ML20101M412
| ML20101M412 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 12/14/1984 |
| From: | Lee O PUBLIC SERVICE CO. OF COLORADO |
| To: | Johnson E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| References | |
| P-84530, NUDOCS 8501030085 | |
| Download: ML20101M412 (25) | |
Text
pr PUBLIC SERVICE COMPANY OF COLORADO P. O.
BOX 840 DENVER, COLORADO 80206 December 14, 1984 OSCAR R. I.EE Fort St. Vrain v,c.....or,a Unit No. 1 P-84530 Regional Administrator
$h$ON-Region IV Nuclear Regulatory Commission E20m 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 J
Attention: Mr. Eric H. Johnson 4
DOCKET NO: 50-267
SUBJECT:
Technical Specification Upgrade Program
REFERENCES:
- 1) NRC Letter, H. R. Denton to R. F. Walker, dated 10/16/84(G-84392)
- 2) PSC Letter, O. R. Lee E. H. Johnson, dated 11/16/84(P-84498)
- 3) NRC/PSC meeting on November 28 - 30, 1984
Dear Mr. Johnson:
?
- Enclosed, for your information, are the Work Specification and a
Schedule that Public Service Company has developed for the Fort St.
Vrain Technical Specification Upgrade Program.
As discussed in References 1, 2 and 3, the program objective is to improve the
- accuracy, completeness, and clarity of the FSV Technical Specifications, and to provide a draft of the upgraded Technical Specifications to the NRC by April 1, 1985.
Attachment 1
is the Work Specification which provides the requirements and guidance for the review of the Final Safety Analysis Report (FSAR), and for the review and upgrading of the Technical Specifications.- is the project schedule which has been developed to i
support the April 1,
1985 draft submittal date.
The Technical Specifications have been grouped into forty-eight (48) subject categories or work packages, and various priorities have been assigned to each one, based on the degree of difficulty and complexity of the subject matter.
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- As discussed in the reference meeting, the overall schedule for
' submitting the upgraded Technical Specifications is as follows:
Provide draft Technical Specifications to NRC April 1, 1985 NRC comments provided to PSC May 1, 1985 Submit for PORC/NFSC approval June 1, 1985 Submit proposed Technical Specifications to NRC -
July 1, 1985 Public Service Company anticipates that further revisions and elaborations to the attachments will be required as the upgrade program aevelops.
If you have any questions or comments about the information contained herein, please contact Mr. M. H. Holmas at (303) 571-8409.
Very truly yours, O. R. Lee, Vice President Electric Production ORL/JMG/kss Attachments 4
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s ATTACHMENT 1 WORK SPECIFICATION FOR TECHNICAL SPECIFICATION UPGRADE PROGRAM to P-84530 PUBLIC SERVICE COMPANY OF COLORADO l{- ~
s FORT ST. VRAIN NUCLEAR GENERATING STATION page I of 16 1
SPECIFICATION COVER SHEET PLANT ITEM NO'S, SPECIFICATION FOR UPGRADE PROGRAM N/A TABLE OF CONTENTS SECTION HEADING PAGE A.
PURPOSE AND SCOPE 2
A.1 Purpose 2
A.2 Scope 2
B.
WORK TO BE PERFORMED 3
C.
TECHNICAL SPECIFICATION REVIEW CRITERIA 4
C.1 General Criteria 4
C.2 Criteria for Bases for Technical Specification 6
C.3 Criteria for Definitions 7
C.4 Criteria for Safety Limits 7
C.5 Criteria for Limiting Safety System Settings 8
C.6 Criteria for Limiting Conditions for Operation 8
C.7 Criteria for Surveillance Requirements 11 C.8 Criteria for Design Features 12 C.9 Criteria for Administrative Controls 12 0.
FSAR REVIEW CRITERIA 13 PAGE SCHEDULE ATTACHMENTS - Overall Program Plan 15 - Project Flow Chart 16 ISSUE
SUMMARY
ISSUE PREPARED BY
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l PUBLIC SERVICE COMPANY OF COLORADO WS-TS-1 FO3T ST. VRAIN NUCLEAR GENERATING STATION Issue A Page 2 of 16 SPECIFICATION CONTINUATION SHEET P: MM 364 22 4043 FORT ST. VRAIN TECHNICAL SPECIFICATION UPGRADE PROGRAM A.
