ML20059D233

Safety Evaluation Supporting Amend 149 to License DPR-72

ML20059D233
Person / Time
Site:	Crystal River
Issue date:	12/20/1993
From:	Office of Nuclear Reactor Regulation
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ML20059D231	List: ML20059D233
References
NUDOCS 9401070080
Download: ML20059D233 (38)

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UNITED STATES

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ij NUCLEAR _ REGULATORY COMMISSION e

WASHINGTON, D.C. 20666-0001

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENOMENT NO. 149 TO FACILITY OPERATING LICENSE NO. DPR-72 FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302 I

1.0 INTRODUCTION

By letter dated August 25, 1989, the Florida Power Corporation-(FPC). proposed to amend Appendix A of Operating License No. DPR-72 to revise, in its-entirety, the Crystal River Unit 3 technical specifications. The' proposed amendment was based on guidance provided in the "NRC-Interim Policy. Statement-on Technical Specification Improvements for_ Nuclear Power Reactors," published on February 6, 1987 (52 FR 3788). During its review, the NRC staff relied on the NRC's Interim Policy _ Statement and later on NUREG-1430, " Standard Technical Specifications - Babcock And Wilcox Plants" which was issued in September-1992.

The NRC's proposeo action on the amendment-request was published in the Federal Reaister on November 8, 1989 (54 FR 46998).

The overall objective of -

the proposed amendment, consistent with the NRC Interim Policy Statement, was to completely rewrite, reformat, and streamline the existing Crystal River technical specifications (TS). The Commission's policy envisioned that the TS conversion process would result in transferring some TS requirements to other licensee-controlled documents.

Emphasis was placed on human factors principles' to add clarity and understanding to the improved Crystal River TS ~

and to define more clearly the appropriate scope of the TS.- In addition, significant changes were proposed to the Bases section of.the Crystal River TS to enhance the clarity and understanding of each specification.

a Crystal River currently operates with TS issued with'the original operating license on: December 3, 1976, as amended from time to time over the years.

FPC's present proposal to revise the Leystal River TS was based on a Babcock &

Wilcox (B&W) Topical Report, "B&WOG Revised. Standard Technical Specifications" (BAW-2076). During 1989 through~1992, tha utility Owners Groups and the NRC/

.i staff developed improved standard technical' specifications (STS) that would. _

l establish.models of the' Commission's policy for each primary reactor _ type. < 0n July 22, 1993, the-Commission issued a " Final Policy Statement on Technical' Specification Improvements for Nuclear Power Reactors" (58 FR 39132)~.

The Commission's policy statement. described the safety benefits of the improved 1

STS and encouraged licensees to use the improved STS as the basis for plant specific TS, particularly complete conversions to the improved STS.. Moreover.

j the policy statement'provided guidance to evaluate the scope'of the technical j

specifications.

The guidance in the interim and final policy statement-were l

used to develop the NUREG-1430, which serves as a model for developing j

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improved technical specifications for B&W plants.

The policy statement reflects the Commission view that satisfying the guidance in the policy statement also satisfies $182a of the Atomic Energy:Act and 10 CFR 50.36..The staff finds that the technical specifications issued with'this license amendment satisfy the guidance in the Commission policy statement, Section 182a of the Atomic Energy Act, and110 CFR 50.36, and;are.in.. accord; with the common defense and security and will provide adequate protection to the health and safety of the public.

NUREG-1430 was established as the model for B&W. plants in general and the improved Crystal River TS specifically.

Portions of the existing TS were also used as a basis. for the, improved Crystal River TS.

Plant-specific issues, including plant-unique design features, plant-unique requirements,. and plant-unique operating practices were discussed with the licensee during a series'of meetings concluding on May 26, 1993.

In addition, meetings were held with the Owners Groups to di.scuss matters of a generic nature that were not incorporated in NUREG-1430.

These generic issues were considered for. specific applications in the Crystal River improved Crystal River TS.

Changes in the licensee's proposed TS that resulted from discussion with the -

licensee during the review are discussed in the.following sections. These plant-specific changes serve to clarify the TS with respect to the' guidance in the Commission's policy statement, the guidelines in NUREG-1430, and'do not affect the intent of the specifications. Therefore, the changes are'within.

the scope of the' action described in the Federal Reaister (54 FR.46998) on November 8, 1989.

This safety evaluation documents the basis for the staff's conclusion that Crystal River can convert its existing TS to those based on NUREG-1430, modified by plant-specific changes, and that the use' of these improved Crystal River TS is acceptable for continued plant operation.

Individual section.

topics and the corresponding section numbers are identical to.those given in NUREG-1430. The staff has identified the changes to the existing Crystal River TS, and has included an explanation of the significant changes in this evaluation. The staff also acknowledges;that, in accordance with the policy statement, the conversion to the STS is a voluntary process.

Therefore,.the-improved Crystal R1ver TS for Crystal River reflect some ' differences that correspond to the existing licensing basis for the plant.

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Since the original August 25, 1989 submittal, FPC has submitted, and the. staff has. accepted, a number of changes to the existing Crystal River TS. The review and approval' of these.TS amendments was independent of the Crystal River improved TS review effort.

These changes are reflected, as appropriate, in the Crystal River improved TS. This safety evaluation describes only those TS amendment changes which affected implementing the Crystal River improved TS.

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2.0 DISCUSSION AND EVALUATION FPC has proposed changes to the existing Crystal River TS using the policy statement and NUREG-1430 as guidance.

Several changes from NUREG-1430 were also proposed by Crystal River due to differences in the plant-specific design and licensing basis.

The proposed changes from that of the present Crystal-River TS can be grouped into four general categories: non-technical (administrative) changes, relocation of requirements to other licensee-controlled documents, more -

restrictive requirements than specified in the existing TS, and less restrictive requirements than specified in the existing TS.

This evaluation describes the substantive changes to the existing TS requirements in the following general areas:

Administrative Chances P

Non-technical, administrative changes were intended to incorporate human-factors principles into the form and structure of the improved Crystal River TS so that they would be easier to use for plant operations personnel. These changes are editorial in nature or involve the reorganization or reformatting i

of requirements without affecting technical content.

Every section of the proposed TS reflect this type of change.

In order to ensure consistency, the

,RC staff and FPC have used NUREG-1430 as guidance to reformat and make other administrative changes.

FPC has proposed changes such as:

(a) providing the appropriate numbers, etc., for NUREG-1430 bracketed information (information which must be supplied on a plant-specific basis, and may change from plant to plant), (b) identifying plant-specific wording for system names, etc., and (c) changing NUREG-1430 section wording to conform to existing Crystal River practices.

The staff, FPC, and the Owners Groups have developed generic administrative and editorial guidelines in the form of a " Writers Guide" for technical specifications which has been used throughout the development of the Crystal River improved TS. The staff believes that this guidance provides for a significant enhancement of the human factors aspects of the overall Crystal River improved TS.

The staff has reviewed all of the administrative and editorial changes proposed by FPC and finds them acceptable since they are compatible with the " Writers Guide" and NUREG-1430, and are consistent with the Commission's regulations. The more significant non-technical administrative changes are discussed individually in this. evaluation'.

Relocated Recuirements Requirements that were in the existing Crystal River TS, but did not meet the guidance set forth in the policy statement for inclusion in TS, would be relocated to other licensee-controlled documents.

The following are the

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l guidance for requirements that must be included in technical specifications, as presented in the Commission's final policy statement:

Criterion 1:

Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor. coolant pressure boundary; Criterion 2:

A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3:

A structure, system, or component that is part of the primary success path :nd which functions or actuates to mitigate a Design Basis Accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; and Criterion 4:

A structure, system, or componc t which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

Requirements that were in the existing Crystal River TS, but do not meet the guidance set forth in the policy statement for inclusion in TS, will be relocated to appropriate licensee-controlled documents.

In general, the licensee has pruposed to relocate these items to the FSAR, plant-specific procedures, programs, and improved Crystal River TS Bases.

These relocated requirements are described in more detail in the following evaluation. Unless otherwise specified in this safety evaluation, the limiting conditions for operation (LCO) portion of the existing TS, which includes the system description, design limits, functional capabilities, and performance levels, will be relocated to the Final Safety Analysis Report (FSAR).

The provisions of the existing TS action statements and surveillance requirements will be relocated to appropriate plant. procedures; i.e., operating procedures, maintenance procedures, surveillance and testing procedures, and work. control procedures, depending on the nature of the requirements being relocated.

These procedures will similarly be described in the FSAR.

Although.the FSAR already includes most of the design information described' above, by letter dated November 23, 1993, the licensee committed to confirm that these details are appropriately reflected in the FSAR or will be included in the next revision to the'FSAR.

The facility and procedures described in the FSAR can only be revised under the provisions of 10 CFR S0.59, which ensures an auditable and appropriate control over the relocated requirements and future changes to these

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. provisions. Other licensee-controlled documents include provisions for making changes, consistent with the applicable regulatory requirements; for example, the Offsite Dose Calculation Manual (0DCM)_can be changed in accordance with 10 CFR 20, the Emergency Plan Implementing Procedures (EPIP) can be changed in-accordance with 10 CFR 50.54(q), and the administrative instructions that.

implement the Quality Assurance Plan (QA Plan) can be changed in accordance with 10 CFR 50.54(a) and 10 CFR 50, Appendix B.

Temporary procedure changes are also controlled by the administrative instructions that implement the QA plan.

The documentation of these changes will be maintained by FPC in accordance with the record retention requirements specified in their QA Plan.

The licensee has committed, in their letter dated November 23, 1993, to maintain an auditable record of and an implementation schedule for the procedure changes associated with the development of the improved Crystal River TS.

The requirements relocated from the existing Crystal. River TS to other licensee-controlled documents are summarized in Attachment 1.

As described in more detail in this evaluation, the staff concludes that appropriate controls have been identified for all of the requirements that are being relocated from the technical specifications to other licensee-controlled documents. Until incorporated in the FSAR and proe.edures, changes to the provisions being relocated from the TS will be controlled in accordance with the applicable existing procedures that control these documents. The NRC will conduct an audit of the relocated requirements following implementation to assure that an appropriate level of control has been achieved.

