ML20108D402
| ML20108D402 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 06/18/1984 |
| From: | Colbert W, Okeefe T, Page E DETROIT EDISON CO. |
| To: | |
| Shared Package | |
| ML20108D387 | List: |
| References | |
| NE-3.1, NUDOCS 8503080352 | |
| Download: ML20108D402 (18) | |
Text
..
J 4
i i~
t.
i i
I
+
i
- .- Nuclear Engineering Procedure
)
NE-3.1, Rev 3,
" Safety Evaluations" I
a i
l 4
i t
i i
I l
l 8503080352 850305 PDR ADOCK 05000341 A
pg
Operating Ordsra
. Detroi! on and ProcedurGO Edis o,m,m,m 6AecO4e k
~ wr-1_1 nou_ g Safety Evaluations C'W *"
- NUCLEAR OPERATIONS 1
~
o NUCLEAR ENGINEERING Dr'
~
is a a a e p
4 W. F. Colbert 7
D**
~
V neoweo or ypy
,g x E. Page: T. O'Keefe EOA d d[###A70.
/o-/A-AJ d
l
~
.0 Purpose 1
r A,r,',; ; _ m ~ -.,....
r~.Ip 1.
bis procedure describes the guidelines established by Nuclear Engi neering to ensure that Safety Evaluations are performed in n' /aff. 3,i
" c~om; liance with 10 CFR 50.59, the Nuclear Quality Assurance
""g" Program, and the Plant Safety Review and Evaluation Program n E..
JUN 2 21934 (References 6.1, 6.4, 6.12, respectively).
b-1.2....Ihi n procedure also stipulates the NSRG review requirement for
...tho ue proposed changes which do not constitute an Unreviewed
- " - 4 m.
ut No.Safoty Question.
2.0 Applicability bC Thi's procedure applies to Nuclear Engineering personn performing Safety Evaluations for all proposed changes in the Fermi-2 plant and to procedures as described in the FSAR and for proposed tests or experiments not described in the FSAR.
06 ED/SO4 der Nuclear Production has an equivalent process for those areas their scope of responsibility. These areas of responsibility wi1L,3 [o be delineated separately in the relevant procedures.
JUN 2 21984 3.0 Responsibility INFOR ON ySTEMS 3.1 The General Supervisor, Nuclear Safety and Plant Engineeri is responsible for the development, control and implementation this procedure.
3.2 Cognizant supervisors responsible for processing change, test, or experiment packages are responsible for assignment, completion, and initial distribution of the Preliminary Safety
^
Review and Safety Evaluations described herein.
3.3 The assigned responsible engineer is responsible for performing the Preliminary Safety Reviews and Safety Evaluations described by this procedure.
l
(
3.4 Information Systems is responsible for permanent filing of Preliminary Safety Reviews and Safety Evaluations and for their subsequent distribution.
100/NE1/1.0 060684
.,s De m>sru sixs
- e. o..
- .a.
.seco 68/h5[OT
?*
o Safety Evaluations NE-3.1. Rev. R f
4.0 Interface Nuclear Safety and Plant Engineering interfaces with the following Edison organizations:
4.1 Nuclear Safety Review Group (NSRG) 4.2 Onsite Review Organization (OSRO) 4.3 Quality Assurance 4.4 Nuclear Production 4.5 Generation Engineering Department (GED) 5.0 Definitions 5.1 Preliminary Safety Review - A technical review of any proposed change, test, or experiment to determine if the proposed change, test, or experiment requires a Safety Evaluation.
5.2 Safety Evaluation - A technical evaluation which provides the bases for determining whether a proposed facility or procedure change, test, or experiment involves an Unreviewed Safety Question as stipulated 10 CFR 50.59.
5.3 Safety Related - Plant features necessary to ensure:
5.3.1.
The integrity of the reactor coolant pressure boundary.
5.3.2.
The capability to shut down the reactor and maintain it in a safe shutdown condition.
5.3.3.
The capability to prevent or mitigate the consequences of accidents that could result in offsite exposure i
comparable to the guidelines of 10 CFR Part 100.
l 5.4 Unreviewed Safety Question - A proposed change in the facility or procedures as described in the FSAR or, tests, or experiments not described in the FSAR involves an Unreviewed Safety Question, if:
5.4.1 The probability of occurrence or the consequences of an accident or malfunction of equipment, important to safety, previously evaluated in the Safety Analysis Report or previously issued License Amendment is increased; or
(
1 100/NEl/1.1 060684
DE 9G5700 8-80CS wa + o2.