PURPOSE AND SCOPE 1.
Purpose The objective of the Fort St. Vrain Technical Specification Upgrade Program is to improve the accuracy, completeness, and clarity of the Fort St. Vrain Technical Specifications consistent with the licensing basis of the Fort St. Vrain plant as embodied in the Fort St. Vrain Final Sifety Analysis Report (FSAR).
The purpose of this Work Specification is to provide requirements and guidance for the review of the FSAR and the review and revision of the Technical Specifications.
2.
Scope The Fort St. Vrain Technical Specification Upgrade Program consists of two parallel review efforts leading to preparation of a more accurate, complete, and clear set of Technical Specifications. The overall program plan is illustrated in Figure 1 and the project flow chart is shown in Figure 2.
The scope of this Work Specification includes the review and revision of the following existing Technical Specifications sections and associated subsections:
Section 1.0 Introduction Section 2.0 Definitions Section 3.0 Safety Limits and Limiting Safety System Settings Section 4.0 Limiting Conditions for Operation Section 5.0 Surveillance Requirements Section 6.0 Design Features Section 7.0 Administrativ^e Controls The Fort St. Vrain Final Safety Analysis Report (FSAR) shall be reviewed to the extent necessary to verify the bases for the existing Fort St. Vrain Technical Specifications. The FSAR shall also be reviewed to identify any omissions from the existing Technical Specifications.
PUBLIC S E R V I C E' COMPANY OF COLORADO WS-TS-1 FORT ST. VRAIN NUCLEAR GENERATING STATION Issus A Page 3 of 16 SPECIFICATION CONTINUATION SHEET PcRM 344 22 4083 The Standard Technical Specifications (STS) for Westinghouse light water reactors shall be considered during the performance of this work as indicated in the following sections of this Work Specification. However, conformance with the Standard Technical Specifications is not intended to be a requirement of this program, nor is substantiation or justification of any differences with STS requirements necessary.
If instances arise whereby Fort St. Vrain's design features and Technical Specification requirements may represent a possible safety concern relative to STS requirements, those instances will be addrcssed as separate licensing issues outside the scope of the Fort St. Vrain Technical Specification Upgrade Program.
Plant modifications and hardware backfits will not be undertaken to permit the adoption of any STS requirement.
It is outside the scope of the Fort St. Vrain Technical Specification Upgrade Program to utilize or consider any Standard Technical Specification requirement which opens the licensing basis of the Fort St. Vrain plant for further justification or analysis.
Significant research and development efforts or analytical investigations beyond those documented in the FSAR will not be undertaken to determine how or whether a Standard Technical Specifications requirement can be utilized at Fort St. Vrain. Questionable Standard Technical Specifications requiring such efforts and investigations will not be utilized or given further consideration.
B.
WORK TO BE PERFORMED 1.
Each existing FSV Technical Specification within the scope of this work specification shall be reviewed using the criteria described in Section C of this work specification.
2.
The FSV Final Safety Analysis Report shall be reviewed using the criteria described in Section D of this work specification.
3.
Upgraded FSV Technical Specifications shall be prepared as necessary according to the criteria described in the following sections and deficiencies identified during the reviews shall be corrected.
PUBLIC SERVICE COMPANY OF COLORADO WS-TS-1 FORT ST. VRAIN NUCLEAR GENERATING STATION Issue A Page 4 of 16 SPECIFICATION CONTINUATION SHEET FORM 344 22 4003 C.
TECHNICAL SPECIFICATION REVIEW CRITERIA 1.
General Criteria a.
The purpose of the technical specifications is to require that the overall facility status is consistent with the assumptions in the safety analysis. These assumptions deal with the following:
(1) Facility Physical Characteristics, i.e., features that are expected to remain constant.
(2) Status of Equipment, i.e., system and component operability.
(3) Operating State of Equipment, i.e., physical equipment parameters which concern system or component actions or the position or running condition of equipment.
(4) Values of Process Parameters i.e., flows, temperatues, pressures, etc..
(5) Condition of Equipment and Structures, i.e., the state of preservation of quality.
(6) Administrative controls (e.g. shift staffing, review and audit) that must be maintained, b.
Prior to establishing the technical specification, the basis shall be defined thereby establishing the rationale for the specification.
c.
Technical specifications shall be provided only for items relied upon in the safety analysis, and for other items specifically required by Federal regulations to be in the technical specifications.
d.
