ML20088A004
| ML20088A004 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 04/03/1984 |
| From: | Gucwa L GEORGIA POWER CO. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| NED-84-017, TAC-54169, TAC-54433, TAC-54607, NUDOCS 8404100385 | |
| Download: ML20088A004 (16) | |
Text
Georgia Power Company f'
533 Piedmcat Menue Attanta, Georgia 30308 Telephone 404 5264526 Maihrig Address:
Post Off,ce Box 4545
. Atlanta, Georgia 30302 Georgia Power L T. Gucwa the southem electre system
- Mer.ager Nuclear Engineertng and Chief Nuclear Eng:neer NE-84-017 i
April 3,1984 Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C.
20555 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 EININ I. HATCH NUCLEAR PLANT UNIT 2 REVISION 10 RBQUEST FOR TECHNICAL SPECIFICATION CHANGES TO SUPiwi ANAIDG 1RANSMInm 1 RIP SYsnN INSTALLATION Gentlemen:
.By letter dated January 23, 1984 Georgia Power Company submitted roposed revisions to the Edwin I. Hatch Unit 2 Technical Specifications p(Appendix _ A to the Operating License).
These changes were to account for and support the modifications to plant design associated with the installation of a General Electric Company (GE) Analog Transmitter Trip System (ATTS).
. Included with that letter was the docment entitled "Edwin I.
Hatch Nuclear Plant Unit. 2, Docket No. 50-366, Proposed Plant Modifications -
Analog - Transmitter Trip System Installation".
Appendix 1 to that document contained a significant hazards review (as required by 10 CFR 50.92) for all of the pmposed Technical Specification changes.
Since the time that this submittal was made, GPC has changed the methodology used for Plant Hatch significant hazards reviews.
Therefore, GPC has completely rewritten these reviews for that submittal, and the new reviews are enclosed with this letter.
Appendix A to the above referenced docment should be replaced in its entirety with the enclosure to this letter. This-revision does not amend or affect the propsed changes to the Hatch 2 Technical Specifications which were submitted on January 23, 1984.
8404100385 840403 DRADOCK05000g
'\\
m; h
Geo'rgiaPower A
- Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief
.0perating Reactors Branch No. 4 April 3,-1984
-Page Two:
~ Pursuant to the - requirements of 10 CFR 50.92, J. L. Ledbetter of the Georgia Department of Natural Resources will be : sent a copy of this letter and all applicable attachments.
Sincerely yours; L. T. Gucwa y
Enclosure
,xc:
H. C. Nix, Jr.
~ Senior Resident Inspector J. P. O'Reilly, (NRC-Region II)
J. L. Ledbetter s
Y t
t t
r
......~-,--.--..,.~m...
_ - - - - -.. ~,. ~. - - - -.. - - -
e e
APPENDIX 1 SIGNIFICANT HAZARDS REVIEW
Overview of ~ the Individual 10 CFR 50.92 Evaluations of the Proposed ATTS Related Technical Specification Changes for the Edwin I.
Hatch Nuclear
. Plant-Unit 2
~ The proposed Technical Specification changes which Georgia Power Company 1
is proposing for use with the new ATTS currently being installed at
. Hatch 2
include new instrument trip setpoints/ allowable values and
- surveillance intervals which take credit for the advantages that the new
-devices have over those currently installed at the plant, in terms of setpoint drift and instrument accuracy.
In addition to these types of
' revisions this submittal also proposes a
number of other types of
. Technical Specification changes including the following :
Changes to plant specific equipment identification (MPL) numbers as the-result of new numbering.which has been assigned to ATTS components.
Changes which account for modifications to instrument loops or trip logic resulting from the new ATTS design.
Chang es which correct minor typographical or descriptive errors found-in the Hatch 2 Technical Specifications during the safety review process for ATTS.
The errors found do not necessarily affect sections covering requirements for ATTS components.
Changes to the Technical Specification Bases Sections to correct existing - errors and to update them with respect to the other proposed ATTS changes-All of these proposed modifications were based on the NRC and industry standards listed in Table 1
to this appendix,.
to the extent practicable.