More Restrictive Reauirements The proposed Crystal River improved TS include certain.more restrictive requirements than the existing TS, which are either more conservative than corresponding requirements in the existing TS, or are additional restrictions which are not in the existing TS, but are contained in NUREG-1430.

Examples of more restrictive requirements include:

placing-an LCO on plant equipment which is not required by the present TS to be operable; more restrictive requirements to restore-inoperable equipment; and more restrictive surveillance requirements. These more restrictive requirements are discussed individually in this evaluation.

Less Restrictive Reauirements Less res_trictive requirements are justified on a case-by-case basis as discussed in Section 2.1.1 through 2.5.0 of this evaluation.

When requirements have been shown to provide little or no safety benefit, their removal from the TS may be justified.

In most cases, relaxations previously granted to individual plants on a plant-specific basis were the result of generic NRC actions, new staff positions that have evolved from technological 1

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advancements and operating experience,.or as a result of the resolution of the Owners Groups comments ~on the draft STS. Generic relaxations contained in NUREG-1430 have been reviewed by the.NRC staff and have-.been found acceptable because they are consistent with current: licensing practices and the Commission's regul.ations.

The Crystal River design was reviewed to determine if the specific design. basis and licensing basis are consistent.with the technical basis for the model requirements.in NUREG-1430, and thus provides: a basis for these revised TS.

One generic change that has been made in STS, and is reflected in the improved-Crystal River TS proposed for. Crystal River, is the change from an 18-month to

l a 24-month surveillance interval associated _withL longer refueling cycles.

These changes are reflected.in NUREG-1430 consistent with the guidance.in Generic Letter 91-04, " Changes in Technical Specification Surveilla'nce Intervals to Accommodate a 24 Month Fuel Cycle."

The following sections explain how the staff has concluded that the conversion of the existing Crystal River TS to those based on NUREG-1430, as modified _ by.

plant specific changes, is consistent with the current Crystal River licensing basis and the requirements and guidance of the policy. statement.

i 2.1 Use and Application (Section 1.0) t The definitions appearing in Section 1 of the Crystal River improved TS have been reorganized from the existing Crystal River TS by deleting the 3

identification numbers associated with each definition and listing them in alphabetical order.

Changes to the defined terms in this section.are.as follows:

The following definitions have all been retained in the Crystal River improved TS. Some editorial changes have been made so that these defined _ terms are consistent with NUREG-1430 and with Crystal River plant-specific terminology.

The modifications have been accepted by the licensee and, based on our review, do not change the -intent of the definitions as. found'in NUREG-1430.

Therefore, we find these definitions acceptable for Crystal. River.

ACTIONS ALLOWABLE THERMAL POWER AXIAL POWER-IMBALANCE

-i CHANNEL CALIBRATION a

CHANNEL CHECK l

CHANNEL FUNCTIONAL TEST CORE ALTERATION

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CORE OPERATING LIMIT REPORT DOSE EQUIVALENT I-131 E AVERAGE DISINTEGRATION ENERGY ENGINEERED SAFETY FEATURE RESPONSE TIME

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. TYPES OF LEAKAGE OPERATIONAL MODE (including Table 1.1)

OPERABLE-0PERABILITY PHYSICS TEST QUADRANT POWER TILT RATED THERMAL POWER REACTOR PROTECTION SYSTEM RESPONSE TIME SHUTDOWN MARGIN.

STAGGERED TEST BASIS THERMAL POWER The following new definitions have been added:

AXIAL POWER SHAPING RODS CONTROL RODS EFFECTIVE FULL POWER DAYS EMERGENCY FEE 0 WATER INITIATION CONTROL. RESPONSE TIME NUCLEAR HEAT FLUX HOT CHANNEL FACTOR NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR PRESSURE AND TEMPERATURE LIMITS REPORT The addition of these new definitions is compatible with changes made throughout the Crystal River improved TS to clarify the related requirements, and to reduce the likelihood of misinterpretation of the new TS.

Each new Crystal River definition was also defined in NUREG-1430.

Plant-specific wording differences have been reviewed and do not change the meaning of these definitions.

P All other definitions in the existing Crystal River TS (1.1,1.7,1.8,1.22, 1.27, 1.28, 1.30, 1.31, 1.33, 1.34, 1.35, 1.37, 1.38, 1.39, and Table 1.2) are-no longer used as defined terms in the' Crystal River ' improved Crystal River TS. However, definitions 1.22 and Table 1.2 have been reformatted-and these concepts are contained in the Crystal River improved TS in-Sections 1.4.

As noted above,-the staff and the licensee have agreed _to minor _ word changes j

throughout the Crystal River definition section.

These word changes are clarifications that do not alter the meaning of the definitions or' change the restrictive level of the TS. The definitions in Section 1.0 of the improved Crystal River TS perform a supporting function for other sections in the Crystal River improved TS. The staff has reviewedithe proposed changes in the definition section for their effect on the Safety Limits-(SL) and SL.

violations that appear in Section 2.0 and the LCOs and Action Statements in Section 3, including the Surveillance Requirements (SR).

The staff finds no adverse effects that would result from the proposed changes and concludes that when the definitions, as modified, are applied in other sections of the TS, the restrictive level of the requirements are not changed and, therefore, the

.. safety margins are not affected.

In addition, the staff concludes that the licensee's proposed changes clarify the definitions and would reduce the tendency for misinterpretation.

The proposed changes to the definitions maintain the restrictive level of the TS.

The staff finds that the proposed Crystal River improved TS has appropriately applied the guidance provided in NUREG-1430. Therefore, we find the changes acceptable.

i 2.1.1 Logical Connectors (Section 1.2)

This is a new section in the Crystal River improved TS. This section explains the meaning and use of " Logical Connectors" through the use of examples so that the entire Crystal River improved TS are clearer from a human factors standpoint. We have reviewed this section and consider this proposed addition and reformatting an enhancement to the Crystal' River improved TS. We find the addition to be consistent with NUREG-1430 and therefore acceptable.

2.1.2 Completion Times (Section 1.3)

This is a new section in the Crystal River improved TS.

This section does not change completion times, but provides guidance through the use of examples on the use of " Completion Times." " Completion Time" is the amount of time allowed to complete an action or the amount of time allowed for a structure, i

system or component to be inoperable. This section is administrative in nature and is provided as an aid to the _ licensee's staff. We have reviewed this section, find it is consistent with NUREG-1430 and therefore acceptable.

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2.1.3 Frequency (Section 1.4) 1 This is a new section in the Crystal River improved TS. This section defines the proper use and application of surveillance frequency practices, through the use of examples. A clear understanding of the correct application of a specified Frequency is necessary to ensure compliance with a surveillance requirement.

i We have reviewed this section and find that the " Frequency Notation" definition and the " Frequency Notation Table" (Definition 1.22 and Table 1.2, respectively) of the existing TS have been adequately-incorporated into the descriptions and examples of this section. We find that this section is consistent with NUREG-1430 and therefore acceptable.

2.2 Safety Limits (Section 2.0) i This section has been renamed from the existing Crystal River TS, Section 2.0

" Safety Limits and Limiting Safety System Settings." Although renamed, this i

Section contains essentially the same information as the existing Section 2.0.

Information not retained in this section is contained elsewhere within the Crystal River improved TS or other licensee documents.

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This section.has been reformatted and reorganized to separate the safety limits and the safety limit violations.

The staff has reviewed FPC's proposed Section 2.0, based on NUREG-1430, as modified to include plant specific limits and terminology, and finds this section is consistent with the Commission's regulations and is acceptable.

2.3 Limiting Conditions for Operation (Section 3.0)

This section has been renamed from the existing TS section " Limiting Conditions for Operation and Surveillance Requirements" to the improved Crystal River TS section entitled " Limiting Condition for Operation (LCO)

Applicability." The following covers changes made throughout Section 3.0.

The licensee has not proposed to relocate any specific LCOs in Section 3.0 out of the improved Crystal River TS to other licensee-controlled documents.

However, the licensee proposed to add two new LCOs from NUREG-1430 (3.0.5_and 3.0.7) to the improved Crystal River TS.

1.

LCO 3.0.5 permits equipment removed from service to be returned under administrative control to perform testing to determine operability.

2.

LC0 3.0.7 is being added to permit certain physics test exceptions.

In addition, clarifying statements have been added to SR 3.0.3 which quantify.

and clarify the maximum time delay or allowance that is permitted to perform a given surveillance.

The staff has reviewed these proposed additions and concludes that the additions will enhance the quality of the Crystal River improved TS and will benefit the operators and others in their understanding of the overall improved TS.

The staff concludes that the proposed Crystal River improved TS have made appropriate application of the guidance provided in NUREG-1430.

Therefore, the changes are acceptable.

2.3.1 Reactivity Control Systems (Section 3.1)

On the basis of. the guidance in the policy statement, FPC has proposed to relocate or reorganize the following existing TS:

Existina TS Number Title 3/4.1.2.1 Boration Flow Path - Shutdown 3/4.1.2.2 Boration Flow Path - Operating 3/4.1.2.3 Makeup Pump - Shutdown 3/4.1.2.4.1 Makeup Pump - Operating (MODES 1,2, and 3) 3/4.1.2.4.2 Makeup Pump - Operating (MODE 4)

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Existino TS Number Title 3/4.1.2.5-Decay Heat Removal Pump - Shutdown 3/4.1.2.6 Boric Acid Pumps - Shutdown 3/4.1.2.7 Boric Acid Pumps - Operating 7

3/4.1.2.8 Borated Water Sources - Shutdown

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3/4.1.2.9_

Borated Water Sources - Operating Normal reactor coolant system (RCS). boron control is needed. to help maintain shutdown margin during both power operation and shutdown. The boration flow path from'the concentrated boric acid storage tank is not~ assumed operable'and is not required-for mitigation of a Design Basis Accident-(DBA). 'The flow paths and pumps required to support emergency core cooling. system'(ECCS).

-l analysis assumptions for Design Basis Accident-(DBA) _ mitigation are controlled by separate retained specification 3.5.2.