&m Safety Evaluations NE-3.1. Rev. 2 b/NE/84 3e 9
(
5.4.2 The possibility of an accident or malfunction, different than any evaluated previously in fthe Safety Analysis Report, or previously issued License Amendment is increased; or 5.4.3 The margin of safety as defined in the basis for the Plant Technical Specifications is reduced.
5.5 Accident Analysis - A formal and documented engineering analysis of the facility's response to postulated disturbances in process variables and equipment malfunctions or failures. The analysis is performed to determine the consequences of such postulated disturbances, malfunctions and failures involved with the safety of plant personnel, the public and the environment. It also evaluates the capability built into the facility to mitigate such f ailures and situations, and/or to identify the limitations of expected performance of such mitigating capability.
5.6 Independent Safety Review - A written revicw of material that describes the safety implications of a proposed change, by a ghg group or committee not responsible for the origination of the material subject to review.
I 5.7 Q-List - Items designated as Safety Related which must conform to QA Level I requirements (Reference 6.11).
5.8 Safe Shutdown - Those systems used to achieve and maintain a safe shutdown condition (hot and cold) of the plant exclusive of the Reactor Protection System and accident mitigation features of Engineered Safeguards (see Section 7.4, FSAR).
5.9 Engineered Safeguards - Those systems provided to mitigate the consequences of postulated accidents. (see Chapter 6, FSAR).
5.10 Reactor Protection System - Those systems and subsystems required to effect a scram if monitored system variables exceed prc-established limits.
6.0 References i
l 6.1 U. S. Codes of Federal Regulations,10 CFR 50.59 6.2 U. S. Codes of Federal Regulations,10 CFR 50, Appendix B 6.3 U. S. NRC Regulatory Guide 1.33, Revision 2, " Quality Assurance Program Requirements (Operation)"
6.4 NOD-14 Nuclear Operations Management Plan i
100/NE1/1.2 060684 l
n,..-
uco %. o.i.
aco.
(
Safety Evaluations NE-1.1; Rev. 3 L o' o
6.5 Fermi-2 FSAR, Chapter 15, " Accident Analysis" f
6.6 Fermi-2 Plant Technical Specifications 4
6.7 NE-1.3,"Staf fing, Training and Qualification of Personnel" 6.8 NE-2.1.4, " Operating License Amendments" 6.9 NE-2.1, " Licensing and Regulatory Requirements" 6.101E-1.4.1, "NSRC's Review of Written Safety Evaluations" 6.11 EF2 Project Q-List, PIS No. A30-00-0-000-QX028 6.12 N0P-103, Program Description, Plant Safety Review and Evaluation Program 6.13 NOP-106, Program Description, Design Change Program 6.14 Plant Administrative Procedure 12.000.53, " Guidelines for Deter-mination of Safety Related Systems, Equipment and Procedures" I
6.15 NOIP 11.000.49, " Document Control and Records Management" A
7.0 Discussion 7.1 This procedure provides for a Safety Evaluation of all proposed changes, tests and experiments in conformance with 10 CFR 50.59, and includes all Engineering Design Packages, Engineering Change Requests, procedure changes, etc.
A two step process is prescribed for such Safety Evaluations.
The first involves a screening process called a Preliminary Safety Review to determine if a Safety Evaluation as described in 10 CFR 50.59 is required. If this review determines that a Safety Evaluation is required, the second step is the performance of the actual Safety Evaluation to determine if an Unreviewed Safety Question is involved.
In performing the Preliminary Safety Review, it should be noted that the intent of 10 CFR 50.59 is to limit the requirement for Safety Evaluations to facility and procedure changes, tests and experiments which could impact the safety of operations includin8 radiation protection for plant personnel. Thus, Safety Evaluations are required for the following categories.
o Changes in the Facility as Described in the FSAR i
This pertains to any changes in the facility which alter the design, function or method of performing the function of a 100/NE1/1.2 060684
DE 9615700 8-80CS
- e. wu sea.
Safety Evaluations NE-3.1. Rev. 3 l 0/IS/84 5 e9 i
component, system or structure described in the FSAR. This would apply to components, systems and structures ' described either in the written portion of the FSAR or in the drawings contained therein.
o Changes in Procedures as Described in the FSAR This pertains not only to procedures discussed in the Initial Operations and Organizational Chapters of the FSAR, but also to other procedural-type commitments such as the emergency plan and modes and sequences of plant operation described in the FSAR.
o Conduct Tests and Experiments Not Described in the FSA?