Technical specifications shall be written in a clear and concise manner with the intent that only one interpretation can be made. The use of vague terms such as "immediately" or "sufficiently" shall be avoided or defined to assure uniform interpretation by all auditors and operators.
i e.
Technical specifications shall be formulated such that compliance is physically possible based on the plant design, including test and measurement limitations.
l f.
Allowance for calculational inaccuracies and dynamic effects shall be considered.
PUBLIC SERVICE COMPANY OF COLORADO WS-TS-1 FORT ST. VRAIN NUCLEAR GENERATING STATION Issue A Page 5 of 16 SPECIFICATION CONTINUATION SHEET P;mM 344 22 4083 g.
Technical specifications shall clearly state the facility operating conditions (e.g. power operation, refueling) to which they appply. The operating conditions selected shall be limited to those conditions for which equipment must be operable or for which parametric limits exist due to assumptions of the safety analysis, h.
Values for parameters shall be specified in units directly available to the operating personnel, shall include allowable tolerances on the specified value, and shall include a'llowance for the effect of any associated instrument error, as appropriate.
i.
Technical specifications shall preserve defense in depth (e.g. multiple barriers, redundancy, backup systems) only to the extent that it nas been rel.ied upon in the safety analysis.
~
J.
Technical specifications shall preserve the single failure criterion to the extent relied upon in the safety analysis and may permit relaxation from this criterion for justifiable periods, for example as based on probability, reliability, previous analyses, or experience.
k.
Adverse impact on plant availability shall be considered in the development of technical specifications consistent with the maintenance of an acceptable level of safety.
1.
Technical specifications shall be developed such that on-site personnel exposure is as low as reasonably achievable while ensuring the health and safety of the public.
m.
Incorporating requirements by references to the Final Safety Analysis Report, Federal regulations, or l ~
industry codes and standards shall be held to a
(
minimum. Where utilized, these references shall be to the subdivision of the document rather than a general reference, n.
The selection of values for technical specifications shall be done by (a) deterministic methods, or (b) probabilistic and reliabilty methods. Probabilistic and reliability methods shall be utilized only when suitable justification is presented, and only on a case-by-case basis.
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PUBLIC S E R V I C E' CdMPANY OF COLORADO WS-TS-1 FORT ST. VRAIN NUCLEAR GENERATING STATION Issue A Page 6 of 16 i
Y SPECIFICATION CONTINUATION SHEET j
Pgmu 344 22 4083 o.
Technical specifications shall be stated in the simplest terms possible to clearly convey their meaning wit'.out ambiguity.
p.
Technical specifications shall be reviewed and expanded, as necessary, to assure accuracy, completeness, and consistency with existing safety analysis documentation.
1 I,
q.
The technical specifications shall account for and utilize existing plant equipment and safety systems.
2.
Criteria for Bases for Technical Specifications a.
Bases for technical specifications shall be summary statements of the reasons for such specifications and shall be provided for safety limits, limiting sa.fety system settings, limiting conditions for operation, and surveillance requirements, b.
The bases shall explicitly correlate the plant design
~
and safety analyses with the technical specification limits and operating conditions, thereby providing a validation of the overall design for the prescribed modes of operation.
c.
The bases for technical specifications shall be developed with appropriate consideration of the
[
following general requirements:
(1) The bases shall not contain requirements over and 3
e i above those in the specification (2) For each technical specification requirement,
. l ',
there shall be a corresponding and clearly identified basis which is solely related to an f /
identified safety requirement.
4 (3) Where applicable, the bases shall identify the specific plant process condition which is controlling for the corresponding specification.
T (4) The. relationship between the values specified in the techr.ical specification and those used in the safety analyses shall be provided in the bases.
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PUBLIC S E R V I C E' COMPANY OF COLORADO US-TS-1 FORT ST. VRAIN NUCLEAR GENERATING STATION Issue A Page 7 of 16 SPECIFICATION CONTINUATION SHEET (5) Errors, from instrumentation or other sources, assumed in the development of the technical specification limits shall be discussed in the bases to provide a clear relationship between the technical specification and the safety analysis values.
(6) The bases shall explain the rationale for the requirements in remedial action statements and the appropriateness of the condition restoration times relative to an acceptable level of safety.
(7) The sources of information summarized in the bases shall be cited.