The borated water storage tank (BWST) is required to supply borated injection water-for ECCS and-can also supply the required boric acid for shutdown. The BWST requirements are in retained specification 3.5.4.

The systems and components supporting the ECCS boration requirements are being retained in the improved ECCS TS 3.5.1 and 1

3.5.4.

The boric acid pumps are typically used to transfer. concentrated boric acid from the concentrated boric acid storage tar.k to the makeup tank as part of planned normal RCS boron control activities.

Normal RCS boron ~ control is needed to help maintain shutdown margin during both power' operation and~

shutdown.

The boric acid pumps are not required to mitigate a DBA. The FSAR analyses for potential reactivity insertion events',. such. as moderator dilution, show that reactor shutdown margin is maintained by control rods.

Normal RCS boron control is needed to help maintain shutdown. margin during both power. operation and shutdown. The concentrated boric ac.id storage tank is not assumed operable and is not required ~to mitigate a.DBA.

-In some areas, the licensee has chosen to follow.the guidance given in-NUREG-1430 with regard to the addition'of new TS with accompanying LCOs, Conditions, Required Actions, Completion Times,.and Surveillance Requirements.

In addition, separate and new Bases have been added for these TS. The following new TS are added to Section 3.1 of thelimproved Crystal ~ River TS:

' 1.

3.1.2 " Reactivity Balance":was contained in existing TS as SR 4.1.1.1.1.2.

The Required' Actions have been updated and new Completion _ Times have been added.

2.

3.1.8 and 3.1.9 " Physics Tests Exceptions" were contained in' existing TS 3/4.10.

NUREG-1430 contains all of-the exceptions in these two -

sections and in a format that is easier to understand..Therefore, the licensee has _ elected to reformat and move all " Physics Tests Exceptions" into improved Crystal River TS Sections 3.1.8 and 3.1.9.

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The BWST requirements to support ECCS analysis assumptions.for DBA mitigation are. covered in separate technical specifications. -The borated water storage i

tank requirements to support ECCS boration are retained in.the ECCS technical.-

specification 3.5.4.

The staff concludes that the proposed Crystal River improved TS provisions described above are. consistent with the policy statement and the guidance provided in.NUREG-1430 and the Commission's

.i regulations, and are, therefore, acceptable.

Except for the retained provisions described'above, the licensee has proposed to relocate the foregoing provisions to the FSAR and appropriate plant procedures, as previously described.

Based on a deterministic. review, the' staff concludes that'these provisions do not need to be controlled by.TS under a

the policy statement or the Commission's regulations.

Therefore, the control.

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of these provisions under 10 CFR 50.59 is acceptable.

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2.3.2 Power Distribution Limits (Section 3.2)

Improved Crystal River TS Section 3.2 contains two less restrictive requirements than the existing TS, as follows:

1.

Improved Crystal River TS Section.3.2.3 has been modified to be consistent with NUREG-1430 which reflects' the current NRC staff j

positions on " Axial Power Imbalance."

l The accident (LOCA) analysis is performed assuming that the value of

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each of the process variables that characterize and control the three-1 dimensional power distribution of the reactor core (regulating rod insertion limits in LC0 3.2.1), axial power shaping rod insertion limits in LC0 -3.2.2, axial power. imbalance operating limit's in LCO.3.2.3, and j!

quadrant power tilt in LC0 3.2.4), are at the LC0 limits at the j

initiation of the event. With this assumption, the power peaking during-the event does not result in fuel cladding damage. An alternative'means.

of demonstrating this level of protection is.to monitor-the power

.i peaking factors (F, and F",), addressed in LC0 3.2.5.

Dire'ctly demonstrating the power peaking factors to be withinclimits ensures. that..

.i the effect ~of one or more of, the power-peaking-related' LCOs being out of limit is bounded by analysis. The staff finds that a completion time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is acceptable given the surveillance requirement of SR 3.2.5.1 for monitoring the. power peaking factors every.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> -

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Improved Crystal River TS Section 3.2.4 has been modified in accordance

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with NUREG-1430. Of significance was a change in applicability from the existing TS so that TS 3.2.4 would not be applicable'until-the plant was-in Mode I with Thermal Power > 20 percent Reactor Thermal Power (RTP)..

j The existing TS applicability was Mode 1, Thermal Power > 15 percent a

RTP.

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This change was made to achieve consistency between the FSAR and the TS.

The minimum value of quadrant power tilt at which action is taken when tilt is out of specification must consider:the ability to measure and to indicate' tilt, as well as safety. The Thermal Power > 20 percent condition is' consistent with the condition described and analyzed in'the-j FSAR, and the-staff finds that this. improved Crystal' River. TS l

requirement is acceptable.

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2.3.3 Instrumentation (Section 3.3) i In accordance with.the guidance in the policy statement, the licensee has j

proposed to relocate the following existing TS to.other licensee-controlled 1

documents Existino TS Section Title

'l 3/4.3.3.1.1.a.1 Fuel Storage Pool Area Rad Monitor 3/4.3.3.1.2.c Condenser Vacuum Pump Exhaust l

3/4.3.3.1.2.e Decay Heat Closed Cooling Water Rad Monitor The radiation monitors listed above and additional monitors in Table 3.3-6..are l

installed instrumentation which are not used to detect degradation of-the reactor coolant system (RCS) boundary, nor do they measure a process variable

j that is an initial assumption in the DBA or have any automatic-isolation a

function to mitigate DBA radioactive releases.

For process and area monitors

.l in the fuel pool storage areas, no indicator monitors are' assumed in the FSAR

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accident analysis for the' fuel handling accident (dropped or damaged fuel j

assembly).

j The Crystal River FSAR evaluation of the_ fuel handling accident assumed that ~

N all radioactive materials released to the environs from the fuel pool. storage.

j area-(auxiliary building) are unfiltered. This analysis was reviewed and i

accepted by the staff as documented in the operating license.SER: dated December 3, 1976 (page 15-2). As a result, no TS on the_ auxiliary building.

l ventilation system for the' fuel.. handling accident is necessary since ;it was not assumed to mitigate the consequences in the licensing basis evaluation of

'j the fuel. handling accident.

The provisions of existing TS 3/4.3.3.1.1.a.1~will be relocated to the FSAR.

and appropriate plant procedures, as previously described.

The provisions of

i existing TS 3/4.3.3.1.2.c and TS 3/4.3.3.1.2.e will be relocated to the FSAR,-

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except for those radiation monitoring practices that will be reflected in the l

ODCM. ~ Based on a deterministic review, the staff concludes that these.

j provisions do not need to be controlled by TS under the policy statement or' the Commission's regulations..Therefore, the control of these provisions ~

j under 10 CFR 50.59 and the ODCM is acceptable.

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The licensee has similarly proposed to modify or relocate the following instrumentation requirements:

1 Existino TS Section Title i

3/4.3.3.2 Incore Detectors 3/4.3.3.~3 Seismic Instrumentation 3/4.3.3.4 Meteorological' Instrumentation 3/4.3.3.11 Chlorine / Sulfur Dioxide Detection 3/4.3.3.8' Radioactive Liquid Effluent l

3/4.3.3.9*

Radioactive Gaseous Effluent Incore Detectors

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The incore detector system is used to provide detailed information on the reactor core neutron flux distribution. This information is used to i

verify that the axial-power distribution and quadrant tilt are within i

their limits. No automatic actions result from'the incore detector system.

The power range neutron flux instrumentation is also available a

to-measure axial power distribution (axial' imbalance)_and quadrant power tilt. The reactor protection system uses the power range neutron flux instrumentation to initiate a. reactor. trip due to unacceptable axial core power distribution. The provisions will be relocated to the FSAR and appropriate plant procedures, as previously' described except for the required-actions.

The existing action. statement is no longer ' applicable -

since the required action for an inoperable system directed the operator ~

1 to use other TS instrumentation. This instrumentation is still:in the TS and would be used to provide this' safety function. Thus, the action' statement'is redundant ~and can be deleted.

s Seismic Instrumentation In the event of.an earthquake, seismic instrumentation is required t'o1 permit comparison of-the measured response to; that~ used in the design 1

basis of the facility to determine if._ plant shutdown is required pur-

- suant to Appendix'"A" of 10 CFR Part 100.

Since this is determined-j after the event has occurred, it has no bearing on!theLmitigation of anyi j

DBA. The provisions will be relocated to the FSAR and appropriate plant,

l procedures, as previously described.

1 Meteoroloaical Instrumentation

' Meteorological instrumentation is used to measure environmental-j parameters which may affect distribution of fission products and gas _es

• Relocated via Amendment #141, 5/4/92.

b

4

, following a DBA; however, it is not a primary success path for the mitigation of a DBA. This provisions will be relocated to the FSAR, as previously described..except for.the accident and radiological practices associated with the meteorological instrumentation that will be reflected in the ODCM and the EPIP.

Chlorine / Sulfur Dioxide Detection Chlorine / sulfur diox.or adection coability is provided to warn personnel in the event of an accidental chlorine or sulfur dioxide release and to isolate the control room and is consistent with the rsemmendations of Regulatory Guide 1.95 to isolate the control room.

Although the toxicity of chlorine or sulfur dioxide presents a personnel hazard, escape or. detection of toxic gas is not part of any design basis event for reactor or containment transients.

The provisions will be relocated to the FSAR and appropriate plant procedures, as previously described.

The licensee has proposed to modify or relocate the foregoing provisions for instrumentation detection systems to the FSAR and appropriate plant procedures, as previously described, except for those provisions that will be controlled by the ODCM and the EPIP.

Based on a deterministic review, the.

staff concludes that these provisions do not satisfy the policy statement guidance for inclusion in TS. Therefore, the control of these provisions-under 10 CFR 50.59, the ODCM, and the EPIP is acceptable.

NUREG-1430 has TS that were not specifically included in the existing TS.

The licensee has evaluated these new TS and has agreed to include the following in the improved Crystal River TS Section 3.3:

1.

The improved Crystal River TS specification for the Emergency Diesel Generator loss-of-Power Start (Section 3.3.8) is new for Crystal River.