It should be noted that use of many of the documents in Table. 1 go beyond the extent of committments made by GPC, including those made in the Hatch 2 FSAR, and that their use in the design and' implementation of ATTS does not represent an extension by GPC of these commitments to other plant systems which are designed to other criteria.
In the case where conflicts arose between the requirements of the FSAR and. those _ contained in the listed standards the requirements. of Hatch 2 FSAR Sections 3.1 and 7.1.2 and also Appendix A were followed by GPC.-
The ind_ividual 10 CFR 50.92 evaluations on the following pages when taken -collectively arovide a complete evaluation for significant hazards re sulting from all of the. proposed ATTS related license changes.
Based on the conclusion.of.'each of the individual reviews, which was that each type - or grouping of changes did not result in a significant hazard as defined in 101CFR 50.92,.GPC has determined that the same conclusion is valid.for this entire ~1icense change poposal.
as Al-1 4
10 CFR 50.~ 92 Evaluation for the Proposed Changes to the Technical Specifications as a Result'of the Installation of the Analog Transmitter Trip' System for Edwin I. Hatch Nuclear Plant-Unit 2 Georgia Power Company- (GPC) has reviewed the requirements of 10 CFR
- 50. 92. as they relate - to the proposed Technical Specifications changes due to the installation of the Analog Transmitter Trip System (ATTS).
'ATTS replaces the p re s su re,
- level, and temperature switches in the
' reactor protection system and emergency core cooling system (ECCS) with analog sensor / trip unit combinations.
The system is designed to improve sensor-intelligence and reliability, while still providing continued monitoring of critical parameters and performing the intended basic logic function. 'Since the ATTS instrumentation is superior in design to the mechanical switches currently used at Plant
- Hatch, certain surveillance intervals may be extended without any significant effect on the expected magnitude of sensor drift or -frequency of instrument malfunction.
GPC proposes to change the surveillance requirements for the ATTS instrumentation to once per shift for channel checks, once per month for channel functional tests, and once ;*r operating cycle for channel calibrations.
These proposed surveillance requi rement s were previously approved on a generic basis for ATTS equipment by the Nuclear Regulatory Commission (NRC) review of the General Electric Company topical report NED0-21617-A.
Additional changes to the nomenclature
' used. in the Technical Specifications are included for clarification and consistency with this proposed change.
GPC has reviewed the proposed changes and considers them not to involve a.significant hazards consideration for the following reasons:
1)
The proposed surveillance-requirement changes would not significantly increase the probability 'or consequences of an accident previously evaluated, because the new ATTS instruments have-been demonstrated to be superior in design to the existing devices in terms of instrument inaccuracy and drift characteristics.
In
- addition, the new setpoints have been rigorously calculated, assuming the proposed surveillance frequencies.
2)
The proposed surveillance requirement changes would not create the.
possibility 'of a
new or different accident from any accident previously evaluated,- because the new surveillance intervals for ATTS were developed to be consistent with the Plant Hatch-Unit 2 FSAR-descriptions.
3)'
The proposed surveillance requirement changes would not involve a significant reduction in a margin of safety, because the new surveillance requirements are tailored to the ATTS instruments, using the methodology of' Regulatory Guide 1.105.
In addition, the basis' for the margins of safety, as described in the FSAR, have been maintained.
Al-2
'10' 'CFR f-50.92 Evaluation :for the Proposed Changes to the Technical Low Low (Level 2)
Trip Specifications Reactor Vessel Water Level
.Setpoint for Edwin I. Hatch Nuclear Plant-Unit 2 Georgia Power Company has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed reactor vessel water level 2 trip setpoint Technical Specifications change.
This proposed change lowers the level 2 trip setpoint/ allowable value f rom -38 in. to -55 in.,
thus decreasing the number of plant transients by decreasing the number of HPCI/RCIC actuations due to normal operational pe:turbations in water level.
GPC has reviewed the proposed change and considers it not to involve a significant hazards consideration for the following reasons:
1)
It will not significantly increase the probability or consequences of an accident previously evaluated, because. a reevaluation of the FSAR analysis showed that the new setpoint
-in conjuction with the new ATTS instrumentation would still provide the same degree of plant protection as described in the FSAR.
2)
It will not create the possibility of a new or different kind of accident from any accident previously evaluated, because the lowered setpoint is still within the bounds of the plant safety analysis and should decrease the number of unnecessary ECCS actuation system. challenges.
