ML19294B221
| ML19294B221 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 02/13/1980 |
| From: | Baer R Office of Nuclear Reactor Regulation |
| To: | Stampley N MISSISSIPPI POWER & LIGHT CO. |
| References | |
| NUDOCS 8002280012 | |
| Download: ML19294B221 (10) | |
Text
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a UNITED STATES y
NUCLEAR REGULATORY COMMISSION j
j WASHINGTON, D. C. 20555 e
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FEB 131980 Docket Nos. 50-416 and 50-417 Mr. N. L. Stampley, Vice President Production and Engineering Mississippi Power and Light Company P. O. Box 1640 Jackson, Mississippi 39205
Dear Mr. Stampley:
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION (Grand Gulf Nuclear Station, Units 1 and 2)
As a, result of our review of the information contained in the Final Safety
" Analysis Report for the Grand Gulf Nuclear Station, Units 1 and 2, we have developed the enclosed requests for additional information.
Included are questions from the Instrumentation and Control Branch concerning Sections 7.2 and 7.3.
We request that you amend your Final Safety Analysis Report to reflect your responses to the enclosed requests by April 30, 1980.
If you cannot meet this date, please advise us of the date you can meet as soon as possible so that we may consider the need to revise our review schedule.
Please contact us if you desire any discussion or clarification of the enclosed requests.
Sincerely, 6&edx.
Robert L. Baer %
, Chief Light Water Reactors Branch No. 2 Division of Project Management
Enclosure:
Requests For Additional Information ces w/ enclosure:
See next page 8002280Q{g
Mr. N. L. Stampley FE6I3$$$$
Mr. N. L. Stanpley Vice President - Production Mississippi Power and Light Canpany P. O. B ox 1640 Jackson, Mississippi 39205 ces: Mr. Robert B. McGehee, Attorney Wise, Carter, Child, Steen and Caraway P. O. Box 651 Jackson, Mississippi 39205 Troy B. Conner, Jr., Esq.
Conner, Moore and Corber 1747 Pennsylvania Avenue, N. W.
Washington, D. C.
20006 Mr. Adrian Zaccaria, Project Engineer Grand Gulf Nuclear Station Bechtel Power Corporation Gaithersburg, Maryland 20760 I
d.
ENCLOSURE QO30.56 The following discrepancies between the actual plant drawings and
[* *' (7.2)
,the FSAR discussion and figures were noted in the review of Section (F7.2-2) 7.2.
(F7.2-3) 1.
The channel 1est Switch shown in Figure 7.2-5 does not appear (F7.205) on the Reactor Protection System Elementary Diagram C71-105C (F7.2-8)
(Revision 7).
This test switch is mentioned a ntnber of times gg (Documents in Section 7.2.2 and is offerred as a " backup to the manual S
C71-1010, scram." The switch is also identified as the first contact in 1050, 1070, the trip channel logic, and 4010) 2.
Section 7.2 and Figures 7.2-5. 8 indicate that there are two isolation valves in each steam line and that they are y,,
connected into the scram logic so that valves in at least three systems must be closing to generate a scram. The trip is key bypassable in all operating. nodes but "Run".
Drawing C71-1050 (sheets 5 thru 9) confirn this trip circuit but also shows a separate unbypassable trip (not mentioned in Section 7.2) connected in 1 out of 2 twice logic that is derived from a third isoletion valve in each steam line. The presence of a third valve in each steam line is verified by the P & I Diagrams in Section 5.2.
3 Section 7.2.1.2.8 indicates that there are no RPS pressure transmitters inside the drywell; however. Drawing C71-1050 (sheet 17) locates the transmitters for reactor pressure, dry-well pressure, reactor level and scram discharge voltme level inside the drywell. Drawing C71-1070 (Hevision 1) agrees with Section 7.2.1.2.8.
4 The FSAR states that both the turbine stop va)ve and the pt
.' turbine control valve protective system trips utilize pressure E
- transmitters that are sensing hydraulic pressure for the valve operating mechanism. The trip point for the two systems are I
indicated to be approximately 40 psig with normal system pressures of 42-70 psig (control valve) and 165 psig (stop valv e).
The system drawings and specifications you provided indicate that the trip signal for the stop valve is generated by position switches mounted on the valve (some of the FSAR discussion appears to corroborate this design).
The drawings indicate that the control valve trip is initiated by a pressure switch. The design specification data sheet indicates that normal hydraulic pressure is 1100-1500 psig and that the trip point is 850 psig.
5.