Mitigation of most plant transients or accidents involves the actuation and operation of electrical supply equipment. Operation of this equipment is dependent upon availability of an adequate source of AC power. While the necessary electrical supply bus power would normally be provided from the 230 kV offsite transmission system, the safety analysis conservatively considered offsite power to be lost coincident with initiation of numerous DBAs. Assumptions relative to power source OPERABILITY.in accident analyses are based on maintaining at least one train of the onsite AC (3.8.1) and DC (3.8.4) power sources and associated distribution systems OPERABLE during accident conditions, i

with an assumed loss of offsite power and. single failure that prevents the operation of one emergency diesel generator (EDG).

In the case of degraded voltage or complete loss of voltage on the 4160. volt electrical supply buses due to faults affecting the 230 kV offsite power supply (no J

, electrical supply signal present), the respective channels will detect the condition and signal automatic start of the EDG associated with the affected electrical supply bus.

Improved Crystal River TS 3.3.8 verifies the OPERABILITY of this system and provides actions to be taken upon the determination of system inoperability.

The staff has reviewed the above added Specification (3.3.8) and believes it results in an enhancement to the improved Crystal River TS.

Therefore, the addition is acceptable.

2.

The licensee proposed to add a new Specification 3.3.15, " Reactor Building Purge Isolation - High Radiation" and to place the specific conditions, required actions, completion times, and surveillance in Specification 3.3.15.

The specification was identified in the existing TS, Table 4.3.2, " Engineered Safety Feature Actuation Systems Instrumentation Surveillance Requirements."

3.

Table item 2.b.i, Control Room Isolation-High Radiation has been removed from the TS Table 3.3-6, " Radiation Monitoring Instrumentation," and included as new Specification 3.3.16.

Following a DBA, LOCA or fuel handling accident (FHA), the high radiation instrumentation is credited with performing the initial Control Complex isolation function and beginning the emergency recirculation mode of operation. This isolation is necessary to limit doses to the Control Room operator.

The principal function of the Control Room Isolation-High Radiation is to provide an enclosed environment for operations personnel from which the plant can be operated to mitigate the consequences of an accident.

The staff concludes that the reorganization of the specifications in paragraphs 2 and 3 above is acceptable.

The licensee has not proposeo to adopt the part of the surveillance requirements in NUREG-1430 for the remote shutdown system (RSS) in SR 3.3.18.2 which reads as follows:

" Verify once every 18 months that each required control circuit and transfer switch is capable of performing the intended function" Currently, the licensee conducts a plant procedure (PT315) to test a portion of the remote control system transfer circuitry to show that affected relay coils will energize and the relays physically close.

The staff concludes that the NUREG-1430 SR 3.3.18.2 is outside of the Crystal River licensing basis, is not in the existing Crystal River TS and the current plant procedure provides an adequate level of safety.

The staff concludes that the existing method of demonstrating that the transfer of the control function is adequate and does not warrant backfitting of the NUREG-1430 surveillance provision. Thus, the i

.m

i

]

- 16 i]

y Crystal. River improved.TS_ requirements for the RSS remain.. effectively-unchanged from the existing TS.

]

2.3.4 Reactor Coolant System (Section 3.4) and Emergency Core Cooling Systems ~

(Section 3.5) i

^

The evaluations of. these two improved Crystal. River TS Sections (3.4 an'd 3.5).

i have been combined'since the proposed change's are mostly administrative and have no-~effect on safety beyond the existing TS. Section 3.5 (ECCS).of the improved Crystal. Rher TS contains the specifications of the' existing TS. 'No specifications were relocated or added. The numbering system of.the improved Crystal River TS is'the same.as that given in NUREG-1430.

1

' The changes. to improved' Crystal River TS Section 3.5 are format. changes, editorial / administrative changes, and minor technical changes to make the 1

improved Crystal River. TS Specifications 3.5.1,,3.5.2, 3.5.3, and 3.5.4

-j consistent with Section 3.5 of NUREG-1430.

Based on its review of the l

improved Crystal River TS.Section 3.5, the staff concludes that these administrative changes are acceptable.

-]

Existin'g TS 3.4.1.1 has been deleted from the improved Crystal. River TS since' l

the safety objective is incorporated into improved Crystal' River TS 3.3.1.. In Modes 1-and-2 the Reactor Protection System (3.3.1) will trip the reactor if:

i 1.

There are three pumps running and THERMAL POWER is not restricted to ti less than 79 percent RTP, or n

2.

There are less than three pumps running.

This is considered an administrative change in the location of the safety-l objective within the TS and is therefore acceptable.

j The requirements for the pressurizer safety valves in the existing TS 3.4.2 '

j!

are not necessary for reactor coolant -system overpressure. protection during plant shutdown conditions.

In MODES 4J and 5,. administrative controls, e.g.,

equipment lockouts, are'used to limit the worst-case overpressure transient,-

. typically a stuck-open-makeup valve. This' ensures that: sufficient time will be: available fo'r operator: action. Overpressure protection is provided-by the pressurizer power operated' relief valve (PORV), Lor the decay heat removal q

system relief valve. The. licensee has-proposed to relocate these provisions.

3 to the FSAR and appropriate plant procedures, as previously. described.

Based.

1 on a-deterministic review, the staff concludes'that these provisions do not l

satisfy the pol _ icy statement guidance for inclusion in TS. Therefore, the

!j control.of these provisions under 10 CFR 50.59 is ' acceptable.

d I

~:

Part of the existing TS 3.4.5 for steam generators, requiring water level maintenance in the once-through steam generator (OTSG), has been transferred to improved Crystal River U Section 3.7.17.

The majority of this existing TS deals with detailed survetiiance requirements for OTSG tube inspection. These inspections, which are prescribed to provide reasonable assurance of SG tube integrity during plant operating conditions, can only be performed during plant shutdown conditions.

Improved Crystal River TS steam generator (OTSG) tube inspection requirements provide the same level of control as those contained in the_ existing TS. Where the existing TS contain an LC0'and numerous surveillance requirements (SR), the improved Crystal River TS contains a single requirement (SR 3.4.12.2) to verify that steam generator tube integrity satisfies the requirements of Specification 5.6.2.10, " Steam Generator Tube Surveillance Program."

Except for minor editorial differences, Specification 5.6.2.10 is identical to the existing TS 4.4. 5.0, 4.4.5.1, 4.4. 5. 2, 4.4.5.3, 4.4. 5. 4, Tabl e 4.4-1, Table 4.4-2, and Table 4.4-6.

Current TS 4.4.5.5 " Reports" is contained in its entirety within improved Crystal River TS 5.7.2 "Special Reports."

Specification 5.6.2.10 contains the information necessary to verify that the OTSGs are OPERABLE. This Specification, in conjunction with improved Crystal River TS SR 3.4.12.2 and the LCO and SR applicability, ensures OTSG tube integrity is restored prior to increasing RCS average temperature above 200* F and provides the tie between the Pro; ram and the RCS Operational Leakage LCO.

Steam generator tube failures during power operation are immediately detectable by steam line or air ejector radiation monitors, while minor tube leaks can be determined by various analyses or evaluations (steam generator chemistry, inventory balances, etc.), as required by improved Crystal River TS LCO 3.4.12 "RCS Operational Leakage" and LC0 3.4.14 "RCS Leakage Detection Instrumentation." This is considered an administrative change in the location of the safety objective within the TS and is therefore acceptable.

Reactor coolant water chemistry is monitored in the existing TS 3.4.7 for a-variety of reasons, one of which is to reduce the possibility of failures in the reactor coolant system pressure boundary caused by corrosion. The monitoring activity is of a preventive nature rather than a mitigative action.

The water chemistry TS do not address any mitigative action for DBAs, and short-term water chemistry anomalies are not expected to affect reactor coolant pressure boundary. integrity. The licensee has proposed' to relocate these provisions to the FSAR and appropriate plant procedures, as previously described.

Based on a deterministic review, the staff concludes that these provisions do not need to be controlled by TS under the policy statement or the Commission's regulations. Therefore, the control of these provisions-under 10 CFR 50.59 is acceptable.

-- 3 L The existing TS 3.4.10 addresses the structural integrity of ASME Code Class 1, 2 and 3 components. ASME Code Class 1, 2, and 3 components are monitored so that the possibility of component structural failure does not degrade the safety function of the system.

The monitoring activity is of a preventive nature rather than a mitigative action.

In addition, surveillances, except for the reactor coolant pump (RCP) flywheel inspection, are already required by 10 CFR 50.55a to be performed in accordance with-Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda.

The RCP flywheel inspection requirement is not covered by other regulatory requirements and is needed for the safe operation of the plant; therefore, this requirement has been included in the Crystal River improved TS..The improved Crystal River TS, Sections 5.6.2.8 and 5.6.2.9 provide a programmatic approach to the requirements relating to the structural integrity of ASME Code Class 1, 2, and 3 components. The staff concludes that the proposed Crystal River improved TS have made appropriate application of the guidance provided in NUREG-1430.

This is considered an administrative change in the location of i

the safety objective within the TS and is therefore acceptable.

The staff finds, based on its review, that Sections 3.4 and 3.5 are consistent with NUREG-1430, the NRC policy statement, and the existing TS, and are therefore acceptable.

2.3.6 Containment Systems (Section 3.6)

The licensee has proposed to reincate or reorganize the following existing TS in sections 3/4.6, " Containment Systems":

Existino TS Sections Title 3.6.4.1 Hydrogen Monitors 3.6.4.2 Hydrogen Purge System 3.6.4.3 Hydrogen Purge Values i

4.6.1.6.1 Containment Tendons 4.6.1.6.2 End Anchorages and Adjacent Concrete Surfaces Hydrocen Monitor _1 Existing TS 3.6.4.1. has been relocated within the Crystal River improved TS, since the safety objective is incorporated into improved

'l TS 3.3.17 " Post Accident Monitoring (PAM) Instrumentation." This is considered an administrative change in the location of the safety i

l objective within the TS and is therefore acceptable.

Hydrocen Purae System The original intent of the hydrogen purge system (existing TS 3.6.4.2) was to control hydrogen in the long term after a large break LOCA.