3)
It will not involve a significant reduction in a margin of safety, because the setpoint still performs its intended safety
- function, as described in the FSAR.
In
- addition, the calculations which determined the new setpoint took credit for the improved drif t characteristics of the ATTS instruments and the criteria of Regulatory Guide 1.105.
a.
See-subsection 4B.1 (page 4-2) for discussion of proposed changes.
Al-3
F 10.-CFR-50.92.
Evaluation for the Proposed Changes to the Technical Specifications due to the Deletion of the High Drywell Pressure Signal for-Residual Heat Removal (Shutdown Cooling Mode),
Reactor Pressure Vessel (RPV) Head Spray Valves, and! Reactor Water Cleanup Isolation for
. Edwin I. Hatch Nuclear Plant-Unit 2 Georgia Power Company has reviewed the requirements of 10 CFR 50.92 as 4
- they relate to the proposed changes to the Technical Specifications due to the deletion of the - high drywell pressure signal for residual heat removal (RHR) (shutdown cooling mode), RPV head spray valves, and RWCU
- isolation.
.The purpose of this change is to stop small steam leaks in the drywell from preventing operation of the RHR and RWCU systems during the shutdown - cooling mode, thereby prohibiting an acceptable normal shutdown procedure..
GPC has r6 viewed this proposed change and considers it not to involve a significant hazards consideration for the following reasons:
1)-
It will not significantly increase the probability or consequences of an accident previously evaluated; b.ecause the requirements of 10 CRF 100 are still met, and the Appendix K calculations are not affected.
2)~ It will not create the possibility of a new or different kind of accident from any accident previously evaluated, because the deletion of the drywell. pressure isolation is only being made on closed-loop ' systems.
In addition, GPC has determined that the reactor vessel low water level trip function which isolates
.the shutdown cooling mode of RHR ~ and RWCU is adequate for reactor protection.
Furthermore, this change does eliminate the possibility for is'olation of the shutdown cooling system, due to high drywell pressure, during periods when its. function is essential for adequate decay heat removal.
3)~.It will not involve a significant reduction in a margin of safety, because the high drywell pressure isolation has little effect in preventing coolant losses'and presently-hinders the operability of the RHR shutdown cooling systems.during certain plant scenarios.
See subsection 4B.2-(page 4.3) for discussion of proposed changes.
a.
Al-4
10 CFR 50.92 Evaluation for the Proposed Changes to the Technical Specifications due to the Lowered Water Level Trip Setpoint for Isolation of Reactor Water Cleanup and Secondary Containment, and Starting of Standby Gas Treatment System for Edwin I.
Hatch Nuclear Plant-Unit 2 Georgia Power Company has reviewed the requirements of 10 CFR 50.92 as they relate to'the proposed changes to the Technical Specifications due to the lowered water level trip setpoint for isolation of RWCU and secondary containment, and starting of the standby gas treatment system (SGTS).
This change proposes to lower the water level trip setpoint for isolation of RWCU and secondary containment, and startup of the standby gas treatment system (SGTS) from level 3 to level 2.
A reactor scram from normal power
(
50-percent rated) usually results in_ a reactor vessel water level transient ' due to a void collapse that causes RWCU isolation at level 3.
This usually results in the dropping of the cleanup filter cake and added radwaste processing.
These problems may be avoided by lowering RWCU isolation to level 2.
Lowering the SGTS actuation and secondary containment isolation from level 3 to level 2 reduc'es the potential for spurious isolations.
GPC has reviewed the proposed changes and considers them not to involve a significant hazards consideration for the following reasons:
1)
They will not significantly increase the probability or
. consequences of an accident from any accident previously evaluated, because the FSAR ECCS analysis already assumes SGTS initiation at level 2.
Secondary containment re quire s a
functioning train of SGTS for full effectiveness, and isolation-of the containment building is assumed to be.-simultaneous with SGTS initiation in the FSAR analysis.
In addition, the changes will reduce operability problems associated with RWCU and secondary containment isolations.
2)
They will not create the possibility of a new or different kind of accident from any accident previously evaluated, because the lower setpoint is within the' bounds of the FSAR analysis and will not change the basic functions of these trips.