This pertains to the performance of an operation not described in the FSAR which could have an adverse effect on safety-related systems.
Accordingly, if the proposed activity affects the safety function of a related system, a Safety Evaluation is required. However, 3
it is important to understand that the term "Unreviewed Safety Questions" is not necessarily limited to those matters which may
(
be considered " Safety Related." Modification to, or addition of, non-safety-related equipment could also constitute an Unreviewed Safety Question. Therefore, all modifications must be reviewed in accordance with Section 8.0 to determine whether or not an Unreviewed Safety Question is involved. Merely stating that an item is not safety related is not sufficient justification to preclude a Safety Evaluation or to exclude the possibility that an Unreviewed Safety Question exists.
Thus, circumstances requiring a Safety Evaluation include:
o A proposed change to any safety-related structure, system, component, or procedure described in the FSAR.
o Any proposed change, test or experiment which does not ordinarily involve safety-related functions, but which, by their introduction or alteration, could create an Unreviewed Safety Question.
o Any proposed change, test or experiment which modifies significant characteristics described in the text and drawings of the FSAR, including, but not limited to, potential for personnel radiation exposure, performance criteria, margins of safety, materials compatibility, methods of control, drawing configurations, operational sequences, and procedural objectives or intent.
100/NEl/1.4 061284
r og,,,,,. 3c3 w.,
w me. or.
Safety Evaluations NE-3.1, Rev. 3 6[IS/NEf 6e 9
Any proposed change, test or experiment which inv,olves o
systems or components required for compliance with the Limiting Conditions for Operation (LOC's) defined in the Technical Specifications.
o Any proposed change, test or experiment which modifies the accepted parameters, assumptions, or analyses described in Chapter 15 of the FSAR, Accident Analyses.
o Any proposed test or experiment which is not described in the FSAR which could impact safety-related systems.
o Any change to the Technical Specifications.
The Safety Evaluation Requirement Checklist and Safety Evaluation Form appended to this procedure are structured to provide guidance in making the determinations required by Preliminary Safety Review and Safety Evaluations respectively, as defined herein. Additional guidance including specific examples of proposed changes and test and experiments that require and do not require a Safety Evaluation as specified by 10 CFR 50.59 has been extracted from the NRC I&E Manual and is appended to this procedure (Attachment 9.3).
}
7.2 The General Supervisor, Nuclear Safety and Plant Engineering may also initiate a Safety Evaluation of a proposed change:
o At the request of the Vice President, Nuclear Operations.
o At the request of the NSRG.
o At the request of the NRC.
o When the General Supervisor independently determines a Safety Evaluation to be necessary.
t 7.3 Accident Analyses may be required to resolve an Unreviewed Safety Question determination.
8.0 Procedure 8.1 Preliminary Safety Review 8.1.1 A cognizant supervisor will assign a responsible engineer to formally identify the proposed change, test, or experiment and to perform a preliminary safety review to determine if a safety evaluation is necessary by completing the Safety Evaluation Requirement l
Checklist (Attachment 9.1).
i 100/NEl/1.5 061284
oc w>sroo s4ces sw me. o.o Safety Evaluations NE-3.1, Rev. 3 k/k8T 7,, 9 j
(
8.1.2 If upon completion of this review, all boxes in Section 3 of the checklist are marked "No," the conclusion is
[
reached that a Safety Evaluation is not necepsary, and the change may be pursued without further evaluation.
1.
A written basis for this conclusion is included and the necessary signoff completed.
2.
The supervisor is responsible for ensuring that the completed checklist is attached to the document package as identified in Section 1 of the checklist and one copy of the checklist is transmitted to Information Systems for filing.
3.
Information Systems, upon receipt of the completed checklist, will immediately assign and affix a file number in the space provided and then process for permanent filing.
8.1.3 If in the course of this review, any boxes in Section 3 of the checklist are marked "Yes," a Safety Evaluation is necessary to determine if an Unreviewed Safety Question exists. The responsible engineer will note I
this conclusion in Section 4 of the checklist and then directly proceed to complete the Safety Evaluation in accordance with Section 8.2 below.
8.2 Safety Evaluation 8.2.1 The responsible engineer will perform and document a Safety Evaluation utilizing the Safety Evaluation Form.
The previously completed checklist becomes part of the Safety Evaluation and is attached to the form.
8.2.2 When compiling the Safety Evaluation, the responsible engineer ensures that the written evaluation contains the following:
i 1.
