ML20038C463
| ML20038C463 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 11/17/1981 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | Davidson D CLEVELAND ELECTRIC ILLUMINATING CO. |
| References | |
| NUDOCS 8112110079 | |
| Download: ML20038C463 (20) | |
Text
- - = -.
Di ribution:
5 cket File (2)
Perry Std.
LB 02 J..!tauch"- ICSB I
DEisenhut/RPurpid ~' F. Rosca.- IC$B ;
~'
RTedesco 4
,N i
ASchwencer tiOV 1 77d ig1 11DHouston x
Docket Hos: 50-440 liService au g
and 50-441 I&E (3) y bcc:
/
k Local PDRg' h 4
flRC PDR
!!r. Dalu/n R. Davidson D-Vice President - Engineering gig NSIC The Cleveland Electric Illu71nating Conpany Post Office Box S000 9
- f.,
TERA TIC P
Cleveland, Ohio 44101 ACRS (16) 4 i
/
Dear "r. Davidso'i:
Subj ect: Request for tieeting - Instrtmentation and Control Systcqs In the perfornance of the Perry licensing review, the staff has identified concerns in regard to instrumentation and control systems. To expedite t
the licensing review, the staff and their consultant request a series of j
neetinas with your staff to discuss these concerns. The information that they wish to discuss is identified in the enclosure.
We suggest that you atte1pt to grobo the items in the enclosure in convenient sets such that each set can be discussed at an individual neeting lasting between one and five working days. Ve also suggest that each individual meeting include the miniquo nunber of participants necessary to fully discuss the topics to be covered. However, you should be prepared to discuss the pertinent details of fluid systels and nechanical equipment with which the instrumentation and controls interface. As can seen fro 7 the attached list,
[
many of the itens are related to the interface areas.
It would be useful if we could be provided with drawing nuabers (and the draw-ings if not already sunbitted to us) of drawings to be used by Perry personnel for discussion of each iteai.
If possible, we uauld like to have this infor-l aation two weeks in advance of the meeting where the drawing will be discussed.
The staff has proposed an initial neeting with you during the week of December 7 or Dec 9ber 14 at a facility specified by you. Af ter reviewing the enclosure, please notify it. Dean Houston, Project ihnager, at (301) 492-8430 of your j
aceting plans.
Sincerely, l
A. Schwencer, Chief bft $$ h o!$$$ 40 A
Division of 1.icensing pg
Enclosure:
Proposed Agenda for !betino omer > D,1;:LB #2 DLc1 MDEF, aa......As '
cer sun, - >
1.lEZl.8)........Wf I.8)....... s................
om>
nac ronu ais ciocon tencu o24o OFFICIAL RECORD COPY uso m ini e.oo
~
'N
..s Mr.' Dalwyn R'. Davidson Vice President, Engineering-The Cleveland Electric 111uminating Company i
P. O. Box.5000
- Cleveland, Ohio 44101
,cc:'IGeraldCairnoff,Esq.
.'Snaw, Pittmen, Potts & Trewbridge
', 1800.M Street, N. W.
Was.hington, D. C. 20006
?-
Donald H. Hauser, Esq.
The Cleveland Electric Illuminating Company P; 0. Box 5000 Cleveland, Onio 44101 Resident Inspector's' 0ffice Da.N.R.C.'
Parmly,a.t Center Road Perry,',0hio 44081 Donald'T. Ezzone, Esq.
'is si s tpr. Prosecuting. Attorney 105 Mein Street
~
Lake Coanty Acainistration Center
, Painesvilla, Ohio 44077 4
Toa J. Kenney 228 South College, Apt. A Bowling Green, Ohie 43402 Daniel D. Wilt
', Wegman, Hesiler & Vanderberg
?201 Chippewa Road, Suite 102 i
i 3recksville, Ohio 44141 Robert Alexander CCRE Interim Representative 2030 Portsmouth Street Suite 2 Houston Texas 77098 i
Terry Lodge, Esq.