1 I

i

.. Assumptions of the LOCA analysis include some hydrogen generation from the zirconium cladding-water reaction, from spray reaction with aluminum and zirconium surfaces in the containment, from radiolytic decomposition of the injection fluid and other sources.

For these analyses, gradual hydrogen accumulation is conservatively calculated'to occur over several-days or weeks after the accident has been terminated and core cooling -

stabilization has occurred. This system is not used for mitigation of the core thermal hydraulic transient and is not directly related to the DBA sequence. Additionally, this system is a backup to the hydrogen recombiner which the licensee has connitted to moving onsite should hydrogen accumulation become a problem. The licensee has proposed to relocate these provisions to the FSAR and appropriate plant procedures, as previously described. Based on a deterministic review, the staff concludes that these provisions 60 not need to be controlled by TS under the policy statement or the Commission's regulations.

Therefore, the control of these provisions under 10 CFR 50.59 is acceptable.

Hydroaen Purae Valves Existing TS 3.6.4.3 has been relocated within the Crystal River improved o

TS, since its safety objective with a minor change is incorporated into improved Crystal River TS 3.6.3.

Since these mini-purge valves are.

designed to meet the requirements for automatic containment isolation valves, a Note in LCO 3.6.3 states that these valves may be opened as-needed in MODES 1, 2, 3, and 4 under administrative controls.

The existing TS requirement only allows the valves to be administratively opened for a total of 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> per year, which had been the staff position to control use of purge systems with large containment penetrations. However, the staff position evolved for mini-purge systems into an administrative control which restricts the use of.the mini-purge system, as it is reflected in the improved STS.

Crystal River restricts use of the mini-purge system by an' administrative control which includes provisions for stationing a dedicated operator at the valve control.e, who is in continuous communication with the control room, and the pent; ration can be rapidly isolated when s need for containment isolation is indicated.

The staff concludes that the licensee has established adequate controls for the containment mini-purge system and, therefore, these changes are acceptable.

Containment Tendons and Anchoraces The licensee has accepted NUREG-1430 specifications for tendon surveillances (existing TS 4.6.1.6.1 and TS 4.6.1.6.2).

The' tendon surveillance requirement is not covered by other regulatory requirements and is needed for the safe operation of the plant.

The improved Crystal River TS Section 5.6.2.7 provides a programmatic approach to the requirements relating to containment tendon surveillances and 2

p n,.

t p

p -,

~I

--20 --

]

i inspections.. This is considered an administrative change in the-location of the safety.. objective within the TS and is therefore I

acceptable.

.I The licensee has proposed a modification to specification 3.6 of improved:

A

~

Crystal River TS with the ' addition of the following:

j 1.

The licensee,has accepted the NUREG-1430 specifications, except that Crystal River no longer relies on the Spray. Additive Systems as described in NUREG-1430. The licensee has replaced the Spray-Additive

-f System with a Containment. Emergency Sump pH Control System (the Sump.pH Control System.) These systems essentially accomplish the same F

i function. ~ The staff had previously authorized FPC to replace theLspray:

additive with the Sump pH Control System aslpart of Amendment 145 to.the-

?!

Crystal River operating. license, dated July 23, 1992. -The staff finds j

the change acceptable.

2.

A modification has been made to improved Crystal River TS 3.6.2 to add to Note 1 the words, "or for emergencies involving personnel safety."

.t There are circumstances where an at-power containment entry would be required during the period of time when one' airlock is inoper ble.

In a

this case, entry would be made through the 0PERABLE airlock. However, i

the containment is a harsh environment with the reactor.at power.

In-the event that something happens to the individual who had entered j

containment, plant personnel. should-be permitted to proceed through the mat expeditious rescue path in order to get'that individual!out of-containment and to provide medical attention. The staff concludes that--

1 this change is an acceptable alternative to invoking the provisions-of.

l 10 CFR 50.54(x) for such' circumstances, q

9 Since ingress is currently permitted to repair the inoperable door, there is no reason to exclude ingress for personnel safety..The staff, therefore, finds the change acceptable.

J 2.3.7 Plant Systems-(Section 3.7)

'l The licensee has elected to relocate or. reorganize the following existing TS:

i Existino TS Sections Title q

3/4.7.2 Steam Generator Pressure / Temperature Limitations-3/4.7.6' Flood Protection 3/4.7.9 Hydraulic Snubbers r-J

Existina TS Sections Title 3/4.7.10 Sealed Source Contamination 3/4.7.11.1**

Fire Suppression Water Systems 3/4.7.11.2,,

Deluge and Sprinkler System 3/4.7.11.3 Halon System 3 / 4. 7.11. 4**

Fire Hose Stations 3 / 4. 7.12**

Fire Barrier Penetrations 3/4.7.13 Waste Gas Decay Tanks 3/4.7.13.5 Waste Gas Decay Tank Explosive Gas Mixture Steam Generator Pressure /Temoerature Limitations The limitations on steam generator pressure and temperature provide protection against non-ductile failure of the secondary side (shell) of the steam generator (OTSG).

The limits are based on the structural analysis of the OTSG and are calculated using the ASME Code for Class 1 components. The limitations do not reflect initial condition assumptions in the DBA.

Except for OTSG wet layup, the conditions which permit violation of these limits are virtually non-existent. The risk of shell failure is small.

The provisions will be relocated to the FSAR and appropriate plant procedures, as previously described.

Flood Protection The existing LC0 provides requirements on water-tight door closure during periods of ultimate heat sink high water level coincident with potential hurricane conditions. The requirements ensure that the plant is shut down and safety-related equipment is sheltered in the event the probable maximum hurricane (PMH) surge level were to occur.. Appropriate plant response to severe weather has been addressed separately from the TS as part of the staff review of the FPC Station Blackout: response.

The provisions will be relocated to the FSAR, as previously described, except for the flood response practices that will be relocated in the Emergancy P1an.

Hydraulic Snubbers The intent of the existing TS is to require various inspectiohs to ensure that snubbers are maintained in good condition and therefore in operational readiness to restrain piping and equipment motion for earthquakes and large break LOCAs. Snubbers are not specifically identified in PRA analyses, although seismic fragility evaluations. are performed for various important equipment.

PRA analyses indicate that

• • Relocated via Amendment #147, 1/22/93

. plant risk is usually only significant for earthquakes considerably stronger than the design basis.

Recent studies for large break LOCAs show that leak-before-break phenomena will likely permit detection of incipient piping failure prior to catastrophic failure, thus indicating that the importance of LOCA-based snubbers is less than previously thought and that some snubbers may be removed from the design.

Since the Design Basis Earthquake is a condition for design, seismic j

snubbers are included in the stress analysis for the RCS and interconnecting piping and are credited for maintaining load below the ASME Code allowable.

Thus, snubber surveillance requirements do not meet the guidance for inclusion in IS and the snubber surveillance requirements are more appropriately controlled in a more integrated and coordinated programmatic surveillance document.

The provisions will be relocated to the FSAR and appropriate plant procedures, as previously described.

Sealed Source Contamination This TS specifies' limitations on fixed contamination for sources requiring leak testing.

Radioactivity released from sealed sources is not a factor in any DBA analysis.

The provisions will be relocated to the Health Physics procedures which can be changed in accordance with 10 CFR 20.

The licensee has proposed to relocate these provisions to the FSAR and appropriate plant procedures, as previously described.

Based on a deterministic review, the staff concludes that these provisions do not satisfy the policy statement guidance for inclusion'in TS. Therefore, the control of these provisions under 10 CFR 50.59, the Emergency Plan Implementing Procedures -(EPIP), and 10 CFR 20 is acceptable.

Waste Gas Decay Tank The purpose of this TS is to limit the amount of gaseous activity inventory available for release in the event of a rupture or other failure of a Waste Gas Decay Tank. The activity limit is established to maintain resultant doses at the site boundary within 10 CFR Part 20 limits.

The improved Crystal River TS address this limit within Specification 5.6.2.13, " Explosive Gas and Storage Tank Radioactivity Monitoring Program," by providing a TS Administrative Control which-limits the amount of activity contained within the tank. Thus, the activity limit in the existing TS is effectively retained within the improved Crystal River TS by specifying the whole body dose to an

L.;

, individual in the unrestricted area in the event of a release of the tanks' contents.

Waste Gas Decay Tank Explosive Gas Mixture The purpose of this TS is to limit the potential for creating an explosive mixture of hydrogen and oxygen within the Waste Gas Decay Tank (WGDT).

This is intended to minimize the potential for a gross rupture of the. tank which would, in turn, result in an uncontrolled release of radioactivity.

Even in the event of such a rupture, the activity limitation (existing TS 3/4.7.13.1) ensures that radiological protection-standards for the public are not exceeded.

I The Crystal River improved TS include Specification 5.6.2.13, " Explosive Gas and Storage Tank Radioactivity Monitoring Program," which prcvides for a TS Administrative Control surveillance program to monitor and maintain the concentration of hydrogen and oxygen in the tank to within limits.

These requirements are included in the TS as required by 10 CFR 50.36a. The changes are considered administrative changes in the location of the requirements within the TS and are therefore acceptable.

The licensee has adpoted a significant portion of the NUREG-1430 provisions for the plant systems. Additional specifications and clarifications beyond the existing TS have been made throughout this section in order to make the section more user-friendly, consistent with current staff positions, and consistent with the current licensing basis for Crystal River.

In particular, the following changes are incorporated in the Crystal River improved TS:

1.

A note has been added to the ACTIONS of LC0 3.7.2, "MSIVs,"-directing that the actions of LCO 3.7.4, " Turbine Bypass Valves," be taken when an inoperable main steam isolation valve (MSIV) results in one or more turbine block valves (TBVs) being rendered inopercle. The TBVs are located downstream of the HSIVs.

Depending on r. tic MSIV is inoperable, the TBV could be incapable of performing _its Steam Generator Tube Rupture accident cooldown function.

LC0 3.0.6 providas that the ACTIONS of 3.7.4 need'not be taken when the inoperability of the TBVs is due to inoperability of the MSIV support system, for which there is a specified -

action.

