.e.(1)]
8.
The plan does not provide for the testing of the analysis of elevated radioactivity levels using the post-accident sampling system.
[N.2.e(2)]
9.
The plan does not describe how exercises and drills will be carried out to allow free play for decisionmaking and how they will be evaluated to deter-mine if the objectives are met.
(N.3)
- 10. The plan does not provide sufficient details on the management control to be used to ensure that corrective actions resulting from the evaluation of drills and exercises will be implemented.
(N.4)
O e
O.
Radiological Emergency Response Training 1.
The plan does not provide for practical drills to demonstrate the a'bility of emergency personnel to perfonn assigned functions with on-the-spot correction of erroneous performance.
(0.2) 2.
The plan does not describe adequately the scope and nature of the training
~
program to determine if it meets the criteria.
The curriculum for the various types of training programs, all the specific categories of personnel receiving specialized training and with the exception of the fire brigade, the frequency of training is not provided.
(0.4)
P.
Responsibility for the Planning Effort:
Development, Periodic Review and Distribution of Emergency Plans 1.
There is no provision to date and mark the pages where the emergency plan has been revised.
(P.5) 2.
The plan does not provide a detailed listing of supporting plans and their sources.
(P.6)
.,y 9
3.
The plan does not provide for the updating of telephone numbers in the emergency procedures on a quarterly basis.
(P.10)
- b y
W
._ -