915 Spitzer Building Toledo, Ohio 43604
PROPOSED AGENDA FOR MEETING (S) WITH PERRY APPLICANT ON INSTRUMENTATION AND CONTROLS Following is a list of items for discussion at one or more meetings with the applicant to provide the NRC staff with information required to ulderstandl the design bas,es and design implementation for the instrumentation and con-trol systems on the Perry project. The applicant should be prepared to use detailed instrumenti control, and fluid system schematics at the meetings in explaining system designs and to provide verification that design bases and regulatory criteria are met. During the meetings, specific items requiring additional documentation in the FSAR will be identi.fied.
Several items where additional documentation is likely to be required are included in the list which follows.
In the review of Chapter 7 of the Perry FSAR, it was apparent that the material presented was considerably abbreviated in comparisen with the discussions provided in Chapter 7 for other plants, i.e., Grand Gulf and Clint on.
In particulare very little information was provided as to how individual systems conform to General Design criteria, Regulatory Guides, and other applicable criteria. This lack of: material has made the review of Chaster 7 of the Perry FSAR quite difficult. Before we complete the review, we will require concise information as to how each of the systems meet the applicable criteria such as the General Design Criteria, Regulatory Guides, IEEE Standards, and Branch Technical Positions. All exceptions to the applicable criteria must be identified.
u -
421.01 Section 7.1 of the FSAR contains no references to the branch technical positions listed in Table 7.1 of the Standard Review P l an.
The FSAR should identify and justify any exceptions taken to these branch technical positions. Alsor Regulatory Guide 1.47 is repeated twice in Table 7.1 whereas Regulatory Guide 1.75 is not in c lude d.
421.02 Several previously reviewed BWR installations, e.g., Grand Gulf, included a start-up transient monitoring system to provide record-ings of selected parameters during the start up and warranty testing. There is no information in the FSAR which describes this system.
If this systemi or any similar systems is intended for use in the Perry unitse provide the following information:
Identify all safety-related parameters which will be monitored a.
with the transient monitoring system during start up.
b.
For each safety parameter identified abover provide a concise description of how the associated circuitry merges or connects (either directlyr or indirectly by means of isolation devices) with the circuitry associated with the transient monitoring system. Where appropriater supplement this description with detailed electrical schematics.
c.
Describe provisions of tha design to prevent failures of this system from degrading safety related systems.
421.03 Various instrumentation and control system circuits in the plant (including the reactor protection systemi engineered safety features actuation system, instrument power supply distribution systea) rely on certain devices to provide electrical isolation capability in order to maintain the independence between redundant safety circuits and between safety circuits and non safety circuits. Thereforer provide the follcwing information:
a.
Identify the types of isolation devices which define the Class 1E boundary for interfaces between the safety circuits.
b.
Provide the acceptance criteria for each isolation device identified in response to part a above.
c.
Describe the type of testing that was conducted on the isolation devices to ensure adequate protection against EMI (i.e., noise)r short-circuit failurese voltage f aults, and/
or surges.
421.04 For several of the protection systems, Table 7.1.3 does not verify compliance with General Design Criteria 19, 20, 21, 22, 23, 24, and 29 as required by Table 7.1 in the Standard Review Plan. Provide justification as to why these criteria do not apply if, indeeds they do not.
421.05 Discuss the design provisions for conducting respense time tests in accordance with R. G. 1.118. Identify safety related systems that do not have provisions for response time testing.
Discuss the
techniques to be used to periodically measure safety related sensor time responses.
~
421.06 Discuss conformance with the following TMI action items as required by NUREG-0737:
a.
II.D.3 - Relief and safety valve position indication b.
II.F.2 - Inadequate Core Cooling c.
II.K.3.18 - ADS actuati on d.
II.K.3.21 - Restart of LPCS and LCPI e.