This note is similar to the one provided in LCO 3.6.3.

" Containment Isolation Valves."

2.

The LC0 for "MFIVs," LCO 3.7.3, has been modified to reflect the current' licensing basis for Crystal River. As discussed in FSAR Chapter 14, Accident Analysis, two isolation valves are provided in each flowpath from the main feedwater (MFW) pump suction valve to the OTSGs. One of these valves was designed to have the capability to isolate the MFW

b i -

within-the response time credited in.the~ MFW line break analysis. Two i

valves are provided for isolation, to provide redundant isolation-devices.

The staff has evaluated the above LCO additions and concludes' that 'they comply -

with the existing Crystal' River licensing basis and are therefore acceptable. '

j FPC.has proposed not to add NUREG-1430 Specifications 3.7.11, " Control. Room Emergency Air Temperature Control System;" 3.7.12. " Emergency Ventilation System;" and 3.7.13, " Fuel Storage Pool Ventilation System," to the improved i

Crystal River TS. The reasons are as follows i

1.

Specification 3.7.11 - The temperature in the Control Complex /Consol 1

Room is maintained at a pre-determined setpoint (~70*F) during normal l

operations.

The Chilled Water System is relied upon to accomplish.this i

function and also fulfills this function post-accident (same system configuration in both cases). A heat. balance analysis was performed "j

which showed that Control Complex heat loads are approximately the same-during normal and accident operations, so that the difference'between the two conditions is negligible. Based upon this line of= reasoning,'if.

the Chilled Water System is capable of maintaining temperature during i

steady-state operations, it is considered capable of controlling temperature post-accident. Any significant increase in Control Compler temperature will be detected by the Control Room operators'well in advance of apprc. aching any real operational limit, be it equipment operability or operator habitability.

2.

Specifications 3.7.12 and 3.7.13 - The Auxiliary Building Ventilation :

System provides the temperature maintenance function of these systems at.'

Crystal River.

For Crystal River, the ventilation filtration aspect' of i

these systems is not credited as part 'of any. accident analysis.

l Based on the licensing basis, the accident analysis, and/or determination of an adequate level of safety, and the above discussion, the staff accepts the deletion of NUREG-1430 recifications 3.7.11, 3.7.12 and 3.7.13 for the improved Crystal River.5.

t 2.3.8 Electrical (Section 3.8)

The licensee did not. include some of the requirements of NUREG-1430, Section 3.8 related to certain diesel generator surveillance tests. :The staff.

concludes that these specifications are outside of.the Crystal River licensing basis, are not in the existing Crystal River TS, and that an adequate level. of safety is ensured by the existing electrical systems surveillances'which-are included in the improved Crystal River TS. The staff concludes that these differences do not warrant backfitting of the NUREG-1430 surveillance.

1 provisions. Therefore, not including these items in the Crystal River-i

-i 6

~

l 3

l P

improved TS-is acceptable.

f The battery surveillance requirements in the Crystal River improved TS have.

1 been revised to_ reflect the battery surveillance requirements:in NUREG-1430, 1

as revised-to be consistent with proposed changes;in the industry standard for q

battery testing (IEEE-450) and to accommodate 24-month refueling cycles. ~The scope and frequency for the battery service. test and performance discharge.

test' are less restrictive in the Crystal River improved TS than the existing.

requirements in TS 4.8.2.3.2.d and e.

However, the Crystal River. improved: TS.

include ' additional. requirements for the performance discharge test, as well as a' provision for an. alternate modified performance discharge test..The-set of.

battery testing: requirements reflected in the Crystal River improved TS will minimize testing that can be detrimental to the life of the batteries, but will provide a more effective measure of the capability of the DC electrical power system.

On this basis, the. staff concludes that the proposed changes a

are acceptable.

2.3.9 Refueling Operations (Section 3.9) l The licensee _ proposed to reloc. ate the following TS to other licensee-controlled. documents:

Existina TS Number Title 3.9.3

. Decay Time j

3.9.5 Communication 3.9.6 Fuel Handling Bridge Operations r

3.9.7 Crane Travel 3.9.11 Missile Shields 3.9.12 Auxiliary Building Ventilation

]

The only design basis accident considered during refueling is the' fuel' handling accident. These specifications-do not involve initial conditions for or mitigation of this DBA.

For example, crane travel is physically limited by-i design provisions and the accident analysis. decay time assumptions;are:

conservative with respect to actual fuel offload sequences. --The licensee has

.propsed to relocate these provisions to the FSAR and' appropriate plant procedures, as previously described.

Based on a deterministic review,.the t

staff concludes that these provisions do not satisfy the policy. statement-

~ guidance for inclusion in TS. Therefore, the control of these provisions t

under 10 CFR 50.59 is acceptable.

1

l 2.4 - Design Features _(Section 4.0)

This.section contains the same material as found in the existing TS, except for the Cyclic' or Transient Limits Table. The licensee has' accepted NUREG-1430 Section 4.0 and Section 5.6-as they relate-to-the Cyclic or

-l b

I

9 Transient Limits Table.

The table will be relocated to the FSAR (FSAR Table 4.8) and a new program, Specification 5.6.2.5 " Component Cyclic and Transient Limit" will be implemented which provides controls to track cyclic and transient occurrences to ensure that components are maintained within the design limits.

The staff finds that the NUREG-1430 change will provide the equivalent safety function and controls as the existing Crystal River' specifications.

Thus the staff has reviewed the proposed relocation.of the Table outside the TS and finds it acceptable. The relocation of the controls for Component Cyclic and Transient Limits to Section 5.6 is considered an administrative change in the location of the safety objective within the TS.

i Therefore this section is acceptable.

2.5 Administrative Controls (Section 5.0)

In August 1987, FPC submitted an amendment request to remove several administrative controls from the existing TS, because they believed that those TS duplicated other regulatory requirements.

The staff rejected the request in May 1988, based on the premise that the STS would address this subject. At the request of FPC, the staff and FPC met on March 3, 1993 to discuss the requirements for TS administrative controls and the implementation of the

.i August 1987 amendment request to the Crystal River improved TS. As a result of that meeting, the staff reopened the review of the August 1987 amendment request and proposed additional changes to the improved Crystal River TS which would relocate other administrative control requirements to licensee-controlled documents.

FPC has proposed to implement Section 5.0 of NUREG-1430, with some plant-specific differences, and relocate some of the specifications in the October 25, 1993 staff letter to the Owners Group on the proposed content of Section 5.0 of the STS.

Specifically, the licensee has proposed to relocate the following existing administrative control provisions to other licensee-controlled documents:

Existina TS Section Title 6.2.2a Minimum Shift Composition Table 6.2.1 Minimum Shift Composition 6.2.2.e Senior Reactor Operator (SRO) Present During Fuel Movement 6.4 Training 6.5 Review and Audits 6.8.1.d Security Plan Implementation 6.8.1.e Emergency Plan Implementation 6.8.2 Review and Approval Process 6.8.3 Temporary Change Process 6.8.4.b Radiological Environmental Monitoring Program 6.9.1 Startup Report 6.9.2.a ECCS Actuation Special Report 4

t Existino TS SectiQD Title 6.10 Record Retention 6.11 Radiation Protection Program 6.14 Process Control Program Minimum Shift composition The licensee proposes the Minimum Shift Crew Composition Specification and associated table not be retained in TS.

10 CFR 50.54(k), (1) and (m) provide the requirements for shift complement regarding licensed operators. The regulations describe the minimum shift composition for operating modes, as well as for cold shutdown and refueling. The requirements in this specification and the associated table.are located in the Crystal River Administrative Instructions for~10 CFR 50.54 and in the Emergency Plan.

Additionally, Specification 5.1.2, new Specifications 5.2.2.a and 5.2.2.b, and Specification 5.2.2.c of the improved Crystal River TS specify when licensed and non-licensed opes tors are required to be in the control room. The staff concludes that the regulatory requirements provide sufficient control of these provisions and removing them from TS is acceptable.

SRO Present Durino Fuel Movement The licensee proposes that the requirement that an SRO to be present during fuel handling and to supervise all core alterations not be retained in the TS.

This is required by 10 CFR 50.54(m)(2)(iv) and need not be controlled by TS to assure safe operation of Crystal River. The current regulation states, "Each licensee shall have present, during alteration of the_ core of a nuclear power unit (including fuel loading or transfer), a-person holding a senior operator license or a senior operator license limited to fuel handling to directly supervise the activity and, during this time, the licensee shall not assign other duties to this person."

This requirement is specified in the Administrative Instructions for 10 CFR 50.54.

The staff concludes that the regulatory requirements l

provide sufficient control of these provisions and removing them from TS-is acceptable.

Trainino i

The licensee proposes that the requirements on training be deleted from the improved Crystal River TS on the basis that they are' adequately addressed by other Section 5.0 administrative controls and Section 12.2

E

"' of the FSAR, as well as in the regulations. The improved Crystal River TS 5.3, " Unit Staff Qualifications," provides adequate requirements to assure an acceptable, competent operating staff.

Each member of the Crystal River staff is required to meet or exceed the minimum qualifications of specific Regulatory Guides.or' ANSI Standards acceptable to the NRC staff. The improved Crystal River TS 5.3 describes the' details of the required qualifications.

FSAR Section 12.2 describes the details of the Crystal River training program.

Additionally, the improved Crystal River TS 5.2, " Organization," details Crystal River staff requirements. New Specifications 5.2.2.a and 5.2.2.b, Specification 5.2.2.c, and 10 CFR 50.54 describe the minimum shift crew composition and delineate which positions require an'R0 or SR0 license. Training and requalification of those positions are as specified in 10 CFR Part 55.

The licensee has proposed to relocate these provisions to the FSAR'and appropriate plant procedures, as previously described.

Based on a deterministic review, the staff concludes that these provisions do not need to be controlled by TS under the policy statement or the Commission's regulations. Therefore, the control.of these provisions under 10 CFR 50.59 is acceptable. The staff also concludes that the regulatory requirements provide sufficient control. of these provisions and removing them from TS is acceptable.