The FSAR states that identification of the specific channels that tripped is obtained from ccmputer typeout or by visual observation of the relay contacts (7.2.2.1.2.3.1.19 ).
The contact positions could not be observed on the relays in the bp ?,
trR unit cabinets that we were shown. Verify that the
.g contact status of relays associated with RPS trip units can be readily observed.
6.
Drawing C71-1050 indicates that armino of the mnual scram is an alarm condition and implies, therefore, that the manual scram is normally disarmed. The FSAR, Section 7.2.1.1.4.2, indicates
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I V, that the manual scram switches in each group are " located close enough to permit one hand motion to initiate a scram", implying that switches are normally armed.
Resolve the discrepancies noted alave and correct the appropriate
m '.
i 3-docunent.
Verify that the instrunents and controls described in
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{p 61, of Chapter 7 are the systems that are being installed.
In the case of the manual scran pushbuttons, justify the use of armed pushbuttons and the logic of operating with a disarmed system.
Include in Section 7.2.1 a complete description of the steps in the actuation of a manual scram, in each reactor mode if any differences exist between modes.
-;., t QO30.57 Section 7.1.3 states that pressure and level transmitters were (7.0) provided so that the improved reliability would not require testing of sensors except at the end of cycles. Sections 7.2 and 7.4 discuss testability and testing of transmitters for specific protective functions and state that tra",ncitters can be and/or are
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valved out and tested during operation. Clarify your position on transmitter testing frequency and revise the FSAR as necessary to provide consistency.
It is the Staff's position that the potential of valving errors disabling a protective channel
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outweighs any advantage of simulating a trip level signal to the
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tl transmitter input, provided that the normal transmitter input can e
b
- 'k be perturbed sufficiently during normal reactor operation to d'emonstrate that the transmitter is reading and responding correctly. Demonstrating the operability of switches and trip units still requires simulating the trip level input to these F
devices.
0030.58 Justify the claim that the containment spray cooling can be (7 3.1.1. 4.1) manually actuated when the drawings indicate that the manual
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i (7.31.1.43) pushbutton must be held depressed continuously for 90 seconds to (h,i HPL E12-1050 in1,tiate spray B.
Identify any other manual pushbuttons that must be held depressed for more than a few seconds to initiate the desired action.
QO30. 59 Justify the claim for diversity in the control circuit for (7. 3.1.1. 4.7 ) containnent spray since both drywell and containnent pressure are y.
required and low water level can neither initiate or prevent system initiation.
QO30. 60 Section 7.3.1.2 identifies and defines " operational Limits" and (7.3.1.2) implies that they are the level at which the trip unit initiates
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the ESF function. " Levels Requiring Protective Action" is not x
defined but is identified as one of the parameters tabulated in the 7.3 tables. "Hargin" is defined as the difference between the
" Operational Limit" and undefined " limiting conditions".
Both
" Operational Limits" and " Levels Requiring Protective Action" are is'*
said to be tabulated in the 7.3 tables but only one value is Y;Y tabulated in some tables and none are identified by either of the
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s.
two defined titles. knend your FSAR to fully define the terms used in specifying the design basis and to utilize consistent terminology throughout the discussion and tabulations.
Indicate the method used to include the ef fect of the. rate of change of the variable fritiating the trip and the transient overshoot as a result of tite incident. For reactor water level transmitters confirm that the setpoints, limits and margins include worst case Effects of drywell and/or containment temperature on the sensed
reactor level. For each case in which the trip setpoint is 10% or les,s from end of scale, provide the actual margin between the trip
- point and the worst case response limits of the measuring circuit (For example, diat would be the higinest signal level that could exist with the water level below the measuring side pressure tap for a low water level trip circuit).
QO30.61 Section 7 3.1.1.2 identifies CRVICS as being ccmprised of 12 (7. 3.1.1. 2) subsystems and discusses the design bases for each subsystem (7. 3. 2. 2 )
individually. This grouping of subsystems is essentially the (T7.1-3) same as the breakdown used in Table 7.1-5 to identify specific (T7.1-5) requirements for the system.
In the analysis for empliance with
^
system requirements the system is divided into 5 groupings. (1) j.
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"CRVICS", (2) MSIV, (3) other Isolation Valves, (4) MSL High Radiation, and (5) PRM Subsystems. No definition of terms is given and it is not clear whether categories 2 thru 5 ccmpletely covers the CRVICS since some compliance statements address
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"CRVICS" and one or more of 2 thru 5 while others only address one
- d.;i),
or more of groups 2 thru 5.
Amend your FSAR to identify the breakdown of the CRVICS into the various analysis groupings.