II.K.3.22 - RCIC automatic switchover 421.07 Describe the secaration criteria for protection channel circuits, protection logic circuits, and non-safety related circuits.
For example are channel circuits and logic circuits separated from one anothe r?
421.08 Table 7.1-2 of the FSAR states that the design of the Perry Reactor l
Protection System is similar to the design of the Grand Gulf f
l Reactor Protection System. Provide a comparative discussion l
identifying specific differences between the two designs.
1 l
421.09 Identify where instrument sensors or transmitters supplying infor-mation to more than one protection channel, to both a protection channel and control, or
. to Dore. than one. control' channel, are located in a common instrument line or connected to a common instrument
[~
_~
tap. The intent of this item is to verify that a single failure in a common instrument line or tap {such,as break or blockage) cannot defeat required protection system redtndancy.
421.10 Provide an evaluation of the effects of high temperatures in reference legs of water level measuring instruments subsequent to high energy line breaks.
421.11 Pevise the discussion concerning compliance with IEEE Standard 279 to verify that all portions of the RPS comply.-
421.12 In the discussion in Section 7.2.2.2 concerning conformance to Criterion 4.15 of IEEE Standard 279, the statement is made that there are no multiple setpoints within the RPS.
Discuss the effects on RPS setpoints of mode switch cperation.
421.13 Discuss the logic used for bypassing the tt:rbine stop valve closure.
Can a single failure in this pressure transmitter system cause a bypas's of this closure to occur. Provide a similar discussion for the turbine control valves.
421.14 A discussion of the Equipment Protection Assembly (EPA) systems is not given in FSAR Section 7.2.
Also, the EPA asser.clies are not shown in Figure 7.2-1, the reactor protection system instrumentation and control diagram.
Discuss the EPA system and how it meets IEEE 279.
---______.____-__m_
421.15 Discuss the differential pressure transmitters that are used to monitor reactor vessel high water level (trip level 8) and the low water level trips.
Discuss the b5 passing of the react,or vessel high water level trip in all operating modes but run. Does this bypass the low level trip at all operating 6
modes but run?
421.16 The statement is made that prudent operational limits for each safety related variable trip setting are selected with sufficient margin so that a spurious scram is avoided. Please provide a detailed discussion and/or reference to the methodology used in determining safety system setpoints.
421.17 Identify any sensors or circuits used to provide input signals to the protection system which are located or routed through non-seismically qualified structures. This should include sensors or circuits providing input for reactor tripi emergency safeguards equipment such as the Emergency Core Cooling system, and safety grade interlocks. Verification should be provided that the sensors and circuits meet IEEE 279 and are seismically and environmentally qualified. Testing or analyses performed to insure that failures of non-seismi c structurese mountings, etc., will not cause f ailures which could interfere with the operation of any other portion of the protection system should be discussed.
421.18 Identify the physical location of the equipment that actuates the reactor trip on turbine trip and indicate whether this
~
equipment and associated circuitry meets the criteria applicable to equipeent performing a safety function.
421.19 Verify that a failure modes and effects analysis has been performed for each of the ESF systems identified in Section 7.3.1.
421.20 It has been noted during past reviews that pressure switches or other devices were incorporated into the final. actuation control circuitry for large horsepover safety-related motors which are used to drive pumps. These switches or devices preclude automatic (safety signal) and manual operation of the motor / pump combination unless permissive conditions, such as tube oil pressure, are satis-fied.
Accordingly, identify any safety related motor / pump combina-tions which are used in the Perry design that operate as noted i
above. Alsor describe the redundancy and diversity which are i
provided for the pressure switches or other permissive devices that are used in this manner.
421.21 Discuss the testing procedures for the pilot solenoid valves which l
control compressed air to the ADS relief valves.
i 421.22 The FSAR st.ates that each ADS trip system has a time delay that can i
be reset manually to delay system initiation. Discuss the conditions l
under which the operator would reset the ADS timers. Alsor discuss the consequences of resetting the timers if the HPCS fails to start.
l s-
~~.