A brief description of the change and reason for j
change (Section 1 of the checklist). This l
information will be used to compile the required annual report to the NRC of changes, tests and experiments.
2.
Identification of the safety functions and corre-sponding FSAR and Technical Specification sections which would be affected by the change.
(Sections 2 and 3 of the checklist). The affected safety functions can be obtained from sources such as;
\\
l 100/NEl/1.6 l
061284
c cs w
.se., au 6.se s%,
Safety Evaluations NE-3.1, Rev. 3 SS$Ff 8e 9
i the FSAR, the Plant Technical Specificapions, outside engineers and vendors, Nuclear Engineering personnel or files, reports, submittalg to the NRC, NRC Safety Evaluation Reports for Fermi 2, Functional System Descriptions and equipment specifications.
3.
A description of the effects if any, of the change, test or experiment, including the systems capability to prevent accidents or mitigate the consequence (s).
4.
The basis or bases for the conclusion. This may be a written argument, calculations, engineering reports, etc., as deemed necessary. Calculations may be performed by Nuclear Engineering and/or an outside agent. The conclusion for a trivial or simple change may be sufficently supported by a written argument in the space provided on the form.
5.
Documentation of any items required to support the evaluation. These documents must be attached or referenced in the written Safety Evaluation.
6.
A conclusion as to whether or not the proposed change, test, or experiment includes an Unreviewed Safety Question by checking appropriate box on the form.
8.2.3 The responsible engineer signs and dates the completed form and submits it, along with any additional, supporting information and/or documents (necessary to support the safety evaluation) to his/her respective supervisor.
8.2.4 The cognizant supervisor is responsible for delegating an engineer to review the written safety evaluation.
This reviewing engineer must be someone other than the engineer who originally prepared the safety evaluation, and must possess some degree of expertise in the area being reviewed. Completion of the review will be noted by the signature of the reviewing engineer prior to returning to the cognizant supervisor.
8.2.5 If the Safety Evaluation shows that an Unreviewed Safety Question is not involved, the proposed change, test or experiment may proceed without NRC concurrence
(
unless a change to the Technical Specifications; con-sidered to be part of the license, is required.
jdfg loo /NE1/1.7 061284
m es>sne s-socs w,,
w
- e. c=
- . a.
[IS/bV 9e 9
(
Safety Evaluations NE-3.1, Rev. 3 8.2.6 If any of the boxes in Sections 1-3 are mark'ed "Yes",
l/h an Unreviewed Safety Question may be involved.
In this event, one of three options must be chosen:
l.
Cancel proposed action.
2.
Revise proposed action and/or analysis, and re-cycle through the review process.
3.
Prepare and submit an application for amendment of the operating license (Reference 6.8).
8.2.7 Disposition and approval of the proposed modification based on the written Safety Evaluation and supporting documentation will be made by the supervisor. The supervisor will be responsible for ensuring that the original of the completed Safety Evaluation (checklist, form, and any attachments) is transmitted to Information Systems for distribution and filing and one copy is attached to the document package as identified in Section 1 of the checklist.
I 8.2.8 Information Systems, upon receipt of the completed Safety Evaluation will immediately assign and affix a single file number to both the checklist and form in the spaces provided.
Single copies are then transmitted to the Secretary, NSRG and Lead Independent Safety Engineer as indicated on the form unless otherwise directed by these two individuals. One copy is processed for permanent filing.
8.3 NSRG Review Requirements 8.3.1 Should the proposed change involve an Unreviewed Safety Question, an independent safety review of the corresponding request for a license ammendment will be performed by the NSRG prior to submittal to the NRC for approval, and prior to implementation.
8.3.2 The Safety Evaluation for a change that did not con-stitute an Unreviewed Safety Question will undergo an l
after-the-fact independent review by the NSRG following l
the guidance provided in Reference 6.10.
9.0 Attachments 9.1 Safety Evaluation Requirement Checklist
{
9.2 Nuclear Safety Evaluation Form 9.3 "10 CFR 50.59 - Changes to Facilities, Procedures and Tests (or Experiments)," exerpt from NRC I&E Manual, 6/1/76.
100/NE1/1.8 061284
NE-3.1, Rev. 3
- .1 SAFETY WALUATION M QUIREMENT CHECKLIST Page 1 of 1 (10CFA 50.59)
File No 1.