3)
They will. not. involve-a significant reduction in a ma rg in of
- safety, because these _ trips still perform their intended functions, as_ described in the FSAR.
See subsection 4B.3 (page 4-4) for discussion of proposed changes.
a.
Al-5
- 10 CFR_ 50.92 Evaluation for the Proposed Changes to the Technical Specifications due to'the Deletion of Ambient Temperature Loops in the Leak Detection System for'Edwin I. Hatch Nuclear Plant-Unit 2
-Georgia Power Company has reviewed the requirements of 10 CFR. 50. 92 as they relate to the - proposed change to the Technical Specifications due to the deletion of ambient'. temperature loops from the leak detection system for the Reactor Water. Clean-Up pump rooms (two) and heat exchanger room.
As part. of the ATTS. modification, the change proposes to use the hot leg of the differential temperature sensor for the high ambient temperature trip rather than using an independent trip element
'and trip-~ device.
This arrangement may cause slight changes in the
-sensitivity of the leak detection system,. depending upon the heating, ventilation, and air-conditioning (HVAC) design, but it will not defeat the intended function of the system.
In general, this new arrangement will create more reliable leakage detection since the HVAC system.will be drawing air across the RTDs.
Therefore, there is no possibility of
. the sensors being located in a dead air space relative to certain break
' locations-in the room.
The proposed Technical Specifications revisions will reference the trip unit loop from which the ambient temperature trip..is taken in place of the existing ambient temperature trip instrument.
GPC has reviewed the proposed changes and considers them. not to involve a significant. hazards consideration for the following reasons:
- 1) -The modification will not significantly increase the probability or consequences of an accident previously evaluated, because this change is consistent with the applicable criteria listed in sections 3.1 and 7.1.2 a'nd in Appendix. A of the FSAR - and in general, is more reliable in detecting leaks.
- 2) -The modification will not create the possibility of a new or different accident. from any accident previously evaluated, because plant trip logic' remains unchanged, and the current
. single-failure criteria are maintained.
t 3)
The modification will not involve a significant reduction-in a L
margin of safety, because -single-f ailure criteria and the level l
of redundancy for each trip-function are maintained.
'Also, in general,.the newl location of the sensors will be more reliable for detecting leaks.
a.
See subsection 4B.4 (page 4-5) for discussion of proposed changes.
i l
Al-6
((
l V
..=
3 10 CFR. 50.92 Evaluation for the Proposed Changes to the Technical Specifications ~due to the Deletion of Drywell Pressure Sensors E11-N011A, B, C, D for Edwin I. Hatch Nuclear Plant-Unit 2 Georgia Power Company has reviewed the requirements of 10 CFR 50.92 as
- they relate to the ~ proposed change to the Technical Specifications due to. the deletion of drywell pressure sensors E11-N011A, B,
C, D.
This change p roposes to make the drywell pressure sensor configuration consistent with that for the water level 2 and 3 sensors.
Drywell
- pressure sensors Ell N010A, B,
C, D may be used to provide signals for all.four systems of the ECCS.
and still maintain single-failure c ri te ri a. -
This change deletes instruments E11-N011A, B,
C.
D and transfers their associated trip function (Core Spray, RHR and HPCI high
.drywell' pressure) to instruments E11-N010A, B,
C, D.
Since these instruments (E11-N010A, B,
C, D) are being incorporated into the ATTS modification,.the instrument number in the Technical Specifications was changed to E11-N694A, B,
C, D.
It should be noted that there is an editorial error in Specification Table 3.3.3-1, item 4a; this item should be listed as E11-N010A, B,
C, D.
The Technical Specifications revision involves changing the instrument numbers from E11-N010A, B,
C, D to E11-N694A, B, C, D.
l G.PC h'a s reviewed the proposed change and considers it not to involve a significant hazards consideration for the following reasons:
1)
It will not significantly increase the probability or consequences of-an accident previously evaluated, because this change is consistent with applicable criteria listed in sections 3.1 and 7.1.2 and in Appendix A of the FSAR.
2)
It will not create the possibility of a new or different accident f rom any accident previously evaluated, because the basic trip functions and trip system redundancies, as described in the FSAR, are unchanged.