421.23
, From' the discussions provided in Sections 6.3.2.2.3, 6. 3.2.2.4,
- 7. 3.1.1.1. 3, an d 7.3.1.1.1. 4, it is not clear whether or not the LPCS and LPCI injection valves are intert'ocked to prevent them from opening unless reactor pressure is low enough for injection to be possible. Provide more information concerning the operation of these valves.
Also, there are discrepancies in the FSAR as to whether dif ferential or gage (absolute) pressure transmitters are used for the interlocks. For example, Secti on 7.3.1.1.1.3 and 7.3.1.1.1.4 imply differential pressure transmitters are used. However, the P&I diagram for the LPCS system, Figure 6.3-8, does not show a differential pressure transmitter near the injection valve.
421.24 Discuss the tesing procedures used to demonstrate that the main steam isolation valve closure time is within the 3 to 5 seconds assumed in Section 15.2.4.3.2.
l 421.25 Discuss how the Main Steamline Isolation Valve Leakage Control System conforms to the requirements of Paragraph 4.1 of IEEE Standard 279 concerning automatic initiation capability.
I
)
421.26 Discuss how the Suppression Pool Cooling Mode of the Residual
\\
Heat Removat System conforms to the requirements of Paragraph 4.1 i
of IEEE Standard 279 concerning automatic initiation capability.
- - - ~~~
e
421.27 The initiating conditions for automatic operation of the RHRS-Containment Spray Cooling Moder discussed in Section 7.3.3.1 of the Construction Permit SERr Gere mot acceptable to the staff. Discuss how the present design complies with the recommendations Listed in the Const ruction Permit SER.
421.28 Discuss how the RHRS-Containment Soray Cooling Mode initiation system conforms to Paragraoh 4.17 of IEEE Standard 279-1971 conce rning manual initiation capability.
Discuss initiation of both loops.
' 421.29 Can the automatic initiation of the emergency recirculation mode of the Control Complex HVAC system be bypassed? If sor des c ribe the design features.
421.30 The P&I diagrams for the Annutus Exhaust Gas Treatment System are shown in Figure 6.5-1, Sheets 1 and 2.
However, Sheets 1 and 2 appear identical except for the FDIB and the FDRB valves. Explain the significance of the two drawings.
421.31 Discuss how the design of the Annulus Exhaust Gas Treatment System conforms to Paragraph 4.11 of IEEE Standard 279, concerning channet l
l bypass or removal f rom operation for purposes of maintenance or testing.
i i
r i
.. ~
.+- -.
1 421.32 As stated in Section 6.2.5.5 of the FSARs proper recombin-er operation after an accident is determined by monitoring a watt meter in the control complex. However, it is important that, the recombiner temperature be maintained above 1150 F for proper ope ration. Discuss why temperature measurements are not used to indicate proper eperation. IEEE 279 criteria require direct measurement of the desired variabler when practical.
421.33 Have hydrogen analyzers, qualified to IEEE ?23 and 344, been procured?
421.34 Table 7.1-2 of the FSAR identifies many ESF systems that are similar to the design of the Grand Gulf ESF systems. Provide a comparative discussion identifying specific differences between designs of similar systems.
421.35 Demonstrate that the Safety Relief Valve (SRV) Low-low setpoint function is adequate assuming a single failure.
421.36 Describe the electrical power supply arrangement, air supply design featureci and any interlocks associated with control. and operation of the safety relief valves.' This should include a discussion of the design bases for the capacity of air reservoirs used to operate the valves.
421.37 Identify the safety related instrumentation used in the suppression pool makeup system and discuss the routing of the safety related ci rcui t ry.
The discussion should identify, the physical location of the transmitters.
421.33 Using detailed system schematics, describe the sequence for periodic testing of the:
a) main steamline isolation valvess b) main feedwater isolation valves, c) main feedwaterocontrol valves (safety features),
d) RCIC system The discussion should include features used to insure the availability of the safety function during test and measures taken to insure that equipment cannot be lef t in a bypassed condition after test completion.