IDENTIFICATION OF PROPOSED G AMCE FROCEDUn f:
OSCOPEDOCunENTf SET POINT CHANGE CEDFf:
J TEST REPORT f OTHER:
Descriptions
[ T T
2.
5 ICATI 4
e oc (s), system (s), equipment or structures involved, or does the acti
'fect No Engineered Safeguards Yes O
l
'! T No Rector Protection System Ye.
O no vn at No Security Syste" O
O "*
3.
PRI V
Deterni Safety Evaluation required)
Yes so the ange t
ability of a safety related item to perf s sa y
?
Yes Bo t
c e
fy ignif haracteristics or procedures Y:.tV6 desc d
2 List Se Yes Eo Does the chang at vol systems or componets required for 11 L a in Technical f
Specifications?
Yes No Does this change mod um 11 sed
. k analyses List Section(s):
N j;]
described in Cbspter 1 f
6 +27 V
V Yes No Is a test or experiment involved which affect
.. safety and is not described in the FSART Yes Bo Is a change to the Technical Specificati red?
List Section(s):
Yes so In the judgment or the evaluator is a Saf tion r (See Section 7.0 of procedure for additional
.)
si%
If the answere to all questions in Section 3 are *No", a Safety Eval 10 required. Complete this checklist iocluding a written basis for the nega nelusi Basis for negative esecimsion
~
If the answer to any goestion in Section 3 is *Tes", check
- Tee
- in Section 4 and l
proceed directly to the Safety Evaluation Form. Imave Section 5. below, blank and l
attach this checklist to the Safety Evaluation Form.
4.
SAFETT WALDATION REQUIRED:
Oyes Ono S.
PREPARED BY Date AFFB0VID BY Title Date Distribution (if no Safety Ivaluation required):
e Attsch to documentation package e
Original to Information Systems: (Bans Ebner)
'~
NE-3.1,Rev.3
~.2 SAFETY EVALUATION F0an Pcge 1 of 1 (10CFR 50.59)
File Nos
{
(Determine if the activity involves an Unreviewed Safety Question.)
~
Ras the probability of occurrence or the consequences of an accident or 1.
malfunction of equipeent important to safety previously evaluated in the FSAR been increased?
Yes No State Basis:
T I
s,.the possibility of an accident or malfunction of a different type than any y
aluated previously in the FSAR been created?
Yes _ 50 Mj State Be e,
s'. -K '.
f fff 1 -.
f I I
.X
- ~I A
XV I I-3 3.
Has t gin defined in the basis for any Technical Specifica-tion the reduce Yes No State Basis 1 If - M - 1 A EF ff -f f
-\\
g fff f :.A -A j
y f... f f.:]
%.A
\\
f:-f f -]
}r' y y
+\\
f=
A
-K%..;
"-. if*
gA
(
If the answer to any of the abo questiuns a "Y
,T Unre' viewed Safety l
Question is involved.
jd JF 4
UNREVIEWED SAFETY QUESTION:
Yes
" " ~
M M..m.
'5.
REVIEWED BT 6.
DISPOSITION:
8 g:
O There is no Unreviewed Safety Question; proceed with d acti O cancel ProFo.e4 etta=.
C Modify proposed action.
O Request NRC approval of proposed action through Licensing Engineer.
Note that if a change in the Technical Specifications is desired, NRC approval is required regardless of the answer to Section 4
)TechnicalSpecificationChange l
l MMr Approved by Title Date j
Distributions o Attach to documentation package e
o Original to Information Systest (Rans Ebner)
-File
-NSRG (Jane Lanart)
-Lead Independent Safety Engineer
l.
.[-
UNITED STATES l
j NUCLEAR REGULATORY COMMISSION NE-3.1, Rev. 3 OFFICE OF INSPECTION AND ENFORCEMENT.3 j
. 4, j
Washington, D.C. 205E5 P.ge 1 of 6
(.;
INSPECTION AND ENFORCEMENT MANUAL DQASIP 10 CFR 50.59
(
i PART 9800 CFR DISCUSSIONS CHANGES TO FACILITIES, PROCEDURES AND TESTS (OR EXPERIMENTS) s
[A.
PURPOSE The purpose of this guidance is to clarify the specific 10 CFR 50.59 language relating to the type of proposed changes, tests, or experi-ments that require a record of the safety evaluation specified in 10 CFR 50.59(b).
It is not intended that this guidance delineate specific licensee review criteria which may be used to identify pro-posed changes, tests, or experiments that require a safety evaluation as specified by 10 CFR 50.59(b).
B.