(8) The justification contained in bases shall not be considered part of the Technical Specifica. tion requirements and may be changed by the licensee without prior NRC approval, providing that the change is evaluated and determined not to involve an unreviewed safety question.
3.
Criteria for Definitions a.
The technical specifications shall include a list of definitions of terms which are frequently used within the document and which are not in general every day use.
In addition, terms which have technical connotations, or terms which are applicable only to Fort St. Vrain should be included. These terms shall be explicit and clearly defined in simple and direct language with the intent that a uniform, unamoiguous interpretation of the technical specifications can be achieved for facility operation and regulatory enforcement.
b.
Relevant standard technical specification definitions shall be adopted where the definitions are consistent with existing plant features and the licensing basis of the plant, i.e.; FSAR terminology and analyses.
4.
Criteria for Safety Limits a.
Safety limits shall be prescribed for selected process variables related to the integrity of barriers to fission product release. Compliance with safety limits shall provide assurance that the barrier will perform as assumed in the safety analysis.
O PUBLIC SERVICE COMPANY OF COLORADO WS-TS-1 FORT ST. VRAIN NUCLEAR GENERATING STATION lssue A Page 8 of 16 SPECIFICATION CONTINUATION SHEET FORM 344 22 4083 b.
The bases shall identify the barrier to fission product release that is being protected by the limit and show why that limit is adequate.
5.
Criteria for Limiting Safety System Settings (LSSS) a.
Limiting safety system settings shall be defined to assure that no safety limit would be violated as a result of a frequent plant process condition and that no infrequent or limiting plant process condition would have consequences which do not meet the acceptance criteria for that condition.
b.
Values for limiting safety system settings shall be based on the assumption that the facility is at or within its limiting conditions for operation when one of these process conditions occurs.
An adequate margin shall be provided between the limiting safety system settings and the safety limits so that safety limits would not be exceeded in the event that protective action is initiated if a limiting safety system setting is exceeded.
c.
Conditions under which channels, features, and interlocks may be bypassed shall be specified either together with the relevant limiting safety system setting or with a relevant limiting condition for operation.
d.
The bases for limiting safety system settings shall identify the safety limit or other safety requirement i
that is being ensured by the LSSS and shall describe l
all allowances included in determining the relationship of the LSSS to the safety limit or other safety requirement. The bases shall discuss the l
conditions under which the bypass of automatic protection associated with an LSSS is permitted.
6.
Criteria for Limiting Conditions for Operation (LCO) a.
The limiting conditions for operation shall define the lowest functional capability or performance levels necessary to assure safe operation of the facility as evaluated in the FSAR accident and safety analyses.
l b.
Limiting conditions for operation shall be provided for the following when they are relied upon in the safety analysis:
(1) Condition, or status, of equipment or systems;
PUBLIC S E R V I C E' COMPANY OF COLORADO tis-TS-1 FORT ST. VRAIM NUCLEAR GENERATING STATION Issue A Page 9 of 16 SPECIFICATION CONTINUATION SHEET (2) Parameter limit with no associated instrument alarm or protective action setpoint; (3)
Instrument setpoints for monitored parameters with no associated automatic protective action (for this case, the LCO limit shall be the limiting value of the parameter, while the SR value shall be the instrument setpoint).
(4)
Instrument setpoints for monitored parameters with associated automatic protective actions.
c.
Each LCO shall include an applicability statement that clearly identifies the operating modes to which the LC0 applies.
d.
Values for limiting conditions for operation shall be consistent with extremes of initial conditions which
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have been shown to result in acceptable consequences for the various plant conditions as demonstrated by the safety analysis.
e.
Included in the limiting conditions for operation shall be an action statement that describes the remedial action to be taken if:
(1) the operable status of equipment or systems is less than the required minimum; (2) the monitored parameters are not within the specified range; or (3) the instrument setpoints are less conservative that the specified value.
f.
Remedial action statements shall specify the condition restoration time and shall require that, unless restoration is accomplished within that time, the facility be taken to a specified mode of operational safety consistent with the safety protection available from the remaining equipment or systems. The time interval allowed for each action shall be specified.
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WS-TS-1 PUBLIC S E R V I C E' COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION Issue A Page 10 of 16 SPECIFICATION CONTINUATION SHEET PCMM 344 22 - 4083 g.