421.39 Demonstrate that the containment isolation system satisfies the single failure criterion and that the redundant instrumentation and control systems provided meet Branch Technical Position ICSB No; 3.
421.40 Clarify the descriptions of the temperature monitoring circuits which l
l initiate containment and reactor vessel isolation. Describe how the system satisfies the requirements of IEEE Standard 338, R. G.1.22
/
and GDC 21.
l l
. _., _.. _. _ ~,
421.41 The statement in Section 7.3.1.1.1.1 that the HPCS provides makeup water to the reactor until the vessel water level reaches the high level (trip level 8) conflicts with the s' tate' ent in Section m
6.3.2.2.1 regarding the HPCS system. Please indicate which discussion is correct.
421.42 Discuss the statement that the HPCS pump motor and injection valves are provided with manual override controls. Does this violate the concern expressed in IE Bulletin 80-06? The statement is made in Section 7.3.1.1.1.4 that once initiatedi the LPCI logic seals-in and can be reset by the control room operator only when initial con-ditions return to normal. This seems to conflict with a statement in the same section that indicates that the operator can manually control the system subsequent to automatic initiation. Explain this conflict.
421.43 Can the manual control switches provided for the ADS safety / relief valves initiate the system without the low pressure pumps operating?
l l
421.44 Can the LPCS suction valve be closed upon containment isolation if keylocked open?
t 421.45 The statement is made that the suppression pool suction valve auto-maticalLy opens if a high water leve t is detected in the suppression pool.
Does this imply that valve F1 will close automatically at this time?
l l
421.46 In the discussions concerning the design of the Standby Liquid Control System in FSAR Sections 9.3.5 and 7.4.1.2, very little information is provided concerning the desisn' of the heating system required to prevent precipitation of the sodium pentaborate from the solution during storage.
Provide a more detailed discussion en the design of the heating system, including associated instrumentation and controls.
Include information concerning the power sources used for the instrumentation and heaters and any alarms used to indicate failure of the heating system.
421.47 Please discuss the following er Lative to the remote shutdewn system $
Design basis for selection of instrumentation and control a.
equipment on the remote shutdown pane L.
b.
Location of manual transfer switches a.d remote shutdown panel (include layout drawings, etc.).
Design criteria for the remote shutdown panet instrumentation, c.
including manual transfer switches.
d.
Description of control of access to the displays a1d controls i
Located outside the control room.
i e.
Description of isolation and separation provisions.
This should include the design basis for preventing electrical i
i interaction between the control room and remote shutdcwn equipment.
1.
Description of control room annunciation of whether devices are under Local or remote control.
Description of any communication systems required to coordinate g.
operator actions.
, Wye 4 e h4.
g h as
h.
Description of testing to be performed to verify the capability of maintaining the plant in a safe shutdown condition f rom outside 15e control room.
421.48 General Design Criterion 19 requires potential capability for subsequent cold shutdown of the reactor.through the use of suitable p ro ce dure s.
Provide a summary of the procedures used to achieve cold shutdown from outside the control room.
Include a list of the systems required for cold shutdown from outside the control room and the location of the panels where these system controls are housed.
Show how the cold shutdown procedures meet the single f ailure criterion.
421.49 The staf f has recently issued Revision 2 to Regulatory Guide 1.97,
" Inst rumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident."
This revision reflects a number of major changes in post-accident inst rumentationi and includes specific implementation requirements for plants in the operating License review stage.
Discuss the schedule for complying with this Regulatory Guide.
421.50 Using detailed system schematics, describe the implementation of the bypassed and inoperable status indication provided for engineered safeguards features. Discuss how the design of the bypass and inoperable status indication systems comply with positions 81 through s
B6 of Branch Technical Position ICSB No. 21.