Review and Audits The licensee proposes that the review and audit. functions be relocated from the improved Crystal River TS on the basis that they are adequately controlled elsewhere. These TS provisions are not necessary.to assure safe operation of Crystal. River, given the requirementsin the Quality Assurance (QA) Program implementing 10 CFR 50.54 and 10 CFR Part 50, Appendix B to control the requirements for all review and audit functions except those associated with the security and emergency plans.

The security and emergency plan review and audit functions are relocated to their respective plans in'accordance with.a staff-proposed Generic Letter.

Such an approach would result in an equivalent level of regulatory authority while providing for a more appropriate change control process.

The net effect of the change is that the level of safety of plant operation is unaffected and NRC and FPC resources associated with processing license amendments to this administrative control are optimized. The following points summarize the. reasons for removing the review and audit requirements from the improved Crystal River TS.

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1 1.

The on-site review function, composition, alternate membership, meeting frequency, quorum, responsibilities, authority and records are all covered in equivalent detail in ANSI N18.7-1976.

These requirements are in the QA Program and change control is provided by 10 CFR 50.54(a).

2.

The off-site review group is also addressed, although with less detail, in ANSI N18.7-1976. The QA Program includes the requirements for the off-site review group.

Therefore, duplicating the review and audit function of the off-site review group in the Crystal River improved TS is unnecessary.

3.

Audit requirements are specified in the QA Program to satisfy 10 CFR 50, Appendix B, Criterion XVIII. Audits are also covered by ANSI N18.7, ANSI N45.2, 10 CFR 50.54(t), 10 CFR 50.54(p), and 10 CFR 73.

Therefore, duplication of these regulatory requirements does not enhance the level of safety of the Crystal River plant, nor are the provisions relating to audits necessary to assure safe operation of Crystal-River.

The licensee has proposed to relocate those provisions that_ are not otherwise covered by regulatory requirements to the QA Plan. The staff concludes that the sufficient regulatory controls exist for the QA Plan E

such that removing these provisions from the TS and relocating them to l

the QA Plan is acceptable.

Security Plan Implementation and Emeraency Plan Imolementation The licensee proposes to relocate the requirements to establish, implement, and maintain procedures related to the Emergency Plan and Security Plan.

Since the Security Plan requirements are specified in 10 CFR 50.54, 73.40, 73.55, and 73.56 and the Emergency Plan requirements are specified in 10 CFR 50.54 and 10 CFR Part 50, Appendix E, Section V, the staff has proposed a Generic Letter to remove the requirements from the STS and relocate them to their respective plans.

The requirements in TS 6.8.1.d for the review of the security program and implementing procedures and in TS 6.8.1.e for the review of the station emergency plan and implementing procedures will be included in their respective plans.

Further changes in these review requirements must be made in accordance with 10 CFR 50.54(p) for the Security Plan and 10 CFR 50.54(q) for the Emergency Plan. The staff concludes that.

in conjunction with this change to the plans, the extensive requirements for emergency planning in 10 CFR 50.47 and 50.54 and for security in 10 CFR 50.54 and 73.55 for drills, exercises, testing, and maintenance of the program, provide. adequate assurance that the objective of the previous TS for a periodic review of the program and changes to the

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t programs will be met. Therefore, duplication of the requirements contained in the regulations does not enhance the level of nuclear safety for Crystal River.

6 The staff concludes that other regulatory requirements provide sufficient control of these provisions and removing them from'TS is acceptable.

Review and Aporoval Process and Temocrary Chanae Process The licensee is proposing to relocate both the review and approval process and the temporary change process for procedures to the QA Plan.

This proposal is based on the existence of the following requirements which duplicate 10 CFR 50.36 in these areas.

The requirement for procedure control is mandated by 10 CFR 50, Appendix B, Criterion II and Criterion V.

ANSI N18.7-1976, which is an-NRC staff-endorsed document used in the development of many' licensee QA plans, also contains specific requirements. related to procedures.

The

' licensee has committed to follow ANSI N18.7-1976 as a means to comply.

with 10 CFR 50, Appendix B.

ANSI N18.7-1976, Section 5.2.2 discusses procedure adherence. This section clearly states that procedures shall be followed, and the. requirements for use of procedures shall be prescribed in writing. ANSI N18.7-1976 'also discusses temporary changes to procedures, and requires review and approval of procedures _to be defined. ANSI N18.7-1976, Section 5.2.15 describes the review,- approval-and control of procedures.

This section describes the requirements for.

the licensee's QA Program to provide measures to control and coordinate the approval and issuance of documents, including changes _thereto, which prescribe' all activities affecting quality. -The-section further. states that each procedure shall be reviewed _ and approved prior to initial use.

The required reviews-are also described. ANSI N45.2-1971, Section 6, also specifies that the QA Program describe procedure. requirements.

The licensee will continue to implement a QA Program _in accordance with the requirements of 10 CFR Part 50, Appendix _B, which provides appropriate controls for the review and approval of procedure changes.

The staff concludes that these regulatory requirements provide sufficient control of these provisions and removing them from TS is acceptable.

Radioloaical Environmental Monitorino Procram i

The Radiological Environmental Monitoring Program requires that procedures be prepared for monitoring the radiation'and radionuclides in the environs of Crystal River consistent with the guidance specified in 10 CFR Part 50, Appendix I.

These procedures are developed'to ensure l

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, that radioactive effluents are restricted to levels as low as reasonably achievable, and have no impact on plant nuclear safety. The details and description of the program are already ~ contained in the ODCM, as specified by existing TS 6.8.4.b.

The staff concludes that these regulatory requirements provide sufficient control-of these provisions and removing them from TS is acceptable.

Startuo Report The requirement to submit a Startup Report has been deleted from the improved Crystal River _ TS.

The report was a summary of plant startup and power escalation testing following receipt of the Operating License, an increase in licensed power level, the installation of nuclear fuel with a different design or manufacturer than the current fuel, and i

modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of Crystal River. The report'provided a mechanism for the staff to review the appropriateness of licensee activities after-the-fact, but contained no requirement for staff l

approval.

The approved 10 CFR Part 50, Appendix B, QA Plan and Startup i

Test Program (FSAR Section 13) provide assurance that the listed 4

activities are adequately performed and that appropriate corrective-l actions, if required, are taken.

Inasmuch as this report was required to be provided to the staff within 90 days following completion of' the respective milestone, it was clearly-not necessary to assure operation of the facility in a safe manner for the interval between completion of the startup testing and submittal of the report. Additionally, because there was no requirement for the staff to approve the report, the Startup Report is not necessary to assure operation of Crystal River in a safe manner. Therefore, the removal of this requirement is acceptable.

ECCS Actuations Special Report The licensee proposes to relocate the TS requirement to submit a special report for ECCS actuations from the improved Crystal River TS.

10 CFR 50.73(a)(2)(iv) provides requirements for the licensee to submit.

a Licensee Event Report in the event of an ECCS actuation. The report is required to be submitted within 30 days and'will contain the same type of information as the special report. The above requirements are included in the Crystal River Administrative Instruction-for 10 CFR 50.72 and 10 CFR S0.73. The staff concludes that these regulatory-requirements provide sufficient control of these provisions and removing them from TS is acceptable.

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b -i Record Retention tThe licensee proposes that the' requirements on record retention be 1

relocated from.the improved Crystal River TS on the basis that they 'are'

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adequately addressed by the QA Program (10 CFR Part 50,. Appendix B,-

Criteria < XVII) and the related QA P1an.

Facility operations are performed in accordance with approved written' procedures. Areas include normal startup, operation and shutdown,_

abnormal conditions and emergencies, refueling,. safety-related -

maintenance, surveillance and testing, and radiation control.

Facility records' document-appropriate station operations;and activities...

1 Retention!of these records provides documentation retrievability for.

review of compliance with requirements and regulations.

Post-compliance.

review of records does not directly assure operation of the facility _in a safe manner, as activities described in these documents have already-been performed.

In addition, numerous other regulations'such as 1

10 CFR Part 20; Subpart L, and 10 CFR 50.71 require the retention of l

certain records related to operation of the nuclear plant. The staff:

concludes that these regulatory requirements provide sufficient: control-of these recordkeeping provisions and removing them from.TS.is

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acceptable.

Radiation Protection Procram The licensee proposes to relocate the TS program description for thel Radiation Protection Program. The Radiation Protection. Program requires:

procedures to be prepared for personnel radiation-protection consistenti e

with the requirements of 10 CFR Part 20.. The requirement to have1 procedures to implement Part 20 is also contained within 10 CFR 20.1101(b).

Periodic review of these procedures;is addressed under-10 CFR 20.1101(c). The program requirements specified above are described in FSAR Section 11.5 and in the' Administrative Instructions.

The licensee has proposed to relocate these provisions to the FSAR and appropriate plant procedures, as previously described. -Based on' a deterministic review, the staff: concludes that these provisions do'not satisfy the policy-statement guidance _ for inclusion"in TS. The~ staff concludes that the requirements of the rule. provide sufficient control' of these provisions, and'that 10 CFR 50.59 provides adequate controls for those. provisions in.the FSAR and plant procedures.-

Process Control Proaram

'I The licensee proposes to relocate the TS program description'for the Process Control Program (PCP). The' PCP is described in the.QA Plan.

j The PCP implements the requirements of 10 CFR.Part 20,10 CFR~ Part 61,

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._ and 10 CFR Part 71. The staff concludes that the regulatory controls for the QA Plan provide sufficient control of these requirements and removing these provisions from the TS is acceptable.

Other differences from the administrative controls, as published in NUREG-1430 in September 1992 are as follows:

In Plant Radiation Monitorino The In Plant Radiation Monitoring Program (NUREG-1430, Section 5.7.2.5) provides controls to ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions.

This program was developed to minimize radiation exposure to plant personnel (post-accident).

The In Plant Radiation Monitoring Program administrative control does not involve monitoring process variables that are initial conditions for a design basis transient or accident, nor does it involve a primary success path to mitigate a DBA.

Therefore, this program has been relocated to a number of different plant procedures.

The training aspect is contained in the training program for personnel. The provisions for monitoring and performing maintenance of sampling and analysis equipment are addressed in the -

chemistry and radiation protection procedures.