In developing remedial action requirtments, consideration shall be given to: the operability of redundant or diverse systems; the probability of an event taking place during the condition restoration time which would be influenced by the limiting condition for operation; the reliability of the redundant or diverse systems; the risk of inducing an undesireable incident while performing the remedial action (for example, the thermal transients induced by a shutdown and cooldown); and the potential cost of complying with the proposed remedial action versus the benefits thus derived.
h.
The allowable condition restoration times shall be established based on level of equipment availability required to assure an acceptable level of safety and should consider events that will reduce the level of availability such as surveillance and maintenance.
i.
When necessary to preserve acceptable channel or train cvailability, condition restoration time requirements shall include establishment of cummulative downtime limits.
J.
Each LCO shall include a cross-reference to the surveillance requirements that support the LCO, except that surveillance shall not be required if the normal operating status of equipment or systems, for the applicable operational modes, equals or exceeds the lowest functional capability of performance level relied upon in the safety analysis.
k.
The bases for limiting conditions for operation shall identify the safety analysis assumption or other safety requirement that establishes the need for the LCO, and shall discuss why the specified lowest functional capability, performance level of equipment, limiting value of a process parameter, or conservative actuation limit for specified automatic protection devices is appropriate. The rationale for deviations from the specified conditiens as allowed by remedial action statements shall also be discussed.
SERVIGE COMPANY OF COLORADO
'D PUBLIC US-TSi FORT ST. VRAIM NUCLEAR GENERATING STATION Zssue A Page 11 of 16 SPECIFICATION CONTINUATION SHEET FORM 344
- 22 4083 7.
Criteria for Surveillance Requirements (SR) a.
Surveillance requirements shall delineate testing, calibration, monitoring, and inspection in sufficient scope, depth, and frequency to provide assurance that equipment, systems and process variables are within limiting conditions for operation.
Each limiting condition for operation shall be supported by a surveillance requirement except where the normal operating status of equipment or systems, for the applicable operational modes, equals or exceeds the lowest functional capability or performance level relied upon in the safety analysis.
Every surveillance requirement shall be cross-referenced to a limiting condition for operation, or to an administrative control.
b.
Minimum disturbance of normal plant operation should be assured by relating surveillance requirements to normal operational cycles such as the refueling period, where practical.
c.
Customary surveillance scopes, depths and frequency which hav,e been found compatible with an acceptable level of safety shall be employed unless sufficient design, operation, or research information suggests alternate approaches. The Standard Technical Specifications may be used for guidance in this regard.
d.
The surveillance program shall demonstrate acceptable availability for equipment for which there is limited experience or reliability data.
(A sliding surveillance frequency can be established by choosing an initial surveillance frequency with provision to lengthen or shorten the time between tests based on experience gained with the equipment involved).
e.
The surveillance shall be consistent with the requirements of recognized and relevant industry codes and standards:
f.
Where it is not obvious that the surveillance supports the LCO, the bases shall describe how the specified surveillance will assure compliance with the LCO.
The rationale for the surveillance frequency shall be identified to facilitate consistent modifications to the frequencies where warranted by plant performance.
PUBLIC S E R V I C E' COMPANY OF COLORADO WS-TS-1 FORT ST. VRAIN NUCLEAR GENERATING STATION Issue A Page 12 of 16 SPECIFICATION CONTINUATION SHEET P c R M 3M. 22 4083 8.
Criteria for Design Features (DF) a.
Design features of the facility which, if altered or modified, could have significant effects on safety and are not covered by the safety limits, limiting conditions for operation, or surveillance requirements shall be incorporated in the design features section of the technical specifications.
b.
Particular sections or criteria of the FSAR may be referenced as an alternative to providing design details in the cechnical specifications; however, such references should be limited and specific since referenced criteria or features will become part of the technical specifications and cannot be changed under the provision of Title 10, Code of Federa.1 Regulations, Part 50, " Licensing of Production and Utilization Facilities," Section 50.59, " Changes, Tests and Experiments," without Commission approval.
References to the FSAR that provide further information but are not intended to be part of the technical specification, should be located in the bases.
c.
Provisions should also be included to allow for normal degradation of design features where applicable.
9.
Criteria for Administrative Controls a.
Administrative controls shall be included in the technical specification to assure that operation of the facility is conducted in a safe manner.
Implicit in this are the requirements for: organization; procedures; record keeping; review; audit; reporting; staffing qualifications and resolution of safety limit violations.
b.