POLICY This revision to this CFR Discussion does not represent a change in IE policy.
The discussion section has been revised to clarify the
]
application of 10 CFR 50.59 to controls for using jumpers / lifted leads
(
and to procedure changes.
Also, the 10 CFR 50.59 flowchart (Appendix
-(
- 1) was updated.
C.
APPLICABILITY:
2515 D.
DISCUSSION 1.
10 CFR 50.59 is composed of three essential parts:
l a.
Paragraph (a)(1) is permissive in that it allows the licen-l see to make changes to the facility and its operation as described in the Safety Analysis Report (SAR) without prior approval, provided a change in Technical Specifications (TS) is not involved or an "unreviewed safety question" does not exist.
Criteria for determining whether an unreviewed l
safety question exists are defined in Paragraph (a)(2).
b.
Paragraph (b) requires that the licensee maintain records of changes made under the authority of Paragraph (a)(1).
These records must include a written safety evaluation which pro-vides the basis for determining Whether an unreviewed safety question exists.
Paragraph (b) also requires that a report (at least annually) of such changes be submitted to the NRC.
~
c.
Paragraph (c) requires that proposed changes in Technical
(
Issue Date:
01/01/84 1
-cm-unm g
Specificaticns ba submitted to tha NRC as an applicaticn for
,/
liccnse (mend:ent.
Likewise, proposed changes to th2 fccil-
.\\
ity or procedures and the proposed conduct of tests which involve an unreviewed safety question must be submitted to I
the NRC as an application for license amendment.
)
2.
It should be noted that the safety evaluation required by 10 CFR 50.59 is only one of the several evaluations and reviewsfrequired by the NRC.
Most Technical Specifications require that onsite review groups review proposed procedures and modifications or 8
changes to plant equipment or components affecting safety.
These f
review requirements are applicable whether or not the equipment i
or component is described in the SAR.
As a result of the TS required reviews, the need for a safety evaluation to meet 10 CFR 50.59 requirements may be identified.
Appendix 1 delineates a typical overall review scheme at a facility.
1 3.
This guidance is to be applied during inspection of facilities holding operating licenses under 10 CFR 50 and is primarily l
directed toward:
i a.
Changes made to those systems and procedures described in l
the SAR, and b.
Performance of tests not described in the SAR.
, [V '
l.
4.
Within the context of this guidance, any proposed change to a system or procedure as described in the SAR either by text or drawings should be reviewed by the licensee to determine whether
(
it involves an unreviewed safety question.
Changes may involve 2
an unreviewed safety question even though they are "beyond the second isolation valves," or they do not serve a normal safety-related function, since alteration may introduce an unreviewed j
safety question.
5.
Maintenance activities which do not result in a change to a system (permanent or temporary), or which replace components with replacement parts procured to the same (or equivalent) purchase i
specification, do not require a written safety evaluation to meet i
10 CFR 50.59 requirements.
However, if components described in the SAR are removed, or their function is altered, or if substi-tute components are utilized, or if changes remain following com-plation of a maintenance activity, a safety evaluation is re-quired to meet the provisions of 10 CFR 50.59 and the change must be reported to the NRC as required by 10 CFR 50.59(b).
6.
In all cases requiring a written safety evaluation, the safety evaluation must provide the basis for deteimination that the pro-i posed change does or does not involve an unreviewed safety ques-tion.
A simple statement of conclusion in itself is not suffi-cient; however, depending upon the significance of the change, the safety evaluation may be quite brief.
(
Issue Date:
01/01/84 NE-3.1, Rev.3
'.3 Page 2 of 6
q 7.
Listed below are exarples of varicus changes to fccilitics, systsas, preceduras, cnd tosts which are typical of those requir-ing a 10 CFR 50.59 safety evaluation and those which do not re-quire a safety evaluation under the requirements of 10 CFR 50.59.
(
a.
Changes in the Facility As Described in the Safety Analysis Report.
This pertains to any changes in the fdcility which alter the design, function, or method of performing the function of a component, system, or structure d6 scribed in the SAR.
This would apply to components, systems, and structures described either in the written portion of the SAR or in the drawings contained therein.
Contrasting k
examples of each case are:
(1) Components.
Replacement of thermocouple in the diesel high-bearing temperature automatic shutdown circuitry (if such a component were described in the SAR) with one made by the same manufacturer, but encompassing different response characteristics, would require a safety evaluation to meet the requirements of 10 CFR 50.59.