3)
It will not involve a significant reduction in a margin of safety,- because single-failure criteria and the level of redundancy for each trip function are maintained, and the new surveillance requirements are consistent with the capabilities of the new ATTS instrumentation.
l a.
See subsection 4B.5 (page 4-8) for discussion of proposed changes.
Al-7 m
10 CFR : 50.92 Evaluation for the Proposed Trip Setpoint/ Allowable Value Modifications to the Technical Specifications for Edwin I.
Hatch Nuclear Plant-Unit 2 Georgia Power Company has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed miscellaneous trip setpoint/ allowable value modifications to the Technical Specifications.
The purpose of this change is to update the Technical Specifications trip setpoints for instruments being replaced by the ATTS.
Since the time that the original setpoints were determined, a better calculational method has been-developed.
This proposed change uses Regulatory Guide 1.105 methodology in updating the setpoints for the instruments being replaced with the new ATTS units, and takes credit for the improved error and drif t characteristics of the new system.
This change replaces the trip setpoints listed in the Technical Specifications with these newly generated allowable values.
~
GPC has reviewed the proposed changes and considers them not to involve a significant hazards consideration for the following reasons:
-1)
They will not significantly increase the probability or consequences of an accident previously evaluated, because the new ATTS instruments are of a superior design as compared to the current instruments.
In
- addition, the setpoints were determined using the criteria of Regulatory Guide 1.105, and therefore still meet the FSAR criteria.
2). They will not create the possibility of a new or different kind of accident from any accident previously evaluated, because the basic trip functions, as described in the FSAR, are unchanged.
3)- They ' will not involve a significant reduction in a ma rg in of safety, because for most trips, the original design basis was maintained.
Any new design bases were fully addressed with regard to the FSAR requirements.
In addition, the criteria of
-Regulatory Guide 1.105 were used in the calculation of the new setpoints.
1 See subsection 4B.6 (page 4-9) for discussion of proposed changes.
a.
t
-Al-8 i
l j
10 CFR 50.92 Evaluation for the Proposed Changes to the Technical High (Level 8)
Specifications due to the Reactor Vessel Water Level Trip Instrumentation Modification for Edwin I. Hatch Nuclear Plant-Unit 2 Georgia Power Company has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed changes to the Technical Sp)ecifications due high (level to the reactor vessel water level 8
trip setpoint change.
This change proposes to replace the current instruments with new ATTS units.
The ATTS instruments are superior in design as compared to the mechanical switches currently used at Plant Hatch.
Using the criteria of Regulatory Guide 1.105, the setpoint/ allowable value will be lowered from 58 in. to 56.5 in.
. GPC has revi ewed this change and considers it not to involve a
significant hazards consideration, because it represents a
more conservative and restrictive Technical Specification requirement than that which is currently in-place.
Consequently, this change is consistent with Item (ii) of the
" Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14,870 of the April 6,
- 1983, issue of the Federal Register and will not result in a significant hazards consideration.
s a.
See subsection 4B.7 (page 4-19) for discussion of proposed changes.
Al-9
- 10. CFR _50.92 Evaluation for the Proposed Change to the Technical Spe'cifications as a Result of the Elimination of the Reactor Pressure Permissive.to the' Bypass of the MSIV Closure Signal Due to Low Condenser Vacuum for Edwin I. Hatch Nuclear Plant-Unit 2 Georgia Power Company has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed change to the Technical Specifications as a re sult of the elimination of the reactor pressure permissive to the bypass of the main steam isolation ~ valve (MSIV) closure signal due to low condenser vacuum.
The change proposes to delete the reactor steam
-dome pressure permissive which prevents the group 1 isolation valves from being bypassed on a low condenser vacuum isolation at reactor pressure above the scram setpoint.
With the permissive deleted, the operator may_ open the ' valves from a hot pressurized condition before clearing a - scram.
Currently the operator must clear the scram signal prior to opening the MSIVs when in this condition.
By eliminating this permissive, the plant protective features and, therefore, plant safety are not compromised in any manner.
GPC has reviewed the proposed changes and considers them not to involve a significant hazards consideration for the following reasons:
1)
The modification will not significantly increase the probability or consequences of an accident previously evaluated, because the permissive being deleted does not perform a safety function.