Specific responsibility and authority shall be delineated for those portions of the organization charged with fulfilling these requirements, c.
The administrative controls shall also require that tne facility procedures include those operator actions relied upon in the safety analysis.
d.
Additional guidance can be found in other standards and regulatory guides for:
l
PUBLIC SERVICE COMPANY OF COLORADO WS-TS-1 FORT ST. VRAIN NUCLEAR GENERATING STATION Issue A Page 13 of 16 SPECIFICATION CONTINUATION SHEET FORM 344
- 22 - 4043 (1) Administrative controls: American National Standard, " Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plant," N18.7-1976/ANS-3.2; (2) Selection and training of personnel: American National Standard, " Selection and Training of Nuclear Power Plant Personnel," N18.1-1971/ANS-3.1; and (3) Reporting of operating informatien: Regulatory Guide 1.16, Revirion 4, August 1975, " Reporting of Operating Information - Appendix A, Technical Specifications.
D.
FSAR REVIEW CRITERIA 1.
The entire Fort St. Vrain Final Safety Analysis Report shall be reviewed to identify the underlying assumptions used to determine that operation of the plant does not present an undue risk to the health and safety of the public.
2.
The essential safety functions that protect the health and safety of the public are those related to:
a.
Protecting the integrity of fission product boundaries.
b.
Controlling reactivity.
c.
Cooling the fuel.
d.
Limiting the release of radioactive fission products, and e.
Mitigating the consequences of accidents and natural and manmade phenomena.
3.
The underlying assumptiors to be identified consist of:
a.
Values of process variables that must be kept within certain bounds.
b.
Operating state of equipment that must be maintained.
l c.
Operating status (or operability) of equipment that l
must be maintained.
d.
Condition (or quality) of equipment and structures I
that must be maintained.
l
PUBLIC S E R V I C E' COMPANY OF COLORADO t!S-TS-1 FORT ST. VRAIN NUCLEAR GENERATING STATION Issue A Page 14 of 16 SPECIFICATION CONTINUATION SHEET froRM 344
- 22
- 4083 e.
Physical characteristics of the plant that must remain fixed, and f.
Administrative controls that must be maintained.
4.
Underlying assumptions that are expected to, or could vary with time or circumstances, throughout the life of the plant shall be identified as being subject to technical specification control. A list of these items shall be forwarded to the Program Coordinator.
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ATTACHMENT 2 i
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ATTACHMENT 2 TECHNICAL SPECIFICATION UPGRADE PROGRAM PROJECT SCHEDULE I
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to P-84530 Page 1 of 5 TECHNICAL SPECIFICATION UPGRADE PROGRAM PROJECT SCHEDULE
- The Technical Specification Upgrade Program includes two parallel reviews, one to review the FSAR for items that should be included in the. Technical Specifications, and the other to review the Technical Specifications for accuracy, ' completeness, and clarity.
The FSAR. review is scheduled to begin on December 17, 1984, and is to
. be completed by February 15, 1985, as follows:
FSAR SECTION COMPLETION DATE 1, 2 & Appendix G 1/04/85 3 & 4 & Appendix A 1/11/85 5 & 6 & Appendix E 1/18/85 7&8 1/25/85 9 & 10 & Appendix H & I 2/01/85 11, 12 & 13 2/08/85 14 & Appendix D 2/15/85 The Technical Specification review has been initiated. The Technical Specifications have been grouped into work packages, based on their subject matter, and priorities have been assigned for their completion. The schedule for work package completion is as follows:
i High Priority work packages are expected to be c9mpleted from January 15, 1985 through March l',.1985; Medium Priority work packages are expected to be completed from February 1, 1985 through March 15, 1985; and Low Priority work packages are expected to be completed from February 15, 1985 through March 20, 1985.
n The work packages, their assigned priority, and the Technical Specification sections that they include are listed below.
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to P-84530 Page 2 of 5
-Work' Package Description Section/LCO SR High Priority 2
Definitions 2.0 4
3 Reactor Core 3.1, part 3.3 5.1.6 14 Reactor Vessel 3.2, part 3.3 5.
Core Irradiation 4.1.1 6
Control Rods 4.1.2 5 ~.1.1 4.1.3 5.1.5 4.1.4 7
Reactivity 4.1.5 5.1.2 4.1.6 5.1.3-4.1.8 5.1.4-8 Inlet Orifice Valves-4.1.7 5.1. 7.