On the other hand, replacement of a thermocouple in the diesel high-bearing temperature automatic shutdown circuitry (if such a component were described in the SAR) with cne encompassing equivalent response charac-teristics, but made by a different manufacturer, would
(
not require a safety evaluation under the requirements of 10 CFR 50.59.
(2) Systems.
Modifications of the diesel shutdown cir-cuitry (described in the SAR) to provide an automatic diesel shutdown on high-bearing temperature (shutdown feature not described in application) would require a safety evaluation to meet the requirements of 10 CFR 50.59.
On the other hand, if the methods of initiating automatic diesel shutdown are not described in the SAR, specific automatic shutdown features may be rendered inoperable without the conduct of a safety evaluation under the requirements of 10 CFR 50.59.
(3) Structures.
The erection of a concrete block shield wall within the containment building (shield wall is not described its the SAR) would require a safety evalu-ation to meet the requirements of 10 CFR 50.59.
On the other hand, deletion of a shield wall within the con-tainment building (shield wall not described in the SAR) would not require a safety evaluation under the requirements of 10 CFR 50.59.
(4) Jumpers / Lifted Leads.
Licensee controls over jump-R ers lif ted/ leads should include a documented review R process consistent with the one presented in Appendix R ee l Issue Date:
01/01/84 NE-3.1, Rev. 3 Attachnent 9.3 Page 3 of 6
~
CHANGES TO FACILITIES, PROCEDURES
- 10 CFR 50.59 (AND TESTS (OR EXPERIMENTS) 1.
If it is determined that use of a jumper / lifted R lead results in a change to the facility as de-R l
(
cribed in the SAR and that the resultant change R will impact on' safety of operation, then a R safety evaluation is required.
This ) approach R should apply to all types of temporary, modifi-R cations.
Generally, if a plant system is R l
changed by use of jumpers /lif ted leads so that R it will function dif ferently than described in R s
g the SAR, a safety evaluation would be required. R l
On the other hand, use of jumpers /lif ted leads R that result in plant conditions already analyzed R and approved by NRC would not require a safety R evaluation.
For example, bypassing protection R channels in a manner already described in the R SAR would not constitute 'an unreviewed safety R question and would not require a safety evalu-R ation under the requirements of 10 CFR 50.59.
R It is expected that only a small percentage of a R licensee's jumpers /lif ted leads will require a R written safety evaluation R
i l
b.
Changes in Procedures As Described in the SAR.
This pertains not only to procedures discussed in the initial operations and
,f organizational chapters of the SAR, but also to other pro-cedural-type commitments, such as the emergency plan and modes
(
and sequences of plant operation described in the SAR.
If a R procedure results in a deviation from the steps listed in the R l
SAR. or will result in a system operation which deviates from R the way that system is described in the SAR, then a a safety R evaluation should be performed.
Contrasting examples of the R above follow.
R i
(1)
If in the description of the radioactive waste system in the l
SAR, the licensee states that the Shift Supervisor will l
authorize all radioactive liquid releases, a safety evalu-ation to meet the requirements of 10 CFR 50.59 would be l
required before assigning this function to another indivi-l dual.
On the other hand, if the SAR merely states that l
l radioactive liquid releases will be authorized as detailed l
by plant procedures, the licensee's redesignation of the authorization function would not require a safety evaluation under the requirements of 10 CFR 50.59.
l l
(2) If the reactor startup procedure, as described in the SAR, contains eight fundamental sequences, the licensee's deci-sion to eliminate one of the sequences would require a safety evaluation to meet the 10 CFR 50.59 requirements.
On the other hand, if the ifcensee consolidated the eight fundamental sequences but did not alter the basic functions
(,
performed, it would not be necessary to conduct a safety evaluation under the requirements of 10 CFR 50.59.
Issue Date:
01/01/84 NE-3.1, Rev. 3 Attachnent 9.3 Page 4 of 6
ANCTESTSIOREXPERINENTSI
^
~~
~ ~10 N R 50.59 c.
Canduct Trasts and Expericents Not Dascrib2d in th2 SAR.
This pertains to the performance of an operation not described in the l
Contrasting examples of such tests or experiments are:
SAR which could have an adverse effect on. safety related systems.
j
~
- (1) Some plants in the startup testing program havepperformed a
, deboration to critical with all rods inserted. Since this test is performed without deference to the "one' stuck rod
'~
' criterion," a safety evaluation to meet the requirements of 10 CFR 50.59 would be required if the test is not delineated in the SAR.