- = -_
l 421.51 Explain why the discussions given in Section 7.5 identify measured variables which are not listed in Table 7.5-1 as providing verification of proper system operation..
421.52 Why are the indicator lights or anntsiciators required by R. G.1.47 included in Table 7.5-1 or any other table in Section 7.5?
not i
421.53 This section i;)dicates that the suppression pc,ol water level and tempe rat ure r as well as the dryweLL and containment pressure and recorders and indicators sill. not perform during and tempe rat ure r af te r a seismic event eveh though these inst ruments are required for post accident monitoring. Provide a justification for the design basis of this instrumentation.
421.54 During normal and emergency conditions it is necessary to keep low pressure systems that are connected to the high pressure coolant systems preperty isolated in order to avoid damage by overpressurizatiois or the potential for loss of integrity of the low pressure systems and l
l possible radioactive releases. Discuss how each of the low pressure t
to high pressure interfaces in your design conform to the requirements of Branch Technical Position ICSB No. 3.
Alsor discuss how the l
associated interlock circuitry conforms to the requirements of IEEE Standard 279. The discussion should include illustrations from applicable draw'ings.
I l
1
- - -.. -. ~.--
- - ~
.--~ ~ ~ ~.., _
n.-.-.
n.
.. ~.. ~. -
t 421.55 Are the positials of the check valves used to interface between the low pressure and high pressure portions of the residual heat removal system annunciated in the control room? Acco'rding to Figure 5.4-13 (Shegt 2)r check valves F019 and F050 do not have their position ann un ci ate d.
How does the design conform with Position 4 of Branch Technical Position IC3B No. 3.
421.56 Describe the periodic self-test mode of the Rod Pattern Control System.
421.57 In the discussion ca,cerning the leak detection instrumentation for fission product monitoring, Section 5.2.5.2.1, reference is made to Section 7.6.
However, no infor=ation is provided in Section 7.6 relative to fission product monitoring.
Discuss this instrumentation and the need to include a description of it in Section 7.6.
421.58 BWR cperating experience has shown that the RHR and RCIC systers have been rendered inoperable because of inadvertent leak detection isolations caused by equipment room area high differential temperature signals. The events occurred when there was a relatively sharp drco in outside temperature.
As noted in Section 7.6.1.3, the Perry N'uclear Power Plant incorporates this type of RCIC and RHR (steam) 'is6lation. Provide a discussion of any modifications that have been or wiLL be made to prevent inadvertent isolations of this type which af fect the availability and reliability of the RCIC and the RHR systems.
. - - ~.
421.59 Several major plant control systems whose functions are not essential for plant safety, i.e., NSSS process computer systemi reactor water cleanup system, gaseous radwaste system, and process sampling systems are not discussed in Section 7.7.
Alsor the discussion on non safety leak detection systems is incomplete.
Discuss the need to amend Secticn 7.7 of the FSAR to include descriptions of the above systems.
~
421.60 Provide a comparison of the systems discussed in Section 7.7 with those of similar approved plants. Identify systems of new design and differences in systems with similar designs.
421.61 Has the Perry design for increasing control rod insertion rate incorporated the prompt relief trip concept or the fast scram concept?
421.62 Discuss the safety aspects of the Perry design for the following trips and interlocks:
a.
Recirculation flow control valve motion interlocks; b.
Low reactor vessel water level and high vessel pressure recirculation pump trips; c.
High reactor vessel water level trips for the feedwater pumps and plant turbine.
421.63 Identify the non-safety grade ecuipment used to mitigate the effects of Anticipated Transients Without Scrar-(ATWS).
Include a discussion of the ATWS recirculation pump trip (ATWSRPT).
wwm+-
e.
.w
=----a,.
. em.-
.. - =
.g=
w-
= - - -+
- is -
421.64 By NRC letter to you dated May 11, 1981, we transmitted questions concerning four ICSB generic issues. Discuss your schedule for formally answering these generic issues as specified in the above let te r.
a
/
9 4
..