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4.1.9 5.4.3 5.4.8 9
Primary Coolant System 4.2.1 5.2.7 4.2.2 5.2.8 4.2.3 5.2.9
,4 4.2.4 5.2.18 i
4.2.5 5.2.23 4.2.19 5.2.24-5.2.27 1
17 Steam Generators /
4~.3.1 5.2.24 Safe Shutdown Cooling-4.3.2 5.3.2 4.3.4 5.3.4
+
4.3.5 5.3.3
~4.3.6 5.3.5 4.3.7' 5.3.6 -
5.3.7 5.3.10
.0 PPS Instrumentation-4.4.1 5. 4.1 -
2 5.4.3 5.4.4
'5.4.5
.4.6 5
5.4.7 5.4.8-p-
a' to P-84530
'Page 3 of 5 ii Work-Package Description Section/LCO SR
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24 Analytical /PPS 4.0.5 5.4.12 Moisture Monitors 4.9.2 26 Auxiliary Electrical 4.6.1 5.6.1 5.6.2 Medium Priority 10 Firewater Systems 4.2.6 5.2.10-4.10.5 5.10.6 4
4.10.7 5.10.8 4.10.8 5.10.9 11 PCRV Pressurization 4.2.7 5.2.1 4.2.9 5.2.13-4.5.2' 5.2.14 4.7.1 5.2.15 5.2.16 5.2.28
~5.4.1 5.3.9 12 Primary / Secondary 4.2.8 5.2.6.
Activity 4.3.8 5.2.11 5.3.7 -
13 Loop Impurity Levels 4.2.10 5.2.12-2 4.2.11-5.2.22 5.2.25 5.4.12-15; PCRV' Liner Cooling 4.2.13 5.4.4 4.2.14 5.4.5.
4.2.15 5.4.11 18' Stetm Water Dump 4.3.3 5.3.1 Tank 5.3.7
-31 Room Isolation Damper /
4.10.1 5.10.1 Halon Fire Suppression 4.10.2 5.10.2 32 Smoke Detectors 4.10.3 5.4.2 5.10.3 33 Fire Barrier 4.10.4 5.10.4 Penetration Seals to P-84530 Page 4 of 5 Work Package Description Section/LCO SR 1-34 CO2 Fire Suppression 4.10.6 5.10.7 System 40 Breathing Air System 5.10.5 Low Priority 1
Introductory (Editorial 1.0 5.0 Material) 3.0 5.1 4.0 5.2 4.1 5.3 4.2 5.4 4.3 5.5 4.4 5.6 4.5 5.7 4.6 5.10 4.7 6.0 4.9 6.2 4.10 7.0 14 LN2 Storage 4.2.12 16 ACM Diesel Generator 4.2.17 5.2.20 5.2.21 19 Shock Suppressors 4.3.10 5.3.8 21 Control Room /480V'-
4.4.2 5.4.7 Room Temperature 4.4.6 5.4.13 22 Area Radiation 4.4.3 5.4.9 Monitors
.23 Seismic Insturmentation 4.4.4
.5.4.10
-25 Reactor Building 4.5.1 5.5.1 5.5.2 5.5.3 27 Fuel Handling Machine 4.7.2 5.7.1
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28 Fuel Storage Facility 4.7.3 5.7.2 29 Spent Fuel Shipping 4.7.4 Container 30
-Fuel Leading & Initial 4.9.1 Rise to Power
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to P-84530 Page 5 of 5 Work Package Description Section/LCO SR i.
35 Tendon Surveillance 5.2.2 5.2.3 36 Concrete Surveillance 5.2.4 37 Liner Specimen 5.2.5 Surveillance 38 RCD Surveillance 5.2.26 39 S/G Bimetallic Welds /
'5.3.11 Tubeleaks 5.3.12 41 Design Features 6.1 6.2.1 6.2.2 6.2.3 6.3 42 Administrative Controls 7.1 7.1.1 7.1.2 7.1.3 43 Safety Limits 7.2 44 Records
.7.3
'45 Procedures 7.4 46 Reporting Requirements 7.5
'7.5.1 7.5.2 7.5.3 47 Environmental Qual-7.6 ification 48
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Depressurization/ Helium 4.2.18 Purification n
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i t