- 2) _The modification will not create the possibility of a new or different-kind of accident from any accident previously evaluated, because the elimination of this permissive has no effect on the reactor protecton system.
- Also, the manual bypass of MSIV closure is performed only when the reactor is not operating at full power.
3)
The modification.will not-involve a significant reduction in a
-margin of safety, because the permissive being deleted does not perform a safety function.
a.
See subsection 4B.8.(page 4-20) for discussion of proposed changes.
Al'-10 i
}
TABLE 1 ATTS CONFORMANCE CRITERIA IEEE STANDARDS IEEE-279-1971:
Criteria for Protection System' for Nuclear Power Generating Station IEEE-323-1974:
Qualifying Class IE Equipment for Nuclear-Power Generating Stations IEEE-336-1977:
Installation, Inspection and Testing Requirements for Instrumentation and Electrical Equipment During the Construction of Nuclear Power Generating Stations IEEE-338-1977:
Criteria for Periodic Testing of Nuclear Power Generating Station Safety Systems IEEE-344-1975:
Recommended Practices for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations IEEE-397-1977:
Application-of the Single-Failure Criterion to Nuclear Power Generating Station Class IE Systems IEEn-420-1973:
Trial-Use Guide for Class IE Control Swi tchboard s for Nuclear Power Generating Stations IEEE-494-1974 :
Method for Identification of Documents Related to Class 1E Equipment and Systems for Nuclear Power Generating Stations NRC REGULATORY GUIDES Regulatory Guide 1.22:
Periodic Testing of Protection System Actuation Functions Regulatory Guide 1.28:
Quality Assurance Program Requirements Regulatory Guide 1.29:
Seismic Design Classification Regulatory Guide 1.30:
Quality Assurance Requirements -for the
. Installation, Inspection,-
and Testing of Instrumentation and Electrical Equipment
.Reflulatory Guide 1.38:
Quality Assurance Requirements for Packing, Sh:.pping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants Regulatory Guide 1.47:
Bypassed and Inoperable Status Indication for Nuclear-Power Plant Safety Systems.
Al-11 o
-. ~ -.
e
,,a e
l TABLE 1 Regulatory Guide 1.53:
Application of the Single-Failure Criterion to Nuclear Power Plant Protection Systems Regulatory Guide 1.61:
Damping Value for Seismic Design of Nuclear
. Power Plants Regulatory Guide 1.62:
Manual Initiation of Protective Actions Regulatory Guide 1.64:
Quality Assurance Requirements for the Design of Nuclear Power Plants Regulatory Guide 1.68:
Initial Test Program for Water-Cooled Reactor Power Plants Regulatory Guide'1.75:
Physical Independence of Electrical Systems Regulatory Guide 1.89:
Qualification of Class IE Equipment for Nuclearl Power Plants Regulatory Guide 1.92:
Combining Modal Responses and Spatial Components in Seismic Response Analysis Regulatory Guide 1.97:
Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident
- Regulatory Guide 1.100:
Seismic Qualification of Electric Equipment fori Nuclear Power Plants Regulatory Guide 1.131:
Qualification Tests of Electric Cables, Field Splices, and Connections for Light-Water-Cooled Nuclear Power Plants NRC REGULATIONS Regulation 10CFR21:
Reporting of Defects and Noncompliance Regulation 10CFR50.49:
Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Flants Regulation 10CFR50.55a:
- Issuance, Limitation and Conditions of Licenses and Construction Permits (CP) - Codes and Standards Regulation 10CFR50, Appendix - A:
General Design Criteria (GDC) for Nuclear Power Plants Al-12 1
1.
TABLE 1 GDC 1:
Quality. Standards and Records GDC 2:
Design Basis for Protection Against Natural Phenomena GDC 5:. Sharing of Structures, Systems, and Components GDC 10:
Reactor Design GDC'13:
Instrumentation and Control GDC 20:
Protection System Functions GDC 21:
Protection System Reliability and Testability GDC 22:
Protection System Independence GDC 23:
Protection System Failure Mode GDC.24:. Separation of Protection and Control Systems GDC 29:
Protection Against Anticipated Operational Occurrences Al-13
,