The licensee has proposed to relocate these provisions to the FSAR and appropriate plant procedures, as previously described.

Based on a deterministic review, the staff concludes that these provisions do not satisfy the policy statement guidance for inclusion in TS. Therefore, the control of these provisions under 10 CFR 50.59 is acceptable.

Fire Protection Procram The Fire Protection Program (NUREG-1430, Section 5.7.2.18) provides controls to ensure that appropriate fire protection measures are maintained to protect the plant from fire and to ensure the capability to achieve and maintain safe shutdown in the event of a fire.

The administrative control was originally developed to ensure the capability to provide for alternate / dedicated safe shutdown in accordance with 10 CFR 50, Appendix R.

As such, it allows for the ability to place the unit in a safe condition in the event of a fire.

The relocation of this administrative control from TS is also consistent with the guidance in NRC Generic Letter (GL) 86-10, " Implementation of Fire Protection Requirements."

In that letter, the staff concluded that the provisions of 10 CFR 50.59 should apply directly to changes the licensee desired to make in the fire protection' program so long as those changes did not adversely affect the ability to achieve and maintain safe shutdown. The standard license condition, included within

".. t GL 86-10, stated that changes which adversely affect the ability to achieve and maintain safe shutdown in the event of a fire ' required prior-approval of the staff..Thus, the license condition established.as part of the NRC GL 86-10 implementation also makes this administrative control unnecessary.

This license condition is License Condition 2.c.9 of Operating License No. DPR-72 for Crystal River. Therefore, the control of these provisions under the terms of the license condition and 10 CFR 50.59 is acceptable.

Shift Technical Advisor The licensee has proposed to relocate NUREG-1430 Specification 5.2.2.g and the changes proposed to this specification in the October 25, 1993 letter to improved Crystal River TS 5.3.1 (existing TS 6.3.1).

FPC is using existing plant-specific wording to describe the Operations Technical Advisor responsibilities and qualifications. This is considered an administrative change in location of the safety objective and is therefore acceptable.

Other Format Chanaes The licensee has also proposed to reorganize specification 5.6.and-specification 5.8 in NUREG-1430 to improved Crystal River TS 5.6,

" Procedures, Programs, and Manuals." Since both of these specifications define programs to be implemented, the licensee has made them part of program specifications instead of separate administrative control specifications. Similarly, the licensee has proposed to reorganize specification 5.7.2.7 (existing TS 6.8.4a) to specification 5.6.2 of the improved Crystal River TS. This program has been combined with the Offsite Dose Calculation Manual (ODCM) (improved TS 5.6.2.3).

These are considered administrative changes and are therefore acceptable.

The licensee did not include other details of the NUREG-1430 admir'strative controls. The specifications or portions of specifications not. inc.luded are:

1.

Specification 5.1.1 relating to the phrase "in accordance with approved administrative procedures," in relation to certain approvals.

2.

Specification 5.1.2 relating to the annual management directive on shift supervisor responsibility, 3.

Specification 5.2.2.f relating to the licensing of the operations

manager, 4.

Specification 5.7.1.1.b relating to NUREG-0737, Emergency Operating Procedures,

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. 5.

Specification 5.7.2.12 relating to the frequency table for inservice

testing, 6.

Specification 5.7.2.16 relating to separately specifying the quantity of radioactivity in unprotected outside liquid storage tanks, 7.

Specification 5.9.1.3 relating to the statements on the NRC Thermoluminesence 00simeter (TLD) Program, and 8.

Specification 5.9.2.b relating to Special Reports on emergency diesel generator failures.

These specifications or portions of specifications are beyond the existing Crystal River licensing basis, are not in the cdsting TS, or are not applicable to the plant design. The staff concludes that these differences do not significantly affect safe operation of the plant or there are other regulatory requirements that ensure adequate controls for these provisions.

Therefore, these differences are acceptable.

Additional clarifications and editorial changes beyond NUREG-1430 TS have been

.made throughout this section in an attempt to make the section more y

understandable and consistent with the current licensing basis.

For example, the Diesel Fuel Oil Testing Program has been modified to conform to the existing surveillances and frequencies specified in existing TS Surveillance 4.8.1.1.2.b.

These changes and clarifications are consistent with the existing licensing basis and are therefore acceptable.

3.0 STATE CONSULTATION

Based upon the written notice 'of the proposed amendment,- the Florida State official had no comments.

4.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact have been prepared and published in.the Federal Reaister on December 10, 1993 (58 FR 64987). Accordingly, based upon the environmental assessment, the staff has determined that the issuance of the amendment will not have a significant effect on the quality of the human environment.

5.0 CONCLUSION

The Crystal River improved technical specifications provide clearer,. more readily understandable requirements to ensure safe operation of the plant.

The staff has concluded that they satisfy the guidance in the NRC policy-statement with regard to the content of technical specifications, and' conform e

A i to the model provided in NUREG-1430 with appropriate modifications for plant-specific considerations. The staff has concluded that the Crystal River improved technical specifications satisfy Section 182a of the Atomic Energy Act and 1-0 CFR 50.36.

Further, the licensee must comply with other applicable regulations. On this basis, the staff concludes that the proposed Crystal River improved technical specifications are acceptable.

Most design features and plant procedures will be described in the FSAR so-that requirements relocated from the technical specifications will be controlled under the provisions of 10 CFR 50.59. Other relocated requirements will be controlled by established programs, for which changes are controlled by related regulatory requirements; these programs include the ODCM, the QA Plan, the EPIP, and RETS.

Specific administrative controls have been removed where regulatory requirements establish adequate controls for the related t

procedures and changes to those procedures. The staff concludes that adequate-controls will exist for all of the requirements relocated from the technical specifications to other licensee-controlled documents.

The staff concludes that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation of Crystal River in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of this amendment will not-be inimical to the common defense and security and will provide adequate protection to the health and safety of the public.

Principal Contributors:

J. Miller R. Giardina R. Lo C. Schulten H. Silver R. Croteau h

ATTACHMENT

SUMMARY

OF CRYSTAL RIVER UNIT 3 RELOCATED TECHNICAL SPECIFICATIONS EXISTING TS TITLE RELOCATION NOTES 3/4.1.2.1 Boration Flow Path - Shutdown FSAR and procedures 3/4.1.2.2 Boration Flow Path - Operating 3/4.1.2.3 Makeup Pump - Shutdown 3/4.1.2.4.1 Makeup Pump - Operating 3/4.1.2.4.2 Makeup Pump - Operating (Mode 4) 3/4.1.2.5 Decay Heat Removal Pump - Shutdown 3/4.1.2.6 Boric Acid Pumps - Shutdown 3/4.1.2.7 Boric Acid Pumps - Operating 3/4.1.2.8 Borated Water Sources - Shutdown 3/4.1.2.9 Borated Water Sources - Operating 3/4.3.3.1.1.a.1 Fuel Storage Pool Area Rad Monitor FSAR and fuel handling procedures 3/4.3.3.1.2.c Condenser Vacuum Pump Exhaust FSAR and ODCM 3/4.3.3.1.2.e Decay Heat Closed Cooling Water FSAR Chapter 11 and ODCM Rad Monitor 3/4.3.3.2 Incore Detectors FSAR and procedures; some instruments retained 3/4.3.3.3 Seismic Instrumentation FSAR and procedures 3/4.3.3.4 Meteorological Instrumentation FSAR, procedures, ODCM, and EPIPs 3/4.3.3.11 Chlorine / Sulfur Dioxide Detection FSAR and procedures 3/4.3.3.8 Radioactive Liquid Effluent Amendment #141 3/4.3.3.9 Radioactive Gaseous Effluent 3.6.4.1 Hydrogen Analyzers FSAR and procedures; some 3.6.4.2 Hydrogen Purge System post accident monitors and 3.6.4.3 Hydrogen Purge Valves containment function retained 3/4.7.2 Steam Generator P/T Limits FSAR and hydro testing procedures 3/4.7.6 Flood Protection FSAR Chapter 2 and EPIPs 3/4.7.9 Hydraulic Snubbers FSAR and procedures; TS Support Function 3/4.7.10 Sealed Source Contamination Part 20 Procedures 3/4.7.11.1 Fire Suppression Water Systems Amendment #147 3/4.7.11.2 Deluge and Sprinkler System 3/4.7.11.3 Halon System 3/4.7.11.4 Fire Hose Stations 3/4.7.12 Fire Barrier Penetrations p

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EXISTING TS TITLE RELOCATION NOTES' l

3/4'7.13 Waste Gas Decay Tanks.

FSAR, procedures, DDCM and 3/4.7.13.5 Waste Gas Decay Tank Explosive Gas admin instructions-i Mixture

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3.9.3 Decay Time FSAR and procedures; decay 1

3.9.5 Communication time not a practical.

j 3.9.6 Fuel Handling Bridge Operations limit,.and aux building 3.9.7 Crane Travel HVAC included in filter 3.9.11 Missile' Shields testing program 3.9.12 Auxiliary Building Ventilation 6.2.2a Minimum Shift Composition 50.54(m) admin instruction.

l Table 6.2.1 Minimum Shift Composition 6.2.2.e SR0 Present During Fuel Movement l

6.4 Training FSAR 12.2 and procedures 6.5 Review and Audits QA Plan 6.8.1.d Security Plan Implementation Consistent with proposed 6.8.1.e Emergency Plan Implementation Generic Letter 6.8.2 Review and Approval Process QA Plan-6.8.3 Temporary Change Process 6.8.4.b Radiological Environmental FSAR and RETS Monitoring Program 6.9.1 Startup Report FSAR Chapter 13 i

6.9.2.a ECCS Actuation Special Report 10 CFR 50.72/50.73 l

admin instruction:

6.10 Record Retention QA Plan 6.11 Radiation Protection Program FSAR 11.5, procedures, and Part 20 admin instruction 6.14 Process Control Prooram QA Plan 1

FSAR - Final Safety Analysis Report ODCM - Offsite Dose Calculation Manual EPIP - Emergency Plan Implementing Procedures l,