Since this test may decrease the margin of I-safety defined in the TS basis, it should, in most in-stances, be classified as an unreviewed safety question.
On the other hand, a test to demonstrate the calibration of the i
nuclear instrumentation system by performance of a secondary plant heat balance would not require a safety evaluation under the requirements of 10 CFR 50.59, even if such a test was not delineated in the SAR, since the test does not in-volve an abnormal mode of operation.
(2) A test to determine if the boric acid evaporator may also be used for concentration of the steam generator blowdown ef-J fluent (function not described in the SAR) would require a safety evaluation to meet the requirements of 10 CFR 50.59, since secondary system chemicals could possibly have a deleterious effect on some components within the reactor I
coolant pressure boundary.
On the other hand, an experiment to determine the decontamination factor of the liquid waste l
f concentrator with influent activities of 10 2 Ci/ml and 10 5 1
(
Ci/m1 would not require a safety evaluation under the requirements of 10 CFR 50.59 since such an experiment would not represent departure from normal operational modes.
d.
General Guidance.
It should be noted that the SARs for a number i
of older facilities contain floor plans of onsite buildings that may include trivial detail such as the locating of dividing walls i
between various offices.
From a rigid reading 10 CFR 50.59, it I
is possible to infer that the removal of a dividing wall between j
two offices constitutes a change from the facility described in the SAR, and therefore requires a safety evaluation.
- However, 4
the intent of 10 CFR 50.59 is to limit the requirement for writ-ten safety evaluations to facility changes, tests, and experi-ments which could impact the safety of operations.
j END 1
L()
l
(
^
i Issue Date:
01/01/84 NE-3.1, Rev. 3.3 Page 5 of 6 c
___.-...---.__,_.-._,..,_.s
. -,,.-._,_,_,- _ __-.-._.,,,_ __ _...~___._,.,.,_, _. _ _, _ _ _,, _..,. _, _. - _. _
9 i : %
CHANGES TO FACILITIES, PROCEDURES 10 CFR 50.59 A_ND TESTS (OR EXPERIMENTS) f (Change proposalj Most Tdchnical Specifications (75) reevire the Onsite Review Group te (1) review all procedures and changes thereto that affect nuclear I
safety. 411 proposed tests and emperiments that af fect* nuclear safety.
and all proposed changes to the facility that affect negf e*ar safety:
i
~ ~ ~ * ~ ~ ~ ~ " " "
and (2) to recommend in writing to the plant Superintendent approval
{
I or disapproval of these proposals.
-' i r is the Safety Analysis Report (SAR) affeCted?
8 pl) Does the proposal change the facility or procedures from their descrip* ion in the 5AR?
- 12) Does the proposat involve a test or esperiment not described in the SAR?
(3) Could the proposal af fect nuclear safety in a way not previously evaluated in the 5AR?
Any answer Yes All answers No 4
[is a enance in the T5 involved?j.
10 CFR 50.59 no longer applies. It is still necessary, however to ask:
No Yes 15 a change in the T5 involved?
Yes No o
1' e-Is an uprevieweg safety question involved?
(1) Is the probability of an occurrence or the conseovences of an accident er malfunction of equipment important to safety previously
/(
evaluated in the SAR increased?
g lk-(2) Is the possibility for an accident er malfunction of a different* type than any previous.ly evaluated in the 5AR created?
(3) is the margin of safety as defined in the basis for any technical soecification reduced?
i Most T5 reovire the Onsite nevtew Group to l
render deterininettons in writing with regard to whether er not tne proposed change constitutes an unreviewed safety euestion.
i Any answer Yes All enswers We <,
Most T5 require the Of fsite Review Group to review proposed changes to procedures. eevipment or systems, and tests or esperiments that involve an unreviewed safety euestion.
Cocument the change. Include in these
O U
records a writtes safety evaluation l Submit the proposal to tne providing the bases for the determination NRC for authorisstion.
t that the change, test er esperiment does not involve an unreviewed safet, euettion.
l Autherisatten receivec.j v
a
[p-stees with the theaewt; Most 15 reestre the Of fsite Review Group to review the safety evaluations for changes to precedures, etvipment er systems. and tests or esperir.ents comeleted under the previsions of SC.59 te vertfy that such actions did not 8-......,_
Q) constitute an unreviewed safety evestion.
t Al-1 NE-3.1, Rev.3
'i Issue Date: 01/01/84.3 Page 6 of 6
-e
,e
,,_-,._c---
-.---,-_a-,_
,