Use either Table 7.1-5 or Section 7.3.1.1.2 to define the subsystems that comprise the CRVICS, but state which.
The cmplete CRVICS should be addressed in each step of the analysis for compliance.
QO30.62 In Section 7.3.2.3.1, it is stated the HSIV-LCS will be able to (7. 3. 2. 3 )
maintain its functional capability assuming a single active
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failure. Appendix A of 10 CFR 50 defines a single failure as a
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failure of an active component assuming all passive components e'
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, function properly or a passive component fails assuming all active components function properly. Section 7.3.2.3.2.3.1.2 states that the MSIV-LCS meets the single failure criteria. This implies that the MSIV-LCS meets the single failure criterion for passive as well as active failures.
Resolve this inconsistency and confirm your
- e
's; -
acceptance of the single failure definition of 10 CFR 50.
QO30. 63 The details presented under paragraph 4.1 and paragraph 4.16 (7 3.2.3.
correctly identify the operation of the MSIV-LCS; however, the 2.3.1) system does not and is not intended to conform to the requirements of IEEE 279 - paragraph 4.1 and paragraph 4.16.
Amend your FSAR to indicate non-compliance and the design basis that supports it.
Similar changes are required for the IEEE 279 analysis for other manually actuated ESF's.
.QO30.64 Throughout the discussions of single failure criteria the terms (7.3)
" credible" and " credible aspects of" are occasionally used to hj"!
modify " single failure". Define these terms and state the y.,,
specific aspects of the single failure criterion that are not credible.
QO30. 65 In the discussion of channel independence for the Suppression (7.3.2.9.2)
Pool Makeup System and for the CRACIS, it is stated that physica1 (7. 3. 2.10. 2 ) separation is maintained where it adds to the reliability of operation.
Identify the particular places where physical separation doesn't add to the reliability and indicate the
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distances over which physical separation is not maintained.
QO30.ss
.Th following inconsistencies and deficiencies have been noted in (7.3)
Section 7.3:
1.
Tables identifying the sensor type, instrument range. accuracy and trip setpoint are provided for 7 of the 10 ESF systems discussed in 7.3 The three systems without tables are the i
HSIV-LCS, the CSCS, and the SSW. The FSAR states the SSW is initiated by other systems and has no ESF instrumentation. No reason is given for omitting the other two systems.
2.
Failure mode and effects analysis are provided for 6 of the 10 systems. The system 1 omitted are the ECCS, CRVICS, HSIV-LCS.
g.
and CSCS.
1.-
3 References to plant drawings range from an actual drawing number to a general reference to Section 1.7.
A general reference to Section 1.7 is inadequate in those cases where the system nanenclature used in the FSAR is not used in the ihl..
drawing titles.
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4 The analysis for conformance generally covers the criteria f
identified in Table 7.1-3 but occasionally emits seme.
The worst case is the Containment Spray Cooling System which only addresses IEEE 279.
Amend your FSAR to provide complete and consistent information for each ESF system.
0030.67 The accident analysis in Chapter 15 takes credit for the pressure (7.3) relief available through the automatic sequencing and operation of
.d
~
' the safety relief valves. Justify the exclusion of the
" ).-
instrumentation and controls for the relief valves from Chapter 7
- in general and Section 7.3 in particular.
QO30.68 The figures in Section 7.1 and 7.3 are inconsistent in the (7 3.1.1.2. 4) separation and logic used in the ESF. The Elementary Diagrams (F7.1-3) agree with F7 3-4 rather than F7.1-3 for Division 1 and Division 2 (F7.1-4 )
but do not agree with either for Division 3 and RCIC. Reference
.?'
(F7.1-5)
B21-1090 agrees with F7.3-5 and disagrees with F7.1-5; however,
( F7. 3-4 )
the discription of the logic in 7.3.1.1.2.4 describes both systems (F7.3-5) in successive paragraphs.
Figures 7.1-4 and 7.3-6 are (F7.3-6) functionally the same but disagree on the assignment of logic HPL B21-1090 between inboard and outboerd valves.
Revise the appropriate HPL E22-1050 docoments to achieve consistency and correctness.
0030.69 The response to Question 211.11 states that the High Differential
-;O
.c (7.0)
Temperature isolation circuit will not be connected (to RCIC and (Q & R 7.6-8) RHR Steam isolation circuits) until a setpoint can be established which will minimize inaGrertent isolation.
Verify that every protective circuit and interlock described in the FSAR will be inplenented before the reactor is started up and will thereafter i,'
be used as described in the FSAR.
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